NUREG-1821

Final Safety Evaluation Report

on the Construction Authorization

Request for the Mixed Oxide Fuel

Fabrication Facility at the Savannah

River Site, South Carolina

Docket No. 70-3098

Duke Cogema Stone & Webster, L.L.C. 

 U.S. Nuclear Regulatory Commission

Office of Nuclear Material Safety and Safeguards

Washington, DC 20555-0001
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 NUREG-1821 

 

Final Safety Evaluation Report

on the Construction Authorization 

Request for the Mixed Oxide Fuel

Fabrication Facility at the Savannah

River Site, South Carolina 

Docket No. 70-3098

Duke Cogema Stone & Webster, L.L.C.

Manuscript Completed: March 2005

Date Published: March 2005

Division of Fuel Cycle Safety and Safeguards 

Office of Nuclear Material Safety and Safeguards 

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001
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iii

ABSTRACT

On February 28, 2001, Duke COGEMA Stone & Webster (DCS or the applicant), a limited

liability company, submitted a request to the U.S. Nuclear Regulatory Commission (NRC) to

construct a Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF or the facility) on the U.S.

Department of Energy’s (DOE) Savannah River Site near Aiken, South Carolina. DCS is a

DOE contractor. If its construction authorization request (CAR), as revised, is approved, DCS

is expected to then submit a license application for authority to possess and use licensed

material at the MFFF. Following an acceptance review of such an application, a notice of

opportunity for hearing would be published in the Federal Register. In support of its revised

CAR, DCS has submitted several items to the NRC, including a quality assurance plan and an

environmental report.

Under the applicable requirements of Title 10, Part 70, “Domestic Licensing of Special Nuclear

Material,” of the Code of Federal Regulations (10 CFR Part 70), before a possession and use

license may be issued, the NRC must first authorize construction of the facility. This final safety

evaluation report (FSER) documents the NRC staff’s review of the revised CAR and

supplemental supporting information provided by the applicant. This FSER pertains only to

approval of construction and does not address operational aspects of the MFFF.

If the MFFF is eventually authorized to operate, it would be a key asset of the DOE Surplus

Plutonium Disposition Program (SPDP), which is being implemented as a result of a bilateral

agreement with the Russian Federation. Pursuant to this agreement, the United States and the

Russian Federation would each convert 34 metric tons (37.5 tons) of weapons-grade plutonium

that has been declared excess to national security needs into forms less usable in nuclear

weapons. As part of the SPDP, surplus U.S. weapons-grade plutonium would be converted into

MOX fuel and irradiated in commercial reactors to produce electricity. Following irradiation, the

resulting spent fuel would contain plutonium in a form less usable in nuclear weapons.
iv

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v

TABLE OF CONTENTS

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

ABBREVIATIONS .......... .......... .......... .......... .......... ........ xii

LIST OF ACRONYMS FOR MFFF BUILDING AND SYSTEM DESIGNATIONS .......... xvii

INTRODUCTION .......... .......... .......... .......... .......... ....... I-1

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

1.1 Facility and Process Overview .......... .......... .......... ........ 1-1

1.1.1 Conduct of Review .......... .......... .......... .......... 1-1

1.1.2 Evaluation Findings .......... .......... .......... ......... 1-6

1.1.3 References .......... .......... .......... .......... ..... 1-6

1.2 Institutional Information .......... .......... .......... .......... ... 1-6

1.2.1 Conduct of Review .......... .......... .......... .......... 1-6

1.2.2 Evaluation Findings .......... .......... .......... ......... 1-8

1.2.3 References .......... .......... .......... .......... ..... 1-8

1.3 Site Description .......... .......... .......... .......... ......... 1-8

1.3.1 Conduct of Review .......... .......... .......... .......... 1-8

1.3.2 Evaluation Findings .......... .......... .......... ........ 1-27

1.3.3 References .......... .......... .......... .......... .... 1-27

2. FINANCIAL QUALIFICATIONS .......... .......... .......... .......... .... 2-1

2.1 Conduct of Review .......... .......... .......... .......... ....... 2-1

2.1.1 Project Costs .......... .......... .......... .......... .... 2-1

2.1.2 Financial Qualifications .......... .......... .......... ...... 2-1

2.1.3 Liability Insurance .......... .......... .......... .......... 2-2

2.2 Evaluation Findings .......... .......... .......... .......... ...... 2-2

2.3 References .......... .......... .......... .......... .......... ... 2-3

3. PROTECTION OF CLASSIFIED MATTER .......... .......... .......... ..... 3-1

3.1 Conduct of Review .......... .......... .......... .......... ....... 3-1

3.2 Evaluation Findings .......... .......... .......... .......... ...... 3-1

3.3 References .......... .......... .......... .......... .......... ... 3-2

4. ORGANIZATION AND ADMINISTRATION .......... .......... .......... ..... 4-1

4.1 Conduct of Review .......... .......... .......... .......... ....... 4-1

4.1.1 Organization .......... .......... .......... .......... ..... 4-1

4.1.2 Administration .......... .......... .......... .......... ... 4-1

4.1.3 Key Management Positions .......... .......... .......... ... 4-2

4.2 Evaluation Findings .......... .......... .......... .......... ...... 4-2

4.3 References .......... .......... .......... .......... .......... ... 4-2

5. SAFETY ASSESSMENT OF THE DESIGN BASES .......... .......... ......... 5-1

5.1 Conduct of Review .......... .......... .......... .......... ....... 5-1
vi

5.1.1 Plant Site Description Relating to Safety Assessment

of the Design Bases .......... .......... .......... ........ 5-2

5.1.2 Safety Assessment Team Description .......... .......... ..... 5-3

5.1.3 Chemical Standards and Consequences .......... .......... ... 5-3

5.1.4 10 CFR 70.61 Performance Requirements .......... .......... . 5-3

5.1.5 Safety Assessment of Design Basis Methodology .......... ...... 5-4

5.1.6 Safety Assessment Results .......... .......... .......... . . 5-11

5.1.7 Description of PSSCs .......... .......... .......... ....... 5-44

5.2 Evaluation Findings .......... .......... .......... .......... ..... 5-44

5.3 References .......... .......... .......... .......... .......... . . 5-45

6. NUCLEAR CRITICALITY SAFETY .......... .......... .......... .......... . 6-1

6.1 Conduct of Review .......... .......... .......... .......... ....... 6-1

6.1.1 Organization and Administration .......... .......... ......... 6-2

6.1.2 Management Measures .......... .......... .......... ...... 6-3

6.1.3 Technical Practices .......... .......... .......... ......... 6-4

6.1.4 Design Bases of the PSSCs .......... .......... .......... . 6-48

6.2 Evaluation Findings .......... .......... .......... .......... ..... 6-55

6.3 References .......... .......... .......... .......... .......... . . 6-55

7. FIRE PROTECTION .......... .......... .......... .......... .......... . . 7-1

7.1 Conduct of Review .......... .......... .......... .......... ....... 7-1

7.1.1 Organization and Conduct of Operations .......... .......... ... 7-2

7.1.2 Features and Systems .......... .......... .......... ....... 7-2

7.1.3 Manual Firefighting Capability .......... .......... .......... 7-13

7.1.4 Preliminary Fire Hazard Analysis .......... .......... ........ 7-14

7.1.5 Design Bases of the PSSCs .......... .......... .......... . 7-14

7.2 Evaluation Findings .......... .......... .......... .......... ..... 7-20

7.3 References .......... .......... .......... .......... .......... . . 7-20

8. CHEMICAL SAFETY .......... .......... .......... .......... .......... . . 8-1

8.1 Conduct of Review .......... .......... .......... .......... ....... 8-1

8.1.1 Background .......... .......... .......... .......... ..... 8-2

8.1.2 Areas of Review and Evaluation Findings .......... .......... . . 8-2

8.1.3 Chemical Process Safety Interfaces .......... .......... ..... 8-58

8.2 Evaluation Findings .......... .......... .......... .......... ..... 8-58

8.3 References .......... .......... .......... .......... .......... . . 8-59

9. RADIATION SAFETY .......... .......... .......... .......... .......... . . 9-1

9.1 Conduct of Review .......... .......... .......... .......... ....... 9-1

9.1.1 Radiation Safety Design Features .......... .......... ........ 9-2

9.1.2 Radiation Protection Program .......... .......... .......... 9-12

9.1.3 Radiation Safety Design Bases .......... .......... ......... 9-17

9.2 Evaluation Findings .......... .......... .......... .......... ..... 9-17

9.3 References .......... .......... .......... .......... .......... . . 9-17

10. ENVIRONMENTAL PROTECTION .......... .......... .......... ......... 10-1

10.1 Conduct of Review .......... .......... .......... .......... ..... 10-1
vii

10.1.1 Radiation Safety .......... .......... .......... .......... 10-2

10.1.2 Effluent and Environmental Monitoring .......... .......... ... 10-5

10.1.3 Safety Assessment of the Design Bases .......... .......... . 10-7

10.2 Evaluation Findings .......... .......... .......... .......... ... 10-10

10.3 References .......... .......... .......... .......... .......... 10-10

11. PLANT SYSTEMS .......... .......... .......... .......... .......... . . 11-1

11.1 Civil Structural Systems .......... .......... .......... .......... . 11-1

11.1.1 Conduct of Review .......... .......... .......... ........ 11-1

11.1.2 Evaluation Findings .......... .......... .......... ...... 11-13

11.1.3 References .......... .......... .......... .......... . . 11-13

11.2 Aqueous Polishing Process and Chemistry .......... .......... ..... 11-15

11.2.1 Conduct of Review .......... .......... .......... ....... 11-15

11.2.2 Evaluation Findings .......... .......... .......... ...... 11-55

11.2.3 References .......... .......... .......... .......... . . 11-55

11.3 Mixed Oxide Process System Description .......... .......... ...... 11-59

11.3.1 Conduct of Review .......... .......... .......... ....... 11-59

11.3.2 Evaluation Findings .......... .......... .......... ...... 11-75

11.3.3 References .......... .......... .......... .......... . . 11-76

11.4 Ventilation and Confinement Systems .......... .......... ......... 11-78

11.4.1 Conduct of Review .......... .......... .......... ....... 11-78

11.4.2 Evaluation Findings .......... .......... .......... ...... 11-94

11.4.3 References .......... .......... .......... .......... . . 11-94

11.5 Electrical Systems .......... .......... .......... .......... .... 11-96

11.5.1 Conduct of Review .......... .......... .......... ....... 11-96

11.5.2 Evaluation Findings .......... .......... .......... ..... 11-104

11.5.3 References .......... .......... .......... .......... . 11-105

11.6 Instrumentation and Control Systems .......... .......... ......... 11-107

11.6.1 Conduct of Review .......... .......... .......... ...... 11-107

11.6.2 Evaluation Findings .......... .......... .......... ..... 11-116

11.6.3 References .......... .......... .......... .......... . 11-117

11.7 Material Transport Systems .......... .......... .......... ...... 11-120

11.7.1 Conduct of Review .......... .......... .......... ...... 11-120

11.7.2 Evaluation Findings .......... .......... .......... ..... 11-131

11.7.3 References .......... .......... .......... .......... . 11-131

11.8 Fluid Transport Systems .......... .......... .......... ......... 11-133

11.8.1 Conduct of Review .......... .......... .......... ...... 11-133

11.8.2 Evaluation Findings .......... .......... .......... ..... 11-149

11.8.3 References .......... .......... .......... .......... . 11-149

11.9 Fluid Systems (Bulk Materials, Reagents, and Gases) .......... ...... 11-151

11.9.1 Conduct of Review .......... .......... .......... ...... 11-151

11.9.2 Evaluation Findings .......... .......... .......... ..... 11-164

11.9.3 References .......... .......... .......... .......... . 11-165

11.10 Heavy Lift Cranes .......... .......... .......... .......... ... 11-166

11.10.1 Conduct of Review .......... .......... .......... ..... 11-166

11.10.2 Evaluation Findings .......... .......... .......... .... 11-168

11.10.3 References .......... .......... .......... .......... . 11-168

11.11 Environmental Qualification .......... .......... .......... ..... 11-169
viii

11.11.1 Conduct of Review .......... .......... .......... ..... 11-169

11.11.2 Evaluation Findings .......... .......... .......... .... 11-170

11.11.3 References .......... .......... .......... .......... . 11-171

11.12 Seismic Qualification of Equipment, Systems, and Components ....... 11-171

11.12.1 Conduct of Review .......... .......... .......... ..... 11-171

11.12.2 Evaluation Findings .......... .......... .......... .... 11-173

11.12.3 References .......... .......... .......... .......... . 11-173

12. HUMAN FACTORS ENGINEERING FOR PERSONNEL ACTIVITIES .......... ... 12-1

12.1 Conduct of Review .......... .......... .......... .......... ..... 12-1

12.1.1 Identification of Personnel Actions .......... .......... ...... 12-1

12.1.2 Human Factors Engineering Design Planning .......... ....... 12-4

12.1.3 Operating Experience .......... .......... .......... ..... 12-5

12.1.4 Function and Task Analysis .......... .......... .......... . 12-5

12.1.5 Human-System Interface Design, Inventory, and Characterization . 12-6

12.1.6 Other Considerations .......... .......... .......... ..... 12-6

12.1.7 Design Bases of the PSSCs .......... .......... .......... . 12-6

12.2 Evaluation Findings .......... .......... .......... .......... .... 12-6

12.3 References .......... .......... .......... .......... .......... . 12-7

13. SAFEGUARDS .......... .......... .......... .......... .......... .... 13-1

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

13.1.1 Conduct of Review .......... .......... .......... ........ 13-1

13.1.2 Evaluation Findings .......... .......... .......... ....... 13-1

13.1.3 References .......... .......... .......... .......... ... 13-1

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

13.2.1 Conduct of Review .......... .......... .......... ........ 13-2

13.2.2 Evaluation Findings .......... .......... .......... ....... 13-3

13.2.3 References .......... .......... .......... .......... ... 13-4

14. EMERGENCY MANAGEMENT .......... .......... .......... .......... . . 14-1

14.1 Conduct of Review .......... .......... .......... .......... ..... 14-1

14.2 Evaluation Findings .......... .......... .......... .......... .... 14-3

14.3 References .......... .......... .......... .......... .......... . 14-3

15. MANAGEMENT MEASURES .......... .......... .......... .......... .... 15-1

15.1 Quality Assurance .......... .......... .......... .......... ..... 15-2

15.1.1 Conduct of Review .......... .......... .......... ........ 15-2

15.1.2 Evaluation Findings .......... .......... .......... ....... 15-3

15.1.3 References .......... .......... .......... .......... ... 15-4

15.2 Configuration Management .......... .......... .......... ........ 15-4

15.2.1 Conduct of Review .......... .......... .......... ........ 15-4

15.2.2 Evaluation Findings .......... .......... .......... ....... 15-7

15.2.3 References .......... .......... .......... .......... ... 15-8

15.3 Maintenance .......... .......... .......... .......... .......... 15-8

15.3.1 Conduct of Review .......... .......... .......... ........ 15-8

15.3.2 Evaluation Findings .......... .......... .......... ...... 15-10

15.3.3 References .......... .......... .......... .......... . . 15-10
ix

15.4 Training and Qualification .......... .......... .......... ......... 15-11

15.4.1 Conduct of Review .......... .......... .......... ....... 15-11

15.4.2 Evaluations Findings .......... .......... .......... ..... 15-14

15.4.3 References .......... .......... .......... .......... . . 15-14

15.5 Plant Procedures .......... .......... .......... .......... ..... 15-15

15.5.1 Conduct of Review .......... .......... .......... ....... 15-15

15.5.2 Evaluation Findings .......... .......... .......... ...... 15-16

15.5.3 References .......... .......... .......... .......... . . 15-16

15.6 Audits and Assessments .......... .......... .......... ......... 15-17

15.6.1 Conduct of Review .......... .......... .......... ....... 15-17

15.6.2 Evaluation Findings .......... .......... .......... ...... 15-19

15.6.3 References .......... .......... .......... .......... . . 15-20

15.7 Incident Investigations .......... .......... .......... .......... . 15-20

15.7.1 Conduct of Review .......... .......... .......... ....... 15-20

15.7.2 Evaluation Findings .......... .......... .......... ...... 15-21

15.7.3 References .......... .......... .......... .......... . . 15-22

15.8 Records Management .......... .......... .......... .......... . 15-22

15.8.1 Conduct of Review .......... .......... .......... ....... 15-22

15.8.2 Evaluation Findings .......... .......... .......... ...... 15-24

15.8.3 References .......... .......... .......... .......... . . 15-24

FIGURES

8.1-1 AP process overview .......... .......... .......... .......... ......... 8-5

8.1-2 Empirical venting relationship .......... .......... .......... .......... . 8-45

11.2-1 Acid Normality and Other Effects Upon Electrolytic Gas Generation .......... . 11-39

11.3-1 Overview of MP .......... .......... .......... .......... .......... . 11-62

11.4-1 Ventilation Confinement Zones .......... .......... .......... ......... 11-79

TABLES

5.1-1 Safety Assessment Summary (Criticality, Natural Phenomena Hazards, External

Manmade Hazards, and Chemical Hazards) .......... .......... .......... 5-47

5.1-2 Safety Assessment Summary (Process Hazards) .......... .......... ...... 5-52

5.1-3 PSSCs and Design Basis Functions and Values Developed from

the Safety Assessment .......... .......... .......... .......... ...... 5-82

6.1-1 Definition of AOA(1) .......... .......... .......... .......... ........ 6-17

6.1-2 Definition of AOA(2) .......... .......... .......... .......... ........ 6-20

6.1-3 Definition of AOA(3) .......... .......... .......... .......... ........ 6-21
x

6.1-4 Definition of AOA(4) .......... .......... .......... .......... ........ 6-29

6.1-5 Definition of AOA(4) .......... .......... .......... .......... ........ 6-33

6.1-6 Definition of AOA(5) .......... .......... .......... .......... ........ 6-35

8.1-1 Summary of Chemicals at the Proposed Facility .......... .......... ........ 8-9

8.1-2 Reaction Products of the AP Process (Normal Operations) .......... ......... 8-12

8.1-3 Potentially Hazardous Chemicals Produced in

AP under Off-Normal Conditions .......... .......... .......... ......... 8-14

8.1-4 Process Chemical Hazardous Characteristics and Incompatibilities

Revised CAR (DCS, 2002e) .......... .......... .......... .......... . . 8-15

8.1-5 Chemical Consequence Limits Proposed by the Applicant for the MFFF

 (mg/m3) .......... .......... .......... .......... .......... ....... 8-20

8.1-6 Applicant’s Use of Chemical Limits for Qualitatively

Determining Chemical Consequence Categories .......... .......... ...... 8-22

8.1-7 Applicant’s Commitments to Chemical Limits for Qualitative Chemical

Consequences (DCS, 2003d, 2003e; NRC, 2003b, 2003e) .......... ......... 8-22

8.1-8 Combustible Characteristics of Chemicals in the AP Area .......... .......... 8-30

8.1-9 PSSCs for Process Units with HAN and without Nitrous Oxide Addition ......... 8-36

8.1-10 Design Basis Values for the Prevention of HAN Explosions .......... ........ 8-36

8.1-11 PSSCs for Process Units with HAN and Nitrous Oxide Addition .......... ..... 8-37

8.1-12 Chemical Limits Proposed by the Applicant for Control Room

Consequence Calculations at the MFFF (mg/m3) .......... .......... ...... 8-56

9.1-1 Mass Fractions of Isotopes in Aged Weapons-Grade Plutonium .......... ...... 9-4

9.1-2 Mass Fractions of Isotopes in Each Source Category

for the Safety Assessment .......... .......... .......... .......... ..... 9-5

9.1-3 Photon Energy Spectrum for MFFF MOX .......... .......... .......... ... 9-6

9.1-4 Neutron Energy Distribution for MFFF MOX .......... .......... .......... . 9-7

9.1-5 ARFs and RFs .......... .......... .......... .......... .......... ... 9-10

9.1-6 Radiation Protection PSSCs .......... .......... .......... .......... . . 9-15
xi

9.1-7 Comparison of the Radiotoxicity of Reactor-Grade Plutonium and

Weapons-Grade Plutonium; TEDE Per Gram of Plutonium Inhaled .......... . . 9-16

10.1-1 Potential for Intermediate-Consequence Events Involving Liquid

Waste Streams .......... .......... .......... .......... .......... . . 10-5

10.1-2 Air Effluent Concentrations from the MFFF .......... .......... .......... . 10-8

11.2-1 Waste Stream Descriptions and Quantities in the Waste Reception Unit ....... 11-31

11.2-2 Design Basis for Chemical Impurities in PDCF Feed Plutonium Dioxide ........ 11-51

11.2-3 Design Basis for Chemical Impurities of

AFS Plutonium Dioxide Feed Material .......... .......... .......... .... 11-52

11.2-4 Radionuclide Impurities in the Feed Plutonium Dioxide .......... .......... . 11-54

11.3-1 Standard Burnback Reaction Enthalpies .......... .......... .......... . . 11-64

11.3-2 Applicant’s Design Basis Temperature Criteria .......... .......... ....... 11-69

11.3-3 Applicant’s Additional Temperature Design Basis Criteria

for Personnel Protection .......... .......... .......... .......... .... 11-69

11.4-1 Confinement System Design Bases .......... .......... .......... ..... 11-84

11.7-1 PSSCs for Loss of Confinement/Dispersal

of Nuclear Materials Events .......... .......... .......... .......... . 11-123

11.7-2 PSSCs for Load-Handling Events .......... .......... .......... ...... 11-125

11.8-1 Categorization of the Fluid Transport System .......... .......... ....... 11-140

11.8-2 Design Basis Codes and Standards for Category 1 Fluid

Transport System Components .......... .......... .......... ........ 11-140

11.8-3 Design Basis Codes and Standards for Category 2 

Fluid Transport System Components .......... .......... .......... ... 11-141

11.8-4 Design Basis Codes and Standards for Category 3

Fluid Transport System Components .......... .......... .......... ... 11-141

11.8-5 Design Basis Codes and Standards for Category 4 Fluid 

Transport System Components .......... .......... .......... ........ 11-142

11.8-6 Design Bases for the Fluid Transport System .......... .......... ....... 11-143
xii

ABBREVIATIONS

ac alternating current

ACI American Concrete Institute

AEC active engineered control

AEGL acute exposure guideline level

AFS alternate feedstock

AHJ authorities having jurisdiction

AISC American Institute of Steel Construction

ALARA as low as reasonably achievable

ALI allowable limit on intake

ALOHA areal locations of hazardous atmospheres

A-MIMAS advanced micronized master blend

ANS American Nuclear Society

ANSI American National Standards Institute 

AOA area of applicability

AP aqueous polishing

ARF atmospheric release fraction

ASHRAE American Society of Heating, Refrigeration, and Air-Conditioning Engineers

ASME American Society of Mechanical Engineers

ASTM American Society for Testing and Materials

AWS American Welding Society

BA Bachelor of Arts

BDC baseline design criterion/criteria

BMF fuel fabrication building

BP&V Boiler and Pressure Vessel Code (ASME)

B.R. breathing rate

BS Bachelor of Science

CAAS criticality accident alarm system

CAM continuous air monitor

CAR construction authorization request

CCU criticality control unit

CEDE committed effective dose equivalent

CFM cubic feet per minute

CFR Code of Federal Regulations

CGA Compressed Gas Association

CM configuration management

CRT cargo restraint transporter

CS conventional seismic

CSE criticality safety evaluation

CWTS clean water treatment system

dc direct current

DCF dose conversion factor

DCP double-contingency principle

DCS Duke Cogema Stone & Webster
xiii

DOE U.S. Department of Energy

DPSG Duke Project Services Group, Inc.

DR damage ratio

DSER draft safety evaluation report

DU depleted uranium

DUO2 depleted uranium dioxide 

EALF energy of average lethargy causing fission

ECR emergency control room

ECRACS emergency control room air conditioning system

EDG emergency diesel generator

EDGFOS emergency diesel generator fuel oil system

EDMS electronic data management system

EIS environmental impact statement

EMMH external manmade hazard

ENDF evaluated nuclear data file

EPA Environmental Protection Agency

EPRI Electric Power Research Institute

EQ environmental qualification

ER environmental report

ERDA U.S. Energy Research and Development Administration

ERPG emergency response planning guidelines

ETF effluent treatment facility

FHA fire hazards analysis

FM Factory Mutual

FNMC fundamental nuclear material control

FOCI foreign ownership, control, or influence

FSER final safety evaluation report

FTS fluid transport system

HA hazards analysis

HAAW high alpha activity waste stream (containing uranium, americium, and other

radioactive decay products)

HAN hydroxylamine nitrate

HAZOP hazard and operability (analysis)

HD high depressurization

HEPA high-efficiency particulate air

HEU high-enriched uranium

HFE human factors engineering

HPT hydrogenated propylene tetramer

HSI human-system interface

HVAC heating, ventilation, and air conditioning

I&C instrumentation and control

ICN immediate control network

ICRP International Commission on Radiation Protection

ICSBEP International Criticality Safety Benchmark Experiments
xiv

IDLH immediate danger to life and health

IEC International Electrotechnical Commission

IEEE Institute of Electrical and Electronics Engineers

IOC individual (or individuals) outside the controlled area

IROFS items relied on for safety

ISA integrated safety analysis

JSHU jar storage and handling unit

LANL Los Alamos National Laboratory

LFL lower flammability limit

LIN local industrial network

LLNL Lawrence Livermore National Laboratory

LLW low-level waste

LPF leak path factor

LWR light-water reactor

MAPE mean annual probability of exceedance

MAR material at risk

MC&A material control and accounting

MCC motor control center

MCNP Monte Carlo Neutron Photon

MDE medium depressurization exhaust

MFFF mixed oxide fuel fabrication facility

MFFP mixed oxide fresh fuel package

MIMAS micronized master blend

MMI Mercalli intensity

MMIS manufacturing management and information system

MOU memorandum of understanding

MOX mixed oxide 

MP mixed oxide process

MPQAP MOX Project Quality Assurance Plan

MSDS material safety data sheet

MTHM metric tons heavy metal

NCS nuclear criticality control

NCSE nuclear criticality safety evaluation

NFPA National Fire Protection Association

NNSA National Nuclear Security Administration

NRC U.S. Nuclear Regulatory Commission

NPH natural phenomena hazards

ORNL Oak Ridge National Laboratory

OSHA Occupational Safety and Health Administration

PAA preliminary accident analysis

PAG protective action guide

PBX public branch exchange
xv

PC performance categories

PDCF pit disassembly and conversion facility

PEC passive engineered control

PEP personnel and equipment protection

PFOD probability of failure on demand

PHA preliminary hazard analysis

PIP plutonium immobilization pit

PLC programmable logic controller

PSCS process safety control system

PSHA probabilistic seismic hazard analysis

PSSC principal structure, system, and component

PTFE polytetrafluorethylene

Pu plutonium

PuO2 plutonium dioxide

PUREX plutonium uranium reduction extraction

QA quality assurance

QL quality level

RACB restricted area boundary concentration

RAI request for additional information

RF respirable fraction

RG Regulatory Guide

SA safety analysis

SAR safety analysis report

SC seismic category

SCALE Standardized Computer Analyses Licensing Evaluation

SER safety evaluation report

SFPE Society of Fire Protection Engineers

SNM special nuclear material

SPDP Surplus Plutonium Disposition Program 

SQ seismic qualification

SRP Standard Review Plan (NUREG-1718)

SRS Savannah River Site

SSC structure, system, and component

SST safe secure transport

STEL short-term exposure limit

S/U sensitive/uncertainty

TBP tributyl phosphate

TEDE total effective dose equivalent

TEEL temporary emergency exposure limit

TLV threshold limit value

TQ threshold quantities

TRU transuranic

UBC Uniform Building Code
xvi

UL Underwriters Laboratories

UO2 uranium dioxide

USL upper subcritical limit

UPS uninterruptible power supply

WAC waste acceptance criteria

WG water gauge

WTA work task agreement

XTN X-terminal network
xvii

LIST OF ACRONYMS FOR MFFF BUILDING AND SYSTEM DESIGNATIONS

Buildings Systems

BAD administration building

BAP aqueous polishing

area

BEG emergency diesel

generator bldg

BMF MOX fuel fabrication

bldg 

BMP MOX fuel fabrication

area (MOX processing

area)

BRP reagent processing

building

BSG standby diesel

generator bldg

BSH safe haven buildings

BSR shipping and receiving

area

BSW secured warehouse

building

BTS technical support

building

BAS breathing air system

CHH HVAC chilled water system

CHP process chilled water system

DCE PuO2 buffer storage unit

DCM PuO2 3013 storage unit

DCP PuO2 receiving unit

DCS decontamination system

DDP U02 drum emptying unit

DMW demineralized water system

DRS U02 receiving and storage unit

EGF emergency diesel generator

fuel oil system

GAH argon/hydrogen system

GDE rod decladding unit

GHE helium system

GME, GMF rod cladding &

decontamination units

GMK rod tray loading unit

GNO nitrogen oxide system

GNS nitrogen system

GOX oxygen system

HDE high depressurization exhaust

system

HWS process hot water system

lAS instrument air system

KCA oxalic precipitation & oxidation

unit

KCB homogenization unit

KCC canning unit

KCD oxalic mother liquor recovery

unit

KDA decanning unit

KDB dissolution unit

KDD dechlorination and dissolution unit

KDM milling unit

KDR recanning unit

KPA purification cycle

KPB solvent recovery cycle

KPC acid recovery unit

KPF silver recovery unit

KWD liquid waste reception unit

KWG offgas treatment unit

KWS waste organic solvent unit

MDE medium depressurization exhaust

system

NBX primary blend ball milling unit

NBY scrap milling unit

NCR scrap processing unit

NDD PuO2 container opening & handling unit

NDP primary dosing unit

NDS final dosing unit

NPE, NPF homogenization & pelletizing unit

NTM jar storage & handling unit

NXR powder auxiliary unit

PAD pellet repackaging unit

PAR scrap box loading unit

PFE, PFF sintering units

PML pellet handling unit

POE process cell exhaust system

PQE quality control and manual sorting units

PRE, PRF grinding units

PSE green pellet storage unit

PSF sintered pellet storage unit

PSI scrap pellet storage unit

PSJ ground and sorted pellet storage unit

PTE pellet inspection and sorting units

PWS plant water system

RDO diluent system 

RHN hydroxylamine nitrate system

RHP hydrogen peroxide system

RHZ hydrazine system

RMN manganese nitrate system

RNA nitric acid system

ROA oxalic acid system

RSC sodium carbonate system

RSH sodium hydroxide system

RSN silver nitrate system

RTP tributyl phosphate system

SAS service air system

SCE rod scanning unit

SDK rod inspection and sorting unit

SEK helium leak test unit

SGF standby diesel generator fuel oil

system

SPS, SPC process steam and process

condensate systems

STK rod storage unit

SXE, SXF x-ray inspection units

TAS assembly handling and storage unit

TCK assembly dry cleaning unit

TCL assembly final inspection unit

TCP assembly dimensional inspection unit

TGM assembly mockup loading unit

TGV assembling mounting unit

TXE assembly packaging unit

VHD very high depressurization exhaust

system

VRM radiation monitoring vacuum system

WSB waste solidification building

WVA vehicle access portal
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1Earlier, on December 19, 2000, DCS submitted an environmental report (ER). The staff’s review of the ER (as

revised July 12, 2001, October 29, 2002, December 10, 2002, January 15, 2003, June 20, 2003, August 13, 2003,

and June 10, 2004), is reflected in NUREG-1767, “Environmental Impact Statement on the Construction and

Operation of a Mixed Oxide Fuel Fabrication Facility at the Savannah River Site, South Carolina,” published in

January 2005.

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INTRODUCTION

On February 28, 2001, Duke Cogema Stone & Webster (DCS or the applicant) submitted a

construction authorization request (CAR) to the U.S. Nuclear Regulatory Commission (NRC)

seeking authority to construct a mixed oxide (MOX) fuel fabrication facility (MFFF or the facility)

on the U.S. Department of Energy’s (DOE) Savannah River Site in South Carolina.1 DCS is

under contract to the National Nuclear Security Administration (NNSA), a semi-autonomous

agency within DOE charged with designing, building, and operating the proposed MFFF. 

Building and operating the MFFF would implement a September 2000 agreement between the

United States and the Russian Federation, in which each nation agreed to the disposition of 34

metric tons (37.5 tons) of surplus weapons-grade plutonium. The MOX fuel manufactured at

the proposed MFFF would be irradiated in U.S. commercial nuclear power plants, thus

converting the surplus plutonium to a form that cannot be readily used to make nuclear

weapons.

The NNSA will own the MFFF. In connection with its operation, the NNSA plans to construct

nearby a pit disassembly and conversion facility (PDCF). The PDCF, which would not be under

the NRC’s jurisdiction, would be used to disassemble nuclear weapon components and convert

them to an unclassified plutonium dioxide form. The plutonium dioxide from the PDCF would

be the largest source of plutonium feedstock for the MFFF.

In 2001, NNSA was planning to build and operate a third surplus plutonium disposition facility, 

the Plutonium Immobilization Plant (PIP). As proposed, the PIP would have immobilized

approximately 8.5 metric tons (9.4 tons) of surplus plutonium into a waste form suitable for

direct disposal in a high-level waste repository, leaving 25.5 metric tons (28.1 tons) of surplus

plutonium for conversion into MOX fuel at the MFFF. But in February 2002, NNSA announced

that it was cancelling the PlP. As a result, NNSA directed DCS to redesign the MFFF to enable

it to convert an additional 8.5 metric tons (9.4 tons) that was no longer slated for conversion at

the PIP. This material, referred to as “alternate feedstock,” represents the balance of the 34

metric tons (37.5 tons) of surplus plutonium that will not come from the proposed PDCF.

On October 31, 2002, DCS submitted a revised CAR, which replaced, in its entirety, the original

CAR. The revised CAR reflects the changes necessary to process plutonium feed materials

from DOE sources other than the PDCF, incorporates information previously provided in the

DCS response to the NRC staff requests for additional information, and provides additional

information addressing open items identified in the NRC draft safety evaluation report (DSER)

dated April 30, 2002. DCS made further changes to the revised CAR, as reflected in CAR

change pages dated December 12, 2002, February 18, 2003, April 1, 2003, and April 8, 2003.

The NRC staff issued another DSER on April 30, 2003, leading to the CAR change pages

dated July 28, 2003. Further CAR revisions were made necessary by the November 2003
I-2

change to the MFFF controlled area boundary, as reflected in the CAR change pages dated

June 10, 2004. Additional CAR revisions were made on January 27, 2005, and February 9,

2005, which were needed to incorporate commitments made previously by DCS in letters to the

NRC that were the bases for closing open items in the DSER. This final safety evaluation

report (FSER) contains the staff’s findings and conclusions on the CAR, revised as described

above. 

MFFF Description

The MFFF would consist of three major functional areas, (1) shipping and receiving, (2)

aqueous polishing where the plutonium dioxide would be purified, and (3) MOX production

where the fuel pellets, fuel rods, and fuel assemblies would be fabricated. The facility would

receive depleted uranium dioxide and plutonium dioxide, purify the plutonium dioxide to remove

impurities such as gallium and americium, fabricate MOX fuel containing uranium and

plutonium dioxides, assemble fuel rods, and fabricate fuel assemblies. The completed fuel

assemblies would be subsequently irradiated in commercial nuclear power plants authorized by

the NRC to use MOX fuel. 

Applicable NRC Regulations

Under the applicable requirements of Title 10, Part 70, “Domestic Licensing of Special Nuclear

Material,” of the Code of Federal Regulations (10 CFR Part 70), before a license to possess

and use licensed material at the MFFF may be issued, the NRC must first authorize

construction of the facility. This FSER documents the NRC staff’s review of the revised CAR

and supplemental supporting information provided by the applicant. This FSER only addresses

regulatory requirements for approval of construction and does not address operational aspects

of the proposed facility. 

The following discussion summarizes the regulatory requirements applicable to the NRC staff’s

review of the revised CAR. In this regard, 10 CFR 70.23(b) states that the NRC will approve

construction of a plutonium processing and fuel fabrication facility if it finds that the design

bases of the principal structures, systems, and components (PSSCs) and the Quality

Assurance (QA) Program, provide reasonable assurance of protection against natural

phenomena and the consequences of potential accidents. The FSER discusses the applicant’s

selection of design basis functions and values and how the applicant determined that the

design bases will provide reasonable assurance of protection against natural phenomena and

the consequences of potential accidents. Each FSER chapter contains one or more of the

NRC staff’s 10 CFR 70.23(b) safety findings on the applicable PSSCs and design bases being

evaluated.

Regarding the term “design bases,” the NRC stated in a letter to DCS dated October 26, 1999,

that the following definition of design basis, which appears in 10 CFR 50.2, “Definitions,” will be

applied to the proposed facility: 

Design basis means that information which identifies the specific functions to be

performed by a structure, system, or component of a facility, and the specific

values or range of values chosen for controlling parameters as reference bounds

for design. These values may be (1) restraints derived from generally accepted
I-3

“state of the art” practices for achieving functional goals, or (2) requirements

derived from analysis (based on calculation and/or experiments) of the effects of

a postulated accident for which a structure, system, or component must meet its

functional goals. 

Other 10 CFR Part 70 requirements applicable to authorizing construction of the facility are

discussed below. Baseline design criteria (BDC), listed in 10 CFR 70.64(a), cover 10 design

issues, (1) quality standards and records, (2) natural phenomena hazards, (3) fire protection,

(4) internal environmental conditions and dynamic effects, (5) chemical protection, (6)

emergency capability, (7) utility services, (8) inspection, testing, and maintenance, (9) criticality

control, and (10) instrumentation and controls. The staff has evaluated the MFFF preliminary

design against these BDC, and the FSER describes how these BDC requirements are satisfied

by the facility’s design bases. The BDC specify design features that are required and

acceptable under the conditions specified in 10 CFR 70.64(a). 

Under 10 CFR 70.64(b), the facility design and layout must be based on defense-in-depth

practices. As used in 10 CFR 70.64, “Requirements for New Facilities or Processes at Existing

Facilities,” defense-in-depth practices at new facilities mean a design philosophy, applied from

the outset and through completion of the design, that is based on providing successive levels of

protection such that health and safety will not be wholly dependent upon any single element of

the facility design. The net effect of incorporating defense-in-depth practices is a conservatively

designed facility that will exhibit greater tolerance to failures and external challenges. Chapter

5 of the FSER documents the NRC staff’s findings that the facility’s preliminary design is

properly based on defense-in-depth practices. 

As further required by 10 CFR 70.64(b), to the extent practicable, the facility design must

incorporate (1) a preference for engineered controls over administrative controls to increase

overall system reliability, and (2) features that will enhance safety by reducing challenges to

items which will be relied upon for safety. Chapter 5 of the FSER documents the NRC staff’s

findings that the facility’s preliminary design adequately incorporates a preference for

engineered controls over administrative controls and that its design features will adequately

enhance safety.

As discussed below, the set of requirements contained in 10 CFR 70.61(b–d), and (f) are also

applicable to authorizing construction of the facility. Pursuant to the 10 CFR 70.61(b–d)

performance requirements, the risk of accidents at the proposed facility must be limited. To

properly determine the nature of such risks, and whether the proposed facility design effectively

controls such risks, requires an evaluation of the safety assessment of the design bases

submitted by DCS for construction approval. In FSER Chapter 5, the staff reviews the

methodology used by DCS in performing the safety assessment of the facility design bases,

and finds that the safety assessment adequately considered all appropriate natural phenomena,

external manmade events, and internal process hazards. Identifying, and later implementing,

the required controls to meet the 10 CFR 70.61(b–d) performance requirements ensures that

an acceptable level of risk will be maintained during any facility operation. DCS may satisfy

these performance requirements through a combination of limiting the chance that accidents at

the facility would occur (prevention) and reducing the consequences of such events (mitigation). 

In approving the revised CAR, the staff finds that the DCS safety assessment describes an
I-4

adequate strategy which, if effectively applied, will ensure that the 10 CFR 70.61 performance

requirements will be met, should the facility later be authorized to operate. 

The performance requirements for “credible high-consequence event[s]” (i.e., potential

accidents involving high levels of radiation or hazardous chemicals) are set forth in 10 CFR

70.61(b). These provisions pertain to the protection of both workers and individuals outside the

controlled area. In particular, 10 CFR 70.61(b) requires the use of controls sufficient to either

make the occurrence of such accidents “highly unlikely,” or make the consequences of such

accidents less severe than (1) an acute 1 Sv (100 rem) total effective dose equivalent (TEDE)

to a “worker,” as defined in 10 CFR 70.4, (2) an acute 0.25 Sv (25 rem) TEDE to an individual

outside the “controlled area,” as defined in 10 CFR 20.1003, which DCS is required to

designate for the proposed facility pursuant to 10 CFR 70.61(f), (3) an intake of 30 mg (6.6 ×

10-5 lb) of soluble uranium to an individual outside the controlled area, or (4) either an acute

chemical exposure that could endanger the life of a worker or an acute chemical exposure that

could lead to irreversible or other serious long-lasting health effects to an individual outside the

controlled area. (See 10 CFR 70.61(b)(1–4)). 

The performance requirements for “credible intermediate-consequence event[s]” (i.e., potential

accidents involving lower levels of radiation or hazardous chemicals than referenced in 10 CFR

70.61(b)), are set forth in 10 CFR 70.61(c). These provisions pertain to environmental

protection, as well as to protecting the health of workers and individuals outside the controlled

area. In particular, 10 CFR 70.61(c) requires the use of controls sufficient to either make the

occurrence of such accidents “unlikely,” or make the consequences of such accidents less

severe than (1) an acute 0.25 Sv (25 rem) TEDE to a worker, (2) an acute 0.05 Sv (5 rem)

TEDE to an individual outside the controlled area, (3) 24-hr average release of radioactive

material outside the “restricted area,” as defined in 10 CFR 20.1003, “Definitions, into the

environment in concentrations exceeding 5000 times the values in Table 2 of Appendix B to 10

CFR Part 20, “Standards for Protection Against Radiation,” or (4) either a chemical exposure

that could lead to irreversible or other serious long-lasting health effects to a worker or a

chemical exposure that could cause mild transient health effects to an individual outside the

controlled area. (See 10 CFR 70.61(c)(1–4)). 

The performance requirements applicable to nuclear criticality accidents, which are discussed

in FSER Chapter 6, are set forth in 10 CFR 70.61(d). Under 10 CFR 70.61(d), the risk of such

accidents is limited by requiring the use of an approved margin of subcriticality for safety. 

Additionally, primary reliance must be placed on the use of preventative, rather than mitigative,

controls and measures to limit the risk of nuclear criticality accidents. 

Two sections of 10 CFR 70.61 are not yet applicable to DCS, as they reference 10 CFR 70.62

safety program requirements which apply to the evaluation of applications for 10 CFR Part 70

licenses to possess and use licensed material. The integrated safety analysis (ISA)

requirements referenced in 10 CFR 70.61(a) are set forth in 10 CFR 70.62(c); under these

provisions, DCS must later show that it would be in compliance with the above-described

performance requirements during any facility operations. Similarly, 10 CFR 70.61(e) also

references the 10 CFR 70.62 safety program and requires that each control identified by DCS

to meet the performance requirements be designated as an item relied on for safety (IROFS). 

Each IROFS must be available and reliable to perform its intended function when needed. 

Pursuant to 10 CFR 70.65, “Additional Content of Applications,” DCS must list all IROFS in its
I-5

ISA summary, which is submitted as part of its application for a 10 CFR Part 70 license to

possess and use licensed material.

In Section 5.5.5 of the CAR, the applicant described its general design philosophy and the

defense-in-depth practices used in formulating the preliminary design of the facility. Pursuant

to 10 CFR 70.64(b), in order to ensure that engineered controls are relied upon over

administrative controls, to the extent practicable, DCS has established a hierarchy of controls. 

In further adherence to 10 CFR 70.64(b), DCS stated that it has incorporated defense-in-depth

practices in its preliminary facility design. In Section 5.5.5.2 of the CAR, DCS stated that, it has

incorporated defense-in-depth practices through use of the double-contingency principle for

protection against criticality events and the use of the single-failure criterion. Under the latter

criterion, DCS stated that for the PSSCs it has identified, each is required to be capable of

carrying out its function in the event any single active component fails, whether such failure

occurs within the applicable system or in an associated system that supports the component’s

operation. In Section 5.5.5.4 of the CAR, DCS stated that the BDC are incorporated into the

facility design. The NRC staff evaluates these DCS statements in the FSER. 

Potential accidents evaluated by the applicant include loss of confinement of licensed nuclear

material, fire, load-handling events, explosions, nuclear criticality, natural phenomena events,

external manmade events, external exposure, and events related to chemical interactions. The

set of natural phenomena hazards identified by the applicant and evaluated by the staff include

earthquakes, high wind, tornadoes and tornado-generated missiles, extreme temperatures,

rain, snow, ice, lightning, and fires external to the MFFF. For most of the postulated hazards,

the applicant has chosen a mitigation strategy, but for the explosion and nuclear criticality

hazards, the applicant has chosen a prevention strategy. In FSER Chapter 5, the staff

reviewed (1) the methodology used by the applicant in performing the safety assessment of the

facility design bases, (2) the applicant’s hazard assessments, (3) the applicant’s formulation of

a safety strategy, and (4) the applicant’s identification of PSSCs to meet the 10 CFR 70.61

performance requirements. The staff finds that the applicant’s safety assessment adequately

considered all appropriate natural phenomena, external manmade events, and internal process

hazards. 

In the revised CAR, the applicant evaluated PSSCs primarily at the systems level, and the

identified PSSCs are primarily design features/administrative controls that are to be

implemented in the final design as IROFS, pursuant to 10 CFR 70.61(e). NUREG-1718,

“Standard Review Plan for the Review of an Application for a Mixed Oxide (MOX) Fuel

Fabrication Facility,” defines PSSCs as “safety controls that are identified in the design bases

as providing protection against the consequences of accidents or natural phenomena.” 

NUREG-1718 further states that designating a control as a PSSC “is effectively synonymous

with designating that control as an IROFS.” The staff used NUREG-1718 as guidance in

performing its review of the CAR and its supporting information. The definition of IROFS in 10

CFR 70.4, “Definitions,” states, in relevant part, that IROFS are “structures, systems,

equipment, components, and activities of personnel that are relied on to prevent potential

accidents at a facility that could exceed the performance requirements in [10 CFR] 70.61 or to

mitigate their potential consequences.” As stated in the rulemaking associated with 10 CFR

Part 70, Subpart H, “Additional Requirements for Certain Licensees Authorized to Possess a

Critical Mass of Special Nuclear Materials,” (NRC, 2001), IROFS may be described at the

systems level, provided that there is enough detail to understand the function of the system in
I-6

relation to the performance requirements. Accordingly, as discussed in subsequent FSER

chapters, the staff finds it acceptable to identify PSSCs at the systems level. Table 5.1-3 of the

FSER summarizes each PSSC identified by DCS, the safety functions of each PSSC, and the

design bases associated with the PSSCs. As indicated above, the IROFS identified in any

subsequent DCS application for a license to possess and use licensed material would be

expected to contain more component-level controls as part of the facility’s final design.

Management measures, as defined in 10 CFR 70.4, will be applied to IROFS to assure that

they are available and reliable to perform their functions when needed. The ISA summary, the

IROFS identified therein, and management measures will be evaluated when the staff reviews

any subsequent DCS application for a license to possess and use licensed material. However,

in determining, pursuant to 10 CFR 70.23(b), that the applicant’s QA program will provide

reasonable assurance of protection against natural phenomena and the consequences of

potential accidents, an evaluation of the management measures described in CAR Chapter 15

is set forth in FSER Chapter 15.

With respect to QA issues, on June 22, 2000, DCS submitted a MOX project quality assurance

plan (MPQAP). The staff’s findings and conclusions on the DCS MPQAP, including Revision 2

dated January 29, 2001, are documented in a safety evaluation report (SER) dated October 1,

2001. On March 26, 2002, DCS submitted to NRC Revision 3 of the MPQAP. Revision 3 of the

MPQAP incorporated all of the DCS commitments noted in the October 1, 2001, SER. By letter

dated January 10, 2003, the NRC approved MPQAP, Revision 3, for the MFFF construction

activities, including design, procurement, and fabrication.

Reference

(NRC, 2001) Nuclear Regulatory Commission (U.S.), Washington, D.C. “Domestic Licensing of

Special Nuclear Material; Possession of a Critical Mass of Special Nuclear Material.” Federal

Register: Vol. 65, No. 181, pp. 56211-56231. September 18, 2001.
1-1

1. GENERAL INFORMATION

1.1 Facility and Process Overview

1.1.1 Conduct of Review

This chapter of the final safety evaluation report (FSER) discusses general information

contained in Chapter 1 of the revised mixed oxide (MOX) fuel fabrication facility (MFFF or the

facility) construction authorization request (CAR) (DCS, 2002). Chapter 1 of the MFFF revised

CAR provides general information about the facility processes and the site. It consists of a

general facility description, material flow, and process overview. The objective of FSER

Chapter 1 is to familiarize the reader with the pertinent features of the proposed facility and the

site.

1.1.1.1 General Facility Description

The proposed facility will be a “plutonium processing and fuel fabrication plant,” as defined in

Title 10, Section 70.4, “Definitions,” of the Code of Federal Regulations (10 CFR 70.4). The

facility will be designed to produce fuel assemblies for commercial nuclear power plants. The

assemblies are composed of fuel rods which contain fuel pellets consisting of a blend of

uranium and plutonium dioxides (i.e., mixed oxides). The plutonium dioxide to be used would

be obtained from weapons-grade plutonium inventories held by the U.S. Department of Energy

(DOE), which are declared surplus to national security needs.

The proposed MFFF would be located in the F-Area of the DOE Savannah River Site (SRS)

near Aiken, South Carolina. The site encompasses approximately 0.17 km2 (41 a), of which

approximately 6.9 × 10-2 km2( 17 a) will be developed with roads, facilities, or buildings. No

roads, railroads, or waterways traverse the facility site. The nearest public transportation route

is South Carolina Route 125, approximately 6.4 km (4 mi) to the west.

1.1.1.1.1 Controlled Area Boundary

With respect to the controlled area boundary, 10 CFR 70.61(f) requires that the applicant

establish a controlled area, as defined in 10 CFR 20.1003, “Definitions.” Section 20.1003 of 10

CFR defines the controlled area as an area outside of the restricted area but inside the site

boundary, access to which can be restricted by the licensee for any reason. Section 20.1003 of

10 CFR defines the term “restricted area” as an area the “access to which is limited by the

licensee for the purpose of protecting individuals against undue risks from exposure to

radiation.” Additionally, 10 CFR 20.1003 defines “site boundary” as the “line beyond which the

land or property is not owned, leased, or otherwise controlled by the licensee.” As discussed

further below, the site boundary line for the proposed facility will correspond with the MFFF site

perimeter, which encompasses 0.17 km2 (41 a). The restricted area and controlled area are

shown in Figure 1.1-2 of the CAR. Pursuant to 10 CFR Section 70.61(f), Duke Cogema Stone

& Webster (DCS or the applicant) will retain authority to exclude or remove personnel and

property from the controlled area. Accordingly, the staff finds that the applicant's proposed

location of the controlled area boundary meets the requirements in 10 CFR 70.61(f) and 10

CFR Part 20, “Standards for Protection Against Radiation.”
1-2

Individuals inside the proposed MFFF controlled area will be employed by the applicant and

other DOE contractors to conduct activities licensed by the U.S. Nuclear Regulatory

Commission (NRC) related to the manufacture of MOX fuel assemblies. As a result, the duties

of individuals inside the proposed MFFF controlled area will involve potential exposures to

radiation or to radioactive material, and, therefore, these individuals could receive an

occupational dose. The staff concludes, therefore, that the applicant may consider these

individuals to be workers for the purposes of demonstrating that the risks of postulated

accidents and natural phenomema hazards are acceptably low.

Individuals outside the controlled area (IOC) are subject to DOE regulations, directives, and

orders issued pursuant to DOE jurisdiction and authority. As such, the applicant committed to

meet the performance requirements for “individuals outside the controlled area,” as that phrase

is used in 10 CFR 70.61, “Performance Requirements.” The staff concludes that this

commitment is consistent with the applicable regulatory requirement and is, therefore,

acceptable.

In this FSER, the staff has adopted the applicant’s usage for names of individuals for whom

accident risks must be limited. That is, “facility workers,” are workers in the restricted area

inside a room of the MFFF near a potential accident release point. “Site workers” are workers

considered to be outside the MFFF located 100 m (328 ft) from the ventilation exhaust stack. 

IOC refers to individuals at or beyond the proposed MFFF controlled area boundary. A fourth

receptor protected under the provisions of 10 CFR 70.61 is the environment, which is

understood by both the applicant and NRC staff to be all areas outside the restricted area.

The regulations in 10 CFR Part 20 define a “member of the public” as any individual, except

when that individual is receiving an occupational dose. SRS employees located outside the

MFFF controlled area (i.e., IOC) include both those who receive an occupational dose and

those who do not. Therefore, the population outside the MFFF controlled area, and on the

SRS, includes both workers and members of the public. As a result, the regulatory terms

“member of the public” (from 10 CFR Part 20) and “IOC” (from 10 CFR Part 70, “Domestic

Licensing of Special Nuclear Material”) may apply to the same individual, depending on the

regulatory context. In this FSER, when staff evaluates the risk of accidents and natural

phenomena to an individual outside the MFFF controlled area, the term “individual outside the

controlled area,” or “IOC,” is used, to be consistent with the use of that phrase in 10 CFR Part

70.61. Alternatively, when the staff evaluates the DCS proposal for protection of the individuals

outside the MFFF controlled area from radiation hazards in accordance with 10 CFR Part 20

(see Chapter 10 of this FSER), the term “member of the public” is used.

1.1.1.1.2 Facility Buildings and Structures

Facility buildings consist of the mixed oxide (MOX) fuel fabrication building (the main building

on the site), the emergency diesel generator building, the standby diesel generator building, the

secured warehouse building, the administration building, the technical support building, and the

reagents processing building. Miscellaneous site structures consist of a gas storage pad;

heating, ventilation, air-conditioning, and process chiller pads; diesel fuel filling stations;

electrical transformers; and other minor structures.
1-3

The main building would be the MOX fuel fabrication building. This building would contain all of

the plutonium dioxide handling, fuel processing, and fuel fabrication operations of the facility. It

would be a reinforced concrete building having a footprint of approximately 91.5 m (300 ft) by

137 m (450 ft) by approximately 22.3 m (73 ft) above grade. The building would be composed

of three major functional areas, (1) the MOX processing area, (2) the aqueous polishing (AP)

area, and (3) the shipping and receiving area. In the AP area, plutonium dioxide feedstock

received from either the pit disassembly and conversion facility (PDCF), or from alternate

feedstock, would be purified to remove impurities, such as gallium and americium. The purified

plutonium dioxide would then be blended with depleted uranium dioxide powder and processed

into MOX fuel, and ultimately fuel assemblies, in the MOX processing area. In the shipping and

receiving area, plutonium and uranium dioxides would be received along with other materials

necessary to produce fuel assemblies. Completed fuel assemblies would be shipped to

commercial nuclear power plants.

Most reagents (e.g., nitric acid, hydrogen peroxide, hydroxylamine nitrate (HAN), hydrazine,

oxalic acid, sodium carbonate, diluent (HPT), and tributyl phosphate (TBP)) would be stored

and solutions would be prepared in the reagent processing building (BRP) for use in the AP

area of the facility. The building would be divided into discrete areas to segregate chemicals

and the associated equipment and vessels to prevent inadvertent chemical interaction. It would

have a below-grade collection tank room that would receive waste chemicals from the building. 

A loading dock at one end of the building would be used for unloading and transfer of chemical

containers and drums. Liquid chemical containers would be located inside curbed areas to

contain accidental spills. The applicant does not intend to store, process, or commingle

radioactive materials or radiochemicals in this building. Chemicals would be transferred to the

AP area from the BRP via piping located in a concrete, below-grade trench between the two

buildings.

1.1.1.2 Material Flow

The facility would receive plutonium dioxide from the PDCF, located on the SRS near the

facility, as well as other DOE sources (i.e., alternate feedstock). The material would be

transported to the shipping and receiving area of the facility in approved shipping containers. 

The material would be unloaded and inspected according to the Material Control and

Accounting (MC&A) and Radiation Protection Programs. The material would then be moved to

the AP or MOX Processing Area. The facility also would receive depleted uranium dioxide at

the material receipt area of the secured warehouse building, where it would be inspected

according to the MC&A and Radiation Protection Programs. The depleted uranium dioxide

would be trucked to the shipping and receiving area of the facility. Fresh MOX fuel assemblies

would be stored in the assembly storage vault in the facility before shipping offsite. For

shipping to commercial power plants, the assemblies would be moved to the shipping and

receiving area of the facility where they would be loaded into a MOX fresh fuel transportation

package that had been approved by the NRC, and then loaded onto a secure transport vehicle

for transport to commercial power plants for irradiation.

Airborne effluents from the MOX fuel fabrication building would be treated, passed through a

final two-stage high-efficiency particulate air filter to remove radioactive particles, and then

discharged through a continuously monitored stack. The exhaust streams would come from
1-4

building ventilation systems; gloveboxes; process vents of tanks, vessels and other equipment;

and the sintering furnaces.

Liquid waste streams containing radioactive materials would be sampled, characterized, and

transferred to the SRS Waste Management Program for final processing and disposal. No

radioactive liquid waste streams would be released from the proposed MFFF to the

environment. Liquid waste streams include high alpha activity waste containing americium,

gallium, and silver from the dissolution process; uranium solutions containing low-enriched

uranium; alkaline solutions; liquid low-level waste (e.g., acid recovery condensate; room

heating, ventilation, and air conditioning (HVAC) condensate; laboratory rinsing; and sanitary

washing); and solvent streams.

Solid radioactive wastes would be typically placed in 55 gal (208.2 L) drums, assayed, and

transferred to SRS for processing and disposal under the SRS Waste Management Program. 

The wastes would be compacted to reduce volume to the extent possible. These wastes

include transuranic and low-level wastes which include uranium and/or plutonium

contamination.

1.1.1.3 Process Overview

The facility would have two main process operations, (1) an AP process that serves to remove

impurities, such as americium and gallium (i.e., polishing), and (2) the MOX fuel fabrication, or

MOX process (MP), which processes the plutonium and depleted uranium dioxides into fuel

pellets, fuel rods, and fuel assemblies. A summary of the major processes in the facility is

provided below. A more detailed discussion of process chemistry and chemical safety is

provided in Chapter 8 and Sections 11.2 and 11.3 of this FSER.

1.1.1.3.1 AP Process Overview

All feedstock, both from the PDCF and from other DOE sources, will be received as plutonium

dioxide. The plutonium dioxide received at the MFFF would contain small amounts of impurities

that must be removed for use of the MOX fuel in reactors. Feedstock from the PDCF will

contain impurities such as gallium, americium, and highly enriched uranium. The diversity of

impurities and the level of impurities will be higher in alternate feedstock. Some of this

alternate feedstock may have higher than normal salt contaminants (other than chlorides),

some will contain chloride contaminants, and some will contain small amounts of uranium. The

AP process is used to remove these impurities. The AP process consists of three major steps,

(1) dissolution, (2) purification, and (3) conversion.

In the dissolution step, the plutonium dioxide powder received from the PDCF and other DOE

sources would be placed into solution by electrolytic dissolution with silver in nitric acid.

The purification step involves purification of the plutonium solution in pulsed columns by solvent

extraction. The solvent mixture would be TPB dissolved in HPT solvent. Nitrate impurities,

such as americium, gallium, and silver, remain in the aqueous phase and would be routed to an

acid recovery unit after solvent washing. The plutonium and uranium solvent stream would be

mixed in a pulsed column with a stripping solution of nitric acid, HAN, and hydrazine, which

would reduce plutonium to a trivalent state and allow its transfer from the solvent phase back to
1-5

the aqueous phase stripping solution. The organic solvent, stripped of plutonium, would be

mixed with an additional stripping solution in a plutonium barrier that assures that nearly all

plutonium is removed before the organic solvent is transferred to the uranium stripping process. 

In the uranium stripping step, the uranium would be removed from the solvent phase using

dilute nitric acid. The uranium stream would be diluted with depleted uranium before being

transferred to SRS as waste. The remaining solvent stream, stripped of plutonium and

uranium, would be routed to solvent recovery mixer-settlers to be recycled. In the purified

plutonium stream, the plutonium valence would be adjusted back to plutonium (IV) by driving

nitrous dioxide fumes through the plutonium solution in an oxidation column. The offgas would

be routed through an offgas treatment system and discharged to the atmosphere.

In the conversion step, the plutonium (IV) would be converted to a PuO2 powder using a

continuous oxalate conversion process. In this step, the plutonium (IV) would react with excess

oxalic acid to precipitate plutonium oxalate. The plutonium oxalate would be collected on a

filter, dried in a screw calciner to produce purified PuO2 powder, which is then blended and

stored in sealed cans. The spent oxalic mother liquors would be concentrated, reacted with

manganese to destroy the oxalic acid, and recycled to the beginning of the extraction cycle.

Figure 1.1-4 in the revised CAR shows the AP process.

1.1.1.3.2 MP Overview

The purified plutonium dioxide powder would be used in the MP where it would be blended with

depleted uranium dioxide powder to make MOX fuel. The applicant proposes to use a

micronized master blend (MIMAS) process, which has been used by COGEMA and

Belgonucleaire to manufacture MOX fuel in Europe. The MOX fuel fabrication process consists

of four major steps, (1) powder blending, (2) pellet production, (3) rod production, and (4) fuel

assembly production.

In the first step, a master blend of plutonium dioxide and depleted uranium dioxide powder and

recycled powder would be produced that consists of approximately 20 wt% plutonium dioxide

The powder mixture would be ground in a ball mill and mixed with additional depleted uranium

dioxide to produce a final mixture consisting of approximately 2–6 percent plutonium dioxide. 

The final blend would be homogenized to assure a uniform distribution of the plutonium dioxide. 

Lubricants and poreformers, to control density, would be added to the final mixture during

homogenization.

In the second step, the final blend would be pressed to form pellets. These pellets are referred

to as “green” pellets because they would not yet have been sintered in the furnace. The green

pellets would then be sintered in a furnace, in which the atmosphere consists of a mixture of

hydrogen and argon, to obtain the required ceramic properties. The sintering would also

remove both organic products from the pellets, the lubricant and the poreformer. The sintered

pellets would be ground to a specified diameter and sorted. Powder from the grinding

operation and from discarded pellets would be recycled through a ball mill and reused in the

powder processing.

In the third step of the MP, fuel rods would be loaded with the pellets to an adjusted pellet

length column. The rods would be welded, pressurized with helium, and decontaminated in
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gloveboxes. The filled rods would then be removed from gloveboxes, placed on racks, and

inspected.

In the fourth and final step, the filled rods would be pulled through the fuel assembly skeleton to

form completed fuel assemblies. Each assembly would consist of a 17 x 17 square grid (fuel

rods, control rod guide tubes, and instrument tubes). There are approximately 264 fuel rods

(uranium or MOX fuel) per assembly. A variety of inspections are performed on the completed

fuel assemblies before shipment. The completed assemblies would be stored for shipment to

the mission reactors. Figure 1.1-5 in the revised CAR shows the MOX fuel fabrication process.

1.1.2 Evaluation Findings

The staff concludes that the facility and process overview descriptions provided by the applicant

in Section 1.1 of the CAR are sufficient for the staff to obtain an introductory understanding of

the facility and the processes. More detailed facility and process descriptions are provided in

other sections of the CAR and are discussed in other chapters of this FSER.

1.1.3 References

(DCS, 2002) Ashe, K.L., Duke COGEMA Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Construction Authorization Request and Environmental Report Change

Pages. October 31, 2002 (including page changes through February 9, 2005).

1.2 Institutional Information

1.2.1 Conduct of Review

This chapter of the FSER contains the staff’s review of institutional information described by the

applicant in Chapter 1 of the revised CAR. The staff used Chapter 1 in NUREG-1718,

“Standard Review Plan for the Review of an Application for a Mixed Oxide (MOX) Fuel

Fabrication Facility,” (NRC, 2000) as guidance in performing the review. The staff evaluated

the institutional information provided by the applicant by reviewing Chapter 1 of the revised

CAR, other sections of the revised CAR, and supplementary information provided by the

applicant (DCS, 2002a; DCS 2002b). 

1.2.1.1 Corporate Identity

The applicant for the proposed MFFF is DCS, a consortium of Duke Project Services Group,

Inc. (DPSG), COGEMA, Inc., and Stone & Webster, Inc. The applicant is a limited liability

company registered in the State of South Carolina and its principal offices are located in

Charlotte, North Carolina.

The applicant is proposing to construct the facility at the DOE SRS near Aiken, South Carolina. 

This facility will produce MOX fuel from weapons-grade plutonium and depleted uranium using

processes adapted from existing plants in France. Licensed activities will involve an AP

process intended to remove impurities from the feed plutonium and fuel fabrication processes

that include material mixing and blending, pelletizing, sintering, and fuel rod loading and

inspection. The fuel is intended for use in commercial reactors in the United States.
1-7

The applicant provided the names of its principal corporate officers, all of whom are U.S.

citizens. The applicant is using technical support from other companies in the design of the

proposed facility. These companies include Belgonucleaire, SGN (a subsidiary of COGEMA),

Framatome ANP, and Nuclear Fuel Services, Inc. The applicant also plans to use other

contractors at the site, but these contractors have not yet been selected (DCS, 2002a; DCS

2002b).

1.2.1.2 Foreign Ownership, Control, or Influence

COGEMA, Inc., a U.S. company, owns 30 percent of DCS. COGEMA, Inc., is wholly owned by

COGEMA, SA, a French company. The remainder of the corporation is owned by DPSG (40

percent) and Stone & Webster, Inc. (30 percent), which are both U.S. companies. The NRC

has confirmed that DOE rendered a favorable foreign ownership, control, or influence

determination (FOCI) for the applicant on March 12, 1999. In addition, DOE rendered a

favorable FOCI determination for DPSG on June 10, 2002. The Department of Defense also

rendered a favorable FOCI determination for Stone & Webster. On February 19, 2004, the

NRC entered into a security cognizance agreement with DOE, whereby DOE will maintain

security oversight of NRC interests at DPSG. The NRC accepts DOE FOCI determinations

based on a memorandum of understanding between the NRC and DOE dated October 9, 1996.

1.2.1.3 Proposed License Information

The applicant provided information on its proposed operations and the type of license (including

possession limits) it will later be requesting. The applicant is requesting authorization to

construct an MFFF pursuant to 10 CFR 70.23(b). The applicant plans to request an NRC

materials license pursuant to 10 CFR Part 30, “Rules of General Applicability to Domestic

Licensing of Byproduct Material,” 10 CFR Part 40, “Domestic Licensing of Source Material,” and

10 CFR Part 70 for the following materials:

[Text removed under 10 CFR 2.390]
1-8

The term of the license to be requested to possess and use these materials is 20 years.

The applicant also identified special exemptions to be requested related to decommissioning

funding and financial protection under the Price-Anderson Act. The applicant will request these

exemptions as part of the license application.

1.2.2 Evaluation Findings

The staff evaluated the institutional information for approval to construct an MFFF at the SRS

according to Section 1.2 of NUREG-1718 (NRC, 2000). This institutional information identifies

the applicant’s corporate structure, favorable FOCI determinations, proposed license

possession limits, and special exemptions to be requested in the license application. This

information is complete and accurate, is consistent with the recommendations in NUREG-1718

(NRC, 2000), and is, therefore, acceptable.

1.2.3 References

(DCS, 2002a) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear

Regulatory Commission, RE: Clarification of Responses to NRC Request for Additional

Information. April 23, 2002.

(DCS, 2002b) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear

Regulatory Commission, RE: Requests for Additional Information, Clarifications, and Open

Item Mapping into the Construction Authorization Request Revision. November 22, 2002.

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility.” Washington,

DC, 2000.

1.3 Site Description

1.3.1 Conduct of Review

This chapter of the FSER contains the staff’s review of the site description provided by the

applicant in Chapter 1 of the revised CAR (DCS, 2002a). The objective of this review is to (1)

ensure that site conditions, including site geography, demographics, meteorology, hydrology,

and geology, are accurately described to properly define potential accident conditions, and (2)

determine whether principal structures, systems, and components (PSSCs) and their design

bases, identified by the applicant, provide reasonable assurance of protection against natural

phenomena and the consequences of potential accidents. The staff evaluated the site

description information provided by the applicant by reviewing Chapter 1 of the revised CAR,

other sections of the revised CAR, supplementary information provided by the applicant, and

relevant documents available at the applicant’s offices but not submitted by the applicant. The

review of the site description was closely coordinated with the natural phenomena accident

sequences described in the safety assessment of the design bases (see Chapter 5 of this

FSER) and the review of other plant systems. 
1-9

The staff reviewed how the information in the revised CAR addresses the following regulations:

• 10 CFR 70.23(b) states, as a prerequisite to construction approval, that the design

bases of the PSSCs and the quality assurance program be found to provide reasonable

assurance of protection against natural phenomena and the consequences of potential

accidents.

• 10 CFR 70.64, “Requirements for New Facilities or New Processes at Existing

Facilities,” requires that baseline design criteria (BDC) and defense-in-depth practices

be incorporated into the design of new facilities. With respect to natural phenomena

hazards,10 CFR 70.64(a)(2) requires that the design of new facilities must provide for

adequate protection against such hazards, with consideration of the most severe

documented historical events for the site.

Section 1.3 of the revised CAR discusses the geographical location of the MFFF and its

environment, including demographic, meteorological, hydrological, geological, seismological,

and geotechnical characteristics of the site and the surrounding area. It describes population

distribution near the site, land and water uses, transportation routes, and nearby industrial

facilities which potentially can affect the site. It also describes and evaluates site characteristics

that affect the magnitude of natural phenomena (e.g., rain, snow, wind, and earthquakes) that

may affect the site. This section also evaluates site characteristics with respect to safety and

identifies assumptions and input that are needed to evaluate safety and the design bases in

other evaluations in the revised CAR.

The staff evaluated site characteristics by reviewing Section 1.3 of the revised CAR, documents

cited in the revised CAR, and other relevant literature. Where appropriate, findings of

regulatory compliance are made for requirements that are fully addressed in Section 1.3 of the

revised CAR. In some cases, regulatory compliance can only be determined by integrating the

information in Section 1.3 of the revised CAR with information in other sections of the revised

CAR. In these cases, evaluations of regulatory compliance are made in Chapter 5, “Safety

Assessment of the Design Basis,” of the revised CAR and findings of technical adequacy are

made in Section 1.3 of this FSER.

1.3.1.1 Site Geography

In the application, the applicant provided information on the site location to include State,

county, municipality, and topographic information; information on public and SRS roads,

railroads, and waterways; nearby bodies of water; and significant geographical features.

The proposed site will be located in the F-Area of the SRS in southwest South Carolina near

Aiken. The site is restricted and has few public roads. There are no unrestricted public roads

in the vicinity of F-Area. A rail system is operated at the SRS by DOE. This rail system

connects to commercial rail lines outside SRS boundaries. Nearby, the principal body of water

is the Savannah River, which forms the SRS southwest boundary. The only river navigation

that takes place is infrequent construction-related barge traffic. The significant physiographic

features at the SRS are the Pleistocene Coastal Terraces and the Aiken Plateau. The

applicant provided supplemental information about aircraft flights and airports in a letter dated

March 8, 2002 (DCS, 2002a). The applicant stated that there are only two airports within 96.5
1-10

km (60 mi) of SRS that provide scheduled air passenger services. Six general aviation airports

were identified by the applicant. Aircraft flight data were presented based on Federal Aviation

Administration data. Aircraft hazards are discussed in Section 11.1 of this FSER.

This geographic information provided in the application is current and accurate, is appropriately

referenced, and is consistent with information used in the safety assessments to support the

design bases of PSSCs.

1.3.1.2 Demographics and Land Use

In the application, the applicant provided information on demographics and land use to include

the 1990 Census data for the area and for minority and low-income populations; a description,

distance, and direction to nearby population centers, public facilities, hospitals, and industrial

facilities that could present potential hazards; residential, industrial, commercial, and

agricultural land use data in the vicinity of the proposed site; and uses of nearby bodies

of water.

There are 621,527 people living within 80 km (50 mi) of the proposed facility site based on the

1990 Census data. The population is expected to grow to slightly more than 1,000,000 in 2030. 

This population includes those living in the two metropolitan areas of Augusta, Georgia, and

Aiken, South Carolina. Because the proposed site is on the SRS, there are no residents within

8 km (5 mi) of the proposed site. Within 8 km (5 mi) and 16 km (10 mi) of the site, 6528 people

reside, the majority being in the towns of New Ellenton and Jackson, South Carolina.

Nearby industrial areas include other DOE SRS operations; several other Federal and State

sponsored activities; Chem-Nuclear Systems, Inc., commercial low-level waste disposal and

waste transportation activities in Barnwell County; Transnuclear, Inc., waste transportation

activities in Aiken County; Carolina Metals, Inc., depleted uranium processing operations in

Barnwell County; the Vogtle nuclear generating station across the Savannah River in Georgia;

a fossil-fired electric generating plant 32 km (20 mi) north of the SRS; and the Fort Gordon

Army post southwest of Augusta, Georgia.

Within the SRS, land use is controlled for the purposes of DOE operations and timber

management. Forested areas within the SRS are managed by the U.S. Forest Service.

The Savannah River is used to supply domestic water, following treatment, for fish propagation,

and for commercial and agricultural uses. Except for limited transportation of construction

equipment, there is no commercial shipping performed on the river. Domestic uses of water

from the Savannah River occur approximately 161 km (100 mi) downstream at treatment plants

near Hardeeville, South Carolina, and Savannah, Georgia.

Ground water extracted near the SRS is used for domestic, industrial, and agricultural activities. 

Smaller communities, schools, and small commercial businesses also use local ground water. 

Nearly 133 million L/d (35 million gal/d) of ground water were pumped in 1985 by 56

communities and industries near the SRS. 
1-11

This demographic and land use information provided in the application is accurate, is

appropriately referenced, and is consistent with information in the safety assessments used to

support the design bases of PSSCs. 

1.3.1.3 Meteorology

In the revised CAR, the applicant provided meteorological information on temperatures; wind

speeds and average and prevailing wind directions; amounts and form of precipitation; design

basis values for maximum snow and ice loads and probable maximum precipitation; and

types, magnitudes, and frequency of severe weather events, such as tornadoes, hurricanes,

and lightning.

Temperature data for the SRS are presented in the revised CAR based on 30 years of

measurements at the site. The annual average temperature is 18.2 °C (64.7 °F). Observed

temperature extremes ranged from 41.7 to !19.4 °C (107 to !3 °F). Data for Augusta, Georgia,

indicate that daytime high temperatures rarely fall below 0 °C (32 °F) during the winter.

Temperatures are above 32.2 °C (90 °F) on more than half the days in the summer months.

Winds near the SRS are generally light to moderate, with the highest wind speeds occurring in

the spring. The lightest winds occur in the summer and fall. The prevailing wind direction

varies throughout the year, coming from the northwest in the winter, from the southeast in the

late spring and early autumn, and from the southwest in the summer. The peak wind gust at

Bush Field in Augusta, Georgia, was 96.5 km/h (60 mph) based on 10 years of data.

The average annual precipitation for the SRS from 1967 to 1996 is 126 cm (49.6 in.). The most

rainfall during a 24-hr period was 19 cm (7.5 in.) in October 1990. During summer

thunderstorms, rainfall rates of up to 5.1 cm/h (2 in./h) can occur. An average of 54

thunderstorm days per year have been observed. Hail storms occur infrequently, an average of

once every 2 years.

Snowfalls of 2.5 cm (1 in.) or greater occur on the average once every 3 years. The greatest

single snowfall recorded from 1951 to 1995 occurred in Augusta, Georgia, in 1973 when

35.6 cm (14.0 in.) of snow fell. The maximum ground snow load for a 100-yr recurrence period

is 0.29 kPa (6 psf). Ice accumulates once every 2 years. The maximum accumulation for a

100-yr recurrence period is 1.7 cm (0.67 in.) or an ice load of 144 Pa (3 psf).

During a 30-yr period (1967–1997), 165 tornadoes occurred in the vicinity of the SRS. Five

Fujita-scale 2 and four Fujita-scale 1 tornadoes have occurred onsite or in close proximity since

site operations began. Damage was primarily to trees. One of these tornadoes produced wind

speeds up to 241 km/h (150 mph). Design basis wind speeds for the DOE moderate hazard

performance category (PC–3) facilities and high hazard performance category (PC–4) facilities

are 290 km/h (180 mph) and 386 km/h (240 mph), respectively. The PSSCs are evaluated for

a tornado recurrence interval of 2 × 10–6 per year and a design basis tornado with a 3-second

tornado speed of 386 km/h (240 mph). For other extreme winds from hurricanes, tropical

weather systems, thunderstorms, and winter storms, PSSCs will be evaluated based on a

recurrence period of 1 × 10-4 per year for a 3-second wind speed of 209 km/h (130 mph). 

These extreme wind speeds are based on SRS meteorological data and data from National
1-12

Weather Service stations in Columbia, South Carolina, and Augusta, Macon, and Athens,

Georgia.

During the period 1700–1992, 36 hurricanes caused damage in South Carolina. However, no

hurricane-force winds of greater than 120 km/h (75 mph) have been measured at the SRS.

Extreme rainfalls generally occur during spring and summer thunderstorms and tropical storms. 

The design basis rainfall for PSSCs is evaluated for a recurrence interval of 1× 10-5 for various

rainfall durations (e.g., 9.9 cm (3.9 in.) for a 15-min rainfall and 58 cm (22.7 in.) for a 24-hr

rainfall).

The number of lightning strikes is estimated at 10 strikes/km2@yr (26 strikes/mi2@yr). From 1989

to 1993, SRS data show an average of 4 strikes/km2@yr (10.4 strikes/km2@yr).

Meteorological information provided in the application was current and accurate, was

appropriately referenced, and was consistent with information in the safety assessments used

to support the design bases of PSSCs.

1.3.1.4 Hydrology

In the revised CAR, the applicant provided information on surface hydrology, including

descriptions of nearby rivers, streams, and other water bodies; subsurface water hydrology,

including water table depths, flow characteristics, potentiometric surfaces, and aquifer

characteristics; and design basis floods.

The Savannah River forms the southwest boundary of the SRS and is the dominant body of

surface water in the nearby area. The Savannah River Basin drains an area 27,394 km2

(10,577 mi2) and extends 465 km (289 mi) from the Atlantic Ocean to the Blue Ridge

Mountains. The principal streams that enter the Savannah River from the SRS are Upper

Three Runs, Fourmile Branch, Pen Branch, Steel Creek, and Lower Three Runs. These

streams discharge water from rainfall, subsurface waters, and various effluent streams from

SRS operations. SRS surface water bodies include Par Pond and L Lake, created as cooling

water reservoirs for production reactors, marshes, and natural basins, including Carolina bays.

The record historical Savannah River flood at Augusta, Georgia, in 1796 had a discharge of

10,000 m3/s (360,000 ft3/s). The peak Savannah River flow recorded by the U.S. Geological

Survey was 9,900 m3/s (350,000 ft3/s) in 1929. There have been no major floods in the

Augusta area since dams were constructed upstream of Augusta beginning in the 1950s. The

estimated 50-yr maximum flow is now 2,100 m3/s (74,600 ft3/s). The probable maximum flood

at the SRS is a water level of 68.4 m (224.5 ft) above mean sea level. The normal Savannah

River flow elevation at the SRS boat dock is 25.9 m (85 ft). The design basis flood for the

MFFF is 63.4 m (207.9 ft) above mean sea level with an annual recurrence interval of 1 × 10-5. 

Because the facility is proposed to be located at an elevation of 82.9 m (272 ft), the probabilities

of flooding the site were calculated to be less than 1× 10-5 per year. A cascading failure of the

Savannah River dams upstream of Augusta, Georgia, was estimated to produce a peak flow in

the Savannah River of 28,000 m3/s (980,000 ft3/s) and a flood elevation of 43 m (141 ft) at the

Vogtle station, which is directly across from the SRS on the Georgia side. Because the MFFF
1-13

is at an elevation of 82.9 m (272 ft), this cascading failure and other events, such as ice

flooding, wave surges, and seiches, will not affect the facility.

The ground water setting at the SRS is characterized by three aquifer systems that overlay the

bedrock formations of the Southeastern Coastal Plain. The Southeastern Coastal Plain

consists of sediments deposited from erosional processes of the Appalachian Mountains that lie

to the west of the SRS. These sediments consist of water-bearing sandy materials and

limestone and clayey confining units. In the F-Area, the confining units of the three aquifer

systems become disjointed and have poor separation that allows flow between aquifer systems. 

In the uppermost Floridan Aquifer System, the Three Runs Aquifer overlays the deeper Gordon

Aquifer. These aquifers are separated by a broken confining unit. Recharge of these aquifers

is primarily through local precipitation, and discharge is primarily through local streams. 

Because Upper Three Runs Creek and the Savannah River incise the Floridan Aquifer System,

there is a head reversal between it and the Crouch Branch Aquifer in the Dublin Aquifer System

which lies just below the Floridan Aquifer System. This means that ground water from the lower

system is under a greater head and flows up into the Floridan. This phenomenon tends to limit

migration of contamination into the lower aquifer systems. The Midville Aquifer System is the

deepest system and lies just above the bedrock formations.

At the proposed MOX site, the ground water table is nearly 15 m (50 ft) below the existing

ground level. Potentiometric surface maps show that ground water in the uppermost Upper

Three Runs Aquifer flows principally toward Upper Three Runs Creek and toward the unnamed

creek located toward the northeast of the proposed site. The underlying Gordon Aquifer flows

horizontally toward the Savannah River. The deeper Dublin and Midville Aquifer Systems flow

to the southeast toward the Savannah River and the coast. The hydraulic conductivity of the

Upper Three Runs Aquifer varies from less than 0.3 m/d (1.0 ft/d) to almost 10 m/d (33 ft/d),

with an average of nearly 3 m/d (10 ft/d). At the MOX site, ground water is abundant, usually

soft, slightly acidic, and low in dissolved solids. Ground water used in site operations from the

F-Area is treated to raise the pH and remove iron.

The F-Area seepage basin, located to the west of the proposed MOX site, was remediated in

2000 according to a hazardous waste Part B, postclosure permit issued by the State of South

Carolina. After remediating the site, boundary wells hydrologically downstream of the seepage

basin were installed and samples were analyzed. The first set of analyses indicated that there

is a contamination plume that exceeds the Environmental Protection Agency drinking water

standards. The DOE staff is beginning an investigation to further evaluate this ground water

contamination and to better understand the local hydrologic conditions that may exist at the

proposed MOX site.

The applicant indicated that there is radioactive contamination in the Upper Three Runs aquifer

from upgradient contamination sources in the F-Area, as well as the F-Area seepage basin. 

This ground water contamination consists of concentrations of gross alpha and beta activity,

uranium, tritium, and trichloroethylene exceeding the maximum contamination limits for drinking

water. The applicant also indicated that ground water contamination occurs at least 9.1 m (30

ft) below the deepest level of expected construction.

During site characterization activities, the applicant measured radioactivity levels of soils using

Geiger-Mueller detector scans and gross alpha and beta measurements of soil samples. The
1-14

applicant indicated that the sensitivity of the gross alpha and beta measurements was

200,000,000 pCi/kg and 100,000 pCi/kg, respectively. In a letter dated February 11, 2003,

(DCS, 2003a), DCS stated that soil radioactivity measurement sensitivity (MDC) in the

preconstruction environmental monitoring report (Fledderman, 2002) was much better than

described in the CY2000 geotechnical investigations.

The NRC staff compared the results. The CY2000 geotechnical value was 200,000,000 pCi/kg

gross alpha. The 2002 preconstruction environmental monitoring report measured values of

actinides in soil which include a mean value of 12.5 pCi/kg 239Pu, and a maximum of 4380

pCi/kg 239Pu, for example. The SRS radiological soil guides for SRS worker protection is

248,000 pCi/kg (Jannik, 1995). Across the depth profile, the values are as follows:

Pu-239 (pCi/kg)

depth mean max 

0–3" 137 690

3–6" 87.1 1590

6–9" 154 4380

9–12" 121 4280

These values correspond to a potential maximum exposure of 0.003 mSv (0.3 mrem) to an

exposed worker using the mean values and a maximum exposure of 0.033 mSv (3.3 mrem)

using the maximum values (Fledderman, 2002). The 0.033-mSv (3.3-mrem) annual projected

dose is acceptable because the NRC annual limit for members of the public in the controlled

area is 1 mSv (100 mrem). The higher 50-mSv (5000-mrem) limit for workers does not apply,

unless construction workers could receive an occupational dose.

The hydrologic information provided in the application is current and accurate, is appropriately

referenced, and is consistent with information in the safety assessments used to support the

design bases of PSSCs.

The planned construction of the facility will not penetrate into the upper ground water table that

exists 15 m (50 ft) below grade level. The ground water contamination in the Upper Three

Runs aquifer is, therefore, not expected to result in hazardous conditions that could

affect construction.

1.3.1.5 Seismic Hazards

To assess the potential seismic hazard at the site, DCS established two sets of ground motion

spectra in the revised CAR; one for the design of the surface facilities and one for soil stability

analyses (liquefaction and dynamic settlements). Although the details of these spectra differ,

analyses presented in the revised CAR show they are comparable. The design spectra (both

vertical and horizontal) for the facility uses a spectrum included in Regulatory Guide (RG) 1.60,

“Design Response Spectra for Seismic Design of Nuclear Power Plants” (NRC, 1973) anchored

at 0.20g peak ground acceleration (PGA). The spectra are also used for the design of the

nearby Vogtle Nuclear Power Plant (licensed under 10 CFR Part 50, ”Domestic Licensing of

Production and Utilization Facilities”). For soil stability, a spectrum was developed based on the

existing DOE uniform hazard spectra developed for the SRS. Because the seismic design and

analyses rely on these established spectra, much of the site-specific seismic hazard information
1-15

presented in the revised CAR was developed to establish that these two proposed design and

soil stability analysis spectra are adequate to meet the regulatory requirements of

10 CFR Part 70 and the performance guidelines in NUREG–1718. 

The following areas concerning the seismic hazards applicable to the safety analysis and

design of the proposed facility were reviewed:

• seismic source characterization

• ground motion attenuation

• seismic hazard calculations

• SRS-wide rock and surface response spectra

• site response and design ground motion

• surface faulting

1.3.1.5.1 Seismic Source Characterization

Geological and Tectonic Setting

The revised CAR provides a detailed description of the local and regional geological and

tectonic settings. The revised CAR noted that the SRS is located on sediments of the Upper

Atlantic Coastal Plain in South Carolina. These sediments consist of stratified, but generally

unconsolidated, sands, silts, clays, and carbonaceous muds deposited in fluvial, deltaic,

near-shore, and marine shelf environments. They range in age between Late Cretaceous

(~100 Ma) and the present and reach a maximum thickness of approximately 1200 m (4000 ft). 

Similar to Coastal Plain sedimentary sequences along the entire Atlantic seaboard, the South

Carolina Coast sediments rest unconformably on Precambrian to Paleozoic (~ 1.1 Ga to

245 Ma) metamorphic, metasedimentary, and igneous rocks of the Appalachian Orogen and on

Triassic to Early Jurassic (~245–180 Ma) siliciclastic rocks associated with early rifting along

the North American continental margin. Age and distribution of the rocks and strata provide an

adequate geologic record to assess faulting and earthquake hazards. 

Earthquakes that could impact safe operation of the proposed facility are associated with two

seismic sources, a repeat of the Charleston 1886 earthquake within the Middle

Place-Summerville Seismic Zone and small shallow earthquakes of the South Carolina

Piedmont. Earthquake source characteristics associated with these seismic zones are

consistent with information used in both the Electric Power Research Institute (EPRI) (EPRI,

1989) and Lawrence Livermore National Laboratory (LLNL) (NRC, 1994) seismic hazard

studies for the eastern United States. As discussed in Section 1.3.1.5.2 of this report, the

bedrock uniform hazard spectra for both the EPRI and LLNL form the basis for the site-wide

DOE PC–3 and PC–4 hazard spectra (WSRC, 1997).

Historical Seismicity

The revised CAR provides a summary of the records of historical seismicity, including those

from the cultural historical record (historical accounts date back to about 1698), as well as more

recent instrumented earthquake records (the South Carolina Seismic network and the SRS

network, both in operation since the mid-1970s). As noted in the revised CAR, the most

significant earthquake source is a repeat of the 1886 Charleston, estimated to have a modified
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Mercalli intensity (MMI) of X at Charleston, South Carolina, and an MMI at the SRS of VI–VII.

Magnitude estimate of the 1886 Charleston earthquake is M 7.3 ± 0.3 (Johnston, 1996;

Ambraseys, 1988). Other significant historical earthquakes felt at the SRS include the 1913

Union County earthquake (an MMI of VII at the epicenter and an MMI of II–III at Aiken, South

Carolina); the 1811–12 New Madrid earthquakes (M > 8.0 at New Madrid, Missouri); and the

1897 Giles County, Virginia, earthquake (MMI of VII, M 5.6 at Pearisburg, Virginia). 

Paleoliquefaction features indicate that the Charleston-type earthquake has reoccurred at least

seven times in the last 6000 years (Talwani, 2001). These prehistoric earthquakes appear to

be restricted to the Carolina Coastal Plain (Talwani, 2001). Two scenarios have been proposed

to explain the distribution of the paleoliquefaction features. In the first scenario, the

earthquakes occurred at Charleston, Georgetown, and Bluffton, South Carolina. In the second

scenario, all the prehistoric earthquakes occurred at Charleston. Hu, et al. (Hu, 2002) conclude

that the paleoliquefaction features were produced by earthquakes with magnitudes between 5.3

and 7.8. 

No definitive geologic evidence has yet been discovered to tie the 1886 Charleston earthquake

to a causative seismogenic fault. Tarr, et al. (Tarr, 1981) defined the Middleton

Place-Summerville Seismic Zone to include the known distribution of seismicity and

paleoseismicity associated with the Charleston-type earthquake. The Middleton

Place-Summerville Seismic Zone is located 20 km (12 mi) northwest of the city of Charleston,

South Carolina. Based on geological and geophysical data, Marple (Marple, 1994),

Madabhushi and Talwani (Madabhushi, 1993), and Marple and Talwani (Marple, 2000) all

inferred that complex and interactive strike slip and reverse faulting associated with the

northwest trending Ashley River fault and the north-northeast trending Woodstock fault were

the most likely causes of the Charleston earthquake. Recently, Weems and Lewis (Weems,

2002) concluded that the region around Charleston, South Carolina, is an active tectonic zone

that accommodates differential movement between the Cape Fear arch and the Southeast

Georgia embayment. All these models are consistent with the source characterization of the

Charleston-type earthquake presented in the EPRI and LLNL probabilistic seismic hazard

assessment (PSHA) seismic hazard studies.

Near the SRS, instrumented historical seismic records indicate that seismicity associated with

the SRS and surrounding region is closely related to the earthquake activity within the South

Carolina Piedmont (Bollinger, 1992). This activity is characterized by shallow, small magnitude,

and infrequent earthquakes. Searches of the National Earthquake Information Center and

Council of National Seismic System show that the vast majority of these earthquakes are M 3 or

less. The largest magnitude earthquakes in the record are the 1974 M 4.9 and M 4.7 events. 

All instrumented earthquakes on the SRS itself were M 2.7 or less. 

Earthquake Recurrence

The long repeat times (> 500 yr) and relatively brief historical record (< 350 yr), coupled with the

absence of active surficial deformation, limit estimates of earthquake recurrence for a

Charleston-type earthquake. The most complete record of the temporal and spatial distribution

of large prehistoric earthquakes comes from identification of earthquake-induced liquefaction

features called sand blows. Numerous sand blows have been identified throughout the South

Carolina coastal area, but few if any outside this region (WSRC, 2000b). Recent reanalysis of
1-17

the paleoliquefaction investigations in South Carolina and recalibrated 14C ages suggest that

there were as many as seven large-magnitude earthquakes in the Charleston region within the

last 6000 years (Talwani, 2001). These results translate to a recurrence interval for the

Charleston-type earthquake of 500 to 600 years. This estimated recurrence interval is

conservative because it assumes the maximum number of possible paleoearthquakes using the

age constraints derived from the 14C age data. Talwani and Schaeffer (Talwani, 2001) used 1F

error ranges to develop their list of age-distinct paleoearthquakes. Overlap of the 14C ages

using 2F error ranges, as advised by Tuttle (Tuttle, 2001) would result in a smaller number of

age-distinct paleoearthquakes during this same 6000-yr interval and thereby increase the

recurrence interval. Nevertheless, the 500–600-yr recurrence interval for the Charleston-type

earthquake is consistent with the LLNL and EPRI PSHA studies. 

Staff Review of Seismic Source Characterization

The staff reviewed the information about seismic sources presented in the revised CAR and

found it sufficient because the applicant identified and assessed all of the potentially significant

seismic sources related to the SRS (including, but not limited to, the Charleston seismic zone). 

The characterization of the tectonic setting and identification of capable seismic sources were

based on extensive review of the published geological literature, regional and site geological

and geophysical data, historical and instrumental seismicity data, regional stress field analysis,

and geological investigations of prehistoric earthquakes. The information follows guidelines

presented in RG 1.165, “Identification and Characterization of Seismic Sources and

Determination of Safe Shutdown Earthquake Ground Motion,” (NRC, 1997) and Section 2.5.2.2

of NUREG–0800, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear

Power Plants” (NRC, 1987). Criteria used to assess capable fault and areal source zones

include those outlined in Appendix A, “Seismic and Geologic Siting Criteria for Nuclear Power

Plants” to 10 CFR Part 100, “Reactor Site Criteria,” as well as those in DOE–STD–1022–94,

“Natural Phenomena Hazards Characterization Criteria” (DOE, 1994). 

Information provided by DCS to determine the tectonic setting of the facility was developed into

a coherent, well-documented discussion that provides an adequate technical basis for

evaluation of the seismic potential of the site. Specifically, documentation in the revised CAR is

sufficient to determine the earthquake potential of geological structures and potential tectonic

zones (i.e., regions of uniform earthquake potential). The information provided in the revised

CAR is also sufficient to evaluate uncertainties associated with seismic source geometry (e.g.,

fault dip, width, segmentation, and depth of seismogenic crust) and recurrence models. Thus,

the staff reviewed the information in the revised CAR and found it acceptable because the basic

geologic and seismic characteristics of the site and vicinity are adequately described in detail to

allow investigation of seismic characteristics at the facility.

1.3.1.5.2 Ground Motion Attenuation

Seismic hazards used to define bedrock uniform hazard spectra at the SRS are based on the

LLNL and EPRI probabilistic seismic hazard studies. The LLNL and EPRI bedrock uniform

hazard spectra were averaged and then broadened using the SRS-specific spectral shapes to

develop bedrock response spectra.
1-18

Ground motion attenuation models contained in the LLNL and EPRI hazard studies

incorporated a number of models developed by individuals and organizations for the

southeastern United States. These models are considered to be representative of the

state-of-the-art studies of ground motion attenuation characteristics in the southeastern United

States and have captured diverse opinions in the scientific community.

The ground motion attenuation model used to develop site-specific spectral shapes was the

Band Limited White Noise/Random Vibration Theory ground-motion model (Hanks, 1981;

Boore, 1983). In applying this stochastic approach, the applicant used the layered crustal

velocity model developed by Herrmann (Herrmann, 1986) with some modifications, the EPRI

median site attenuation model (Q-model), and the range of the EPRI site-dependent parameter

Kappa values (EPRI, 1989). 

Staff Review of Ground Motion Attenuation

Ground motion attenuation models used in the LLNL and EPRI studies are representative of the

current scientific understanding of ground motion attenuation in the southeastern United States. 

These attenuation models adequately capture uncertainty in ground motion estimates, including

the potential for Moho bounce effects. For example, a recent ground motion attenuation model

for the eastern United States (Campbell, 2003), which accounts for Moho bounce effects, yields

ground motion estimates consistent with those derived using the LLNL and EPRI studies.

Application of the LLNL and EPRI models to the SRS and, consequently, to the facility is

considered acceptable. The NRC staff has previously accepted the LLNL and EPRI ground

motion modeling (NRC, 1997) for sites in the central and eastern United States.

The use of the stochastic model or numerically simulated ground motions in the central and

eastern United States, instead of recorded ground motions, is consistent with common practice

and the state of knowledge because sufficient strong motion data are lacking in this tectonic

regime as a result of low seismicity rates. The approach was accepted by the staff in its review

of the PSHA for the Paducah Gaseous Diffusion Plant (CNWRA, 1999). In addition, the

random vibration theory model has been shown to yield conservative results for eastern U.S.

crustal conditions (Silva, 1989). Thus, the staff has determined that the applicant’s ground

motion attenuation modeling is acceptable because it provides reasonable assurance that

ground motion attenuation modeling is accurate. 

1.3.1.5.3 Seismic Hazard Calculations

The applicant used the seismic hazard results from the LLNL and EPRI probabilistic seismic

hazard to define bedrock uniform hazard spectra at the Savannah River. No other probabilistic

seismic hazard calculations were conducted specifically for the SRS or the facility. The LLNL

and EPRI hazard studies include site-specific hazard calculations for the SRS. 

Staff Review of Seismic Hazard Calculations

The LLNL and EPRI studies represent the state-of-the-art probabilistic hazard studies in the

southeastern United States. Application of these results to the SRS and, consequently, to the

facility, is considered acceptable. The NRC staff previously accepted the LLNL and EPRI data,

seismic sources, seismic hazard methods, and results (NRC, 1997) for sites in the central and
1-19

eastern United States. Thus, the staff determined using the LLNL and EPRI hazard results is

technically sound. 

1.3.1.5.4 SRS-Wide Rock and Surface Response Spectra

The SRS-wide rock response spectra were developed by Westinghouse Savannah River

Company for the entire SRS (WSRC, 1997). These are site-specific uniform hazard spectra for

bedrock from the LLNL and EPRI seismic probabilistic hazard studies, broadened by using sitespecific spectral shapes. The rock response spectra were used as the bases for developing

bedrock time histories as input into site response analyses for the facility and the SRS-wide

surface response spectra.

The SRS-wide surface response spectra are not directly used in the design of structures or in

soil stability analyses for the facility. However, they were used by the applicant to justify the

sufficiency of the selected design spectra for the facility.

The SRS-specific rock uniform hazard spectra for bedrock were developed following the

guidance and methodologies outlined in DOE–STD–1023-95, Change Notice #1, “Natural

Phenomena Hazards Assessment Criteria,” (DOE, 1995). Probabilistic hazards were

developed according to DOE PC–3 and PC–4 spectra. The DOE PC–3 and PC–4 spectra were

developed following seismic design and evaluation criteria in the DOE STD–1020–94, Change

Notice No. 1, “Natural Phenomena Hazards Design and Evaluation Criteria for Department of

Energy Facilities” (DOE, 1996) and DOE STD–1020–2002 (DOE, 2002). In DOE

STD–1020–94, PC–3 and PC–4 categories have mean annual probabilities of exceedance for

design ground motions at 5 × 10!4 and 1 × 10!4, respectively. In terms of the annual return

period ground motions, mean annual probabilities of exceedances of 5 × 10

!4

 and 1 × 10!4

correspond to mean 2,000-yr and 10,000-yr return period ground motions, respectively. 

The development of the rock response spectra included the following procedures: 

• The mean bedrock uniform hazard spectra were computed for two mean annual

probabilities of exceedances, 5 × 10!4 and 1 × 10!4 (corresponding to performance

categories of PC–3 and PC–4, respectively), by averaging the LLNL and EPRI mean

uniform hazard spectra for the SRS.

• Site-specific spectral shapes were generated using EPRI mean magnitude and mean

distance values based on the magnitude and distance deaggregation results at each

probability of exceedance.

• The spectral shapes were then scaled to the corresponding mean bedrock uniform

hazard spectrum at frequencies 1–2.5 and 5–10 Hz.

• The resulting three spectra (the averaged LLNL and EPRI uniform hazard spectrum and

the 1–2.5-Hz- and 5–10-Hz-scaled site-specific spectra) were then enveloped and

smoothed to obtain the broadened bedrock response spectra for the PC–3 and PC–4

hazards. 
1-20

Site-Wide Surface Response Spectra

Site-wide surface response spectra were obtained by multiplying the broadened bedrock

uniform hazard spectra by frequency-dependent site amplification factors to account for soil

effects. In deriving site amplification factors, hypothetical bedrock spectra were vertically

propagated through soil columns representative of the site soil conditions using the

one-dimensional equivalent linear analysis procedure developed by Silva (Silva, 1989). The

procedure was considered to be equivalent to SHAKE analyses summarized in Idriss and Sun

(Idriss, 1992).

The hypothetical bedrock spectra were power spectral density functions and spectral

accelerations for a suite of PGAs at the soil/bedrock interface (bedrock motions described

previously) and were developed using the random vibration theory model (Boore, 1983). Three

magnitude and distance dependent spectra were developed for each control motion

acceleration representing the 5

th, 50th, and 95th percentile contributions to the probability of

exceedance. Again, the magnitude and distance pairs were obtained from the EPRI

deaggregated hazard results.

The calculation of the site amplification factors considered SRS-wide variability in velocity

profile, soil column thickness, bedrock velocity, and dynamic properties (WSRC, 1997). Soil

conditions characterized in the most recent study of the site’s geotechnical properties (DCS,

2003b) are consistent with subsurface conditions reported in the previous geotechnical reports

of the site. The site-wide, uniform-hazard based response spectrum was taken as the envelope

of all of the soil response spectra obtained by multiplying the broadened mean bedrock uniform

hazard spectra by the site amplification factors for different soil/bedrock categories, scaling

frequencies, and magnitude levels. As with the design ground motions, the site-specific soil

spectra were shown to envelope the Charleston earthquake spectra.

Staff Review of SRS Rock and Surface Response Spectra

The LLNL and EPRI studies represent the state-of-the-art probabilistic hazard studies in the

southeastern United States. Application of these probabilistic hazard results to the SRS and,

consequently, to the facility is considered acceptable. The NRC staff previously accepted the

LLNL and EPRI ground motion modeling (NRC, 1997) for sites in the central and eastern United

States. In addition, broadening the LLNL and EPRI bedrock uniform hazard spectral shapes

and the development of surface response spectra are consistent with the methodologies of

DOE–STD–1023-95 (DOE, 1995). These methodologies and procedures are well established

within the ongoing seismic program at the SRS. These site-specific adjustments have been

extensively reviewed by the Westinghouse Savannah River Company and DOE. Thus, the staff

determined that SRS-wide rock and surface response spectra are acceptable because they

provide reasonable assurance that potential seismic hazards are sufficiently estimated.

1.3.1.5.5 Design Spectra and Site Response Analyses

Design Spectra

The applicant-proposed design basis ground motions for the surface facilities are an RG 1.60

spectra (NRC, 1973) anchored at 0.20g PGA, which is the same spectra used for the design of
1-21

the nearby Vogtle Nuclear Power Plant (licensed under 10 CFR Part 50). More recently,

regulations at 10 CFR 100.23, “Geologic and Seismic Siting Criteria,” for nuclear power plants

have been updated to include the application of probabilistic methods to the assessment of

seismic hazards. RG 1.165 (NRC, 1997) provides general guidance for determining the safeshutdown earthquake for new nuclear reactors based on a PSHA, consistent with the regulatory

requirements of 10 CFR 100.23. RG 1.165 recommends a reference median annual probability

of exceedance of 1 × 10!5. As shown by a similar analysis in Appendix C to DOE

STD–1020–2002 (DOE, 2002), a median annual probability of exceedance of 1 × 10!5

corresponds approximately to a mean annual probability of exceedance of 1 × 10

!4

. 

Evaluations performed by the applicant in the safety analysis report (SAR) show that the 0.20g

RG 1.60 spectra have mean annual exceedance probabilities that range between 1.6 × 10!4

and 4.5 × 10!5 (or equivalent return periods that range between 6,300 and 22,000 years; see

Table 1 of Enclosure B of Ihde, 2001b). For frequencies between 2 and 10 Hz, the mean

annual probabilities of exceedance are equal to or less than 1 × 10!4 (or equivalent return

periods greater than 10,000 years). For higher frequencies up to the PGAs, the mean annual

probabilities of exceedance are equal to or sightly greater than 1 × 10

!4

. These mean annual

exceedance probabilities are based on ground motions from the averaged EPRI (EPRI, 1989)

and LLNL (NRC, 1994) seismic hazard results for the eastern United States. 

RG 1.60 spectra anchored at 0.20g were selected by the applicant because they were deemed

to be conservative. This seismic design spectrum is shown by the applicant to lie between the

SRS-wide DOE PC–3 and PC–4 spectra. The current PC–3 and PC–4 site-wide spectra are

based on ground motion spectra developed by Westinghouse Savannah River Company

(WSRC, 1997) for the entire SRS. The PC–3 and PC–4 spectra were developed following

seismic design and evaluation criteria in DOE STD–1020–94 (DOE, 1996) and DOE

STD–1020–2002 (DOE, 2002), as discussed in more detail in Section 1.3.1.5.4 of this FSER.

To ensure safe operation of the structures, systems, and components (SSCs) beyond the

design ground motions, DOE STD–1020–94 (DOE, 1996) and DOE STD–1020–2002 (DOE,

2002) developed performance goals associated with each performance category. The

performance goals are defined in terms of the ability of the SSCs to perform essential safety

functions during and after the natural hazard phenomena (in this case an earthquake). The

acceptable behavior limit for normal use SSCs, such as buildings, is major damage, but limited

in extent such that the occupants can safely exit the building. For more critical SSCs, such as

nuclear containment structures, damage at the performance goal should be limited such that

the containment is not compromised. In DOE STD–1020–94 (DOE, 1996), as well as in DOE

STD–1020–2002 (DOE, 2002), the seismic ground motion performance goals for PC–3 and

PC–4 SSCs were established with a mean annual probability of exceedance of 1 × 10!4 and 1 ×

10!5, respectively.

On page 27 of the MOX Fuel Fabrication Facility Site Geotechnical Report (DCS, 2001a), the

applicant indicated that the desired performance goal probability is based on the approach

recommended in DOE–STD–1020–94 (DOE, 1996) and DOE STD–1020–2002 (DOE, 2002). 

That assertion is supported by performance calculations (Enclosure B of DCS, 2002a) which

show that many of the SSCs performed their safety functions to ground motion levels with a

mean annual probability of exceedance of 1 × 10!5 or less. These calculations support the

conclusion that the design criteria—RG 1.60 (NRC, 1973) spectra anchored to the 0.20g PGA,
1-22

which is significantly greater than the site-wide PC–3 spectra—is adequate for safe design of

the facility. 

The design spectra were also shown by the applicant to envelop the deterministic spectra for a

repeat of the Charleston-type earthquake. This deterministic check analysis follows the 

requirements of DOE–STD–1023-95 (DOE, 1995) using the largest historic earthquakes within

121 km (75 mi) having a moment magnitude greater than 6. In this analysis, the deterministic

median bedrock and soil spectra were generated for the 1886 Charleston earthquake using

median source parameters, a source-to-site distance of 200 km (124 mi), and other parameters

used in generating uniform-hazard-based response spectra.

In response to the staff’s request for additional information, the applicant evaluated the

vertical-to-horizontal seismic spectral ratios for the facility (DCS, 2002c). The results show that

the vertical-to-horizontal spectral ratios could exceed the standard generally used at the SRS

(normally the vertical is assumed to be two-thirds of the horizontal), particularly for frequencies

greater than approximately 3 Hz. Thus, the applicant has agreed to use both the horizontal and

vertical spectra in RG 1.60 (NRC, 1973) anchored at 0.20g PGA.

Site Response Analyses

The applicant indicated that the site-wide response spectra are intended for simple response

analysis and are not appropriate for soil-structure interaction and soil stability analyses. It

further indicated that the site-wide response spectra represent a surface response, not an

embedded response. For soil stability and soil-structure interaction analyses, a onedimensional, free-field site response analysis procedure was established by the applicant (DCS,

2001a). The control ground motions for site response analyses included the modified PC–3

motion and the 1886 Charleston motion. The modified PC–3 motion is the SRS-wide PC–3

rock response spectrum increased by a factor of 1.25 (PC–3+ rock spectrum) to yield a bedrock

PGA of 0.14g, one that would achieve the design surface PGA of 0.20g at the facility through

site response analyses. The 1886 Charleston motion is the 50th percentile attenuated rock

motion at the actinide packaging and storage facility site. It was used by the applicant to

evaluate the liquefaction potential associated with large, distant earthquakes. The spectrumcompatible acceleration time histories for both of these design motions were developed by

Westinghouse Savannah River Company.

Site response analyses were conducted using PROSHAKE, a Windows version of SHAKE91

(Idriss, 1992). The design motion time histories were applied at the base of the soil column.

Properties for the soil column were developed from geotechnical studies specific to the facility

(WSRC, 2000a; DCS, 2001b; DCS, 2003b). Soil conditions characterized in the most recent

study of the site geotechnical properties (DCS, 2003b) are consistent with subsurface

conditions reported in the previous geotechnical reports of the site. The cyclic stress ratios

computed from the site-response analyses were input into the dynamic soil-structure interaction

analyses of the critical structures and into the liquefaction analyses.

Results from site response analyses show that the PC–3+ bedrock time history produces a

surface PGA of 0.20g and a surface spectrum that correlates well to the RG 1.60 (NRC, 1973)

surface spectrum anchored at 0.20g PGA. Thus, the applicant concluded that the PC–3+
1-23

bedrock spectrum satisfies the requirement for a bedrock time history that can be used for

dynamic analysis at the facility. 

Staff Review of Site Response and Design Ground Motion

Use of the RG 1.60 spectrum (NRC, 1973) anchored at 0.20 g PGA is deemed acceptable by

the staff. Analyses performed by the applicant of several SSCs using the RG 1.60 design

spectrum demonstrate that the performance objectives necessary for highly unlikely events with

potentially high consequences, set forth in NUREG–1718 (NRC, 2000) are met. Similarly, the

analysis procedures, input bedrock time histories, and soil column properties for soil stability

analyses and soil-structure interaction analyses are also deemed acceptable by the staff. The

resulting surface response spectrum exceeds the DOE PC–3 spectrum and is comparable to

the design spectrum. The applicant further verified that the SRS-wide PC–3 spectrum is

applicable to the design of the MFFF through examinations of the soil stratigraphy, soil column

thickness, bedrock type, velocity profile, and geologic formations at the MFFF. Using response

spectra that envelop the SRS-wide PC–3 spectrum for the analysis of soil and subsurface

stability of the MFFF is therefore conservative.

1.3.1.5.6 Surface Faulting Hazard

The revised CAR summarizes the tectonic structures of interest in the SRS and surrounding

region, including faults, folds, arches, basins, and paleoliquefaction features that resulted from

past earthquakes. Many of these features are vestiges of the contractional tectonism that

characterized the Appalachian Orogen from the Late Precambrian through the Late Paleozoic

(~1.1 Ga to ~245 Ma) and rifting and extensional tectonism that characterized the break up of

Pangea and the opening of the Atlantic Ocean in the Triassic and Early Jurassic periods

(~245 to 180 Ma). Although reactivation of some of these features has been proposed to

explain the origin of the Charleston-type earthquake (see discussion in Section 1.3.1.5.6 of the

FSER), none of these features impact direct-faulting hazards at the SRS. 

Faulting of the Atlantic Coastal Plain sediments is evident from geologic and geophysical data

(e.g., Prowell, 1991). Most of the faults are moderately to steeply dipping reverse faults,

although some small normal faults were noted in the Late Cretaceous and Early Tertiary strata

(100 to 37 Ma). Maximum displacements are less than 80 m (250 ft), and displacements

become progressively smaller in younger sediments suggesting that faulting was coeval with

deposition. 

At the SRS, the Pen Branch fault has been identified as the primary structural feature of interest

to a potential faulting hazard. This fault appears to be an upward propagation of the boundary

fault on the northern side of buried Dunbarton Basin, a Triassic to Early Jurassic rift feature.

This boundary fault was originally a down-to-the-southeast normal fault, but was reactivated as

an up-to-the-southeast reverse fault in the Late Cretaceous and Early Tertiary (100 to 37 Ma).

Extensive geological and geophysical evidence summarized in the revised CAR documents that

the Pen Branch fault was not active in the last 500,000 years and probably was not active in the

Quaternary (last ~ 2 Ma). Thus, the Pen Branch fault is not deemed capable according to

criteria established in Appendix A to 10 CFR Part 100.
1-24

Staff Review of Surface Faulting Hazard

The staff reviewed the information in the revised CAR and found it acceptable because the

potential for surface faulting of the site and vicinity has been adequately assessed. There is

sufficient evidence to conclude, with reasonable assurance, that surface faulting hazards do not

exist at the SRS.

1.3.1.6 Stability of Subsurface Materials

The objective of the staff review in this section is to determine, with reasonable assurance,

whether characterization of the stability of the subsurface materials for the facility is adequate

for foundation design for the civil structural systems. The following areas concerning the

subsurface material stability which were applicable to the safety analysis and design of the

proposed facility were reviewed:

• soil liquefaction potential assessment

• soft zone characterization

• slope stability assessment

1.3.1.6.1 Soil Liquefaction Potential Assessment

The information regarding the paleoliquefaction at the SRS where the proposed facility will be

located was provided in Section 1.3.5.3.4.3, “Post-Rift and Cenozoic Structures,” of the revised

CAR. The revised CAR indicates that no systematic reconnaissance surveys in search of

paleoliquefaction evidence within the geomorphic and geologic environment of the SRS were

performed in the past because of limited access, high water table conditions, dense vegetative

cover, and few exposures. 

For seismically induced liquefaction to occur and be identified, the following conditions have to

be met (DCS, 2002a):

• presence of Quaternary-age unconsolidated deposits

• presence of a shallow ground water table

• proximity to potential seismogenic features

• quality and extent of exposure

According to these conditions, young fluvial terraces at or slightly above the level of the modern

flood plain and Carolina bays may have the highest potential for generating and recording

Holocene (last 10,000 years) and Quaternary (~ last 2 Ma) seismically induced liquefaction. 

Limited investigation of the exposed young fluvial terraces along the Savannah River adjacent

to the SRS suggests that most of the exposed deposits were clay and silt and thus have a low

liquefaction potential. Although local clean sand deposits with a high liquefaction potential

exist, evidence about the seismically induced liquefaction is not observed (DCS, 2002c). In

general, these young fluvial deposits are historical in age. In historical times, no strong ground

motions occurred in the SRS area. Consequently, evidence for seismically induced liquefaction

in the young fluvial deposits may not exist.
1-25

According to the revised CAR, potential paleoliquefaction for the flood plain deposits at depth is

likely. Evaluation of postdepositional features associated with the upland areas at the SRS,

however, suggests that they are not related to seismically induced liquefaction (DCS, 2002a).

Liquefaction susceptibility at the facility was discussed in Section 1.3.7.1 of the revised CAR. 

The discussion was supported by detailed soil geotechnical testing data, as documented in

three facility site geotechnical reports (DCS, 2001b; DCS 2001a; and DCS, 2003b). The site

geotechnical reports present properties of soils, including soil classifications, particle-size

distributions, water contents, plasticity indices, liquid limits, blow counts from standard

penetration tests, tip shear resistances from cone penetration tests, and shear wave velocities.

The liquefaction potential of the facility site within the proximity of the MFFF and emergency

generator buildings was evaluated using the cyclic stress approach described by the National

Center for Earthquake Engineering Research (NCEER, 1997). This approach is acceptable to

the staff for a liquefaction potential investigation because it represents the state-of-the-art

procedure. This procedure is suited for evaluating liquefaction resistance of soils under level to

gently sloping ground; the surface gradient at the proposed facility is gently sloping (as shown

in Figure 1.3.1-2 of the revised CAR).

In the third geotechnical report (DCS, 2003b), liquefaction potential was evaluated for the 95

soil columns from cone penetration tests and 14 soil columns from standard penetration tests. 

Cyclic stress ratio and cyclic resistance ratio are two important parameters for assessing

liquefaction. The cyclic stress ratios for the aforementioned penetration tests were estimated

directly using the PROSHAKE computer program with a generalized soil profile. The

liquefaction potential for 18 of the cone penetration tests was also estimated using the

test-hole-specific soil profiles to compare with the results using the generalized soil profile. The

revised CAR assumed full liquefaction was triggered if the factor of safety (cyclic stress

ratio/cyclic resistance ratio) was equal to or smaller than 1.1. For factors of safety between

1.1 and 1.4, soil settlement may result because of the excessive water pressure buildup that

reduces soil strength and stiffness.

The analysis results indicate that the liquefaction potential at the facility site is low. Only a few

localized areas have been identified to be liquefiable or to have soil settlement potential

because of excessive pore water pressure. The potentially liquefiable soils identified at the site

are located in the lower Tertiary (~65 to 33 Ma) Tobacco Road, Dry Branch, and Santee

formations. The revised CAR indicated that the analysis results were conservative because the

effect of soil aging and the cohesiveness of the soils for the cone-penetration-based results are

not considered in the analysis.

Staff Review of Soil Liquefaction Potential Assessment

The staff reviewed the information presented in the revised CAR and found reasonable

assurance that paleoliquefaction at the SRS was sufficiently discussed to support the design of

the PSSCs of the proposed facility. The staff review also concurred that the analysis of

liquefaction potential at the proximity of the MFFF and emergency generator buildings

demonstrated a conservative approach and is acceptable. Consideration of the effect of the

seismically induced settlements caused by either liquefaction or excessive water pressure
1-26

buildup in the development of design criteria for the PSSCs of the facility are evaluated in

Section 11.1 of this FSER.

1.3.1.6.2 Soft Zone Characterization

Soft zones in the soils are unique features for the SRS. The origin of the soft zones is

discussed in Section 1.3.5.1.5.5, “Carolina Bays,” of the revised CAR. The discussion about

the characterization of the soft zones at the facility is provided in Section 1.3.7.2, “Evaluation of

Soft Zones,” of the revised CAR and supported by site geotechnical data (DCS, 2001a; DCS

2003b).

The soft zones are often found in the Tinker/Santee Formation, particularly in the upper third of

this section. These soft zones consist of weak material zones interspersed in stronger

carbonate-rich matrix materials. The presence of soft zones may pose a concern for foundation

design by developing undesirable soil settlement not accounted for in the design. In

engineering terms, a soft zone is defined as a zone with a cone penetration test corrected tip

resistance less than 1.44 MPa (15 tsf) or blow counts from a standard penetration test less

than 5 over a continuous interval of at least 0.6 m (2 ft) (DCS, 2001a). In characterizing the

soft zones, the applicant used these criteria to identify soft material zones not located in the

Tinker/Santee Formation. The staff considered this approach prudent and acceptable.

The results of the site exploration program related to identifying soft material zones were

documented in three site geotechnical reports (DCS, 2001b; DCS, 2001a; DCS, 2003b). The

exploration hole spacing in the vicinity of an identified soft zone was generally 27 m (90 ft) or

less. The lateral extent of soft zones was conveniently estimated to be half of the exploration

spacing. The exploration program identified soft zones in the vicinity and beneath the MOX fuel

fabrication building with limited lateral extent. The thickness of these soft zones ranges from

0.91 to 2.13 m (3 to 7 ft).

Staff Review of Soft Zone Characterization

Based on the review of the information concerning soft zones, the staff concluded the

exploration program conducted by the applicant sufficiently characterized the soft zones at the

facility to support design of the PSSCs. Consideration of the effect of the soft zones in the

development of design criteria for the PSSCs of the facility is evaluated in Section 11.1 of this

FSER.

1.3.1.6.3 Slope Stability

Slope stability was not specifically discussed in Section 1.3, “General Site Description,” of the

revised CAR. In evaluating the natural phenomena applicable to the site, however, debris

avalanching and landslides were determined not to be applicable to the site because the site is

relatively flat and no significant quantities of soil or rock are available in the surrounding area

(see Table 5.5-5 in the revised CAR). An examination of topographic contours provided in

Figure 1.3.4-6 confirms that the slopes at the facility site are relatively gentle in nature and

therefore pose no threat for instability or landslide. The staff site visit further confirmed that

slope stability is not a safety concern for the site.
1-27

1.3.2 Evaluation Findings

In Chapter 1 of its revised CAR, and in supplementary information, the applicant provided the

geographic, demographic and land use, meteorologic, hydrologic, geologic, and seismic

information relevant to the MFFF site. This information is generally current, appropriately

referenced, and consistent with information in the safety assessments used to support the

design bases of the PSSCs. The staff finds that the applicant has accurately described the site

so as to allow potential accident conditions to be properly defined. Based on its review of the

revised CAR and the relevant supplementary information provided by the applicant, the staff

further finds that DCS has met the BDC set forth in 10 CFR 70.64(a)(2) for natural phenomena

hazards. The staff concludes, pursuant to 10 CFR 70.23(b), that the design bases of the

PSSCs evaluated in this FSER will provide reasonable assurance of protection against natural

phenomena and the consequences of potential accidents. 

1.3.3 References

(Ambraseys, 1988) Ambraseys, N.N. “Engineering Seismology,” Earthquake Engineering and

Structural Dynamics 17:1–105. 1988.

(Bollinger, 1992) Bollinger, G.A. “Specification of Source Zones, Recurrence Rates, Focal

Depths, and Maximum Magnitudes for Earthquakes Affecting the Savannah River Site in South

Carolina.” U.S. Geological Survey Bulletin, 2017:1992.

(Boore, 1983) Boore, D.M. “Stochastic Simulation of High-Frequency Ground Motions Based

on Seismological Models of the Radiated Spectra,” Bulletin of the Seismological Society of

America, 73:1865–1894. 1983.

(BNL, 1996) Brookhaven National Laboratory, “Description and Validation of the Stochastic

Ground Motion Mode,” Upton, New York, 1996.

(Campbell, 2003) Campbell, K. “Prediction of strong ground motion using the hybrid empirical

method and its use in the development of ground-motion (attenuation) relations in eastern

North America.” Bulletin of the Seismological Society of America, 93:1012–1033. 2003.

(CNWRA, 1999) Center for Nuclear Waste Regulatory Analyses. “Review of the Probabilistic

Seismic Hazard Analyses for the Paducah Gaseous Diffusion Plant—Final Report.”

San Antonio, Texas. 1999.

(DCS, 2001a) Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory Commission,

RE: MOX Fuel Fabrication Facility CAR Supplemental Information—Site Geotechnical Report

(DCS01–WRS–DS–NTE–G–00005– A). April 24, 2001.

(DCS, 2001b) Ihde, R., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: MOX Fuel Fabrication Facility Site Geotechnical Report

(DCS01–WRS–DS–NTE–G–00005–C). August 10, 2001. 

(DCS, 2002a) Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory Commission,

RE: Clarification of Responses to NRC Request for Additional Information. March 8, 2002.
1-28

(DCS, 2002b) Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory Commission,

RE: Clarification of Responses to NRC Request for Additional Information. April 23, 2002.

(DCS, 2002c) Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory Commission,

RE: Submitting Revised Request for Authorization of Construction of the Mixed Oxide Fuel

Fabrication Facility. October 31, 2002 (including page changes through February 9, 2005).

(DCS, 2003a) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear

Regulatory Commission, RE: Responses to Open Items/Additional NRC Questions on

Construction Authorization Request Revision. February 11, 2003.

(DCS, 2003b) Duke Cogema Stone & Webster. MOX Fuel Fabrication Facility Site

Geotechnical Report (DCS01–WRS–DS–NTE–G–00005–E). Report prepared for the U.S.

Department of Energy, Chicago Operations Office. June 30, 2003. 

(DOE, 1994) U.S. Department of Energy. DOE–STD–1022–94, “Natural Phenomena Hazards

Characterization Criteria.” Washington, DC, 1994.

(DOE, 1995) U.S. Department of Energy. DOE–STD–1023–95 Change Notice #1, “Natural

Phenomena Hazards Assessment Criteria.” Washington, DC, 1995.

(DOE, 1996) U.S. Department of Energy. DOE-STD–1020–94 Change Notice #1, “Natural

Phenomena Hazards Design and Evaluation Criteria for Department of Energy Facilities.” 

Washington, DC, 1996.

(DOE, 2002) U.S. Department of Energy. DOE–STD–1020–2002, “Natural Phenomena

Hazards Design and Evaluation Criteria for Departments of Energy Facilities.” Washington, DC,

2002.

(EPRI, 1989) Electric Power Research Institute. “Probabilistic Seismic Hazard Evaluations at

Nuclear Power Plants in the Central and Eastern United States: Resolution of the Charleston

Earthquake Issue,” NP–6395–D. Palo Alto, California, 1989.

(Fledderman, 2002) Fledderman, P. D., Westinghouse Savannah River Company. “Plutonium

Disposition Program (PDP) Preconstruction Environmental Monitoring Report,”

ESH–EMS–2002–1141. June 26, 2002.

(Hanks, 1981) Hanks, T.C. and R.K. McGuire. “The Character of High-Frequency Strong

Ground Motion.” Bulletin of the Seismological Society of America. 71:2071–2095, 1981.

(Herrmann, 1986) Herrmann, R.B. “Surface-Wave Studies of Some South Carolina

Earthquakes.” Bulletin of the Seismological Society of America. 76:111–121, 1986.

(Hu, 2002) Hu, K., S.L. Glassman, and P. Talwani. “Magnitudes of prehistoric earthquakes in

the South Carolina Coastal Plain from geotechnical data.” Seismological Research Letters.

73:979–991, 2002.
1-29

(Idriss, 1992) Idriss, I.M. and J.I. Sun. User’s Manual for SHAKE-91: A Computer Program for

Conducting Equivalent Linear Seismic Response Analysis of Horizontally Layered Soil Deposits.

Center for Geotechnical Modeling, Department of Civil and Environmental Engineering,

University of California, Davis, California, 1992.

(Jannik, 1995) Jannik, G.T., Westinghouse Savannah River Company. “Concentration

Guidelines for the Initial Screening of Soil in Determining Onsite ‘Soil Concentration Areas,’” 

WSRC–SRT–ETS–950155. 1995.

(Johnston, 1996) Johnston, A.C. “Seismic Moment Assessment of Earthquakes in Stable

Continental Regions, III, New Madrid 1811–1812, Charleston 1886, and Lisbon 1755.” 

Geophysical Journal International. 126:314–344, 1996.

(Madabhushi, 1993) Madabhushi, S. and P. Talwani. “Fault Plane Solutions and Relocations of

Recent Earthquakes in Middleton Place Summerville Seismic Zone Near Charleston, South

Carolina,” Bulletin of the Seismological Society of America. 83:1442–1466, 1993.

(Marple, 1994) Marple, R.T. Discovery of a Possible Seismogenic Fault System Beneath the

Coastal Plain of South and North Carolina from an Integration of River Morphology and

Geological and Geophysical Data. Ph.D. Dissertation. University of South Carolina. Columbia,

South Carolina, 1994.

(Marple, 2000) Marple, R.T. and P. Talwani. “Evidence for a Buried Fault System in the Coastal

Plain of the Carolinas and Virginia—Implications for Neotectonics in the Southeastern United

States.“ Geological Society of America Bulletin. 112:200–220, 2000.

(NCEER, 1997) National Center for Earthquake Engineering Research. Technical Report

No. NCE ER–97–002. ”Proceedings of the National Center for Earthquake Engineering

Research Workshop on Evaluation of Liquefaction Resistance of Soils.” State University of New

York at Buffalo. Buffalo, New York, 1997.

(NRC, 1973) U.S. Nuclear Regulatory Commission., Regulatory Guide 1.60. “Design

Response Spectra for Seismic Design of Nuclear Power Plants.” Washington, DC, 1973.

(NRC, 1987) U.S. Nuclear Regulatory Commission. NUREG–0800. “Standard Review Plan for

the Review of Safety Analysis Reports for Nuclear Power Plants.” Washington, DC, 1987. 

(NRC, 1994) U.S. Nuclear Regulatory Commission. NUREG–1488 “Revised Livermore

Seismic Hazard Estimates for Sixty Nine Nuclear Power Plant Sites East of the Rocky

Mountains.” Washington, DC, 1994.

(NRC, 1997) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.165. “Identification

and Characterization of Seismic Sources and Determination of Safe Shutdown Earthquake

Ground Motion.” Washington, DC, 1997.

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG–1718. “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility.” Washington,

DC, 2000.
1-30

(Prowell, 1991) Prowell, D.C. and S. F. Obermeier. Evidence of Cenozoic Tectonism. The

Geology of the Carolinas. University of Tennessee Press, Knoxville, Tennessee, 1991.

(Silva, 1989) Silva, W., et al. Site Dependent Specification of Strong Ground Motion in Dynamic

Soil Property and Site Characterization. Workshop sponsored by the National Science

Foundation and the Electric Power Research Institute. Palo Alto, California, 1989.

(Talwani, 2001) Talwani, P. and W.T. Schaeffer. “Recurrence Rates of Large Earthquakes in

South Carolina Costal Plain Based on Paleoliquefaction Data.” Journal of Geophysical

Research. 106:6621–6642, 2001. 

(Tarr, 1981) Tarr, A., et al. “Results of Recent South Carolina Seismological Studies.” 

Seismological Research Letters. 71:1883–1902, 1981.

(Tuttle, 2001) Tuttle, M.P. “The use of liquefaction features in paleoseismology: lessons

learned in the New Madrid seismic zone, central United States.” Journal of Seismology.

5:361–380, 2001.

(Weems, 2002) Weems, R.E. and W.C. Lewis. “Structural and Tectonic Setting of the

Charleston, South Carolina, Region: Evidence from the Tertiary Stratigraphic Record.” 

Geological Society of America Bulletin. 114:24–42, 2002.

(WSRC, 1997) Westinghouse Savannah River Company. WSRC–TR–97–0085, Revision 0,

“SRS Seismic Response Analysis and Design Basis Guidelines.” Aiken, South Carolina, 1997.

(WSRC, 2000a) Westinghouse Savannah River Company. WSRC–TR–2000 00454, Revision

0, “Natural Phenomena Hazards (NPH) and Design Criteria and Other Characterization

Information for the Mixed Oxide (MOX) Fuel Fabrication Facility at Savannah River Site (U).”

Aiken, South Carolina, 2000.

(WSRC, 2000b) Westinghouse Savannah River Company. Report WSRC–TR–2000–00039,

Revision 0. “Probability of Liquefaction for F-Area, Savannah River Site.” Aiken, South Carolina,

2000.

(WSRC, 2001) Westinghouse Savannah River Company. WSRC–TR–2001–00342, Revision

0, “Development of MFFF-Specific Vertical-to-Horizontal Seismic Spectral Ratios.” Aiken, South

Carolina, 2001.
2-1

2. FINANCIAL QUALIFICATIONS

2.1 Conduct of Review

This chapter of the final safety evaluation report contains the staff’s review of the financial

qualifications presented by Duke Cogema Stone & Webster (DCS or the applicant) in Chapter 2

of the revised construction authorization request (CAR). Additionally, the staff evaluated the

supplemental financial qualification information provided by the applicant (DCS, 2001, 2002, and

2003).

DCS is a limited liability company registered in the State of South Carolina and headquartered in

Charlotte, North Carolina. DCS is a consortium of Duke Project Services Group, Inc. (DPSG),

COGEMA, and Stone & Webster, Inc. The applicant is proposing to construct the Mixed Oxide

(MOX) Fuel Fabrication Facility (MFFF or the facility) at the U.S. Department of Energy’s (DOE)

Savannah River Site near Aiken, South Carolina, under a cost-plus-fixed-fee contract with DOE.

COGEMA, a French company, owns 30 percent of DCS. The remainder of the corporation is

owned by DPSG (40 percent) and Stone & Webster, Inc. (30 percent), which are both U.S.

companies. 

2.1.1 Project Costs

In an August 31, 2001, proprietary response (DCS, 2001) to a staff request for additional

information (RAI), the applicant provided financial statements for its fiscal year ending December

31, 2000, and an estimate of project costs. These costs include costs for facility design and

construction. Design costs include licensing costs, contingencies, and escalation. The

applicant, however, did not detail the project design cost estimate, which was provided as a

single value. Although DCS is not a publicly held company—and is thus not required to submit

Report 10-K to the U.S. Security and Exchange Commission—it provided, as part of its

proprietary filing, an independent accountant’s review report for the fiscal year ending December

31, 2000, and for the period of corporate inception (March 22, 1999) to December 31, 1999

(DCS, 2001). Balance sheets, cash flows, income statements, and equity statements were

included in the accountant’s report. The independent accountant concluded that it was not

aware of any material modifications that should be made to the accompanying financial

statements in order for them to be in conformity with accounting principles generally accepted in

the United States.

In its March 8, 2002, clarification to the RAI responses (DCS, 2002), DCS stated that revised

proprietary design cost information would be provided at a later date when available. In its initial

draft safety evaluation report (NRC, 2002), the staff identified this as Open Item FQ-1. On

February 13, 2003, NRC staff requested current financial statements and a commitment to

provide annual updates of this information (NRC, 2003). In its subsequent February 18, 2003,

filing (DCS, 2003), the applicant retracted its earlier commitment to provide revised proprietary

design cost information.

2.1.2 Financial Qualifications

As indicated above, the applicant has a contract with DOE to design, construct, and operate the

facility. The base scope of work of the contract covers design and engineering of the facility. An
2-2

option, which has not yet been exercised, includes construction, functional testing, and

preliminary startup. A second option, also not yet exercised, includes final startup and full

operations. All funding for the facility will be from DOE. The applicant does not plan to selffinance the facility or seek external funding other than that under the DOE contract. In the event

of cost overruns or funding shortfalls, the applicant would seek additional funding from DOE. If

such funds were not provided, the applicant would stop any ongoing engineering, design, and

construction activities.

As noted by DCS (DCS, 2003), the proposed MFFF is part of a fully funded U.S. Government

project undertaken to carry out the national security mission to dispose of surplus plutonium. 

Other than temporary startup funding provided as part of DCS organizing as a company, DCS is

not relying on its financial resources to provide money to cover engineering or construction

costs. Instead, the funds necessary to cover such costs are provided directly through the DCS

contract with the U.S. Government. Accordingly, DCS now takes the position that its financial

qualifications are not relevant to the ability of the U.S. Government to carry out and fund the

mission of surplus plutonium disposition (DCS, 2003). 

2.1.3 Liability Insurance

Under Title 10, Section 140.13a, of the Code of Federal Regulations (10 CFR 140.13a), a holder

of a license to possess and use plutonium at a plutonium processing and fuel fabrication plant is

required to have and maintain public liability insurance in the amount of $200 million. This

insurance is not required for MOX facility construction approval. The applicant stated that as a

DOE contractor it is fully covered by the DOE nuclear liability protection under the PriceAnderson Act, as amended. Under Section 170d of the Atomic Energy Act, 42 U.S.C. 2210(d),

the applicant and DOE have entered into an agreement that fully indemnifies the applicant and

its contractors up to the statutory limit of liability. The applicant intends to request an exemption

to the requirements of 10 CFR 140.13a in any possession and use license application it may

later submit, and would use the DOE indemnification to meet its public liability insurance

requirements.

2.2 Evaluation Findings

Based on its review of the requirements of 10 CFR Part 70, “Domestic Licensing of Special

Nuclear Material,” the staff has concluded that a finding regarding the adequacy of the DCS

financial qualifications is not required in deciding whether the CAR should be approved. As a

general matter, the issue of an applicant’s financial qualifications may be relevant in considering

applications for licenses to possess and use special nuclear material (see 10 CFR 70.23(a)(5)).

However, this requirement is not applicable in deciding whether to approve the CAR because the

CAR is not an application for a license to possess and use special nuclear material. Similarly,

the financial assurance for decommissioning requirements, found in 10 CFR 70.22(a)(9), 10

CFR 70.25(b)(2), and 10 CFR 70.25(f)(4–5), are not applicable in deciding whether to approve

the CAR. These requirements also apply only to applications for licenses to possess and use

special nuclear material. 

Pursuant to the above requirements, the staff may consider financial qualification issues as part

of reviewing any application DCS may later submit for a license to possess and use licensed

material. However, as noted by DCS (DCS, 2003), any funds necessary to cover MFFF-related
2-3

costs would ultimately come from the U.S. Government. The staff need not decide these

financial qualification issues now.

As discussed above in Section 2.1.3, as part of any possession and use license application it

may later submit, DCS plans to request an exemption from the liability insurance requirements of

10 CFR 140.13a. The staff would consider any such exemption request in reviewing a DCS 

application for a license to possess and use licensed material at the proposed MFFF.

2.3 References

(DCS, 2001) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Proprietary DCS Financial Information. August 31, 2001.

(DCS, 2002) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Clarification of Responses to NRC Request for Additional Information. March

8, 2002.

(DCS, 2003) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Responses to Financial Qualification, Fire Safety, Chemical Safety, Aqueous

Processing, Material Processing and Ventilation Open Items/Additional NRC Questions on

Construction Authorization Request (CAR) Revision. February 18, 2003.

(NRC, 2002) U.S. Nuclear Regulatory Commission. “Draft Safety Evaluation Report on the

Construction Authorization Request for the Mixed Oxide Fuel Fabrication Facility.” Washington,

DC, April 30, 2002.

(NRC, 2003) Persinko, A., U.S. Nuclear Regulatory Commission. Letter to Duke COGEMA

Stone & Webster, RE: February 2003 Monthly Open Item Status Report. February 13, 2003.
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3. PROTECTION OF CLASSIFIED MATTER

3.1 Conduct of Review

This chapter of the final safety evaluation report pertains to protection of classified matter and

whether the applicant’s treatment of classified matter will be sufficient to assure that it is

adequately protected. The staff evaluated the information provided by the applicant in Chapter 3

of the revised construction authorization request (CAR), as well as supplementary information

provided by the applicant. 

The staff reviewed how the applicant has addressed the following regulations:

• Title 10, Part 25, “Access Authorization for Licensee Personnel,” of the Code of Federal

Regulations (10 CFR Part 25) specifies the requirements granting and terminating

access authorizations to licensee personnel and its contractors who may require access

to classified information. 

• 10 CFR Part 95, “Facility Security Clearance and Safeguarding of National Security

Information and Restricted Data,” specifies the requirements for obtaining security facility

approval and for safeguarding secret and confidential national security information and

restricted data.

The staff used Chapter 3 in NUREG-1718, “Standard Review Plan for the Review of an

Application for a Mixed Oxide Fuel Fabrication Facility,” as guidance in performing the review.

Chapter 3 of the revised CAR states that the applicant is handling classified matter in

accordance with applicable U.S. Department of Energy (DOE) requirements and that it will

submit its standard practice procedures plan for the protection of classified matter along with its

application for a license to possess and use licensed material.

The staff reviewed the applicant’s commitment to provide its standard practice procedures plan

for the protection of classified matter along with its license application and concludes that the

revised CAR does not need to address procedures for the protection of classified matter

because the design basis of the mixed oxide fuel fabrication facility, need not be evaluated for

the protection of classified matter. However, the applicant will be required to submit a standard

practice procedures plan for the protection of classified matter in accordance with the

requirements of 10 CFR Part 25 and 10 CFR Part 95 as part of its license application.

The foreign ownership, control, or influence determination is discussed in Section 1.2.1.2 of the

FSER.

3.2 Evaluation Findings

The staff concludes that it is acceptable for the applicant to exclude its standard practice

procedures plan for the protection of classified matter as part of the revised CAR because it will

be submitted along with the license application. The staff also concludes that it is acceptable for

the applicant to handle classified matter based on a memorandum of understanding between the

U.S. Nuclear Regulatory Commission and DOE dated October 9, 1996.
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3.3 References

(NRC, 1999) U.S. Nuclear Regulatory Commission. “Standard Practice Procedures Plan

Standard Format and Content for the Protection of Classified Matter for NRC Licensee,

Certificate Holder and Others Regulated by the Commission.” Washington, DC, October 1999,

as revised.
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4. ORGANIZATION AND ADMINISTRATION

4.1 Conduct of Review

This chapter of the final safety evaluation report (FSER) reviews the organization and

administration information presented in Chapter 4 of the revised construction authorization

request (CAR). The staff used Chapter 4 in NUREG-1718, “Standard Review Plan for the

Review of an Application for a Mixed Oxide Fuel Fabrication Facility,” (NRC, 2000), as guidance

in performing the review. The objective of the review is to determine whether organizational and

administrative functions have been identified which will enable the applicant to plan, implement,

and control site activities in a manner that adequately ensures the safety of workers and

individuals outside the controlled area, and which will also protect the environment. This review

ensures that the qualifications for key management positions are adequate. Organizational

information is also described in the applicant’s quality assurance plan, which is discussed in

FSER Chapter 15.

4.1.1 Organization

The applicant proposed a functional organization for engineering, design, and construction that

includes lines of responsibility and control of engineering, design, construction, procurement,

administrative services, environmental health and safety, licensing, and quality assurance. 

Proposed organization charts showing lines of responsibility and communications were provided.

The President of Duke Cogema Stone & Webster (DCS or the applicant) has overall

responsibility for the project. Reporting to the president are managers responsible for licensing

and safety analysis; environment, safety, and health; facility engineering; construction

management; procurement; plant operations and startup; project services and administration;

site integration; and quality assurance. The quality assurance manager has a direct line of

communication to the president and is unencumbered with responsibilities for costs or

schedules. 

Management of process and facility design will be provided directly by the applicant. The

applicant will be responsible for planning, managing, and controlling construction activities. 

Construction subcontractors will be managed by the construction management manager. At this

time, no subcontractors have been selected for construction activities.

The organizational information provided by the applicant sets forth clear and unambiguous

controls and communications between organizational groups responsible for designing and

constructing the facility. Lines of communication, responsibility, and authority are clearly

delineated in the organization chart. The president has overall responsibility for all design and

construction activities.

4.1.2 Administration

As construction of the mixed oxide fuel fabrication facility proceeds, the applicant will transition

from the design and construction phase of the proposed facility to the operating phase. During

any construction of the proposed facility, the construction management manager would be

assigned a greater work scope and level of resources. Concomitant with the increased scope of
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work for the construction management manager, a decreased work scope and level of resources

would be assigned to the facilities design and process design managers. Toward the completion

of any potential construction activities, the construction organization would shift to a

manufacturing organization consisting of a corporate-level plant manager responsible for safety,

plant operations, maintenance, and quality assurance. Reporting to the plant manager would be

an operations manager, a regulatory management manager, a security manager, an

administration manager, and a quality assurance manager. If plant systems are completed,

acceptance tests would be performed followed by turnover to the operations organization. The

operations organization would then be responsible for maintenance and configuration

management of the systems. 

Formal management controls described in Chapter 15, of this FSER would be applied to ensure

that there is reasonable assurance that design bases are maintained during any transition

between design and construction activities and operations.

4.1.3 Key Management Positions

The management positions described in Section 4.1.1, above, have responsibilities for activities

involving the proposed facility. The applicant also provided the minimum qualifications for each

of these positions. 

The scope and number of each key management position are described appropriately for every

management function involving the proposed facility. The qualification requirements for key

management positions provide an adequate breadth and level of experience for their respective

responsibilities and authorities. The staff filling key management positions will be available

during the design and potential construction phases of the project.

4.2 Evaluation Findings

The staff reviewed the organization and administration for construction approval of the proposed

facility in accordance with Chapter 4 of NUREG-1718 (NRC, 2000). The staff evaluated the

proposed organization for design and construction; the administration of the project, including

how the project would transition from design and construction to operations; and the

responsibilities, qualifications, and authorities of key management positions. The proposed

organization, administration, and key management position descriptions and qualifications are

consistent with guidance in NUREG-1718 and are, therefore, acceptable.

The issue of whether the design bases of the principal structures, systems, and components of

the proposed facility would provide reasonable assurance of protection against natural

phenomena and the consequences of potential accidents are addressed in other chapters of this

FSER.

4.3 References

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide Fuel Fabrication Facility.” Washington, DC,

2000.
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5. SAFETY ASSESSMENT OF THE DESIGN BASES

5.1 Conduct of Review

This chapter of the final safety evaluation report (FSER) contains the U.S. Nuclear Regulatory

Commission (NRC) staff’s review of the safety assessment of the design bases of the principal

structures, systems, and components (PSSCs) performed by Duke Cogema Stone & Webster

(DCS or the applicant) in Chapter 5 of the revised construction authorization request (CAR)

(DCS, 2002b). The objective of this review is to determine whether the PSSCs and their design

bases identified by the applicant provide reasonable assurance of protection against natural

phenomena and the consequences of potential accidents. The staff evaluated the information

provided by the applicant by reviewing Chapter 5 of the revised CAR, other sections of the

revised CAR, supplementary information provided by the applicant, and relevant documents

available at the applicant’s offices, but not submitted by the applicant. The review of PSSCs and

their design bases and strategies was closely coordinated with the review of evaluations

performed in other chapters of the FSER. 

The staff reviewed how the safety assessment information in the revised CAR addresses or

relates to the following regulations:

• Title 10, Section 70.23(b), of the Code of Federal Regulations (10 CFR 70.23(b)) (CFR,

2004b) states, as a prerequisite to construction approval, that “the design bases of the

PSSCs and the quality assurance program be found to provide reasonable assurance of

protection against natural phenomena and the consequences of potential accidents.”

• 10 CFR 70.61(b) and 10 CFR 70.61(c) set forth performance requirements addressing

specified high-consequence events and intermediate-consequence events. For such

events, controls must be identified and eventually implemented sufficient to either lessen

the likelihood that such events will occur, and/or make the consequences of such events

less severe. 

• 10 CFR 70.62(c)(2) requires the applicant to have a team with expertise in engineering

and process operations. The team must have at least one person who has experience

and knowledge specific to each process being evaluated, as well as persons who have

experience in nuclear criticality safety, radiation safety, fire safety, and chemical process

safety. One member of the team must be knowledgeable in the specific analysis

methodology being used. The guidance in Section 5.4.3.1 of NUREG-1718, “Standard

Review Plan for the Review of an Application for the Construction of a Mixed Oxide Fuel

Fabrication Facility,” (NRC, 2000) recommends that a review of team qualifications be

made during the safety assessment of the design bases (construction authorization

review), as well as during review of the application for a license to possess and use

licensed material.

• 10 CFR 70.64, “Requirements for New Facilities or New Processes at Existing Facilities,”

requires that baseline design criteria (BDC) and defense-in-depth practices be

incorporated into the design of new facilities. It specifically addresses quality standards;

natural phenomena hazards; fire protection; environmental conditions and dynamic
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effects; chemical protection; emergency capability; inspection, testing, and maintenance;

criticality control; and instrumentation and controls. 

The staff used Chapter 5 in NUREG-1718 as guidance in performing the review. NUREG-1718

states the following:

The steps the applicant follows to develop the safety assessment for the design

bases should be analogous to the steps that the applicant will use to develop the

ISA; however the reviewer should expect the application of these steps to be

adjusted according to the level of design when the applicant applies for

construction approval. 

NUREG-1718 also states that the description of PSSCs should include “the functional

relationship of each principal SSC to the top-level safety function for a process....”

The review for this construction approval focused on the design basis of systems, their

components, and other related information. For each PSSC, the staff reviewed information

provided by the applicant for the safety function, system description, and safety analysis. The

review also encompassed proposed design basis considerations, such as redundancy,

independence, reliability, and quality. The staff reviewed descriptions of the systems to assure

that the facility can be designed to meet the performance requirements of 10 CFR 70.61,

Performance Requirements,” during operation of the mixed oxide (MOX) fuel fabrication facility

(MFFF or the facility). Much of the review was directed at the applicant’s hazard assessment

(including natural phenomena and external manmade events), the formulation of a strategy and

identification of PSSCs to meet the performance requirements, and assuring that the design

bases of these PSSCs are adequate with regard to the performance requirements of the

regulation.

The safety assessment review was an integrated team approach. Team members with

expertise in the various areas of technical review, such as engineering, nuclear science, and

other disciplines, reviewed their respective revised CAR chapters, as well as the revised CAR

Chapter 5. These revised CAR chapters or discipline reviews often identified issues that were

referred to the safety assessment review at which point they were either resolved or carried as

open items based on the hazard assessment or performance strategies. 

5.1.1 Plant Site Description Relating to Safety Assessment of the Design Bases

The plant site description includes information to support the safety assessment of the design

bases, including the following:

• Site description. The level of detail should be sufficient to allow an evaluation of natural

phenomena and other external accidents. The site description is discussed in Section

1.3 of the FSER.

• Facility description. The level of detail should allow an understanding of the relationship

between the design bases of the PSSCs and the facility. The facility description is

discussed in Section 1.1 of the FSER.
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• Process description. The process description should provide sufficient detail to allow the

evaluation of the process design as it is established through the design bases. The

process description is discussed in Section 1.1 of the FSER.

5.1.2 Safety Assessment Team Description

The safety assessment team is described in Section 5.2 of the revised CAR. The safety

assessment team is described as a team of individuals experienced in hazard identification,

hazard evaluation techniques, accident analysis, including dose consequence assessment, and

probabilistic analysis. The team members possess operational experience at similar facilities,

specific discipline knowledge (e.g., mechanical; electrical; and heating, ventilation, and air

conditioning) and specific knowledge of the processes to be used in the facility. In addition, the

team has safety analysis experience that is MOX process and aqueous polishing (AP) specific. 

The integrated safety analysis (ISA) manager is described as having overall responsibility for

preparation of the safety assessment. The ISA manager reports to the facility’s licensing and

safety analysis manager. The ISA manager provides overall direction for the analysis, organizes

and executes safety analysis activities, and facilitates team meetings. The technical analysis

which supports the safety assessment is the responsibility of the ISA team leader who reports to

the ISA manager. The ISA team leader is knowledgeable in the specific safety assessment

methodologies chosen for the hazard and accident analyses and has an understanding of

process operations and the hazards under evaluation. Based on the above description, the staff

concludes that the applicant’s safety assessment team satisfies the criteria of Section 5.4.3.1 of

NUREG-1718 (NRC, 2000) at the construction authorization stage.

5.1.3 Chemical Standards and Consequences

DCS provided chemical concentration limits to evaluate the potential consequences to

individuals outside the controlled area and workers for an accidental release of chemicals. The

applicant based these limits on the acute exposure guideline level (AEGL) values and the

emergency response planning guideline (ERPG) values. For chemicals which do not have

AEGL or ERPG value, limits are based on temporary emergency exposure limits (TEELs)

adopted by the U.S. Department of Energy (DOE), Subcommittee on Consequence Assessment

and Protective Action. A discussion of the chemical consequences and the applicant’s

consequence analysis is provided in Section 8.1.2.3 of the FSER. A summary of the staff’s

review of chemical events is provided in Section 5.1.6.3.6 of the FSER.

5.1.4 10 CFR 70.61 Performance Requirements

As discussed in the introduction to the FSER, 10 CFR 70.61(b) sets forth the performance

requirements for “credible high-consequence events” (i.e., potential accidents involving high

levels of radiation or hazardous chemicals);, and its provisions pertain to the protection of both

onsite workers and offsite individuals. As required by 10 CFR 70.61(b), the use of controls

must be sufficient to either make the occurrence of such accidents “highly unlikely,” or make the

consequences of such accidents less severe than (1) an acute 1 Sv (100 rem) total effective

dose equivalent (TEDE) to a “worker,” as defined in 10 CFR 70.4, “Definitions,” (2) an acute 0.25

Sv (25 rem) TEDE to a person outside the “controlled area,” as defined in 10 CFR 20.1003,

“Definitions,” (3) an intake of 30 mg of soluble uranium to a person outside the controlled area,

or (4) either an acute chemical exposure that could endanger the life of a worker or an acute
5-4

chemical exposure that could lead to irreversible or other serious long-lasting health effects to a

person outside the controlled area. (See 10 CFR 70.61(b)(1–4)). 

As set forth in 10 CFR 70.61(c), the performance requirements for “credible intermediateconsequence events,” (i.e., potential accidents involving lower levels of radiation or hazardous

chemicals than referenced in 10 CFR 70.61(b)), and its provisions pertain to environmental

protection, as well as to protecting the health of onsite workers and offsite individuals. This

regulation also requires the use of controls sufficient to either make the occurrence of such

accidents “unlikely,” or to make the consequences of such accidents less severe than (1) an

acute 0.25 Sv (25 rem) TEDE to a worker, (2) an acute 0.05 Sv (5 rem) TEDE to a person

outside the controlled area, (3) a 24-hr average release of radioactive material outside the

“restricted area,” as defined in 10 CFR 20.1003, into the environment in concentrations

exceeding 5000 times the values in Table 2 of Appendix B, “Radiation Protection Programs,” to

10 CFR Part 20, “Standards for Protection Against Radiation,” or (4) either a chemical exposure

that could lead to irreversible or other serious long-lasting health effects to a worker or a

chemical exposure that could cause mild transient health effects to a person outside the

controlled area. (See 10 CFR 70.61(c)(1–4)). 

Under both 10 CFR 70.61(b) and 10 CFR 70.61(c), properly identifying and implementing the

required controls—which are later designated as items relied on for safety (IROFS) pursuant to

10 CFR 70.61(e) and 10 CFR 70.65(b)—ensures that an acceptable level of risk will be

maintained during any operation of the proposed facility. This goal is met through a combination

of limiting the chance that high-consequence or intermediate-consequence events would occur

(prevention) and reducing the consequences of such events (mitigation). 

The starting point for the applicant’s demonstration of acceptable control over the risk of credible

high-consequence and intermediate-consequence events, as well as the risk of nuclear criticality

accidents, is its safety assessment of the facility design bases. In Section 5.1.6 of the FSER,

the staff evaluates the hazards that have been addressed by DCS, and finds that these hazards

are adequately controlled by the PSSCs designated by DCS.

5.1.5 Safety Assessment of Design Basis Methodology

The objective of the staff’s review of the methodology was to determine if the safety assessment

was complete by assuring that all appropriate natural phenomena and external manmade and

internal process hazards were considered. The review of natural phenomena and external

manmade hazards consisted of evaluating the DCS screening criteria to determine if they were

appropriate for identifying all credible events. To evaluate whether the internal process hazards

were sufficiently addressed, the staff reviewed the proposed plant processes, reviewed the

operating experience and hazard analyses of other similar facilities, and considered feedback

from the discipline-specific FSER reviews. 

The DCS safety assessment of the design bases consisted of the identification and assessment

of natural phenomena hazards, external manmade hazards, and internal process hazards. 

Pursuant to 10 CFR 70.61, high-consequence events must be highly unlikely and intermediateconsequence events must also be unlikely. In the revised CAR, the applicant has provided

qualitative definitions of the terms “not unlikely,” “unlikely,” “highly unlikely,” “credible,” and “not

credible”. (Quantitative likelihood values are not required in 10 CFR Part70, “Domestic
5-5

Licensing of Special Nuclear Materials.”) All initiating events were assumed to have a likelihood

of “not unlikely,” which the applicant has defined as events that may occur during the life of the

facility. Because of the high probability associated with postulated events, PSSCs will be

selected and designed such that the accident sequences with regard to the above threshold

doses to the facility worker, site worker, and individuals outside the controlled area will be highly

unlikely. The applicant has defined “highly unlikely” as “events originally classified as not

unlikely or unlikely to which sufficient principal SSCs are applied to further reduce their likelihood

to an acceptable level.” The applicant has proposed deterministic design criteria to assure that

the consequences from postulated events that will exceed the threshold limits of 10 CFR

70.61(c) are highly unlikely. These criteria include the following:

• application of the single-failure criterion or double-contingency principle

• application of NQA-1 in Appendix B, “Quality Assurance Criteria for Nuclear Power Plants

and Fuel Reprocessing Plants,” to 10 CFR Part 50, “Domestic Licensing of Production

and Utilization Facilities”

• application of industry codes and standards

• management measures including IROFS failure detection (IROFS failure detection and

repair or process shutdown capability) 

In addition, the applicant committed to a supplemental likelihood assessment for event

sequences that could exceed the 10 CFR 70.61(c) criteria for persons outside the controlled

area boundary. Only the deterministic criteria would apply to facility workers and the

environment. In the revised CAR, Section 5.4.3 (DCS, 2002b), the applicant stated that, “This

supplemental likelihood assessment will be based on the guidance provided in the NUREG-1718

[NRC, 2000] and will demonstrate a target likelihood index comparable to a ‘score’ of -5 as

defined in Appendix A of the SRP.” The staff finds the applicant’s definition of “highly unlikely” to

be acceptable.

In regard to the environmental protection requirements of 10 CFR 70.61(c)(3), PSSCs will be

selected and designed so as to ensure that the accident sequence is unlikely, as discussed in

FSER Chapter 10. In the revised CAR, “unlikely” is defined as “events that are not expected to

occur during the lifetime of the facility but may be considered credible.” In that there were no

accident sequences that resulted in over-the-threshold consequences for the environment only,

this classification of events (i.e., unlikely) was not used in the safety assessment to demonstrate

a compliance strategy with the performance requirements of 10 CFR 70.61.

Another definition provided by DCS in the revised CAR to support the safety assessment is “not

credible.” The applicant defined this term as “natural phenomena or external man-made events

with extremely low initiating frequency and process events that are not possible.” The

application of this definition is explained in the revised CAR in the discussions of screening

criteria for natural phenomena and external manmade external events. The staff found the

applicant’s definition of credible to be acceptable in its review of the natural and manmade event

screening criteria in FSER Sections 5.1.5.1 and 5.1.5.2, respectively.
2Five rem total TEDE to any individual outside the controlled area, 25 rem TEDE to a facility worker, 24-hr

concentrations exceeding 5000 times the values in Table 2 of Appendix B to 10 CFR Part 20, or the chemical safety

criteria.

5-6

As indicated above, controlling facility risks entails the identification and assessment of potential

facility hazards (i.e., accident scenarios). Based on this hazards assessment, PSSCs and the

safety functions of the PSSCs can then be identified. The applicant’s methodology for

developing the PSSCs and their functions is presented graphically in the flowchart in the revised

CAR Figure 5-4.1 The basic inputs to the selection process are the site description from which

credible natural phenomena and external manmade hazards are determined from a screening

process and preliminary design information from which credible internal hazards are identified. 

The results of the external event screening and internal hazard screening are inputs to a

preliminary accident analysis.

An early step in the preliminary hazard analysis (PHA), the correlation of process units with

facility workshops and process support units, is shown in the revised CAR Tables 5.5-1 and 5.5-

2. The radioactive material inventory in each facility location is shown in the revised CAR Table

5.5-3a; the radioactive material inventory by fire area is shown in the revised CAR Table 5.5-3b. 

The summary hazard identification matrix of hazards versus workshops and process support

groups is shown in the revised CAR Table 5.5-4. This segmentation and correlation with

hazards allowed a comprehensive hazard identification for each individual area. A consequence

analyses was then performed by the applicant to evaluate the bounding unmitigated

consequences for each type of accident within a workgroup. If the unmitigated consequences

exceeded the dose thresholds for 10 CFR 70.61(c)2, then the group was further evaluated. For

the event scenarios which exceed the 10 CFR 70.61(c) thresholds, a safety strategy for

prevention or mitigation was established and PSSCs at the structure and system level were

identified. The selection of safety strategies was facilitated by segregating events which had

common features that would allow similar prevention or mitigation strategies into event groups. 

This simplified the analysis by allowing for the development of common safety strategies and

PSSCs for multiple events, such that the PSSCs that cover bounding events also cover

nonbounding events. In the context of the applicant’s analysis, a bounding event is the event

which results in the largest consequence in each group and the greatest risk because the

likelihood of all of the events is considered to be the same. 

After the PSSCs have been determined, their design bases are developed and, if the accident

consequence is mitigated, the resulting bounding mitigated consequence is compared against

the performance requirements of 10 CFR 70.61(c). If mitigation is successful at sufficiently

reducing the consequences, or if the accident scenario is prevented, the developed PSSCs and

support functions become input to the final design. If not, the evaluation is repeated with a

different set of PSSCs (or a change in design-bases values). The applicant, in the revised CAR,

Section 5.4.4.3, pursuant to 10 CFR 70.61(c)(3), also performed analyses of the potential

radioactive release to the environment by calculating the 24-hr average effluent concentration of

each radionuclide released in an accident sequence and comparing this to 5000 times the

values specified in Table 2 of Appendix B, “Annual Limits on Intake (ALIs) and Derived Air

Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations;

Concentrations for Release to Sewerage,” to 10 CFR Part 20, “Standards for Protection Against

Radiation” (CFR, 2004a). This is required to show compliance with 10 CFR 70.61(c)(3). 
5-7

5.1.5.1 Natural Phenomena Hazards Methodology

Natural phenomena having a credible potential effect on facility operations were identified

through a screening process in which natural phenomena hazards (NPHs) having a frequency of

occurrence of less than 10-6 per year were designated as incredible and screened from further

consideration. Deterministic methods were also used to screen out events that would not be

physically appropriate for the site. For example, debris avalanching was ruled out because of

the relatively level nature of the surrounding topography. The staff considers the NPHs

screening methodology to be in accordance with the criteria of NUREG-1718, Section 5.4.5.2

(NRC, 2000) and is therefore acceptable for the purpose of meeting the performance

requirements of 10 CFR 70.61. A comprehensive list of NPHs were initially evaluated and the

rationale for further consideration or exclusion of each NPH is provided in the revised CAR Table

5.5-5.

5.1.5.2 External Manmade Events Methodology

The applicant considered external manmade hazards (EMMHs) to be those hazards that are

caused by events originating from the operation of nearby public, private, government, industrial,

chemical, nuclear, and military facilities and vehicles. The major categories of events that could

result from EMMHs that were considered by DCS are as follows:

• a release of radioactive material resulting in exposures to facility personnel

• a release of hazardous chemicals resulting in exposures to facility personnel

• explosions or other events that directly impact facility PSSCs

• events that result in a loss of offsite power

• events that result in a fire (and/or resulting smoke) that spreads to the facility

Events in these categories were screened using applicable criteria from NUREG/CR-4839,

“Methods for External Event Screening Quantification: Risk Methods Integration and Evaluation

Program (RMIEP) Methods Development,” Regulatory Guide (RG) 1.91, “Evaluations of

Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plants,” RG 1.78,

“Assumptions for Evaluating the Habitability of a Nuclear Power Plant Control Room during a

Postulated Hazardous Chemical Release,” and NUREG-0800, “Standard Review Plan.” A

summary of the EMMH screening is provided in Table 5.5-8 of the revised CAR.

External manmade events that were evaluated and screened out as not applicable to the site

or of having too low a probability for consideration include the following:

• roadway accidents

• rail accidents

• aircraft accidents

• barge/shipping traffic accidents

• industrial facility accidents, except for F-Area

• military facility accident

The staff performed an in-office review of the applicant’s calculation of future flight activities over

the life of the facility to confirm the aircraft accident analysis. The staff found these calculations

and previously submitted calculations (DCS, 2002a; Enclosure A, DCS, 2002c) to be acceptable. 
5-8

The staff considers the EMMHs screening methodology to be in accordance with NRC guidance

and is acceptable for the purpose of meeting the performance requirements of 10 CFR 70.61.

5.1.5.3 Process Hazards Methodology

The applicant evaluated the potential for and consequences of process-related internal events. 

These events were divided into six major categories:

(1) loss of confinement/dispersal of nuclear material events

(2) fire events

(3) load-handling events

(4) explosion events

(5) criticality

(6) chemical

In the revised CAR, DCS only presented numerical radiological consequence values for the most

severe event in each of the above major categories (except for the chemical category). 

Chemical consequences are discussed in Chapter 8 of the revised CAR and the staff’s

evaluation is provided in Chapter 8 of the FSER. The staff reviewed the accident scenarios

developed by the applicant and has determined that these are a complete and bounding set

based on the preliminary design and description of the processes projected for use at the

proposed facility and described in Chapters 1, 7, 8, 9, 10, and 11 of the revised CAR (DCS,

2002b). 

5.1.5.4 Baseline Design Criteria

As set forth in 10 CFR 70.64(a)(1–10), the following 10 BDC are applicable to the proposed

facility: 

(1) quality standards and records, requiring that the facility design be developed and

implemented in accordance with management measures and IROFS, and that records of

the IROFS be maintained for the life of the facility

(2) natural phenomena hazards, requiring that the facility design adequately protects against

natural phenomena and takes into consideration the most severe historical events

documented at the facility site

(3) fire protection, requiring that the facility design adequately protects against fires and

explosions

(4) environmental and dynamic effects, requiring that the facility design adequately protects

against internal environmental conditions and dynamic effects from normal operations,

maintenance, testing, and accidents

(5) chemical protection, requiring that the facility design adequately protects against

specified chemical risks and internal conditions affecting the safety of licensed material
5-9

(6) emergency capability, requiring that the facility be designed to provide emergency

capability to control (i) licensed material and hazardous chemicals produced from

licensed material, (ii) evacuation of onsite personnel, and (iii) specified onsite emergency

facilities and services

(7) utility services, requiring that the facility be designed so that essential utility services will

continue to operate

(8) inspection, testing, and maintenance, requiring that IROFS designs provide for adequate

inspection, testing, and maintenance, to ensure that IROFS will be available and will

reliably perform their functions when needed

(9) criticality control, requiring that the facility be designed to provide for criticality control,

including adherence to the double-contingency principle

(10) instrumentation and controls, requiring that the facility be designed to provide for

inclusion of instrumentation and control systems to monitor and control the behavior of

IROFS

In the FSER sections referenced below, the staff states whether or not the facility’s preliminary

design satisfies the BDC outlined above, pursuant to 10 CFR 70.64(a). In the review of the

revised CAR, meeting the 10 CFR 70.61 performance requirements was the primary criterion

considered by the staff for satisfying the BDC.

(1) Quality Standards and Records. The design must be developed and implemented in

accordance with management measures to provide adequate assurance that IROFS will

be available and reliable to perform their functions when needed. Appropriate records of

these items would have to be maintained by or under the control of the licensee

throughout the life of the proposed facility. This BDC is satisfied at the construction

authorization stage, as discussed in FSER Section 15.1.

(2) Natural Phenomena Hazards. The design must provide for adequate protection from

natural phenomena with consideration of the most severe documented historical events

for the site. This BDC is satisfied at the construction authorization stage, as discussed in

FSER Sections 1.3, 5.1.6.1, 11.1, and 11.12.

(3) Fire Protection. The design must provide for adequate protection against fires and

explosions. This BDC is satisfied for fires at the construction authorization stage, as

discussed in FSER Section 7.1.2. This BDC is also met for explosions, as discussed in

Chapters 8 and 11.

(4) Environmental and Dynamic Effects. The design must provide for adequate protection

from environmental conditions and dynamic effects associated with normal operations,

maintenance, testing, and postulated accidents that could lead to loss of safety functions. 

This BDC is satisfied at the construction authorization stage, as discussed in FSER

Section 11.11. 
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(5) 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. This BDC

is met, as discussed in FSER Sections 8.2 and 11.2.2.

(6) Emergency Capability. The design must provide for emergency capability to maintain

control of the following:

(i) licensed material and hazardous chemicals produced from licensed material

(ii) evacuation of onsite personnel

(iii) onsite emergency facilities and services that facilitate the use of available offsite

services

This BDC is satisfied at the construction authorization stage, as discussed in FSER

Chapter 14.

(7) Utility Services. The design must provide for continued operation of essential utility

services. Based on the safety assessment of the proposed facility, the only utility with

safety significance is electrical power. This BDC is satisfied at the construction

authorization stage, as discussed in FSER Section 11.5. 

(8) Inspection, Testing, and Maintenance. The design of the IROFS must provide for

adequate inspection, testing, and maintenance to ensure their availability and reliability to

perform their function when needed. This BDC is satisfied at the construction

authorization stage, as discussed in FSER Sections 15.1 and 15.3.

(9) Criticality Control. The design must provide for criticality control including adherence to

the double-contingency principle. This BDC is met at the CAR stage, as discussed in

FSER Section 6.1.4.2.

(10) Instrumentation and Controls. The design must provide for inclusion of instrumentation

and control systems to monitor and control the behavior of the IROFS. This BDC is

satisfied at the construction authorization stage, as discussed in FSER Section 11.6.

5.1.5.5 Defense-in-Depth

Under 10 CFR 70.64(b), the facility design and layout must be based on defense-in-depth

practices. As used in 10 CFR 70.64, defense-in-depth practices at new facilities means a design

philosophy, applied from the outset and through completion of the design, that is based on

providing successive levels of protection, such that health and safety will not be wholly

dependent upon any single element of the facility design. The net effect of incorporating

defense-in-depth practices is a conservatively designed facility that will exhibit greater tolerance

to failures and external challenges. As required by 10 CFR 70.64(b), to the extent practicable,

the facility design must incorporate (1) a preference for engineered controls over administrative

controls to increase overall system reliability, and (2) features that will enhance safety by

reducing challenges to items which will be relied upon for safety.
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In Section 5.5.5 of the revised CAR, the applicant described its general design philosophy and

defense-in-depth practices. In this section, the applicant described a hierarchy of controls in its

general philosophy of design that has been established as follows (the most favored control

listed first):

• protection by a single passive safety device, functionally tested on a predetermined basis

• independent and redundant active engineered features, functionally tested on a

predetermined basis

• single hardware system/engineered feature, functionally tested on a predetermined basis

• enhanced administrative controls

• simple administrative controls or normal process equipment

The staff has determined that the above hierarchy of controls demonstrates a preference for

engineered controls, as required by 10 CFR 70.64(b)(1). In addition, the applicant’s

incorporation of additional protection features into the design of the proposed facility will

enhance safety by reducing challenges to items which will be relied upon for safety, as required

by 10 CFR 70.64(b)(2). 

In addition, the applicant has described its defense-in-depth practices as consisting primarily of

meeting double contingency (for protection against criticality events) and the single-failure

criterion. The staff’s evaluation of the applicant’s implementation of double contingency is

provided in Section 6.1.4.2 of the FSER. The applicant’s implementation of the single-failure

criterion, as described in Section 5.5.5.2 of the revised CAR, consists of (1) the use of redundant

equipment or systems, (2) independence, (3) separation, and (4) the fail-safe principle.

The staff concludes that the applicant’s strategy for defense-in-depth meets the requirements of

10 CFR 70.64.

5.1.6 Safety Assessment Results

The safety assessment methodology, as described above, resulted in the identification of

accident scenarios and PSSCs and their functions. The PSSCs identified by the applicant

through its safety assessment are summarized in Tables 5.1-1 and 5.1-2 of the FSER. Table

5.1-3 lists the design bases associated with the safety functions of the PSSCs and references

the FSER sections that contain more detailed staff evaluations of the PSSCs. 

5.1.6.1 Natural Phenomena Design Basis Events and Related PSSCs

As stated in FSER Section 5.1.5.1, the staff has determined the screening methodology for

NPHs to be acceptable. The screening methodology identified design basis natural events and

their related PSSCs. The likelihood of any such design basis event occurring should be

sufficiently low to assure that any adverse consequences are highly unlikely because structural

failures caused by natural phenomena were assumed to have the potential for high
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consequences. The adequacy of the PSSCs to prevent releases are evaluated using normally

accepted industry practice as a criterion.

Natural phenomena that were not screened out included the following:

• extreme wind

• earthquake (including liquefaction)

• tornado (including tornado missiles)

• external fire

• rain, snow, and ice

• lightning

• temperature extremes

The design basis wind selected for the facility has an annual exceedance probability of 10-4 per 

year (DCS, 2002b, Section 5.5.2.6.5.1). The PSSCs identified to provide protection against the

design basis wind are the MOX fuel fabrication building, emergency diesel generator building,

associated missile barriers, and the waste transfer line. The safety function of the structures

and missile barriers are to withstand design basis wind loads and wind-driven missiles and to

provide protection for internal structures, systems, and components (SSCs). Although the

exceedance probability of 10-4 does not, by itself, preclude a consequence which is highly

unlikely, the design for wind loadings is controlled by the tornado at low frequencies of

occurrence. Hence, the staff concludes that the design basis established by the applicant for

extreme wind satisfies the performance requirements of 10 CFR 70.61. 

The design basis earthquake selected for the facility also has an annual exceedance probability

of 10-4 (DCS, 2002b, Section 5.5.2.6.5.2). The PSSCs identified to provide protection against

the design basis earthquake are (1) the waste transfer line, (2) the MFFF building, (3) the

emergency diesel generator building, (4) the fluid transport systems, as necessary, and (5) the

seismic monitoring and associated isolation valves. The primary safety function of the first four

PSSCs is to withstand the effects of the design basis earthquake and to assure that seismic

effects on non-PSSCs will not result in the prevention of PSSCs from performing their safety

function. The safety function of the seismic monitoring and isolation valves is to prevent fire

and/or criticality as a result of an uncontrolled release of chemicals and water within the MFFF

building. For NRC-licensed facilities, such as nuclear power plants, the conservatism between

design and performance arises from factors such as prescribed analysis methods, specification

of material strengths, and limits on inelastic behavior following nuclear design criteria and NRC

Standard Review Plans (SRPs). Conservatism in the NRC seismic SRPs are not explicitly keyed

to risk-reduction values. Nevertheless, the risk-reduction factors achieved by applying NRC

guidelines to the evaluation of commercial reactor SSCs have been shown to be equal to or

even higher than those prescribed by DOE STD-1020-94, Change Notice No. 1, “Natural

Phenomena Hazards Design and Evaluation Criteria for Department of Energy Facilities” (DOE,

1994). For example, the average mean annual probability of exceedance (MAPE) for the design

ground motions at existing nuclear power plants is approximately 1×10-4 (NRC, 1997), yet the

mean annual seismic core damage frequency of nuclear power plants is estimated to range

between 6×10-6 and 1×10-5 (Chen, 1991). Thus, an effective risk reduction for nuclear power

plants is 10 times or greater. In additional information provided to NRC (DCS, 2002a; Enclosure

B, DCS, 2002c), DCS provided the results of calculations which showed that, taking into account

the building and component designs, the performance of SSCs in the facility will meet the
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availability (or failure) criteria necessary to make high consequences highly unlikely. The staff

concludes that the applicant’s selection of the design basis earthquake satisfies the performance

requirements of 10 CFR 70.61. 

The design basis tornado selected for the facility has an annual exceedance probability of

2×10-6 (DCS, 2002b, Section 5.5.2.6.5.3). The PSSCs identified to provide protection against

the design basis tornado are the MFFF building, the emergency diesel generator building, the

associated missile barriers, the waste transfer line, and the tornado dampers. The safety

functions of these PSSCs are to withstand the design basis tornado wind loads, tornadogenerated missiles, and differential pressure, as well as to provide protection for internal SSCs. 

The staff concludes that the applicant’s selection of the design basis tornado satisfies the

performance requirements of 10 CFR 70.61. 

The design basis for external fire was assumed to be a forest fire near the facility (DCS, 2002b,

Section 5.5.2.6.5.4). The plant exterior is designed to withstand a fire duration of at least 2

hours (further information may be found in Section 7.1.5.4 of the FSER). This is considered by

the staff to be adequate, based on the availability of an onsite fire brigade and the fuel loading

provided by natural growth around the building. The PSSCs identified to provide protection

against the external fire are the MFFF building structure, the emergency generator building

structure, the emergency control room air-conditioning system, and the waste transfer line. The

safety functions of the PSSCs are to withstand the effects of the external fire, to provide

protection for internal SSCs, and to ensure habitable conditions for operators as necessary. The

staff concludes that the applicant’s selection of the design basis external fire is an acceptable

strategy for meeting the performance requirements of 10 CFR 70.61. 

The design basis rainfall has an annual exceedance probability of 10-5 (DCS, 2002b, Section

5.5.2.6.5.5). This will meet the likelihood requirements for high- and intermediate-consequence

events. The snow and ice loadings have an annual exceedance probability of 102 (DCS, 2002b,

Section 5.5.2.6.5.5). The applicant has stated that effects of snow and ice loads with a lower

annual exceedance probability are bounded by the design for other live loads. DCS determined

that a 10,000-yr snow and ice load would be less than one-half of the design load for live loads. 

The PSSCs and design basis safety functions associated with rain, snow, and ice are the MOX

fuel fabrication building structure and the emergency diesel generator building structure, which

will be designed to withstand the effects of rain, snow, and ice without failing and will protect

internal SSCs from the effects of rain, snow, and ice. The staff considers the applicant’s

strategy and selection of PSSCs to be acceptable for meeting the performance requirements of

10 CFR 70.61.

The design basis for lightning protection was in accordance with National Fire Protection

Association (NFPA) 780-1997 (DCS, 2002b, Section 5.5.2.6.5.6). Design basis temperature

extremes for the ventilation system were based on observed temperatures at the Savannah

River Site (SRS) over a 35-yr period (1961 to 1996). Both of these design bases are appropriate

because neither lightning nor severe temperature are expected to cause a significant

consequence by themselves. No PSSCs are required for protection against lightning or extreme

temperatures. The staff agrees with the applicant’s rationale for not requiring additional PSSCs

to protect against these events.
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The staff considers the results of the evaluation of NPHs to be acceptable for meeting the

performance requirements of 10 CFR 70.61. The PSSCs identified by DCS to control the risks

of natural phenomena at the proposed facility will, if properly implemented, ensure that such

risks will be acceptably low during any facility operation. In addition, the staff considers the

applicant’s evaluation to be adequate to satisfy the requirements of 10 CFR 70.64(a)(2) (BDC,

NPHs) which state that the design must provide for adequate protection against natural

phenomena with consideration of the most severe documented historical events for the site.

5.1.6.2 External Manmade Events and Related PSSCs

As stated in FSER Section 5.1.5.2, the staff has determined the screening methodology for

external manmade hazards to be acceptable. The screening methodology identified manmade

external design basis events and their related PSSCs. The likelihood of any such design basis

event occurring should be sufficiently low to assure that any adverse consequences are highly

unlikely because structural failures caused by such events were assumed to have the potential

for high consequences. The adequacy of the PSSCs to prevent releases are evaluated using

normally accepted industry practice as a criteria.

Manmade external events that were not screened out include:

• Potential hazardous chemical or radioactive releases from SRS facilities or vehicles. 

SRS documentation provides the radiological/chemical consequences of accidents at

existing facilities. The applicant has reviewed these analyses and determined that there

are no credible accidents that could potentially impact facility operations personnel.

However, personnel in the emergency control room will be protected by the emergency

control room air-conditioning system which is considered a PSSC for other evaluated

events. In addition, based on existing DOE requirements, it is not expected that facilities

to be designed and operated by DOE to support the MOX facility will present a significant

risk for the facility. The staff will consider possible risks from these facilities as part of its

review of any later application for a license to possess and use licensed material, when

more information about the proposed DOE facilities is expected to be available.

• Potential explosions at a nearby facility or an explosion involving a vehicle, particularly

one in the F-Area. The applicant stated that the main MOX building (BMF) and the

emergency diesel generator buildings can withstand the impacts of explosions in the FArea. The staff performed an in-office review of the applicant’s calculations of vapor

cloud explosions resulting from a tanker truck accident on an F-Area road nearest to the

MFFF and an explosion from a large storage area more distant from the MFFF. The

explosion calculations were based on conservative assumptions from RG 1.91. The staff

made an independent check on the applicant’s reflected overpressure calculations using

Figure 2-5.14 on page 2-101 of the Society of Fire Protection Engineers (SFPE)

Handbook of Fire Protection Engineering, First Edition, published September 1988. The

staff determined that the applicant’s calculations were close to the staff’s values obtained

from Figure 2-5.14 and were safely within the design overpressure limits of the

structures. 

• Loss of offsite power from EMMHs is considered similar in potential for consequences as

loss of offsite power from NPHs. PSSCs requiring power are supplied with emergency
1

In its evaluation, the staff has considered administrative controls that consist of two or more independent actions,

where the success of either will prevent the consequences, as enhanced administrative controls.

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power upon loss of offsite power and failure of the standby alternating current (ac) power

system. The adequacy of the power supply in terms of the baseline design criteria and

the performance requirements has been evaluated in FSER Section 11.5.1.3. The staff

considers the applicant’s strategy and selection of PSSCs to be acceptable for meeting

the performance requirements of 10 CFR 70.61 with regard to loss of offsite power.

• External manmade fires are fires resulting from a vehicle crash, train crash/derailment,

barge/shipping accident, or SRS facility fire that engulf neighboring grasslands or forests. 

This event has the same consequences as the design basis external fire listed as an

NPH. The ability of the facility to withstand the effects of external fires is discussed in

FSER Section 7.1.5.4. The staff also considers the applicant’s strategy and selection of

PSSCs to be acceptable for meeting the performance requirements of 10 CFR 70.61.

Based on its evaluation of the screening methodology (FSER Section 5.1.5.2) and the evaluated

events as described above, the staff considers the results of the evaluation of manmade events

to be acceptable for meeting the performance requirements of 10 CFR 70.61. The PSSCs

identified by DCS to control the risks of such manmade events at the proposed facility will, if

properly implemented, ensure that such risks will be acceptably low during any facility operation. 

5.1.6.3 Internal Process Hazard Design Basis Events and Related PSSCs

As stated in FSER Section 5.1.5.3, the staff found that the screening methodology for internal

process hazards was adequate. With respect to the internal process hazards which DCS

evaluated, the staff’s review of those hazards was primarily an evaluation of the strategy and

PSSCs at a conceptual level with regard to their potential to guide the development of a design

which will meet the 10 CFR 70.61 performance requirements. Criteria used in the staff

evaluation consisted of a comparison against normally accepted industry practice, consideration

of the applicant’s design criteria, consideration of a probability index using the NUREG-1718,

Table A-5 descriptions (primarily for protection of individuals outside the controlled area

boundary (IOC)), and/or deterministic arguments primarily for protection of facility workers

and/or the environment. Table A-5 of NUREG-1718 provides a table equating types of controls

to approximate probabilities of failure on demand (PFOD). The following descriptions of controls

were provided:

• exceptionally robust passive engineered control (PEC) or an inherently safe process

(index -4 or -5; PFOD 10-4–10-5)

• a single PEC or an active engineered control (AEC) with high dependability (index -3 or

• -4; PFOD 10-3–10-4 )

• a single AEC, an enhanced1 administrative control, or an administrative control for routine

planned operations (index -2 or -3; PFOD 10-2–10-3 )

• an administrative control that must be performed in response to a rare, unplanned

demand (index -1 or -2; PFOD 10-1–10-2)
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For the purposes of this review, the staff considered the revised CAR description of a high

dependability AEC (such as the C4 confinement system) and assigned it an index of -4 or -5. In

addition to the base dependability indices that the staff determined for the PSSCs based on their

descriptions in the revised CAR, the staff also took into account the impact of surveillance

intervals on the overall reliability. The assumption was made that as long as the PSSC was

capable of being part of a surveillance program, surveillance intervals would be adjusted at the

ISA preparation stage to achieve the desired dependability. In addition, enhanced administrative

controls, such as combustible loading controls, are assigned a lower PFOD in the staff review

because of the incorporation of features such as fire modeling, quantifiable margins, and

surveillances in their implementation.

In addition to the dependability of the safety strategy, the staff also independently evaluated the

applicant’s consequence assessment for those event sequences in which the consequences for

one or more of the potential receptors was determined by the applicant to be less than the 10

CFR 70.61(c) threshold value. These independent calculations are described in FSER Section

9.1.1.4.

Feedback from the technical reviews was also used to evaluate the practicality and

appropriateness of the PSSCs or safety strategy to the event being evaluated. This feedback

was used to assure that the proposed strategies did not significantly deviate from accepted

nuclear industry practice, taking into account historical events as well as successful operation at

other chemical or nuclear facilities.

An area of discussion with DCS was the protection of facility workers during accident events. It

was the position of DCS that the index method and its implied numerical probability may not be

applicable to protection of the facility worker. Reliance on worker actions for mitigation in many

of the worker protection scenarios requires a deterministic, rather than a probabilistic evaluation. 

The accident scenarios which rely on prevention (such as most explosions, some fires, and

some of the materials-handling accidents) do not require an evaluation for the facility worker

separate from the one performed for protection of the environment and persons outside the

controlled area boundary. If the specific prevention measures are considered sufficient to make

the release of radioactive material from the accident sequence highly unlikely, dose to the

worker does not need to be evaluated. Also, some of the fires are not prevented to a low

probability of occurrence, but are considered sufficiently slow-growing events such that a worker

may take a course of action such as leaving the area and/or donning a respirator that would

make the worker’s dose effectively zero or negligible. For some of the load-handling events, the

staff questioned the ability of the worker to don a mask or vacate the area in sufficient time to

keep the worker dose below the 10 CFR 70.61(c) threshold levels. In these cases, the staff

requested the applicant to perform dose calculations which were reviewed on site (NRC, 2002). 

These were found to be acceptable, because the doses were low enough to allow the worker to

take protective action within a reasonable time.

The following six subsections (5.1.6.3.1 through 5.1.6.3.6) discuss the accident sequences

developed to evaluate internal process hazards. The staff finds that the evaluations of internal

process hazards at the proposed facility meet the performance requirements of 10 CFR 70.61.
2NUREG-1718, Table A-5, is used to estimate a probability index number for the strategy based on the

type of control(s) being described. The applicant's deterministic design criteria comm itments were also

considered. The staff considers these comm itments sufficient to lower the PFOD and the probability of

the accident sequence, often by at least an order of m agnitude based on the type of control(s) proposed. 

If the probability index of the sequence can achieve a value of -5 or less, the staff considered the strategy

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5.1.6.3.1 Confinement Events

Confinement of radioactive material at the facility is provided by static confinement boundaries in

conjunction with ventilation systems and sealed confinement barriers (e.g., containers and fuel

rods).

Thirty-one separate events with potentially significant consequences were analyzed by DCS to

determine the bounding consequences from a potential loss of confinement event. These

events were assigned to the following 12 groups with a unique prevention or mitigation strategy:

(1) overtemperature

(2) corrosion

(3) glovebox breaches or backflows

(4) leaks in the AP process vessels or pipes

(5) backflow from a process vessel through utility lines

(6) rod-handling operations

(7) breaches in containers outside of gloveboxes as a result of handling 

(8) over- or underpressurization of glovebox

(9) excess temperature caused by radioactive decay

(10) glovebox dynamic exhaust failure

(11) process fluid line leak in a C3 area outside a glovebox

(12) sintering furnace confinement boundary failure

These 12 groups are discussed below:

Overtemperature (Confinement)

The bounding event for the overtemperature event group in the confinement events accident

category was determined by the applicant to be excessive temperature of the AP electrolyzer

resulting in high-temperature damage to, and breach of, the AP electrolyzer and damage to

glovebox panels and dispersal of radioactive material. The material at risk was the maximum

inventory of radioactive material in the electrolyzer glovebox. Such an event could be caused by

control system failure, electrical isolation failure, or loss of cooling to process equipment. The

applicant determined this to be an above the 10 CFR 70.61(c) threshold consequence event for

facility workers, site workers, IOC, and the environment, and has opted to protect potential

receptors through a strategy of prevention and mitigation. The PSSC identified for protection of

the facility worker and the environment for this event is the process safety control subsystem,

which will shut down process equipment before exceeding temperature safety limits. The PSSC

for protection of site workers and IOC is the C3 confinement system, which will provide filtration

to mitigate dispersions from the C3 area. In addition, the process safety control subsystem

provides defense-in-depth for protection of the site worker and IOC. Based on the applicant’s

deterministic criteria and the NUREG-1718 Table A-5 descriptions2 (high availability AEC), the
to be acceptable. This will be dem onstrated by the applicant in a supplemental likelihood assessment to

be conducted in conjunction with the ISA when specific IROFS are developed.

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staff considers this to be an acceptable strategy for meeting the 10 CFR 70.61 performance

requirements.

Corrosion (Confinement)

The corrosion event group is defined as catastrophic failure of a primary confinement boundary

(i.e., a laboratory or an AP glovebox containing corrosive chemicals, AP fluid transport systems,

a pneumatic transfer line, or ducting of the C4 confinement system) postulated to result from

corrosion. Loss-of-confinement events caused by corrosion within process cells are included in

the leaks of AP process vessels or pipes within process cells event group. Loss-of-confinement

events caused by corrosion of pipes containing process fluids within C3 areas not enclosed

within a glovebox are discussed in the process fluid leak in C3 area outside of glove box event

group. The bounding event for the corrosion event group in the confinement events accident

category was determined by the applicant to be corrosion of the pneumatic pipe automatic

transfer system from corrosive chemicals resulting in a breach of confinement and dispersal of

radioactive materials. The material at risk was the maximum inventory in the pneumatic pipe

automatic transfer system. The applicant determined this to be an above the 10 CFR 70.61(c)

threshold consequence event for the facility worker and the environment and a below the 10

CFR 70.61(c) threshold consequence event for the individuals outside the controlled area

boundary. The staff independently evaluated this accident sequence and agrees to its

categorization. The applicant has opted to protect the facility worker and the environment

through a strategy of prevention and mitigation. The PSSCs identified for facility worker

protection and protection of the environment are the material maintenance and surveillance

programs, which will detect and limit the damage resulting from corrosion. No PSSCs are

identified by the applicant as being necessary to protect the site worker and IOC. However, the

C4 and C3 confinement systems, as well as the C2 confinement system passive boundary,

provide defense-in-depth protection for site workers and IOC. Based on the nature of this event

(mitigated by slow development or prevented by administrative controls), the staff considers this

to be an acceptable strategy for meeting the 10 CFR 70.61 performance requirements.

Small Breaches in Glovebox Confinement Boundary (Confinement)

The bounding event for small breaches in a glovebox confinement boundary or backflow in the

confinement events accident category was determined to be backflow through the interfacing

gas line (e.g. nitrogen, helium) to the interfacing system followed by the opening of this

interfacing system during a maintenance operation. The material at risk was the maximum

inventory of radioactive material in a glovebox. Loss of gasflow through a supply line was listed

as a possible cause. The applicant determined this to be an above the 10 CFR 70.61(c)

threshold consequence event for the facility worker and the environment and a below the 10

CFR 70.61(c) threshold consequence event for the site worker and IOC. The staff

independently evaluated this accident sequence and agrees to its categorization. The applicant

has opted to protect the facility worker through a strategy of mitigation. The PSSC identified for

protection of the facility worker and the environment is the C4 confinement system which

maintains a negative glovebox pressure differential between the glovebox and the interfacing

systems and will also maintain a minimum inward flow through small glovebox breaches. No
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PSSCs are identified by the applicant as being necessary to adequately protect the site worker

and IOC. In addition, the C3 confinement system provides defense-in-depth protection for the

site worker and IOC. Based on the nature of this event (mitigated by a high-availability AEC),

the staff considers this to be an acceptable strategy for meeting the 10 CFR 70.61 performance

requirements.

Leaks of AP Process Vessels or Pipes within Process Cells (Confinement)

The bounding event for leaks in the AP process vessels or pipes within process cells in the

confinement events accident category is a break or leakage of a tank/vessels inside the process

cell containing the liquid waste reception unit. The material at risk was the maximum inventory

of radioactive materials in the affected equipment in the AP process cell. Corrosion and

mechanical failure were listed as potential causes. The applicant determined this to be an above

the 10 CFR 70.61(c) threshold consequence event for the facility worker, environment, site

worker, and IOC. The applicant has opted to protect the facility worker through a strategy of

prevention. The PSSCs identified for facility worker protection are the process cells, which

contain fluid leaks within the cells, and the process cell entry controls, which prevent the entry of

personnel into process cells during normal operation. The environment, site worker, and IOC

are protected through a strategy of mitigation. The PSSC identified for protection of the

environment, site worker, and IOC is the process cell exhaust system. The function of the

process cell exhaust system is to ensure that a negative pressure exists between the process

cell areas and the C2 area and to ensure that the C2 exhaust is effectively filtered. The C2

confinement system passive boundary provides defense-in-depth protection for the environment,

site worker, and IOC. Based on the applicant’s deterministic criteria and the NUREG-1718

Table A-5 descriptions (AECs), the staff considers this to be an acceptable strategy for meeting

the 10 CFR 70.61 performance requirements.

Backflow from a Process Vessel through Utility Lines (Confinement)

The bounding event for backflow from a process vessel through the utility lines confinement

events accident category is backflow of radioactive material from a waste tank containing

americium. This backflow was postulated to flow through an interfacing supply line that is

subsequently breached or opened during a maintenance operation. The material at risk was the

maximum radioactive material in the waste tank. Loss of gas flow through the supply lines and

failure of pipes and/or valves are identified as potential causes. The applicant determined this to

be an above the 10 CFR 70.61(c) threshold consequence event for the facility worker,

environment, site worker, and IOC. The staff independently evaluated this accident sequence

and agrees to its categorization. The applicant has opted to protect the facility worker,

environment, site worker, and IOC through a strategy of prevention. The PSSC identified for

protection of the facility worker and the environment, site worker, and IOC are backflow

prevention features which will prevent process fluids from backflowing into interfacing systems. 

However, the C2 confinement system passive boundary provides defense-in-depth protection for

the environment, site worker, and IOC. Based on the applicant’s deterministic criteria and the

NUREG-1718 Table A-5 descriptions (PEC), the staff considers this to be an acceptable strategy

for meeting the 10 CFR 70.61 performance requirements.
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Rod-Handling Operations (Confinement)

The bounding event for rod-handling operations in the confinement events accident category

was determined to be the fracture of one or more fuel rods while utilizing fuel-rod-handling

equipment resulting in a breach of confinement and dispersal of radiological materials. The

material at risk was the maximum inventory of radioactive material in a tray of fuel rods. Human

error or equipment failure were listed as potential causes. The applicant determined this to be

an above the 10 CFR 70.61(c) threshold consequence event for the facility worker and a below

the 10 CFR 70.61(c) threshold consequence event for the environment, site worker, and IOC. 

The staff independently evaluated this accident sequence and agrees to its categorization. The

applicant has opted to protect the facility worker through a strategy of prevention and mitigation. 

The PSSCs identified for protection of the facility worker are facility worker actions which ensure

that facility workers take proper actions to limit radiological exposure, materials-handling controls

which ensure proper handling of primary confinements outside of gloveboxes, and materialshandling equipment to limit damage to fuel rods/assemblies during handling operations. No

PSSCs were identified by the applicant as being necessary to adequately protect the

environment, site worker, and IOC. The combination of rod cladding (primary confinement),

materials-handling controls, and facility worker actions are intended to make the likelihood of

above the 10 CFR 70.61(c) threshold consequences from the rod-handling accident sequence

highly unlikely. However, because a release could occur without warning, the applicant provided

dose calculations which were reviewed on site and found to be acceptable (NRC, 2002). The C2

confinement system passive boundary provides defense-in-depth protection for the environment,

site worker, and IOC. Based on the nature of this event (mitigated by limited initial release and

immediate worker responses), the staff considers this to be an acceptable strategy for meeting

the 10 CFR 70.61 performance requirements.

 

Breaches in Containers outside of Gloveboxes (Confinement)

The bounding event for breaches in containers outside of gloveboxes caused by handling

operations in the confinement events accident category was the failure of a 3013 canister,

transfer container containing plutonium-bearing waste, or other primary confinement types within

the C2 or C3 areas outside of a glovebox. The material at risk was the maximum inventory of

radioactive material in the container. The applicant determined this to be an above the 10 CFR

70.61(c) threshold consequence event for the facility worker, environment, site worker, and IOC. 

The applicant has opted to protect the facility worker, environment, site worker, and IOC using a

strategy of prevention and mitigation. PSSCs identified to protect the facility worker are

materials-handling controls, a 3013 canister to withstand the effects of design basis drops

without breaching, transfer container which will also withstand the effects of design basis drops

without breaching, and facility worker controls for bag-out operations in C3 areas. The staff has

requested the applicant to perform a dose calculation to determine the unmitigated dose to the

worker from the drop of a container other than a 3013 canister. This calculation was provided

and reviewed on site and found to be acceptable. PSSCs identified as required for protection of

the environment, site worker, and IOC are materials-handling controls to ensure proper handling

of primary confinement types outside of gloveboxes, as well as a transfer container, a 3013

canister, and the C3 confinement system, which will provide filtration to mitigate dispersions from

the C3 areas. The C2 confinement system passive boundary and the preventive features

utilized to reduce the risk to the facility worker and the environment provide defense-in-depth

protection for the IOC and site worker. Based on the applicant’s deterministic criteria and the
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NUREG-1718 Table A-5 descriptions, (PEC, AEC, and administrative controls), the staff

considers this to be an acceptable strategy for meeting the 10 CFR 70.61 performance

requirements. 

Over/Under Pressurization of Glovebox (Confinement)

The bounding event for over/under pressurization of a glovebox (i.e., C4 dynamic confinement)

in the confinement events accident category was determined by the applicant to be a rapid

overpressurization of the calcining furnace glovebox. The material at risk was the maximum

inventory of radioactive material in the glovebox. Potential causes of this event were identified

as the rupture of a high-flow or a high-pressure supply line or a clogged outlet high-efficiency

particulate air (HEPA ) filter. The applicant determined this to be an above the 10 CFR 70.61(c)

threshold consequence event for the facility worker and the environment, site worker, and IOC. 

For a rapid over- or underpressurization event, a prevention strategy is used. The PSSC

identified for protection of the facility worker is glovebox pressure controls, which will maintain

glovebox pressure within design limits. For a slow pressurization event, a mitigation strategy is

used. The PSSCs identified for facility worker protection are facility worker actions and the

process safety control subsystem to warn operators of glovebox pressure discrepancies before

exceeding differential pressure limits. The PSSCs identified for protection of the environment,

site worker, and IOC are the C3 and C4 confinement systems. The C2 confinement system will

provide defense-in-depth protection for the site worker and IOC. Based on the nature of this

event (mitigated by a warning system and immediate worker responses or prevented by an

AEC), the staff considers this to be an acceptable strategy for meeting the 10 CFR 70.61

performance requirements.

Excess Temperature before Decay Heat from Radioactive Materials (Confinement)

Thermal calculations have been performed to evaluate the effects of temperature on

confinement structural materials. Thermal sources considered in the calculation include

radioactive decay of nuclear materials, spontaneous heating of uranium dioxide caused by

oxidation (burnback), operation of electrical/mechanical equipment, and process equipment

(calcining furnace). However, only the 3013 storage area was found to require long-term cooling

to mitigate the effects of decay heat. The material at risk was the maximum inventory of

radioactive material in the powder storage area in the plutonium dioxide 3013 storage unit. The

applicant determined this to be an above the 10 CFR 70.61(c) threshold consequence event for

the facility worker, environment, site worker, and IOC. The applicant has opted to protect these

receptors through a strategy of prevention. The PSSC identified for protection of the facility

worker, environment, site worker, and IOC is the high depressurization exhaust system (part of

the C3 confinement system) which will ensure that temperatures in the 3013 canister storage

structure are maintained within design limits. Based on the applicant’s deterministic criteria and

the NUREG-1718 Table A-5 descriptions (high-availability AEC), the staff considers this to be an

acceptable strategy for meeting the 10 CFR 70.61 performance requirements.

Glovebox Dynamic Exhaust Failure (Confinement)

The bounding event for the glovebox dynamic exhaust failure event group in the confinement

event accident sequence was a loss of negative pressure or a flow perturbation involving the C4

dynamic confinement system resulting in a ventilation airflow reversal into a C3 area. The
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material at risk was the maximum inventory of airborne radioactive material in all connected

gloveboxes. Potential causes of this event are loss of normal control system, loss of all power,

or mechanical failure of the ventilation system. The applicant determined this to be an above the

10 CFR 70.61(c) threshold consequence event for the facility worker, environment, site worker,

and IOC. The applicant has opted to protect all receptors through a strategy of prevention. The

staff independently evaluated this accident sequence and agrees to its categorization. The

PSSC identified for protection of the facility worker, environment, site worker, and IOC is the C4

confinement system. The C3 and C2 confinement system passive boundaries provide defensein-depth protection of the environment and individuals outside the controlled area boundary. 

Based on the applicant’s deterministic criteria and the NUREG-1718 Table A-5 descriptions

(high-availability AEC), the staff considers this to be an acceptable strategy for meeting the 10

CFR 70.61 performance requirements.

Process Fluid Line Leak in a C3 Area outside of a Glovebox (Confinement)

This event was postulated to result from a leak from a line carrying a process fluid in a C3 area

outside of a glovebox or process cell caused by corrosive chemicals or mechanical failure of AP

piping. The material at risk was the maximum inventory of radiological material in a single AP

vessel. The applicant determined this to be an above the 10 CFR 70.61(c) threshold

consequence event for facility workers and the environment and a below the 10 CFR 70.61(c)

threshold consequence event for the site workers and IOC. The staff independently evaluated

this accident sequence and agrees to its categorization. The applicant has opted to use a

strategy of prevention to protect the facility worker and the environment. The PSSC identified

for this strategy is a double-walled pipe. No PSSCs were identified by the applicant as being

necessary to adequately protect individuals outside the controlled area boundary. However, the

C3 confinement system provides defense-in-depth protection for individuals outside the

controlled area boundary. Based on the nature of this event (prevented by a PEC), the staff

considers this to be an acceptable strategy for meeting the 10 CFR 70.61 performance

requirements.

Sintering Furnace Confinement Boundary Failure

This event was postulated to result from a breach in the sintering furnace confinement boundary. 

The furnace is postulated to fail either through a slow leak through the seals or a rapid

overpressurization event. These events could be caused by failure of the control system for the

hydrogen/argon supply line, a failure in the sintering furnace exhaust system, or a sintering

furnace seal failure. The material at risk was the maximum inventory of radiological material in

both sintering furnaces. The applicant determined this to be an above the 10 CFR 70.61(c)

threshold consequence event for the facility worker, environment, site workers, and IOC. For

the rapid overpressurization event, the applicant has chosen a strategy of prevention to reduce

the risk to the facility worker and the environment. The PSSCs for this event are the sintering

furnace pressure controls and the sintering furnace which functions as a confinement boundary. 

For the seal failure event, the applicant has chosen a strategy of mitigation. The PSSC for this

event is the sintering furnace which is designed to limit any postulated leakage. The PSSC

identified by the applicant to adequately protect the site worker and IOC is the C3 confinement

system. The C3 confinement system’s passive boundary also provides defense-in-depth

protection for the environment. The C2 confinement system’s passive boundary provides

defense-in-depth protection to the site worker and IOC. Based on the nature of this event
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(mitigated by a combination of PECs and AECs), the staff considers this to be an acceptable

strategy for meeting the 10 CFR 70.61 performance requirements.

5.1.6.3.2 Fire Events

The potential consequences of fire events at the facility as listed by DCS include the following:

• destruction of confinement barriers

• destruction of civil structures

• destruction of equipment contributing to dynamic confinement

• failure or damage to utility equipment

• loss of criticality controls

• loss of other PSSCs

All of the above can lead to the release of nuclear and chemical material to the environment.

Potential causes for fire events within the facility identified by DCS include the following:

• short circuits or equivalent event involving electrical equipment

• ignition or combustion of fixed or transient combustibles

• equipment that operates at high temperatures

• ignition of a solvent or other flammable/reactive chemical

Thirty-five separate events with potentially significant consequences were analyzed by DCS to

determine the bounding consequences from a potential fire event. These events were assigned

to 13 groups as follows with a unique prevention or mitigation strategy:

The AP process cells

AP/MP C3 glovebox area

C1 and/or C2 areas

— 3013 canister

— 3013 transport cask

— Fuel rod

— MOX fuel transport cask

— Waste container

— Transfer container

— Final C4 HEPA filter

Outside MOX fuel fabrication building

Facility-wide systems

Facility

Electrolyzer

These 13 groups are discussed below:

AP Process Cell (Fire)

The bounding event for the AP process cells events group in the fire events accident category

was determined to be a fire in the cell containing the liquid waste reception unit tanks. The
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material at risk was taken to be the maximum inventory of radioactive material in the cell

containing the tanks. A fire was postulated to occur in the process cell and consequences were

evaluated. The applicant determined this to be an above the 10 CFR 70.61(c) threshold

consequence event for the facility worker, environment, site worker, and IOC. The applicant has

opted to protect these receptors through a strategy of prevention. The PSSC for protection of

the facility worker was the process cell fire prevention features, the purpose of which is to ensure

that fires in the process cells are highly unlikely. The process cell fire prevention features

consist of the following:

• elimination of ignition sources, including electrical equipment and static electricity

• fire barriers to protect process cell areas

• elimination of all combustible materials from process cells containing aqueous solutions

• elimination of combustibles outside of process equipment in cells containing solvents

• maintenance of temperatures at levels to prevent creation of flammable vapors.

The process cell fire prevention features are also identified as the PSSC for protection of the

environment, site worker, and IOC. In addition, the process cell ventilation system’s passive

boundary and the C2 confinement system’s passive boundary provide defense-in-depth

protection to mitigate the potential consequences to the environment, site worker, and IOC.

Based on the applicant’s deterministic criteria and the NUREG-1718 Table A-5 descriptions

(enhanced administrative control with active and passive features), the staff considers this to be

an acceptable strategy for meeting the 10 CFR 70.61 performance requirements. 

AP/MP C3 Glovebox Area (Fire)

The bounding event for a fire in the AP/mixed oxide process (MP) C3 glovebox area in the fire

events accident category was a fire within the plutonium dioxide buffer storage area. The

material at risk was the maximum inventory of radioactive material within the fire area. The

specific cause of a fire in this area was not addressed, but the bounding event in this event

group was identified as a fire originating in a glovebox. The applicant determined this to be an

above the 10 CFR 70.61(c) threshold consequence event for the facility worker, environment,

site worker, and IOC. The applicant has opted to protect potentially affected workers, IOC, and

the environment through a strategy of mitigation. The PSSCs identified for protection of the

facility worker were facility worker actions and facility worker controls. The PSSCs identified for

protection of the environment, site worker, and IOC were the C3 confinement system, the active

portion of the C4 confinement system, and fire barriers. In addition, combustible loading

controls is also identified as a PSSC for protection of the the environment, site worker, and IOC

for fires involving storage gloveboxes. The primary protection of the worker will be early

detection of the fire and the ability to evacuate the area before a release. Although not credited

by the accident analysis, there will also be a fire suppression system in areas with dispersable

radioactive material. This suppression system will be classified as a PSSC. Consideration of

the warning time available from a fire before a breech in containment allows facility worker action

to be an acceptable strategy for meeting the 10 CFR 70.61 performance requirements with

regard to the facility worker. Based on the applicant’s deterministic criteria and the NUREG1718 Table A-5 descriptions (AEC, enhanced administrative control), the staff considers this to

be an acceptable strategy for meeting the 10 CFR 70.61 performance requirements.
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3013 Canister (Fire)

The bounding event for a fire affecting a 3013 canister event group in the C2 area is a fire in the

3013 storage area. The material at risk for this fire is the maximum inventory of radioactive

material in the fire area. The cause of the fire was ignition of transient combustibles. The

applicant determined this to be an above the 10 CFR 70.61(c) threshold consequence event for

the facility worker, environment, site worker, and IOC. The applicant has opted to protect the

facility worker, environment, site worker, and IOC through a strategy of prevention. The PSSC

identified for protection of the facility worker, environment, site worker, and IOC is combustible

loading controls, which are intended to limit the quantity of combustibles in a fire area containing

3013 canisters to ensure that the canisters are not adversely impacted by a fire. Based on the

applicant’s deterministic criteria and the NUREG-1718 Table A-5 descriptions (enhanced

administrative control), the staff considers this to be an acceptable strategy for meeting the 10

CFR 70.61 performance requirements.

3013 Transport Cask (Fire)

The bounding event for a fire affecting a 3013 transport cask in the C1 or C2 area was

determined to be a fire in the truck bay involving transport packages resulting in an energetic

breach of the containers and the dispersal of radioactive materials. The type of fire postulated

would be a fuel fire involving a truck. The material at risk was determined to be the maximum

inventory in the transport packages. The applicant determined this to be an above the 10 CFR

70.61(c) threshold consequence event for the facility worker, environment, site worker, and IOC. 

The applicant has opted to protect the facility worker, environment, site worker, and IOC 

through a strategy of prevention. The PSSCs identified for protection of the facility worker,

environment, site worker, and IOC are the 3013 transport cask, which will withstand the design

basis fire without breaching, and combustible loading controls, which will limit the quantity of

combustibles in a fire area containing 3013 transport casks to ensure that the cask design basis

fire is not exceeded. In addition to the identified PSSCs, there will also be a fire suppression

system which is considered an additional protective feature. Based on the applicant’s

deterministic criteria and the NUREG-1718 Table A-5 descriptions (PEC and enhanced

administrative control), the staff considers this to be an acceptable strategy for meeting the 10

CFR 70.61 performance requirements.

Fuel Rods (Fire)

The bounding event for a fire affecting fuel rods in the fire event accident category is a fire in the

fuel assembly storage area. The material at risk for this fire is the maximum inventory of

radioactive materials in the assembly storage area. Combustible loading in this area is low, but

the fire is still assumed to involve all of the radioactive materials in the storage area. The source

of the fire is considered to be electrical equipment and transient combustibles. The applicant

determined this to be an above the 10 CFR 70.61(c) threshold consequence event for the facility

worker, the environment, site worker, and IOC and has opted to protect the potential receptors

through a strategy of prevention. The PSSC identified for protection of these receptors is

combustible loading controls which will limit the quantity of combustibles in a fire area containing

fuel rods to ensure that the fuel rods are not adversely impacted by a fire. Based on the

applicant’s deterministic criteria and the NUREG-1718 Table A-5 descriptions (enhanced
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administrative control), the staff considers this to be an acceptable strategy for meeting the 10

CFR 70.61 performance requirements.

MOX Fuel Transport Cask (Fire)

The bounding event for a fire affecting the MOX fuel transport cask in the fire event accident

scenario was determined to be a fire in the fuel assembly truck bay. The source of the fire was

considered to be electrical equipment and transient combustibles. The material at risk was the

radioactive material in the transport casks. The applicant determined this to be an above the 10

CFR 70.61(c) threshold consequence event for the facility worker, environment, site worker, and

IOC. The staff independently evaluated this accident sequence and agrees to its categorization. 

The applicant has opted to protect potential receptors through a strategy of prevention. The

PSSCs identified for protection of the facility worker, environment, site worker, and IOC are the

MOX fuel transport cask, which will withstand the design basis fire without breaching, and

combustible loading controls, which are intended to limit the quantity of combustibles in a fire

area containing MOX fuel transport casks to ensure that the cask design basis fire is not

exceeded. In addition to the identified PSSCs, there will also be a fire suppression system which

is considered an additional protective feature. Based on the applicant’s deterministic criteria and

the NUREG-1718 Table A-5 descriptions (PEC and enhanced administrative control), the staff

considers this to be an acceptable strategy for meeting the 10 CFR 70.61 performance

requirements. 

Waste Container (Fire)

The bounding event for a fire affecting a waste container in the C1, C2, or C3 area event group

in the fire event accident category was determined to be a fire located in the assembly

packaging area. The material at risk was the maximum inventory of radioactive material in the

waste container. The source of the fire was considered to be electrical equipment and transient

combustibles. The applicant determined this to be an above the 10 CFR 70.61(c) threshold

consequence event for the facility worker, but a below the 10 CFR 70.61(c) threshold

consequence event for the environment, site worker, and IOC. The staff independently

evaluated this accident sequence and agrees to its categorization. The applicant has opted to

limit the dose to the facility worker using a strategy of mitigation. The PSSC identified for

protection of the facility worker is facility worker action to ensure that facility workers take proper

actions to limit dose. No PSSCs were identified by the applicant as being necessary to

adequately protect the environment, site worker, and IOC. Based on the nature of this event

(mitigated by rapid detection of fire and immediate worker responses), the staff considers this to

be an acceptable strategy for meeting the 10 CFR 70.61 performance requirements.

Transfer Container (Fire)

The bounding event for a fire affecting a transfer cask within the C1, C2, or C3 areas event

group in the fire event accident category was determined to be a fire in either the air locks,

corridors, stairways, safe areas, or liquid waste reception areas. The material at risk was the

maximum inventory in a transfer container. The source of the fire was identified as electrical

equipment, transient combustibles, or a HEPA filter. The applicant determined this to be an

above the 10 CFR 70.61(c) threshold consequence event for the facility worker, environment,

and site worker, but a below the 10 CFR 70.61(c) threshold consequence event for the IOC.
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The staff independently evaluated this accident sequence and agrees to its categorization. The

applicant has opted to limit dose to the facility worker, environment, and site worker using a

strategy of prevention. The PSSC identified for protection of the facility worker, environment,

and the IOC is combustible loading controls, which limit the quantity of combustibles in a fire

area containing transfer containers to ensure that the containers are not adversely impacted by

a fire. No PSSCs were identified by the applicant as being necessary to adequately protect the

site worker. Based on the nature of this event (prevented by an enhanced administrative

control), the staff considers this to be an acceptable strategy for meeting the 10 CFR 70.61

performance requirements.

C4 HEPA Filter (Fire)

The bounding event for a fire affecting the final C4 HEPA filter in the fire event accident category

is a fire which breaches the HEPA filter housing and allows material from the HEPA filters to

pass directly to the stack. The material at risk for this event is based on a conservative estimate

of material present on the C4 HEPA filters. The applicant determined this to be an above the 10

CFR 70.61(c) threshold consequence event for the facility worker, environment, site worker, and

IOC. The applicant has opted to limit dose to the facility worker, environment, site worker, and

IOC using a strategy of prevention. The PSSC identified for protection of the facility workers,

environment, site worker, and IOC are combustible loading controls. Based on the applicant’s

deterministic criteria and the NUREG-1718 Table A-5 descriptions (enhanced administrative

control), the staff considers this to be an acceptable strategy for meeting the 10 CFR 70.61

performance requirements. 

Fire outside of MFFF Building (Fire)

The bounding event for a fire originating outside of the MFFF building event group in the fire

event accident category was determined to be a fire involving diesel fuel storage, gasoline

storage, or the reagents processing building, such that the MFFF building structure is damaged

and radioactive material is released. The material at risk was the maximum inventory of

radioactive material in the MFFF that is susceptible to the effects of external fires. The applicant

determined this to be an above the 10 CFR 70.61(c) threshold consequence event for the facility

worker, environment, site worker, and IOC and has opted to meet the performance requirements

using a strategy of prevention. The PSSCs identified to protect all receptors are the MFFF

building structure, which is designed to maintain structural integrity and prevent damage to

internal PSSCs from external fires, the emergency diesel generator building structure, which is

designed to maintain structural integrity and prevent damage to internal PSSCs from fires

external to the structure, the emergency control room air-conditioning system, which will ensure

habitable conditions for operators, and the waste transfer line, which will prevent damage to the

line from external fires. Based on the applicant’s deterministic criteria and the NUREG-1718

Table A-5 descriptions (PECs and an AEC), the staff considers this to be an acceptable strategy

for meeting the 10 CFR 70.61 performance requirements.

Fire Affecting Facility-Wide Systems (Fire)

The bounding event for a fire affecting facility-wide systems (fires involving systems that cross

fire areas) in the fire event accident category was determined to be a fire involving the

pneumatic pipe automatic transfer system which results in a breach of confinement and the
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dispersal of radioactive material. The material at risk was the maximum inventory of radioactive

material in the pneumatic pipe automatic transfer system. The fire was postulated to be caused

by electrical equipment and transient combustibles. The applicant determined this to be an

above the 10 CFR 70.61(c) threshold consequence event for the facility worker and the

environment, but a below the 10 CFR 70.61(c) threshold consequence event for the site worker

and IOC. The staff independently evaluated this accident sequence and agrees to its

categorization. The applicant has opted to limit the dose to the facility worker using a strategy of

mitigation. The PSSCs identified for protection of facility workers are facility worker actions and

combustible loading controls. The primary protection of the worker will be early detection of the

fire and evacuation of the area before a release. The PSSC identified for protection of the

environment is combustible loading controls. No PSSCs were identified by the applicant as

being necessary to adequately protect the site worker and IOC. Based on the nature of this

event (mitigated by enhanced administrative controls and immediate worker responses), the

staff considers this to be an acceptable strategy for meeting the 10 CFR 70.61 performance

requirements.

Facility (Fire)

The bounding event for a facility fire which involves more than one fire area in the fire event

accident scenario was determined to be a fire in all process units and support units with

radioactive materials present. The source term is the maximum inventory in the facility

susceptible to a facility-wide fire. The applicant determined this to be an above the 10 CFR

70.61(c) threshold consequence event for the facility worker, environment, and IOC and has

opted to meet the performance requirements using a strategy of mitigation and prevention. The

PSSCs identified for protection of the facility worker are worker actions and fire barriers that will

contain the fires within the fire area. The PSSC identified for protection of the the environment,

site worker, and IOC is fire barriers. In addition to the identified PSSCs, there will also be a fire

suppression system designated as a PSSC where dispersable radioactive material is present.

Based on the applicant’s deterministic criteria and the NUREG-1718 Table A-5 descriptions

(PEC), the staff considers this to be an acceptable strategy for meeting the 10 CFR 70.61

performance requirements.

AP Electrolyzer (Fire)

The bounding event for a titanium fire in the AP electrolyzer was determined to be the energetic

breach of the AP electrolyzer and the dispersal of radioactive materials. The material at risk was

the maximum inventory in the AP dissolution units. The applicant determined this to be an

above the 10 CFR 70.61(c) threshold consequence event for the facility worker, environment,

and IOC and has opted to meet the performance requirements using a strategy of prevention. 

The PSSCs identified for the protection of the facility worker, environment, site worker, and IOC

are maintenance activity controls, the process safety control subsystem, and specified

electrolyzer components, such as the electrolyzer structure, sintered silicon nitride barrier, guide

sleeves, and polytetrafluorethylene insulator (PTFE). The safety function of the maintenance

activity controls is to isolate power from the electrolyzer when the electrolyzer is drained. The

safety function of the process control subsystem is to monitor the electrolyzer for faults that

could result in arcing or other imparting of electrical energy with the risk of titanium fire. The

safety function of the specified electrolyzer components is to provide physical separation and

electrical insulation between the electrolyzer components and structural integrity and stability to
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the electrolyzer system. The C3 confinement system, the C4 confinement system, and the fire

suppression and detection system provide defense in depth protection to mitigate potential

consequences to the environment, site worker, and IOC. Based on the applicant’s deterministic

criteria and the NUREG-1718 Table A-5 descriptions (AECs), the staff considers this to be an

acceptable strategy for meeting the 10 CFR 70.61 performance requirements.

5.1.6.3.3 Load-Handling Events

Load-handling events may occur during the operation of load-handling or lifting equipment

during normal operations or maintenance activities. Load-handling events may occur because

of the failure of handling equipment to lift or support the load, failure to follow designated load

paths, or toppling of loads. Consequences of load-handling events include possible damage to

handled loads, resulting in dispersal of radioactive and/or chemical materials, possible damage

to nearby equipment or structures, resulting in a loss of confinement and/or a loss of subcritical

conditions, and possible damage to process equipment or structures relied on for safety.

Twenty-eight separate events with potentially significant consequences were analyzed by DCS

to determine the bounding consequences from a potential load-handling event. These events

were assigned to 12 groups as follows with a unique prevention or mitigation strategy:

The AP process cells

AP/MP C3 glovebox area

C1 and/or C2 areas

— 3013 Canister in the C2 confinement area

— 3013 Transport Cask

— Fuel rod in the C2 confinement area

— MOX fuel transport cask 

— Transfer container 

— Waste container

— Final C4 HEPA filter

C4 Confinement

Outside MFFF building

Facility-wide systems

These 12 groups are discussed below:

AP Process Cells (Load Handling)

The bounding event for load-handling events in the AP process cells event group in the loadhandling event accident category was determined to be an event in the cell containing the liquid

waste reception unit. The material at risk was the maximum inventory of radioactive material in

the AP process cell containing the liquid waste reception unit. The load-handling event is

postulated to result in a breach of the americium reception tank and subsequent release of

americium in solution because of vessels in the process cell being impacted by a lifting device or

a lifted load. The applicant determined this to be an above the 10 CFR 70.61(c) threshold

consequence event for the facility worker, environment, site workers, and IOC. The applicant

has opted to meet the performance requirements using a strategy of mitigation. The PSSCs for

the protection of facility workers are the process cells which contain fluid leaks using drip trays
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and the process cell entry controls which will prevent entry during normal conditions and assure

that worker dose limits are not exceeded during maintenance operations. The PSSC for

protection of the environment, site worker, and IOC is the process cell exhaust system. The

safety function of the process cell exhaust system is to ensure that a negative pressure exists

between the process cell areas and the C2 areas, as well as to ensure that the process cell

exhaust system is effectively filtered. The C2 confinement system’s passive boundary provides

defense-in-depth protection for the environment, site worker, and IOC. Based on the applicant’s

deterministic criteria and the NUREG-1718 Table A-5 descriptions (administrative control, AECs,

and PECs), the staff considers this to be an acceptable strategy for meeting the 10 CFR 70.61

performance requirements. 

AP/MP C3 Glovebox (Load Handling)

The bounding event for load-handling events in the AP/MP C3 glovebox area event group in the

load-handling accident category was determined to be an event which occurs within the

gloveboxes that contain jar storage and handling of the MOX powder workshop from a breach of

the glovebox. The material at risk was the maximum inventory of radioactive material in the

glovebox. The breach of the glovebox is from a lifting device or a lifted load. The applicant

determined this to be an above the 10 CFR 70.61(c) threshold consequence event for the facility

worker, environment, site worker, and IOC and has opted to meet the performance requirements

using a strategy of prevention and mitigation. The PSSCs for protection of the facility worker

and the environment are materials-handling controls, which are intended to prevent impacts to

the glovebox during normal operations from loads outside or inside the glovebox that could

exceed the glovebox design basis, materials-handling equipment, which is engineered to prevent

impacts to the glovebox, the glovebox, which maintains confinement integrity for design basis

impacts, and facility worker controls (facility worker during maintenance operations). An

additional safety function of the materials-handling controls is to prevent potential

overpressurization of the reusable plutonium dioxide cans caused by radiolysis or oxidation of

plutonium (III) oxalate and its subsequent impact to the glovebox. The PSSC for protection of

the site worker and IOC is the C3 confinement system. The C2 confinement system also

provides defense-in-depth protection for the site worker and IOC. Based on the applicant’s

deterministic criteria and the NUREG-1718 Table A-5 descriptions (AEC, PEC, and

administrative controls), the staff considers this to be an acceptable strategy for meeting the 10

CFR 70.61 performance requirements. 

3013 Canister (Load Handling)

The bounding event for the 3013 canister event group (C2 area) in the load-handling accident

category was the drop of one 3013 container onto another 3013 container, each containing

unpolished plutonium dioxide in powder form. The material at risk was the amount of radioactive

material in two 3013 canisters. The cause of the event would likely be human error or

equipment failure during a hoisting operation. The applicant determined this to be an above the

10 CFR 70.61(c) threshold consequence event for the facility worker, environment, site worker,

and IOC. The applicant has opted to meet the performance requirements using a strategy of

prevention. The staff independently evaluated this accident sequence and agrees to its

categorization. The PSSCs identified for protection of the facility worker, environment, site

worker, and IOC are the 3013 canister and materials-handling controls. The C2 confinement

system’s passive boundary provides defense-in-depth for the environment, site worker, and IOC. 
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Based on the applicant’s deterministic criteria and the NUREG-1718 Table A-5 descriptions

(administrative control and PEC), the staff considers this to be an acceptable strategy for

meeting the 10 CFR 70.61 performance requirements. 

3013 Transport Cask (Load Handling)

The bounding event for the 3013 transport cask event group (C1 or C2 area) in the loadhandling accident category was the drop of a 3013 transport cask containing unpolished

plutonium dioxide in powdered form onto another 3013 transport cask. The material at risk was

the maximum inventory of radioactive material in two 3013 transport canisters. The applicant

determined this to be an above the 10 CFR 70.61(c) threshold consequence event for the facility

worker, environment, site worker, and IOC. The applicant has opted to meet the performance

requirements using a strategy of prevention. The staff independently evaluated this accident

sequence and agrees to its categorization. The PSSCs identified for protection of the facility

worker, environment, site worker, and IOC are the 3013 transport cask and materials-handling

controls. The C2 confinement system’s passive boundary provides defense-in-depth for the site

worker, IOC, and the environment. Based on the applicant’s deterministic criteria and the

NUREG-1718 Table A-5 descriptions (PEC and administrative controls), the staff considers this

to be an acceptable strategy for meeting the 10 CFR 70.61 performance requirements.

Fuel Rods in C2 Area (Load Handling)

The bounding event for the fuel rods in the C2 area event group in the load-handling accident

category was the drop of a strongback containing three fuel assemblies containing MOX (6

percent). The material at risk was the maximum inventory of three fuel rod assemblies. The

cause of this event would probably be human error or equipment failure. The applicant

determined this to be an above the 10 CFR 70.61(c) threshold consequence event for the facility

worker, but a below the 10 CFR 70.61(c) threshold consequence event for the environment, site

worker, and IOC. The staff independently evaluated this accident sequence and agrees to its

categorization. The applicant has opted to use a strategy of mitigation to protect the facility

worker. The PSSC identified for protection of the facility worker is facility worker actions. No

PSSCs were identified by the applicant as being necessary to adequately protect the

environment, site worker, and IOC. However, the C2 confinement system passive boundary

provides defense-in-depth for the environment, site worker, and IOC. Because a release could

occur without warning, the applicant provided dose calculations which were reviewed on site and

found to be acceptable (NRC, 2002). Based on the nature of this event (limited initial release

and immediate worker response), the staff considers this to be an acceptable strategy for

meeting the 10 CFR 70.61 performance requirements.

MOX Fuel Transport Cask (Load Handling)

The bounding event for the MOX fuel transport cask event group (C1 or C2 areas) in the loadhandling accident category was determined to be the drop of one MOX fuel transport cask

containing up to three MOX fuel assemblies. The cause of this event would probably be human

error or equipment failure. The material at risk was determined to be the maximum inventory of

one fuel assembly transport package. The applicant determined this to be an above the 10 CFR

70.61(c) threshold consequence event for the facility worker and the environment, but a below

the 10 CFR 70.61(c) threshold consequence event for the site worker and IOC. The staff
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independently evaluated this accident sequence and agrees to its categorization. The applicant

has decided to meet the performance requirements of 10 CFR 70.61 for the facility worker and

the environment using a strategy of prevention. The PSSCs identified for protection of the

facility worker and the environment are the MOX fuel transport cask and materials-handling

controls. No PSSCs were identified by the applicant as being necessary to adequately protect

the site worker and IOC. However, the MOX fuel transport cask provides defense-in-depth

protection for the site worker and IOC. Based on the nature of this event (prevented by a PEC

and administrative controls), the staff considers this to be an acceptable strategy for meeting the

10 CFR 70.61 performance requirements.

Waste Container (Load Handling)

The bounding event for the waste container event group (C1, C2, or C3 area) in the loadhandling accident category is a damaged waste drum in the assembly packaging (truck bay)

area caused by human error or equipment failure. The material at risk was determined to be the

maximum inventory of radiological material in a waste container. The applicant determined this

to be an above the 10 CFR 70.61(c) threshold consequence event for the facility worker, but a

below the 10 CFR 70.61(c) threshold consequence event for the environment, site worker, and

IOC. The staff independently evaluated this accident sequence and agrees to its categorization. 

The applicant has decided to meet the performance requirements of 10 CFR 70.61 for the

facility worker using a strategy of mitigation. The PSSCs identified for protection of the facility

worker are worker actions. No PSSCs were identified by the applicant as being necessary to

adequately protect the environment, site worker, and IOC. For drops in the C2 area, the C2

confinement passive boundary provides defense-in-depth protection for the environment, site

worker, and IOC. However, because a release could occur without warning, the applicant

provided dose calculations which were reviewed on site and found to be acceptable (NRC,

2002). Based on the nature of this event (mitigated by limited initial release and immediate

worker responses), the staff considers this to be an acceptable strategy for meeting the 10 CFR

70.61 performance requirements.

Transfer Container (Load Handling)

The bounding event for the transfer container event group (C2 area) in the load-handling

accident category was the drop of a transfer container containing a HEPA filter with plutonium

dioxide in powdered form. The material at risk was determined to be the maximum inventory in

a HEPA filter. The applicant determined this to be an above the 10 CFR 70.61(c) threshold

consequence event for the facility worker, environment, site worker, and IOC. The applicant has

opted to meet the performance requirements using a strategy of prevention. The staff

independently evaluated this accident sequence and agrees to its categorization. The PSSCs

identified for protection of the facility worker, environment, site worker, and IOC are the transfer

container and materials-handling controls. Based on the applicant’s deterministic criteria and

the NUREG-1718 Table A-5 descriptions (PEC and administrative control), the staff considers

this to be an acceptable strategy for meeting the 10 CFR 70.61 performance requirements.

Final C4 HEPA Filter (Load Handling)

The bounding event in the final C4 HEPA filter event group in the load-handling accident

category was determined to be the impacting of the final C4 filters by a load that breaches the
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HEPA filter housing and allows material from the HEPA filters to pass directly to the stack. The

cause of this event would probably be human error or equipment failure around the ventilation

system. The material at risk was determined to be the radiological material contained in the

HVAC system and filters. The applicant determined this to be an above the 10 CFR 70.61(c)

threshold consequence event for the facility worker, environment, site worker, and IOC. The

staff independently evaluated this accident sequence and agrees to its categorization. The

applicant has decided to meet the performance requirements of 10 CFR 70.61 for the facility

worker, environment, site worker, and IOC using a strategy of prevention. Materials-handling

controls were identified as the PSSC for protection of these receptors. In addition, the applicant

stated that in the current design and operations, there are no cranes or heavy equipment in the

vicinity of the C4 final filters that could cause a load-handling event. Thus, there are no credible

load-handling events during normal operations. During maintenance operations, maintenance

will only be performed on out-of-service trains, which will prevent a release to the stack. The C2

confinement system’s passive boundary provides defense-in-depth protection for the

environment, site workers, and IOC for load-handling events that occur in the C2 areas where

the final C4 filters are located. Based on the applicant’s deterministic criteria and the NUREG1718 Table A-5 descriptions (enhanced administrative control), the staff considers this to be an

acceptable strategy for meeting the 10 CFR 70.61 performance requirements. 

 

C4 Confinement (Load Handling)

The bounding event in the C4 confinement event group in the load-handling accident category

was determined to be a spill of unpolished plutonium powder that occurs inside the glovebox, but

does not result in a breach of the glovebox. The cause of this event would probably be human

error or equipment failure during load-handling operations inside the glovebox. The material at

risk would be the maximum inventory in the glovebox. The applicant determined this to be an

above the 10 CFR 70.61(c) threshold consequence event for the facility worker, environment,

site worker, and IOC. The applicant has decided to meet the performance requirements of 10

CFR 70.61 for the facility worker, environment, site worker, and IOC using a strategy of

mitigation. The C4 confinement system is identified as the PSSC for protection of the facility

worker, environment, site worker, and IOC. The safety functions of the C4 confinement system

in this event are to ensure that the C4 exhaust is effectively filtered and to maintain a negative

glovebox differential pressure between the glovebox and the interfacing systems. The C3

confinement system provides defense-in-depth protection for the environment, site worker, and

IOC. Based on the applicant’s deterministic criteria and the NUREG-1718 Table A-5

descriptions (high dependability AEC), the staff considers this to be an acceptable strategy for

meeting the 10 CFR 70.61 performance requirements.

Outside MFFF Buildings (Load Handling)

The bounding event in the load-handling event category outside the MOX fuel fabrication

building is an event involving the waste transfer line. The cause of this event would probably be

human error or equipment failure. The material at risk was determined to be the maximum

inventory in the waste tank. The applicant determined this to be an above the 10 CFR 70.61(c)

threshold consequence event for the facility worker, environment, site worker, and IOC and has

opted to meet the performance requirements using a strategy of prevention. The PSSC

identified for protection of the facility worker, environment, site worker, and IOC is the waste

transfer line which is double walled and buried. Based on the applicant’s deterministic criteria
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and the NUREG-1718 Table A-5 descriptions (robust PECs), the staff considers this to be an

acceptable strategy for meeting the 10 CFR 70.61 performance requirements.

Facility Wide (Load Handling)

The bounding event in the load-handling event category for the facility-wide event class is the

breach of the facility structure from a heavy load resulting in a breach of primary confinement or

in a breach of a container holding nuclear materials. The cause of this event would probably be

human error or equipment failure. The material at risk was determined to be the maximum

inventory in a container or primary confinement. The applicant determined this to be an above

the 10 CFR 70.61(c) threshold consequence event for the facility worker, environment, site

worker, and IOC and has opted to meet the performance requirements using a strategy of

prevention. The PSSCs identified for protection of the facility worker, environment, site worker,

and IOC are the MOX fuel fabrication building structure, which is designed to withstand the

effects of load drops that could potentially impact radiological material, and materials-handling

controls that would prevent load-handling events that could breach primary confinements. Based

on the applicant’s deterministic criteria and the NUREG-1718 Table A-5 descriptions (PEC and

administrative control), the staff considers this to be an acceptable strategy for meeting the 10

CFR 70.61 performance requirements.

5.1.6.3.4 Explosion Events

Explosions are postulated to occur inside of the MOX fuel fabrication building from process

operations, outside the MOX fuel fabrication building from nearby support facilities and the

storage of chemicals on the facility site, and from laboratory operations. The following are

considered to be the major consequences of explosions:

• release of radioactive materials or chemicals to the environment

• damage to a confinement boundary

• damage to equipment contributing to dynamic confinement

• loss of subcritical conditions

• damage to civil structures

• damage to other PSSCs

All of the above may result in the release of nuclear materials or chemicals to the environment.

Twenty-three separate events with potentially significant consequences were analyzed by DCS

to determine the bounding consequences from a potential explosion event. These events were

assigned to the following 15 groups with a unique prevention or mitigation strategy:

Hydrogen explosion

Steam over-pressure explosion

Radiolysis induced explosion

HAN explosion

Hydrogen peroxide explosion

Solvent explosion

Tributyl phosphate (TBP)-Nitrate (Red oils) explosion

AP vessel over-pressurization explosion
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Pressure vessel over pressurization explosion

Hydrazoic acid explosion

Metal azide explosion

Pu (VI) Oxalate explosion

Electrolysis related explosion

Laboratory explosion

Outside explosion 

These 15 groups are discussed below:

Hydrogen (Explosion)

The bounding event for the hydrogen explosion event group is the explosion of hydrogen and

oxygen in a sintering furnace or sintering furnace room. The cause of this event would probably

be excessive hydrogen in the furnace and air leakage into the furnace or hydrogen accumulation

into the room. The applicant determined this to be an above the 10 CFR 70.61(c) threshold

consequence event for the facility worker, environment, site worker, and IOC and has opted to

meet the performance requirements using a strategy of prevention. The PSSC identified for

prevention of this event is the process safety control subsystem with the safety function of

preventing an explosive accumulation of hydrogen vapors. Specific control approaches may

consist of hydrogen monitors outside the furnace, oxygen monitors inside the furnace, limiting

the hydrogen content in the hydrogen-argon mixture, and crediting dilution flow from the high

depressurization exhaust or very high depressurization exhaust systems to prevent an explosive

mixture of hydrogen. The specific functions and systems selected will be identified as IROFS in

the ISA. Based on the applicant’s deterministic criteria and the NUREG-1718 Table A-5

descriptions (AEC), the staff considers this to be an acceptable strategy for meeting the 10 CFR

70.61 performance requirements. 

Steam (Explosion)

A steam explosion is postulated to occur in the sintering furnace as a result of humidifier water in

the inlet gas stream. The cause of the event is expected to be a failure of the water level

controller in the humidifier. The material at risk was determined to be the maximum inventory of

radiological materials in the sintering furnace. The applicant determined this to be an above the

10 CFR 70.61(c) threshold consequence event for the facility worker, environment, site worker,

and IOC and has opted to meet the performance requirements using a strategy of prevention.

The PSSC identified to protect the facility worker, environment, site worker, and IOC is the

process safety control subsystem which will ensure the isolation of sintering furnace humidifier

water flow on high-water level. Because of the design of the water cooling coils of the sintering

furnace, water will not enter the sintering furnace through any other path. Based on the

applicant’s deterministic criteria and the NUREG-1718 Table A-5 descriptions (AEC), the staff

considers this to be an acceptable strategy for meeting the 10 CFR 70.61 performance

requirements. 
5-36

Radiolysis (Explosion)

The bounding events in the radiolysis-induced explosion event group were explosions caused by

radiolysis-induced hydrogen buildup in the vapor space of an AP vessel tank or piping,

radiolysis-induced hydrogen buildup in the vapor space of a raffinates tank (in an AP process

cell), and radiolysis-induced hydrogen accumulation in a waste container containing

hydrocarbons. The applicant determined all three event sequences to be above the 10 CFR

70.61(c) threshold consequence events for the facility worker, environment, site worker, and

IOC. The applicant has opted to meet the performance requirements using a strategy of

prevention. The PSSCs identified for protection of the facility worker, environment, site worker,

and IOC are the offgas treatment system, the instrument air system (scavenging system), and

waste containers. The waste container contains a filter to allow the escape of radiolytic

hydrogen before explosive mixtures occur as the PSSC to prevent an explosion inside the waste

container. Based on the applicant’s deterministic criteria and the NUREG-1718 Table A-5

descriptions (AECs and PEC), the staff considers this to be an acceptable strategy for meeting

the 10 CFR 70.61 performance requirements. 

HAN (Explosion)

The hydroxyl nitrate (HAN) explosions that could potentially occur within the facility may be

characterized by one of the following two cases:

(1) process vessels containing HAN and hydrazine nitrate without NOx addition

(2) process vessels containing HAN and hydrazine nitrate with NOx addition

The applicant provided a hazard analysis and selected a PSSC for each case separately. The

material at risk, however, was the same for both events and consisted of the maximum

radiological inventory in AP vessels, tanks, and piping. The applicant has determined that the

consequences from this event sequence for each of the cases are an above the 10 CFR

70.61(c) threshold consequence event for the facility worker, environment, site worker, and IOC

and has opted to meet the performance requirements using a strategy of prevention. The

applicant identified two PSSCs as required for the first case. These are the process safety

control subsystem, which ensures that the temperature of the solution containing HAN is limited

to temperatures that are within safety limits, and the chemical safety control, which ensures that

the concentration of nitric acid, metal impurities, hydrazine nitrate, and HAN introduced in the

process are within safety limits. In the second case, the chemical safety control is identified to

limit the concentration of HAN, hydrazine nitrate, and hydrazoic acid and the offgas treatment

system is identified to provide an exhaust path for offgases. Based on the applicant’s

deterministic criteria and the NUREG-1718 Table A-5 descriptions (enhanced administrative

control and AEC), the staff considers this to be an acceptable strategy for meeting the 10 CFR

70.61 performance requirements.

Hydrogen Peroxide (Explosion)

The bounding event in the hydrogen peroxide explosion class was determined to be an event

involving hydrogen peroxide in AP vessels, tanks, and piping which results in an energetic

breach of the vessels, tanks, and piping and results in a loss of confinement and dispersal of

nuclear materials. The applicant determined this to be an above the 10 CFR 70.61(c) threshold
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consequence event for the facility worker, environment, site worker, and IOC and has opted to

meet the performance requirements using a strategy of prevention. The PSSC identified to

protect the facility worker, environment, site worker, and IOC is the chemical safety control to

ensure that explosive concentrations of hydrogen peroxide do not occur. This administrative

control consists of a certified analysis by the manufacturer under an approved quality assurance

plan, an analysis upon receipt at the facility, and an analysis after mixing and diluting. Based on

the applicant’s deterministic criteria and the NUREG-1718 Table A-5 descriptions (enhanced

administrative control), the staff considers this to be an acceptable strategy for meeting the 10

CFR 70.61 performance requirements.

Solvent (Explosion)

The bounding event in the solvent explosion class was determined to be a process-related

explosion involving solvents in AP vessels, tanks, and piping which results in an energetic

breach of the vessels, tanks, and piping and a loss of confinement and dispersal of nuclear

materials. The applicant determined this to be an above the 10 CFR 70.61(c) threshold

consequence event for the facility worker, environment, site worker, and IOC. The applicant has

opted to meet the performance requirements using a strategy of prevention. The staff

independently evaluated this accident sequence and agrees to its categorization. The PSSCs

identified to protect the facility worker, environment, site worker, and IOC are the process safety

control subsystem, to ensure that the temperature of the solutions containing solvents do not

exceed the temperature at which the gaseous phase becomes flammable, the process cell fire

prevention features, to ensure that fires in process cells are highly unlikely; and the offgas

treatment system, which will provide exhaust to ensure that an explosive buildup of explosive

vapors does not occur. Based on the applicant’s deterministic criteria and the NUREG-1718

Table A-5 descriptions (AECs and PECs), the staff considers this to be an acceptable strategy

for meeting the 10 CFR 70.61 performance requirements. 

TBP-Nitrate (Red Oils) (Explosion)

The bounding event in the TBP-nitrate (red oil) explosion class is a process-related chemical

explosion involving red oil formation in the AP boiler, vessel, or tank and results in a loss of

confinement and dispersal of nuclear materials. The applicant determined this to be an above

the 10 CFR 70.61(c) threshold consequence event for the facility worker, environment, site

worker, and IOC and has opted to meet the performance requirements using a strategy of

prevention. The PSSCs identified for prevention of this event are the process safety control

subsystem, chemical safety control, and offgas treatment system. The purpose of the process

safety control subsystem is to ensure that the evaporator process temperature is maintained

within safe limits and to control the residence time of organics in the presence of oxidizers,

radiation fields, and high temperatures. The chemical safety control ensures that quantities of

organics are limited from entering process vessels containing oxidizing agents and at potentially

high temperatures and ensures that a diluent is used that is not very susceptible to either

nitration or radiolysis. The offgas treatment system provides an exhaust path for the removal of

gases in process vessels. Based on the applicant’s deterministic criteria and the NUREG-1718

Table A-5 descriptions (PECs, AECs, and administrative controls), the staff considers this to be

an acceptable strategy for meeting the 10 CFR 70.61 performance requirements.
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AP Vessel Overpressurization (Explosion)

The bounding events in AP vessel overpressurization explosion class were determined to be the

overpressurization of AP tanks, vessels, and piping postulated to result from increases in the

temperature of exothermic chemical reactions of solutions into tanks or vessels within the facility. 

The applicant determined this to be an above the 10 CFR 70.61(c) threshold consequence event

for the facility worker, environment, site worker, and IOC and has opted to meet the performance

requirements using a strategy of prevention. The PSSCs identified for prevention of this event

are the fluid transport systems, which will insure that vessels, tanks, and piping are designed to

prevent process deviations from creating overpressurization events, the offgas treatment

system, which will provide an exhaust path for the removal of gases in process vessels, and

chemical safety controls to ensure control of the chemical makeup of the reagents and ensure

segregation/separation of vessels/components from incompatible chemicals. Based on the

applicant’s deterministic criteria and the NUREG-1718 Table A-5 descriptions (PECs, AECs, and

administrative controls), the staff considers this to be an acceptable strategy for meeting the 10

CFR 70.61 performance requirements.

Pressure Vessel Overpressurization (Explosion)

The bounding event in the pressure vessel overpressurization explosion class is an explosion

related to the overpressurization of gas bottles, tanks, or receivers which could impact primary

confinements and result in a release of radioactive material. The applicant determined this to be

an above the 10 CFR 70.61(c) threshold consequence event for the facility worker, environment,

site worker, and IOC and has opted to meet the performance requirements using a strategy of

prevention. The PSSCs identified for prevention of this event are the pressure vessel controls,

which ensure that primary confinement is protected from the impact of pressure vessel failures. 

Pressure vessels will be located away from PSSCs or otherwise protected so that a failure of

any vessel would have no impact on the ability of the PSSC to perform its safety function. 

Based on the applicant’s deterministic criteria and the NUREG-1718 Table A-5 descriptions

(PECs and administrative controls), the staff considers this to be an acceptable strategy for

meeting the 10 CFR 70.61 performance requirements.

Hydrazoic Acid (Explosion)

The bounding event in the hydrazoic acid explosion sequence was a process-related chemical

explosion involving HAN/nitric acid in the AP vessels, tanks, and piping (in AP process cells or

gloveboxes) which results in a breach of the AP vessels, tanks, and piping. The material at risk

was the maximum inventory of radiological material in AP vessels, tanks, and piping. The

applicant determined this to be an above the 10 CFR 70.61(c) threshold consequence event for

the facility worker, environment, and individuals outside the controlled area boundary and has

opted to meet the performance requirements using a strategy of prevention. The PSSCs

identified for protection of the facility worker, environment, site worker, and IOC are the chemical

safety control and the process safety control subsystem. The function of the chemical safety

control is to (1) assure that the proper concentration of hydrazine nitrate is introduced to the

system, limiting the quantity of hydrazoic acid produced, and (2) ensure that hydrazoic acid is

not accumulated in the process or propagated into the acid recovery and oxalic mother liquors

recovery units by either taking representative samples in upstream units or by crediting the

neutralization process within the solvent recovery unit. The safety function of the process safety
5-39

control subsystem is to limit the temperature of the solution, thereby limiting the evaporation rate

and resulting vapor pressure of hydrazoic acid so that an explosive concentration of hydrazoic

acid does not occur. Based on the applicant’s deterministic criteria and the NUREG-1718 Table

A-5 descriptions (enhanced administrative control and AEC), the staff considers this to be an

acceptable strategy for meeting the 10 CFR 70.61 performance requirements.

Metal Azide Explosions

The bounding event in the metal azide explosion category was a process-related chemical

explosion involving an azide (other than hydrazoic acid) in an AP boiler, vessel, or tank (in an AP

cell or glovebox) that results in an energetic breach of the AP boiler, vessel, or tank. The

material at risk was the maximum inventory of radiological material in AP vessels, tanks, and

piping. The applicant determined this to be an above the 10 CFR 70.61(c) threshold

consequence event for the facility worker, environment, site worker, and IOC and has opted to

meet the performance requirements using a strategy of prevention. The PSSCs identified for

protection of the receptors are the chemical safety control and the process safety control

subsystem. The safety functions of the chemical safety control are to (1) ensure that metal

azides are not added to high temperature process equipment, and (2) ensure that the sodium

azide has been destroyed before transfer of the alkaline waste into the acidic high alpha waste

of the waste recovery unit. The safety function of the process safety control subsystem is to

ensure that metal azides are not exposed to temperatures that would supply sufficient energy to

overcome the activation energy needed to initiate the energetic azide decomposition and limit

and control conditions under which dryout can occur. Based on the applicant’s deterministic

criteria and the NUREG-1718 Table A-5 descriptions (AEC and enhanced administrative

control), the staff considers this to be an acceptable strategy for meeting the 10 CFR 70.61

performance requirements. 

Plutonium (VI) Oxalate Explosion

The bounding event for the plutonium (VI) oxalate explosion category was a process-related

chemical explosion involving plutonium (VI) in the calcining furnace results in an energetic

breach of the furnace and glovebox and the dispersal of radiological materials. The material at

risk was the maximum inventory of radiological material in the AP vessels, tanks, and piping. 

The applicant determined this to be an above the 10 CFR 70.61(c) threshold consequence event

for the facility worker, environment, site worker, and IOC and has opted to meet the performance

requirements using a strategy of prevention. The PSSC identified for protection of the facility

worker, site worker, environment, and IOC is the chemical safety control. The safety function of

the chemical safety control is to perform a measurement of the valency of the plutonium before

the addition of oxalic acid to the oxalic precipitation and oxidation unit to ensure that plutonium

(IV) cannot be formed. In addition, the design basis for the calciner will assure that the rapid

decomposition of any plutonium (VI) oxalate that may enter the calciner will not challenge the

calciner vessel’s integrity. Based on the applicant’s deterministic criteria and the NUREG-1718

Table A-5 descriptions (enhanced administrative control), the staff considers this to be an

acceptable strategy for meeting the 10 CFR 70.61 performance requirements. 
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Electrolysis-Related Explosion

The bounding event for the electrolysis-related explosion category was the explosion of

hydrogen in the vapor space of the electrolyzer. The material at risk was the maximum

inventory of radiological material in the AP vessels, tanks, and piping. The applicant determined

this to be an above the 10 CFR 70.61(c) threshold consequence event for the facility worker,

environment, site worker, and IOC and has opted to meet the performance requirements using a

strategy of prevention. The PSSCs identified for protection of the facility worker, environment,

site worker, and IOC are the process safety control subsystem and specified electrolyzer

components, such as the electrolyzer structure, sintered silicon nitride barrier, and

polytetrafluorethylene insulator (PTFE). The function of the process safety control subsystem is

to ensure that the normality of the acid is sufficiently high to ensure that the offgas is not

flammable and to limit excessive generation of hydrogen. The process safety control subsystem

also has the function of monitoring the electrolyzer for electrical faults that could result in arcing

or other imparting of electrical energy with the risk of a titanium fire and hydrogen explosion. 

The safety function of the specified electrolyzer components is to provide physical separation

and electrical insulation between the electrolyzer components and structural integrity and

stability to the electrolyzer system. Based on the applicant’s deterministic criteria and the

NUREG-1718 Table A-5 descriptions (AECs), the staff considers this to be an acceptable

strategy for meeting the 10 CFR 70.61 performance requirements. 

Laboratory (Explosion)

The bounding event for the laboratory explosion class is an explosion within the MFFF laboratory

involving flammable, explosive, or reactive chemicals which results in a dispersal of radiological

material. The radiological material assumed to be dispersed is the maximum inventory in the

laboratory. The applicant determined this to be an above the 10 CFR 70.61(c) threshold

consequence event for the facility worker, environment, site worker, and IOC and has opted to

meet the performance requirements using a strategy of prevention and mitigation. The PSSCs

for protection of the facility worker are the chemical safety control, laboratory material controls,

and facility worker actions. The function of the chemical safety control is to ensure control of the

chemical makeup of the reagents and ensure segregation/separation of vessels/components

from incompatible chemicals. The safety function of the laboratory material controls is to

minimize quantities of hazardous chemicals in the laboratory and to minimize quantities of

radioactive materials in the laboratory. The function of facility worker actions is to ensure that

facility workers take proper actions to limit radiological/chemical exposure. The PSSC identified

for protection of the environment, site worker, and IOC is the C3 confinement system which

provides filtration to mitigate dispersions from the C3 areas. Based on the applicant’s

deterministic criteria and the NUREG-1718 Table A-5 descriptions (enhanced administrative

controls and AEC), the staff considers this to be an acceptable strategy for meeting the 10 CFR

70.61 performance requirements.

Outside (Explosion)

The bounding events for the outside explosion class were determined to be explosions in the

reagent processing building, gas storage area, emergency diesel generator building, standby

diesel generator building, and the access control building. The applicant determined this to be

an above the 10 CFR 70.61(c) threshold consequence for the facility worker, environment, site
5-41

worker, and IOC and has opted to meet the performance requirements using a strategy of

prevention of a release. The PSSCs identified are the waste transfer line, which is designed to

prevent damage to the line during an explosion, the MOX fuel fabrication building structure,

which is designed to maintain structural integrity and prevent damage to internal PSSCs from

explosions external to the structure, and the emergency diesel generator building structure,

which is designed to maintain structural integrity and prevent damage to internal PSSCs from

explosions external to the structure. Also identified as a PSSC is the hazardous material

delivery controls which ensure that the quantity of delivered hazardous material and its proximity

to the fuel fabrication building structure, emergency generator building structure, and the waste

transfer line are controlled to within the bounds of the values used to demonstrate that the

consequences of outside explosions are acceptable. Based on the applicant’s deterministic

criteria and the NUREG-1718 Table A-5 descriptions (robust PECs and administrative controls),

the staff considers this to be an acceptable strategy for meeting the 10 CFR 70.61 performance

requirements.

5.1.6.3.5 Criticality Events

A criticality event is characterized by a self-sustaining fission chain reaction and can potentially

release a large amount of energy over a short period of time. When fissionable materials, such

as 235 U or 239 Pu are present in sufficient quantities, a self-sustaining fission chain reaction may

be attained depending on the size and shape of the fissionable materials, the nature of solvents

or diluent, and the proximity of potential reflectors.

The most immediate potential consequences from a criticality event is direct radiation exposure

to the facility worker. Distance from the event normally protects persons beyond the controlled

area boundary. Shielding materials can also reduce the dose. 

Criticality accidents may be caused by violation of safety limits such as the following:

• geometry control

• mass control

• density control

• isotopics control

• reflection control

• moderation control

• concentration control

• interaction control

• neutron absorber control

• volume control

• heterogeneity control

• process variable control

The applicant will use a strategy of prevention to protect the facility worker from a criticality

accident. The applicant identified criticality control as the only PSSC for its prevention strategy. 

The staff’s review and evaluation of the applicant’s analysis of criticality events is discussed in

Chapter 6 of the FSER.
5-42

5.1.6.3.6 Chemical Events

The applicant evaluated chemical event scenarios having the potential to affect radiological

safety by considering a range of initial conditions and failure modes of storage containers and

associated systems. The revised CAR provided the following release scenarios for chemical

events:

• leaks and ruptures involving equipment vessels and piping

• evaporating pools formed by spills and tank failures

• flashing and evaporating liquified gases from pressurized storage

Fourteen chemical events with potentially significant radiological consequences were analyzed

by DCS to determine the bounding consequences of such events. These events were assigned

to two groups with a unique prevention or mitigation strategy. These two groups are discussed

below.

Chemical Hazards

Events involving chemical hazards are chemical releases from vessels, tanks, pipes, or transport

containers, either internal or external to the MOX fuel fabrication building. The primary potential

for external releases would be from the reagent processing building (BRP). For this group of

events, only chemical consequences that impact radiological safety or MFFF operations, and

which may result in a radioactive material release, were considered. The applicant determined

that the only chemical releases that could result in radiological releases would be chemical

releases that could affect workers providing monitoring functions in the control room. The

applicant has assumed that such radiological releases could be above threshold levels for

receptors and has opted to meet the performance requirements through a strategy of

prevention. The PSSC identified to prevent a radioactive release is the emergency control room

air-conditioning system. 

In addition to the potential for radiological releases through the incapacitation of workers

performing critical functions (highly unlikely), there is also the potential for other facility workers

to receive a chemical dose which could be injurious to their health and exceed the 10 CFR 70.61

performance requirements. These releases could occur from pipes and process vessels in the

areas containing the dechlorination and dissolution unit electrolyzer and the dissolution unit

chlorine offgas scrubbing column. The applicant has opted to meet the performance

requirements through a strategy of mitigation. The PSSCs identified to implement this safety

strategy are process cell entry controls, the C4 confinement system, and facility worker action. 

Based on the nature of this event (mitigated by ventilation system and administrative

procedures), the staff considers this to be an acceptable strategy for meeting the 10 CFR 70.61

performance requirements.

Radiochemical Hazards

Events involving radiochemical hazards are postulated to occur inside the MFFF and involve the

AP system. The applicant performed bounding consequence analysis to determine the

unmitigated consequences to persons beyond the controlled area boundary. The applicant

determined that under certain circumstances (e.g., a process failure), gases which would
5-43

normally react with hydrazine, HAN, and hydrazoic acid could be released and result in an

unacceptable dose to persons beyond the controlled area boundary. The applicant has opted to

satisfy the performance requirements using a strategy of mitigation. The PSSC chosen to

protect the site worker is the process safety control subsystem which will ensure that the flow of

nitrogen dioxide/nitrogen tetroxide is limited to the oxidation column. With regard to the facility

worker, the applicant stated that the same PSSCs that protect the worker from radioactive

releases will also protect the facility worker from any radiochemical releases. Based on the

applicant’s deterministic criteria and the NUREG-1718 Table A-5 descriptions (AEC), the staff

considers this to be an acceptable strategy for meeting the 10 CFR 70.61 performance

requirements.

A fire inside the secured warehouse could also result in unacceptable consequences to the

facility worker and persons beyond the controlled area boundary through the release of depleted

uranium dioxide. The applicant has opted to meet the performance objectives using a strategy

of mitigation. The PSSCs identified to protect the facility worker and persons beyond the

controlled area boundary are facility worker action and combustible loading controls. The safety

function of the combustible loading controls is to limit the quantity of combustibles in the secured

warehouse to ensure that any fire that may occur will not encompass a large fraction of the

stored depleted uranium dioxide. The function of the facility worker action is to ensure that

facility workers take proper actions to limit chemical consequences as a result of a fire. Based

on the applicant’s deterministic criteria and the NUREG-1718 Table A-5 descriptions

(administrative control and an enhanced administrative control), the staff considers this to be an

acceptable strategy for meeting the 10 CFR 70.61 performance requirements.

5.1.6.4 Consequence Assessment

The applicant has performed an analysis of the bounding mitigated consequences of each event

type. These analyses are derived from the hazard assessment performed to establish the

PSSCs and represent the bounding accident from each event type. The event types considered

are the same as those discussed earlier and consist of loss of confinement, internal fire, loadhandling, explosion, and criticality. The calculated mitigated consequences from each event

type were found to be below 10 CFR 70.61(c) threshold levels. 

The bounding loss of confinement event is an event caused by a load-handling accident

involving a glovebox in the jar storage and handling unit. 

[Text removed under 10 CFR 2.390]

The bounding internal fire event is a fire in the fire area containing the final dosing unit. This unit

contains polished plutonium powder for the purpose of down-blending the MOX powder to the

desired blend for fuel rod fabrication.

[Text removed under 10 CFR 2.390]
5-44

The bounding explosion event is an event that involves the entire material at risk within a

process cell. The cause of the explosion was not postulated.

[Text removed under 10 CFR 2.390]

The staff’s independent evaluation of the applicant’s bounding consequence determinations are

provided in FSER Sections 9.1.1.4 and 10.1.3. 

5.1.7 Description of PSSCs

The acceptability of the various strategies for preventing and/or mitigating identified hazards is

also dependent on the design bases of the PSSCs. The specific designs or “design bases” of

the PSSCs are determined through the DCS safety assessment and are discussed in the

appropriate sections of this FSER. The FSER sections in which the design bases of each of the

PSSCs are discussed is identified in Table 5.1-3.

5.2 Evaluation Findings

In Section 5 of the revised CAR, DCS provided a description of the safety analysis that it

performed and the identified PSSCs for the proposed facility. Based on the staff's review of the

revised CAR and supporting information provided by the applicant relevant to the safety analysis

and the identified PSSCs, the staff concludes, pursuant to 10 CFR 70.23(b), that the design

bases of the PSSCs identified by the applicant will provide reasonable assurance of protection

against natural phenomena and the consequences of potential accidents.

A summary of the staff’s evaluation findings in regards to the safety assessment portion of the

review is as follows:

• The plant site description relating to safety assessment was found to be adequate.

• The safety assessment team description was found to be adequate.
5-45

The purpose of the safety assessment methodology review, as discussed in FSER Section

5.1.5, was to determine if the safety assessment was complete by assuring that all appropriate

natural phenomena, external manmade hazards, and internal process hazards were considered. 

The methodology for the safety assessment of the design basis was found to be adequate.

The purpose of the safety assessment results review, as discussed in FSER Section 5.1.6, was

to evaluate the appropriateness of natural phenomena and external manmade hazards selected

for design, the adequacy of the PSSCs selected to protect against these events, and the

adequacy of the strategy and identified PSSCs at a conceptual level for the internal process

hazards. The results of the safety assessment of the design basis were found to be adequate

and all open items have been closed.

5.3 References

(Chen, 1991) Chen, J.T, et al. “Procedural and Submittal Guidance for the Individual Plant

Examination of External Events (IPEEE) for Severe Accident Vulnerabilities.” Final Report 1-26. 

June 1991.

(CFR, 2004a) Code of Federal Regulations. Title 10, Energy, Part 20, “Standards for Protection

Against Radiation.”

(CFR, 2004b) Code of Federal Regulations. Title 10,Energy, Part 70, “Domestic Licensing of

Special Nuclear Material.”

(DCS, 2002a) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear

Regulatory Commission, RE: Clarification of Response to “Request for Additional Information.” 

March 8, 2002.

(DCS, 2002b) Ihde, R., Duke Cogema Stone & Webster. Letter to Document Control Desk,

U.S. Nuclear Regulatory Commission. Docket Number 070-03098 Duke Cogema Stone &

Webster, Mixed Oxide (MOX) Fuel Fabrication Facility Construction Authorization Request,

DCS-NRC-000114. October 31, 2002.

(DCS, 2002c) Ihde, R., Duke Cogema Stone & Webster. Letter to Document Control Desk,

U.S. Nuclear Regulatory Commission, Docket NUMBER 070-03098, Duke Cogema Stone &

Webster, Mixed Oxide (MOX) Fuel Fabrication Facility Construction Authorization Request. 

DCS-NRC-000120. November 22, 2002.

(DOE, 1994) U.S. Department of Energy. DOE-STD-1020-94, Change Notice No. 1, “Natural

Phenomena Hazards Design and Evaluation Criteria for Department of Energy Facilities.” 

Washington, DC, April 1994.

(NRC, 1997) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.165, “Identification and

Characterization of Seismic Sources and Determination of Safe Shutdown Earthquake Ground

Motion.” Washington, DC, 1997.
5-46

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility.” Washington,

DC, 2000.

(NRC, 2002) Brown, D. U.S. Nuclear Regulatory Commission. Memorandum to Eric Leeds,

U.S. Nuclear Regulatory Commission, “February 21–22, 2002, In-Office Review Summary: 

Review of Duke Cogema Stone and Webster Construction Authorization Request Supporting

Documents for the Mixed Oxide Fuel Fabrication Facility.” March 11, 2002.
Table 5.1-1 Safety Assessment Summary (Criticality, Natural Phenomena Hazards, External Manmade Hazards, and

Chemical Hazards)

5-47

Event Event Type PSSC Safety Function

Criticality Criticality Criticality Controls—Features required to ensure

that design bases are fulfilled.

Maintain subcritical conditions in the

process.

Natural Phenomena

Hazards

Extreme W ind MOX Fuel Fabrication Building W ithstand design basis wind loads and

wind-driven missiles.

Provide protection for internal SSCs.

Emergency Generator Building W ithstand design basis wind loads and

wind-driven missiles.

Provide protection for internal SSCs.

Missile Barriers W ithstand design basis wind loads and

wind-driven missiles.

Provide protection for internal SSCs.

W aste Transfer Line W ithstand design basis wind loads and

wind-driven missiles.

Earthquake MOX Fuel Fabrication Building W ithstand the effects of design basis

earthquake.

Emergency Generator Building W ithstand the effects of design basis

earthquake.

W aste Transfer Line W ithstand the effects of design basis

earthquake.

Fluid Transport Systems W ithstand the effects of design basis

earthquake, as necessary.

Seism ic Monitoring and Associated Seismic

Isolation Valves

Prevent fire and criticality as a result of an

uncontrolled release of chemicals and

water within the MFFF building.
Table 5.1-1 Safety Assessment Summary (Criticality, Natural Phenomena Hazards, External Manmade Hazards, and

Chemical Hazards)

Event Event Type PSSC Safety Function

5-48

Natural Phenomena

Hazards

(cont.)

Tornado MOX Fuel Fabrication Building W ithstand design basis tornado wind loads

and tornado-generated missiles.

Provide protection to internal PSSCs.

Emergency Generator Building W ithstand design basis tornado wind loads

and tornado-generated missiles.

Provide protection for internal PSSCs.

Missile Barriers W ithstand design basis tornado wind loads

and tornado-generated missiles.

Provide protection for internal PSSCs.

W aste Transfer Line W ithstand design basis tornado wind loads

and tornado-driven missiles.

Tornado Dampers Protect MFFF ventilation systems from

differential pressure effects of the tornado.

External Fire MOX Fuel Fabrication Building W ithstand the effects of the design basis

external fire.

Provide protection for internal SSCs from

the effects of heat, fire, and smoke.

Emergency Generator Building W ithstand the effects of the design basis

external fire.

Provide protection for internal SSCs from

the effects of heat, fire, and smoke.

Emergency Control Room Air-Conditioning

System

Ensure habitable conditions for operators.

W aste Transfer Line W ithstand the effects of external fires.
Table 5.1-1 Safety Assessment Summary (Criticality, Natural Phenomena Hazards, External Manmade Hazards, and

Chemical Hazards)

Event Event Type PSSC Safety Function

5-49

Natural Phenomena

Hazards

(cont.)

Rain, Snow, and

Ice

MOX Fuel Fabrication Building W ithstand the effects of the design basis

rain, snow, and ice loads.

Provide protection for internal SSCs.

Emergency Generator Building W ithstand the effects of the design basis

rain, snow, and ice loads.

Provide protection for internal SSCs.

W aste Transfer Line W ithstand the effects of design basis rain,

snow, and ice loads.

Lightning None N/A

Tem perature

Extremes

None N/A

External Manmade Release of

Radioactive

Material or

Hazardous

Chemicals

Emergency Control Room Air-Conditioning

System

Ensure that the emergency control rooms

are habitable.

Direct Damage to

PSSCs

MOX Fuel Fabrication Building

(To be described in the ISA)

W ithstand overpressure of explosions

external to the MFFF area.

Emergency Generator Building

(To be described in the ISA)

W ithstand overpressure of explosions

external to the MFFF area.

Loss of Offsite

Power

Section 5.5.2.9, Section 5.4, Support

Systems—Table 5.5-22 (Revised CAR)

Supply emergency power upon loss of

offsite power.

Fire MOX Fuel Fabrication Building W ithstand the effects of the design basis

external fire.

Provide protection for internal SSCs from

the effects of heat, fire, and smoke.
Table 5.1-1 Safety Assessment Summary (Criticality, Natural Phenomena Hazards, External Manmade Hazards, and

Chemical Hazards)

Event Event Type PSSC Safety Function

5-50

External Man-Made

(cont.)

Fire (cont.) Emergency Generator Building W ithstand the effects of the design basis

external fire.

Provide protection for internal SSCs from

the effects of heat, fire, and smoke.

Emergency Control Room Air-Conditioning

System

Ensure habitable conditions for operators.

W aste Transfer Line W ithstand the effects of external fires.

External Exposure Operator is

inadvertently

exposed to

excessive direct

radiation.

Not Required N/A

Chemical A release of

hazardous

chemicals not

produced from

licensed

materials.

Emergency Control Room Air-Conditioning

System

Ensure habitable conditions for operators

in the control room.

A release of

hazardous

chem icals

produced from

licensed

materials.

Process Cell Entry Controls Prevent the entry of personnel into process

cells during normal operations.

Ensure that workers do not receive a

chemical consequence in excess of limits

while performing maintenance in the AP

process cells.

C4 Confinement Systems Contain a chemical release within a

glovebox and provide an exhaust path for

removal of chemical vapors.
Table 5.1-1 Safety Assessment Summary (Criticality, Natural Phenomena Hazards, External Manmade Hazards, and

Chemical Hazards)

Event Event Type PSSC Safety Function

5-51

Chemical

(cont.)

A release of

hazardous

chem icals

produced from

licensed

materials

(cont.).

Facility Worker Action Ensure that facility workers take proper

actions to limit chemical consequences for

leaks occurring in C3 ventilated areas.

A release of

hazardous

chemicals and

radioactive

materials.

Process Safety Control Subsystem

(NO2/N2O4)

Ensure that the flow rate of nitrogen

dioxide/dinitrogen tetroxide is limited to the

oxidation column of the purification cycle.

Com bustible Loading Controls

(UO2)

Limit the quantity of combustibles in the

secured warehouse to ensure that any fire

that may occur will not encompass a large

fraction of the stored depleted uranium.

Facility Worker Actions

(UO2)

Ensure that facility workers take proper

actions to limit chem ical consequences in

the secured warehouse.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

5-52

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

Loss of

Confinement/

Dispersal of

Nuclear Material

Over

Temperature

Excessive

temperature of

the AP

electrolyzer in a

glovebox

C3 Confinement Provide filtration

to mitigate

dispersions from

C3 areas.

Process Safety

Control

Subsystem

Shutdown

process

equipment before

exceeding a

temperature

safety limit.

Process Safety

Control

Subsystem

Shutdown process

equipment before

exceeding a

temperature safety

limit.

Corrosion Corrosion

involving

pneumatic

transfer of

corrosive

chemicals

None N/A Material

Maintenance

and

Surveillance

Program

Detect and limit

damage resulting

from corrosion.

Material

Maintenance

and

Surveillance

Program

Detect and limit

damage resulting

from corrosion.

Small Breaches

in a Glovebox

Confinement

Boundary or

Backflow from a

Glovebox

through Utility

Lines

Backflow

through the

interfacing gas

line to the

interfacing

system followed

by the opening

of this

interfacing

system during a

maintenance

operation.

None N/A C4

Confinement

System

Maintain a

negative

glovebox

differential

pressure between

glovebox and

interfacing

systems.

Maintain a

minimum inward

flow through

small glovebox

breaches.

C4

Confinement

System

Maintain a negative

glovebox

differential

pressure between

glovebox and

interfacing

systems.

Maintain a

minimum inward

flow through small

glovebox breaches.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-53

Loss of

Confinement/

Dispersal of

Nuclear Material

(cont.)

Leaks of AP

Process Vessels

or Pipes Within

Process Cells

Leak of

tanks/vessels

inside the

process cell

containing a

portion of the

purification

cycle.

Process Cell

Exhaust System

Ensure that a

negative pressure

exists between

the process cells

and the C2 area.

Ensure that the

process cell

exhaust is

effectively filtered.

Process Cells Contain leaks

within the process

cells.

Process Cell

Exhaust

System

Ensure that a

negative pressure

exists between the

process cells and

the C2 area.

Ensure that the

process cell

exhaust is

effectively filtered.

Process Cell

Entry Controls

Prevent the entry

of personnel into

process cell

during normal

operation.

Backflow from a

Process Vessel

Through Utility

Lines

Backflow of

radioactive

material from a

waste tank

containing

americium

through an

interfacing

supply line that

is subsequently

breached or

opened during a

maintenance

operation.

Backflow

Prevention

Features

Prevent process

fluids from

backflowing into

interfacing

systems.

Backflow

Prevention

Features

Prevent process

fluids from

backflowing into

interfacing

systems.

Backflow

Prevention

Features

Prevent process

fluids from

backflowing into

interfacing

systems.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-54

Loss of

Confinement/

Dispersal of

Nuclear Material

(cont.)

Rod-Handling

Operations

Fracture of one

or more fuel

rods while

utilizing fuel

handling

equipment

resulting in a

breach of

confinement.

None N/A Facility Worker

Action

Ensure that

facility workers

take proper

actions to limit

radiological

exposure.

None N/A

Material

Handling

Controls

Ensure proper

handling of

primary

confinements

outside of

gloveboxes.

Material

Handling

Equipment

Limit damage to

fuel

rods/assemblies

during handling

operations.

Breaches in

Containers

Outside the

Gloveboxes

from Handling

Operations in C2

and C3 Areas

Container

containing filters

is breached

while in C2 area.

MaterialsHandling and

Control

(C2 Areas)

Ensure proper

handling of

primary

confinement

types outside of

gloveboxes.

outside of

gloveboxes.

Material

Handling

Controls

Ensure proper

handling of

primary

confinement

types outside the

gloveboxes

outside of

gloveboxes.

MaterialsHandling and

Control

(C2 Areas)

Ensure proper

handling of primary

confinement types

outside of

gloveboxes outside

of gloveboxes.

3013 Canister

(C2 Areas)

Withstand the

effects of design

basis drops

without

breaching.

3013 Canister Withstand the

effects of design

basis drops

without

breaching.

3013 Canister

(C2 Areas)

Withstand the

effects of design

basis drops without

breaching.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-55

Loss of

Confinement/

Dispersal of

Nuclear Material

(cont.)

Breaches in

Containers

Outside the

Gloveboxes

from Handling

Operations in C2

and C3 Areas

(cont.)

Container

containing filters

is breached

while in C2 area

(cont.)

Transfer 

Container

(C2 Areas)

Withstand the

effect of design

basis drops

without

breaching.

Transfer

Container

Withstand the

effects of design

basis drops

without

breaching.

Transfer

Container

(C2 Areas)

Withstand the

effect of design

basis drops without

breaching.

C3 Confinement

System

(C3 Areas)

Provide filtration

to mitigate

dispersion from

the C3 areas.

Facility Worker

Controls (C3

Areas)

Ensure that

facility workers

take proper

actions before

bag-out

operations to limit

radiological

exposure.

C3

Confinement

System

(C3 Areas)

Provide filtration to

mitigate dispersion

from the C3 areas.

Over- or

Underpressuriza

tion of Glovebox

Rapid

overpressurizati

on of the

calcining

furnace

C3/C4

Confinement

System

Provide filtration

to mitigate

dispersion from

C3/C4 areas.

Facility Worker

Action

(slow

pressurization)

Ensure that

facility workers

take proper

actions to limit

radiological

exposure.

C3/C4

Confinement

System

Provide filtration to

mitigate dispersion

from C3/C4 areas.

Process Safety

Control

Subsystem

(slow

pressurization)

Warn operators of

glovebox

pressure

discrepancies

before exceeding

differential

pressure limits.

Glovebox

Pressure

Controls

(rapid or slow

pressurization)

Maintain

glovebox

pressure within

design limits.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-56

Loss of

Confinement/

Dispersal of

Nuclear Material

(cont.)

Excessive

Temperature

from Decay

Heat from

Radioactive

Materials

Excessive

temperature

(from decay

heat) of C2

storage area

(PuO2 powder

3013 storage

unit)

High

Depressurization

Exhaust System

(C3

Confinement

System)

Provide exhaust

to ensure that

temperatures in

the 3013 canister

storage structure

are maintained

within design

limits. 

High

Depressurizati

on Exhaust

System (C3

Confinement

System)

Provide exhaust

to ensure that

temperatures in

the 3013 canister

storage structure

are maintained

within design

limits. 

High

Depressurizati

on Exhaust

System (C3

Confinement

System)

Provide exhaust to

ensure that

temperatures in the

3013 canister

storage structure

are maintained

within design limits. 

Glovebox

Dynamic

Exhaust Failure

Loss of negative

pressure or a

flow perturbation

involving the C4

dynamic

confinement

system resulting

in a ventilation

airflow reversal

into a C3 area

C4 Confinement

System

Operate to ensure

that a negative

pressure

differential exists

between the C4

glovebox and the

C3 areas.

Ensure that the

C4 exhaust is

effectively filtered.

C4

Confinement

System

Operate to ensure

that a negative

pressure

differential exists

between the C4

glovebox.

Ensure that the

C4 exhaust is

effectively filtered.

C4

Confinement

System

Operate to ensure

that a negative

pressure

differential exists

between the C4

glovebox.

Ensure that the C4

exhaust is

effectively filtered.

Process Fluid

Line Leak in a

C3 Area Outside

the Glovebox

A leak from a

line carrying a

process fluid in

a C3 area

outside of a

glovebox or

process cell

caused by

corrosive

chemicals or

mechanical

failure of AP

piping 

None N/A Double- Walled

Pipe

Prevent leaks

from pipes

containing

process fluids

from leaking into

C3 areas.

Double- Walled

Pipe

Prevent leaks from

pipe containing

process fluids from

leaking into C3

areas.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-57

Loss of

Confinement/Dis

persal of

Nuclear Material

(cont.)

Sintering

Furnace

Confinement

Boundary

Failure

Rapid

overpressurizati

on of the

sintering furnace

and a slow

leakage through

seals

C3 Confinement

System

Provide filtration

to mitigate

dispersion from

the C3 areas

Sintering

Furnace

Provide primary

confinement

boundary against

leaks into C3

areas.

Minimize

consequences of

leak from seal

failure.

Sintering

Furnace

Provide primary

confinement

boundary against

leaks into C3 areas 

Sintering

Furnace

Pressure

Controls

Maintain sintering

furnace pressure

within

Sintering

Furnace

Pressure

Controls

Maintain sintering

furnace pressure

within design limits.

design limits.

Fire AP Process

Cells

Fire in the AP

process cells

containing the

dissolution tanks

Process Cell

Fire Prevention

Features

Ensure that fires

in the process

cells are highly

unlikely.

Process Cell

Fire Prevention

Features

Ensure that fires

in the process

cells are highly

unlikely.

Process Cell

Fire Prevention

Features

Ensure that fires in

the process cells

are unlikely.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-58

Fire

(cont.)

AP/MP C3

Glovebox Areas

Fire inside or

outside the

glovebox

(fire areas

containing

process

gloveboxes)

C3/C4

Confinement

Systems

Remain operable

during design

basis fire and

effectively filter

any release.

Facility Worker

Action

Ensure that

facility workers

take proper

actions to limit

radiological

exposure.

C3/C4

Confinement

Systems

Remain operable

during design basis

fire and effectively

filter any release.

Fire within the

PuO2 buffer

storage area

(Gloveboxes

that store

radiological

materials)

Combustible

Loading

(For storage

gloveboxes

only)

Limit the quantity

of combustibles in

fire areas

containing a

storage glovebox

such that any fire

that may occur

will not

encompass a

large fraction of

the stored

radiological

material.

Facility Worker

Controls

Ensure that

facility workers

take proper

actions before

maintenance

activities to limit

radiological

exposure.

Combustible

Loading

(For storage

gloveboxes

only)

Limit the quantity of

combustibles in fire

areas containing a

storage glovebox

such that any fire

that may occur will

not encompass a

large fraction of the

stored radiological

material.

C1 and/or C2-

3013 Canister

Fire affecting

3013 canister

Combustible

Loading

Controls

Limit the quantity

of combustibles in

a fire area

containing 3013

canisters to

ensure that the

canisters are not

adversely

impacted by a

fire.

Combustible

Loading

Controls

Limit the quantity

of combustibles in

a fire area

containing 3013

canisters to

ensure that the

canisters are not

adversely

impacted by a

fire.

Combustible

Loading

Controls

Limit the quantity of

combustibles in a

fire area containing

3013 canisters to

ensure that the

canisters are not

adversely impacted

by a fire.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-59

Fire

(cont.)

C1 and/or C2-

3013 Transport

Cask

Fire in the truck

bay involving

transport

packages

3013 Transport

Cask

Withstand the

design basis fire

without

breaching.

3013 Transport

Cask

Withstand the

design basis fire

without

breaching.

3013 Transport

Cask

Withstand the

design basis fire

without breaching.

Combustible

Loading

Controls

Limit the quantity

of combustibles in

a fire area

containing 3013

transport casks to

ensure that the

cask design basis

fire is not

exceeded.

Combustible

Loading

Controls

Limit the quantity

of combustibles in

a fire area

containing 3013

transport casks to

ensure that the

cask design basis

fire is not

exceeded.

Combustible

Loading

Controls

Limit the quantity of

combustibles in a

fire area containing

3013 transport

casks to ensure

that the cask

design basis fire is

not exceeded

C1 and/or C2

Areas-Fuel Rod

Fire in the fuel

assembly

storage area

Combustible

Loading

Controls

Limit the quantity

of combustibles in

a fire area

containing fuel

rods to ensure

that the fuel rods

are not adversely

impacted by a

fire.

Combustible

Loading

Controls

Limit the quantity

of combustibles in

a fire area

containing fuel

rods to ensure

that the fuel rods

are not adversely

impacted by a

fire.

Combustible

Loading

Controls

Limit the quantity of

combustibles in a

fire area containing

fuel rods to ensure

that the fuel rods

are not adversely

impacted by a fire.

C1 and/or C2

Areas-MOX Fuel

Transport Cask

Fire affecting 

the MOX fuel

transport cask

MOX Fuel

Transport Cask

Withstand the

design basis fire

without

breaching.

MOX Fuel

Transport Cask

Withstand the

design basis fire

without

breaching.

MOX Fuel

Transport Cask

Withstand the

design basis fire

without breaching.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-60

Fire

(cont.)

C1 and/or C2

Areas-MOX Fuel

Transport Cask

(cont.)

Fire Affecting 

the MOX fuel

transport cask

(cont.)

Combustible

Loading

Controls

Limit the quantity

of combustibles in

a fire area

containing MOX

fuel transport

casks to ensure

that the cask

design basis fire

is not exceeded.

Combustible

Loading

Controls

Limit the quantity

of combustibles in

a fire area

containing MOX

fuel transport

casks to ensure

that the cask

design basis fire

is not exceeded.

Combustible

Loading

Controls

Limit the quantity of

combustibles in a

fire area containing

MOX fuel transport

casks to ensure

that the cask

design basis fire is

not exceeded.

C1 and/or C2

Areas—Waste

Container

Fire in the

assembly

packaging area

None N/A Facility Worker

Action

Ensure that

facility workers

take proper

actions to limit

radiological

exposure.

None N/A

C1 and/or C2

Areas—Transfer

Container

Fire in e the

airlocks,

corridors,

stairways, safe

areas, or liquid

waste reception

areas

Combustible

Loading

Controls

(For site worker

only—no PSSCs

required for IOC)

Limit the quantity

of combustibles in

a fire area

containing

transfer

containers to

ensure that the

containers are not

adversely

impacted by a

fire.

Combustible

Loading

Controls

Limit the quantity

of combustibles in

a fire area

containing

transfer

containers to

ensure that the

containers are not

adversely

impacted by a

fire.

Combustible

Loading

Controls

Limit the quantity of

combustibles in a

fire area containing

transfer containers

to ensure that the

containers are not

adversely impacted

by a fire.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-61

Fire

(cont.)

C1 and/or C2

Areas-Final C4

HEPA Filter

Fire impacting

final C4 HEPA

filters breaches

the HEPA filter

house allowing

the material to

pass directly to

the stack.

Combustible

Loading

Controls

Limit the quantity

of combustibles in

the filter area to

ensure that the

final C4 HEPA

filters are not

impacted by a fire

in the filter room.

Combustible

Loading

Controls

Limit the quantity

of combustibles in

the filter area to

ensure that the

final C4 HEPA

filters are not

adversely

impacted by a fire

in the filter room.

Combustible

Loading

Controls

Limit the quantity of

combustibles in the

filter area to ensure

that the final C4

HEPA filters are

not impacted by a

fire in the filter

room.

Outside the

MOX

Fabrication

Building

Fire involving

diesel fuel

storage,

gasoline

storage, or the

reagents

processing

building such

that the MFFF

building

structure is

damaged and

radioactive

material is

released.

MOX Fuel

Fabrication

Building

Structure

Maintain

structural integrity

and prevent

damage to

internal PSSCs

from external

fires.

MOX Fuel

Fabrication

Building

Structure

Maintain

structural integrity

and prevent

damage to

internal PSSCs

from external

fires.

MOX Fuel

Fabrication

Building

Structure

Maintain structural

integrity and

prevent damage to

internal PSSCs

from external fires.

Emergency

Generator

Building

Structure

Maintain

structural integrity

and prevent

damage to

internal PSSCs

from fires external

to the structure.

Emergency

Generator

Building

Structure

Maintain

structural integrity

and prevent

damage to

internal PSSCs

from fires external

to the structure.

Emergency

Generator

Building

Structure

Maintain structural

integrity and

prevent damage to

internal PSSCs

from fires external

to the structure.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-62

Fire

(cont.)

Outside the

MOX

Fabrication

Building

(cont.)

Fire involving

diesel fuel

storage,

gasoline

storage, or the

reagents

processing

building such

that the MFFF

building

structure is

damaged and

radioactive

material is

released

(cont.)

Emergency

Control Room

Air-Conditioning

System

Ensure habitable

conditions for

operators.

Emergency

Control Room

AirConditioning

System

Ensure habitable

conditions for

operators.

Emergency

Control Room

AirConditioning

System

Ensure habitable

conditions for

operators.

Waste Transfer

Line

Prevent damage

to line from

external fires.

Waste Transfer

Line

Prevent damage

to line from

external fires.

Waste Transfer

Line

Prevent damage to

line from external

fires.

Facility-Wide

Systems

Fire Involving

the pneumatic

pipe automatic

transfer system

None N/A Facility Worker

Action

Ensure that

facility workers

take proper

actions to limit

radiological

exposure.

Combustible

Loading

Controls

Limit the quantity of

combustibles in

areas containing

the pneumatic

transfer system to

ensure this system

is not adversely

impacted.

Combustible

Loading

Controls

Limit the quality

of combustibles in

a fire area

containing a

pneumatic

system to ensure

that this system is

not adversely

impacted by a

fire.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-63

Fire

(cont.)

Facility Fire in all

process cell

units and

support units

with radioactive

material present

Fire Barriers Contain fires

within a single fire

area.

Fire Barriers Contain fires

within a single fire

area.

Fire Barriers Contain fires within

a single fire area.

Facility Worker

Action

Ensure that

facility workers

take proper

actions to limit

radiological

exposure.

AP Electrolyzer Energetic

breach of the AP

electrolyzer and

the dispersal of

radioactive

materials

Maintenance

Activity Controls

Isolation of power

to the electrolyzer

when the

electrolyzer is

drained.

Maintenance

Activity

Controls

Isolation of power

to the electrolyzer

when the

electrolyzer is

drained.

Maintenance

Activity

Controls

Isolation of power

to the electrolyzer

when the

electrolyzer is

drained.

Electrolyzer

structure,

sintered silicon

nitride barrier,

guide sleeves,

and

polytetrafluorethylene insulator

(PTFE)

To provide

physical

separation and

electrical

insulation

between

electrolyzer

components and

structural integrity

and stability to

the electrolyzer

system.

Electrolyzer

structure,

sintered silicon

nitride barrier,

guide sleeves,

and PTFE

To provide

physical

separation and

electrical

insulation

between

electrolyzer

components and

structural integrity

and stability to

the electrolyzer

system.

Electrolyzer

structure,

sintered silicon

nitride barrier,

guide sleeves,

and (PTFE)

To provide physical

separation and

electrical insulation

between

electrolyzer

components and

structural integrity

and stability to the

electrolyzer

system.

Process Safety

Control

Subsystem

Monitor the

electrolyzer for

electrical faults

that could result

in arcing or other

imparting of

electrical energy

with the risk of

titanium fire.

Process Safety

Control

Subsystem

Monitor the

electrolyzer for

electrical faults

that could result

in arcing or other

imparting of

electrical energy

with the risk of

titanium fire.

Process Safety

Control

Subsystem

Monitor the

electrolyzer for

electrical faults that

could result in

arcing or other

imparting of

electrical energy

with the risk of

titanium fire.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-64

Load Handling AP Process

Cells

AP process cell

containing the

dissolution tanks

(Event results in

a breach of the

AP dissolution

tanks and

subsequent

release of

unpolished PuO2

solution.)

None N/A Process Cells Contain fluid

leaks within the

process cells.

Process Cell

Ventilation

System

Passive

Boundary

Provide filtration to

limit the dispersion

of radioactive

material.

Process Cell

Entry Controls

Prevent the entry

of personnel into

process cells

during normal

operations.

Ensure that

facility workers do

not receive a

radiological dose

in excess of limits

while performing

maintenance.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-65

Load Handling

(cont.)

AP/MP C3

Glovebox Areas

Gloveboxes that

contain jar

storage and

handling of the

MOX powder

workshop 

(Event results in

a breach of a

glovebox and

subsequent

release of

radiological

material.) 

C3 Confinement

System

Provide filtration

to mitigate

dispersion from

the C3 Areas.

Material

Handling

Controls

Prevent impacts

to the glovebox

during normal

operations from

loads handled

either outside or

inside the

glovebox that

could exceed the

glovebox design

basis.

Material

Handling

Controls

Prevent impacts to

the glovebox during

normal operations

from loads handled

either outside or

inside the glovebox

that could exceed

the glovebox

design basis.

Glovebox Maintain

confinement

integrity for

design basis

impacts.

Glovebox Maintain

confinement

integrity for design

basis impacts.

Material

Handling

Equipment

Prevent impacts

to the glovebox

through the use

of engineered

equipment.

Material

Handling

Equipment

Prevent impacts to

the glovebox

through the use of

engineered

equipment.

Facility Worker

Controls

Ensure that

facility workers

take proper

actions before

maintenance

activities to limit

radiological

exposure.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-66

Load Handling

(cont.)

C1 and/or C2

Areas—3013

Canister

Drop of one

3013 canister

onto another

3013 canister

each containing

unpolished PuO2

in powder form

3013 Canister Withstand the

effects of the

design basis drop

without

breaching.

3013 Canister Withstand the

effects of the

design basis drop

without

breaching.

3013 Canister Withstand the

effects of the

design basis drop

without breaching.

Material

Handling

Controls

Ensure that the

design basis lift

height of the 3013

canister is not

exceeded.

Material

Handling

Controls

Ensure that the

design basis lift

height of the 3013

canister is not

exceeded.

Material

Handling

Controls

Ensure that design

basis lift height of

the 3013 canister is

not exceeded.

C1 and/or C2

Areas-3013

Transport Cask

Drop of a 3013

transport cask

containing

unpolished PuO2

in powder form

onto another

transport cask

3013 Transport

Cask

Withstand the

effects of design

basis drops

without release of

radioactive

materials.

3013 Transport

Cask

Withstand the

effects of design

basis drops

without release of

radioactive

materials.

3013 Transport

Cask

Withstand the

effects of design

basis drops without

release of

radioactive

materials.

Material

Handling

Controls

Ensure that the

design basis lift

height of the 3013

transport cask is

not exceeded.

MaterialsHandling

Controls

Ensure that the

design basis lift

height of the 3013

transport cask is

not exceeded.

Material

Handling

Controls

Ensure that the

design basis lift

height of the 3013

transport cask is

not exceeded.

C1 and/or C2

Areas-Fuel Rod

Drop of a fuel

assembly onto

another fuel

assembly each

containing MOX

(6%)

None N/A Facility Worker

Action

Ensure that

facility workers

take proper

actions to limit

radiological

exposure.

None N/A
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-67

Load Handling

(cont.)

C1 and/or C2

Areas- MOX

Fuel Transport

Cask

Drop of one

MOX fuel

transport cask

containing up to

three MOX fuel

assemblies

None N/A MOX Fuel

Transport Cask

Withstand the

effects of design

basis drops

without release of

radioactive

material.

MOX Fuel

Transport Cask

Withstand the

effects of design

basis drops without

release of

radioactive

material.

Material

Handling

Design

Controls

Ensure that the

design basis lift

height of MOX

fuel transport

cask is not

exceeded.

Material

Handling

Design

Controls

Ensure that design

basis lift height of

MOX fuel transport

cask is not

exceeded.

C1, C2, and/or

C3

Areas—Waste

Container

A damaged

waste drum in

the assembly

packaging (truck

bay) Area

None N/A Facility Worker

Action

Ensure that

facility workers

take proper

actions to limit

radiological

exposure.

None N/A

C1 and/or C2

Areas—Transfer

Container

Drop of a

transfer

container

containing a

HEPA filter with

PuO2 in powder

form

Transfer

Container

Withstand the

effects of design

basis drops

without

breaching.

Transfer

Container

Withstand the

effects of design

basis drops

without

breaching.

Transfer

Container

Withstand the

effects of design

basis drops without

breaching.

Material

Handling

Controls

Ensure that the

design basis lift

height of the

transfer container

is not exceeded.

Material

Handling

Controls

Ensure that the

design basis lift

height of the

transfer container

is not exceeded.

Material

Handling

Controls

Ensure that the

design basis lift

height of the

transfer container

is not exceeded.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-68

Load Handling

(cont.)

C1 and/or C2

Areas-Final C4

HEPA Filter

Impact to final

C4 HEPA filters

breaching the

HEPA filter

housing and

allowing the

material from

the HEPA filters

to pass directly

to the stack

Material

Handling

Controls

Ensure that loadhandling activities

that could

potentially lead to

a breach in the

final C4 HEPA

filters do not

occur.

Material

Handling

Controls

Ensure that loadhandling activities

that could

potentially lead to

a breach in the

final C4 HEPA

filters do not

occur.

Material

Handling

Controls

Ensure that loadhandling activities

that could

potentially lead to a

breach in the final

C4 HEPA filters do

not occur.

C4 Confinement Spill of

unpolished

plutonium

powder inside a

glovebox but

does not result

in a breach of

the glovebox

C4 Confinement

System

Maintain a

negative

glovebox

pressure

differential

between the

glovebox and the

interfacing

systems.

C4

Confinement

System

Maintain a

negative glove

box pressure

differential

between the

glovebox and the

interfacing

systems.

C4

Confinement

System

Maintain a negative

glove box pressure

differential between

the glovebox and

the interfacing

systems.

Outside the

MOX Fuel

Fabrication

Building

Load-handling

event involving

the waste

transfer line

Waste Transfer

Line

Ensure that the

waste transfer

line is protected

from activities

taking place

outside the MOX

fuel fabrication

building.

Waste Transfer

Line

Ensure that the

waste transfer

line is protected

from activities

taking place

outside the MOX

fuel fabrication

building.

Waste Transfer

Line

Ensure that the

waste transfer line

is protected from

activities taking

place outside the

MOX fuel

fabrication building.

Facility wide Breach of the

MFFF structure

from a heavy

load resulting in

a breach of

confinement or

in a breach of

container

holding nuclear

materials 

MOX Fuel

Fabrication

Building

Structure

Withstand the

effects of load

drops that could

potentially impact

radiological

material.

MOX Fuel

Fabrication

Building

Structure

Withstand the

effects of load

drops that could

potentially impact

radiological

material.

MOX Fuel

Fabrication

Building

Structure

Withstand the

effects of load

drops that could

potentially impact

radiological

material.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-69

Load Handling

(cont.)

Facility wide

(cont.)

Breach of the

MFFF structure

from a heavy

load resulting in

a breach of

confinement or

in a breach of

container

holding nuclear

materials

(cont.)

Material

Handling

Controls

Prevent loadhandling events

that could breach

primary

confinements.

Material

Handling

Controls

Prevent loadhandling events

that could breach

primary

confinements.

Material

Handling

Controls

Prevent loadhandling events

that could breach

primary

confinements.

Explosion Hydrogen

Explosion

Explosion of

hydrogen and

oxygen in a

sintering furnace

or sintering

furnace room

Process Safety

Control

Subsystem

Prevent formation

of an explosive

mixture of

hydrogen within

the MFFF

associated with

the use of the

hydrogen-argon

gas.

Process Safety

Control

Subsystem

Prevent formation

of an explosive

mixture of

hydrogen within

the MFFF

associated with

the use of the

hydrogen-argon

gas.

Process Safety

Control

Subsystem

Prevent formation

of an explosive

mixture of

hydrogen within the

MFFF associated

with the use of the

hydrogen-argon

gas.

Steam

Overpressurizati

on Explosion

Water entry into

the sintering

furnace from

failure of the

water level

controller in the

humidifier, from

the sintering

furnace, results

in a steam

explosion

Process Safety

Control

Subsystem

Ensure isolation

of the sintering

humidifier water

flow on highwater level.

Process Safety

Control

Subsystem

Ensure isolation

of the sintering

humidifier water

flow on highwater level.

Process Safety

Control

Subsystem

Ensure isolation of

the sintering

humidifier water

flow on high-water

level.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-70

Explosion 

(cont.)

RadiolysisInduced

Explosion

Buildup in the

vapor space of

an AP vessel

tank or piping

Hydrogen

buildup in the

vapor space of a

raffinates tank

Hydrogen

accumulation in

waste container

containing

hydrocarbons

Offgas

Treatment

System

Provide an

exhaust path for

the removal of the

diluted hydrogen

gas in process

vessels.

Offgas

Treatment

System

Provide an

exhaust path for

the removal of the

diluted hydrogen

gas in process

vessels.

Offgas

Treatment

System

Provide an exhaust

path for the

removal of the

diluted hydrogen

gas in process

vessels.

Instrument Air

System

(Emergency

Scavenging Air)

Provide sufficient

scavenging air to

dilute the

hydrogen

generated during

radiolysis.

Instrument Air

System

(Emergency

Scavenging

Air)

Provide sufficient

scavenging air to

dilute the

hydrogen

generated during

radiolysis.

Instrument Air

System

(Emergency

Scavenging

Air)

Provide sufficient

scavenging air to

dilute the hydrogen

generated during

radiolysis.

Waste

Containers

Ensure that

hydrogen buildup

in excess of

explosive limits

does not occur

while providing

appropriate

confinement of

radioactive

material.

Waste

Containers

Ensure that

hydrogen buildup

in excess of

explosive limits

does not occur

while providing

appropriate

confinement of

radioactive

material.

Waste

Containers

Ensure that

hydrogen buildup in

excess of explosive

limits does not

occur while

providing

appropriate

confinement of

radioactive

material.

Hydroxylamine

Nitrate (HAN)

Explosion—Proc

ess Vessels

containing HAN

and Hydrazine

Nitrate without

NOx Addition

Explosion in AP

process vessels

containing HAN

Process Safety

Control

Subsystem

Ensure that the

temperature of

the solution

containing HAN is

limited to

temperatures that

are within safety

limits.

Process

Safety Control

Subsystem

Ensure that the

temperature of

the solution

containing HAN is

limited to

temperatures that

are within safety

limits.

Process Safety

Control

Subsystem

Ensure that the

temperature of the

solution containing

HAN is limited to

temperatures that

are within safety

limits.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-71

Explosion 

(cont.)

Hydroxylamine

Nitrate (HAN)

Explosion—Proc

ess Vessels

containing HAN

and Hydrazine

Nitrate without

NOx Addition

(cont.)

Explosion in AP

process vessels

containing HAN

(cont.)

Chemical Safety

Control

Ensure that

concentration of

nitric acid, metal

impurities, and

HAN introduced

in the process are

within safety

limits.

Chemical

Safety Control

Ensure that

concentration of

nitric acid, metal

impurities, and

HAN introduced

in the process are

within safety

limits.

Chemical

Safety Control

Ensure that

concentration of

nitric acid, metal

impurities, and

HAN introduced in

the process are

within safety limits.

Hydroxylamine

Nitrate

Explosion—Proc

ess Vessels

Containing HAN

and Hydrazine

Nitrate with NOx

Addition

Explosion in AP

process vessels

containing HAN 

Chemical Safety

Control

Ensure that

concentrations of

HAN, hydrazine

nitrate, and

hydrazoic acid

are within the

safety limits.

Chemical

Safety Control

Ensure that

concentrations of

HAN, hydrazine

nitrate, and

hydrazoic acid

are within the

safety limits.

Chemical

Safety Control

Ensure that

concentrations of

HAN, hydrazine

nitrate, and

hydrazoic acid are

within the safety

limits.

Offgas

Treatment

System

Provide an

exhaust path for

the removal of

offgases

generated during

the

decomposition of

HAN, hydrazine

nitrate, and

hydrazoic acid.

Provide heat

transfer/ pressure

relief for affected

process vessels. 

Offgas

Treatment

System

Provide an

exhaust path for

the removal of

offgases

generated during

the

decomposition of

HAN, hydrazine

nitrate, and

hydrazoic acid.

Provide heat

transfer/ pressure

relief for affected

process vessels. 

Offgas

Treatment

System

Provide an exhaust

path for the

removal of

offgases generated

during the

decomposition of

HAN, hydrazine

nitrate, and

hydrazoic acid.

Provide heat

transfer/ pressure

relief for affected

process vessels. 
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-72

Explosion

(cont.)

Hydrogen

Peroxide

Explosion

Hydrogen

peroxide

explosion in AP

vessels, tanks,

and piping which

results in an

energetic breach

of the vessels,

tanks, and

piping and in a

loss of

confinement and

dispersal of

nuclear

materials

Chemical Safety

Control

Ensure that

explosive

concentrations of

hydrogen

peroxide do not

occur.

Chemical

Safety Control

Ensure that

explosive

concentrations of

hydrogen

peroxide do not

occur.

Chemical

Safety Control

Ensure that

explosive

concentrations of

hydrogen peroxide 

do not occur.

Solvent

Explosion

Solvents in AP

vessels, tanks,

and piping which

results in an

energetic breach

of the vessels,

tanks, and

piping and in a

loss of

confinement and

dispersal of

nuclear

materials

Process Safety

Control

Subsystem

Ensure the

temperature of

the solution

containing

solvents is limited

within the safety

limits.

Process Safety

Control

Subsystem

Ensure the

temperature of

the solution

containing

solvents is limited

within the safety

limits.

Process Safety

Control

Subsystem

Ensure the

temperature of the

solution containing

solvents is limited

within the safety

limits.

Process Cell

Fire Prevention

Features

Ensure that fires

in process cells

are highly

unlikely.

Process Cell

Fire Prevention

Features

Ensure that fires

in process cells

are highly

unlikely.

Process Cell

Fire Prevention

Features

Ensure that fires in

process cells are

highly unlikely.

Offgas

Treatment

System

Provide an

exhaust path for

removal of gases

in process

vessels.

Offgas

Treatment

System

Provide an

exhaust path for

removal of gases

in process

vessels.

Offgas

Treatment

System

Provide an exhaust

path for removal of

gases in process

vessels.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-73

Explosion

(cont.)

TBP-Nitrate

Explosion (Red

Oil)

Process-related

chemical

explosion

involving red oil

formation in the

AP boiler,

vessel, or tank

results in loss of

confinement and

dispersal of

nuclear

materials

Offgas

Treatment

System

Provide an

exhaust path for

aqueous phase

evaporative

cooling in 

process vessels,

thereby providing

a mechanism for

heat removal.

Provide venting of

vessels/equipmen

t that potentially

contain TBP and

its associated byproducts to

prevent over

pressurization in

the case of

excessive

oxidation of TBP

and/or its

degradation

products.

Offgas

Treatment

System

Provide an

exhaust path for

aqueous phase

evaporative

cooling in 

process vessels,

thereby providing

a mechanism for

heat removal.

Provide venting of

vessels/equipmen

t that potentially

contain TBP and

its associated byproducts to

prevent over

pressurization in

the case of

excessive

oxidation of TBP

and/or its

degradation

products.

Offgas

Treatment

System

Provide an exhaust

path for aqueous

phase evaporative

cooling in process

vessels, thereby

providing a

mechanism for

heat removal.

Provide venting of

vessels/ equipment

that potentially

contain TBP and its

associated byproducts to prevent

over pressurization

in the case of

excessive oxidation

of TBP and/or its

degradation

products.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-74

Explosion

(cont.)

TBP-Nitrate

Explosion (Red

Oil)

(cont.)

Process-related

chemical

explosion

involving red oil

formation in the

AP boiler,

vessel, or tank

results in loss of

confinement and

dispersal of

nuclear

materials

(cont.)

Process Safety

Control

Subsystem

Ensure that the

temperature of

solutions

containing

organics is

restricted to

temperatures

within safety

limits in order to

limit the rate of

energy

generation.

Ensure that the

design basis

heatup rate is not

exceeded.

Limit the

residence time of

organics in

process vessels

containing

oxidizing agents

and potentially

exposed to high

temperatures and

in radiation fields.

Process Safety

Control

Subsystem

Ensure that the

temperature of

solutions

containing

organics is

restricted to

temperatures

within safety

limits in order to

limit the rate of

energy

generation.

Ensure that the

design basis

heatup rate is not

exceeded.

Limit the

residence time of

organics in

process vessels

containing

oxidizing agents

and potentially

exposed to high

temperatures and

in radiation fields.

Process Safety

Control

Subsystem

Ensure the

temperature of

solutions

containing organics

is restricted to

temperatures within

safety limits in

order to limit the

rate of energy

generation.

Ensure that the

design basis

heatup rate is not

exceeded.

Limit the residence

time of organics in

process vessels

containing

oxidizing agents

and potentially

exposed to high

temperatures and

in radiation fields.

Chemical Safety

Control

Ensure a diluent

is used that does

not contain cyclicchain

hydrocarbons.

Chemical

Safety Control

Ensure a diluent

is used that does

not contain cyclicchain

hydrocarbons.

Chemical

Safety Control

Ensure a diluent is

used that does not

contain cyclic-chain

hydrocarbons.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-75

Explosion

(cont.)

AP Vessel

Overpressurizati

on Explosion

Overpressurizati

on of AP tanks,

vessels, and

piping as a

result of an

increase of

exothermic

chemical

reactions of

solutions

Fluid Transport

Systems

Ensure that

process vessels,

tanks, and piping

are designed to

prevent process

deviations from

creating

overpressurizatio

n events.

Fluid Transport

Systems

Ensure that

process vessels,

tanks, and piping

are designed to

prevent process

deviations from

creating

overpressurizatio

n events.

Fluid Transport

Systems

Ensure that

process vessels,

tanks, and piping

are designed to

prevent process

deviations from

creating

overpressurization

events.

Offgas

Treatment

Systems

Provide an

exhaust path for

removal of gases

in process

vessels.

Offgas

Treatment

Systems

Provide an

exhaust path for

removal of gases

in process

vessels.

Offgas

Treatment

Systems

Provide an exhaust

path for removal of

gases in process

vessels.

Chemical Safety

Controls

Ensure control of

the chemical

makeup of the

reagents.

Ensure

segregation/

separation of

vessels/

components from

incompatible

chemicals.

Chemical

Safety Controls

Ensure control of

the chemical

makeup of the

reagents.

Ensure

segregation/

separation of

vessels/

components from

incompatible

chemicals.

Chemical

Safety Controls

Ensure control of

the chemical

makeup of the

reagents.

Ensure

segregation/

separation of

vessels/

components from

incompatible

chemicals.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-76

Explosion

(cont.)

Pressure Vessel

Over

pressurization

Explosion

Overpressurizati

on of gas

bottles, tanks, or

receivers which

could impact

primary

confinement and

result in a

release of

radioactive

material

Pressure

Vessels

Controls

Ensure that

primary

confinements are

protected from

the impact of

pressure vessels

failures.

Pressure

Vessels

Controls

Ensure that

primary

confinements are

protected from

the impact of

pressure vessels

failures.

Pressure

Vessels

Controls

Ensure that

primary

confinements are

protected from the

impact of pressure

vessels failures.

Hydrazoic Acid 

(HN3) Explosion

Process-related

chemical

explosion

involving

HAN/nitric acid

vessels, tanks,

and piping which

results in the

breach of AP

vessels, tanks,

and piping

Chemical Safety

Control

Assure the proper

concentration of

hydrazine nitrate

is introduced into

the system.

Ensure that

hydrazoic acid is

not accumulated

in the process or

propagated into

the acid recovery

and oxalic mother

liquors recovery

units.

Chemical

Safety Control

Assure the proper

concentration of

hydrazine nitrate

is introduced into

the system.

Ensure that

hydrazoic acid is

not accumulated

in the process or

propagated into

the acid recovery

and oxalic mother

liquors recovery

units.

Chemical

Safety Control

Assure the proper

concentration of

hydrazine nitrate is

introduced into the

system.

Ensure that

hydrazoic acid is

not accumulated in

the process or

propagated into the

acid recovery and

oxalic mother

liquors recovery

units.

Process Safety

Control

Subsystem

Ensure that the

temperature

solutions

potentially

containing

hydrazoic acid is

limited to prevent

an explosive

concentration of

hydrazoic acid.

Process Safety

Control

Subsystem

Ensure that the

temperature

solutions

potentially

containing

hydrazoic acid is

limited to prevent

an explosive

concentration of

hydrazoic acid.

Process Safety

Control

Subsystem

Ensure that the

temperature

solutions

potentially

containing

hydrazoic acid is

limited to prevent

an explosive

concentration of

hydrazoic acid.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-77

Explosion

(cont.)

Metal Azide

Explosions

Chemical

explosion

involving an

azide (other than

hydrazoic acid in

an AP boiler,

vessel, or tank

that results in an

energetic breach

of the AP boiler,

vessel, or tank

Chemical Safety

Control

Ensure that metal

azides are not

added to high

temperature

process

equipment. 

Ensure that the

sodium azide has

been destroyed

before transfer of

the alkaline waste

into the high

alpha waste of

the waste

recovery unit.

Chemical

Safety Control

Ensure that metal

azides are not

added to high

temperature

process

equipment. 

Ensure that the

sodium azide has

been destroyed

before transfer of

the alkaline waste

into the high

alpha waste of

the waste

recovery unit.

Chemical

Safety Control

Ensure that metal

azides are not

added to high

temperature

process

equipment. 

Ensure that the

sodium azide has

been destroyed

before transfer of

the alkaline waste

into the high alpha

waste of the waste

recovery unit.

Process Safety

Control

Subsystem

Ensure that metal

azides are not

exposed to

temperatures that

can allow the

energetic

decomposition of

azides.

Limit and control

conditions in

which dryout can

occur.

Process Safety

Control

Subsystem

Ensure that metal

azides are not

exposed to

temperatures that

can allow the

energetic

decomposition of

azides.

Limit and control

conditions in

which dryout can

occur.

Process Safety

Control

Subsystem

Ensure that metal

azides are not

exposed to

temperatures that

can allow the

energetic

decomposition of

azides.

Limit and control

conditions in which

dryout can occur.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-78

Explosion

(cont.)

Pu (IV) Oxalate

Explosion

Chemical

explosion

involving Pu (IV)

oxalate in the

calcining

furnace results

in an energetic

breach of the

furnace and

glovebox and

the dispersal of

radioactive

material

Chemical Safety

Control

Measure the

valence of

plutonium before

adding oxalic acid

to the oxalic

precipitation and

oxidation unit to

ensure that

plutonium (IV)

cannot be

formed.

Chemical

Safety Control

Measure the

valence of

plutonium before

adding oxalic acid

to the oxalic

precipitation and

oxidation unit to

ensure that

plutonium (IV)

cannot be

formed.

Chemical

Safety Control

Measure the

valence of

plutonium before

adding oxalic acid

to the oxalic

precipitation and

oxidation unit to

ensure that

plutonium (IV)

cannot be formed.

ElectrolysisRelated

Explosion

Explosion of

hydrogen in the

vapor space on

the electrolyzer

Process Safety

Control

Subsystem

Limit the

generation of

hydrogen to

ensure that the

normality of the

acid is sufficiently

high to ensure

that the offgas is

not flammable.

Monitor the

electrolyzer for

electrical faults

that could result

in arcing or other

imparting of

electrical energy

with the risk of

titanium fire.

Process Safety

Control

Subsystem

Limit the

generation of

hydrogen to.

ensure that the

normality of the

acid is sufficiently

high to ensure

that the offgas is

not flammable.

Monitor the

electrolyzer for

electrical faults

that could result

in arcing or other

imparting of

electrical energy

with the risk of

titanium fire.

Process Safety

Control

Subsystem

Limit the

generation of

hydrogen to ensure

that the normality

of the acid is

sufficiently high to

ensure that the offgas is not

flammable.

Monitor the

electrolyzer for

electrical faults that

could result in

arcing or other

imparting of

electrical energy

with the risk of

titanium fire.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-79

Explosion

(cont.)

ElectrolysisRelated

Explosion

(cont.)

Explosion of

hydrogen in the

vapor space on

the electrolyzer

(cont.)

Electrolyzer

Structure,

Sintered Silicon

Nitride Barrier,

and PTFE

insulator

To provide

physical

separation and

electrical

insulation

between

electrolyzer

components and

structural integrity

and stability to

the electrolyzer

system.

Ensures that the

normality of the

acid is sufficiently

high to ensure

that the offgas is

not flammable.

Electrolyzer

Structure,

Sintered

Silicon Nitride

barrier, and

PTFE insulator 

To provide

physical

separation and

electrical

insulation

between

electrolyzer

components and

structural integrity

and stability to

the electrolyzer

system.

Ensures that the

normality of the

acid is sufficiently

high to ensure

that the offgas is

not flammable.

Electrolyzer

Structure,

Sintered

Silicon Nitride

Barrier, and

PTFE insulator

To provide physical

separation and

electrical insulation

between

electrolyzer

components and

structural integrity

and stability to the

electrolyzer

system.

Ensures that the

normality of the

acid is sufficiently

high to ensure that

the offgas is not

flammable.

Laboratory

Explosion

Explosion within

the BMF

laboratory

involving

flammable,

explosive, or

reactive

chemicals which

results in a

dispersal of

radiological

material

C3 Confinement Provide filtration

to mitigate

dispersions from

the C3 areas.

Chemical

Safety Control

Ensures control

of the chemical

makeup of the

laboratory

reagents.

Ensures

segregation/

separation of

vessels/

components from

incompatible

chemicals.

C3

Confinement

Provides filtration

to mitigate

dispersions from

the C3 areas.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-80

Explosion

(cont.)

Laboratory

Explosion

(cont.)

Explosion within

the BMF

laboratory

involving

flammable,

explosive, or

reactive

chemicals which

results in a

dispersal of

radiological

material

(cont.)

Laboratory

Materials

Controls

Minimize the

quantities of

hazardous

chemical/

radiological

materials in the

laboratory

Facility Worker

Action

Ensures that

facility workers

take proper

actions to limit

radiological/

chemical

exposures.

Outside

Explosion

Explosions in

the reagent

processing

building, gas

storage area,

emergency

diesel generator

building, and the

access building

MOX Fuel

Fabrication

Building

Structure

Maintain

structural integrity 

and prevent

damage to

internal SSCs.

MOX Fuel

Fabrication

Building

Structure

Maintain

structural integrity 

and prevent

damage to

internal SSCs.

MOX Fuel

Fabrication

Building

Structure

Maintain structural

integrity and

prevent damage to

internal SSCs.

Emergency

Generator

Building

Structure

Maintain

structural integrity

and prevent

damage to

internal SSCs

from external

explosions.

Emergency

Generator

Building

Structure

Maintain

structural integrity

and prevent

damage to

internal SSCs

from external

explosions.

Emergency

Generator

Building

Structure

Maintain structural

integrity and

prevent damage to

internal SSCs from

external

explosions.

Waste Transfer

Line

Prevent damage

to the line from

outside

explosions.

Waste Transfer

Line

Prevent damage

to the line from

outside

explosions.

Waste Transfer

Line

Prevent damage to

the line from

outside explosions.
Table 5.1-2 Safety Assessment Summary (Process Hazards)

Event Type Event Group Bounding

Event

PSSC 

(SW & IOC)

Safety

Function

PSSC (facility

worker)

Safety

Function

PSSC

(environment)

Safety

Function

5-81

Explosion

(cont.)

Outside

Explosion

(cont.)

Explosions in

the reagent

processing

building, gas

storage area,

emergency

diesel generator

building, and the

access building

(cont.)

Hazardous

Materials

Delivery

Controls

Ensures that the

quantity of

delivered

hazardous

material and its

proximity to the

outside

explosion’s

PSSCs are

controlled within

the bounds of the

values used to

demonstrate that

the

consequences of

outside

explosions are

acceptable.

Hazardous

Materials

Delivery

Controls

Ensures that the

quantity of

delivered

hazardous

material and its

proximity to the

outside

explosion’s

PSSCs are

controlled within

the bounds of the

values used to

demonstrate that

the

consequences of

outside

explosions are

acceptable.

Hazardous

Materials

Delivery

Controls

Ensures that the

quantity of

delivered

hazardous material

and its proximity to

the outside

explosion’s PSSCs

are controlled

within the bounds

of the values used

to demonstrate that

the consequences

of outside

explosions are

acceptable.
Table 5.1-3 PSSCs and Design Basis Functions and Values Developed from the Safety Assessment

5-82

PSSC Design Basis Safety Function Design Basis Values FSER

Section

3013 Canister Withstand the effects of design

basis drops without breaching

DOE-STD-3013-2000

Outer canister designed to withstand 9.14-m (30-ft) drop while remaining leak tight. Inner container designed

to withstand 1.22-m (4-ft) drop while remaining leak tight. Outer container designed to withstand 4927 kPa

(699 psig), inner container withstands 790 kPa (100 psig). Qualified 50-yr life.

11.7.1.2

3013 Transport

Cask

Withstand design basis fire Thermal design per 10 CFR 71.73, 800 °C (1472 °F) for 30 minutes. 7.1.5.2

Withstand design basis drop Designed for free drop, crushing, and puncture per 10 CFR 71.73 11.7.1.2

Backflow

Prevention

features

Prevent process fluids from

backflowing into interfacing

systems

ASME B31.3 11.8

C2 Confinement

System Passive

Barrier

Limit the dispersion of

radioactive material

Two HEPA filter banks before discharge

HEPA filter design temperature of 232 °C (450 °F)

Automatic and manual fire-rated dampers between designated fire areas

In-place HEPA filter testing for final discharge filtration assemblies

System design in accordance with RG 3.12, except heat removal is by airflow dilution

HEPA filter design, HEPA filter housing design, construction and testing, and HEPA filter housing isolation 

dampers in accordance with ASME N509;

HEPA filter design and testing, HEPA filter housing design and testing;,ductwork and pipe flexible 

connections, and fan design, construction, and testing in accordance with ASME AG-1

Sheet metal ductwork design, construction, and testing; “bubble tight” isolation damper construction; and

testing; HEPA filter housing testing; and HEPA filter testing in accordance with ERDA 76-21;

Filter testing in accordance with ASME N510 with each HEPA stage having a leakage efficiency of 99.95%

Tornado dampers

Final filters and downstream ductwork remain structurally intact during and after tornadoes and design basis

earthquakes

11.4.1.3
Table 5.1-3 PSSCs and Design Basis Functions and Values Developed from the Safety Assessment

PSSC Design Basis Safety Function Design Basis Values FSER

Section

5-83

C3 Confinement

System

Provide filtration to mitigate

dispersions from the C3 areas

C3 zone pressure maintained at negative pressure with respect to atmosphere during normal operation and

transients

Two 100% capacity fans in C3 confinement system

System design in accordance with RG 3.12, except heat removal is by airflow dilution

HEPA filter design; HEPA filter housing design, construction and testing; and HEPA filter housing isolation

dampers in accordance with ASME N509

HEPA filter design and testing; HEPA filter housing design and testing; ductwork and pipe flexible connections;

and fan design, construction, and testing in accordance with ASME AG-1

Sheet metal ductwork design, construction, and testing; “bubble tight” isolation damper construction and

testing; HEPA filter housing testing; and HEPA filter testing in accordance with ERDA 76-21

Filter testing in accordance with ASME N510 with each HEPA stage having a leakage efficiency of 99.95%

Tornado dampers

Fan power from normal (non-PSSC), standby (non-PSSC), and emergency (PSSC) supplies

Remains operational after facility fires, tornadoes, and design basis earthquakes

11.4.1.3

Remain operable during design

basis fire and effectively filter

any release

Spark arrestors (roughing filters) in each filtration assembly upstream of HEPA filters

Fire-rated dampers between designated fire areas

In-place HEPA filter testing for final discharge filtration assemblies

HEPA filter design temperature of 232 °C (450 °F)

 7.1.5.5

11.4.1.3

Limit the dispersion of

radioactive material

Designed to maintain exhaust safety function assuming single active component failure

HEPA filter assembly release fraction of 1 × 10-4

Two, 100%-capacity redundant assemblies of two HEPA filter banks before discharge

11.4.1.3

Provide exhaust to ensure that

temperature in the 3013

canister storage structure is

maintained within design limits.

Maintain ambient temperatures with sufficient airflow in the canister storage structure

Reliability design bases are as described above

11.4.1.3

Provide cooling air exhaust

from designated electrical

rooms

Maintain ambient temperatures with sufficient airflow in the designated electrical rooms

Reliability design bases are as described above

11.4.1.3

C4 Confinement

System

Remain operable during design

basis fire and effectively filter

any release

Fire isolation valves where process room exhausts exit fire areas

In-place HEPA filter testing for final discharge filtration assemblies

HEPA filter design temperature of 232 °C (450 °F)

7.1.5.5

11.4.1.3
Table 5.1-3 PSSCs and Design Basis Functions and Values Developed from the Safety Assessment

PSSC Design Basis Safety Function Design Basis Values FSER

Section

5-84

C4 Confinement

System (cont.)

Maintain negative glovebox

pressure between glovebox and

interfacing systems

Same as above, as appropriate

C4 zone pressure maintained at negative pressure with respect to C3 zone during normal operation and

transients

Redundant pressure sensors to maintain C4 pressures

11.4.1.3

Maintain minimum inward flow

through small glovebox

releases

Same as above, as appropriate

High-capacity flow system [38.1 m/min (125 ft/min) in the event of glovebox breach to maintain negative

pressure

11.4.1.3

Ensure that C4 exhaust is

effectively filtered

Same as above, as appropriate

In-place HEPA filter testing for final discharge filtration assemblies

System design in accordance with RG 3.12, except heat removal is by airflow dilution

HEPA filter design; HEPA filter housing design, construction and testing; and HEPA filter housing isolation

dampers in accordance with ASME N509

HEPA filter design and testing; HEPA filter housing design and testing, ductwork and pipe flexible connections;

and fan design, construction, and testing in accordance with ASME AG-1

Filter testing in accordance with ASME N510 with each HEPA stage having a leakage efficiency of 99.95%

11.4.1.3

Operate to ensure that a

negative pressure differential

exists between the C4 glovebox

and the C3 area.

Fan power from normal (non-PSSC), standby (non-PSSC), emergency (PSSC), and uninterruptible (PSSC)

supplies

Remains operational during facility fires and tornadoes and design basis earthquakes

Four, 100%-capacity fans in C4 discharge system

Piping, valves, and fittings associated with gloveboxes in accordance with ASME B31.3

11.4.1.3

Contain a chemical release

within a glovebox and provide

an exhaust path for removal of

the chemical vapors

Contain chemicals within C4 and exhaust so that moderate chemical consequence limits are not exceeded

outside

8.1.2.4.1

Chemical

Safety Controls

Ensure that explosive

concentrations of hydrogen

peroxide do not occur.

Limit the received H2O2 solution concentrations to 35% or less 8.1.2.5.4

Ensure a diluent is used that is

not very susceptible to either

nitration or radiolysis

Diluent does not contain cyclic hydrocarbons 8.1.2.5.5
Table 5.1-3 PSSCs and Design Basis Functions and Values Developed from the Safety Assessment

PSSC Design Basis Safety Function Design Basis Values FSER

Section

5-85

Chemical

Safety controls

(cont.)

Ensure that hydrazoic acid is

not accumulated in the process

or propagated to units that

might lead to explosive

conditions.

Maximum hydrazine concentration of 0.14M. Hydrazine yield of 39.3% or less. No hydrazoic acid

accumulation into acid recovery and OML recovery units.

8.1.2.5.3

Ensure metal azides are not

introduced into high

temperature process equipment

No addition to high-temperature equipment. Tanks potentially containing azides not allowed to dry out. 8.1.2.5.3.3

Ensure the sodium azide has

been destroyed before the

transfer of the alkaline waste to

the waste recovery unit.

Azides completely destroyed (OM) before acidification 8.1.2.5.3.3

Ensure the valance of the

plutonium before oxalic acid

addition is not VI.

Pu(VI) concentration will be low—actual value to be derived at ISA stage 8.1.2.5.6

Ensure that nitric acid, metal

impurities, and HAN

concentrations are controlled

and maintained to within safety

limits.

[HNO3] < 6 M

[N2H4] $ 0.1 M

[HAN] < 2.5 M

8.1.2.5.3.1

Ensure concentrations of HAN,

hydrazine nitrate, and hydrazoic

acid are controlled to within

safety limits.

8.1.2.5.3

Ensure the proper

concentration of hydrazine

nitrate is introduced into the

system.

Hydrazine is not added at concentrations exceeding 35% (as N2H4) 8.1.2.5.3.1
Table 5.1-3 PSSCs and Design Basis Functions and Values Developed from the Safety Assessment

PSSC Design Basis Safety Function Design Basis Values FSER

Section

5-86

Chemical

Safety controls

(cont.)

Ensure control of the chemical

makeup of the reagents and

ensure segregation/separation

of vessels/components from

incompatible chemicals.

Separation and prevention of mixing of incompatible reagents 8.1.2.5

Combustible

Loading

Controls

Limit combustibles in C2 filter

area to ensure that the C4 final

HEPA filters are not adversely

impacted by a filter room fire.

Based on defense-in-depth principles and multiple layers of protection. Includes control of fixed combustibles

by design and control of transient combustibles by design and during operations (through worker training,

regular surveillance, and postings). Utilizes NFPA 801.

7.1.5.1

Limit the quantity of

combustibles in fire areas

containing a storage glovebox

and the secured warehouse

such that any fire that may

occur will not encompass a

large fraction of the stored

radiological material. 

Same as above 7.1.5.1

Limit combustible in areas

containing 3013 canisters

Same as above 7.1.5.1

Limit combustibles in a fire area

containing 3013 transport casks

Same as above 7.1.5.1

Limit combustibles in a fire area

containing fuel rods.

Same as above 7.1.5.1

Limit combustibles in areas

containing MOX fuel transport

casks.

Same as above 7.1.5.1

Limit the quantity of

combustibles in areas

containing transfer containers.

Same as above 7.1.5.1
Table 5.1-3 PSSCs and Design Basis Functions and Values Developed from the Safety Assessment

PSSC Design Basis Safety Function Design Basis Values FSER

Section

5-87

Combustible

Loading

Controls (cont.)

Limit the quantity of

combustibles in areas

containing the pneumatic

transfer system to ensure this

system is not adversely

impacted.

Same as above 7.1.5.1

Criticality

controls 

Prevent criticality events 1. Design of facility operations shall comply with the double contingency principle, as stated in ANSI/ANS-8.1. 

Nuclear criticality shall be made “highly unlikely” and the failure of each leg of double contingency shall be

“unlikely.”

2. Computer calculations shall not exceed a maximum keff, taking all uncertainties and biases into account. 

Description of calculational methods and their validation, or means of establishing subcritical margins if

parameter limits are not based on computer calculations.

3. Facility operations shall be designed to be subcritical under both normal and credible abnormal conditions. 

Normal conditions will be considered to be those when all controlled parameters are at their controlled values

and uncontrolled parameters at their worst credible values. Abnormal conditions shall consider the worst case

upset for each loss of a control or controlled parameter.

4. Dominant nuclear criticality safety controlled parameters shall be specified for each major process and in

their order of preference.

5. Design approach shall prefer engineered over administrative controls and passive over active engineered

controls.

6. The facility shall have a criticality accident alarm system that complies with the requirements of 10 CFR 

70.24. The detection system and its operating characteristics shall be described.

7. The management measures and how they are applied to each controlled parameter shall be described,

along with the safety grades for criticality IROFS and the criteria used to assign these IROFS to individual

safety grades.

8. A description of the organization and administration for NCS and the key elements of the NCS Program

(including those in SRP Section 6.4.3.2).

9. A description of the technical practices used to determine limits and controls on each controlled parameter,

in criticality safety evaluations, including what ANSI/ANS standards are being committed to in whole or in part.

10. Where moderation control is required for subcriticality, a description of the approach to designing the

facility to meet both fire safety and criticality safety requirements (including presence and type of fire

suppression).

 6.1.3.4.1

 6.1.4.3

Double-Walled

Pipe

Prevent leaks from pipes

containing process fluids from

leaking into C3 areas.

ANSI/ASME B31.3 11.8
Table 5.1-3 PSSCs and Design Basis Functions and Values Developed from the Safety Assessment

PSSC Design Basis Safety Function Design Basis Values FSER

Section

5-88

Electrolyzer

Structure

Provides support for the

electrolysis pot.

Seismically designed, designed to withstand turbulent flow, maintains geometry 11.2.1.3.4

Emergency AC

Power System

Provide ac power to emergency

dc system battery charger

Overall design per IEEE Std 308-1991 and RG 1.32 (Rev.2). Environmental and seismic qualification per

ANSI/AISC N690-1994, ASCE 4-98, IEEE Std 323-1983, IEEE Std 344-1987, RG 1.61, and RG 1.100 (Rev.2).

Designed for single failure per IEEE Std 379-1994. Electrical independent and separation per IEEE 384-1992

and RG 1.75 (Rev. 2). Periodic testing per IEEE Std 338-1987 and RG 1.118 (Rev. 3). Electrical cables in

open trays qualified per IEEE Std 383-1974. Equipment protection based on IEEE Std 741-1997. Battery

design and installation per IEEE Std 484-1996. Emergency diesel generators with overall design per IEEE Std

387-1995 and RG 1.9 (Rev. 3) and fuel oil per ANSI/ASTM D975-94. Overall design of uninterruptible power

supplies per IEEE Std 944-1986.

11.5.1.3.1

Provide ac power to emergency

diesel generator fuel oil system

Same as above 11.5.1.3.1

Provide ac power to high

depressurization exhaust

system

Same as above 11.5.1.3.1

Provide ac power to emergency

control room air-conditioning

system

Same as above 11.5.1.3.1

Provide ac power to C4

confinement system

Same as above 11.5.1.3.1

Provide ac power to emergency

diesel ventilation system

Same as above 11.5.1.3.1

Provide ac power to emergency

control system

Same as above 11.5.1.3.1

Provide ac power to seismic

monitoring and trip system and

seismic isolation valves.

Same as above 11.5.1.3.1

Provide ac Power to process

cell exhaust system

Same as above 11.5.1.3.1
Table 5.1-3 PSSCs and Design Basis Functions and Values Developed from the Safety Assessment

PSSC Design Basis Safety Function Design Basis Values FSER

Section

5-89

Emergency

Control Room

Air-Conditioning

System

Ensure habitable conditions for

operators

One, 100%-capacity filtration stage (using prefilter stage, two HEPA filter stages, and chemical filters) for each

control room air supply 

One, 100%-capacity air-conditioning unit for each engineering control room and for each emergency electrical

battery room

Two, 100%-capacity exhaust fans for each emergency battery room

Designed to maintain protection assuming single component failure

HEPA filter design temperature of 232 °C (450 °F) 

Tornado dampers prevent pressurization in supply air system

In-place HEPA filter testing for final discharge filtration assemblies

System design in accordance with RG 3.12

HEPA filter design; HEPA filter housing design, construction and testing; and HEPA filter housing isolation

dampers in accordance with ASME N509

HEPA filter design and testing; HEPA filter housing design and testing; ductwork and pipe flexible connections;

and fan design, construction, and testing in accordance with ASME AG-1

Sheet metal ductwork design, construction, and testing; “bubble tight” isolation damper construction and

testing; HEPA filter housing testing; and HEPA filter testing in accordance with ERDA 76-2;

Filter testing in accordance with ASME N510 with each HEPA stage having a leakage efficiency of 99.95%

Fan power from normal (non-PSSC), standby (non-PSSC), and emergency (PSSC) supplies

Remains operational during and after facility fires and after tornadoes and design basis earthquakes

Fresh air inlets are located so that the presence of contaminants are minimized (NFPA 801)

11.4.1.3

8.1.2.6

7.1.5.4

Emergency

Control System

Provide controls for high

depressurization exhaust

system

Two redundant, separate, and independent trains. Fundamental design as per IEEE 603-1998. Electrical

independence and separation as per IEEE 384-1992 and RG 1.75 (Rev. 2). Single failure criteria as per IEEE

379-1994. Instrument setpoints as per ANSI/ISA 67.04.01-2000, and RG 1.105 (Rev. 3). Designed to function

during design basis event as per ANSI/AISC N690-1994, ASCE 4-98, IEEE Std 323-1983, IEEE Std 344-1987,

and RG 1.100 (Rev.2). Software programmable electronic systems per EPRI Topical Report TR-106439 (with

NRC safety evaluation), IEC 61131-3 (1993-03), IEEE Std 7-4.3.2-1993, IEEE Std 730-1998, IEEE Std 828-

1998, IEEE Std 830-1998, IEEE Std 1012-1998, IEEE Std 1028-1997, IEEE Guide 1042-1987, IEEE Std 1074-

1997, IEEE Std 1228-1994, NUREG/CR-6090, NUREG/CR-6463, RG 1.168, RG 1.169, RG 1.172, and RG

1.173. Human-system interface per IEEE Std 1023-1988 and NUREG-0700. Seismic monitoring per RG 3.17

and periodic testing per IEEE Std 338-1987, NUREG-0800 (Branch Technical Position HICB-17), and RG

1.118 (Rev. 3). Reduction of electromagnetic and radiofrequency interference per IEEE Std 518-1982, IEEE St

Std 1050-1996, and RG 1.180 with the design of data communications networks per ANSI/IEEE 802.3.

11.6.1.3.1
Table 5.1-3 PSSCs and Design Basis Functions and Values Developed from the Safety Assessment

PSSC Design Basis Safety Function Design Basis Values FSER

Section

5-90

Emergency

Control System

(continued)

Provide controls for C4

confinement system

Same as above 11.6.1.3.1

Provide controls for emergency

control room air-conditioning

system

Same as above 11.6.1.3.1

Provide controls for emergency

ac system

Same as above 11.6.1.3.1

Provide controls for emergency

dc system

Same as above 11.6.1.3.1

Provide controls for emergency

generator ventilation system

Same as above 11.6.1.3.1

Provide controls for emergency

diesel generator fuel oil system

Same as above 11.6.1.3.1

Shutdown process on loss of

power

Same as above 11.6.1.3.1

Shut down and isolate process

and systems (as necessary) in

response to an earthquake

Same as above for seismic monitoring and trip system 11.6.1.3.1

Provide controls for process cell

exhaust system

Same as above 11.6.1.3.1

Emergency DC

System

Provide dc power for high

depressurization exhaust

system

Overall design per IEEE Std 308-1991, IEEE Std 946-1992, and RG 1.32 (Rev.2). Environmental and seismic

qualification per ANSI/AISC N690-1994, ASCE 4-98, IEEE Std 323-1983, IEEE Std 344-1987, and RG 1.100

(Rev.2). Designed for single failure per IEEE Std 379-1994. Electrical independent and separation per IEEE

384-1992 and RG 1.75 (Rev. 2). Periodic testing per IEEE Std 338-1987, IEEE Std 450-1995, and RG 1.118

(Rev. 3). Battery design and installation per IEEE Std 484-1996, IEEE Std 485-1997, and NFPA 111. 

11.5.1.3.2

Provide dc power for C4

confinement system

Same as above 11.5.1.3.2
Table 5.1-3 PSSCs and Design Basis Functions and Values Developed from the Safety Assessment

PSSC Design Basis Safety Function Design Basis Values FSER

Section

5-91

Emergency DC

System

(continued)

Provide dc power for

emergency ac power system

controls

Same as above 11.5.1.3.2

Provide dc power for

emergency control room airconditioning system

Same as above 11.5.1.3.2

Provide dc power for

emergency control system

Same as above 11.5.1.3.2

Provide dc power for

emergency generator

ventilation system

Same as above 11.5.1.3.2

Provide dc power to process

cell exhaust system

Same as above 11.5.1.3.2

Emergency

Diesel

Generator

Structure

Maintain structural integrity and

prevent damage to internal

SSCs from external fires,

external explosions,

earthquakes, extreme winds,

tornadoes, missiles, rain, and

snow and ice loadings

[Text removed under 10 CFR 2.390]

7.1.5.4

11.1.1.3

Emergency

Generator

Ventilation

System

Provide emergency diesel

generator ventilation

One, 100%-capacity air-conditioning unit for each switchgear room

One ,100%-capacity roof ventilator for engine room cooling during standby (engine fan cools room during

engine operation)

Fan power from normal (non-PSSC), standby (non-PSSC), and emergency (PSSC) supplies

Remains operational after facility fires, tornadoes, and design basis earthquakes;

11.4.1.3

Emergency

Diesel

Generator Fuel

Oil System

Provide emergency diesel

generator fuel oil for the

emergency diesels

7 days plus margin fuel storage tank, day tanks 2498 L (660 gal), dual 100% transfer pumps, strainers, dual

cartridge filters, isolation and maintenance valves. IEEE 344-1987, RG 1.100 (Rev. 2), IEEE 308-1991, ANS

59-51-1997, ASTM D75-94, NFPA-37, NFPA-110

11.9.1.1, 

11.9.1.3

Facility Worker

Action

Ensure that facility workers take

proper action to limit chemical

and/or radiological exposure

Facility worker response to exit the affected area. 9.1.2.4
Table 5.1-3 PSSCs and Design Basis Functions and Values Developed from the Safety Assessment

PSSC Design Basis Safety Function Design Basis Values FSER

Section

5-92

Facility Worker

Controls

Ensure that facility workers take

proper actions before bagout

operations to limit radiological

exposure

Facility worker prejob preparation to prevent and/or limit dose during tasks involving transient primary

confinements or maintenance

9.1.2.4

Ensure that facility workers take

proper actions before

maintenance activities to limit

radiological exposure

Same as above 9.1.2.4

Fire Barriers Contain fires within fire area Minimum rating of 2 hours. Constructed in accordance with NFPA 221-1997. Fire doors are designed in

accordance with NFPA 80-1999. Fire damper per UL 555-1995. Barrier selection and penetration seal

program per NFPA 801-1998.

 7.1.5.6

Fire Detection

and

Suppression

Support fire barriers as

necessary

Detection and alarm per NFPA 72-1996

Suppression per NFPA 2001-1996 (clean agent) where dispersable fissile material is present

7.1.5.7

Fluid Transport

System

Prevent overpressurization ASME Section Boiler & Pressure Vessel Code VIII

ASME B31.3

Effectiveness for reactive chemicals (HAN, red oil) not specified

11.8.1.3

Withstand as necessary the

effects of the DBE such that

confinement of radionuclides is

maintained

 Seismic Category I design as per seismic qualification program 11.8

Glovebox Maintain confinement integrity

for design basis impacts

Leak integrity 2.5 × 10-3 vol/hr @ 500 Pa. Impact resistant windows, glovebox floor designed to withstand load

drops. Internal guides and barriers to prevent fall of containers. Have pressure relief devices.

Welding per AWS D9.1-1998.

11.7.1.2

Glovebox

Pressure

controls

Maintain glovebox pressure

within design limits

Redundant pressure sensors to monitor differential pressures and provide alarm

Remains operational after facility fires in nonaffected areas, tornadoes, and design basis earthquakes

11.4.1.3

Guide Sleeves Provides insulation/separation

between anode and titanium

shell of electrolyzer

Capable of withstanding environmental conditions in electrolyzer 11.2.1.3.4
Table 5.1-3 PSSCs and Design Basis Functions and Values Developed from the Safety Assessment

PSSC Design Basis Safety Function Design Basis Values FSER

Section

5-93

Hazardous

Material

Delivery

Controls

Ensure that the quantity of

delivered hazardous material

and its proximity to the MFFF

building structure, EDG building

structure, and the waste

transfer line are controlled to

within the bounds of the values

used to demonstrate that the

consequences of outside

explosions are acceptable

Limit quantities and distances of deliveries so that they are within the bounds found acceptable in the safety

analyses performed for potential explosives in F-area

5.1.6.2

Instrument Air

System

(Emergency

Scavenging Air)

Provide sufficient scavenging

airflow to dilute the hydrogen

produced by radiolysis such

that an explosive condition

does not occur

Limiting hydrogen concentration to 1% or less. Two, 100%-capacity banks of compressed air available. Will

be constructed to ASME B&PV and B31.3 standards. Also RG 1.100 or IEEE 344.

11.9.1.1,

11.9.1.2,

11.2.1.11,

8.1.2.5.1.2

Laboratory

Material

Controls

Minimize quantities of

hazardous chemicals in the

laboratory

Procedures will be established to limit sample size, number, and reagent quantity, in accordance with safe

laboratory operating practices

8.1.2.1.3

Minimize quantities of

radioactive materials in the

laboratory

Procedures will be established to limit sample size, number, and reagent quantity, in accordance with safe

laboratory operating practices

8.1.2.1.3

Maintenance

Activity Controls

Isolation of power to the

electrolyzer when electrolyzer is

drained

Administrative controls 11.2.1.3.4

Material

Handling

Controls

Ensure proper handling of

primary confinement types

outside of gloveboxes

Management measures include training and qualification of personnel, approved procedures, including

precautions and limitations, use of proper equipment, testing, and surveillances

 11.7.1.2

Ensure that design bases lift

heights are not exceeded

MOX fresh fuel casks—9.14-m (30-ft) drop

MOX waste containers—1-m (3.28-ft) drop

3013 outer can—9.14-m (30-ft) drop

11.7.1.2
Table 5.1-3 PSSCs and Design Basis Functions and Values Developed from the Safety Assessment

PSSC Design Basis Safety Function Design Basis Values FSER

Section

5-94

MaterialHandling

Controls (cont.)

Prevent load-handling activities

that could potentially lead to a

breach in the final C4 HEPA

filters

PSSC for structural protection is C4 confinement system 11.4.1.3

Prevent impacts to the inside or

outside of glovebox during

normal operations

Engineered equipment used to reduce likelihood of failures causing glovebox breaches 11.7.1.2

Prevent potential

overpressurization of the

reusable plutonium dioxide

cans caused by radiolysis or

oxidation of Pu (III) oxalate and

its subsequent impact to the

glovebox

The reusable can is designed to withstand the maximum pressure attainable from radiolysis and plutonium (III)

oxalate reactions, plus 10%

11.3.1.2.3

Prevent load-handling events

that could breach primary

confinements

During normal operations; material handling equipment, material handling controls, and the glovebox will

prevent breaches. During maintenance operations the above plus training and procedures will be used.

11.7.1.2

Material

Handling

Equipment

Limit damage to fuel

rods/assemblies during

handling

Designed using hardware stops, limit switches, speed controllers, bumpers to limit travel of equipment. Will fail

to safe condition upon loss of power.

11.7.1.1,

11.7.1.2

Prevent impacts to the

glovebox through the use of

engineered equipment

Same as above same

Material

Maintenance

and

Surveillance

Programs

Detect and limit the damage

resulting from corrosion

Corrosion allowance in accordance with rational code practices 11.8.1.3

BMF Tornado

Dampers

Protect BMF ventilation

systems from differential

pressure effects of the tornado

Designed per ASME AG-1 11.4.1.3
Table 5.1-3 PSSCs and Design Basis Functions and Values Developed from the Safety Assessment

PSSC Design Basis Safety Function Design Basis Values FSER

Section

5-95

Missile Barriers Protect BMF, BEG, and UEF

building internal SSCs from

damage caused by tornado or

wind-driven missiles

Designed for impacts from:

[Text removed under 10 CFR 2.390]

11.1.1.3.2.1

MOX Fuel

Fabrication

Building

Structure

(including vent

stack)

Maintain structural integrity and

prevent damage to internal

SSCs from external events

[Text removed under 10 CFR 2.390]

7.1.5.4

11.1.1.3.2.2

Withstand the effects of load

drops that could potentially

impact radiological material

NUREG-0612 11.1.1.3.2.1

MOX Fuel

Transport Cask

Withstand the design basis fire Thermal design per 10 CFR 71.73, 800 °C (1,472 °F) for 30 minutes 7.1.5.2

Withstand the effects of design

basis drops without release of

radioactive material

Mechanical design per 10 CFR 71.73, certified to withstand 9.14-m (30-ft) drop 11.7.1.2

Offgas

Treatment

System

Provide an exhaust path for the

removal of gases in process

vessels

Process vessels do not pressurize

Final filtration units in accordance with Table 11.4-1 of FSER 11.4.1

Provide an exhaust path for

aqueous phase evaporative

cooling in process vessels,

thereby providing a mechanism

for heat removal

Relieve 1.2 times the combination of energy generation and energy input to the system 8.1.2.5.5

Provide venting of

vessels/equipment that

potentially contain TBP and its

associated byproducts to

prevent overpressurization in

the case of excessive oxidation

of TBP and/or its degradation

products

Vent sizes > 8 × 10-3 mm2/g (5.6 × 10

-3 in2

/lb)of organic material present in the system 8.1.2.5.5
Table 5.1-3 PSSCs and Design Basis Functions and Values Developed from the Safety Assessment

PSSC Design Basis Safety Function Design Basis Values FSER

Section

5-96

Polytetrafluoroethylene

Insulator

(PTFE)

Provides insulation/separation

between cathode and anode

structures, and anode and

ground of electrolyzer

Capable of withstanding environmental conditions in electrolyzer 11.2.1.3.4

Pressure

Vessel Controls

Ensure that primary

confinements are protected

from the impact of pressure

vessel failures (bulk gas,

breathing air, service air, and

instrument air systems)

Limited by ASME Section VIII and ASME B31.3 code design 11.8.1.3

11.9.1.2

Process Cells Contain fluid leaks within

process cells

Fully welded, designed to handle maximum inventory of largest vessel in cell 11.7

Process Cell

Entry Controls

Prevent the entry of personnel

into process cells during normal

operations

Engineered and administrative controls, including radiation work permits, signs and postings, and barricades to

restrict access.

9.1.2.3

Ensure that workers do not

receive a dose in excess of

limits while performing

maintenance

Same as above and procedures which implement 10 CFR 20.1602 controls for very high radiation areas 9.1.2.3

Process Cell

Exhaust System

Effectively filter process cell

exhaust

Fire protection and filtration design bases as per Table 11.4 of this FSER 11.4.1.3

Operation to ensure that a

negative pressure exists

between the process cell areas

and the C2 areas

Process cell pressure maintained subatmospheric during normal and transient operation. Designed to maintain

safety function assuming single component failure.

11.4.1.3

Process Cell

Fire Prevention

Features

Ensures that fires in process

cells are highly unlikely

Combustible loading fire controls per NFPA 801

Ignition source controls

7.1.5.3
Table 5.1-3 PSSCs and Design Basis Functions and Values Developed from the Safety Assessment

PSSC Design Basis Safety Function Design Basis Values FSER

Section

5-97

Process Safety

Control

Subsystem

Prevent the formation of an

explosive mixture of hydrogen

within the MFFF associated

with the use of the hydrogenargon gas.

I & C aspects same as emergency control system

25% of the LFL

11.6.1.3.1

8.1.2.5.1.1

Ensure isolation of sintering

furnace humidifier waterflow on

high-water level.

I & C aspects same as emergency control system

 No overflow or liquid water into furnace

11.6.1.3.1

11.3.1.2.4

Ensure the temperature of

solutions containing HAN is

limited to temperatures within

safety limits

I & C aspects same as emergency control system

Control flow of HAN/hydrogen solution such that chemical consequence limits are not exceeded in the

oxidation column

Maintain temperature of HAN within safety limits

11.6.1.3.1

11.3.1.2.4

8.1.2.5.3.1

Ensure the temperature of

solutions containing organics is

limited to temperatures within

safety limits

I & C aspects same as emergency control system

Temperature limited to less than 135 °C (275 °F)

11.6.1.3.1

 8.1.2.5.5 

Limit the residence time of

organics in process vessels

containing oxidizing agents and

potentially exposed to high

temperatures and in radiation

fields

I & C aspects same as emergency control system

Residence time of concern is on the order of years. Actual limit will be developed in the ISA. 

11.6.1.3.1

8.1.2.5.5

Ensure the temperature of

solutions potentially containing

hydrazoic acid is limited to

prevent an explosive

concentration of hydrazoic acid

from developing

I & C aspects same as emergency control system

Temperature not exceeding 60 °C (140 °F)

11.6.1.3.1

 8.1.2.5.3

Limit and control conditions

under which dryout can occur

I & C aspects same as emergency control system

Tanks potentially containing azides not left dry

11.6.1.3.1

 8.1.2.5.3
Table 5.1-3 PSSCs and Design Basis Functions and Values Developed from the Safety Assessment

PSSC Design Basis Safety Function Design Basis Values FSER

Section

5-98

Process Safety

Control

Subsystem

(cont.)

Ensure the temperature of

solutions potentially containing

metal azides is insufficient to

overcome the activation energy

needed to initiate the energetic

decomposition of the azide

I & C aspects same as emergency control system

Temperature is not to exceed 140 °C (284 °F).

11.6.1.3.1

8.1.2.5.3.3

Ensure the normality of the

nitric acid is sufficiently high to

ensure that the offgas is not

flammable and to limit

excessive hydrogen production

I & C aspects same as emergency control system

Limit hydrogen concentration to under 25% of LFL; nitric acid normality not specified

Electrical design basis not identified

11.2.1.5

Warn operators of glovebox

pressure discrepancies before

exceeding differential pressure

limits

I & C aspects same as emergency control system

Redundant pressure sensors monitor differential pressure with respect to the process room and alert the

operators to upset conditions. The instruments remain operational following facility fires in unaffected areas,

tornadoes, and design basis earthquakes.

11.6.13.1

11.4.1.3

Shut down process equipment

before exceeding temperature

safety limits

I & C aspects same as emergency control system

Temperature safety limits specified in CAR

11.6.1.3.1

8.1.2.5.2.3.1

Ensure the temperature of

solutions containing solvents is

limited to temperatures within

safety limits

I & C aspects same as emergency control system

Temperature limits will control maximum vapor concentrations

11.6.1.3.1

8.1.2.5.2

Monitor the electrolyzer for

electrical faults that could result

in arcing or other imparting of

electrical energy with the risk of

titanium fire

Detection of 10 ma 11.2.1.3.4

Ensure the flow rate of nitrogen

dioxide/dinitrogen tetroxide is

limited to the oxidation column

of the purification cycle

I & C aspects same as emergency control system

Concentration of NOx at site worker location < TEEL-2

11.6.1.3.1

8.1.2.4.1
Table 5.1-3 PSSCs and Design Basis Functions and Values Developed from the Safety Assessment

PSSC Design Basis Safety Function Design Basis Values FSER

Section

5-99

Seismic

Monitoring

System and

Associated

Seismic

Isolation valves

Prevent fire and criticality as a

result of an uncontrolled release

of hazardous material and

water within the BMF building in

the event of an earthquake

Seismic monitoring and trip system are same as emergency control system 11.6.1.3

Sintered Silicon

Nitride Barrier

Physically separate the cathode

from the anode in the nitric acid

solution by serving as a

dielectric barrier in the

electrolyzer

Nonconductive, high dielectric strength, sufficient porosity to allow ionic current transfer but insufficient porosity

to allow mixing of the anode and cathode solutions

11.2.1.3.4

Sintering

Furnace

Provide a primary confinement

boundary against leaks into the

C3 areas

Seals designed for peak temperature of 316 °C (601 °F).

Furnace shell and airlocks designed to withstand an overpressure of 2,552 g/cm2 (36.3 psi).

Leak tightness is 5 × 10-5 leaked vol/hr at 155 g/cm2 (2.2 psi).

Controls prevent overpressure.

Furnace is designed to maintain confinement function during design basis earthquake.

11.4.1.3

Sintering

Furnace

Pressure

Controls

Maintain sintering furnace

pressure within design limits

Same as process safety control subsystem 11.6.1.3

Supply Air

System

Provide unconditioned

emergency cooling air to the

storage vault and designated

electrical rooms

Provide supply air for emergency cooling

HEPA filter stages for static confinement

HEPA filter design temperature of 232 °C (450 °F) 

System design in accordance with RG 3.12

Sheet metal ductwork design, construction, and testing; “bubble tight” isolation damper construction and

testing; HEPA filter housing testing; and HEPA filter testing in accordance with ERDA 76-21

Filter testing in accordance with ASME N150 with each HEPA stage having a leakage efficiency of 99.95%

11.4.1.3

Transfer

Container

Withstand the effects of designbasis drops without breachingDesigned to withstand 9.14-m (30-ft) drop; DOE -STD-3013-2000 11.7.1.27
Table 5.1-3 PSSCs and Design Basis Functions and Values Developed from the Safety Assessment

PSSC Design Basis Safety Function Design Basis Values FSER

Section

5-100

Waste

Container

Ensure that hydrogen buildup in

excess of limits does not occur

while providing appropriate

confinement of radioactive

particles

Meet 49 CFR 178.350 requirements for certification; withstand 1-m (3.3-ft) drop 11.7.1.3

Waste Transfer

Line

Ensure that the waste transfer

line is protected from activities

taking place outside of the BMF

Double walled SS piping with leak detection designed to DBE (0.2g hor, 0.13g vertical)

RG 3.10, ANSI N13.10-1974, ANSI N317-1980, ASME B31.3

11.8.1.3

Prevent damage to the line from

external events

Same as above 11.8.1.3
6-1

6. NUCLEAR CRITICALITY SAFETY

6.1 Conduct of Review

This chapter of the final safety evaluation report (FSER) contains the staff’s review of the

nuclear criticality safety (NCS) analysis performed by the applicant, Duke Cogema Stone &

Webster (DCS), in Chapter 6 of the revised construction authorization request (CAR) (DCS,

2002d). The objective of this review is to (1) ensure that special nuclear material (SNM) storage

and processing remains subcritical under normal and credible abnormal conditions during all

operations, transfers, and storage at the mixed oxide (MOX) fuel fabrication facility (MFFF or the

facility) and (2) determine whether the principal structures, systems, and components (PSSCs)

and their design bases identified by the applicant provide reasonable protection against natural

phenomena and the consequences of potential accidents. The staff evaluated the information

provided by the applicant for NCS by reviewing Chapter 6 of the revised CAR, other sections of

the revised CAR, and supplementary information provided by the applicant. The review of NCS

design bases and strategies was closely coordinated with the review of chemical process safety,

the safety assessment of the design bases (see Chapter 5 of this FSER), and the review of

other plant systems. 

The staff reviewed how the NCS information in the revised CAR addresses or relates to the

following regulations, which are the top-level criticality safety requirements to be factored into the

technical practices used in designing the facility:

• Title 10, Section 70.23(b), of the Code of Federal Regulations (10 CFR 70.23(b)) states,

as a prerequisite to construction approval, that the design bases of the PSSCs and the

quality assurance program be found to provide reasonable assurance of protection

against natural phenomena and the consequences of potential accidents.

• 10 CFR 70.24, “Criticality Accident Requirements,” requires licensees authorized to

possess specified quantities of SNM to have a criticality accident alarm system (CAAS).

• 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, including use of an approved margin of subcriticality for safety. Preventive

controls and measures must be the primary means of protection against nuclear criticality

accidents.

• 10 CFR 70.64, “Requirements for New Facilities or New Processes at Existing Facilities,”

requires that baseline design criteria (BDC) and defense-in-depth practices be

incorporated into the design of new facilities. With respect to NCS, the design of new

facilities must provide for criticality control, including adherence to the double

contingency principle (DCP) pursuant to 10 CFR 70.64(a)(9).

The review for this construction approval focused on the design bases of the aqueous polishing

(AP) and MOX process (MP) systems, their components, and other related information. The

review also encompassed proposed design basis considerations, such as redundancy and

independence. The staff used Chapter 6.0 in NUREG-1718, “Standard Review Plan for the

Review of an Application for Mixed Oxide (MOX) Fuel,” issued August 2000 (NRC, 2000), as
6-2

guidance in performing the review. Sections 6.3.1 to 6.3.3 of NUREG-1718 discuss NCS

organization and administration, management measures, and technical practices.

6.1.1 Organization and Administration

Sections 6.3.1 and 6.5.1 of NUREG-1718 state that the staff should review the applicant’s

commitment to establish organization and administration methods to determine whether the

applicant has identified the responsibilities and authorities for organizations and individuals

implementing the NCS program. For construction authorization, the design bases include a

description of the roles and responsibilities of the NCS function related to the design of the

facility, as well as NCS staff education and experience levels. Certain aspects of the NCS

program would only be relied on during operation of the facility, such as audits and

assessments, event investigations, and emergency response, while other aspects are relied on

during design. However, because specific items relied on for safety (IROFS) need not be

identified at the construction authorization stage, significant reliance is placed on the NCS

program to provide reasonable assurance that adequate controls will be established to protect

against an accidental criticality, in the event that the facility is licensed to operate. 

Revised CAR Section 6.1.1 (DCS, 2002d) contains information relating to the organization and

administration of the NCS program during the design phase of the proposed facility, including a

description of its relevant roles and responsibilities. The roles and responsibilities during design

basically consist of establishing design criteria, supporting the integrated safety analysis (ISA)

process, validating calculational methodologies, performing analyses, and designing criticality

control systems. Section 6.4.3.1 of NUREG-1718 states that the NCS function should be

independent of operations to the extent practical. However, during the construction phase, the

staff considers it appropriate that the NCS function be part of the design team to ensure that

criticality safety features are designed into the proposed facility.

Revised CAR Section 6.1.2 only discusses the NCS program during the operations phase, not

the construction phase. The staff considered only commitments relating to the NCS program

during the construction phase to be part of the design bases of the facility and therefore did not

review the material in CAR Section 6.1.2.

Commitments to specific American National Standards Institute/American Nuclear Society

(ANSI/ANS) NCS standards (ANSI/ANS-8.1-1983 (R1988), “Nuclear Criticality Safety in

Operations with Fissionable Materials Outside Reactors,” revised 1988 (ANSI, 1988), and

ANSI/ANS-8.19-1996, “Administrative Practices for Nuclear Criticality Safety,” issued 1996

(ANSI, 1996)) are discussed in FSER Section 6.1.4. 

The applicant provided qualifications for key NCS positions during design (DCS, 2002d, 2003c).

The staff evaluated the proposed qualifications (education and experience levels) for the NCS

function manager, senior NCS engineer, and NCS engineer. The staff concluded that the

education levels for these positions are sufficient to provide the requisite skills and knowledge to

perform their technical duties and equivalent to the education level of many of the practicing

specialists in the NCS community.

Much of criticality safety is practiced by “skill-of-the-craft” and requires an intuitive understanding

of the neutron physics and margins of safety for the materials and systems being evaluated.
6-3

Moreover, comparable industry experience is likely to consist of experience at low- and highenriched uranium (HEU) fuel fabrication and enrichment plants. The staff also recommended

that plutonium- or MOX-specific experience be sought for at least senior NCS positions. The

applicant provided a summary of its criticality safety experience base involved in the design work

(DCS, 2003b). This included significant participation by COGEMA, Inc., and SGN (a subsidiary

of COGEMA) staff with over 20 years of experience with plutonium and MOX processing at

MELOX and La Hague, France, as well as staff with experience in the domestic nuclear industry. 

In addition, the applicant has committed to training the criticality staff in the appropriate

characteristics of plutonium and MOX materials. This commitment, combined with the significant

in-house experience with criticality safety associated with plutonium and MOX processing,

should ensure a sufficient knowledge base suitable for the design of the facility. Therefore, the

staff finds the qualifications and training requirements for new NCS staff, and the experience of

current NCS staff, adequate for the design of the facility.

6.1.2 Management Measures

Sections 6.3.2 and 6.5.1 of NUREG-1718 (NRC, 2000) state that the staff should review the

applicant’s commitment to establish management measures in support of the applicant’s ability

to implement and maintain the NCS program and to ensure the continued availability and

reliability of IROFS. The applicant is not required to identify specific IROFS in the revised CAR

(DCS, 2002d); therefore, the specific management measures to be applied to them cannot yet

be specified. However, the staff reviewed the quality level definitions and the associated

statements pertaining to quality levels of individual IROFS in Section 2.2, “Graded Quality

Assurance,” of the applicant’s MOX Process Quality Assurance Plan (MPQAP). As discussed in

FSER Chapter 15, the staff approved Revision 2 of the MPQAP in its safety evaluation report

issued on October 1, 2001, and approved Revision 3 of the MPQAP in its letter to DCS dated

January 10, 2003. The staff considered a description of the management measures and quality

levels to be part of the design basis of the facility. These quality-level definitions were approved

by the U.S. Nuclear Regulatory Commission (NRC) (NRC, 2001b). Questions remained,

however, concerning the examples of how these quality levels would be applied in practice (i.e.,

application to different classes of IROFS), as discussed below. 

The controlled parameters have been identified as part of the design basis of the proposed

facility for NCS, but the specific controls to be relied on have not. They will be specified later as

IROFS by the applicant and identified in the ISA summary as part of the license application (to

be submitted if the revised CAR is approved). Therefore, the means of providing for criticality

control (as required by 10 CFR 70.64(a)(9)) are described at the parameter level for the

construction review. The only specific structure, system, or component (SSC) addressed in the

revised CAR is the criticality alarm system required by 10 CFR 70.24 (discussed in FSER

Section 6.1.3.2). Because there are no specific controls identified, the application of specific

management measures (with quality levels as approved in Section 2.2 of the MPQAP) to IROFS

will be reviewed later as part of the license application. 

Specific administrative management measures applicable to NCS (e.g., training, procedures,

and audits and assessments) are not applicable to the design and construction of the facility. 

Commitments to specific NCS ANSI/ANS-8 standards (ANSI/ANS-8.1-1983 (R1988) (ANSI,

1988) and ANSI/ANS-8.19-1996 (ANSI, 1986)) are discussed in Section 6.1.4 of this FSER.
6-4

6.1.3 Technical Practices

Sections 6.3.3 and 6.5.1 of NUREG-1718 (NRC, 2000) state that the staff should review the

applicant’s commitment to designing and operating the proposed facility in accordance with NCS

technical practices to ensure that, if the facility is later authorized to operate, NCS requirements

will be met. For construction authorization, only those technical practices related to the design

of criticality safety PSSCs are applicable. Design basis information includes (1) commitments

describing the design philosophy for meeting the performance requirements of 10 CFR 70.61(d)

and BDC of 10 CFR 70.64, (2) the technical practices related to the determination of criticality

safety limits, including calculational methods and criticality code validation methods, and (3) the

technical practices related to the determination of controls, including the preferred hierarchy of

controls and measures to ensure control reliability and availability. Many of these technical

practices are contained in ANSI/ANS-8 standards; therefore, the revised CAR section on

technical practices (DCS, 2002d) contains a description of the codes and standards to which the

applicant is committing. Commitments to specific ANSI standards (ANSI/ANS-8.1-1983 (R1988)

and ANSI/ANS-8.19-1996) are discussed in Section 6.1.4 of this FSER.

6.1.3.1 Commitment to Baseline Design Criteria

The revised CAR states that the design of the proposed facility will adhere to the DCP as

required by 10 CFR 70.64(a)(9). The DCP will be used to meet the performance requirements in

an ISA summary supporting a license application. This revised CAR section also describes the

process for flowing down NCS evaluation (NCSE) requirements into the ISA as IROFS. These

commitments are part of the design basis of the facility.

Compliance with the DCP will be demonstrated in an ISA summary supporting a license

application by identifying two or more process conditions which are relied on to ensure a

subcritical configuration. The facility NCSEs will evaluate both normal and credible abnormal

conditions. Common-mode failures and system interactions will also be considered. The

NCSEs will use geometry controls (with fixed neutron absorbers, as necessary) as the preferred

controlled parameter for plant systems. To enhance the reliability of criticality controls, the

following preferred hierarchy of controls will be used:

• Passive engineered features will be preferred over active engineered features.

• Engineered features will be preferred over administrative controls.

• Enhanced administrative controls will be preferred over simple administrative controls.

Where controlled parameters rely on physical measurements, representative sampling and

analysis will be used. The sampling and analysis requirements will be identified as IROFS along

with appropriate management measures. The staff concludes that these programmatic

commitments provide reasonable assurance that an adequate safety basis will be established

and documented.

6.1.3.2 MFFF CAAS

The staff reviewed the facility CAAS described in revised CAR Section 6.3.2 based on the

requirements in 10 CFR 70.24 and guidance in Section 6.3.3 of NUREG-1718 (NRC, 2000). 

Although the applicant has not listed the CAAS as a PSSC, it stated that the CAAS will be
6-5

designed in accordance with 10 CFR 70.24, and the staff thus finds the CAAS to be part of the

design basis of the facility. Specific technical commitments regarding the CAAS are found in the

discussion of ANSI/ANS-8.3-1997, “Criticality Accident Alarm System,” issued 1997 (ANS,

1997), in FSER Section 6.1.4.3. 

The revised CAR describes the CAAS as a monitoring system composed of groups of detectors

called monitoring units (detector network, data processing, and alarm actuation units) that will

activate audible and visual alarms (network of audible and visual alarms, data processing

capability) in case of a criticality accident. The CAAS will be designed to detect both gamma

and neutron radiation and to actuate within 0.5 second of detector recognition of a criticality

accident. The range and design features of the alarm will also follow the guidance provided in

ANSI/ANS-8.3-1997. Each area in which alarms are required will be covered by two alarm units. 

The applicant committed that any areas requiring exemption from CAAS coverage would be

identified and justified in the license application or separate exemption request (DCS, 2003c).

These commitments are consistent with the requirements of 10 CFR 70.24 and ANSI/ANS-8.3-

1997 (as endorsed by Regulatory Guide (RG) 3.71, “Nuclear Criticality Safety Standards for

Fuels and Material Facilities,” issued 1972 (NRC, 1972)) and are, therefore, acceptable.

6.1.3.3 Criticality Safety Control Design Criteria

Section 6.3.3 of NUREG-1718 (NRC, 2000) includes the expectation that “the technical practices

to ensure that sufficient NCS controls, developed in the criticality safety evaluations (CSEs) and

flowed into the ISA, are identified for each process.” The staff considers the description of the

technical practices for each controlled parameter and the design criteria, including the preferred

hierarchy of controls, to be part of the design basis of the proposed facility. 

Revised CAR Section 6.3.3.1 defines passive and active engineered controls and enhanced and

simple administrative controls and describes the preferred hierarchy of controls. Tolerances will

be conservatively taken into account and neutron interaction fully evaluated. Revised CAR

Section 6.3.3.2, “Available Methods of Control,” discusses the criticality control modes and

available methods of control to be used in the proposed facility.

Revised CAR Section 6.3.3.2 also describes the requirements for geometry, mass, density,

isotopics, reflection, moderation, concentration, interaction, neutron absorber, volume,

heterogeneity, and process variable controls. Technical practices associated with these

controlled parameters were found to be acceptable, based on standard industry practice and a

comparison with the acceptance criteria of NUREG-1718. The following identifies additional

clarifications regarding specific controlled parameters:

• Revised CAR Section 6.3.3.2.1, “Geometry Control,” states that “tolerances on nominal

design dimensions are treated conservatively.” The applicant expanded on this, stating

the following:

The design approach with respect to criticality...for each controlled

parameter, assumes the credible optimal condition (i.e., most reactive

condition physically possible) for the parameter, or calculates the allowed

range for the parameter. Criticality calculations and nuclear criticality

safety evaluations are performed assuming the most reactive physical
6-6

condition to ensure that the process remains subcritical under all normal

and abnormal conditions, in accordance with 10 CFR 70.61(d). (DCS,

2001a, 2002e)

The NRC staff considers the applicant’s commitment acceptable to ensure that the most

reactive combination of tolerances will be used.

• Revised CAR Section 6.3.3.2.4 states that “isotopics control includes both the 235U/U

concentration (enrichment) and the concentration of fissile and nonfissile plutonium

isotopes...as well as the relative abundance of plutonium to uranium.” The plutonium

isotopics will be represented as 96 wt% 239Pu and 4 wt% 240Pu in criticality models. The

revised CAR also states the following:

The presence of 240Pu (5% to 9%) and 242Pu (<0.02%) offsets any

contribution from 241Pu (<1%) such that it can be neglected for 239Pu

ranges from 90% to 95%....This will be demonstrated in the criticality

calculation to be referenced in the Nuclear Criticality Safety Evaluations

(NCSEs). Justification will be provided in NCSEs and ISA summary.

The applicant also indicated, in a footnote to revised CAR Table 6-1, that the bounding

isotopics would be sufficient to bound the slight amount (

235U/Pu < 2 wt%) of HEU in the

incoming plutonium stream. To confirm that the bounding assumption (96 wt% 239Pu, 4

wt% 240Pu) is more reactive than any combination of incoming feed material, the staff

performed a series of independent calculations. These calculations were done using a

near-critical sphere of plutonium nitrate solution with full water reflection. The isotopic

abundance of 239Pu, 240Pu, and 241Pu was varied within the specification range above (the

small presence of 242Pu was neglected). The calculations show that the most reactive

combination occurs when 240Pu and 241Pu are at their maximum values (i.e., 94 wt%

239Pu, 5 wt% 240Pu, and 1 wt% 241Pu). This is still less reactive than the bounding

assumption. Further, the effect of increasing both 240Pu and 241Pu by the same amount

(1 wt%) is a net decrease in keff (|Mk/M

 240Pu| > |Mk/M 241Pu|), again indicating that the

assumed plutonium isotopics are bounding. Subsequent to completion of its

independent evaluation, the staff reviewed an additional analysis prepared by the

applicant in the supplement to the validation report, DCS 01 ZJJ DS CALC H 35075 A,

“MFFF Validation Report Supplemental Information,” dated January 31, 2005 (DCS,

2005; NRC, 2005). The NRC considers this to be part of the licensing basis in support of

the validation effort for the MFFF. In this analysis, DCS modeled an infinite cylinder

containing a plutonium dioxide-water mixture at five different H/X ratios, assuming four

fissile media with different plutonium isotopic vectors. The isotopic vectors include both

the bounding reference isotopics used in criticality calculations (96 wt% 239Pu, 4 wt%

240Pu) and three different combinations of on-spec material, including the combination

determined to be the most reactive by the NRC staff (94 wt% 239Pu, 5 wt% 240Pu, 1 wt%

241Pu). This combination contains the minimum 240Pu and maximum 241Pu content and is

expected to be conservative based on the relative values of the cross sections. The

applicant’s calculations concluded that the bounding reference case was as reactive or

more reactive than the three combinations of on-spec material across the entire

spectrum of H/X values. Therefore, the staff concludes that DCS has demonstrated that

the bounding reference case is conservative.
6-7

• Revised CAR Section 6.3.3.2.5 states that a minimum 1-in. tight-fitting water reflector

would be assumed, with the amount of reflection justified in the NCSE. In addition, the

applicant stated the following:

It is MFFF criticality calculation practice that the ranges of water reflection

up to and including 12 inches (30 cm) be employed. The calculations as

referenced by the NCSEs shall provide a demonstration of the full range of

reflection. Any exceptions to this principle, such as in the moderation

controlled, normally dry areas, shall be fully justified in the NCSEs. (DCS,

2001a, 2002e)

This approach is conservative and corresponds to common industry practice for modeling

both incidental and full reflection.

• Revised CAR Section 6.3.3.2.6, “Moderation Control,” states that, in moderationcontrolled areas, hydrogenous firefighting materials will not be allowed. The applicant’s

approach to balancing the combined risk of criticality and fire during the facility design is

part of the design basis of the facility. This requires an integrated approach to safety,

and, while this is not limited conceptually to the interaction between fire protection and

criticality safety, this has historically been one of the most significant areas of overlap. 

The commitment to avoid moderating fire suppressants is acceptable to the staff

because it removes a credible source of water from plant areas relying on moderation

control for criticality safety.

• Revised CAR Section 6.3.3.2.9 contains commitments regarding the use of fixed neutron

absorbers. The applicant did not consider other types of neutron absorbers as inherently

reliable as fixed neutron absorbers and did not envision them to be employed in any

facility process unit or area (see also FSER Section 6.1.4 on ANSI standards). 

Moreover, the applicant stated that, wherever neutron absorber control will be used, it will

be part of the geometry and fixed by design. Thus, this excludes the use of removable

(raschig rings) and soluble absorbers. Because fixed neutron absorbers are inherently

more reliable than removable or soluble absorbers, the NRC staff finds this acceptable.

The staff considers the information in this section of the revised CAR to be in broad agreement

with the regulatory acceptance criteria in the standard review plan and finds, in accordance with

10 CFR 70.23(b), that it provides an acceptable basis for design.

6.1.3.4 Criticality Safety Process Description

Revised CAR Section 6.3.4 (DCS, 2002d) contains a description of the applicable safety

principles to be applied to the design, as well as a preliminary description of how the controlled

parameters will be applied to the proposed facility. The staff reviewed this material, as well as

the process descriptions in revised CAR Sections 11.2, “MOX Process Description,” and 11.3,

“Aqueous Polishing Process Description,” along with the tables of associated criticality control

units (CCUs), revised CAR Tables 6-1 and 6-2. Revised CAR Chapter 11 provides an overview

of both the MP and AP process. Revised CAR Tables 6-1 and 6-2 list each of the 55 AP and 25

MP CCUs and their controlled parameters.
3

 While this commitment does not preclude the use of administrative controls or control modes other than geometry,

these are to be preferred over other types of controls; the process descriptions in revised CAR Sections 11.2 and

11.3, and the revised CAR Tables 6-1 and 6-2, were reviewed to confirm that the facility will be designed in

accordance with this design philosophy to the greatest extent practical (e.g., passive geometry control used on a

majority of processes). 

4Pu(NO3

)4 is the most common nitrate, but the applicant stated that it will assume Pu(NO3)3 in the analysis because it

is more reactive.

6-8

In reviewing the controlled parameters for the MP and AP process, the staff reviewed the

applicant’s use of the preferred design approach. Revised CAR Section 6.3.3.1, “Criticality

Control Modes,” commits to the preferred use of engineered over administrative controls and the

preferred use of geometry control, where practicable. Revised CAR Section 6.3.4.2, “Applicable

Safety Principles,” commits to the preferred use of passive engineered over active engineered

controls, engineered over administrative controls, and enhanced over simple administrative

controls, as well as dual- over single-parameter controls.3 Reliance on diverse control modes is

important to minimize the potential for common-mode failure. The applicant’s commitment to the

DCP ensures that, regardless of the parameter or type of control, at least two independent and

unlikely process changes must occur before criticality is possible. However, accident sequences

in which there are no credible means to achieve criticality (as may occur in certain cases

involving geometry control) do not require dual independent controls to comply with the DCP.

Although specific controls (i.e., IROFS) have not been identified in the revised CAR, the design

bases include the dominant controlled parameters. Specific values for these parameters were

not considered necessary in most cases, although some assumed values were provided (e.g.,

powder density and isotopics) when they represented bounding constraints on the design. This

information is presented in revised CAR Tables 6-1 and 6-2. The staff’s review in the following

sections is based on revised CAR Tables 6-1 and 6-2 (DCS, 2004d).

6.1.3.4.1 NCS—AP Process

The AP process, as described in revised CAR Table 6-1 and revised CAR Section 11.3.2 (DCS,

2004d), consists of 16 major areas that have been subdivided into 55 CCUs. The plutonium

dioxide powder is received from off site in cans and then batched to the dissolution unit and/or

dissolution and dechlorination unit, where it is electrolytically dissolved into plutonium nitrate

(primarily Pu(NO3)4) solution. Following this, it is purified in pulsed solvent extraction, scrubbing,

and stripping columns. During these stages, the plutonium nitrate chemical form and process

equipment favorable geometry are the dominant controlled parameters. Following purification,

the plutonium nitrate is precipitated and calcined to form a purified plutonium dioxide powder

used as feed for the MP, all still in favorable geometry process equipment. Unprecipitated

plutonium dioxide is filtered and then recovered in the oxalic mother liquor recovery unit. 

Additional process units recover solvent, nitric acid, offgases, and liquid waste and recycle

raffinate produced in the solvent extraction. These latter units primarily rely on dual independent

concentration controls because they are designed not to contain plutonium under normal

process conditions.

For each CCU in revised CAR Table 6-1, the dominant controlled parameters are listed along

with some parameter ranges. The chemical form through most of this process is primarily

aqueous plutonium oxide, plutonium nitrate solution,4 or plutonium oxalate. Plutonium oxide is
6-9

considered more reactive than plutonium nitrate. (The applicant also determined that plutonium

oxalate is bounded by plutonium oxyfluoride.) The solution is converted to plutonium dioxide

during calcination. For those areas in which plutonium nitrate is the chemical form, the applicant

identified physicochemical form as a criticality control mode. For units relying on chemical form,

the applicant has committed to take into account both nominal conditions and possible process

upsets in accordance with the DCP. The bounding nature of these compounds will be

demonstrated in NCSEs. 

According to revised CAR Table 6-1, most of the process is conducted in favorable geometry

equipment, with the general geometrical shape but not the dimensions specified. Typically, units

in the AP process will consist of cylindrical columns or slab/annular tanks. Dimensions will be

based on validated calculational methods, standards (such as ANSI/ANS-8.1-1983 (R1988)

(ANSI, 1988)), or handbook data. This will ensure that favorable geometry units will be

subcritical under normal and all credible abnormal conditions. When using methods that

determine favorable dimensions for a specific material composition (e.g., chemical form,

isotopics), the commitment to the DCP provides assurance that process upsets capable of

changing the form of the material will be considered.

Revised CAR Table 6-1 also assumes specific powder densities and isotopic compositions. The

AP process downstream of the plutonium dioxide filter assumes a chemical form of Pu(NO3)3,

which is less reactive than plutonium dioxide. Dual independent filters preclude the presence of

plutonium dioxide beyond this point. For powder density, different bounding values are

presented. In some CCUs within the decanning and milling units, powder density is controlled

by upstream measurement by a variety of means (revised CAR Sections 11.3.2.1.2 and

11.3.2.2.2). For these units, density must be confirmed to be less than 7 g/cm3. This is

considered very high for plutonium dioxide because of its highly porous nature. 

Downstream of the calcination furnace in the oxalic precipitation and oxidation unit, density is

limited to less than 3.5 g/cm3. These values were derived based on La Hague experience and

will be confirmed during startup testing. The applicant stated that, wherever the physical form of

the material changes, density will be confirmed by direct measurement or otherwise justified

(DCS, 2003b). Based on domestic nuclear industry experience, the staff accepts that a density

of 7 g/cm3 is very high for plutonium dioxide or MOX powder and that the described

measurement techniques are reasonable to confirm this. For lower density values, appropriate

controls on density, and on any process variables that can affect powder density

nonconservatively, will be specified and justified in the NCSEs. The staff will review the

justification in the NCSEs to be submitted as part of any later DCS application for a license to

possess and use licensed material.

In several CCUs, the applicant identified neutron absorption as a means of control. The staff

reviewed the application of neutron absorption and concludes that it is appropriate. For several

cases, neutron absorption is combined with geometry control (i.e., the presence of fixed

absorbers is assumed in determining subcritical geometric limits). The use of fixed absorbers is

recognized in industry practice as a highly reliable means of control, especially in conjunction

with geometry control, and thus complies with the preferred hierarchy of controls and is

acceptable.
6-10

The applicant proposed concentration control for those units not relying on favorable geometry,

where fissile material is not expected under normal conditions. Revised CAR Table 6-1

indicates only that concentration is controlled upstream of these processes. However, a

previous communication from the applicant had stated that this would be through inline monitors

and/or sampling measurements (DCS, 2001a, RAI 84; DCS, 2002e). Revised CAR Section

6.3.3.2.7 contains the requirements for both sampling and instrumentation, including the use of

dual independent sampling methods. This method for preventing plutonium intrusion into these

units is in accordance with established industry practice and the DCP and, therefore, is

acceptable to the staff.

Revised CAR Table 6-1 also describes the design bases for auxiliary systems connected to the

main AP process. The offgas treatment unit removes plutonium from the gaseous effluents in

the AP process. The liquid waste reception unit treats liquid effluents, including high-alpha

waste. High-enriched uranium is present in the incoming plutonium stream (assumed to be 100

wt% 235U). This is isotopically diluted with depleted uranyl nitrate in two steps, first to less than

or equal to 30 wt% 235U (evaluated at 35 wt% 235U), and then to less than or equal to 1 wt% 235U,

before being separated from the plutonium in the purification unit. Control of uranium assay is

required because the high-alpha waste assay must be less than 2 wt% 235U for criticality safety. 

The applicant identified uranium isotopics as a control mode for the downstream processes in

revised CAR Table 6-1; the staff, therefore, finds this acceptable. 

For other auxiliary systems for which plutonium intrusion must be precluded (such as chemical

and water addition), the applicant stated that a similar approach will be used (DCS, 2003b). The

control methods will include passive design features and dual, independent sampling. These

were not included in the revised CAR tables as they are not process units but will be addressed

in an auxiliary system NCSE. This conforms to common industry practice and is, therefore,

acceptable to the staff. (This will apply to auxiliary systems supporting both the MP and AP

process.)

The staff reviewed the application of the preferred design approach (i.e., the preferred use of

passive over active and engineered over administrative controls, and the preferred reliance on

favorable geometry) in the AP process and concluded that it was appropriate because of the

observed predominance of passive engineered controls. In addition, two or more parameters

were identified for almost every CCU. This diversity in the control modes is desirable because it

reduces the potential for common-mode failure. Given the higher inherent risk of the AP

process relative to the MP, because of the forms and types of materials being processed, the

favorable geometry design of the facility is appropriate. Most of the process relies on favorable

geometry, with the exception of the concentration-controlled units. Moreover, fixed neutron

absorbers are used in the process as part of the design. The identified parameters provide

reasonable assurance that the design of the facility will be based primarily on passive design

features, in accordance with the preferred control hierarchy. The only notable exceptions to this

are the density and isotopic composition, which have bounding values defined that are not

passively controlled (the basis for bounding densities will be provided with any license

application). Coupled with the commitment to follow the preferred design approach during the

design of actual IROFS, revised CAR Table 6-1 provides reasonable assurance that the design

of the AP process will be in agreement with the regulatory acceptance criteria. 
5There are some inconsistencies between revised CAR Table 6-2 and revised CAR Section 11.2.2 (e.g., primary

dosing and plutonium dioxide container handling are in the receiving area according to revised CAR Table 6-2, but in

the powder area according to revised CAR Section 11.2.2). This is not significant because all plant areas are

covered in both sections. The areas defined in revised CAR Section 11.2.2 are used in this discussion.

6-11

6.1.3.4.2 NCS—MP

The MP, as described in revised CAR Table 6-2 and revised CAR Section 11.2.2 (DCS, 2004d),

consists of five major areas that have been subdivided into 25 CCUs. In the receiving area,

depleted uranium dioxide powder, as well as purified plutonium dioxide powder (in 3013

containers) is received and stored before being used as feed in the powder area. Depleted

uranium poses no criticality concerns. The plutonium dioxide powder is stored in closed 3013

containers.5 Geometry and spacing in the plutonium dioxide 3013 storage pit and buffer storage

unit are controlled, along with the moderation inside containers. In the powder area, the

plutonium dioxide containers are emptied inside a glovebox onto a conveyor supplying the

dosing unit, where the depleted uranium dioxide and the plutonium dioxide powder is blended,

homogenized, and milled to form the master blend. The relative proportion of uranium dioxide

and plutonium dioxide is carefully controlled for product specification and criticality purposes. 

The master blend has a composition of less than or equal to 22 wt% plutonium; the final blend

consists of less than or equal to 6.3 wt% plutonium. These values are the conservative modeled

values for maximum plutonium content that are used in the criticality safety analyses. During

this process, mass, isotopic composition, and moderation are the dominant controlled

parameters. The powder process ends with the addition of a controlled amount of poreformer

and pelletizing of the powder into green pellets. The remaining process areas (pellet process

area, fuel rod process area, and assembly area) consist of sintering and handling the green

pellets, loading the pellets into rods, and loading the rods into assemblies, in a fashion similar to

other fuel manufacturing processes. During these latter processes, the chemical and geometric

form, isotopic composition, and moderation are the dominant controlled parameters.

For each CCU in revised CAR Table 6-2, the dominant controlled parameters are listed along

with parameter ranges for various parameters. The chemical form throughout this process is

either purified plutonium dioxide or MOX. Powder in the receiving area is handled in fixed

geometry 3013 containers and stored in fixed storage arrays. Once the powder is removed from

the containers, criticality safety is based on mass and moderation control in lieu of geometry

control. Favorable geometry is also a dominant control in the processing and handling of pellets,

rods, and fuel assemblies once the fuel has been processed into a fixed configuration.

Following primary dosing, the isotopic composition of the MOX powder is controlled for criticality

safety. The primary blend consists of MOX with less than or equal to 22 wt% plutonium and is

stored in J60 jars; the final blend consists of MOX with less than or equal to 6.3 wt% plutonium

and is stored in J80 jars. These are controlled by geometry; in addition, the MOX operations are

batch processes that rely largely on mass control for criticality safety.

Moderation is relied on for criticality safety in much of the MP, including primarily the powder

handling and storage operations. The moderator is limited to less than or equal to 1 wt% water

for incoming plutonium dioxide powder, is limited by the controlled addition of organic additives

following the primary dosing CCU, and is limited further downstream by controlling the amount of

poreformer added. 
6-12

Heterogeneity is controlled by ensuring a homogeneous mixture of uranium dioxide and

plutonium dioxide powder to prevent the formation of high-plutonium hot spots in the fuel. 

Revised CAR Table 6-2 also assumes specific powder densities and isotopic compositions. For

powder densities, different bounding values are presented. Powder density is controlled by

upstream measurement. 

There are several other units, however, where the density of powder is limited to less than

3.5 g/cm3 or 5.5 g/cm3, or the density of scrap to less than 4.6 g/cm3. These values were

derived based on MELOX experience and will be confirmed during startup testing. Following

pelletizing, full theoretical density (11.46 g/cm3) is assumed for NCS purposes. The applicant

stated that, wherever the physical form of the material changes, density will be confirmed by

direct measurement or otherwise justified (DCS, 2003b). Appropriate controls on density, and

on any process variables that can affect powder density nonconservatively, will be specified and

justified in the NCSEs. The staff will review the justification in the NCSEs to be submitted as

part of any later DCS application for a license to possess and use licensed material.

The staff reviewed the application of the preferred design approach (i.e., the preferred use of

passive over active and engineered over administrative controls, and the preferred reliance on

favorable geometry) in the MP and concluded that it was appropriate because of the observed

predominance of passive engineered controls. In addition, two or more parameters were

identified for almost every CCU. This diversity in the control modes is desirable because it

reduces the potential for common-mode failure. The majority of CCUs in the MP utilize

favorable geometry for criticality control. Where geometry control is not used, mass and

moderator control is typically used. As stated above, these operations consist of batch

processes inside gloveboxes designed to exclude moisture. The relative risk of the MP is low

compared with that of the AP process (the risk is especially low following downblending of the

plutonium dioxide into MOX powder and subsequent pelletizing). The identified parameters

provide reasonable assurance that the design of the facility will be based primarily on passive

design features, in accordance with the preferred control hierarchy. Coupled with the

commitment to follow the preferred design approach during the design of actual IROFS, the

information presented in revised CAR Table 6-2 provides reasonable assurance that the design

of the MP will be in agreement with the regulatory acceptance criteria.

Although diverse control modes were described for both the MP and AP process in revised CAR

Tables 6-1 and 6-2, the extent to which diverse control modes were used in meeting the DCP

could not be determined. This was simply because the tables contain no information on the

structure of individual accident sequences. While it appeared that dual parameter control had

been implemented, during the January 16, 2003, public meeting on criticality safety, the

applicant made the following statement:

For the MFFF, double contingency, in most cases, is based on 2 controls or

barriers to prevent a change in one controlled parameter. A loss of one of these

controls or barriers does not cause a change in the controlled parameter and

therefore does not change the keff value.

The applicant further clarified that the use of two-parameter control is not typically practicable for

MFFF operations because of the assumed 239Pu content (DCS, 2003e). As clarified in the July
6-13

31, 2003, public meeting, for systems with high 239Pu content (96 wt% 239Pu/Pu and greater than

or equal to 6.3 wt% PuO2/(UO2 + PuO2)), mass limits are sufficiently low that limiting operations

to the minimum critical mass would be infeasible. At the masses used in the process, only a

loss of moderation control is needed for criticality. In addition, upon loss of mass control,

criticality can be possible without additional moderation beyond what is required by the process.

Therefore, criticality may occur if control over either mass or moderation is lost, and two

independent controls are required on each of these parameters in order to meet the DCP. The

staff acknowledges that this may be the case in large portions of the plant but that a

determination of whether dual-parameter control is practical must be made on a case-by-case

basis. Dual-parameter control is recognized as generally more reliable than single-parameter

control because of the increased susceptibility to common-mode failures when controlling only

one parameter. However, in CAR Section 6.3.4.2, the applicant committed to the use of dualparameter control where practical, and it further committed that the determination of the

practicality of following the design principles will be made in the development of the NCSEs and

subjected to numerous technical and managerial reviews (DCS, 2001a, RAI 67). Based on

these commitments, the staff understands that the applicant will provide justification in the

NCSEs whenever single-parameter control is necessary. While this is less desirable than dualparameter control for meeting the DCP, the applicant has also stated (DCS, 2003e) that, in both

cases, controls must be demonstrated to be sufficiently robust and independent to prevent

criticality. The staff therefore finds these commitments to be acceptable for the purpose of

construction authorization. The staff will review the justification for the use of single-parameter

control and the adequacy of specific control measures to be submitted as part of any later DCS

application for a license to possess and use licensed material. 

6.1.3.4.3 Single-Parameter Limits

In revised CAR Tables 6-3 and 6-4, the applicant provided a list of the single-parameter limits for

various parameters, such as subcritical masses, dimensions, and volumes, for use in the facility

design. Revised CAR Section 6.3.4.5 and the footnote to these tables indicate that they are not

to be used in support for criticality calculations or NCSEs. These are only intended as typical,

order-of-magnitude estimates. Therefore, the staff did not evaluate the adequacy or derivation

of the values in these revised CAR tables as part of the revised CAR review.

6.1.3.5 Nuclear Criticality Analysis and Safety Evaluation Methods

The staff considers a description of the calculational methodology, including the computer code

validation methodology, to be part of the design basis of the facility. The staff also considers the

design basis to include the maximum keff or upper subcritical limit (USL). The applicant has

committed to ANSI/ANS-8.1-1983 (R1988) (ANSI, 1988) as part of the design basis of the

facility. The standard requires that processes be shown to be subcritical under both normal and

credible abnormal conditions and also contains requirements for the validation of calculational

methodologies. The staff evaluated the applicant’s submitted validation report (DCS, 2003f,

2003h), which sets forth the applicant’s proposed methodology for determining the margin of

subcriticality for safety. To approve the revised CAR, the staff must, pursuant to 10 CFR

70.61(d), approve the margin of subcriticality for safety. The applicant has committed to the

methodology described below to ensure that there will be an adequate margin of subcriticality for

safety during any operation of the proposed facility. 
6-14

The ANSI/ANS-8.1-1983 (R1988) standard provides single and multiparameter limits to be

referenced in criticality safety calculations, as well as guidance used in the performance of

criticality analysis method validation. In revised CAR Section 6.3.5, the applicant described the

criticality analysis methodology to be used in facility design activities and facility safety

programs, as also discussed in ANSI/ANS-8.1-1983 (R1988). The applicant stated that the

chosen computational methods are those using the evaluated nuclear data file (ENDF)

cross-section libraries (i.e., the Standardized Computer Analyses Licensing Evaluation-(SCALE)-

4.4 and 4.4a code system with the 238-group cross-section library derived from the ENDF/B-V

file of evaluated nuclear physics data and the Monte Carlo Neutron Photon code, with ENDF60

cross-section libraries). Use of these codes and cross section libraries is accepted industry

practice and has been recognized by the NRC as generally acceptable (NUREG/CR-6361,

“Criticality Benchmark Guide for Light-Water-Reactor Fuel in Transportation and Storage

Packages”).

In the revised CAR, the applicant discussed the method of validation and calculated keff design

limits to be used in the facility. Consistent with ANSI/ANS-8.1-1983 (R1988), the applicant’s

code validation methodology uses critical experiment benchmarks to determine bias and bias

uncertainties. Critical experiments are selected to represent the neutronic and physical

characteristics of the systems analyzed in specific design applications. The derived biases are

then applied to calculations on systems with neutronic and physical characteristics that are

bounded by those of the selected set of experimental benchmarks. When the neutronic and

physical characteristics of the system analyzed are not bounded by those of the experimental

benchmarks, additional adjustments for bias and uncertainties will be employed as appropriate,

by making use of trends in the bias. The applicant stated that criticality safety calculations would

be performed for each of the stages of process operations at the facility (ranging from receipt of

plutonium dioxide and uranium dioxide powders through fabricated MOX fuel assembly storage

and shipment). 

The validation approach to be used by the applicant is similar to those already approved by the

NRC in previous license application approvals for a broad range of applications. This validation

approach is described in NUREG/CR-6361, NUREG/CR-6689, “Guide for Validation of Nuclear

Criticality Safety Calculational Methodology,” and NUREG/CR-6655, “Sensitivity and Uncertainty

Analyses Applied to Criticality Safety Validation.” The International Handbook of Evaluated

Criticality Safety Benchmark Experiments of the International Criticality Safety Benchmark

Evaluation Project provides a large compilation of benchmark criticality experiment descriptions. 

The staff reviewed the information contained in revised CAR Section 6.3.5, “Nuclear Criticality

Analysis and Safety Evaluation Methods,” and the code validation report (DCS, 2003f) submitted

separately to determine that an acceptable subcritical margin will be provided for the design of

the facility. The following sections describe the validation methodology and the application of

that methodology to determine design basis keff limits for each of five areas of applicability

(AOAs).

6.1.3.5.1 NCS Validation Report

Pursuant to 10 CFR 70.64(a)(9), the design of the proposed MFFF must provide for criticality

control, including adherence to the DCP. The ANSI/ANS-8.1 standard (ANSI, 1988) requires

both adherence to the DCP and that processes be subcritical under both normal and credible
6-15

abnormal conditions. The applicant has chosen to commit to this standard to meet, in part,

10 CFR 70.64 and has identified it as a part of the design basis of the MFFF. Determining the

bias, uncertainty in the bias, and an administrative margin are all necessary tasks in setting a

USL that ensures an adequate margin of subcriticality. For the purpose of showing compliance

with the USL, ks + )ks # USL, where ks is the calculated keff for the specific application. As the

applicant stated (DCS, 2003g), for criticality calculations using the SCALE code, )ks = 2Fk,

where Fk is the Monte Carlo standard deviation.

The applicant used the statistical methods in NUREG/CR-6361, which describes the two

methods for determining the USL, including (1) confidence band with administrative margin and

(2) lower tolerance band, also known as the single-sided uniform width closed interval approach. 

These approaches are summarized in the validation report.

Method 1 requires the choice of an administrative margin, which the applicant set to 0.05. 

Method 2 requires the choice of the confidence " that some fraction P of all future calculations

below the USL are subcritical. The applicant used values of " = 0.95 and P = 0.999 throughout

the validation. These values are conservative with respect to the guidance in NUREG/CR-6698,

“Guide for Validation of Nuclear Criticality Safety Calculation Methodology,” and therefore

acceptable to the staff. The choice of administrative margin is discussed in FSER Section

6.1.3.5.2. The linear regression and statistical analysis required by these methods is performed

by the USLSTATS code distributed with the SCALE code package.

An important caution is that the aforementioned techniques can only be used on data that are

normally distributed. To verify normality, the applicant’s validation methodology uses the P

2

 (chisquared) test for normality included with the USLSTATS code. The guidance in NUREG/CR6361 advises that normality can usually only be assured when there are more than 25 data

points. In those cases when data are not normally distributed, the applicant committed to using

nonparametric methods, as discussed in NUREG/CR-6698. This requires basing the USL on

the lowest observed value, with additional margin based on the number of benchmark

experiments.

The aforementioned methodology is acceptable to the staff, as it entails techniques that have

been endorsed in various NUREG publications and have been used to validate criticality codes

at fuel facilities and across the domestic nuclear industry.

The applicant submitted its validation report in three separate parts by letter dated January 8,

2003 (DCS, 2003a). The latest revision to Parts I and III was submitted by letter dated July 2,

2003 (DCS, 2003f). The latest revision to Part II was submitted by letter dated October 10, 2003

(DCS, 2003h). Each part of the validation report addresses one or more separate AOAs. Part I

includes AOA(1), covering plutonium nitrate solutions, and AOA(2), covering MOX fuel pellets,

rods, and assemblies. Part II includes AOA(3), covering plutonium dioxide powder systems, and

AOA(4), covering MOX powder systems. Part III includes AOA(5), covering solutions of

plutonium compounds, such as plutonium oxalate and plutonium oxyfluoride. For each AOA, the

applicant used the validation methodology described in this section of the FSER.

In a previous revision to Part II of the validation report, the applicant made use of the

sensitivity/uncertainty (S/U) methodology developed recently at Oak Ridge National Laboratory

(ORNL). This method provides a more quantitative basis for choosing benchmark experiments
6-16

than the traditional qualitative approach, by comparing dependencies on the underlying cross

section data. The sensitivity portion of the code calculates the sensitivity of keff to changes in the

underlying cross section data for a particular system. The uncertainty portion of the code

determines the degree of correlation between candidate benchmarks and hypothetical design

applications by combining the sensitivity information with known uncertainties in the cross

section data. Following the staff’s questions on the use of the S/U method, the applicant

provided Revision 3 to Part II of the validation report by letter dated October 10, 2003 (DCS,

2003h). In Revision 3 to Part II, the applicant used a more traditional approach, relying on

parametric screening criteria rather than on the S/U approach. However, the staff did use the

TSUNAMI modules (TSUNAMI-1D, TSUNAMI-3D, and TSUNAMI-IP) of the SCALE 5 code

package to test the physics assumptions used in each part of the validation report and to obtain

additional information on the applicability of chosen benchmarks. As a result of these

independent TSUNAMI calculations, the NRC staff was able to conclude that the applicant’s

selected benchmarks and statistical analyses were appropriate, with the exception of certain

benchmarks in AOA(4). At the time the NRC staff performed these analyses, ORNL had not yet

released the SCALE 5 code for general use. While SCALE 5 was made available to the NRC, it

was not made available to the applicant; therefore, the applicant was unable to perform the type

of S/U analysis relied on by the NRC staff to reach these conclusions. 

Because of the developing nature of the TSUNAMI methodology and the staff’s questions on its

use, the staff does not consider it appropriate to base acceptability solely on the application of

the S/U method.

Except for remaining questions on the AOA(4) benchmarks used in Revision 3 to Part II, the

NRC staff determined that the code validation methodology described generally in CAR Section

6.3.5 was appropriate. The validation of specific AOAs intended to cover anticipated AP and MP

operations is discussed in the following sections. For the purpose of construction authorization

review, only the results of the validation (i.e., determination of the benchmark experiments and

parametric range for each AOA, and computation of the bias and uncertainty) were reviewed. 

The staff expects that the applicant will demonstrate the applicability of the validation results to

specific design applications in NCSEs and will ensure that the design calculations use the code

in a manner consistent with that used in the validation (e.g., similar statistics and use of code

options, such as albedo and biasing). This will be reviewed as part of the staff’s review of any

later DCS application for a license to possess and use licensed material. 

6.1.3.5.1.1 AOA(1)—Plutonium Nitrate Solutions. To validate AOA(1), the applicant modeled

191 critical experiments using the KENO-VI code with the 238-group cross section library and

the INFHOMMMEDIUM cross section treatment option. The bias was determined for both the

SCALE-4.4 code package on the SGN Sun hardware platform and the SCALE-4.4a code

package on the SGN PC hardware platform. The applicant’s chosen experiments included

plutonium nitrate systems modeled as spheres, cylinders, and slabs with a variety of reflector

conditions (no reflector, water, concrete, and cadmium). The staff examined the chosen

benchmarks and determined that they were appropriate for a validation of systems covered by

AOA(1), as they consist of plutonium nitrate benchmarks with parameters close to the

parametric range of AOA(1).

The applicant defined the validated range of parameters for AOA(1) in Table 5-2 of Part I of the

validation report. The validated AOA is described in Table 6.1-1.
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Table 6.1-1 Definition of AOA(1)

Geometry Cylinder, slab, annular cylinders, arrays of cylinders

Reflectors Full water, cadmium*/water, borated concrete*

(1) Cd limited to 0.05-cm- (0.02-in.-) thick sheet surrounding

4.5–9.5-cm- (1.8–3.7-in.-) thick slab tanks.

(2) Borated concrete limited to 15 cm (5.9 in.) inside and outside

7–7.5-cm- (2.76–2.95-in.-) thick annulus and separated from tank by

1.8–2-cm (0.71–0.79-in.-) gap, with the composition below:

10B = 1.59×10-3 atoms/bAcm (2.61×1022 atoms/in.3

) 

11B = 7.04×10-3 atoms/bAcm (1.15×1023 atoms/in.3

)

Ca = 4.65×10-3 atoms/bAcm (7.62×1022 atoms/in.3)

Fe = 5.01×10-4 atoms/bAcm (8.21×1021 atoms/in.3)

Si = 1.66×10-4 atoms/bAcm (2.72×1021 atoms/in.3)

H = 2.17×10-2 atoms/bAcm (3.56×1023 atoms/in.3)

Al = 1.96×10-3 atoms/bAcm (3.21×1022 atoms/in.3)

O = 3.25×10-2 atoms/bAcm (5.33×1023 atoms/in.3 )

Chemical form Plutonium nitrate solution

Pu/(U + Pu) 100 wt%

240Pu 4 wt%

H/Pu 100–200

gPu/l 125–237

EALF 0.14–0.25 eV

*Subject to the limitations of this section

The applicant ran SCALE/KENO-VI for the chosen experiments, calculated keff, and plotted the

values of USL-1 and USL-2 (determined by USLSTATS) versus the trending parameters of

energy of average lethargy causing fission (EALF), the moderator to fuel ratio (H/Pu), and the

240Pu content. Over the range of applicability, the minimum USL-1 and USL-2 were calculated

as a function of these parameters and the lowest such value chosen as the USL for the entire

AOA. Because the data were found not to be normally distributed, nonparametric estimates of

the USL were also made. Based on the number of benchmarks included, no additional

nonparametric margin (NPM) was needed. The lowest value determined using USLSTATS was

more conservative than that using the NPM, so this was used as the final USL for the AOA. 

The staff compared the range of parameters covered by the benchmark data to the applicant’s

AOA and determined that there was a sufficient number of experiments and that they provided

adequate coverage of the defined AOA across the entire range, except for cases involving some

absorber materials.

For certain absorber materials, sufficient benchmarks were not available. Therefore, in

Attachments 5 and 6 to Part I of the validation report, the applicant included an analysis that

attempted to show that the presence of borated concrete and cadmium has a negligible effect on

the bias. Because the applicant did not identify any plutonium nitrate systems containing boron,

it analyzed several uranyl nitrate benchmarks with borated concrete, plaster, or water to

determine if the presence of the boron had an effect on the bias for these systems. Although

the applicant did not observe any effect on the bias, the staff considered the number of
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benchmarks containing boron insufficient to make a definitive conclusion. The applicant also

stated that the conclusions regarding the effect of these absorbers on the bias for uranyl nitrate

systems could be applied to plutonium nitrate systems. This determination was based on

similarities in the neutron absorption spectra between the uranium benchmarks and the same

cases with plutonium substituted for uranium. The applicant also analyzed several additional

plutonium nitrate cases containing cadmium to demonstrate that cadmium has a negligible effect

on the bias. These cases were not included in determining the bias for AOA(1) because the

240Pu content was outside the target range. They also did not show any statistically significant

effect of the absorber on the bias. Therefore, the applicant concluded that the existing

benchmarks were sufficient to validate calculations for plutonium nitrate systems containing

borated concrete and cadmium. 

The staff had concerns regarding the number of experiments containing boron and cadmium, as

well as using uranyl nitrate benchmarks as part of the basis for this conclusion. The staff,

therefore, performed an independent analysis to evaluate the applicant’s conclusions regarding

the use of absorbing materials in AOA(1). Perturbation of the number densities of 10B, 1H, and

239Pu show that the most highly poisoned benchmarks are relatively insensitive to the 10B cross

sections compared to the sensitivity to 1H and 239Pu (i.e., a small change in the 10B content

produces a negligible change in the overall system keff). Therefore, any bias resulting from

errors in the boron cross section is expected to be correspondingly small. Subsequent to

completion of its independent evaluation, the staff reviewed an additional analysis prepared by

the applicant in the supplement to the validation report , DCS01 ZJJ DS CALC H 35075 A,

“MFFF Validation Report Supplemental Information,” dated January 31, 2005 (DCS, 2005; NRC,

2005). The NRC considers this to be part of the licensing basis in support of the validation effort

for the MFFF. In this analysis, DCS modeled three different configurations containing

colemanite concrete as a neutron absorber. The three cases considered (two fissile solution

tanks and a drip tray) correspond to actual process calculations. The applicant varied the boron

content in the concrete to determine the keff sensitivity to boron number density and concluded

that a 50-percent reduction in the boron density would produce only an approximately 1 percent

)keff for these three cases. These cases correspond to thicknesses and arrangements of the

colemanite concrete reflector described in Table 6.1-1 above. Therefore, the staff concludes

that even a very large error in the 10B cross section will not be sufficient to significantly affect the

code bias. The staff concludes that DCS has demonstrated that cases involving borated

concrete reflectors are within the scope of AOA(1). In addition because of the high sensitivity to

the 239Pu cross sections, the staff expects that there is little relevance of the uranyl nitrate

benchmarks to the behavior of plutonium nitrate systems.

Because the experiments considered appear to show little sensitivity to the absorbers, the staff

concluded that the lack of an observed trend in the bias could result from either the poor

statistics or low reactivity worth of the absorbers in the cases considered. Because the absorber

worth can depend strongly on the geometrical configuration and size of the absorbers, system

leakage, and other physical characteristics of the system, these results may not be indicative of

the behavior of systems with much higher absorber worth. Therefore, the NRC requested the

applicant to describe and justify “the range of boron and cadmium absorber loading for which

AOA(1) is considered valid” (NRC, 2003a). The footnotes to Table 5-2 in Part I of the validation

report provide this detailed information on the dimensions and compositions of borated concrete

and cadmium to be used in anticipated design applications. The applicant confirmed that the

footnotes to Table 5-2 represent the maximum loading of these materials (DCS, 2003g). The
6-19

behavior of these design applications with regard to their sensitivity to perturbations in the

absorber number density is comparable to that in the uranyl nitrate study discussed above. In

addition, the applicant stated that, for AOA(1), no more than 5 percent of the borated concrete is

required to ensure subcriticality (maintain keff below the USL) (DCS, 2003g). This information is

useful because it indicates the amount of error that would be required to cause the keff limits to

be exceeded. While comparable information was not initially provided for cadmium, the staff’s

independent calculations indicate that it would require at least a 50-percent decrease in the

cadmium number density before a significant change in the system keff occurred, assuming a

fully reflected slab tank with a cadmium sheet as described in the footnote to Table 5-2 of Part I. 

This appears to occur because cadmium is a nearly black absorber of thermal neutrons at the

thicknesses described. The staff considers the likelihood of the 10B and cadmium cross sections

being in error by such a large proportion (95 percent for 10B and approximately 50 percent for

cadmium) to be very low, as both boron and cadmium are commonly used absorbing materials,

and errors of this magnitude would have been noticed during the long history of using these

materials in criticality calculations. Subsequent to completion of its independent evaluation, the

staff reviewed an additional analysis prepared by the applicant in the supplement to the

validation report, DCS01 ZJJ DS CALC H 35075 A, “MFFF Validation Report Supplemental

Information,” dated January 31, 2005 (DCS, 2005; NRC, 2005). The NRC considers this to be

part of the licensing basis in support of the validation effort for the MFFF. In this analysis, DCS

modeled two different configurations containing cadmium sheets as neutron absorbers. The

cases considered (slab tanks) correspond to actual process calculations. The applicant

performed two series of calculations in which it (1) maintained the cadmium sheet thickness but

varied the cadmium density and (2) maintained the cadmium density but varied the cadmium

sheet thickness. For both applications, the two methods yield nearly identical results and show

that a 60-percent decrease in the cadmium density would be required to produce an

approximately 1 percent )keff for these cases. These cases correspond to thicknesses and

arrangements of the cadmium reflector described in Table 6.1-1 of the FSER. Therefore, the

staff concludes that even a very large error in the cadmium cross section will not be sufficient to

significantly affect the code bias.

Based on the foregoing considerations, the staff has reasonable assurance that systems

meeting the criteria in Table 6.1-1 are within the scope of AOA(1) and will be safely subcritical

with a USL as defined in Part I of the validation report (USL = 0.9370).

6.1.3.5.1.2 AOA(2)—MOX Pellets, Fuel Rods, and Fuel Assemblies. To validate AOA(2), the

applicant modeled 36 critical experiments using the KENO-VI code with the 238-group cross

section library and the LATTICECELL cross section treatment option. The bias was determined

for both the SCALE-4.4 code package on the SGN Sun hardware platform and the SCALE-4.4a

code package on the SGN PC hardware platform. The applicant’s chosen experiments include

lattices of MOX fuel rods in water, with a variety of different moderator-to-fuel ratios (v

m

/vf) and

plutonium contents. The staff examined the chosen benchmarks and determined that they are

appropriate for the validation of systems covered by AOA(2), as they consist of water-moderated

MOX fuel rod lattices with parameters close to the parametric range of AOA(2).

The applicant defined the validated range of parameters for AOA(2) in Table 5-4 of Part I of the

validation report. The validated AOA is described in Table 6.1-2.
6-20

Table 6.1-2 Definition of AOA(2)

Geometry Heterogeneous, rectangular lattices

Reflectors Water

Chemical form MOX fuel

Pu/(U + Pu) 6.3 wt%

240Pu 4 wt%*

v

f

/vm 1.9–10

EALF 0.1–0.66 eV

*See discussion below

The applicant ran SCALE/KENO-VI for the chosen experiments, calculated keff, and plotted the

values of USL-1 and USL-2 (determined by USLSTATS) versus the trending parameters of

EALF, the moderator to fuel ratio (vm/vf), and the plutonium content (PuO2/(PuO2 + UO2)). Over

the range of applicability, the minimum USL-1 and USL-2 were calculated as a function of these

parameters, and the lowest such value was chosen as the USL for the entire AOA. The

benchmark data tested normal, although USLSTATS indicated that the amount of data may not

be sufficient for this normality test. However, no anomalies in the keff histogram indicating a

deviation from normality were apparent.

The staff compared the range of parameters covered by the benchmark data to the applicant’s

AOA and determined that there was a sufficient number of experiments to provide adequate

coverage of the defined AOA across the entire range, except for the 240Pu content.

The staff noted that the isotopic composition considered validated in Table 5-4 (4 wt% 240Pu)

was outside the range of the benchmark experiments (8–22 wt% 240Pu), and no trending of the

bias as a function of 240Pu content was performed for AOA(2). The applicant justified this on the

basis that LA-12683, “Forecast of Criticality Experiments and Experimental Programs Needed to

Support Nuclear Operations in the United States of America: 1994–1999,” supports a ±4 wt%

extension to the defined AOA with regard to 240Pu content. However, while this extrapolation

may be permissible, the applicant has committed to extend the bias beyond the range of

experimental data using the guidance in NUREG/CR-6698. This guidance states that

extensions to the AOA should take into account trends in the bias and any increased uncertainty

in the bias. This was not done.

Therefore, staff used the USLSTATS program to determine the effect of extrapolating the bias

and the uncertainty in the bias to systems with 4 wt% 240Pu content. This was done to evaluate

whether additional margin should have been included to account for the 240Pu content being

outside the screening range. This analysis shows that there is a very weak correlation between

the bias and the 240Pu content and no significant change in the bias as a result of extending the

AOA down to 4 wt% 240Pu. Subsequent to completion of its independent evaluation, the staff

reviewed an additional analysis prepared by the applicant in the supplement to the validation

report, DCS01 ZJJ DS CALC H 35075 A, “MFFF Validation Report Supplemental Information,”
6-21

dated January 31, 2005 (DCS, 2005; NRC, 2005). The NRC considers this to be part of the

licensing basis in support of the validation effort for the MFFF. In this analysis, DCS performed

a statistical analysis of the benchmark keff values as a function of 240Pu content, using the ORNL

code USLSTATS. Using the results from the confidence band technique generated by

USLSTATS, the applicant determined the functional form of both the calculated keff and its

uncertainty, k(x) and w(x), as a function of 240Pu content. The applicant then used this

information to extend the trend, including allowance for uncertainty, down to 4 wt% 240Pu

content, and showed that the originally determined USL is still valid. Therefore, the staff

concludes that the definition of AOA(2) may be extended down to 4 wt% 240Pu content without

the need for additional margin.

Therefore, based on the foregoing considerations, the staff has reasonable assurance that

systems meeting the criteria in the table above are within the scope of AOA(2) and will be safely

subcritical with a USL, as defined in Part I of the validation report (USL = 0.9321).

6.1.3.5.1.3 AOA(3)—Plutonium Oxide Powder. To validate AOA(3), the applicant modeled 70

critical experiments using the KENO-VI code with the 238-group cross section library. The

applicant modeled some of the benchmarks using the INFHOMMEDIUM cross section

treatment option and some using the LATTICECELL treatment option. The validation report

does not clearly state which code options (including cross section treatment) were considered

validated. The bias was determined for the SCALE-4.4a code package on the SGN PC

hardware platform. The applicant’s chosen experiments include reflected and bare plutoniummetal systems and arrays of plutonium dioxide- and MOX-polystyrene compacts. The applicant

selected this set of experiments from a much larger suite of 308 candidate benchmarks by

comparing the most important physical and neutronic parameters of the candidate benchmark

experiments to those of anticipated design applications. To do this, the applicant chose a set of

screening criteria consisting of allowable ranges of important physical parameters, as described

in NUREG/CR-6698. The staff examined the chosen benchmarks to determine whether they

were appropriate for a validation of systems covered by AOA(3). The evaluation of the chosen

benchmarks is discussed below.

The applicant defined the validated range of parameters for AOA(3) in Table 5-2 of Part II of the

validation report. The validated AOA is described in Table 6.1-3.

Table 6.1-3 Definition of AOA(3)

Geometry Parallelepipeds, arrays of cylinders, spheres

Reflectors Water

Chemical form PuO2 powder

Pu/(U + Pu) 100 wt%

240Pu 4 wt%

H/Pu 0–15

EALF 5 eV–266 keV
6-22

The applicant ran SCALE/KENO-VI for the chosen experiments, calculated keff, and plotted the

value of keff versus the trending parameters of EALF, the moderator to fuel ratio (H/Pu), and the

240Pu content. Because the data were not normally distributed, a non-parametric estimate of the

USL was made, assuming an administrative margin of 0.05. Based on the number of

benchmarks included, no additional NPM was needed. The non-parametric estimate was used

as the USL for AOA(3).

The staff compared the range of parameters covered by the benchmark data to the applicant’s

AOA and determined that the benchmarks provide adequate coverage of the defined AOA

across the entire range.

To select the benchmarks, the applicant followed the steps in Section 2.5 of NUREG/CR-6698,

which consist of (1) identifying key parameters associated with normal and upset conditions of

systems to be evaluated, (2) based on these key parameters, establishing a screening AOA for

critical experiments, (3) identifying critical experiments that are within this screening AOA, and

(4) from the scope of selected critical experiments, determining the detailed AOA covered by the

experiments. A final step, consisting of showing that design applications are within the detailed

AOA, will be conducted as part of the overall final design process (as indicated in Figure 3-1 of

Part II of the validation report) and is outside the scope of the construction validation review.

With regard to the implementation of this methodology, the staff noted that the screening criteria

used were significantly broader in some cases than those in Table 2.3 of NUREG/CR-6698. The

values in this table are the result of consensus among criticality experts and acknowledged to be

conservative. The staff questioned the ranges of parameters in the screening criteria and the

justification for extending them beyond the ranges in Table 2.3 of NUREG/CR-6698. In addition,

the parameters used in the primary screening criteria (239Pu content, 240Pu content, H/Pu, and

EALF) did not include the total plutonium content (wt% of plutonium in MOX), which is known as

one of the parameters to which keff is most sensitive. Since it could not resolve all the technical

concerns with the applicant’s implementation of the NUREG/CR-6698 methodology, the staff

examined all the chosen benchmarks on a case-by-case basis to make an independent

determination regarding their inclusion for a validation of AOA(3). There were two broad

categories of benchmarks, those involving plutonium metal and those involving plutonium oxide

powder, which needed additional justification. The staff’s review of the applicability of these

benchmarks is described below.

Plutonium-Metal Benchmarks

The staff included several sets of plutonium-metal benchmarks (PU-MET-FAST-001, -003, -016,

-017, -033, and -037) in the benchmarks for AOA(3). At the November 13, 2003, meeting, the

applicant stated that these were included to validate plutonium dioxide powder storage array

applications at the front end of the AP process (NRC, 2003b). The applicant stated that, for

some array cases, dry plutonium dioxide powder at full theoretical density (i.e., with H/Pu = 0)

very closely resembles plutonium metal and that it is the most reactive case. This is relevant

because the most reactive normal condition case primarily determines the subcritical limits for

the proposed process. During the in-office review (NRC, 2004a), the staff examined proposed

calculations for the 3013 storage and the plutonium buffer storage units and determined that

these plutonium dioxide powder storage arrays are physically very similar to the configuration of
6-23

several of the plutonium-metal benchmarks (particularly, PU-MET-FAST-003, -016, -017, and -

037). For both these plutonium dioxide storage applications and the plutonium-metal

benchmarks, the configuration consists of bare or reflected arrays of high-density plutonium

encased in steel cans, with similar isotopics, absorber materials, geometric arrangement, and

neutron fission spectra. Thus, the most important parameters characterizing these systems are

very similar to those of the chosen benchmarks.

To confirm that the dry case is the most reactive, the applicant performed a sensitivity study

showing that plutonium dioxide powder at full theoretical density and an H/Pu = 0 results in a

higher keff than plutonium dioxide powder with lower densities and a moisture content up to 1

wt% water (H/Pu approximately 0.3) (NRC, 2004a). The staff performed its own confirmatory

calculations that show that the dry case is the most reactive, provided that the moisture content

does not exceed approximately 1.25 wt% (at which point keff will approach that of the dry case).

Because normal-condition moisture content is limited to 1 wt% water content (as stated in CAR

Section 6.3.4.3.2.6), the dry case is expected to be bounding for some array calculations under

normal conditions, and thus it is appropriate to include H/Pu = 0 cases in the definition of

AOA(3).

There are two differences between the material composition of plutonium metal and plutonium

dioxide powder that can theoretically affect the bias—(1) differing material density and (2)

differing chemical composition. The staff investigated the importance of these two effects

relative to other differences between the systems. The staff performed independent calculations

with different-sized spheres containing various plutonium dioxide-water mixtures, using the

XSDRNPM discrete ordinates code. To gain further insights, the staff also performed S/U

calculations using the TSUNAMI sequences of the SCALE 5 code package. These results all

show that keff is much less sensitive to density than to other system parameters; what sensitivity

exists appears because of changes in the energy spectrum and the amount of leakage that

occurs with changing density. To confirm this, the staff performed an additional sensitivity study

of the variation of kinf as a function of water content and EALF for several different values of

density (3–10 g/cm3(0.11–0.36 lb/in.3)). The model consists of an infinite fissile medium

modeled with the KENO-Va Monte Carlo code. When kinf is plotted as a function of H/Pu or

EALF at several different densities, the curves at different densities coincide, indicating that keff

is relatively insensitive to powder density when leakage can be neglected. (The leakage is

related to the neutron energy and system geometry, which the TSUNAMI codes show could be

considered parameters of secondary importance for criticality code validation.) The S/U

calculations discussed above corroborate this, as there was a very high correlation coefficient

between the plutonium-metal benchmarks and dry plutonium dioxide powder cases with several

different densities. Subsequent to completion of its independent evaluation, the staff reviewed

an additional analysis prepared by the applicant in the supplement to the validation report,

DCS01 ZJJ DS CALC H 35075 A, “MFFF Validation Report Supplemental Information,” dated

January 31, 2005 (DCS, 2005; NRC, 2005). The NRC considers this to be part of the licensing

basis in support of the validation effort for the MFFF. In this analysis, DCS modeled 83 different

plutonium dioxide-water mixtures, ranging from 0–0.999 weight fraction of water, for 20 different

values of powder density, ranging from theoretical density down to 5 percent of theoretical

density. For each density value, the powder and water density were both scaled by the same

density factor to maintain a constant H/X ratio. The Kinf values were chosen for the comparison

because there is a strong correlation between density, neutron leakage, and energy for a finite

system. The Kinf values were used to isolate the direct effect of density on keff, in the absence of
6-24

leakage. When kinf is plotted on a three-dimensional graph as a function of both water content

and density, kinf appears to be totally independent of density. As further demonstration, the

residuals (keff–kREF)/kREF were plotted on a quantile-quantile plot and observed to be normally

distributed, such that the variation can be inferred to be caused only by Monte Carlo statistical

fluctuation. The neutron spectra causing fission for cases with the same water fraction are also

identical. Therefore, the staff concludes that variations in powder density have an insignificant

effect on the bias and may be ignored for the purpose of choosing validation benchmarks.

In addition, differences in the chemical form between plutonium dioxide powder and plutonium

metal are not expected to be significant because 16O is essentially transparent to neutrons

across most of the energy range. In addition to evaluating these differences in composition, the

staff also evaluated the effects of heterogeneous geometry. The applicant performed

calculations showing that the effect of resonance self-shielding in the plutonium-metal array

calculations is small (NRC, 2004a). When the LATTICECELL cross section treatment option

was used instead of INFHOMMEDIUM, the result was a very small difference in keff (less than or

equal to 0.3 percent) (NRC, 2004a). This confirmed the staff’s expectation that the crosssection treatment option would have a small effect. This expectation resulted from the presence

of a very fast spectrum and therefore very little absorption in the resonance region.

Based on these foregoing considerations, the staff has reasonable assurance that the assumed

design applications of AOA(3) include storage applications involving dry, full-density plutonium

dioxide powder and that the plutonium-metal benchmarks are applicable to the validation of

these cases in AOA(3).

Plutonium Dioxide Powder Benchmarks

PU-COMP-MIXED-001: This benchmark set consists of bare arrays of plutonium dioxidepolystyrene compacts. These benchmarks are moderated internally by polystyrene, which has

similar moderating effects to water. The staff found the geometrical arrangement (a single

homogeneous fuel region with vacuum boundary conditions) comparable to that of anticipated

design applications. The staff compared the values of benchmark parameters to the screening

criteria and concluded that they were sufficiently close to those of anticipated design

applications. The staff noted that, in some cases, the 240Pu content is somewhat outside the

range endorsed in NUREG/CR-6698. This is not considered significant because the staff’s

independent calculations and bias trending show that the bias is relatively insensitive to the 240Pu

proportion. Based on this review, the staff considers the three experiments chosen from the PUCOMP-MIXED-001 benchmark set to be applicable for validation of AOA(3).

PU-COMP-MIXED-002: This benchmark set consists of water-reflected arrays of plutonium

dioxide-polystyrene compacts. These benchmarks are moderated internally by polystyrene,

which has similar moderating effects to water. The staff found the geometrical arrangement (a

single homogeneous fuel region with external hydrogenous reflector) comparable to that of

anticipated design applications. The staff compared the values of benchmark parameters to the

screening criteria and concluded that they are sufficiently close to those of anticipated design

applications. The staff noted that, in some cases, the 240Pu content is somewhat outside both

the applicant’s screening criteria and the range endorsed in NUREG/CR-6698. This is not

considered significant because the staff’s independent calculations and bias trending show that

the bias is relatively insensitive to the 240Pu proportion. Subsequent to completion of its
6-25

independent evaluation, the staff reviewed an additional analysis prepared by the applicant in

the supplement to the validation report, DCS01 ZJJ DS CALC H 35075 A, “MFFF Validation

Report Supplemental Information,” dated January 31, 2005 (DCS, 2005; NRC, 2005). The NRC

considers this to be part of the licensing basis in support of the validation effort for the MFFF. In

this analysis, DCS performed a statistical evaluation of the benchmark keff values as a function of

240Pu content, using the ORNL code USLSTATS. This was done for the PU-COMP-MIXED-001

and -002 benchmark experiments in AOA(3), which cover the range in 240Pu content from

approximately 2–18 wt%. Therefore, no extrapolation was needed, and this analysis shows that

the originally determined USL is still valid. The staff concludes that the definition of AOA(3)

adequately covers cases down to 4 wt% 240Pu content. The staff considers the 22 experiments

chosen from the PU-COMP-MIXED-002 benchmark set to be applicable for validation of AOA(3). 

Low Plutonium Assay Powder Benchmarks: Combining the metal and powder benchmarks

discussed above gives a minimum of 49 applicable critical experiments. Based on statistical

techniques employed in the validation report with 49 benchmarks, no additional NPM is required. 

The applicant included additional low-plutonium benchmarks; however, the staff did not make

any determination as to the applicability of these benchmarks (experiment sets BNWL2129T4,

NSE55T5, and PU-29) for validation of AOA(3). The non-parametric technique bases

calculation of the USL on the lowest observed keff; thus, including more experiments can only

decrease the USL. Therefore, the acceptance of these experiments is irrelevant to the

determination of whether there is sufficient margin of subcriticality.

Combination of Metal and Powder Benchmarks in AOA(3)

The staff questioned, however, whether the plutonium-metal benchmarks were sufficiently

similar to the plutonium dioxide-polystyrene benchmarks to be combined in the same AOA. 

Because the non-parametric method applies a statistical margin that is a function only of the

total number of experiments, combining disparate benchmarks together could cause the margin

to be artificially decreased. There is also almost no guidance on when it is appropriate to

combine different types of benchmarks in a single AOA. The staff, therefore, had to develop a

criterion for determining when it is appropriate to combine different types of benchmarks in a

single AOA. Many of the powder benchmarks have been shown to apply to slightly moderated

plutonium dioxide applications, and the metal benchmarks have been shown to apply to certain

dry plutonium dioxide storage applications. The finding that each of these sets of experiments is

acceptable implies that the range of parameters covered by each group is not too broad for the

experiments to be evaluated together in validation. Therefore, the staff determined that it would

consider combining the two sets of benchmarks in a single AOA to be acceptable if the

difference between the two sets is less than or comparable to the difference within each set. 

Conversely, if the difference between the two sets is much greater than the difference within

each set, they should be considered as constituting two separate AOAs. To make this

determination, the staff used several different types of comparisons between the plutoniummetal benchmarks and plutonium dioxide powder benchmarks.

The staff first compared the ranges of the most important physical parameters covered by each

set of benchmarks (i.e., H/(U + Pu), plutonium content, and EALF), including trends of keff as a

function of these parameters. The staff noted that there is significant overlap between many of

the parametric ranges covered by the two sets of benchmarks. Where the parametric ranges do

not overlap, they are contiguous (i.e., there are no discernable gaps between them). The staff
6-26

did not note any significant differences in trends, with the exception that there appears to be a

different trend in EALF for the plutonium-metal and plutonium dioxide powder benchmarks. The

somewhat different behavior of the two sets may be responsible for their not following a normal

distribution. This uncertainty in the distribution is taken into account in the nonparametric

treatment.

Because of the apparent difference in bias trends, the staff compared the spectra for the two

sets of benchmarks in more detail. The staff collapsed the 238 energy groups into 30 groups

and divided the entire energy range into three regions, based on the three categories in Table

2.3 of NUREG/CR-6698 (i.e., thermal = 0–1 eV, intermediate = 1 eV–100 keV, and fast = 100

keV–20 MeV). The rebinning was done to improve statistical accuracy. The staff compared

histograms of the fast, intermediate, and thermal fission fractions for the two benchmark sets. 

The distributions of fast, intermediate, and thermal fission fractions showed a significant degree

of overlap or contiguity. The staff then compared the fission spectra of the two benchmark sets

across the entire energy range to determine how much difference there was in the fission

spectra within each benchmark set and between the benchmark sets. For each benchmark set,

the maximum, minimum, and average fission contribution in each energy group was then

tabulated and graphed. The results show that the spread within the plutonium-metal and within

the plutonium dioxide powder benchmark sets is comparable to or greater than (for each energy

group) the spread between the averages of the two sets. That is, there is as much or more

variation in the fission spectra when comparing different plutonium-metal benchmarks (or

different plutonium dioxide powder benchmarks) than when comparing the typical plutoniummetal and typical plutonium dioxide powder benchmark. The bands formed by the minimum and

maximum fission contributions in each energy group show a significant degree of overlap or

contiguity. Subsequent to completion of its independent evaluation, the staff reviewed an

additional analysis prepared by the applicant in the supplement to the validation report, DCS01

ZJJ DS CALC H 35075A,“MFFFValidation Report Supplemental Information,” dated January 31,

2005 (DCS, 2005; NRC, 2005). The NRC considers this to be part of the licensing basis in

support of the validation effort for the MFFF. In this analysis, DCS compared the neutron

spectra causing fission for the two groups of benchmarks in AOA(3) (the 24 plutonium-metal

benchmarks and the 25 plutonium dioxide-polystyrene benchmarks). The spectra corresponding

to plutonium-metal benchmarks were graphed in red and the spectra corresponding to

plutonium-powder benchmarks in blue. When these were overlaid, there was significant overlap

across the entire neutron energy spectrum (i.e., fission fraction as a function of the EALF). In

addition, the applicant plotted the maximum, minimum, and average fission fraction within each

energy group for each of the two classes of benchmarks. In addition to considering the spectra

using the 238-group structure, the applicant also collapsed the data into 30 energy groups to

obtain better statistics within each group. These analyses all show that, on an energy-group

basis, the spread in the fission fraction within either the metal or powder benchmark set is at

least as great as the difference between the average values for each benchmark set. Therefore,

the staff concludes that the difference between these two sets is smaller than or comparable to

the variation within each set and that there is sufficient overlap in the spectra that they may be

combined for validation purposes.

In addition, use of the TSUNAMI sequences in the SCALE 5 code package shows that, for fast

systems, the reactivity is dominated by 239Pu fission. This observation is particularly important

for the acceptance of plutonium-metal benchmarks for this AOA. Subsequent to completion of

its independent evaluation, the staff reviewed an additional analysis prepared by the applicant in
6-27

the supplement to the validation report, DCS01 ZJJ DS CALC H 35075 A, “MFFF Validation

Report Supplemental Information,” dated January 31, 2005 (DCS, 2005; NRC, 2005). The NRC

considers this to be part of the licensing basis in support of the validation effort for the MFFF. In

this analysis, DCS compared the macroscopic cross sections for dominant reactions that

contribute to the reactivity of a plutonium dioxide-water system in the high-energy range. The

nuclides considered were 239Pu, 240Pu, 16O, and 1H. The reactions evaluated were fission (n,f),

radiative capture (n,(), elastic scattering (n,n), and neutron knockout (n,2n). Three different

fissile media were considered, corresponding to different plutonium dioxide densities and

moderation conditions. It is not necessary to consider higher values of moderation because this

demonstration is limited to the high-energy (defined as 100 keV–10 MeV) range. Based on a

comparison of the kinematics for the neutron energy loss in elastic scattering, it was determined

that elastic scattering is only an important contribution from 16O and 1H. The microscopic cross

sections for these nuclide-reaction pairs were multiplied by appropriate number densities to

determine the macroscopic cross sections. This shows that, in the highest energy range,

239Pu(n,f) dominates all other reactions by at least an order of magnitude. In addition, the

relative ranking of nuclide-reaction pairs is consistent with the ranking of sensitivity coefficients

determined by the NRC staff using TSUNAMI. Therefore, the staff concludes that 239Pu fission

is the dominant reaction for AOA(3) systems in the high-energy limit.

As systems become increasingly moderated, however, absorption becomes increasingly

important and the sensitivity to other materials (e.g., 240Pu, hydrogen, and absorbers) increases. 

The applicant provided justification as to why the fission spectrum is more sensitive to

differences between physical systems than the absorption or leakage spectrum (NRC, 2004a). 

The basis for concluding that the leakage spectrum is relatively insensitive to differences

between systems is that (1) for fully reflected systems, there is essentially no leakage and (2) for

bare or nominally reflected systems, the leakage spectrum is preferentially skewed towards fast

neutron energies (and thus fission dominates). The staff considers this reasonable based on the

known behavior of systems with these boundary conditions. In addition, for fully reflected

systems, the absorption spectrum is greatly dominated by parasitic absorption in the external

reflector and thus relatively insensitive to changes in the interior physical system.

With regard to the absorption spectrum, the applicant provided graphs of the fission, absorption,

and leakage spectra as a function of benchmark number (NRC, 2004a). These graphs show

that there is considerably more variation in the fission spectrum than in either the absorption or

leakage spectrum. This leads to the conclusion (at least for these experiments) that

comparisons of the fission spectra are more indicative of similarity between systems than

comparisons of the absorption or leakage spectra. As an additional check, the staff performed

the same analysis described above for the absorption spectrum (i.e., comparison of the fast,

intermediate, and thermal absorption fractions, and comparison of the spectra on a group-bygroup basis). The staff analysis shows very similar results to the analysis done on the fission

spectra because the difference in absorption characteristics between the two sets of

benchmarks is comparable to or less than the difference within each set.

The staff noted that there was a weak correlation between the metal and powder benchmarks

when the TSUNAMI code sequences were used to compare all the benchmarks in pair-wise

fashion. In light of the results of the spectral comparison described above, the staff surmises

this is likely because of parametric differences (most importantly H/Pu ratio and plutonium
6-28

content). As previously noted, however, the ranges in these parameters covered by the two

benchmark sets have considerable overlap or are reasonably contiguous.

Therefore, based on the relative insensitivity of keff to density and chemical form (presence of

16O), physical considerations about the relative importance of fission, absorption, and leakage,

and a comparison of parametric ranges and energy spectra for the two sets of benchmarks, the

staff concludes that it is appropriate to validate AOA(3) using both the plutonium-metal and

plutonium dioxide powder benchmarks. The combination of these two benchmark sets results in

49 benchmarks found acceptable by the staff for validating AOA(3). In accordance with the nonparametric method described in Part II of the validation report and NUREG/CR-6698, no

additional NPM is required.

Therefore, based on the foregoing considerations, the staff has reasonable assurance that the

systems meeting the criteria in the table above are within the scope of AOA(3) and will be safely

subcritical with a USL, as defined in Part II of the validation report (USL = 0.9345).

6.1.3.5.1.4 AOA(4)—Mixed Oxide Powder. To validate AOA(4), the applicant modeled 59

critical experiments using the KENO-VI code with the 238-group cross-section library. The

applicant modeled some of the benchmarks using the INFHOMMEDIUM cross section treatment

option, and some using the LATTICECELL treatment option. The validation report does not

clearly state which code options (including cross section treatment) were considered validated. 

The bias was determined for the SCALE-4.4a code package on the SGN PC hardware platform. 

The applicant’s chosen experiments included reflected arrays of plutonium dioxide- and MOXpolystyrene compacts. The applicant selected this set of experiments from a much larger suite

of 308 candidate benchmarks by comparing the most important physical and neutronic

parameters of candidate benchmark experiments to those of anticipated design applications. To

do this, the applicant chose a set of screening criteria, consisting of allowable ranges of

important physical parameters, as described in NUREG/CR-6698. The staff examined the

chosen benchmarks to determine whether they were appropriate for a validation of systems

covered by AOA(4). The evaluation of the chosen benchmarks is discussed below.

The applicant defined the validated range of parameters for AOA(4) in Table 5-4 of Part II of the

validation report. The validated AOA is described in Table 6.1-4.

The applicant ran SCALE/KENO-VI for the chosen experiments, calculated keff, and plotted the

value of keff versus the trending parameters of EALF, the moderator to fuel ratios (H/(U + Pu)

and H/Pu), and the 240Pu and total plutonium content. Because the data were not normally

distributed, a non-parametric estimate of the USL was made, assuming an administrative margin

of 0.05. Based on the number of benchmarks included, no additional NPM was needed. The

non-parametric estimate was used as the USL for AOA(4).

The staff compared the range of parameters covered by the benchmark data to the applicant’s

AOA and determined that the benchmarks provided adequate coverage of the defined AOA

across the entire range (except as discussed below).
6-29

Table 6.1-4 Definition of AOA(4)

Geometry Parallelepipeds, spheres

Reflectors Water, depleted uranium*

Chemical form MOX powder

Pu/(U + Pu) 8–22 wt% 

240Pu 4 wt%

H/(U + Pu) 2.8–15*

EALF 0.63–92.6 eV*

*Except as described below

To select the benchmarks, the applicant followed the steps in Section 2.5 of NUREG/CR-6698,

which consist of (1) identifying key parameters associated with normal and upset conditions of

systems to be evaluated, (2) based on these key parameters, establishing a screening AOA for

critical experiments, (3) identifying critical experiments that are within this screening AOA, and

(4) from the scope of selected critical experiments, determining the detailed AOA covered by the

experiments. The final step, showing that design applications are within the detailed AOA, will

be conducted as part of the overall final design process (as indicated in Figure 3-1 of Part II of

the validation report) and is outside the scope of the construction validation review.

With regard to the implementation of this methodology, the staff noted that the screening criteria

are significantly broader in some cases than those in Table 2.3 of NUREG/CR-6698. The values

in Table 2.3 of NUREG/CR-6698 are the result of consensus among criticality experts and may

be considered to be conservative (i.e., they are narrowly defined to provide a very high degree of

confidence of the benchmark applicability). The staff questioned the ranges of parameters in

the screening criteria and the justification for extending them beyond the ranges in Table 2.3 of

NUREG/CR-6698. In addition, the parameters used in the primary screening criteria (239Pu

content, 240Pu content, H/(U + Pu), and EALF) do not include the total plutonium content (wt% of

plutonium in MOX), which is known to be one of the parameters to which keff is most sensitive. 

Since the staff could not resolve all the technical concerns with the applicant’s implementation of

the NUREG/CR-6698 methodology, the staff examined all the chosen benchmarks on a caseby-case basis to evaluate whether they should be included for a validation of AOA(4). The

staff’s review of the applicability of these benchmarks is described below.

NSE55T5: This benchmark set consists of plexiglass-reflected arrays of MOX-polystyrene

compacts. These benchmarks are moderated internally by polystyrene and reflected by

plexiglass, which have similar moderating and reflecting effects to water. The staff found the

geometrical arrangement (a single homogeneous fuel region with external hydrogenous

reflector) comparable to that of anticipated design applications. The staff compared the values

of benchmark parameters to the screening criteria and concluded that, in most cases, they are

sufficiently close to those of anticipated design applications. The staff noted that the H/(U + Pu)
6-30

and 240Pu content are somewhat outside the range endorsed in NUREG/CR-6698. The H/(U +

Pu) was considered slightly high. Despite this, the EALF of these benchmarks, which is strongly

correlated to the H/(U + Pu) ratio, remain within appropriate bounds. Bias trending showed that

there is an unexpectedly strong correlation between keff and the 240Pu content. This was

unexpected because calculations using the TSUNAMI code sequences have shown that the

system sensitivity to the 240Pu content is typically low. The trend is such that including high 240Pu

benchmarks results in a lower keff and therefore reduces the overall USL. Subsequent to

completion of its independent evaluation, the staff reviewed an additional analysis prepared by

the applicant in the supplement to the validation report, DCS01 ZJJ DS CALC H 35075 A,

“MFFF Validation Report Supplemental Information,” dated January 31, 2005 (DCS, 2005; NRC,

2005). The NRC considers this to be part of the licensing basis in support of the validation effort

for the MFFF. In this analysis, DCS performed a statistical analysis of the benchmark keff values

as a function of 240Pu content, using the ORNL code USLSTATS. Using the results from the

confidence band technique generated by USLSTATS, the applicant determined the functional

form of both the calculated keff and its uncertainty, k(x) and w(x), as a function of 240Pu content. 

The applicant then used this information to extend the trend, including allowance for uncertainty,

down to 4 wt% 240Pu content and showed that the originally determined USL is still valid. 

Therefore, the staff concludes that the definition of AOA(4) may be extended down to 4 wt%

240Pu content without the need for additional margin. Because of outstanding questions

concerning the H/(U + Pu) ratio and 240Pu content, the staff considers the 10 experiments

chosen from the NSE55T5 benchmark set to be reasonably applicable for validation of AOA(4).

PU-29: This benchmark set consists of plexiglass-reflected arrays of MOX-polystyrene

compacts. These benchmarks are moderated internally by polystyrene and reflected by

plexiglass, which have similar moderating and reflecting effects to water. The staff found the

geometrical arrangement (a single homogeneous fuel region with external hydrogenous

reflector) comparable to that of anticipated design applications. The staff compared the values

of benchmark parameters to the screening criteria and concluded that, in most cases, they are

sufficiently close to those of anticipated design applications. The staff noted that the H/(U + Pu)

and 240Pu content are somewhat outside the range endorsed in NUREG/CR-6698, as discussed

above for NSE55T5. Because of outstanding questions concerning the H/(U + Pu) ratio and

240Pu content, the staff considers the nine experiments chosen from the PU-29 benchmark set to

be reasonably applicable for validation of AOA(4).

BNWL2129T4: This benchmark set consists of plexiglass-reflected arrays of MOX-polystyrene

compacts. These benchmarks are moderated internally by polystyrene and reflected by

plexiglass, which have similar moderating and reflecting effects to water. The staff found the

geometrical arrangement (a single homogeneous fuel region with external hydrogenous

reflector) comparable to that of anticipated design applications. The staff compared the values

of benchmark parameters to the screening criteria and concluded that, in most cases, they are

sufficiently close to those of anticipated design applications. The staff noted that the H/(U + Pu)

and 240Pu content are somewhat outside the range endorsed in NUREG/CR-6698, as discussed

above for NSE55T5. Because of outstanding questions concerning the H/(U + Pu) ratio and

240Pu content, the staff considers the 19 experiments chosen from the BNWL2129T4 benchmark

set to be reasonably applicable for validation of AOA(4).

PU-8: This benchmark set consists of plexiglass-reflected arrays of MOX-polystyrene compacts. 

These benchmarks are moderated internally by polystyrene and reflected by plexiglass, which
6-31

have similar moderating and reflecting effects to water. The staff found the geometrical

arrangement (a single homogeneous fuel region with external hydrogenous reflector)

comparable to that of anticipated design applications. The staff compared the values of

benchmark parameters to the screening criteria and concluded that, in most cases, they are

sufficiently close to those of anticipated design applications. The staff noted that the H/(U + Pu)

and 240Pu content are outside the ranges endorsed by NUREG/CR-6698. The H/(U + Pu) was

considered well outside the screening range in NUREG/CR-6698, although the EALF of these

benchmarks, which is strongly correlated to the H/(U + Pu) ratio, remains within appropriate

bounds. The situation with regard to the 240Pu content is as discussed above for NSE55T5. 

Because of the high H/(U + Pu) values, the staff does not consider the four experiments chosen

from the PU-8 benchmark set to be demonstrated applicable for validation of AOA(4).

PU-COMP-MIXED-002: This benchmark set consists of water-reflected arrays of plutonium

dioxide-polystyrene compacts. These benchmarks are moderated internally by polystyrene,

which has similar moderating effects to water. The staff found the geometrical arrangement (a

single homogeneous fuel region with external hydrogenous reflector) comparable to that of

anticipated design applications. The staff compared the values of benchmark parameters to the

screening criteria and concluded that they are significantly different from those of anticipated

design applications. The staff found that, in several cases, the H/(U + Pu) ratio is significantly

outside the range endorsed in NUREG/CR-6698. The total plutonium content (100 wt%

plutonium) is also significantly outside the range endorsed in NUREG/CR-6698. The applicant’s

screening criteria specified a range of allowable plutonium content of 1.2 to 52 wt% Pu/(U + Pu). 

The staff performed independent calculations using the TSUNAMI sequences of the SCALE 5

code package to obtain more information about the bias sensitivity to plutonium content. This

was done by comparing a reference case with a fixed plutonium content to a series of test cases

with varying plutonium content. The calculations consist of series of MOX-water spheres at the

same H/X but with different plutonium content. These calculations show that the correlation

between MOX reference cases with the same plutonium content as AOA(4) (6.3–22 wt%

plutonium) and test cases with arbitrary plutonium content drops off substantially for test cases

in the high plutonium range (greater than or equal to 60 wt% plutonium). Similarly, the

correlation between pure plutonium dioxide reference cases and MOX test cases drops off

substantially for test cases in the low plutonium range (less than or equal to 30 wt% plutonium).

To justify the inclusion of 100 wt% plutonium benchmarks in the data set for AOA(4), the

applicant provided two sets of calculations (NRC, 2004a). The first consists of a comparison of

the fission spectra for 30 and 100 wt% plutonium cases, and the second consists of a sensitivity

study of the effect on keff of 238U removal. With regard to fission spectra, the results show that

there is a high degree of similarity between the spectra for the 30 and 100 wt% cases. The staff

remarked during the in-office review that the TSUNAMI results suggest that similarity would

decrease significantly below approximately 30 wt%. In response to this, the applicant provided

some additional calculations comparing 22 and 100 wt% plutonium cases that showed similar

behavior. The staff examined the TSUNAMI results described above in greater detail to gain

additional insights into the similarities and differences between low-plutonium MOX and 100 wt%

plutonium systems. These calculations showed that, for very fast (low-moderation) plutonium

dioxide systems, the sensitivity is dominated by 239Pu fission, but that as systems become

increasingly moderated and the plutonium content decreases, other nuclides and reactions

become increasingly important. For example, for systems with plutonium content in the range

covered by AOA(4), the dominant nuclide-reaction pairs include 238U fission and absorption. 
6-32

Comparing only the 239Pu fission spectra does not take into account the effect of these

competing nuclide-reaction pairs.

In the 238U-removal study, the applicant performed calculations to determine the change in keff

when all of the 238U was removed from 30 wt% plutonium cases. This shows that the 238U

produces a ) keff up to approximately 8 percent. Based on this, the applicant concluded that the

presence of 238U at these levels is not significant to validation. However, the applicant did not

justify that ) keff corresponds to an acceptable degree of similarity. Because both 238U fission

and absorption were shown to be significant in the low plutonium range and they produce

opposite effects on keff, simply studying their combined effect on keff may mask the significance

of either fission or absorption; an error in either the absorption or fission cross section could

produce a bias in keff. Because of the outstanding questions concerning the sufficiency of the

comparison of the 239Pu fission spectra and the 238U-removal study, the staff does not consider

the applicability of 100 wt% plutonium content benchmarks to be demonstrated for applications

in AOA(4).

In addition, for some of the cases, the 240Pu content is somewhat outside the range endorsed in

NUREG/CR-6698, as discussed above for NSE55T5. Because of the high H/(U + Pu) values

and high plutonium content, the staff does not consider the 17 experiments chosen from the PUCOMP-MIXED-002 benchmark set to be demonstrated applicable for validation of AOA(4).

Based on the preceding review, the staff determined that, out of the 59 benchmarks selected for

AOA(4), only 38 are marginally acceptable without further justification (i.e., those in the

NSE55T5, PU-29, and BNWL2129T4 benchmark sets). The non-parametric technique from

NUREG/CR-6698 followed by the applicant specifies that, with 38 applicable benchmarks that

are not normally distributed, an NPM of 1 percent in keff is appropriate. The staff therefore

considers an NPM of 1 percent in keff necessary to provide reasonable assurance that MOX

powder systems will be subcritical under normal and credible abnormal conditions. 

In accordance with the methodology of NUREG/CR-6698, the staff examined the parametric

range covered by the 38 benchmark experiments and determined that they do not adequately

cover the entire validated AOA defined in Table 5-2 of Part II of the validation report. In

particular, the 38 benchmarks do not cover the highest and lowest EALF ranges or the high H/(U

+ Pu) range. Based on the span of the data, the staff considers the range of parameters in

Table 6.1-5 validated with margin as described above.

In addition to restrictions on H/(U + Pu) and EALF, the staff noted that the 38 benchmarks did

not contain depleted uranium reflectors. However, the staff’s independent calculations using the

TSUNAMI sequences of the SCALE 5 code package show a reasonably high correlation

between bare systems and those reflected with 60 cm of depleted uranium (with the same H/(U

+ Pu) ratio). The staff also notes that all design applications covered by AOA(4) contain a large

proportion of depleted uranium as uranium dioxide in the fissile region (78 and 93.5 wt%

uranium dioxide). This is suggestive that cases containing depleted uranium reflectors may be

within the scope of AOA(4). Subsequent to completion of its independent evaluation, the staff

reviewed an additional analysis prepared by the applicant in the supplement to the validation

report, DCS01 ZJJ DS CALC H 35075 A, “MFFF Validation Report Supplemental Information,”

dated January 31, 2005 (DCS, 2005; NRC, 2005). The NRC considers this to be part of the 
6-33

Table 6.1-5 Definition of AOA(4)

Geometry Parallelepipeds, spheres

Reflectors Water, depleted uranium

Depleted uranium up to a reflector of 60 cm thickness

Chemical form MOX powder

Pu/(U + Pu) 8–22 wt% 

240Pu 4 wt%

H/(U + Pu) 2.77–2.79

EALF 1.49–43.5 eV

licensing basis in support of the validation effort for the MFFF. In this analysis, DCS performed

a series of sensitivity calculations to determine the effect of depleted uranium reflectors on the 

calculated keff. The first series of runs consisted of spheres containing MOX powder with 0, 1, 3,

and 5 wt% water content, surrounded by spherical shells of depleted uranium of varying

thicknesses. The results of these calculations show that the presence of depleted uranium

results in a significant increase in the system reactivity up to approximately 20 cm thickness, but

beyond that, the system keff is nearly constant. A reduction of 67 percent in the depleted

uranium cross sections would thus be required to have a significant effect on the system keff.

DCS then performed a series of sensitivity calculations applicable to the ball-milling operation in

the MP. This series consisted of a heterogeneous fissile region containing a spherical core of

plutonium dioxide and water surrounded by a spherical shell of depleted uranium dioxide and

water. This arrangement was reflected by a shell containing 500 kg of depleted uranium, the

same mass as in the ball milling operation. These results show that, unlike the homogeneous

case, there is only a )keff of approximately 5 percent in going from 2 to 6 cm of depleted

uranium. The staff reviewed the calculation for the ball milling operation, DCS01 NBX CG CAL

H 05629D, “Criticality Safety of Ball Milling Units NBX and NBY,” which demonstrates that the

heterogeneous two-zone configuration is more reactive than the homogeneous case and thus

was used to determine criticality safety limits for this operation. The applicant also performed a

calculation in which a depleted uranium packing fraction of approximately 74 percent

(corresponding to a hexagonal close-package array) was used instead of 100 percent. This is

because the depleted uranium will not realistically form a contiguous shell around the fissile

material. This calculation shows that there is an approximately 2-percent conservatism from

assuming that the depleted uranium formed a uniform shell.

DCS also provided a comparison of the fission spectra for the bare, water-reflected, and

depleted uranium-reflected calculations and showed that the spectra were nearly identical,

except with regard to the small contribution under the thermal peak. The fact that systems

covered under AOA(4) will have a significant quantity of depleted uranium mixed in with the

fissile medium (78 percent depleted uranium for the master blend) and that the spectra are

nearly identical shows that the depleted uranium cross sections are well-represented by MOX

benchmarks applicable to AOA(4). The staff believes that the presence of 238U and the similarity
6-34

in the spectra are responsible for the high degree of correlation evidenced in its TSUNAMI

calculations, even though keff exhibits significant sensitivity to the presence of depleted uranium

reflectors of up to 20-cm thickness. 

The staff concluded from the analysis provided by the applicant that (1) it would take a 2/3 error

in the depleted uranium cross sections to produce a significant effect on keff, (2) the

heterogeneous case is bounding and is used as the basis for criticality safety limits, (3) the

heterogeneous case is much less sensitive to the presence of depleted uranium than the

homogeneous case, (4) the highest-plutonium MOX applications still have 78 percent depleted

uranium in the fissile medium, and (5) there is considerable overlap between the spectra of bare,

water-reflected, and depleted uranium-reflected cases covered by AOA(4). Therefore, based on

the staff’s review of the additional analysis prepared by the applicant and the TSUNAMI results

described above, the staff has reasonable assurance that design applications containing

depleted uranium with no greater worth in keff than a 60-cm external reflector are within the

scope of AOA(4). 

Based on the foregoing review, the staff has reasonable assurance that systems meeting the

criteria in the preceding table are within the scope of AOA(4) and will be safely subcritical with a

USL of 0.9249. This is based on a value of 0.9349 as determined in Part II of the validation

report, plus an additional NPM of 0.01 in keff .

6.1.3.5.1.5 AOA(5)—Solutions of Plutonium Compounds. To validate AOA(5), the applicant

modeled 119 critical experiments using the KENO-VI code with the 238-group cross-section

library and the INFHOMMEDIUM cross-section treatment option. The bias was determined for

the SCALE-4.4a code package on the SGN PC hardware platform. The applicant’s chosen

experiments include arrays of bare and reflected plutonium dioxide-polystyrene compacts and

plutonium nitrate solutions. The applicant separated these benchmark experiments into two

groups and determined the bias for each group. Group 1 consists of intermediate energy

plutonium dioxide-polystyrene compacts, while Group 2 consists of highly thermal plutonium

nitrate solution systems. Thus, Group 1 (with 32 experiments) covers the higher energy, lowmoderated portion of AOA(5), while Group 2 (with 87 experiments) covers the low-energy, highly

moderated portion of AOA(5). The staff examined the chosen benchmarks to determine

whether they are appropriate for the validation of systems covered by AOA(5). The evaluation of

the chosen benchmarks is discussed below.

The applicant defined the validated range of parameters for AOA(5) in Table 5-2 of Part III of the

validation report. The validated AOA is described in Table 6.1-6.

The applicant ran SCALE/KENO-VI for the chosen experiments, calculated keff, and plotted the 

values of USL-1 and USL-2 (determined by USLSTATS) versus the trending parameters of

EALF and the moderator to fuel ratio (H/Pu) for each group. Over the range of applicability, the

minimum USL-1 and USL-2 were calculated as a function of these parameters and the lowest

such value chosen as the USL for the entire AOA. The lowest USL (which occurred for Group 1)

was used as the USL for the entire AOA. The benchmark data tested normal, though

USLSTATS indicated that the amount of data may not be sufficient for this normality test. The

staff noted that the histogram in Figure 6-1 shows a double-humped distribution of keff values. 

FSER Reference 6.3.18 calculated the USL using the non-parametric technique, with an

additional margin of 0.01 based on the number of experiments in Group 1. This results in
6-35

essentially the same limit as that derived in the validation report, and, therefore, the staff

considers the USL to be adequate.

The staff compared the range of parameters covered by the benchmark data to the applicant’s

AOA and determined that there are a sufficient number of experiments and that they provide

adequate coverage of the defined AOA across the entire range, except for the cases discussed

below. Additional justification of portions of this table was required for the chemical form of the

fissile material, the 240Pu content, and the reflecting materials.

Table 6.1-6 Definition of AOA(5) 

Geometry Parallelepipeds, arrays of cylinders, spheres

Reflectors Water, cadmium*, borated concrete*

(1) Cd limited to 0.05 cm (0.02 in.) thick sheet surrounding 4.5–9.5-

cm- (1.8–3.7-in.-) thick slab tanks.

(2) Borated concrete limited to 15 cm (5.9 in.) inside and outside

7–7.5-cm- (2.76–2.95-in.-) thick annulus and separated from tank by

1.8–2-cm (0.71–0.79-in.) gap, with the composition below:

10B = 1.59×10-3 atoms/bAcm (2.61×1022 atoms/in.3

) 

11B = 7.04×10-3 atoms/bAcm (1.15×1023 atoms/in.3

)

Ca = 4.65×10-3 atoms/bAcm (7.62×1022 atoms/in.3)

Fe = 5.01×10-4 atoms/bAcm (8.21×1021 atoms/in.3)

Si = 1.66×10-4 atoms/bAcm (2.72×1021 atoms/in.3)

H = 2.17×10-2 atoms/bAcm (3.56×1023 atoms/in.3)

Al = 1.96×10-3 atoms/bAcm (3.21×1022 atoms/in.3)

O = 3.25×10-2 atoms/bAcm (5.33×1023 atoms/in.3)

Chemical form PuO2F2 solution

Pu/(U + Pu) 100 wt%

240Pu 4 wt%

H/Pu 30–50, 85–210

EALF 0.685–4900 eV, 0.135–0.551 eV

*Except as described below

The applicant ran SCALE/KENO-VI for the chosen experiments, calculated keff, and plotted the 

values of USL-1 and USL-2 (determined by USLSTATS) versus the trending parameters of

EALF and the moderator to fuel ratio (H/Pu) for each group. Over the range of applicability, the

minimum USL-1 and USL-2 were calculated as a function of these parameters and the lowest

such value chosen as the USL for the entire AOA. The lowest USL (which occurred for Group 1)

was used as the USL for the entire AOA. The benchmark data tested normal, though

USLSTATS indicated that the amount of data may not be sufficient for this normality test. The

staff noted that the histogram in Figure 6-1 shows a double-humped distribution of keff values. 

FSER Reference 6.3.18 calculated the USL using the non-parametric technique, with an

additional margin of 0.01 based on the number of experiments in Group 1. This results in

essentially the same limit as that derived in the validation report, and, therefore, the staff

considers the USL to be adequate.
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The staff compared the range of parameters covered by the benchmark data to the applicant’s

AOA and determined that there are a sufficient number of experiments and that they provide

adequate coverage of the defined AOA across the entire range, except for the cases discussed

below. Additional justification of portions of this table was required for the chemical form of the

fissile material, the 240Pu content, and the reflecting materials.

The applicant stated that a chemical form of PuO2F2 solution would be used in any criticality

calculations. To demonstrate that this compound is sufficient to bound the various compounds

present (plutonium oxalates and nitrates), the applicant performed sensitivity studies to

determine whether PuO2F2 is more reactive than these compounds over the entire H/Pu range. 

The study shows that PuO2F2 is more reactive than any compound considered except plutonium

dioxide in water. Therefore, the staff considers modeling the various miscellaneous compounds

covered by AOA(5) as PuO2F2 solution to be conservative within the limits on H/Pu, provided that

the presence of plutonium dioxide is precluded.

In addition, the validated range of 4 wt% 240Pu is somewhat outside the range of the Group 2

(plutonium nitrate) benchmarks (which cover 4.23–4.67 wt% 240Pu) but within the range of the

Group 1 (plutonium dioxide-polystyrene) benchmarks. It was not apparent that the validated

AOA therefore covered highly thermal systems (with an H/Pu = 85–210) with less than 4.23 wt%

240Pu content. The staff used the USLSTATS program to determine the effect of extrapolating

the bias and the uncertainty in the bias to systems with 4 wt% 240Pu content. This showed that

extending the AOA down to 4 wt% 240Pu does not result in exceeding the USL of 0.9328 and that

the USL is sufficient to account for any necessary extrapolation. Subsequent to completion of its

independent evaluation, the staff reviewed an additional analysis prepared by the applicant in

the supplement to the validation report, DCS01 ZJJ DS CALC H 35075 A, “MFFF Validation

Report Supplemental Information,” dated January 31, 2005 (DCS, 2005; NRC, 2005). The NRC

considers this to be part of the licensing basis in support of the validation effort for the MFFF. In

this analysis, DCS performed a statistical analysis of the benchmark keff values as a function of

240Pu content, using the ORNL code USLSTATS. Using the results from the confidence band

technique generated by USLSTATS, the applicant determined the functional form of both the

calculated keff and its uncertainty, k(x) and w(x), as a function of 240Pu content, performing this

analysis for the Group II benchmarks in AOA(5). The benchmarks in Group I cover the range in

240Pu content from approximately 2–18 wt%, while those in Group II require a slight

extrapolation down to 4 wt% 240Pu content. The applicant then used this information to extend

the trend, including allowance for uncertainty, down to 4 wt% 240Pu content and showed that the

originally determined USL is still valid. Therefore, the staff concludes that the definition of

AOA(5) may be extended down to 4 wt% 240Pu content without the need for additional margin.

Finally, the applicant referenced the analysis in Attachments 5 and 6 to Part 1 of the validation

report in justifying the inclusion of cadmium and borated concrete absorbers in AOA(5). 

However, no justification for applying the results of that analysis to cover PuO2F2 systems,

especially in the low H/Pu range, was provided. Therefore, further justification for including

these absorbers in the materials covered by AOA(5) was reviewed, as described below.

To demonstrate the applicability of the selected benchmarks in these two groups to postulated

design applications, the applicant conducted several sensitivity studies to determine the effect of

different chemical forms, geometrical shapes, and reflector conditions, on the average neutron

energy (EALF). The chemical compositions considered include plutonium dioxide, hydrides of
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Pu (III) nitrate (Pu(NO3)3@5H2O) and plutonium oxalate (Pu(C2O4)2@6H2O), and plutonium

oxyfluoride (PuO2F2), which the applicant also referred to as standard salt. The geometrical

shapes considered include infinite critical slabs and cylinders. The reflectors considered include

water, plexiglass, colemanite (borated) concrete, regular concrete, and cadmium/steel. For each

of these sensitivity studies, the applicant plotted EALF versus H/Pu for the different

configurations. The applicant concluded from these sensitivity studies (i.e., comparison of EALF

values) that the plutonium dioxide-polystyrene benchmarks are sufficient to validate PuO2F2

design applications at low H/Pu and that the plutonium nitrate benchmarks are sufficient at high

H/Pu. The applicant further concluded that AOA(5) adequately encompasses both cylindrical

and slab geometries, as well as systems with a variety of external reflectors. 

While the difference in EALF between the models was considered diminished at high H/Pu (i.e.,

for highly thermal, dilute systems), an increasing divergence in EALF was observed at low H/Pu.

The range over which this effect could be seen extended significantly higher than the minimum

H/Pu of 30 in some graphs. The applicant has not addressed the impact of this on the validation

in the low H/Pu limit. The applicant also has not justified using this technique as a test for

benchmark applicability or provided a criterion as to what constitutes an acceptable difference in

EALF. The use of an integral parameter such as EALF may not account for all important

characteristics of the systems modeled; systems with the same EALF can have significantly

different material specification and energy spectra.

To test the appropriateness of using the EALF comparison as a sufficiency test for benchmark

applicability, the staff performed an independent analysis to evaluate the effect on the bias to

changes in the chemical form of the fissile material, geometry, and reflectors. 

Chemical Form

The staff constructed several models consisting of fully water-reflected spheres containing

(PuO2 or PuO2F2)-H2O mixtures at various H/Pu ratios and calculated keff using the XSDRNPM

discrete ordinates code. The largest difference in keff for the two systems is expected at the

lowest H/Pu because the concentration of nuclides in the fissile compound is at a maximum in

the low H/Pu limit. This is consistent with the observed behavior in the sensitivity studies

performed by the applicant. The difference between the plutonium dioxide and PuO2F2 cases at

the same H/Pu is caused by the presence of 19F in the design application, as well as differences

in the number densities of plutonium, hydrogen, and oxygen. The substitution of 19F for 16O is

expected to have a small effect on the system keff. These nuclides have very small absorption

and scattering cross sections over the entire energy range of interest (up to about 104 eV), with

elastic scattering the dominant mechanism. Because these nuclides have similar scattering

cross sections and similar atomic masses, they are expected to result in similar neutron

moderation. Because the mass is large and the scattering cross section small compared to that

for hydrogen, moderation by hydrogen (which is common to both cases) will dominate. 

Differences in keff and neutron energy primarily result from differences in density, rather than the

presence of different chemical elements. Therefore, little significant difference in the underlying

physics between the plutonium dioxide benchmarks and the PuO2F2 design applications is

expected.

Perturbation of the 19F and 16O number densities confirms that the presence of these nuclides

has a small effect on the bias. Removing all of the 19F from the model resulted in a )keff of just
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over 2 percent; removing all the 16O resulted in a )keff of about 1.5 percent. Subsequent to

completion of its independent evaluation, the staff reviewed an additional analysis prepared by

the applicant in the supplement to the validation report, DCS01 ZJJ DS CALC H 35075 A,

“MFFF Validation Report Supplemental Information,” dated January 31, 2005 (DCS, 2005; NRC,

2005). The NRC considers this to be part of the licensing basis in support of the validation effort

for the MFFF. In this analysis, DCS modeled plutonium dioxide-water and PuO2F2-water

mixtures at six different H/X ratios to cover the full range in moderation covered by AOA(5) (H/X

= 30–210). For each modeled configuration, DCS performed sensitivity studies in which the

dioxide and diflouride in the fissile medium was removed and the change in keff determined. The

applicant used the CIGALES V2.0 code to derive the number densities of the solutions. The

staff examined the report entitled Sec/T/03.023, “Density of Actinides Nitric Solutions and their

Mixtures: Uranium (VI), Uranium (IV), Plutonium (IV), Plutonium (III), Americium (III) and

Thorium (IV),” and determined that the report demonstrates good agreement between the code’s

predictions and experimental measurements. 

The results show that the largest change in keff occurs for PuO2F2 at low H/X, consistent with the

results of staff’s calculations. The largest change is a )keff of approximately 1.6 percent,

demonstrating that the presence or absence of oxygen and fluorine in the fissile medium has a

small effect on the calculated keff. Removal of all the oxygen and fluorine represents a condition

which greatly exceeds the effect of the worst credible error in the material cross sections. 

Therefore, the staff concludes that plutonium dioxidebenchmarks may be used to validate

PuO2F2 design applications in the parametric range covered by AOA(5).

Geometry

Historically, geometry has been recognized as less important to code validation than material

specification or neutron energy (NUREG/CR-6698, Table 2.3). The system geometry is only

important to the extent that it affects the problem boundary conditions and neutron energy

spectrum. Table 2.3 of NUREG/CR-6698 contains the guidance that the neutron energy

spectrum should cover the same energy range as the design application (i.e., thermal,

intermediate, or fast spectrum). This table also contains broad tolerances on the dimensions of

geometric bodies. The staff expects that systems with similar geometric shapes and sizes, but

with otherwise identical material and neutronic characteristics, will be mutually applicable. In

addition, system geometry is not one of the parameters ordinarily used to trend the bias. The

staff’s calculations show that system shape and dimensions have a direct correlation to the

energy spectrum. Because the bias is trended as a function of EALF, the potential effect of

system geometry on the bias has been taken into account. Therefore, the staff has reasonable

assurance that the benchmark experiments selected, which consist of arrays of plutonium

dioxide-polystyrene compacts, can be used to validate design applications consisting of

cuboidal, cylindrical, or spherical arrangements of the same materials. 

Reflector Materials

The results of the staff’s evaluation of different absorbers in Part I of the validation report for

AOA(1) indicate that the effect of external reflectors on benchmark applicability is expected to

depend strongly on the type, geometry, and thickness of the reflector (see FSER Section

6.1.3.5.1.1). Unreflected benchmark experiments are generally considered to be applicable to
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applications containing low-worth, relatively thin absorbers. Based on this, the staff expected

similar qualitative behavior for externally reflected applications covered by AOA(5). 

A comparison of the EALF values in these modeled cases shows that the reflector materials can

be divided into two categories, including (1) hydrogenous reflectors, consisting of water,

plexiglass, and regular concrete, and (2) strongly absorbing reflectors, including cadmium,

boron, and carbon steel. The energy of applications with the first type of reflector is significantly

lower than those with the second type because of hydrogenous moderation. Hydrogenous

reflectors return a large proportion of the incident neutrons back into the fissile region with

significant softening of the neutron spectrum. Strong absorbers result in a much smaller

proportion of neutrons being returned to the fissile region and with less softening of the

spectrum. Since the benchmark set contains both bare and reflected cases, in general, the bare

benchmarks are expected to be most applicable to cases with strong absorbers and the waterreflected benchmarks to cases with hydrogenous absorbers. Borated concrete is expected to

exhibit characteristics of both groups, as it is a mixture of a strong absorber and hydrogenous

reflector. 

To determine the potential effect of the presence of different absorber materials on the bias, the

staff performed several independent calculations using the XSDRNPM discrete ordinates code. 

The models consist of slab or cylindrical regions with reflectors of the same thickness and

composition as those in Table 5-1 of Part I of the validation report. The applicant stated that the

borated concrete and cadmium reflectors used are similar to those for AOA(1) (DCS, 2003g). 

The fissile material consists of a plutonium dioxide-water mixture at an H/Pu of 30. As with the

sensitivity studies for chemical form and geometry, the greatest differences are expected at the

low H/Pu range of the AOA. Perturbation of the number densities of the absorber (10B or

cadmium), 1H, and 239Pu shows that, for cases with these absorber loadings, the models are

relatively insensitive to the 10B or cadmium cross sections compared to the sensitivity to 1H and

239Pu (i.e., a small change in the 10B or cadmium cross sections produces a small change in the

overall system keff). While the sensitivity to the absorber is greater than in AOA(1), direct

perturbation shows that the absorber is responsible for only a few percent of the total system

sensitivity. The staff notes that AOA(5) has a much lower minimum H/Pu ratio (H/Pu = 30) than

AOA(1) (H/Pu = 100), which is likely to be responsible for the difference in sensitivity. Therefore,

any bias resulting from errors in the boron or cadmium cross sections is expected to be

correspondingly small.

Because the absorber worth can depend strongly on the geometrical configuration and size of

the absorbers, system leakage, and other physical characteristics of the system, these results

may not be indicative of the behavior of systems with much higher absorber worth. Therefore,

the specification of the applicable absorber loading is necessary in the definition of the AOA. 

The staff also performed additional sensitivity calculations in which the number density of the 10B

in the borated concrete and cadmium were reduced. These calculations indicate that, in both

cases, it would require a significant decrease in the absorber number density before there would

a significant change in the system keff, assuming reflector conditions as described above. 

Especially in the case of cadmium, it would require at least a 50-percent decrease in the number

density for a 1-percent change in keff. The applicant stated that, for AOA(5), no more than 30

percent of the borated concrete is required to ensure subcriticality (maintain keff below the USL)

(DCS, 2003g). While comparable information was not provided for cadmium, calculations show
6-40

that the change in keff is less than that observed for boron for a proportional change in number

density. The staff considers the likelihood of the 10B and cadmium cross sections being in error

by such a large proportion to be very low because both boron and cadmium are commonly used

absorbing materials, and errors of this magnitude would have been noticed in other applications. 

Subsequent to completion of its independent evaluation, the staff reviewed an additional analysis

prepared by the applicant in the supplement to the validation report, DCS01 ZJJ DS CALC H

35075 A, “MFFF Validation Report Supplemental Information,” dated January 31, 2005 (DCS,

2005; NRC, 2005). The NRC considers this to be part of the licensing basis in support of the

validation effort for the MFFF. In this analysis, DCS modeled three different configurations

containing borated concrete and cadmium as neutron absorbers. The three cases considered

(an annular tank with borated concrete inside and outside, and a slab tank and precipitator with

cadmium sheets) corresponds to actual process calculations. The applicant varied the boron

and cadmium content in increments of 5 percent from 95 percent to 0 percent of their nominal

values. These nominal values correspond to the thicknesses and arrangements of the

absorbers in Table 6.1-6 above. The applicant concluded that a 40-percent reduction in the

boron content would be required to produce an approximately 1 percent )keff. In addition, a 70-

percent and 55-percent reduction in the cadmium density for the precipitator and slab tank,

respectively, would be required for a noticeable change in keff. Therefore, the staff concludes

that even a very large error in the 10B and cadmium cross sections will not be sufficient to

significantly affect the code bias.

The staff determined that the chosen benchmarks are appropriate for the validation of systems

covered by AOA(5) based on the considerations presented above. Therefore, based on the

foregoing considerations, the staff has reasonable assurance that systems meeting the criteria

in Table 6.1-1 are within the scope of AOA(5) and will be safely subcritical with a USL as defined

in Part III of the validation report (USL = 0.9328).

6.1.3.5.2 Determination of Normal and Abnormal Subcritical Margins

The applicant selected an administrative margin of 0.05 for both normal and credible abnormal

conditions for each AOA covered in the validation report. The administrative margin is used to

determine, along with the bias and uncertainty (as documented in FSER Section 6.1.3.5.1), the

maximum keff or USL, which is part of the design basis of the MFFF. This section discusses the

selection of an appropriate administrative margin, which is part of the margin of subcriticality for

safety as required in 10 CFR 70.61(d).

This margin represents a minimum margin of subcriticality to account for any uncertainties not

explicitly treated in the validation process and is applied in Method 1 (confidence band with

administrative margin) and in the nonparametric method. This margin is not included as a

separate term in Method 2 (lower tolerance band). However, applying the criterion that USL-1 is

less than USL-2 practically means that the USL is determined on the basis of USL-1 (or on the

basis of the nonparametrically determined USL). Justification for the chosen administrative

margin is provided in each part of the validation report and is based on three arguments,

including (1) historical nuclear industry practice, (2) NRC guidance, and (3) a comparison of

USL-1 (including a 0.05 administrative margin) with USL-2. The staff reviewed these

justifications and determined that they are insufficient by themselves to support using an

administrative margin of 0.05 for all potential operations involving plutonium or MOX fuel, for the

reasons stated below.
6Historically, some applicants have been licensed with separate limits for normal and abnormal conditions, while

others have been licensed with a single limit. The use of a single limit is acceptable if judged sufficient to cover both

normal and abnormal conditions. Either approach would in principle have been acceptable for the MFFF.

6-41

Historical fuel facility subcritical margin, as put forth by the applicant, does not support a

minimum subcritical margin of 0.05 for the MFFF because this precedent is not applicable to

systems at the MFFF. Systems involving high-239Pu SNM (e.g., 96 wt% 239Pu oxide and

solutions) exhibit a higher sensitivity of keff to changes in the underlying physical and neutronic

parameters than systems encountered at many other fuel facilities. The effect of any unknowns

can reasonably be expected to be magnified in systems displaying this higher degree of

sensitivity. Therefore, in general, a larger margin of subcriticality for safety is appropriate for

operations with high-239Pu SNM than those with lower fissile content (e.g., 6 wt% plutoniumcontent MOX, low-enriched uranium (LEU)).

Guidance from the NRC (NUREG-1718) does state that an administrative margin of 0.05 is

“generally considered to be acceptable without additional justification when both the bias and its

uncertainty are determined to be negligible.” However, the applicant has not demonstrated that

these conditions have been met. When the bias and its uncertainty are not negligible, it is

difficult to justify that the underlying physics is sufficiently well-understood to justify such a low

administrative margin.

Finally, the comparison of USL-1 with USL-2 does not demonstrate the adequacy of the

administrative margin. This condition that USL-1 is less than USL-2 is necessary, but not

sufficient, to show that adequate margin as used in Method 1 has been provided. Method 1 and

Method 2 are two different statistical treatments of the data, and a comparison between them

can only demonstrate whether the administrative margin is sufficient to bound statistical

uncertainties included in Method 2 but not included in Method 1. There may also be other

statistical or non-statistical errors in the calculation of keff that are not handled in the statistical

treatments. 

For the reasons stated above, the staff did not agree with the applicant’s stated justification for

the use of a 0.05 administrative margin. 

Therefore, staff considered other factors, which are described below, before approving the

margin. In general, an administrative margin of 0.05 has been historically accepted for most fuel

cycle operations. However, additional margin has historically been considered necessary for

operations involving HEU (and, by extension, high-239Pu SNM) than that for those involving LEU. 

In addition, more margin has historically been considered necessary for normal conditions than

for abnormal conditions6 (NRC, 2001c). Therefore, the staff considers an additional

administrative margin of subcriticality for safety to be necessary for proposed operations at the

MFFF.

This margin may be provided in terms of either margin in keff or margin in system parameters. 

The staff has concluded that the applicant’s proposed keff limits (including an administrative

margin of 0.05), together with margin inherent in the parameters associated with proposed

MFFF operations, provide reasonable assurance that there will be a sufficient margin of

subcriticality for safety for proposed operations at the MFFF.
7During the public meeting held on July 31, 2003, the applicant clarified that ”normal condition cases” in this context

refers to the actual physical state of plant operations (”normal operating” conditions), as opposed to conditions

defined by the controlled parameter limits.

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This determination regarding margin in proposed MFFF operations is based on specific

commitments in the CAR and associated documentation, as described below.

Normal Conditions

As stated above, the staff considers an additional margin of subcriticality for safety necessary for

proposed operations at the MFFF. This margin is provided by the following commitments:

• The applicant stated the following: 

All criticality applications in the design of the MFFF show that the

abnormal conditions cases are bounding—by an appreciable equivalent

margin in keff–over the normal condition cases....Operations are rarely

expected to be conducted at the subcritical value.7 (DCS, 2003g). 

The staff reviewed the description of technical practices contained in CAR Section 6.3.4

to verify this statement. 

The staff determined, based on this review, that the technical practices provide considerable

margin in many of the criticality parameters (especially with regard to isotopics, reflection,

density, concentration, and physicochemical form). With regard to these specific criticality

parameters—

• CAR Section 6.3.4.3.2.4 states the following: 

The assumptions used in the criticality calculations are, typically, as

follows:

239Pu/Putotal = 96%, 240Pu/Putotal = 4% and 235U/Putotal = 0%

This is also stated throughout CAR Tables 6-1 and 6-2 (“240Pu $ 4%”). CAR Section

6.3.4.3.2.4 also states the following: 

Incoming plutonium will respect the following conditions (for the main

plutonium isotopes): 

90% #

239Pu/Putotal # 95%, 5% #240Pu/Putotal # 9%

This CAR section also states that there may be up to 1 percent 241Pu and up to 2 percent

U/Pu total present in the incoming material and that “calculations have demonstrated that

increasing the 239Pu content by 1.0 wt% while decreasing the 240Pu content by a

corresponding amount is sufficient to offset any reactivity effect from 241Pu and 235U.” 

The staff’s confirmatory calculations show that the assumed composition is sufficient to
6-43

bound the worst-case combination of possible isotopics for incoming plutonium, with a )

keff between 0.3 and 0.8 percent. 

• CAR Section 6.3.3.2.5 states the following: 

At a minimum, reflection conditions equivalent to 2.54 cm (1 in)...tightfitting water jacket are assumed to account for personnel and other

transient incidental reflectors....In cases when reflection control is not

indicated, full water reflection of process stations or fissile units is

represented by a minimum of 30.48 cm (12 in)...of tight-fitting water

jacket.

The applicant also stated that “there is no active, controllable reflection...control used in

the MFFF” (DCS, 2001a, RAI 101). Based on these two quotations, the use of full water

reflection is expected in the majority of the MFFF criticality calculations (except when

some other level of interstitial moderation is more reactive). This is supported by a

review of CAR Tables 6-1 and 6-2. Tight-fitting full water reflection is known to be an

extremely conservative way to model hydrogenous reflection around process equipment. 

The staff’s confirmatory calculations show that there is a significant )keff resulting from

even a small gap between the fissile material and reflector, as well as from small

changes in reflector thickness. The staff expects, based on these calculations, that the

use of tight-fitting full water reflection will result, in most cases, in conservatism of at least

several percent in )keff.

• CAR Section 6.3.3.2.3 states, “Conservative assumptions are always made about the

density of the fissile material.” CAR Section 6.3.4.3.2.3 states that plutonium dioxide

incoming to the dissolution unit will be limited to less than or equal to 7 g/cm3 and

polished plutonium dioxide will be limited to less than or equal to 3.5 g/cm3. In addition,

this section states the following: 

The assumed density of PuO2 powder being dissolved (of # 7 gm/cm3) is

quite high and, based upon experience, would not actually be

expected...the storage of material initially received will be shown to be

sub-critical for maximum theoretical density material—11.46 g/cm3. 

Moseley (1965) states that powder density depends primarily on the chemical reactions

and temperature of decomposition used in its formation. The density is typically lowest

for powder prepared from plutonium oxalate, as in the MFFF. Based on the density

measurements in this reference, the staff concludes that the applicant’s assumed

densities for most processes appear conservative. The staff’s confirmatory calculations

show that there is a significant )keff in going from the densities reported in Moseley

(1965) (1–2.3 g/cm3) up to the applicant’s assumed densities (ranging from

3.5–11.46 g/cm3). The difference in density produced conservatism in )keff from

approximately 5 percent up to approximately 75 percent.

• CAR Section 6.3.3.2.7 states the following:
6-44

Concentration control...involves the use of concentration-based singleparameter limits established based upon worst-case geometry (i.e.,

spherical) and SNM fissile composition unless these parameters are

controlled by IROFS....Single-parameter limits for concentration are

established in a manner that ensures an adequate margin of subcriticality

(including margins to protect against uncertainties in process variables

and against limits being accidentally exceeded)....

CAR Sections 6.3.3.2.7 and 6.3.4.3.2.7 state that concentration control is generally used

only in systems with “very low fissile material concentrations (secondary streams).” 

Systems with very low concentration will also be deeply subcritical. For those systems,

the commitments above are expected to provide a large margin of subcriticality.

For the large majority of wet processes not relying on concentration control, CAR Section

6.3.5.5 applies. This section states the following: 

Credible optimum conditions (i.e., most reactive conditions physically

possible) for each controlled parameter are assumed in criticality safety

calculations and NCSEs unless specified controls are implemented to limit

the controlled parameter to a specified value or range. 

Based on these statements, the staff expects that optimal concentration will be assumed

in most of the wet portions of the AP process. To gauge the degree of conservatism, the

staff compared nominal concentration values (based on the plutonium inventories and

vessel volumes in CAR Section 11.3) with the optimum values for Pu(NO3)4 solutions

contained in ARH-600, “Criticality Handbook.” The ARH-600 curves for the subcritical

cylindrical diameter (i.e., the primary controlled parameter in the process) as a function of

the plutonium concentration show that the assumption of optimal rather than nominal

concentration results in conservatism of at least several percent in )keff.

In addition to commitments with regard to specific criticality parameters, there is conservatism in

the physicochemical form:

• CAR Section 6.3.4.3.1.1 states, “In the MP process, no chemical transformations take

place. As a consequence, the oxide form of the fissile medium (PuO2 or UO2 as

applicable) is always assumed.” It is well known from calculations and criticality

handbooks that the uranium dioxide and plutonium dioxide oxide forms are typically the

most reactive physicochemical form besides metal.

CAR Section 6.3.4.3.1.1 also states the following: 

For the AP process, a conservative assumption concerning the chemical

form of the fissile matter is made in each step of the process, taking into

account not only the nominal conditions but also the possible process

upsets....The different chemical forms used in the criticality analyses are

as follows...PuO2...Pu(NO3)4...Pu(NO3)3...plutonium oxalate. 
6-45

For certain wet processes in the AP process, Footnote 3 to CAR Table 6-1 states,

“Actual chemical form of Pu Nitrate is Pu(NO3)4, for most process steps, which is less

reactive than Pu(NO3)3.” The staff calculated the magnitude of this effect by changing

the number densities of atoms in the fissile medium and showed that the effect of this on

keff is negligible. However, the process description in CAR Chapter 11 shows that much

of the AP process (such as dissolution and the plutonium purification cycle) involves the

presence of excess nitric acid, which acts as a soluble neutron poison. CAR Table 6-1

shows that excess nitric acid is not considered in the criticality calculations. The

previously referenced graphs in ARH-600 show that there is a varying degree of margin

in the critical cylinder diameter resulting from neglecting the nitric acid. The actual

amount of margin depends on the amount of nitric acid present and was not quantified by

the staff.

In addition, following oxalic precipitation, PuO2F2 is assumed as the reference fissile

medium. For these units, Footnote 4 to CAR Table 6-1 states, “Actual chemical form is a

mixture of Pu Oxalate and Pu Nitrate. Either chemical form is less reactive than PuO2F2.” 

As documented in FSER Section 6.1.3.5.1.5, the staff’s analysis has determined that

PuO2F2 bounds both plutonium oxalate (Pu(C2O4)2) and plutonium nitrate (assumed

Pu(NO3)4). As shown by sensitivity calculations in support of the validation of AOA(5),

over the range of interest in H/Pu, PuO2F2 bounds pure plutonium oxalate by

approximately 0.5–4 percent in )keff and bounds pure plutonium nitrate by approximately

1–6 percent in )keff.

In addition to these specific commitments, the CAR contains preliminary process information

(primarily in CAR Tables 6-1 and 6-2, and CAR Chapter 11) that shows how this margin is

applied in each of the proposed process steps, as described below.

The following relates to the AP process.

• The decanning, milling, and recanning units rely primarily on mass and moderation

control modes. Because the geometry is not controlled, the staff expects that spherical

geometry will be assumed in the derivation of mass limits. CAR Section 6.3.3.2.2 states,

“Mass control involves the use of mass-based single-parameter limits established based

upon worst-case geometry (i.e., spherical)...unless these parameters are controlled by

principal SSCs/IROFS.” Typically, this provides a large degree of conservatism, as there

is no known mechanism for loose powder to spontaneously take on a spherical shape. 

Conservative assumptions regarding plutonium isotopics, reflection, and density are also

made. In particular, CAR Table 6-1 states that before measurement of density, full

theoretical density is assumed. 

• For dissolution/dechlorination and dissolution units, conservative assumptions are made

with regard to plutonium isotopics, reflection, density, and physicochemical form (i.e.,

neglecting of excess nitric acid). In addition, the actual nominal plutonium concentration

will be less than optimal.

• For the purification cycle, conservative assumptions are made with regard to plutonium

isotopics, reflection, and physicochemical form (i.e., neglecting excess nitric acid). In

addition, the actual nominal plutonium concentration will be significantly less than
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optimal. In addition, for the uranium stripping and diluent washing mixer settlers,

conservative uranium isotopics are assumed (i.e., 35 wt% 235U used in calculations,

whereas processes will be limited to less than or equal to 30 wt% 235U). The staff’s

calculation indicate that this will result in a )keff of approximately 2 percent.

• For the oxalic precipitation and oxidation unit, conservative assumptions are made with

regard to plutonium isotopics, reflection, and physicochemical form (i.e., Pu(NO3)3

without excess nitric acid before precipitation; plutonium dioxide diflouride following

precipitation). The calcination furnace is treated below.

• For the homogenization unit, canning unit, and calcination furnace, conservative

assumptions are made with regard to plutonium isotopics and reflection. These units rely

primarily on geometry in combination with neutron absorber control, mass, and density. 

The density still appears to be conservative relative to the values in Moseley (1965),

although it is lower (3.5 g/cm3) than for upstream units. Moderation is not controlled, but

powder should be dry under normal conditions following calcining. 

• For the oxalic mother liquor recovery unit, conservative assumptions are made with

regard to plutonium isotopics, reflection, and physicochemical form (i.e., PuO2F2).

• The remaining units in the AP process contain only low levels of plutonium under normal

conditions, relying mainly on concentration control. Under normal conditions, they are

expected to be deeply subcritical.

The following relates to the MP:

• For the receiving area, conservative assumptions are made with regard to plutonium

isotopics and reflection. In addition, some processes in this area rely on mass control

without controlling geometry and therefore must assume spherical geometry. There is

also a variety of densities assumed; all of these densities are conservative relative to the

values in Moseley (1965). 

• For the remaining Units in the MP, the plutonium content of powder and fuel is limited to

22 and 6.3 wt% plutonium dioxide. The neutronic behavior of these systems (especially

at 6.3 wt%) is more indicative of LEU fuel than that of other processes at the MFFF. 

Especially for systems with 6.3 wt% plutonium dioxide content (comprising AOA(2) and

part of AOA(4)), an administrative margin comparable to that historically allowed at

licensed LEU facilities is appropriate.

There is also some conservatism in density and reflection. While the density is much greater

following pelletizing and sintering, pellets are all produced at the lower assay of 6.3 wt%

plutonium dioxide. There is some conservatism in the plutonium isotopics for the powder and

pellet areas. CAR Section 6.3.4.3.2.4 states the following regarding the MP: 

At each step of the process (pure PuO2, master blend, final blend, and pellets), a

conservative assumption is made based on process values. The process values

are as follows...20% plutonium for the master blend...2% to 6% for the final blend

and pellets. 
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The CAR section also states, however, that “the plutonium content used for criticality

calculations is as follows (actual design values are less than these)...22% for the master

blend...6.3% for the final blend.” The difference between the nominal values and bounding

values used in the criticality calculations produces some additional conservatism in the

calculated keff. 

In addition to the lower plutonium content, there are other considerations that provide additional

assurance of an adequate margin of subcriticality for MOX pellets, rods, and assemblies under

AOA(2). The neutronic characteristics of finished MOX fuel have been well studied in order to

enable it to be used as reactor fuel. This is possible because the final configuration is fixed and

not easily subject to change. In addition, the chosen benchmarks used to validate AOA(2) are

physically unusually, similar to anticipated design applications (i.e., they are both MOX fuel rod

lattices). The benchmarks are also all well-moderated thermal systems for which the neutron

cross sections are well behaved (i.e., smoothly varying), have been measured with a high

degree of confidence, and have been tested in many applications over the years. For these

systems, there is relatively little apparent margin in physical parameters; density, material form,

and geometry are expected to be modeled fairly accurately. However, this is offset by the

reduced sensitivity to the underlying parameters, the lower risk, and the smaller uncertainty in

cross sections and the physical configuration. 

As discussed in FSER Section 6.1.3.4.3, CAR Tables 6-3 and 6-4 also contain “typical, order-ofmagnitude values” of volume, dimension, and mass limits characteristic of operations at the

MFFF. Although the applicant stated that these values will not be referenced in NCSEs, they

provide some information as to the degree of margin in proposed MFFF operations. As stated

by the applicant, these values represent “‘preliminary best estimate’ nominal Pu values typical of

those expected to be found in the AP and MP processes” (DCS, 2001a, RAI 104). The staff

compared the values in CAR Tables 6-3 and 6-4 to corresponding published values in

ANSI/ANS-8.1-1983 (R1988) (ANSI, 1988) and criticality handbooks TID-7028 and TID-7016. 

Mass values were compared to those in ANSI/ANS-8.12-1987 and TID-7016. The staff noted

that the numbers included in Tables 6-3 and 6-4 are based on conservative assumptions of

density, reflection (for the fully reflected cases), and geometry (for mass and volume limits). 

Based on this, the staff has reasonable assurance that these preliminary best-estimate values

are satisfactorily conservative to provide for an adequate margin of subcriticality for safety.

Therefore, based on the proposed administrative margin of 0.05 and the margin in the system

parameters defined by the commitments and preliminary process information above, the staff

has reasonable assurance that there will be acceptable margin of subcriticality for safety for

normal conditions at the MFFF.

Credible Abnormal Conditions

Section 6.4.3.3.4 of NUREG-1718 (NRC, 2000) states that the fact that a condition is abnormal

may be used as “justification for having a lower margin of subcriticality than would be

permissible for normal conditions.” In such cases, “the abnormal conditions should meet the

standard of being at least ‘unlikely’ from the standpoint of the double contingency principle” and,

in addition, “the increased risk associated with the less conservative margin should be

commensurate with and offset by the unlikelihood of achieving the condition.” Based on the

applicant’s commitment to follow the DCP, the staff has reasonable assurance that achieving an
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abnormal condition will be at least unlikely (FSER Section 6.1.4.2). In addition, an administrative

margin of 0.05 has historically been found acceptable by the NRC for abnormal conditions at

those facilities most similar to the MFFF. Therefore, the staff finds an administrative margin of

0.05 acceptable under abnormal conditions for proposed MFFF operations.

The staff has reasonable assurance that an administrative margin of 0.05, together with the

commitments and preliminary process information described above, will provide an acceptable

margin of subcriticality for safety, for both normal and credible abnormal conditions at the MFFF.

6.1.3.6 NCSE Criticality Controls

The applicant stated that the criticality controls credited in NCSEs will be identified and

evaluated during development of the ISA and will later be identified as IROFS in its ISA

summary, to be submitted as part of any later DCS application for a license to possess and use

licensed material. Since the specific controls relied on as NCS IROFS are not design

information, the identification of such controls is not needed for the review of the revised CAR;

therefore, this approach is acceptable to the staff.

6.1.4 Design Bases of the PSSCs

6.1.4.1 Description of PSSCs

Pursuant to 10 CFR 70.23(b), before approving construction of the facility, the NRC must find

that design bases of the PSSCs provide reasonable assurance of protection against natural

phenomena and the consequences of potential accidents. Revised CAR Section 6.4, “Design

Bases,” states that the “[p]rincipal SSCs are described in Chapter 5 of this document. Specific

IROFS associated with criticality safety will be identified in the ISA.” Revised CAR Table 5.6-1,

“MFFF Principal SSCs,” identifies the PSSCs for criticality hazards as ”criticality control,” with

specific controls to be determined during the design of the facility.

Because the applicant can choose any of the criticality control methods described in this revised

CAR chapter to prevent criticality, the staff did not evaluate any specific controls during review of

the revised CAR. However, based on the commitments in this chapter, the staff has reasonable

assurance that the design bases of the PSSCs will provide adequate protection against the

consequences of a criticality accident (as required by 10 CFR 70.23(b)), even though the

specific IROFS have not yet been identified.

6.1.4.2 Commitment to DCP

The commitment to the DCP, including the corresponding definition of unlikely, is part of the

design basis. The DCP requires (ANSI/ANS-8.1-1983 (R1988)) that “at least two unlikely,

independent, and concurrent changes in process conditions” must occur before criticality is

possible. Additionally, 10 CFR 70.64(a)(9) requires that the design of the proposed facility

adhere to the DCP. As discussed below, the staff considers the commitment to the DCP

acceptable and, in this regard, finds that the applicant has met the double contingency part of

the 10 CFR 70.64(a)(9) requirement for construction authorization. The commitments contained

in the remainder of Chapter 6 provide the additional measures necessary for criticality control,

which satisfies the requirements of 10 CFR 70.64(a)(9) for construction authorization. 
6-49

To provide reasonable assurance that the design will result in controls that are sufficiently

reliable and available when needed to perform their safety function, the meaning of “unlikely” for

each contingency must be unambiguously defined. As discussed in FSER Section 6.1.4.3,

“unlikely” for meeting the DCP is defined as “not expected...during the facility lifetime.” This is

qualitatively consistent with a failure probability on the order of once in 100 years.

The applicant has stated that criticality will be made highly unlikely. An accidental criticality has

the potential to exceed the threshold under 10 CFR 70.61, “Performance Requirements,” for

doses to facility workers. Dose mitigation in the form of shielding will not be credited in reducing

the consequence level, as discussed in FSER Section 6.1.4.3. In revised CAR Section 5.4.3,

the use of deterministic methods for demonstrating that criticality hazards are highly unlikely is

described. This consisted of application of double contingency, Appendix B, “Quality Assurance

Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,” to 10 CFR Part 50, “Domestic

Licensing of Production and Utilization Facilities,” industry codes and standards, and

management measures for failure detection. The applicant, however, did not explain in the

revised CAR how it will determine when an acceptable level of likelihood has been reached or

how these generic criteria will be applied to specific controls. The applicant provided additional

details on its method for ensuring that criticality hazards would be made highly unlikely (DCS,

2003b, 2003d). The description of the accident sequences leading to criticality and the controls

relied on to prevent criticality will be described in NCSEs. The description of accident

sequences would include the initiating event, methods of prevention, including at least two

independent controls and their safety function, redundancy and diversity, the safety margin,

failure modes, and failure detection and surveillance requirements. The description of controls

would include the safety function, quality classification (Ql-1a or Ql-1b), operating range and

limits, testing and maintenance requirements, environmental design factors, natural phenomena

response, required instrumentation, and applicable codes and standards. 

In accordance with the DCP, a least two independent controls whose failure is unlikely must be

specified. For passively controlled units whose failure is not credible, it is considered sufficient

to place them under the plant’s configuration management program and assign the highest

quality level (QL-1a), as done for sole IROFS. To show incredibility, it must be demonstrated

that the system will remain subcritical under all credible process conditions. As discussed in

FSER Section 6.1.3.4, accident sequences in which there are no credible means to achieve

criticality do not require dual independent controls to comply with the DCP. Therefore, this

approach is acceptable to the staff. 

For systems with credible failure paths leading to criticality, active or passive engineered controls

must be classified at least QL-1b, and administrative controls must be simple and unambiguous. 

The likelihood determination will be based on consideration of all the applicable availability and

reliability qualities, consistent with Section 5.4.3.2 of NUREG-1718 (NRC, 2000), which include

those specified above. The staff considers this sufficient to ensure that all factors that affect

accident sequence and control likelihood will be taken into consideration. In addition, for each

independent and unlikely-to-fail control, there must be either a means to detect the failure of the

control (on a period to be justified in the NCSEs) or safety margin (sufficient to ensure that

repeated failure of the controls does not lead to criticality). Based on this, the staff has

reasonable assurance that adherence to the DCP (in which each control failure is sufficiently

reliable and available to be made unlikely), combined with failure detection or additional safety
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margin as described above, provides reasonable assurance that accident sequences leading to

criticality will be highly unlikely. 

For the first case, the requirement that a means to detect the control failure be provided can

result in a significant reduction in the likelihood that both independent controls will be in a failed

state at the same time, provided that this time period is sufficiently short compared to the

expected time between failures of the control. This period will be specified and justified in the

NCSEs. The staff has reasonable assurance that the combination of two independent and

unlikely-to-fail controls (each of which is not expected during the facility lifetime) with adequately

justified failure detection will ensure that criticality is highly unlikely. 

For the second case, the applicant has stated that the safety margin must be sufficiently robust

that multiple failures of the control will not lead to a loss of subcriticality. Specifically, there must

be sufficient margin to accommodate at least three independent failures of the controls. In

evaluating abnormal conditions, the most reactive credible change in the controlled parameter

must be evaluated. For processes relying on dual-parameter control, this typically leads to

abnormal conditions that have considerable margin, which would ensure considerable safety

margin for the sequence. In the case of both systems relying on dual- and single-parameter

control, however, the requirement that the system must accommodate multiple independent

failures provides a significant reduction in the likelihood of criticality. The staff considers the

occurrence of multiple independent failures, each of which is not expected during the facility

lifetime, to be highly unlikely.

Instead of monthly failure detection or additional safety margin, the applicant stated that it may

use other means to ensure that criticality is highly unlikely. In this case, the applicant will show

that these alternate means provide comparable assurance to the two methods described above. 

A list of any systems utilizing methods other than failure detection or additional margin will be

provided in an ISA summary. Regardless of the means of demonstrating that criticality is highly

unlikely, the rationale will be provided in NCSEs supporting a license application and ISA

summary. 

Section 5.4.3.2 of NUREG-1718 contains the following acceptance criterion:

A purely qualitative method of defining “unlikely” and “highly unlikely” is

acceptable if it incorporates all of the applicable availability and reliability qualities

to an appropriate degree...one acceptable definition of “highly unlikely” is a

system of IROFS that possesses double contingency protection with each of the

applicable qualities to an appropriate degree.

In addition, the Commission has adopted a strategic goal of no inadvertent criticality. Pursuant

to 10 CFR 70.64(a)(9), the staff, therefore, finds that the design basis of the facility adheres to

the DCP for construction authorization. Based on this and the other commitments contained in

Chapter 6 of the CAR, the staff therefore finds that the criticality BDC set forth in

10 CFR 70.64(a)(9) have been met for construction authorization.
8The term “unlikely” as used in the DCP should not be confused with the term “unlikely” in the performance

requirements of 10 CFR 70.61(c) for intermediate-consequence events.

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6.1.4.3 Commitment to ANSI Standards

The staff reviewed the applicant’s commitment to the ANSI/ANS-8 series consensus standards

in criticality safety, as described in revised CAR Sections 6.4 and 6.5. In revised CAR Section

6.4, the applicant committed to the use of the ANSI/ANS-8 series standards endorsed in RG

3.71 (NRC, 1972) in the design of the facility. Revised CAR Section 6.4 commits, in general, to

“comply with the guidance (shall statements) and implement the recommendations (should

statements)” of the applicable standard but identifies several clarifications to specific

commitments within certain of the standards. Thus, in the context of this section, “guidance” is

equivalent to the “requirements” of the ANSI standards. Moreover, RG 3.71 endorses several of

these standards conditionally. In those cases in which the regulations or regulatory guidance

disagreed with the standard, the applicant was requested to clarify or modify the commitment.

Revised CAR Section 6.5 contains a discussion of commitments to ANSI/ANS-8.23-1997, which

the applicant does not consider part of the design basis of the facility. Specific commitments to

these standards are summarized below:

• ANSI/ANS-8.1-1983 (R1988) (ANSI, 1988)—The applicant committed to comply with the

guidance (shall statements) of ANSI/ANS-8.1-1983 (R1988) and implement the

recommendations (should statements), with clarification of three provisions. The first

clarification is that, for Section 4.2.2 of the ANSI standard, the applicant committed to

follow the DCP, which requires that at least two unlikely, independent, and concurrent

changes in process conditions must occur before criticality is possible. For the purposes

of meeting this commitment, “unlikely” is defined as “events or event sequences that are

not expected during the facility lifetime, but are considered credible.” This commitment

will be met for those processes and areas in which criticality is determined to be credible. 

The staff notes that assessment of credibility will be determined during the performance

of the ISA (covered in revised CAR Chapter 5). The staff further notes that a definition of

unlikely that is qualitatively consistent with a probability of failure on the order of 10-2 per

year is considered to be acceptable in NUREG-1718 (NRC, 2000). The phrase “not

expected during the facility lifetime” is, therefore, acceptable provided the lifetime of the

facility is assumed to be greater than approximately 100 years.

8

 The applicant clarified

that, while this is a qualitative determination, it is consistent with a failure probability on

the order of once in 100 years (DCS, 2002c, 2002e). Therefore, this is acceptable to the

staff. The second clarification is that, for Section 4.2.3 of the ANSI standard, the

applicant committed to follow the standard (which lists several different types of control

methods including engineered and administrative controls) but committed to relying on

engineered features whenever practical and to justify the use of administrative controls. 

This is consistent with the preferred design approach and is, therefore, acceptable to the

staff. The third clarification is that, for Section 4.3.2 of the ANSI standard, the applicant

committed that, where an extension to the area(s) of applicability is required, the

calculational method will be supplemented by other calculational methods to provide a

better estimate of bias in the extended area(s) or through an increase in the margin of

subcriticality. In making the extension, trends in the bias and any additional uncertainty

must be considered in determining the appropriate amount of margin. Section 4.3.2 of

the ANSI standard states that the AOA may be extended by making use of trends in the
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bias, and Section 4.3.3 states that the uncertainty in the bias shall contain allowances for

extensions to the AOA. These commitments are consistent with the standard. The

standard has been endorsed, and so this process is acceptable to the staff. However,

this should not be construed as allowing redefinition of the applicant’s AOA more broadly

than that described in the NRC-reviewed validation reports. The proposed keff limits are

acceptable to the staff with the AOAs as defined in the validation reports (and

documented in this FSER). As stated in the authorization condition, the proposed keff

limits and definition of the AOAs cannot be changed without NRC review and approval.

With regard to the subcritical limits in ANSI/ANS-8.1-1983 (R1988), the staff considers

the use of subcritical limits from the standard instead of explicit calculation to be an

acceptable practice. These results have been endorsed in RG 3.71 (NRC, 1972) and, as

single parameter limits, are very conservative. In addition, since the standard was

merely reaffirmed in the 1988 version, the staff considers the use of the 1988 version

acceptable.

• ANSI/ANS-8.3-1997, “Criticality Accident Alarm System”—The applicant committed to

comply with the guidance (shall statements) of ANSI/ANS-8.3-1997 and implement the

recommendations (should statements), as modified by RG 3.71, with clarification of one

provision. The clarification is that, for Section 4.1.3 of this standard, the applicant

committed to evaluate the overall risk to personnel, specifically with respect to the risk

from operational interruption and relocation following false alarms. The staff considers it

appropriate to consider overall risk but notes that 10 CFR 70.24(a) requires a criticality

alarm in all areas of the facility where more than the specified quantities of fissionable

materials are handled or stored. A request for exemption from the requirements of

10 CFR 70.24(a) will be handled as described in revised CAR Section 6.3.2 and may

include risk arguments. This is consistent with the rule and regulatory practices and,

therefore, is acceptable to the staff.

• ANSI/ANS-8.7-1975, “Guide for Nuclear Criticality Safety in the Storage of Fissile

Materials”—The applicant did not consider ANSI/ANS-8.7-1975 to be part of the design

basis of the facility. The general commitment to ANSI/ANS-8.1-1983 (R1988) and

technical practices as described in revised CAR Sections 6.3.3.2.8 and 6.3.4.3.2.8 are

sufficient to ensure that criticality safety is appropriately provided for fissile material

storage areas.

• ANSI/ANS-8.9-1975, “Guide for Nuclear Criticality Safety for Steel-Pipe Intersections

Containing Aqueous Solutions of Fissile Materials”—The applicant stated that ANSI/ANS8.9-1975 has been withdrawn by the ANS-8 working group and will not be used in the

design of the facility. Piping configurations containing aqueous solutions of fissile

material will be evaluated by calculation in accordance with ANSI/ANS-8.1-1983 (R1988). 

Because using validated methods to determine subcritical limits is an acceptable

methodology, the staff determined that this approach is acceptable.

• ANSI/ANS-8.10-1983, “Criteria for Nuclear Criticality Safety Controls in Operations with

Shielding and Confinement”—The applicant did not consider ANSI/ANS-8.10-1983 to be

part of the design basis of the facility because the approach used for the facility is to

prevent criticality in accordance with the DCP, rather than rely on shielding and
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confinement for dose mitigation. In addition, the applicant stated that events with the

potential to exceed 10 CFR 70.61 requirements will be made highly unlikely, “including

criticality events, without regard to actual dose consequences” (DCS, 2003b). Because

shielding will not be credited in this way (it must still be considered for detector

coverage), the staff considers it appropriate to exclude ANSI/ANS-8.10-1983 as a design

basis for the facility. 

• ANSI/ANS-8.12-1987, “Nuclear Criticality Control and Safety of Plutonium-Uranium Fuel

Mixtures Outside Reactors”—The applicant did not consider ANSI/ANS-8.12-1987 to be

part of the design basis of the facility. The staff notes that this standard does not contain

any administrative requirements to which the applicant should commit. This standard

only contains subcritical limits for certain plutonium-uranium mixtures. In the absence of

a commitment to this standard, the commitment to ANSI/ANS-8.1-1983 (R1988) ensures

the use of validated methods in computer calculations to demonstrate subcriticality. 

Therefore, the staff considers it appropriate to exclude ANSI/ANS-8.12-1987 as a design

basis for the facility. 

• ANSI/ANS-8.15-1981, “Nuclear Criticality Control of Special Actinide Elements”—The

applicant did not consider ANSI/ANS-8.15-1981 to be part of the design basis of the

facility. Criticality control of special actinide nuclides will be explicitly evaluated by

calculation in accordance with ANSI/ANS-8.1-1983 (R1988). Because using validated

methods to determine subcritical limits is an acceptable methodology, the staff

determined that this approach is acceptable.

• ANSI/ANS-8.17-1984, “Criticality Safety Criteria for the Handling, Storage, and

Transportation of LWR Fuel Outside Reactors”—The applicant committed to comply with

the guidance (shall statements) of ANSI/ANS-8.17-1984 and implement the

recommendations (should statements) with clarification of two provisions. The first

clarification is that, for Section 4.11 of the standard, the applicant committed to use of the

DCP for the handling, storage, and transportation of fuel units and rods. Compliance

with the DCP is required for the facility by 10 CFR 70.64(a)(9), and, thus, this

commitment is acceptable to the staff. The second clarification is that, for Section 5.1 of

the standard, the applicant committed that, where an extension to the area(s) of

applicability is required, the calculational method will be supplemented by other

calculational methods to provide a better estimate of bias in the extended area(s) or

through an increase in the margin of subcriticality. This is consistent with the endorsed

standard and is, therefore, acceptable to the staff.

• ANSI/ANS-8.19-1996, “Administrative Practices for Nuclear Criticality Safety”—The

applicant committed to comply with the guidance (shall statements) of ANSI/ANS-8.19-

1996 and implement the recommendations (should statements), with the exception that

no commitments are made to Section 10 of the standard regarding emergency response

to criticality accidents. ANSI/ANS-8.19 is currently undergoing review by the working

group and has the potential to be changed in the near term. However, the staff notes

that the applicant has committed to ANSI/ANS-8.23-1997, which contains many of the

same requirements.
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Additionally, the staff notes that emergency response procedures are not at issue in

deciding whether to approve the revised CAR. These procedures will be evaluated by

the staff if DCS submits an application for a license to possess and use licensed

material. 

• ANSI/ANS-8.20-1991, “Nuclear Criticality Safety Training”—The applicant committed to

comply with the guidance (shall statements) of ANSI/ANS-8.20-1991 and implement the

recommendation (should statements) without exception or clarification. This is,

therefore, acceptable to the staff.

• ANSI/ANS-8.21-1995, “Use of Fixed Neutron Absorbers in Nuclear Facilities Outside

Reactors”—The applicant committed to comply with the guidance (shall statements) of

ANSI/ANS-8.21-1995 (this standard contains no recommendations) without exception or

clarification. This is, therefore, acceptable to the staff.

• ANSI/ANS-8.22-1997, “Nuclear Criticality Safety Based on Limiting and Controlling

Moderators”—The applicant committed to comply with the guidance (shall statements) of

ANSI/ANS-8.22-1997 and implement the recommendations (should statements), with

clarification of one provision. This clarification is that, for Section 4.1.7 of the standard,

the applicant committed to administrative controls to limit combustible loading for fire

protection and to justify fire protection provisions in all fissile material processing,

handling, or storage areas. This approach is acceptable to the criticality staff since this

mainly affects fire protection, although the effects on criticality safety of a fire or initiation

of fire protection measures (including both engineered systems and administrative

responses) should be evaluated in NCSEs supporting a license application.

• ANSI/ANS-8.23-1997, “Nuclear Criticality Accident Emergency Planning and

Response”—As stated in revised CAR Section 6.5, the applicant committed to comply

with the guidance (shall statements) and implement the recommendations (should

statements) of ANSI/ANS-8.23-1997 without exception or clarification. While the

applicant did not consider this part of the design basis of the PSSCs, this commitment

provides reasonable assurance that consideration will be given to emergency planning

during design. This is, therefore, acceptable to the staff. 

• In addition to those standards to which the applicant has committed above, the following

criticality safety standards are referenced in NUREG-1718 (NRC, 2000):

— ANSI/ANS-8.5-1996, “Use of Borosilicate-Glass Raschig Rings as a Neutron

Absorber in Solutions of Fissile Material”

— ANSI/ANS-8.6-1983 (R1987), “Safety in Conducting Subcritical NeutronMultiplication Measurements In Situ”

Revised CAR Section 6.3.4.3.2.9 states that, wherever neutron absorber control is used, it will

be part of the geometry and fixed by design. In addition, the applicant stated (DCS, 2001a, RAI

77; and DCS, 2002e) that it did not envision using raschig rings for criticality control in facility

operations but will instead rely only on fixed neutron absorbers in accordance with ANSI/ANS8.21-1995. The applicant also did not intend to conduct subcritical neutron multiplication
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measurements at the facility. Therefore, the staff concurs that commitments to these two

standards are not applicable to the design of the facility and, therefore, are not part of the design

basis of the facility.

Therefore, the following standards have been identified, in whole or in part, as part of the design

basis of the facility—ANSI/ANS-8.1-1983 (R1988), ANSI/ANS-8.3-1997, ANSI/ANS-8.17-1984,

ANSI/ANS-8.19-1996, ANSI/ANS-8.20-1991, ANSI/ANS-8.21-1995, and ANSI/ANS-8.22-1997. 

The applicant has also committed to ANSI/ANS-8.23-1997. The other standards are either not

applicable to the design of the facility or are adequately covered by other commitments. This,

therefore, represents an acceptable set of design bases of the facility. However, nothing

precludes the use of any other standards endorsed in RG 3.71.

6.2 Evaluation Findings

In Section 6.4 of the revised CAR, DCS provided design basis information for NCS PSSCs that it

identified for the facility. Based on the staff’s review of the revised CAR and supporting

information provided by the applicant relevant to NCS, the staff finds that, with the following

conditions contained in this chapter of the FSER, DCS has met the 10 CFR 70.61(d)

performance requirement and the BDC set forth in 10 CFR 70.64(a)(9) to the extent required for

construction authorization. Therefore, pursuant to 10 CFR 70.23(b), the staff finds that the

design bases of the PSSCs identified by the applicant will provide reasonable assurance of

protection against natural phenomena and the consequences of potential accidents. The

following construction authorization condition (S-1) applies:

Construction Authorization Condition

Notwithstanding commitments in the Construction Authorization Request, the

facility will be designed so that keff shall not exceed an Upper Subcritical Limit of

0.9249 for normal and credible abnormal conditions covered by AOA(4). DCS shall

not increase the keff limits or change the AOA boundaries beyond those contained

in Chapter 6 of the Final Safety Evaluation Report (FSER) without prior NRC review

and approval.

The following open items in the April 30, 2002, draft safety evaluation report have been

closed—NCS-1, NCS-2, NCS-3, NCS-4, NCS-5, NCS-6, NCS-7, and NCS-8. 

6.3 References

(ANSI, 1988) American National Standards Institute/American Nuclear Society. ANSI/ANS-8.1-

1983 (R1988), “Nuclear Criticality Safety in Operations with Fissionable Materials Outside

Reactors.” 1988.

(ANSI, 1996) American National Standards Institute/American Nuclear Society. ANSI/ANS8.19-1996, “Administrative Practices for Nuclear Criticality Safety.” 1996. 

(ANSI, 1997) American National Standards Institute/American Nuclear Society. ANSI/NAS-8.3-

1997, “Criticality Accident Alarm System.” 1997.
6-56

(DCS, 2001a) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Response to Request for Additional Information—Construction Authorization

Request. August 31, 2001.

(DCS, 2001b) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Clarification of Responses to NRC Request for Additional Information. 

December 5, 2001.

(DCS, 2002a) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Clarification of Responses to NRC Request for Additional Information. 

January 7, 2002.

(DCS, 2002b) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Clarification of Responses to NRC Request for Additional Information. 

February 11, 2002.

(DCS, 2002c) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Clarification of Responses to NRC Request for Additional Information. March

8, 2002.

(DCS, 2002d) Ihde, R., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Mixed Oxide Fuel Fabrication Facility Construction Authorization Request

Revision. October 31, 2002 (including page changes through February 9, 2005). 

(DCS, 2002e) Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory Commission,

RE: Requests for Additional Information, Clarifications, and Open Item Mapping into the

Construction Authorization Request Revision. November 22, 2002.

(DCS, 2002f) Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory Commission,

RE: Mixed Oxide Fuel Fabrication Facility Construction Authorization Request Change Pages. 

December 20, 2002.

(DCS, 2003a) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Duke Cogema Stone & Webster Mixed Oxide Fuel Fabrication Facility

Criticality Validation Report—Criticality Validation Report—Revision 1 of Part I and Original Issue

of Part II. January 8, 2003.

(DCS, 2003b) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Responses to Site Description, Instrumentation, and Criticality Open

Items/Additional NRC Questions on Construction Authorization Request (CAR) Revision. 

February 11, 2003.

(DCS, 2003c) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Mixed Oxide (MOX) Fuel Fabrication Facility Construction Authorization

Request Change Pages. February 18, 2003.
6-57

(DCS, 2003d) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Mixed Oxide (MOX) Fuel Fabrication Facility Construction Authorization

Request Change Pages. April 10, 2003.

(DCS, 2003e) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Response to DSER Open Item NCS-04. June 13, 2003.

(DCS, 2003f) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Criticality Validation Report–Revision 3 of Part I, Revision 2 of Part II, and

Revision 1 of Part III, Response to DSER Open Item NCS-04. July 2, 2003.

(DCS, 2003g) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Response to Request for Additional Information—MFFF Criticality Validation

Report (DSER Open Item NCS-04). July 29, 2003.

(DCS, 2003h) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Criticality Validation Report—Revision 3 of Part II: Response to DSER Open

Item NCS-04. October 10, 2003.

(DCS, 2004a) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Mixed Oxide Fuel Fabrication Facility (MFFF), Response to Questions on the

Criticality Validation Report—Revision 3 of Part II. February 12, 2004.

(DCS, 2005) Ashe, K., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Criticality Review. February 28, 2005.

(Moseley, 1965) Moseley and Wing. “Properties of Plutonium Dioxide,” RFP-503. 1965.

(NRC, 1972) U.S. Nuclear Regulatory Commission. Regulatory Guide 3.71, “Nuclear Criticality

Safety Standards for Fuels and Material Facilities,” 1972.

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility.” August 2000.

(NRC, 2001a) Giitter, J.G., U.S. Nuclear Regulatory Commission. Letter to P.S. Hastings, Duke

Cogema Stone & Webster, RE: Mixed Oxide Fuel Fabrication Facility Construction

Authorization—Request for Additional Information. June 21, 2001.

(NRC, 2001b) Persinko, A., U.S. Nuclear Regulatory Commission. Letter to P.S. Hastings,

Duke Cogema Stone & Webster, RE: Duke Cogema Stone & Webster Quality Assurance

Program for Construction of the Mixed Oxide Fuel Fabrication Facility. October 1, 2001.

(NRC, 2001c) Persinko, A., U.S. Nuclear Regulatory Commission. Letter to P.S. Hastings,

Duke Cogema Stone & Webster, RE: Nuclear Criticality Safety Staff Qualifications and

Administrative Margins for Fuel Fabrication Facilities. November 9, 2001.
6-58

(NRC, 2003a) Persinko, A., U.S. Nuclear Regulatory Commission. Letter to P.S. Hastings,

Duke Cogema Stone & Webster, RE: Request for Additional Information—Mixed Oxide (MOX)

Fuel Fabrication Facility Nuclear Criticality Safety. June 25, 2003.

(NRC, 2003b) Persinko, A., U.S. Nuclear Regulatory Commission. Memorandum to J.G. Giitter,

U.S. Nuclear Regulatory Commission, RE: November 13, 2003, Meeting Summary: Meeting

with Duke Cogema Stone & Webster to Discuss Nuclear Criticality Safety Related to Mixed

Oxide Fuel Fabrication Facility Revised Construction Authorization Report. December 3, 2003.

(NRC, 2004a) Tripp, C.S., U.S. Nuclear Regulatory Commission. Memorandum to J.G. Giitter,

U.S. Nuclear Regulatory Commission, RE: December 17–19, 2003, In-Office Review Summary: 

Duke Cogema Stone & Webster Construction Authorization Request Supporting Documentation

for Criticality Code Validation. January 8, 2004.

(NRC, 2004b) Persinko, A., U.S. Nuclear Regulatory Commission. Letter to P.S. Hastings,

Duke Cogema Stone & Webster, RE: Mixed Oxide (MOX) Fuel Fabrication Facility Nuclear

Criticality Safety—Review of Validation Report. April 16, 2004.

(NRC, 2005) Brown, D.D., U.S. Nuclear Regulatory Commission. Memorandum to J.G. Giitter,

U.S. Nuclear Regulatory Commission, RE: February 1–2, 2005, In-Office Review Summary: 

Duke Cogema Stone & Webster Construction Authorization Request Supporting Documentation

for Criticality Code Validation. March 14, 2005.
7-1

7. FIRE PROTECTION

7.1 Conduct of Review

This chapter of the final safety evaluation report (FSER) contains the staff’s review of fire

protection issues described by the applicant, Duke Cogema Stone & Webster (DCS), in Chapter

7 of the revised construction authorization request (CAR). The objective of this review is to

determine whether the applicant’s commitments and goals related to fire protection are adequate

to meet the regulatory acceptance criteria referenced below. The review will also determine

whether the design of the proposed facility adequately protects against external and internal fires

and whether the design bases of the principal structures, systems, and components (PSSCs)

identified by the applicant adequately protect against natural phenomena and the consequences

of potential accidents. The staff evaluated the information provided by the applicant for fire

protection by reviewing Chapter 7 of the revised CAR, other sections of the revised CAR, and

supplementary information provided by the applicant. The review of fire protection was closely

coordinated with fire protection aspects of accident sequences described in the safety

assessment of the design basis (see Chapter 5 of this FSER), the review of explosion protection

aspects (see Chapter 8 of this FSER), and the review of other plant systems (see Chapter 11 of

this FSER). 

The staff reviewed how the fire protection information in the revised CAR addresses the

following regulations:

• Title 10, Section 70.23(b), of the Code of Federal Regulations (10 CFR 70.23(b))

requires, as a prerequisite to construction approval, that the design bases of the PSSCs

be found to provide reasonable assurance of protection against natural phenomena and

the consequences of potential accidents.

• 10 CFR 70.64, “Requirements for New Facilities or New Processes at Existing Facilities,”

requires that the baseline design criteria (BDC) and defense-in-depth practices are

incorporated into the design of new facilities. With respect to fire protection, 10 CFR

70.64(a)(3) requires that the facility design “provide for adequate protection against

fires.”

The review focused on the design basis of fire protection systems, their components, and other

related information. For fire protection systems, the staff reviewed information provided by the

applicant for the safety function, system description, and safety assessment. The review also

encompassed proposed design basis considerations, such as redundancy, independence,

reliability, and quality. The staff used Chapter 7 in NUREG-1718 (NRC, 2000) as guidance in

performing the review. 

Sections 7.1.1 through 7.1.4 of this FSER are the staff’s evaluation of how the applicant

addressed the fire protection acceptance criteria in NUREG-1718. Section 7.1.5 of this FSER is

the staff’s evaluation of the design bases PSSCs related to fire protection.
7-2

7.1.1 Organization and Conduct of Operations

The organization and conduct of operations are the management measures that assure that fire

safety is administered appropriately at a licensed facility. Section 7.1 of the revised CAR

describes the applicant’s commitment to assure that PSSCs, as identified in the revised CAR,

are available and reliable; fire protection organizational responsibilities are defined; transient

ignition sources and combustibles are controlled; and the facility maintains a readiness to

extinguish or limit the consequences of a fire. The applicant committed to develop a fire

protection program and administrative controls in order to meet organizational and operational

guidance of NUREG-1718. The conclusion of the review is that the applicant’s commitment to

provide fire protection organizational management, and to develop conduct of operations at the

facility site, meets the regulatory acceptance criteria and fire protection BDC of 10 CFR Part

70.64(a)(3) for the construction approval stage. Therefore, it is acceptable. The following

paragraphs provide more details on the applicant’s plans for a fire protection program and

administrative controls. 

7.1.1.1 Fire Protection Program

In Section 7.1.1 of the revised CAR, the applicant committed to develop a fire protection

program which establishes the fire protection policies for the facility site. The applicant stated

that the objectives of the Fire Protection Program are to prevent fires from starting and to detect,

control, and extinguish fires that do occur. The applicant will identify management and

organizational responsibilities for the Fire Protection Program. The fire protection organization

will develop prefire plans and fire safety training, procedures, and considerations. It will be

responsible for maintenance, inspection, and quality assurance of fire protection hardware;

design change control; recordkeeping; and administrative controls. An effective fire protection

program extends the concept of defense-in-depth to fire protection in areas with PSSCs.

7.1.1.2 Administrative Controls

Administrative controls establish procedures primarily for the storage, handling, and use of

combustible materials and ignition sources. In Section 7.1.2 of the revised CAR, the applicant

committed to develop procedures that govern the handling of transient combustibles in buildings

containing PSSCs, permit systems that control ignition sources, formal periodic surveillance

inspections, compensatory measures for detection and suppression systems, and emergency

action plans. The applicant committed to develop and implement testing, inspection, and

maintenance procedures for facility fire protection systems and equipment. Systems that are

designed and installed in accordance with National Fire Protection Association (NFPA) codes

and standards are maintained according to NFPA requirements. The staff finds that the

applicant’s commitment to maintain all fire protection equipment, as recommended by

NUREG-1718 (NRC, 2000), increases the overall availability and reliability of fire protection

equipment and the performance of fire protection personnel. The applicant designated some

administrative controls as fire safety PSSCs, which are discussed in Section 7.1.5 of the FSER.

7.1.2 Features and Systems

Plant fire protection features and systems include building construction, fire area determination,

electrical installation, ventilation, detection and alarm, and suppression. Section 7.4 of
7-3

NUREG-1718 provides acceptance criteria for fire protection features and systems. For the

reasons set forth below, the staff finds that the applicant’s proposed fire protection features and

systems meet the 10 CFR 70.64(a)(3) BDC for fire protection. 

7.1.2.1 Construction

Section 7.2.2 of the revised CAR states that buildings where radioactive materials would be

used, handled, or stored at the facility are Type I or Type II construction. Thus, the structural

members, including walls, columns, beams, girders, trusses, arches, floors, and roofs are of

approved noncombustible or limited-combustible materials and will have fire resistance ratings

as specified by NFPA 220, “Standard on Types of Building Construction,” issued 1995 (NFPA,

1995b). Buildings at the facility that contain PSSCs associated with hardware are Type I

construction. Type I construction requires that exterior bearing walls are rated at least 3 hours

and that interior bearing walls, trusses, beams, girders, and columns are rated at least 2 hours. 

This type of construction has the ability to remain structurally sound during anticipated fires. The

exterior walls of the mixed oxide (MOX) fuel fabrication building (BMF) structure carry a

minimum 2-hr fire rating in order to provide exposure protection. However, they would carry a

minimum 2-hr fire rating in order to provide protection from exterior exposure fires. Because of

the amount of depleted uranium it contains, the secured warehouse building (BSW) contains

PSSCs—combustible loading controls and facility worker actions. The BSW construction would

be classified as Type II. The staff concludes that the BSW does not need the fire resistance

rating of a Type I building because it does not contain hardware PSSCs that need to be

protected. In addition, buildings are protected from exterior fires by observing fire safety criteria

recommended in NFPA 80A, “Recommended Practice for Protection of Buildings from Exterior

Fire Exposures,” issued 1996 (NFPA, 1996e). By following guidelines found in NFPA 80A,

buildings and building components are able to withstand anticipated exterior fire events, such as

from ruptured fuel tank, trucks, other buildings, and forestry or wildland. Thus, the staff finds

that the preliminary construction features at the facility are adequate to meet the BDC of 10 CFR

70.64(a)(3) for fire safety. 

7.1.2.2 Interior Surface 

Section 7.2.2 of the revised CAR indicated that exposed interior walls or ceilings and any

factory-installed facing material would have a Factory Mutual-approved or Underwriters

Laboratories- (UL-) listed flame-spread rating of 25 or less, and a smoke-developed rating of 50

or less, in accordance with American Society for Testing and Materials (ASTM)-E-84, “Standard

Test Method for Surface Burning Characteristics of Building Material,” issued 1998 (ASTM,

1998). If carpets and rugs are used, they would be tested in accordance with NFPA 253,

“Standard Test Method for Critical Radiant Flux of Floor Covering Systems Using a Radiant Heat

Energy Source,” issued 1995 (NFPA, 1995a). Interior finish materials with flame-spread and

smoke-developed ratings, as stated, prevent fires from attaining full-room fire involvement. 

Thus, the staff finds that interior finish materials would not contribute to fire or smoke

development. 

7.1.2.3 Storage Racks

Revised CAR Section 7.2.3.6 states that racks for the storage of plutonium dioxide, uranium

dioxide, or MOX in powder, pellet, or rod form would be noncombustible. The applicant will
7-4

provide combustible loading controls to prevent the buildup of combustibles in areas where

storage racks would be located. Limiting combustible materials in areas where special nuclear

materials would be stored reduces the intensity of potential fires if they occur and minimizes

structural damage to the storage racks. Noncombustible storage racks do not contribute to the

fixed fire load; consequently, they provide stability to their contents during a fire. The staff finds

that the provisions for storage racks are acceptable because they meet the guidance of

NUREG-1718 (NRC, 2000).

7.1.2.4 Electrical Considerations

Section 11.5 of the revised CAR discusses the electrical systems proposed for use at the facility. 

In addition, the applicant provided details on fire and electrical safety interfaces during a briefing

to the Advisory Committee on Reactor Safeguards (NRC, 2002). To prevent fires from initiating

and to protect cables from fire exposure hazards, the electrical systems at the facility are

designed with the following considerations:

• Cables in redundant electrical trains are separated in accordance with Institute of

Electrical and Electronics Engineers (IEEE) 384-92, “Standard Criteria for Independence

of Class 1E Equipment and Circuits,” issued 1992 (IEEE, 1992a), electrical separation

criteria.

• The main electrical trains (for ventilation and for the emergency control rooms (ECRs))

enter the BMF from different sides of the building (approximately 45.7 m (150 linear ft

apart) but on the same floor. However, immediately after entering the building, they are

routed to separate floors of the MOX process (MP) area (DCS, 2002).

• Other cables, backup power supplies, and equipment feeds are kept as far apart as

practical and are routed separately to the extent practical.

• Cables supplying emergency electrical power to PSSCs are routed in conduit.

• Systems are designed so that electrical faults do not initiate fires. When the breaker trips

or the fuse blows, the electrical circuit is interrupted, which prevents damage to electrical

components and wiring or cable insulation. Faults that could result in the inability of the

system to perform its intended safety function are detected and removed.

• Electrical characteristics of equipment and/or power systems are continuously monitored

to assure adequate protection and operation.

• The number of trip devices that could shut down electrical systems that are designated

PSSCs is kept to a minimum.

• Exposed electrical cable is qualified under IEEE 383, “Standard for Type Test of Class

1E Electric Cables, Field Splices and Connections for Nuclear Power Generation

Stations,” issued 1992 (IEEE, 1992b), (i.e., ignition resistant and self-extinguishing).

• The normal and standby alternating current (ac) power systems are designed according

to NFPA 70, “National Electric Code,” issued 1999 (NFPA, 1999b).
7-5

• Where electrical cables enter glovebox or decanning rooms, they are installed in solid

stainless steel wireways or ducts (revised CAR Section 7.2.3.1).

The applicant’s proposed electrical features follow the applicable provisions of NFPA 70-1999,

IEEE 383-92, and IEEE 384-92 that provide reasonable assurance that electrical installations

are free from fire hazards. This strategy allows for redundancy and independence of electrical

wiring. As a result, automatic functioning of the detection, alarm, suppression, and barrier

systems would be maintained; electrical cables would not pose a source of fire hazard to other

systems; and cables are protected from anticipated fires. Thus, the staff finds that the protection

of electrical wiring from fire is acceptable and adequate.

7.1.2.5 Fire Alarm and Detection Systems

Section 7.2.3.2 of the revised CAR states that the fire alarm systems are designed according to

NFPA 72, “National Fire Alarm Code,” issued 1996 (NFPA, 1996f). The BMF detection systems

are tied into a central alarm panel that would be located in the aqueous polishing (AP) control

room. The central fire alarm panel has a graphical display which would assist the BMF

operators or the emergency response team in identifying and responding to alarms. Upon

detection of a fire, audible and visual alarms would be heard or seen in the affected parts of the

BMF.

The alarm systems are capable of annunciating and differentiating alarm or fire conditions,

supervisory indicators, or trouble signals. Alarm signals are transmitted to the monitored alarm

center at the U.S. Department of Energy Savannah River Site fire department and the AP control

room. Initiating circuits are capable of transmitting an alarm under circuit fault conditions of

single ground, open, or both. 

Heat and smoke detectors supplement or can actuate fire-extinguishing systems, fire dampers,

and door closure devices. Detection systems would be located throughout the facility in

accordance with the principles of NFPA 72 (revised CAR Section 7.2.3.2). Automatic fire

detection systems are selected on the basis of the fire hazard and their reliability in each area. 

Section 7.5.3 of the revised CAR states that detection systems in areas containing dispersable

radioactive materials are identified as PSSCs for defense-in-depth reasons (see FSER Section

7.1.5.7). Each glove box is provided with a minimum of two detectors. Generally, smoke

detectors would be deployed. However, heat detectors would be installed in gloveboxes that are

prone to dusty conditions.

The primary power supply for the fire detection/alarm system is the normal power system, which

has two sources of offsite ac power. In the event that both sources of normal power are lost, the

detection/alarm system can be powered by the standby ac power systems and then by normal

battery backup. The staff finds the design of the fire alarm and detection systems acceptable

because it reduces malfunctions and assures prompt detection and notification of fires and

trouble conditions.
7-6

7.1.2.6 Means of Egress Protection

The proposed facility layout largely complies with the 1997 version of NFPA 101, “Life Safety

Code,” issued 1997 (NFPA, 1997c). According to revised CAR Section 7.2.2, the means of

egress are arranged and maintained to provide free and unobstructed illuminated egress. 

Buildings at the facility are designed to provide means of egress that are adequate in number,

location, and capacity for emergencies.

[Text removed under 10 CFR 2.390]

[Text removed under 10 CFR 2.390]

Emergency lighting is provided for means of egress and for critical operations areas where

manual operations must be performed during an outage of normal ac power sources. Standby

generators support the emergency egress lighting. The proposed design calls for two standby

generators, each of which could operate continuously for 24 hours. The staff finds that the

preliminary emergency lighting provisions provide the necessary illumination in the event that

power to normal lighting is interrupted. The provisions for means of egress and emergency

lighting are adequate and acceptable.

7.1.2.7 Lightning Protection

The applicant designated the BMF and the emergency diesel generator (BEG) building as PSSC

structures. Revised CAR Section 7.2.2 states that PSSC structures are provided with lightning

protection in accordance with the applicable provisions of NFPA 780, “Lighting Protection Code,”

issued 1997 (NFPA, 1997a). Such lightning protection provides a means of directing a lightning

discharge to the earth without damage to the property. The staff finds that the protection

schemes adequately protect the PSSC structures from a lightning strike. Therefore, the

provisions for lightning protection are acceptable.

7.1.2.8 Ventilation System

The areas of the BMF that contain PSSCs are provided with ventilation systems that minimize

the spread of fire and the products of combustion. Section 11.4 of the revised CAR describes

the four confinement zones that the BMF would contain; the staff’s evaluation of this design

feature is provided in FSER Section 11.4. Essentially, pressure gradients between the

confinement zones ensure that leakage airflows from the zones of lowest contamination risk to

zones of increasing contamination risk. During a fire, the main objective is to maintain
7-7

differential pressure between the room of fire origin and the surrounding areas. The ventilation

systems operate continuously to ensure that combustion products flow through the exhaust

stacks of the gloveboxes, the process rooms, or the process cells. 

The BMF design incorporates airlocks that offer access to the process rooms. The airlocks are

separated from the process rooms by fire barriers and are exhausted by the secondary

confinement system. The airlocks have minimum-leakage doors and are maintained at a

positive pressure with respect to the process room, thereby reducing the spread of radioactive

contaminants and combustion products from the process room. Stairwells are pressurized with

respect to corridors, to minimize smoke infiltration during a fire in the MP or AP areas. The

applicant’s treatment of acceptance criteria for nuclear filter plenums is discussed below. The

staff‘s evaluation of the PSSC confinement systems is provided in Sections 11.4.1.3 and 7.1.5.5

of the FSER.

7.1.2.8.1 Filter Plenum Construction and Protection

The final filter plenum enclosures are 2-hr fire-rated construction with appropriate doors and are

located as far as practical from the postulated fires. The design calls for multiple separate filter

housings and redundant trains of filter systems in separate fire areas. No ignition sources and

only limited amounts of combustible materials would be permitted within the filter housings or

filter housing rooms. 

7.1.2.8.2 Ventilation Ductwork and Dampers 

The proposed ductwork in the ventilation systems incorporates manual and automatic dampers

and controls to distribute and regulate the movement of air. The ductwork would be welded

stainless steel or welded galvanized pipe. Additional fire-rated protective features are provided

when the ductwork passes through a fire barrier into another fire area. The fire resistance of the

dampers is suitable for the fire barriers they penetrate. Dampers are manually or automatically

operated depending on the stage of fire development and/or location of the fire. As a result, the

operation of the exhaust and supply dampers would be pre-planned.

7.1.2.8.3 Filter Design and Protection

At the BMF, high efficiency particulate air (HEPA) filters would be used to prevent the release of

radioactive materials. The HEPA filters meet the requirements of American Society of

Mechanical Engineers (ASME)-AG-1, “Code on Nuclear Air and Gas Treatment,” issued 1991

(ASME, 1991). The exhaust for the C3 and C4 confinement systems contains two sets of final

filtration units. Each final filtration unit consists of a filter assembly housing, a two-stage spark

arrestor, and two stages of HEPA filters. The spark arrestors are located upstream of the HEPA

filter exhaust plenum. The first-stage spark arrestor is made of stainless steel wire mesh; it is

designed to eliminate large, burning embers that may pass through the ductwork. The secondstage spark arrestor is noncombustible fiberglass and steel mesh designed to remove particles

larger than 1.0 micron. The first-stage spark arrestor is composed of stainless steel fiber, and

the second stage is composed of high-strength steel and glass fiber, while HEPA filters are

fabricated of glass media with metallic frames and silicone gaskets. The filter housing

assemblies are designed and fabricated to the same temperature ratings as the duct materials in

which they are installed. 
7-8

7.1.2.8.4 Protection of the Final HEPA Filters 

Fire protection of the final HEPA filter systems would be provided by heat detectors, spark

arrestors, and dilution of high-temperature exhaust streams. Exhaust systems are designed to

operate continuously so that hot gases generated from one fire area will be diluted with air from

unaffected fire areas in order to prevent prolonged exposure of the HEPA filters to temperatures

above their maximum service temperature. The HEPA filters are rated to 232 °C (450 °F). Heat

detectors located in the ductwork upstream of each final filtration unit would alarm in the event of

high temperatures. The applicant performed analyses to demonstrate that maximum

temperatures, soot, and pressures are not exceeded during credible fires (DCS, 2003b). In

addition, the preliminary results indicated that the HEPA filters at the MOX fuel fabrication facility

(MFFF) could perform their safety function under hypothetical severe heat and soot conditions. 

By the same letter, the applicant committed to provide demonstration (through testing) of the

final design’s function as part of the integrated safety analysis (ISA). Based on the applicant’s

design functions and the preliminary analyses and commitment, the staff concluded that the

ventilation system is adequate to meet the BDC of 10 CFR 70.64(a)(3) for fire safety. 

7.1.2.9 Fire Areas and Barriers

For facility design and operational purposes, the BMF is subdivided into fire areas. The

applicant used guidance from NFPA 801, “Standards for Facilities Handling Radioactive

Material,” issued 1998 (NFPA, 1998a), to determine fire area boundaries. Section 7.2.3.1 of the

revised CAR indicates that fire areas separate manufacturing from material storage areas,

control or computer rooms from adjacent areas, emergency generators from each other,

redundant trains of PSSCs as required, and electrical equipment or battery rooms from adjacent

areas. There are more than 350 fire areas at the BMF, including 169 in the MP, 128 in the AP,

and 56 in the shipping and receiving areas. Fire areas are separated by noncombustible,

reinforced concrete walls with a minimum 2-hr fire rating; 3-hr fire barriers separate some

hazardous areas. Hourly ratings are based on ASTM E-119, “Standard Test Methods for Fire

Test of Building Construction and Materials,” issued 1995 (ASTM, 1995), definitions. 

Construction details are in accordance with NFPA 221, “Fire Walls and Fire Barrier Walls,”

issued 1997 (NFPA, 1997b). 

Openings in fire barriers would have appropriately rated self-closing fire doors, fire dampers, and

fire-rated penetration seals. Fire doors are designed according to NFPA 80, “Standard for Fire

Doors and Fire Windows,” issued 1999 (NFPA, 1999a). The staff finds that the general

selection of fire areas is appropriate for the hazards at the BMF. The fire area selection

minimizes the potential size of a fire. Therefore, it potentially limits the fire exposure to PSSC

equipment, material, and personnel. The applicant designated BMF and BEG fire barriers as

PSSCs. Because of the quantity and combustibility of fuels found in the BMF, the applicant

performed zone fire modeling and one-dimensional finite difference analysis to determine heat

transfer to the fire barriers (NRC, 2003b; DCS, 2003c). The analyses indicate that, in most

cases, the room temperatures would not exceed the ASTM E-119-95 temperature profile (to

which the barriers would be rated). In scenarios where the predicted peak room temperature

exceeded the ASTM E-119-95 peak temperature, the applicant demonstrated that the transient

excessive temperatures would not lead to barrier failure caused by thermal shock, thermal

stress, or spalling. The staff finds that the selection of the fire barriers provides an adequate
7-9

margin of safety against potential flashover and/or rapid temperature rise. Thus, the design

function of the barriers meets the BDC of 10 CFR 70.64(a)(3) for fire safety.

7.1.2.10 Water Supply and Drainage 

The facility design incorporates a water-supply system in accordance with NFPA 801 (NFPA,

1998a) requirements. Section 7.2.3.4 of the revised CAR states that the fire protection water

supply system consists of an underground loop around the facility site, fire hydrants, fire pumps,

and two firewater storage tanks. One of the storage tanks is seismically qualified. The water

supply system is designed to handle the largest sprinkler demand plus 500 gpm (1893 L/min) for

hose streams. The distribution systems are hydraulically designed to meet NFPA 13, “Standard

for the Installation of Sprinkler Systems,” issued 1996 (NFPA, 1996k); NFPA 14, “Standard for

the Installation of Standpipes and Hose Systems,” issued 1996 (NFPA, 1996i); and NFPA 24,

“Standard for the Installation of Private Service Mains and their Appurtenances,” issued 1995

(NFPA, 1995d).

[Text removed under 10 CFR 2.390]

The staff determined that the facility water supply system accommodates the requirements for

automatic and manual suppression activities at the facility. The provisions for water capacity are

based on defense-in-depth and industry practices; therefore, they are adequate.

7.1.2.11 Fire Suppression

A combination of automatic suppression systems, fire hose stations, exterior hydrants, and

manual extinguishing devices would provide fire suppression at the facility. Automatic

suppression is provided in areas where potentially significant fire loading is present. 

Because of criticality concerns, water-based fire suppression systems are not planned in areas

where fissile material would be handled (revised CAR Section 7.2.3.3.1). Where fissile materials

would normally be present, clean-agent suppression systems will be provided. In such cases,

clean-agent suppression is provided as a defense-in-depth PSSC to the fire barriers. The staff

evaluated clean-agent suppression as a PSSC in FSER Section 7.1.5.7. The clean-agent

system will also be provided in areas of the facility that contain electrical and/or electronic

equipment, such as computer rooms, motor control centers, and control rooms. Where needed,

the clean-agent system would protect under floor areas as well. All clean-agent systems will be

designed, installed, and maintained according to NFPA 2001, “Standard on Clean Agent

Extinguishing Systems,” issued 1996 (NFPA, 1996a).

Otherwise, water-based suppression is provided throughout the facility. To avoid possible

ingress of water into areas where fissile material is handled, pre-action-type sprinklers are used. 

Pre-action systems reduce the chance of accidental discharge by requiring independent actions

for water discharge. The secured warehouse and reagent processing buildings would deploy

wet-pipe sprinkler systems (discharge water when elevated temperatures are detected). The

truck bays of the shipping and receiving building would have an automatic deluge system (all

sprinklers activate at once). Sprinklers are designed according to NFPA 13-1996. All waterbased systems will be periodically inspected, tested, and maintained in accordance with NFPA
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25, “Standard for the Inspection, Testing, and Maintenance of Water-Based Fire Protection

Systems,” issued 1998 (NFPA, 1998c).

Not all fire areas in the BMF would have automatic suppression. Suppression is not provided in

some airlocks, solvent cells, plenums, and chases, as well as areas that are not normally

occupied, have low combustible loading, or have no ignition sources. Plenums, chases, and

solvent cells are difficult to inspect on a routine basis. Similarly, automatic suppression features

are not provided in areas such as rod storage, where there are minimal ignition sources and

combustible loading.

The standpipe and hose systems allow manual firefighting capabilities throughout the proposed

facility. Because of criticality concerns, a dry-standpipe system, instead of the normally

pressurized wet-standpipe system, is provided in the MP, AP, and shipping and receiving areas.

The standpipe systems are designed for Class II service, so that both fire department and fire

brigade personnel can use them. The standpipe system is designed per NFPA 14-1996.

Portable fire extinguishers would be present throughout the BMF and inside all buildings at the

facility so occupants could extinguish small fires. Extinguishers are selected and located

according to fire hazards and to their effectiveness. A combination of multipurpose dry

chemical, metal-use, and carbon dioxide extinguishers would be used. The use of portable

extinguishers would be in accordance with NFPA 10, “Standard for Portable Fire Extinguishers,”

issued 1998 (NFPA, 1998e). Specially configured portable carbon dioxide bottles would be

present in rooms with gloveboxes. These portable bottles can be quickly connected to the

glovebox to suppress fires within without overpressurizing the glovebox.

The design calls for packets of extinguishing powder to be present in gloveboxes and in the

rooms where a zircaloy fire risk exists. In addition, zirconium swarfs are collected in covered

metal containers and removed daily, stored, and disposed.

Based on the variety and redundancy of suppression features, the applicant did not place total

reliance on a single fire-suppression feature at the BMF. Standpipe and hose systems (which

both facility-trained personnel and the fire department can use) provide backup fire suppression

to the automatic suppression systems. An additional backup suppression is the specially

configured portable carbon dioxide bottles with quick disconnect fittings. According to revised

CAR Section 7.3, the facility fire brigade would respond to firefighting emergencies. 

Development and training of the facility fire brigade would follow the requirements of NFPA 600,

“Standard on Industrial Fire Brigades,”issued 1996 (NFPA, 1996b).

Based on information provided to date, the staff finds that the fire suppression strategy provides

diversity and defense-in-depth and is therefore acceptable. Automatic and manual fire

suppression controls the spread of fire, thereby reducing the challenge to fire barriers. 

7.1.2.12 Combustible and Pyrophoric Metals

When plutonium dioxide is fully oxidized, it is not pyrophoric. Uranium dioxide does not oxidize

in an inert atmosphere and oxidizes very slowly in air under process temperature conditions.

DCS maintains that uranium oxide is not combustible under process conditions (DCS, 2000, p.

45). However, U.S. Nuclear Regulatory Commission (NRC) Information Notice 92-14, “Uranium
7-11

Oxide Fires at Fuel Cycle Facilities,” issued February 1992 (NRC, 1992), indicates that normally

stable uranium dioxide may be pyrophoric or oxidize rapidly even at room temperatures when in

very fine powder form. Section 7.2.5 of the revised CAR indicates that, for process reasons,

some gloveboxes containing plutonium and uranium oxides are inerted with nitrogen. Thus, the

nitrogen systems provide additional fire protection as they limit the potential combustion of

special nuclear materials and ordinary combustibles. The pyrophoricity of plutonium and

uranium dioxides is discussed in FSER Section 11.3.

Zirconium is stored and handled in accordance with NFPA 482, “Production, Processing,

Handling and Storage of Zirconium,” issued 1996 (NFPA, 1996c). The applicant intends to use

titanium for the electrolyzer circuit and associated equipment that could be exposed to silver(II)

ions. Finely divided or powdered forms and thin metal sections of titanium exposed to hot

sparks can ignite and burn. The staff’s evaluation of titanium fires is discussed in FSER Section

11.2. 

7.1.2.13 Glovebox Protection

Gloveboxes provide physical and visual access to internal equipment, processes, and material. 

Glovebox details are given in revised CAR Section 11.4.7.1 and the applicant’s polycarbonate

report (DCS, 2000). A typical glovebox is a large, stainless steel enclosure mounted on a

structural stainless steel stand. Glovebox windows consist of rectangular polycarbonate panels

(10-mm thick) that fit into frames in the glovebox walls and ceilings. Lead-impregnated polymer

sheets or lead glass panels overlay windows where radiation shielding is required to reduce

operator exposures. 

Light fixtures are generally installed outside of the gloveboxes, and they provide illumination for

the interior spaces through windows located in glovebox ceilings. The lenses of the fluorescent

light fixtures are noncombustible.

The design of the proposed facility calls for each glovebox to be provided with a minimum of two

detectors. Generally, smoke detectors would be deployed. However, if conditions such as dust

limit their effectiveness, heat detectors would be installed. Actuation of the detectors transmits

an alarm to the control room, operations utilizing hydraulic fluids are automatically shut down,

the glovebox process fire doors are actuated to close, and operators can manually close the

glovebox dampers if warranted. 

The proposed facility design provides additional fire protection features in the process rooms

containing gloveboxes, such as manual-injection carbon dioxide for each glovebox, inert

atmospheres for some gloveboxes with dispersable materials, fire doors between interconnected

process gloveboxes, and minimal ignition sources near glovebox windows. Clean-agent

automatic suppression is also provided and designated a defense-in-depth PSSC (see FSER

Section 7.1.5.7). 

Fire areas are separated by fire barriers, which help to prevent fires from spreading throughout

the proposed facility. The BMF fire barriers are rated a minimum of 2 hours and are designated

PSSCs. All associated dampers and doors are fire rated for use in the barriers. To further

reduce the fire risk, combustible loading controls are provided as a PSSC for areas with
7-12

gloveboxes that store radiological materials. This PSSC limits the amount of fixed and transient

combustibles by design and through the use of administrative controls. 

The applicant’s proposed use of polycarbonate is not in compliance with the requirements of

NFPA 801 (NFPA, 1998a) requirements to use noncombustible materials. However, the

applicant demonstrated that an acceptable level of fire protection is achievable with the use of

polycarbonate. Compared to other plastic glovebox materials, polycarbonate is relatively difficult

to sustain combustion or ignite (its reported flashpoint is 466 °C (871 °F), and the spontaneous

ignition temperature is 577 °C (1071 °F), as obtained by the test method described in ASTM D1929, “Standard Test Method for Determining Ignition Temperature of Plastics,” issued 1986

(ASTM, 1986)).

The ISA and ISA summary to be prepared by DCS in support of its later application for a license

to possess and use licensed material will evaluate whether the range of the reported properties

for the polycarbonate are bounding for its expected use or conditions. In addition, the fire

hazard analysis (FHA) will account for the polycarbonate fuel. Revised CAR Section 7.4 states

that the design-basis fire scenarios will include an evaluation of the types of potential fires based

on the combustible form and quantity, combustible type (including polycarbonate), and the fire

severity, intensity, and duration. Based on the applicant’s commitments to provide further

evaluations in support of its later application and the information in the DCS polycarbonate

report (DCS, 2000), the NRC considers polycarbonate to be a potential candidate material for

use in glovebox window panels.

7.1.2.14 Flammable and Combustible Liquids and Gases

Chapter 8 of the revised CAR identifies facility fire areas with the potential for large spills of

flammable or combustible liquids. Section 7.2.2 of the revised CAR states that flammable and

combustible liquids would be stored and handled in accordance with NFPA 30, Flammable and

Combustible Liquids Code,” issued 1996 (NFPA, 1996h). Means for containing spills, such as

dikes, and for drainage systems are provided for in accordance with NFPA 30. Most vessels

containing flammable or combustible liquids would be placed in process cells. Process cells

would have fire prevention features, such as eliminating ignition sources, earth grounding of

vessels and pipes, and combustible loading controls (see FSER Section 7.1.5.3). The

purification and solvent recovery cycles involve a solvent-diluent mixture, which is combustible. 

In addition to the PSSC process cell fire prevention features, the facility design incorporates the

following additional measures to limit fires: 

• Welded equipment and pipes are connected to the offgas treatment system to preclude

overpressurization.

• Combustible chemicals are stored in welded tanks or equipment to avoid leakage.

• Chemicals are supplied through leak-tight welded lines.

• Vapor phase concentration of flammable constituents or mixtures below limits is

maintained, as specified in NFPA 69, “Standard on Explosion Prevention Systems,”

issued 1997 (NFPA, 1997d).
7-13

• A minimum clearance is maintained from the tank bottoms for feed lines to preclude

splashing or vapor formation.

• Offgases from process equipment are collected and adequately treated.

The NRC concludes that these additional protective features minimize the amount of fuel and

consequently the potential heat output of a fire initiating in the process cell. Therefore, these

features provide defense-in-depth for fire protection and reduce challenges to the process cell

PSSC.

Flammable and combustible gases would be stored and handled in accordance with NFPA 50A,

“Standard for Gaseous Hydrogen Systems at Consumer Sites,” issued 1999 (NFPA, 1999c), and

NFPA 55, “Standard for Portable Fire Extinguishers,” issued 1998 (NFPA, 1998b). Flammable

gas generation, delivery systems, and protection are evaluated in FSER Section 8.1.2.

The sintering furnaces in the MP areas would use hydrogen/argon atmospheres, while the

calcining furnace in the AP area would use an air/oxygen atmosphere. The fire safety measures

for the sintering furnaces, such as inert-gas purge and automatic shutoffs, are in accordance

with NFPA 86C, “Industrial Furnaces Using a Special Processing Atmosphere,” issued 1995

(NFPA, 1995c). Concerns regarding the adequacy of the overall safety provisions for the

furnaces are discussed in FSER Section 11.3.

7.1.2.15 Special Hazards

The BMF battery rooms would be separated from other areas by 3-hr fire walls and have

automatic suppression and provisions for ventilation and hydrogen gas detection. The

ventilation rate is a minimum of two air changes per hour, and the potential hydrogen

accumulation is limited to less than 2 percent of room volume. These requirements are

adequate according to NFPA 111, “Stored Electrical Energy Emergency and Standby Power

Systems,” issued 1996 (NFPA, 1996d), and IEEE 484-1996, “Recommended Practice for

Installation Design and Installation of Vented-Lead Acid Batteries for Stationary Applications,”

issued 1996 (IEEE, 1996). 

There would be several laboratories in the BMF to provide physical and chemical analyses of

samples from the MP or AP process. Fire protection for laboratories is designed to meet NFPA

45, “Standard for Fire Protection for Laboratories Using Chemicals,” issued 1996 (NFPA,

1996g). The isolation of these special hazards reduces the potential fire damage, while escape

routes are safeguarded. Therefore, they enhance safety by reducing challenges to the PSSCs. 

7.1.3 Manual Firefighting Capability

The applicant performed a baseline needs assessment that determined the minimum necessary

capabilities of the facility firefighting forces. The assessment evaluated the minimum staffing for

the firefighting forces, organization and coordination of onsite and offsite resources, personal

protective and firefighting equipment, and training and emergency planning. The assessment

determined that an onsite emergency response team is required at the facility (revised CAR

Section 7.3). The facility fire brigade will be developed in accordance with NFPA 600 (NFPA,
7-14

1996b). The staff finds that these plans for manual firefighting are adequate for approving

construction of the facility.

7.1.4 Preliminary Fire Hazard Analysis

The preliminary FHA for the facility consists of a systematic analysis of the fire hazards, an

identification of fire areas, and an evaluation of anticipated consequences given the features

proposed to control the hazards.

The applicant provided a summary of the methodology and assumptions of the preliminary FHA

in Section 7.4.1 of the revised CAR. The preliminary FHA concludes that potential fires are

typically small and will not propagate beyond the fire area of origin, and that fire barriers could

prevent facility-wide fire spread. Furthermore, programs to maintain the fire barriers and fire

detection/suppression features increase the reliability of the fire barriers. The staff concludes

that the preliminary FHA begins the process of documenting the adequacy of facility fire safety. 

It meets the regulatory acceptance criteria and fire protection BDC of 10 CFR Part 70.64(a)(3)

for the construction authorization stage. Based on the applicant’s commitment to perform an

FHA, staff finds that the preliminary FHA is acceptable. 

7.1.5 Design Bases of the PSSCs

In revised CAR Sections 5.5.2.2.6, 7.5.3, and 11.4.11, the applicant identified the design bases

and safety functions of the PSSCs that are intended to provide adequate protection against fires. 

Pursuant to 10 CFR 70.23(b), this FSER documents the staff’s determination whether those

design bases provide reasonable assurance of protection against natural phenomena and the

consequences of potential accidents.

7.1.5.1 Combustible Loading Controls

Revised CAR Tables 5.5-13b, 5.5-13a, and 5.5-14 identify combustible loading controls as a fire

event group PSSC to protect the following:

• confinement barriers in the C1, C2, and C3 areas

• radiological material storage gloveboxes

• the pneumatic transfer system

• the secured warehouse

The specific confinement barriers include 3013 canisters, 3013 transport casks, fuel rods, MOX

fuel transport casks, transfer containers, and the final C4 HEPA filters. Combustible loading

controls are identified by the applicant as a PSSC for gloveboxes that store radiological material. 

For example, combustible loading controls would be provided for the sintered and green pellet

glovebox stores and the plutonium dioxide buffer storage area.

[Text removed under 10 CFR 2.390]

The combustible-loading control PSSC limits the quantity of combustibles in specific fire areas in

order to ensure that the various confinement barriers, the facility-wide system, and large stores
7-15

of radiological material would not be adversely impacted by fire. In revised CAR Section 5.6.2.2,

the applicant identified combustible loading controls to include control of the fixed combustible

loads by design and control of transient combustibles by design and during operations. 

Transient combustible controls can be achieved through worker training, regular surveillance,

and postings. Even when engineered features are relied on to mitigate potential fires,

combustible loading controls are essential to limit the maximum fire growth rate. In addition to

combustible loading controls, the applicant identified additional protective features for fire areas

that contain the various plutonium confinement barriers. They are (1) ignition source controls,

(2) training of facility workers in manual suppression of incipient stage fires, (3) automatic fire

detection in all areas having confinement barriers, (4) automatic fire suppression in some areas,

and (5) the passive boundary of the C2 confinement system.

The design bases for the combustible loading controls identified by the applicant as PSSCs are

the guidelines within NFPA 801 (NFPA, 1998a). The administrative controls in NFPA 801 limit

the unnecessary accumulation of combustibles in facilities handling radioactive materials. Based

on industry experience, these controls restrict the occurrence and development of fires. Thus,

credible fires do not significantly impact primary confinement barriers. The staff concludes that

PSSC combustible loading controls and their associated design bases are adequate because

they help to ensure that structures, systems, and components are not adversely impacted by

fire.

7.1.5.2 Confinement Barriers

The applicant designated 3013 transport cask and MOX fuel transport cask confinement barriers

as PSSCs. The function of transport cask confinement barriers is to withstand the design basis

fire without breaching. The transportation casks are designed to meet the requirements of

10 CFR 71.73, “Hypothetical Accident Conditions.”

[Text removed under 10 CFR 2.390]

However, because of the combustible-loading controls PSSC and the additional protection

provided by automatic suppression

[Text removed under 10 CFR 2.390]

 The staff concludes

that the design bases for the confinement barriers of the MOX transport cask and the 3013

transport cask help to provide reasonable assurance of protection against the consequences of

fires.

7.1.5.3 AP Process Cell Fire Prevention Features

One objective of a fire protection program is to prevent fires from starting. However, fires are

postulated to occur in AP process cells because of the presence of solvents and other chemicals

with low flash points. The applicant designated the AP process cell fire prevention features as

PSSCs to ensure that fires in the process cells are highly unlikely. The safety strategy utilizes

the following ignition source controls:
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• elimination of ignition sources (including electrical equipment) within the AP process cells

• earth grounding of vessels and pipes to avoid static electricity

• the use of controls to ensure that potential chemical reactions that may result in a fire are

made highly unlikely

• limiting combustible materials and the processes required combustible products

In addition, the following are provided:

• ventilation to prevent accumulations of vapors in the flammable range

• fire barriers rated at 2 hours or more (evaluated in FSER Section 7.1.5.6) to isolate

process cells from each other and limit the effect of external fires

• all materials at risk contained in sealed vessels and pipes and isolated from any hazards

inside the process cell

• lightning protection

Furthermore, the applicant identified procedures that apply to process cells:

• permit systems to control ignition sources 

• procedures to prohibit leak testing using open flames or combustion-generated smoke.

The design bases for AP process cell fire prevention are practices promoted in NFPA 801

(NFPA, 1998a) and NFPA 30 (NFPA, 1996h), which are based on industry experience, and rely

on the use of fundamental safety measures and controls regarding storage, exposure, ignition

sources, combustibles, and separation. The staff concludes that the design bases for AP fire

prevention features provide reasonable protection against fire initiation and development.

7.1.5.4 MOX Fuel Fabrication Structures

[Text removed under 10 CFR 2.390]
7-17

[Text removed under 10 CFR 2.390]

7.1.5.5 C3 and C4 Confinement Systems

The applicant identified the C4 (primary) and C3 (secondary) systems (revised CAR Section

11.4) as PSSCs that are designed to limit the release of radioactive materials to the

environment. These systems are designed to remain operational during a design basis fire to

ensure that any potential releases caused by a fire are filtered. Additionally, the design function

of the C4 confinement system is to ensure that the final C4 HEPA filters are not impacted by fire. 

The design basis is Regulatory Guide 3.12, “General Design Guide for Ventilation Systems of

Plutonium Processing and Fuel Fabrication Plants,” issued August 1973 (NRC, 1973), which

allows for the cooling of hot gas from a fire by mixing it with air from rooms that are not involved

in the fire. However, if the fire is large enough to exceed set thresholds in the final filtration units

or exhaust, the C3 and/or C4 systems may be isolated.

The HEPA filters are fabricated of glass media with metallic frames and silicone gaskets. The

continuous service rating is 232 °C (450 °F). Preliminary analyses indicate that the exhaust flow

dilution would reduce extremely high room temperatures of 1260 °C (2300 °F) to less than 204

°C (400 °F) in the final filter plenum. Thus, the final C3 and C4 HEPA filters are not expected to

be affected by the maximum temperatures anticipated to result from credible fires within the

BMF. Additionally, preliminary analyses indicated that the resulting soot and pressure would not

exceed the capacity of the final HEPA filters.

The applicant committed to provide demonstration of the final design function (through testing)

as part of the ISA. The ability of the final filtration units to withstand potential fires is described in

FSER Section 11.4.1.3. The staff concludes that the ability of the final C4 and C3 HEPA filters

to perform their safety function has been adequately demonstrated. Thus, the HEPA filters

provide reasonable protection against the release of radioactive material during a fire.
7-18

7.1.5.6 Fire Barriers

The BMF fire barrier systems (including walls, doors, fire dampers, and penetration seals) are

PSSC that prevent fires from spreading from one fire area to another. In addition, the applicant

has identified fire barriers as PSSCs to ensure that facility-wide systems, such as pneumatic

transfer tubes that contain or handle radioactive materials, do not spread fire. The design basis

for fire barriers is NFPA 801 (NFPA, 1998a), NFPA 221 (NFPA, 1997b), NFPA 80 (NFPA,

1999a), and UL 555, “Standard for Fire Dampers and Ceiling Dampers,” issued 1995 (UL, 1995).

To minimize the potential for fire to spread between fire areas, the applicant implemented the

following measures:

• Fire barriers within the MOX BMF are designated PSSCs.

• Detection and suppression systems in areas with significant quantities of dispersable

radioactive material are designated defense-in-depth PSSCs to provide a diverse means

of mitigating risks.

• Buildings containing PSSCs are designed as Type I construction. In addition, fire

barriers in those structures are rated a minimum of 2 hours.

• Fire barriers are noncombustible, reinforced concrete.

• Fire doors between fire areas are normally closed or self-closing.

• Barriers have adequate penetration seals; a penetration seal tracking program will record

pertinent information regarding the emplacement and modification of fire barrier

penetration seals. 

• Airlocks are provided at the doors between secondary and tertiary confinement areas. 

The airlock is maintained at a positive pressure with respect to the secondary

confinement area.

• Fire doors on gloveboxes are normally closed or close upon fire detection. Automatic

closure is inhibited momentarily during operations when the collision zone is not clear.

• Transfer gloveboxes that pass through a fire barrier are equipped with fire doors that are

rated at 2 hours.

• Fire dampers to the supply and exhaust systems are manually or automatically closed,

as necessary to maintain the effectiveness of fire barriers.

• Combustible-loading controls are used as a PSSC to reduce the potential spread of hot

gases through pneumatic transfer tubes (see FSER Section 7.1.5.1).

The applicant evaluated fire severity based on the maximum quantities of combustibles

anticipated in each fire area. The quantities and types of combustibles and the ignition sources

were based on those found in comparable spaces in the MELOX facility. The estimate of the
7-19

combustible quantity included transient fuels (revised CAR Section 7.4.1.3). After accounting for

all combustibles within a fire area and using reasonable estimates of fire growth, the applicant’s

calculations show that, in a few cases, the room temperatures could exceed temperatures that

the barriers are qualified to withstand. Because of the type of fire growth rate assumed for the

combustibles, the temperature profile results in steeper temperatures at the early stages of fire

growth (DCS, 2003b). However, in no case is the peak temperature of the ASTM-E119 (ASTM,

1995) fire curve (to which the fire barriers would be rated) exceeded. The applicant reevaluated

those scenarios with the early rapid temperature rise and determined (using finite difference

analysis) that the first 2 inches of the concrete cover on the fire barriers would not exceed 93 °C

(200 °F) when exposed to the ASTM E-119 time-temperature curve. The temperature

necessary to cause spalling of concrete is 93 °C (200 °F). Thus, the analysis concludes that the

fire barrier members would not fail because of thermal shock or stress (NRC, 2003b). In

addition, the applicant will demonstrate fire barrier performance under credible fire conditions

during the ISA. While fire barrier failure is not expected, if the ISA shows that such failure is

credible, appropriate items relied on for safety (IROFS) would be identified (DCS, 2003a). 

Further, as part of its ISA, the applicant committed to evaluate the impact of hot gas transport

within the transfer tubes and to identify isolation valves as IROFS if warranted by the ISA results. 

For the reasons discussed above, the staff finds that designating the fire barriers as PSSCs as

part of the MFFF design provides adequate protection against internal fires. The applicant’s

commitments to perform further analyses as part of its ISA and to identify appropriate IROFS if

warranted constitute an acceptable fire safety strategy.

7.1.5.7 Suppression and Detection

Sections 7.4.2 and 7.5.3 of the revised CAR state that, where dispersable fissile material is

present, fire detection and suppression is provided as a PSSC. The PSSC

detection/suppression systems further ensure that a propagating fire cannot result in an

unacceptable release of radioactive material to the environment and, therefore, provide defensein-depth. Because of criticality concerns when using water, clean-agent suppression is provided.

However, clean-agent suppression is not credited in the safety assessment or the preliminary

FHA. According to revised CAR Section 7.5.3, the PSSC clean-agent suppression systems are

designed to be operable after an earthquake. Therefore, if a postseismic fire were to occur in

fire areas containing radioactive materials, postseismic firefighting capability will be available. All

clean-agent systems will be designed, installed, and maintained according to NFPA 2001 (NRC,

1996a). To assure the reliability of clean-agent suppression systems, the effect of ventilation

and agent soak time will be considered in their design. They will include 100-percent reserve. 

The reserve is not automatically connected. Instead, the facility fire brigade manually connects

the reserve capacity when needed. 

As discussed in FSER Section 7.1.2.5, smoke detection systems would be present in all areas of

the BMF. Fire detection and alarm systems are designed according to NFPA 72 (NFPA, 1996f). 

Detection systems associated with automatic suppression in fire areas containing significant

quantities of dispersable radioactive materials are designated as PSSCs. Detection and cleanagent suppression systems allow initiating fires to be quickly detected and controlled. As

PSSCs, they reduce the impact of fire so that fire barriers can be more effective. Thus, the staff

concludes that fire detection/suppression systems provide defense-in-depth to the fire barriers to

help assure protection against fires.
7-20

7.1.5.8 Emergency Control Room Air-Conditioning System

To protect operators against external fires, the ECR air-conditioning system (ECRAS) is

identified as a PSSC. The ECRAS is a collection of heating, ventilation, and air-conditioning

components located within the shipping and receiving area. The safety function of the ECRAS is

to ensure that conditions are habitable for the control room operators. According to revised CAR

Section 11.4.2.7.2, the location of control room fresh air inlets minimizes the presence of

contaminants. Section 11.4.1.3 of the FSER provides more detail on the ECRAS. 

7.2 Evaluation Findings

In Section 7.5 of the revised CAR, the applicant provided design-basis information for the fire

protection systems that were identified as PSSCs for the proposed MOX facility. Based on the

staff’s review of the revised CAR and supporting information provided by the applicant relevant

to fire protection systems, the staff finds that DCS has met the BDC set forth in

10 CFR 70.64(a)(3) for fire safety. Further, the staff concludes, pursuant to 10 CFR 70.23(b),

that the design bases of the PSSCs evaluated in this FSER will provide reasonable assurance of

protection against natural phenomena and the consequences of potential accidents. 

7.3 References

(ASME, 1991) American Society of Mechanical Engineers. ASME AG-1, “Code on Nuclear Air

and Gas Treatment.” 1991.

(ASTM, 1986) American Society for Testing and Materials. ASTM D-1929, “Standard Test

Method for Determining Ignition Temperature of Plastics.” 1986.

(ASTM, 1995) American Society for Testing and Materials. ASTM E-119, “Standard Test

Methods for Fire Test of Building Construction and Materials.” 1995.

(ASTM, 1998) American Society for Testing and Materials. ASTM E-84, “Standard Test Method

for Surface Burning Characteristics of Building Material.” 1998.

(DCS, 2000) Hastings P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Choice of MFFF Process Glovebox Window Material (DCS-NRC-000030).

December 15, 2000.

(DCS, 2002) Hastings P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Evaluation of the Draft Safety Evaluation Report (DSER) on Construction of a

Mixed Oxide Fuel Fabrication Facility. July 9, 2002.

(DCS, 2003a) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Responses to Financial Qualification, Fire Safety, Chemical Safety, Aqueous

Processing, Material Processing and Ventilation Open Items/Additional NRC Questions on

Construction Authorization Request (CAR) Revision. February 18, 2003.

(DCS, 2003b) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Construction Authorization Request Change Pages. April 10, 2003.
7-21

(DCS, 2003c) Hastings, P., Duke Cogema Stone & Webster. Letter to the U.S. Nuclear

Regulatory Commission, RE: Response to DSER Open Item FS-2, DCS-NRC-000134. May 14,

2003.

(DOE, 1997) U.S. Department of Energy. DOE-STD-1066-97, “Fire Protection Design Criteria.” 

Washington, DC, March 1997.

(IEEE, 1992a) Institute of Electrical and Electronics Engineers, Inc. Standard 384-92,

“Standard Criteria for Independence of Class 1E Equipment and Circuits.” 1992.

(IEEE, 1992b) Institute of Electrical and Electronics Engineers, Inc. IEEE Standard 383,

“Standard for Type Test of Class 1E Electric Cables, Field Splices and Connections for Nuclear

Power Generation Stations.” 1992.

(IEEE, 1996) Institute of Electrical and Electronics Engineers, Inc. Standard 484-1996,

“Recommended Practice for Installation Design and Installation of Vented-Lead Acid Batteries

for Stationary Applications.” 1996.

(NFPA, 1995a) National Fire Protection Association. Standard 253, “Standard Test Method for

Critical Radiant Flux of Floor Covering Systems Using a Radiant Heat Energy Source.” 1995.

(NFPA, 1995b) National Fire Protection Association. Standard 220, “Standard on Types of

Building Construction.” 1995.

(NFPA, 1995c) National Fire Protection Association. Standard 86C, “Industrial Furnaces Using

a Special Processing Atmosphere.” 1995.

(NFPA, 1995d) National Fire Protection Association. Standard 24, “Standard for the Installation

of Private Service Mains and their Appurtenances.” 1995.

(NFPA, 1996a) National Fire Protection Association. Standard 2001, “Standard on Clean Agent

Extinguishing Systems.” 1996.

(NFPA, 1996b) National Fire Protection Association. Standard 600, “Standard on Industrial Fire

Brigades.” 1996.

(NFPA, 1996c) National Fire Protection Association. Standard 482, “Production, Processing,

Handling and Storage of Zirconium.” 1996.

(NFPA, 1996d) National Fire Protection Association. Standard 111, “Stored Electrical Energy

Emergency and Standby Power Systems.” 1996.

(NFPA, 1996e) National Fire Protection Association. Standard 80A, “Recommended Practice

for Protection of Buildings from Exterior Fire Exposures.” 1996.

(NFPA, 1996f) National Fire Protection Association. Standard 72, “National Fire Alarm Code.” 

1996.
7-22

(NFPA, 1996g) National Fire Protection Association. Standard 45, “Standard for Fire Protection

for Laboratories Using Chemicals.” 1996.

(NFPA, 1996h) National Fire Protection Association. Standard 30, “Flammable and

Combustible Liquids Code.” 1996.

(NFPA, 1996i) National Fire Protection Association. Standard 14, “Standard for the Installation

of Standpipes and Hose Systems.” 1996.

(NFPA, 1996k) National Fire Protection Association. Standard 13, “Standard for the Installation

of Sprinkler Systems.” 1996.

(NFPA, 1997a) National Fire Protection Association. Standard 780, “Lightning Protection

Code.” 1997.

(NFPA, 1997b) National Fire Protection Association. Standard 221, “Fire Walls and Fire Barrier

Walls.” 1997.

(NFPA, 1997c) National Fire Protection Association. Standard 101, “Life Safety Code.” 1997.

(NFPA, 1997d) National Fire Protection Association. Standard 69, “Standard on Explosion

Prevention Systems.” 1997.

(NFPA, 1998a) National Fire Protection Association. Standard 801, “Standards for Facilities

Handling Radioactive Material.” 1998.

(NFPA, 1998b) National Fire Protection Association. Standard 55, “Standard for Compressed

and Liquefied Gases in Portable Cylinders.” 1998.

(NFPA, 1998c) National Fire Protection Association. Standard 25, “Standard for the Inspection,

Testing, and Maintenance of Water-Based Fire Protection Systems.” 1998.

(NFPA, 1998e) National Fire Protection Association. Standard 10, “Standard for Portable Fire

Extinguishers.” 1998.

(NFPA, 1999a) National Fire Protection Association. Standard 80, “Standard for Fire Doors and

Fire Windows.” 1999.

(NFPA, 1999b) National Fire Protection Association. Standard 70, “National Electric Code.” 

1999.

(NFPA, 1999c) National Fire Protection Association. Standard 50A, “Standard for Gaseous

Hydrogen Systems at Consumer Sites.” 1999.

(NRC, 1973) U.S. Nuclear Regulatory Commission. Regulatory Guide 3.12, “General Design

Guide for Ventilation Systems of Plutonium Processing and Fuel Fabrication Plants.” August

1973.
7-23

(NRC, 1992) U.S. Nuclear Regulatory Commission. Information Notice 92-14, “Uranium Oxide

Fires at Fuel Cycle Facilities.” February 1992.

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide Fuel Fabrication Facility.” August 2000.

(NRC, 2002) U.S. Nuclear Regulatory Commission. Briefing to the Advisory Committee on

Reactor Safety on the MFFF, “Electrical and Instrument and Control Systems Overview.” April

10, 2002.

(NRC, 2003a) Persinko, A., U.S. Nuclear Regulatory Commission. Meeting Summary: 

February 6–7 Meeting with Duke Cogema, Stone & Webster to Discuss the Mixed Oxide Fuel

Fabrication Facility Revised Construction Authorization Report. March 5, 2003.

(NRC, 2003b) Persinko, A., U.S. Nuclear Regulatory Commission. July 29–August 1, 2003,

Meeting Summary: Meeting with Duke Cogema, Stone & Webster to Discuss Open Items

Related to the Mixed Oxide Fuel Fabrication Facility. August 29, 2003.

(UL, 1995) Underwriters Laboratories, Inc. Standard 555, “Standard for Fire Dampers and

Ceiling Dampers.” 1995.
7-24

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8-1

8. CHEMICAL SAFETY

8.1 Conduct of Review

This chapter of the final safety evaluation report (FSER) contains the staff’s review of chemical

and process safety described by the applicant, Duke Cogema Stone & Webster (DCS), in

Chapter 8 of the revised construction authorization request (revised CAR) (DCS, 2002e), with

supporting process safety information from Chapters 5 and 11 of the revised CAR. The

objective of this review is to determine whether the chemical process safety principal structures,

systems, and components (PSSCs) and their design bases identified by the applicant provide

reasonable assurance of protection against the consequences of potential accidents. The staff

evaluated the information provided by the applicant for chemical process safety by reviewing the

revised CAR, supplementary information provided by the applicant, and relevant documents

available at the applicant’s offices but not submitted by the applicant. The staff also reviewed

technical literature as necessary to understand the process and safety requirements. The

review of aqueous polishing (AP) and mixed oxide (MOX) process (MP) safety design bases and

strategies was closely coordinated with the review of the radiation and chemical safety aspects

of accident sequences described in the safety assessment of the design bases (see Chapter 5

of this FSER), the review of fire safety aspects (see Chapter 7 of this FSER), and the review of

plant systems (see Chapter 11 of this FSER). 

The staff reviewed the chemical process and safety information in the revised CAR and

supporting documents against the following regulations:

• Title 10, Section 70.23(b), of the Code of Federal Regulations (10 CFR 70.23(b)) states,

as a prerequisite to construction authorization, that the design bases of the PSSCs and

the Quality Assurance Program must be found to provide reasonable assurance of

protection against natural phenomena and the consequences of potential accidents.

• 10 CFR 70.64, “Requirements for New Facilities or New Processes at Existing Facilities,”

requires that baseline design criteria (BDC) and defense-in-depth practices be

incorporated into the design of new facilities. With respect to chemical protection, 10

CFR 70.64(a)(5) requires that the MOX fuel fabrication facility (MFFF or the facility)

design 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. Related to chemical protection, 10 CFR

70.64(a)(3) requires that the facility design provide for adequate protection against fires

and explosions, such as those that could be initiated by or involve chemicals at the

proposed facility.

The review for this construction authorization focused on the design bases of chemical process

safety systems and components. For each chemical process safety system, the staff reviewed

information provided by the applicant for the safety function, system description, and safety

analysis. The review also encompassed proposed design basis considerations, such as

redundancy, independence, reliability, and quality. The staff used Chapter 8.0 of NUREG-1718,

“Standard Review Plan for the Review of an Application for a Mixed Oxide (MOX) Fuel

Fabrication Facility,” issued August 2000 (NRC, 2000), as guidance in performing the review. 

As stated on page 8.0-2 of NUREG-1718, “information contained in the application should be of
8-2

sufficient quality and detail to allow for an independent review, assessment, and verification by

the NRC reviewers.”

8.1.1 Background

As stated in the memorandum of understanding (MOU) between the U.S. Nuclear Regulatory

Commission (NRC) and the Occupational Safety and Health Administration (OSHA), “Worker

Protection at NRC-licensed Facilities,” (NRC, 1998b), the NRC oversees chemical safety issues

related to (1) radiation risk produced by radioactive materials, (2) chemical risk produced by

radioactive materials, and (3) plant conditions that affect the safety and safe handling of

radioactive materials, and, thus, represent an increased radiation risk to workers. The NRC

does not oversee facility conditions that result in an occupational risk unless they affect the safe

use of licensed radioactive material. The MOU provisions applicable to the proposed MOX

facility are now codified in 10 CFR 70.64(a)(5).

The NRC staff reviewed the revised CAR submitted by the applicant (DCS, 2002e) for the

following areas applicable to process safety at the construction authorization stage:

• chemistry description

• hazardous chemicals and potential interactions affecting licensed materials

• chemical accident sequences

• chemical accident consequences

• safety controls

Additional documentation from the applicant and the literature was reviewed as necessary to

understand the process and safety requirements. DCS stated that the BDC of 10 CFR 70.64(a)

are incorporated into the facility design and that applicable sections of the revised CAR are

intended to demonstrate compliance with these BDCs (revised CAR, p. 5.5-67). 

In performing its review of the revised CAR and supporting documents, the staff used the

applicable guidance of Section 8.4 of NUREG-1718 for determining acceptance under 10 CFR

Part 70, “Domestic Licensing of Special Nuclear Material,” consistent with the construction

authorization stage and the level of the design. The evaluation is summarized in the sections

that follow.

8.1.2 Areas of Review and Evaluation Findings

8.1.2.1 Chemical Processes

The plutonium feedstock to the facility requires purification before it can be used in nuclear fuel. 

This purification process is termed AP. A solvent extraction process is used in the AP process

to separate plutonium from gallium, americium, uranium, and other minor impurities. The

solvent extraction process is similar to the plutonium-uranium recovery extraction (PUREX)

process used domestically and worldwide in nuclear fuel processing (Bennedict, 1983). The

resulting purified plutonium, sometimes referred to as polished plutonium, is converted into

plutonium dioxide. Subsequently, powder processing operations combine the polished

plutonium dioxide with uranium dioxide to form the nuclear fuel. This is often referred to as MP. 
8-3

Chemical processes occur as part of the AP and MP operations, supported by reagent

preparation in the reagent processing building.

The MOX fuel fabrication building and the reagent processing building form the core group of

buildings for plutonium polishing (i.e., AP process) and MOX fuel fabrication (i.e., MP). These

are described in Section 1.1.1.1.2 of this FSER.

The reagent processing building would be located adjacent to the main MFFF building. Most of

the initial reagent dilution and mixing operations would occur here. The design has a

below-grade collection tank room that would receive waste chemicals from the reagent and AP

buildings. The AP waste chemicals could be slightly radioactive. In addition, the liquid solvent

area will be located on the northwest side of the building for the collection of liquid waste solvent

from the AP process and its subsequent transfer to the U.S. Department of Energy (DOE)

Savannah River Site (SRS) for waste processing. The waste solvent would be slightly

radioactive. The reagent processing building will be divided into discrete rooms/areas to

segregate chemicals and the associated equipment and vessels to prevent inadvertent chemical

interactions. Most reagents (e.g., nitric acid, hydrogen peroxide, hydroxylamine nitrate (HAN),

hydrazine, oxalic acid, sodium carbonate, diluent (C10–C13 isoalkanes, hydrogenated

tetrapropylene (HTP), or tetrapropylene hydrogenate (TPH)), and tributyl phosphate (TBP)) are

stored and solutions would be prepared in the reagent processing building for use in the AP

area. Sodium hydroxide and nitrates of silver and manganese are stored and prepared in the

AP area. Liquid chemical containers would be located inside curbed areas or sumps would be

provided to contain accidental spills. Chemicals are transferred to the AP area from the reagent

processing building via piping located in a concrete, below-grade trench between the two

buildings.

In addition, there would be a separate, secure warehouse building (referred to as BSW in Figure

1.1-2 of the revised CAR) that also stores depleted uranium dioxide, as a fine powder, in steel

drums. This building would also contain several drums of depleted uranium nitrate solution. 

These drums would be stored in a room with a 2-hour fire resistance rating. The uranium

dioxide powder is used to form the matrix for the MOX fuel pellets. The process adds depleted

uranium nitrate solution as an isotopic diluent for enriched uranium within the process. 

[Text removed under 10 CFR 2.390]

 Section 8.1.2.3.3 provides a detailed discussion of depleted

uranium.

The staff review identified the need for additional information regarding mass, energy, and

radioactivity balances relative to what was provided in the revised CAR. The applicant has

provided supplemental information in specific correspondence (DCS, 2001, 2002b, 2002d,

2002f). Mass balances have been found to generally agree to within 5–10 percent around

process units. The staff notes that these values may change as the design progresses. The

staff concludes that the additional information provided on the major balances and overall

process meets the standard review plan (SRP) (NRC, 2000) acceptance criteria and is adequate

and acceptable for the construction authorization stage.

Fluid transport system (FTS) categories and their applicable codes and standards are discussed

further in Section 11.8 of this FSER. The FTS categories 1 and 2 represent PSSCs, with FTS-1
8-4

representing a higher qualification level because of higher anticipated quantities and

concentrations of radionuclides (primarily plutonium and americium). Based upon the qualitative

descriptions of FTS categories and radionuclide concentrations, the staff concludes that the

majority of the AP equipment would have a PSSC and an FTS-1 or FTS-2 designation. Specific

PSSC and design basis findings for chemical safety are discussed later in this chapter and in

Sections 11.2 and 11.3 of this FSER.

8.1.2.1.1 AP Process

The AP process can be segmented into the following four operational areas (unit symbols are

provided in parentheses and discussed in Section 11.2):

• plutonium purification process—includes the decanning unit, dissolution unit,

dechlorination and dissolution unit, purification cycle, oxalic precipitation and oxidation

unit, homogenization unit, milling unit, canning unit, and recanning unit

• recovery processes—includes the solvent recovery cycle, oxalic mother liquor recovery

unit, and acid recovery unit

• waste storage—includes the liquid waste reception unit and waste organic solvent unit

• offgas treatment—includes the offgas treatment unit

The applicant will purchase a uranium nitrate solution as a reagent; consequently, the plutonium

purification process design no longer includes a uranium oxide dissolution unit. Likewise, the

applicant has eliminated the recycle of silver nitrate solutions from the recover processes. Thus,

the design no longer includes a silver recovery unit. An automatic sampling system also exists

for providing materials for radiochemical analyses.

Overviews of these operational areas are discussed in FSER Section 1.1. Figure 8.1-1 provides

a process overview. Section 11.2 of this FSER provides a more detailed description and

evaluation of the AP process.

8.1.2.1.2 MOX Process 

The MP involves dry subprocesses (i.e., blending of uranium and plutonium dioxide powders,

poreformer and lubricant additions, pelletizing, sintering, and rod and assembly processing). 

The MFFF uses the advanced micronized master blend (A-MIMAS) process for the manufacture

of MOX fuel assemblies. The A-MIMAS process uses a two-step mixing procedure. In the first

step, the plutonium dioxide powder is mixed with depleted uranium dioxide and recycled scrap

powder to form a primary blend (master blend) with approximately 20-percent plutonium dioxide

content of the total mass. This mix is then micronized (i.e., ball milled into a very fine powder). 

In the second step, the primary blend is forced through a sieve, poured into a jar, and mixed with

depleted uranium dioxide and scrap powder to obtain the final blend with the specified plutonium

content. The maximum plutonium dioxide content in the final blend is 6 percent of the total metal

content (normally referred to as heavy metal). The two-step mixing process is used to ensure a

consistent product. The MP areas include the following:
8-5

Figure 8.1-1 AP process overview

• The receiving area includes truck unloading, plutonium dioxide container handling,

counting, and storage before and after transfer to the AP line. The function of the

receiving area is to receive, unload, and store plutonium dioxide and uranium dioxide

powder.

• The powder area has equipment for dosing MOX powder at the specified plutonium

content in two steps for homogenizing and for pelletizing. The powder area receives

uranium dioxide and plutonium dioxide powders and produces a mixture of specific

plutonium content suitable for the production of MOX fuel pellets

• In the pellet process area, MOX pellets are sintered, ground, and sorted. The function of

the pellet process area is to receive, store, process, and handle fuel pellets. 

• In the fuel rod process area, pellets are loaded into rods, and the rods are inspected.

The function of the fuel rod process area is to assemble, inspect, and store fuel rods. 

• In the assembly area, rods are loaded into assemblies, and the assemblies are inspected

and stored. The functions of the assembly area are to receive fuel rods and the required

fuel assembly components and to assemble, inspect, and store completed MOX fuel

assemblies. 

• In a separate waste area, solid radioactive waste generated during the MP is processed,

stored, and packaged for shipment. 
8-6

The MP is reviewed in detail in Section 11.3 of this FSER.

8.1.2.1.3 Laboratory

Chemical and physical analyses of samples from the MP and AP areas will be conducted in the

laboratory. The MFFF laboratory operations would also include laboratory liquid and solid waste

management, temporary storage of scrap materials from the laboratory, MP adjustment in a test

line, calibration, and document storage.

As shown in revised CAR Appendix 5A, the applicant’s safety assessment notes that laboratory

operations involve hazards associated with the major event types described in revised CAR

Section 5.4.1.2.1. These events include loss of confinement/dispersal, fire, explosions,

criticality, natural phenomena, external man-made events, external exposure, and chemical

release. For this reason, the applicant stated that the following PSSCs are relied upon to reduce

risks to the facility worker, the individual outside the controlled area (IOC), and the environment

from potential hazards in the laboratory:

• C3/C4 confinement systems

• chemical safety controls

• laboratory material controls

• combustible loading controls

• criticality controls 

• fire barriers

• fire detection and suppression

• fluid transfer systems

• gloveboxes

• glovebox pressure controls

• material handling controls

• material handling equipment

• Material Maintenance and Surveillance Programs

• process control safety subsystem

• facility worker action

As discussed in Section 5.6.2.7 of the revised CAR, laboratory material controls consist of

administrative procedures that would be used to control the quantity of radiological and chemical

materials in the laboratory. The safety function is to limit the extent of any potential explosion by

limiting the quantity of hazardous chemicals that may be involved in any explosion and to limit

the quantity of radiological/chemical material available for dispersion following a potential

explosion. The applicant intended to develop procedures to establish limits on sample size, the

number of samples that may be stored and used in the entire laboratory and in any one location,

and the quantity of chemicals, reagents, or other hazardous materials that may be stored or

used in a laboratory. The applicant will also develop procedures to ensure that laboratory

operations are performed in accordance with safe laboratory operating practices. Procedures

will be developed by the applicant and reviewed by the NRC staff as part of the license

application phase.

As discussed in FSER Section 5, the staff independently evaluated the consequences of the

controlling events within each event group for the laboratory. The staff’s analysis confirms the
8-7

applicant’s assessment that the PSSCs listed above would be adequate to reduce the risk to all

receptors to acceptable levels. Section 11.11 and Tables 11.3-34, 11.11-1, and 11.11-2 of the

revised CAR provide information on the laboratory and the samples. The proposed safety

strategies are similar to practices at many radiochemical facilities. The staff finds this approach

acceptable for the construction authorization stage.

8.1.2.1.4 Chemical Reagents

Chemicals are received in various forms (solid, liquid, and gas) for use in the MFFF process.

Most chemicals are stored in the reagent processing building, while some would be stored in the

AP or MP areas. The main chemicals potentially stored in these areas have been identified by

the applicant as the following:

• Solids

— secured warehouse building—uranium dioxide

— reagent processing building—oxalic acid and sodium carbonate

— AP area—silver nitrate, sodium sulfite, plutonium dioxide, and manganese nitrate

— MP area—azodicarbamide, zinc stearate, plutonium dioxide, and uranium dioxide

• Liquids

— secured warehouse building—uranium (uranyl) nitrate solutions

— reagent processing building—hydrazine hydrate, HAN, nitric acid, TBP,

solvent/diluent (C10–C13 isoalkane), hydrogen peroxide, sodium hydroxide,

hydrazine nitrate, silver nitrate, sodium carbonate, sodium sulfite, and zirconium

nitrate

— AP area—fresh and recovered nitric acid, aluminum nitrate solution, silver nitrate

solution, manganese nitrate solution, solvent/diluent, HAN, hydrogen peroxide,

oxalic acid, sodium carbonate, sodium hydroxide, sodium nitrate, sodium sulfite,

TBP, uranyl nitrate, and zirconium nitrate solution

— MP area—isopropanol

• Gases

— reagent processing building—nitrogen tetraoxide (stored in a liquefied form in

large cylinders and forced by pressurized air through an evaporator in the reagent

processing building to form the gas, which is a mixture of nitrogen tetraoxide and

nitrogen dioxide)

— site—nitrogen, oxygen, hydrogen, argon, P10 (10 percent methane + 90 percent

argon), helium, and 95 percent argon/5 percent hydrogen
8-8

— AP area—nitrogen, oxygen, P10 (10 percent methane + 90 percent argon), and

95 percent argon/5 percent hydrogen

— MP area—nitrogen, hydrogen, argon, P10 (10 percent methane + 90 percent

argon), helium, and 95 percent argon/5 percent hydrogen

Storage facilities in the reagent processing building would contain the following:

• drums or tote tanks of reagents, including 13.6 N nitric acid, TBP, solvent/diluent, HAN

(1.9 M), hydrazine hydrate (N2H4.H2O; 22 percent hydrazine in water), hydrazine nitrate

(30 percent solution in water), and sodium hydroxide (10N)

• cylinders of liquid nitrogen tetraoxide

• containers of hydrogen peroxide (30 wt%)

• storage of material for dissolving solid reagents, such as oxalic acid, silver nitrate,

sodium sulfite, and sodium carbonate 

A list of process chemicals can be found in Tables 8-1a to 8-1e and 8-2a to 8-2c of the revised

CAR. Table 8.1-1 of this FSER provides a summary.

Chemicals and chemical mixtures in the process are usually used at lower concentrations than

the reagent-grade chemicals potentially stored in the reagent processing building or AP area. 

The more dilute reagents usually pose reduced hazards.

The applicant provided information about the delivery of chemicals to the proposed facility and

the potential impacts from spills and explosions. The applicant concluded that potential releases

from the delivery of chemicals would not affect the IOC. The applicant stated that the

emergency control room air conditioning system (ECRACS) is designated as a PSSC to ensure

habitable conditions for all potential workers in the emergency control rooms (ECRs). Based

upon completion of the chemical consequence analyses, measures will be provided, if

necessary, to ensure that ECR workers are protected from potential chemical releases. 

Habitability is discussed further in FSER Section 8.1.2.6. The applicant stated that no chemical

delivery scenarios have been identified that exceed the performance criteria of 10 CFR 70.61,

“Performance Requirements.” The applicant has identified an administrative PSSC entitled,

“Hazardous Material Delivery Controls,” which has the safety function of ensuring that the

quantity of delivered hazardous material and its proximity to the MOX fuel fabrication building

structure, the emergency generator building structure, and the waste transfer line are controlled

to within the bounds shown in the safety analysis to produce acceptable results. Currently, this

is expressed qualitatively because the designs are still evolving and numeric values may

change; the specific values, detailed administrative controls, and procedures will be developed

by the applicant and reviewed by the staff as part of its review of any later DCS application for a

license to possess and use licensed material. The staff notes that controlling the deliveries of

materials at nuclear facilities usually involves administrative controls. The staff finds this

qualitative approach acceptable for the construction authorization stage because the potential 
8-9

Table 8.1-1 Summary of Chemicals at the Proposed Facility

Chemical Anticipated Annual Usage Typical Estimated

Onsite Quantity

[Text removed under 10 CFR 2.390]
8-10

hazard is identified and controlled by a method that has the capability to meet the regulatory

requirements.

The staff review noted that sufficient general information on the chemical process description

was provided by the revised CAR and the applicant’s responses to Requests for Additional

Information 111 and 113 (DCS, 2001, 2002f), recognizing that some of the information

represents a snapshot in time as the design has continued to evolve. This includes information

on mass balances, inventories, process physical parameters, chemical reactions, equipment,

potential hazards, and proposed controls. The staff review noted several issues on the process

chemistry that are discussed in Sections 8, 11.2, and 11.3 of this FSER. The staff finds that the

overall process description itself is adequate and acceptable for the construction approval stage.

The applicant selected specific codes and standards that are typically used for similar purposes

in the chemical and nuclear industries. The selected specific codes and standards are

presented in general terms, and specific values are not mentioned and associated with specific

structures, systems, and components (SSCs). FSER Sections 8.1.2.5, 8.1.2.6. 11.2, 11.3, and

11.8 discuss the equipment, SSCs, and codes and standards in more detail.

8.1.2.1.5 Chemical Process Inventories

The applicant provided chemical inventory information (revised CAR Sections 8.1, 8.2, 11.2, and

11.3). Common hazardous materials (e.g., vehicle fuel) and commonly used small quantities of

solvents or gases are also used on site, and the applicant stated that specific inventories will be

identified in the detailed design. 

The staff reviewed the list of chemicals and their inventories. For construction, the chemical

listing appears sufficiently complete, and the quantities appear consistent with the proposed

activities and for preliminary safety assessments. The staff finds that this information is

adequate and acceptable for the proposed activities and the construction authorization stage.

8.1.2.1.6 Chemical Process Ranges

The applicant provided information on ranges in Sections 8, 11.2, and 11.3 of the revised CAR

and in supplementary submittals. In general, the focus is on design and steady-state conditions;

ranges are not well addressed. Specific information on process ranges that might impact safety

design bases are discussed in FSER Sections 8.1.2.5, 11.2, and 11.3. Additional information on

the ranges will be provided by the applicant and reviewed by the staff as part of its review of any

later DCS application for a license to possess and use licensed material.

8.1.2.1.7 Chemical Process Limits

The applicant provided information on chemical process limits in the revised CAR. In general,

the focus is on design and steady-state conditions; limits are not well addressed. Specific

information on process limits that might impact safety design bases are discussed in FSER

Sections 8.1.2.5, 11.2, and 11.3. Additional information on the limits will be provided by the

applicant and reviewed by the staff as part of its review of any later DCS application for a license

to possess and use licensed material.
8-11

8.1.2.2 Hazardous Chemicals and Potential Interactions

This section identifies the major chemicals that would be stored and used at the facility, identifies

potential interactions among these chemicals, and discusses potential unusual and unexpected

reactions. 

8.1.2.2.1 Chemicals

Plutonium would be received as impure plutonium dioxide in the AP area and transformed into

various compounds in different solutions at various stages of the process to remove the

impurities (e.g., gallium, americium, uranium). Purified plutonium will be converted back into

plutonium dioxide before being transported to the MP plutonium dioxide buffer storage unit. 

During some AP operations, new chemical compounds would be produced. Table 8.1-2

identifies chemicals produced during normal operations. Several of these chemicals are

potentially hazardous or are the precursors to potentially hazardous compounds, including

hydrogen, nitric acid, nitrous acid, and nitrogen oxides. Table 8.1-3 in this FSER identifies

hazardous and reactive chemicals potentially produced in significant quantities under off-normal

conditions. These are hydrazoic acid, azides, and red oil. These compounds and identified

controls (PSSCs and design bases) are discussed in detail in Section 8.1.2.5 of this FSER.

Nitrogen and helium gases at appropriate pressures would be used in the MP. Azodicarbamide

(a poreformer) and zinc stearate are used in powder form in the pelletizing process. Other

chemicals (liquids, solids, and gases) used in the MP area are present mainly in the laboratory

for inspection and analysis purposes. Chemicals would be stored and used in the MFFF

laboratory in small quantities and would thus be a smaller hazard than the chemicals referenced

in the previous paragraph. In the revised CAR, the applicant stated that laboratory chemicals

will be identified in the license application. The applicant provided a preliminary list of

anticipated chemicals and quantities in the laboratory area (revised CAR Table 8-2c). Typical

quantities are less than 10 kg (22 lb) of each reagent. 

8.1.2.2.2 Chemical Interactions

Human error or equipment malfunction may result in inadvertent chemical interactions and

initiate hazardous reactions. Hazardous chemical characteristics and incompatibilities with the

associated materials/process conditions have been identified by the applicant and are

summarized in Table 8.1-4. The staff concluded that this list was derived from standard

guidelines. The applicant would conduct hazard and operability analyses (HAZOPs) as part of

the integrated safety analysis (ISA) during detailed design and submit a complete chemical

interaction matrix as part of the license application. As stated in the revised CAR, the applicant

would control chemical preparation in accordance with operating procedures by trained

personnel in order to minimize the potential for unexpected interactions. To minimize the risk to

PSSCs associated with inadvertent chemical interactions, the applicant would prepare most

chemical reagents for the AP process in the reagent processing building (a non-radiological

building), with subsequent distribution to the AP area. 

Sections 8.4.3.2 (B) and (C) of the SRP list the acceptance criteria for potential chemical

interactions, with the understanding that some generalities are acceptable at the construction

authorization stage. The staff notes that the applicant identified general precautions and several 
8-12

Table 8.1-2 Reaction Products of the AP Process (Normal Operations)

Chemical Formula Comment

Alkaline Wastes

(including dibutyl phosphate

and monobutyl phosphate)

Various Alkaline wastes generated in the solvent

recovery unit as a result of washing solvent

with sodium carbonate and sodium hydroxide

solutions (Note 3)

Carbon Dioxide CO2 Reaction product when plutonium oxalate is

transformed into plutonium dioxide in the

oxalic precipitation and oxidation unit (Note 3)

Carbon Monoxide CO Reaction product when plutonium oxalate is

transformed into plutonium dioxide in the

oxalic precipitation and oxidation unit (trace

quantities only) (Note 3)

Chlorine Cl2 Reaction product from dissolution of AFS

material in the dechlorination dissolution unit

(subsequently treated in the dechlorination

scrubbing column) (Note 2)

Hydrogen H2 Produced from radiolysis and electrolysis

reaction (Note 4)

Nitrogen Oxides NOx Reaction product of dissolution units and

excess reactant from purification cycle

oxidation column (Note 2)

Nitrogen N2 Reaction product of several reactions in the

purification cycle oxidation column; reaction

product in dechlorination scrubbing column of

the dechlorination dissolution unit (Note 3)

Nitric Acid HNO3 Reformed in nitric oxide scrubbing columns

(offgas treatment unit) (Note 1)

Nitrous Acid HNO2 Always present in nitric acid solutions (Note 3)

Nitrous Oxide N2O Reaction product of several reactions in the

purification cycle oxidation column (Note 2)

Oxygen O2 Reaction product of hydrogen peroxide

decomposition during plutonium dioxide

dissolution in the dissolution unit (Note 3)

Plutonium Dioxide PuO2 Reformed in the calcining furnace of the oxalic

precipitation and oxidation unit from the

plutonium oxalate feed (Note 1)
8-13

Table 8.1-2 (cont.) Reaction Products of the AP Process (Normal Operations)

Chemical Formula Comment

Plutonium Oxalate Pu(C2O4)2 Precipitated in the oxalic precipitation and

oxidation unit from the reaction of plutonium

nitrate with oxalic acid (Note 1)

Plutonium (III, IV, VI) Nitrate Pu(NO3)3,

Pu(NO3)4,

PuO2(NO3)2

Plutonium (VI) nitrate formed from the

dissolution of plutonium dioxide in the

dissolution unit and in the evaporator of the

oxalic mother liquor recovery unit (Note 1)

Plutonium (IV) nitrate formed from the addition

of hydrogen peroxide to the plutonium (VI)

nitrate solution in the dissolution unit (Note 1)

Plutonium (III) nitrate formed from the

reduction of plutonium (IV) nitrate solution with

HAN in the purification unit (Note 1)

Sodium Chloride NaCl Reaction product in the dechlorination

scrubbing column of the dechlorination

dissolution unit (Note 3)

Uranyl Nitrate UO2(NO3)2 Formed from the dissolution in nitric acid of

the impure plutonium dioxide received at the

proposed facility (Note 1)

Water H2O Reaction product of several reactions in the

purification cycle oxidation column; reaction

product of hydrogen peroxide decomposition

during plutonium dioxide dissolution in the

dissolution unit; reaction product of uranium

dioxide dissolution reaction in the uranium

dioxide dissolution unit (Note 3)

Notes:

(1) The applicant performed chemical consequence analyses for nitric acid, uranyl nitrate,

and the plutonium compounds, using inventories from Table 8.2a of the revised CAR.

(2) The applicant performed chemical consequence analyses for chlorine, nitric acid, and

nitrogen oxides in selected areas of the facility based upon their projected quantities from

reactions.

(3) Because of the low rate of production and/or lack of toxicity, inventories are not

quantified for the purposes of calculating chemical consequences to the site worker or

the IOC from spills or releases.

(4) The generation of hydrogen is considered in the design of the scavenging air system.
8-14

Table 8.1-3 Potentially Hazardous Chemicals Produced in

AP under Off-Normal Conditions

Chemical Formula Revised CAR Comment

Hydrazoic Acid HN3

Interaction of hydrazine nitrate and nitrous acid could

initiate, under certain conditions, the formation of

hydrazoic acid. However, any accumulation of

hydrazoic acid, which occurs in the AP process, is 10

times lower than the explosive limits. In the gaseous

phase, hydrazoic acid may decompose to nitrogen

and hydrogen. (See FSER Section 8.1.2.5.3.2.)

Hydrazoic Salts (i.e., azides) NaN3

AgN3

Interaction of hydrazine nitrate and nitrous acid could

initiate, under certain conditions, the formation of

hydrazoic salts (i.e., azides). However, since the

solubility limits of these azides are not exceeded,

precipitation and the potential for an explosion are

prevented. (See FSER Section 8.1.2.5.3.3.)

Red Oil

(Various)

Red oil is an organic mixture, consisting of TBP and

its complexes with plutonium nitrate and nitric acid,

degradation products of TBP (e.g., dibutyl phosphate),

and possibly various nitrated hydrocarbons. These

are compounds that could react exothermically at

temperatures higher than 130 °C (266 °F). Process

unit design prevents the process fluid temperature

from exceeding 130 °C (266 °F) by providing

adequate margin between the maximum operating

temperature and 130 °C (266 °F). (See FSER Section

8.1.2.5.5.)

specific hazards associated with the interaction of chemicals and also considered radiolysis

effects. Specific hazards include red oil (solvent), HAN/hydrazine, azides, plutonium (VI)

oxalate, hydrogen, and pyrophoric effects from certain metals and oxides. Thus, the staff finds

that the overall process description itself is generally adequate and acceptable for the

construction authorization stage. Significant specific chemical-related risks are discussed

further in FSER Sections 8.1.2.5, 11.2, and 11.3.

8.1.2.2.3 Unusual and Unexpected Reactions

Additional chemical compounds are produced in some AP operations. The applicant stated that

the behavior of these mixtures is well understood from experience at the La Hague facility in

France and is included in the preliminary chemical process safety evaluation (e.g., Tables 8.1-2

to 8.1-4 of this FSER).

In general, for vapor and gaseous species, AP chemical interactions produce nitrogen oxides,

carbon dioxide, carbon monoxide, and hydrogen, as well as possible plutonium, americium, and

uranium mixtures entrained in nitric acid vapors. These chemicals would be generated in tanks

and equipment and would be collected by the offgas treatment system. Solvent-diluent vapors

would also be collected in a separate stream and treated by the offgas system before being 
8-15

Table 8.1-4 Process Chemical Hazardous Characteristics and Incompatibilities

Revised CAR (DCS, 2002e)

Form

Chemical

CorrosiveFlam

mable

ExplosiveChemical BurnToxic

Incompatibilities

Liquid

Nitric Acid (13.6 N) x x x Organics, Hydrogen Peroxide, HAN, Hydrazine

Monohydrate, Sodium Carbonate, Sodium Hydroxide

Hydrogen Peroxide x x x Organics, Nirtic Acid, Manganese (metal), Hydrazine,

Sodium Carbonate, Metallic Salts

Tributyl Phosphate (solvent) x x x x Ammonia, Nitric Acid, Oxidizing Agents, Strong Bases

Diluent (C10–C13 isoalkane) x x x Oxidizing Agents, Oxygen

Sodium Carbonate x Aluminum, Acids, Hydrogen Peroxide

Hydroxylamine Nitrate x x x x Bichromate and Permanganate of Potassium, Copper

Sulfate, Zinc, Strong Oxiders, Strong Reducing Agents,

Nitric Acid, Combustible Materials

Hydrazine Monohydrate x x x x Oxidizing Agents (Nitric Acid), Metals, Asbestos

Sodium Hydroxide x x x Acids, Aluminum and Other Metals, Organic Halogens

(especially Trichlorethylene), Sugars

Aluminum Nitrate x x x Combustible Materials, Strong Reducing Agents,

Metals, Water

Hydrazine Nitrate x x x Acids, Strong Oxidizers, Metal Salts

Isopropanol x x Oxidizing Agents

Zirconium Nitrate x x x Combustible Materials, Strong Reducing Agents,

Metals

Gas

Dinitrogen Tetraoxide/Nitrogen

Dioxide

x x x x Reducing Agents, Organics, Metals

Hydrogen x x

Oxygen x Organics

Solid

Silver Nitrate (also present as

liquid)

x x x x Ammonia, Carbonates, Chlorides

Manganese Nitrate (also

present as liquid)

x x x x Strong Reducing Agents, Combustible Materials

Oxalic Acid (also present as

liquid)

x x Silver, Sodium Chloride, Sodium Hypochlorite

Azodicarbonamide Strong Oxidizing Agents

Zinc Stearate x x Strong Oxidizing Agents, Acids
8-16

released to the stack.

The staff review notes that the list of chemicals produced in the process, as identified by the

applicant, is reasonably comprehensive but that additional reactions and compounds may be 

added as the design progresses. The applicant’s approach to identifying hazardous chemicals,

interactions, and incompatibilities is based on experience and process understanding and

appears capable of identifying any new potential hazards as the design evolves. The overall

approach is reasonable and acceptable for the CAR stage.

FSER Section 8.1.2.5 discusses process safety information and specific concerns in more detail,

and FSER Sections 11.2 and 11.3 discuss specific chemical concerns by process unit. The staff

findings are presented in those sections.

8.1.2.3 Chemical Accident Sequences 

The applicant provided information on chemical-related events in Sections 5.5 and 8 of the CAR. 

The applicant provided supplementary information in public meetings (NRC, 2003a, 2003b,

2003c), correspondence on the docket (DCS, 2003a), and in-office reviews of the preliminary

hazards analysis, preliminary accident analysis, and chemical accident modeling reports (NRC,

2001c, 2002d). This section presents the staff’s assessment of accident sequence bases,

unmitigated sequences, estimated concentrations, and concentration limits for chemical safety.

This section also provides the methodology and results for the evaluation of chemical

consequences that may be associated with a release of radiochemical materials.

As reflected in 10 CFR 70.64(a)(5), the design must provide adequate protection from the

following:

(1) chemical risks produced from licensed radioactive material

(2) facility conditions which affect the safety of licensed material 

(3) hazardous chemicals produced from licensed materials

Staff evaluations and conclusions are presented in subsequent sections.

8.1.2.3.1 Chemical Consequence Limits

This section evaluates the chemical risks to workers and members of the public from events that

release chemicals and radionuclides. The applicant stated that the chemical consequences are

based on bounding analyses.

The 10 CFR 70.61 performance requirements define high- and intermediate-consequence

events as including acute chemical exposures to individuals. The applicant primarily based the

chemical consequence limits for accident categories on “Temporary Emergency Exposure Limits

(TEELs),” Revision 18. The proposed limits are listed in Tables 8.1-5 and 8.1-6 of this FSER. 

The following are the TEEL definitions:
8-17

• TEEL—0 is the threshold concentration below which most people will experience no

appreciable risk of health effects.

• TEEL—1 is the maximum concentration in air below which it is believed that nearly all

individuals could be exposed without experiencing other than mild transient adverse

health effects or perceiving a clearly defined objectionable odor.

• TEEL—2 is the maximum concentration in air below which it is believed that nearly all

individuals could be exposed without experiencing or developing irreversible or other

serious health effects or symptoms that could impair their abilities to take protective

action.

• TEEL—3 is the maximum concentration in air below which it is believed that nearly all

individuals could be exposed without experiencing or developing life-threatening health

effects.

TEELs were developed by the DOE Subcommittee on Consequence Assessment and Protective

Action (WSMS, 2000; DOE, 2004) to serve as temporary guidance until the American Industrial

Hygiene Association (AIHA) publishes emergency response planning guideline (ERPG)

concentrations. The TEEL values are used by DOE, the U.S. Department of Defense, and some

other government agencies as ERPG approximations until peer-reviewed ERPGs become

available. As noted in Table 8.1-5, the applicant used ERPG values for several chemicals. 

However, NRC regulations and guidance do not mention TEELs, TEEL methodology, or values

derived from TEELs.

The TEEL definitions are similar to the ERPG definitions except for the 1-hour term because

concentration-dependent chemical limits are based on a 15-minute exposure time and dose

dependent chemicals are based on a 1-hour exposure time. The concentrations are then

adjusted to the same 1-hour exposure timeframe by an exponential formula. 

The ERPGs were developed by the AIHA ERPG committee (AIHA, 2004) as planning guidelines

to anticipate adverse health effects from the exposure to toxic chemicals. The ERPGs do not

contain safety factors that are usually incorporated into exposure guidelines.

The TEEL limits are subject to change over time. The TEEL limits are revised if new or

additional toxicity data are found or new concentration limits are issued. For example, within the

last 3 years, the TEEL-3 limits for nitric acid, hydrazine hydrate, and nitrogen tetraoxide, three

important chemicals used in bulk quantities in the AP system, varied by more than 50 percent.

The NRC does not promulgate its own chemical consequence limits but relies on values from

other government agencies and organizations that have a clear toxicological and regulatory

basis. There are multiple limits available for assessing the impacts from potential chemical

releases. The U.S. Environmental Protection Agency (EPA) is developing acute exposure

guideline levels (AEGLs) as a government standard for emergency response (EPA, 2004). The

AEGLs are defined as follows:

• AEGL-1 is the airborne concentration (expressed as ppm or mg/m3) of a substance at or

above which it is predicted that the general population, including susceptible but
8-18

excluding hyper-susceptible individuals, could experience notable discomfort. Airborne

concentrations below AEGL-1 represent exposure levels that could produce mild odor,

taste, or other sensory irritations. 

• AEGL-2 is the airborne concentration (expressed as ppm or mg/m3) of a substance at or

above which it is predicted that the general population, including susceptible but

excluding hyper-susceptible individuals, could experience irreversible or other serious

long-lasting effects or impaired ability to escape. Airborne levels below AEGL-2 but

above AEGL-1 represent exposure levels which may cause notable discomfort.

• AEGL-3 is the airborne concentration (expressed as ppm or mg/m3) of a substance at or

above which it is predicted that the general population, including susceptible but

excluding hyper-susceptible individuals, could experience life-threatening effects or

death. Airborne levels below AEGL-3 but at or above AEGL-2 represent exposure levels

that may cause irreversible or other serious, long-lasting effects or impaired ability to

escape.

Currently, AEGLs have been established for several chemicals. Interim AEGLs are available for

additional chemicals, including chlorine, hydrazine, nitric acid, and nitrogen dioxide, which are

chemicals of concern in the AP process. There are some overlaps in the definitions of the

AEGLs and TEELs. However, the high-consequence (Level 3 in Table 8.1-5) values proposed

by the applicant, which come from TEEL Revision 18 and ERPGs, are generally higher than the

AEGL values. 

The staff comparison of the applicant’s proposed values for selected chemicals to the limits

contained in the National Institute for Occupational Safety and Health (NIOSH) immediately

dangerous to life and health (IDLH) values (NIOSH, 1994, 2004; Casarett, 2001; Genium, 1999;

Lewis, 1996) (of which 10 percent is sometimes used for members of the public to

accommodate a wider range of population groups) found differences from the applicant values.

The chemical literature also includes other sources of limits. Those sources include military air

guidelines for short-term chemical exposure to deployed military personnel, short-term public

emergency guideline levels (a minimum-effects level), ceiling-limit values (also a minimumeffects level), and EPA toxic endpoints as part of its risk management program (USACHPPM,

1999; CFR, 2004). Many of the published values are different from those proposed by the

applicant.

The staff has conducted significant reviews of chemical consequence limits because of their

importance for meeting the performance requirements of 10 CFR 70.61 (NRC, 2004b). The

SRPs for MOX and fuel cycle facilities (NRC, 2000; NRC, 2002b) mention AEGL and ERPG

values as guidance. Other NRC guidance for control room habitability (discussed in Section

8.1.2.6 of this FSER) and NRC fuel cycle practice have used IDLH values. Licensing actions by

the NRC on other facilities have used IDLH and TEEL values (e.g., NFS, 2002). In general, the

ERPG/TEEL values tend to be higher when multiple guidance documents exist. There is a body

of scientific information that is subject to data availability, interpretation, uncertainty, and

discussion, and some values may have more conservatism and margin than others for chemical

exposures. Chemical exposure limits do not correlate well with risk levels for different types of

exposed populations (e.g., the general population as compared to a (healthy) worker
8-19

population). 

The staff notes that the differences are greater for the high-consequence limits (e.g., Level 3

values from Table 8.1-5, as compared to other values for similar purposes), and, for many

chemicals listed in Table 8.1-5, there are significant differences between the Level 2 and Level 3

values. 

The staff also considered the following:

• The NRC January meeting summary, dated March 5, 2003 (NRC, 2003b), states the

DCS position as using TEEL-2 limits for site worker limits and that site worker impacts

bound facility worker impacts.

• The May 30, 2003, letter from DCS (DCS, 2003c) discusses the offgas system and

identifies a DCS commitment not to exceed a TEEL-2 limit for site workers.

• The June 5, 2003, letter from DCS (DCS, 2003d) discusses chemical spills in cells and

states that, if necessary, existing MFFF features could be credited to ensure that

chemical consequences are low. “Low” is understood to mean less than TEEL-2 for

workers and less than TEEL-1 for the IOC.

• The applicant stated at a June 2003 meeting, as documented in the NRC June meeting

summary, dated July 18, 2003 (NRC, 2003e), that, if low consequences are exceeded for

any cell release, the applicant will upgrade existing devices to PSSCs.

In addition, page 8-12a of the revised CAR mentions comparing calculated concentrations of

chemicals to TEEL-2 values for the site worker and to TEEL-1 values for the IOC. Thus, while

the limits for determining chemical consequence categories are shown in Table 8.1-6, based

upon the numerical values in Table 8.1-5, the applicant is committing to not exceeding Level 2

values for the site worker (the 100-meter receptor) and not exceeding Level 1 values for the IOC

(the 160-meter receptor), as shown in Table 8.1-7. The staff finds that the Level 2 numerical

values in Table 8.1-5 are below the corresponding IDLH values for each chemical, and the

differences with other sets of values (e.g., AEGL-2s) are much smaller and would correspond to

little or no health impact. Further, as stated above, the applicant committed to maintaining low

chemical consequences to the site worker and IOC from events regulated by the NRC. Thus,

the staff concludes that the use of the numerical values of Table 8.1-5 with the commitment to

maintain low consequences to the site worker and IOC (Table 8.1-7) is a reasonable and

consistent approach that provides for adequate protection of site worker and IOC health and

safety and will result in a consistent designation of SSCs for site workers below that considered

to be IDLH. Thus, the staff finds that the applicant’s proposed approach is acceptable for

addressing the applicable 10 CFR 70.61 performance requirements for chemical protection. 

This closes Open Item CS-05b, previously identified in the staff’s revised draft safety evaluation

report (DSER) (NRC, 2003d).

8.1.2.3.2 Chemical Accident Assumptions

The applicant considered a range of initial conditions, as well as the failure modes of storage 
8-20

Table 8.1-5 Chemical Consequence Limits Proposed by the Applicant for the MFFF

 (mg/m3)

Name LowConsequence

Level

Level 1

IntermediateConsequence

Level

Level 2

HighConsequence

Level

Level 3

Aluminum Nitrate (TEEL) 15 15 500

Azodicarbonamide (TEEL) 125 500 500

Chlorine (ERPG) 3 7.5 60

Chromic (III) Acid (TEEL) 1 2.5 25

Diluent (C10–C13 Isoalkanes)

(TEEL—see Note 1)

5 35 200

Decane (C10) 5 35 25000

Undecane (C11) 6 40 200

Dodecane (C12) 15 100 750

Tridecane (C13) 60 400 500

Ferrous Sulfate (Lab) (TEEL) 7.5 12.5 350

Fluorine (Lab) (ERPG) 0.75 7.5 30

Hydrazine Monohydrate (TEEL) 0.0075 0.06 50

Hydrazine Nitrate (TEEL) 3 5 5

Hydrofluoric Acid (ERPG) 1.5 15 40

Hydrochloric Acid (ERPG) 4 30 200

Hydrogen Peroxide (ERPG) 12.5 60 125

Hydroxylamine Nitrate (TEEL) 15 26 125

Iron (TEEL) 30 50 500

Isopropanol (TEEL) 1000 1000 5000

Manganese Nitrate (TEEL) 10 15 500

Manganous Sulfate (TEEL) 7.5 12.5 500

Nitric Acid (ERPG) 2.5 15 200

Nitric Oxide (TEEL) 30 30 125

Nitrogen Dioxide (TEEL) 7.5 7.5 35
8-21

Table 8.1-5 (cont.) Chemical Consequence Limits Proposed by the Applicant for the

MFFF

(mg/m3)

Nitrogen Tetraoxide (TEEL) 15 15 75

Oxalic Acid (TEEL) 2 5 500

Potassium Permanganate 

(TEEL)

7.5 15 125

Silver Nitrate (TEEL) 0.03 0.05 10

Silver Oxide (TEEL) 30 50 75

Sodium (TEEL) 0.5 5 50

Sodium Carbonate (TEEL) 30 50 500

Sodium Hydroxide (ERPG) 0.5 5 50

Sodium Nitrite (TEEL) 0.125 1 60

Sulfuric Acid (ERPG) 2 10 30

Sulfamic Acid (TEEL) 40 250 500

Thenoyl Trifluoroacetone (TEEL) 3.5 25 125

Tributyl Phosphate (TEEL) 6 10 300

Uranium Dioxide (TEEL) 0.6 1 10

Uranyl Nitrate (TEEL) 1 1 10

Xylene (TEEL) 600 750 4000

Zinc Stearate (TEEL) 30 50 400

Zirconium Nitrate (TEEL) 35 35 50

(1) The TEEL value for the diluent represents the most conservative value in each category among the

primary constituents n-decane, n-undedcane, n-dodecane, and n-tridecane.

containers and associated systems. The following release scenarios are stated in the revised

CAR:

• leaks and ruptures involving equipment vessels and piping

• evaporating pools formed by spills and tank failures

• flashing and evaporating liquefied gases from pressurized storage

The applicant modeled releases using the total material at risk from the largest single tank or

container. No credit was taken for process equipment installed to scrub and remove gases and

vapors (i.e., installed for normal operations; assumed not to function during an accident involving

a release). The applicant stated that more detailed accident sequences will be developed as

necessary as part of the ISA during final design. The staff notes that the approach is consistent

with similar facilities and the handling of similar chemical hazards (AIHA, 2004; Casarett, 2001;
8-22

DCS, 2002d; DOE, 2004; EPA, 1999, 2004; Genium, 1999; Lewis, 1996; NIOSH, 1994; NRC,

1998a, 2001b; Stumpf, 2000; USACHPPM, 1999; WSRC, 1994a, 1994b, 1995, 1999).

The distance of the possible receptors was also considered. A 100-m (328-ft) distance was

used for calculations involving the site worker. The IOC receptor was assumed to be

approximately 160 m (525 ft) from the point of release. 

Table 8.1-6 Applicant’s Use of Chemical Limits for Qualitatively

Determining Chemical Consequence Categories

Consequence

Category

Site or Facility Worker IOC

High Concentration $ Level 3 Concentration $ Level 2

Intermediate Level 3 > Concentration $ Level 2 Level 2 > Concentration $ Level 1

Low Level 2 > Concentration Level 1 > Concentration

Table 8.1-7 Applicant’s Commitments to Chemical Limits for Qualitative Chemical

Consequences (DCS, 2003d, 2003e; NRC, 2003b, 2003e)

NRC Regulated Chemical

Event Type

Site Worker* IOC

 All Level 2 > Concentration Level 1 > Concentration

* Site worker impacts bound facility worker impacts.

The applicant used the largest evaporation rate calculated from two models, as discussed in

revised CAR Section 8.3.3, as input to the dispersion models for gases and vapors. The

applicant used the dispersion coefficients derived from the ARCON96 model (NRC, 1999) (see

FSER Sections 5.4 and 9) for estimating concentrations for the site worker (100-m (328-ft)

receptor) and the IOC (160-m (525-ft) receptor). For chemicals that do not have appreciable

vapor pressures and are released as fine particles (dust), the applicant used the five-factor

formula for the release rate with dispersion coefficients calculated by the ARCON96 model (see

FSER Section 9). The applicant no longer uses the EPA ALOHA code for chemical releases

(EPA, 1999).

The staff notes that the ARCON96 code has been found acceptable (see FSER Section

9.1.1.4.3). 

The analyses used the following input parameters and assumptions:

• release duration of 1 hour

• pool depth of 1 cm (0.4 in.)
8-23

• F atmospheric stability, with a 2.2 m/s (7.2 ft/s) windspeed, based on the 95th percentile

P/Q value from hourly measurements over a 5-year period (NRC, 1998a, 2001b) 

• largest single-process vessel (common-mode failures that could affect adjacent or

connected vessels were not considered)—477 L (126 gal) (each) for nitric acid,

dodecane, TBP, and hydrazine hydrate; 681 L (180 gal) for HAN; and 908 L (240 gal) for

nitrogen tetraoxide

[Text removed under 10 CFR 2.390]

• gas storage area gases will not reduce oxygen in personnel areas

• a temperature of 25 °C (77 °F)

• a ground-level release

• no mechanical or buoyancy plume rise

• a rural, flat-terrain topography

Section 5.5.2.10 of the revised CAR also indicates that direct chemical consequences with no

direct radiological effects are not regulated by 10 CFR Part 70 because they do not impact or

directly involve licensed material. In the case of a chemical release that has the potential to

impact a facility worker and prevent the performance of a safety function, Section 5.5.2.10.6.1 of

the revised CAR states that workers in the ECR perform a monitoring role during emergency

conditions. Consequently, the ECRACS is a PSSC with a safety function of ensuring that

habitable conditions are maintained in the ECR. The staff’s evaluation of this issue is described

in FSER Section 8.1.2.6.

Section 12.1 of the revised CAR states that the control of the facility would rely to a great extent

on automated systems to ensure facility safety. It also mentions that the operations staff would

be expected to monitor and confirm the status of confinement systems, fluid systems, and other

facility systems and to recover from off-normal conditions. The applicant expected that few

personnel actions would be relied upon for safety. At the present time, the applicant has

identified the closing of dampers and the valving in of emergency scavenging air as two operator

actions required outside of the ECR as part of a longer term response, and the applicant

expected that the effects of a chemical release would have dissipated by the time these operator

actions might become necessary (NRC, 2002d). The applicant stated that no safety-related

operator or worker actions are necessary outside of the ECR that affect radiological safety,

during or after a chemical release. Regarding PSSCs that are administrative controls, the

applicant indicated in the revised CAR (Section 5.5.2.10.6.1) that these are either permissive in

nature (e.g., no action until a sample is analyzed) or fail safe (i.e., crane stops). 
8-24

8.1.2.3.3 Chemical Modeling Approach

The applicant has estimated hazardous chemical concentrations using techniques and models

that are generally consistent with industry practice and generally follow the guidance on

atmospheric and consequence modeling found in NUREG/CR-6410, “Nuclear Fuel Cycle

Accident Analysis Handbook,” issued 1998 (NRC, 1998a). The staff has reviewed the chemical

modeling approach and finds that the basic models, approaches, and assumptions are

consistent with NUREG/CR-6410 and are, therefore, acceptable.

As part of its review of any license application that the applicant may later submit, when more

detailed design information is expected to be available, the staff will review the applicant’s

assumptions to ensure that they remain appropriate for the postulated application and physical

phenomena and that input values lead to a conservative estimate of potential consequences. 

Items that the staff will evaluate include release quantities, spill temperatures, chemical

concentrations, and other factors affecting potential hazardous chemical release rates.

8.1.2.4 Chemical Accident Consequences

This section presents the staff’s assessment of accident consequences for chemical safety.

8.1.2.4.1 Analysis

The applicant evaluated potential chemical consequences for an IOC at the controlled area

boundary, approximately 160 m (525 ft) from the point of release (i.e., the 160-m (525-ft)

receptor).

For the site worker and facility worker, the applicant estimated the chemical consequences to be

low. In the revised CAR, the applicant stated that the chemical consequence analyses for the

individual at 100 m (328 ft) (site worker) bound the consequences for the facility worker. As

stated in Section 11.2 of the FSER, the staff concurs with the applicant’s qualitative distance

analysis that, from the point of an in-cell release, the distance to a worker would be at least 100

m (328 ft). This closes a portion of the previously identified Open Item AP-13 from the revised

DSER (NRC, 2003d).

The staff evaluation notes that the ECRACS is identified for chemical safety (protecting

operators from chemical releases that could impact the safe handling of radioactive materials). 

As discussed in Section 8.1.2.6 of this FSER, the applicant provided design basis values for

habitability in the ECR.

The applicant identified events involving the release of hazardous chemicals (primarily chlorine)

produced from radioactive materials from the dechlorination and dissolution unit equipment and

associated areas that can affect the facility worker (revised CAR Section 5.5.2.10.6.2). The

applicant identified a mitigation strategy, and, as discussed in Section 8.1.2.3.1 above, the

applicant must meet the chemical consequence limits set forth in Tables 8.1-5 and 8.1-7. The

PSSCs are process entry controls for leaks occurring in cells, the C4 confinement system for

leaks occurring in gloveboxes, and facility worker action for leaks occurring in C3 ventilated

areas. The staff notes that administrative and ventilation control strategies similar to those

proposed by the applicant are routinely used in similar hazardous situations in industry and
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should be capable of reducing consequences and/or frequencies as needed. Process entry

controls would prevent worker access and hence exposure. The C4 confinement system would

maintain negative pressure in gloveboxes and would direct any chemical release away from

workers and, ultimately, would disperse it, thus preventing chemical exposure. Additionally,

facility worker actions, such as evacuating affected areas or donning protective equipment,

would mitigate chemical effects produced by site worker actions. Thus, the staff finds the

applicant’s approach for addressing chemical effects to be acceptable for the construction

authorization stage. The applicant will be required to demonstrate the effectiveness of the

safety strategies as part of any later application for a license to possess and use licensed

material.

The applicant identified an event involving the release of hazardous chemicals and radioactive

material (revised CAR Sections 5.5.2.10.6.3 and 8.5.1.10). This scenario involves the flow of

NO2/N2O4 at an abnormally high rate through the oxidation column. This has the potential for

excessive chemical consequences for the worker, site worker (individual receptor at 100 m (328

ft)), and the IOC (160-m (525-ft) receptor). The applicant identified a safety strategy based upon

mitigation. The PSSC is the process safety control subsystem (PSCS). The safety function is to

ensure that the flow of NO2/N2O4 is limited to under 44 kg/hr (97 lb/hr) (e.g., by active flow

controls) to the oxidation column such that chemical consequences do not exceed the chemical

consequence limits in Tables 8.1-5 and 8.1-7 (i.e., of low consequence) and, thus, are

acceptable. The staff finds the applicant’s fundamental approach of flow limitation for addressing

these chemical effects will enable it to perform the required mitigative safety function with the

necessary range of reliability; thus, it is acceptable for the construction authorization stage.

The staff had identified a concern with the airspeed and phenomenological models used to

estimate evaporation rates from spills within the facility. Section 8.3 of the revised CAR states

that the airspeed is calculated from the volumetric airflow rate through the room where the spill

occurs divided by the minimal vertical cross-sectional area in that room; this has been identified

by the applicant as an airspeed of approximately 0.1 m/s (0.33 ft/s) (NRC, 2003e). Similarly,

Section 8.4.1 of the revised CAR states that the airspeed is calculated from the volumetric

airflow rate through the room where the nitric acid spill occurs divided by the minimal vertical

cross-sectional area in that room; this has been identified by the applicant as an airspeed of

approximately 0.01 m/s (0.033 ft/s) (NRC, 2003e) for nitric acid spills or leaks in the AP area,

with spill temperatures modeled up to the boiling point of nitric acid. The applicant used these

air speeds as input into the evaporation models, which provide a release rate or source term. 

The staff notes that this use of such very low air velocities results in a significant reduction of the

estimated consequences by factors of between 20 and 200 because the evaporation rate model

used by the applicant is approximately linearly dependent with air velocity. Using these

airspeeds for evaporation and source term estimation, combined with the 2.2-m/s (7.2-ft/s)

windspeed for dispersion, the applicant concluded that the consequences would be low for the

workers and IOC. The staff noted the following:

• The approach of using different airspeeds for evaporation and dispersion is an

uncommon practice (e.g., NRC, 1998a).

• Use of the 2.2-m/s (7.2-ft/s) windspeed would likely result in exceeding the lowconsequence level for nitric acid (Tables 8.1-5 and 8.1-7), currently identified by the

applicant as 15 mg/m3 for the workers and 2.5 mg/m3 for the IOC. The applicant stated
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that the principal parameters which result in low consequences for all indoor releases are

the spill/room size and the low indoor windspeed. The applicant did not consider these

input airspeed parameters to be design basis values (DCS, 2003a). Thus, an

appearance of a safety effect is credited without identification of PSSCs and design

bases.

• Experience has shown that airflow velocities are usually uneven in cells, with higher

velocities as the air exhaust inlets are approached. Most facility designs include exhaust

inlets near the floor, where spills could likely occur. These real effects would likely result

in much higher velocities near spills than the 0.01–0.1 m/s (0.033–0.33 ft/s) calculated on

an average basis. For example, typical cell designs could produce air velocities of 0.5–1

m/s (1.6–3.3 ft/s) in the vicinity of the ventilation system inlet duct(s) based upon an

average cell airspeed of 0.01 m/s (0.033 ft/s). Typical nuclear and chemical ventilation

systems are designed for around 0.5–1 m/s (1.6–3.3 ft/s) at the inlet. The applicant’s

analysis does not account for these facts.

• There is no margin in the assumed air velocities of 0.01–0.1 m/s (0.033–0.33 ft/s).

• The selected models may no longer be suitable (e.g., laminar versus turbulent flow) or

may not be supported by experimental data for the low velocities assumed by the

applicant. 

• Other phenomena may need to be considered and modeled by the applicant. These

could include flashing and bulk convection effects. These would likely exceed the

straight wind velocity effects upon evaporation. For example, bulk correction factors

would increase release estimates by a factor of at least 3 at temperatures of

approximately 110 °C (230 °F) and trend to very high values as the boiling point

(approximately 118–120 °C (244–248 °F)) is approached (i.e., the flashing condition).

The applicant provided supplemental information (NRC, 2003e) indicating that there would be

drip pans in the AP process cells (not currently credited for chemical consequence safety) that

would reduce evaporation rates and keep the chemical consequence levels low, in accordance

with the levels in Tables 8.1-5 and 8.1-7. The applicant also committed (DCS, 2003d, NRC,

2003e) to performing a detailed analysis of each applicable process cell as part of its ISA. If

necessary, these drip pans could be credited at the license application stage to ensure that

chemical consequences do not exceed the low level of concern (Level 2 for the site worker and

Level 1 for the IOC, in accordance with Tables 8.1-5 and 8.1-7). The staff finds that the detailed

design and as-built information to be developed as part of the ISA process are not required for

CAR approval. The approach provides assurance of adequate safety and is found to be

acceptable to the staff. This part of the previously identified Open Item CS-05b from the revised

DSER (NRC, 2003d) is now considered closed.

8.1.2.4.2 Uncertainty

Estimates of risks are often accompanied by uncertainty because of the complexity of the

postulated scenarios, the physical models used to describe them, and the design itself

(particularly if it is preliminary). As stated in Section 8.4.3 of the revised CAR, the applicant

believed that conservative modeling has been utilized for the chemical releases, and this bounds
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any uncertainty. The applicant will perform more detailed consequence analyses as part of any

later application for a license to possess and use licensed material (DCS, 2003d; NRC, 2003e). 

As discussed in Section 8.1.2.4.1 above, the applicant has made a commitment to conduct a

cell-by-cell analysis as part of the ISA and credit features, as necessary, to ensure that chemical

consequences do not exceed the low level of concern. This should adequately address

uncertainty. 

8.1.2.5 Process Safety Controls

The applicant provided process safety information in Section 8.5 of the revised CAR. Additional

information has also been obtained from the applicant in correspondence on the docket, in-office

reviews of documentation, and public meetings (e.g., DCS, 2003a; NRC, 2002d, 2003a, 2003b,

2003c, 2003e, 2003f). This is reviewed and evaluated in the sections that follow.

Each of the MFFF facilities would have control requirements that would be incorporated into the

overall design of the control system for process safety control. More information on the control

system is available in FSER Section 11.6. 

Reagents are stored and chemical mixtures are prepared in the reagent processing building and

the reagent storage area of the AP area. The AP facility is broken down into process functional

units, which are made up of one or more subassemblies performing consistent and elementary

tasks. The applicant stated that the breakdown into control functional units allows each entity to

be operated relatively independently in the given operating mode. The applicant intended to

control process storage and operation conditions in order to prevent exothermic and potential

autocatalytic reactions in the reagent processing building and the AP area. Autocatalytic and

exothermic reactions of chemicals would be prevented through the control of the process

parameters (e.g., reactant concentration, temperature, catalyst concentration in solution, and

pressure) that affect the reactions.

The applicant stated that there is reasonable assurance that the PSSCs will be sufficiently

reliable and available, and this assurance will be provided through the use of standard nuclear

industry engineering practices. These practices are incorporated into the facility general design

philosophy, design bases, system design, and commitments to applicable management

measures. These practices ensure that applicable industry codes and standards are utilized,

adequate safety margins are provided, engineering features are utilized to the extent practicable,

the defense-in-depth philosophy is incorporated into the design, and PSSCs will be appropriately

maintained. Significant chemical-related risks and associated process safety information are

discussed in the following sections.

8.1.2.5.1 Safety Strategies for Events Involving Hydrogen

The staff notes that flammable liquids and gases would be delivered to and used in the proposed

MOX facility. At the proposed facility, hydrogen is used as a reagent, generated by 

radiolysis, and generated electrolytically. The applicant’s general strategies for the three

respective cases are to limit the hydrogen concentration below its lower flammability limit (LFL)

by dilution with ventilation air, dilute with scavenging air, and inhibit generation. The applicant

identified the design basis for hydrogen as National Fire Protection Association (NFPA) 69,

“Standard on Explosion Prevention Systems,” issued 1997 (NFPA, 1997), with the application of
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25 percent of the LFL (DCS, 2003c). The staff notes that a design-basis value of 25 percent of

the LFL is endorsed by standards (e.g., NFPA 69 (NFPA, 1997), 801, “Standard for Fire

Protection for Facilities Handling Radioactive Materials,” issued 1998 (NFPA1998), 50A,

“Gaseous Hydrogen Systems at Consumer Sites,” issued 2000 (NFPA, 2000a), and 86C,

“Industrial Furnaces Using a Special Processing Atmosphere,” issued 2000 (2000b)), prior NRC

activities (NRC, 1999), and DOE practices. The NRC inspects existing fuel cycle facilities using

this criterion. The staff evaluation of this design basis is discussed in Section 11.2 of this FSER.

8.1.2.5.1.1 Argon-Hydrogen Mixture in Sintering Furnace and Hydrogen Storage. The sintering

furnaces and the hydrogen storage unit/mixing station are two places associated with the use of

an argon-hydrogen mixture (see FSER Section 11.3). The nominal mixture is nonexplosive with

a minimum argon level of 95 percent and a maximum hydrogen concentration of 5 percent. 

Mixtures of argon and hydrogen containing more than about 5.8-percent hydrogen are

flammable in air (DCS, 2001, 2002f). The applicant intended to minimize explosion risk by the

following preventive measures:

• Storage tanks containing argon and hydrogen and the mixing station are located

approximately 60 m (197 ft) outside the MFFF building.

• The composition of the argon-hydrogen mixture is monitored and controlled.

• Welded supply tanks and feed lines are designed in accordance with the guidelines of

the Compressed Gas Association (CGA).

• Ignition sources (e.g., electricity, fire, and lightning) are either eliminated or precluded by

design for the area containing the hydrogen storage system.

The applicant described its comprehensive strategy regarding potential explosions associated

with the sintering furnace (DCS, 2002a, 2002c; NFPA, 2000a, 2000b; NRC, 2002a, 2003a,

2003b). This is discussed in FSER Section 11.3.1.2.4.

8.1.2.5.1.2 Radiolytic Hydrogen Production. Hydrogen production by radiolytic dissociation of

hydrogenous molecules occurs because of the presence of radioactive material in solution. The

applicant intended to preclude radiolytic-hydrogen explosion hazards by preventing the

accumulation of hydrogen exceeding the LFL. This will be accomplished by adding dilution

(scavenging) air and ventilation. 

Hydrogen generation will occur in the AP process. For the air-hydrogen system under ambient

conditions, the LFL is 4-percent hydrogen. The applicant stated in the revised CAR that

hydrogen concentration in the free volume of AP equipment would be maintained below 1 vol%

(i.e., less than 25 percent of the LFL) during normal operation by maintaining an adequate

dilution airflow and ensuring that an exhaust path exists. Air for dilution would be provided by

the scavenging air system. Emergency scavenging air would be supplied to those systems and

vessels in which flammable gas concentrations could exceed the LFL within 7 days after a loss

of ventilation. In such circumstances, emergency scavenging air would be supplied to maintain

the hydrogen concentration below 25 percent of the LFL at all times. The applicant stated that

the design basis for hydrogen is NFPA 69 (NFPA, 1997), with the application of 25 percent of

the LFL (DCS, 2002e). The staff evaluation of this strategy is presented in Section 11.2.1.3.3.
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Hydrogen production and accumulation may occur in the waste and byproducts, such as

contaminated organic waste or organic-additive-bearing waste containing significant amounts of

plutonium, scraps in transuranic waste containers, and other liquid waste. The applicant

intended to use container filters to limit hydrogen accumulation while maintaining the

confinement of radioactive materials. The waste containers have been identified as PSSCs with

the safety function to ensure that hydrogen accumulation in excess of limits does not occur. The

design basis is identified in Section 11.4.11.6 of the revised CAR and is a U.S. Department of

Transportation standard for waste containers. This approach is consistent with regulated and

successful nuclear industry experience and is acceptable to the staff at the construction

authorization stage.

8.1.2.5.1.3 Hydrogen Generated by Electrolysis. In an efficient process and with relatively low

voltages, hydrogen production by electrolysis is usually 5–10 percent of the passed current. At

higher voltages, the cathode may blow hydrogen, and the majority of the current is involved with

hydrogen generation. Using the applicant’s stated electrical conditions, the rate could approach

10 cc/s or more, which can be an order of magnitude or more larger than that generated by

radiolysis. The applicant identified an approach that inhibits hydrogen generation and, thus,

prevents its accumulation. The applicant stated that nitric acid normality and voltage limits may

be used to prevent hydrogen generation and keep the hydrogen concentration below the design

basis of 25 percent of the LFL (DCS, 2002e, 2003f; NRC, 2003a). The applicant also indicated

that acid concentration controls may be used. As discussed further in FSER Section 11.2.1.3.3,

the applicant provided information that indicates that the proposed strategy has the ability to

control hydrogen generated from electrolysis.

8.1.2.5.2 Solvents

The purification cycle and the solvent recovery cycle are two units in which solvent and diluent

are used in processing. The purification cycle uses the solvent-diluent mixture for the extraction

of plutonium. The solvent recovery cycle regenerates the solvent-diluent mixture by removing

the degradation products and adjusting the TBP content and stores the treated solvent at a

slightly acidic condition to prevent degradation by hydrolysis. The aqueous stream is washed

with the diluent to remove traces of entrained solvent. The solvent-diluent utilized in these

processes is combustible. According to NFPA 30 definitions, the diluent and solvent (TBP) are

Class IIIa and IIIb combustible liquids, respectively. Combustible characteristics of these

chemicals are presented in FSER Table 8.1-8.

The PSSCs and design bases associated with fire and explosion hazards within the process

cells are addressed in Chapters 5 and 7 of this FSER, respectively. This discussion focuses on

in-process areas, such as tank ullage spaces, piping, and ventilation systems (e.g., the offgas

system).

With respect to potential explosions, the applicant identified a preventive strategy. A preventive

strategy renders the potential event highly unlikely. This meets the intent of the performance

requirements of 10 CFR 70.61 and is acceptable to the staff. Acceptable approaches include

maintaining a temperature margin below the flashpoint, dilution of the flammable vapors by

active ventilation, rendering the vapor nonflammable (i.e., above the upper flammability limit or

inerting), or combinations thereof.
8-30

Table 8.1-8 Combustible Characteristics of Chemicals in the AP Area

Combustible Location Flashpoint, °C (°F)

Solvent—Diluent only Pulsed column, diluent washing, and

storage

> 57 (135)

Solvent—TBP only Storage tanks and solvent recovery > 120 (248)

Solvent—Diluent and TBP

mixture

Purification cycle and solvent recovery > 55 (131)

Solvent—Diluent, TBP, and

aqueous

Purification cycle and solvent recovery > 66 (151)

For each solvent or solvent mixture, there exists a relationship between the flashpoint of the

liquid and the LFL of vapors above the liquid-vapor interface. To a close approximation, the

flashpoint temperature of the liquid corresponds to 100 percent of the LFL in the vapor-air

mixture near the surface of the liquid. Therefore, two acceptable control strategies for

preventing solvent fires include (1) control of the solvent temperature below the liquid flashpoint

and (2) prevention of flammable vapor-air mixtures above the surface of the liquid by maintaining

vapor concentrations below some fraction of its LFL.

The applicant proposed to use the guidance provided in NFPA 69 (NFPA, 1997) as the design

basis for the control of flammable mixture concentrations within the AP process vessels (DCS,

2003f). As noted by the applicant, NFPA 69 provides various options to accomplish this,

including one of the following:

• The combustible concentration shall be maintained at or below 60 percent of the LFL

when automatic instrumentation with safety interlocks is provided.

• The combustible concentration shall be maintained at or below 25 percent of the LFL.

The applicant identified six groups of flammable gas concentration limits and safety strategies,

corresponding to six types of process units throughout the AP process. The limits, AP process

areas of applicability, and corresponding safety strategy are described below.

Group No. 1—Solvent Extraction, Solvent Recovery, and Wastes

Inside process vessels in the solvent extraction, solvent recovery, and waste units, the applicant

proposed a design-basis value of 60 percent of the LFL. This limit is in accordance with NFPA

69, as long as the system is equipped with automatic instrumentation with safety interlocks.

The PSSC, which performs the safety interlock function, is the PSCS. The PSCS will monitor

the process temperature and terminate heat sources before exceeding a temperature limit. For

example, under certain conditions, a temperature of 50 °C (122 °F) for HTP will ensure that

flammable solvent vapors remain below 60 percent of the LFL. A second administrative PSSC,

chemical safety controls, is also proposed by the applicant to control chemical concentrations

and flow rates of reagents. The control of reagents is necessary to maintain process
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temperatures below those which would cause a temperature limit to be exceeded from

exothermic chemical reactions.

Group No. 2—Oxalic Precipitation and Oxalic Mother Liquor Units

Inside process vessels in the oxalic precipitation and oxalic mother liquor units, the applicant

proposed a design-basis value of 25 percent of the LFL. This limit is in accordance with NFPA

69 (NFPA, 1997).

The safety function of the PSCS is to prevent diluent from entering equipment at elevated

temperatures. This will be implemented by online redundant measurements (e.g., density or

conductivity).

Group No. 3—Equipment at Elevated Temperatures in the Acid Recovery Unit

Inside equipment at elevated temperatures in the acid recovery unit, the applicant proposed a

design-basis value of 25 percent of the LFL. This limit is in accordance with NFPA 69.

The safety function of the PSCS is to prevent diluent from entering equipment at elevated

temperatures. This will be implemented by online redundant measurements (e.g., density or

conductivity).

Group No. 4—Equipment Not at Elevated Temperatures in the Acid Recovery Unit

Inside equipment not at elevated temperatures in the acid recovery unit, the applicant proposed

a design-basis value of 60 percent of the LFL. This limit is in accordance with NFPA 69, as long

as the system is equipped with automatic instrumentation with safety interlocks.

The PSSC, which performs the safety interlock function, is the PSCS. The PSCS will monitor

the process temperature and terminate heat sources before exceeding a temperature limit. For

example, under certain conditions, a temperature of 50 °C (122 °F) for HTP will ensure that

flammable solvent vapors remain below 60 percent of the LFL. A second administrative PSSC,

chemical safety controls, is also proposed by the applicant to control chemical concentrations

and flowrates of reagents. The control of reagents is necessary to maintain process

temperatures below those which would cause a temperature limit to be exceeded from

exothermic chemical reactions.

Group No. 5—Hydrogen Concentration Associated with Radiolysis

Inside equipment that has the potential for hydrogen concentration associated with radiolysis,

the applicant proposed a design-basis value of 25 percent of the LFL.

The safety function of the waste unit offgas treatment system is to provide an exhaust path for

the removal of diluted hydrogen gas in process vessels. The instrument air system will also

provide sufficient scavenging air to dilute the hydrogen that is generated during radiolysis. This

is discussed and evaluated in Section 11.2.1.3.3 of the FSER.
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Group No. 6—Hydrogen Generated by Electrolysis within the Dissolution Process

Inside equipment that has the potential for hydrogen gas concentration associated with

electrolysis within the dissolution unit, the applicant proposed a design-basis value of 25 percent

of the LFL.

The safety function of the PSCS is to limit the generation of hydrogen by ensuring that the

normality of nitric acid is sufficiently high to ensure that the offgas is not flammable. This is

discussed and evaluated in Section 11.2.1.3.3 of the FSER.

Staff Evaluation

The staff have evaluated the applicant’s proposal to adopt the guidance of NFPA 69 as the

design bases for AP process vessels containing flammable gases.

The guidance in NFPA 69 includes the following several key parts:

3-2 Basic Design Considerations: the following factors shall be considered in

the design of a system to reduce the combustible concentration below the

lower flammabilty limit:

(a) Required reduction in combustible concentration

(b) Variations in the process, process temperature and pressure, and

materials being processed

(c) Operating controls

(d) Maintenance, inspection, and testing

3-3.1 The combustible concentration shall be maintained at or below 25% of the

lower flammability limit. Exception: When automatic instrumentation with

interlocks is provided, the combustible concentration shall be permitted to

be maintained at or below 60% of the lower flammability limit.

3-4.1 Suitable instrumentation shall be provided to monitor the control of the

concentration of combustible components.

The standard requirement of NFPA 69 is 25 percent of the LFL, and the use of automatic

interlocks is an exception. The context of Section 3 is direct control of the combustible

concentration. The code also requires instrumentation to monitor the control of the

concentration of combustible components. 

The staff review noted that the 25 percent limit is endorsed in the MOX SRP (NRC, 2000), other

NRC guidance, DOE guidance, and NFPA codes for hydrogen and flammable vapors. The staff

has drawn a conclusion regarding the acceptability of a flammable hydrogen concentration limit

equivalent to 25 percent of the LFL in Section 11.2.1.3.3 of this FSER.

The staff has reviewed the applicant’s proposed design basis of NFPA 69. This is a top-level,

consensus standard often used in the chemical and nuclear industries. It has the ability to

address the potential hazard of flammable solvent vapors. It is used as the basis for several
8-33

acceptance criteria in Chapter 7 of the SRP. It provides for the determination of ranges of 

acceptable values. Consequently, the staff finds the general use of NFPA 69 by the applicant

for solvents to be acceptable. 

The staff concludes that the applicant has several different control strategies to pursue in the

ISA. Some of these strategies have similarities to approaches used to control solvent

flammabilities in other facilities and, in combination, may be capable of achieving very high

reliabilities to render the flammability event highly unlikely. The applicant will need to

demonstrate this high reliability as part of its later application for a license to possess and use

licensed material.

If the applicant intends to pursue the use of interlocks in accordance with NFPA 69, then, in the

ISA supporting the possession and use license application, the applicant must demonstrate that

the proposed interlocks can perform their required safety function(s). Such a demonstration will

likely require a clear calculational and experiential basis and setpoint analysis to meet the

setpoint design basis. This may result in setpoint values below those allowed by NFPA 69. 

However, the staff concludes, on the basis of the foregoing evaluation, that the applicant has

proposed a safety strategy and design basis for preventing solvent explosions in AP equipment

which meet the requirements for construction authorization of the proposed MFFF. This closes

Open Items CS-9 and AP-9 identified in the staff’s revised DSER (NRC, 2003d).

8.1.2.5.3 Hydroxylamine Nitrate and Hydrazine in Nitrate Media

The AP process uses a mixture of HAN and nitric acid to strip (i.e., recover in the aqueous

phase) plutonium from the solvent after removal of americium and gallium at the extraction step. 

The HAN reduces the plutonium (IV) to the plutonium (III) state. It possesses the proper

plutonium (IV) to plutonium (III) reduction potentials and produces volatile reaction

products—nitrogen, nitrous oxide, and water. These products do not contribute to the wastes

generated by the process.

Hydrazine is used in conjunction with HAN to impede the HAN reaction on nitrous acid (always

present in nitric acid solutions) and, thus, increase HAN availability for plutonium reduction. 

However, interaction of hydrazine nitrate and nitrous acid could initiate, under certain conditions,

the formation of hydrazoic acid or azides (hydrazoic salts), which are hazardous compounds.

Besides limiting the plutonium losses in the process, hydrazine can enhance safety by keeping

the nitrous acid concentration at a very low level. Nitrous acid is the main intermediate chemical

that can lead to the formation of unstable compounds and conditions in the solvent extraction

process.

The staff notes that energetic hydroxylamine-nitric acid reactions can occur under the right

conditions, as evidenced by a DOE investigation of an accident at Hanford (DOE, 1997a, 1997b,

1998a). As a result of this explosion, DOE investigated the situation and concluded that the

HAN phenomena involved the interdependence between at least the following five parameters:

(1) chemical concentration of each reactant

(2) molar ratio of nitric acid to HAN
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(3) temperature of the mixture

(4) concentration of metal ions (as catalysts)

(5) pressure of the system (appears to influence the severity of the reactions but not the

initial autocatalytic initiation)

The DOE derived an instability index and a graph to link these parameters and generally account

for the behavior of the system. The DOE also identified other precautions, such as frequent

monitoring of HAN solutions, and mentioned that only dilute nitric acid should be added to HAN

and that the addition should be performed slowly and in a well-ventilated tank. Dilute nitric acid

was not specified further.

The material safety data sheet (MSDS) for hydroxylamine (which may be mixed with nitric acid to

make HAN) also lists safety precautions that include temperature limits, venting/volume increase

concerns, crystallization concerns, and hazardous reactions with oxidants, such as nitrates. The

MSDS documentation identifies large volume increases (up to 2000 times the initial volume of

the solution) during modest heating and violent decomposition at temperatures above 75 °C 

(167 °F), with the liberation of ammonia, nitrogen, and nitric oxide. Metallic impurities and

surfaces, such as iron, can accelerate the reactions. 

The use of HAN and hydrazine in nitric acid is a complex, multiparameter chemical system. This

system involves the interrelationship of the chemical concentrations of each material, molar

ratios (particularly for HAN and nitric acid), temperature, and concentration of metals (i.e.,

potential catalytic effect). These systems generally present three main hazards, including HAN

autocatalytic decomposition, hydrazoic acid, and azides, which correspond to overpressurization

and explosion events. 

8.1.2.5.3.1 Hydroxylamine Nitrate Decomposition. The applicant provided its safety strategy in

the revised CAR (Section 5.5.2.4.6.4). The staff evaluation and conclusions (NRC, 2004c) are

summarized in this section of the FSER. In the revised CAR (DCS, 2002e), the applicant

revised and explained its proposed safety strategy to prevent HAN explosions for the two

categories—without and with nitrous oxide addition. 

Category I—Process Vessels Containing HAN and Hydrazine Nitrate without Nitrous Oxide

Addition

The applicant proposed a preventive safety strategy to reduce the risk to the worker, site worker,

IOC, and environment, as summarized in Table 8.1-9. The PSSCs are the following:

• The PSCS has the safety function of maintaining temperatures of these solutions to

within safety limits.

• Chemical safety controls have the safety function of ensuring that the concentrations of

nitric acid, metal impurities, and HAN introduced into the process are within safety limits. 

The design basis safe limits were developed by the applicant using a kinetic model comprised of

five nonlinear ordinary differential equations and a number of rate constants and boundary
8-35

conditions (NRC, 2003a; DCS, 2003c, 2003e, 2003g). The kinetic model includes generation

and destruction terms for plutonium (IV), nitrous acid, HAN, hydrazine hydrate, and hydrazoic

acid. The applicant used a commercial software package to solve the five-equation set. For a

given set of initial conditions (i.e., starting temperature and initial concentrations of plutonium

(IV), ionic strength, nitric acid, nitrous acid, HAN, hydrazine hydrate, and hydrazoic acid), the

numerical solution to the problem shows the time-dependent changes in the concentrations of

plutonium (IV), nitrous acid, HAN, hydrazine hydrate, and hydrazoic acid. The boundaries of

safe operations are those values of limiting initial conditions for which the model calculates wellbehaved reduction of plutonium (IV) and nearly stoichiometric consumption of HAN. In this

mathematical model, the autocatalytic runaway condition is exhibited as a rapid consumption of

HAN (usually in fractions of a second) and the generation of large amounts of nitrous acid.

In addition to the model, DCS has committed to conducting confirmatory testing to further

substantiate the results of the model (DCS, 2003e). This testing will examine the reaction

mechanism and kinetics of nitric and nitrous acid with nitrogen tetraoxide. DCS also plans to

conduct testing to examine the reaction kinetics and mechanism for reactions between nitrous

acid, hydrazine, and HAN. The investigations will be performed under a range of temperatures

(0–70 °C (32–158 °F)), HAN concentrations (1 mM–2.5 M), and hydrazine concentration (0.1

mM–0.2 M). In addition, the plutonium redox behavior in nitric acid will be examined. Lastly,

integral experiments that confirm the kinetic model predictions will be performed.

The staff reviewed the system of equations and rate constants. These have been derived from

several different literature sources published over two decades. Rate constants were presented

for conditions different from those anticipated at the proposed facility, and, consequently, the

applicant used correlations to extrapolate the data. The staff used a different commercial

software package to solve the series of differential equations and to evaluate the margin in the

applicant’s approach. The results are shown in Table 8.1-10. The staff evaluated the system of

equations and results and found that they predicted a region of stability in process temperature,

nitric acid concentration, and HAN concentration that is consistent with the experimental data

used to develop the DOE instability index method (DOE, 1998a). 

As a base case, the initial conditions were set equal to either the design basis values, where

they had been proposed by the applicant, or the normal values expected for the purification

process. Then, the value of each parameter was adjusted individually to find the region of

instability with respect to that parameter. The staff assumed a nitrous acid concentration of 0.01

M. The system is insensitive to plutonium (IV) initial concentration, but a nominal value of 0.17

M (corresponding to 40 g Pu/L) is used. Ionic strength is adjusted automatically in the model as

a function of plutonium ion and nitric acid concentrations.

Category II—Process Vessels Containing HAN and Hydrazine Nitrate with Nitrous Oxide

Addition

In the revised CAR (DCS, 2002e) and by letters dated May 30, 2003, July 28, 2003, and October

6, 2003 (DCS, 2003c, 2003e, 2003g), the applicant revised and explained its proposed safety

strategy to prevent HAN explosions in units with nitrous oxide addition. Nitrous oxide is added to

some AP units to destroy excess HAN, hydrazine, and hydrazoic acid. In the revised strategy

(summarized in Table 8.1-11), the overproduction of noncondensible gases resulting from HAN,

hydrazine, and hydrazoic acid destruction by nitrous oxide would be prevented.
8-36

The purpose of the two PSSCs in Table 8.1-11 is to ensure that the pressure rating of process

vessels is not exceeded. The applicable codes and standards for process vessels with fluids

containing significant quantities of americium or plutonium are those for FTS-1, as described by

the applicant in Section 11.8 of the revised CAR. The design basis values for pressure,

temperature, flow, and volumetric capacities of process columns are provided in Table 11.8.2 of

the revised CAR. Section 11.8 of this FSER contains the staff review and evaluation of these

FTS codes and standards.

Table 8.1-9 PSSCs for Process Units with HAN and without Nitrous Oxide Addition

PSSC Safety Function Controlled

Parameter

Design

Basis

PSCS Maintain temperature below safe

limits

Temperature < 50 °C

(122 °F)

Chemical Safety

Controls

Maintain maximum nitric acid

concentration

[HNO3] < 6 M

Maintain minimum hydrazine

concentration

[N2H4] $ 0.1 M

Maintain maximum HAN

concentration

[HAN] < 2.5 M

Limiting residence time of nitric

acid, HAN, hydrazine with nitric

acid, and plutonium-bearing

solution

time as low as

reasonable

(probably

several

months)

Table 8.1-10 Design Basis Values for the Prevention of HAN Explosions

Controlled

Parameter

Design Basis Value Stable Value Margin (%)

Temperature < 50 °C (122 °F) < 63 °C (145 °F) 13 °C (55 °F), (25%)

HNO3 < 6.0 M < 7.1 M 1.1 M, (18%)

HAN < 2.5 M * *

N2H4 $ 0.1 M > 0.018 M 0.082 M, (550%) 

HN3 0.0 M N/A N/A

* The kinetic model with hydrazine = 0.1 M exhibits stability at all HAN concentrations

below the design basis value.
8-37

Table 8.1-11 PSSCs for Process Units with HAN and Nitrous Oxide Addition

PSSC Safety Function Controlled

Parameter

Design basis

Offgas Treatment

System

Exhaust path for removal of offgases,

which provides a means for heat

transfer/pressure relief for affected

process vessels

Heat transfer,

pressure relief

Revised CAR

Table 11.8-2

Chemical Safety

Controls

Limit nitric acid concentration [HNO3] Revised CAR

Table 11.8-2

Limit hydrazine concentration [N2H4] Revised CAR

Table 11.8-2

Limit HAN concentration [HAN] Revised CAR

Table 11.8-2

Limit hydrazoic acid concentration [HN3] Revised CAR

Table 11.8-2

The applicant also indicated that the tank in the reagent building would be inerted but has not

determined whether the tank in the MOX building would be inerted. Inerting was intended for

process control and not for safety purposes. The applicant did not anticipate either of these

becoming PSSCs but planned HAZOPs at a later time to further investigate the hydrazine and

HAN areas.

Staff Evaluation and Conclusions on HAN/Hydrazine

With regard to Category I HAN/hydrazine controls, the staff notes that its independent testing of

the model equations indicates that significant margin is present in the safety strategy. In

addition, the presence of multiple controls and measurements indicates the potential for high

reliabilities. The applicant’s commitments to confirmatory experimental testing to further

substantiate the kinetic model and understand the HAN/hydrazine system provide additional

confidence that the proposed strategy has the ability to address the safety concerns. The staff

has also found literature analogues for HAN/hydrazine control that are consistent with the

proposed approach (Biddle, 1968; WSRC, 2001b). Therefore, the staff concludes that the

overall strategy has the ability to meet the performance requirements and is acceptable for

construction authorization.

With regard to Category II HAN/hydrazine controls, the staff finds that the proposed codes and

standards are consistent with industry approaches and provide for a value or range of values

that protect the integrity of the vessels with a nitrous oxide addition from overpressurization. 

Therefore, the staff concludes that the overall strategy has the ability to meet the performance

requirements and is acceptable for construction authorization.

Therefore, the applicant has provided sufficient information to close Open Item CS-02 from the

revised DSER (NRC, 2003d).
8-38

Finally, the applicant intended to use hydrazine as a scavenger for nitrous acid and as a means

to improve the efficacy of the plutonium stripping. Hydrazine can also react strongly with

oxidants, such as nitric acid and peroxide. In the revised CAR, the applicant proposed the use

of administrative controls to control hydrazine concentrations within safety limits. The applicant

provided additional information indicating that the hydrazine would be received as hydrazine

hydrate, at a certified concentration of 35 percent (22 percent as hydrazine). The applicant

stated that solutions below 40-percent hydrazine (60 percent as the hydrate) possess no flash or

fire point and are thus nonflammable. The applicant stated that the concentration would be

confirmed by independent testing not to exceed 35-percent hydrazine (DCS, 2001). The

applicant summarized the approach as follows: 

• a certified or lot analysis by the manufacturer under a facility-approved quality assurance

plan

• an analysis upon receipt at the facility

• an analysis after mixing/diluting for facility use

The staff evaluation concludes that the applicant’s proposed approach using multiple

independent administrative controls at different times to control hydrazine concentrations

provides the potential for high reliability and, thus, reasonable assurance that the additions do

not exceed 35-percent hydrazine. The staff finds this acceptable for the construction

authorization stage.

8.1.2.5.3.2 Hydrazine and Hydrazoic Acid. Hydrazoic acid is also called hydrogen azide. It is

formed when hydrazine reacts with the relatively small quantities of nitrous acid present in nitric

acid solutions:

N2H5NO3 + HNO2 = > HN3 + 2 H2O + HNO3

Hydrazoic acid can undergo further oxidation to yield gases and water:

HN3 + HNO2 = > N2O + N2 + H2O

The first reaction generally proceeds faster than the second reaction under most of the

conditions in the AP process. Thus, there is a net accumulation of hydrazoic acid, and, over

time, an equilibrium concentration can be established. Hydrazoic acid exists in the aqueous

phase and organic phases that have contacted the aqueous solutions. Thus, the process at the

proposed facility is designed to destroy aqueous hydrazoic acid in the oxidation column (with

nitrous/nitrous oxide fumes) and to remove the organic hydrazoic acid in the solvent recovery

unit (as sodium azide, in the aqueous alkaline waste stream). The sodium azide is subsequently

destroyed by sodium nitrite in the liquid waste reception unit.

Hydrazoic acid is a colorless liquid and extremely soluble in water. The multiple nitrogen bonds

produce a molecule containing significant energy that can be released by decomposition

reactions as follows: 

2HN3 = >3 N2 + H2 + 526KJ
8-39

Energetic decomposition of hydrazoic acid can occur if sufficient concentrations form in either

the liquid or gaseous phases. The applicant has identified the following two type of hazards

associated with hydrazoic acid:

(1) an explosion related to a mixture of hydrazoic acid in air

(2) an explosion related to the distillation and condensation of hydrazoic solutions

The applicant intended to relate the first type to the liquid phase concentration via the Henry’s

Law correlation.

The applicant adopted a preventive safety strategy to protect the worker, site worker, IOC, and

environment. The PSSCs include the following:

• For the chemical safety controls, the first safety function is to assure that the proper

concentration of hydrazine nitrate is introduced into the system, thus limiting the quantity

of hydrazoic acid produced. The applicant has identified the design bases as a

maximum hydrazine concentration of 0.14 mol/L and a hydrazine to hydrazoic acid yield

of 39.3 percent or less.

The second safety function is to ensure that hydrazoic acid does not accumulate or

propagate into the acid recovery and oxalic mother liquor recovery units. This is

accomplished by neutralization in the solvent recovery unit. The applicant intended to

demonstrate the effectiveness of hydrazoic acid neutralization as part of the ISA at the

time of the possession and use license application (NRC, 2003a). Sampling controls

have been identified as a PSSC to confirm the destruction of the hydrazoic acid (revised

CAR Section 8.5.1.8, p. 8-33).

• For the PSCS, the safety function is to limit the temperature of the solution, thereby

limiting the vapor pressure and evaporation of hydrazoic acid. The applicant identified a

design basis temperature not exceeding 60 °C.

The staff reviewed information on hydrazoic acid and concluded that it is a potential explosion

hazard. The staff concurs with a prevention strategy for this hazard. The revised CAR does not

provide adequate justification for the design bases. Supplemental information from the literature

and the applicant (WSRC, 2001b; DCS, 2003b) indicates the following:

• The Henry’s Law coefficient is sometimes used to estimate vapor concentrations of

hydrazoic acid in HAN/hydrazine systems.

• A typical hydrazoic yield is around 33 percent.

The staff finds that limiting the hydrazine concentration and hydrazoic acid yield directly limits the

hydrazoic acid concentrations in the liquid and gaseous phases, in a manner consistent with the

literature. The use of neutralization prevents recycling of the hydrazoic acid into other parts of

the process. In addition, vapor pressure is a strong function of temperature, and, thus,

temperature control provides another direct control that can be very effective. The staff
8-40

concludes that the three safety functions are reasonably independent and, in combination, have

the ability to make the explosion event highly unlikely, in accordance with the performance

requirements.

The staff concludes that the selected strategy, PSSCs, and design bases have the ability to

meet the performance requirements and are acceptable for the construction authorization stage. 

This closes Open Item CS-03 from the revised DSER (NRC, 2003d). 

8.1.2.5.3.3 Hydrazine and Azides. Hydrazoic acid forms metal azides with cations in nearneutral and alkaline solutions. These are slightly soluble or insoluble, depending on the cation. 

Some metal azides can become unstable under certain conditions and present a potential

explosion hazard.

The applicant has adopted a preventive safety strategy to protect the worker, site worker, IOC,

and environment. The PSSCs include the following:

• For the chemical safety controls, the first safety function is to assure that azides are not

added to high-temperature process equipment (e.g., the calciner). The applicant

identified the design bases as no addition of azides to high-temperature process

equipment.

The second safety function is to ensure that azides have been destroyed before

acidification of the alkaline wastes and their transfer to the high alpha waste of the liquid

waste reception unit. Sodium nitrite is added as the scavenging agent for the azides. 

Sampling controls have been identified as a PSSC to confirm the destruction of the

azides (revised CAR Section 8.5.1.9, p. 8-33) before acidification.

The third safety function is to ensure that tanks potentially containing azides are not left

dry (revised CAR Section 8.5.1.9, p. 8-32).

• For the PSCS, the first safety function is to ensure that metal azides are not exposed to

temperatures that would supply sufficient energy to initiate energetic azide

decomposition. The temperature design basis has been identified as not exceeding 140

°C (284 °F).

The second safety function is to limit and control conditions under which dryout can

occur.

Silver azide can also form upstream in the process according to the reaction: 

HN3 + AgNO3 = > AgN3 + HNO3 

However, the process separates silver from hydrazoic acid containing vessels and does not

recycle hydrazoic acid. In addition, silver concentrations are estimated to be very low in vessels

containing hydrazoic acid. Finally, the temperature design basis of 140 °C is below the

decomposition temperature of silver azide. 
8-41

The staff review found that the applicant identified the potential hazards of metal azides, the

process includes steps to reduce these hazards, and PSSCs and design bases have been

proposed to reduce potential hazards and risks even further. Two PSSCs with multiple safety

functions have sufficient independence and, in combination, should have the ability to render the

potential event highly unlikely. Consequently, the staff concludes that the approach provides

adequate assurances of safety for construction authorization. 

8.1.2.5.4 Hydrogen Peroxide

A 35-wt% hydrogen peroxide solution is received and diluted to 10 wt% for use in the dissolution

units and the oxalic mother liquor recovery unit. Explosive vapors can be produced if a

concentration of hydrogen peroxide solution exceeding 75 wt% is boiled. The applicant stated

that administrative controls would be used to specify and dilute the hydrogen peroxide

concentration before use in the process in order to prevent explosions. The PSSC is identified

as chemical safety controls. 

The applicant provided supplemental information (DCS, 2002f) stating that the peroxide would

be received as 35-wt% solution and diluted to form a 10-wt% solution. The applicant

summarized the approach as follows:

• a certified or lot analysis by the manufacturer under a facility approved quality assurance

plan

• an analysis upon receipt at the facility

• an analysis after mixing/diluting for facility use

The staff evaluation concludes that the applicant’s proposal to use multiple independent

administrative controls at different times to control peroxide concentrations is acceptable.

8.1.2.5.5 TBP—Nitrate (Red Oil)

Tributyl phosphate is a phosphate ester containing three butyl groups that is added to the

solvent phase as the extractant for uranium and plutonium nitrate complexes. A TBP-solvent

mixture preferentially moves uranium and plutonium from the aqueous (acid) phase to the

organic phase, leaving most of the impurities in the acid stream. In a separate stripping column,

the solvent-TBP-uranium-plutonium mixture is contacted with nitric acid containing a reductant

(HAN). The HAN changes the oxidation state of the plutonium to a form that is only soluble in

the nitric acid, thus allowing its recovery and separation from uranium. The organic phase

containing the uranium is contacted with dilute nitric acid in a separate step that reverses the

equilibria, and the uranium nitrate moves back into the aqueous phase. In the absence of TBP,

no uranium and plutonium would be removed from the aqueous phase, and purification could not

be accomplished.

In the presence of a nitrate of heavy metals or in nitric acid solutions, the TBP will form nitrate

complex compounds that could react exothermically. Exothermic TBP-nitrate reactions are

frequently referred to as red oil reactions because of the reddish color that has been observed in

nitrated TBP/diluent mixtures and residues found during experiments and after the incidents. 
8-42

Red oil is an organic mixture, consisting of TBP and its complexes with plutonium nitrate and

nitric acid, degradation products of TBP (e.g., dibutyl phosphate), and possibly various nitrated

hydrocarbons. The applicant analyzed the risk associated with red oil in units where the

aqueous phase is likely to contain traces of these products above ambient temperatures or

come into contact with a heating source (e.g., acid recovery unit, oxalic mother liquor recovery

unit).

8.1.2.5.5.1 Applicant’s Proposed Approach for Red Oil. The applicant adopted a preventive

safety strategy to protect the worker, site worker, IOC, and environment (revised CAR Section

5.5.2.4.6.7; NRC 2003a, 2003c). Three sets of PSSCs to prevent the overpressurization event 

are identified by the applicant in Chapter 5 of the revised CAR. The PSSCs include (1) the

offgas treatment system, (2) the PSCS, and (3) chemical safety controls. These PSSCs apply to

open and closed systems as described below:

The applicant defined a system’s ability to accommodate a red oil reaction in terms of the

system being open or closed. An open system is defined as one in which the vent area

associated with the offgas treatment system is sufficient to prevent overpressurization, if the

runaway reaction occurs. The offgas treatment system design assumes that the system is filled

with100-percent solvent. A closed system is one in which the vent area is not sufficient to

prevent overpressurization, if the runaway reaction occurs. A closed system will have

substantial quantities of solvent present (estimated at about 40–60 percent; the exact values

would be determined for specific vessels at the ISA stage) and must use an evaporative cooling

correlation.

Offgas Treatment System

For closed systems, the safety function of the offgas treatment system is to provide an exhaust

path for aqueous phase evaporative cooling in process vessels. The following provides the

design basis:

evaporative cooling = 1.2 (steam energy input at 133 °C (271 °F) + energy generated

chemically)

where 1.2 is a safety factor

For open systems, an additional function of the offgas treatment system in open systems is to

provide venting of vessels and equipment that potentially contain TBP and its associated

byproducts to prevent overpressurization, should the runaway reaction occur. The design-basis

value for the vent size is 12.5 kg/cm2 (0.008 mm2/g) of organic material present in the vessel.

Process Safety Control Subsystem

For open and closed systems, there are three safety functions of the PSCS, including to (1)

control the energy generation rate by restricting the bulk temperature of solutions that may

contain degraded organic material to within safety limits, (2) limit the residence time of organic

compounds in the presence of oxidizers (e.g., nitric acid) and radiation fields, and (3) for closed

systems only, ensure an adequate aqueous phase inventory to provide evaporative cooling.
8-43

The design basis for the first safety function is to limit the temperature of steam used to heat

process vessels to 133 °C (271 °F). The proposal to limit steam temperature to 133 °C (271 °F)

should not be interpreted as a small operational margin from the red oil runaway initiation

temperature of 137 °C (279 °F). Steam at this temperature is required to efficiently heat the

nitric acid-water azeotrope to its boiling point of 120 °C (248 °F). The boiling point temperature

should not be exceeded under normal conditions (vented). Therefore, the effective temperature

margin is 17 °C (63 °F). 

The residence time will be a controlled parameter of one or more items relied on for safety

(IROFS) to be identified by the applicant in its ISA summary submitted as part of any later

application for a license to possess and use licensed material.

The design basis for the third safety function is a bulk fluid maximum temperature of 125 °C

(257 °F) with a maximum heatup rate of 2 °C (3.6 °F)/min after startup. Aqueous phase

availability would be assured through continuous feed or injection if either the temperature limit

or ramp rate is exceeded. Detailed evaluation of startup operations will be conducted as part of

the ISA to ensure that an appropriate margin of safety exists to account for regimes where the

temperature rate of change during startup may exceed 2 °C (3.6 °F)/min.

Chemical Safety Controls

The safety function of the chemical safety controls is to ensure that the diluent does not contain

cyclic chain hydrocarbons. The exclusion of cyclic chain hydrocarbons minimizes the

involvement of the diluent in the red oil reactions. The exclusion of degraded organic

compounds prevents the phenomena from initiating at lower temperatures. 

DCS has committed to further evaluate the red oil phenomena, including continuing analyses

and experiments which could result in an increase or decrease of the temperature at which

action is required to remain below the design basis value. DCS is also evaluating the effect of

impurities on the initiation temperature in closed systems (NRC, 2003a).

8.1.2.5.5.2 NRC Review of Red Oil. The staff has reviewed the proposed use of evaporators

for waste processing and reagent recovery at the planned facility. Contact between TBP with

aqueous solutions of nitric acid and acidic solutions of metal nitrates is intrinsic to the PUREX

and similar solvent extraction processes, and evaporator treatment of PUREX streams has been

performed safely since the 1950s. However, several events have occurred involving a range of

conditions. Extended contact between heated mixtures of TBP and nitric acid, nitrates, and/or

heavy metal nitrate salts can form degradation products and intermediates (including nitrated

esters and adducts) that may lead to violent exothermic reactions under certain conditions (Hou,

1996; ORNL, 1988; SRS, 1976; WSRC, 1994a, 1994b, 1994c, 1995). 

Several explosive incidents have occurred in the United States (Savannah River in 1953 and

1975, Hanford in 1953) (ORNL, 1988; WSRC, 1994a, 1994b) and the former Soviet Union

(Tomsk-7 in 1993) (DOE, 1993; IAEA, 1998; OECD, 1994). The Hanford and first Savannah

River (1953) event involved evaporators undergoing heating with low-pressure steam;

temperature control was not precise but ranged from 135–140 °C (275–284 °F). The second

Savannah River incident (1975) involved a denitrator operating at greater than 150 °C (302 °F). 

The pretreatment step for the denitrator is an evaporator operated with controls and a design
8-44

basis temperature not exceeding 135 °C (275 °F). Thus, the TBP and solvent mixture safely

traveled through the evaporator before the event in the denitrator. The incident at Tomsk did not

involve heated equipment; the reactions started in a tank at a nominal temperature of 50 °C (122

°F) with an organic layer estimated at 90 °C (194 °F). Other, less serious incidents have also

occurred. 

The staff notes from its review that these events occurred because of the unanticipated

presence (either carryover or accumulation) of the TBP, solvent, and degradation products. The

applicant indicated that solvent carryover can be considered as an anticipated event in the

facility.

Westinghouse Savannah River Company has summarized much of the experimental information

that is available to understand the phenomena and establish reasonable controls (WSRC,

1994a, 1994b). Experimental work has replicated portions of the observed phenomena. The

important specific reactions are summarized as follows:

• hydrolysis reactions (occur slowly–rates increase with temperatures)

• dealkylation and nitration reactions (occur slowly–rates increase with temperature)

• pyrolysis (becomes significant at about 150 °C (302 °F) with little or no water present)

• butyl alcohol and butyl nitrate nitration/oxidation reactions (become significant at

90–100 °C (194–212 °F))

• TBP nitration/oxidation (becomes significant around 135 °C (275 °F))

• uranium and plutonium adduct reactions (become significant at 15–175 °C (59–347 °F))

These reactions evolve large quantities of non-condensible gases, including the reddish-brown

nitrous oxide fumes observed in several red oil events and experiments. The chemical reactions

are outlined in revised CAR Section 8.5.1.5.2, with the degradation mechanisms in revised CAR

Figure 8.5-2 and the degradation products in revised CAR Table 8-8.

In general, these reactions can occur in aqueous or organic phases. Organic compounds have

a greater affinity for and tend to concentrate in the organic phase, if it is present. Organic

phases also tend to have poorer heat transfer characteristics than aqueous phases. Radiolysis

also produces similar reactions and products, albeit at slower rates. Finally, the staff review

found three basic routes to the red oil phenomena. If volatile species do not evaporate, the

reactions become significant with heating at 90–100 °C (194–212 °F) and reaction runaway

occurs at 135 °C (275 °F). Pyrolysis dominates at higher temperatures if little water is present.

8.1.2.5.5.3 NRC Review of Controls, Limits, and Safety Design Bases Proposed in the

Literature. After a review of the Tomsk accident and additional experimental studies in the early

1990s, DOE researchers recommended conditions for controlling red oil effects and preventing

related explosions at DOE facilities. Paddleford and Fauske (WSRC, 1995) conclude that a

runaway reaction is possible in an open system (i.e., at atmospheric pressure), and self-heating

was observed at temperatures in the vicinity of 130 °C (266 °F). They mention the following
8-45

preventive measures:

• prevent accumulation of significant masses of TBP

• prevent significant interaction with strong nitric acid

• stir (mix) during additions and sampling

• maintain low temperatures

• maintain cooling system effectiveness

• keep TBP/nitric acid mixed with or thermally coupled to aqueous layers

• implement practical vent sizing (0.005 in.2/kg of TBP in nitric acid) 

The derived venting relationship is shown as Figure 8.1-2 and relates vent cross-sectional area

to the mass of organic material present and the internal pressure generated from the red oil

reactions. The recommended safe range is below a value of 32 kg organic compounds/cm2 of

vent area, which essentially prevents pressurization of the vessel containing the red oil/nitric acid

mixture. For comparison, the applicant’s design basis for the offgas treatment system in open

systems is 12.5 kg/cm2 (0.008 mm2/g) of organic material present in the vessel. This compares

to a 32-kg/cm2 limit for rapid pressurization, as shown in Figure 8.1-2. Thus, the applicant’s

design basis for open systems is in the recommended safe range and provides for a safety

margin of about 60 percent below the initiation conditions.

Figure 8.1-2 Empirical venting relationship

The DOE team recommendations include the following (DOE, 1993):

• adequate venting and relief disks on vessels
8-46

• adequate mixing

• vessel containment

• controls with interlocks and independent verification

• minimization of organic compound accumulations in vessels containing high

concentrations of nitric acid

• presence of alarms on temperature and pressure controls

• controls and limits on strong oxidants and reductants

The Tomsk engineers mentioned the following lessons learned (DOE, 1993):

• use tanks capable of withstanding high pressures (20 atm (284 psi))

• limit the concentration of added nitric acid (e.g., for pH adjustments) to 6 M or less

• limit the total nitric acid concentration to 3 M or less

• perform continual air sparging (for mixing)

• maintain a temperature rise of no more than 7 °C (13 °F) in any organic layers

• keep the tank contents below 70 °C (158 °F)

• provide adequate relief valves and rupture disks

Hyder (WSRC, 1994a) mentioned the following lessons learned from the incidents and the

experimental studies:

• Natural processes provide adequate cooling at temperatures below 80 °C (176 °F).

• Above this temperature, care should be taken to ensure that adequate cooling is

available for the amount of TBP present.

• In well-ventilated systems, n-butanol evaporation and convective cooling are sufficient to

control the reactions up to about 120 °C (248 °F) and the decomposition reactions

become very rapid in the range of 130–150 °C (266–302 °F). 

• Confined systems and circumstances that can give rise to confined systems should be

avoided.

• Limits on TBP mass, concentration, or mass per unit surface area should be established. 

The SRS currently uses a concentration limit of 0.5 percent in the agitated (total) feed to

an evaporator.

Hyder (WSRC, 1994b) further elaborated upon potential limits:

• Tanks in which TBP-nitrate complexes are or may be present should be adequately

vented. A value of 0.032 cm2/kg (0.005 in.2/kg or 31 kg/cm2) of TBP is mentioned. This

is based upon Figure 8.1-2.
8-47

• Chemically degraded TBP, or TBP that has sat for a long time in the presence of acids or

radiation, should be purified before use in solvent extraction.

• Evaporators in which TBP might be introduced should be operated at a controlled

temperature and their TBP content should be limited. Local high temperatures should be

avoided. No portion or surface of the evaporator should exceed 130 °C (266 °F), and

operating temperatures substantially lower than this are desirable. If evaporators or

heating units must operate at higher temperatures, the means for preventing the

introduction of TBP must be correspondingly rigorous.

• Evaporator bottoms and other solutions that may contain TBP should be cooled under

conditions that ensure heat removal. Sparging and agitation may also be appropriate.

• Process design and operations should consider the potential for red oil reactions,

particularly for nonroutine operations and situations.

The SRS currently operates evaporators in H Canyon (Jones, 2003). These evaporators are

subject to DOE safety class (i.e., for the public) and safety significant (i.e., for the workers)

controls to prevent a potential red oil runaway reaction and explosion (i.e., frequency under

1.0×10-6/yr). The controls are generally divided into two categories, those that prevent excessive

amounts of TBP entering the evaporators and those that prevent overheating. The latter include

a 120 °C (248 °F) temperature safety limit (used in an analogous manner to the design basis), a

high steam coil pressure interlock, and an alarm for the operator to manually check that the

steamflow has been terminated. The staff notes that the 120 °C (248 °F) temperature limit

corresponds to the normal boiling point of the water-nitric acid azeotrope (Lange, 1999).

The Defense Nuclear Facilities Safety Board (DNFSB) issued a technical report on the control of

red oil explosions (DNFSB, 2003). This report emphasizes controls on temperature, pressure,

mass, and concentration (acid), and the use of effective defense-in-depth practices for the

prevention of a red oil explosion. The following paragraphs summarize the discussions in the

report regarding the controls.

Temperature Control

The report identifies an initiation temperature of 130 °C (266 °F) for the runaway red oil

reactions. Lower temperatures are needed for operationally protected temperatures and

setpoints. H Canyon has an operational protected temperature (technical safety requirement,

similar to the design basis) of 120 °C (248 °F) and an overtemperature safety setpoint of 117 °C

(243 °F) (i.e., 3 °C is allowed for instrument errors and biases). Steam interlocks are set not to

exceed 25 psig, which corresponds to 132 °C (269 °F).

Pressure Control 

The report mentions passive vents in accordance with the Fauske correlation (> 0.063 mm2/g of

TBP (< 15.9 g TBP/mm2 or < 1.59 kg TBP/cm2)). It also mentions 312 g (11 oz) of red oil/mm2

(31.2 kg/cm2). With some safety margin, it recommends no more than 208 g/mm2 (20.8 kg/cm2)

of red oil (DNFSB, 2003, p. 4-3). Page 5-2 of the DNFSB report mentions 41.5 cm2 (6.44 in.2) as

the minimum vent for 1361 kg (3000 lb) of TBP. This corresponds to less than 32.7 kg TBP/cm2
8-48

(465 lb TBP/in.2) of vent. The applicant used 0.008 mm2/g of organic compound (12.5 kg/cm2).

Mass Control 

Mass control is applied to the organic phase to keep it from entering heated equipment

(prevention) or by omitting TBP (or equivalent) mass/concentration (mitigation). 

The applicant used 30 percent TBP in branched dodecane (HPT). There are no organic mass

limits needed for open systems. The applicant stated that organic mass limits for closed

systems will be developed in the license application stage and will be a substantial fraction of the

volume (e.g., 40–60 percent). Organic material entering an open or closed system is considered

an anticipated event.

Concentration Control 

Concentration control applies primarily to nitric acid (less than 10 M) and, to a lesser extent,

nitrate salts/uranyl nitrate (less than 20 percent). The applicant planned to reconcentrate nitric

acid to over 13.6 N. There are no limits on acid concentration.

8.1.2.5.5.4 NRC Staff Findings and Conclusions on Red Oil. The NRC staff has evaluated the

applicant’s proposed controls for the red oil phenomena (NRC, 2004a). For open systems, the

applicant provided PSSCs and design bases. Significantly, the applicant’s design basis for the

offgas treatment system in open systems is 12.5 kg/cm2 (0.008 mm2/g) of organic material

present in the vessel. By comparison to Figure 8.1-2, this is well within the safe range. Thus,

the applicant’s design basis for venting open systems provides for a safety margin of 60 percent

from the initiation conditions. In addition, since the system cannot pressurize easily, the

temperature will not exceed the normal boiling point of the azeotrope (120.4 °C (249 °F)), which

is below the initiation temperature of the runaway red oil reactions. Therefore, the staff finds the

applicant’s approach for open systems acceptable. The remainder of the discussion focuses on

the evaluation of the PSSCs and design bases for closed systems. 

Staff Evaluation of Temperature Limit

The staff evaluated whether the average solution temperature limit of 125 °C (257 °F) is

adequate. The staff evaluated additional literature from SRS which supports a red oil initiation

limit of no greater than 130 °C (266 °F) for evaporators containing nitric acid and TBP. The SRS

H-Canyon Safety Analysis Report (WSRC, 2001c) recommends an ”always safe” limit of less

than 130 °C (266 °F) but cites a temperature of 135 °C (275 °F) at which the initiation of a

runaway reaction between TBP and concentrated (70 wt%) nitric acid was observed to occur. 

Using contemporary methods but similar experimental conditions, others (WSRC, 2001d) have

concluded that some of the Colven data from 1956, which support initiation temperatures as low

as 129 °C (264 °F), are outliers. The minimum initiation temperature found by these authors is

137 °C (279 °F) for 15 M nitric acid.

The applicant has committed to design the acid recovery evaporator system such that the bulk

fluid temperature will not exceed 125 °C (257 °F). This will be accomplished by shutting off the

steam and injecting aqueous phase material into the system. As long as the pressure is

maintained at a pressure of approximately 1 atmosphere, the bulk fluid cannot exceed the nitric
8-49

acid/water azeotrope boiling point, which is 120.4 °C (249 °F). Establishing a vent path that can

provide sufficient mass flow to remove 1.2 times the heat input from the steam and the heat

generated from hydrolysis and an adequate aqueous phase inventory provides the physical

conditions necessary to limit the bulk temperature to approximately 120.4 °C (249 °F). 

Therefore, the applicant has proposed a shutdown temperature margin of 12 °C (137–125 °C)

for the evaporators, plus an additional 20-percent margin for heat removal capacity by the offgas

treatment system.

The 125 °C (257 °F) bulk fluid limit provides operational flexibility. The only way to increase the

system temperature to 130 °C (266 °F) or above is to (1) boil off all of the aqueous phase and/or

(2) pressurize the evaporator by blocking the vent path. With respect to system pressurization,

staff evaluated the change in system pressure which would raise the azeotrope boiling point of

120.4 °C (249 °F) to 125 °C (257 °F). The pressure increase required to raise the temperature

4.6 °C (8.3 °F) is about 10 percent of the ambient offgas treatment system pressure. However,

the availability and reliability of the vent path is not in question, as it will be addressed during the

license application phase. With the steam temperature limited to 133 °C (271 °F) and the bulk

fluid temperature at 125 °C (257 °F), the total heat input from the steam will only be

approximately 3 percent of the heat generated by the chemical reaction because of the small

delta-T driving force. Cool aqueous phase will be injected into the system should the bulk

temperature or heatup rates be exceeded. 

Past DOE experiments (WSRC, 1998) have analyzed steam heating of TBP/nitrogenparaffin/nitric acid mixtures to determine whether a mixture of steam and air from an empty tank

could heat the organic layer until an exothermic, runaway reaction could take place. A jet of

superheated steam (90 percent) and air (10 percent) at 139 °C (282 °F) was used. The jet was

cooled to the organic layer temperature, which reached a maximum temperature of 128 °C (262

°F). The organic layer was effectively limited to this temperature through evaporative cooling. 

This work demonstrates that red oil reactions are not susceptible to source point initiation

caused by minor temperature variations that may be encountered during operations.

The staff inquired about the likelihood for preventing organic compounds inadvertently entering

nitric acid recovery systems (e.g., Nomura, 1994). Since the applicant has not committed to

PSSCs to prevent the introduction of organic mass into the evaporators, the staff requested

information concerning the potential for foaming, two-phase flow and, ultimately, pressurization

of the evaporator by blockage of the offgas treatment system vent (NRC, 2003a). The applicant

committed to developing a fundamental understanding of the system by evaluating the

mechanism and behavior of such events through modeling and experimentation, as needed

(NRC 2003a). This fundamental understanding is intended to allow a determination of the

appropriateness of the relationship of the vent area-to-mass organic ratio, including the potential

for two-phase flow.

The staff considered the safety approach used for the SRS H-Canyon evaporators. The DOE

identified the red oil initiation temperature for 14–15 M nitric acid in closed systems as 137 °C

(279 °F) and has identified an always safe solution temperature limit of 130 °C (266 °F) for its

systems (WSRC, 2001c). An always safe value of 130 °C (266 °F) ensures that the red oil

initiation temperature lies above this value for all foreseeable acid concentrations in either open

or closed systems. To protect this always safe value, operational limits (safety class set points,

as specified in the SRS technical safety requirements) of 120°C (248 °F) are placed on solutions
8-50

in H Canyon evaporators. However, the nitric acid solution concentrations are typically less than

50 wt%. This is lower than the 67 wt% azeotrope concentration that applicant proposed as part

of its nitric acid recovery system. As a result, the 120 °C (248 °F) limit is sufficiently above the

actual boiling point of the H Canyon nonazeotropic solutions, which allows the evaporators to

both operate efficiently and maintain an adequate safety margin. Since the applicant proposed

to recover the azeotrope, an always safe (not to exceed) value of 125 °C (257 °F) allows the

recovery of the azeotrope (B.P. = 120.4 °C (249 °F)) while maintaining a 12 °C (22 °F) margin

from the 137 °C (279 °F) initiation temperature for 14–15 M nitric acid in closed systems. The

applicant will establish safety set points, with margin, less than the 125 °C (257 °F) design-basis

value as part of any later application it may file for a license to possess and use licensed

material.

The Shape of the Evaporator

The staff evaluated the shape of the proposed acid recovery evaporators. Specifically, the staff

evaluated the hypothesis that a tall and thin (high aspect ratio) thermosiphon evaporator would

have a high likelihood of inducing the separation of organic and aqueous phases. This potential

would be highest during startup and shutdown modes of operation. However, the condition of

two separate phases is already assumed in the safety assessment (i.e., the red oil phenomena

would occur in a separate organic phase after its contact with nitric acid). This organic phase

could be lighter than the aqueous phase and float on the surface. Alternatively, the organic

phase may be heavier than the aqueous phase (because of complexation of the TBP with heavy

metals, such as uranium or plutonium) and be located on the bottom, which is a condition

referred to as phase inversion. In either case, the safety functions of the proposed red oil

controls do not rely on the miscibility of the organic phase with aqueous solutions. The applicant

assumed that a distinct and separate organic phase may be present that requires temperature

controls, venting, active cooling, and antifoaming controls. In its November 2003 report, the

DNFSB (DNFSB, 2003) noted that “if solution temperature sensors are used it is important that

they be located such that the organic phase temperature can be measured with or without phase

inversion.” The staff agrees in principle with this recommendation as important to the final

design of temperature measurement IROFS. Information on the placement of temperature

measurement devices is not required at the CAR stage.

November 2003 DNFSB Technical Report

On November 13, 2003, the DNFSB issued a technical report, “Control of Red Oil Explosions in

Defense Nuclear Facilities, 2003” (DNFSB, 2003). In the report, DNFSB recommends that two

or more of four controls should be used to prevent a runaway red oil reaction and explosion of

the detonable gases produced by the reaction, as described above in Section 8.1.2.5.5.3.

In describing the temperature control of the evaporators (Section 4.1 of the report), the DNFSB

staff provided the following specific recommendation:

...to be assured that red oil conditions are not present in an evaporator, controls

for temperature, pressure, and concentration should all be utilized. (p. 4-3)
8-51

MFFF Evaporators

Of the four controls listed above, the applicant identified only temperature controls for the closed

acid recovery evaporators at the MOX facility. Under the definition of a closed system, the

pressure controls in these vessels are not adequate to prevent an overpressurization event. In

addition, mass controls, though present in the design, are not credited in the safety assessment

as PSSCs, and there are no limits on nitric acid concentration.

However, the staff concludes that the applicant provided sufficient defense-in-depth provisions

by committing to a multitiered approach that includes (1) a combination of multiple, independent

temperature controls, (2) adequate aqueous phase evaporative cooling provided by the offgas

treatment system, (3) the exclusion of cyclic chain hydrocarbons, and (4) the commitment to

perform additional research on the runaway initiation temperature and the effect of impurities on

the initiation temperature.

For the CAR stage, the applicant provided sufficient controls and margin such that the bulk

temperature will not exceed 125 °C (257 °F). This temperature is 9–15 °C (16–27 °F) below the

134–140 °C (273–284 °F) range of experimentally measured runaway initiation temperature data

(which average 137 °C (279 °F)). The applicant’s proposed aqueous injection system extends

beyond the safety requirements at DOE facilities and the operating French MOX facility. Other

possible operational concerns related to the evaporator startup, shutdown, and possible

abnormal conditions are best addressed in the ISA when more specific design information is

expected to be available. 

Consequently, the staff finds the applicant’s approach acceptable for the CAR stage. This

closes Open Item CS-01 from the revised DSER (NRC, 2003d).

8.1.2.5.6 Plutonium (VI) Oxalate

Plutonium can exist in several different oxidation states in aqueous solutions. The most

common are the +3, +4, and +6 states. The applicant intended to precipitate plutonium (IV)

oxalate and expeditiously calcine it to the plutonium dioxide powder, in alignment with good

handling practices for plutonium (DOE, 1998b), for subsequent MOX fuel manufacture. 

Plutonium (VI) forms slowly in nitric acid solutions and may also be present if the valence

adjustments are not performed correctly. Upon heating, plutonium (VI) oxalate (PuO2C2O4)

exhibits an endothermic peak around 142 °C (288 °F) because of dehydration (it is usually

present as the trihydrate) and a subsequent rapid exothermic peak at approximately 219 °C (426

°F) associated with oxidation of the oxalate. Section 8.5.1.7 of the revised CAR mentions a

potential explosive hazard with plutonium (VI) oxalate introduction into the calcining furnace and

its subsequent rapid decomposition. As described in revised CAR Section 5.5.2.4.6.12, the

safety strategy is prevention using the PSSC of chemical safety control. The safety function is

to perform a measurement of the valency of the plutonium before adding oxalic acid to it in the

oxalic precipitation and oxidation unit. If plutonium (VI) is detected, feed would be terminated.

The revised CAR also indicates that the design basis is to prevent the introduction of

plutonium (VI) oxalate into heated equipment where temperatures in excess of 219 °C (426 °F)

are credible. In addition, controls will be in place to ensure that temperatures do not exceed 219

°C (426 °F) where plutonium (VI) oxalate might be present (e.g., the oxalic mother liquor

recovery unit and the oxalic precipitation and oxidation unit). 
8-52

The applicant will develop a measurement system for plutonium (VI) as part of the final design 

process. Analytical methods under consideration include ultraviolet-visible spectrum and alpha

spectrum spectroscopy, with detection/resolution limits of 1×10-5 and 1×10-9 M for plutonium (VI),

respectively (NRC, 2003a, 2003c).

The applicant also provided information on the design basis for the calciner (NRC, 2003a,

2003c). According to Section 11.8 of the revised CAR, the design basis pressure criteria for

storage tanks, process columns, and exchangers are the greater of maximum pressure acting at

the top of the vessel in normal operating conditions or the maximum pressure acting at the top of

the vessel during transient or accident conditions, plus 10 percent. The applicant committed to

the same design-basis criteria for the calciner furnace.

The staff notes the detection limits of plutonium (VI) are very low regardless of the method used

and effectively prevent significant quantities from entering the calciner (i.e., the safety strategy of

prevention). In addition, the design basis of the calciner (the maximum pressure plus

10 percent) assures that pressure increases from the rapid decomposition of any plutonium (VI)

oxalate that remains after detection will not challenge the vessel’s integrity. The staff review

also found that the highest temperature in any piece of AP equipment other than the calciner is

135 °C (275 °F); this provides a significant margin vis-a-vis the exotherm temperature of

approximately 219 °C (426 °F). The staff finds the approach acceptable for the construction

authorization stage.

8.1.2.6 Habitability Issues

The purpose of this section is to review the criterion used by the applicant to ensure the

habitability of the facility during routine operations and during an accident or event. 

8.1.2.6.1 Design Approach

The applicant stated that facility workers mainly perform a monitoring role during emergency

conditions. The applicant identified the ECRACS as the PSSC to allow the performance of the

monitoring safety role. The description of the ECRACS is given in Section 11.4.2.7 of the

revised CAR. The following safety functions are attributed to the ECRACS:

• maintain a habitable environment in each of the two ECRs for the facility personnel

• provide cooling to the emergency electrical rooms

The system will include two 100-percent capacity air filter trains, one for each ECR. Each unit

consists of a filtration unit and a booster fan for each intake. Each filter housing contains a

hazardous gas removal cartridge and/or an organic vapor cartridge, and high-efficiency

particulate air (HEPA) filter cartridges. The filter housings are manufactured from stainless

steel.

The control strategy is provided as follows:

• Each ECR intake is continuously monitored for hazardous chemicals.
8-53

• Upon detection of a hazardous chemical above allowable limits, the intake is

automatically isolated and switched to the recirculation mode using a filtration unit with

HEPA filtration and hazardous gas removal elements. 

• An alarm sounds if hazardous chemical concentrations are detected at both intakes.

• The alarm alerts operators to don emergency self-contained breathing apparatuses

(SCBAs).

The applicant stated that monitoring would be performed for those chemicals for which

unmitigated release could result in control room concentrations exceeding the limits in

Table 8.1-12. Specific setpoints for actuation of air inlet isolation valves and ventilation

recirculation would be determined during final design. Initial calculations by the applicant have

identified releases of hydrazine or nitrogen tetraoxide as having the potential for resulting in

concentrations at or above these limits. Calculations would be made during final design to verify

the list of chemicals monitored.

The table contains limits from two different sources. The preferred limit is the IDLH value from

NIOSH/OSHA. If a TEEL-3 value is lower than the IDLH value for a given chemical, the

applicant intended to apply the TEEL-3 limit for that chemical. The applicant will use TEEL-2

limits where no IDLH value exists. 

The applicant has also proposed standards for breathing air (revised CAR Section 11.9.1.11)

that would be used as design information for the chemical processes that involve chemicals

which could displace oxygen (i.e., the chemicals are not toxic or hazardous), thus decreasing its

concentration in the air. For operating procedures to avoid asphyxiating atmospheres, the

applicant specified that the high ventilation rates preclude the creation of an asphyxiating

atmosphere (DCS, 2001). The applicant referenced the CGA publication, “Accident Prevention

in Oxygen Rich and Oxygen-Deficient Atmospheres,” issued 1992 (CGA, 1992b), as guidance.

The applicant proposed two standards to assure the quality of breathable air:

• CGA G-7.1, “Commodity Specification for Air,” issued 1992 (CGA, 1992a)

• American National Standards Institute (ANSI) Z88.2-92, “American National Standard for

Respiratory Protection,” issued 1992 (ANSI, 1992)

8.1.2.6.2 Asphyxiating and Hazardous Atmospheres

The excess or lack of oxygen may become the initiators of additional accidents. For example,

nonflammable materials can become flammable with more oxygen present, while operating

personnel can suffer anoxia and experience difficulties controlling the plant with less oxygen

present. If the atmospheric concentration of oxygen exceeds 23 percent, it is called an oxygenrich atmosphere (CGA, 1992b, CGA P-14, Section 2.1). An oxygen-deficient atmosphere is one

that contains less than 19.5 percent of oxygen (CGA, 1992b, CGA P-14, Section 2.2). An

atmosphere containing less than 19.5-percent oxygen can cause problems in the cognitive

abilities and impair the operational capacity of personnel. Moreover, the exposure to insufficient

concentrations of oxygen for a prolonged time may cause convulsions, unconsciousness, and
8-54

death. It is important to consider that the exposure effects may vary depending on the individual

(CGA, 1992b, 2001).

With regard to design basis values for chemicals that can cause a hazardous atmosphere, the

applicant will use the values in Table 8.1-12 based upon IDLHs, TEEL-3s, and TEEL-2s. The

applicant committed (DCS, 2002e) to actively monitor the ECRACS air intakes for hazardous

chemicals that could exceed Table 8.1-12 values at the intakes. 

8.1.2.6.3 Habitability Standards

Regulatory guidance on control room habitability and ventilation systems is summarized as

follows:

• Regulatory Guide (RG) 1.78, “Assumptions in Evaluating the Habitability of a Nuclear

Power Plant Control Room During Postulated Hazardous Release,” issued 2001 (NRC,

2001a), provides general guidance for the evaluation of control room habitability. This

document has information about hazardous chemicals and the toxic limits of some

components. Although the information included in this document is more qualitative than

quantitative, it depicts some information that would be acceptable to the NRC as control

room design bases. The document concludes that a chemical exposure period not

exceeding 2 minutes at the IDLH value provides sufficient margin (i.e., as compared to

the maximum 30-minute exposure and different functions (donning SCBAs as compared

to evacuation) assumed in the derivation of IDLH values).

• ANSI/ASHRAE 62-1989, “Ventilation for Acceptable Indoor Air Quality,” issued 1990

(ASHRAE, 1990), includes tables with information about outdoor air requirements to

achieve the stipulated indoor air quality. These requirements only apply to some parts of

the facility (e.g., offices, lobbies, reception areas, cafeterias, conference rooms,

restrooms, lockers and dressing rooms, and elevators). Appendix C includes some

standards and dose limits that can be used as bases for air pollutants (e.g., carbon

monoxide, nitrogen dioxide, and particulates). The dose limits are depicted in terms of

short-term exposure limits, threshold limit values, and weighted average concentrations. 

Although this information could be used for air ventilation design, it is important to

remember that the document specifies that these tables “are not part of the standard,”

but are “included in the standard only for information purposes.”

• The National Ski Patrol document, “Components and Hazardous Atmospheres

(Prehospital Emergency Care & Crisis Intervention),” issued 1992 (NSP, 1992), provides

guidance for constituents in air considered dangerous for human beings, based upon a

30-minute or longer exposure.

Most of the limits shown in ANSI/ASHRAE 62-1989 are included in 29 CFR 1910.1000, “Air

Contaminants.”

8.1.2.6.4 Staff Conclusions on Habitability

The staff notes that the applicant identified a safety function for workers in the ECRs. To protect

workers in the ECRs, the applicant identified a safety function for the ECRACS to maintain
8-55

habitable conditions. The staff concludes that it is reasonable and prudent to staff the ECRs and

maintain habitable conditions, such that the facility can be monitored and, if necessary,

emergency actions can be taken (NRC, 2001a, 2003g). Such approaches are often used in the

nuclear and process industries. The staff finds the top-level safety strategy acceptable.

The revised CAR states that the safety function is to maintain habitable conditions for operators. 

The ECRACS is identified as the PSSC. The revised CAR, page 8-11, states, “Monitoring will be

performed for those chemicals whose unmitigated release could result in control room

concentrations above the limits specified in Table 8-12.” Table 8.1-12 contains concentrations

that NRC RG 1.78 indicates are only suitable for 2-minute exposures while the control room staff

dons protective clothing and SCBA.

Consequently, these limits (Table 8.1-12, the design basis for ECR habitability) do not represent

habitable concentrations. However, the applicant also identified chemical concentration levels

for accident consequence categorization (FSER Table 8.1-5). Level 1 on this table represents

“[t]he maximum concentration in air below which it is believed nearly all individuals could be

exposed without experiencing other than mild transient adverse health effects or perceiving a

clearly defined objectionable odor.” The staff notes that Level 1 would not impair the activities of

operators in the ECR and approximates a habitable condition. The staff concludes the applicant

should maintain hazardous chemical concentrations below Level 1 of FSER Table 8.1-5 for the

duration of potential hazardous chemical events and proposes the following condition:

Construction Authorization Condition

In addition to the safety functions and design bases specified in the revised CAR,

the facility will be designed so that a safety function of the Emergency Control

Room (ECR) Air Conditioning System shall be to maintain hazardous chemical

concentrations in each ECR below CAR Table 8-5 Level 1 concentrations for the

duration of credible hazardous chemical release events. 

The CAR Table 8-5 Level 1 values use an exposure duration basis of 1 hour. As noted in

Section 8.1.6.1.2 of this FSER, longer ECR durations might require lower chemical

concentrations. The staff review of the revised CAR (Section 11.4) found that the ECRACS will

include specific components for removing hazardous chemicals. The staff found that the current

state-of-the-art technology of scrubbers and absorbers (e.g., Perry, 1997) likely has the ability to

design and fabricate these components to meet the desired chemical concentration levels

corresponding to habitability. In addition, technology that is considered to be state-of-the-art

implies the ability for these systems to be built with redundant and diverse components and

independent trains and, thus, the ability to meet high reliabilities corresponding to highly unlikely

failure rates. The staff expects the specific details of meeting the FSER Table 8.1-5 Level 1

values (or lower values for longer durations, if appropriate) to be demonstrated by the applicant

in the ISA summary, which would be submitted as part of any later application for a license to

possess and use licensed material. The staff concludes that the applicant has now provided

suitable habitability design bases (for oxygen levels, impurities, and hazardous materials) that

are consistent with its proposed approach and provide sufficient margin. This closes Open Item

CS-10 from the staff’s revised DSER (NRC, 2003d).
8-56

Table 8.1-12 Chemical Limits Proposed by the Applicant for Control Room Consequence

Calculations at the MFFF (mg/m3)

Chemical Name Applicant’s Proposed

Concentration Limit (mg/m3)

Source of Limit NRC Staff

Observation

Aluminum Nitrate 15 TEEL-2 No IDLH

Azodicarbonamide 500 TEEL-2 Rev 18; Rev 19

TEEL-2 is 200

Chromic (III) Acid 25 (as CrIII) IDLH IDLH

Chlorine 29 IDLH IDLH

Diluent (C10–C13 Isoalkanes) 35 TEEL-2 No IDLH;

dodecane

TEEL-2 = 80

Ferrous Sulfate 12.5 TEEL-2 No IDLH

Fluorine 30 TEEL-3 IDLH = 40

Hydrazine Monohydrate 0.06 TEEL-2 No IDLH

Hydrazine Nitrate 5 TEEL-2 No IDLH

Hydrofluoric Acid 25 IDLH IDLH

Hydrochloric Acid 75 IDLH IDLH

Hydrogen Peroxide 106 IDLH IDLH

Hydroxylamine Nitrate 26 TEEL-2 No IDLH

Iron 500 TEEL-3 IDLH = 2,500

Isopropanol 5000 IDLH IDLH (also 10%

of LEL)

Manganese Nitrate 500 (as Mn) IDLH IDLH

Manganous Sulfate 500 (as Mn) IDLH IDLH

Nitric Acid 66 IDLH IDLH

Nitric Oxide 125 IDLH IDLH

Nitrogen Dioxide 35 TEEL-3 IDLH = 40

Nitrogen Tetraoxide 15 TEEL-2 No IDLH

Oxalic Acid 500 IDLH IDLH

Potassium Permanganate 15 TEEL-2 No IDLH

Silver Nitrate 10 (as Ag) IDLH IDLH

Silver Oxide 10 (as Ag) IDLH IDLH

Sodium 5 TEEL-2 No IDLH

Sodium Carbonate 50 TEEL-2 No IDLH

Sodium Hydroxide 10 IDLH IDLH

Sodium Nitrite 1 TEEL-2 No IDLH

Sodium Sulfite 50 TEEL-2 No IDLH

Sulfuric Acid 15 IDLH IDLH

Sulfamic Acid 250 TEEL-2 No IDLH

Thenoyl Trifluoroacetone 25 TEEL-2 No IDLH

Tributyl Phosphate 300 TEEL-3 IDLH = 330

Uranium Dioxide 10 (as U) IDLH IDLH

Uranyl Nitrate 10 (as U) IDLH IDLH

Xylene 3900 IDLH IDLH

Zinc Stearate 50 TEEL-2 No IDLH

Zirconium Nitrate 10 (as Zr) IDLH IDLH
8-57

In the license application, the applicant must demonstrate that habitable conditions can be

maintained in the ECRs (i.e., the CAR Table 8-5 Level 1 values) or propose and demonstrate an

alternative.

8.1.2.7 Baseline Design Criteria

The applicant stated that the BDC specified in 10 CFR 70.64(a) are incorporated into the design

and operation of the proposed facility (revised CAR Section 5.5.5.4). The applicant stated that

information demonstrating compliance with these criteria is provided in the applicable chapters

of the revised CAR. For chemical protection, 10 CFR 70.64(a)(5) states the following:

Chemical protection. The design must provide for adequate protection against

chemical risks produced from licensed material, facility conditions which affect the

safety of licensed material, and hazardous chemicals produced from licensed

material.

Chapter 8 of the SRP (NRC, 2000) contains guidance and references to other peer-reviewed

work on the subject of chemical safety. The applicant indicated that reagents are stored and

chemical mixtures are prepared in the reagent processing building and the reagent storage part

of the AP area. They are generally separated from each other and radioactive materials. The

applicant will avoid mixing incompatible materials by design, controls, and procedures. The AP

and MP facilities are broken down into process functional units, which are made up of one or

more subassemblies performing consistent and elementary tasks. The applicant stated that the

breakdown into control functional units allows each entity to be operated relatively independently

in the given operating mode. The staff review notes that this separation and independence is

consistent with accepted practices for safe operations.

The applicant will control process storage and operation conditions in order to prevent

exothermic and potential autocatalytic reactions in the reagent processing building and the AP

and MP areas. Autocatalytic and exothermic reactions of chemicals would be prevented through

control of the process parameters (e.g., reactant concentration, temperature, catalyst

concentration in solution, and pressure) that affect the reactions. Many of these controls have

been identified as PSSCs with design bases. Codes and standards have been applied on a

specific level, often as design bases of PSSCs.

The applicant stated that there is reasonable assurance that the PSSCs will be sufficiently

reliable and available. This assurance will be provided through the use of standard nuclear

industry engineering practices (e.g., reasonably and generally accepted good engineering

practices (RAGAGEP)). These practices are incorporated into the facility general design

philosophy, design bases, system design, and commitments to applicable management

measures. These practices ensure that applicable industry codes and standards are utilized,

adequate safety margins are provided, engineering features are utilized to the extent practicable,

the defense-in-depth philosophy is incorporated into the design, and PSSCs will be appropriately

maintained.

The staff review finds that the applicant has provided sufficient information to meet the

requirements of 10 CFR 70.64(a)(5). The staff review using the SRP notes that the previously

identified open items in MP have been closed by additional safety strategies and information
8-58

from the applicant, and the staff concludes that the applicant has satisfied this BDC.

Related to chemical protection, the explosion protection BDC is stated as part of the fire

protection BDC in 10 CFR 70.64(a)(3):

Fire protection. The design must provide for adequate protection against fires

and explosions.

Chapters 7 and 8 of the SRP (NRC, 2000) describe the fire protection/explosion BDC and

include guidance and references to other peer-reviewed work on the subject. Guidance, codes,

and standards have been applied on a specific level, often as design bases of PSSCs.

The applicant stated that there is reasonable assurance that the PSSCs will be sufficiently

reliable and available. This assurance will be provided through the use of standard nuclear

industry engineering practices (e.g., RAGAGEP). These practices are incorporated into the

facility general design philosophy, design bases, system design, and commitments to applicable

management measures. These practices ensure that applicable industry codes and standards

are utilized, adequate safety margins are provided, engineering features are utilized to the extent

practicable, the defense-in-depth philosophy is incorporated into the design, and PSSCs will be

appropriately maintained.

The staff review finds that the applicant has provided sufficient information to meet the

requirements of 10 CFR 70.64(a)(3). The staff review using the SRP notes that the previously

identified open items related to MP have been closed by additional safety strategies and

information from the applicant, and the staff concludes that the applicant has satisfied this BDC.

8.1.3 Chemical Process Safety Interfaces

The applicant will make controls established for chemical safety consistent with those

established for radiological safety and criticality safety, along with the associated management

measures. Accordingly, the Chemical Safety Program will be conducted under the same

elements of programmatic infrastructure described in Chapters 4, 12, and 14 of the revised CAR

and interfaces with the management measures discussed in Chapter 15 of the revised CAR.

8.2 Evaluation Findings

In revised CAR Chapters 5, 8, and 11, DCS provided design basis information for chemical

process safety PSSCs identified for the proposed facility. Based on the staff’s review of these

revised CAR chapters and supporting information provided by the applicant relevant to chemical

process safety, the staff finds, for the reasons discussed above, that DCS has met the BDC set

forth in 10 CFR 70.64(a)(3) for explosions and 10 CFR 70.64(a)(5) for chemical safety. Further,

the staff concludes, pursuant to 10 CFR 70.23(b), that the design bases of the PSSCs identified

by the applicant will provide reasonable assurance of protection against natural phenomena and

the consequences of potential accidents.

The following previously identified open items in the April 30, 2002, and April 30, 2003, DSERs

(NRC, 2003d) have now been closed in a satisfactory manner, including CS-01, CS-02, CS-03,

CS-05b, CS-9, and CS-10.
8-59

8.3 References

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Commission, RE: Clarification of Responses to NRC Request for Additional Information. 

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NRC-000083. February 11, 2002.

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Commission, RE: Clarification of Responses to NRC Request for Additional Information. 

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Commission, RE: Clarification of Responses to NRC Request for Additional Information. April

23, 2002.

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Commission, RE: Requests for Additional Information, Clarifications, and Open Item Mapping

into the Construction Authorization Request Revision. November 22, 2002.

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Commission, RE: Responses to Financial Qualification, Fire Safety, Chemical Safety, Aqueous

Processing, Material Processing and Ventilation Open Items/Additional NRC Questions on

Construction Authorization Request (CAR) Revision. February 18, 2003.

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Commission, RE: Response to DSER Open Items AP-3 and CS-3. May 23, 2003.

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Commission, RE: Response to DSER Open Items AP-10 and CS-2. May 30, 2003.

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Commission, RE: Meeting Minutes from the 02–04 June 2003 Public Meeting on Chemical

Safety Open Items. June 5, 2003.

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Commission, RE: Response to DSER Open Item CS-2. July 28, 2003.

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Commission, RE: Response to Request for Additional Information—Chemical Safety Open

Items CS-09, AP-02, AP-08, and AP-09. September 29, 2003.

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Commission, RE: Response to Request for Additional Information—Chemical Safety Open

Items CS-01 and CS-02. October 6, 2003.

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Russia, June 19–29, 1993.” Washington, DC, September 1993.

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Investigation Board Report on the May 14, 1997 Chemical Explosion at the Plutonium

Reclamation Facility, Hanford, Washington.” Washington, DC, July 26, 1997.

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Investigation Board Report on the May 14, 1997 Chemical Explosion at the Plutonium

Reclamation Facility, Hanford, Washington.” Washington, DC, July 26, 1997.

(DOE, 1998a) U.S. Department of Energy. DOE/EH-0555, “Technical Report on

Hydroxyloamine Nitrates.” Washington, DC, February 1998.

(DOE, 1998b) U.S. Department of Energy. DOE Std 1128-98, “Guide of Good Practices for

Occupational Radiological Protection in Plutonium Facilities.” Washington, DC, June 1, 1998.

(DOE, 2004) U.S. Department of Energy. “SCAPA.” http://www.orau.gov/emi/scapa/

(November 15, 2004)

(EPA, 1999) U.S. Environmental Protection Agency. “ALOHA: Areal Locations of Hazardous

Atmospheres, User’s Manual.” Washington, DC, January 1999.

(EPA, 2004) U.S. Environmental Protection Agency. “Acute Exposure Guideline Levels.” July

7, 2004. http://www.epa.gov/oppt/aegl/chemlist.htm (November 15, 2004)

(Genium, 1999) Genium’s Handbook of Safety, Health, and Environmental Data for Common

Hazardous Substances. McGraw Hill: New York, 1999.

(Hou, 1996) Hou, Y., E.K. Barefield, D.W. Tedder, and S.I. Abdel-Khalik. “Thermal

Decomposition of Nitrated Tributyl Nitrate,” Nuclear Technology. Vol. 113, pp. 304–315. March

31, 1996.

(IAEA, 1998) International Atomic Energy Agency. “The Radiological Accident in the

Reprocessing Plant at Tomsk.” Vienna, Austria, 1998.

(Jones, 2003) Jones, R.P. “A Potential Inadequacy Related to Operator Response Time

Potentially Inadequate t Effectively Respond to an Analyzed Condition.” U.S. Department of

Energy, Occurrence Report Number SR-WSRC-HCAN-2003-0006. March 21, 2003.

(Lange, 1999) Lange’s Handbook of Chemistry, 15th Edition. McGraw Hill: New York, 1999.

(Lewis, 1996) Lewis, R.J., Sr. Sax’s Dangerous Properties of Industrial Materials, 9th Edition. 

Van Nostrand Reinhold: New York, 1996.

(Malachowski, 1999) Malachowski, M.J. Health Effects of Toxic Substances, 2nd Edition. 

Government Institutes, a Division of ABS Group, Inc. Rockville, Maryland, 1999.

(NFPA, 1997) National Fire Protection Association. NFPA Standard 69, “Standard on Explosion
8-62

Prevention Systems.” 1997.

(NFPA, 1998) National Fire Protection Association. NFPA Standard 801, “Standard for Fire

Protection for Facilities Handling Radioactive Materials.” 1998.

(NFPA, 2000a) National Fire Protection Association. NFPA Standard 50A, “Gaseous Hydrogen

Systems at Consumer Sites.” 2000.

(NFPA, 2000b) National Fire Protection Association. NFPA Standard 86C, “Industrial Furnaces

Using a Special Processing Atmosphere.” 2000. 

(NFS, 2002) Moore, B.M., Nuclear Fuel Services, Inc. Letter to G. Janosko, U.S. Nuclear

Regulatory Commission, RE: ISA Summary for the BLEU Preparation Facility Processes. 

October 14, 2002.

(NIOSH, 1994) National Institute for Occupational Safety and Health. “NIOSH Pocket Guide to

Chemical Hazards.” June 1994.

(NIOSH, 2004) National Institute for Occupational Safety and Health. 

http://www.cdc.gov/niosh/homepage.html (November 15, 2004) 

(Nomura, 1994) Nomura, Y., R. Leicht, and P. Ashton. “Fault Tree Analysis of System Anomaly

Leading to Red Oil Explosion in Plutonium Evaporator,” Journal of Nuclear Science and

Technology. Vol. 31, No. 8, pp. 850–860. August 1994.

(NRC, 1992) U.S. Nuclear Regulatory Commission. Information Notice 92-14, “Uranium Oxide

Fires at Fuel Cycle Facilities.” Washington, DC, February 21, 1992.

(NRC, 1998a) U.S. Nuclear Regulatory Commission. NUREG/CR-6410, “Nuclear Fuel Cycle

Facility Accident Analysis Handbook.” Washington, DC, March 1998.

(NRC, 1998b) U.S. Nuclear Regulatory Commission. “Memorandum of Understanding between

the Nuclear Regulatory Commission and the Occupational Safety and Health Administration: 

Worker Protection at NRC-licensed Facilities.” Federal Register, Vol. 53, No. 210, pp.

43950–43951. October 31, 1998.

(NRC, 1999) Pierson, R., U.S. Nuclear Regulatory Commission. Memorandum to D. Clark

Gibbs, U.S. Department of Energy, RE: Estimates of Hydrogen Generation From Wastes at the

Proposed, TWRS-P Facility. April 21, 1999.

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility.” Washington,

DC, August 2000. 

(NRC, 2001a) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.78, “Assumptions in

Evaluating the Habitability of a Nuclear Power Plant Control Room During Postulated Hazardous

Release.” Washington, DC, 2001.
8-63

(NRC, 2001b) Brown, D., U.S. Nuclear Regulatory Commission. Memorandum to E.J. Leeds,

U.S. Nuclear Regulatory Commission, RE: 2/21–22/02 In-Office Review Summary of DCS

Construction Authorization Request Supporting Documents for the MFFF. March 11, 2001.

(NRC, 2001c) Persinko, A., U.S. Nuclear Regulatory Commission. Memorandum to E.J. Leeds,

U.S. Nuclear Regulatory Commission, RE: 11/27–29/01 In-Office Review Summary of DCS

Construction Authorization Request Supporting Documents for the MFFF. December 18, 2001.

(NRC, 2002a) Persinko, A., U.S. Nuclear Regulatory Commission. Memorandum to E.J. Leeds,

U.S. Nuclear Regulatory Commission, RE: 2/22/02 Phone Call Summary: DCS Construction

Authorization Request Supporting Documents for the MFFF. February 28, 2002.

(NRC, 2002b) U.S. Nuclear Regulatory Commission. NUREG-1520, “Standard Review Plan for

the Review of a License Application for a Fuel Cycle Facility.” Washington, DC, March 2002.

(NRC, 2002c) U.S. Nuclear Regulatory Commission. Draft Regulatory Guide 1114, “Control

Room Habitability at Nuclear Power Reactors.” Washington, DC, March 2002.

(NRC, 2002d) Persinko, A., U.S. Nuclear Regulatory Commission. Memorandum to M.N.

Leach, U.S. Nuclear Regulatory Commission, RE: August 28–30, 2002, In-Office Review

Summary of DCS Construction Authorization Request Supporting Documents for the MFFF. 

December 31, 2002. 

(NRC, 2003a) Persinko, A., U.S. Nuclear Regulatory Commission. Memorandum to M.N.

Leach, U.S. Nuclear Regulatory Commission, RE: December 10–12 Meeting Summary: 

Meeting with DCS to Discuss Mixed Oxide Fuel Fabrication Facility Revised Construction

Authorization Report. January 31, 2003.

(NRC, 2003b) Persinko, A., U.S. Nuclear Regulatory Commission. Memorandum to M.N.

Leach, U.S. Nuclear Regulatory Commission, RE: January 15–16 Meeting Summary: Meeting

with DCS to Discuss Mixed Oxide Fuel Fabrication Facility Revised Construction Authorization

Report. March 5, 2003.

(NRC, 2003c) Persinko, A., U.S. Nuclear Regulatory Commission. Memorandum to M.N.

Leach, U.S. Nuclear Regulatory Commission, RE: February 6–7 Meeting Summary: Meeting

with DCS to Discuss Mixed Oxide Fuel Fabrication Facility Revised Construction Authorization

Report. March 5, 2003.

(NRC, 2003d) Pierson, R., U.S. Nuclear Regulatory Commission. Letter to R. Idhe, Duke

Cogema Stone & Webster, RE: Draft Safety Evaluation Report on Construction of Proposed

Mixed Oxide Fuel Fabrication Facility, Revision 1. April 30, 2003.

(NRC, 2003e) Persinko, A., U.S. Nuclear Regulatory Commission. Memorandum to K. Gibson,

U.S. Nuclear Regulatory Commission, RE: June 2–4, 2003, Meeting Summary: Meeting with

Duke Cogema Stone & Webster to Discuss Chemical Safety Related to Mixed Oxide Fuel

Fabrication Facility. July 18, 2003.

(NRC, 2003f) Persinko, A., U.S. Nuclear Regulatory Commission. Memorandum to K. Gibson,
8-64

U.S. Nuclear Regulatory Commission, RE: July 29–August 1, 2003, Meeting Summary: 

Meeting with Duke Cogema Stone & Webster to Open Items Related to Mixed Oxide Fuel

Fabrication Facility. August 29, 2003.

(NRC, 2003g) U.S. Nuclear Regulatory Commission. Advisory Committee on Reactor

Safeguards Full Committee Meeting Transcripts. 

http://www.nrc.gov/reading-rm/doc-collections/acrs/tr/fullcommittee/2003/ November 6, 2003

(NRC, 2004a) Giitter, J., et al, U.S. Nuclear Regulatory Commission. Memorandum to H.

Peterson, U.S. Nuclear Regulatory Commission, RE: Determination on Position for Closure of

Chemical Safety Open Item CS-1 Pertaining to Red Oil Explosions. April 12, 2004.

(NRC, 2004b) Giitter, J., et al, U.S. Nuclear Regulatory Commission. Memorandum to H.

Peterson, U.S. Nuclear Regulatory Commission, RE: Safety Evaluation and Staff Positions on

the Closure of Remaining Chemical Safety Open Items CS-5b Pertaining to the MOX Fuel

Fabrication Facility Applicant’s Use of Temporary Emergency Exposure Limits. April 16, 2004.

(NRC, 2004c) Giitter, J., et al, U.S. Nuclear Regulatory Commission. Memorandum to S.

Magruder, U.S. Nuclear Regulatory Commission, RE: Determination on Position for Closure of

Chemical Safety Item CS-2 Pertaining to HAN Explosions. August 23, 2004.

(NSP, 1992) National Ski Patrol. “Components and Hazardous Atmospheres (Prehospital

Emergency Care & Crisis Intervention),” 4th Edition, p. 349. Brady Morton Series. 1992. 

http://www.patrol.org/resource/petri/airgas.html

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Agency. Proceedings at a Topical Meeting on Safety of the Nuclear Fuel, “Tomsk-7 Nuclear

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Impact on Purex Safety.” Oak Ridge, Tennessee, June 16, 1988.

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Water, and Soil. Noyes Publications: New Jersey, 1994.

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USACHPPM TG 230A, “Short-Term Chemical Exposure Guideline for Deployed Military

Personnel.” May 1999.
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and TEELs for Chemicals of Concern: Rev. 17.” Aiken, South Carolina, August 20, 2000.

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Conditions for Contacting Nitric Acid or Nitrates with Tri-n-Butyl Phosphate (TBP).” Aiken, South

Carolina, January 1994.

(WSRC, 1994b) Westinghouse Savannah River Company. WSRC-TR-94-0372, “Safe Handling

of TBP and Nitrates in the Nuclear Process Industry.” Aiken, South Carolina, July 1994.

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of “Red Oil” Runaway Reactions.” Aiken, South Carolina, December 1994.

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Savannah River Company. WSRC-TR-98-00171, “Analysis of Steam Heating of a Two-Layer

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00443, Revision 0, “Hydrozoic Acid Controls and Risks when Processing Plutonium Solutions in

HB-Line Phase II.” Aiken, South Carolina, January 31, 2001.

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Reaction.” Aiken, South Carolina, September 14, 2001.
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9-1

9. RADIATION SAFETY

9.1 Conduct of Review

This chapter of the final safety evaluation report (FSER) contains the U.S. Nuclear Regulatory

Commission (NRC) staff’s review of radiation safety features described by Duke Cogema Stone

& Webster (DCS or the applicant) in Chapter 9 of the revised construction authorization request

(CAR) (DCS, 2002e). The objective of this review is to determine whether the applicant’s

principle structures, systems, and components (PSSCs) and their design bases provide

reasonable assurance of worker protection against natural phenomena and the consequences of

potential accidents.

The radiological safety of workers is the subject of this FSER chapter. Further, for the purposes

of this FSER, the staff has adopted the applicant’s usage for names of individuals for whom

accident risks must be limited. That is, facility workers are workers in the restricted area inside a

room of the mixed oxide (MOX) fuel fabrication facility (MFFF) near a potential accident release

point. Site workers are workers considered to be outside the MFFF located 100 meters from the

ventilation exhaust stack. An individual outside the controlled area (IOC) is an individual at or

beyond the proposed MFFF controlled area boundary. A fourth receptor protected under the

provisions of Title 10, Section 70.61, “Performance Requirements,” of the Code of Federal

Regulations (10 CFR 70.61) is the environment, which is the area outside the proposed

restricted area. The staff’s evaluation of radiation safety protection of the IOC and the

environment is provided in Chapter 10 of this FSER.

The staff evaluated the information provided by the applicant for radiation safety by reviewing

Chapter 9 of the revised CAR, other sections of the revised CAR, supplementary information

provided by the applicant, and relevant documents available at the applicant’s offices but not

submitted by the applicant. The review of the radiation safety design bases and strategies was

closely coordinated with the review of accident sequences described in the safety assessment of

the design bases (see Chapter 5 of this FSER).

The staff reviewed how the radiation safety information in the revised CAR addresses or relates

to the following regulations:

• 10 CFR 70.23(b) states that, as a prerequisite to construction approval, the design bases

of the PSSCs and the Quality Assurance Program must be found to provide reasonable

assurance of protection against natural phenomena and the consequences of potential

accidents.

• Pursuant to 10 CFR 70.61(b)(1), the risk of credible high-consequence events to workers

must be limited. For workers, a high-consequence event is an internally or externally

initiated event that results in an acute 1 sievert (Sv) (100 rem) total effective dose

equivalent (TEDE). Controls must be used which either make the occurrence of such

events highly unlikely or make their consequences less severe than an acute 1 Sv (100

rem) TEDE.

• Pursuant to 10 CFR 70.61(c)(1), the risk of credible intermediate-consequence events to

workers must be limited. For workers, an intermediate-consequence event is an
9-2

internally or externally initiated event that is not a high-consequence event that results in

an acute 0.25 Sv (25 rem) TEDE. Controls must be used which either make the

occurrence of such events unlikely or make their consequences less severe than an

acute 0.25 Sv (25 rem) TEDE. 

The review of this revised CAR focused on the design basis of radiation safety systems,

components, and other related information. For radiation safety systems, the staff reviewed

information provided by the applicant for the safety function, system description, and safety

analysis. The staff used Chapter 9.0 in NUREG-1718, “Standard Review Plan for the Review of

an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility,” issued 2000, as guidance in

performing the review (NRC, 2000).

9.1.1 Radiation Safety Design Features

The objective of the staff’s review under this section was to determine whether the applicant's

design for construction and operation of the proposed facility is adequate to protect the

radiological health and safety of workers (including site workers) from accidents and credible

natural phenomena hazards (NPHs). Specifically, the staff’s review of the applicant’s proposed

radiation safety design features of the facility focused on the PSSCs required to reduce the risk

from intermediate-consequence events (those producing an acute 0.25 Sv (25 rem) TEDE to a

worker) and high-consequence events (those producing an acute 1 Sv (100 rem) TEDE to a

worker).

9.1.1.1 ALARA Design Considerations

The applicant has committed to designing the MFFF in accordance with as low as reasonably

achievable (ALARA) considerations. Below, the staff discusses the applicant’s commitment to

ALARA design considerations that might affect the design of radiation safety PSSCs.

The staff evaluated organizational relationships and responsibilities with respect to the

performance of reviews of radiation safety design features, application of ALARA into designstage collective dose estimates, descriptions and elements of the design review process for

radiation safety design features, and the applicant’s plan to use experience from past designs

and from operating plants to develop improved radiation safety design features.

The applicant identified the organizational relationships and responsibilities for performing

radiological design and design reviews in Section 9.1.1.1 of the CAR. These relationships and

responsibilities are consistent with those described in the previously approved MOX Project

Quality Assurance Plan, Revision 3 (DCS, 2002c; NRC, 2003a), and are thus acceptable to the

staff.

The applicant’s approach to design-stage collective dose estimates includes performing dose

assessments in accordance with NRC Regulatory Guides (RGs) 8.19, “Occupational Radiation

Dose Assessment in Light Water Reactor Power Plants Design State Man-Rem Estimates,”

issued 1979, and 8.34, “Monitoring Criteria and Methods to Calculate Occupational Radiation

Doses,” issued 1992 (NRC, 1979; NRC, 1992), using the ABAQUES method and radiation

zoning criteria for minimizing the direct (external) component of collective radiation dose, and

using reviews of experience at the MELOX facility and considerations of facility design to
9-3

evaluate and minimize collective internal doses.

The staff evaluated the applicant’s approach to design-stage collective dose estimates. The

staff finds that this approach will reduce the time spent in radiation areas, improve accessibility

to components requiring periodic maintenance or inservice inspection, reduce the retention of

radioactivity throughout plant systems, reduce contamination, facilitate decommissioning and

minimize secondary waste production, and incorporate experience from operating plants and

past designs. This approach is consistent with the acceptance criteria in Section 9.1.4.1.3 of

NUREG-1718 (NRC, 2000) and is, therefore, acceptable.

With respect to external doses, the applicant did not identify sources of direct radiation that pose

an unacceptable accident or natural phenomena risk. Therefore, the ABAQUES method and

radiation zoning criteria are not design bases for PSSCs identified in the safety assessment. 

The applicant’s ALARA design philosophy to prevent or mitigate the hazard of radioactive

material intake will rely primarily on experience at the MELOX facility in Marcoule, France. 

General measures and design philosophies for reducing the risk posed by a loss of material

confinement that are described in the revised CAR include remote operations and automation,

use of gloveboxes with negative pressure relative to occupied spaces, preventive maintenance,

and airborne radioactivity monitoring. The NRC staff has reviewed and evaluated the sources of

radioactive material that the applicant proposes to use, namely weapons-grade plutonium and

depleted uranium. The staff agrees that direct radiation from these sources will not pose an

unacceptable accident or natural phenomena risk.

9.1.1.2 Facility Design Features

The information required to support the staff’s review of the applicant’s safety assessment

includes facility and process drawings and descriptions with radiation safety design features that

are PSSCs relied upon in the safety assessment. In the revised CAR, the applicant provided

facility and process drawings and descriptions, including ventilation system drawings, for the

locations and access control points for restricted areas, the controlled area, and passive and

active material confinement boundaries. This information facilitated a clear understanding of the

intended function of radiation safety PSSCs, commensurate with the expected level of design

detail available at the construction authorization stage. The level of information provided by the

applicant is, therefore, acceptable to the staff.

9.1.1.3 Source Identification

The applicant proposed that the plutonium oxide that would be shipped to an operating facility

would contain mostly plutonium isotopes (e.g., 238Pu, 239Pu, 240Pu) and small quantities of other

impurities. The impurities will include elements and radionuclides such as gallium, isotopes of

uranium and americium, and radioactive decay products of uranium, plutonium, and americium. 

The mass distribution of the radionuclides used by the applicant in its safety assessment is

provided in FSER Table 9.1-1 (DCS, 2002e, Table 9-3).

The staff confirmed by calculation that the radioactive decay progeny of the plutonium that the

applicant proposed to use would not make a significant contribution to the dose to workers

resulting from accidents at the proposed facility. The staff calculated that the radionuclides

listed in FSER Table 9.1-1 would account for greater than 99.99 percent of the dose resulting
9-4

from inhalation. Therefore, for the purposes of the safety assessment, the staff considers the

applicant’s proposed list of radionuclides and mass fractions to be complete.

The aqueous polishing (AP) process would remove the impurities from the plutonium oxide as

the first step in the fuel fabrication process. The purified form of plutonium oxide would be

transferred to the powder blending and fuel fabrication process. The impurities would be

temporarily stored in tanks in solution form before being transferred to the proposed waste

solidification building for further treatment.

The fuel fabrication process would use depleted uranium, which is less than 0.7 percent by mass

235U. A depleted uranium nitrate solution (DU(NO3

)2) would be used to dilute high-enriched

uranium that would be separated from plutonium oxide feedstock in the AP process. All tanks

containing depleted uranium nitrate would be in the AP process. 

Depleted uranium is also the major component of fuel that would be manufactured at the facility. 

Depleted uranium dioxide (DUO2) powder would be used directly in the powder blend and pellet

manufacturing process. The powder would comprise approximately 94 percent of the mass of

each fuel pellet.

Table 9.1-1 Mass Fractions of Isotopes in Aged Weapons-Grade Plutonium

Isotope 0-yr-old WGPu,

gm/gm WGPu

40-yr-old WGPu,

gm/gm WGPu

70-yr-old WGPu,

gm/gm WGPu

236Pu 1.00×10-9 5.98×10-14 4.06×10-17

238Pu 6.86×10-4 5.00×10-4 3.95×10-4

239Pu 9.21×10-1 9.20×10-1 9.19×10-1

240Pu 6.18×10-2 6.15×10-2 6.13×10-2

241Pu 1.00×10-2 1.00×10-2 (1) 1.00×10-2 (1)

242Pu 1.00×10-3 1.00×10-3 1.00×10-3

235U 0.00×10+00 1.83×10-2 (2) 1.90×10-2 (2)

238U 0.00×10+00 1.26×10-3 (2) 1.26×10-3 (2)

241Am 0.00×10+00 7.00×10-3 7.92×10-3 (1)

1. The applicant has made a conservative assumption that 241Pu, which has a 14.4-

year half-life, does not decay.

2. The m ass fraction of uranium isotopes represents ingrowth from the radioactive

decay of 239Pu and 242Pu and isotopes that would be present in MFFF feedstock for

other reasons.

Given a facility design production rate of 70 metric tons heavy metal of fuel per year, this

corresponds to an annual demand for 65.8 tons of depleted uranium dioxide feedstock.
9-5

[Text removed under 10 CFR 2.390]

In summary, the applicant identified six categories of radioactive sources at the facility that

would be present in quantities that could pose unacceptable risks to workers from accidents and

NPHs. These are the impure plutonium oxide (PuO2(U)), the purified plutonium oxide (PuO2(P)),

the AP high alpha activity waste stream containing mostly americium and other radioactive

decay products, depleted uranium oxide, the master blend, and the final blend. The purified

plutonium oxide source, master blend, and final blend categories represent three distinct

concentrations of plutonium oxide (in decreasing concentration of the plutonium oxide)—(1) a

100-percent plutonium oxide powder, (2) a 20-percent plutonium oxide-depleted uranium oxide

(MOX) mixture, and (3) a 6-percent MOX final blend. The isotopic mass distribution of

radionuclides in each source category is provided in FSER Table 9.1-2.

The applicant provided an estimate of the amount of licensed material from each source

category that could be present in each processing unit throughout an operating facility. For units

in which multiple sources may be present, only the source likely to result in the bounding

accident consequences was identified.

Table 9.1-2 Mass Fractions of Isotopes in Each Source Category for the Safety

Assessment

RadionuclidePuO2(U)

 (1)

PuO2(P) 

(2)

High Alpha

Activity

Waste

(Liquid)

Depleted

Uranium

(3)

20% MOX 6% MOX

236Pu 4.07×10-17 4.10×10-17

----- ----- 8.20×10-18 2.46×10-18

238Pu 3.95×10-4 3.98×10-4

----- ----- 7.96×10-5 2.39×10-5

239Pu 9.20×10-1 9.27×10-1

----- ----- 1.85×10-1 5.56×10-2

240Pu 6.14×10-2 6.18×10-2

----- ----- 1.24×10-2 3.71×10-3

241Pu 1.00×10-2 1.01×10-2

----- ----- 2.02×10-3 6.06×10-4

242Pu 1.00×10-3 1.00×10-3

----- ----- 2.00×10-4 6.00×10-5

235U (4) ----- 2.500×10-3 2.00×10-3 2.35×10-3

238U (4) ----- 9.975×10-1 7.98×10-1 9.38×10-1

241Am 7.00×10-3

----- 0.78 g/L ----- ----- -----

1. PuO2(U) is unpolished plutonium oxide feed.

2. PuO2(P) is polished plutonium oxide.

3. Depleted uranium is both a feed material and a waste stream. These values are for feed.

4. Uranium isotopes do not contribute significantly to dose from inhalation events as compared to

isotopes of plutonium and am ericium.

The applicant also provided the photon and neutron energy spectra and intensities per unit mass
9-6

of licensed material associated with both purified and impure plutonium oxide (FSER Tables 9.1-

3 and 9.1-4). The photon energy spectra are based on the decay of the isotopes present in

freshly separated weapons-grade plutonium that would occur over a 70-year period. This

accounts for the ingrowth of beta- and gamma-emitting radionuclides that would contribute to

direct gamma radiation exposures. The applicant also included in these spectra the contribution

from (",n) interactions. The staff independently verified the intensity and energy distribution of

these spectra using simple spreadsheet models and the mass distributions of isotopes provided

by the applicant. The staff finds that these spectra provide an acceptable basis for the proposed

shielding design. 

Table 9.1-3 Photon Energy Spectrum for MFFF MOX

Photon Energy

(MeV)

Intensity

(gamma/sec/gram)

20% MOX 5% MOX

0.015 4.79×107 1.20×107

0.025 7.03×101 1.76×101

0.038 4.94×104 1.23×104

0.058 1.27×105 3.17×104

0.085 1.36×104 3.40×103

0.125 4.21×104 1.06×104

0.225 5.93×103 1.57×103

0.375 2.72×104 6.80×103

0.575 1.02×103 2.55×102

0.85 1.39×102 3.47×101

1.25 1.25×101 3.13×100

1.75 5.67×100 1.42×100

2.25 3.24×100 8.14×10-1

2.75 1.86×100 4.67×10-1

3.5 1.65×100 4.13×10-1

5 6.94×10-1 1.74×10-1

7 7.84×10-2 1.97×10-2

9.5 8.91×10-3 2.24×10-3

Total 4.82×107 1.20×107
9-7

Table 9.1-4 Neutron Energy Distribution for MFFF MOX

Neutron

Source

Intensity

(neutrons/sec/gram)

20% MOX 5% MOX

Spontaneous

Fission

11.8 2.96

(", n) 12.0 3.0

Total 23.8 5.96

Energy spectra and source intensities per unit mass of depleted uranium were not provided by

the applicant. However, given the low specific activity of this material, the staff does not

anticipate that depleted uranium compounds would pose a direct radiation hazard to workers

that would challenge the performance requirements in 10 CFR Part 70, “Domestic Licensing of

Special Nuclear Material.”

Another consideration in the hazard evaluation of radioactive sources is the solubility of inhaled

radioactive compounds. In general, inhalation of soluble compounds of plutonium results in a

committed dose approximately twice as high as the inhalation of insoluble compounds, namely

plutonium oxide. However, soluble plutonium compounds, such as plutonium nitrate, tend to

have higher molecular weights, which result in lower radioactivity per gram of molecular weight

of the compound. To simplify the safety assessment, the applicant demonstrated a conservative

assumption that the licensed nuclear material in all process units of the proposed facility would

be insoluble plutonium oxide. 

The applicant proposed the following relation:

[(specific activity)x(DCFY)]applicant $ [(specific activity)x(DCFW)]nitrate, oxalate, et al.

In this relation, the specific activity refers to the quantity of radioactivity per gram of molecular

weight of the compound. DCFY and DCFW are the dose conversion factors (DCFs) for

transportability class Y and class W, respectively, from Federal Guidance Report No. 11,

“Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for

Inhalation, Submersion and Ingestion,” issued 1988 (EPA, 1988). This relation demonstrates

that, for worker radiation dose calculations in the safety assessment, the applicant’s assumption

that all plutonium compounds are plutonium dioxide is conservative. This is because a mass

unit of plutonium in the insoluble applicant form (class Y) results in a higher dose than a mass

unit of plutonium in higher molecular weight soluble compounds (class W). The staff evaluated

this assumption and find that it holds for the types of licensed nuclear materials proposed for use

in the facility. Therefore, for the purposes of the safety assessment, the applicant’s use of DCFs

for the least soluble compounds is acceptable to the staff.

The depleted uranium compounds also pose an inhalation hazard to workers. However, the

limiting hazard in the case of depleted uranium dioxide is chemical toxicity, not radiological dose. 
9-8

For this reason, the staff’s evaluation of the applicant’s safety assessment for this and other

hazardous chemicals is provided in Chapter 8 of this FSER.

9.1.1.4 Safety Assessment of the Design Bases for PSSCs

The methodology presented by the applicant for its safety assessment of the design bases is

consistent with the methodology presented in the NUREG/CR-6410, “Nuclear Fuel Cycle Facility

Accident Analysis Handbook,” issued 1998 (NRC, 1998). The following sections describe the

staff’s evaluation of the applicant’s methodology for radiological consequence assessment in

support of the safety assessment of the design bases.

With the exception of the postulated criticality accident, which involves a significant direct

radiation hazard to facility workers, the radiation hazard of most concern in the safety

assessment is the release of airborne radioactive material to occupied spaces. Therefore, the

staff focused its review on the applicant’s estimates of potential accidental releases of

radioactive material from engineered confinement systems.

9.1.1.4.1 Facility Worker Consequence Assessment

To reduce the risk to the facility worker, the applicant proposed a methodology in the safety

assessment, described in Section 5.4.4.2 of the revised CAR, in which event consequences to

facility workers are qualitatively estimated. Since the applicant assumed that the facility worker

would be at the location of the release, the risks from events involving plutonium and americium

are deemed to be qualitatively unacceptable and PSSCs are deterministically applied. Further,

the applicant assumed that unmitigated consequences from the release of uranium are low and

PSSCs are not applied.

However, the applicant’s safety assessment includes several events in which the proposed

PSSC for protection of the facility worker is facility worker action to mitigate the consequences. 

For example, in the event of a loss of radioactive material confinement involving a fuel assembly

drop, waste container drop, or a small loss of confinement at a glovebox, the worker is expected

to recognize the event and leave the affected area. For these events, the staff evaluated the

applicant’s quantitative dose consequence estimates in order to evaluate the adequacy of this

safety strategy. The staff’s evaluation of this approach is discussed further in FSER Section

9.1.2.3.

9.1.1.4.2 Accident Release Estimates for Site Worker Radiation Safety Features

The applicant used a five-factor approach to estimate accidental releases of radioactivity to the

atmosphere that would result in a site worker receiving a dose. The total quantity released, or

the release rate over time, is referred to as the source term.

The source terms derived in the applicant’s assessment are the products of five parameters. 

The first parameter is the quantity of radioactive material that is postulated to be involved in the

event. The remaining four parameters are reduction factors that are applied to account for

different physical phenomena that would limit the amount of licensed nuclear material to which a

facility or site worker would be exposed. The following is the general form of the source term

formula:
9-9

Source Term, kilograms = MAR × DR × ARF × RF × LPF Eq. 9-1

where MAR is the quantity of material at risk during the event, in kilograms,

DR is the unitless damage ratio of material actually impacted by the event

(0<DR<1),

ARF is the atmospheric release fraction, which is the unitless fraction of impacted

material (MAR × DR) that can become airborne (0< ARF<1),

RF is the respirable fraction, which is the unitless fraction of airborne material that

can be inhaled into the human respiratory system (0<RF<1), and

LPF is the unitless fraction of airborne material that breaches the confinement

barrier.

The material at risk was provided by the applicant for more than 200 individual process units

throughout the proposed facility. These include the six major types of radioactive sources as

described in FSER Section 9.1.1.3. The applicant’s hazard analysis considered many different

types of events that could cause an adverse human health or environmental effect as a result of

accidental exposure to these radioactive sources. These types of accidents are categorized into

the major events of fires, explosions, loss of confinement, load drops, and nuclear criticality. 

The staff evaluation of the applicant’s hazard assessment, including event types that were

screened from further consideration, is provided in Chapter 5 of this FSER.

The applicant set the damage ratio (DR) equal to 1, with the exception that the DRs assumed for

pellets exposed to overpressurization gas flows and pressurized rods that have been breached

are 0.01 and 0.001, respectively. If rods are dropped, the DR is assumed by the applicant to be

equal to 0.02. The applicant also set the DR equal to 0.1 for a fire in the secured warehouse

building. The staff has verified that the applicant’s assumptions are consistent with staff’s

guidance provided in NUREG/CR-6410 (NRC, 1998). Therefore, these assumptions are

acceptable to the staff.

The applicant chose values for the atmospheric release fraction (ARF) and respirable fraction

(RF) based on values in NUREG/CR-6410 (NRC, 1998). The values chosen by the applicant for

these parameters are provided in FSER Table 9.1-5.

The staff reviewed the applicant’s bases for choosing ARFs and RFs and found them to be

consistent with those in NUREG/CR-6410. The applicant selected a value of 3.0×10-5 for the

ARF for dropped fuel rods. The basis for this value is a study performed by Sandia National

Laboratory (SNL, 1991). These assumptions are acceptable to the staff because they are based

either on data previously used and accepted by the staff for safety assessments or on data

derived from experiments related to the actual hazard.
9-10

Table 9.1-5 ARFs and RFs

Release

Form

Release

Fraction

Explosive

Detonation

Explosive OverpressurizationFire/Boil Drop Entrainment

Solution ARF 1.0 5.0×10-5 2×10-3 2.0×10-5 4.0×10-7

RF 0.01 0.8 1.0 1.0 1.0

ARF × RF 0.01 4.0×10-5 2×10-3 2.0×10-5 4.0×10-7

Powder ARF 1.0 5.0×10-3 6.0×10-3 2.0×10-3 4.0×10-5

RF 0.2 0.3 0.1 0.3 1.0

ARF × RF 0.2 1.5×10-3 6.0×10-4 6.0×10-4 4.0×10-5

Pellet ARF 0.01 5.0×10-3 5.0×10-4 1.0 NA

RF 1.0 0.3 0.5 1.1×10-5

ARF × RF 0.01 1.5×10-3 2.5×10-4 1.1×10-5

Rod ARF 0.01 3.0×10-5 0.0 3.0×10-5 NA

RF 1.0 1.0 1.0 1.0

ARF × RF 0.01 3.0×10-5 0.0 3.0×10-5

Filter

(unencased)

ARF 2.0×10-6 0.01 1.0×10-4 1.0×10-2 NA

RF 1.0 1.0 1.0 1.0

ARF × RF 2.0×10-6 0.01 1.0×10-4 1.0×10-2

As used by the applicant, the leak path factor generally accounts for the particulate matter

removal efficiency of the high-efficiency particulate air (HEPA) filters used in the ventilation

confinement systems. When relied upon to mitigate the effects of an accident, the applicant

assumed a 99-percent removal efficiency (i.e., 1-percent leak path factor) per stage. Each

HEPA system relied upon for safety includes two banks or stages of HEPA filters in series. 

Assuming that the effective leak path factor for a system of staged HEPA filters is the product of

the individual leak path factors for successive filter stages, the applicant applied a leak path

factor of 10-4 for systems relied upon in its safety assessment. The combination of efficiencies in

this manner is acceptable to the staff because it is consistent with the guidance in Section

F.2.1.3 of the staff’s accident analysis handbook (NRC, 1998).

When prevention alone, rather than mitigation, was the applicant’s preferred safety strategy, the

applicant applied a leak path factor equal to 0. The applicant used a leak path factor equal to 1

when the HEPA filters were either unlikely to function as needed or not required to mitigate the

event consequences (see FSER Section 9.1.1.4.4).

NRC RGs 3.71, “Nuclear Criticality Safety Standards for Fuels and material Facilities,” issued
9-11

1972, and 3.35, “Assumptions Used for Evaluating the Potential Radiological Consequences of

Accidental Nuclear Criticality in a Plutonium Processing and Fuel Fabrication Plant,” issued 1972

(NRC,1972c; NRC,1972b), were used by the applicant to develop source terms for direct

radiation and airborne releases resulting from a criticality accident. However, since the NRC has

withdrawn these guides, the staff used the current guidance in NUREG/CR-6410 (NRC, 1998) to

estimate the downwind consequences to a site worker of a criticality accident. By so doing, the

staff independently evaluated the applicant’s source terms. The staff finds that the applicant’s

analysis is consistent with the current guidance and is, therefore, acceptable.

9.1.1.4.3 Dose Assessment for the Site Worker

The applicant’s methodology for dose assessment relies on an assumption that the principle

human health hazard posed by releases of radioactive material from the proposed MOX facility

is inhalation of radioactive material downwind of the facility. Other pathways of exposure would

include direct radiation from the passing plume and exposure to ground surfaces contaminated

by material depositing on the ground as the plume passes. However, the staff confirmed by

calculation that, with the exception of the postulated criticality event, the direct radiation and

ground contamination pathways are negligible as compared to the inhalation pathway.

To calculate the 50-year committed effective dose equivalent (CEDE) from inhalation doses from

passing plumes, the applicant applied a simple formula involving the source term (Eq. 9.1), the

atmospheric dispersion factor (P/Q), a human receptor’s breathing rate (B.R.), and the DCF

(from EPA, 1988):

CEDEi [rem] = Source Termi [kg] × P/Q [s m-3] × B.R. [m3 s-1] × DCFi [rem :Ci-1] × Ci [:Ci kg-1]

where CEDEi is the committed dose from the ith radionuclide, and

Ci is the specific activity of the ith radionuclide.

Atmospheric dispersion factors were calculated by the applicant using site-specific

meteorological data from the U.S. Department of Energy’s Savannah River Site (SRS) H-Area

meteorological tower collected from 1987 through 1996. The ARCON96 model (NRC, 1999)

was used to estimate factors for the site worker located 100 meters from the plant stack. The

value calculated by the applicant is 6.1×10-4 s m-3.

The use of ARCON96 to estimate near-field atmospheric dispersion in the vicinity of buildings

and structures is described in NUREG/CR-6410 (NRC, 1999) and RGs 1.78, “Evaluating the

Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical

Release,” issued December 2001, and 1.194, “Atmospheric Relative Concentrations for Control

Room Radiological Habitability Assessments at Nuclear Power Plants,” issued June 2003 (NRC,

2001b; NRC, 2003b).

The staff verified by independent calculations that the meteorological data used by the applicant

in its safety assessment is consistent with data published by the SRS for the H-Area

meteorological tower (DOE, 1999). The staff also performed independent calculations for the

site worker atmospheric dispersion factor and calculated a value of 6.1×10-4 s m-3.

The breathing rate of 3.47×10-4 m3 s-1 assumed by the applicant is consistent with guidance

provided by the NRC in RG 1.25 (NRC 1972a) and is equivalent to a volume of 10 cubic meters
9-12

inhaled during an 8-hour workday. This assumption is based on NRC guidance applicable to

fuel handling and is, therefore, acceptable to the staff for use in the applicant’s safety

assessment.

The U.S. Environmental Protection Agency DCFs used by the applicant (EPA, 1988) are based

on the recommendations of the International Commission on Radiation Protection. These are

the same recommendations that form the basis for NRC radiation protection standards in 10

CFR Part 20. Therefore, these factors are acceptable to the staff.

9.1.1.4.4 Verification of Low-Consequence Events

Unmitigated event consequences result from an accident sequence when mitigative controls

either fail or do not exist. Unmitigated event consequences are those consequences calculated

by the applicant before determining and taking credit for PSSCs that would reduce the risk of the

event. However, in some cases, the unmitigated event consequence is so low that it falls below

the intermediate-consequence threshold values for workers specified in 10 CFR 70.61(c)(1). 

These events, referred to as low-consequence events, require no PSSCs to lower the risk. The

applicant identified 22 hazard assessment events as low-consequence events.

The staff performed independent calculations to verify the applicant’s assertion that some events

would be low-consequence events and would not require PSSCs to further reduce the accident

risk. Based on the staff’s confirmatory analysis, the staff accepts the applicant’s categorization.

9.1.2 Radiation Protection Program

The purpose of this review is to determine whether the applicant’s Radiation Protection Program

is adequate to protect the radiological health and safety of the workers and to comply with the

regulatory requirements of 10 CFR Part 19, “Notices, Instructions and Reports to Workers: 

Inspection and Investigations”; 10 CFR Part 20, “Standards for Protection Against Radiation”;

and 10 CFR Part 70, to the extent that such programmatic elements are relied upon to provide

reasonable assurance of protection against natural phenomena and the consequences of

potential accidents.

The staff reviewed the revised CAR using the guidance in NUREG-1718, Section 9.2

(NRC,2000).

9.1.2.1 ALARA

The applicant’s commitment to an ALARA program at the construction authorization stage

includes a management commitment to this policy, an ALARA committee, administrative control

levels and dose limits for design, and internal audits and assessments.

The staff has reviewed the applicant's brief description of the ALARA program and its related

safety assessment of the design bases, in which the applicant did not identify any PSSCs or

management measures within the purview of its ALARA program. The staff finds that, at the

construction authorization stage, no such PSSCs or management measures are required. 

However, the regulations in 10 CFR Part 20 and 10 CFR Part 70 contain specific requirements

for an ALARA program that would have to be fully and adequately addressed as part of any later

application for a license to possess and use licensed material.
9-13

9.1.2.2 Radiation Safety Procedures and Radiation Work Permits

The applicant identified seven major components of its Radiation Safety Procedures and Work

Control Program in the CAR. These are work planning, entry and exit control, radiological work

controls, posting and labeling, release of materials and equipment, sealed radioactive source

accountability and control, and radioactive material receipt.

Of these controls, the entry and exit controls, specifically process cell entry controls, are relied

on in the safety assessment of the design bases as a PSSC to protect the facility worker, as

shown in FSER Table 9.1-6. The applicant proposes several controlling parameters in the

design basis for the process cell entry controls. To prevent access during normal operations,

the applicant proposed radiation work permits, signs and postings, and barricades. The reliance

on a combination of administrative and engineered controls for this safety function provides

adequate assurance that it will be highly unlikely for facility workers to inadvertently enter

process cells upon failure of an administrative control. Therefore, this safety strategy is

acceptable to the staff.

The applicant proposed to further control radiation dose during any maintenance activities at the

proposed facility by committing to meet the requirements of 10 CFR 20.1601 and

10 CFR 20.1602 for high and very high radiation areas, respectively. The applicant’s

commitment to meet these regulatory requirements is acceptable to the staff.

9.1.2.3 Training

In the CAR, the applicant identified facility worker action and facility worker controls as PSSCs

for several internal hazards at the proposed facility. As shown in Table 9.1-6, design basis

controlling parameters for facility worker actions would include evacuation of the affected area. 

Facility worker controls are those actions taken by the facility worker before commencing an

activity that could result in an event with unacceptable dose consequences. In any subsequent

license application, in accordance with 10 CFR 70.62(d), the applicant would be required to

specify the management measures (such as training and procedures) which would adequately

ensure that facility worker actions and controls are reliable to serve their intended safety function

(i.e., the facility worker understands what to do and when to do it).

The applicant’s calculations demonstrate that dose consequences would be adequately

mitigated by the facility worker’s response, which is to evacuate the affected area within

30 seconds. The applicant’s calculation demonstrates an adequate margin of safety in that the

dose to the facility worker would be at least 10 times lower than the 0.25 Sv (25 rem)

intermediate consequence threshold for workers specified in 10 CFR 70.61(c)(1). Based on the

staff’s review of the applicant’s calculations, the staff finds that self-protective actions, including

escape from the affected area, would be sufficient to mitigate the consequences for those

events for which this action is credited.

9.1.2.4 Air Sampling

The applicant has committed to air sampling for facility worker exposure monitoring and control. 

The applicant planned to use measurements of airborne radioactivity concentrations to estimate
9-14

internal dose in accordance with 10 CFR 20.1204. However, the Airborne Radioactivity

Monitoring Program was not identified as a PSSC by the applicant in its safety assessment.

The staff has reviewed the applicant’s brief description of the Airborne Radioactivity Monitoring

Program and its related safety assessment of the design bases, in which the applicant did not

identify any PSSCs or management measures within the purview of this program. The staff

finds that, at the construction authorization stage, no such PSSCs or management measures

are required. However, the regulations in 10 CFR Part 20 and 10 CFR Part 70 contain specific

requirements for such a program that would have to be fully and adequately addressed in any

subsequent license application.

9.1.2.5 Contamination Control

The staff has reviewed the applicant’s brief description of the Contamination Control Program

and its related safety assessment of the design bases, in which the applicant did not identify

PSSCs or management measures within the purview of this program. The staff finds that, at the

construction authorization stage, no such PSSCs or management measures are required. 

However, the regulations in 10 CFR Part 20 and 10 CFR Part 70 contain specific requirements

for such a program that would have to be fully and adequately addressed in any subsequent

license application.

9.1.2.6 External Exposure, Internal Exposure, and Summing Internal and External Exposures

The applicant provided a brief description of its Direct Exposure Control Program, Internal

Exposure Control Program, and procedures for summing dose contributions from both direct

exposure and the intake of radioactivity in the revised CAR. These programs are not identified

as either PSSCs or management measures in its safety assessment.

Though not applicable to safety assessments for potential accidents, the applicant provided

estimates of total radiation doses to facility workers during normal operations based on

experience during operation of a similar facility, the MELOX plant in Marcoule, France.

The applicant’s estimates of external exposure hazards at the proposed MFFF involve an

adjustment of the external dose rates at the MELOX plant to account for the different isotopic

distribution of plutonium isotopes in the reactor-grade plutonium used at MELOX, as compared

to the weapons-grade plutonium proposed for use at the MFFF. This adjustment includes both a

20:1 MELOX:MFFF adjustment of the photon intensity and an 11:1 adjustment in the neutron

intensity per unit mass of plutonium. With these adjustments, the external collective dose is

estimated to be 12 person-rem per year, including a 10 person-rem contribution from the MOX

process (MP) units and 2 person-rem from the AP area.

For internal dose estimates, the applicant surveyed recent records at the MELOX MFFF in

Marcoule, France. From 1996 through July 2001, 41 individuals received an internal dose. 

Thirty workers received less than 10 percent of the allowable limit on intake (ALI) (based on a

0.02 Sv (2 rem) annual limit (ICRP, 1991)), 10 workers received between 10 and 33.3 percent 
9-15

Table 9.1-6 Radiation Protection PSSCs

PSSC Safety Function Controlling

 Parameters 

Process cell

entry controls

(administrative)

Prevent the entry of personnel into

process cells during normal operations.

Radiation work

permits

Signs and postings

Barricades

Ensure that facility workers do not receive

a dose in excess of limits while performing

maintenance in the AP process cells.

Radiation work

permits

Signs and postings

Commitment to

10 CFR 20.1602 for

very high radiation

areas

Facility worker

action

(administrative)

Ensure that facility workers take proper

actions to limit chemical and radiological

exposure.

Facility worker

response to exit the

affected area

Facility worker

controls

(administrative)

Ensure that facility workers take proper

actions before bagout operations to limit

radiological exposure.

Ensure that facility workers take proper

actions during maintenance activities to

limit radiological exposure.

Facility worker

prejob preparation

to prevent and/or

limit dose during

tasks involving

transient primary

confinements or

maintenance in

AP/MP C3 areas

of the ALI, and 1 worker received between 33.3 and 100 percent of the ALI. Using these data,

the staff calculated a collective dose of approximately 13 person-rem. Given the 5-year period

over which these intakes occurred, the average annual internal dose is approximately 3 personrem for workers at the MELOX facility. The applicant estimated a value of 4.5 person-rem per

year. The MELOX facility is larger than the proposed MFFF and has a higher throughput of

material. In addition, MELOX processes reactor-grade plutonium, which delivers about 10 times

more dose from inhalation than the weapons-grade plutonium proposed for use in the MFFF. 

Table 9.1-7 presents a comparison of the inhalation doses per unit mass of plutonium dioxide for

reactor-grade and weapons-grade plutonium. Therefore, the staff concurs that this collective

dose is a reasonable upper bound of doses expected at the MFFF.

Therefore, the applicant’s estimate that the TEDE among facility workers at the proposed MFFF

would be less than 20 person-rem is reasonable to the staff. The applicant further assumed that

the total workforce will number approximately 400 facility workers. Therefore, the average TEDE

per facility worker is less than 0.5 mSv (50 mrem) per facility worker per year, well below the

occupational dose limits of 50 mSv (5000 mrem) set forth in 10 CFR 20.1201(a). 
9-16

Table 9.1-7 Comparison of the Radiotoxicity of Reactor-Grade Plutonium and WeaponsGrade Plutonium; TEDE Per Gram of Plutonium Inhaled

Pu

Isotope

DCF

(rem/:Ci)

Specific

Activity of

Pu isotope

(:Ci/g)

ReactorGrade

Isotopic

Mass

Fraction

ReactorGrade

Dose

Fraction

(rem/g Pu)

W eaponsGrade

Isotopic

Mass

Fraction

W eaponsGrade

Dose

Fraction

(rem/g Pu)

238Pu 4×10-3 1.71×107

0.0149 1.0×1030.000395 2.7×101

239Pu 4×10-3 6.13×104

0.595 1.5×1020.92 2.3×102

240Pu 4×10-3 2.28×105

0.240 2.2×1020.0614 5.6×101

241Pu 2×10-1 1.03×108

0.103 2.1×1060.01 2.1×105

242Pu 4×10-3 3.93×103

0.04 6.3×10-1 0.001 1.6×10-2

TOTAL 2.1×106

2.1×105

9.1.2.7 Respiratory Protection

As discussed above in Section 9.1.2.4, the applicant proposed that facility worker actions be

relied upon to mitigate the consequences of some facility hazards (DCS, 2002b, page 16). The

applicant did not explicitly identify design basis controlled parameters and values associated with

facility worker respiratory protection equipment for this purpose. Rather, the applicant stated

that facility workers are assumed to be in the impacted area for no more than 30 seconds and

that they do not use their personnel protective equipment. The staff has reviewed the

applicant’s calculations and agrees that facility worker action, an administrative PSSC that

requires facility workers to escape the affected area, provides an adequate level of protection for

the facility worker for those events where this PSSC is applied.

9.1.2.8 Instrumentation

The applicant has committed to implement its Instrumentation Calibration and Maintenance

Program to ensure that radiation protection instrumentation is calibrated and maintained in

accordance with American National Standards Institute N323-1978 (R1993), “Radiation

Protection Instrumentation Test and Calibration.”

The staff has reviewed the applicant's brief description of the Instrumentation Calibration and

Maintenance Program and its related safety assessment of the design bases, in which the

applicant did not identify PSSCs or management measures within the purview of this program.

The staff finds that, at the construction authorization stage, no such PSSCs or management

measures are required. However, the regulations in 10 CFR Part 20 and 10 CFR Part 70

contain specific requirements for such a program that would have to be fully and adequately

addressed in any subsequent license application.
9-17

9.1.3 Radiation Safety Design Bases

The PSSCs for facility worker radiological protection, which are elements of the applicant’s

proposed Radiation Protection Program and were identified in the applicant’s safety assessment

of the design bases, are listed in FSER Table 9.1-6. This table lists the radiation safety PSSCs,

functions, and controlled parameters for the design described in the revised CAR. The staff has

reviewed whether these functions and parameters provide reasonable assurance that the

applicant’s design for construction and operation of the facility would be adequate to protect the

radiological health and safety of workers and to comply with the regulatory requirements of 10

CFR Part 20 and 10 CFR Part 70 during routine and nonroutine operations, including anticipated

events. The staff’s evaluation findings are provided in FSER Section 9.2.

9.2 Evaluation Findings

In Chapter 5 of the revised CAR, DCS provided design basis information for radiation protection

elements that it identified as PSSCs for the proposed facility. Based on the staff’s review of the

revised CAR and supporting information provided by the applicant relevant to radiation

protection, the staff finds that the design bases of the PSSCs identified by the applicant will

provide reasonable assurance of protection against natural phenomena and the consequences

of potential accidents, in accordance with 10 CFR 70.23(b).

As discussed above in FSER Section 9.1, pursuant to 10 CFR 70.61(b)(1) and

10 CFR 70.61(c)(1), the risk of credible high-consequence and intermediate-consequence

events must be limited. For workers, a high-consequence event is an internally or externally

initiated event that results in an acute 1 Sv (100 rem) TEDE, and an intermediate-consequence

event is one which results in an acute 0.25 Sv (25 rem) TEDE. For such high-consequence

events, controls must be used which either make the occurrence of these events highly unlikely

or make their consequences less severe than an acute 1 Sv (100 rem) TEDE. For such

intermediate-consequence events, controls must be used which either make the occurrence of

these events unlikely or make their consequences less severe than an acute 0.25 Sv (25 rem)

TEDE. Based on staff’s review of the safety assessment methodology presented in the revised

CAR, the staff finds that the design bases of the PSSCs identified by the applicant will reduce

facility worker dose consequences to acceptable levels for both high-consequence and

intermediate-consequence events. Accordingly, the staff concludes that, for worker doses, the

applicant has met the 10 CFR 70.61(b)(1) and 10 CFR 70.61(c)(1) performance requirements. 

9.3 References

(DCS, 2002a) Hastings, P.S., Duke Cogema Stone & Webster. Letter to U.S. Nuclear

Regulatory Commission, RE: DCS-NRC-000081, “Clarification of Responses to NRC Request

for Additional Information.” January 7, 2002. 

(DCS, 2002b) Hastings, P.S., Duke Cogema Stone & Webster. Letter to U.S. Nuclear

Regulatory Commission, RE: DCS-NRC-000083, “Clarification of Responses to NRC Request

for Additional Information.” February 11, 2002.

(DCS, 2002c) Hastings, P.S., Duke Cogema Stone & Webster. Letter to U.S. Nuclear

Regulatory Commission, RE: MOX Project Quality Assurance Program Plan. March 26, 2002.
9-18

(DCS, 2002d) Hastings, P.S., Duke Cogema Stone & Webster. Letter to U.S. Nuclear

Regulatory Commission, RE: Clarification of Responses to NRC Request for Additional

Information. April 23, 2002.

(DCS, 2002e) Ihde, R., Duke Cogema Stone & Webster. U.S. Nuclear Regulatory Commission,

RE: MOX Fuel Fabrication Facility Construction Authorization Request. October 31, 2002

(including page changes through February 9, 2005).

(DOE, 2000) U.S. Department of Energy. DOE-STD-3009-2000, “DOE Standard: Stabilization,

Packaging, and Storage of Plutonium-Bearing Materials.” Washington, DC, 2000.

(EPA, 1988) U.S. Environmental Protection Agency. EPA-520/1-88-020, Federal Guidance

Report No. 11, “Limiting Values of Radionuclide Intake and Air Concentration and Dose

Conversion Factors for Inhalation, Submersion and Ingestion.” Washington, DC, 1988. 

(ICRP, 1983) International Commission on Radiation Protection. ICRP Publication 38,

“Radionuclide Transformations: Energy and Intensity of Emissions.” 1983.

(ICRP, 1991) International Commission on Radiation Protection. ICRP Publication 60, “1990

Recommendations of the International Commission on Radiological Protection.” 1991.

(NRC, 1972a) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.25, “Assumptions

Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the

Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors (Safety Guide

25).” Washington, DC, 1972.

(NRC, 1972b) U.S. Nuclear Regulatory Commission. Regulatory Guide 3.35, “Assumptions

Used for Evaluating the Potential Radiological Consequences of Accidental Nuclear Criticality in

a Plutonium Processing and Fuel Fabrication Plant.” Washington, DC, 1972.

(NRC, 1972c) U.S. Nuclear Regulatory Commission. Regulatory Guide 3.71, “Nuclear Criticality

Safety Standards for Fuels and Material Facilities.” Washington, DC, 1972.

(NRC, 1979) U.S. Nuclear Regulatory Commission. Regulatory Guide 8.19, “Occupational

Radiation Dose Assessment in Light Water Reactor Power Plants Design Stage Man-Rem

Estimates.” Washington, DC, 1979.

(NRC, 1992) U.S. Nuclear Regulatory Commission. Regulatory Guide 8.34, “Monitoring Criteria

and Methods to Calculate Occupational Radiation Doses.” Washington, DC, 1992.

(NRC, 1998) U.S. Nuclear Regulatory Commission. NUREG/CR-6410, “Nuclear Fuel Cycle

Facility Accident Analysis Handbook.” Washington, DC, March 1998.

(NRC, 1999) U.S. Nuclear Regulatory Commission. “ARCON96 Code System to Calculate

Atmospheric Relative Concentrations in Building Wakes.” CCC-664 Radiation Safety

Information Computational Center. Washington, DC, 1999.
9-19

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility.” Washington,

DC, 2000.

(NRC, 2001a) U.S. Nuclear Regulatory Commission. NUREG-1747, “Overview and Summary

of NRC Involvement with DOE in the Tank Waste Remediation System-Privatization (TWRS-P)

Program.” Washington, DC, 2001.

(NRC, 2001b) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.78, “Evaluating the

Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical

Release.” Washington, DC, December 2001.

(NRC, 2003a) Persinko, A., U.S. Nuclear Regulatory Commission. Letter to P. Hastings, Duke

COGEMA Stone & Webster, RE: Safety Evaluation Report: Quality Assurance Program for

Construction of MOX Fuel Fabrication Facility. January 10, 2003.

(NRC, 2003b) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.194, “Atmospheric

Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear

Power Plants.” Washington, DC, June 2003.

(SNL, 1991) Sandia National Laboratories. SAND87-7082, “Effects of a Potential Drop of a

Shipping Cask, a Waste Container, and a Bare Fuel Assembly During Waste-Handling

Operations; Yucca Mountain Site Characterization Project.” 1991.
9-20

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10-1

10. ENVIRONMENTAL PROTECTION

10.1 Conduct of Review

This chapter of the final safety evaluation report (FSER) contains the staff’s review of

environmental protection measures described by Duke Cogema Stone & Webster (DCS or the

applicant) in Chapter 10 of the revised construction authorization request (CAR) (DCS, 2002e). 

The objective of this review is to determine whether principle structures, systems, and

components (PSSCs) and their design bases identified by the applicant provide reasonable

assurance of protection of an individual outside the controlled area (IOC) and the environment

against natural phenomena and the consequences of potential accidents. The staff evaluated

the information provided by the applicant for environmental protection by reviewing Chapter 10

of the revised CAR, other sections of the revised CAR, supplementary information provided by

the applicant, and relevant documents available at the applicant’s offices but not submitted by

the applicant. In some cases, the staff also performed independent calculations. The review of

environmental protection design bases and strategies was closely coordinated with the review of

the accident sequences described in the safety assessment of the design bases (see Chapter 5

of this FSER).

As previously discussed in FSER Chapter 1, the set of U.S. Nuclear Regulatory Commission

(NRC) radiation safety requirements applicable to an individual depends on whether that

individual is a worker exposed to radiation as part of his assigned employment duties or is a

member of the public. The radiological safety of workers is discussed in FSER Chapter 9, and

the radiation protection standards applicable to the public, or IOC, are covered in this FSER

chapter.

The staff reviewed how the environmental protection information in the revised CAR addresses

or relates to the following regulations:

• Title 10, Section 20.1101(b), of the Code of Federal Regulations (10 CFR 20.1101(b))

states that a licensee shall use, to the extent practical, procedures and engineering

controls based upon sound radiation protection principles to achieve doses to members

of the public that are as low as reasonably achievable (ALARA).

• 10 CFR 20.1301(b) states, “If the licensee permits members of the public to have access

to controlled areas, the limits for members of the public continue to apply to those

individuals.”

• 10 CFR 20.1302(c) states the following: 

Upon approval from the Commission, the licensee may adjust the effluent

concentration values in appendix B to part 20, Table 2, for members of the

public, to take into account the actual physical and chemical

characteristics of the effluents (e.g., aerosol size distribution, solubility,

density, radioactive decay equilibrium, chemical form).

• 10 CFR 70.23(b) states that, as a prerequisite to construction approval, the design bases

of the PSSCs and the Quality Assurance Program must be found to provide reasonable
10-2

assurance of protection against natural phenomena and the consequences of potential

accidents.

• Pursuant to 10 CFR 70.61(b)(2), the risk of credible high-consequence events to any IOC

must be limited. For such individuals, a high-consequence event is an internally or

externally initiated event that results in an acute 0.25 sievert (Sv) (25 rem) total effective

dose equivalent (TEDE). Controls must be used which either make the occurrence of

such events highly unlikely or make their consequences less severe than an acute 0.25

Sv (25 rem) TEDE.

• Pursuant to 10 CFR 70.61(c)(2), the risk of credible intermediate-consequence events to 

any IOC must be limited. For such individuals, an intermediate-consequence event is an

internally or externally initiated event that is not a high-consequence event and that

results in an acute 0.05 Sv (5 rem) TEDE. Controls must be used which either make the

occurrence of such intermediate-consequence events unlikely or make their

consequences less severe than an acute 0.05 Sv (5 rem) TEDE.

• For environmental protection purposes, an intermediate-consequence event is one which

produces a 24-hour averaged release of radioactive material outside the restricted area

in concentrations exceeding 5000 times the values in Table 2 of Appendix B to

10 CFR Part 20, “Standards for Protection Against Radiation.” Pursuant to

10 CFR 70.61(c)(3), controls must be used which either make the occurrence of such

intermediate-consequence events unlikely or make their environmental consequences

less severe than those described above. 

In the revised CAR, the applicant described its commitment to environmental protection in the

areas of (1) radiation safety (ALARA) goals for effluent control and waste minimization, (2)

design of effluent and environmental monitoring for normal and off-normal operations, and (3)

design bases for PSSCs derived from the safety assessment, as necessary to ensure

environmental protection. The staff focused its review on the applicant’s safety assessment of

the design bases for environmental protection PSSCs, which is discussed below in FSER

Section 10.1.3. The staff also evaluated preliminary information provided by the applicant on

ALARA goals and the Effluent and Environmental Monitoring Program. With regard to the

design bases for environmental and IOC protection, this FSER chapter addresses the

applicant’s consequence methodology and results used to identify PSSCs that are relied upon to

meet the IOC and environmental performance requirements of 10 CFR 70.61(b)(2) and

10 CFR 70.61(c)(2–3). The staff used Chapter 10.0 in NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility,” issued August

2000, as guidance in performing the review (NRC, 2000).

10.1.1 Radiation Safety

The staff evaluated the applicant’s radiation safety measures for environmental protection based

on the applicant’s methods to maintain public doses at ALARA levels in accordance with

10 CFR 20.1101 and the applicant’s waste minimization practices.

10.1.1.1 ALARA Design Goals for Effluent Control

The applicant defined ALARA design goals for effluent control in revised CAR Section 10.1.1. 
10-3

The ALARA design goal is for airborne radioactive effluents released from the proposed mixed

oxide (MOX) fuel fabrication facility (MFFF). This goal is 20 percent of the effluent concentration

limits in column 1 of Table 2 in Appendix B to 10 CFR Part 20, as determined at the restricted

area boundary. The human dose corresponding to this goal, assuming an individual were

present continuously over 1 year at the restricted area boundary, is 0.1 mSv (10 mrem) TEDE.

Therefore, this goal affords an initial level of protection for members of the public in the

controlled area of 10percent of the 1 mSv (100 mrem) TEDE limit described in

10 CFR 20.1301(b). This fraction is consistent with staff expectations that an initial goal of 10 to

20 percent of Appendix B values or less can be achieved by almost all materials facility

licensees, as stated in Regulatory Guide (RG) 8.37, “ALARA Levels for Effluents from Materials

Facilities,” issued 1973 (NRC, 1973b); therefore, it is acceptable to the staff.

The applicant has not defined liquid effluent ALARA goals because the proposed facility will not

discharge liquid effluent directly to the environment. This is acceptable because the applicant’s

proposal is to transfer low-level waste containing NRC-licensed material to the U.S. Department

of Energy (DOE) at the Savannah River Site (SRS). The DOE would perform additional

treatment before discharging this material. Therefore, these discharges would be regulated by

DOE and would be subject to DOE ALARA considerations.

10.1.1.2 Air Effluent Controls to Maintain Public Doses at ALARA Levels

The confinement ventilation systems are described in FSER Section 11.4. In summary, there

are five major sources of ventilation exhaust that would contribute to air emissions from the

facility stack. The following ventilation and air-conditioning systems are PSSCs or have

individual components that are PSSCs:

• process cell off-gas treatment system

— final filters

— pressure boundary downstream of the final filters

• process cell ventilation system

— final filters

— pressure boundary downstream of the final filters

— tornado dampers

• medium depressurization exhaust system (which maintains the C2 confinement zone)

— final filters

— pressure boundary downstream of the final filters

— tornado dampers

• high depressurization exhaust system (C3 confinement zone)

• very high depressurization exhaust system (C4 confinement zone)

The applicant’s proposed design bases for these systems relies on NRC RG 3.12, “General

Design Guide for Ventilation Systems of Plutonium Processing and Fuel Fabrication Plants,”

issued 1973 (NRC, 1973a), and the American Society of Heating, Refrigerating and AirConditioning Engineers report, “Design Guide for Department of Energy Nuclear Facilities”

(ASHRAE, 1993), for guidance. Additional design bases are described in FSER Section 11.4.
10-4

The staff concludes that a commitment to these codes and standards should ensure that

engineered effluent controls will be designed and constructed to meet the requirements for

capacity, compartmentalization, safe shutdown, corrosion resistance, and efficiency required

during normal and likely facility conditions to maintain public doses at ALARA levels. The staff

will review the applicant’s process controls and procedures that will augment these engineered

controls and form the basis for a complete ALARA program during review of any license

application that DCS may later submit.

10.1.1.3 Liquid Effluent Controls to Maintain Public Doses at ALARA Levels

The proposed facility would not have liquid effluents that discharge directly to the environment. 

There are, however, five categories of liquid waste that must be managed. These waste

streams would be transferred to SRS before final treatment and disposal. The five waste

streams are the high alpha activity waste, stripped uranium waste, solvent wastes, low-level

liquid waste (including chlorine wastes), and nonhazardous liquid waste. Estimated volumes of

each waste type are provided in FSER Table 10.1-1. Waste minimization practices identified by

the applicant are discussed in FSER Section 10.1.1.5. The PSSCs required to safety handle,

store, and transfer liquid wastes are discussed in FSER Chapter 5 and FSER Section 11.8.

10.1.1.4 ALARA Review and Reports to Management

The applicant has committed to a program of measuring trends in environmental monitoring and

surveillance data against the effluent ALARA goals on a quarterly basis. The goals will be

reevaluated and new goals will be established for the following year.

The staff has reviewed the applicant’s brief description of the operational ALARA program and

its related safety assessment of the design bases, in which the applicant did not identify PSSCs

or management measures within the purview of this program. The staff finds that, at the

construction authorization stage, no such PSSCs or management measures are required. 

However, the regulations in 10 CFR Part 20 and 10 CFR Part 70 contain specific requirements

for such a program that would have to be fully and adequately addressed in any subsequent

license application.

10.1.1.5 Waste Minimization

The applicant provided an overview of its commitment to waste minimization practices in Section

10.1.4 of the revised CAR. The applicant’s proposal for incorporating waste minimization

practices into the design process focuses on recycling and reuse of materials. During

operations, the applicant proposed to rely on waste management procedures to separate and

segregate solid and liquid wastes and remove packaging and shipping materials before entry

into contaminated areas.

The applicant would use active and passive confinement systems and vacuum systems inside

gloveboxes. These systems are designed to allow recycling of materials from the secondary

waste streams in the aqueous polishing (AP) process and MOX process (MP) scraps back to the

main processes. Specific AP process waste minimization steps include acid recovery and

solvent regeneration.
10-5

Table 10.1-1 Potential for Intermediate-Consequence Events Involving Liquid Waste

Streams

W aste

Category

Maximum

Volume

(liters/ yr)

(gallons/

yr)

Radionuclide

Maximum

Concentration

(:Ci/mL)

5000 times

Table 2,

Appendix B,

to 10 CFR

Part 20

(:Ci/mL)

Potential to

Exceed

Environmental

Performance

Requirement?

Disposition

(5)

High

Alpha

Activity

W aste

39,000

(10,300)

241Am 1,550

(1)

1×10-4 Yes

(6)

W SB

Stripped

Uranium

167,000

(44,000)

238U 0.01

(2)

1.5×10-3 Yes

(7)

W SB

Solvent

LLW

10,600

(2,800)

239Pu 1.2×10-4

(3)

1×10-4 Yes

(8)

SRS

Solvent

Recovery

Low-Level

Aqueous

W aste

1,080,000

(285,000)

alpha 3×10-6

(4)

1×10-4 No W SB

Nonhazar

dous

Liquid

W aste 

33,300,000

(8,800,000)

Radionuclide concentrations in nonhazardous liquid waste are

far below levels that could cause an interm ediate

consequence event.

Approved

NPDES

Facilities at

SRS

(1) Based on 3.1×1015 Bq (84,000 Ci) of 241Am per year (DCS, 2004, Table 3-3) and normal

volume

(2) Based on 5,000 kg (11,000 lb) of uranium per year (DCS, 2004, Table 3-3) and normal volume

(3) Based on 17.2 mg (3.8×10-5 lb) per year of plutonium (DCS, 2004, Table 3-3) and normal

volume

(4) Based on 1.12×108 Bq (3.03×10-3 Ci) alpha-emitting radioactivity per year (DCS, 2004, Table 3-

3) and normal volume

(5) W SB = waste solidification building

(6) W aste transfer line is a PSSC for this waste

(7) W aste transfer line is a PSSC for this waste

(8) PSSCs are applied to ensure safe transfer of this waste to DOE without spillage

The applicant’s safety assessment of the design bases considered hazards in these areas. 

However, specific waste minimization practices and program commitments are not relied upon in

the safety assessment to reduce the risk of these hazards. The PSSCs relied upon to reduce

the risks of hazards in the acid recovery and solvent regeneration systems are described in

Chapters 5, 8, and 11 of this FSER.

10.1.2 Effluent and Environmental Monitoring

The staff has reviewed the applicant’s brief description of the Effluent and Environmental

Monitoring Programs and its related safety assessment of the design bases, in which the
10-6

applicant did not identify PSSCs or management measures within the purview of these

programs. The staff finds that, at the construction authorization stage, no such PSSCs or

management measures are required. However, the regulations in 10 CFR Part 20 contain

specific requirements for such programs that would have to be fully and adequately addressed in

any subsequent license application.

The following sections describe the staff’s evaluation of the Effluent and Environmental

Monitoring Programs, to the extent that such information was provided by the applicant in the

environmental report and revised CAR (DCS, 2002b; DCS, 2002e).

10.1.2.1 Concentrations of Radionuclides in Air Effluents and Public Doses

The applicant provided an estimate of maximum controlled area radionuclide concentrations,

which is based on annual releases described in its environmental report (DCS, 2004, Table D-7),

a 50-percent atmospheric dispersion parameter value (P/Q) of 2.5×10-4 s/m3, a distance to a

receptor from the plant stack of 52 meters (171 feet), and the assumption that releases occur

from ground level. This calculation demonstrates that the average controlled area concentration

immediately outside the restricted area would be less than 34 percent of the ALARA goal. The

staff performed an independent calculation using the methodology described in the National

Council on Radiation Protection and Measurements Report 123, as described in NUREG-1718. 

The staff assumed a 28-meter (92-foot) stack height, no plume rise, and a site-specific 3.6-m/s

(8 mph) annual average windspeed and that the wind blows downwind toward a receptor 100

percent of the time. The staff’s estimate of the P/Q is 5×10-5 s m-3 at a distance of 400 meters

(1312 feet). The staff’s calculation demonstrates that the concentration outside the controlled

area would be less than 7 percent of the applicant’s ALARA goal. The staff’s analysis is

summarized in FSER Table 10.1-2.

The applicant’s estimate of the maximum potential dose to an individual member of the public

from SRS is 3.3×10-5 mSv (3.3×10-3 mrem) per year. The staff performed an independent

analyses using GENII, the Hanford environmental radiation dosimetry software system. The

staff’s result is 5.2×10-6 mSv (5.2×10-4 mrem) per year, which is in good agreement with the

applicant’s value, given the uncertainties inherent in the assumptions related to environmental

dosimetry analyses. The staff’s value is 0.00052 percent of the NRC’s 10 CFR Part 20 1 mSv

(100 mrem) standard for members of the public.

Based on the staff’s independent calculation, the known or expected concentrations of

radioactive material in airborne effluents from the proposed facility would be below the limits in

Table 2 of Appendix B to 10 CFR Part 20. This forms an acceptable basis for future ALARA

evaluations as required by 10 CFR 20.1101(b).

10.1.2.2 Physical and Chemical Characteristics of Radionuclides in Discharges

With regard to the provisions of 10 CFR 20.1302(c), the applicant did not propose to adjust the

effluent concentration values that appear in Table 2 of Appendix B to 10 CFR Part 20 for

members of the public by taking into account the actual physical and chemical characteristics of

the effluents (e.g., aerosol size distribution, solubility, density, radioactive decay equilibrium,

chemical form). Instead, the applicant intended to demonstrate compliance with the annual dose

limit of 10 CFR 20.1301 using the dose methodology provided for in 10 CFR 20.1302(b)(1) and

not the concentration-based methodology provided for in 10 CFR 20.1302(b)(2).
10-7

This approach is consistent with the requirements of 10 CFR Part 20 and, therefore, is

acceptable to the staff.

10.1.2.3 Air Effluent Discharge Location and Effluent Monitoring

Revised CAR Section 10.2.1.4 indicates that the discharge location for radioactive air effluents

from the proposed facility would be the facility stack located on the roof of the MP process

building. This stack would be 12.2 m (40 ft) tall and would be located on the top of the MOX fuel

fabrication building, which would be 22.3 m (73 ft) tall. The stack would discharge 5,419 m3/min

(191,360 cubic feet per minute) during normal operations. The applicant has committed to

maintaining an operable continuous sample collection system in accordance with RG 4.16,

Revision 1, “Monitoring and Reporting Radioactivity in Releases of Radioactive Materials in

Liquid and Gaseous Effluents from Nuclear Fuel Processing and Fabrication Plants and Uranium

Hexafluoride Production Plants,” issued 1985 (NRC, 1985). 

The applicant will be expected to provide additional information (including, for example, sample

collection and analysis procedures, a description of action levels, pathway analyses for public

doses, and recording and reporting procedures) in any subsequent license application.

10.1.2.4 Environmental Monitoring Program

The applicant performed preconstruction environmental monitoring measurements, as described

in revised CAR Section 10.3. These measurements were conducted in accordance with DOE

Order 5400.1 and established a baseline of existing radiological, chemical, physical, and

biological conditions in the area of the site. These measurements were used to evaluate

whether contaminants could pose a potential safety concern for construction personnel. The

data may also be used by the applicant and/or DOE to apply for environmental permits.

To accomplish these goals, the applicant made use of the data provided from the extensive SRS

Environmental Monitoring Program and augmented the SRS environmental studies with

additional sample collections.

With regard to expected impacts, the nonradiological impacts to the environment from the

construction of the facility, if authorized, are expected to be minimal. The applicant expected

that nonradiological monitoring prescribed through various environmental permits will be

sufficient through construction and operation of the facility.

The staff reviewed the Plutonium Disposition Program Preconstruction Environmental Monitoring

Report (Fledderman, 2002) and finds that the report is consistent with the guidance in the

Standard Review Plan (NRC, 2000); therefore, it is acceptable to the staff.

10.1.3 Safety Assessment of the Design Bases

The staff’s evaluation of the applicant’s environmental protection measures at the construction

authorization stage focuses on the potential accident sequences that result in radiological

releases to the environment and the PSSCs relied upon for safety that are specified by the

applicant to reduce the risk of these accidents.
10-8

Table 10.1-2 Air Effluent Concentrations from the MFFF

Radionuclide Annual

Releases 1

NRC

Average

RABC 2

DCS

Average

RABC3

20% of

Appendix B

to 10 CFR

Part 20

NRC

Ratio

DCS

Ratio

236Pu 1.3×10-8 2.1×10-26 1.0×10-25 1×10-14 1.7×10-12 1.0×10-11

238Pu 8.5×100 1.3×10-17 6.8×10-17 4×10-15 3.4×10-3 1.7×10-2

239Pu 9.1×101 1.4×10-16 7.6×10-16 4×10-15 3.6×10-2 1.8×10-1 

240Pu 2.3×101 3.6×10-17 1.8×10-16 4×10-15 9.1×10-3 4.6×10-2 

241Pu 1.01×102 1.6×10-16 8.1×10-16 2×10-13 8.0×10-4 5.0×10-3

242Pu 6.1×10-3 9.7×10-21 4.9×10-20 4×10-15 2.4×10-6 1.2×10-5

241Am 4.8×101 7.6×10-17 3.8×10-16 4×10-15 1.9×10-2 9.6×10-2

234U 5.1×10-3 8.1×10-21 4.1×10-20 1×10-14 8.1×10-7 4.1×10-6

235U 2.1×10-4 3.3×10-22 1.7×10-21 1×10-14 2.8×10-8 1.4×10-7

238U 1.2×10-2 1.9×10-20 9.6×10-20 1×10-14 1.6×10-6 8.0×10-6

Totals 0.07 0.34

(1) DCS, 2002b

(2) RABC = restricted area boundary concentration. The atmospheric dispersion factor (P/Q)

estimated by the staff for this table is based on a 28-m (92-ft) stack height, no plume rise,

and a site-specific 3.6 m/s (8 mph) annual average windspeed (ASHRAE, 1993). The

maximum value is 5x10-5 s/m3 at a distance of 400 m (1312 ft).

(3) The atmospheric dispersion factor (P/Q) estimated by the applicant is based on 50 percent

meteorology and a distance to the receptor of 52 m (171 ft) (DCS, 2004). The maximum

value is 2.5×10-4 s/m3.

10.1.3.1 Consequence Assessment Methodologies

In its safety assessment, the applicant calculated committed doses to an IOC and

concentrations of radioactive material in the environment outside the restricted area from each

accident to demonstrate that risks from event consequences were reduced to acceptable levels. 

The consequence assessment methodology used by the applicant for dose consequences at the

controlled area boundary is the same methodology used for the site worker, as described in

Chapter 9 of this FSER, with the exception of the value of the atmospheric dispersion factor.

The atmospheric dispersion factor that the applicant derived for the distance from the facility to

the controlled area boundary is 2.5×10-4 s/m3.
10-9

The consequence assessment methodology used to comply with the 10 CFR 70.61(c)(3)

performance requirement is also similar to the methodology presented for site workers in

Chapter 9 of this FSER. The principal difference is that the applicant did not use the respirable

fraction (RF) as a reduction factor in calculations demonstrating that concentrations fall below

the 10 CFR 70.61(c)(3) intermediate-consequence threshold. This is because the 

10 CFR 70.61(c)(3) concentration pertains to total concentrations in the environment and not to

doses to human receptors resulting from intake of respirable particles. This approach is

acceptable to the staff.

The proposed restricted area boundary for the facility would be approximately 52 meters from

the facility discharge stack (DCS, 2004). The atmospheric dispersion factor that the applicant

derived for this location is 2.79×10-4 s/m3.

As a result, the following equation is used to calculate environmental consequences:

[EC]x = {[Source Term / RF] × [P/Q]RA × [f]

x

} / (3600 s hr-1 × 24 hr)

The source term is the same as that described in Chapter 9 of this FSER, the RF is divided

back into the source term to negate the reduction applied for human consequence source terms,

f

x

 is the specific activity and the fraction of the total quantity of the material at risk that is the

radionuclide X, and [P/Q]RA is the value of the atmospheric dispersion factor for the facility stack,

as described above.

The use of this equation is consistent with the staff’s guidance in NUREG/CR-6410, “Nuclear

Fuel Cycle Facility Accident Analysis Handbook,” issued 1998 (NRC, 1998), and the regulations

in 10 CFR 70.61(c)(3); therefore, it is acceptable to the staff.

10.1.3.2 Radiation Doses to Individuals Outside the Controlled Area from Accidents

In Chapter 5 of the revised CAR, the applicant presented the mitigated bounding event

consequences for the five major categories of events (i.e., fire, explosion, loss of confinement,

load handling events, and criticality). The methodology that the applicant used to estimate

radiation doses to the IOC from potential accidents is substantially similar to that described in

FSER Section 9.1.1.4, with the exception that the postulated distance to the IOC is 160 meters,

as compared to the 100-meter distance to the site worker. As stated in FSER Chapter 9, the

staff finds that the applicant used acceptable methods for estimating the consequences from

accident sequences that result in radiological releases that could result in doses to the IOC. The

applicant’s estimates are bounding and its assumptions are appropriately conservative, given the

available level of design at the construction authorization stage.

The mitigated consequences fall below the performance requirements of 10 CFR 70.61,

“Performance Requirements.” Therefore, the staff finds that the applicant’s estimates of IOC

accident consequences are acceptable.
10-10

10.1.3.3 Environmental Consequences

The staff reviewed the applicant’s estimates of environmental consequences of bounding events

provided in the revised CAR. The staff finds that the consequences are below the intermediateconsequence level described in 10 CFR 70.61(c)(3); therefore, they are acceptable.

10.1.3.4 PSSCs for Protection of the IOC and the Environment

The PSSCs required for protection of the IOC and the environment for each of the controlling

events are discussed in Chapter 5 of this FSER and in other sections as referenced in FSER

Table 5-1. As discussed above, the applicant used acceptable methods for estimating

consequences from accident sequences that result in radiological releases to the environment.

10.2 Evaluation Findings

In Section 10 of the revised CAR, DCS provided information on public and environmental

protection ALARA goals and methodologies to demonstrate compliance with NRC standards for

radiation safety applicable to members of the public in accordance with 10 CFR 20.1101(b),

10 CFR 20.1301(b), and 10 CFR 20.1302(c). Based on the staff’s review of this information, it

finds that the applicant has established acceptable ALARA goals and methodologies to ensure

public and environmental radiation safety.

In Section 10.5 of the revised CAR, DCS provided design basis information for the IOC and

environmental protection PSSCs that it identified for the proposed facility. Based on the staff’s

review of the revised CAR and supporting information provided by the applicant relevant to

environmental protection, the staff finds that the design bases of the PSSCs identified by the

applicant will provide reasonable assurance of protection against natural phenomena and the

consequences of potential accidents, in accordance with 10 CFR 70.23(b).

Further, pursuant to 10 CFR 70.61(b)(2), 10 CFR 70.61(c)(2), and 10 CFR 70.61(c)(3), and

based on staff’s review of the safety assessment methodology presented in the revised CAR,

the staff finds that the design bases of the PSSCs identified by the applicant will reduce IOC

dose and environmental contamination consequences to acceptable levels.

10.3 References

(ASHRAE, 1993) American Society of Heating, Refrigeration and Air-Conditioning Engineers.

“Design Guide for Department of Energy Nuclear Facilities.” 1993.

(DCS, 2002a) Hastings, P.S., Duke Cogema Stone & Webster. Letter to U.S. Nuclear

Regulatory Commission, RE: Clarification of Responses to NRC Request for Additional

Information. 2002.

(DCS, 2002b) Hastings, P.S., Duke Cogema Stone & Webster. Letter to U.S. Nuclear

Regulatory Commission, MFFF Environmental Report. 2002.
10-11

(DCS, 2002c) Hastings, P.S., Duke Cogema Stone & Webster. Letter to U.S. Nuclear

Regulatory Commission, DCS-NRC-000083, Clarification of Responses to NRC Request for

Additional Information. 2002.

(DCS, 2002d) Hastings, P.S., Duke Cogema Stone & Webster. Letter to U.S. Nuclear

Regulatory Commission, Clarification of Responses to NRC Request for Additional Information. 

2002.

(DCS, 2002e) Idhe, R.H., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Construction Authorization Request Revision. October 31, 2002 (including

page changes through February 9, 2004).

(Fledderman, 2002) Fledderman, P.D. Plutonium Disposition Program Preconstruction

Environmental Monitoring Report, Westinghouse Savannah River Company, ESH-EMS-2002-

1141, Revision 0. 2002.

(Napier, 1988) Napier, B.A. et al. “GENII—The Hanford Environmental Radiation Dosimetry

Software System.” PNL-6584. Prepared by Pacific Northwest Laboratory, Richland, WA, for

U.S. Department of Energy. 1988.

(NCRP, 1996) National Council on Radiation Protection and Measurements. NCRP No. 123,

“Screening Models for Releases of Radionuclides to Atmosphere, Surface Water and Ground,

Recommendations of the National Council on Radiation Protection and Measurements.” 1996.

(NRC, 1973a) U.S. Nuclear Regulatory Commission. Regulatory Guide 3.12, “General Design

Guide for Ventilation Systems of Plutonium Processing and Fuel Fabrication Plants.” 

Washington, DC, 1973.

(NRC, 1973b) U.S. Nuclear Regulatory Commission. Regulatory Guide 8.37, “ALARA Levels

for Effluents from Materials Facilities.” Washington, DC, 1973.

(NRC, 1985) U.S. Nuclear Regulatory Commission. Regulatory Guide 4.16, Revision 1,

“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.” Washington, DC, 1985.

(NRC, 1990) U.S. Nuclear Regulatory Commission. NUREG/CR-4691, “MELCOR Accident

Consequence Code System (MACCS).” Washington, DC, 1990.

(NRC, 1998) U.S. Nuclear Regulatory Commission. NUREG/CR-6410, “Nuclear Fuel Cycle

Facility Accident Analysis Handbook.” Washington, DC, 1998.

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility.” Washington,

DC, August 2000.
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11-1

11. PLANT SYSTEMS

11.1 Civil Structural Systems

11.1.1 Conduct of Review

This chapter of the final safety evaluation report (FSER) contains the U.S. Nuclear Regulatory

Commission (NRC) staff’s review of the civil structural systems described by the applicant, Duke

Cogema Stone & Webster (DCS) in Chapter 11 of the revised construction authorization request

(CAR) (DCS, 2002b). The objective of this review is to determine whether the principal

structures, systems, and components (PSSCs) and their design bases identified by the applicant

provide reasonable assurance of protection against natural phenomena and the consequences

of potential accidents. The staff evaluated the information provided by the applicant for the civil

structural systems by reviewing Chapter 11 of the revised CAR, other sections of the revised

CAR, supplementary information provided by the applicant, and relevant documents available at

the applicant’s offices but not submitted by the applicant. The review of the civil structural

systems design bases and strategies was closely coordinated with the review of the civil

structural aspects of accident sequences described in the safety assessment of the design

bases (see Chapter 5 of this FSER) and the review of other plant systems. 

The staff reviewed how the information in the revised CAR addresses the following regulations:

• Title 10, Section 70.23(b), of the Code of Federal Regulations (10 CFR 70.23(b)) states,

as a prerequisite to construction approval, that the design bases of the PSSCs and the

Quality Assurance Program be found to provide reasonable assurance of protection

against natural phenomena and the consequences of potential accidents.

• 10 CFR 70.64, “Requirements for New Facilities or New Processes at Existing Facilities,”

requires that baseline design criteria (BDC) and defense-in-depth practices be

incorporated into the design of new facilities. With respect to structural systems, 10 CFR

70.64(a)(2) requires that the mixed oxide (MOX) fuel fabrication facility (MFFF, or the

facility) design “provide for adequate protection against natural phenomena with

consideration of the most severe documented historical events for the site.”

The review for this construction approval focused on the design basis of the civil structural

systems and other related information. For each civil structural system, the staff reviewed

information provided by the applicant for the safety function, system description, and safety

analysis. The review also encompassed proposed design basis considerations such as

redundancy, independence, reliability, and quality. The staff used Chapter 11.0 in NUREG1718, “Standard Review Plan for the Review of an Application for a Mixed Oxide (MOX) Fuel

Fabrication Facility” (NRC, 2000), as guidance in performing the review.

The following three general areas of Section 11.1, “Civil Structural Systems,” of the revised CAR

are reviewed for this section of the FSER:

• classification of civil structural systems

• codes and standards

• structural design criteria
11-2

11.1.1.1 System Description

The civil structural systems for the proposed facility include the buildings; support structures; and

facilities that house, support, confine, or contain various plant systems, components, and

equipment associated with licensed nuclear materials, or hazardous chemicals associated with

licensed nuclear materials, as well as support buildings.

11.1.1.1.1 Function

As described in Sections 11.1.1 and 11.1.7 of the revised CAR, the safety functions for the civil

structural systems would accomplish the following:

• Support PSSCs during normal, severe, and extreme environmental conditions.

• Provide confinement functions as part of secondary and tertiary confinement systems.

• Protect PSSCs from the effects of normal, severe, and extreme environmental loads.

• Protect PSSCs from the effects of temperature extremes, including design basis internal

and external fires.

• Protect PSSCs from the effects of design basis man-induced events, including potential

load drops.

11.1.1.1.2 Major Components

The civil structural systems considered to be seismic Categories (SCs) I and II or conventional

seismic structures were identified in Section 11.1.3 of the revised CAR. The major components

for each of the seismic categories are as follows:

• SC-I structures include the MOX fuel fabrication building (which includes the MOX

processing area, aqueous polishing (AP) area, and the shipping and receiving area),

emergency fuel storage vault, and the emergency generator building.

• SC-II structures include the safe haven buildings. 

• Conventional seismic (CS) structures include the reagent process building, administration

building, secured warehouse building, technical support building, standby generator

building, receiving warehouse building, and miscellaneous site structures (e.g., the bulk

gas storage pad; the heating, ventilation, and air conditioning (HVAC) and process chiller

pads; and the electric transformer pads) specified in Section 11.1.3.11 of the revised

CAR. 

A detailed description for each major component is provided in Section 11.1.3 of the revised

CAR.
11-3

11.1.1.1.3 Control Concepts

This section is not applicable to civil structural systems.

11.1.1.1.4 System Interfaces

Civil structural systems interface with the site and all plants systems because they provide

protection and support for structures, systems, and components (SSCs).

11.1.1.2 Classification of Civil Structural Systems

The classification outlined in Section 11.1, “Civil Structural Systems,” of the revised CAR

consists of three levels—(1) SC-I, (2) SC-II, and (3) CS structures. The design loadings

considered for the civil structures in each category are as follows:

• SC-I—normal, severe, and extreme environmental loads, including the design basis

earthquake and tornado

• SC-II—normal, severe, and design basis earthquake 

• CS—normal, severe, and conventional seismic loads as specified by the Uniform

Building Code

11.1.1.3 Design Basis of the PSSCs

11.1.1.3.1 Codes and Standards

This section contains a review of Sections 11.1.6.1.3, “Codes and Standards for SC-II

Structures,” and 11.1.7.3, “Codes and Standards for SC–I Structures,” of the revised CAR.

The designs of the SC-1 civil structural systems and the associated steel and concrete

components conform to standard engineering practice. A comprehensive list of the applicable

codes and standards is provided in Section 11.1.7.3 of the revised CAR. The American

Concrete Institute (ACI) report, ACI 349-97, “Code Requirements for Nuclear Safety-Related

Concrete Structures,” issued 1997 (ACI, 1997), and American National Standards

Institute/American Institute of Steel Construction (ANSI/AISC) N690-1994, “Specification for the

Design, Fabrication, and Erection of Steel Safety-Related Structures for Nuclear Facilities,”

issued 1994 (AISC, 1994), including Supplement 1, would be supplemented by specific

provisions. In Section 11.1.6.1.3 of the revised CAR, the applicant referenced the same codes

and standards as those for the SC-I structures for the design of the SC-II civil structural

systems, except for ACI 349.1R–91, “Reinforced Concrete Design for Thermal Effects on

Nuclear Power Plant Structures, issued 1996 (ACI, 1996); ACI 349-97; ACI 349.2R–97,

“Embedment Design Examples,” issued 1998 (ACI, 1998); and ANSI/AISC N690-1994, including

Supplement 1. Additionally, ACI 318-99, “Building Code Requirements for Structural Concrete,”

issued 1999 (ACI, 1999), was included for SC-II civil structures. 

The staff reviewed the codes and standards for the designs of SC-I and SC-II civil structural

systems and concluded that the cited codes and standards are consensus standards that

provide reasonable guidance consistent with the categorization assigned to the buildings.
11-4

11.1.1.3.2 Structural Design Criteria

This section contains a review of the structural design criteria and load combinations for the civil

structural systems discussed in the revised CAR. 

[Text removed under 10 CFR 2.390]

 Structural design loads for SC-I and SC-II structures were

discussed in Sections 11.1.7.4, “Loads and Loading Combinations for SC-I Structures,” and

11.1.6.1.4, “Values for SC-II Structures,” of the revised CAR. Design criteria and loads

anticipated for the structures were divided into (1) normal loads, (2) severe environmental loads,

and (3) extreme environmental loads. The normal loads include dead, live, hydrostatic fluid

pressure, lateral soil pressure, thermal, and component reaction loads. The severe

environmental loads include wind and flood loads. The extreme environmental loads include

seismic, tornado, and explosive loads, and post-earthquake settlements. The only extreme

environmental load considered in the design of the SC-II structures was the seismic load. The

staff finds that the applicant’s evaluation satisfies the requirements of 10 CFR 70.64(a)(2), which

states that structures must be designed to provide for adequate protection against natural

phenomena with consideration of the most severe documented historical events for the site. 

11.1.1.3.2.1 Normal Loads.

• Dead Loads

Dead loads are gravity loads induced by the mass of the structure, permanent

equipments, and any permanent hydrostatic loads with constant fluid levels. This

definition for dead loads is consistent with the American Society of Civil Engineers

(ASCE) Standard 7-98, “Minimum Design Loads for Buildings and Other Structures,”

issued 1998 (ASCE, 1998), and is acceptable to the staff. The CAR also indicates that

effects of differential settlement would be considered in determining dead loads.

• Live Loads

Live loads are loads produced by the use and occupancy of the building. The live loads

considered for the civil structures of the proposed MFFF include floor, rain, snow and

ice, transportation vehicle and heavy floor, as well as crane, monorail, hoist, and elevator

loads.

— Floor Live Loads

The minimum uniformly distributed live loads for the civil structures were

established by the applicant in accordance with the ASCE Standard 7-98 (ASCE,

1998). Specifically, the floor live loads identified include the following:
11-5

platform and work area 6.0 kN/m2 (125 psf)

light storage 6.0 kN/m2 (125 psf)

heavy storage 12.0 kN/m2 (250 psf)

heavy operation 12.0 kN/m2 (250 psf)

office 4.79 kN/m2 (100 psf)

computer room 7.18 kN/m2 (150 psf)

dining/meeting rooms 4.79 kN/m2 (100 psf)

laboratory 9.58 kN/m2 (200 psf)

toilet areas 4.79 kN/m2 (100 psf)

mechanical (utility) rooms 7.18 kN/m2 (150 psf)

electrical rooms 7.18 kN/m2 (150 psf)

stairs, fire escapes, and corridors 4.79 kN/m2 (100 psf)

roof 2.4 kN/m2 (50 psf)

transportation vehicle loads 14.37 kN/m2 (300 psf) or forklift truck

of 26.69 kN (6 kips) capacity

The staff reviewed the design basis floor live loads discussed in Section

11.1.7.4.1.1 of the revised CAR and additional information provided in the

applicant’s March 8, 2002, letter (DCS, 2002a) and determined that the floor live

loads and the roof loads are appropriate and acceptable for the design of the

facility civil structures.

— Rain Loads

The design basis rain loads for the civil structures were determined by the

applicant in accordance with the requirements of ASCE Standard 7-98 (ASCE,

1998). The design rain load for the roof system of the SC-I structures is 2.4

kN/m2(50 psf), which is equivalent to more than 24.4 cm (9.6 in.) of standing

water on the roof because of deflection of the roof or blockage of the primary roof

drains. The CAR further states that “parapets or other structures, which could

potentially contribute to significant ponding, are not used on the roofs of SC-1

structures,” and that the rain load does not combine with the roof live load in the

load combinations. The staff reviewed the design basis rain load and determined

that it is appropriate and acceptable.

— Snow and Ice Loads

The design basis snow and ice loads value was determined by the applicant to be

0.48 kN/m2 (10 psf). This value was estimated based on the 100-year maximum

ground snow and ice loads. The importance factor for the snow and ice loads

value is 1.2. The staff reviewed the design basis maximum ground snow and ice

loads value and found that it was based on acceptable methods. 
11-6

— Transportation Vehicle Loads and Heavy Floor Loads

The design basis load for transportation vehicular truck traffic in designated

building areas was determined in accordance with the standard loadings defined

by the American Association of State Highway and Transportation Officials in

1996 (AASHO, 1996). The minimum truck loading of HS 20-44 is used for wheelloading design. The heavy floor loading considered in areas used for

transportation, transfer, and storage of finished fuel assemblies is 14.37 kN/m2

(300 psf), or forklift truck of 26.69 kN (6 kips) capacity. The staff reviewed the

design basis transportation vehicle loads and heavy floor loads and determined

that they are appropriate and acceptable.

— Crane, Elevator, and Hoist Loads

These design loads apply to structural members and components to support

permanently installed cranes, monorails, hoists, and elevators. Section

11.1.7.4.1.1 of the revised CAR states that the design basis crane, monorail,

hoist, and elevator loads would envelop the full-rated capacity of the cranes,

monorails, hoists, and elevators, including impact loads and test load

requirements. The effects of a crane load drop were evaluated in accordance

with the guidance provided in NUREG-0612, “Control of Heavy Loads at Nuclear

Plants,” issued 1980 (NRC, 1980), for SC-I structures. The staff reviewed the

design basis crane, monorail, hoist, and elevator loads and found that these

design basis loads are appropriate and acceptable.

— Hydrostatic Fluid Pressure Loads

The revised CAR indicates that no hydrostatic fluid pressure loads caused by fluid

held inside the buildings of the facility are currently identified. The staff agrees

with this statement. However, if the applicant later determines that there is fluid

inside the building of the facility, hydrostatic fluid pressure loads should be

included in the design.

— Lateral Soil Pressure Loads

Section 11.1.7.4.1.1 of the revised CAR indicates that a determination of the

lateral soil pressure loads on structures, elements of structures, or both because

of retaining soil would be based on the density of the soil and any surcharge load,

plus the hydrostatic pressure caused by ground water or soil saturation. 

The minimum lateral soil pressure loads on structures and/or elements of

structures resulting from retaining soil are defined in ASCE Standard 7-98 (ASCE,

1998). Earthquake-induced soil pressure on structures or embedded wall design

would be developed in accordance with ASCE Standard 4-98, “Seismic Analysis

of Safety-Related Nuclear Structures and Commentary,” issued 1999 (ASCE,

1999). No hydrostatic pressure is expected because the ground water table at

the site is below the MFFF.
11-7

The staff reviewed the approach presented by the applicant for determining

lateral soil pressure loads and found that the approach is based on acceptable

standards and engineering practices.

— Thermal Loads

Thermal loads consist of thermally induced forces and moments on the structural

components of buildings. These loads would result from operating and

environmental conditions. The thermally induced loads would be design

dependent. Consequently, determination and consideration of these thermally

induced loads would be an integral part of a design. For the design of civil

structural systems, the applicant would consider the effects of thermal expansion

loads caused by axial restraint of the structural components, as well as loads

resulting from thermal gradients. The applicant also indicated that it would

determine these thermally induced loads based on the most critical transient or

steady-state condition. The staff reviewed the information that the applicant

presented in the revised CAR and found that its consideration of thermal

expansion loads and thermal gradient loads for the design of civil structural

systems is acceptable.

— Equipment Reaction Loads

The equipment reaction loads included those from pipes, HVAC ducts, conduits,

and cable trays. These loads would be design dependent and need to be

assessed during design. The applicant stated that the equipment reaction loads

would be determined based on the most critical transient or steady-state

condition. The applicant further indicated that it would make sure that the final

designs envelop the actual equipment reaction loads. The staff reviewed the

information presented regarding consideration of equipment reaction loads in the

design and found that the bounding/enveloping approach is acceptable.

11.1.1.3.2.2 Severe Environmental Loads.

• Wind Loads

Figure 6-1b in ASCE Standard 7-98 (ASCE, 1998) identifies a design basis wind speed

of 160.93 km/h (100 mph) for the region. Information provided in Table 11.1-2 of the

revised CAR for the Savannah River Site (SRS) identifies a design basis wind speed of

209.22 km/h (130 mph), which is higher than the value provided in ASCE 7-98. The

revised CAR also indicates that the wind loads calculated based on the design basis wind

would be determined by the procedures provided in ASCE 7-98. The approach for

determining the wind loads is acceptable to the staff because the approach used is the

same or similar to consensus standards approved by the staff.

The applicant also included the consideration of windborne missiles in the design of the

civil structural systems. 

[Text removed under 10 CFR 2.390]
11-8

Considering the effects of windborne missiles in the

design is consistent with ASCE Standard 7-98 (ASCE, 1998), which requires the

inclusion of windborne debris in areas where the basic windspeed is equal to or greater

than 193.12 km/h (120 mph). The inclusion of windborne missiles in the design is

acceptable to the staff because it meets the guidance provided in ASCE 7-98.

• Flood Loads

The maximum probable flood level for the site is at elevation 68.43 m (224.5 ft) above

mean sea level. The corresponding site grade level is approximately 82.91 m (272 ft)

above mean sea level. Because the site grade level is much higher than the maximum

probable flood level at the site, the proposed facility is considered to be a flood-dry site

and will be free from the adverse effects of the maximum probable flood. Consequently,

the loads resulting from flood water do not need to be considered in the design of civil

structural systems. Analysis of the maximum probable flood level was based on the

surface hydrology of the region and potential dam failure resulting from seismic events

given in Section 1.3.4.2, “Floods,” of the revised CAR.

The staff reviewed the discussion regarding the flood loads and concludes that the facility

design is consistent with the design criteria of Regulatory Guide (RG) 3.40, “Design

Basis Floods for Fuel Reprocessing Plants and for Plutonium Processing and Fuel

Fabrication Plants,” issued 1977 (NRC, 1977), and that the approach used for conducting

the flood analysis is consistent with that outlined in Section 5.1.3 of ANSI/American

Nuclear Society (ANS) Standard 2.8-1992, “Determining Design Basis Flooding at Power

Reactor Sites,” issued 1992 (ANS, 1992), which determined design basis flooding at

power reactor sites. Therefore, the information provided is acceptable to the staff.

11.1.1.3.2.3 Extreme Environmental Loads.

• Seismic Loads

Table 11.1-2 and Section 1.3.6, “Seismology,” of the revised CAR list the design basis

earthquake ground motions based on probabilistic seismological studies specific to SRS. 

The applicant’s design basis horizontal and vertical response spectra for the facility were

developed based on RG 1.60, “Design Response Spectra for Seismic Design of Nuclear

Power Plants,” issued 1973 (NRC, 1973a), horizontal and vertical spectrum shapes

anchored at a peak ground acceleration of 0.20 g. A more detailed discussion on the

applicant’s design spectra is provided in Section 1.3.1.5.5 of this FSER. 

According to the revised CAR, the design seismic loads for the SC-I structures of the

facility would be determined by considering the design basis earthquake accelerations in

the three orthogonal directions (two horizontal and one vertical).

To support soil-structure interaction analysis, synthetic time histories for the three

components of the design basis ground motion would be generated. The response

spectra of the synthetic time histories would envelop the design response spectra and

meet the minimum power spectral density requirement in accordance with the criteria in

Section 3.7.1, “Seismic System Analysis,” of NUREG-0800, “Standard Review Plan for
11-9

the Review of Safety Analysis Reports for Nuclear Power Plants,” issued 1987 (NRC,

1987), according to the revised CAR.

A simplified three-dimensional finite-element model that ignored the embedment and

simulated intact slabs would be used for the soil-structure interaction analysis of the MOX

fuel fabrication building. A three-dimensional lumped mass stick model representing the

building characteristics and the assumption that the roofs and floors are rigid diaphragms

would be used for the soil-structure interaction analysis of the emergency generator

building. 

According to the revised CAR, the soil model for the soil-structure interaction analysis

would include a sufficient number of idealized soil layers from the ground surface to the

bedrock. This soil model was developed using the information from the soil exploration

and site-response analysis. The structural damping values used in the analysis would be

in accordance with RG 1.61, “Damping Values for Seismic Design of Nuclear Power

Plants,” (NRC, 1973b) for a safe-shutdown earthquake and are acceptable to the staff. 

From the soil-structure interaction analysis, the response spectra at the foundation and

each floor and roof level would be obtained in order to develop acceleration profiles for

building design. 

Three soil conditions (lower, best, and upper bound) accounting for material property

uncertainties would be considered in the soil-structure interaction analysis to develop the

in-structure response spectra at each direction and for a given structural level. A

broadening of in-structure spectrum peaks would be applied to the spectrum envelope to

account for structural material properties. Specific extents of spectrum-peak broadening

for the MOX fuel fabrication and emergency generator buildings were discussed in

Section 11.1.7.4.1.3, “Extreme Environmental Loads,” of the safety analysis report. The

staff determined that the approach for consideration of uncertainties associated with

structural and soil material properties is consistent with the design guidance in RG 1.122,

“Development of Floor Design Response Spectra for Seismic Design of Floor-Supported

Equipment or Components,” issued 1978 (NRC, 1978). 

For the MOX fuel fabrication and emergency generator buildings, a three-dimensional

finite-element model would be used to conduct a static analysis using the equivalent

accelerations of each level developed from the three-dimensional soil-structure analysis. 

The acceleration profiles that envelop the acceleration profiles from the three soil

conditions would be used for the finite-element analysis. Results from the equivalent

static three-dimensional finite-element analysis based on the acceleration profiles applied

in the three directions would be combined using the 100-40-40 percent rule as described

in Section 3.2.7.1.2 of ASCE Standard 4-98 (ASCE, 1999). 

For other structures, the approach of applying the applicable seismic response to the

base of the finite element models may be used. Guidance provided by RG 1.92,

“Combining Modal Responses and Spatial Components in Seismic Response Analysis,”

issued 1976 (NRC, 1976), and ASCE 4-98 would be used for combining modal

responses and collinear responses from the individual earthquake components.

The staff reviewed the approach presented by the applicant for determining the seismic

loads and found the approach to be acceptable.
11-10

• Tornado Loads for SC-I Structures

Table 11.1-2 of the revised CAR provides the design basis tornado windspeed,

atmospheric pressure change, and rate of pressure drop. In determining design tornado

loads, the procedure provided in ASCE Standard 7-98 (ASCE, 1998) would be used. 

The staff reviewed the approach presented by the applicant and found the approach to

be acceptable based on the guidance provided in ASCE 7-98. The three types of

tornado loads on the facility structures are as follows:

— Tornado Wind Pressure Loads

The tornado wind pressure loads are defined in Table 11.1-2 of the revised CAR. These

pressure loads are based on Section 6 of ASCE 7-98, which is an industry consensus

standard that has been accepted by the staff. Therefore, the information presented is

acceptable to the staff.

— Tornado-Created Differential Pressure Loads

The tornado-created differential pressure loads are defined in Table 11.1-2 of the revised

CAR. These pressure loads are based on guidance provided in Section 3.3.2 of

NUREG-0800 (NRC, 1987), and, therefore, it is acceptable to the staff.

— Tornado-Generated Missile Loads

The design basis for tornado-generated missiles is presented in Table 11.1-2 of the CAR. 

In Section 11.1.7.2.1.2 of the CAR, the applicant stated that the analysis for tornadogenerated missiles is complete. However, in its March 8, 2002, letter, the applicant

stated that the CAR incorrectly implies that the analysis for tornado-generated missiles

has been completed. The applicant revised the CAR to say that the analysis will be

provided as part of its license application. This deferral is acceptable to the staff based

on the guidance provided in (1) Section 11.4.6.2 of NUREG-1718 (NRC, 2000), which

states that the adequacy of civil structural system designs pertaining to tornado missiles

may be determined based on the integrated safety analysis (ISA), the summary of which

need not be submitted as part of the revised CAR (see 10 CFR 70.65(b)), and (2)

Sections 3.5.2 and 3.5.3 of NUREG-0800 (NRC, 1987).

[Text removed under 10 CFR 2.390]

• Site Proximity Missile (Except Aircraft) Load

The applicant indicated that the distance from the nearest transportation route to the

facility and the emergency diesel generator (EDG) buildings is greater than the safe

distance established based on the available inventories of hazardous materials. 
11-11

Therefore, the effects of explosion-generated missiles from the transportation route are

not a concern for design. Furthermore, the applicant stated that analyses to identify site

proximity missiles that might be generated from events within the safe distance will be

submitted as part of its license application (DCS, 2002a, Enclosure A, p. 49). This

deferral is acceptable to the staff based on the guidance provided in Section 11.4.6.2 of

NUREG-1718 (NRC, 2000) stating that the adequacy of civil structural system designs

may be determined based on the ISA, the summary of which need not be submitted as

part of the revised CAR (see 10 CFR 70.65(b)).

• Foundation Design Criteria

The foundation design criteria are discussed in MFFF site geotechnical reports (DCS,

2001; DCS, 2003) supporting the revised CAR. The MOX fuel fabrication building has

two foundation levels. The base level of the MOX processing area is at grade. The base

level of the AP area is approximately 5.3 m (17.5 ft) below grade. A portion of the base

level of the shipping and receiving area is at grade, and the remainder is 5.3 m (14 ft)

below grade. The base mat for the MOX fuel fabrication building will be 2 m (6.5 ft) thick,

and it will be 0.91 m (3 ft) thick for the emergency generator building. Some soils at the

foundation locations will be removed and replaced with engineered structural fill. 

Approximately 3.05 m (10 ft) of the engineered fill material will be placed beneath the

main foundation level of the MOX fuel fabrication building and approximately 1.54 m (5 ft)

beneath the emergency generator building. The engineered fill will be extended at least

3.05 m (10 ft) beyond the edges of the foundations and will be compacted to at least 95

percent of the maximum dry density determined in accordance with American Society for

Testing and Materials (ASTM) DI557. 

The ultimate bearing capacity is approximately 3.4 MPa (71.2 ksf) for the MOX fuel

fabrication building and approximately 0.65 MPa (13.5 ksf) for the EDG building. The

third MFFF site geotechnical report (DCS, 2003) indicates that the ultimate bearing

capacities were determined based on conservative soil strength properties. For the final

design, a minimum factor of safety of more than 6 will be maintained for both buildings. 

The staff reviewed the estimated bearing capacities and found that the values provided

are acceptable. 

• Settlements

The first MFFF site geotechnical report (DCS, 2001) correctly states that the critical

foundation load condition that limits the allowable foundation bearing capacity will be

settlement. The settlements at the site come from two potential sources—(1)

compaction of soft soil materials, including soft zones, and (2) localized liquefaction.

Two methods (DCS, 2003) were used to estimate settlements for the SC-I

structures—recompression indices and numerical modeling. The estimated settlement

based on the recompression indices for the MOX fuel fabrication building is

approximately 4.32 cm (1.7 in.) at the edge of the foundation and 7.62 cm (3.0 in.) at the

center of the foundation. The numerical modeling incorporating the presence of soft

zones estimates the primary settlements ranging from 6.86–8.89 cm (2.7–3.5 in.). 

Including the secondary consolidation, the total estimated settlements of the MOX fuel

fabrication building are approximately 8.13–10.16 cm (3.2–4.0 in.). These primary
11-12

settlement estimates were determined assuming that the in situ soils were

preconsolidated to pressures in excess of the sum of the initial overburden pressures

and the pressure induced by structure loads, but that the soft zone soils were normally

consolidated under the existing overburden pressures (DCS, 2003, Table 7-1).

The appropriateness of the use of this preconsolidation pressure was justified by the third

MFFF site geotechnical report (DCS, 2003) based on the measured settlements for

Building 221-S (vitrification building) at the SRS site. The soil conditions at the Building

221-S site are similar to those at the MFFF site except that no soft zones are present at

the Building 221-S site (DCS, 2003). The average foundation pressure for Building 221-

S (239.4–263.3 kPa (5–5.5 ksf)) is approximately 10–20 percent smaller than that of the

MOX fuel fabrication building (292.5 kPa (6.11 ksf)). The dimensions of Building 221-S

(110.3×35.6 m (362×117 ft)) are relatively smaller than those of the MOX fuel fabrication

building (123.1 m and 140.2×91.4 m (408 ft and 460×300 ft)). Measured primary

settlements at Building 221-S range from 3.1–6.6 cm (1.22–2.6 in.). These settlements

are comparable to the 6.86–7.34 cm (2.7–2.9 in.) settlements estimated for the MOX fuel

fabrication building if no soft zones are included in the analysis. The difference may be

attributed to relatively larger structural dimensions and expected foundation pressures for

the MOX fuel fabrication building than those for Building 221-S (DCS, 2003).

The staff reviewed the information presented in the site geotechnical reports regarding

settlement analysis and determined that the approach for estimating structural

settlements is acceptable because it is based on current industry guidance and

analysis techniques.

The postearthquake dynamic settlement of the potentially liquefiable soil was estimated

in the third MFFF site geotechnical report (DCS, 2003) based on the PC-3+ design basis

ground motion and the 1886 Charleston motion. The settlement ranges from 0.66–3.73

cm (0.26–1.47 in.) for the design basis ground motion and 1.55–5.64 cm (0.61–2.22 in.)

for the 1886 Charleston motion. The applicant indicated that these dynamic settlements

might occur in loose or soft strata below the ground water level, at a depth of 18.29 m

(60 ft) or more. There are two significantly stiffer soil layers more than 12.19 m (40 ft)

thick between the potentially liquefiable zones and the foundations. According to the

MFFF site geotechnical reports (DCS, 2001; DCS, 2003), these two soil layers would

tend to redistribute the estimated dynamic settlement such that no significant differential

settlement would occur at the foundation level.

The staff reviewed the information presented in the site geotechnical report regarding the

dynamic settlements and concur that the postearthquake-induced dynamic settlements

resulting from localized liquefaction will not create stability problems for the foundations

for the SC-I structures.

• Aircraft Crash Hazard

As discussed in Section 1.1.1.9 of this FSER, the estimated total aircraft crash probability

per year, including those induced by the federal airways and SRS helicopters for the

MFFF building, is 2.99×10!8 and 6.67×10!8 for the emergency generator building, based

on the applicant’s calculation. Each of the estimated probabilities is smaller than the 10!7

annual probability of unacceptable radiological consequences indicated in Section 3.5.1.6
11-13

of NUREG-0800 (NRC, 1987). As a result, the applicant concluded that aircraft crash

hazards at the facility are not a design concern. 

In its review of the aircraft hazard analysis, the staff also included an in-office

review of the applicant’s calculation of projected future flight information (Section 5.1.5.2

of this FSER). The staff found these calculations to be acceptable.

• Load Combinations

The load combinations for both SC-I and SC-II civil structures were determined using

Section 3.8.4, “Other Seismic Category I Structures,” of NUREG-0800 (NRC, 1987) as a

guide. Tornado, tornado missile, and explosion loads will not be considered in the load

combinations of SC-II structures. The staff reviewed the various load combinations

presented in Sections 11.1.7.4.2.1, “Loading Combinations for SC-I Concrete

Structures,” and 11.1.7.4.2.2, “Loading Combinations for SC-I Steel Structures,” of the

revised CAR and determined that these load combinations are the same as those

suggested in NUREG-0800 for the design of structures. Therefore, the load

combinations are acceptable to the staff.

11.1.2 Evaluation Findings

In Section 11.1 of the revised CAR, DCS provided design basis information for civil structural

systems that it identified as PSSCs for the proposed facility. Based on the staff’s review of the

revised CAR and supporting information provided by the applicant relevant to civil structural

systems, the staff finds that DCS has met the BDC set forth in 10 CFR 70.64(a)(2). Further, the

staff concludes, pursuant to 10 CFR 70.23(b), that the design bases of the PSSCs identified by

the applicant will provide reasonable assurance of protection against natural phenomena and the

consequences of potential accidents.

11.1.3 References

(AASHO, 1996) American Association of State Highway and Transportation Officials. “Standard

Specifications for Highway Bridges,” 16th Edition. Washington, DC, 1996.

(ACI, 1996) American Concrete Institute. ACI 349.1R-91, “Reinforced Concrete Design for

Thermal Effects on Nuclear Power Plant Structures.” Reapproved, 1996. Detroit, Michigan,

1996.

(ACI, 1997) American Concrete Institute. ACI 349-97, “Code Requirements for Nuclear SafetyRelated Concrete Structures.” Detroit, Michigan, 1997.

(ACI, 1998) American Concrete Institute. ACI 349.2R-97, “Embedment Design Examples.”

Detroit, Michigan, 1998.

(ACI, 1999) American Concrete Institute. ACI 318-99, “Building Code Requirements for

Structural Concrete.” Detroit, Michigan, 1999.
11-14

(AISC, 1994) American National Standards Institute/American Institute of Steel Construction. 

ANSI/AISC N690–1994, “Specification for the Design, Fabrication, and Erection of Steel SafetyRelated Structures for Nuclear Facilities.” Chicago, Illinois, 1994.

(ANS, 1992) American National Standards Institute/American Nuclear Society. 

ANSI/ANS-2.8-1992, “Determining Design Basis Flooding at Power Reactor Sites.” La Grange

Park, Illinois, 1992.

(ASCE, 1998) American Society of Civil Engineers. ASCE 7-98, “Minimum Design Loads for

Buildings and Other Structures.” Reston, Virginia, 1998.

(ASCE, 1999) American Society of Civil Engineers. ASCE 4-98, “Seismic Analysis of SafetyRelated Nuclear Structures and Commentary.” Reston, Virginia, 1999.

(DCS, 2001) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: MOX Fuel Fabrication Facility Site Geotechnical Report

(DCS01–WRS–DS–NTE–G–00005–C). August 10, 2001. 

(DCS, 2002a) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Clarification of Responses to NRC Request for Additional Information

(DCS–NRC–000085). March 8, 2002.

(DCS, 2002b) Ihde, R., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, Mixed Oxide Fuel Fabrication Facility Construction Authorization Request Revision. 

October 31, 2002.

(DCS, 2002c) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, Requests for Additional Information, Clarification, and Open Item Mapping into the

Construction Authorization Request Revision (DCS01–NRC–000120). November 22, 2002. 

(DCS, 2003) Duke Cogema Stone & Webster. MOX Fuel Fabrication Facility Site Geotechnical

Report (DCS01-WRS-DS-NTE-G-00005-E), report prepared for the Department of Energy,

Chicago Operations Office. June 30, 2003. 

(DCS, 2004) Ihde, R, Duke Cogema Stone and Webster. Letter to W. Kane, U.S. Nuclear

Regulatory Commission, RE: Mixed Oxide Fuel Fabrication Facility—Construction Authorization

Request. 2004.

(NRC, 1973a) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.60, “Design

Response Spectra for Seismic Design of Nuclear Power Plants.” Washington, DC, 1973.

(NRC, 1973b) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.61, “Damping Values

for Seismic Design of Nuclear Power Plants.” Washington, DC, 1973.

(NRC, 1976) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.92, “Combining Modal

Responses and Spatial Components in Seismic Response Analysis.” Washington, DC, 1976.
11-15

(NRC, 1977) U.S. Nuclear Regulatory Commission. Regulatory Guide 3.40, “Design Basis

Floods for Fuel Reprocessing Plants and for Plutonium Processing and Fuel Fabrication Plants.” 

Washington, DC, 1977.

(NRC, 1978) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.122, “Development of

Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or

Components.” Washington, DC, 1978.

(NRC, 1980) U.S. Nuclear Regulatory Commission. NUREG-0612, “Control of Heavy Loads at

Nuclear Power Plants.” Washington, DC, 1980.

(NRC, 1987) U.S. Nuclear Regulatory Commission. NUREG-0800, “Standard Review Plan for

the Review of Safety Analysis Reports for Nuclear Power Plants.” Washington, DC, 1987.

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility.” Washington,

DC, 2000.

(NRC, 2001) Giitter, J.G., U. S. Nuclear Regulatory Commission. Letter to P. Hastings, Duke

Cogema Stone & Webster, RE: Mixed Oxide Fuel Fabrication Facility Construction

Authorization—Request for Additional Information. June 21, 2001.

11.2 Aqueous Polishing Process and Chemistry

11.2.1 Conduct of Review

This section of the FSER contains the staff’s review of the AP process safety described by the

applicant in Section 11.3 of the revised CAR (DCS, 2002e), with supporting process safety

information from Chapters 5, 8, and 11 of the revised CAR, supplementary information provided

by the applicant, and relevant documents available at the applicant’s offices but not submitted by

the applicant. The staff also reviewed technical literature as necessary to understand the

process and safety requirements. The objective of this review is to determine whether the

chemical process safety PSSCs and their design bases provide reasonable assurance of

protection against natural phenomena and the consequences of potential accidents. The review

of AP safety design bases and strategies was closely coordinated with the review of the radiation

and chemical safety aspects of accident sequences described in the safety assessment of the

design bases (see Chapter 5 of this FSER), the review of fire safety aspects (see Chapter 7 of

this FSER), and the review of plant systems (see Chapter 11 of this FSER). 

The staff evaluated the AP process and chemistry information in the revised CAR against the

following regulations:

• 10 CFR 70.23(a) states, as a prerequisite to construction approval, that the design bases

of the PSSCs and the Quality Assurance Program be found to provide reasonable

assurance of protection against natural phenomena and the consequences of potential

accidents.

• 10 CFR 70.64 requires that BDC and defense-in-depth practices be incorporated into the

design of new facilities or new processes at existing facilities. With respect to chemical
11-16

protection, 10 CFR 70.64(a)(5) requires that the MFFF design 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. Related to chemical protection, 10 CFR 70.64(a)(3) requires that the

facility design provide for adequate protection against fires and explosions, such as those

that could be initiated by or involve chemicals at the proposed facility.

The review of the revised CAR focused on the design basis of chemical process safety systems,

their components, and other related information. For each chemical process safety system, the

staff reviewed information provided by the applicant for the safety function, system description,

and safety analysis. The review also encompassed proposed design basis considerations such

as redundancy, independence, reliability, and quality. The staff used Chapter 8.0 of

NUREG-1718 as guidance in performing the review (NRC, 2000).

As stated in the memorandum of understanding (MOU) between the NRC and the Occupational

Safety and Health Administration, “Worker Protection at NRC-licensed Facilities,” (Federal

Register, Vol. 53, No. 210, October 31, 1998, pp. 43950–43951), the NRC oversees chemical

safety issues related to (1) radiation risk produced by radioactive materials, (2) chemical risk

produced by radioactive materials, and (3) plant conditions that affect the safety and safe

handling of radioactive materials. These chemical safety issues are important because they

represent an increased radiation risk to workers. The NRC does not oversee facility conditions

that result in an occupational risk but do not affect the safe use of licensed radioactive material. 

The MOU provisions applicable to the proposed MOX facility are now codified in 10 CFR

70.64(a)(5). 

The NRC staff reviewed the following areas of the revised CAR applicable to process safety at

the construction approval stage and consistent with the level of design (NRC, 2000, p. 8.0-8):

• AP description

• hazardous chemicals and potential interactions affecting licensed materials

• AP chemical accident sequences

• AP chemical accident consequences

• AP safety controls

Additional documentation from the applicant and the literature was reviewed as necessary to

understand the process and safety requirements. In addition, the revised CAR incorporates the

BDC of 10 CFR 70.64(a) into the design and operations of the proposed facility (see revised

CAR, p. 5.5-53), and applicable sections of the revised CAR are intended to demonstrate

compliance with these BDCs. The staff’s detailed evaluation of the proposed AP process is

presented in the sections that follow.

11.2.1.1 System Description of the AP Process

This section provides a description and overview of the AP process, including design,

operational, and process flow information. A summary of the major components and functions is

discussed in Section 11.2.1.2 of this FSER.

The AP process is designed to receive weapons-grade plutonium from the proposed pit

disassembly and conversion facility (PDCF) and alternate feedstock (AFS) at SRS and to
11-17

remove the impurities from the feed plutonium from the PDCF and AFS for use in the MOX

process (MP). There are four major steps in AP:

(1) The first step is the dissolution of plutonium dioxide powder by electrically generated

silver (II) ions. The dissolution step involves the use of silver as a catalyst in a nitric acid

medium. This electrolytically assisted route is independent of the powder characteristics. 

Dissolution is complete, with the rate determined by the generation of the silver (II) ions,

at relatively mild temperatures (20–40 °C (68–104 °F); nominal temperature 30 °C (86

°F)). Dechlorination is required before dissolution for those powders containing chlorides

(i.e., for AFS).

(2) The second step involves plutonium purification by solvent extraction. Purification

extracts and recovers plutonium, regenerates the solvent, and recycles nitric acid. 

Solvent extraction removes impurities such as gallium, uranium, and americium (CAR

Section 11.3). The proposed facility would use a modified plutonium uranium reduction

extraction (PUREX) process (e.g., Bennedict, 1981).

(3) Third, the process includes conversion into plutonium dioxide by continuous oxalate

calcination. Precipitation and calcination of the plutonium oxalate is a standard process

used to prepare plutonium dioxide fuels.

(4) Finally, recovery and recycle efforts are completed. Process side streams are treated to

recover and recycle chemical reagents and reduce waste generation, as appropriate.

The proposed design of the AP process is similar to the design currently employed at La

Hague’s plutonium finishing facilities. Departures from the La Hague design result from U.S.

regulatory requirements, lessons learned at La Hague, or manufacturing and throughput

requirements specific to the facility.

11.2.1.2 Major Components and Functions

The AP process can be segmented into the following four operational areas (unit symbols are

provided in parentheses):

(1) plutonium purification process—includes the decanning unit (KDA), dissolution unit

(KDB), dechlorination and dissolution unit (KDD), purification cycle (KPA), oxalic

precipitation and oxidation unit (KCA), homogenization unit (KCB), milling unit (KDM),

canning unit (KCC), and recanning unit (KDR), but no longer includes a uranium oxide

dissolution unit

(2) recovery processes—includes the solvent recovery cycle (KPB), oxalic mother liquor

recovery unit (KCD), and acid recovery unit (KPC), but no longer includes a silver

recovery unit.

(3) waste storage—includes the liquid waste reception unit (KWD) and waste organic solvent

unit (KWS)

(4) offgas treatment—includes the offgas treatment unit (KWG)
11-18

11.2.1.2.1 Decanning Unit

The decanning unit consists of a series of workstations and gloveboxes distributed between the

MP and AP areas. All cans entering the facility via this unit are initially packaged in 3013

standard containers and meet the 3013 standard. For this reason, the staff notes that such

powders have been stabilized and are not pyrophoric and do not contain significant volatile

species.

The three main functions of the decanning unit include the following: 

(1) density measurement (PDCF powders only)

(2) opening of outer, inner, and convenience cans

(3) transfer of powder to a dissolution dosing hopper (PDCF powder) or to a reusable can for

the milling unit (AFS powder)

Density is determined from an x-ray level measurement on the entire, sealed 3013 container.

Cans containing PDCF powders with a density less than 7 g/cc are sent to the dissolver for

processing. Any PDCF powders with densities near or greater than 7 g/cc require ball milling

and a subsequent, confirmatory density measurement to meet the less-than-7-g/cc limit. All

AFS powders are ball milled. In addition, AFS powders are analyzed as follows:

• If chlorinated species are detected, the reusable can is emptied into the dosing hopper of

one of the electrolyzers in the dechlorination and dissolution unit.

• If the powder does not contain chlorinated species, the reusable can is fed into the

dosing hopper of the dissolution unit’s electrolyzer.

• If the powder contains chemical species not compatible with the AP process, the powder

is transferred to the recanning unit for repackaging.

The plutonium feed design basis is discussed in Section 11.2.1.4 of this FSER.

11.2.1.2.2 Milling Unit

The milling unit consists of a series of gloveboxes. The functions of the milling unit are as

follows: 

• Mill AFS powders to homogenize the plutonium dioxide and decrease the particle size to

a level compatible with the dissolution process.

• Perform density measurements on the milled powders.

• Sample the powder for chemical analysis.

• Store powders while awaiting analytical results. 
11-19

As described in Section 11.2.1.2.1, the AFS powders are analyzed in the laboratory. Depending

on the results of those analyses, the powder is transferred to different processes. If the chloride

concentration is found to exceed 500 micrograms/gram of plutonium (500 ppm), the powder is

directed to the feeding hopper for one of the two dechlorination electrolyzers (dechlorination and

dissolution unit). If the chloride level exceeds the design basis values discussed in Section

11.2.1.4, the reusable can is sent to the recanning unit for repackaging in a 3013 container.

11.2.1.2.3 Recanning Unit

The recanning unit consists of two gloveboxes and two workstations designed to repackage

powders that are not compatible with the AP process into 3013 containers. The powder is

analyzed before transfer out of any unit in the facility. The operations include welding, tightness,

and contamination checks. 

11.2.1.2.4 Dissolution Unit

The function of the dissolution unit is to dissolve plutonium dioxide powders containing less than

500 ppm chlorides. The plutonium dioxide is electrolytically dissolved in the dissolution unit as a

precursor for separating impurities (specifically americium, gallium, and uranium) in the

purification cycle. Samples from the dilution and sampling tank are analyzed to determine the

fissile material content and the required degree of dilution before being sent to the purification

cycle feed tank.

The dissolution kinetics are improved by augmenting the reaction with a strong oxidizing agent

(in this case, by electrolytic dissolution with silver ions (Ag[II])). Silver ions (Ag[II]) are

electrolytically produced in a cylindrical compartment. The electrolytic dissolution takes place in

a 6 N nitric acid solution at nominally 30 °C (86 °F). The general dissolution process may be

described as follows:

 

Electrolytic production of Ag2+ (at the anode):

Ag+ = Ag2+ + e-(11.2-1)

Dissolution of PuO2 powder:

PuO2 (solid) + Ag2+ = PuO2

+

 (solid) + Ag+ (11.2-2)

 PuO2

+

 (solid) + HNO3 = PuO2

+

 (solution) (11.2-3)

PuO2

+

 (solution) + Ag2+ = PuO2

2+ (solution) + Ag+

 (11.2-4)

This gives the following general reaction:

PuO2 (solid) + 2Ag2+ = (PuO2)

2+ (solution) + 2Ag+

(11.2-5)

Ag+ ions are oxidized at the anode. The staff review notes that electrolytic dissolution of the

plutonium dioxide is indirect; electrolysis produces silver (II), which affects the actual dissolution

of the plutonium dioxide. If a sufficient concentration of silver is not available, other anode (and

near-anode) reactions might occur, such as the following:

2 H2O = O2 (gas) + 4H+ + 4e- (water electrolysis/oxidation) (11.2-6)

2 Ag2+ + H2O = 2 Ag+ + 2 H+ + (1⁄2) O2 (gas) (11.2-7)
11-20

(water oxidation by silver)

Water oxidation by silver (II) is promoted at temperatures of 40 °C (104 °F) and above (Bourges,

1985). The production of oxygen may have safety concerns. For example, the presence of

oxygen can lead to explosions with hydrogen. 

The applicant stated that the following reduction reaction takes place at the cathode:

NO3

-

 + 3H+ + 2e- = HNO2 + H2O (11.2-8)

The literature also identifies the following cathode reactions (Bray, 1987, 1992; Bourges, 1985):

2 H+ + NO3

-

 + e- = NO2 (gas) + H2O (11.2-9)

4 H+ + NO3

-

 + 3e- = NO (gas) + 2 H2O (11.2-10)

Hydrogen generation can also occur at the cathode:

2 H+ + 2 e- = H2 (gas) (11.2-11)

The staff review notes hydrogen generation usually occurs at a low rate at all times. Under offnormal conditions, such as overvoltage or reduced acid concentrations, hydrogen generation

has the potential to increase substantially and become the dominant cathode reaction.

Dissolution occurs when a current is applied. The joule heating effect of the electrical current

supplied is reduced by cooling the anolyte and the catholyte. The staff review notes that

electrolytic processes usually operate at 90–95 percent efficiency (i.e., 90–95 percent of the

current goes towards the intended reaction) under the best conditions. The literature indicates

that the silver-assisted plutonium dissolution reaction occurs with 30–70 percent efficiencies

(Bray, 1987, 1992; Bourges, 1985). Side reactions almost invariably occur and likely involve the

evolution of gases, such as hydrogen, oxygen, and nitrogen oxides. At higher and lower

electrode voltages, the electrolyzer would operate in a different regime, and a higher percentage

of the current could produce gas evolution. 

The dissolution unit is operated in batches. The dissolution unit and the companion

dechlorination and dissolution unit are designed to treat 22 kg/day (48.51 lb/day) of plutonium

dioxide. The operating ranges depend on the quantity of AFS powder processed. During the

processing of AFS powder (which may occur over the first 3 years of operation), the dissolution

units are anticipated to process between 4 kg/week (8.8 lbs/week) of powder (i.e., one batch

from the dechlorination and dissolution unit) up to a maximum of 129 kg/week (284 lbs/week) of

powder. This correlates to six dissolutions per week for both lines of the dechlorination and

dissolution unit and six batches of the single line in the dissolution unit (13.5 kg/batch (29.8

lb/batch)). During the processing of PDCF powders, the minimum processed is expected to be

13.54 kg/week (29.8 lbs/week) of powder (i.e., one batch through the single electrolyzer of the

dissolution unit). The maximum processed is estimated as 243 kg/week (536 lbs/week) of

powder, corresponding to six dissolutions per week for all three electrolyzers. The nominal flow

rate to the purification cycle is approximately 14.4 L/h (3.8 gal/h).
11-21

Hydrogen peroxide is added to the receiving tank used for interim storage in order to adjust the

oxidation state of the plutonium from (VI) to (IV). A plutonium (IV) oxidation state allows for

better extraction and separations processing. The peroxide also reduces any excess silver (II)

to silver (I). The uranium impurity exists as the 235U isotope (from radioactive decay of 239Pu and

nonseparable portions of the original pit—essentially more than 93 percent uranium enrichment

or assay). Consequently, an isotopic dilution to around 30 percent assay is made by feeding the

appropriate quantity of depleted uranium nitrate solution (0.25 percent 235U) to the dilution and

sampling tank. Other adjustments (e.g., acidity) may also be made to the solution in the dilution

and sampling tank to optimize subsequent purification of the plutonium.

11.2.1.2.5 Dechlorination and Dissolution Unit

The dechlorination and dissolution unit includes two electrolyzers, each the same size as the

single electrolyzer in the dissolution unit. However, as noted previously, batch sizes are different

depending on the feed material.

The functions of the dechlorination and dissolution unit are to remove chloride ions from the

plutonium dioxide powder and then dissolve the material. The unit processes powders with

chloride levels exceeding 500 ppm. The unit can also be used to process nonchlorinated feed

materials (e.g., the PDCF feed). 

Each electrolyzer line includes a scrubbing column for the removal of chlorine. The electrolyzers

are operated in two modes. The first mode is dechlorination—electricity is applied and the

chlorine gas is evolved at the anode as follows:

2 Cl- = Cl2 (g) + 2 e-(11.2-12)

The applicant indicated that the process removes greater than 99 percent of the initial chloride in

the powder, based on operation with 6 N nitric acid and a temperature of 60 °C (140 °F).

In the second mode of operation (i.e., after chloride removal), the electrolyzer is operated in a

manner completely analogous to the dissolution unit electrolyzer. Silver ions are fed into the

electrolyzer, and the powder is dissolved.

11.2.1.2.6 Purification Cycle

The purification cycle uses a modification of the standard PUREX process (Bennedict, 1981). 

Plutonium nitrate from the dissolution unit is received, and plutonium is solvent extracted and

scrubbed for impurities. The plutonium with uranium left in the stream is stripped after

adjustment of the plutonium valence to the trivalent state. The purification cycle includes

plutonium reception, recycle, and transfer to the oxalic precipitation and oxidation unit. The

purification cycle also sends the solvent/diluent stream to the solvent recovery cycle and the

raffinate stream to the acid recovery unit.

The main goal of the purification cycle is to separate plutonium from impurities contained in the

solution coming from the dissolution unit. In the revised CAR, the applicant identified the main

functions of the purification cycle as follows:
11-22

• Receive plutonium nitrate from the dissolution unit and the dechlorination and dissolution

unit.

• Receive recycled plutonium nitrate from the oxalic mother liquor recovery unit.

• Receive solutions with high plutonium content from the laboratories.

• Perform plutonium extraction and impurities scrubbing.

• Perform plutonium stripping and uranium scrubbing.

• Perform further plutonium purification by additional plutonium stripping and diluent

washing.

• Adjust plutonium to the tetravalent state.

• Perform plutonium stripping in a plutonium barrier.

• Perform scrub bed uranium stripping and diluent washing.

• Control purified plutonium and transfer to the oxalic precipitation and oxidation unit or

recycle to the beginning of the purification cycle.

• Control and dilute scrubbed uranium and prepare uranium for transfer to the liquid waste

reception unit.

• Wash, control, and transfer raffinates to the acid recovery unit.

• Wash, control, and transfer solvent/diluent to the solvent recovery cycle.

• Destroy residual hydroxylamine nitrate (HAN)/hydrazine and hydrazoic acid in the

aqueous phase transferred to the oxalic precipitation and oxidation unit.

The extraction process is continuous, but the feed solutions from the dissolution unit are

received in batches. The raffinate and the plutonium nitrate solutions are transferred

continuously to the acid recovery unit inlet buffer storage and to the oxalic precipitation and

oxidation unit inlet buffer storage, respectively.

Plutonium nitrate solution is batched to the feed tank for plutonium extraction and impurities

scrubbing. Plutonium (IV) in the aqueous solution (4.5 N HNO3) is extracted by the solvent (30

percent tributyl phosphate (TBP) in a branched dodecane—the latter is sometimes referred to as

hydrogenated propylene tetramer (HPT)) in a pulsed extraction column. The impurities remain

primarily in the aqueous phase. The solvent stream is scrubbed by 1.5 N nitric acid in a pulsed

scrubbing column to ensure good decontamination. The aqueous raffinates are washed by the

diluent in a pulsed column and transferred to the raffinate reception tank.

Plutonium (IV) is reduced to plutonium (III) by HAN ([NH3OH+][NO3

-

]), and plutonium (III) is

stripped in a pulsed stripping column (Reactions 11.2-13 and 11.2-14). Hydrazine nitrate is

introduced to prevent parasitic reoxidation of plutonium (III) back to plutonium (IV) (Reactions
11-23

11.2-15 and 11.2-16). The stripped plutonium is washed with diluent in a pulsed diluent washing

column before the final valence adjustment. Unstripped plutonium is extracted in the plutonium

barrier mixer-settler bank. Hydrazine nitrate and HAN are introduced in the last stage of the

plutonium barrier. The solvent from the plutonium barrier flows to the uranium-stripping

mixer-settler bank. These reactions are written as follows:

Plutonium reduction by HAN (NH3OH[NO3]):

2[NH3OH]+ + 4Pu+4 = 4Pu+3 + N2O(g) + H2O + 6H+(11.2-13)

2[NH3OH]+ + 2Pu+4 = 2Pu+3 + N2(g) + 2H2O + 4H+(11.2-14)

Plutonium reduction by hydrazine:

4Pu4+ + N2H5

+

 + H2O = 4 Pu3+ + N2O + 5 H+(11.2-15)

Parasitic reoxidation of Pu(III) to Pu(IV):

2Pu3+ + 2HNO2 + 3H+ + NO3- = 2Pu4+ + 3HNO2 + H2(11.2-16)

Uranium is stripped (recovered from the organic phase) in a slightly acidic 0.02 N nitric acid

solution in a uranium-stripping, mixer-settler bank. The stripped uranium stream is washed with

diluent in a three-stage, diluent-washing, mixer-settler bank. Isotopic dilution with depleted

uranium nitrate solution is performed as necessary. The stripped solvent from the

uranium-stripping mixer-settler is directed to the solvent recovery cycle. The aqueous phase

from the uranium diluent washing is directed to the liquid waste reception unit.

The final valence adjustment of plutonium (III) to plutonium (IV) is achieved by oxidizing the

plutonium (III) solution with nitrous fumes (essentially a nitrogen dioxide/nitrogen tetraoxide

mixture). In this process, excess HAN and hydrazine are eliminated, and air stripping of the

plutonium solution in an air-stripping column destroys the nitrous acid. The plutonium nitrate

solution is received in the plutonium reception tank from where it is transferred to the batch

constitution tanks of the oxalic precipitation and oxidation unit.

 

The selected aqueous-to-organic ratios in the plutonium extraction and plutonium stripping

operations enable the process to obtain a plutonium concentration close to 40 g/L (0.34 lb/gal) at

the outlet of the purification cycle.

The purification cycle operates continuously. The feeding solutions from the dissolution unit are

received in batches. Thus, the nominal flow rate to purification is approximately 15.1 L/h (4

gal/h). This cycle is designed to process 14.5 kg/day (32 lb/day) of plutonium. The operating

range of the purification cycle is 11–19 kg/h (24.3–42 lb/h) of plutonium.

11.2.1.2.7 Solvent Recovery Cycle

The solvent recovery cycle operates continuously in conjunction with the purification cycle. 

Standard PUREX methods are used to wash the solvent and remove the degradation products. 

In the revised CAR, the applicant identified the functions of the solvent recovery cycle as follows:

• Recover the used solvent from the purification cycle to prevent the accumulation of

degradation products.
11-24

• Renew the solvent and adjust its TBP content.

• Store the treated solvent and continuously feed the purification cycle.

• Perform a diluent wash operation on the aqueous effluents produced by this operation to

remove traces of entrained solvent. Effluent in this section refers to effluent from

individual process units to other process units; the facility discharges no radioactive liquid

effluent directly into the environment.

Washing removes undesirable byproducts, such as hydrazoic acid and TBP degradation

products, from the solvent to prevent their accumulation and associated deleterious impact upon

process and safety performance. The solvent is washed by contact with an aqueous solution of

sodium carbonate and sodium hydroxide in mixer-settler banks. Washing removes hydrazoic

acid (as sodium azide) and TBP degradation products (mainly monobutyl phosphate and dibutyl

phosphate) from the solvent to the aqueous stream. The washed solvent is collected in a buffer

tank where it is cooled. The purification cycle is continuously fed at a controlled flow rate using a

dosing pump. The excess solvent, generated by the diluent wash and the content adjustment

TBP wash, is transferred to the liquid waste reception unit. The aqueous effluents generated by

washing undergo a diluent wash in a mixer-settler battery (one stage) at ambient temperature to

remove traces of entrained solvent. The aqueous-to-organic phase ratio for this operation is

around 100:1.

The diluent is recycled in the mixer-settler with a specific system including an airlift and two 

pots. The recycling flow rate equals the incoming aqueous flow rate from the mixer-settler bank. 

The aqueous-to-organic ratio is close to 1 when recycling is in operation.

The unit is designed to treat solvents at a flow rate of 17.4 L/h (4.6 gal/h), which corresponds to

14.2 kg/day (31.7 lb/day) of plutonium.

11.2.1.2.8 Oxalic Precipitation and Oxidation Unit

In the revised CAR, the applicant identified the functions of the oxalic precipitation and oxidation

unit as follows:

• Receive purified plutonium nitrate concentrated to approximately 40 g/L (0.34 lb/gal )

from the purification cycle and prepare uniform batches.

• Precipitate out the plutonium nitrate as oxalate.

• Produce plutonium dioxide after filtering, drying, and calcining the oxalate. The filtering

operation includes drawing off the mother liquors, washing, and dewatering the plutonium

oxalate cake.

• Transfer the plutonium dioxide to the homogenization unit and transfer the mother liquors

and the filter washing solutions to the oxalic mother liquor recovery unit. Solutions are

transferred by air lifts and flow control valves as described in Section 11.3.2.8.2 of the

revised CAR.
11-25

• Ensure reducing agents, hydrazoic acid, and plutonium (VI) do not propagate into

downstream processing units (e.g., the oxalic mother liquor recovery unit).

The following depicts the precipitation reaction:

Pu(NO3)4 + 2 H2C2O4 = Pu(C2O4)2 (s) + 4 HNO3(11.2-17)

The conversion line is rated for the processing of 1.1 kg/h (2.3 lb/h) (25.7 kg/day (56.6 lb/day))

of plutonium. The operating range of the oxalic precipitation and oxidation unit is 0–1.25 kg/h

(2.75 lb/h) of plutonium. Plutonium nitrate solutions arrive from the purification cycle where

acidity and valency are adjusted. They are received in alternate batches in two annular tanks to

form a batch with a volume of 0.6 m3 (21.2 ft

3

 or about 158.5 gal). Solutions are transferred by a

pump to two dosing wheels, which supply one precipitator each. The solutions flow by gravity to

the precipitators.

Precipitation takes place in two precipitators which are connected in parallel. The oxalate

reagents are injected into each precipitator. The plutonium oxalate precipitate carried by the

mother liquors escapes via the precipitator overflows and flows by gravity to a rotary filter.

Rotation ensures that dewatering and cake removal are continuously performed. The filter

removes the plutonium oxalate cake, plate by plate, with a scraper. The removed cake is

collected by a screw mechanism and falls into a chute. It enters the next processing operation

(the calcination furnace) by gravity.

The furnace consists of two main parts—a drying zone where the plutonium oxalate is dried, and

a calcining zone where the oxalate is transformed into plutonium dioxide in an oxidizing

atmosphere of oxygen. The following shows the overall reaction:

Pu(C2O4)2 + O2 = PuO2 + 4 CO2(11.2-18)

The staff notes that this is plutonium (IV) oxalate, which generally requires a higher temperature

(400–500 °C) for conversion to the oxide than either plutonium (III) or plutonium (VI) oxalates

(around 200 °C). Potential concerns and controls for plutonium (VI) oxalate are discussed

further in Section 8.1.2.5.1.6 of this FSER. Potential concerns and controls with plutonium (III)

oxalate are discussed in Section 11.3 under plutonium pyrophoricity.

The oxalic mother liquors, which are collected in separator pots, flow by gravity to the oxalic

mother liquor recovery unit. The filtered mother liquors are adjusted to approximately 3.3 N with

recovered 13.6 N nitric acid to avoid any risk of precipitation of plutonium oxalate caused by

residual oxalic acid. 

The gases produced during drying and calcination of the plutonium oxalate (e.g., carbon dioxide

and steam, nitric oxides, and trace organics), the excess of oxygen, and the air from upstream

and downstream of the process are removed by a negative-pressure circuit. This system

comprises a filter, a condenser, a demister, an electric heater, two high-efficiency particulate air

(HEPA) filters, and two fans.
11-26

11.2.1.2.9 Homogenization Area

The plutonium dioxide produced in the oxalic precipitation and oxidation unit is continuously fed

by gravity from the calcination furnace into one of the two separating hoppers installed in

parallel. The plutonium material balance is determined by weighing the filled cans (canning unit)

and by determining the plutonium content of the hopper by powder sampling. Sampling ensures

that all the finished product specifications are met in each batch of plutonium dioxide in each

hopper and checks the isotopic composition of the plutonium dioxide for the finished product of

each batch in each hopper. 

In the revised CAR, the applicant identified the functions of the homogenization unit as follows:

• Receive and homogenize the plutonium dioxide produced in the oxalic precipitation and

oxidation unit.

• Fill cans with plutonium dioxide in such a manner that the mass of plutonium per can is

constant.

• Prepare samples for laboratory analysis to characterize the batch.

• Perform sample-based residual moisture measurement.

• Perform thermogravimetry analysis.

• Store reference samples.

11.2.1.2.10 Canning Unit

The canning unit is designed to package plutonium dioxide powder in reusable stainless steel

cans and transfer them one by one to the MP plutonium dioxide buffer storage unit to prepare

the MOX powder. It is also used to establish the plutonium dioxide powder material balance. 

The nominal capacity is about 10 cans of plutonium dioxide per day, each filled with

approximately 2.4 kg (5.3 lb) of plutonium dioxide. The plutonium dioxide powder is fed by

gravity from the homogenizer at a temperature not exceeding 150 °C (302 °F). Full plutonium

dioxide cans are transferred pneumatically in a shuttle to the MP plutonium dioxide buffer

storage unit. Cans that are rejected because of overfilling (as indicated by weighing) or

unsatisfactory laboratory results are transferred and recycled to the appropriate upstream

process. The nominal flow rates are as follows:

• plutonium dioxide inlet from the homogenization unit—1.2 kg/h (2.6 lb/h)

• plutonium dioxide outlet—approximately 10 full reusable cans per day

11.2.1.2.11 Oxalic Mother Liquor Recovery Unit

In the revised CAR, the applicant identified the functions of the oxalic mother liquor recovery unit

as follows:

• Continuously receive oxalic acid mother liquors adjusted to 3.3 N with nitric acid from the

oxalic precipitation and oxidation unit.
11-27

• Continuously receive ventilation effluent droplets from the oxidation and degassing

columns.

• Concentrate the oxalic mother liquors in a subcritical evaporator to destroy the oxalic acid

and remove plutonium from the distillates.

• Check and transfer the distillates to the acid recovery unit.

• Monitor and recycle, batchwise, the concentrates to the top of the purification cycle.

The nominal capacity corresponds to the processing of the materials generated by the

precipitation of 24 kg (HM)/day (52.9 lb/day) of plutonium. This translates into a liquor flow rate

of approximately 250 L/h (66 gal/h). The oxalic acid mother liquor recovery unit operates

continuously, unlike the oxalic acid precipitation and oxidation unit, which produces the oxalic

mother liquors. Consequently, the design includes buffer tanks with more than 3 days capacity. 

This allows for the independent operation of the two units. The mother liquor solution flows by

gravity into the buffer tanks (two tanks, about 1 m3 (264 gal) each). After sampling for plutonium

concentration, an airlift transfers the solution into a feed tank (also of 1 m3 (264 gal) capacity). 

These tanks have a geometrically safe, annular design. A double airlift transfers the solution

from the feed tank into an evaporator. The evaporator concentrates the liquor and generates a

relatively clean overhead product (distillate). In the evaporator, residual plutonium oxalate is

converted into plutonium nitrate and oxalic acid. In the presence of manganese (II) ions (added

as a catalyst) and excess nitric acid, the resulting free oxalic acid decomposes into carbon

dioxide, water, and nitric oxide. The following equations depict these reactions:

Pu(C2O4)2 + 4 HNO3 = Pu(NO3)4 + 2 H2C2O4(Mn(II) as catalyst) (11.2-17)

H2C2O4 + 2 HNO3 = 2 CO2 + 2 NO2 + 2 H2O (Mn(II) as catalyst) (11.2-18)

The evaporator exposes the plutonium nitrate to prolonged boiling (100-135 °C (212-275 °F)) in

a highly acidic and oxidizing environment. Consequently, plutonium (IV) and plutonium (III) are

oxidized to plutonium (VI) nitrate by reactions such as the following:

3 Pu+3 + HNO3 + 3H+ = 3 Pu+4 + NO + 2 H2O

-

(11.2-19)

3 Pu+4 + 2 H2O = 2 Pu+3 + PuO2

+2 + 4 H+

(Mn(II) as catalyst) (11.2-20)

The distillate is condensed and cooled, and a small percentage is returned to the

evaporator/column system to supply reflux via a pot. The net distillate product is analyzed for

the plutonium concentration by an online neutron counter. In the revised CAR, the applicant

stated that, if the concentration is sufficiently low, the distillate is routed to the acid recovery unit.

If the plutonium specification is exceeded, the distillate is transferred to the buffer tanks for

recycling and retreatment.

The concentrates are removed from the evaporator by an airlift and placed in small buffer tanks. 

Because of the oxidation reactions, the plutonium is present in the hexavalent oxidation state (as

PuO2

+2). The applicant noted in the revised CAR that, if the residual oxalate concentration

meets requirements, then the concentrates are returned to the purification unit via an airlift.
11-28

11.2.1.2.12 Acid Recovery Unit

In the revised CAR, the applicant identified the following functions for the acid recovery unit:

• Receive extraction raffinates from the purification cycle, oxalic mother liquor distillates

from the oxalic mother liquor recovery unit, and effluents from laboratories in batches and

continuously receive active liquid effluents from the offgas treatment unit equipment

ventilation.

• Concentrate the radioactivity contained in the effluents and send it to the liquid waste

reception unit.

• Recover concentrated acid for recycling in the process.

• Recover distillates from the rectification column for use in the offgas treatment unit and

the purification cycle, with excess liquid to the liquid waste reception unit.

The acid recovery unit uses evaporation as the principal treatment method. The feed tank

receives the following materials:

• raffinates from the purification cycle in 1500 L (396 gal) batches

• oxalic mother liquor distillates (oxalic mother liquor recovery unit evaporator 3000) in

2500 L (660 gal) batches

• recombined acid from the offgas treatment unit (continuously)

• effluents from laboratories in batches

The solution is transferred by double-stage airlifts from the feed tank to the boiler of the first

evaporator, a natural recirculation, thermosiphon design. The heating power is kept constant by

regulating the steam pressure in the boiler. The concentrates are drained off into a 500 L (132

gal) tank discontinuously, several times a day. The concentrates contain the americium and

gallium impurities and are sent to the liquid waste reception unit. The overhead product is fed to

the second evaporator, which has a similar design and includes a rectification column on the

overhead product. The reflux system at the top of the column can be used to spray the upper

trays and improve decontamination. The recovered acid is drawn off as a concentrate product

by airlift and cooled. The acid drawoff flow rate is regulated to maintain the desired acidity. The

acid is recycled within the facility. The distillate product is continuously transferred by pump for

AP water recycling. Any excess recycle water is analyzed and temporarily stored before transfer

by pump to the liquid waste reception unit.

11.2.1.2.13 Silver Recovery Unit

The applicant decided to eliminate the silver recovery unit for economic reasons. The silver

nitrate used in the dissolution unit is part of the high alpha waste stream that will be sent to the

acid recovery unit.
11-29

11.2.1.2.14 Offgas Treatment Unit

The offgas treatment unit ventilation system is dedicated to maintaining a system negative

pressure for AP process confinement and for removal vapors and gases from processing

equipment. In the revised CAR, the applicant identified the following functions of this unit:

• Remove plutonium from offgases collected from the dissolution unit, the dechlorination

and dissolution unit (during the dissolution step), the oxalic precipitation and oxidation

unit, the oxalic mother liquors recovery unit, the acid recovery unit, the liquid waste

reception unit, and the oxidation and degassing columns (purification cycle).

• Recombine the nitrous fumes in a specific nitric oxide scrubbing column.

• Clean, by water scrubbing, the offgases collected from all AP units.

• Treat the offgas flow by HEPA filtration before releasing to the stack.

• Maintain negative pressure in the tanks and equipment connected to the process

ventilation system.

A specific offgas treatment unit extraction system is dedicated to the pulsed purification

columns, with similar functions:

• Treat offgases by HEPA filtration before releasing to the stack.

• Maintain negative pressure in the pulsation system and the pulsed columns legs.

The offgases containing nitric oxide (from dissolution/oxidation and degassing columns) are

gathered downstream of a cap impactor to remove droplets. The collected effluent stream is

recycled, by gravity, to the oxalic mother liquor recovery. Offgases are then scrubbed with

recycled effluents and with recovered distillates from the acid recovery unit. The offgases pass

through a demister, using an air ejector to provide the motive force. The extraction rate is

regulated based on the pressure in the scrubbing column.

Normal ventilation gases (i.e., process vents) are combined with the treated nitric oxide gas

streams. These gases are scrubbed with recycled effluents and then with water. The washed

gases successively pass through a cooler, a demister, an electric heater, a double HEPA

filtering line (2×2), and an exhauster before release through the stack.

The pulsation air from solvent extraction is treated in a separate (extraction) line. The air

successively passes through an electric heater, a HEPA filtering line (2×2), and an exhauster

before release through the stack.

The offgas treatment unit operates continuously. The nitric oxide scrubbing column is designed

to treat approximately 62 N m3/h (36 cubic feet per minute (cfm)). The main ventilation line

(offgas scrubbing and filters) is designed to process approximately 600 N m3/h (350 cfm). The

designed capacity of the column pulsation air extraction is 150 N m3/h (88 cfm).

The applicant stated that specific operating limits and the associated items relied on for safety

(IROFS) will be provided in the ISA.
11-30

11.2.1.2.15 Liquid Waste Reception Unit

The liquid waste reception unit receives liquid waste from the AP process for temporary storage

before sending it to SRS for treatment and processing (see Sections 11.3.2.14 and 10.1.4 of the

revised CAR). Waste stream descriptions and quantities are summarized in Table 11.2-1. The

functions of this unit are to treat the following liquid waste streams:

• The low-level liquid waste stream, which comprises of (1) the room HVAC condensate,

rinse water from laboratories, and washing water from sanitaries which are potentially

noncontaminated and are collected as low-low-level liquid waste, (2) the distillate stream

from the acid recovery unit which is contaminated and slightly acidic, (3) miscellaneous

floor washes from C2/C3 rooms and overflows or drip tray material from some of the

reagent tanks in the AP building, and (4) the chloride stream from the scrubbers used

during the dechlorination step for AFS feeds (i.e., from the dechlorination and dissolution

unit).

[Text removed under 10 CFR 2.390]

• The high alpha waste stream is a combination of three waste streams—americium,

alkaline waste, and excess acid. The americium stream collects americium and gallium

nitrates and all of the silver used in the dissolution unit, along with traces of plutonium. 

The alkaline waste stream from the solvent recovery area contains dilute caustic soda

(NaOH), sodium carbonate, sodium azide, and traces of uranium and plutonium. The

excess acid stream from the acid recovery unit contains high alpha activity excess acid. 

The high alpha storage tank along with the high alpha buffer storage tank are a holding

point for high alpha wastes and provide 90 days of storage.

[Text removed under 10 CFR 2.390]

• The stripped uranium (less than 1 percent 235U) waste stream receives the contents of

the uranium dilution tanks in the purification cycle. 

[Text removed under 10 CFR 2.390]

11.2.1.2.16 Waste Organic Solvent Unit

The function of the waste organic solvent unit is to handle the organic solvent generated by the

process and transfer it to SRS for handling and disposal.

The unit consists of an intermediate solvent tank, a final solvent tank, and a carboy filling station. 

It is designed to collect waste solvent for sampling in order to assure compliance with the SRS

waste acceptance criteria (WAC).

The intermediate solvent tank is located in a closed process cell and has mixing and automatic

sampling capabilities. Sampling is conducted to determine compliance with the SRS WAC. If

the waste solvent meets the WAC, it is transferred to the final waste solvent tank.

The final waste solvent tank receives solvent from the intermediate tank and from the solvent

recovery cycle unit solvent tank. This final tank is located in an open process cell and has

mixing and manual sampling capabilities. Waste solvent is transferred from this tank to a
11-31

Table 11.2-1 Waste Stream Descriptions and Quantities in the Waste Reception Unit

Waste Stream

Designation

Annual

Volume,

Liters (gal)

Main Chemical

or Isotope

Disposition,

Liters (gal)

Liquid americium

stream

(concentrated stream

from acid recovery)

11,400 (3,000)

(PDCF)

16,700 (4,400)

(AFS)

241Am < 24.5 kg/y (54 lb/y) (3.1x1015 Bq

(84,000 Ci))

Pu <150 g/y (0.33 lb/y)

Hydrogen ions—9.4 N

Nitrate salts—2200 kg/y (4850 lb/y) from

silver nitrate

Silver < 300 kg/y (661 lb/y)

Trace quantities of thallium, lead, and

mercury

High alpha waste to WSB

39,000

(10,300) (PDCF)

31,900

(8,421) (AFS)

Excess acid Stream 5,000 (1,321)

(AFS)

15,500 (4,100)

(PDCF)

Am < 90 mg/y (2×10-4 lb/y)

(rectification step after two evaporation

steps)

Hydrogen ions—13.6 N

Alkaline stream 12,100 (3,200)

(PDCF)

10,200 (2,700)

(AFS)

Pu < 16 g/y (0.035 lb/y)

U < 13 g/y (0.029 lb/y)

Na < 110 kg/y (243 lb/y)

Stripped uranium

stream

167,000

(44,000)

(PDCF)

151,000

(40,000) (AFS)

Pu < 0.1 mg/L

U < 3100 kg/y (6834 lb/y)(1% assay)

Hydrogen ions—0.14 N

Stripped uranium to WSB

167,000 (44,000) (PDCF)

151,000 (40,000) (AFS)

Excess low-level

radioactive solvent

wastes

10,600

(2,800) (PDCF)

9,800

(2,600) (AFS)

Solvent—30% TBP in dodecane

Pu < 17.2 mg/y (3.8×10-5 lb/y)

LLW solvent to SRS

solvent recovery

10,600 (2,800) (PDCF)

9,800 (2,600) (AFS)

Distillate waste 386,000

(102,000)

(PDCF)

265,000

(70,000) (AFS)

241Am-241 < 0.85 mg/y (1.9×10-6 lb/y)

Activity—2.39x108 Bq/y

Hydrogen ions—0.02 N

Liquid LLW

1,080,000

(285,000) (PDCF)

1,030,000

Chloride removal (273,000) (AFS)

waste

75,700

(20,000) (AFS)

Only when processing AFS materials

Chloride < 5 g/L (0.04 lb/gal)

Rinsing water 598,000

(158,000)

(PDCF)

598,000

(158,000)

(AFS)

Alpha activity < 4 Bq/L

Hydrogen ions—0–0.6 N

Internal HVAC

condensate

94,600

(25,000)

Trace contamination
11-32

carboy. The slightly contaminated solvent is anticipated to be a low-level waste (LLW). This

waste is stored in a 300-gallon carboy or other suitable vessel and transferred by truck to SRS

for disposition.

11.2.1.2.17 Uranium Dissolution Unit

The applicant decided to purchase depleted uranium nitrate solution as a reagent from an

outside supplier. Consequently, the uranium dissolving unit has been eliminated from the

proposed facility (DCS, 2002e).

11.2.1.2.18 Sampling System

The sampling system is used for radioactive and chemical solutions. Section 11.3.2.16 of the

revised CAR discusses the following three liquid sampling system approaches that the applicant

intends to use at the facility:

(1) In direct sampling, the solution is directly extracted from the process equipment by

gravity flow or with a recycling pump into a vial. Direct sampling is limited to

nonaggressive reagents or effluents of suspect origin. A large sample volume provides a

lower detection limit. 

(2) In suction sampling, a vial is filled by suction through a needle by the vacuum in the vial. 

Aggressive reagents can be sampled manually but with vacuum vial filling. Particles are

not expected in the sampling system because all AP process solutions are expected to

be clear. A moving cask is used for suction filling of active liquids. 

(3) With remote sampling, the solution is lifted up by an airlift head from which direct vacuum

sampling is carried out. For concentrated radioactive liquid waste, remote sampling

under a box is required. 

The applicant stated that all sampling systems will be qualified using engineering studies and/or

evaluations.

11.2.1.3 Staff Review of the AP Process System

11.2.1.3.1 Chloride Concentration of AFS Powder (KDM)

The staff evaluated the risk to the site worker and the individual outside the controlled area

(IOC) from chlorine emissions resulting from AP processes into the environment. Alternate

plutonium dioxide feedstock that contains greater than 500 ppm chloride salts would be

processed in the dechlorination electrolyzers. The maximum concentration of chloride salts in

alternate feedstock is expected to be a mass fraction of 20–33 percent chloride. If such a high

chloride-containing stream is directed to the normal (nondechlorination) dissolution electrolyzer,

which does not have a chlorine scrubber, by mistake, up to 3–5 kg (6.6–11 lbs) of chlorine gas

could be released. This would result in an intermediate-level consequence to the IOC if the

release were to occur in less than 1 hour. The staff identified this as Open Item AP-07 in the

revised draft safety evaluation report (DSER) (NRC, 2003c).
11-33

The staff has analyzed this event and concludes that the electrical current necessary to evolve

this rate of material would be significantly larger than the steady-state load of the electrolyzer,

and that the electrical current would have to exist for a relatively long time (15–60 minutes or

more) and would likely exceed standard protection devices, such as circuit breakers, and/or the

capacity of the electrolyzer itself. The controls would likely terminate the current before

significant quantities of chlorine could be evolved. Thus, the staff concludes that this is an

unlikely or highly unlikely event, and that no additional controls are necessary to meet the

performance requirements in 10 CFR 70.61. “Performance Requirements.” This closes AP-07

from the revised DSER (NRC, 2003c).

11.2.1.3.2 Electrolyzers (KDB and KDD)

The applicant provided information (DCS, 2001) that discussed a loss of confinement scenario

for the electrolyzer, based on an overtemperature situation caused by a control system failure,

electric isolation failure, or a loss of cooling

[Text removed under 10 CFR 2.390]

The applicant concluded that the event must be either prevented or mitigated and selected a

prevention strategy to reduce the risk to the facility worker and the environment, based on

shutdown of the electrolyzer and natural cooling. The applicant identified the safety design basis

as the detection of the high temperature (identified as greater than 70 °C (158 °F)) and

shutdown of the electrolyzer and related processes without exceeding any design limits or

chemical control limits, using assigned channels on the emergency control system. Shutdown

was understood to be a termination of the electrical current. The PSSCs identified by the

applicant are the temperature and shutdown controls, as well as the process safety

instrumentation and control (I&C) system.

The applicant’s response further noted that the electrolyzer is geometrically safe to preclude

potential criticality events. The applicant mentioned isolation of the anode and cathode and an

isolation monitoring system; these are not identified as PSSCs. The applicant also stated that

the scavenging and emergency air systems would be used to preclude the possibility of

explosions, based on the rate of hydrogen generated by radiolysis. Consequently, the applicant

indicated that the voltage to the electrolyzer would be limited. 

The safety strategy for the site worker and the IOC relies upon mitigation features. The PSSC is

the C3 confinement system, which has the safety function of filtering radioactive materials from

the C3 air before release into the atmosphere. 

In subsequent discussions (DCS, 2002d; NRC, 2002c), the applicant indicated that there will be

multiple, redundant temperature sensors located at the highest temperature location(s). These

would provide temperature signals to the process safety I&C system (a PSSC), which would

terminate electrical power. The applicant has also stated the temperature design basis of 70 °C

(125 °F) incorporates the effects of potential reaction exotherms. The power is actually

terminated before the temperature reaches 70 °C (125 °F). 

The NRC staff conducted a literature review on the electrical dissolution of plutonium dioxide

(Bray, 1987, 1992; Bourges, 1985). Key points from this review are as follows:
11-34

• Plutonium dioxide from different sources and calcination conditions (e.g., high-fired) can

be effectively and rapidly dissolved.

• Multiple reactions are involved.

• Current efficiencies are in the 30–70 percent range.

• Nitric oxide is formed at the cathode; ammonia/ammonium ions can also be produced.

• Hydrogen is produced over a wide range of acidities and cathodic materials; at catholyte

concentrations below about 2 N nitric acid, measured hydrogen gas concentrations

exceeded 1 percent in the evolved gases. Potential reactions between evolved hydrogen

and nitric oxide were not discussed.

• Hydrogen is a concern because of potential flammability in the gas space and hydrogen

embrittlement in certain metals used in construction (e.g., titanium).

• Polarization curves were not available.

Several safety or operational limits are mentioned in the literature, which would require shutting

off the power if exceeded. These limits include catholyte density (e.g., an indirect measure of

acid concentration), catholyte flow, offgas hydrogen concentration, offgas flow, cell temperature,

anolyte cooling water flow, catholyte cooling water flow, cell applied voltage, and cell current. 

Conductivity and radioactivity monitoring for the cooling water were mentioned as a possible

alarm condition. The staff also notes that cooling water inflow into the catholyte (e.g., from a

leak) would change acid concentrations and potentially increase hydrogen generation rates.

In the revised CAR, an overtemperature event with the electrolyzer is identified in the category of

a loss of confinement. The safety strategy for the facility worker and the environment utilizes

prevention features. The PSSC is the process safety control subsystem. The safety function is

to shut down process equipment before exceeding a temperature safety limit (i.e., the 70 °C

(125 °F) design basis). The applicant stated that the temperature limit will be established by

considering all material limits associated with the glovebox (NRC, 2002c). The applicant will

perform final calculations and identify specific temperature setpoints in support of its ISA, based

on the codes and standards identified in Section 11.6.7 of the revised CAR. This will assure

that, subsequent to the shutdown of process equipment, normal convective cooling is sufficient

(NRC, 2002c).

On page 11.6-16 of the revised CAR, setpoint design bases for the process safety control

subsystem are identified as ANSI/ISA-67.04.01-2000. From clarifications provided during the

August in-office review (NRC, 2002c), the applicant indicated that the setpoint analysis will

consider electrolysis, potential exotherms from reactions, and natural cooling effects. The staff

concludes that this approach provides assurance that the design basis temperature of 70 °C

(125 °F) will not be exceeded, provides margin (i.e., vis-a-vis the boiling point of the solution,

around 110 °C (230 °F)), and provides assurance that the control strategy will adequately

prevent the overtemperature event. The staff has determined this to be acceptable for the CAR

stage. Thus, Open Item AP-01 from the original DSER (NRC, 2002b) is now considered closed.
11-35

11.2.1.3.3 Hydrogen Production (KDB, KDD, KPA, KCA, KWG, and KWD)

The staff review noted that electrolytically generated hydrogen from overvoltage conditions and

off-normal concentrations could produce hydrogen concentrations exceeding the lower

flammability limit (LFL) if the scavenging airflow is based only upon radiolysis. In addition,

overvoltage conditions could produce other undesirable effects, such as different reaction

products and gases, flow oscillations, sparking, and greater heating. 

In the revised CAR, the applicant identified the design basis for radiolytic hydrogen production

as 25 percent of the LFL. The LFL is identified as 4 percent of hydrogen in air under normal

conditions. A prevention strategy has been selected (see revised CAR p. 5.5-33), utilizing the

PSSCs of the instrument air system (supplying scavenging air) and the offgas system (providing

an exhaust path). The staff review notes that hydrogen is principally generated by radiolysis (on

the anode side of the electrolyzer) and electrolysis (on the cathode side of the electrolyzer). The

anode and cathode sides of the electrolyzer may be separated by membranes (i.e., the liquid

phase) but will likely share a common header for venting to the offgas system.

Page 8-17 of the revised CAR discusses hydrogen produced by electrolysis. The design basis is

stated as 25 percent of the LFL. A prevention strategy has been selected (see Section

5.5.2.4.6.13 of the revised CAR, p. 5.5-40), utilizing the process safety control subsystem as the

PSSC. The safety function is to limit the generation of hydrogen from electrolysis by ensuring

that the acid normality is sufficiently high and produces an offgas that does not exceed the

design basis (25 percent of the LFL). Acid normality measurements would be instrumented, and

normality would be maintained by the PSSC of chemical safety controls (NRC, 2003a). The

applicant indicated that voltage control might also be used. 

The applicant identified approaches for calculating LFLs for flammable gas mixtures and

conditions removed from normal temperatures and pressures (NRC, 2003b; DCS, 2003f). The

methodology for mixtures is based on a sum of fractions approach. Graphs were used to show

the impact of temperature and pressure—pressure was not thought to be a concern for the

MFFF because of the lack of high pressures within the buildings. 

The staff notes that the general industrial practice is to use 25 percent of the LFL or less as the

design basis for flammable gases in air (NRC, 1999; WSRC, 2000). A 25-percent limit is

mentioned in Section 11.9.5.1 of the revised CAR for radiolytically generated hydrogen in and

around process-type systems, using scavenging air. During off-normal or accident conditions,

scavenging air is provided by the PSSC of the emergency scavenging air subsystem of the

instrument air system. This emergency scavenging air is provided to those vessels where

radiolytic hydrogen generation could exceed 4-percent hydrogen (the ambient LFL) in 7 days or

less if venting and dilution were not maintained; the design basis for the scavenging air would

still be 1-percent hydrogen (i.e., 25 percent of the ambient LFL).

As part of its review, the staff considered the following facts:

• The MOX standard review plan (SRP), NUREG-1718, mentions a 25-percent LFL limit for

hydrogen in noninerted systems (NRC, 2000).

• The NRC used 25 percent of the LFL for hydrogen in a manner analogous for tank

ullages in a radiochemical plant review (NRC, 1999). Consideration for this limit included
11-36

National Fire Protection Association (NFPA) code evaluation, radiolysis, chemical

reactions, and uncertainties in ventilated process equipment in radiochemical facilities. 

• The NRC inspects fuel cycle facilities based on not exceeding 25 percent of the LFL.

• The Department of Energy (DOE) uses 25 percent of the LFL for waste tanks and

facilities at SRS and Hanford in a manner analogous to design basis.

• The SRS has hydrogen monitors installed in many ventilated tanks; these are set to

alarm at 10 percent of the LFL (WSRC, 2000). 

National Fire Protection Association (NFPA) 69, “Standard on Explosion Prevention Systems,”

issued 1997 (NFPA, 1997) and 801, “Standard for Fire Protection for Facilities Handling

Radioactive Materials,” issued 1998 (NFPA, 1998), emphasize 25 percent of LFL as an

acceptable safety limit. The NFPA allows exceptions if they are justified (e.g., interlocks and

well-defined systems and chemistry, multiple continuous monitoring); such exceptions can allow

concentrations up to 60 percent of the LFL. 

In the revised CAR, the applicant identified the design basis for hydrogen as NFPA 69 (NFPA,

1997). For hydrogen, the applicant identified the controls described below for groups of process

equipment.

Group No. 5. Hydrogen Concentration Associated with Radiolysis

For inside equipment that has the potential for hydrogen concentration associated with

radiolysis, the applicant proposed a design basis value of 25 percent of the LFL. This limit is in

accordance with NFPA 69 (NFPA, 1997).

The safety function of the waste reception unit offgas treatment system is to provide an exhaust

path for the removal of diluted hydrogen gas in process vessels. The instrument air system

(scavenging air) has the safety function to provide sufficient scavenging air to dilute the

hydrogen that is generated during radiolysis.

Group No. 6. Hydrogen Generated by Electrolysis within the Dissolution Process

For inside equipment that has the potential for hydrogen gas concentration associated with

electrolysis within the dissolution unit, the applicant proposes a design basis value of 25 percent

of the LFL. This limit is in accordance with NFPA 69 (NFPA, 1997).

The safety function of the process safety control system (PSCS) is to limit the generation of

hydrogen by ensuring that the normality of nitric acid is sufficiently high to ensure that the offgas

is not flammable.

The applicant’s refined design basis of NFPA 69 for Groups 5 and 6 (maintain the combustible

concentration at or below 25 percent of the LFL) is consistent with the previously cited NRC,

DOE, and industry practice for similar applications, addresses combustible vapors present in

these groups, and should provide adequate margin. The staff also interprets the design basis of

NFPA 69 to apply to normal, off-normal, loss of power, and accident conditions. Consequently,

the staff finds the 25 percent of LFL design basis acceptable for the CAR stage. 
11-37

The control strategy for radiolysis is based on active dilution and active ventilation. This strategy

has the ability to maintain concentrations of hydrogen below the design basis of 25 percent of

the LFL and is acceptable to the staff for the construction authorization.

The strategy for electrolysis is based on acid normality control. The staff notes that acid

normality can be monitored by density or other techniques that may be instrumented, with

concentration adjusted by chemical additions. Additional administrative controls may be

necessary; however, the applicant already has a PSSC entitled chemical safety controls for

chemical makeup that could be used if needed. Specific details would be supplied by the

applicant as part of any later license application. The staff conducted a review of the public

literature and found curves relating hydrogen production to acid normality. Figure 11.2-1 gives

two examples. Hydrogen can be produced over a wide range of acidities and cathodic materials;

at catholyte concentrations above approximately 2 N HNO3, measured hydrogen gas

concentrations are below 1 percent (25 percent of the ambient LFL) in the evolved gases and

thus would also be below 25 percent of the LFL in the ullage spaces above the electrolyzer. 

Consequently, the staff concludes that acid normality has the ability to be monitored and

controlled in order to reduce combustible gases below 25 percent of the LFL, and this strategy is

acceptable for construction authorization. 

Therefore, Open Item AP-02 from the revised DSER (NRC, 2003c) has been adequately

addressed and is now considered closed for the CAR phase.

The staff has reviewed the standards cited as design bases for combustible/hydrogen gas

detectors (see Section 11.6.7 of the revised CAR). The staff notes that the proposed standards

are used extensively in the industry in similar situations for combustible and hydrogen gas

monitoring, as part of control, alarm, and emergency control systems. The staff finds the

standards acceptable for the construction authorization stage.

11.2.1.3.4 Titanium Reactions (KDB and KDD)

The staff evaluation found that the applicant’s proposed AP process uses oxidation-reduction

chemistry based on the silver (I) to silver (II) couple. Silver (II) is corrosive, and special alloys

are necessary for the electrolyzer equipment. The applicant stated that it intends to use titanium

for the electrolyzer circuit and associated equipment that could be exposed to silver (II) ions

(DCS, 2001). The applicant identified a negligible corrosion rate for titanium in the presence of

silver (II) and nitric acid. The applicant intended to destroy silver (II) (i.e., by conversion into

silver (I)) before the solutions contact other equipment in the process that is fabricated out of

300-series stainless steels. Destruction would be accomplished by the addition of peroxide,

which reduces the silver (II) back to silver (I).

The staff finds that a higher alloy material, such as titanium (Bray, 1987, 1992; Bourges, 1985),

is needed for adequate corrosion resistance in the presence of aggressive conditions that are

likely to exist in this electrolyzer. However, titanium is a reactive metal, and the industry has

developed guidelines for the safe use of such alloys, particularly for protection during wet/dry

cycling and heating (DOE, 1994; NFPA, 2000; Mahnken, 2000; Poulson, 2000; PSC, 2001; TI,

2004). In addition, each electrolyzer operates with several hundred amps of current and multiple

tens of volts, which could serve as an ignition source for the titanium in certain accident

sequences. The staff notes that, under certain conditions, titanium fines or thin components

exposed to hot sparks can ignite and burn, and incidents have been reported of unanticipated
11-38

fires in titanium heat exchanger tube bundles, packing, and turbine blades (Mahnken, 2000). 

However, the staff also recognizes that bulk titanium components can burn only under extreme

conditions.

The applicant described its safety approach for titanium events in the electrolyzer in its revised

CAR (DCS, 2002e) and in supporting letters (DCS, 2003d, 2003g, 2004a; NRC, 2003d). The

electrolyzer is a vertical cylinder with separate anode and cathode compartments. The

compartments are concentric with the vertical axis. The cathode compartment is in the center

while the anode compartment forms an annulus around it. The two compartments are separated

from each other by a sintered silicon nitride (Si3N4) barrier (sometimes called a porous frit, or

frit). Elastomeric materials are used to insulate and separate the cathode and anode structures,

the anode and the ground, and the anode and the titanium shell. 

The safety strategy is to prevent the titanium events during normal, abnormal, and shutdown and

maintenance of the electrolyzer. During shutdown and maintenance, the applicant proposed to

use administrative controls associated with the isolation of power to the electrolyzer when the

electrolyzer is drained. For normal and abnormal operations, the applicant proposed a

combination of active and passive engineered features to prevent the titanium events. 

As noted previously, the anode and anodic compartment circuit connect to the positive terminal

of the rectifier (direct current (dc) power) system. At the anode (a platinum electrode), the

electricity oxidizes the silver (I) ions to silver (II) in the anolyte solutions, which subsequently

react with and dissolve the plutonium dioxide powders. The cathode and cathodic compartment

circuit connect to the negative terminal of the rectifier system. At the cathode (a tantalum

electrode), reduction reactions occur, primarily forming nitrous acid from nitric acid.

The silicon nitride barrier between the anode and cathode compartments is nonconductive, has

a high dielectric strength, and has sufficient porosity to allow ionic current transfer, but

insufficient porosity to allow mixing of the anode and cathode solutions. Thus, the silicon nitride

prevents the passage of material between the compartments while facilitating ionic electron

transfer (the ions transfer the electric current, thus completing the circuit). 

Teflon, other fluoropolymers, and polyethylene analogues have been mentioned as potential

elastomeric materials. The applicant stated that insulation material used as guide sleeves

between the anode and the titanium shell will be identified during the ISA process and will be

capable of withstanding the environmental conditions of being submerged in the electrolyzer

fluid. If appropriate, a maintenance/changeout frequency will be established, in accordance with

Section 5.4.3 of the CAR, p. 5.4.10, regarding environmental design considerations and

environmental qualification (EQ) requirements from the ISA.

The applicant identified the silicon nitride barrier and elastomeric materials as PSSCs—these

represent passive engineered controls. In addition, the electrolyzer is geometrically safe from a

criticality perspective, seismically designed (as per Section 11.12.3.3 of the revised CAR, which

addresses B2 elements for seismic structural integrity), and can withstand turbulent flow without

inducing vibration.
11-39

Figure 11.2-1 Acid Normality and Other Effects Upon Electrolytic Gas Generation

The applicant also noted that the nitric acid provides a liquid heat sink for cooling during

operations, and that protection features exist to limit operations. There will be a current leakage

detection system and a trip circuit on the rectifier, which represent the active engineered

controls. The current leakage detection system is designed as a permissive signal and will stop

the startup process (i.e., not allow the power to be turned on during startup) or cease operations

(i.e., turn the power off to the electrolyzer) if the leakage current setpoint of 10 mA is exceeded. 

The trip circuit of the rectifier is another protective feature during operations. The rectifier

normally supplies the electrolyzer with 400 A at 30 V. The trip setpoint is 420 A, and the rectifier

is physically incapable of providing more than 900 A. The current leakage detection system and

the trip circuit are part of the PSCS, an existing PSSC for active engineered controls.
11-40

The staff review finds that the overall safety strategy of prevention is appropriate given the

postulated consequences.

Shutdown and Maintenance

The use of administrative controls that isolate power to electrolyzers during shutdown and

maintenance is well established in the nuclear and chemical industries. In addition, there could

be other requirements added to the procedures in the administrative control as a result of the

ISA, such as hot-work limitations around titanium. However, these would be identified as part of

any later DCS license application and would not change the fundamental strategy of using

administrative controls. Thus, for the CAR stage, the PSSC of administrative control provides

adequate assurances of safety and is acceptable. 

Seismic Event During Operations

The electrolyzer structure is identified as the PSSC, with the safety functions of resisting seismic

events (i.e., seisimically designed), withstanding turbulent flow, not inducing vibrations, and

maintaining geometry for criticality purposes. This represents a passive engineered control. 

The applicant subsequently clarified seismic design in Section 11.12.3.3 of the revised CAR. In

addition, the applicant confirmed that the function of the seismic trip system would include

isolation of power to the electrolyzers. The seismic trip system represents an active engineered

control (part of the PSCS PSSC) that is in addition to the seismic design of the electrolyzers. 

The trip system isolates power to the electrolyzers, thus removing the potential energy and root

cause for a titanium fire event. The staff concludes the electrolyzer structure and seismic trip

system are independent controls and, combined with the low initiating frequency of the seismic

event, have the ability to achieve the very high reliabilities necessary to render the titanium event

highly unlikely. Thus, the staff finds this approach acceptable.

Non-seismic Event During Operations

The applicant identified passive and active engineered controls. The engineered controls

consist of the following:

• The first passive PSSC is the sintered silicon nitride barrier, with the safety function of

physically separating the cathode from the anode in the nitric acid solution by serving as

a dielectric barrier.

• The second passive PSSC is the polytetrafluoroethylene (PTFE, sometimes called

Teflon) insulator, which has the two safety functions of providing insulation/separation

between the cathode and anode structures, and providing insulation/separation between

the anode and the ground.

• The third passive PSSC consists of the guide sleeves, which have the safety function of

insulating/separating the anode from the titanium shell. The actual material would be

identified by the applicant during the ISA process and would be capable of withstanding

the environmental conditions associated with being submerged in the electrolyzer fluid. 

As appropriate, the applicant indicated that a maintenance/changeout frequency will be

established and documented as part of any later license application DCS may submit.
11-41

• The fourth PSSC (active engineered control) is the PSCS, with the safety functions of

detecting current leakage (shut down if the leakage exceeds the design basis of 10 mA)

and tripping the rectifier circuit if the design basis current of 420 A is exceeded.

The staff notes that the applicant has committed to the EQ of electrical and mechanical

components, as outlined in Section 11.11 of this FSER. The performance of elastomeric and

ceramic materials as highly reliable passive engineered controls in the environment of the

electrolyzer will be demonstrated during the license application phase, or the applicant will

identify other PSSCs. Inspection, maintenance, and changeout frequencies would also be

identified by the applicant as part of any later license application DCS may submit.

The active engineered control has direct analogues with electrolyzers in the industry. The staff

review found that control/leakage current and control/maximum current control combinations are

frequently used and can be designed and fabricated to meet the 10 CFR 70.61 performance

requirements. The applicant will provide design information that implements the design basis in

any subsequent license application it may later submit. However, the leakage design basis of 10

mA is consistent with the detection of small currents in the mA range. Instrumentation is

available for detecting smaller currents than this, and current trip instrumentation is available for

almost any amperage rating. The staff also has found that trip currents in the 5–10 mA range

are generally used for trip instrumentation for the protection of industrial workers and the

prevention of sparks and component damage. The staff anticipates that the actual design at the

license application stage will involve consideration of protective device characteristics, such as

let-through currents and time/temperature effects in the titanium resulting from electrical faults

and arcing. Thus, the staff concludes that the applicant’s proposed design bases are consistent

with available ranges of values for the protection of industrial workers, in that sparks and

component damage would be prevented. The design bases are thus found to be acceptable.

There is an interrelationship between the passive and active controls. The first active safety

function detects leakage current, which essentially results from the initiation of failure

mechanisms in the passive controls. Failure of the active leak current detection may not prevent

the complete failure of one or more of the passive controls, which in turn would need the trip

circuit to function. The staff notes that safety control systems can be designed and constructed

such that safety monitoring and safety trip systems can be fully independent. Actual

independence of the controls will have to be demonstrated by the applicant as part of any later

license application DCS may submit.

In conclusion, the staff finds that the passive engineered controls appear capable of rendering

the titanium event unlikely, and the separate active engineered controls also appear capable of

rendering the event at least unlikely. In addition, the safety controls have the ability to be

independent from each other and not subject to common mode failure. Therefore, the

combination of passive and engineered controls proposed by the applicant should be capable of

rendering the event highly unlikely, and this is acceptable for the CAR stage. The staff further

notes that the applicant may develop additional information and submit it as part of its later

license application, demonstrating that an electrical-induced titanium fire is not credible or is

highly unlikely. This closes Open Item AP-03 from the staff’s revised DSER (NRC, 2003c).
11-42

11.2.1.3.5 Loss of Confinement of Process Solutions 

The applicant discussed the control strategy for leaks and loss of confinement of process

solutions in process cells and outside gloveboxes in Sections 5.5.2.1.6.2 and 5.5.2.1.6.11 of the

revised CAR. In the revised CAR, the applicant identified a control strategy for loss of

confinement events (leaks) of AP process vessels and pipes in process cells based on

mitigation. This control strategy uses the process cell to protect the worker, and it uses the

associated ventilation system to protect the site worker, the IOC, and the environment. The

applicant intended to contain fluid leaks within the cell, and any airborne contamination would be

treated with HEPA filtration before exhaust. For facility workers, the PSSC of process cell entry

controls prevents the entry of personnel into process cells during normal operations and ensures

that workers do not receive a dose in excess of limits while performing maintenance. The PSSC

of the process cell has the safety function of containing leaks within the cell. The actual fluid

leaks would not be prevented.

The process cell exhaust (ventilation) system is identified as the PSSC for protecting the site

worker, the IOC, and the environment with the safety functions to provide filtration to limit the

dispersion of radioactive material and to ensure that a negative pressure exists between the

process cell areas and the C2 areas of the facility. 

The staff review identified a potential event involving an acute chemical exposure to the facility,

site workers, and the IOC from hazardous chemicals produced from licensed materials that leak

from AP process vessels during such a loss of confinement event. Such a leak could occur

because of erosion/corrosion of the vessels and piping. As discussed in Section 8.1.2.4.1 of this

FSER, the applicant proposed controls for chlorine and nitrogen tetraoxide. Potential liquid

phase leaks would consist of radioactive nitrate solutions which, once released from the vessels

and pipes, would expose a large liquid surface area that allows a nitric acid and nitric oxide

release into the cell’s atmosphere. This material would not be removed by the HEPA filters on

the exhaust system and would be released to the atmosphere. 

The applicant indicated that the distance from the point of one of these in-cell releases to the

worker is approximately equal to the 100-meter distance to the site worker; therefore

consequence estimates of such releases for the site worker should bound any consequences for

the facility worker (NRC, 2003a). The staff agrees with this qualitative distance analysis. This

closes part of Open Item AP-13 from the original DSER (NRC, 2002b). However, for 379–757 L

(100–200 gal) of radioactive nitrate solutions, IOC limits could be exceeded for several hundred

meters. Some of the solutions might be at temperatures above ambient which could result in

worker limits being exceeded for larger distances. Thus, the performance requirements of 10

CFR 70.61(b)(4) and 10 CFR 70.61(c)(4) might not be met. This potentially impacts the

dissolution, dechlorination and dissolution, oxalic precipitation and oxidation, oxalic mother

liquor, acid recovery, and liquid waste reception units. As discussed in Section 8.1.2.4.1 of this

FSER, the applicant used lower airspeeds with a velocity-dependent evaporation model; some of

the airspeeds are as low as 0.01 m/s (0.033 ft/s) (for comparison, a slight breeze is 1–2 m/s

(3.3–6.6 ft/s)). As discussed in Section 8.1.2.4.1 of this FSER, the applicant has committed to a

low chemical consequence limit and, if necessary, will identify planned features as IROFS to

address the safety concern as part of any later license application it may submit. The staff

concludes that this is an acceptable strategy, and Open Items CS-5b and AP-13 from the

revised DSER (NRC, 2003c) are now closed.
11-43

The applicant identified a prevention strategy for process leaks in a C3 area outside of a

glovebox. The PSSC is double-walled pipe for containing process fluids, and this protects the

facility worker and the environment. The design basis is the fluid transport system (FTS),

identified in Section 11.8.7 of the revised CAR. No PSSCs are anticipated for the site worker

and the IOC because the estimated consequences are below the performance requirements of

10 CFR Part 70, “Domestic Licensing of Special Nuclear Material.” However, any release would

be mitigated by the C3 ventilation system. 

The staff review finds that this approach relies upon practices and codes and standards typically

used in the nuclear and chemical process industries to control similar hazards. By analogy, the

approach should have the ability to address the potential concerns and is acceptable to the staff.

11.2.1.3.6 Oxalic Precipitation Concerns (KCA)

The applicant identified six categories of hazard events associated with the oxalic precipitation

and oxidation unit. The types of events postulated in this unit include fire, explosion, loss of

confinement, external exposure, load handling, and criticality. The safety strategy, including the

PSSCs and design basis safety functions for controlling events within these categories, is

discussed in Chapter 5 of this FSER.

The staff notes that the applicant’s description mentions acidification of the residual mother

liquors to avoid the precipitation and unanticipated accumulation of residual plutonium by the

oxalate. This indicates a potential for a safety function (i.e., avoid plutonium precipitation and

potentially related accident scenarios, such as erosion or plugging that could lead to loss of

confinement). The staff notes that the design does not rely on concentration control to prevent a

nuclear criticality in this unit and that the unit’s calciner uses oxygen. The applicant has

committed to standards for oxygen use and furnace applications (see Section 11.9.4 of the

revised CAR). However, the calciner is likely to include components, such as bearings and

seals, that have requirements to maintain their integrity. These components may be adversely

affected and lose confinement integrity if operated at above-ambient temperatures in the

presence of air or oxygen. The applicant identified nitrogen cooling of the calciner bearings as a

means to protect them, presumably from the oxygen-rich environment, but has not identified this

as a safety function. The issue of whether the nitrogen system is a PSSC because of its bearing

cooling function has been discussed in Section 11.9 of this FSER. 

Nitrogen cooling is identified as a means to extend bearing life; confinement is provided by the

PSSC of the C4 confinement system. 

As discussed in Section 8.1.2.4.1 of this FSER, the applicant proposed a control strategy for

hazardous chemical releases from the potential loss of confinement of radioactive materials in

this unit. Such a loss of confinement would result in an untreated flowpath (bypassing the offgas

treatment system) for chemical releases from the nitric acid solutions, nitrate/oxalate mixture,

and calciner gases. But any release of potentially hazardous chemicals from the calciner area

would likely be bounded by potential releases from tanks in process cells because the source

term in the process cell tanks would be larger. The staff finds that the applicant’s proposed

control strategy is acceptable. 

The applicant identified a design basis for the plutonium dioxide powder produced by oxalic acid

precipitation and oxidation. Without controls, the oxide powder may be substoichiometric or
11-44

have entrained or absorbed solutions subject to radiolysis and thus be pyrophoric and present a

hazard. The applicant proposed controls, which are discussed and evaluated below in Section

11.3.1.2.3.

11.2.1.3.7 Oxalic Mother Liquor Recovery

Prior experience with evaporators indicates the potential for the unintended accumulation of

either solvent or plutonium, or both. The accumulation of material can result from changes in

system chemistry (WVNS, 1999). Such unintended accumulation can pose the hazards of (1)

inadvertent criticality, (2) erosion-corrosion from accumulated solids, and (3) the potential for

red-oil events.

The applicant identified six categories of hazard events associated with the oxalic mother liquor

recovery unit. The types of events postulated in this unit include fire, explosion, loss of

confinement, external exposure, load handling, and criticality. The safety strategy, including the

PSSCs and design basis safety functions for controlling events within these categories, is

discussed in Chapter 5 of this FSER.

The staff notes that the applicant’s description mentions acidification of the residual mother

liquors in order to avoid the precipitation and unanticipated accumulation of residual plutonium

by the oxalate. This indicates a potential safety function (i.e., avoid plutonium precipitation and

potentially related accident scenarios, such as erosion or plugging that could lead to loss of

confinement). The staff notes that the applicant is not relying on concentration control to prevent

a nuclear criticality in this unit. A control strategy, PSSCs, and design bases have been

identified for hazardous chemical releases from the potential loss of confinement of radioactive

materials in this unit. At a minimum, this applies to the following:

• the distillate product stream

• the plutonium-containing stream returned to purification

• the evaporator itself and associated vessels

Section 8.1.2.4.1 of this FSER discusses the potential chemical releases from loss of

confinement events. The staff finds that the control strategy and its related PSSCs and design

bases constitute an acceptable approach.

11.2.1.3.8 Acid Recovery Unit 

The acid recovery evaporators operate on the stream containing americium, uranium, and traces

of plutonium. This is essentially a high alpha contaminated stream, and effective

decontamination between the concentrates (bottoms products) and the distillate (overheads

product) has safety implications. The NRC would anticipate separation requirements and/or

specifications for these evaporators and their products. This is related to 10 CFR Part 20,

“Standards for Protection Against Radiation,” and will be reviewed as part of the review of any

later license application DCS may submit.

11.2.1.3.9 Offgas Treatment Unit

The applicant identified the offgas treatment unit as a PSSC with the safety function of providing

an exhaust path for the removal of gases in process vessels (Section 5.6 of the revised CAR). 
11-45

Mechanical design bases are identified in Section 11.4.11.1.11 of the revised CAR. These

include the following:

• two stages of HEPA filters

• spark arrestors and prefilters in each final filtration assembly

• each HEPA stage field tested to have an efficiency of 99.95 percent

• fire-rated dampers between designated fire areas

• in-place HEPA filter testing capability in accordance with ASME N510, “Nuclear Power

Plant Air Cleaning Units and Components,” issued 1980 (ASME, 1980), for the final

filtration assemblies

• final filters and ductwork remain structurally intact during and after design basis

earthquakes and withstand the effects of tornados

The staff requested information on the filtering line. The applicant provided additional

information and identified the following design features of the offgas unit (DCS, 2001; RAI 142;

Sections 11.3.2.11 and 11.3.2.13 of the revised CAR):

• Bubbling air scavenges tank ullage to maintain hydrogen concentrations at 1 percent or

less.

• The system operates below the flash point of solvent vapors.

• Supplemental air is added to the system in order to further dilute any potential

combustible concentration of gases and to maintain minimum volumetric throughput for

the scrubbing and washing columns.

• The material of construction is stainless steel to resist the corrosive atmosphere.

• The HEPA filters are constructed of acid-resistant materials.

The offgas system handles vapors and gas mixtures that are potentially combustible in air

streams, such as hydrogen, hydrazine, and TBP/dodecane (the solvent/diluent

mixture—dodecane will likely be a petroleum distillation fraction and its constituents are more

volatile than TBP). The applicant identified a setpoint/design basis/safety limit of 25 percent of

the LFL in air for hydrogen (DCS, 2001; DCS, 2002f, RAI response 142). In RAI response 122

(DCS, 2001, 2002e), the applicant also identified a limit of 25 percent of the LFL for hydrogen in

air from radiolysis in vessels containing plutonium. The staff notes that a limit of 25 percent of

the LFL in air is routinely used by designers and operating facilities and is embodied in codes

and standards. The applicant subsequently identified a design basis limit for hydrogen and other

flammable gases and vapors, based on NFPA 69. This is discussed in Section 11.2.1.3.3 of this

FSER. This should maintain potentially flammable gases and vapors at safe concentrations

below their LFLs at all times, with adequate margin, and thus satisfactorily addresses the staff

concerns. Consequently, Open Item AP-08 from the revised DSER is now considered closed

(NRC, 2003c).
11-46

In response to RAI 142 (DCS, 2002f), the applicant did not identify temperatures below the

flashpoint of solvent vapors. However, the applicant identified a safety strategy using a

commitment to NFPA 69. This is discussed and found to be acceptable in Section 8.1.2.5 of this

FSER. Consequently, Open Item AP-09 from the revised DSER is now considered closed

(NRC, 2003c).

The process handles gases and vapors that are potentially reactive and toxic, such as nitrogen

tetraoxide, nitric acid, nitric oxide, and hydrazine. The unplanned evolution of these

gases—from and through radioactive solutions—via the offgas treatment unit could have

potentially detrimental consequences that would likely exceed the performance requirements of

10 CFR Part 70 at considerable distances from the proposed facility. In addition, the applicant

identified the removal of hazardous chemicals as a function of the offgas unit for protection of

the HEPA filters (DCS, 2003a; NRC, 2003b). As discussed in Section 8.1.2.3.1 of this FSER,

the applicant has agreed not to exceed low chemical consequence levels (see Tables 8-5 and 8-

7). In addition, the offgas system includes components that have the ability to remove

potentially toxic and reactive gases. In order to maintain the confinement barrier, the applicant

implemented administrative controls to replace HEPA filters, such as those in the offgas system,

after identified exposures to water or chemicals (DCS, 2003e). Therefore, the staff concludes

that the applicant has provided strategies and design bases that have the ability to address the

safety concerns regarding the offgas system, and this is acceptable for the CAR stage. 

Consequently, Open Item AP-10 from the revised DSER has been adequately addressed and is

now considered closed (NRC, 2003c).

11.2.1.3.10 Corrosion Control (KDB and KDD)

Lower alloys can be inadvertently exposed to aggressive conditions. For example, stainless

steel would likely experience uneven pitting corrosion that could lead to premature leaks and

failures if it is routinely exposed to low concentrations of silver (II) ions. In a letter to the NRC,

the applicant mentioned the use of stainless steel to resist corrosion in the offgas system and

the use of acid resistant materials in the HEPA filters (DCS, 2001). The applicant proposed a

generic corrosion control program as a PSSC. This appears to be based on general corrosion. 

The pitting corrosion that could occur from silver (II) ions might not be detected before failure by

the proposed PSSC of a general corrosion control program; therefore, the potential exists for the

corrosion leak to release plutonium compounds (i.e., a loss of confinement). 

Corrosion-resistant materials would be needed to maintain confinement of radioactive and

chemical species. The revised CAR identifies two administrative PSSCs that apply to corrosion. 

The first is chemical safety control. One of its safety functions is to ensure control of the

chemical makeup of the reagents and to ensure segregation/separation of vessels/components

from incompatible chemicals (i.e., planned corrosion exposures). The second PSSC involves

the Material Maintenance and Surveillance Programs (Sections 5.5.2.1.6.2 and 5.6.2.4 of the

revised CAR). The safety function of this PSSC is to detect and limit the damage resulting from

corrosion and thus mitigate the effects to the facility worker and the environment. As discussed

in Section 5.6.2.4 of the revised CAR, this PSSC can identify corrosion problems within the

facility before the occurrence of catastrophic failures. 

The administrative PSSCs are not required to prevent corrosion that could result in small leaks. 

Small is not defined but inferred to be a small percentage of inventory. However, the staff

concludes that the design basis function of the corrosion function of the FTS PSSC should
11-47

address instrumentation and/or monitoring of components that could be exposed to aggressive

species in the offgas and other units and thus address small leaks. The FTS design bases are

discussed in more detail in Section 11.8 of this FSER. Specific details would be provided in any

later license application DCS may submit.

The staff concludes that this information on corrosion controls provided in the revised CAR

follows established codes and practices in similar industries, has the ability to address the

potential safety concerns, and, consequently, adequately addresses the staff’s concerns about

corrosion monitoring for preventing major failures. The staff finds this to be acceptable for the

CAR stage. Therefore, Open Items AP-04 and AP-11 from the revised DSER have been

adequately addressed and are now considered closed (NRC, 2003c).

11.2.1.3.11 Liquid Waste (KWD)

In Section 10.1.4 of the revised CAR, the applicant discussed waste minimization and waste

management. Liquid and solid wastes produced at the proposed facility will be transferred to

SRS for processing and disposal. The applicant has worked closely with SRS during the facility

design phase and provided SRS with waste characterization information. The SRS has

reviewed and evaluated the information in the context of the existing WAC. The applicant is

committed to meeting the SRS WAC or providing a stream that qualifies for a WAC deviation

and exemption. The WAC for the SRS waste solidification building has not been issued, but the

applicant stated that the interface between it and SRS will ensure that the waste solidification

building is designed to manage the facility high alpha waste stream and the depleted uranium

stream. This was part of Open Item AP-05 in the original DSER. This portion of Open Item AP05 (now designated as AP-05a) is considered closed.

The applicant stated that the alkaline waste stream will be acidified in a separate neutralization

tank before being mixed in the high alpha waste tanks. Neutralization and acidification are

performed to eliminate the potential for an explosion from azide formation that may form under

alkaline conditions. In acidic media, the azides have a solubility limit greater than their

concentration. Because the solubility limits of azides in alkaline media are lower, the alkaline

media is acidified in order to increase the solubility limits. This ensures that the azides do not

precipitate and create an explosion potential. Chapter 8 of the revised CAR and supplemental

information provided by the applicant identify pH control as serving a safety function. Section

8.1.2.5 of this FSER provides further discussion of azide formation and an evaluation of the

controls proposed by the applicant. 

The applicant identified the high alpha activity and stripped uranium waste transfer lines as

PSSCs (CAR Sections 5.5.2.3.6.5 and 10.5.2). These are double-walled stainless steel pipes

that are seismically qualified and designed for leak detection and collection. The lines will be

designed to accommodate mechanical and seismic loads. For load-handling events, the safety

strategy relies on prevention. The PSSCs are the waste transfer lines. The safety function is to

protect the lines from activities taking place outside the MFFF building. For external events

(e.g., external fires, explosions, extreme winds, tornadoes, missiles, rain, and snow/ice

loadings), the safety function is to prevent damage to the line. The design basis for both

functions is ASME B31.3, “Process Piping,” issued 1998 (ASME, 1998). This standard requires

consideration of loads in the design of piping. The staff analysis notes that the code, the

proposed approach with the waste transfer lines (i.e., double walled with leak detection), and the

prevention strategy provide reasonable assurance that the design will not be damaged and
11-48

release radionuclides outside of the MFFF building. The staff finds this to be acceptable. This

item, identified as Open Item AP-06, is now closed.

The staff notes that an explicit inventory limit on waste is not specified in the revised CAR.

Currently, the facility is designed to accommodate up to 90 days equivalent of most waste

solutions (e.g., of the values in Table 11.2-1, because the storage of the LLW destined for the

waste solidification building will likely be less than 90 days equivalent), although the applicant

anticipated that there will be transfers of liquid wastes every 2 weeks. The applicant indicated

that the facility will shut down before exceeding the liquid waste storage capacity. The staff

interprets this to mean active waste generating operations would be curtailed at some setpoint

before the tankage is completely full, until the potential backlog of waste at the facility is cleared. 

Actual setpoints would be defined by DCS as part of any license application it may later submit. 

The staff finds this approach acceptable for construction authorization.

[Text removed under 10 CFR 2.390]

 Both are bounded by the

safety assessment; therefore the staff finds this to be acceptable for the construction

authorization. This item, identified as Open Item AP-05a, is now closed.

11.2.1.3.12 Sampling Systems

Table 11.3-34 in the revised CAR categorizes the sampling systems. Section 11.11 of the

revised CAR provides information on the laboratory and indicates that a significant number of

samples are required. The staff notes that laboratory personnel will most likely conduct the

sampling. The applicant provided information on a PSSC entitled laboratory material controls

that limits quantities of hazardous and radioactive materials in the laboratory. Section 5.6.2.7 of

the revised CAR also states that procedures will be developed to establish limits on sample size

and to ensure that laboratory operations are performed in accordance with safe laboratory

operating practices. The staff finds that the outline of the sampling approaches appears to

follow typical practices used in the chemical and nuclear industries. 

In addition, DCS indicated that all sampling, with the exception of samples with very low levels of

radioactivity, will be conducted within gloveboxes, and that there are no bag-in/bag-out

operations because the sample containers would be pneumatically transferred from stations

within the gloveboxes to the laboratory. The revised CAR identifies gloveboxes as PSSCs, with

the safety function of maintaining confinement integrity for design basis impacts (Section

11.4.11.2). Laboratory material controls address the handling of the samples within the

laboratory. Only the LLW samples would not be handled within gloveboxes; qualitatively, the

staff concludes this does not present a challenge to the performance requirements of 10 CFR

70.61, “Performance Requirements.” The staff finds that the applicant has identified PSSC and
11-49

design basis information and finds that this issue has been adequately addressed for the

construction authorization stage. This closes Open Item AP-12.

11.2.1.4 Design Basis of the PSSCs and Applicable Baseline Design Criteria

The plutonium isotopic composition design basis for AFS and PDCF/ARIES feeds is identified by

the applicant as follows (see Section 11.3.7 of the revised CAR):

•

236Pu < 1 ppb, at the origin of pit

•

238Pu < 0.05%

• 90% < 239Pu < 95%

• 5% < 240Pu < 9%

•

241Pu < 1% during lifetime of plant

•

242Pu < 0.1%

The feed chemical impurities design bases were identified by the applicant and are listed in

Tables 11.2-2 and 11.2-3 of this FSER. The radionuclide impurities design bases are listed in

Table 11.2-4.

The americium content design basis for PDCF/ARIES feed material is identified by the applicant

as follows:

 

241Am

 ------------------------- < 0.7% during the lifetime of the plant

 Pu total + 241Am

The feed plutonium dioxide powder to the AP process in the facility has a maximum density of

less than 7 g/cc. Plutonium dioxide powder entering the facility can have a density up to 11.46

g/cc. Therefore, after receipt and storage, a density measurement is made. If necessary, a

milling step is performed to ensure the density is below 7 g/cc.

The applicant stated that feed materials which have an impurity content that exceeds a value

listed in Tables 11.2-2, 11.2-3, and 11.2-4 can, in some cases, be processed at the proposed

facility. The applicant will evaluate these batches for safety before accepting them for

processing at the facility. 

The staff noted that these parameters and the values listed in Tables 11.2-2, 11.2-3, and 11.2-4

for the plutonium feed to the facility may affect the design and the safe operation of the facility. 

The applicant identified these as design bases. The PSSC is chemical safety controls, and the

safety function is to “ensure control of the chemical makeup of the reagents and ensure

segregation/separation of vessels/components from incompatible chemicals.” According to the

staff review, this is an administrative control based on the analysis of the feed to the facility. The

staff also concludes that exceptional batches of plutonium dioxide that have impurity contents

exceeding the design bases will be evaluated according to the facility change process in 10 CFR

70.72. The staff concludes that this approach provides assurance that the design basis

plutonium feed parameters will not be exceeded and finds this to be acceptable for the

construction authorization stage. Consequently, this part of Open Item AP-07 from the revised

DSER has been adequately addressed and is now considered closed (NRC, 2003c). Section

11.2.1.3.1 of this FSER discusses closure of the other part of AP-07.
11-50

The applicant stated that the BDC specified in 10 CFR 70.64(a) are incorporated into the design

and operation of the facility (Section 5.5.5.4 of the revised CAR). The applicant stated that

information demonstrating compliance with these criteria is provided in the applicable chapters

of the revised CAR. For chemical protection, 10 CFR 70.64(a)(5) states the following:

Chemical protection. The design must provide for adequate protection against

chemical risks produced from licensed material, facility conditions which affect the

safety of licensed material, and hazardous chemicals produced from licensed

material.

The applicant stated that reagents would be stored, and chemical mixtures would be prepared,

in the reagent processing building and the reagent storage area of the AP area, and they are

generally separated from each other and from radioactive materials. The applicant will avoid

mixing incompatible materials by design, controls, and procedures. The AP and MP facilities are

broken down into process functional units, which are made up of one or more subassemblies

performing consistent and elementary tasks. The applicant stated that the breakdown into

control functional units allows each entity to be operated relatively independently in the given

operating mode (DCS, 2002e). The staff review notes that this separation and independence

are consistent with accepted practices for safe operations.

The applicant will control process storage and operation conditions in order to prevent

exothermic and potential autocatalytic reactions in the reagent processing building, as well as in

the AP and MP areas. Autocatalytic and exothermic reactions of chemicals would be prevented

through control of the process parameters (e.g., reactant concentration, temperature, catalyst

concentration in solution, and pressure) that affect the reactions. Many of these controls have

been identified as PSSCs with design bases. Codes and standards have been applied on a

specific level, often as design bases (DCS, 2002e).

The applicant stated that there is reasonable assurance that the PSSCs will be sufficiently

reliable and available based on the use of standard nuclear industry engineering practices. 

These practices are incorporated into the facility general design philosophy, design bases,

system design, and commitments to applicable management measures. These practices ensure

that applicable industry codes and standards are utilized, adequate safety margins are provided,

engineering features are utilized to the extent practicable, the defense-in-depth philosophy is

incorporated into the design, and PSSCs will be appropriately maintained (DCS, 2002e).

The staff review finds that the applicant has provided sufficient information to meet the

requirements of 10 CFR 70.64(a)(5). 

Related to chemical protection, the explosion protection BDC is stated as part of the fire

protection BDC in 10 CFR 70.64(a)(3):

Fire protection. The design must provide for adequate protection against fires

and explosions.

The applicant stated that there is reasonable assurance that the PSSCs will be sufficiently

reliable and available based on the use of standard nuclear industry engineering practices. 

These practices are incorporated into the facility general design philosophy, design bases,

system design, and commitments to applicable management measures. These practices ensure
11-51

that applicable industry codes and standards are utilized, adequate safety margins are provided,

engineering features are utilized to the extent practicable, the defense-in-depth philosophy is

incorporated into the design, and PSSCs will be appropriately maintained.

Table 11.2-2 Design Basis for Chemical Impurities in PDCF Feed Plutonium Dioxide

Chemical 

Component

Maximum

Content

(:g/g Pu)

Maximum

Exceptional

Content

(:g/g Pu)

Chemical 

Component

Maximum

Content

(:g/g Pu)

Maximum

Exceptional

Content

(:g/g Pu)

Ag NA 10,000 Mg 500 10,000

Al 150 10,000 Mn 100 1,000

B 100 1,000 Mo 100 1,000

Be 100 2,500 N 400 400

Bi 100 1,000 Na 300 10,000

C 500 1,500 Nb 100 3,500

Ca 500 10,000 Ni 200 2,500

Cd 10 1,000 P 200 1,000

Cl (+F < 250) 500 Pb 200 1,000

Co 100 10,000 S 250 1,000

Cr 100 500 Si 200 200

Cu 100 500 Sm 2 1,000

Dy 1 1,000 Sn 100 2,500

Eu 1 1,000 Ti 100 2,500

F (+Cl < 250) 350 Th 100 100

Fe 500 2,500 V 300 2,500

Ga 12,000 12,500 W 200 2,500

Gd 3 250 Zn 100 1,000

In 20 1,000 Zr 50 1,000

K 150 10,000 Boron 

Equivalent

NA

Li 400 10,000 Total

Impurities

18,837

NA = Not applicable or not available

Maximum exceptional value means the maximum anticipated value for that element, with all others

at the maximum value.

The applicant stated that the BDC specified in 10 CFR 70.64(a) are incorporated into the design

and operation of the facility (Section 5.5.5.4 of the revised CAR). The applicant further stated

that information demonstrating compliance with these criteria is provided in the applicable

chapters of the revised CAR. For chemical protection, 10 CFR 70.64(a)(5) states the following:

Chemical protection. The design must provide for adequate protection against

chemical risks produced from licensed material, facility conditions which affect the 
11-52

Table 11.2-3 Design Basis for Chemical Impurities of

AFS Plutonium Dioxide Feed Material

Chemical 

Component

Maximum

Content for

most (~75%)

of Items

(:g/g Pu)

Maximum

Content

exceeded only

by 2% of Items

(:g/g Pu)

Chemical 

Component

Maximum

Content for

most (~75%)

of Items

(:g/g Pu)

Maximum

Content

exceeded only

by 2% of Items

(:g/g Pu)

Ag NA 10,000 I NA 100

Al 4,000 15,000 In 20 2,500

Am 7,000

(100% 241Am)

7,000

(100% 241Am)

K 220,000 (Ca+K+Mg+Na)

#40% Net weight

As NA 100 La NA 5,000

Au NA 100 Li 5,000 10,000

B 100 1,000 Mg 70,000 (Ca+K+Mg+Na)

#40% Net weight

Ba 5,000 10,000 Mn 1,000 2,000

Be 100 5,000 Mo 100 (Mo+Zr)<5,000

Bi 1,000 1,000 N 400 5,000

C 2,000 10,000 NO3 NA 5,000

Ca 120,000 (Ca+K+Mg+Na)

#40% Net weight

Na 130,000 (Ca+K+Mg+Na)

#40% Net weight

Cd 1,000 1,000 Nb 100 3,500

Ce NA 500 Ni 5,000 15,000

Cl 200,000 330,000 Np 500 1,000

Co 5,000 10,000 P 1,000 (P+S)#10,000

Cr 3,000 8,000 Pb 200 5,000

Cu 500 3,000 Pd NA 100

Dy NA NA Pt NA 100

Er NA 500 Rb 100 5,000

Eu NA NA S 330 (P+S)#10,000

F 1,000 7,000 SO4 1,000 (P+S)#10,000

Fe 5,000 18,000 Sb NA 100

Ga 12,000 15,000 Si 5,000 10,000

Gd 250 250 Sm NA NA

Ge NA 100 Sn 1,000 10,000

Hf 50 1,000 Sr 5,000 10,000

Hg NA 100 Ta 4,000 10,000
11-53

Table 11.2-3 (cont): Design Basis for Chemical Impurities of

AFS Plutonium Dioxide Feed Material

Chemical 

Component

Maximum

Content for

most (~75%)

of Items

(:g/g Pu)

Maximum

Content

exceeded only

by 2% of Items

(:g/g Pu)

Chemical 

Component

Maximum

Content for

most (~75%)

of Items

(:g/g Pu)

Maximum

Content

exceeded only

by 2% of Items

(:g/g Pu)

Ti 100 3,000 V 300 1,000

Th 100 300 W 4,000 10,000

Tl NA 100 Y 200 10,000

Enriched

Uranium

(EU)

EU#30% Net

weight

EU#30% Net

weight

Annual max.

value—50 kg

(110 lbs)

(

235U—93.2%)

Zn 1,000 10,000

Depleted

Uranium,

Natural

Uranium

[TBD] 500,000 Zr 50 (Mo+Zr)#5,000

safety of licensed material, and hazardous chemicals produced from licensed material.

The applicant stated that reagents would be stored, and chemical mixtures would be prepared,

in the reagent processing building and the reagent storage area of the AP area, and they are

generally separated from each other and from radioactive materials (DCS, 2002e). The

applicant intends to avoid mixing incompatible materials by design, controls, and procedures. 

The AP and MP facilities are broken down into process functional units, which are made up of

one or more subassemblies performing consistent and elementary tasks. The applicant stated

that the breakdown into control functional units allows each entity to be operated relatively

independently in the given operating mode (DCS, 2002e). The staff review notes that this

separation and independence are consistent with accepted practices for safe operations.

The applicant will control process storage and operation conditions in order to prevent

exothermic and potential autocatalytic reactions in the reagent processing building and in the AP

and MP areas. Autocatalytic and exothermic reactions of chemicals would be prevented through

control of the process parameters (e.g., reactant concentration, temperature, catalyst

concentration in solution, and pressure) that affect the reactions. Many of these controls have

been identified as PSSCs with design bases (DCS, 2002e). Codes and standards have been

applied on a specific level, often as design bases.

The applicant stated that there is reasonable assurance that the PSSCs will be sufficiently

reliable and available based on the use of standard nuclear industry engineering practices. 

These practices are incorporated into the facility general design philosophy, design bases,

system design, and commitments to applicable management measures. These practices ensure

that applicable industry codes and standards are utilized, adequate safety margins are provided,
11-54

engineering features are utilized to the extent practicable, the defense-in-depth philosophy is

incorporated into the design, and PSSCs will be appropriately maintained (DCS, 2002e).

Table 11.2-4 Radionuclide Impurities in the Feed Plutonium Dioxide

PDCF Type

Impurity Isotope Maximum Content

:g/g Pu

Americium 241Am—100% 7,000 :g/g Pu (Note 1)

Uranium (HEU) 235U—93.2% Standard value—5,000 :g/g Pu

Maximum value—20,000 :g/g Pu

for 10% of the delivered cans

during 1 year

Annual maximum value = 17 kg

(37 lbs) (Note 2)

(1) At the plutonium design basis feed rate of 3.5 MTHM/yr, the americium

annual quantity becomes 24.5 kg/yr (54 lb/yr).

(2) The uranium standard maximum value corresponds to 17.5 kg/yr (39

lb/yr), while 10% at 20,000 and 90% at 5,000 (micrograms U/g Pu)

correspond to 22.75 kg/yr (50 lb/yr).

AFS Type

Impurity Isotope Maximum Content

:g/g Pu

Americium 241Am—100% 11,000 :g/g Pu

Enriched

Uranium

235U—93.2% Maximum value—30% of can net

weight

Annual maximum value = 50 kg

(110 lb)

Depleted

Uranium and

Natural Uranium

238U Maximum value—42% of can net

weight (with 5% of enriched

uranium)

The staff review finds that the applicant has provided sufficient information to meet the

requirements of 10 CFR 70.64(a)(5). Using the SRP, the staff review notes that the previously

identified open items in AP have been closed by additional safety strategies and information

from the applicant, and the staff concludes that the applicant has satisfied this BDC.

Related to chemical protection, the explosion protection BDC is stated as part of the fire

protection BDC in 10 CFR 70.64(a)(3):
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Fire protection. The design must provide for adequate protection against fires

and explosions.

The applicant states that there is reasonable assurance that the PSSCs will be sufficiently

reliable and available based on the use of standard nuclear industry engineering practices. 

These practices are incorporated into the facility general design philosophy, design bases,

system design, and commitments to applicable management measures. These practices ensure

that applicable industry codes and standards are utilized, adequate safety margins are provided,

engineering features are utilized to the extent practicable, the defense-in-depth philosophy is

incorporated into the design, and PSSCs will be appropriately maintained (DCS, 2002e).

The staff review finds the applicant has now provided sufficient information to meet the

requirements of 10 CFR 70.64(a)(3). Using the SRP, the staff review notes that the previously

identified open items related to MP have been closed by additional safety strategies and

information from the applicant, and the staff concludes that the applicant has now satisfied this

BDC.

11.2.2 Evaluation Findings

In Section 11.3 and Chapters 5, 8, and 11 of the revised CAR, DCS provided design basis

information for chemical process safety PSSCs identified for the proposed facility. Based on the

staff’s review of these revised CAR chapters and on supporting information provided by the

applicant that is relevant to AP and chemical process safety, the staff finds that, for the reasons

discussed above, DCS has now met the BDC set forth in 10 CFR 70.64(a)(3) for explosions, and

10 CFR 70.64(a)(5) for chemical safety. Further, the staff concludes, pursuant to

10 CFR 70.23(b), that the design bases of the PSSCs identified by the applicant will provide

reasonable assurance of protection against natural phenomena and the consequences of

potential accidents.

Several previously identified open items in the April 30, 2002, and the April 30, 2003, DSERs

have now been closed in a satisfactory manner—AP-02, AP-03, AP-07, AP-08, AP-09, and AP10.

11.2.3 References

(ASME, 1980) American Society of Mechanical Engineers. ANSI/ASME N50, “Testing of

Nuclear Air-Cleaning Systems.” New York, 1980.

(ASME, 1998) American Society of Mechanical Engineers. ASME B31.1, “Process Piping.” 

New York, 1998.

(Bennedict, 1981) Bennedict, Pigfood, and Levi. Nuclear Chemical Engineering. McGraw-Hill,

New York, 1981.

(Bourges, 1985) Bourges, J., C. Madic, G. Koehly, and M. Lecomte. “Dissolution of Plutonium

Dioxide in Nitric Acid Medium by Electrogenerated Silver (II),” Y/TR-91/4, translated from the

French Journal of the Less Common Metals, 122:303–311. 1986. Paper originally presented at

Actinides 85, Aix-en-Provence, France, September 2–6, 1985 (CONF-8509147-28-Trans).
11-56

(Bray, 1987) Bray, L.A., J.L. Ryan, and E.J. Wheelwright. “Electrochemical Process for

Dissolving Plutonium Dioxide and Leaching Plutonium from Scrap or Wastes.” AIChE

Symposium Series 254, Volume 83, Electrochemical Engineering Applications. R.E. White, R.F.

Savinell, and A. Schneider eds. (CONF-861146-10), 1987.

(Bray, 1992) Bray, L.A., J.L. Ryan, E.J. Wheelwright, and G.H. Bryan. “Catalyzed Electrolytic

Plutonium Oxide Dissolution: The Past 17 Years and Future Potential.” Chapter 30 in

Transuranium Elements: A Half Century. L.R. Morss and J. Fuger, eds. American Chemical

Society, Washington DC, 1992.

(DCS, 2001) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Response to Request for Additional Information. August 31, 2001.

(DCS, 2002a) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Clarification of Responses to NRC Request for Additional Information. 

January 7, 2002.

(DCS, 2002b) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Clarification of Responses to NRC Request for Additional Information. March

8, 2002.

(DCS, 2002c) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Clarification of Responses to NRC Request for Additional Information. April

23, 2002.

(DCS, 2002d) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Evaluation of the Draft Safety Evaluation Report (DSER) on the Construction

of a Mixed Oxide Fuel Fabrication Facility. July 9, 2002.

(DCS, 2002e) Ashe, K., Duke Cogema Stone & Webster. Letter to Docket Number 070-03098,

U.S. Nuclear Regulatory Commission, RE: Mixed Oxide Fuel Fabrication Facility—Construction

Authorization Request. October 31, 2002 (page changes through January 27, 2005).

(DCS, 2002f) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Requests for Additional Information, Clarifications, and Open Item Mapping

into the Construction Authorization Request Revision. November 22, 2002.

(DCS, 2003a) Hastings, P., Duke Cogema Stone & Webster, Letter to U.S. Nuclear Regulatory

Commission, RE: Responses to Financial Qualification, Fire Safety, Chemical Safety, Aqueous

Processing, Material Processing and Ventilation Open Items/Additional NRC Questions on

Construction Authorization Request (CAR) Revision. February 18, 2003.

(DCS, 2003b) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: NRC to DCS Letter dated 13 February 2003, February 2003 Monthly Open

Item Status Report. February 18, 2003.

(DCS, 2003c) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Construction Authorization Request Change Pages. April 10, 2003.
11-57

(DCS, 2003d) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, “Response to DSER Open Items AP-3 and CS-3.” May 23, 2003.

(DCS, 2003e) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, “Response to DSER Open Items AP-10 and CS-2.” May 30, 2003.

(DCS, 2003f) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, “Response to Request for Additional Information—Chemical Safety Open Items

CS-09, AP-02, AP-08, and AP-09.” September 29, 2003.

(DCS, 2003g) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, “Response to Request for Additional Information—DSER Open Items MP-01 (UO2)

and AP-03 (Titanium Fires).” October 10, 2003.

(DCS, 2004a) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, ”Response to Request for Additional Information—DSER Open Item AP-03

(Titanium Fires).” March 12, 2004.

(DOE, 1994) U.S. Department of Energy. “Primer on Spontaneous Heating and Pyrophoricity,”

DOE-HDBK-1081-94. December 1994.

(Mahnken, 2000) Mahnken, G. “Watch Out for Titanium Tube Bundle Fires,” Chemical

Engineering Progress, Vol. 96, No. 4, pp 47–52. April 2000.

(NFPA, 1997) National Fire Protection Association. Standard 69, “Standard on Explosion

Prevention Systems.” 1997.

(NFPA, 1998) National Fire Protection Association. NFPA Standard 801, “Standard for Fire

Protection for Facilities Handling Radioactive Materials,” 1998.

(NFPA, 2000) National Fire Protection Association. NFPA Standard 481, “Standard for the

Production, Handling, and Storage of Titanium.” 2000.

(NFPA, 2001) National Fire Protection Association. Standard 2001, “Standard on Clean Agent

Extinguishing Systems.”

(NRC, 1999) Pierson, R.C., U.S. Nuclear Regulatory Commission. Memorandum to D. Clark

Gibbs, U.S. Department of Energy, RE: Estimates of Hydrogen Generation From Wastes at the

Proposed, TWRS-P Facility. April 21, 1999.

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility.” Washington,

DC, August 2000.

(NRC, 2001) Persinko, A., U.S. Nuclear Regulatory Commission. Memorandum to E.J. Leeds,

U.S. Nuclear Regulatory Commission, RE: 11/27-29/01 In-Office Review Summary of DCS

Construction Authorization Request Supporting Documents for the MFFF. December 18, 2001.
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(NRC, 2002a) Persinko, A., U.S. Nuclear Regulatory Commission. Memorandum to E.J. Leeds,

U.S. Nuclear Regulatory Commission, RE: 2/13/02 Meeting Summary: MFFF Program

Changes and Applicant Reorganization. February 27, 2002.

(NRC, 2002b) Pierson, R. U.S. Nuclear Regulatory Commission. Letter to R. Idhe, Duke

Cogema Stone & Webster, “Draft Safety Evaluation Report on Construction of Proposed Mixed

Oxide Fuel Fabrication Facility.” April 30, 2002.

(NRC, 2002c) Persinko, A., U.S. Nuclear Regulatory Commission. Memorandum to M.N.

Leach, U.S. Nuclear Regulatory Commission, RE: August 28–30, 2002 In-Office Review

Summary of DCS Construction Authorization Request Supporting Documents for the MFFF. 

December 31, 2002. 

(NRC, 2003a) Persinko, A., U.S. Nuclear Regulatory Commission. Memorandum to M.N.

Leach, U.S. Nuclear Regulatory Commission, RE: December 10–12 Meeting Summary: Meeting

with DCS to Discuss Mixed Oxide Fuel Fabrication Facility Revised Construction Authorization

Report. January 31, 2003.

(NRC, 2003b) Persinko, A., U.S. Nuclear Regulatory Commission. Memorandum to M.N.

Leach, U.S. Nuclear Regulatory Commission, RE: January 15–16 Meeting Summary: Meeting

with DCS to Discuss Mixed Oxide Fuel Fabrication Facility Revised Construction Authorization

Report. March 5, 2003.

(NRC, 2003c) Pierson, R., U.S. Nuclear Regulatory Commission. Letter to R. Idhe, Duke

Cogema Stone & Webster, “Draft Safety Evaluation Report on Construction of Proposed Mixed

Oxide Fuel Fabrication Facility, Revision 1.” April 30, 2003.

(NRC, 2003d) Brown, D., U.S. Nuclear Regulatory Commission. Memorandum to J. Giitter,

U.S. Nuclear Regulatory Commission, “October 16 & 21, 2003 Summary of Phone Call with the

Applicant: Chemical Safety Open Items for the Mixed Oxide (MOX) Fuel Fabrication Facility.” 

October 22, 2003.

(Poulson, 2000) Poulson, E. “Safety-Related Problems in the Titanium Industry in the Last 50

Years,” Journal of Metals, Vol. 52, No. 5, pp 13–17. 2000.

(PSC, 2001) O’Connor, M.K., Process Safety Center. “A Fire in Titanium Structured Packing

Involving Thermite Reactions.” August 3, 2001. http://www.process-safety.tamu.edu/safetyalert/08_03_01.htm.

(TI, 2004) Titanium Industries, Inc. “Fire Prevention.” November 16, 2004. 

http://www.titanium.com/titanium/tech_manual/tech15.cfm.

(WSRC, 2000) Hobbs, D.T. “Possible Explosive Compounds in the Savannah River Site Waste

Tank Farm Facilities,” WSRC-TR-91-444, Revision 3. February 15, 2000.

(WVNS, 1999) Baker, M.N. and H.M. Houston, West Valley Nuclear Services Company. “Liquid

Waste Treatment System: Final Report,” DOE/NE/44139-88. June 1999.
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11.3 Mixed Oxide Process System Description

11.3.1 Conduct of Review

This section of the FSER contains the staff’s review of MP safety described by the applicant in

Section 11.2 of the revised CAR, with supporting process safety information from Chapters 5, 8,

and 11 of the revised CAR (DCS, 2002d). The objective of this review is to determine whether

the chemical process safety PSSCs and their design bases identified by the applicant provide

reasonable assurance of protection against natural phenomena and the consequences of

potential accidents. The staff evaluated the information provided by the applicant for chemical

process safety by reviewing Chapter 8 of the revised CAR, other sections of the revised CAR,

supplementary information provided by the applicant, and relevant documents available at the

applicant’s offices but not submitted by the applicant. The staff also reviewed technical literature

as necessary to understand the process and safety requirements. The review of MP safety

design bases and strategies was closely coordinated with the review of the radiation and

chemical safety aspects of accident sequences described in the safety assessment of the

design bases (see Chapter 5 of this FSER), the review of fire safety aspects (see Chapter 7 of

this FSER), and the review of plant systems (see Chapter 11 of this FSER). 

The staff reviewed how MP and chemistry information in the revised CAR addresses or relates

to the following regulations:

• 10 CFR 70.23(b) states, as a prerequisite to construction approval, that the design bases

of the PSSCs and the Quality Assurance Program be found to provide reasonable

assurance of protection against natural phenomena and the consequences of potential

accidents.

• 10 CFR 70.64 requires that BDC and defense-in-depth practices be incorporated into the

design of new facilities or new processes at existing facilities. With respect to chemical

protection, 10 CFR 70.64(a)(5) requires that the MFFF design 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. Related to chemical protection, 10 CFR 70.64(a)(3) requires that the

facility design provide for adequate protection against fires and explosions, such as those

that could be initiated by or involve chemicals at the proposed facility.

The review for this CAR stage focused on the design basis of chemical process safety systems,

their components, and other related information. For each chemical process safety system, the

staff reviewed information provided by the applicant for the safety function, system description,

and safety analysis. The review also encompassed proposed design basis considerations such

as redundancy, independence, reliability, and quality. The staff used Chapter 8.0 of

NUREG-1718 (NRC, 2000), as guidance in performing the review. As stated on page 8.0-2 of

NUREG-1718, information contained in the application should be of sufficient quality and detail

to allow for an independent review, assessment, and verification by NRC reviewers.

At NRC-licensed facilities, as stated in the “Memorandum of Understanding between the

Nuclear Regulatory Commission and the Occupational Safety and Health Administration: Worker

Protection at NRC-Licensed Facilities,” (Federal Register, Vol. 53, No. 210, October 31, 1998,

pp. 43950–43951), the NRC oversees chemical safety issues related to (1) radiation risk
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produced by radioactive materials, (2) chemical risk produced by radioactive materials, and (3)

plant conditions that affect the safety and safe handling of radioactive materials. These

chemical safety issues are important because they represent an increased radiation risk to

workers. The NRC does not oversee facility conditions that result in an occupational risk but do

not affect the safe use of licensed radioactive material.

The staff reviewed the revised CAR for the following areas applicable to process safety at the

CAR stage and consistent with the level of design (NRC, 2000, p. 8.0-8):

• MP description

• hazardous chemicals and potential interactions affecting licensed materials

• MP chemical accident sequences

• MP chemical accident consequences

• MP safety controls

Additional documentation from the applicant and the literature was reviewed as necessary to

understand the process and safety requirements. In addition, the revised CAR incorporates the

BDC of 10 CFR 70.64(a) into the design and operations of the proposed facility (see Section

5.5.5.4 of the revised CAR, p. 5.5-67 (DCS, 2004)), and applicable sections of the revised CAR

are intended to demonstrate compliance with these BDCs. 

The staff utilized the guidance provided by Chapter 8.0 of NUREG-1718 for assistance in

determining the compliance of the application with the regulation. The evaluation used the

guidance in Section 8.4 of NUREG-1718 for determining acceptance with 10 CFR Part 70,

consistent with the CAR stage and the level of the design. The evaluation is summarized in the

sections that follow.

11.3.1.1 System Description of the MP Process

This section provides a description and overview of the MP, including design, operational, and

process-flow information. This information is provided to support the hazard and accident

analysis provided in Chapter 5 of this FSER, as well as to assist in understanding the overall

design and function of the MP.

The MP area receives polished plutonium dioxide from the AP process, depleted uranium

dioxide (i.e., uranium depleted in the 235U isotope below the natural assay of 0.71 percent), and

the required components for assembling light-water reactor MOX assemblies. The process

mixes the plutonium and uranium dioxides to form MOX fuel pellets. The pellets are loaded into

fuel rods, which are then assembled into MOX fuel assemblies for use in commercial reactors. 

The MP area is designed to process up to 70 metric tons heavy metal (uranium plus plutonium)

annually. The safety functions of the PSSCs associated with the MP are discussed in Chapter 5

of the revised CAR.

The facility uses the advanced micronized master blend (A-MIMAS) process for the manufacture

of MOX fuel assemblies. The A-MIMAS process uses a two-step, dry mixing procedure. In the

first step, the plutonium dioxide powder is mixed with depleted uranium dioxide and recycled

scrap powder to form a primary blend (master blend) with approximately 20 percent plutonium

dioxide content of the total mass. This master blend is then micronized—reduced in particle size

into a very fine powder. In the second step, the primary blend is forced through a sieve, poured
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into a jar, and mixed with more depleted uranium dioxide and scrap powder to obtain the final

blend with the specified plutonium content (typically around 6 percent of the heavy metal

content). The two-step mixing process is used to ensure a consistent product.

The MP consists of 43 process units or systems divided into six areas corresponding to the

different segments of the process (see Figure 11.3-1).

Receiving Area

This area includes truck unloading, plutonium dioxide container handling, counting, and storage

before and after transfer to the AP line. The function of the receiving area is to receive, unload,

and store plutonium dioxide and uranium dioxide powder. The receiving area comprises the

following units:

• uranium dioxide receiving and storage unit

• uranium dioxide drum emptying unit

• plutonium dioxide receiving unit

• plutonium dioxide 3013 storage unit

• plutonium dioxide buffer storage unit

Powder Area

This area has equipment for dosing MOX powder at the specified plutonium content in two steps

for homogenizing and for pelletizing. The powder area receives uranium dioxide and plutonium

dioxide powders and produces a mixture of specific plutonium content suitable for the production

of MOX fuel pellets. The powder area is composed of the following units:

• plutonium dioxide can receiving and emptying unit

• primary dosing unit

• primary blend ball billing unit

• final dosing unit

• homogenization and pelletizing units

• scrap processing unit

• scrap ball milling unit

• powder auxiliary unit

• jar storage and handling unit

Pellet Process Area

In this area, MOX pellets are sintered, ground, and sorted. The function of the pellet process

area is to receive, store, process, and handle fuel pellets. The pellet process area is composed

of the following units:

• green pellet storage unit

• sintering units

• sintered pellet storage unit

• grinding units
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Figure 11.3-1 Overview of MP

• ground and sorted pellet storage unit

• pellet inspection and sorting units

• quality control and manual sorting units

• scrap box loading unit

• pellet repackaging unit

• scrap pellet storage unit

• pellet handing unit

Fuel Rod Process Area

In this area, pellets are loaded into rods, and the rods are inspected. The function of the fuel rod

process area is to assemble, inspect, and store fuel rods. The fuel rod process area is

composed of the following units:

• rod cladding and decontamination units

• rod tray loading unit

• rod storage unit

• helium leak test unit

• x-ray inspection units

• rod scanning unit
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• rod inspection and sorting unit

• rod decladding unit

Assembly Area

In this area, rods are loaded into assemblies, and the assemblies are inspected and stored. The

functions of the assembly area are to receive fuel rods and the required fuel assembly

components and to assemble, inspect, and store completed MOX fuel assemblies. The

assembly area is composed of the following units:

• assembly mockup loading unit

• assembly mounting unit

• assembly dry cleaning unit

• assembly dimensional inspection unit and assembly final inspection unit

• assembly handling and storage unit

• assembly packaging unit

Waste Area

In a separate waste area, solid radioactive waste generated during the MP is processed, stored,

and packaged for shipment. The waste area consists of the following units:

• filter dismantling unit

• maintenance and mechanical dismantling unit

• waste storage unit

• waste nuclear counting unit

A detailed description of the main units is provided in Section 11.2 of the revised CAR.

11.3.1.2 Staff Review of MP Process Safety

11.3.1.2.1 Potential Uranium Dioxide Pyrophoricity and Burnback Concerns

The MP will blend the depleted uranium dioxide powder with the plutonium dioxide to form the

matrix for the MOX fuel. Uranium dioxide powders are handled in conventional nuclear fuel

fabrication facilities. The staff review noted a potential concern regarding the pyrophoric nature

(sometimes referred to as burnback) of some fine uranium dioxide powders that can result in

oxidation, damage to equipment (essentially a thermal oxidation and heating effect), and a

potential release path resulting from the damage of confinement and filter systems (NRC, 1992). 

This is a known hazard, as such rapid oxidations have occurred in NRC-licensed facilities. 

Those events involved burnback reactions that started in process equipment and caused

localized damage, and then spread through the ventilation system. After those events, relatively

large quantities of uranium dioxide powders were found on the filters and equipment (including

polycarbonate barriers and filters (prefilters and primary HEPA) filters)) was damaged. The hot

uranium dioxide particles were carried through the ventilation system to the filters, where a

combination of the hot particles and continued oxidation reactions damaged the HEPA filters.

The health consequences were low because of rapid response by personnel and the relatively

low radiological hazard of uranium (i.e., as compared to plutonium). 
11-64

Burnback reactions evolve heat. The following depicts a typical reaction:

3 UO2 + O2 => U3O8

The reaction enthalpy is approximately -75 Kcal/mole U3O8 formed at 25 °C (77 °F) (i.e.,

exothermic—heat is released), using information from standard sources (e.g., Perry, 1997;

Cordfunke, 1969). This equates to approximately 440 J/g of uranium. Usually, mixtures of

oxides (e.g., U3O8 and UO3) are formed and the actual heat release may vary slightly—the

formation of UO3 actually increases the heat released by about 50 percent. The heat released

by the reactions increases slightly with increasing temperature. Table 11.3-1 summarizes the

burnback reaction enthalpies.

Table 11.3-1 Standard Burnback Reaction Enthalpies

Temperature

(°C)

UO3

Reaction

(Kcal/mol)

UO3

Reaction

(J/g U)

U3O8

Reaction

(Kcal/mol)

U3O8

Reaction

(J/g U)

25 -35.00 -615.59 -75.30 -441.46

100 -35.03 -616.12 -75.30 -441.45

150 -35.07 -616.78 -75.30 -441.47

200 -35.11 -617.54 -75.32 -441.57

250 -35.15 -618.31 -75.35 -441.75

300 -35.20 -619.08 -75.40 -442.03

350 -35.24 -619.81 -75.46 -442.40

400 -35.28 -620.50 -75.54 -442.87

Note: the negative sign indicates that the reaction is exothermic and releases heat.

The reaction initiation temperature and rate of heat release may also vary with the specific

surface area of the powder. Finely divided uranium dioxide (surface area greater than 10 m2/g)

can oxidize rapidly at room temperature. Coarser powders undergo oxidation at higher

temperatures (e.g., 60 °C (140 °F), and several steps and intermediates may be involved. The

burnback phenomena also requires optimal dispersion of a mixture of the particles in air.

Uranium dioxide powders in a container or loaded on filters and exposed to air oxidize slowly

(i.e., oxidant limited) unless this dispersion occurs. If the powder is too dispersed, no

propagation/powder heating occurs, and the uranium dioxide oxidizes slowly (i.e., analogous to a

fuel limited condition).

The applicant stated that the depleted uranium dioxide particle size will be approximately 100

microns before ball milling and greater than 2 microns after ball milling (DCS, 2003d). 

Therefore, from the time depleted uranium dioxide is received at the MFFF to when it is loaded

into the homogenization unit, there is a substantially lower risk of depleted uranium dioxide

burnback. However, after ball milling, the surface area of the depleted uranium dioxide is larger

(closer to 10 m2/g), and the risk of burnback is higher. 

The applicant discussed oxidation reactions of uranium dioxide in Section 8.5.1.6.1 of the

revised CAR and identified this as a potential cause for a fire in a glovebox. The applicant also

provided additional information on uranium dioxide powders in response to requests for
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additional information (RAIs) (DCS, 2001; DCS, 2002f) and identified temperature design bases

for materials of construction in the facility. The temperature design bases have been reviewed

by the staff and found to be acceptable (see Section 11.3.1.2.2 of this FSER). 

Documentation for powder processing areas was reviewed during an in-office review, and no

PSSCs or design bases were found to address the potential loss of confinement hazard (NRC,

2001). The applicant provided supplemental information on the subject (DCS, 2002b). The

applicant stated that uranium dioxide is processed as a fine powder at low temperatures and

within inert atmospheres; thus burnback does not occur during normal operations. The applicant

further indicated that burnback could occur during off-normal conditions if the inert atmosphere

has been replaced by air (the applicant has not currently identified a safety function for the inert

atmosphere). The applicant stated that burnback has been taken into account in the thermal

analysis of the facility during off-normal conditions and cites the RAI (DCS, 2001; DCS, 2002f)

responses. No PSSCs or design bases were identified by the applicant to address loss of

confinement concerns caused by burnback. In Section 8.5.1.6.1, the revised CAR does identify

the following as additional protective features (DCS, 2004, pp. 8-26):

• Uranium dioxide is delivered to the facility site and stored in sealed, 114-L (30-gal)

drums. Uranium dioxide is double bagged within the drums, under a nitrogen

atmosphere.

• Uranium dioxide is maintained in a nitrogen atmosphere throughout the process.

• Fire detection and suppression systems are provided for gloveboxes (carbon dioxide

injection) and process rooms (clean agent). 

• Noncombustible or nonflammable materials will be used for process equipment

construction and furnishing.

• Combustible materials will be controlled.

The staff concludes that a potential pyrophoric reaction or burnback of uranium dioxide cannot

be dismissed, because it has occurred previously during fine uranium dioxide fuel powder

processing and PSSCs had not been identified to address its potential effects upon confinement,

such as the entrainment of the (potentially hot) powder into the ventilation system, deposition on

filters, and damage to the filters by the hot powder particles and continuing oxidation reactions. 

This could potentially impact several units in the MP area that handle fine uranium dioxide

powder by itself or blended with plutonium dioxide. Such a burnback event could result in

damage to the final HEPA filters and loss of confinement, the release of large quantities of

uranium oxides (a chemical toxicity concern), the release of plutonium powders from the

commingled blend, and/or the initiation of other loss of confinement events such as fires. 

The applicant provided additional information on the filter system (DCS, 2003a). The applicant

again acknowledged that burnback reactions can occur but did not expect this to affect the final

HEPA filters for the following reasons:

• During normal operations, uranium dioxide is not expected to be present on the final

HEPA filters because the glovebox and very high depressurization exhaust system (VHD)

intermediate HEPA filters prevent any significant quantities of uranium dioxide powder
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from reaching the final HEPA filter housing and the first and second stage spark arrestor

(previously identified as the high-strength roughing filter and prefilter) remove nearly all of

the remaining particles before they reach the final HEPA filters (DCS, 2004). 

• During a fire, uranium dioxide powder is one of many potential embers. Large embers

are removed by the first and second stage spark arrestors. The remaining micron-size

particles are either cooled by the time they reach the final HEPA filters or do not contain

enough energy to degrade the performance of the final filters. The staff notes that the

uranium dioxide powder is likely to start and remain as small particles.

• The energy associated with the burnback phenomena is small when compared to the

energies involved in a facility area fire. 

[Text removed under 10 CFR 2.390]

The applicant provided the following design information on the final filtration unit and housing

design, which is identified as part of the filter system PSSC (DCS 2003a, Attachment 1; DCS

2003d; NRC 2003c; DCS, 2003b, Section 11.4.9):

• All of the ductwork is constructed of stainless steel.

• The first-stage roughing filter comprises stainless steel wire mesh in a stainless steel

frame. 

• The second-stage spark arrestor is stainless steel wire mesh interwoven with fiberglass. 

The filtration efficiency is 99 percent for particles greater than 1 micron in diameter. 

• A two-stage HEPA filter completes the facility HEPA filter train.

The applicant expected to credit the following engineered features and administrative controls on

the final filtration unit to ensure that the 10 CFR 70.61 performance requirements are met (DCS,

2003a, Attachment 1; DCS, 2003b, Sections 11.4.9 and 11.4.11.1.4):

• high-strength stainless steel mesh first-stage spark arrestor

• high-efficiency, high-strength second-stage spark arrestor (stainless steel/glass fiber

prefilter)
11-67

• protected two-stage final HEPA filters with a structural integrity of greater than 10 inches

of water (i.e., pressure drop)

• multiple redundant ventilation fan systems

• ventilation system design which ensures adequate airflow dilution

• ventilation system design which ensures a pressure drop of less than 10 inches of water

across the HEPA filter elements

• preventive maintenance to ensure HEPA filter integrity

• low combustible loads

• fire areas protected by 2-hour-minimum rated fire barriers

The staff concludes that potential burnback events are not prevented and that fine uranium

dioxide powders and uranium oxide embers from such events could travel through the ventilation

system to the final HEPA filter housing. 

The staff analyzed the potential impact from burnback events (NRC, 2004), summarized as

follows:

• The burnback event can be initiated wherever uranium dioxide is stored or processed,

provided that the uranium dioxide is finely divided, and sufficient energy (air temperature)

is present to start the reaction. At the MFFF, the staff assumes that only the depleted

uranium oxide which has been ball-milled is at risk.

• The sequence which results in damage to final HEPA filters from burnback is postulated

to start in a process glovebox downstream of the ball milling unit. The depleted uranium

dioxide could either be spilled or exposed to fire. Convective air currents would carry the

unreacted, finely divided depleted uranium dioxide into the glovebox ventilation exhaust

system. The unreacted, finely divided depleted uranium dioxide would travel to the final

C4 HEPA filters plenum. The burnback event is postulated to occur after, or just before,

the depleted uranium dioxide strikes the first stage of final C4 HEPA filters. The sudden

release of reaction energy (heat) could damage the first stage of final C4 HEPA filters,

resulting in a loss of pressure drop and loss of filter efficiency, across the first stage. 

Operations personnel would be notified of the loss by differential pressure-indicating

switches. A second C4 train would be brought on line, and the affected train would be

removed from service for maintenance.

• The consequences of both the spill-initiated and fire-initiated burnback events are low. 

However, the potential event results in damage to a PSSC and loss of its safety function.

The staff evaluated the hazard to final C3 and C4 HEPA filters that would be posed by a uranium

burnback event occurring on or near those filters. The staff postulated that both a glovebox spill

and a glovebox fire involving depleted uranium dioxide are credible means to disperse depleted

uranium dioxide into the ventilation systems. 
11-68

[Text removed under 10 CFR 2.390]

Multiple filters are present in the C3 and C4 ventilation systems; therefore the margins could be

higher.

The consequence analysis demonstrates that a HEPA filter would survive uranium burnback

after the maximum credible dispersal of depleted uranium dioxide powder in a glovebox spill or

fire. The margin of safety is at least a factor of 16 against spills, and a factor of 5 against fires. 

This margin is more than sufficient, given the conservative assumptions in the staff’s evaluation.

Consequently, the staff finds the PSSC and design basis information associated with the

ventilation system and final filtration units to be sufficient to address the potential burnback

phenomena associated with the pyrophoric nature of some uranium dioxide powders. This

closes the previously identified Open Item MP-1. The staff considers this to be related to the

filter soot loading issue discussed in Chapter 7 of this FSER (former Open Item FS-1).

11.3.1.2.2 Potential Plutonium Dioxide Heating Effects

The staff has reviewed plutonium-handling areas for potential chemical safety concerns. The

review noted concerns related to the potential heat generation by the plutonium

dioxide—plutonium in the glovebox environment can easily reach equilibrium temperatures of

80 °C (176 °F) (DOE, 1998, Section 2.6.3.1). 

Section 8.5.1.6.2 of the revised CAR and RAI responses (DCS, 2001; DCS, 2002f) provide

supplementary information on plutonium thermal effects, as well as a summary of the design

bases for decay heat and temperatures. The specific heat loads for plutonium were identified by

the applicant as follows:

• unpolished plutonium—2.899 W/kg of unpolished plutonium dioxide powder

• polished plutonium—2.181 W/kg of polished plutonium dioxide powder

Using values from the literature (DOE, 1998), the staff review estimates heat loads of 2.5–3.5

W/kg of plutonium dioxide, depending on the isotopic ranges used. These values generally

overlap the applicant’s heat load estimates. Thus, the applicant’s values are reasonable.

The applicant identified temperature design bases for the materials of construction in the facility,

which are reproduced as Tables 11.3-2 and 11.3-3 (DCS, 2001; DCS, 2002f). The response

identified the storage rooms, storage gloveboxes, and larger production units as having

potentially large heat loads. The applicant’s response mentions that the temperature design

bases in Tables 11.3-2 and 11.3-3 will be met during normal operations, but might be exceeded

during incidents where ventilation is not maintained for the plutonium dioxide storage area and

the handling and storage tunnel. Consequently, in Section 5.5.2.1.6.9 of the revised CAR, the

applicant identified the high depressurization exhaust system (HDE) (part of the C3 confinement

system) as the PSSC with the safety function of providing exhaust so that temperatures in the

3013 cannister storage area are maintained within the design basis values. The staff notes that

these design bases and approaches are consistent with accepted practice for steels

(e.g.,ASME), concrete (e.g., ACI), and most plastics (e.g., Perry, 1997) and finds this approach

to be acceptable. Polycarbonate is discussed separately in Section 11.4 of this FSER. The staff
11-69

finds this approach to be acceptable for the construction authorization stage according to the

acceptance criteria in Section 8.4.3.5 of the SRP.

Table 11.3-2 Applicant’s Design Basis Temperature Criteria

Material

Situation

Normal Operating

Temperature, °C (°F)

Hypothetical Maximum

Operating Temperature, °C

(°F)

Ordinary Concrete 60 (140) 100 (212)

Stainless Steel 425 (797) 425 (797)

BPP #9 80 (176) 100 (212)

BPP #10 100 (212) 100 (212)

NS41 Silicone Elastomer 180 (356) 180 (356)

Polycarbonate (Lexan) 35 (95) (thermal cycling)

50 (122)

70 (158)

BPP = borated polyethylene plaster

Table 11.3-3 Applicant’s Additional Temperature Design Basis Criteria

for Personnel Protection

Material Normal Operating

Temperature Limit, °C (°F)

Borate (colemanite) concrete 80 (176)

Kyowaglas—storage

 —operating

80 (176)

35 (95)

Fuel rods, pellets, and cladding 60 (140)

11.3.1.2.3 Potential Plutonium Dioxide Pyrophoricity and Burnback Concerns

The AP process produces plutonium (IV) oxalate, which generally requires a higher calcination

temperature (about 400–500 °C (752–932 °F)) than either plutonium (III) or plutonium (VI)

oxalates (about 200 °C (392 °F)). The staff found that plutonium dioxide powder from the

calciner in the oxalic precipitation and oxidation unit does not have a design basis or

specification, and the calcining operation does not have a design basis related to plutonium

dioxide quality. The staff concerns fall into two areas:

(1) The first concern involves oxidation reactions and powder dispersion because of the

presence of substoichiometric oxides of plutonium dioxide (essentially PuO2-x) that are

ignited by the decay heat. Depending on conditions, plutonium can form varying oxides,

some of which can be pyrophoric. In general, plutonium oxides with oxygen contents

lower than the dioxide are potentially pyrophoric (DOE, 1998, Section 2.6.3.2).
11-70

(2) The presence of water or other volatile species and impurities can increase pressure

inside containers because of radiolysis and decay heat and ultimately cause

overpressurization, resulting in an explosion and a loss of confinement. 

The applicant discussed plutonium dioxide pyrophoricity in Section 8.5.1.6.2 of the revised CAR. 

This section states that plutonium dioxide supplied to the facility will meet the requirements of

DOE-STD-3013-2000, “Stabilization, Packaging, and Storage of Plutonium-Bearing Materials,”

issued November 1999 (DOE, 1999), in order to ensure stability. For plutonium dioxide within

the facility, a standard or specification is not identified. Supplemental information provided by

the applicant (DCS, 2002b) states that plutonium dioxide is stable in air. The applicant stated

that the formation of substoichiometric oxides is not a concern, as reduction of the plutonium (IV)

oxalate does not occur. The staff notes that oxide formation depends on the process conditions,

and, without a specification or basis, some substoichiometric plutonium dioxide from impurity

reactions cannot be discounted. The powder will also be handled under a nitrogen atmosphere,

but this was not identified as a PSSC. 

The staff conducted a brief literature review and found that plutonium dioxide is often present as

a substoichiometric oxide (i.e., PuO2-x) and prone to absorb moisture unless it has been calcined

and held at a temperature of about 900 °C (1652 °F) for 2 hours to stabilize (ceramicize) the

material (DOE, 1998, Section 2.7). Unstabilized plutonium oxides may exhibit pyrophoric

reactions in air because of their substoichiometry or the radiolysis of absorbed water, which

could lead to a loss of confinement and release or initiate other events, such as fires.

The DOE experience with plutonium dioxide indicates that the pyrophoricity and stability

(including entrained, attached, and absorbed compounds, such as water) of plutonium dioxide

depend on the calcination conditions (e.g., time, temperature, and atmosphere), and that

substoichiometric and unstable plutonium dioxide can initiate events (DOE, 1999). The staff

review indicates that a lower calcination temperature may be involved at the facility, and

standards and limits may be needed to control plutonium dioxide substoichiometry and

instability. The staff did not find standards or limits described by the applicant that will assure

that the formation of significant amounts of substoichiometric or unstable plutonium dioxide

(from the calciner and in MP operations) is prevented. 

The applicant identified nominal values of humidity (water) in the powders entering the MP area. 

For plutonium dioxide, 1 percent is used for normal situations and 3 percent is used for offnormal situations. These were not identified as design bases and PSSCs were not specified. 

The DOE 3013 standard (DOE, 1999) mentions a loss on ignition of 0.5 percent (i.e., weight loss

from water vaporizing caused by heating of the material). Moderation (i.e., water content) is not

controlled by the applicant in the calciner, homogenization, and MP buffer storage units. Water

and volatile content have not been identified as design bases. 

Furthermore, at the February 13, 2003, public meeting (NRC, 2003b), the applicant stated that a

review was underway to determine if unstabilized plutonium dioxide would be received by the

facility. The staff review of the calcining section of the AP process (see Section 11.2 of this

FSER) did not identify any PSSCs or design bases for ensuring that stabilized plutonium dioxide

powder would be produced. At the December 10–12, 2002, public meeting (NRC, 2003a), the

applicant indicated the plutonium dioxide storage containers would be convenience cans, with

gas vent filters and spring-loaded tops for pressure relief. These were not identified as PSSCs. 
11-71

The applicant subsequently submitted additional information on the subject (NRC, 2001; DCS,

2003a). Leakage from the can is covered by the spill event discussed in Section 5.5.2.3.6.4 of

the revised CAR on load-handling controls and C4 confinement. For pressure accumulation

within the can, the lid is deterministically assumed to impact the glovebox. The materialhandling control PSSC is used to mitigate this potential event. For this case, this PSSC may

include the control of moisture content of the material, residence time of the canned material

(e.g., in the range of months and years), and/or design pressure of the reusable can. The

specific IROFS will be identified by DCS in its ISA summary and evaluated by the staff as part of

any license application DCS may later submit.

In addition, the applicant indicated that overpressurization from the oxidation of plutonium (III)

oxalate contained within the stored cans may be prevented through one of the following means:

• controls on plutonium oxalate furnace (calciner) parameters, such as residence time and

minimum temperature to ensure complete oxidation and moisture content of plutonium

oxalate entering the furnace

• experimental confirmation of the minimum water content accompanying plutonium (III) at

the exit of the furnace to prevent any overpressurization because of the energy liberated

during reoxidation (to plutonium (IV))

• measurement of the plutonium (III) content in the plutonium dioxide powder

The specific IROFS will be identified by DCS in its ISA Summary and evaluated by the staff as

part of any license application DCS may later submit.

The staff review found that leakage or spillage from the can that does not challenge confinement

is adequately addressed by the C4 confinement system, as discussed in Section 5.5.2.3.6.4 of

the revised CAR. For the case with pressure accumulation, the material-handling control PSSC

has the safety function to prevent impacts to the glovebox from (1) overpressurization from

residual volatile species (and their radiolysis) and (2) overpressurization from the oxidation of

plutonium (III). The description of the material handling control PSSC in Section 5.6.2.3 of the

revised CAR did not mention any design bases. The staff notes that the applicant appeared to

be considering both preventive and mitigative approaches for overpressurization events.

The applicant provided supplemental information on this subject (DCS, 2003b). This information

identifies an additional safety function of the material handling controls PSSC to prevent

potential overpressurization of the reusable plutonium dioxide cans resulting from radiolysis or

oxidation of plutonium (III) oxalate, as well as its subsequent impact to the glovebox. The

associated design basis is to ensure that the reusable can is designed to the maximum internal

pressure calculated for these events, plus an additional 10 percent as the margin. The staff

notes that this provides an approach with a defined margin and is consistent with the safety

approach for FTS components and code requirements, as discussed in Section 11.8 of this

FSER. The staff concludes that this provides an approach and range of design basis values

consistent with normal industry practice and finds this acceptable for the construction

authorization stage.
11-72

11.3.1.2.4 Sintering Furnace Concerns

The staff requested clarification and more information on the controls around the sintering

furnaces, including the hydrogen detectors, as they appear to involve a complex mixture of

hydrogen detectors, oxygen sensors, and pressure controls. In response (DCS, 2001; DCS,

2002f), the applicant provided a diagram that showed that part of the control range of hydrogen

in argon is flammable in air and stated that the sensors would detect hydrogen and, at 25

percent of the LFL, would terminate hydrogen flow at the hydrogen/argon mixing station. In

addition, fire detector(s) in the room would detect any fire and alarm but would not terminate

hydrogen flow. The applicant explained that the sintering furnace would not be in a glovebox,

and, as the room functioned as secondary confinement, the sintering room and the furnace

would become the PSSCs for confinement (DCS, 2001; DCS, 2002f). 

The sintering furnace has water-cooled walls and a moisture conditioning system for the furnace

gases (the hydrogen/argon mixture). The in-office review of the preliminary hazard analysis and

preliminary accident analysis did not find a potential steam explosion included (NRC, 2001). The

applicant provided supplemental information on potential steam explosions (DCS, 2002b). The

applicant stated that steam explosions have been identified during the facility safety analysis as

a credible event. Ongoing safety analyses by the applicant have identified three types of

scenarios that can lead to a steam explosion—(1) entry of water from the water cooling loop, (2)

entry of water from the humidifying loop, and (3) steam generation within the water cooling

systems. The applicant mentioned that a steam explosion involving a water-cooled furnace had

previously occurred at Los Alamos National Laboratory, but this involved internal cooling coils,

while the proposed furnaces would have external cooling coils. The applicant further stated that

a cooling water leak will be demonstrated to be highly unlikely, specific IROFS features will be

identified for the humidifying loop, and relief valves will render steam pressurization of the

cooling water loop highly unlikely. Supporting information to demonstrate an even frequency of

highly unlikely events is not included in the response. 

The staff reviewed Sections 5.5.2.4.6.2 and 11.2.2.16 of the revised CAR for additional

information. These sections identify a prevention strategy to address potential steam

explosions. The PSSC is the PSCS with the safety function of isolating humidifier water flow on

a high water level. Thus, the water supply to the humidifier would be terminated before the

humidifier overflowing and potentially allowing liquid water to enter the sintering furnace via the

gas supply side. This would be an active, engineered control. Such active engineered controls

can be designed with multiple layers, subsystems, and components in order to achieve very high

reliabilities for level control. The staff concludes that such an active engineered control could be

designed and implemented to prevent water intrusion and a subsequent steam explosion;

therefore, it finds it acceptable at the CAR stage. Sections 11.2.2.16 and 11.4.11.8 identify an

additional function of the PSCS to shut down the sintering furnace (by electrical cutoff) upon a

loss of cooling water flow, and to shut down zone heating if the related surface temperature is

excessive (greater than 60 °C (140 °F))—this is identified as the design basis. There is also a

backup cooling water supply, and the cooling water coils are on the outside of the furnace. The

staff reviewed preliminary view and section drawings of the sintering furnace during an in-office

review and concluded that (1) the coils are outside the sintering furnace shell, (2) the shell is

thicker than the cooling water tube thickness, and (3) the coils are not confined within additional

metal shells. The staff therefore concludes that the applicant can demonstrate that a steam

explosion from a cooling water leak is either unlikely or highly unlikely and, if it were to occur,

unlikely to penetrate the significantly thicker sintering furnace shell (i.e., if the event were to
11-73

occur at all, it would be directed toward the unconfined areas which are in the opposite direction

from the furnace itself). The staff finds that the applicant’s approach for addressing steam

explosions is acceptable for the CAR stage, according to Section 8.4.3.5 of the SRP. 

The applicant did not perform any coverage or location/distance analyses for sensors and

detectors. The applicant stated that it would verify that a hydrogen leak from the furnace would

be detected and terminated by pressure detection. The applicant expected that, between the

hydrogen monitors and pressure sensors, a hydrogen leak would be detected and the flow

terminated (DCS, 2002a; NRC, 2001). 

Section 11.6.7 of the revised CAR identifies the design bases for I&C PSSCs. This section

states that combustible and hydrogen detectors will be selected in accordance with ISA-S12.13-

Part I-1995, “Performance Requirements, Combustible Gas Detectors,” issued 1995. 

Installation, operation, and maintenance of combustible gas detectors would be in accordance

with ISA RP12.13-Part II-1987, “Installation, Operation, and Maintenance of Combustible Gas

Detection Equipment,” issued 1987. The staff has reviewed these standards for applicability. 

They include considerations of gas properties, sources, and detector locations, and they are a

routine approach for the industry. A setpoint methodology is also included, based on ANSI/ISA67.04.01-2000, “Setpoints for Nuclear Safety-Related Instrumentation,” issued 2000 (ANSI,

2000). This is also a standard approach for the nuclear industry. The staff concludes that the

design bases for hydrogen detection have been adequately described for the CAR stage.

The applicant indicated that hydrogen sensors in the room would detect any leaks and would

terminate the flow of hydrogen flow to the furnace. In addition, pressure controls would detect

any loss of pressure in the furnace caused by a leak and also terminate the hydrogen flow

(NRC, 2001). The staff expressed concerns about the potential for small leaks to result in

hydrogen burning that might go undetected and exacerbate radionuclide releases. Regarding

the airlocks, the applicant stated that there would be interlocks to prevent both doors (inner and

outer) from opening at the same time. In addition, oxygen sensors in the furnace would detect

atmospheric intrusion and terminate the hydrogen flow; these have been identified as PSSCs. 

The applicant provided supplemental information on PSSCs and design bases in the sintering

furnace area that identified additional PSSCs and design bases (DCS, 2002c). In addition,

Section 11.2.2.16 of the revised CAR states that, if hydrogen is detected in the furnace room,

then the gas supply is automatically shifted to pure argon (i.e., the hydrogen flow is terminated). 

In Section 5.5.2.4.6.1, the revised CAR discusses a prevention strategy for hydrogen explosions. 

The PSSC is identified as the PSCS, with the safety function of preventing the formation of an

explosive mixture of hydrogen within the facility associated with the use of the hydrogen-argon

mixture. The revised CAR indicates that the applicant is performing detailed analyses of the

hydrogen-argon system and associated furnace design and operations as part of the final design

and license application. The revised CAR notes that potential specific controls for preventing

hydrogen explosions around the sintering furnace (such as limiting the hydrogen content in the

hydrogen-argon mixture, monitoring for oxygen within the furnace, monitoring for hydrogen

outside of the furnace, and crediting dilution airflow associated with the HDE or VHD) have

already been identified as PSSCs in other safety strategies; thus there would be little or no

impact of the specific control selection upon design at the later stage. The staff finds the

proposed approach to be an acceptable strategy for preventing explosions around the sintering

furnace. The staff concludes that combinations of the specific controls already identified as

PSSCs should be capable of performing the safety function of preventing explosive mixtures. 
11-74

The staff notes that a specific approach based upon limiting the hydrogen content in the

hydrogen-argon mixture may be the simplest and most effective approach.

In Section 8.5.1.1.1, the revised CAR states that the hydrogen design basis is 25 percent of the

LFL of hydrogen in air. This value will not be exceeded during normal and off-normal conditions. 

Actual setpoints will be identified by DCS in its ISA summary and evaluated by the staff as part

of any license application DCS may later submit.

The applicant provided a methodology for LFL determination, which is discussed in Section 11.2

of the FSER.

The staff review indicates that 25 percent of the LFL in a chemical flow system is normally

considered a hydrogen limit analogous to a design basis (NRC, 1999). Other NRC-licensed fuel

fabrication facilities typically use 25 percent of the LFL as an administrative limit. The SRS

waste storage tank facilities alarm at 10 percent but use 25 percent of the LFL as their design

basis (Hobbs, 2000). The staff concludes that a value of 25 percent of the LFL is usually used

as a design basis or equivalent value in similar nuclear facility applications. The staff finds this

to be acceptable for the CAR stage. This also closes Open Item AP-2 in Section 11.2 of the

April 2003 DSER. 

11.3.1.2.5 Design Basis of the PSSCs and Applicable Baseline Design Criteria

The design bases of MP associated with chemical processing have been discussed in Sections

11.3.1.2.1–11.3.1.2.4 of this FSER. 

The applicant stated that the BDC specified in 10 CFR 70.64(a) are incorporated into the design

and operation of the facility (Section 5.5.5.4 of the revised CAR). The applicant stated that

information demonstrating compliance with these criteria is provided in the applicable chapters

of the revised CAR. For chemical protection, 10 CFR 70.64(a)(5) states the following:

Chemical protection. The design must provide for adequate protection against

chemical risks produced from licensed material, facility conditions which affect the

safety of licensed material, and hazardous chemicals produced from licensed

material.

The applicant indicated that reagents are stored, and chemical mixtures are prepared, in the

reagent processing building and the reagent storage area of the AP area, and they are generally

separated from each other and from radioactive materials. The applicant will avoid mixing

incompatible materials by design, controls, and procedures. The AP and MP facilities are

broken down into process functional units, which are made up of one or more subassemblies

performing consistent and elementary tasks. The applicant stated that the breakdown into

control functional units allows each entity to operate relatively independently in the given mode. 

The staff review notes that this separation and independence are consistent with accepted

practices for safe operations.

The applicant will control process storage and operation conditions in order to prevent

exothermic and potential autocatalytic reactions in the reagent processing building and the AP

and MP areas. Autocatalytic and exothermic reactions of chemicals would be prevented through

the control of the process parameters (e.g., reactant concentration, temperature, catalyst
11-75

concentration in solution, and pressure) that affect the reactions. Many of these controls have

been identified as PSSCs with design bases. Codes and standards have been applied on a

specific level, often as design bases.

The applicant stated that there is reasonable assurance that the PSSCs will be sufficiently

reliable and available based on the use of standard nuclear industry engineering practices. 

These practices are incorporated into the facility general design philosophy, design bases,

system design, and commitments to applicable management measures. These practices ensure

that applicable industry codes and standards are utilized, adequate safety margins are provided,

engineering features are utilized to the extent practicable, the defense-in-depth philosophy is

incorporated into the design, and PSSCs will be appropriately maintained.

The staff review finds that the applicant has provided sufficient information to meet the

requirements of 10 CFR 70.64(a)(5). Using the SRP, the staff review notes that the previously

identified open items in MP have been closed by additional safety strategies and information

from the applicant, and the staff concludes that the applicant has satisfied this BDC.

Related to chemical protection, the explosion protection BDC is stated as part of the fire

protection BDC in 10 CFR 70.64(a)(3):

Fire protection. The design must provide for adequate protection against fires

and explosions.

The applicant stated that there is reasonable assurance that the PSSCs will be sufficiently

reliable and available based on the use of standard nuclear industry engineering practices. 

These practices are incorporated into the facility general design philosophy, design bases,

system design, and commitments to applicable management measures. These practices ensure

that applicable industry codes and standards are utilized, adequate safety margins are provided,

engineering features are utilized to the extent practicable, the defense-in-depth philosophy is

incorporated into the design, and PSSCs will be appropriately maintained.

The staff review finds that the applicant has provided sufficient information to meet the

requirements of 10 CFR 70.64(a)(3). Using NUREG-1781, the staff review notes that the

previously identified open items regarding the MP process have been closed by additional safety

strategies and information from the applicant, and the staff concludes the applicant has satisfied

this BDC.

11.3.2 Evaluation Findings

In Section 11.3.7 of the revised CAR, DCS provided design basis information for the MP process

that it identified as PSSCs for the facility. Based on the staff’s review of the revised CAR and

supporting information provided by the applicant relevant to the MP process, the staff finds that,

for the reasons discussed above, DCS has met the BDC set forth in 10 CFR 70.64(a)(3) for

explosions and 10 CFR 70.64(a)(5) for chemical safety. Further, the staff concludes, pursuant

to 10 CFR 70.23(b), that the design bases of the PSSCs identified by the applicant will provide

reasonable assurance of protection against natural phenomena and the consequences of

potential accidents.
11-76

Several previously identified open items in the April 30, 2002, and April 30, 2003, DSERs have

now been closed in a satisfactory manner—MP-1, MP-2, MP-3, and MP-4. 

11.3.3 References

(ANSI, 2000) American National Standards Institute/Instrument Society of America. ANSI/ISA

67.04.01-2000, “Setpoints for Nuclear Safety-Related Instrumentation.” Research Triangle Park,

North Carolina, 2000.

(Cordfunke, 1969) Cordfunke, E.H.P. The Chemistry of Uranium (Including its Applications in

Nuclear Technology). Elsevier, New York, 1969.

(DCS, 2001) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Response to Request for Additional Information. August 31, 2001.

(DCS, 2002a) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Clarification of Responses to NRC Request for Additional Information. 

January 7, 2002.

(DCS, 2002b) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Clarification of Responses to NRC Request for Additional Information, DCSNRC-000083. February 11, 2002.

(DCS, 2002c) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Clarification of Responses to NRC Request for Additional Information. March

8, 2002.

(DCS, 2002d) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Clarification of Responses to NRC Request for Additional Information. April

23, 2002.

(DCS, 2002e) Ashe, K., Duke Cogema Stone & Webster. Letter to Docket Number 070-03098,

U.S. Nuclear Regulatory Commission, RE: Mixed Oxide Fuel Fabrication Facility—Construction

Authorization Request. October 31, 2002 (including page changes through February 9, 2005).

(DCS, 2002f) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Requests for Additional Information, Clarifications, and Open Item Mapping

into the Construction Authorization Request Revision (DCS-NRC-000120). November 22, 2002.

(DCS, 2003a) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Mixed Oxide (MOX) Fuel Fabrication Facility Construction Authorization

Request Change Pages. February 18, 2003.

(DCS, 2003b) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Mixed Oxide (MOX) Fuel Fabrication Facility Construction Authorization

Request, DCS-NRC-000129, February 18, 2003 (DCS-NRC-000131). April 1, 2003. 
11-77

(DCS, 2003c) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Response to Request for Additional Information—Chemical Safety Open

Items CS-9, AP-02, AP-08, and AP-09, DCS-NRC-000157. September 29, 2003. 

(DCS, 2003d) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Response to Request for Additional Information—DSER Open Items MP-01

(UO2) and AP-03 (Titanium Fires), DCS-NRC-000161. October 10, 2003. 

(DCS, 2004) Swiegart, R., Duke Cogema Stone & Webster. Letter to Docket Number 070-3098,

U.S. Nuclear Regulatory Commission, Mixed Oxide Fuel Fabrication Facility, Response to

Request for Additional Information. October 11, 2004.

(DOE, 1998) U.S. Department of Energy. DOE Std 1128-98, “Guide of Good Practices for

Occupational Radiological Protection in Plutonium Facilities.” Washington, DC, June 1, 1998.

(DOE, 1999) U.S. Department of Energy. DOE STD 3013-2000, “Stabilization, Packaging, and

Storage of Plutonium-Bearing Materials.” Washington, DC, November 1999.

(Hobbs, 2000) Hobbs, D.T. “Possible Explosive Compounds in the Savannah River Site Waste

Tank Farm Facilities,” WSRC-TR-91-444, Revision 3. February 15, 2000.

(NRC, 1992) U.S. Nuclear Regulatory Commission. Information Notice 92-14, “Uranium Oxide

Fires at Fuel Cycle Facilities.” Washington, DC, February 21, 1992.

(NRC, 1999) Pierson, R., U.S. Nuclear Regulatory Commission. Memorandum to D. Clark

Gibbs, U.S. Department of Energy, RE: Estimates of Hydrogen Generation from Wastes at the

Proposed TWRS-P Facility. April 21, 1999.

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility.” Washington,

DC, August 2000.

(NRC, 2001) Persinko, A., U.S. Nuclear Regulatory Commission. Memorandum to E.J. Leeds,

U.S. Nuclear Regulatory Commission, RE: November 27–29, 2001 In-Office Review Summary

of DCS Construction Authorization Request Supporting Documents for the MFFF. December

18, 2001.

(NRC, 2003a) Persinko, A., U.S. Nuclear Regulatory Commission. Memorandum to M.N.

Leach, U.S. Nuclear Regulatory Commission, RE: December 10–12 Meeting Summary: 

Meeting with DCS to Discuss Mixed Oxide Fuel Fabrication Facility Revised Construction

Authorization Report. January 31, 2003.

(NRC, 2003b) Persinko, A., U.S. Nuclear Regulatory Commission. Memorandum to M.N.

Leach, U.S. Nuclear Regulatory Commission, RE: February 6-7 Meeting Summary: Meeting

with DCS to Discuss Mixed Oxide Fuel Fabrication Facility Revised Construction Authorization

Report. March 5, 2003.

(NRC, 2003c) Persinko, A., U.S. Nuclear Regulatory Commission. Memorandum to K. Halvey

Gibson, U.S. Nuclear Regulatory Commission, RE: July 29–August 1, 2003, Meeting Summary: 
11-78

Meeting with Duke Cogema Stone & Webster to Discuss Open Items Related to Mixed Oxide

Fuel Fabrication Facility. August 29, 2003.

(NRC, 2004) Giitter, J., J. Holonich, and R. Pierson, U.S. Nuclear Regulatory Commission.

Memorandum to H. Peterson and M.N. Leach, U.S. Nuclear Regulatory Commission, RE: 

Safety Evaluation and Staff Positions on the Closure of Remaining Chemical Safety Open Items

MP-1 Pertaining to the Potential for Uranium Burnback to Damage Final HEPA Filters. April 15,

2004.

(Perry, 1997) Perry’s Chemical Engineers’ Handbook, Seventh Edition. McGraw-Hill, New York,

1997.

11.4 Ventilation and Confinement Systems

11.4.1 Conduct of Review

This section of the FSER contains the staff’s review of ventilation and confinement systems

described by the applicant in Chapter 11 of the revised CAR (DCS, 2002). The objective of this

review is to determine whether the ventilation and confinement PSSCs and their design bases

identified by the applicant provide reasonable assurance of protection against natural

phenomena and the consequences of potential accidents. The staff evaluated the information

provided by the applicant for ventilation and confinement systems by reviewing Chapter 11 of the

revised CAR, other sections of the revised CAR, and supplementary information provided by the

applicant. The review of ventilation and confinement systems design bases and strategies was

closely coordinated with the review of accident sequences described in the safety assessment of

the design bases (see Chapter 5 of this FSER) and the review of other plant systems. 

The staff reviewed how the information in the revised CAR addresses the following regulation:

• 10 CFR 70.23(b) states, as a prerequisite to construction approval, that the design bases

of the PSSCs and the Quality Assurance Program be found to provide reasonable

assurance of protection against natural phenomena and the consequences of potential

accidents.

The review of the CAR focused on the design basis of ventilation and confinement systems, their

components, and other related information. For ventilation and confinement systems, the staff

reviewed information provided by the applicant for the safety function, system description, and

safety analysis. The review also encompassed proposed design basis considerations such as

redundancy, independence, reliability, and quality. The staff used Chapter 11.0 in NUREG-1718

(NRC, 2000) as guidance for performing the review.

11.4.1.1 System Description

11.4.1.1.1 Functions and Major Components 

In the revised CAR, the applicant proposed a ventilation and confinement system to confine

radioactive materials within process areas and gloveboxes and to ensure minimum dispersal of

radioactive materials during routine operations and under accident conditions. The system was

intended to meet the recommendations in RG 3.12, “General Design Guide for Ventilation
11-79

Systems of Plutonium Processing and Fuel Fabrication Plants,” issued 1973 (NRC, 1973), and

consists of the following:

• Ventilation zones are operated at pressure differentials designed such that air leakage

occurs from areas of low radiation hazard into areas with the greater radiation hazard

(NRC, 2000; Section 11.4.1.2 of the revised CAR) (see Figure 11.4-1).

• A system of final filter assemblies is consistent with ASME N509, “Nuclear Power Plant

Air Cleaning Units and Components,” issued 1980 (ASME, 1980a), consisting of HEPA

filters, stainless steel/glass fiber prefilters, and stainless steel roughing filters intended to

remove radioactive materials from process areas and occupied areas during routine

operations and under accident conditions. The stainless steel/glass fiber prefilters and

stainless steel roughing filters are designed to remove hot embers and a large

percentage of the soot in order to protect the final HEPA filters from fire damage and

excessive soot plugging (DCS, 2003c; NRC, 2000; Section 11.4.9 of the revised CAR).

Figure 11.4-1 Ventilation Confinement Zones

• The design is in accordance with ASME N510 (ASME, 1980b) to allow testing and

inservice surveillance in order to ensure operability and required functional performance

(NRC, 2000; Section 11.4.11 of the revised CAR).

• Redundant PSSCs ensure that the performance requirements in 10 CFR 70.61 are met

(NRC, 2000).

• The design provides sufficient capacity and capability under routine operations. Under

accident conditions involving fires, the proposed design has sufficient capacity to

maintain air temperatures at the HEPA filters less than 232 °C (450 °F) and to ensure

that HEPA filters will remain operational. The applicant proposed to use air dilution of hot

air from a fire area with air from other areas to protect the HEPA filters (see Section

7.1.5.5 of this FSER ). The HEPA filters are capable of operating at temperatures up to

232 °C (450 °F) without severe degradation effects (NRC, 2000; Section 11.4.9 of the

revised CAR).

• Under accident conditions, the applicant proposed to only take credit for the final filtration

assemblies and not HEPA filters located on gloveboxes or at confinement zone

boundaries (NRC, 2000; Section 11.4.9 of the revised CAR). The applicant proposed to
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use a release factor of 1×10-4 for the final filter assemblies in its accident safety analyses

(Sections 5.5.3.2 and 11.4.9 of the revised CAR). The applicant based its proposal on

having redundant HEPA filter banks that are protected by stainless steel roughing filters,

and stainless steel/glass fiber prefilters in each redundant filter assembly with HEPA

filters that have been tested to have an efficiency for the removal of 0.3 micron particles

of at least 99.97 percent (DCS, 2003c). After installation, the HEPA filter banks will be

leak tested in accordance with ASME N510 (ASME, 1980b) to ensure that leakage

efficiency is at least 99.95 percent. The stainless steel roughing filters and stainless

steel/glass fiber prefilters are noncombustible and have sufficient structural strength to

collect greater than 90 percent of the soot generated in a design basis fire at the design

flow rate to withstand the full differential pressure generated by the fans when completely

clogged (DCS, 2003c). Under severe conditions, the applicant considered that each

bank of HEPAs will remove at least 99 percent of the particulates and will provide an

overall efficiency of 99.99 percent for the two combined banks.

• Monitoring instrumentation, alarms, and controls ensures that pressure differentials in

confinement zones are maintained, alternative power supplies are actuated when

needed, and the consequences of accidents are mitigated (NRC, 2000; Sections 11.4.2,

11.4.3, 11.4.4, and 11.4.7 of the revised CAR).

• The design provides for a safe air supply to the emergency control rooms, consisting of

heaters, coolers, prefilters, HEPA filters, and acid gas/organic vapor cartridges to remove

chemicals. This system will maintain a safe environment during emergencies for

personnel and equipment (NRC, 2000; Section 11.4.2.6 of the revised CAR).

• The design provides for the removal and replacement of filters and other expected

maintenance activities in order to minimize personnel exposures. The design allows for

in-place testing of HEPAs in accordance with ASME N510 (ASME, 1980b) in order to

ensure that HEPAs have been properly installed and are undamaged (NRC, 2000;

Section 11.4.9.1 of the revised CAR).

• Gloveboxes consist of welded stainless steel enclosures with windows, alone and in

interconnected groups, that act as a primary barrier to confine hazardous (radioactive,

toxic, or flammable) materials and to provide structural support capable of protecting

process equipment during a postulated seismic event. The MFFF personnel’s access to

equipment inside the gloveboxes is provided through access holes in the glovebox

windows fitted with gloves that maintain the confinement boundary. Gloveboxes that

contain powder or pellet forms are inerted with nitrogen gas to eliminate adverse effects

of atmospheric oxygen on the process or fuel (NRC, 2000; Section 11.4.7 of the revised

CAR).

• Glovebox window panels, viewing ports, or video cameras give visual access to the

gloveboxes. Light fixtures provide illumination and are generally located outside the

glovebox. The windows are clear rectangular panels that fit into gasketed frames that

cover specifically designed openings in the glovebox shell. The windows proposed by

the applicant are polycarbonate (Lexan®) that may have lead-impregnated polymer

sheets or lead-glass panels to provide additional radiation protection. The gasket

material is Neoprene, and the window panels range from small portholes to large, 1×1.5

m (3.3×4.9 ft) polycarbonate panels that are 10 mm (0.39 in.) thick. A panel of this size
11-81

would weigh a maximum of 18 kg (40 lb), not including frames, gaskets, glove hardware,

gloves, or panel-handling equipment. A panel of laminated safety glass of similar size

would weigh approximately 36 kg (79 lb) (NRC, 2000; Section 11.4.7 of the revised

CAR).

• Gloveboxes have pass-through connectors in glovebox shells that are used to bring

processes and utilities (e.g., air, electricity, and water) inside the glovebox. These

connectors are designed and tested to ensure that glovebox pressure integrity stays

within the maximum leakage criteria. At mechanical interfaces in the glovebox shells,

pass-throughs are designed to allow motion generated outside of the glovebox to be

transferred to equipment inside of the glovebox while maintaining confinement boundary

leak-tight criteria. Primary process equipment contains MOX product in various forms

(i.e., powder, pellets, trays, and rods). These MOX forms are manufactured, transferred,

stored, and maintained inside of gloveboxes (NRC, 2000; Section 11.4.7 of the revised

CAR). Gloveboxes designated as PSSCs and relied on to maintain the confinement

boundary include the following features (NRC, 2000; Section 11.4.11.2 of the revised

CAR):

— They use highly automated processes with hard-wired logic to avoid challenging

the glovebox confinement.

— They provide seismic support of process equipment and resist impacts and

potential load drops outside of the gloveboxes.

— They provide a stable base for barriers, stops, and guides that would prevent the

impact from equipment and material moving inside of the gloveboxes.

• Gloveboxes are designed, fabricated, and welded in accordance with national codes and

standards appropriate for their use (NRC, 2000; Section 11.4.7 of the revised CAR). The

glovebox system uses continuously welded construction with the seams ground smooth,

minimizes the holdup of powder, minimizes leakage paths, and facilitates the

decontamination of equipment. Fabrication and welding codes include ANSI N690,

“Specification for the Design, Fabrication, and Erection of Safety Related Steel

Structures for Nuclear Facilities,” issued 1994 (ANSI, 1994), and American Welding

Society (AWS) D1.1, “Structural Welding Code,” issued 1998 (AWS, 1998).

• The ability to safely shut down the primary process is facilitated by the seismic design for

the glovebox and similar equipment, as well as structural support members (NRC, 2000;

Section 11.4.7 of the revised CAR). Equipment geometry and alignment must be

maintained in order to maintain confinement. The glovebox system is designed to

prevent physical interaction with confinement boundary elements or PSSCs under

worst-case loading conditions associated with normal, off-normal, accident, and design

basis events in accordance with ANSI N690 (ANSI, 1994). The system will also be

designed to meet the criteria provided in RG 1.100, Revision 2, “Seismic Qualification of

Electric and Mechanical Equipment for Nuclear Power Plants,” issued 1988 (NRC, 1988).

• Ductwork is designed, fabricated, and tested in accordance with ASME B31.3, “Process

Piping,” issued 1998 (ASME, 1998); ASME N509 (ASME, 1998); and U.S. Energy
11-82

Research and Development Administration (ERDA) 76-21, “Nuclear Air Cleaning

Handbook,” issued 1976 (ERDA, 1976). 

11.4.1.1.2 Control Concepts

As described in Sections 11.4.2.2.4, 11.4.2.3.4, 11.4.2.4.4, 11.4.2.5.4, 11.4.2.6.4, 11.4.2.7.4,

11.4.3.4, 11.4.4.4, and 11.4.7.1.4 of the revised CAR, the ventilation and confinement system

provides for I&C systems to monitor and control the ventilation and confinement PSSCs. These

I&Cs include the following:

• Pressure I&Cs maintain proper negative pressures in each of the separate confinement

zones. Negative pressures are established such that airflows from areas having a lower

radiation hazard to areas having a higher radiation hazard. These controls include

pressure monitoring instrumentation and associated interlocks with individual

confinement systems to ensure that the proper negative pressures are maintained. Fan

controls are also provided to activate redundant fans, if needed, to ensure proper

negative pressures.

• Manual and automatic damper controls regulate air and gas flows within gloveboxes and

confinement zones. These controls maintain proper confinement zone pressures and

ensure proper airflows or isolation of gloveboxes in the event of glovebox breaks, in

addition to the isolation of fire areas in the event of fires. Tornado dampers are provided

to ensure that overpressures do not occur in the event of tornadoes or high winds.

• Controls are provided to transfer alternate power supplies. Switchgear controls ensure

that standby, emergency, and uninterruptible power supplies (UPSs) are properly

actuated if normal power supplies are lost.

• Instrumentation measures differential pressures across filter banks to alert operators if

filters need to be replaced.

• Variable-speed controls for fan operation maintain proper pressure control in each

confinement zone. Individual confinement systems are interlocked to ensure that

airflows from zones of higher radiation hazard cannot flow to zones of lower radiation

hazard.

• Air temperature instrumentation is provided for fire protection and to ensure that area

temperature requirements are properly maintained for habitability and for meeting

equipment requirements. Airflow instrumentation is provided to ensure proper system

operation.

• Nitrogen and dry air supply flow control is provided to ensure that proper confinement

zone negative pressures are maintained.

• Sintering furnace I&Cs are provided to ensure that proper reducing conditions are

maintained in the sintering furnace and that potential explosive or flammable conditions

are detected and prevented. In the event of leakage from a sintering furnace seal, the

staff calculated that a worker exposed to plutonium releases from a failed furnace seal

would receive a dose less than 0.1 Sv (10 rem). This dose is less than the limit for
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intermediate consequence events in 10 CFR 70.61 for a facility worker (0.25 Sv (25

rem)).

11.4.1.2 System Interfaces

Individual confinement zone ventilation systems are interconnected to ensure that the proper

negative pressures are maintained within confinement zones. Interlocks are provided to ensure

that airflows from zones of higher radiation hazard cannot flow to zones of lower radiation

hazard. The ventilation and confinement system also interfaces with the normal, standby,

emergency, and uninterruptible power supplies so that systems can properly function in the

event of power loss. Gloveboxes have functional and physical interfaces with ventilation

systems; electrical systems; air, gas, chilled water, and demineralized water systems; chemical

processing systems; and fire suppression systems (NRC, 2000; Sections 11.4.1, 11.4.1.2.3,

11.4.2.2.4, 11.4.2.2.5, 11.4.2.3.4, 11.4.2.3.5, 11.4.2.4.4, 11.4.2.4.5, 11.4.2.5.4, 11.4.2.5.5,

11.4.2.6.4, 11.4.2.6.5, 11.4.2.7.4, 11.4.2.7.5, 11.4.3.5, 11.4.4.4, 11.4.4.5, 11.4.7.1.4, and

11.4.7.1.5 of the revised CAR).

11.4.1.3 Design Bases of the PSSCs

Design bases for the ventilation and confinement system PSSCs are discussed in Section

11.4.11 of the revised CAR. Based on the applicant’s safety assessment, PSSCs for the

ventilation and confinement system are shown in Table 11.4-1.

The staff reviewed the design bases for the ventilation and confinement system PSSCs to

ensure that there is reasonable assurance of protection against natural phenomena and the

consequences of potential accidents. The design bases are consistent with RG 3.12 (NRC,

1973) and other industry air cleaning standards such as ASME N509, (ASME, 1980a), ASME

N510 (ASME, 1980b), and ASME AG-1, “Code on Nuclear Air and Gas Treatment,” issued 1991

(ASME, 1991). The staff also reviewed the design bases against the BDC in 10 CFR 70.64. 

The following discusses how the BDC are met.

The staff reviewed the revised CAR to ensure that the design bases would meet the quality

standards and that management measures would be appropriately applied. Application of

quality standards and management measures is discussed in Chapter 15 of this FSER.

For natural phenomena hazards, the staff reviewed the proposed design bases consistent with 

Chapter 5 of the revised CAR, which discusses the analyses performed to show that the facility

is adequately designed against natural phenomena with consideration of the most severe

documented historical events for the SRS. The ventilation and confinement system gloveboxes,

ductwork, and filter assemblies are designed to withstand the design basis earthquake. In

addition, tornado dampers ensure that the effects of tornadoes, hurricanes, and high winds

applicable to the proposed site do not degrade ventilation and confinement system PSSCs.

The staff reviewed the design bases of the ventilation and containment system to ensure that it

can withstand the effects of environmental conditions and the dynamic effects associated with

normal operations, maintenance, testing, and postulated accidents. The ventilation and

confinement system is designed to withstand fire and chemical effects. In-place testing and

maintenance of HEPA filters are performed in accordance with ASME N509 (ASME, 1980a) and

ASME N510 (ASME, 1980b). HEPA filters are designed to withstand pressure transients 
11-84

Table 11.4-1 Confinement System Design Bases

PSSC System Function Controlling Parameters Status

C4

Confinement

System

Control releases of

plutonium

Remain operable

during fires

Maintain differential

pressure between

glovebox and C3 areas

Maintain negative

confinement airflow

after small breaches

C4 zone pressure maintained at negative pressure with respect to C3 zone during normal

operation and transients

Redundant pressure sensors to maintain C4 pressures

Designed to maintain exhaust safety function assuming single active component failure

Final HEPA filter assembly release fraction—1×10-4

Final HEPA filter design temperature of 232 °C (450 °F)

Two 100% capacity redundant assemblies of two HEPA filter banks before discharge

Stainless steel roughing filters and stainless steel/glass fiber prefilters in each filtration

assembly upstream of HEPA filters

Four 100% capacity fans in C4 discharge system

Manually activated fire-isolation valves between designated fire areas

High-capacity flow to provide 0.635 m/s (125 ft/min) airflow velocity into a breach in the

glovebox primary confinement

In-place HEPA filter testing for final discharge filtration assemblies

System design in accordance with RG 3.12, except heat removed upstream of the final

HEPA filters during a fire is accomplished by airflow deletion, not water spray

HEPA filter design; HEPA filter housing design, construction, and testing; and HEPA filter

housing isolation dampers in accordance with ASME N509

HEPA filter design and testing; HEPA filter housing design and testing; ductwork and pipe

flexible connections; and fan design, construction, and testing in accordance with ASME

AG-1

Ductwork piping in accordance with ASME B31.3 

Bubble-tight isolation valve construction and testing, HEPA filter housing testing, and HEPA

filter testing in accordance with ERDA 76-21 

Filter testing in accordance with ASME N510 with each HEPA stage having a leakage

efficiency of 99.95%

Piping, valves, and fittings associated with gloveboxes in accordance with ASME B31.3

Fan power from normal (non-PSSC), standby (non-PSSC), emergency (PSSC), and

uninterruptible (PSSC) supplies

Remains operational during and after facility fires, tornadoes, and design basis earthquakes

HEPA filter

removal

efficiency may

be inadequate

for severe

accident

conditions such

as fires

HEPA filter soot

loading capacity

may be

inadequate

during fires
Table 11.4-1 Confinement System Design Bases

PSSC System Function Controlling Parameters Status

11-85

C3

Confinement

System

Control releases of

plutonium

Remain operable

during fires

Ensure 3013 canister

storage area

temperature

Provide cooling air to

electrical rooms that

support C3 and C4

exhaust system fans

C3 zone pressure maintained at negative pressure with respect to atmosphere during

normal operation and transients

HEPA filter release fraction—1×10-4

HEPA filter design temperature of 232 °C (450 °F)

Designed to maintain exhaust safety function assuming single active component failure 

Two 100% capacity redundant assemblies of two HEPA filter banks before discharge

Stainless steel roughing filters and stainless steel/glass fiber prefilters in each filtration

assembly upstream of HEPA filters

Two 100% capacity fans in C3 confinement system

Automatic or manual fire-rated dampers between designated fire areas

In-place HEPA filter testing for final discharge filtration assemblies

System design in accordance with RG 3.12, removed upstream of the final HEPA filters

during a fire, accomplished by airflow deletion, not water spray

HEPA filter design; HEPA filter housing design, construction, and testing; and HEPA filter

housing isolation dampers in accordance with ASME N509

HEPA filter design and testing; HEPA filter housing design and testing; ductwork and pipe

flexible connections; and fan design, construction, and testing in accordance with ASME

AG-1

Sheet metal ductwork design, construction, and testing; bubble-tight isolation damper

construction and testing; HEPA filter housing testing; and HEPA filter testing in accordance

with ERDA 76-21

Filter testing in accordance with ASME N510 with each HEPA stage having a leakage

efficiency of 99.95%

Tornado dampers

Fan power from normal (non-PSSC), standby (non-PSSC), and emergency (PSSC) supplies

Provides emergency cooling capability for selected areas

Remains operational after facility fires, tornadoes, and design basis earthquakes

HEPA filter

removal

efficiency may

be inadequate

for severe

accident

conditions such

as fires

HEPA filter soot

loading capacity

may be

inadequate

during fires
Table 11.4-1 Confinement System Design Bases

PSSC System Function Controlling Parameters Status

11-86

C2

Confinement

Passive

Barrier

System

Limit the dispersion of

radioactive material

Two HEPA filter banks before discharge

Stainless steel roughing filters and stainless steel/glass fiber prefilters in each filtration

assembly

HEPA filter design temperature of 232 °C (450 °F)

Automatic fire-rated dampers between designated fire areas

In-place HEPA filter testing for final discharge filtration assemblies

System design in accordance with RG 3.12, except heat removal is by airflow dilution

HEPA filter design; HEPA filter housing design, construction, and testing; and HEPA filter

housing isolation dampers in accordance with ASME N509

HEPA filter design and testing; HEPA filter housing design and testing; ductwork and pipe

flexible connections; and fan design, construction, and testing in accordance with ASME

AG-1

Sheet metal ductwork design, construction, and testing; bubble-tight isolation damper

construction and testing; HEPA filter housing testing; and HEPA filter testing in accordance

with ERDA 76-21

Filter testing in accordance with ASME N510 with each HEPA stage having a leakage

efficiency of 99.95%

Tornado dampers

Fan power from normal (non-PSSC) and standby (non-PSSC) supplies

Final filters and downstream ductwork remain structurally intact during and after tornadoes

and design basis earthquakes
Table 11.4-1 Confinement System Design Bases

PSSC System Function Controlling Parameters Status

11-87

Process Cell

Exhaust 

System

Effectively filter process

cell exhaust 

Operate to ensure that

a negative pressure

exists between the

process cell areas and

C2 areas 

HEPA filter release fraction—1×10-4

Two 100% capacity filtration units with two HEPA filter banks before discharge

Stainless steel roughing filters and stainless steel/glass fiber prefilters in each final filtration

assembly

HEPA filter design temperature of 232 °C (450 °F) 

Two 100% capacity fans

System design in accordance with RG 3.12, except heat removal is by airflow dilution

Fan power from normal (non-PSSC), standby (non-PSSC), and emergency (PSSC) supplies

Fire-rated dampers between designated fire areas

Process cell pressure maintained subatmospheric during normal and transient operation

Designed to maintain safety function assuming single component failure

HEPA filter design; HEPA filter housing design, construction, and testing; and HEPA filter

housing isolation dampers in accordance with ASME N509

HEPA filter design and testing; HEPA filter housing design and testing; ductwork and pipe

flexible connections; and fan design, construction, and testing in accordance with ASME

AG-1

Sheet metal ductwork design, construction, and testing; bubble-tight isolation damper

construction and testing; HEPA filter housing testing; and HEPA filter testing in accordance

with ERDA 76-21

Filter testing in accordance with ASME N510 with each HEPA stage having a leakage

efficiency of 99.95%

Remains operational after fires and tornadoes and design basis earthquakes

HEPA filter

removal

efficiency may

be inadequate

for severe

accident

conditions such

as fires
Table 11.4-1 Confinement System Design Bases

PSSC System Function Controlling Parameters Status

11-88

Emergency

Control

Room AirConditioning

System

Ensure habitable

conditions for operators

Provide cooling to the

emergency electrical

rooms

Redundant systems each with an outside air intake with continuous monitoring for

hazardous chemicals

One 100% capacity filtration stages (using prefilter stage, two HEPA filter stages, and

chemical filters) for each control room air supply and one 100% capacity booster fan for

each intake

One 100% capacity air handling unit per control room

One 100% capacity pressurization fan and one 100% capacity booster fan

One 100% capacity air conditioning unit for each emergency control room and for each

emergency electrical and battery room

Two 100% capacity exhaust fans for each emergency battery room

HEPA filter design temperature of 232 °C (450 °F)

Tornado dampers prevent overpressurization

In-place HEPA filter testing for final discharge filtration assemblies

HEPA filter design; HEPA filter housing design, construction and testing; and HEPA filter

housing isolation dampers in accordance with ASME N509

HEPA filter design and testing; HEPA filter housing design, and testing; ductwork and pipe

flexible connections; and fan design, construction, and testing in accordance with ASME

AG-1

Sheet metal ductwork design, construction, and testing; bubble-tight isolation damper

construction and testing; HEPA filter housing testing; and HEPA filter testing in accordance

with ERDA 76-21

Filter testing in accordance with ASME N510 with each HEPA stage having a leakage

efficiency of 99.95%

Fan power from normal (non-PSSC), standby (non-PSSC), and emergency (PSSC) supplies

Remains operational during and after facility fires and after tornadoes and design basis

earthquakes

Emergency

Diesel

Generator

Ventilation

System

Provide emergency

generator ventilation

Two 50% capacity air conditioning units for each switchgear room

One 100% capacity roof ventilator for engine room cooling during standby (engine fan cools

room during engine operation)

Fan power from normal (non-PSSC), standby (non-PSSC), and emergency (PSSC) supplies

Remains operational after facility fires, tornadoes, and design basis earthquakes
Table 11.4-1 Confinement System Design Bases

PSSC System Function Controlling Parameters Status

11-89

Offgas

Treatment

System

Provide an exhaust

path for the removal of

gases in process

vessels

Provide an exhaust

path for aqueous phase

evaporative cooling in

process vessels,

thereby providing a

mechanism for heat

removal 

Provide venting of

vessels/equipment that

potentially contain TBP

and its associated

byproducts to prevent

overpressurization in

the case of excessive

oxidation of TBP and/or

its degradation

products

HEPA filter release fraction—1×10-4

HEPA filter design temperature of 232 °C (450 °F)

Two 100% capacity redundant assemblies of two HEPA filter banks before discharge with

coolers and electric reheating coils on the filter inlets to reduce moisture carryover on the

filters

Two 100% capacity fans

In-place HEPA filter testing for final discharge filtration assemblies in accordance with ASME

N509 and N510

Piping and valves in accordance with ASME B31.3

Filter testing in accordance with ASME N510 with each HEPA stage having a leakage

efficiency of 99.95%

Fan power from normal (non-PSSC), standby (non-PSSC), and emergency (PSSC) supplies

Remains operational during and after tornadoes and design basis earthquakes

Gloveboxes

and

Glovebox

Pressure

Controls

Maintain glovebox

pressure

Gloveboxes are designed to remain functional during and after a design basis earthquake

Breach of confinement from load drops protected by automated processes

Sintering

Furnace

Confinement

Boundary

Control releases of

plutonium

Seals designed for peak operating temperature of 316 °C (601 °F)

No damage if furnace is shut down or if overheating or low cooling flow conditions occur

Furnace shell and airlocks designed to withstand overpressures of 250 kPa (2.5 bar)

Furnace shell leak tightness of 5×10-5 leakage at 15 kPa (2.2 psi)

Moisture carryover controls to prevent steam generation

Controls in place to minimize hydrogen hazards

Redundant pressure controls to maintain furnace pressure

Furnace designed to function during design basis earthquake
Table 11.4-1 Confinement System Design Bases

PSSC System Function Controlling Parameters Status

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Tornado

Dampers

Protect ventilation

system from tornado

effects

Resists design basis tornado effects

Remains operational after facility fires and design basis earthquakes

Canisters,

casks, and

containers

Control releases of

plutonium

Designed to standards to prevent releases underapplicable handling and accident

conditions

Supply Air

System

Supply emergency

cooling air to canister

storage area and

electrical rooms

Maintain confinement

zone differential

pressures

Provide supply air for emergency cooling

HEPA filter stages for static confinement

System design in accordance with RG 3.12

Sheet metal ductwork design, construction, and testing; bubble-tight isolation damper

construction and testing; and HEPA filter testing in accordance with ERDA 76-21

Filter testing in accordance with ASME N510 with each HEPA stage having a leakage

efficiency of 99.95%
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considering filter loadings and fan suction pressures. Filter replacement will be performed using

bag-in, bag-out procedures to reduce the possibility of spreading contamination.

For potential accidents involving fires, the staff reviewed the proposed design bases consistent

with Chapters 5 and 7 of the revised CAR, which describe the analyses and design features

applicable to fire protection. Design features of the ventilation and confinement system for fire

protection include filter assembly redundancy, use of dilution to mitigate the high-temperature

effects of a fire, use of stainless steel roughing filters and stainless steel/glass fiber prefilters to

prevent hot particles greater than 1 micron in size from contacting and starting fires on filters,

and use of two redundant banks of HEPA filters in each filter assembly having a temperature

rating of 232 °C (450 °F) and meeting the standards set by Underwriters Laboratories in UL 586,

“High-Efficiency, Particulate, Air Filter Units,” issued 1990 (DCS, 2003c; UL, 1990). 

The applicant considered the effects of fire on the ventilation and confinement system. The

applicant assumed that fires are restricted to single fire areas and will not spread (NRC, 2000;

Section 11.4.8 of the revised CAR). The staff reviewed the Fire Protection Program and, as

described in Section 7.1.5.5 of this FSER, concludes that during fires temperatures at the HEPA

filters can be maintained less than 204 °C (400 °F) and that HEPA filters can be protected

without severe temperature effects.

Under accident conditions, the applicant proposed to take credit only for the final filtration

assemblies and not the HEPA filters located on gloveboxes or at confinement zone boundaries

(NRC, 2000; Section 11.4.7.1.5 of the CAR). The applicant proposed to use a release factor of

1×10-4 for the final filter assemblies in its accident safety analyses (Sections 5.5.3.2 and 11.4.9

of the CAR). The applicant based its proposal on having redundant HEPA filter banks that are

protected by stainless steel roughing filters and stainless steel/glass fiber prefilters in each

redundant filter assembly with HEPA filters that have been tested to have an efficiency for the

removal of 0.3 micron particles of at least 99.97 percent (DCS, 2003c). After installation, the

HEPA filter banks will be leak tested in accordance with ASME N510 (ASME, 1980b) to ensure

that leakage efficiency is at least 99.95 percent. The stainless steel roughing filters and

stainless steel/glass fiber prefilters are noncombustible and have sufficient structural strength to

collect greater than 90 percent of soot generated in a design basis fire at the design flow rate to

withstand the full differential pressure generated by the fans when completely clogged (DCS,

2003c). Under severe conditions, the applicant considered that each bank of HEPAs will remove

at least 99 percent of particulates and will provide an overall efficiency of 99.99 percent for the

two combined banks. The staff reviewed this assumption in light of recommendations in

NUREG/CR-6410, “Nuclear Fuel Cycle Facility Accident Analysis Handbook,” issued 1998

(NRC, 1998), that a release factor for each filter assembly of no more than 1×10-2 should be

used under severe conditions such as a fire. Because of the following factors, the staff

considers that the proposed filtration system design has sufficient robustness to ensure that the

HEPA filters will be protected during a fire (DCS, 2003a; DCS, 2003b):

• The filter assembly contains stainless steel roughing filters and stainless steel/glass fiber

prefilters to minimize the possibility of hot embers contacting filters (DCS, 2003c). The

roughing filters use a first-stage stainless steel wire mesh in stainless steel frames with a

filtration efficiency of 60–70 percent for particles from 3–10 microns in size and 10–20

percent of 1–3 micron particles. The stainless steel/glass fiber prefilter uses a stainless

steel and fiberglass mesh with an efficiency of 99 percent for particles greater than 2

microns. These filters are capable of preventing burning embers and ash from
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contacting the HEPA filters or starting fires on dust-loaded prefilters which could

adversely affect the performance of the HEPA filters. Hot embers and ash particles that

are smaller than 10 microns—the particle sizes not removed by the stainless steel

roughing filters and stainless steel/glass fiber prefilters—would be unlikely to be carried

through the ventilation system ductwork and remain hot enough to ignite collected dust

on prefilters or HEPA filters (DCS, 2003c).

• HEPA filters have stainless steel frames in accordance with ASME AG-1 (ASME, 1991),

have stainless steel face guards, meet UL 586 fire resistance standards (UL, 1990), and

can operate at temperatures up to 232 °C (450 °F) and at differential pressures up to

2.49 kPa (10 kPa in water).

• The applicant’s Fire Protection Program is designed to contain fires within a single fire

area without spreading to adjacent areas. This program is based on using at least

2-hour fire barriers and combustible material controls (see Chapter 7 of this FSER). The

applicant also analyzed uncertainties in the spread of fires by evaluating the

temperatures and soot loadings if fires spread to the two worst-case fire areas. In these

analyses, the applicant found that temperatures at the final filters of the C4 and C3

systems are less than 188 °C (370 °F), and the differential pressures across the HEPA

filters would be less than 2.49 kPa (10 kPa in water).

Therefore, the staff considers that the filtration assemblies can perform their intended function

under conditions of fire and that use of a removal efficiency of 99.99 percent is acceptable.

The applicant also performed a soot generation analysis to ensure that filter loadings during a

fire would not adversely affect filter integrity. Using the Ballinger correlation (Ballinger, 1988)

corrected for the proposed filter sizes, the applicant computed a maximum HEPA filter loading

assuming the soot generation from the two fire areas producing the maximum amounts of soot. 

In these cases, greater than 90 percent of the soot generated during a fire is expected to be

collected on the stainless steel roughing filters and the stainless steel/glass fiber prefilters,

resulting in soot loadings at the HEPA filters of less than 350 g (12 oz) (DCS, 2003c). Under

these loads, the differential pressures across the HEPA filters are less than 2.49 kPa (10 kPa in

water) and would be sufficient to ensure that HEPA filters will not rupture from high particulate

loading. The applicant also committed to performing tests to determine the specific soot loads

applicable to the combustion of materials to be used at the MOX facility. Depending on the

results, the number of filters could be increased to maintain adequate soot loads (DCS, 2003c).

The staff reviewed the design bases of the ventilation and confinement system to ensure that it

provides adequate protection against chemical risks produced from licensed material, facility

conditions that affect the safety of licensed material, and hazardous chemicals produced from

licensed material. Gloveboxes are constructed of welded stainless steel to resist the corrosive

effects of chemicals used in AP and MOX fuel fabrication processes. In addition, ductwork and

filter assemblies upstream of the final filters are stainless steel, and filter materials will be

designed to withstand the chemical effects resulting from normal operations. Accident analyses

are further discussed in Chapter 5 of the revised CAR. Chemical safety is discussed in Chapter

8 of the revised CAR.

The staff reviewed the design bases of the ventilation and confinement system to ensure that it

provides for emergency capability to control the release of licensed material during normal
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operations and under postulated accident conditions. Release of licensed material is controlled

by the use of redundant HEPA filter banks in redundant filter assemblies. Individual HEPA filters

are tested to ensure that individual HEPA filters are capable of removing at least 99.97 percent

of 0.3 micron particles. Following installation, PSSC HEPA filters are in-place tested in

accordance with ASME N510 (ASME, 1980b) to ensure that leakage around filter banks is less

than 0.05 percent.

The staff reviewed the design bases of the ventilation and confinement system with respect to

the electrical power supply. The C4 confinement system is supplied by normal, standby,

emergency, and uninterruptible power supplies. The C3 exhaust system, process cell exhaust

system, the emergency control room, and EDG systems are supplied by normal, standby, and

emergency power supplies. The C2 confinement system and the supply air system are supplied

by normal and standby power supplies. These diverse power supply systems will ensure the

continued operation of ventilation and confinement system PSSCs. The staff’s review of the

electrical systems is provided in Section 11.5 of this FSER.

The staff reviewed the proposed design bases of ventilation and confinement system PSSCs to

ensure that they provide for adequate inspection, testing, and maintenance to ensure availability

and reliability to perform their function when needed. Redundant filter assemblies are provided

so that single filter assemblies can be taken offline for maintenance, testing, and filter

replacement. Dampers can be used to isolate individual filter assemblies and fans. The filter

assembly design includes provisions for in-place testing of HEPA filters in accordance with

ASME N510 (ASME, 1980b). Filter assemblies use bag-in/bag-out designs for filter replacement

to minimize the possibility of spreading contamination. 

The staff reviewed the proposed design bases of the ventilation and confinement system to

ensure that it provides for criticality control and adherence to the double contingency principle. 

Based on experience from the MELOX site, the applicant assumed that up to 3 kg of plutonium

dioxide could exist in the glovebox HEPA filter located in the pellet grinding glovebox, where

material becomes airborne at a rate of 0.3 g/h (0.01 oz/h), assuming that the HEPA filters are

replaced at 450-day intervals. This amount would be subcritical, as a quantity of 3 kg (6.6 lbs) of

plutonium dioxide is substantially less than the minimum critical mass. The ANSI/ANS standard

8.1, “Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors,”

issued 1988 (ANSI, 1988), contains single-parameter (i.e., always safe) subcritical limits for

239PuO2

 containing not more than 1.5 wt% water. At full density, the subcritical limit is 10.2 kg

(22.5 lb); at half density, the subcritical limit is 27 kg (59.5 lb). This would bound the worst-case

conditions that could be found in the HEPA filters, because the ANSI limits conservatively

assume that all of the plutonium is 239Pu (MOX plutonium will have at least 4 wt% 240Pu), and the

maximum density for unsintered plutonium dioxide powder (according to revised CAR Table 6-2)

falls within the density range covered by the limits in ANSI/ANS-8.1.

The moisture level of powder is limited to 1 wt% water, according to Section 6.3.4.3.2.6 and

Table 6-2 of the revised CAR. The MP gloveboxes are under an inert atmosphere, and this

quantity of water (approximately 10 percent by volume) would be readily noticeable to operators.

The amount of organic additives in the final blend is neutronically equivalent to 2 wt% water

(discussed in Section 6.3.4.3.2.6 of the revised CAR). Although this slightly exceeds the

bounding moderation level assumed in the ANSI/ANS-8.1 limits, this would not be sufficient

moderation to exceed the subcritical mass limits. In addition, the plutonium dioxide would be

spread over the surface of the filter medium and would not be accumulated in a spherical
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geometry. Therefore, application of the subcritical limits in ANSI/ANS-8.1 is considered

appropriate. The only remaining source of water would be the possible intrusion of

condensation via the ventilation system. Because the ventilation system is designed as a dry

system, condensation would not be present. Moreover, the ventilation system complies with the

double contingency principle because the design incorporates multiple confinement zones

before accumulation outside the ventilation system is possible. It would require multiple upset

conditions to accumulate an unsafe mass of plutonium oxides outside the process gloveboxes

through airborne migration. Therefore, the staff has reasonable assurance that the MFFF

ventilation system will be criticality safe. The staff’s review of nuclear criticality safety is provided

in Chapter 6 of this FSER.

The staff reviewed the proposed design bases of the ventilation and confinement system to

ensure that it provides for I&C systems to monitor and control the ventilation and confinement

PSSCs. These I&Cs include (1) pressure I&Cs to maintain proper negative pressures in each of

the separate confinement zones, (2) manual and automatic damper controls to regulate air and

gas flows within gloveboxes and confinement zones, (3) controls for the transfer of alternate

power supplies, (4) instrumentation to measure differential pressures across filter banks, (5)

variable-speed controls for fan operation, (6) air temperature and airflow instrumentation, and (7)

nitrogen and dry air supply controls. The staff’s review of I&C systems is provided in Section

11.6 of this FSER.

11.4.2 Evaluation Findings

In Section 11.4 of the revised CAR, DCS provided design basis information for the Ventilation

and Confinement Systems that it identified as PSSCs for the proposed MFFF. Based on the

staff’s review of the revised CAR and supporting information provided by the applicant relevant

to the ventilation and confinement systems, the staff finds that it can conclude, pursuant to

10 CFR 70.23(b), that the design bases of the PSSCs identified by the applicant will provide

reasonable assurance of protection against natural phenomena and the consequences of

potential accidents.

11.4.3 References

(ANSI, 1988) American National Standards Institute/American Nuclear Society. Standard

ANSI/ANS 8.1, “Nuclear Criticality Safety in Operations with Fissionable Materials Outside

Reactors.” New York, 1988.

(ANSI, 1994) American National Standards Institute. Standard ANSI N690, “Specification for

the Design, Fabrication, and Erection of Safety Related Steel Structures for Nuclear Facilities.”

New York, 1994.

(ASME, 1980a) American Society of Mechanical Engineers. Standard ANSI/ASME N509,

“Nuclear Power Plant Air Cleaning Units and Components.” New York, 1980.

(ASME, 1980b) American Society of Mechanical Engineers. Standard ANSI/ASME N510,

“Testing of Nuclear Air-Cleaning Systems.” New York, 1980.

(ASME, 1991) American Society of Mechanical Engineers. Standard ASME AG-1, “Code on

Nuclear Air and Gas Treatment.” New York, 1991.
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(ASME, 1998) American Society of Mechanical Engineers. Standard ASME B31.3, “Process

Piping.” New York, 1998.

(AWS, 1998) American Welding Society. Standard AWS D1.1, “Structural Welding Code.” 

Miami, Florida, 1998

(Ballinger, 1988) Ballinger, M.Y., et al., “Aerosols Released in Accidents in Reprocessing

Plants.” Nuclear Technology, Vol. 81. May 1988.

(DCS, 2002) Ihde, R., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, Mixed Oxide Fuel Fabrication Facility Construction Authorization Request Revision. 

October 31, 2002 (including page changes through February 9, 2005).

(DCS, 2003a) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: NRC to DCS letter dated 13 February 2003, “February 2003 Monthly Open

Item Status Report.” February 18, 2003.

(DCS, 2003b) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Construction Authorization Request Change Pages. April 10, 2003.

(DCS, 2003c) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Evaluation of Draft Safety Evaluation Report (DSER) on Construction of

Proposed Mixed Oxide Fuel Fabrication Facility, Revision 1. May 30, 2003.

(ERDA, 1976) U.S. Energy Research and Development Administration. ERDA 76-21, “Nuclear

Air Cleaning Handbook.” Oak Ridge, Tennessee, 1976.

(NRC, 1973) U.S. Nuclear Regulatory Commission. Regulatory Guide 3.12, “General Design

Guide for Ventilation Systems of Plutonium Processing and Fuel Fabrication Plants.” 

Washington, DC, 1973.

(NRC, 1988) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.100, Revision 2,

“Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants.” 

Washington, DC, 1988.

(NRC, 1998) U.S. Nuclear Regulatory Commission. NUREG/CR-6410, “Nuclear Fuel Cycle

Facility Accident Analysis Handbook.” Washington, DC, 1998.

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility.” Washington,

DC, 2000.

(UL, 1990) Underwriters Laboratories. Standard UL 586, “High-Efficiency, Particulate, Air Filter

Units.” Northbrook, Illinois, 1990.

(UL, 1995) Underwriters Laboratories. Standard UL 900, “Air Filter Units.” Northbrook, Illinois,

1995.
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11.5 Electrical Systems

11.5.1 Conduct of Review

This section of the FSER contains the staff’s review of electrical systems described by the

applicant in Chapter 11 of the revised CAR (DCS, 2002b). The objective of this review is to

determine whether the electrical PSSCs and their design bases identified by the applicant

provide reasonable assurance of protection against natural phenomena and the consequences

of potential accidents. The staff evaluated the information provided by the applicant for electrical

systems by reviewing Chapter 11 of the revised CAR, other sections of the revised CAR,

supplementary information provided by the applicant, and relevant documents available at the

applicant’s offices but not submitted by the applicant. The review of electrical systems design

bases and strategies was closely coordinated with the review of the I&C and electrical aspects of

accident sequences described in the safety assessment of the design bases (see Chapter 5 of

this FSER), as well as the review of other plant systems. 

The staff reviewed how the information in the revised CAR addresses the following regulations:

• 10 CFR 70.23(b) states, as a prerequisite to construction approval, that the design bases

of the PSSCs and the Quality Assurance Program must be found to provide reasonable

assurance of protection against natural phenomena and the consequences of potential

accidents.

• 10 CFR 70.64 requires that BDC and defense-in-depth practices be incorporated into the

design of new facilities. With respect to electrical systems, 10 CFR 70.64(a)(7) requires

that the MFFF design “provide for continued operation of essential utility services,” which

includes electrical power.

The review for this construction approval focused on the design bases of electrical systems, their

components, and other related information. For each electrical system, the staff reviewed

information provided by the applicant for the safety function, system description, and safety

analysis. The review also encompassed proposed design basis considerations such as

redundancy, independence, reliability, and quality. The staff used Chapter 11.0 in NUREG-1718

(NRC, 2000) as guidance for performing the review.

11.5.1.1 System Description

The facility electrical systems include the normal power, standby alternating current (ac) power,

and emergency power systems. The facility communications system, although not an electrical

system, is briefly described herein. 

11.5.1.1.1 Function

The electrical systems (as described in Section 11.5, “Electrical Systems,” of the revised CAR)

are to provide reliable electrical power for normal operation and safe shutdown and monitoring

when normal electrical power is lost. Specifically, electrical systems provide support to and are

essential for the operation of PSSCs such as the HDE, emergency control room air-conditioning

system, C4 confinement system (including very high depressurization), emergency control

system, EDG fuel oil system (EDGFOS), and emergency generator ventilation system.
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11.5.1.1.2 Major Components

11.5.1.1.2.1 Normal Power System. The normal power system distributes both ac and dc

power throughout the facility but is not a PSSC. Control of the breakers for the normal power

system is provided from the utility control system or locally.

Normal AC Power System

The normal ac power system consists of two separate 13.8-kV feeds from the F-area substation. 

Each feeder with a separate 13.8-kv/4.16-kV transformer and an associated distribution

switchgear bus is capable of supplying 100 percent of the facility’s loads. Each 4.16-kv

switchgear bus supplies power to five 4.16-kV/480-v load center transformers, one emergency

bus, and various 4.16-kV motors. Small loads are supplied power from 480-v to 120/240-v

transformers and associated distribution panels throughout the facility.

Normally, the two 4.16-kV switchgear buses are isolated from each other. If an offsite feed is

lost and the other feeder is available, an automatic delayed transfer connects the two switchgear

buses together to supply power from the one remaining offsite feeder. Additional crossties

utilizing series breakers are provided between the corresponding buses of most 480-v load

centers and are operated manually to facilitate maintenance.

Normal DC Power System

The normal dc system supplies power for the trip/close coils of low- and medium-voltage circuit

breakers in the normal power distribution system and other dc loads. The ungrounded system

consists of two 125-vdc lead-acid batteries, each with an associated charger and distribution

panel. Each battery has sufficient capacity to supply its loads for 1 hour, and each charger can

supply its loads plus recharge its associated battery to fully charged within 24 hours. 

Annunciators for low battery voltage, low charging current, and battery charger failure along with

battery voltage indication are provided in the utilities control rooms.

11.5.1.1.2.2 Standby AC Power System. The standby ac power system supplies power to loads

that are required for safe shutdown of the process on the loss of normal power but is not a

PSSC. The system also provides for a quick resumption of production to minimize the economic

impact to the facility. Upon the loss of an offsite feeder and failure of the automatic bus transfer,

loads are shed for the deenergized 4.16-kV switchgear bus and power is supplied to standby

loads by the standby ac power system. The system consists of two separate diesel generators

with associated auxiliary equipment and two 120/208-v UPSs.

Standby Diesel Generators

Both generators are connected, when required, to the deenergized 4.16-kV switchgear bus via

the 4.16-kV paralleling bus and are sized to carry all the facility life-safety loads together (critical

loads (non-IROFS) and loads on the associated emergency bus), plus other loads that are

important to facility production. The standby diesel generators are automatically started in about

10 seconds, and loads are sequenced on the bus based on standby power requirements. The

diesel generators can be stopped and/or started locally or from the utility and remote utility

control rooms. Synchronizing capability is provided for full-load testing. Each diesel is provided

with its own fuel oil day tank which stores fuel for 8 hours of operation, up to 2500 L (660 gal). 
11-98

An additional fuel oil supply tank is provided with enough fuel for both standby diesel generators

to operate for 24 hours at full load. Starting capability is provided by redundant batteries sized

for five cranking cycles, each with a charger capable of recharging its associated battery within

12 hours following a duty-cycle discharge.

Uninterruptible Power Supplies

The two 120/208-vac UPSs provide power to the programmable logic controllers (PLCs) used for

process control and instrumentation loads. Each UPS feeds an associated essential bus by a

battery/charger through parallel static inverters. When the static inverters are unavailable, the

corresponding essential bus is fed from a 480-v normal bus via a regulating transformer. In

addition, a maintenance bypass is provided.

11.5.1.1.2.3 Emergency Power System. The emergency power system distributes both ac and

dc power to PSSCs and is itself also a PSSC. Control of the breakers for the emergency power

system is provided in the emergency control room or locally.

11.5.1.1.2.3.1 Emergency AC Power System. The emergency ac power system supplies power

to redundant PSSCs when an offsite power source is lost, the bus transfer fails, and the standby

diesel generators do not start. Under those conditions or a degraded voltage condition, the

4.16-kv emergency bus is isolated from the normal 4.16-kV bus, loads are stripped from the bus

as required, the associated EDG is started, and emergency loads are automatically sequenced

onto the emergency bus. The system consists of two redundant and independent EDGs

(including their support systems) with their associated 4.16-kV emergency bus, a 480-v

emergency system, 480-v UPSs, and a 120-v vital ac power system. The system’s equipment is

qualified for the design basis seismic event and for the expected normal, off-normal, and design

basis accident environmental conditions. 

Emergency Diesel Generators

Each EDG is connected, when required, to its associated separate and independent 4.16-kV

emergency bus to provide emergency power to large PSSC loads, such as the HDE fans, and to

lower voltage levels for other PSSC loads, such as the UPSs for the VHD fans and the vital 120-

vac power system. Each diesel generator is automatically started on the loss of normal and

standby power or degraded voltage to its associated emergency bus and accepts loads in

approximately 10 seconds. The generators can be stopped and/or started locally or from their

associated emergency control room. Synchronizing capability is provided for full-load testing. 

Each diesel is sized to start and accelerate its connected loads and is provided with its own fuel

oil day tank which stores fuel for 8 hours of operation, up to 2500 L (660 gal). A separate fuel oil

supply tank is provided for each diesel generator with enough fuel for 7 days. The EDGs are the

same type as the standby diesel generators. Starting capability is provided by redundant

batteries sized for five cranking cycles, each with a charger capable of recharging its associated

battery within 12 hours following a duty-cycle discharge. An EDG can be taken out of service for

maintenance while the standby diesel generators and the redundant emergency ac power

system (including the redundant EDG) remain available.

Emergency 480-v System
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The 480-v emergency system supplies power to small 480-v loads and to emergency motor

control centers (MCCs). The system consists of two redundant 4.16-kv/480-v transformers,

each connected to a separate and independent 480-v emergency bus for power distribution. 

Redundant buses and MCCs permit maintenance of one train’s equipment with the opposite

train in service.

Emergency 480-v UPS

A dedicated 480-v UPS (two per each train of 4.16-kv emergency ac power) provides

uninterruptible power to one of the four VHD glovebox extraction fans. Each UPS consists of a

rectifier and an inverter with a battery bank capable of carrying a VHD fan for 1 hour. Power is

assumed to be restored to the UPS from either the associated emergency or the standby diesel

generators in 1 hour or less.

Vital 120 VAC Power System

The 120-v vital ac power system provides I&C power to PSSC loads. The system consists of

two independent and redundant UPSs, each supplying an associated vital bus by a

battery/charger (part of the emergency dc power system) through a static inverter. When the

static inverter is unavailable, the corresponding vital bus is fed from an associated 480-v

emergency bus via a regulating transformer. A maintenance bypass is also provided. Each

UPS battery has a 1-hour capacity and is located in a separate emergency battery room. The

remaining 120-v vital ac power system components for each UPS are located in separate rooms.

11.5.1.1.2.3.2 Emergency DC Power System. The emergency dc system supplies power for

the trip/close coils of low- and medium-voltage circuit breakers in the emergency power

distribution system, for some emergency lighting, and for other PSSC dc loads. The

ungrounded system consists of two redundant and independent 125-vdc lead-acid batteries,

each with an associated charger and distribution panel. Each battery has sufficient capacity to

supply its loads for 1 hour, and each charger can supply its loads plus recharge its associated

battery to fully charged from a minimum charged state within 24 hours. Physical separation and

electrical isolation prevent a fault in one train from affecting the opposite train. Annunciators for

low battery voltage, low charging current, and battery charger failure along with battery voltage

indication are provided in the utility control rooms.

11.5.1.1.2.4 Communication Systems. Communication systems provide voice and data

communications for normal operations, emergency response, and security and are supplied

power from the essential 120/208-vac buses. The applicant has stated that the communications

systems are not PSSCs because communication systems are not required to perform safety

functions during design basis events.

Normal Communication System

Part of the normal communication system consists of the telephone voice systems, including the

public switched telephone system and dedicated nonpublic branch exchange (PBX) telephone

lines, that are installed throughout the facility site. A public address system, consisting of central

amplifiers and electronics, remote microphones, and building speakers, will be installed to

provide intrasite and F-Area public address announcements. A wireless trunk system will be
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used for communication between operations and maintenance personnel, and a fiber optic data

highway will be provided throughout the facility.

Emergency Communication System

The emergency communication system consists of dedicated, non-PBX telephone lines between

the following:

• facility operations support center and SRS area emergency coordinators

• facility fire alarm system and SRS operations center

• facility operations support center and SRS safety system/alarm public announcement

system

• facility operations support center and SRS operations center, emergency operations

center, and technical support center.

In addition, each safe haven location is provided with telephone, intercom, and public address

systems.

11.5.1.2 System Interfaces

The normal, standby, and emergency power systems provide electrical power to equipment

throughout the facility and thus interface with most facility systems and components, especially

PSSCs such as the HDE fans and systems listed in Section 11.5.1.1.1 of this FSER. 

Environmental conditions for proper operation of the electrical systems are maintained by the

HVAC systems, and protection against fires is provided by the fire protection systems. The

electrical systems and their components are controlled by various manual and automatic

controls in the process, utility, and emergency control systems. Fuel oil is provided to the diesel

generators in the electrical systems via interfaces with the EDGFOS and standby diesel

generator fuel oil system.

11.5.1.3 Design Bases of the PSSCs and Applicable Baseline Design Criteria

For design basis requirements, the electrical systems designated as PSSCs are to be available

and provide reliable electrical power for normal operation and safe shutdown and monitoring

during credible external events. The electrical systems designated as PSSCs should remain

functional when subjected to severe natural phenomena and environmental and dynamic effects,

consistent with the BDC in 10 CFR 70.64(a)(2) and 10 CFR 70.64(a)(4), respectively, and be

adequately protected from fires in accordance with 10 CFR 70.64(a)(3). Additionally, the

electrical systems must support the continued operation of essential utility services, consistent

with the BDC in 10 CFR 70.64(a)(7).

The electrical systems should also be designed with I&Cs for monitoring and controlling each

bus and major component, consistent with the BDC in 10 CFR 70.64(a)(10). Pursuant to 10

CFR 70.64(b), the electrical systems should be designed using defense-in-depth practices, with

a preference for engineered controls over administrative controls. 
11-101

To ensure that the design basis requirements and BDC are met, the applicant has committed to

specific industry standards and staff guidance as discussed in the following sections.

11.5.1.3.1 Emergency AC Power System

The emergency ac power system (as a PSSC) will be designed with redundancy, independence,

physical separation, no single-failure vulnerability, and failsafe performance with sufficient

capacity, capability, periodic testing and maintenance, and adequate protective relaying to

perform its safety function and maintain qualification for natural phenomena and environmental

and dynamic effects (see Section 11.11 of this FSER). It will also be designed using guidance

from the following sources (to the extent described in the revised CAR and the applicant’s letters

dated August 31, 2001, December 5, 2001, and November 22, 2002 (DCS, 2001b, 2001c,

2002c)):

• for overall system design:

— Institute of Electrical and Electronics Engineers (IEEE), IEEE Std 308-1991,

“IEEE Standard Criteria for Class 1E Power Systems for Nuclear Generating

Stations,” (IEEE, 1992b)

— RG 1.32, Revision 2, “Criteria for Safety-Related Electric Power Systems for

Nuclear Power Plants,” issued 1977 (NRC, 1977)

• for equipment seismic qualification (SQ):

— IEEE Std 344-1987, “IEEE Recommended Practices for Seismic Qualification of

Class 1E Equipment for Nuclear Generating Stations,” issued 1987 (IEEE, 1987)

— RG 1.100, 1988 (NRC, 1988)

• for periodic testing:

— IEEE Std 338-1987, “IEEE Standard Criteria for Periodic Testing of Nuclear

Power Generating Station Class 1E Power and Protection Systems,” issued 1987

(IEEE, 1987)

— RG 1.118, Revision 3, “Periodic Testing of Electric Power and Protection

Systems,” issued 1994 (NRC, 1994)

• for single failure:

— IEEE Std 379-1994, “IEEE Standard Application of the Single Failure Criterion to

Nuclear Power Generation Station Safety Systems,” issued December 16, 1994

(IEEE, 1994)

• for qualification of cables installed in open trays:
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— IEEE Std 383-1974, “IEEE Standard for Type Test of Class 1E Electric Cables,

Field Splices and Connections for Nuclear Power Generating Stations,” issued

April 15, 1974 (IEEE, 1974)

• for electrical independence and separation:

— IEEE Std 384-1992, “Standard Criteria for Independence of Class 1E Equipment

and Circuits,” issued December 1, 1992 (IEEE, 1992a)

— RG 1.75, Revision 2, “Physical Independence of Electric Systems,” issued

September 1978 (NRC, 1978)

• for equipment protection:

— IEEE Std 741-1997, “IEEE Standard Criteria for the Protection of Class 1E Power

Systems and Equipment in Nuclear Power Generating Stations,” issued

September 25, 1997 (IEEE, 1997b)

• for design and installation of batteries:

— IEEE Std 484-1996, “IEEE Recommended Practice for Installation Design and

Installation of Vented Lead-Acid Batteries for Stationary Applications,” issued July

17, 1996 (IEEE, 1996b)

The EDGs will be designed using guidance from the following sources (to the extent described in

the revised CAR and the applicant’s letters dated August 31, 2001, December 5, 2001, and

November 22, 2002 (DCS, 2001b, 2001c, 2002c)):

• for fuel oil quality:

— ANSI/ASTM D975-94, “Standard Specification for Diesel Fuel Oils,” issued

October 10, 1995 (ANSI, 1995)

• for overall design and qualification:

— IEEE Std 387-1995, “IEEE Standard Criteria for Diesel Generator Units Applied

as Standby Power Supplies for Nuclear Power Generating Stations,” issued April

12, 1996 (IEEE, 1996a)

— RG 1.9, Revision 3, “Selection, Design, Qualification, and Testing of Emergency

Diesel Generator Units Used as Class 1E Onsite Electric Power Systems at

Nuclear Power Plants,” issued July 1993 (NRC, 1993)

The emergency 480-v UPS and 120-v vital ac power system will be designed using guidance

from the following sources (to the extent described in the revised CAR and the applicant’s letter

dated August 31, 2001, and November 22, 2002 (DCS, 2001b, 2002c)):

• For overall design:
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— IEEE Std 944-1986, “IEEE Recommended Practice for the Application and

Testing of Uninterruptible Power Supplies for Power Generating Stations,” issued

June 30, 1986 (IEEE, 1986)

11.5.1.3.2 Emergency DC Power System

The emergency dc power system (as a PSSC) will be designed with redundancy, independence,

physical separation, no single-failure vulnerability, and failsafe performance with sufficient

capacity; capability; periodic testing and maintenance; adequate protective relaying to perform

its safety function; and qualification for natural phenomena, environmental, and dynamic effects

(see Section 11.11 of this FSER). It will also be designed using guidance from the following (to

the extent described in the revised CAR and the applicant’s letters dated August 31, 2001,

December 5, 2001, March 8, 2002, and November 22, 2002 (DCS, 2001b, 2001c, 2002a,

2002c)):

S for overall system design:

— IEEE Std 308-1991 (IEEE, 1992b)

— IEEE Std 946-1992, “IEEE Recommended Practice for the Design of SafetyRelated DC Auxiliary Power Systems for Nuclear Power Generating Stations,”

issued February 25, 1993 (IEEE, 1993)

— RG 1.32, Revision 2 (NRC, 1977)

• for equipment SQ:

— IEEE Std 344-1987 (IEEE, 1987)

— RG 1.100, Revision 2 (NRC, 1988)

• for periodic testing:

— IEEE Std 338-1987 (IEEE, 1988)

— IEEE Std 450-1995, “IEEE Recommended Practice for Maintenance, Testing, and

Replacement of Vented Lead-Acid Batteries for Stationary Applications,” issued

May 31, 1995 (IEEE, 1995a)

— RG 1.118, Revision 3 (NRC, 1994)

• for single failure:

— IEEE Std 379-1994 (IEEE, 1994)

• for electrical independence and separation:

— IEEE Std 384-1992 (IEEE, 1992a) 
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— RG 1.75, Revision 2 (NRC, 1978)

• for design and installation of batteries:

— IEEE Std 484-1996 (IEEE, 1996b)

— IEEE Std 485-1997, “IEEE Recommended Practice for Sizing Lead-Acid Batteries

for Stationary Applications,” issued September 3, 1997 (IEEE, 1997a)

— NFPA 111, “Standard for Stored Electrical Energy Emergency and Standby

Power Systems,” issued 1996 (NRC, 1996)

• for qualification of cables:

— IEEE Std 383-1974 (IEEE, 1974)

In Section 5.5.5 of the revised CAR, the applicant described its general design philosophy used

in formulating the preliminary design of the facility. Pursuant to 10 CFR 70.64(b), in order to

ensure that, to the extent practicable, engineered controls are relied upon over administrative

controls, DCS has established a hierarchy of controls. In further adherence to 10 CFR 70.64(b),

the applicant stated that it has incorporated defense-in-depth practices at new facilities in its

preliminary facility design. Defense-in-depth is defined in the 10 CFR 70.64 footnote. In Section

5.5.5.2 of the revised CAR, DCS stated that it has incorporated defense-in-depth practices

through use of the single-failure criterion. Under this criterion, PSSCs are required to be

capable of carrying out their functions in the event that any single active component fails,

whether such failure occurs within the applicable system or in an associated system that

supports the component’s operation. 

Accordingly, the facility electrical systems should be designed using defense-in-depth practices

employing the single-failure criterion, including redundancy, independence, separation, and failsafe for electrical systems identified as PSSCs through the use of the industry standards listed

above. For example, the emergency ac power system, which provides power if offsite power is

lost, is designed with redundant and independent EDGs that are physically separated from each

other. This is discussed in Sections 11.5.1.3.1 and 11.5.1.3.2 of this FSER. 

11.5.2 Evaluation Findings

In Section 11.5.7 of the revised CAR, DCS provided design basis information for electrical

systems that it identified as PSSCs for the facility. Based on the staff’s review of the revised

CAR and the supporting information provided by the applicant, as well as the applicant’s

commitments to the industry standards and guidance discussed in the sections above for

electrical systems, the staff finds that DCS has met the BDC set forth in 10 CFR 70.64(a)(7). 

The staff concludes, pursuant to 10 CFR 70.23(b), that the design bases of the PSSCs

evaluated in this FSER section will provide reasonable assurance of protection against natural

phenomena and the consequences of potential accidents. The staff further concludes, from its

review of the revised CAR and supporting information submitted by DCS, that the preliminary

design of the facility electrical systems meet the defense-in-depth provisions and the preference

to engineered controls over administrative controls stated in 10 CFR 70.64(b).
11-105

11.5.3 References

(ANSI, 1995) American National Standards Institute/American Society for Testing Materials. 

D975-94, “Standard Specification for Diesel Fuel Oils.” Philadelphia, Pennsylvania. October 10,

1995.

(DCS, 2001a) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: MOX Fuel Fabrication Facility Site Geotechnical Report

(DCS01–WRS–DS–NTE–G–0005–C). August 10, 2001. 

(DCS, 2001b) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Response to Request for Additional Information (DCS-NRC-000059). August

31, 2001.

(DCS, 2001c) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Clarification of Responses to NRC Request for Additional Information

(DCS–NRC–000074). December 5, 2001.

(DCS, 2002a) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Clarification of Responses to NRC Request for Additional Information

(DCS–NRC–000085). March 8, 2002.

(DCS, 2002b) Ihde, R, Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Mixed Oxide Fuel Fabrication Facility Construction Authorization Request

Revision (DCS-NRC-000114). October 31, 2002.

(DCS, 2002c) Hastings, P., Duke Cogema Stone& Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Requests for Additional Information, Clarifications and Open Item Mapping

into the Construction Authorization Request Revision (DCS-NRC-000120). November 22, 2002.

(IEEE, 1974) Institute of Electrical and Electronics Engineers. Std 383-1974, “IEEE Standard

for Type Test of Class 1E Electric Cables, Field Splices and Connections for Nuclear Power

Generating Stations.” New York, April 15, 1974.

(IEEE, 1986) Institute of Electrical and Electronics Engineers. Std 944-1986, “IEEE

Recommended Practice for the Application and Testing of Uninterruptible Power Supplies for

Power Generating Stations.” New York, June 30, 1986. 

(IEEE, 1987) Institute of Electrical and Electronics Engineers. Std 344-1987, “IEEE

Recommended Practices for Seismic Qualification of Class 1E Equipment for Nuclear

Generating Stations.” New York, August 3, 1987.

(IEEE, 1988) Institute of Electrical and Electronics Engineers. Std 338-1987, “IEEE Standard

Criteria for Periodic Testing of Nuclear Power Generating Station Class 1E Power and

Protection Systems.” New York, August 22, 1988.

(IEEE, 1992a) Institute of Electrical and Electronics Engineers. Std 384-1992, “Standard

Criteria for Independence of Class 1E Equipment and Circuits.” New York,, December 1, 1992.
11-106

(IEEE, 1992b) Institute of Electrical and Electronics Engineers. Std 308-1991, “IEEE Standard

Criteria for Class 1E Power Systems for Nuclear Generating Stations.” New York, August 26,

1992.

(IEEE, 1993) Institute of Electrical and Electronics Engineers. Std 946-1992, “IEEE

Recommended Practice for the Design of Safety-Related DC Auxiliary Power Systems for

Nuclear Power Generating Stations.” New York, February 25, 1993.

(IEEE, 1994) Institute of Electrical and Electronics Engineers. Std 379-1994, “IEEE Standard

Application of the Single Failure Criterion to Nuclear Power Generation Station Safety Systems.” 

New York, December 16, 1994.

(IEEE, 1995a) Institute of Electrical and Electronics Engineers. Std 450-1995, “IEEE

Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid

Batteries for Stationary Applications.” New York, May 31, 1995.

(IEEE, 1995b) Institute of Electrical and Electronics Engineers. Std 765-1995, “IEEE Standard

for Preferred Power Supply (PPS) for Nuclear Power Generating Stations.” New York, July 11,

1995.

(IEEE, 1996a) Institute of Electrical and Electronics Engineers. Std 387-1995, “IEEE Standard

Criteria for Diesel Generator Units Applied as Standby Power Supplies for Nuclear Power

Generating Stations.” New York, April 12, 1996.

(IEEE, 1996b) Institute of Electrical and Electronics Engineers. Std 484-1996, “IEEE

Recommended Practice for Installation Design and Installation of Vented Lead-Acid Batteries for

Stationary Applications.” New York, July 17, 1996.

(IEEE, 1997b) Institute of Electrical and Electronics Engineers. Std 485-1997, IEEE

Recommended Practice for Sizing Lead-Acid Batteries for Stationary Applications.” New York,

September 3, 1997.

(IEEE, 1997a) Institute of Electrical and Electronics Engineers. Std 741-1997, “IEEE Standard

Criteria for the Protection of Class 1E Power Systems and Equipment in Nuclear Power

Generating Stations.” New York, September 25, 1997. 

(NFPA, 1996) National Fire Protection Association Inc. NFPA 111, “Standard for Stored

Electrical Energy and Standby Power Systems.” Quincy, Massachusetts, 1996.

(NRC, 1977) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.32, Revision 2,

“Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants.” Washington,

DC, February 1977.

(NRC, 1978) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.75, Revision 2,

“Physical Independence of Electric Systems.” Washington, DC, September 1978.

(NRC, 1988) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.100, Revision 2,

“Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants.” 

Washington, DC, June 1988.
11-107

(NRC, 1993) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.9, Revision 3,

“Selection, Design, Qualification, and Testing of Emergency Diesel Generator Units Used as

Class 1E Onsite Electric Power Systems at Nuclear Power Plants.” Washington, DC, July 1993.

(NRC, 1994) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.118, Revision 3,

“Periodic Testing of Electric Power and Protection Systems.” Washington, DC, September

1994.

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility.” Washington,

DC, 2000.

11.6 Instrumentation and Control Systems

11.6.1 Conduct of Review

This chapter of the FSER contains the staff’s review of I&C systems described by the applicant

in Chapter 11 of the revised CAR (DCS, 2002b). The objective of this review is to determine

whether the I&C PSSCs and their design bases identified by the applicant provide reasonable

assurance of protection against natural phenomena and the consequences of potential

accidents. The staff evaluated the information provided by DCS for I&C systems by reviewing

Chapter 11 of the revised CAR, other sections of the revised CAR, supplementary information

provided by the applicant, and relevant documents available at the applicant’s offices but not

submitted by the applicant. The review of I&C system design bases and strategies was closely

coordinated with the review of the electrical and I&C aspects of accident sequences described in

the safety assessment of the design bases (see Chapter 5 of this FSER) and the review of other

plant systems. 

The staff reviewed how the I&C information in the revised CAR addresses the following

regulations:

• 10 CFR 70.23(b) states, as a prerequisite to construction approval, that the design bases

of the PSSCs and the Quality Assurance Program be found to provide reasonable

assurance of protection against natural phenomena and the consequences of potential

accidents. 

• 10 CFR 70.64 requires that BDC and defense-in-depth practices be incorporated into the

design of new facilities. With respect to I&C, 10 CFR 70.64(a)(10) requires that the

MFFF design “provide for inclusion of instrument and control systems to monitor and

control the behavior of items relied on for safety.”

The review for this construction approval focused on the design basis of I&C systems, their

components, and other related information. For each I&C system, the staff reviewed information

provided by the applicant for the safety function, system description, and safety analysis. The

review also encompassed proposed designbasis considerations such as redundancy,

independence, reliability, and quality. The staff used Chapter 11 in NUREG-1718, “Standard

Review Plan for the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility,”

issued 2000, as guidance in performing the review.
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11.6.1.1 System Description

The facility I&C systems include the MP and AP process control, utility control, and emergency

control systems. 

11.6.1.1.1 Function

The I&C systems (as described in Section 11.6, “Instrumentation and Control Systems,” of the

revised CAR) are to monitor and control the manufacturing process systems, the plant utility

systems, and the plant safety and emergency systems. Specifically, I&C systems are to monitor

and control plant parameters during normal and transient conditions to ensure the quality of the

produced product and ensure that limits (including safety limits) are not exceeded. They also

provide control signals to plant equipment to prevent the occurrences of faulted conditions and

mitigate the consequences of faulted conditions should they occur.

11.6.1.1.2 Major Components

11.6.1.1.2.1 MP and AP Process Control Systems. The MP and AP process control systems

use an automated, distributed processing control system strategy with the manufacturing

process translated into control algorithms for each process step. The systems include the

normal, protective, and safety control subsystems which ensure the final product conforms to

specifications and reduce plant waste and risk. Specifically, the normal control subsystem

controls the manufacturing process, the protective control subsystem protects personnel

(industrial safety) and equipment, and the safety control subsystem ensures that safety limits will

not be exceeded and that undesirable operational conditions are prevented or mitigated. 

The process control systems are built around PLCs which, in turn, are built around softwarecontrolled microprocessors. There are two independent safety channels in the safety control

subsystem, both of which are separate and independent from the other two control subsystems.

All channels in the safety control subsystem in the MP or AP process control system will be

designated PSSCs. The design basis for the safety control subsystem is described in FSER

Section 11.6.1.3.1 for a PSSC.

11.6.1.1.2.2 Utility Control Systems. The utility control systems also use a PLC/

microprocessor-based automated, distributed processing control strategy and include separate

normal, protective, and auxiliary control subsystems which ensure that the plant support (utility)

systems operate within specifications. Specifically, the normal control subsystem controls

operation of the support systems, the protective control subsystem protects personnel and

equipment, and the auxiliary control subsystem is a backup to the normal control system and

prevents challenges to the emergency control system. The single channels of the normal and

auxiliary control subsystems are separate and independent from each other and the protective

control subsystem. The auxiliary control subsystems that are part of the utility control systems

are not credited by DCS towards meeting the performance requirements of 10 CFR Part 70,

“Domestic Licensing of Special Nuclear Material,” and have not been identified as PSSCs.

11.6.1.1.2.3 Emergency Control System. The emergency control system is a PSSC and

consists of manual and some automatic controls for power, ventilation, and required safety

functions. Specifically, the emergency control system ensures particular plant support systems
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operate when needed to mitigate the consequences of hazardous occurrences. The system is a

hard-wired system with no software control and is designed to continue to operate during and

following specific design basis events. It is divided into two separate and independent trains,

each with a separate control room.

11.6.1.1.2.4 Data Communications Networks. The data communications networks employ

Ethernet technology and provide for communications among various systems and components

such as normal controllers, workstations, the manufacturing management and information

system (MMIS), the manufacturing status system, and the computer-aided diagnosis system. 

For the MP and AP process control systems, Fieldbus technology or hard-wired methods are

used to link normal controllers to motor control centers and sensor/instruments. The X-terminal

network (XTN) links display terminals, workstations, the MMIS, manufacturing status system,

and the computer-aided diagnosis system. The local industrial network (LIN) links workstations,

controllers, the MMIS, manufacturing status system, and the computer-aided diagnosis system. 

The immediate control network (ICN) links the workstations and normal controllers. Separate

connections are used to link each device to each network.

For the utility control systems, Fieldbus technology or hard-wired methods are used to link

normal controllers to MCCs and sensor/instruments. Dual redundant networks link the

workstations and normal controllers. Separate connections are used to link each device to each

network. 

11.6.1.1.2.5 Control Rooms. The facility includes various control rooms with workstations

(industrial personal computers) to monitor and control the operation of the AP and MP functional

units and utility system. Access to the control system data communication network is provided in

the control rooms, and emergency stop switches typically provide manual control capability. 

Human factors engineering principles are used in the design of the control rooms (see FSER

Chapter 12). 

The major control rooms are discussed in the following sections:

• AP and MP control rooms

The AP and MP control rooms provide central locations for monitoring, supervising, and

controlling the manufacturing and processing operations. In these control rooms,

operators access production control information to verify that process automation is

performing satisfactorily or to receive notification of a problem. Video displays of the

actual functional unit conditions are provided. If an AP or MP manufacturing process

functional unit is not operating, manning of the associated control room, for the particular

manufacturing process steps, is not required.

In addition to providing for the control of AP systems, the AP control room is always

manned and provides for the control of the normal and safety utilities systems, the fire

detection systems, and the health physics systems. 
11-110

• utilities control room

In addition to the controls for the utilities systems located in the AP control room, the

utilities control room provides support system monitoring and controlling capability when

the controls in the AP control room for the utilities systems are unavailable. Support

systems monitored and controlled from these two redundant locations include the HVAC

systems, the electrical distribution system, and various other process support systems

such as the hot and chilled water system.

• emergency control rooms

Two separate emergency control rooms (Train A and Train B) each with an independent

ventilation system are located in the shipping and receiving building. Solid state or

traditional electromechanical control devices in these control rooms are hard wired to

PSSCs in the emergency control system. These controls have priority over the normal,

auxiliary, and protective controls. 

11.6.1.1.2.6 Sensors. The facility functional units are provided with sensors and instruments

(parts of the process control systems) to monitor and measure operational conditions and

parameters such as temperature, pressure, mass, component identification bar codes, machine

tool and valve positions, etc. Some of the information is provided as inputs for the automatic

control of actuators (fans, pumps, heaters, conveyers, etc.) and other devices which perform

steps in the manufacturing process.

The utility control systems are also provided with sensors and instruments to monitor the

performance of support systems such as the electrical power and HVAC systems. These

sensors also provide input signals for the automatic controllers for the various utility control

systems.

11.6.1.1.2.7 Controllers. The AP process, MP, and the utility control systems have controllers

which receive inputs from sensors and instruments. Specific controllers are discussed below:

• normal controllers

The PLCs provide for the control of normal operations of the functional and utility units. 

These normal controllers, located in rooms close to the process they control, contain

communication devices and are connected to the control system networks so that data

flow between the controllers and the distributed input/output, the normal controllers of

other functional and utility units, the workstations, the MMIS, and the computer-aided

diagnosis system. This data communication between controllers and systems allows

coordination of activities. These controllers are not PSSCs. 

• protective controllers 

The protective controllers are dedicated controllers that protect operations personnel

from injury and functional unit equipment from damage resulting from inappropriate

operation such as overspeed, overtorque, and overtravel. These controllers are

autonomous and are not connected to the control system networks. They may be PLCs

or traditional electromechanical relays and are physically located in or near MCCs. 
11-111

Protective sensors are hard-wired to the controllers, with output control signals hardwired

to control circuits in MCCs or power panels. Protective controllers have priority over

control signals from the normal controllers and provide performance data to the normal

controllers which, in turn, provide data to workstations via the ICN. Operators cannot

directly access the protective controllers and cannot routinely intervene in their operation. 

These controllers are not PSSCs.

• safety controllers

Safety controllers in the process control of the functional units serve specific and limited

functions for preventing and mitigating certain accidents. These controllers are

autonomous and are not connected to the control system network. Performance data

from the safety controllers are provided to the normal controllers which, in turn, provide

data to workstations via the ICN. Any safety controller in the MP or AP process control

system may be designated a PSSC as a result of the facility safety analysis. The design

basis for that controller will then be in accordance with FSER Section 11.6.1.3.1. 

The safety controllers for the functional units may be PLCs or traditional relays. They are

physically located in rooms separate from the normal controllers and are hard-wired to

the sensors and actuator control circuits in MCC or power distribution panels. The

control signals from the safety controllers have priority over signals from the normal

controllers. If a safety controller does not detect appropriate functional unit conditions, it

will block commands issued by the normal controllers. Operators cannot directly access

the safety controllers and cannot routinely intervene in their operation with the exception

of the scrap jar isotopic concentration data which are manually loaded into the safety

controllers. 

Upon receipt of a fire condition signal from the facility fire detection system, fire safety

controllers direct the normal controllers to immediately close the fire doors. Following a

delay, the fire safety controller will close the fire doors if the normal controllers fail to take

action. The fire barrier is a PSSC.

• auxiliary controllers

Auxiliary controllers in selected utility systems provide backup monitoring and control

capability if the normal control systems are unavailable. They are also installed in

separate locations from the normal controllers and are hard-wired to dedicated

workstations located in the AP control room. Auxiliary controllers in the utility control

system have not been identified as PSSCs. 

• emergency controllers

The facility emergency controllers in the emergency control system are solid-state

controllers or traditional relay logic circuits with control switches located on emergency

control panels in the emergency control rooms. Inputs to emergency controllers from the

sensors and outputs from the controllers to MCCs are hard wired. The emergency

controllers provide manual control for selected systems. These controllers are PSSCs.
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11.6.1.1.2.8 MMIS. The MMIS is a realtime system that tracks the product inventory and

maintains/updates the manufacturing status system database files as the product moves

through the manufacturing process. It tracks quantities of product that are allowed in any area

at a given time and authorizes the movement of product into or out of a functional unit. The

MMIS also is a server for the XTN.

11.6.1.1.2.9 Manufacturing Status System. The manufacturing status system is identical to the

MMIS and maintains a copy of the MMIS database. The system is used to sort, analyze, and

report on data collected by the MMIS.

11.6.1.1.2.10 Computer-Aided Diagnosis System. The computer-aided diagnosis system

monitors the operation of the normal controllers for the function units and the ICN. The system

is independent from the controllers and has its own software. It provides a primary diagnostic

capability when there is a problem with the production process and works with the software

used in the PLCs to determine the state of the PLC when the problem occurred. 

11.6.1.1.2.11 Process Computers. Process computers are dedicated microprocessor-based

computers that control and handle data for the more complex process measurement systems

such as the laser optical micrometer. These computers operate the measurement systems,

provide data signal conditioning, and send data to the PLCs controlling the associated functional

unit.

11.6.1.1.2.12 Seismic Monitoring System. The seismic monitoring system provides data for the

evaluation of the confinement structure and other PSSCs and automatically shuts down the MP

and AP process during a high seismic event. 

11.6.1.2 System Interfaces

The facility I&C systems are primarily electronic systems that interface with mechanical,

electrical, and process systems. The I&C systems provide signals to control electromechanical

and pneumatic actuators and data signals for information display, processing, and storage. The

control systems and their associated components, such as electromechanical actuators, are

supplied electrical power by the facility electrical systems. The pneumatic actuators are

provided air by either the instrument air or plant air systems. The various control rooms provide

the primary human-system interface for all the facility systems. Environmental conditions for

proper operation of the I&C systems are maintained by the HVAC systems, and protection

against fires is provided by the fire protection systems.

11.6.1.3 Design Bases of the PSSCs and Baseline Design Criteria

For design basis requirements, the I&C systems are to be available and reliable for normal

operation and safe shutdown and monitoring (see FSER Section 11.6.1.1.1 for a discussion of

the function performed by the I&C systems). With respect to I&C, 10 CFR 70.64(a)(10) requires

that the facility design “provide for inclusion of instrument and control systems to monitor and

control behavior of items relied on for safety.” The I&C systems designated as PSSCs are

designed to remain functional when subjected to severe natural phenomena and environmental

and dynamic effects, consistent with the baseline design criteria in 10 CFR 70.64(a)(2) and

10 CFR 70.64(a)(4), respectively, and be adequately protected from fires per
11-113

10 CFR 70.64(a)(3). Additionally, the I&C systems must support continued operation of

essential utility services, consistent with the baseline design criterion in 10 CFR 70.64(a)(7).

Pursuant to 10 CFR 70.64(b), the I&C systems are to be designed using defense-in-depth

practices, with a preference for engineered controls over administrative controls. 

To ensure that the design basis requirements and the baseline design criteria are met, DCS has

committed to specific industry standards and staff guidance as discussed in the following

section.

11.6.1.3.1 I&C Systems (PSSCs)

The I&C systems identified by DCS as PSSCs are the emergency control system and the AP

and MP safety control subsystems (DCS, 2002b). These PSSCs will be designed with

provisions for periodic testing, redundancy, independence, no single-failure vulnerability, failsafe failure mode, proper instrument spans/setpoints/control ranges, status monitoring, and

qualification for natural phenomena and environmental and dynamic effects (see FSER Section

11.11). Additionally, these systems will be designed using guidance (to the extent described in

the revised CAR and DCS letters dated August 31, 2001, December 5, 2001, January 7, 2002,

November 22, 2002, and February 11, 2003 (DCS, 2001a, 2001b, 2002a, 2002c, 2003) from the

documents listed below.

• for overall system design (including seismic monitoring and trip system):

— IEEE Std 603-1998, “IEEE Standard Criteria for Safety Systems for Nuclear

Power Generating Stations,” issued 1998 (IEEE, 1998f)

• for single failure:

— IEEE Std 379-1994 (IEEE, 1994a)

• for software programmable electronic systems:

— Electric Power Research Institute (EPRI) TR-106439, “Guideline on Evaluation

and Acceptance of Commercial Grade Digital Equipment for Nuclear Safety

Applications,” issued October 1996 (EPRI, 1996)

— International Electrotechnical Commission (IEC) 61131-3 (1993-03),

“Programmable Controllers -Part 3: Programming Languages,” issued 1993 (IEC,

1993) 

— IEEE Std 7-4.3.2-1993, “IEEE Standard Criteria for Digital Computers in Safety

Systems of Nuclear Power Generating Stations,” issued 1993 (IEEE, 1993)

— IEEE Std 730-1998, ”Software Quality Assurance Plans,” issued 1998 (IEEE,

1998e) 

— IEEE Std 828-1998, “IEEE Standard for Software Configuration Management

Plans,” issued 1998 (IEEE, 1998d)
11-114

— IEEE Std 830-1998,“IEEE Standard Recommended Practice for Software

Requirements Specifications,” issued 1998 (IEEE, 1998c)

— IEEE Std 1012-1998, “IEEE Standard for the Software Verification and

Validation,” issued 1998 (IEEE, 1998b)

— IEEE Std 1028-1997, “IEEE Standard for Software Reviews,” issued 1997 (IEEE,

1997a)

— IEEE Guide 1042-1987, ”Software Configuration Management,” issued 1987

(IEEE, 1987c) 

— IEEE Std 1074-1997, “IEEE Standard for Developing Software Life Cycle

Processes,” issued 1997 (IEEE, 1997b)

— IEEE Std 1228-1994, “IEEE Standard for Software Safety Plans,” issued 1994

(IEEE, 1994b)

— NUREG/CR-6090, “The Programmable Logic Controller and Its Application in

Nuclear Power Plants,” issued February 1999 (NRC, 1999a)

— NUREG/CR-6463, “Review Guidelines on Software Languages for Use in Nuclear

Power Plant Safety Systems,” issued June 1996 (NRC, 1996)

— RG 1.168, “Verification, Validation, Reviews, and Audits for Digital Computer

Software Used in Safety Systems of Nuclear Power Plants,” issued September

1997 (NRC, 1997a)

— RG 1.169, Configuration Management Plans for Digital Computer Software Used

in Safety Systems of Nuclear Power Plants,” issued September 1997 (NRC,

1997b)

— RG 1.172, “Software Requirements Specifications for Digital Computer Software

Used in Safety Systems of Nuclear Power Plants,” issued September 1997 (NRC,

1997c)

— RG 1.173, “Developing Software Life Cycle Processes for Digital Computer

Software Used in Safety Systems of Nuclear Power Plants,” September 1997

(NRC, 1997d)

— Safety Evaluation by the Office of Nuclear Reactor Regulation, “EPRI Topical

Report TR-106439,” issued May 1997 (NRC, 1997e)

• for electrical independence and separation:

— IEEE Std 384-1992 (IEEE, 1992)
11-115

— NUREG-0800, Standard Review Plan, Branch Technical Position HICB-11,

“Guidance on the Application and Qualification of Isolation Devices,” issued June

1997 (NRC, 1997f)

— RG 1.75, Revision 2 (NRC, 1978)

• for equipment SQ:

— IEEE Std 344-1987 (IEEE, 1987b)

— RG 1.100, Revision 2 (NRC, 1988) 

• for setpoints:

— ANSI/ISA-67.04.01-2000 (ANSI, 2000a)

— RG 1.105, Revision 3, “Setpoints for Safety-Related Instrumentation,” issued

December 1999 (NRC, 1999b).

• for human-system interface:

— IEEE Std 1023-1988, “IEEE Guide for the Application of Human Factors

Engineering to Systems, Equipment, and Facilities of Nuclear Power Generating

Stations,” issued 1988 (NRC, 1988)

— NUREG-0700, Revision 2, “Human System Design Review Guidelines,” issued

May 2002 (NRC, 2002) 

• for the seismic monitoring and trip system (recording function):

— RG 3.17-1974, “Earthquake Instrumentation for Fuel Reprocessing Plants,”

issued 1974 (NRC, 1974)

• for periodic testing:

— IEEE Std 338-1987 (IEEE, 1987a)

— NUREG-0800, Standard Review Plan, Branch Technical Position HICB-17,

“Guidance on Self-Test and Surveillance Test Provisions,” issued June 1997

(NRC, 1997g)

— RG 1.118, Revision 3 (NRC, 1995)

• for reduction of electromagnetic and radio frequency interference:

— IEEE Std 518-1982, “IEEE Guide for the Installation of Electrical Equipment to

Minimize Electrical Noise Inputs to Controllers from External Sources,” issued

1982 (IEEE, 1982)
11-116

— IEEE Std 1050-1996, “Guide for Instrumentation and Control Equipment

Grounding in Generating Stations,” issued 1996 (IEEE, 1996)

— RG 1.180, “Guidelines for Evaluating Electromagnetic and Radio-Frequency

Interference in Safety-Related Instrumentation and Control Systems,” issued

January 2000 (NRC, 2000)

• for design of data communications networks:

 — ANSI/IEEE 802.3 Standards Series, “IEEE Standards for Local Area Networks: 

Carrier Sense Multiple Access with Collision Detection (CSMA/CD) Access

Method and Physical Layer Specifications,” issued 2000 (ANSI, 2000b)

• for design of combustible gas detectors:

— ISA-S12.13-Part 1-1995 (ISA, 1995)

— ISA RP12.13-Part II-1987 (ISA, 1987)

In Section 5.5.5 of the revised CAR, the applicant described its general design philosophy used

in formulating the preliminary design of the facility. Pursuant to 10 CFR 70.64(b), in order to

ensure that engineered controls are relied upon over administrative controls, to the extent

practicable, DCS has established a hierarchy of controls. In further adherence to

10 CFR 70.64(b), DCS stated that it has incorporated defense-in-depth practices in its

preliminary facility design. Defense in depth is defined in the 10 CFR 70.64 footnote. In Section

5.5.5.2 of the revised CAR, DCS stated that it has incorporated defense-in-depth practices

through use of the single-failure criterion. Under this criterion, PSSCs are required to be

capable of carrying out their functions in the event that any single active component fails,

whether such failure occurs within the applicable system, or in an associated system that

supports the component’s operation. 

Accordingly, the facility I&C systems are to be designed using defense-in-depth practices using

the single-failure criterion including redundancy, independence, separation, and fail safe for the

I&C systems identified as PSSCs through use of the industry standards listed above. As an

example, the emergency control system is designed with redundant, separate, and independent

trains with fail safe failure modes (see FSER Section 11.6.1.1.2.3).

11.6.2 Evaluation Findings

In Section 11.6.7 of the revised CAR, DCS provided design basis information for I&C systems

that it identified as PSSCs for the facility. Based on the staff’s review of the revised CAR and

supporting information provided by the applicant and the applicant’s commitments to the industry

standards and guidance discussed in the sections above for I&C systems, the staff finds that

DCS has met the BDC set forth in 10 CFR 70.64(a)(10). The staff concludes, pursuant to

10 CFR 70.23(b), that the design bases of the PSSCs evaluated in this FSER section will

provide reasonable assurance of protection against natural phenomena and the consequences

of potential accidents. The staff further concludes, from its review of the revised CAR and

supporting information submitted by DCS, that the preliminary design of the facility I&C systems
11-117

meet the defense-in-depth provisions and the preference for engineered controls over

administrative controls as stated in 10 CFR 70.64(b).

11.6.3 References

(ANSI, 2000a) American National Standards Institute/Instrument Society of America. 

67.04.01-2000, “Setpoints for Nuclear Safety-Related Instrumentation.” Research Triangle Park,

North Carolina, 2000.

(ANSI, 2000b) American National Standards Institute/Institute of Electrical and Electronics

Engineers. 802.3 Standards Series, “IEEE Standards for Local Area Networks: Carrier Sense

Multiple Access with Collision Detection (CSMA/CD) Access Method and Physical Layer

Specifications.” New York, 2000.

(DCS, 2001a) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Response to Request for Additional Information (DCS-NRC-000059). August

31, 2001.

(DCS, 2001b) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Clarification of Responses to NRC Request for Additional Information

(DCS–NRC–000074). December 5, 2001.

(DCS, 2002a) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Clarification of Responses to NRC Request for Additional Information. 

January 7, 2002.

(DCS, 2002b) Ihde, R., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Mixed Oxide Fuel Fabrication Facility Construction Authorization Request

Revision (DCS-NRC-000114). October 31, 2002.

(DCS, 2002c) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Requests for Additional Information, Clarifications, and Open Item Mapping

into the Construction Authorization Request Revision (DCS-NRC-000120). November 22, 2002.

(DCS, 2003) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Responses to Site Description, Instrumentation, and Criticality Open

Items/Additional NRC Questions on Construction Authorization Request (CAR) Revision. 

February 11, 2003.

(EPRI, 1996) Electric Power Research Institute. TR-106439, “Guideline on Evaluation and

Acceptance of Commercial Grade Digital Equipment for Nuclear Safety Applications.” Palo Alto,

California, October 1996.

(IEC, 1993) International Electrotechnical Commission. 61131-3 (1993-03), “Programmable

Controllers—Part 3: Programming Languages.” Geneva, Switzerland, 1993.

(IEEE, 1982) Institute of Electrical and Electronics Engineers. Std 518-1982, “IEEE Guide for

the Installation of Electrical Equipment to Minimize Electrical Noise Inputs to Controllers from

External Sources.” New York, 1982.
11-118

(IEEE, 1987a) Institute of Electrical and Electronics Engineers. Std 338-1987, “IEEE Standard

Criteria for Periodic Testing of Nuclear Power Generating Station Class 1E Power and

Protection Systems.” New York, 1987.

(IEEE, 1987b) Institute of Electrical and Electronics Engineers. Std 344-1987, “IEEE

Recommended Practices for Seismic Qualification of Class 1E Equipment for Nuclear

Generating Stations.” New York, 1987.

(IEEE, 1987c) Institute of Electrical and Electronics Engineers. Guide 1042-1987, “Software

Configuration Management.” New York, 1987.

(IEEE, 1988) Institute of Electrical and Electronics Engineers. Std 1023-1988. “IEEE Guide for

the Application of Human Factors Engineering to Systems, Equipment, and Facilities of Nuclear

Power Generating Stations.” New York, 1998.

(IEEE, 1992) Institute of Electrical and Electronics Engineers. Std 384-1992, “Standard Criteria

for Independence of Class 1E Equipment and Circuits.” New York, 1992.

(IEEE, 1993) Institute of Electrical and Electronics Engineers. Std 7-4.3.2-1993, “IEEE

Standard Criteria for Digital Computers in Safety Systems of Nuclear Power Generating

Stations.” New York, 1993.

(IEEE, 1994a) Institute of Electrical and Electronics Engineers. Std 379-1994, “IEEE Standard

Application of the Single Failure Criterion to Nuclear Power Generation Station Safety Systems.” 

New York, 1994.

(IEEE, 1994b) Institute of Electrical and Electronics Engineers. Std 1228-1994, “IEEE Standard

for Software Safety Plans.” New York, 1994.

(IEEE, 1996) Institute of Electrical and Electronics Engineers. Std 1050-1996, “Guide for

Instrumentation and Control Equipment Grounding in Generating Stations.” New York, 1996.

(IEEE, 1997a) Institute of Electrical and Electronics Engineers. Std 1028-1997, “IEEE Standard

for Software Reviews.” New York, 1997.

(IEEE, 1997b) Institute of Electrical and Electronics Engineers. Std 1074-1997, “IEEE Standard

for Developing Software Life Cycle Processes.” New York, 1997.

(IEEE, 1998b) Institute of Electrical and Electronics Engineers. Std 1012-1998, “IEEE Standard

for the Software Verification and Validation.” New York, 1998.

(IEEE, 1998c) Institute of Electrical and Electronics Engineers. Std 830-1998, “IEEE Standard

Recommended Practice for Software Requirements Specifications.” New York, 1998.

(IEEE, 1998d) Institute of Electrical and Electronics Engineers. Std 828-1998, “IEEE Standard

for Software Configuration Management Plans.” New York, 1998.

(IEEE, 1998e) Institute of Electrical and Electronics Engineers. Std 730-1998, “Software Quality

Assurance Plans.” New York, 1998.
11-119

(IEEE, 1998f) Institute of Electrical and Electronics Engineers. Std 603-1998, “IEEE Standard

Criteria for Safety Systems for Nuclear Power Generating Stations.” New York, 1998.

(ISA, 1987) Instrument Society of America. ISA RP12.13-Part II-1987, “Installation, Operation,

and Maintenance of Combustible Gas Detection Instruments.” 1987.

(ISA, 1995) Instrument Society of America. ISA-S12.13-Part 1-1995, “Performance

Requirements, Combustible Gas Detectors.” 1995.

(NRC, 1974) U.S. Nuclear Regulatory Commission. Regulatory Guide 3.17-1974, “Earthquake

Instrumentation for Fuel Reprocessing Plants.” Washington, DC, 1974.

(NRC, 1978) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.75, Rev. 2, “Physical

Independence of Electric Systems.” Washington, DC, September 1978.

(NRC, 1988) U.S. Nuclear Regulatory Commission (NRC). Regulatory Guide 1.100, Rev. 2,

“Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants.” 

Washington, DC, June 1988.

(NRC, 1995) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.118, Rev. 3, “Periodic

Testing of Electric Power and Protection Systems.” Washington, DC, April 1995.

(NRC, 1996) U.S. Nuclear Regulatory Commission. NUREG/CR-6463, “Review Guidelines on

Software Languages for Use in Nuclear Power Plant Safety Systems.” Washington, DC, June

1996. 

(NRC, 1997a) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.168, “Verification,

Validation, Reviews, and Audits for Digital Computer Software Used in Safety Systems of

Nuclear Power Plants.” Washington, DC, September 1997.

(NRC, 1997b) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.169, “Configuration

Management Plans for Digital Computer Software Used in Safety Systems of Nuclear Power

Plants.” Washington, DC, September 1997.

(NRC, 1997c) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.172, “Software

Requirements Specifications for Digital Computer Software Used in Safety Systems of Nuclear

Power Plants.” Washington, DC, September 1997.

(NRC, 1997d) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.173, “Developing

Software Life Cycle Processes for Digital Computer Software Used in Safety Systems of Nuclear

Power Plants.” Washington, DC, September 1997.

(NRC, 1997e) U.S. Nuclear Regulatory Commission. Safety Evaluation by the Office of Nuclear

Reactor Regulation, “EPRI Topical Report TR-106439.” Washington, DC, May 1997.

(NRC, 1997f) U.S. Nuclear Regulatory Commission. NUREG-0800, Standard Review Plan,

Branch Technical Position HICB-11, “Guidance on the Application and Qualification of Isolation

Devices.” Washington, DC, 1997. 
11-120

(NRC, 1997g) U.S. Nuclear Regulatory Commission. NUREG-0800, Standard Review Plan,

Branch Technical Position HICB-17, “Guidance on Self-Test and Surveillance Test Provisions.” 

Washington, DC, 1997.

(NRC, 1999a) U.S. Nuclear Regulatory Commission. NUREG/CR-6090, “The Programmable

Logic Controller and Its Application in Nuclear Power Plants.” Washington, DC, February 1999.

(NRC, 1999b) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.105, Revision 3,

“Setpoints for Safety-Related Instrumentation.” Washington, DC, December 1999.

(NRC, 2000) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.180, “Guidelines for

Evaluating Electromagnetic and Radio-Frequency Interference in Safety-Related Instrumentation

and Control Systems.” Washington, DC, January 2000.

(NRC, 2001) U.S. Nuclear Regulatory Commission. NUREG/CR-6597, “Results and Insights on

the Impact of Smoke on Digital Instrumentation and Control.” Washington, DC, January 2001. 

(NRC, 2002) U.S. Nuclear Regulatory Commission. NUREG-0700, Rev. 2, “Human System

Design Review Guidelines.” Washington, DC, May 2002.

11.7 Material Transport Systems

11.7.1 Conduct of Review

This chapter of the FSER contains the staff’s review of the material transport systems described

by the applicant in Chapter 11.0 of the revised CAR (DCS, 2002a). The objective of this review

is to determine whether the material transport systems’ PSSCs and their design bases identified

by the applicant provide reasonable assurance of protection against natural phenomena and the

consequences of potential accidents. The staff evaluated the information provided by DCS for

material transport systems by reviewing Chapter 11.0 of the revised CAR, other sections of the

revised CAR, supplementary information provided by the applicant, and relevant documents

available at the applicant’s offices but not submitted by the applicant. The review of material

transport systems’ design bases and strategies was closely coordinated with the review of

accident sequences described in the safety assessment of the design bases (see Chapter 5 of

this FSER) and the review of other plant systems. 

The staff reviewed how the information in the revised CAR addresses the following regulation:

• 10 CFR 70.23(b) states, as a prerequisite to construction approval, that the design bases

of the PSSCs and the Quality Assurance Program must be found to provide reasonable

assurance of protection against natural phenomena and the consequences of potential

accidents.

The review of the revised CAR focused on the design bases of material transport systems, their

components, and other related information. For material transport systems, the staff reviewed

information provided by the applicant for the safety function, system description, and safety

analysis. The review also encompassed proposed design basis considerations such as

redundancy, independence, reliability, and quality. The staff used Chapter 11.0 in NUREG-1718

(NRC, 2000) as guidance in performing the review.
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11.7.1.1 System Description

Revised CAR Section 11.7 describes the functional requirements and design bases for

equipment designed to transfer MOX fuel production material that is in a dry, solid form. 

Examples of such forms are plutonium dioxide and uranium oxide powders, master blends of

MOX powder, production batches of MOX powder, green pellets, sintered pellets, and fuel

rods/assemblies. The equipment described in this section is located inside the MP area and

shipping and receiving area. Because of the nature of the equipment (i.e., it is inside and

attached to, or supported by, gloveboxes), certain parts of this review relate to gloveboxes. The

staff’s review of the confinement system and gloveboxes can be found in Section 11.4 of this

FSER. Descriptions of the MP and AP process are provided in revised CAR Sections 11.2 and

11.3, respectively, and are discussed in the corresponding FSER sections. A description of the

process control system is given in FSER Section 11.6.

Different material-handling equipment is used to transport material in the facility, depending on

the form of the product, the container used to carry it, and the configuration of the process

equipment that receives the container. Fuel material in the MP is in one of five forms—powder,

pellets, rods, assemblies, or waste. The material-handling equipment is used in (1) the receipt

and opening of plutonium dioxide and uranium dioxide containers in the powder area, (2) pellet

processing, (3) fuel rod and fuel assembly fabrication, (4) fuel assembly inspection and storage,

(5) fresh fuel cask loading, and (6) loading the fresh fuel casks onto over-the-road trailers.

Material-handling equipment that can be used during these processes includes, but is not limited

to, scales, pallet trucks, fork lifts, drum-tilting devices, storage frames, handling monorails,

pouring stations, feeding lines and control valves, hoppers, monorail cranes, standard and

vibrating conveyors, turntables, traveling cranes, bridge cranes, gloveboxes, storage arrays,

pneumatic transfer stations, airlocks, hoppers, impactors, funnels, can opening/closing devices,

support frames, elevators, clamping devices, grippers, inspection stands, cleaning stations, jars,

molybdenum boats, sintering furnaces, three-dimensional stackers, trolleys, winches, transfer

tunnels, grinders, tray stackers, tilting tables, and air pallets. Various containers are also

included in the list of material-handling equipment, including DOE Standard 3013 (DOE, 2000)

containers and transport casks, transfer containers, waste containers, and MOX fresh fuel

casks.

DOE Standard 3013 containers provide primary and secondary confinement for plutonium

received at the facility. This standard applies to plutonium-bearing metals and oxides containing

at least 30 wt% plutonium and uranium, as well as to plutonium-oxide materials with significant

chloride contamination. The 3013 container is made up of an outer can, inner can, and

convenience can. The outer and inner cans make up the primary and secondary containment,

respectively.

Waste containers will be used to hold and ship MOX transuranic wastes. The waste transfer

containers hold waste drums that provide the primary and secondary confinement for the waste. 

The drums are bag-lined before loading and sealed with a gasketed cover. The drums are also

provided with filters to prevent pressurization and prevent the release of wastes from the drum.

MOX fresh fuel casks will contain multiple fresh MOX fuel assemblies for shipping. The fresh

fuel casks will be qualified to meet the requirements of 10 CFR Part 71, “Packaging and
11-122

Transportation of Radioactive Material.” The NRC safety review of the fresh fuel casks for

transportation is being performed separately from the facility review.

11.7.1.1.1 Function

The functions that the material transport systems are designed to perform include the following:

• transferring MOX fuel material and components from one point in the process to another,

in accordance with process throughput, positioning tolerance, mechanism reliability, and

radiological shielding requirements

• maintaining structural integrity and control of process containers to ensure that the

confinement boundary is not breached

• maintaining structural integrity and control of process containers to ensure that criticality

control functions are performed

• working with fire barriers, as required, to transfer material across process atmosphere or

fire barrier boundaries

• transferring tooling and equipment spare parts during maintenance operations from point

to point within the glovebox system

The material-handling equipment operation during those processes begins when

the plutonium powder arrives in a DOE safe, secure trailer at the shipping and receiving area at

the facility. The 3013 containers, on a shipping pallet, are transferred from the truck to the

shipping bay laydown area by forklift. The pallet is unpacked on a turntable and transported by

roller conveyor to the 3013 storage area. Likewise, for the depleted uranium dioxide receiving

and storage unit, uranium dioxide is delivered to the secured warehouse building in palletized

drums. From there, the drums are sent to the MOX process area and are staged in a buffer

area near the uranium dioxide drum-emptying room. The 3013 containers are transferred to the

transfer cask opening area to remove the overpack. A hoist lifts the 3013 package (of

approximately 9.1 kg (20 lb)) onto a small roller conveyor. From there, the plutonium

dioxide/3013 storage crane transfers the 3013 package to plutonium dioxide/3013 storage racks. 

When removed from storage, the package rides by conveyor to the decanning unit, which is fully

enclosed in a glovebox. Inside this glovebox, the 3013 can is moved both horizontally and

vertically, and the outer can is removed. The inner can is transferred by pneumatic transfer tube

to Level 4 of the AP building where the inner can is opened. Following this operation, the

convenience can is opened. The opened convenience can is rotated and emptied into a

homogenizer located in a glovebox immediately below the can opening glovebox. From there,

the homogenized plutonium is transferred by pneumatic lift to the electrolyzer on Level 3 of the

AP building. The electrolyzer marks the beginning of the AP chemical processing of the

plutonium, which is discussed in detail in Sections 8 and 11.2 of this FSER.

Following AP chemical processing, the “wet” MOX material is returned to the MP. The MOX

material will remain within gloveboxes from this point until it emerges in a completed, sealed fuel

rod. The MP is described and evaluated in Section 11.3 of this FSER.
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Table 11.7-1 PSSCs for Loss of Confinement/Dispersal

of Nuclear Materials Events

Loss of Confinement/

Dispersal of Nuclear Material

Events Related to MaterialHandling Equipment

Identified PSSC

For the protection of the...

Facility

Worker

Site

Worker

*

IOC* Environment

Corrosion Material Maintenance

& Surveillance

Program†

U - - U

Small breaches in glovebox

boundary or backflow from utility

lines

C4 confinement

system

U - - U

Leaks of AP process vessels or

pipes within process cells

Process cell U - - -

Process cell entry

controls

U - - -

Process cell exhaust

system

- U U U

Rod-handling operations

 

Material-handling

equipment**

U - - -

Material-handling

controls**

U - - -

Facility worker action U - - -

Breaches in containers outside

gloveboxes caused by handling

operations in C2 areas

3013 canister** U U U U

Transfer container** U U U U

Material-handling

controls**

U U U U

Breaches in containers outside

gloveboxes caused by handling

operations in C3 areas

3013 canister** U - - -

Transfer container** U - - -

Facility worker

controls

U - - -

Material-handling

controls**

U - - -

C3 confinement

system

- U U U

* There may be confinement systems or barriers not listed in this table that provide defense-indepth protection for the site worker, IOC, or the environm ent for which no credit is technically

being taken by DCS.

** These items are material transport system PSSCs evaluated in this section of the FSER. 

The rem aining items are discussed in other sections, as appropriate, of this document.

†

 Material Maintenance and Surveillance Programs have been identified by the applicant as a

PSSC (see Section 5.6.2 of the MFFF revised CAR, Revision 1).
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11.7.1.1.2 Major Components

Major components include transfer containers, process equipment, confinement systems, MOX

fuel transport casks, and waste containers. The applicant stated in revised CAR Section 5.0 that

the material-handling system may have a plutonium dispersal hazard if the static barrier of the

primary confinement system is damaged because of a loss of confinement/dispersal of nuclear

material event or a load-handling event. The applicant has identified the material-handling

PSSCs, shown in Tables 11.7-1 and 11.7-2, which reduce the risk of the postulated events to

ensure the protection of facility workers, site workers, individuals outside the MFFF controlled

area (IOC), and the environment.

The applicant stated in Chapter 5 of the revised CAR that the material-handling system may

have a load-handling hazard from the presence of lifting or hoisting equipment used during

normal operations or maintenance activities in the facility. A load-handling event could occur

when either a lifted load containing radioactive materials is dropped, or the load or lifting

equipment impacts equipment containing radioactive material. Heavy load drops and other loadhandling events, as specifically defined in NUREG-1718 (NRC, 2000) and NUREG-0612, are

discussed in the revised CAR Section 11.10 and are evaluated in Section 11.10 of this FSER. 

11.7.1.1.3 Control Concept

The MP and AP process control systems use a distributed processing control system strategy,

with the manufacturing process translated into control algorithms for each process step. The

systems include normal, protective, and safety control subsystems that ensure the final product

conforms to manufacturing specifications and minimize plant waste and risk. The normal control

subsystem controls the manufacturing process, and the protective control subsystem maintains

industrial safety (protects personnel) and protects equipment. The safety control subsystem

ensures that safety limits will not be exceeded and that undesirable operational conditions are

prevented or mitigated. 

11.7.1.2 Design Bases of PSSCs

This section describes the PSSC design bases contained in the revised CAR.

Material-Handling Equipment and Controls 

The applicant states in Section 11.7.7 of the revised CAR that material-handling equipment and

support structural members will be designed to prevent physical interaction with confinement

boundary elements or PSSCs under worst-case loading associated with normal, upset, and

design basis events. To achieve this design objective, the applicant intends to apply design

principles such as redundant brakes, with fail-safe design, on lifting equipment; structural

oversizing of mechanical drive equipment; overspeed detection; mechanical stops; overtorque

detection; electrical interlocks; component sizing; magnetic grippers; glovebox hoods; and

shielding. The applicant also states that, as an additional safety function, material- handling

controls will be designed to prevent the potential overpressurization of reusable plutonium oxide

cans resulting from radiolysis or oxidation of plutonium (III) oxalate, thereby preventing potential

glovebox impacts from such overpressurizations. Material-handling controls are also discussed

in FSER Chapter 5.
11-125

Table 11.7-2 PSSCs for Load-Handling Events

Load-Handling Events Related to

Material-Handling Equipment Identified PSSC

For the protection of the...

Facility

Worker 

Site

Worker* IOC*

Environment

AP Process Cells Process cell U - - -

Process cell entry

controls

U - - -

Process cell exhaust

system

- U U U

AP/MP C3 Glovebox Areas Material-handling

controls**

U - - U

Material-handling

equipment**

U - - U

Glovebox U - - U

Facility worker controls U - - -

C3 confinement system - U U -

C1 and/or C2 Areas/3013 Canister 3013 canister** U U U U

Material-handling

controls**

U U U U

C1 and/or C2 Areas/3013 Transport

Cask

3013 transport cask** U U U U

Material-handling

controls**

U U U U

C1 and/or C2 Areas/Fuel Rod

 

Facility worker action U - - -

C1 and/or C2 Areas/MOX Fuel

Transport Cask

MOX fuel transport

cask**

U - - U

Material-handling

controls**

U - - U

C1 and/or C2 Areas/Waste Container Facility worker action U - - -

C1 and/or C2 Areas/Transfer

Container

Transfer container** U U U U

Material-handling

controls**

U U U U

C1 and/or C2 Areas/Final C4 HEPA

Filter

Material-handling

controls**

U U U U

C4 Confinement/Spill inside Glovebox C4 confinement system U U U U

C4 Confinement/Outside of MFFF

Building

Waste transfer line U U U U

C4 Confinement/Facilitywide MFFF building structure U U U U

Material-handling

controls**

U U U U

* There may be confinement systems or barriers that provide defense-in-depth protection for the site worker, IOC, or

the environment for which no credit is technically being taken by DCS.

** These items are material transport system PSSCs evaluated in this section of the FSER. The remaining items are

discussed in other sections, as appropriate, of this document.
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The design bases of material-handling PSSCs are qualified by the following codes and

standards:

C ANSI/AISC N690-1994 (AISC, 1994) for design of components required to maintain

structural integrity

C ASME B30.2-1996, “Top Running Bridge, Single or Multiple Girder, Top Running Trolley

Hoist Overhead and Gantry Cranes,” issued 1996 (ASME, 1996), for design of overhead

cranes

C ASME B30.16-1998, “Overhead Hoists,” issued 1998 (ASME, 1998), for design of

hoisting equipment

C ASME NOG-1-1998, “Rules for Construction of Overhead and Gantry Cranes (Top

Running Bridge, Multiple Girder),” issued 1998 (ASME 1998b), for seismic design

C AWS D1.1-1998 (AWS, 1998a) for structural integrity and fabrication

C AWS D1.6-1999, “Structural Welding Code—Stainless Steel,” issued 1999 (AWS, 1999),

for structural integrity and fabrication

C AWS D14.1-1997, “Specification for Welding Industrial and Mill Cranes and Other

Material Handling Equipment,” issued 1997 (AWS, 1997), for structural integrity and

fabrication

C Crane Manufacturers Association of America, CMAA-70-1994, “Specifications for Top

Running Bridge and Gantry Type Multiple Girder Electric Overhead Traveling Cranes,”

issued 1994 (CMAA, 1994a), for design of bridge cranes

C CMAA-74-1994, “Specification for Top Running and Under Running Single Girder Electric

Overhead Traveling Cranes Utilizing Under Running Trolley Hoist,” issued 1994 (CMAA,

1994b), for design of bridge cranes

C NUREG-0554, “Single Failure Proof Cranes at Nuclear Power Plants,” issued May 1979

(NRC, 1979), for design of bridge cranes

3013 Canister and Transport Cask 

The 3013 canister consists of individually sealed outer and inner containers which isolate the

stored materials from the environment. The 3013 canister is designed to meet the requirements

of DOE-STD-3013-2000, “Stabilization, Packaging, and Storage of Plutonium-Bearing Materials”

(DOE, 2000). The outer can is designed and qualified for a 9-m (30-ft) drop onto a flat,

unyielding surface while remaining leak-tight, and the inner can is designed to remain leak-tight

after a drop of 1.3 m (4 ft) onto a flat unyielding surface. The outer and inner cans are designed

to withstand pressures of 4927 kPa (699 psig) and 790 kPa (100 psig) and are hydrostatically

tested before use at a pressure 1.5 times that of the design pressure. Both containers must be

fabricated from ductile, corrosion-resistant materials, such as 300 series stainless steel or better. 

Closure welding of the stainless steel must be done in such a way as to minimize the

sensitization of stainless steel to stress-corrosion cracking. Heat generation limits the mass of
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plutonium contained in the containers to less than or equal to 19 watts (1.1 Btu/min). Both of the

containers are designed to hold the material for a maximum of 50 years.

The 3013 transport cask is a PSSC designed and qualified to protect the 3013 canister against

transportation accidents. It will be certified to meet the applicable requirements contained in

10 CFR 71.73, “Hypothetical Accident Conditions.”

Transfer Containers

Transfer containers are designated as PSSCs that will be used to hold and ship MOX

transuranic wastes. The waste transfer containers are designed, constructed, and qualified to

meet the requirements of the U.S. Department of Transportation Specification 7A of

49 CFR 178.350, “Specification 7a; General Packaging, Type A.”

MOX Fuel Transport Casks

Fresh fuel assemblies will be transported in MOX fuel transport casks. The casks may be

stacked in storage frames in the truck shipping bay. To accommodate the load, the frames will

be designed for the full weight of the shipping package including seismic effects. The fresh fuel

casks will be qualified to meet the requirements of 10 CFR Part 71. The casks will not be used

in the facility before being licensed by the NRC, and, if approved, they will be handled in the

proposed facility according to the certification.

Seismic Design 

Material-handling equipment designated as PSSCs is designed and qualified according to

national codes and standards, enabling the equipment to perform its safety function during

normal operations, upset conditions, and design basis events. The ability to safely shut down

the primary process is facilitated by the seismic design for the material-handling equipment and

structural support members. Equipment geometry and alignment must be maintained in order to

have an orderly shutdown of the system. The system is designed to prevent physical interaction

with confinement boundary elements or PSSCs under worst-case loading associated with

normal, off-normal, accident, and design basis events. The system will be designed to meet the

criteria provided in RG 1.100, Revision 2 (NRC, 1988); IEEE Standard 344-1987 (IEEE, 1987);

and ANSI/AISC N690-1994 (AISC, 1994). By adhering to IEEE Standard 344-1987, DCS is

satisfying the NRC’s regulations pertaining to SQ of mechanical equipment. Thus, thermal

loads, attached piping loads, and live loads, such as fluid sloshing, are considered for equipment

operability. In addition, applied loads are required to meet or exceed accelerations

corresponding to their installed locations.

The seismic monitoring system is designed to satisfy the criteria provided in RG 3.17-1974

(NRC, 1974). The design basis of the seismic monitoring system is that it provides sufficient

data to evaluate the response of the confinement structure and other PSSCs to a seismic event

and initiate a shutdown of process systems in the event of a high seismic event. The seismic

system will meet the requirements of IEEE 603-1998 (IEEE, 1998).

The staff has reviewed the DCS commitment to these codes and standards and finds that they

provide an acceptable basis for the seismic design of the MFFF material transport system.
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Evaluation of Capacity

The staff evaluated the information provided by DCS in its revised CAR regarding the capacity of

the proposed material-handling equipment. The material-handling equipment is designed such

that, in the event of accident or off-normal conditions, the equipment is designed to de-power or

return to a shutdown condition. The throughput, or capacity, of the equipment peaks during

normal operations or maintenance. The revised CAR describes the design of the equipment as

being sized to handle the required throughput of shipping packages, containers, canisters,

drums, casks, cans, powder, pellets, scrap, rods, and assemblies necessary for normal and offnormal operating conditions and maintenance. The revised CAR also discusses that the active

systems, such as motors, power transmission systems and pass-throughs, carriers, actuators,

end effectors, structural supports, sensors, and control systems will be based on the material

throughput requirements for each process unit.

Capacities of the material-handling equipment vary based on the operational throughput needed

for the equipment, the design and qualification of the equipment, and other specific design

criteria. The material-handling equipment designated as PSSCs must also retain its loads under

all credible accidents and design basis natural phenomena events. Therefore, the capacity of

these PSSCs may be greater than non-PSSCs. To accomplish these design requirements,

equipment will be designed to (1) maintain clearance between equipment and the confinement

boundary under all conditions, (2) include physical stops to prevent uncontrolled motion of

payloads from breaching containment in the event of over-travel or seismic conditions, (3)

ensure that actuating mechanisms, such as grippers, are designed to retain the payload under

all conditions including loss-of-power and credible seismic events, and (4) maintain appropriate

margins of safety in hoisting equipment. Capacity of equipment is not directly discussed by

design codes. However, the staff accepts that, if PSSC equipment is built to the design codes

referenced in the revised CAR, it will be designed to handle all loads, events, and configurations

while maintaining its safety function.

Material-handling equipment intended to suspend loads from flexible cables is designed using

codes for cranes, monorails, and hoists, which include appropriate minimum factors of safety. 

The capacities of hoisting equipment identified as PSSCs are further de-rated according to

safety factors from NUREG-0554 applicable to single-failure-proof cranes. Process equipment

used during maintenance is further de-rated to 65 percent of capacity according to projectspecific design criteria. Hoists will be designed in accordance with ASME B30.16-1998 (ASME,

1998a). Material-handling equipment that runs on fixed tracks is designed using structural

design codes. Equipment classified as PSSCs is qualified in accordance with load combinations

and acceptance criteria provided in the ANSI/AISC N690-1994 structural design code (AISC,

1994).

The staff has reviewed the description of the capacity of the material-handling equipment. 

Based on the description of the system provided in the revised CAR and the use of industry

codes and standards, the staff finds the design to be acceptable to handle the expected volume

of radioactive material during normal operating and accident conditions.
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Evaluation of Redundancy and Diversity

The revised CAR describes the passive design of the equipment to handle shipping packages,

containers, canisters, drums, casks, cans, powder, pellets, scrap, rods, and assemblies

necessary for normal and off-normal operating conditions and maintenance. The staff reviewed

the system design basis including the provision of active systems, such as motors, power

transmission systems and pass-throughs, carriers, actuators, end effectors, structural supports,

sensors, and controls. Material-handling equipment includes devices that suspend loads from

flexible cables and material-handling equipment that runs on fixed tracks. Some of this

equipment is designed to work external to a glovebox, and some is designed to work internal to

a glovebox environment.

Redundancy and diversity in the design are accomplished by various factors of safety, types of

equipment, and the layering of active and passive controls that protect the confinement

boundary. The facility design for material-handling equipment inside gloveboxes includes

redundant brakes, with fail-safe design, on lifting equipment; structural oversizing of mechanical

drive equipment; overspeed detection; mechanical stops; overtorque detection; electrical

interlocks; and component sizing based on worst-case loading combinations. Various containers

and casks are also discussed in the facility design. Each of these containers or casks is

designed for different applications and is certified under Federal regulations before use.

The staff has reviewed the applicant’s description of the material-handling equipment regarding

system redundancy and diversity. On the basis of standard industry practices, the staff finds the

design to be acceptable to prevent the release of radioactive materials to the environment. The

staff notes that casks and canisters are likewise acceptable if they are used within their

certification basis or if appropriate compensatory measures are taken with consideration to the

hazard to the facility and site worker, the IOC, and the environment.

Evaluation of Safe Shutdown

The staff evaluated the information provided by DCS in the revised CAR regarding the ability to

safely shut down the proposed material-handling equipment during normal, accident, and

maintenance conditions. The material-handling equipment is designed such that in the event of

an accident or off-normal condition, the equipment is designed to de-power in a fail-safe

condition. Emergency power is not provided to the material-handling equipment. For example,

during a loss of power, hoist brakes passively activate, and end effectors, such as magnetic

grippers, passively fail to a closed or retain load condition. In this way, the design ensures that

in any accident or off-normal condition, all system loads are maintained and the confinement

boundaries are not challenged by dropped or unrestrained loads. The staff has reviewed the

applicant’s description of the material-handling equipment’s ability to safely shut down and finds

this approach to be acceptable.

Evaluation of Welded Construction 

In NUREG-1718 (NRC, 2000), the regulatory acceptance criteria for the material transport

system specifically mentions that tank and piping systems should be of welded construction to

the fullest extent possible. For the purposes of this review, this guidance is applied to

gloveboxes (a material- handling PSSC). Continuously welded construction means the seams

are ground smooth, which facilitates cleaning and minimizes holdup of powder, pellets, dust, or
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debris. The specification of welded construction also minimizes leakage paths and facilitates

decontamination of gloveboxes. The applicant’s proposed design for gloveboxes is evaluated in

FSER Section 11.4. Based on the use of industry codes for the design and construction of

welded material-handling equipment, the staff finds this design basis to be acceptable.

Evaluation of Passive Features/Remote Operation 

As discussed previously, the material-handling equipment is designed such that, in the event of

an accident or off-normal condition, the equipment will de-power in a fail-safe condition. 

Emergency power is not provided to the material-handling equipment. For example, during a

loss of power, hoist brakes passively activate, and end effectors, such as magnetic grippers,

passively fail to a closed or retain load condition. In this way, the design ensures that in any

accident or off-normal condition, all system loads are maintained and the confinement

boundaries are not challenged by dropped or unrestrained loads. The material- handling

equipment is designed with engineered features to prevent active failures from impacting the

glovebox walls. Based on the applicant’s commitment in CAR Section 11.7.7 to these passive

engineered features, the staff finds that the design provides an adequate level of protection

against active failures.

For most operations, the process control system is designed to control the material-handling

equipment during normal process conditions. In the event of an off-normal or accident condition,

additional control system elements are capable of overriding the normal process controllers to

mitigate the potential hazardous condition. This equipment and its functions are described in

further detail in Section 11.6 of this FSER. During maintenance, process equipment and

controllers are deenergized, and equipment may be selectively energized under manual control

of facility personnel engaged in the maintenance activities. Based on the control system

design’s being for remote operation of process equipment, the staff finds this design for remote

operation to be acceptable.

Evaluation of Radiation Safety

The staff review and evaluation of the radiation safety program is discussed in detail in Section 9

of this FSER. In general, the design basis for the material-handling equipment hardware for

radiation safety entails the use of design configurations to minimize powder/dust or debris;

mounting of stainless steel casings on structural supports to prevent powder/dust retention; easy

visibility and accessibility of parts for cleaning; use of sealed bearings or leak-free coupling

mechanisms; use of appropriate surface quality or coatings for equipment in contact with

powder; limiting lubricant use to the extent practical; continuous and smoothly ground internal

welds; re-entrant corners of large relative radius; and sealed powder-handling channels. Based

on commitments to industry standards for surface finish and the general equipment design for

minimization of powder holdup in equipment (e.g., Section 11.7.6 of the CAR), the staff finds

these provisions to be acceptable.

Evaluation of Corrosion Resistance

The material-handling equipment proposed by the applicant is to be made primarily of stainless

steel with the appropriate surface finishes to resist corrosion. The corrosion of carbon steel

parts that cannot be painted will be prevented by a glovebox environment of nitrogen or dry air. 

For other areas, components may be coated or painted for corrosion resistance and ease of
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decontamination. Outside gloveboxes, painting systems will be used for materials located in

C3b rooms to facilitate decontamination. In addition, the material condition of the equipment will

be monitored by the Material Maintenance and Surveillance Programs. On the basis of industry

codes and standards that specify system design accounting for corrosion as a standard

industrial practice, the staff finds the design to be acceptable.

Evaluation of Personnel Protection 

Section 11.4.7.2 of NUREG-1718 (NRC, 2000) states that the need for hoods, gloveboxes, and

shielding for personnel protection should be evaluated. These systems are generally required

for processing operations involving more than gram quantities of plutonium or general

operations involving 50 micrograms or more of plutonium in respirable form. In Section 11.8.2 of

the revised CAR, the applicant states that personnel protection for these materials is provided by

process cells and welded equipment confinement. Process cells contain equipment that handles

radioactive materials in chemical solutions, and that equipment is of a fully welded construction

and does not require routine maintenance. Equipment containing radioactive materials in the

powder process (MP) is contained in gloveboxes in process rooms that provide equivalent

confinement to fully welded equipment in process cells. In Section 11.4.7.1.3 of the revised

CAR, the applicant documented its analysis of the accident scenarios for fire and impact events

with gloveboxes. Based on the applicant’s analyses and commitments described above, the

staff finds this design for personnel protection to be acceptable.

11.7.2 Evaluation Findings

In Chapter 11.7 of the revised CAR, DCS provided design basis information for the material

transport systems that it identified as PSSCs for the proposed facility. Based on the staff's

review of the revised CAR and supporting information provided by the applicant relevant to the

material transport systems, the staff concludes, pursuant to 10 CFR 70.23(b), that the design

bases of the PSSCs evaluated in this FSER section will provide reasonable assurance of

protection against natural phenomena and the consequences of potential accidents.

11.7.3 References

(AISC, 1989) American National Standards Institute/American Institute of Steel Construction.

“Manual of Steel Construction, Allowable Stress Design,” 9th Edition. 1989.

(AISC, 1994) American National Standards Institute/American Institute of Steel Construction

(ANSI/AISC). N690, “Specification for the Design, Fabrication, and Erection of Steel SafetyRelated Structures for Nuclear Facilities.” 1994.

(ASME, 1995) American Society of Mechanical Engineers. B46.1, “Classification and

Designation of Surface Qualities.” 1995.

(ASME, 1996) American Society of Mechanical Engineers. B30.2, “Top Running Bridge, Single

or Multiple Girder, Top Running Trolley Hoist Overhead and Gantry Cranes.” 1996.

(ASME, 1998a) American Society of Mechanical Engineers. B30.16, “Overhead Hoists.” 1998.
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(ASME, 1998b) American Society of Mechanical Engineers. NOG-1, “Rules for Construction of

Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder).” 1998.

(AWS, 1997) American Welding Society. D14.1, “Specification for Welding Industrial and Mill

Cranes and Other Material Handling Equipment.” 1997.

(AWS, 1998a) American Welding Society. D1.1, “Structural Welding Code—Steel.” 1998.

(AWS, 1998b) American Welding Society. D1.3, “Structural Welding Code—Sheet Steel.” 

1998.

(AWS, 1998c) American Welding Society. D9.1, “Sheet Metal Welding Code.” 1998.

(AWS, 1999) American Welding Society. D1.6, “Structural Welding Code—Stainless Steel.”

1999.

(CMAA, 1994a) Crane Manufacturers Association of America. 70, “Specifications for Top

Running Bridge and Gantry Type Multiple Girder Electric Overhead Traveling Cranes.” 1994.

(CMAA, 1994b) Crane Manufacturers Association of America. 74, “Specification for Top

Running and Under Running Single Girder Electric Overhead Traveling Cranes Utilizing Under

Running Trolley Hoist.” 1994.

(DCS, 2001) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission RE: Clarification of Responses to NRC Request for Additional Information (DCSNRC-000074). December 5, 2001.

(DCS, 2002a) Ihde, R., Duke Cogema Stone & Webster. Letter to Document Control Desk,

U.S. Nuclear Regulatory Commission, RE: Mixed Oxide Fuel Fabrication Facility—Construction

Authorization Request. October 31, 2002 (including page changes through February 9, 2005).

(DCS, 2002b) Hastings, P., Duke Cogema Stone & Webster. Letter to Document Control Desk,

U.S. Nuclear Regulatory Commission, RE: Clarification of Responses to NRC Request for

Additional Information. February 11, 2002.

(DOE, 2000) U.S. Department of Energy. Standard-3013-2000, “Stabilization, Packaging, and

Storage of Plutonium-Bearing Materials.” Washington, DC, 2000.

(IEEE, 1987) Institute of Electrical and Electronics Engineers. Standard 344, “IEEE

Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear Power

Generating Stations.” 1987.

(IEEE, 1998) Institute of Electrical and Electronics Engineers. Standard 603, “IEEE Standard

Criteria for Safety Systems for Nuclear Power Generating Station.” 1998.

(NRC, 1974) U.S. Nuclear Regulatory Commission. Regulatory Guide 3.17, “Earthquake

Instrumentation for Fuel Reprocessing Plants.” Washington, DC, 1974.
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(NRC, 1979) U.S. Nuclear Regulatory Commission. NUREG-0554, “Single Failure Proof

Cranes at Nuclear Power Plants.” Washington, DC, May 1979.

(NRC, 1988) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.100, Rev. 2, “Seismic

Qualification of Electric and Mechanical Equipment for Nuclear Power Plants.” Washington, DC,

1988.

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility.” Washington,

DC, August 2000.

11.8 Fluid Transport Systems

11.8.1 Conduct of Review

This chapter of the FSER contains the staff’s review of the fluid transport systems described by

the applicant in Chapter 11.0 of the revised CAR (DCS, 2002a). The objective of this review is

to determine whether the fluid transport PSSCs and their design bases identified by the

applicant provide reasonable assurance of protection against natural phenomena and the

consequences of potential accidents. The staff evaluated the information provided by the

applicant for fluid transport systems by reviewing Chapter 11.0 of the revised CAR, other

sections of the revised CAR, supplementary information provided by the applicant, and relevant

documents available at the applicant’s offices but not submitted by the applicant. Additional

documentation from the literature was reviewed as necessary to understand the process and

safety requirements. The review of fluid transport systems design bases and strategies was

closely coordinated with the review of other chapters in this FSER (fire protection in Chapter 7,

chemical safety in Chapter 8, and safety assessment of the design bases in Chapter 5).

The staff reviewed how the information in the revised CAR addresses the following regulation:

• 10 CFR 70.23(b) requires that the design bases of the PSSCs and the Quality Assurance

Program must provide reasonable assurance of protection against natural phenomena

and the consequences of potential accidents before construction of the PSSCs is

approved.

The review for this construction approval focused on the design bases of fluid transport systems,

their components, and other related information. For fluid transport systems, the staff reviewed

and evaluated information provided by the applicant for the safety function, system description,

and safety analysis. The review also encompassed proposed design basis considerations such

as redundancy, independence, reliability, and quality. The staff used Chapter 11.0 in

NUREG-1718 (NRC, 2000), particularly Section 11.4.7, and industry codes and standards as

guidance in performing the review.

The MFFF fluid transport systems include systems that handle process and utility fluids. Other

fluid-containing support systems are discussed in Section 11.9 of this FSER. 

The NRC staff reviewed the revised CAR submitted by DCS for the following areas applicable to

the fluid transport systems at the construction approval stage and consistent with the level of

design:
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• system description

• system function

• major components

• control concepts

• system interfaces

• design bases

Section 11.4.7 of NUREG-1718 was the primary guidance used for this review. Regarding the

proposed facility fluid transport systems, specific design considerations given in the revised CAR

are to demonstrate the following:

• Capacity is adequate to handle the expected volume of radioactive material during

normal operating and accident conditions.

• Redundancy or diversity of components prevents the release of radioactive materials to

the environment or contributes to the safe operation of the fluid transport systems.

• The fluid transport system can be safely shut down during normal and accident

conditions. Provisions for emergency power are included for critical process

components.

• Tank and piping systems are of welded construction to the fullest extent possible.

• Tank and piping systems are designed to take advantage of gravity flow to reduce the

potential for contamination associated with pumping and pressurization.

• Criticality will not occur under normal and credible accident conditions.

• All system components expected to be in contact with strong acids or caustics are

corrosion resistant.

• Piping is designed to minimize entrapment and buildup of solids in the system.

• The systems are evaluated to determine the need for hoods, gloveboxes, and shielding

for personnel protection. Generally, wet processing operations involving gram quantities

of plutonium and any operations involving 50 micrograms of respirable plutonium are

conducted in a glovebox.

• Surface finishes of materials in the work areas have satisfactory decontamination

characteristics for their particular application.

• Fluid transport systems maintain functionality when subjected to tornadoes, tornado

missiles, earthquakes, floods, and any other natural phenomena deemed to be credible

as further established in the ISA to be performed by the applicant.

In the FSER discussions that follow, the system descriptions are provided as well as function,

major components, control concepts, and system interfaces. These discussions include but are

not limited to PSSCs, to provide an understanding of the system. The design bases of PSSCs

are discussed in Section 11.8.1.3 of this FSER.
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11.8.1.1 System Description

The fluid transport system contains the hardware portion of the AP wet process that contains the

dissolution unit, the purification cycle, solvent recovery, oxalic precipitation, and precipitation and

drying for the various stages of plutonium polishing. All wastes generated are stored and sent

by pipeline to DOE SRS for disposition.

In general, the fluid transport system consists of components that contain radioactive process

fluids, radioactive waste products, and utility fluids. The PSSCs in the process equipment

include vessels, tanks, pulsed columns, heat exchangers, pumps, piping, and valves. Other

process equipment includes electrolyzers, airlifts, drip pots, sampling lines and pots, spargers,

gravity feeds, dosing wheels, magnetic stirrers, extraction screws, and stationary and rotary

filters. 

Fluid transport components are designed for the most severe credible service conditions. Fluids

containing radioactive materials are designed to always be within at least two levels of

confinement according to the general principles described in revised CAR Section 11.4 and

reviewed in the FSER. Fluid-bearing components within process cells that are not easily

accessed for routine inspection will be designed with corrosion allowances and provided with

drip trays sized to hold the contents of the largest vessel in a criticality-safe condition. Sump

pumps on the drip trays are monitored for activity signaling a leak. The system is designed to

accommodate flushing and high-pressure decontamination to remove sediment buildup and any

blockages and to maintain ALARA principles, before system maintenance.

The fluid transport system for the AP process is diverse in the control concepts for the process

control systems that govern the fluid transport system. It is composed of the normal, protective,

and safety control subsystems. The normal control subsystem controls the facility

manufacturing and processing operations. The protective control subsystem provides protection

for equipment and personnel. The safety control subsystem is designed to ensure that safety

limits will not be exceeded and that undesired operational conditions or events will not occur or

will be properly mitigated. For more information on the AP process, refer to Section 11.2 of this

FSER. Section 11.6 of the FSER provides more details on the design and operation of the I&C

system.

Fluid transfer systems containing hazardous fluids are contained within trenches, rooms, or

double-walled piping systems or are accessible for inspection and are of a fully welded

construction. All piping components designated as PSSCs are designed to withstand the design

basis earthquake loads. Isolation valves may be of the butterfly, gate, plug, or ball type. The

valves are specified for service after consideration of the chemical characteristics of the fluid,

piping material of construction, and operating conditions.

Section 11.4.7 of NUREG-1718 (NRC, 2000), used for the review of the fluid transport systems,

specifically mentions that tank and piping systems should be of welded construction to the fullest

extent possible. For process equipment, radiological fluids are maintained within at least two

levels of confinement. Components containing fluids that are located in process cells are

specified with corrosion allowances, and welded joints are radiographed, as appropriate. 

Components that are not of fully welded construction are installed in a glovebox. Piping

components carrying radiological fluids between two confinements are either of fully welded

double-wall construction or are installed in gloveboxes. Welding requirements are contained in
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the ASME Boiler and Pressure Vessel (B&PV) Code, Section VIII, 1996 Edition through the 1998

Addenda (ASME, 1998), ASME B31.3 (ASME, 2001a), and their referenced AWS codes. 

Welding work will be performed according to the fluid transport system (FTS) category (see

FSER Table 11.8-1) and quality assurance plan quality level of the system being welded. 

Materials used for the construction of this equipment are specified in accordance with ASME and

ASTM specifications. ASME materials are used in the fabrication of equipment and piping

components built to the requirements of ASME B&PV Code, Section VIII, 1996 Edition through

the 1998 Addenda, and ASME B31.3. In general, design of equipment to these standards

means that the components are designed for the most severe service conditions. Included in

the severe service conditions are pressure, temperature, stress, material compatibility, and

corrosion.

11.8.1.1.1 Function

The AP process can be segmented into the following four operational areas:

(1) plutonium purification process—includes the decanning unit, dissolution unit, purification

cycle, oxalic precipitation and oxidation unit, homogenization unit, and canning unit

(2) recovery processes—includes the solvent recovery cycle, oxalic mother liquor recovery

unit, and acid recovery unit

(3) waste storage—includes the liquid waste reception unit

(4) offgas treatment—includes the offgas treatment unit

This section concentrates on the MOX equipment located in the plutonium purification process. 

The plutonium purification process separates impurities from the fissile material. This process is

a radiochemical process where the fissile material commingles with inorganic and organic

solutions at various concentrations. The plutonium purification process is divided into six

discrete steps:

(1) The decanning unit is a mechanical operating unit where can opening and powder

transfer operations are automatically performed, primarily in gloveboxes.

(2) Plutonium dioxide from the decanning unit is electrochemically dissolved with silver (Ag2+)

in nitric acid. The plutonium valence of the resulting plutonium nitrate solution is altered

by hydrogen peroxide from (VI) to (IV). The plutonium nitrate solution is transferred to

the purification cycle feed tank.

(3) The plutonium nitrate solution from the dissolution unit is extracted into the organic phase

from impurities (e.g., gallium, americium) in the flux. The organic stream is scrubbed

with dilute nitric acid, its valence is further reduced to Pu (III), and it is stripped back into

the aqueous solution. The plutonium nitrate is oxidized to the tetravalent state by

nitrogen oxide fumes.

(4) In the oxalic precipitation and oxidation unit, plutonium is precipitated with excess oxalic

acid in vortex precipitators.
11-137

(5) In the homogenization unit, the plutonium oxalate precipitate and the mother liquors flow

from the precipitator and are channeled to a flat filter where they are filtered, washed,

and vacuum-dewatered. The oxalate is dried and calcined.

(6) The canning unit is a mechanical operating unit where the calcined plutonium dioxide is

sampled and packaged for use in the MP.

These steps are described and evaluated in detail in Chapter 8 and Section 11.3 of this FSER.

11.8.1.1.2 Major Components

The major components of the fluid transport systems are part of the primary process and are

located in the AP area of the facility. In addition to piping and valves, the major components of

the fluid transport systems include the following: 

• Welded process equipment includes vessels, tanks, process columns, and heat

exchangers. In general, fully welded process equipment is located in process cells. 

Storage tanks vary in design at different stages of the primary process. Storage tanks

include annular tanks, stab tanks, and conventional tanks. These tanks are fabricated

using fully welded construction. Other welded process equipment includes various small

tanks used in the AP process, such as separating pots, leak detection pots, barometric

seal pots, pulse column pots, drip pots, condensate pots, and demisters. The AP

process columns are also of a fully welded construction. Examples of these columns are

pulsed, rectification, packed or scrubbing columns, and tray columns for process

distillation. Various AP process heat exchangers used in radiological service are also of

a fully welded construction. These heat exchangers may be evaporators, condensers,

and jacketed heaters/coolers designed to transfer process heat. 

• Partially welded process equipment and prime movers include filters, mixing tanks, and

precipitators. Partially welded equipment classified as FTS Category 1 is housed inside

gloveboxes (see FSER Section 15.1 for a more detailed discussion). Other process

prime movers include pumps, low-pressure airlifts, ejectors, and siphons. Pump types

include centrifugal and positive displacement dosing pumps.

 .

11.8.1.1.3 Control Concept

The AP process control systems are designed to ensure that the product of the manufacturing

process will conform to the product specifications while simultaneously minimizing risk to the

facility worker, site worker, and IOC. They are composed of the normal, protective, and safety

control subsystems. The normal control subsystem controls the facility’s normal manufacturing

and processing operations. The protective control subsystem provides protection for personnel

and equipment. The safety control subsystem is designed to ensure that safety limits will not be

exceeded and that undesired operational conditions or events will not occur or will be mitigated. 

Section 11.6 of the revised CAR discusses the facility instrumentation and control systems in

more detail.

In general, each unit is controlled by one or several PLCs associated with a monitoring

workstation located in the AP control room. All units are operated in an automatic mode. The
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operator may also intercede via a manual mode in which the interlocks are active in case of

trouble in the automatic mode or for maintenance operations. The MMIS collects the information

coming from all process units to control the position and the exchange of special nuclear

material (SNM), as well as the traceability and the quality of the products.

Process storage and operation conditions are controlled to prevent exothermic and potential

autocatalytic reactions in the AP area. Autocatalytic and exothermic reactions of chemicals,

precipitation, and criticality are prevented through control of the process parameters (e.g.,

reactant concentration, temperature, catalyst concentration in solution, and pressure) that affect

the reactions.

11.8.1.2 System Interfaces

The individual systems that interface with the fluid transport systems include the mechanical

utility systems, bulk gas system, and reagent systems as described in the revised CAR Section

11.9, “Fluid Systems.”

The mechanical utility systems that interface with the fluid transport systems include the process

chilled water system, the demineralized water system, the process hot water system, the

process steam and condensate systems, and the instrument air system. The bulk gas systems

that interface with the fluid transport systems include the nitrogen system, the argon/helium

system, the helium system, and the oxygen system. The reagent systems that interface with the

fluid transport system include the nitric acid system, the silver nitrate system, the TBP system,

the hydroxylamine system, the sodium hydroxide system, the oxyalic acid system, the diluent

system, the sodium carbonate system, the hydrogen peroxide system, the hydrazine system, the

manganese nitrate system, the decontamination system, and the nitrogen oxide system.

11.8.1.3 Design Bases of PSSCs

This section describes the design basis commitments made by DCS for the PSSCs of the fluid

transport system. The fluid transport system PSSCs identified by the applicant include backflow

prevention features, material maintenance and surveillance programs, seismic isolation valves,

waste transfer line, and double-walled pipe.

The applicant established an FTS categorization to describe the combination of component and

material codes, SCs, and quality levels. These categories are summarized in Table 11.8-1. 

Design, engineering, construction, and operation of the components classified in each category

must conform to applicable industry codes and standards. Fluid transport system PSSCs are

categorized as either FTS 1 or 2, while non-PSSCs are either FTS 3 or 4. Tables 11.8-2 through

11.8-5 list the design basis codes and standards that will apply to each of the FTS categories. 

Design basis parameters for the fluid transport system are presented in Table 11.8-6.

By adhering to industry codes and standards, the design bases of the fluid transport system

PSSCs conform to the regulatory acceptance criteria established in Section 11.4.7.2 of NUREG1718 (NRC, 2000). These criteria are summarized in Section 11.8.1 of this FSER.

The DCS safety assessment, in Section 5.5 of the revised CAR, lists two failure modes related

to fluid transport equipment that could lead to a loss of confinement/dispersal of nuclear material

events. They are (1) corrosion occurring from within or from the outside of process equipment
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and (2) leaks of AP process vessels or pipes within process cells. The material maintenance

and surveillance programs and the qualifications of the fluid transfer systems to limit corrosion

have been identified as PSSCs that reduce the radiological risk to facility workers from

corrosion. The safety function of the material maintenance and surveillance programs is to

detect and limit the damage resulting from corrosion (principally associated with failures from

corrosion occurring outside of process equipment). The safety function of the fluid transfer

systems is to limit corrosion through the use of materials compatible with the environment and

system fluids (principally associated with failures from corrosion occurring inside process

equipment). No PSSCs are required to protect the IOC and site worker from this type of

accident because of the low unmitigated consequences. The revised CAR lists the design

corrosion allowances for equipment not accessible to inspection, and therefore not available for

the material maintenance and surveillance program, which can provide additional protection for

the facility worker against the consequences of corrosion.

The process cell has been identified to reduce the risk to facility workers from leaks of AP

process vessels or pipes within process cells. The safety function of the process cell is to

contain leaks within process cells. The process cell entry controls have also been identified as

PSSCs. The safety function of the process cell entry controls is to prevent the entry of

personnel to process cells during normal operations and to maintain worker radiation doses

within limits while performing maintenance in process cells. No PSSCs are required to protect

the IOC and site worker from this type of accident because of the low unmitigated

consequences. However, the C3 confinement system provides defense-in-depth protection to

the IOC and site worker.

Evaluation of Capacity 

The staff evaluated the information provided by DCS in its revised CAR regarding the capacity of

the proposed fluid transport systems. The staff notes that fluid transport systems with safety

functions are identified as PSSCs. The PSSCs in the fluid transport systems are designed to

have adequate capacity to handle the volume of radioactive material during normal operation

and design basis accident conditions. The components’ design bases are defined from thermal

and hydraulic calculations and will consider the physical and chemical properties of the process

fluid. The design flow and volumetric capacity design criteria were provided and discussed in

Table 11.8-6. The nominal capacity of the process is 24 kg/day (52.9 lb/day). Solvents are

recovered from the process as they are generated. Process heat is removed by intermediate

cooling loops to the process chilled water system. Process heat is provided by the process

steam system through the process hot water system. Both the process heating and cooling

systems are sized to provide reliable and sufficient heating and cooling to the facility process

systems. Drip trays in process cells are designed to contain radiological fluids, sized to contain

the contents from the largest vessel in the cell, and shaped to maintain the leaked fluid in a

criticality-safe geometry.
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Table 11.8-1 Categorization of the Fluid Transport System

FTS

Category Description of Components in the Category

Category 1 Includes components that are PSSCs that contain process fluids with significant quantities of

plutonium or americium. See Table 11.8-2 for applicable codes and standards. The application

of the specific criteria for the material, fabrication, examination, testing, and installation were

derived from applicable codes and standards and augmented by operating experience at the

French La Hague facility.

Category 2 Includes components that are PSSCs that contain process fluids with trace quantities of

plutonium or americium or nonradiological fluids. See Table 11.8-3 for applicable codes and

standards. Positive material identification, inspection, and test requirements are used in

engineering and procurement specifications for Category 2 components.

Category 3 Includes components that are non-PSSCs that may contain process fluids with trace quantities of

plutonium or americium or nonradiological fluids that play a significant role for plant production

reliability. See Table 11.8-4 for applicable codes and standards. 

Category 4 Includes components as well as facility services that maintain production reliability. See Table

11.8-5 for applicable codes and standards.

Table 11.8-2 Design Basis Codes and Standards for Category 1 Fluid

Transport System Components

FTS

Category 1

Design Basis Codes and Standards

Process

Vessels

ASME B&PV Code, Section VIII, 1996 Edition through the 1998 Addenda (ASME, 1998), Division

1 or 2, for lethal service with enhanced positive material identification and test and inspection

requirements

Pumps ASME B73.1, “Specifications for Horizontal End Suction Centrifugal Pumps for Chemical

Process,” issued 2001 (ASME, 2001b), and ASME B73.2, “Specifications for Vertical In-line

Centrifugal Pumps for Chemical Process,” issued 2001 (ASME, 2001c), enhanced design

specification of ASME materials with enhanced positive material identification and test and

inspection requirements 

Specialty pumps per manufacturer’s standards (e.g., submerged rotor seal-less pumps)

API Standard 610, “Centrifugal Pumps for Petroleum, Heavy Duty Chemical and Gas Industry

Services,” issued 1995 (API, 1995b)

Piping ASME B31.3 (ASME, 2001a) Category M, with enhanced positive material identification, and test

and inspection requirements

Valves ASME B31.3 (ASME, 2001a) Category M

Other Criteria SC-1 and Quality Level-1 for all PSSCs
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Table 11.8-3 Design Basis Codes and Standards for Category 2 

Fluid Transport System Components

FTS Category 2 Design basis Codes and Standards

Process Vessels ASME B&PV Code, Section VIII, 1996 Edition through the 1998 Addenda (ASME,

1998), Division 1 or 2

Compressed Gas Association (CGA) S-1.1, “Pressure Relief Device

Standards—Part 1—Cylinders for Compressed Gases,” issued 2001 (CGA, 2001a),

and CGA S-1.3, “Pressure Relief Device Standards—Part 3—Stationary Storage

Containers for Compressed Gases,” issued 2001 (CGA, 2001b)

Pumps ASME B73.1 (ASME, 2001b) and ASME B73.2 (ASME, 2001c)

Specialty pumps per manufacturer’s standards (e.g., submerged rotor seal-less

pumps)

API Standard 610 (API, 1995b)

NFPA 20, “Standard for the Installation of Stationary Pumps for Fire Protection,”

issued 2003 (NFPA, 2003)

Piping ASME B31.3 (ASME, 2001a) with enhanced test and inspection requirements

Valves ASME B31.3 (ASME, 2001a)

Other Criteria SC-1 and Quality Level-1 for all PSSCs

Table 11.8-4 Design Basis Codes and Standards for Category 3

Fluid Transport System Components

FTS Category 3 Design basis Codes and Standards

Process Vessels ASME B&PV Code, Section VIII, 1996 Edition through the 1998 Addenda (ASME, 1998),

Division 1 or 2

CGA S-1.1 - (CGA, 2001a) and CGA S-1.3 (CGA, 2001b)

UL-142, “Standard for Safety for Steel Aboveground Tanks for Flammable and

Combustible Liquids”

Pumps ASME B73.1 (ASME, 2001b) and ASME B73.2 (ASME, 2001c) and Hydraulic Institute

standards

Specialty pumps per manufacturer’s standards (e.g., submerged rotor seal-less pumps)

NFPA 20 (NFPA, 2003)

Piping ASME B31.3 (ASME, 2001a)

Valves ASME B31.3 (ASME, 2001a)

Other Criteria CS or SC-2, as applicable and Quality Level 2, 3, or 4, as applicable
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Table 11.8-5 Design Basis Codes and Standards for Category 4 Fluid 

Transport System Components

FTS Category 4 Design Basis Codes and Standards

Process Vessels ASME B&PV Code, Section VIII, 1996 Edition through the 1998 Addenda (ASME,

1998), Division 1 or 2

CGA S-1.1 (CGA, 2001a) and CGA S-1.3 (CGA, 2001b)

UL-142

Pumps ASME B73.1 (ASME, 2001b) and ASME B73.2 (ASME, 2001c) and Hydraulic Institute

standards

Specialty pumps per manufacturer’s standards (e.g., submerged rotor seal-less

pumps)

Piping ASME B31.3 (ASME, 2001a)

Valves ASME B31.3 (ASME, 2001a)

Other Criteria CS or SC-2, as applicable and Quality Level 2, 3, or 4, as applicable

Normal electrical power is removed from process prime movers during an earthquake event. 

Those electrical loads requiring power for safe shutdown are supplied by emergency

uninterrupted power. (See Section 11.5 of this FSER for additional discussion and the staff

evaluation of the facility electrical system.) Based on the commitment to design-bases codes

and standards (e.g., Tables 11.8-1 and 11.8-2 of the CAR), the staff has determined that fluid

transport systems will provide adequate capacity to handle the expected volume of radioactive

material during normal operating and accident conditions.

Evaluation of Redundancy and Diversity 

Section 11.4.7.2 of NUREG-1718 (NRC, 2000) states that the applicant is to describe the

redundancy and diversity of components required to prevent the release of radioactive materials

to the environment or needed for safe operation of the fluid transport system. The fluid transport

system is designed with multiple layers of confinement and is supplemented by administrative

programs designed to monitor the integrity of these systems. Radiological fluids are maintained

within at least two layers of confinement. Piping systems are double walled with leak detection

systems if they are not located in process cells or gloveboxes. Drip trays with sump monitors

are designed to detect leakage.

[Text removed under 10 CFR 2.390]

Various types of methods are used to transfer fluids, such as pumps, gravity, steam jet lifts,

siphons, and airlifts. Hydraulic seals as well as check valves and isolation valves of various

types are used to isolate systems. Redundancy and diversity in the design are the result of the

various safety factors and types of equipment provided in the design and the layering of active

and passive controls that protect the fluid transport system.
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Table 11.8-6 Design Bases for the Fluid Transport System

FTS Component Design Pressure Design Temperature Design Flow & Volumetric

Capacity

Storage tanks The highest value of the

following:

(1) max. pressure—

normal operations + 10%

(2) max. pressure— 

transient condition 

The highest value of the

following:

(1) max. temp.—normal

operations

(2) max. temp.—transient

condition

Maximum flow in normal

operating conditions +

20%

Volumetric capacity for

vessels determined on the

basis of process safety

and operation reliability

requirements

Process columns Same as above Same as above Same as above

Heat exchangers Same as above Same as above Same as above

Pumps Same as above Same as above Same as above

Prime movers such as air

lifts, ejectors, jets, siphons

Same as above Same as above Process hydraulic

calculation

Piping and valves The highest value of the

following:

(1) max. pressure—

normal operations + 10%

(2) max. pressure in

normal operating

condition +0.9 bar

(13 psig) 

(3) max. pressure— 

transient condition 

The highest value of the

following:

(1) max. temp.—normal

operations + 15 °C 

(27 °F) 

(2) max. temp.—transient

condition

Pipe sizing per line

velocity and pressure drop

requirements

The fluid transport system is diverse in the control concepts for the process control systems that

govern the fluid transport systems. The AP process control systems are designed to ensure that

the product of the manufacturing process will conform to the product specifications with minimal

waste and risk. They are composed of the normal, protective, and safety control subsystems. 

The normal control subsystem controls the facility normal manufacturing and processing

operations. The protective control subsystem provides protection for personnel and equipment. 

The safety control subsystem is designed to ensure that safety limits will not be exceeded and

that undesired operational conditions or events will not occur or will be mitigated. The staff has

reviewed the DCS description of the fluid transport equipment and finds these systems to be

diverse. On the basis of standard industry practices, the staff finds the design to be acceptable.

Evaluation of Safe Shutdown 

The staff evaluated the information provided by DCS in its revised CAR regarding the ability to

safely shut down the proposed fluid transport systems during normal and accident conditions

and maintenance. The AP process control system, the electrical power system, and basic

system design criteria primarily control the ability of the proposed systems to shut down safely. 

The AP process instrumentation and control systems are designed to ensure that the product of

the manufacturing process will conform to the product specifications with minimal waste and risk. 

They are composed of the normal, protective, and safety control subsystems. The normal

control subsystem controls the facility’s normal manufacturing and processing operations. The

protective control subsystem provides protection for personnel and equipment. The safety
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control subsystem is designed to ensure that safety limits will not be exceeded and that

undesired operational conditions or events will not occur or will be mitigated. The PSSCs for the

I&C systems are the safety subsystems of the MP and AP process control systems and utility

control system and the hard-wired emergency control system. The I&C systems monitor and

control plant parameters during normal and transient conditions to ensure that limits are not

exceeded and to ensure the required quality of the product. They also provide signals to control

equipment to prevent and mitigate faulted conditions. All emergency control equipment is

qualified for design basis seismic events and normal, off-normal, and design basis accident

environmental conditions. (See Section 11.6 of this FSER for additional detail on I&C system

safety.)

Normal electrical power is removed from process prime movers during an earthquake event. 

Those electrical loads requiring power for safe shutdown are supplied by emergency

uninterrupted power. (See the discussion of electrical systems safety in Section 11.5 of this

FSER for more detail.)

Basic system design criteria are applied to provide the first line of protection against events or

hazards posed by process fluids. Radiological fluids are transferred using gravity flow, airlifts,

air jets, steam jets, and siphons when practicable. Separator or “knockout pots” are needed in

lines in which fluid transfer is made by air or vacuum lift. The separated fluid is allowed to flow

by gravity into the desired component while the airflow vents at the top of the pot. This design

prevents backflow siphoning. Steam jet lift transfer system piping is terminated in the receiving

vessel vent space to provide an air gap that prevents backflow siphoning. Siphons are used to

initiate gravity transfer of fluids in applications where flow rate is not critical. The siphon

transfers liquid from the higher upstream tank to the lower downstream tank. The elevation

difference between tanks prevents backflow.

Hydraulic seals are used to prevent backflow of process fluid to auxiliary systems during reagent

addition. The liquid seal or plug is maintained by the piping configuration. The seal is

implemented by a “U” bend in piping or by hydraulic seal pots. The hydraulic seal design

ensures that the seal remains filled with liquid at all times, that the seal withstands internal

pressure differences between connected vessels, and that siphon action does not occur.

Check valves are used only in the process fluid pressure boundary. The check valve design is

based on effective pressure drop, type of seating material, pressure and flow reversal response

time, mounting requirements, and reliability and maintainability. Redundant isolation valves that

are PSSCs are used to automatically isolate utility and reagent fluids in the process area when

earthquake conditions are detected. These PSSC isolation valves close in the event of valve or

actuator failure. Isolation valve selection is based on process hydraulics, control system

characteristics, mounting requirements, and other valve specifications. These valves may be of

the butterfly, gate, plug, or ball type. The valves will be specified for service after consideration

of the chemical characteristics of the fluid, piping material of construction, and operating

conditions. The valves will be designed and constructed according to good engineering

practices and in accordance with applicable codes, such as the following:

• ASME B16.10, “Face-to-Face and End-to-End Dimensions of Valves,” issued 2000

(ASME, 2000) 

• API 598, “Valve Inspection and Testing,” issued October 1996 (API, 1996)
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• API 600, “Bolted Bonnet Steel Gate Valves for Petroleum and Natural Gas

Industries—Modified National Adoption of ISO 10434:1998,” issued October 2001 (API,

2001a)

• API 602, “Compact Steel Gate Valves—Flanged, Threaded, Welding, and Extended Body

Ends,” issued October 1998 (API, 1998)

• API 603, “Corrosion-Resistant, Bolted Bonnet Gate Valves—Flanged and Butt-welding

Ends,” issued April 2001 (API, 2001b)

• API 608, “Metal Ball Valves—Flanged, Threaded and Butt-welding Ends,” issued

September 1995 (API, 1995a)

• API 609, “Butterfly Valves: Double Flanged, Lug- and Water-Type,” issued May 1997

(API, 1997) 

The valves and their supports will also be designed to withstand and remain operable during the

design basis earthquake, as they are intended to prevent uncontrolled flooding of the fuel

fabrication building as a result of a seismic event. The safety function of the isolation valves is

to maintain safe isolation between controlled areas and uncontrolled areas that may contain

radioactive materials. Based on the proposed design and use of industry practices, the staff

finds that the redundancy and diversity of the fluid transport systems adequately prevent the

release of radioactive materials to the environment.

Evaluation of Welded Construction 

The regulatory acceptance criteria used for the review of this system specifically mentions that

tank and piping systems be of welded construction to the fullest extent possible. For process

equipment, radiological fluids are maintained within at least two levels of confinement. 

Components containing fluids that are located in process cells are specified with corrosion

allowances, and welded joints are radiographed, as appropriate. Fluid transfer systems

containing hazardous fluids are contained within trenches, rooms, or double-walled piping

systems or are accessible for inspection and are of a fully welded construction. Components

that are not of fully welded construction are installed in a glovebox. Piping components carrying

radiological fluids between two confinements are either of fully welded double-wall construction

or are installed in gloveboxes. Welding requirements are contained in the ASME B&PV Code,

Section VIII, 1996 Edition through the 1998 Addenda (ASME, 1998), ASME B31.3 (ASME,

2001a), and their referenced AWS codes. Welding work will be performed according to the FTS

category and quality assurance plan quality level of the system being welded. Requirements for

the qualification of welders and the welding process are specified in both the ASME B&PV Code,

Section VIII, Division 1 and 2, 1996 Edition through the 1998 Addenda, and the ASME B31.3

codes. Other general design and technical specifications that will be used for welding are ASME

Sections II, V, and IX; ASTM codes; and ANSI/AWS D-10.4 and B-3.0.

Material used for the construction of this equipment is specified in accordance with ASME and

ASTM material specifications. ASME materials are used in the fabrication of equipment and

piping components built to the requirements of ASME B&PV Code, Section VIII, 1996 Edition

through the 1998 Addenda, and ASME B31.3. ASTM materials are also used for the fabrication

of other components. In general, design of equipment to these standards means that the
11-146

components are designed for the most severe service conditions. Included in the severe service

conditions are pressure, temperature, stress, material compatibility, and corrosion. The staff has

reviewed the DCS design basis for welding and finds it acceptable, based on the information

submitted referencing appropriate codes and standards for the design and construction of the

fluid transport system.

Evaluation of Passive Features That Address Cross Contamination 

The staff evaluated the information provided by DCS in its revised CAR regarding the passive

features designed to prevent contamination of the fluid transport systems. The facility design

basis for the fluid transport systems contains a significant number of passive features that help

to prevent cross contamination.

Fluid transport components are hydraulically designed for the most severe service conditions. 

Radiological fluids are maintained within at least two levels of confinement. Fluid transfer

systems containing hazardous fluids are contained within trenches, rooms, or double-walled

piping systems or are accessible for inspection and are of a fully welded construction. All piping

components designated as PSSCs are designed to withstand the design basis earthquake

loads.

Separator or knockout pots are specified for piping in which fluid transfer is made by air or

vacuum lift. The separated fluid is allowed to flow by gravity into the desired component, while

the airflow vents at the top of the pot. This design prevents backflow siphoning. Steam jet lift

transfer system piping is terminated in the receiving vessel vent space to provide an air gap that

prevents backflow siphoning. Siphons are used to initiate gravity transfer of fluids in applications

where flow rate is not critical. The siphon transfers liquid from the higher upstream tank to the

lower downstream tank. The elevation difference between tanks prevents backflow. Knockout

pots, steam jet lifts, and elevation differences between tanks are passive features that help to

prevent cross contamination.

Hydraulic seals are used to prevent backflow of process fluid to auxiliary systems during reagent

addition. The seal is implemented by a “U” bend in piping or by hydraulic seal pots. The

hydraulic seal design ensures that the seal remains filled with liquid at all times, that the seal

withstands internal pressure differences between connected vessels, and that siphon action

does not occur. Hydraulic seals are additional passive features that help to prevent cross

contamination.

Check valves are used only in the process fluid pressure boundary. Redundant isolation valves,

which are PSSCs, are used to automatically isolate utility and reagent fluids in the process area

when earthquake conditions are detected. The isolation valves close in the event of valve or

actuator failure. Check valves in the process fluid pressure boundary and isolation valves that

fail to the safe position are examples of passive safety features that help to prevent cross

contamination. DCS has committed to the design basis previously discussed. The staff has

reviewed the facility design basis for cross contamination and, based on nuclear industry

experience, concludes that the design is acceptable.
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Evaluation of Radiation Safety 

The staff review and evaluation of the Radiation Safety Program are discussed in detail in

Chapter 9 of this FSER. In general, the design basis for the fluid transport systems hardware for

radiation safety is as follows:

 

• Process equipment is designed to handle fissile material in accordance with radiation

safety principles. 

• Stainless steel and other material designed to be compatible with process fluids will be

used to prevent corrosion. 

• Parts are easily visible and accessible for cleaning, and the material specified has

appropriate surface quality or coatings, if needed. 

• Process cell drip trays will be designed to contain the contents of the largest vessel in the

process cell in a criticality-safe configuration. 

• Decontamination will be done by flushing and high-pressure washing to remove sediment

buildup or blockages, should they occur. 

• The fluid transport systems layout is intended to minimize the potential for entrapment and

buildup of radioactive materials. 

Based on the DCS commitment to design the system to minimize the entrapment and buildup of

radioactive materials (CAR Section 11.8.3.4), the staff finds this design to be acceptable.

Evaluation of Corrosion Resistance 

The facility fluid transport systems construction materials are selected based on compatibility

with the physical and chemical characteristics of the process fluids. In general, stainless steel of

type 304L or 316L and alloys of titanium and zirconium are used for FTS Category 1

components. Components of FTS Category 2 and 3 that handle acidic or alkaline fluids are

generally constructed from 304L or 316L stainless steel. Material used for the construction of

this equipment is specified in accordance with ASME and ASTM material specifications. ASME

materials are used in the fabrication of equipment and piping components built to the

requirements of ASME B&PV Code, Section VIII, 1996 Edition through the 1998 Addenda

(ASME, 1998), and the ASME B31.3 (ASME, 2001a) code. ASTM materials are also used for

the fabrication of other components. In general, design of equipment to these standards means

that the components are designed for the most severe service conditions. Included in the

severe service conditions are pressure, temperature, stress, and corrosion. Based on the DCS

commitment to design the system to be resistant to corrosion, and on the commitment to use

material maintenance and surveillance programs to detect and limit the damage from corrosion

(as discussed in revised CAR Section 5.5.2.1.6.2), the staff finds this design to be acceptable.
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Evaluation of Personnel Protection

NUREG-1718 (NRC, 2000) requires the evaluation of the need for hoods, gloveboxes, and

shielding for personnel protection. These items are generally required for wet processing

operations involving more than gram quantities of plutonium or general operations involving 50

micrograms or more of plutonium in respirable form. Process cells contain equipment that

handles radioactive materials in chemical solutions; this equipment is of a fully welded

construction and does not require routine maintenance. Equipment containing radioactive

materials in the powder process (MP) is contained in gloveboxes in process rooms that provide

equivalent confinement to fully welded equipment in process cells. The staff has reviewed the

DCS list of equipment and agrees that the equipment involving more than gram quantities of

plutonium or general operations involving 50 micrograms or more of plutonium in respirable form

are properly contained in either gloveboxes or fully welded process equipment. Therefore,

based on the facility design following the guidelines of NUREG-1718, the staff finds this design

basis to be acceptable.

Evaluation of Functionality During Severe Natural Phenomena 

Fluid transport systems designated as PSSCs are designed and qualified according to national

codes and standards enabling them to perform their safety function during normal operations,

upset conditions, and design basis events. These codes and standards are delineated in Tables

11.8-1, 11.8-2, and 11.8-3 of this FSER. The seismic monitoring system is designed to satisfy

the criteria provided in RG 3.17-1974 (NRC, 1974). The design basis of the seismic monitoring

system is that it provides sufficient data to evaluate the response of the confinement structure

and other PSSCs to a seismic event and initiate a shutdown of process systems in the event of a

high seismic event. The seismic system will meet the requirements of IEEE Standard 603,

“IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations,” issued

1998 (IEEE, 1998).

The SQ requirements are applied using a graded classification program that considers the

relative importance of the safety function and the structural behavior of the PSSC. (See the

discussion of quality assurance classes and grading in Section 15.1 of this FSER for details on

implementation.) This classification system is defined in RG 3.14, “Seismic Design

Classification for Plutonium Processing and Fuel Fabrication Plants,” issued 1973 (NRC, 1973). 

The SQ of MFFF mechanical equipment will meet the guidance of RG 1.100, Revision 2 (NRC,

1998).

The SC-I classification is applicable to facility SSCs and the supporting SSCs that are required

to withstand the effects of an earthquake and remain functional to the extent that they will

prevent the uncontrolled release of radioactive materials. The SC-I classification applies to all

PSSCs that must perform a safety function during and/or after the design basis earthquake to

comply with the facility safety assessment as discussed in Chapter 5 of this FSER. The SC-II

designation applies to the portions of systems whose continued function is not required but

whose failure could reduce the functioning of any plant feature of an SC-I SSC. Items that are

neither SC-I or SC-II are not classified with respect to SC. Components that form an interface

between SC-I and non-SC-I components are to be classified as SC-I. The quality assurance

(QA) plan will apply to these components as previously discussed. Based on the applicant’s

commitments to design and qualify fluid transport system PSSCs to perform their safety

functions as discussed in this section, the staff finds this design basis to be acceptable.
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11.8.2 Evaluation Findings

In Section 11.8 of the revised CAR, DCS provided design basis information for the fluid transport

systems that it identified as PSSCs for the proposed facility. Based on the staff's review of the

revised CAR and supporting information provided by the applicant relevant to the fluid transport

systems, the staff concludes, pursuant to 10 CFR 70.23(b), that the design bases of the PSSCs

evaluated in this FSER section will provide reasonable assurance of protection against natural

phenomena and the consequences of potential accidents.

11.8.3 References

(AISC, 1994) American National Standards Institute/American Institute of Steel Construction. 

N690, “Specification for the Design, Fabrication, and Erection of Safety Related Steel Structures

for Nuclear Facilities.” 1994.

(API, 1995a) American National Standards Institute/American Petroleum Institute. Standard

608, “Metal Ball Valves—Flanged, Threaded and Butt-welding Ends.” September 1995.

(API, 1995b) American National Standards Institute/American Petroleum Institute. Standard

610, “Centrifugal Pumps for Petroleum, Heavy Duty Chemical and Gas Industry Services.”

August 1995.

(API, 1996) American National Standards Institute/American Petroleum Institute. Standard 598,

“Valve Inspection and Testing.” October 1996.

(API, 1997) American National Standards Institute/American Petroleum Institute. API Standard

609, “Butterfly Valves: Double Flanged, Lug- and Wafer-Type.” May 1997.

(API, 1998) American National Standards Institute/American Petroleum Institute. Standard 602,

“Compact Steel Gate Valves—Flanged, Threaded, Welding, and Extended Body Ends.” 

October 1998.

(API, 2001a) American National Standards Institute/American Petroleum Institute. Standard

600, “Bolted Bonnet Steel Gate Valves for Petroleum and Natural Gas Industries— Modified

National Adoption of ISO 10434:1998.” October 2001.

(API, 2001b) American National Standards Institute/American Petroleum Institute. Standard

603, “Corrosion-Resistant, Bolted Bonnet Gate Valves—Flanged and Butt-welding Ends.” April

2001.

(ASME, 1998) American Society of Mechanical Engineers Boiler and Pressure Vessel Code

(B&PV Code). Section VIII, “Design and Fabrication of Pressure Vessels.” 1996 Edition through

the 1998 Addenda.

(ASME, 2000) American Society of Mechanical Engineers. ASME B16.10, “Face-to-Face and

End-to-End Dimensions of Valves.” 2000.

(ASME, 2001a) American Society of Mechanical Engineers. ASME B31.3, “Process Piping.” 

2001.
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(ASME, 2001b) American Society of Mechanical Engineers. ASME B73.1, “Specifications for

Horizontal End Suction Centrifugal Pumps for Chemical Process.” 2001.

(ASME, 2001c) American Society of Mechanical Engineers. ASME B73.2, “Specifications for

Vertical In-line Centrifugal Pumps for Chemical Process.” 2001.

(CGA, 2001a) Compressed Gas Association. Standard S-1.1, “Pressure Relief Device

Standards, Part 1—Cylinders for Compressed Gases.” 2001.

(CGA, 2001b) Compressed Gas Association. Standard S-1.3, “Pressure Relief Device

Standards, Part 3—Stationary Storage Containers for Compressed Gases.” 2001.

(DCS, 2002a) Ihde, R., Duke Cogema Stone & Webster. Letter to Document Control Desk,

U.S. Nuclear Regulatory Commission, RE: Mixed Oxide Fuel Fabrication Facility—Construction

Authorization Request. October 31, 2002.

(DCS, 2002b) Hastings, P., Duke Cogema Stone & Webster. Letter to Document Control Desk,

U.S. Nuclear Regulatory Commission, RE: Clarification of Responses to NRC Request for

Additional Information. February 11, 2002.

(IEEE, 1987) Institute of Electrical and Electronics Engineers. Standard 344, “IEEE

Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear Power

Generating Stations.” 1987. 

(IEEE, 1998) Institute of Electrical and Electronics Engineers. Standard 603, “IEEE Standard

Criteria for Safety Systems for Nuclear Power Generating Stations.” 1998.

(NFPA, 2003) National Fire Protection Association. Standard 20, “Standard for the Installation

of Stationary Pumps for Fire Protection.” 2003.

(NRC, 1973) U.S. Nuclear Regulatory Commission. Regulatory Guide 3.14, “Seismic Design

Classification for Plutonium Processing and Fuel Fabrication Plants.” Washington, DC, 1973.

(NRC, 1974) U.S. Nuclear Regulatory Commission. Regulatory Guide 3.17, “Earthquake

Instrumentation for Fuel Reprocessing Plants.” Washington, DC, 1974.

(NRC, 1998) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.100, Rev. 2, “Seismic

Qualification of Electric and Mechanical Equipment for Nuclear Power Plants.” Washington, DC,

1988.

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility.” Washington,

DC, August 2000.
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11.9 Fluid Systems (Bulk Materials, Reagents, and Gases)

11.9.1 Conduct of Review

This chapter of the FSER contains the staff’s review of the fluid systems described by the

applicant in Chapter 11.0 of the revised CAR (DCS, 2002a). The objective of this review is to

determine whether the fluid systems’ PSSCs and their design bases identified by the applicant

provide reasonable assurance of protection against natural phenomena and the consequences

of potential accidents. The staff evaluated the information provided by the applicant for fluid

systems by reviewing Chapter 11.0 of the revised CAR, other sections of the revised CAR,

supplementary information provided by the applicant, and relevant documents available at the

applicant’s offices but not submitted by the applicant. The review of fluid systems’ design bases

and strategies was closely coordinated with the review of other chapters in this FSER (fire

protection in Chapter 7, chemical safety in Chapter 8, and safety assessment of the design

bases in Chapter 5).

The staff reviewed how the information in the revised CAR addresses the following regulation:

• 10 CFR 70.23(b) requires that the design bases of the PSSCs and the Quality Assurance

Program provide reasonable assurance of protection against natural phenomena and the

consequences of potential accidents before construction of the PSSCs is approved.

The review for this construction approval focused on the design bases of fluid systems, their

components, and other related information. For fluid systems, the staff reviewed and evaluated

the information provided by the applicant for the safety function, system description, and safety

analysis. The review also encompassed proposed design basis considerations such as

redundancy, independence, reliability, and quality. The staff used Chapter 11.0 in NUREG-1718

(NRC, 2000) as guidance in performing the review.

The discussions that follow provide the system descriptions, as well as function, major

components, control concepts, and system interfaces. These discussions include but are not

limited to PSSCs to provide an understanding of the system. Design bases of the PSSCs are

provided in Section 11.9.1.2 of this FSER.

The NRC staff reviewed the revised CAR submitted by DCS for the following areas applicable to

the fluid systems at the construction approval stage and consistent with the level of design:

• system description

• system function

• major components

• control concepts

• system interfaces

• design bases

Regarding the MFFF fluid systems, specific design considerations given in the CAR are to

demonstrate adherence to regulations, and the acceptance criteria from Section 11.4.4.2 of

NUREG-1718, set forth below:
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• transfer of heat loads to an appropriate heat sink under normal, off-normal, and accident

conditions

• adequate water supply under normal, off-normal, and accident conditions

• adequate component redundancy; the capability to isolate components, systems, or piping

for maintaining system safety function under varying system configuration; and the

capability of integrated system control

• the applicant’s commitment to supporting management measures (including tests and

other verification methods) that ensure the structural integrity and system leak-tightness

(including the prevention of cross contamination (radioactive and chemical)), the

operability and adequate performance of active system components, and the capability of

the system to perform required functions during normal and accident situations

• capability for withstanding environmental hazards resulting from pipeline breaks and

dynamic effects associated with flow instability and attendant loads and measures to

prevent such dynamic conditions from occurring

• capacity and capability for detecting leaks and cross contamination (radioactive and

chemical), for inservice component inspection and system maintenance, and for

operational functional testing of the system and its components

• system capability to maintain functionality when subject to tornadoes, tornado missiles,

earthquakes, floods, and any other appropriate severe natural phenomena

11.9.1.1 Description (Bulk Material, Reagent, and Gas Systems)

The facility fluid systems consist of bulk material, reagent, and gas systems designed to support

the production function of the facility. DCS has identified the seismic isolation valves, EDGFOS,

and the emergency scavenging air system as PSSC fluid systems. Non-PSSC systems may

contain PSSC components, such as seismic isolation valves. Because these non-PSSC

systems contain and could interact with PSSCs, they are summarized but not evaluated in this

section. Only PSSC systems are evaluated. The fluid systems are divided into three

generalized categories—mechanical utility systems, reagent systems, and bulk gas systems.

Mechanical Utility Systems 

The mechanical utility systems consist of the HVAC chilled water system, process chilled water

system, demineralized water system, process hot water system, process steam and condensate

systems, plant water system, the EDGFOS, standby diesel generator fuel oil system, service air

system, instrument air system (which includes the emergency scavenging air subsystem),

breathing air system, and the radiation monitoring vacuum system. Besides those systems

mentioned at the beginning of this section, most of these systems are non-PSSCs but contain

components, such as seismic isolation valves, that are PSSCs.

The HVAC chilled water system consists of an external loop to supply chilled water to the HVAC

supply air system fan coils and an internal cooling loop to provide cooling water to individual

area fan coils. The primary ventilation cooling coils are located upstream of the building intake
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HEPA filters, thereby placing them outside of the building radiological boundary. Placing them in

this location allows the condensate formed to be discharged directly. The process chilled water

system consists of an external cooling loop that provides cooling water to multiple intermediate

heat exchangers. In-leakage to the system will be monitored for contamination by radiation

detectors located on a continuous bypass flowpath in the common chilled water return line of

each internal cooling loop. The demineralized water system receives, stores, and transfers

pressurized and gravity-fed demineralized water to process equipment and utility systems. The

process hot water system is a closed-loop circulating system that supplies hot water to process

equipment in the solvent recovery and purification cycles and the oxalic precipitation and

oxidation units. The process steam system transfers and regulates the primary steam supplied

by SRS and the acid recovery unit secondary steam to the AP area. The process condensate

system collects and transfers primary condensate and acid recovery unit secondary condensate. 

Radiation detection is included in the design for the secondary steam/condensate closed loop

systems. The plant water system consists of a supply header that supplies industrial-grade

water from SRS to utility, MP, and AP users. The standby diesel generator fuel oil system

supplies fuel oil to the standby generators that are used to provide an onsite power source for

major electrical loads. The service air system supplies pressurized air to the facility service air

headers for maintenance and utility use and makeup to the instrument air system. The

breathing air system is an independent air supply system that is used for personnel breathing in

an emergency. The radiation monitoring vacuum system consists of multiple, parallel vacuum

pumps, a common header, and associated piping connected to parallel continuous air monitors

and samplers designed to evaluate airborne radioactivity in the facility.

The HVAC chilled water system, process chilled water, demineralized water, process steam and

condensate systems, and plant water system contain seismic isolation valves, designated as

PSSCs, that close during a seismic event to prevent flooding in the facility.

The EDGFOS is a PSSC that supports the EDGs. This system receives, samples, stores, and

supplies fuel oil to the EDGs. The EDGFOS consists of two independent fuel oil subsystems,

each capable of fuel receipt, long-term storage, and transferring an adequate supply of fuel oil to

its associated EDG operating at 100-percent load continuously for 7 days. The tanks will be

protected in a below-grade concrete vault adjacent to the EDG building. Each supply line has its

own storage tank, transfer pump, day tank, purification system, strainers, filters, instrumentation

and control elements, and piping. The supply lines will be installed in a protective enclosure

located adjacent to the EDG. The EDGFOS will supply No. 2 fuel oil to the day tanks that are

sized to store 2500 L (660 gal) each. The EDG exhaust system is a PSSC and is made up of

piping and silencers supplied in accordance with the manufacturer’s general design and supplied

as part of the vendor-supplied skid-mounted generator set. The function of the EDGFOS is to

receive, store, and transfer fuel oil to the EDGs used to provide emergency power to PSSCs in

the facility. The function of the EDG exhaust system is to remove fumes and dampen noise

produced by the EDGs. The major components of the EDGFOS are the fuel oil storage tank,

the fuel oil day tanks, the fuel oil transfer pumps, isolation valves, and the EDG exhaust system. 

The EDGFOS employs fuel oil day tank level control and various level instrumentation and

alarms. The EDGFOS interfaces with the EDGs and the emergency electrical power supply.

The instrument air system is a non-PSSC and is designed to supply instrument air to the

instruments, air-operated pneumatic valves and HVAC dampers, scavenging air for plutonium

vessel level measurement and equipment atmosphere scavenging, cooling air for gloveboxes

and pelletizing press bellows, and nitrogen system backup. Instrument air is a dried and filtered
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air specifically designed for instrumentation, valves, and emergency scavenging air; the

emergency scavenging air subsystem is a PSSC. Instrument air is further dried for use as

bubbling air for scavenging AP gloveboxes and miscellaneous equipment. A portion of the

instrument air will be further dried to get super-dry process air for various instruments and for air

being supplied directly to the AP process stream and in glovebox applications. The instrument

air system will be designed to meet or exceed the standards of ANSI/ISA S7.0.01-1996, “Quality

Standard for Instrument Air.” 

The facility seismic isolation valves, which are PSSCs, automatically isolate during a seismic

event. This system interfaces with the service air system, very high depressurization exhaust

system, I&C system, and seismic detectors. The applicant has considered related industry

experience in developing the design basis of non-safety-related air systems that support safetyrelated equipment at the facility.

Reagent Systems 

The reagent systems consist of the nitric acid system, the silver nitrate system, TBP system,

HAN system, sodium hydroxide system, oxalic acid system, diluent system, sodium carbonate

system, hydrogen peroxide system, hydrazine system, manganese nitrate system,

decontamination system, nitrogen oxide system, aluminum nitrate system, zirconium nitrate

system, and uranyl nitrate system. In general, these systems are not PSSCs but contain

components, such as isolation valves, that are PSSCs.

The nitric acid system provides 13.6 N nitric acid to the AP process and for the preparation of

hydrazine, oxalic acid, manganese nitrate, and silver nitrate reagents. The acid is stored in tote

tanks in the reagent processing building. The system includes isolation valves, designated

PSSCs, that automatically isolate during a seismic event.

The silver nitrate system provides silver nitrate to the electrolyzers in the dissolution unit. Silver

nitrate, demineralized water, and 13.6 N nitric acid are mixed in a preparation tank in the reagent

processing area and pumped to a distribution tank in the AP area.

The TBP system provides TBP for solvent extraction in the purification cycle of the AP process

and solvent washing in the solvent recovery cycle. The system includes isolation valves,

designated PSSCs, that automatically isolate during a seismic event.

The HAN system supports the stripping of plutonium in the AP or purification process. The

system includes isolation valves, designated PSSCs, that automatically isolate during a seismic

event. The HAN system interfaces with the normal power system, seismic detectors, and the

purification cycle.

The sodium hydroxide system makes the sodium hydroxide soda for washing in the solvent

recovery cycle. The sodium hydroxide system interfaces with the solvent recovery cycle and the

normal power supply.

The oxalic acid system provides oxalic acid for converting plutonium nitrate to plutonium oxalate

in the oxalic precipitation and oxidation unit. The oxalic acid system design includes isolation

valves, designated PSSCs, that automatically isolate during a seismic event. The oxalic acid
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system interfaces with the normal power system, seismic detectors, nitric acid system,

demineralized water system, and oxalic precipitation and oxidation unit.

The diluent system supplies diluent (branched dodecane) for washing in the purification and

solvent recovery cycles and diluent for preparation of the 30-percent TBP solvent solution. This

system interfaces with the purification cycle, solvent recovery cycle, TBP system, and normal

power system. The diluent system design includes isolation valves, designated PSSCs, that

automatically isolate during a seismic event.

The sodium carbonate system is used in the solvent recovery cycle to wash and adjust the pH in

the liquid waste reception unit in the AP process. Isolation valves, designated PSSCs,

automatically isolate during a seismic event. The sodium carbonate system interfaces with the

normal power system, seismic detectors, liquid waste reception unit, demineralized water

system, and solvent recovery cycle.

The hydrogen peroxide system is designed to provide hydrogen peroxide for valence adjustment

of the dissolution solution in the dissolution unit. Isolation valves, designated PSSCs,

automatically isolate during a seismic event. The hydrogen peroxide system interfaces with the

normal power system, seismic detectors, and the silver recovery unit.

The hydrazine system supplies hydrazine hydrate to be used in the preparation of hydrazine

nitrate. The hydrazine nitrate is mixed with HAN and used in the purification cycle of the AP

process. Tanks are vented to a hydrogen peroxide scrubber system before release to the

atmosphere. All components of this system are located in the reagent processing building and

do not require seismic isolation from the material or AP buildings.

The manganese nitrate system supplies manganese nitrate to the oxalic mother liquor recovery

unit. The manganese nitrate is used as a catalyst. The manganese nitrate system interfaces

with the normal power supply, oxalic mother liquor recovery unit, and the nitric acid system.

The decontamination system supplies decontamination solution to the AP process. The

decontamination system interfaces with the nitric acid system, the demineralized water system,

the oxalic mother liquor recovery unit, the normal power supply, purification cycle, solvent

recovery cycle, homogenization unit, acid recovery unit, silver recovery unit, liquid waste

reception unit, dissolution unit, and offgas treatment unit.

The nitrogen oxide system supplies nitrous fumes (nitrogen dioxide and dinitrogen tetroxide) to

the AP process to assist in the oxidation of plutonium nitrate in the purification cycle. Nitrogen

oxide detectors will be installed in rooms where fumes could be present. Isolation valves,

designated PSSCs, automatically isolate during a seismic event. The system interfaces with the

purification cycle, normal power supply, instrument air system, and seismic detectors.

The aluminum nitrate system supplies aluminum nitrate to the laboratory where it is mixed with

1.5 N nitric acid. The mixture is added to the aluminum nitrate buffer tank and transferred by

dosing pump to a purification cycle seal pot. The aluminum nitrate system interfaces with the

normal power supply, nitric acid system, and purification cycle.

The zirconium nitrate system supplies zirconium nitrate to avoid fluoride corrosion of titanium

vessels in the purification cycle and acid recovery unit. The zirconium nitrate penetrations into
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the MFFF are provided with double isolation valves to automatically isolate during a seismic

event.

The uranyl nitrate system supplies depleted uranyl nitrate to the plutonium feed stream and

uranium waste stream for reducing the isotopic composition of uranium. This system interfaces

with the nitric acid system, demineralized water system, dechlorination dissolution and

dissolution units, purification unit, normal power supply, and decontamination system.

Bulk Gas Systems 

The bulk gas systems consist of the nitrogen system, the argon/hydrogen system, helium

system, oxygen system, and methane/argon (P10) system. In general, these systems are not

PSSCs but contain components, such as seismic isolation valves, that are PSSCs.

The nitrogen system is a non-PSSC system suppled by an onsite nitrogen production system.

Liquid nitrogen is stored in a tank in the MP area to supply germanium to the operating and

reserve detectors. Nitrogen gas is also used for MP area glovebox inerting to eliminate the

adverse effects of atmospheric oxygen on the process or fuel, and also for scavenging the

sintering furnace airlock. The sintering furnace airlock transfers molybdenum boats with

sintered pellets from an argon/hydrogen atmosphere, and the nitrogen flow in the airlock is

intended to eliminate the argon/hydrogen atmosphere before returning to the MP gloveboxes. 

The graphite rear bearing of the calcination furnace is cooled and the atmosphere inerted with

nitrogen to lengthen the life of the bearing. DCS stated that this bearing provides a containment

function. Nitrogen gas is also used to scavenge the hydrazine and hydroxylamine tanks and is

designed to eliminate the concern for the flammability of hydrazine vapors in these tanks. A

liquid nitrogen storage tank and ambient air vaporizer, designed to supply at least 2 days supply

of nitrogen, provide backup for the onsite nitrogen production system. The nitrogen system is

designed to operate at ambient temperature, at a pressure of 300–790 kPa (3–7.9 bar) and a

flow rate of 495–561 m3/h (17,500–19,800 ft3/h). The estimated annual usage of nitrogen is

103,900 m3 (3.67 million ft3). The nitrogen system must supply gas to the gloveboxes to meet

the required oxygen impurity level inside the glovebox of less than 2000 ppm. Two nitrogen

buffer tanks are included in the design with 34.2 m3 (1209 ft3) and 0.31 m3 (11 ft3) capacities. 

The backup nitrogen storage tank is 34 m3 (9000 gal) capacity. In CAR Section 11.9.2.1.1, DCS

stated that none of the functions of the nitrogen system are credited in the facility safety

analyses, and the system is therefore not designated as a PSSC.

The argon/hydrogen system is a non-PSSC system that provides argon/hydrogen mixture to an

operating electric sintering furnace, to furnace airlocks, and to the laboratory. The system will

be designed to NFPA 50A, “Standard for Gaseous Hydrogen Systems at Consumer Sites.” The

argon/hydrogen system is designed to mix the gases to make a 95:5 mixture (a proprietary

range is given) used in scavenging the operating electric sintering furnace and airlocks in the MP

area. It is also used in the laboratory. An emergency system, made of argon/hydrogen in

compressed cylinders at 4.3 to 8 percent hydrogen, serves as backup to the normal supply

system. This emergency supply of pre-mixed argon and hydrogen is stored in a tube trailer with

a capacity of 1500 m3 (53,000 ft3) for a 24-hour backup supply. Pure argon provides secondary

backup and pure nitrogen provides tertiary backup to the argon/hydrogen system. The argon is

supplied from two argon vaporizing packages sourced from two 11 m3(3000 gal) bulk liquid

argon storage tanks, for a 2-week supply. Hydrogen gas is supplied from a tube trailer holding

1200 m3 (43,000 ft3) located in the facility gas storage area. The gas is supplied at 20,000 kPa
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(200 bar) and pressure-reduced in two stations to normal operating pressure before being mixed

with argon. The mixture of the two gases is controlled by PSSC proportioning control valves and

a PSSC static mixer. The hydrogen concentration in the buffer tank is continuously monitored

by redundant detectors, identified as PSSCs, to ensure that the lower explosion limit is not

exceeded. The argon/hydrogen mixture is supplied to the operating furnace at 200 kPa (2 bar). 

The argon/hydrogen system is designed to operate at ambient temperature, at a pressure of

20,000 kPa (200 bar) and a flow rate of 68 m3/h (2400 ft3/h). The estimated annual usage of

argon in the sintering furnaces and laboratory is 462,000 m3 (16.3 million ft3), and the estimated

annual usage of hydrogen in the sintering furnace is 11,400 m3 (403,000 ft3). The major

components of the argon/hydrogen system are the liquid argon bulk storage system, the

hydrogen tube trailer, the backup argon/hydrogen cylinders, inline mixing stations, and the buffer

tanks, pressure transmitters, pressure control valves, alarms, piping, and isolation valves. The

argon/hydrogen system is controlled by argon and hydrogen levels after mixing, automatically

switching from normal supply to backup supply on low pressure. Gas flow is maintained by flow

control valves. Pressure control valves regulate gas pressure to the sintering furnaces and

laboratory, redundant hydrogen monitors, and the facility seismic isolation valves, designated as

PSSCs, that actuate in the event of a seismic event. The argon/hydrogen system interfaces with

the sintering furnaces, normal power supply, standby power supply, laboratory, and facility

seismic detectors.

The helium system is supplied from tube trailers located in the gas storage area. The safety

functions of this system are to isolate during a seismic event and prevent backflow into the

system. The major components of the helium system are the tube trailer, pressure regulator,

pressure switches, alarms, relief valves, isolation valves, piping, and switchover circuit. 

Penetrations into the facility have seismic isolation valves, designated as PSSCs, that

automatically isolate the system during a seismic event. The helium system interfaces with the

rod cladding and decontamination unit, rod welding scavenging, laboratory, and seismic

detectors.

The oxygen system provides oxygen from cylinders, stored in the north end of the loading dock

at the reagent processing building, to the calcination furnace in the homogenization unit in the

AP area and to the laboratory in the MP area. The major components of this system are the

cylinders, pressure regulators, pressure switches, alarms, three-way valve, relief valves,

isolation valves, piping, and instrumentation. The safety functions of this system are to isolate

during a seismic event and prevent backflow into the system. The oxygen system interfaces

with the calcination furnace, the MP area laboratory, and seismic detectors that actuate the

seismic isolation valves, designated as PSSCs, during seismic events.

The methane/argon (P10) system is composed of 10 percent methane and 90 percent argon. 

DCS states that the system is supplied by one tube trailer holding a 6-week supply (1300 m3

(45,000 ft3) of argon-methane gas). The gas is found in the gas storage area and the

laboratory.

11.9.1.2 Design Bases of the PSSCs

This section lists the referenced design basis codes and standards for proposed facility fluid

systems identified as PSSCs.
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Design Bases for PSSCs 

Vessels/components are segregated/separated from incompatible chemicals. The chemical

makeup of the reagent introduced into cells or AP reagent rooms is controlled to prevent

explosions caused by chemical reactions. All piping, chemicals, and equipment will be clearly

marked to prevent errors. Pressure vessels will be located away from PSSCs or otherwise

protected so that a failure of any vessel will have no impact on the ability of the PSSC to perform

its safety function.

Design Basis for Seismic Isolation Valves

Piping penetrations in the facility process building walls that could pose a fire, explosion,

confinement, or flooding risk, or that could leak radioactive materials into the environment, will

be isolated by redundant, automatic, seismically qualified isolation valves. The redundant

isolation valves are designed to receive isolation signals from redundant seismic monitors. An

isolation will occur when a signal is sent to the pneumatic operator of the isolation valve to vent. 

Valve position indicators will be included in the design of the process and utility control room. 

The seismic monitoring system is designed to satisfy the criteria provided in RG 3.17-1974

(NRC, 1974). The design basis of the seismic monitoring system is that it provides sufficient

data to evaluate the response of the confinement structure and other PSSCs to a seismic event

and initiate a shutdown of process systems in the event of a high seismic event. The seismic

system will meet the requirements of IEEE Standard 603 (IEEE, 1998). The SQ of these valves

will be done in accordance with IEEE Standard 344 (IEEE, 1987). (See discussion of the

acceptability of this standard in Section 11.5 of this FSER and NRC RG 1.100, Revision 2 (NRC,

1988).) These valves have been identified as PSSCs and will be handled according to the

facility Quality Assurance Program quality level appropriate to each application of an isolation

valve.

The seismic isolation valves can be of the butterfly, gate, plug, or ball type. The design is

determined based on the chemical characteristics of the fluid, the construction material for

associated piping, and operating conditions. The valve design will use good engineering

practices and will be in accordance with applicable code practices. The valves will be supported

in such a way as to remain operable during design basis events.

Instrument air quality is very important to maintaining the reliability of pneumatic isolation valves. 

DCS has committed to design, operate, and maintain the instrument air system to the standard

given in ANSI/ISA S7.0.01-1996 or better.

Evaluation of Seismic Isolation Valves 

During this evaluation, the staff reviewed applicable Federal regulations, national codes and

standards, the MOX SRP, other industry and NRC staff guidance, and available operational

history. The staff’s consideration and use of these documents are discussed in the following

evaluation.

As discussed previously, piping penetrations in the facility process building walls that could pose

a fire, explosion, confinement, or flooding risk, or that could leak radioactive materials into the

environment, will be isolated by redundant, automatic, seismically qualified isolation valves. 

These valves have been designated as PSSCs. They will receive signals to isolate from
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redundant seismic monitors. The valves will be ordered, built, delivered, installed, and used in

accordance with the facility Quality Assurance Program quality level appropriate to each

application. 

The staff evaluation of the design basis for the seismic isolation valves began with a

review of the DCS proposed design basis against national codes and standards and industry

practices. The staff also applied the guidance of the MOX SRP to the review.

The isolation of fluid systems from the MFFF building is achieved by redundant seismic isolation

valves that are designed to fail to a closed position. An isolation will occur when a signal is sent

to the pneumatic operator of the isolation valve to vent. The venting of the air from the operator

removes the power from the system, resulting in the operator moving to the closed position. 

Valve position indicators will be included in the design of the process and utility control room. 

The SQ of these valves will be done in accordance with IEEE Standard 344 (IEEE, 1987). (See

discussion of the acceptability of this standard in Section 11.5 of this SER and NRC RG 1.100,

Revision 2 (NRC, 1988). DCS has committed, in its mechanical equipment SQ program (CAR

Section 11.12), to consider attached piping loads, thermal loads, and live loads such as tank

sloshing in its design. The applicant has committed to using standards for the SQ of isolation

valves which are commensurate with standards applicable to commercial nuclear power stations. 

The staff finds these commitments acceptable.

The isolation valves will be designed to passively return to a closed, or isolated, position in the

event of a failure of the valve actuator or the air supply, or a loss of power. Based on the varying

sizes, designs, materials of construction, methods of operation, and fail-safe design, the staff

finds that the seismic isolation valves have the redundancy and diversity of components required

for isolation and to prevent release of radioactive material to the environment.

DCS has committed to design fluid systems, including the seismic isolation valves and other

components (CAR Section 11.8.7), with consideration to the process fluids contained therein

and will specify materials with the appropriate corrosion resistance. The applicant will also

consider proper surface finishes and decontamination characteristics for their particular

application in a system. Based on these considerations and their alignment with the

requirements of the ASME B&PV Code, Section VIII (ASME, 1995), for materials and design

methodology, the staff finds this design basis to be acceptable.

The seismic isolation valves will be designed and qualified to maintain functionality when

subjected to severe natural phenomena such as tornadoes, tornado missiles, earthquakes,

floods, and any other appropriate phenomena as established in the ISA. The seismic isolation

valves will actuate on a seismic event that is one-third of the design basis earthquake ground

motion, that is one-third of the safe-shutdown earthquake ground motion. This design basis

meets the requirements for nuclear power generation facilities given in Section IV(a)(2)(i)(A) of

Appendix S, “Earthquake Engineering Criteria for Nuclear Power Plants,” to 10 CFR Part 50,

“Domestic Licensing of Production and Utilization Facilities.” The staff finds this design basis to

be acceptable, since the facility design for actuation of the seismic isolation valves is based on

the ground motion found to be acceptable at nuclear power generation facilities.

The seismic monitoring system is designed to satisfy the criteria provided in RG 3.17-1974. The

design basis of the seismic monitoring system is that it provides sufficient data to evaluate the

response of the confinement structure and other PSSCs to a seismic event and initiate a
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shutdown of process systems in the event of a high seismic event. Because the seismic system

will meet the requirements of IEEE Standard 603 (IEEE, 1998), the staff finds this design basis

to be acceptable.

Instrument air quality is important to maintaining the reliability of pneumatic isolation valves. 

DCS has committed to design, operate, and maintain the instrument air system to the standard

given in ANSI/ISA S7.0.01-1996 (as described in CAR Section 11.9.1.10). In general, the staff

agrees with the use of the instrument air system as the motive force for the seismic isolation

valves. On the basis of the applicant’s commitment to ANSI/ISA S7.0.01-1996, and the

applicant’s consideration of relevant industry experience with nonsafety-related air systems

(discussed in Section 11.9.1.1 of this FSER), the staff finds this design basis to be acceptable.

Design Basis of the Emergency Diesel Generator Fuel Oil System 

Both the main emergency fuel oil supply vault and the EDG building are designed to withstand

hazardous natural phenomena such as earthquakes, floods, and tornadoes. The EDG piping

and equipment are seismically qualified to prevent failures from causing a loss of emergency

electrical power supply. The SQ will be performed in accordance with IEEE Standard 344-1987

(IEEE, 1987) (see discussion of the acceptability of this standard in Section 11.5 of this FSER,

and NRC RG 1.100, Revision 2 (NRC, 1988).) Each EDGFOS has a bulk tank that is sized for 7

days of use (plus extra operational margin for testing). The storage capacity was sized for an

EDG running at 100- percent power for 7 days plus some reserve capacity. The EDG design,

with redundant transfer pumps, fuel oil lines, associated piping, and exhaust equipment, is

intended to prevent a single failure from causing a loss of the emergency electrical power

supply. This equipment has been identified as a PSSC and will be handled according to the

facility Quality Assurance Program quality level appropriate to the importance of each piece of

equipment. The equipment will comply with the requirements of IEEE Standard 308-1991

(IEEE, 1991); ANS 59-51-1997, “American Nuclear Society Fuel Oil Systems for Safety Related

Emergency Diesel Generators”; and ASTM D75-94, “Standard Specification for Diesel Fuel

Oils,” issued 1974 (ASTM, 1974). The EDG exhaust system function is to collect and remove

exhaust efficiently and silently. The engine silencer will be capable of attenuating noise from a

running EDG to acceptable Occupational Safety and Health Administration (OSHA) levels for

worker hearing protection in an industrial environment. The EDG exhaust system will be

included in the generator set from the vendor as a packaged unit and will comply with NFPA-37,

“Standard for the Installation and Use of Stationary Combustion Engines and Gas Turbines,”

issued 1998 (NFPA, 1998a), and NFPA 110, “Standard for Emergency and Standby Power

Systems,” issued 1999 (NFPA, 1999).

Fuel filtration or conditioning is proposed at five places in the EDGFOS, including (1) on receipt

of new fuel from the remote filling station, (2) at the transfer pump suction, (3) in a bypass flow

purification system, (4) between the day tank and the engine, and (5) at the engine itself. The

filter on the remote filling station for the main tank will be a multi-element, differential pressure

instrumented, inlet filter rated at 2 μm (0.08 mils). The basis for this filter is to prevent

coagulated fuel and particulates from entering the storage tank assuming a low-pressure input

flow of 5.0–6.3 L/s (80–100 gpm). The transfer system will use a rotary screw positive

displacement pump to filter and circulate fuel oil from a main storage tank to the associated day

tank. The transfer system filters will be differential pressure instrumented, dual basket

switchable stainless steel suction strainer with automatic transfer and mesh filters rated to

remove 100 percent of 1.8 mils (45 μm) particles and 98 percent of particles at 0.7 mils (17 μm). 
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The flow rate for the transfer pump is approximately 1.6 L/s (25 gpm). The bypass flow

purification system will redirect approximately 0.9–1.3 L/s (15–20 gpm) of fuel through the

purification filters and dewatering devices. The bypass flow purification system is a vendor

package containing filters and water-removal equipment that is connected between the transfer

system supply and return lines. The transfer pump will provide the flow through the system. 

The transfer pump will be operated on a periodic basis to maintain fuel oil condition. The

purification package will condition the full bypass flow of fuel oil while the transfer pump is

operating.

The immediate use day tank includes an engine supply line with a differential pressure

instrumented, switchable stainless steel mesh dual-element strainer with automatic transfer

valve. The mesh filters will be rated to remove 100 percent of particles at 0.7 mils (18 μm) and

98 percent of particles at 0.2 mils (5 μm). These filters are sized to prevent the clogging of the

engine-mounted fuel filters, that are also sized at 0.2 mils (5 μm). The engine fuel pump

provides motive force for this fuel flow. The typical flow in the fuel oil system between the

engine and the day tank is approximately 21 L/min (330 gph), with the EDG using 4.5 L/min (72

gph) at 100-percent load and 16.3 L/min (258 gph) returning to the day tank.

The entire system of filters and screens is designed to turn over, filter, and purify the entire

volume of the main storage and day tanks every 12 hours. The filtration system is designed to

remove abrasive particulates and water that could damage the system pumps and the EDG.

Evaluation of Emergency Diesel Generator Fuel Oil System 

To complete this evaluation, the staff reviewed national codes and standards, such as IEEE

Standard 308 (IEEE, 1991), NFPA-37 (NFPA, 1998a), and NFPA 110 (NFPA, 1999). The staff

reviewed the seismic design of the EDGs. The staff also reviewed the design basis and

provisions for maintaining a clean, reliable source of fuel oil for the EDGs by comparing the

proposed design to standard industry practices and historical diesel generator failure information

published by the NRC.

As discussed in the system description, the EDGFOS is to be designed to supply adequate

capacity to handle the expected volume of material during normal, off-normal, and accident

conditions. Each EDGFOS bulk tank capacity is based on an EDG running at 100-percent load

for 7 days (plus extra operational margin for testing). This design meets the requirements of

IEEE Standard 308. The flow rate for the transfer pump is approximately 1.6 L/s (25 gpm). The

bypass flow purification system will redirect approximately 0.9–1.3 L/s (15–20 gpm) of fuel

through the purification filters and dewatering devices. The immediate-use day tank includes an

engine supply line with a differential pressure instrumented, switchable stainless steel mesh

dual-element strainer with automatic transfer valve. The typical flow in the fuel oil system

between the engine and the day tank is approximately 21 L/min (330 gph), with the EDG using

4.5 L/min (72 gph) at 100-percent load and 16.3 L/min (258 gph) returning to the day tank. The

engine silencer will be capable of attenuating noise from a running EDG to acceptable OSHA

levels for worker hearing protection in an industrial environment. The EDG exhaust system will

be included in the generator set from the vendor as a packaged unit. Based on this review, the

staff believes that the designs of the EDGFOS and the EDG exhaust system are acceptable. 

The EDGFOS and EDG exhaust systems do not interface with systems that carry radioactive

material; therefore, the consideration of the redundancy and diversity of components required to
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prevent release of radioactive material to the environment is not required. The evaluation of the

adequacy of the EDG system for its intended safety purpose is found in Section 11.5 of this

FSER.

The seismic design of the EDGFOS and the EDG exhaust systems assures that these systems

will be able to safely shut down during normal operations and accident conditions. Based on the

description of the system and the structures that will contain the system, the EDGFOS is

adequately designed to maintain functionality when subjected to severe natural phenomena

such as tornadoes, tornado missiles, earthquakes, floods, and any other appropriate

phenomena as established in the ISA. The EDG engine fuel oil pump is a PSSC that will

support the EDGs starting from rest and accepting a load within 10 seconds at a temperature of

-7 °C (20 °F). Based on the above, the staff finds that the design provisions for the EDGFOS

and the EDG exhaust system are acceptable to support continued operation of the EDGs.

The staff has reviewed the design and configuration of the EDGFOS system tanks and piping. 

Because of the nature and location of the system, the staff finds that it cannot be designed to

take advantage of gravity flow in order to perform its primary function. However, the redundancy

and diversity of the system and the seismic design, applied to the design as standard industry

practices, will assure a positive flow of fuel oil into, and waste gas removal from, the EDGs. The

EDGFOS and EDG exhaust systems are designed to minimize the likelihood that nuclear

material can enter these systems. Neither of these systems interfaces with systems that contain

nuclear material; they are, in fact, completely external to the facility buildings. This provides

reasonable assurance that contamination or criticality will not occur in these systems. DCS has

committed to design these systems and components with consideration to the process fluids

contained and to specify materials with the appropriate corrosion resistance where necessary. 

The staff finds the design of the system tanks, piping, and pumping systems to be acceptable. 

The staff has evaluated the EDG exhaust system to determine the provisions for personnel

protection. The engine silencer will be capable of attenuating noise from a running EDG to

acceptable OSHA levels for worker hearing protection in an industrial environment. The EDG

exhaust system will be included in the generator set from the vendor as a packaged unit and will

comply with NFPA-37 (NFPA, 1998a) and NFPA 110 (NFPA, 1999). On the basis of the system

design, industry codes and standards, and standard industry practices, the staff finds the EDG

exhaust system to be acceptable.

The electrical capacity of the EDGs has been reviewed by the staff, and Section 11.5 of this

FSER presents this evaluation.

Design Basis for Emergency Scavenging Air Subsystem 

The emergency scavenging air subsystem of the instrument air system is designed to provide

scavenging air for dilution of hydrogen that may be produced by radiolysis. The equipment most

likely to produce hydrogen by radiolysis is primarily those tanks containing a high plutonium

concentration solution. The emergency pressurized air supply tanks and piping are seismically

qualified in accordance with IEEE Standard 344-1987 (IEEE, 1987) (see the discussion of the

acceptability of this standard in Section 11.5 of this SER) and NRC RG 1.100, Revision 2 (NRC,

1988), and will be manually operated following a design basis event. The emergency

scavenging air subsystem, a PSSC, is physically separated from the instrument air system and

is automatically activated following the loss of the normal instrument air supply by opening one
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of the two parallel, fail-open, air-operated valves. The system is intended to be fitted to vessels

and equipment in the AP areas that will produce hydrogen by radiolysis and that could reach the

LFL of hydrogen in air (4 percent) in 7 days. The emergency scavenging air system is designed

to provide a 7-day supply of emergency scavenging air, based on consideration of a reasonable

time to repair or restore the normal bubbling air supply following failure of the instrument air

system. The 1-hour supply for the receiver/buffer tanks is sized based on the estimated time to

facilitate a normal process shutdown following loss of instrument air. 

Evaluation of Emergency Scavenging Air Subsystem 

In order to complete this evaluation, the staff reviewed national codes and standards, such as

IEEE Standard 308 (IEEE, 1991), NFPA 801 (NFPA, 1998b), RG 3.12 (NRC, 1973b), RG 1.100,

Revision 2 (NRC, 1988), and IEEE Standard 344-1987 (IEEE, 1987).

The emergency scavenging air subsystem is a PSSC designed to provide emergency

scavenging air for dilution of hydrogen that may be produced by radiolysis. The equipment

susceptible to this condition is primarily that containing a high plutonium concentration solution. 

Ordinarily, extra-dry instrument air is used by the bubbling air system to scavenge the hydrogen

produced in equipment. The bubbling air system is designed to prevent the system from

exceeding 25 percent of the LFL (or 1 percent) for hydrogen in the tanks. In the event of a

failure of any of these systems, the emergency scavenging air subsystem is designed to

compensate for the loss of the bubbling air system. The emergency scavenging air subsystem

is automatically activated following a loss of instrument air. The emergency scavenging air

system is designed to supply emergency scavenging air to all process vessels that could reach a

hydrogen concentration of 4 percent within 7 days. Each bank of the emergency scavenging air

system pressurized air supply main and backup banks of compressed air cylinders and piping

are designed for 100-percent capacity for 7 days. The two banks are connected by an air

pressure controller and transfer valves that switch from the primary bank to the secondary bank

on low system pressure. The emergency scavenging air subsystem is not physically connected

to the normal scavenging air supply.

The NRC staff reviewed the requirements for other safety-related backup systems, including

national codes and standards such as IEEE Standard 308. In that standard, the required

storage of fuel for the standby power supplies (such as EDGs) is either 7 days or the time

required to replenish the fuel and test it. DCS has committed to a 7-day fuel supply for the

EDGs, as described in CAR Section 11.5.7. Therefore, based on the IEEE standard and the

DCS commitment to providing a 7-day fuel supply for the EDGs, the staff finds that a 7-day

supply of emergency scavenging air is acceptable. The design basis for the maximum hydrogen

level allowed by the emergency scavenging air subsystem is calculated to prevent the hydrogen

concentration from exceeding 1 percent by volume. This level is equivalent to 25 percent of the

LFL for hydrogen in air and, as discussed in Chapter 7 of this FSER, is generally an accepted

limit. The applicant has committed to NFPA-801, which specifies that hydrogen concentrations

be kept at or below 25 percent of the LFL in air. Therefore, the staff finds the emergency

scavenging air subsystem design basis for the maximum allowed hydrogen concentration in

vessels susceptible to hydrogen generation to be acceptable based on the DCS commitment to

NFPA-801.

The emergency scavenging air subsystem is designed to provide two trains each capable of

supplying 100 percent of the scavenging air requirements for 7 days. Based on the provision of
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two 100-percent capacity trains of emergency scavenging air, and the design of the system air

pressure control to switch supply banks on low system pressure, the staff finds that the

subsystem has the adequate redundancy and diversity of components required to prevent

release of radioactive material to the environment and that its design is adequate for safe

operation of the system.

The emergency scavenging air subsystem is independent of the instrument air system and is

automatically actuated following loss of the scavenging air function in certain tanks susceptible

to the buildup of flammable levels of hydrogen resulting from radiolysis. The subsystem does

not rely on emergency power and is designed to operate during accident conditions. The staff

concludes that, because of the passive nature of the subsystem, it can be safely shut down

during any normal or accident conditions.

The emergency scavenging air subsystem is isolated from process fluids by the air spaces at the

top of the tanks it is designed to serve. In accident situations, the system is pressurized above

the pressure in the process equipment, so the flow of material is expected to be into the tank. 

The staff concludes that, based on the positive pressure of the subsystem, and the location of

the supply nozzles relative to process fluids, the subsystem is adequately designed to minimize

the likelihood of contamination and holdup of hazardous nuclear materials.

The design of the emergency scavenging air subsystem falls into FTS Category 2, discussed in

FSER Section 11.8. Category 2 is for PSSC components that may contain trace quantities of

plutonium or americium or nonradiological fluids that play a significant role in plant production

reliability. This means that the system is designed to ASME B&PV and B31.3 codes standards.

This includes consideration of piping material for corrosion protection and piping layout. 

Enhanced positive material identification, inspection, and test requirements are used in the

procurement specifications for FTS Category 2 components. The staff concludes that these

design criteria adequately address equipment design to resist corrosion, piping design, and

layout.

The design basis of the emergency scavenging air subsystem includes layout and design of the

system to minimize the potential for entrapment and buildup of radioactive materials. Therefore,

the staff concludes that the equipment is made of materials with the proper surface finishes and

decontamination characteristics for their particular application.

The SRP guidance indicates that equipment is to be adequately designed to maintain

functionality when subjected to severe natural phenomena such as tornadoes, tornado missiles,

earthquakes, floods, and any other appropriate phenomena as established in the ISA. The

emergency scavenging air system is seismically designed and qualified in accordance with NRC

RG 1.100, Revision 2 (NRC, 1988), and IEEE Standard 344-1987 (IEEE, 1987). Because the

design conforms to that RG and industry standard, the staff concludes that the applicant

adequately addressed the features that apply to this PSSC.

11.9.2 Evaluation Findings

In revised CAR Section 11.9, DCS provided design basis information for the fluid systems that it

identified as PSSCs for the proposed facility. Based on the staff’s review of the revised CAR

and supporting information provided by the applicant relevant to the fluid systems, the staff

concludes, pursuant to 10 CFR 70.23(b), that the design bases of the PSSCs evaluated in this
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FSER section will provide reasonable assurance of protection against natural phenomena and

the consequences of potential accidents. 

11.9.3 References

(ASME, 1995) American Society of Mechanical Engineers. Boiler and Pressure Vessel Code. 

Section VIII, “Rules for Construction of Division 1 Pressure Vessels.” 1995 Edition through the

1996 Addenda.

(ASME, 2000) American Society of Mechanical Engineers. ASME B16.10, “Face-to-Face and

End-to-End Dimensions of Valves.” 2000.

(ASME, 2001) American Society of Mechanical Engineers. ASME B31.3, “Process Piping.” 

2001.

(ASTM, 1974) American Society of Testing and Materials. ASTM D75, “Standard Specification

for Diesel Fuel Oils.” 1974.

(DCS, 2002a) Ihde, R., Duke Cogema Stone & Webster. Letter to Document Control Desk,

U.S. Nuclear Regulatory Commission, RE: Mixed Oxide Fuel Fabrication Facility—Revised

Construction Authorization Request. October 31, 2002 (including page changes through

February 9, 2005).

(DCS, 2002b) Hastings, P., Duke Cogema Stone & Webster. Letter to Document Control Desk,

U.S. Nuclear Regulatory Commission, RE Mixed Oxide Fuel Fabrication Facility —Evaluation of

the Draft Safety Evaluation Report (DSER) on Construction of a Mixed Oxide Fuel Fabrication

Facility. July 9, 2002.

(IEEE, 1987) Institute of Electrical and Electronics Engineers. IEEE Standard 344, “IEEE

Recommended Practice for Seismic Qualification of Class IE Equipment for Nuclear Power

Generating Stations.” 1987.

(IEEE, 1991) Institute of Electrical and Electronics Engineers. IEEE Standard 308, “IEEE

Standard Criteria for Class IE Power Systems for Nuclear Power Generating Stations.” 1991.

(IEEE, 1998) Institute of Electrical and Electronics Engineers. IEEE Standard 603, “IEEE

Standard Criteria for Safety Systems for Nuclear Power Generating Stations.” 1998. 

(NFPA, 1998a) National Fire Protection Association. NFPA 37, “Standard for the Installation

and Use of Stationary Combustion Engines and Gas Turbines.” 1998.

(NFPA, 1998b) National Fire Protection Association. NFPA 801, “Standard for Fire Protection

for Facilities Handling Radioactive Materials.” 1998.

(NFPA, 1999) National Fire Protection Association. NFPA 110, “Standard for Emergency and

Standby Power Systems.” 1999.

(NRC, 1973a) U.S. Nuclear Regulatory Commission. Regulatory Guide 3.14, “Seismic Design

Classification for Plutonium Processing and Fuel Fabrication Plants.” Washington, DC, 1973.
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(NRC, 1973b) U.S. Nuclear Regulatory Commission. Regulatory Guide 3.12, “General Design

Guide for Ventilation Systems of Plutonium Processing and Fuel Fabrication Plants.” 

Washington, DC, 1973.

(NRC, 1974) U.S. Nuclear Regulatory Commission. Regulatory Guide 3.17, “Earthquake

Instrumentation for Fuel Reprocessing Plants.” Washington, DC, 1974.

(NRC, 1988) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.100, Rev. 2, “Seismic

Qualification of Electric and Mechanical Equipment for Nuclear Power Plants.” Washington, DC,

1988.

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility.” Washington,

DC, August 2000.

(NRC, 2002) Persinko, A., U.S. Nuclear Regulatory Commission. Letter to P. Hastings, Duke

COGEMA Stone & Webster, RE: Response to DCS Letter Dated July 9, 2002 (DCS-NRC0010000). October 30, 2002.

11.10 Heavy Lift Cranes

11.10.1 Conduct of Review

This chapter of the FSER contains the staff’s review of the heavy lift cranes described by the

applicant in the revised CAR (DCS, 2002). The objective of this review is to determine whether

heavy lift crane PSSCs and their design bases identified by the applicant provide reasonable

assurance of protection against natural phenomena and the consequences of potential

accidents. The staff evaluated the information provided by the applicant for heavy lift cranes by

reviewing Chapter 11.0 and other sections of the revised CAR, supplementary information

provided by the applicant, and relevant documents available at the applicant’s offices but not

submitted by the applicant. The review of the design bases and strategies of the heavy lift

cranes was closely coordinated with the review of other chapters in the FSER (fire protection in

Chapter 7, chemical safety in Chapter 8, and safety assessment of the design bases in Chapter

5) and the review of other plant systems. 

The review for this construction approval focused on the design basis of heavy lift cranes, their

components, and other related information. For heavy lift cranes, the staff reviewed information

provided by the applicant on the safety function, system description, and safety analysis. The

review also encompassed proposed design basis considerations such as redundancy,

independence, reliability, and quality. The staff used Section 11.4.8 in NUREG-1718 (NRC,

2000) as guidance in performing the review of heavy lift cranes.

The FSER discussions that follow provide the system descriptions, as well as functions, major

components, control concepts, and system interfaces.

11.10.1.1 System Description

The applicant has not identified cranes, heavy lift or otherwise, as PSSCs in the revised CAR for

the MFFF. However, material-handling equipment and controls have been identified as PSSCs. 
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DCS will identify IROFS in its ISA summary to be submitted as part of any later license

application it may submit.

For the facility, heavy lift cranes are designated as those cranes designed to lift greater than 816

kg (1800 lb), as defined in NUREG-0612, “Control of Heavy Loads at Nuclear Power Plants.” 

This load definition limits, to a very few, the number of cranes or lifting devices designated as

heavy lift cranes. Other cranes or lifting devices may be identified as the design matures. The

currently identified applications of heavy lift cranes in the facility are bridge cranes in the fresh

fuel cask shipping truck bay, the assembly loading area, and the fresh fuel cask handling area; a

bridge crane stacker for waste drum handling; a bridge crane for handling empty plutonium

dioxide shipping package pallets; and a maintenance hoist in the EDG building.

The facility general design philosophy is stated as preventing lifts of heavy loads above PSSCs.

Heavy lift cranes are not designated as PSSCs because equipment being lifted would be

qualified for drops from the maximum lift height of the cranes. For example, the qualification of

the MOX fresh fuel package is to maintain confinement integrity of the package for a drop from a

height of 9.1 m (30 ft) onto an unyielding surface. DCS states that the maximum lift height for

any piece of equipment in the facility is 4.9 m (16 ft). However, it is possible for cranes to impact

PSSCs without a load drop. In this case, the PSSCs identified in the revised CAR are the

material-handling controls discussed in Section 11.7 of this FSER. The term “material-handling

controls” includes controls on material-handling equipment and administrative controls. 

Examples of these controls include safe travel paths, procedures and training to limit crane

operations or to properly prepare loads or nearby equipment before crane use, and a radiation

protection program to ensure workers are protected during maintenance activities. Specific

material-handling controls will be identified in the ISA summary.

11.10.1.1.1 Function

The function of the facility heavy lift cranes is to lift and move critical loads in a manner

consistent with the loads’ qualification, to limit inadvertent movement, and to retain loads under

normal, accident, and design basis natural phenomena conditions, such as earthquake and loss

of power events. Critical loads are loads of a type that, if dropped and the contents released,

could result in unacceptable radiological dose consequences to the worker, the IOC, or the

environment. Other noncritical loads may also be moved by this equipment at any time. The

heavy lift cranes’ functions include controlling movement of the cranes and loads during

operation and maintenance so they will not impact other PSSCs. 

11.10.1.1.2 Major Components

The major components of the heavy lift cranes are the crane structures, drive motors for

positioning and hoisting, the operator’s cab or local operating station, and the lights and control

systems for the cranes. Crane equipment consists of slings, lifting frames, and other below-thehook lifting devices. Cranes may also be equipped with auxiliary hoists to perform routine lifting

of smaller loads. The capacity of each crane is such that it should meet or exceed the lifted

load; the lifted load includes the equipment to be moved plus the weight of the slings and other

lifting devices plus applicable design margins specified in the crane design codes and standards.
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11.10.1.1.3 Control Concepts

Most of the facility’s heavy lift cranes will be operated locally by crane operators who are in

visual contact with the crane and the load. The crane control stations are local, and the controls

are conventional and required in the industry codes and standards, such as ASME B30.2

(ASME, 1996b). The stacker crane in the waste drum stacker area will be designed to operate

automatically, with provisions for manual control as necessary. Similarly, the crane in the

assembly loading area will function in automatic mode for some of its operations.

11.10.1.2 System Interfaces

The heavy lift cranes interface with the building structure; the rails on which the cranes run; the

girders that span the work area; the electric drives for the bridge, trolley, and hoist; local

operating stations; and the lights and control systems for the cranes. The crane may be

controlled by radio communication with the operator. In these cases, the interactions between

the crane controls and the building security, process monitoring, and control systems will be

considered.

11.10.1.3 Design Bases of PSSCs

Heavy lift cranes or hoists at the MFFF have not been identified as PSSCs in Section 11.10 of

the revised CAR. There are only two PSSCs over which heavy loads may be lifted—(1) the

building floor, which has been evaluated for dynamic loads, including drop loads, and (2) the

emergency generator during maintenance operations. Any nonheavy loads over PSSCs will be

handled in accordance with material-handling controls, which are PSSCs. The applicant stated

in Section 11.10.7 of the revised CAR that the drop of any load handled by a heavy lift crane

would not result in consequences exceeding those presented in the 10 CFR 70.61 performance

requirements. 

Section 5.0 of the revised CAR discusses load-handling events. A load-handling hazard is

postulated to occur from the presence of lifting or hoisting equipment used in normal or

maintenance activities. Examples of load-handling events include a dropped lifted load, a lifted

load impacting other equipment during the lift, or the load-handling equipment impacting other

nearby items. An event of this type could damage handled loads, thereby dispersing nuclear

material or chemicals, damaging nearby equipment resulting in a loss of confinement or loss of

subcritical conditions, or damaging IROFS. Such load-handling events are evaluated in Chapter

5 of this FSER.

11.10.2 Evaluation Findings

In Section 11.10 of the revised CAR, the applicant did not designate heavy lift cranes as PSSCs. 

Based on the staff’s review of the revised CAR and supporting information provided by the

applicant relevant to the heavy lift cranes, the staff concurs that heavy lift cranes need not be

designated as PSSCs.

11.10.3 References

(ANSI, 1993) American National Standards Institute. ANSI N14.6, “Radioactive

Materials—Special Lifting Devices for Shipping Containers Weighing 10,000 lbs or More.” 1993.
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(ASME, 1996a) American Society of Mechanical Engineers. ASME B30.9, “Slings.” 1996.

(ASME, 1996b) American Society of Mechanical Engineers. B30.2, “Top Running Bridge,

Single or Multiple Girder, Top Running Trolley Hoist Overhead and Gantry Cranes.” 1996.

(ASME, 1998) American Society of Mechanical Engineers. ASME B30.16, “Overhead Hoists

(Underhung).” 1998.

(CMAA, 1994) Crane Manufacturers Association of America. 70, “Specifications for Top

Running Bridge and Gantry Type Multiple Girder Electric Overhead Traveling Cranes.” 1994.

(DCS, 2002) Ihde, R., Duke Cogema Stone & Webster. Letter to Document Control Desk, U.S.

Nuclear Regulatory Commission, RE: Mixed Oxide Fuel Fabrication Facility—Construction

Authorization Request. October 31, 2002 (including page changes through February 9, 2005).

(DOE, 2000) U.S. Department of Energy(DOE). DOE-STD-3013, “Stabilization, Packaging, and

Storage of Plutonium-Bearing Materials.” September 2000.

(NRC, 1980) U.S. Nuclear Regulatory Commission. NUREG-0612, “Control of Heavy Loads at

Nuclear Power Plants.” Washington, DC, July 1980.

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility.” Washington,

DC, August 2000. 

11.11 Environmental Qualification

11.11.1 Conduct of Review

This chapter of the FSER contains the staff’s review of EQ described by the applicant in Chapter

11.0 of the revised CAR (DCS, 2002). The objective of this review is to determine whether the

EQ of electrical and mechanical equipment designated as PSSCs and the EQ design bases

identified by the applicant provide reasonable assurance of protection against natural

phenomena and the consequences of potential accidents. The staff evaluated the information

provided by the applicant for EQ of electrical and mechanical equipment by reviewing Chapter

11 of the revised CAR, other sections of the revised CAR, supplementary information provided

by the applicant, and relevant documents available at the applicant’s offices but not submitted by

the applicant. The review of the EQ of electrical and mechanical equipment design bases and

strategies was closely coordinated with the review of the electrical, I&C, and mechanical aspects

of accident sequences described in the safety assessment of the design bases (see Chapter 5

of this FSER) and the review of other plant systems. 

The staff reviewed how the information in the revised CAR addresses the following regulations:

• 10 CFR 70.23(b) and the Quality Assurance Program state, as a prerequisite to

construction approval, that the design bases of the PSSCs must be found to provide

reasonable assurance of protection against natural phenomena and the consequences of

potential accidents.
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• 10 CFR 70.64 requires that BDC and defense-in-depth practices be incorporated into the

design of new facilities. With respect to the EQ of electrical and mechanical equipment,

10 CFR 70.64(a)(4) requires that the MFFF design “must provide for adequate protection

from environmental conditions and dynamic effects associated with normal operations,

maintenance, testing, and postulated accidents that could lead to loss of safety functions.”

The review for this construction approval focused on the design basis of the electrical and

mechanical equipment that are PSSCs specifically related to the EQ requirement. The staff

used Chapter 11.0 in NUREG-1718 as guidance in performing the review.

11.11.1.2 System Description

The electrical and mechanical equipment evaluated in this section are PSSCs and are major

components of systems (also PSSCs) such as the confinement systems (described in FSER

Section 11.4), the electrical systems (described in FSER Section 11.5), the I&C systems

(described in FSER Section 11.6), the material transport system (described in FSER Section

11.7), the fluid transport systems (described in FSER Section 11.8), the fluid systems (described

in FSER Section 11.9), and the heavy lift cranes (described in FSER Section 11.10).

11.11.1.3 Design Basis of the PSSCs and Applicable Baseline Design Criteria 

In addition to the design basis requirements for the electrical and mechanical equipment

designated as PSSCs, which encompass the PSSCs within the systems listed in FSER Section

11.11.2, the electrical and mechanical equipment that are designated PSSCs are to remain

functional when subjected to the environmental and dynamic effects referenced in the 10 CFR

70.64(a)(4) BDC. For EQ purposes, this electrical and mechanical equipment will be designed

as described in the revised CAR and the staff’s memorandum dated January 31, 2003 (NRC,

2003), based on IEEE Standard 323-1983, “IEEE Standard for Qualifying Class 1E Equipment

for Nuclear Power Generating Stations” (IEEE, 1983).

DCS stated that electrical and mechanical PSSCs will be qualified to anticipated environments,

including qualification to environments if the event causing the harsh environment is not

prevented (NRC, 2003). Pursuant to this commitment, DCS will be required to demonstrate that

each piece of electrical and mechanical equipment identified as a PSSC will perform its safety

functions under the environmental and dynamic service conditions in which it will be required to

function and for the length of time its function is required. In addition, non-PSSCs will be

required to be able to withstand environmental stresses caused by environmental and dynamic

service conditions under which their failure could prevent satisfactory accomplishment of safety

functions by PSSCs. The staff finds that these commitments satisfy the requirements of 10 CFR

70.64(a)(4) and notes that this approach is consistent with how EQ requirements are applied at

nuclear power plants (NRC, 1984).

11.11.2 Evaluation Findings

Based on the staff’s review of the revised CAR, the supporting information provided by the

applicant, and the applicant’s commitments to the industry guidance referenced above, the staff

finds, pursuant to 10 CFR 70.64(a)(4), that for EQ, the design basis of the proposed facility

provides for adequate protection against environmental condition hazards. The staff concludes,

pursuant to 10 CFR 70.23(b), that the design basis of the PSSCs relevant to EQ of electrical and
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mechanical equipment will provide reasonable assurance of protection against natural

phenomena and the consequences of potential accidents.

11.11.3 References

(DCS, 2002) Ihde, R., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, Mixed Oxide Fuel Fabrication Facility Construction Authorization Request Revision. 

October 31, 2002 (including page changes through February 9, 2005).

(IEEE, 1983) Institute of Electrical and Electronics Engineers. Std. 323-1983, “IEEE Standard

for Qualifying Class 1E Equipment for Nuclear Power Generating Stations.” New York,

September 30, 1983.

(NRC, 1984) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.89, Revision 1,

“Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power

Plants.” Washington, DC, June 1984.

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility.” Washington,

DC, August 2000.

(NRC, 2003) Persinko, Andrew, U.S. Nuclear Regulatory Commission. Memorandum to Melvyn

N. Leach, U.S. Nuclear Regulatory Commission, RE: December 10–12, 2002, Meeting

Summary: Meeting with Duke Cogema Stone & Webster to Discuss Mixed Oxide Fuel

Fabrication Facility Revised Construction Authorization Request. January 31, 2003.

11.12 Seismic Qualification of Equipment, Systems, and Components

11.12.1 Conduct of Review

This chapter of the FSER contains the staff’s review of SQ described by the applicant in Section

11.12 of the revised CAR (DCS, 2002a). The objective of this review is to determine whether

the SQ of electrical and mechanical equipment and their SQ design bases identified by the

applicant provide reasonable assurance of protection against natural phenomena and the

consequences of potential accidents. Seismic issues related to the ground motion are evaluated

in FSER Section 1.3.1.5. The staff evaluated the information provided by the applicant for SQ of

electrical and mechanical equipment by reviewing Chapter 11 of the revised CAR, other sections

of the revised CAR, supplementary information provided by the applicant, and relevant

documents available at the applicant’s offices but not submitted by the applicant. The review of

the SQ of electrical and mechanical equipment design bases and strategies was closely

coordinated with the review of the electrical, I&C, and mechanical aspects of accident

sequences described in the safety assessment of the design bases (see Chapter 5 of this

FSER) and the review of other plant systems.

The staff reviewed how the information in the revised CAR addresses the following regulations:

• 10 CFR 70.23(b) states, as a prerequisite to construction approval, that the design bases

of the PSSCs must be found to provide reasonable assurance of protection against

natural phenomena and the consequences of potential accidents.
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• 10 CFR 70.64 requires that BDC and defense-in-depth practices be incorporated into the

design of new facilities. With respect to natural phenomena hazards, 10 CFR 70.64(a)(2)

requires that the MFFF design provide for adequate protection against such hazards, with

consideration of the most severe documented historical events for the MFFF site.

The review for this construction approval focused on the design basis of the electrical and

mechanical equipment related to the SQ requirement. The staff used Chapter 11.0 in NUREG1718 as guidance in performing the review.

11.12.1.1 System Description

The electrical and mechanical equipment evaluated in this section are PSSCs and are major

components of systems (also PSSCs) such as the confinement systems (described in FSER

Section 11.4), the electrical systems (described in FSER Section 11.5), the I&C systems

(described in FSER Section 11.6), the material transport system (described in FSER Section

11.7), the fluid transport systems (described in FSER Section 11.8), the fluid systems (described

in FSER Section 11.9), and the heavy lift cranes (described in FSER Section 11.10).

11.12.1.2 Design Basis of the PSSCs and Applicable Baseline Design Criteria 

The applicant’s methodology for seismic classification of SSCs consists of SC and seismic

performance requirements. The SC classification contains SC-I which applies to all PSSCs that

must perform safety functions during and/or after the design basis earthquake to comply with the

MFFF safety assessment as described in Chapter 5 of the revised CAR; and SC-II which applies

to systems which are not required after a design basis earthquake but whose failure could

adversely impact the ability of an SC-I system to perform its safety function. The seismic

performance requirements segregate SC-I and SC-II by the safety functions they must perform

during or after a seismic event such as remaining active during or after a seismic event,

maintaining a pressure boundary or structural integrity, and not failing in a way that

compromises a PSSC (SC-II). 

The analysis requirements provided in the revised CAR consist of seismic design parameters

that were developed with the determination of the design basis earthquake and procedures for

the use of seismic inertial response techniques. For the seismic parameters, in-structure

response spectra will be generated in accordance with one of the methods cited in Section 3.4 of

ASCE 4-98. The applicant’s qualification methodology allows the inertial response of PSSCs to

be determined using dynamic analysis or equivalent static analysis for elements that can be

represented by a single-degree-of-freedom or a simple multiple-degree-of-freedom model. 

The SQ requirements as specified in the revised CAR allow either analysis or shake-table

testing for most seismic performance categories. Shake table testing is required for

components where analysis alone is insufficient to ensure operability after a seismic event (e.g.,

electrical components). 

In the CAR (DCS, 2002a) and by letter dated August 31, 2001 (DCS, 2002b), DCS stated that

electrical and mechanical equipment will be seismically qualified using IEEE Standard 344-1987

(IEEE 1987) and NRC RG 1.100, Revision 2 (NRC 1988). 
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DCS stated (DCS, 2002b) that mechanical equipment will be qualified to IEEE 344-1987,

including NRC additions to the 1987 IEEE standard stated in RG 1.100. Therefore, mechanical

equipment qualification will consider attached piping loads, thermal loads, and live loads such as

fluid sloshing. In addition, applied loads will meet or exceed accelerations corresponding to their

installed locations.

The staff concludes that these commitments are consistent with the application of SQ

requirements at nuclear power plants. The staff further concludes that the facility design

provides for adequate protection against seismic hazards with consideration of the most severe

documented historical events for the facility site and satisfies the requirements of 10 CFR

70.64(a)(2).

11.12.2 Evaluation Findings

Based on the staff’s review of the revised CAR, the supporting information provided by the

applicant, and the applicant’s commitments to the guidance referenced above, the staff finds,

pursuant to 10 CFR 70.64(a)(2), that for SQ of equipment, the design basis of the proposed

facility provides for adequate protection against seismic hazards, and that the applicant has

adequately considered the most severe documented historical seismic events for the MFFF site.

The staff concludes, pursuant to 10 CFR 70.23(b), that the design basis of the PSSCs relevant

to the SQ of electrical and mechanical equipment will provide reasonable assurance of

protection against natural phenomena and the consequences of potential accidents.

11.12.3 References

(DCS, 2002a) Ihde, R.H., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Mixed Oxide Fuel Fabrication Facility Construction Authorization Request

Revision. October 31, 2002 (including page changes through February 9, 2005). 

(DCS, 2002b) Hastings, P.S., Duke Cogema Stone & Webster. Letter to U.S. Nuclear

Regulatory Commission, RE: Construction Authorization Request: Clarification of Responses to

NRC Request for Additional Information. February 11, 2002.

(IEEE, 1987) Institute of Electrical and Electronics Engineers. Std 344-1987, “IEEE

Recommended Practices for Seismic Qualification of Class1E Equipment for Nuclear Generating

Stations.” 1987.

(NRC, 1988) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.100, Rev. 2, “Seismic

Qualification of Electric and Mechanical Equipment for Nuclear Plants.” Washington, DC,

August 2000.

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility.” Washington,

DC, August 2000.
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12. HUMAN FACTORS ENGINEERING FOR PERSONNEL ACTIVITIES

12.1 Conduct of Review

This chapter of the final safety evaluation report (FSER) contains the staff’s review of the human

factors engineering (HFE) plans, processes, and analyses performed by Duke Cogema Stone &

Webster (DCS or the applicant) in Chapter 12 of the revised construction authorization request

(CAR), dated October 31, 2002 (DCS, 2002c). The objectives of this review are to (1) establish

that HFE is being applied to personnel activities identified as a principal structure, system,

component (PSSC) (PSSCs and items relied on for safety (IROFS) include activities of

personnel that are relied on to prevent potential accidents that could exceed the performance

requirements in Title 10, Section 70.61, “Performance Requirements,” of the Code of Federal

Regulations (10 CFR 70.61)), consistent with the findings of the revised CAR, and (2) determine

whether PSSCs and their design bases identified by DCS provide reasonable assurance of

protection against natural phenomena and the consequences of potential accidents. The staff

evaluated the information provided by the applicant for HFE by reviewing Chapter 12 of the

revised CAR, other sections of the revised CAR, and supplementary information provided by the

applicant in its letter dated November 22, 2002 (DCS, 2002d). The review of HFE was closely

coordinated with the review of the instrumentation and control (I&C) and electrical aspects of

accident sequences described in the safety assessment of the design bases (see Chapter 5 of

this FSER) and the review of other plant systems.

The staff reviewed how the information in the revised CAR addresses the following regulation:

• 10 CFR 70.23(b) states, as a prerequisite to construction approval, that the design bases

of the PSSCs and the Quality Assurance Program must be found to provide reasonable

assurance of protection against natural phenomena and the consequences of potential

accidents.

The scope of the HFE review included (1) a description of the safety-significant personnel

actions, the associated human system interfaces, and the consequences of incorrectly

performing or omitting actions for each personnel activity, (2) the applicant’s plans for the HFE

design review, (3) review of operating experience at existing facilities that are similar to the

proposed mixed oxide (MOX) fuel fabrication facility (MFFF or the facility), (4) function and task

analysis, (5) human-system interface (HSI) design, inventory, and characterization, (6) staffing,

(7) procedure development, (8) training program development, and (9) verification and

validation.

The staff used applicable portions of Chapter 12.0 in NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility,” issued August

2000, as guidance in performing the review.

12.1.1 Identification of Personnel Actions

The applicant discussed the nature of personnel actions at the proposed facility in Section 12.1

of the revised CAR. Control of the operations of the facility relies to a great extent on automated

systems to ensure production quality and facility safety. In general, the operations staff is

expected to perform the following types of tasks:
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• initiate batch or continuous operations

• monitor the progress of the operations

• perform or initiate performance of quality control checks at preprogrammed hold points in

the process

• monitor and confirm the status of confinement systems, fluid systems, and other facility

systems

• respond to or recover from off-normal conditions

In the revised CAR, DCS discussed the human factors/human performance activities associated

with maintenance of automated systems which would be used in the facility and did not identify

any safety-significant human-system maintenance interfaces. The applicant stated that the

integrated safety analysis (ISA) process will identify the sensors, instruments, and actuators that

are classified as IROFS. The appropriate HSI requirements and the human performance

requirements will be established as part of the DCS application for an operating license under

10 CFR Part 70, “Domestic Licensing of Special Nuclear Material.” Activities associated with the

maintenance or operation of the instruments, sensors, and actuators which are later classified as

IROFS will be evaluated for human factors attributes using the criteria of Institute of Electrical

and Electronics Engineers (IEEE) Standard 1023, “IEEE Guidelines for the Application of Human

Factors Engineering to Systems and Equipment, and Facilities of Nuclear Power Generating

Facilities,” issued 1988 (IEEE, 1988), with the recognition that there are conditions, systems,

operating requirements, and consequences unique to a nuclear power plant and not found in a

fuel fabrication facility. The applicant also committed to use and review the criteria of the Design

Review Checklist in NUREG/CR-6636, “Maintainability of Digital Systems: Technical Basis and

Human Factors Guidelines,” issued March 2000 (NRC, 2000d), for applicability to the digital

controls of the PSSCs as part of its application for a license to possess and use licensed

material. The applicant also stated that NUREG-0700 and all of the NUREG/CR references in

Chapter 12.0 of NUREG-1718 would be used as guidance as part of its application for a

possession and use license for human performance activities associated with maintenance of

the facility automated systems. The staff finds this information acceptable because it provides

the clarification requested regarding the human performance activities associated with

maintenance at the facility.

In the revised CAR, the applicant described the criteria and basis used for determining that the

protective control subsystem does not constitute a significant HSI and defined what “significant”

means. The applicant stated that the protective control subsystem is designed to satisfy

industrial safety requirements and is not a PSSC and provided additional information describing

the design of the protective control subsystem’s HSI. In the revised CAR, DCS more explicitly

defined what is meant by “significant human-system interface” for the protective control

subsystem, evaluated the potential for personnel errors of commission that might result in

overriding or defeating safety systems, and provided cross-references to appropriate parts of

Chapter 11 of the revised CAR. The staff finds this information acceptable because it provides

the clarification requested regarding the protective control system, its human system interface,

and potential personnel errors of commission.
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The applicant stated in its revised CAR Section 12.1 that the facility would have a high level of

automation with operators mainly monitoring the operation of systems and exercising

supervisory control only when necessary. In a June 21, 2001, letter to the applicant, the U.S.

Nuclear Regulatory Commission (NRC) staff asked the applicant to describe how its staff are

alerted to undesirable conditions at control stations that are not normally staffed, and what

criteria are used to decide when appropriate operations staff need to be at these remote

locations for appropriate and timely response (NRC, 2001). By letter dated August 31, 2001

(DCS, 2001), DCS provided supplemental information stating that the performance of systems in

automated areas would be constantly monitored by automated supervisory systems. One of the

attributes of the functional units which would be monitored by the supervisory systems is the

state of an automated activity. If the activity is not concluded in an anticipated state or within an

expected time, or if a continuous process is not within allowed limits, an alarm would be

generated in the control room for that functional unit.

The design of the facility establishes several different control rooms, and controls of the various

functional units of the facility are grouped together into these control rooms. If a functional unit

is in operation, the control room associated with that functional unit would be occupied. If none

of the functional units assigned to a particular control room is operating, that control room would

probably not be occupied. For example, control outputs for the fissile material mass accounting

system would not be needed if there are no movements into, out of, or within a glovebox;

similarly, the mass measurement system and mass limit alarms would not be meaningful in this

situation. Signals for functions appropriate only to an operational unit would be transmitted to

the control room that is assigned to that function.

[Text removed under 10 CFR 2.390]

Staffing evaluations will be completed as part of the DCS application for a 10 CFR Part 70

operating license and will be derived from the staffing requirements that exist in the La Hague

and MELOX facilities. The applicant will provide this information in the license application for the

facility. The staff finds this information acceptable because it provides the clarification requested

regarding alerting staff to undesirable conditions at control stations not normally staffed, which

control rooms would be staffed, and the development of staffing requirements derived from La

Hague and MELOX. By letter dated November 22, 2002, the applicant stated that it was not

necessary to incorporate this information in the revised CAR (DCS, 2002d).

In revised CAR Section 12.1, the applicant stated that, in general, omission of an operator action

would not result in adverse conditions, and that errors in operator actions would generally be

bounded by the other deterministic design basis-assumptions. The applicant clarified what is

meant by “in general” and described by example the other deterministic design basis

assumptions. The applicant stated that no scenario has been identified where omission of an

operator action would result in adverse conditions, and errors in operator actions have been

anticipated in the system design while considering other deterministic design basis accident

assumptions and scenarios. The applicant also more explicitly defined what is meant by “other
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deterministic design basis accident assumptions and scenarios” and considered and evaluated

the potential for personnel errors of commission that might result in overriding or defeating

safety systems. The staff finds this information acceptable because it provides the clarification

requested regarding other deterministic design basis assumptions and potential for personnel

errors of commission.

12.1.2 Human Factors Engineering Design Planning

The applicant discussed HFE design planning in Section 12.2 of the revised CAR. The design of

HFE includes the identification of HFE programmatic goals and scope and a description of the

plans for HFE review, including HFE team makeup and processes for conducting HFE reviews. 

The HFE principles are applied to the facility design based on the guidelines of IEEE-1023

(IEEE, 1988). In a request for additional information dated June 21, 2001 (NRC, 2001), the

NRC staff asked the applicant to verify a commitment made in an April 25, 2001, meeting with

the NRC staff, to use NUREG-0711 to further guide its human factors design basis development

work during construction and evaluate a subsequent revision to IEEE-1023. In letters dated

August 31, 2001, and January 7, 2002 (DCS, 2001; DCS, 2002a; RAI 228), the applicant

provided supplemental information which stated that NUREG-0711, Revision 2, “Human Factors

Engineering Program Review Model,” issued May 2002 (NRC, 2002b), would continue to be

reviewed for HFE criteria that may be applicable to the design of the HSI for the control systems

of PSSCs in the facility. DCS also stated that it would evaluate any future revision to IEEE-1023

for applicability to the design of the facility. The NRC staff finds this response acceptable

because it provides the clarification requested on the use of IEEE-1023 and NUREG-0711. By

letter dated November 22, 2002, the applicant stated that it was not necessary to incorporate

this information in the revised CAR (DCS, 2002d).

In the revised CAR, DCS identified and described the meaning of “facility baseline design” and

cross-referenced other appropriate chapters of the revised CAR. The applicant stated that

facility baseline design is synonymous with the technical basis defined in the configuration

management policies in Section 15.2.1 of the revised CAR. The staff finds this information

acceptable because it defines the meaning of facility baseline design.

In the revised CAR, the applicant identified and described the aspects of the facility design that

reduce the risk of errors or challenges to PSSCs and how these aspects are evaluated. DCS

stated that the facility is designed to maximize the use of automation, thus minimizing human

operations and interactions with the facility SSCs. By reducing these interactions, the probability

of a human-caused error being introduced is reduced. The applicant also stated that it would

consider both human errors of omission and commission in its evaluation of the probability of

human error and describe these results as part of the license application. The staff finds this

information acceptable because it provides both a process and rationale for maximizing the use

of automation to reduce the probability of human errors of omission or commission.

In revised CAR Section 12.1.2, DCS also described how the HFE team will conduct its activities

and where the team resides within the organization, with organizational roles and responsibilities

clearly defined. The applicant discussed the activities of each of three phases of the HFE

process—preliminary design, final design, and construction and startup. In the preliminary

design phase, the facility control system architecture, control philosophies, and HSIs were

developed with emphasis on the proven control methods from MELOX and La Hague. The

original design and ongoing evolution of these facilities incorporated various degrees of human
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factors methods and reflect several years of safe operation. To supplement their use as a

reference design, operational experience is incorporated into the facility design through a

combination of lessons-learned evaluations (focusing on operability and maintainability issues

and involving current operations and maintenance personnel) and review of the facility design on

an ongoing basis by experienced operations staff. The applicant described, by example, how

operating experience of the La Hague and MELOX facilities is incorporated in the facility design

process, and provided examples of lessons-learned evaluations that show how the facility, as a

proposed next generation facility, effectively incorporates this operating experience. DCS

provided a presentation with examples of significant MELOX and La Hague operating events

involving human performance to the NRC staff in a meeting at the NRC on October 11, 2001. 

The applicant submitted supplemental information to document its October 11, 2001, meeting

presentation and provided additional examples of significant MELOX and La Hague operating

events involving human performance. The staff finds this information acceptable because it

provides the clarification requested regarding specific examples of operating experience at

MELOX and La Hague which are being incorporated in facility design.

In the revised CAR, DCS also stated that, as part of the license application, criteria for HFE will

be identified in facility design basis documents and will be applied throughout the final design for

aspects of operation and maintenance of the facility. The task analysis, which will be completed

as part of the application for a 10 CFR Part 70 operating license, will reflect the personnel

activities relied on for safety as identified as part of the development of the ISA. During the

detailed design of the HSIs, inventory and characterization of the interfaces will be performed. 

Evaluation of the characteristics of the HSIs will use the review criteria of NUREG-0700,

Revision 2, “Human-System Interface Design Review Guidelines,” issued May 2002 (NRC

2002a), as the basis.

12.1.3 Operating Experience

The applicant discussed operating experience in Section 12.3 of the revised CAR, as well as in

Section 12.2 of the revised CAR (see Section 12.1.2 of this FSER above). In Section 12.3 of the

revised CAR, the applicant stated that as a result of selection of existing facilities with successful

operating histories as a reference design for the facility, and the ongoing involvement of

operations and engineering personnel from these facilities in the development of facility design,

no additional formal operating experience review is anticipated. The applicant clarified what is

meant by “no additional formal operating experience review is anticipated” for the facility based

on the operational experience at the La Hague and MELOX facilities previously incorporated in

the facility design. Lessons learned from operating experience will be a continuing activity

throughout construction, detailed design, and operation. The applicant also stated that there

would be ongoing involvement of operations and engineering personnel from the MELOX and La

Hague facilities in the development and design of the facility, thus providing a capability for

evaluating and including results of operating experience as appropriate for the facility. The staff

finds this an acceptable way to incorporate ongoing operational experience into the facility

design.

12.1.4 Function and Task Analysis

The applicant discussed function and task analysis in Section 12.4 of the revised CAR, as well

as in Sections 12.2.3.1 and 12.2.3.2. Operational tasks are well established for the MELOX and

La Hague facilities for the purposes of preliminary design of the facility. The facility is an
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automated facility, and the tasks assigned to humans involve primarily initiating, verifying, and

monitoring system status. The task analysis will be completed as part of the license application

and will reflect the personnel activities relied on for safety identified as part of the ISA.

12.1.5 Human-System Interface Design, Inventory, and Characterization

The applicant discussed HSI design, inventory, and characterization in Section 12.5 of the

revised CAR, as well as in Section 12.2.3. The HSI design, inventory, and characterization for

the facility are initially based on the MELOX and La Hague designs. As part of the license

application, detailed design of the HSI, inventory, and characterization of the interfaces will be

completed.

12.1.6 Other Considerations

The applicant discussed staffing, procedure development, and training in Section 12.6 of the

revised CAR and stated that these issues will be addressed in the HFE plan to be developed as

part of the license application. The HFE verification and validation is discussed in Sections

12.2.3, 12.2.3.2, and 12.2.3.3 of the revised CAR. As part of the license application, HFE

verification and validation activities will be conducted to support construction and startup. The

HSI design will be verified in accordance with the configuration management and design control

processes discussed in Chapter 15 of the revised CAR. A final personnel activities review will be

performed during startup testing. This review will be an integrated system validation of

personnel activities relied on for safety including, but not limited to, HSIs, procedure

development, training development, staffing, and maintenance tasks. The human performance

activities identified in the functional allocations and task analysis will be updated in the license

application to reflect the results of the ISA.

12.1.7 Design Bases of the PSSCs

In Chapter 5 of the revised CAR, DCS identified administrative controls and HSIs as PSSCs, to

be implemented by appropriate procedures, training, and management measures. These

PSSCs are the human factors PSSCs for the facility. The applicant has stated that the facility is

being designed to maximize the use of automation, thus minimizing human operations and

interactions with the facility SSCs. By reducing these interactions, the probability of a humancaused error being introduced is reduced. The applicant has also committed to using guidance

as appropriate from the following standards and NUREGs as additional design bases for

reducing human error:

• IEEE Std. 1023 (IEEE, 1988)

• NUREG-0700, Revision 2 (NRC, 2002a)

• NUREG-0711, Revision 2 (NRC, 2002b)

12.2 Evaluation Findings

In Chapter 12 of the revised CAR, the applicant described the general design philosophy

(hierarchy of controls) and defense-in-depth practices (double contingency protection for

criticality events; single-failure criterion including redundancy, independence, separation, and fail
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safe for PSSCs; plus other noncredited PSSCs) applied during the preliminary design of the

facility. Based on that information and the discussion provided in the sections above for HFE for

personnel activities, the staff concludes that the applicant’s HFE plans, processes, and analyses

provide reasonable assurance that the design bases of the relevant PSSCs identified by DCS

will protect against natural phenomena and the consequences of potential accidents, and are

thus adequate to approve the revised CAR, pursuant to 10 CFR 70.23(b). The applicant will be

required to submit more detailed evaluation of HFE as part of any later license application it may

submit.

12.3 References

(DCS, 2001) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Response to Request for Additional Information—Construction Authorization

Request. August 31, 2001.

(DCS, 2002a) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Clarification of Responses to NRC Request for Additional Information. 

January 7, 2002.

(DCS, 2002b) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Clarification of Responses to NRC Request for Additional Information. 

February 11, 2002.

(DCS, 2002c) Ihde, R., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, Mixed Oxide Fuel Fabrication Facility Construction Authorization Request Revision. 

October 31, 2002 (including page changes through February 9, 2005).

(DCS, 2002d) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Response to Request for Additional Information, Clarification, and Open Item

Mapping into the Construction Authorization Request Revision. November 22, 2002.

(DOD, 1989) U.S. Department of Defense. MIL-STD-1472D, “Human Engineering Design

Criteria for Military Systems, Equipment and Facilities.” Washington, DC, March 1989.

(IEEE, 1988) Institute of Electrical and Electronics Engineers. Std. 1023, “IEEE Guidelines for

the Application of Human Factors Engineering to Systems and Equipment and Facilities of

Nuclear Power Generating Facilities.” 1988.

(NRC, 1999) U.S. Nuclear Regulatory Commission. “Domestic Licensing of Special Nuclear

Material (10 CFR Part 70).” Federal Register: Vol. 64, No. 146. pp. 41338–41357. July 30,

1999.

(NRC, 2000a) U.S. Nuclear Regulatory Commission. NUREG/CR-6633, “Advanced Information

Systems Design: Technical Basis and Human Factors Review Guidance.” Washington, DC,

March 2000.

(NRC, 2000b) U.S. Nuclear Regulatory Commission. NUREG/CR-6634, “Computer-Based

Procedure Systems: Technical Basis and Human Factors Review Guidance.” Washington, DC,

March 2000.
12-8

(NRC, 2000c) U.S. Nuclear Regulatory Commission. NUREG/CR-6635, “Soft Controls: 

Technical Basis and Human Factors Review Guidance.” Washington, DC, March 2000.

(NRC, 2000d) U.S. Nuclear Regulatory Commission. NUREG/CR-6636, “Maintainability of

Digital Systems: Technical Basis and Human Factors Review Guidance.” Washington, DC,

March 2000.

(NRC, 2000e) U.S. Nuclear Regulatory Commission. NUREG/CR-6637, “Human Systems

Interface and Plant Modification Process: Technical Basis and Human Factors Review

Guidance.” Washington, DC, March 2000.

(NRC, 2001) Giitter, J., U.S. Nuclear Regulatory Commission. Letter to P. Hastings, Duke

Cogema Stone & Webster, RE: MOX Fuel Fabrication Facility Construction

Authorization—Request for Additional Information. June 21, 2001.

(NRC, 2002a) U.S. Nuclear Regulatory Commission. NUREG-0700, Rev. 2, “Human-System

Interface Design Review Guidelines.” Washington, DC, May 2002.

(NRC, 2002b) U.S. Nuclear Regulatory Commission. NUREG-0711, Rev. 2, “Human Factors

Engineering Program Review Model.” Washington, DC, May 2002.
13-1

13. SAFEGUARDS

13.1 Physical Protection

13.1.1 Conduct of Review

This chapter of the final safety evaluation report (FSER) reviews the physical protection

commitments made by Duke Cogema Stone & Webster (DCS or the applicant) in Chapter 13 of

the revised construction authorization request (CAR) (DCS, 2002b) and supplemental

information provided by the applicant. The staff used applicable portions of NUREG-1718,

“Standard Review Plan for the Review of an Application for a Mixed Oxide (MOX) Fuel

Fabrication Facility,” issued August 2000, as guidance in performing the review (NRC, 2000). 

The staff also reviewed the applicant’s letter of March 8, 2002 (DCS, 2002a), which responded

to a request for information. The objective of the review is to determine whether the applicant

has committed to developing and implementing a physical protection plan that meets the

requirements in Title 10, Section 73.20, “General Performance Objectives and Requirements,” of

the Code of Federal Regulations (10 CFR 73.20); 10 CFR 73.45, “Performance Capabilities for

Fixed Site Physical Protection Systems”; and 10 CFR 73.46, “Fixed Site Physical Protection

Systems, Subsystems, Components, and Procedures,” and Appendices B, C, G, and H to

10 CFR 73.46.

13.1.2 Evaluation Findings

The applicant committed to submit, as part of any later license application, (1) a physical security

plan that meets the requirements of 10 CFR 73.20, 73.45, and 73.46 and Appendix B, “General

Criteria for Security Personnel”, Appendix C, “Licensee Safeguards Contingency Plans,”

Appendix G, “Reportable Safeguards Events,” and Appendix H, “Weapons Qualification Criteria,”

to 10 CFR Part 73, “Physical Protection of Plants and Materials,” and (2) a design that meets the

design basis threat (NRC, 2003) (note that the attachments to this letter contain classified

information and are not publicly available).

The staff concludes that the applicant has provided adequate commitments to submit, as part of

its license application, a physical security plan that will meet the applicable physical protection

requirements in 10 CFR Part 73. 

The U.S. Nuclear Regulatory Commission (NRC) is conducting a comprehensive review of

safeguards programs and design basis threats as a result of the September 11, 2001, events. 

When this review is completed, a determination will be made with respect to the effect on the

mixed oxide fuel fabrication facility design.

13.1.3 References

(DCS, 2002a) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Clarification of Responses to NRC Request for Additional Information. March

8, 2002.
13-2

(DCS, 2002b) Ihde, R., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Mixed Oxide Fuel Fabrication Facility Construction Authorization Request

Revision. October 31, 2002 (including page changes through February 9, 2005).

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility.” Washington,

DC, January 2000.

(NRC, 2003) Pierson, R., U.S. Nuclear Regulatory Commission. Letter to P. Hastings, Duke

Cogema Stone & Webster, RE: Design Basis Threat. July 28, 2003.

13.2 Material Control and Accounting

13.2.1 Conduct of Review

The staff reviewed the applicant’s treatment of material control and accounting (MC&A) issues in

the revised CAR (DCS, 2002) for the MFFF using the guidance in Section 13.2 of NUREG-1718

(NRC, 2000). The purpose of this review was to establish whether the applicant’s design basis

information for systems relevant to MC&A, and related commitments, will lead to an acceptable

Fundamental Nuclear Material Control (FNMC) Plan that will meet the requirements of 10 CFR

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

The material control system must be designed to protect against, detect, and respond to the loss

or diversion of nuclear materials. The material accounting system must be designed to

determine the quantities of nuclear materials in a licensee’s possession, maintain knowledge of

such materials, verify the presence of such materials, and detect the loss or diversion of such 

materials. The staff reviewed and evaluated information provided by the applicant that is related

to the design bases for the proposed MC&A program. The following physical aspects of the

MC&A design were reviewed:

• process monitoring

• item monitoring

• alarm resolution

• quality assurance and accounting

• international safeguards

13.2.1.1 Process Monitoring

The applicant identified a detailed segmentation of the main processes into process control

subunits which are capable of monitoring the status of licensed material in process. The

applicant also integrated MC&A features into the proposed facility’s manufacturing management

information system (MMIS). The staff evaluated the applicant’s process monitoring program and

finds that this program consists of adequate design features in its abrupt loss detection

capability, including process subdivision and measurement points, material control tests, location

categorization, material substitution, trend analysis, material exemptions, and research and

development operations.
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13.2.1.2 Item Monitoring

The applicant committed to an item monitoring program, which establishes item identification

and the basis for verifying the presence and integrity of licensed nuclear materials. The

applicant also applied MC&A aspects of this program to the facility design and the highly

automated remote-controlled process and manufacturing facility features, including its MMIS. 

The staff has reviewed the elements of this item monitoring program and finds that the

applicant’s program is capable of providing timely plant-wide detection of the loss of items and

real-time status of nuclear materials, including a system of item identification, item classification,

tamper-safing procedures, material accessibility, item accounting and control procedures, item

measurements, sample items, and item verification tests.

13.2.1.3 Alarm Resolution

The applicant identified features of an alarm resolution system which is capable of resolving the

nature and cause of any MC&A alarm within an approved time period, notifying the NRC of any

unresolved alarms, and establishing and maintaining the ability to respond to any suspected

thefts of licensed nuclear material. The staff has reviewed the elements of this alarm resolution

program and finds that the applicant’s program commitments demonstrate it will have the ability

to respond to and resolve MC&A alarms of potential loss of nuclear materials, including alarm

resolution procedures, decision rules, response time, item discrepancies, alarm reporting

responsibilities, and any suspected thefts. 

13.2.1.4 Quality Assurance and Accounting

The applicant identified feasible approaches and methods with respect to the 11 elements of its

MC&A Quality Assurance Program, which includes management structure, personnel training

and qualification, measurement systems, measurement control, physical inventory, records

system and maintenance, shipments and receipts, scrap material control, human errors,

independent assessment, and material custodial responsibilities. The staff has reviewed the

elements of these quality assurance and accounting programs and finds that these program

commitments are appropriate and acceptable.

13.2.1.5 International Safeguards

The applicant outlined design features for potential future interaction involving the International

Atomic Energy Agency and related international safeguards agreements and regulatory

requirements. The staff finds that these design features are adequate.

13.2.2 Evaluation Findings

The staff concludes that the applicant provided adequate commitments to submit, as part of its

application for a license to possess and use licensed materials, an FNMC Plan that will meet the

10 CFR Part 74 requirements. The staff finds that, at the construction authorization approval

stage, MC&A issues have been adequately addressed.
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13.2.3 References

(DCS, 2002) Ihde, R., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Mixed Oxide Fuel Fabrication Facility Construction Authorization Request

Revision. October 31, 2002 (including page changes through February 9, 2005).

(NRC, 1995) U.S. Nuclear Regulatory Commission. NUREG-1280, “Standard Format and

Content Acceptance Criteria for the Material Control and Accounting (MC&A) Reform

Amendment, Rev. 1.” Washington, DC, April 1995. 

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility.” Washington,

DC, January 2000.
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14. EMERGENCY MANAGEMENT

14.1 Conduct of Review

This chapter of the final safety evaluation report (FSER) reviews the emergency management

and related information provided by Duke Cogema Stone & Webster (DCS or the applicant)

during a meeting with the staff on January 4, 2001 (NRC, 2001), further related information

presented by the applicant in Chapters 5, 11, and 14 of the revised construction authorization

request (CAR) (DCS, 2002), and information in the revised CAR Table 5.6-1 (principal

structures, systems, and components (PSSCs) related to emergency conditions). The staff used

the applicable portions of Chapter 14.0 in NUREG-1718, “Standard Review Plan for the Review

of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility,” issued 2000 (NRC, 2000),

as guidance in performing the review. The staff reviewed how the information in the revised

CAR, and the other information referenced above, addresses the following regulations:

• 10 CFR 70.23(b) states, as a prerequisite to construction approval, that the design bases

of the PSSCs and the Quality Assurance Program must be found to provide reasonable

assurance of protection against natural phenomena and the consequences of potential

accidents.

• 10 CFR 70.64, “Requirements for New Facilities or New Processes at Existing Facilities,”

requires that baseline design criteria (BDC) and defense-in-depth practices be

incorporated into the design of new facilities. With respect to emergency capability, 10

CFR 70.64(a)(6) requires that the design of the proposed facility provide for emergency

capability to maintain control of (1) licensed material and hazardous chemicals produced

from licensed material, (2) evacuation of onsite personnel, and (3) onsite emergency

facilities and services that facilitate the use of available offsite services.

As discussed further below, the staff evaluated the applicant’s safety assessment provided in

Chapter 5 of the revised CAR (including the PSSCs related to emergency conditions referenced

in revised CAR Table 5.6-1) to ensure that the applicant considered all appropriate accident

initiators and to ensure that the BDC for emergency capability were met. The staff found that

the applicant considered hazards from internal, external, and natural phenomena. Those

accident initiators that were not used as parts of the design basis were adequately explained. 

While DCS did not specifically cite 10 CFR 70.64(a)(6) in the revised CAR, it did address how it

planned to meet all of the BDC in the slides it presented at a meeting on January 4, 2001 (NRC,

2001). In the presentation, the applicant indicated that it would meet the BDC for emergency

capability through emergency planning design criteria and the site work task agreement (WTA),

including integration with the U.S. Department of Energy (DOE) Savannah River Site (SRS)

Emergency Plan; the Emergency Onsite Mixed Oxide Fuel Fabrication Facility Evacuation Plan;

utilization of existing onsite facilities and services at SRS, coordinated through WTA; and safe

havens provided to personnel while controlling potential losses of licensed material. 

The applicant’s assessment of potential loss of control of radioactive material is discussed in 

Section 5.5.2.1, "Loss of Confinement/Dispersal of Nuclear Material Events,” of the revised CAR. 

In Section 5.5.2.10.6.2, “Events Involving Hazardous Chemicals and Radioactive Material,” the

applicant concluded that the chemical events would be bounded by the radioactive release

events and no additional PSSCs in this area would be required. The staff finds that the applicant
14-2

has adequately considered all of the event categories in its analysis, for emergency planning

purposes.

In revised CAR Section 5.5.2.1.6, the applicant identified the following 10 potential loss-ofconfinement events:

(1) over-temperature

(2) corrosion

(3) small breaches in glovebox confinement boundary or backflow from a glovebox through

utility lines

(4) leaks of aqueous polishing process vessels or pipes within process cells

(5) canister handling operations

(6) rod handling operations

(7) breaches in containers outside gloveboxes caused by handling operations

(8) over/underpressurization of glovebox

(9) excessive temperature caused by decay heat from radioactive materials 

(10) glovebox dynamic exhaust failure 

The applicant identified the following 12 PSSCs either to prevent these hypothetical events from

occurring, or to mitigate the potential consequences:

(1) the safety instrumentation and controls system

(2) C3 confinement system

(3) material maintenance and surveillance programs

(4) the C4 confinement system

(5) the process cell

(6) process cell entry controls

(7) 3013 canister outer can opening device

(8) M 3013 canister

(9) material-handling equipment and material-handling controls

(10) glove box pressure controls

(11) the high-depressurization exhaust system

(12) training and procedures 

The staff’s evaluation of the applicant’s methodology concerning loss of confinement is

presented in Chapter 5 of this revised FSER. The staff notes that 10 CFR 20.1801, “Security of

Stored Material,” and 10 CFR 20.1802, “Control of Material Not in Storage,” have specific

requirements for control of licensed material, and the applicant will need to address these

requirements as part of any later license application it may submit.
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[Text removed under 10 CFR 2.390]

[Text removed under 10 CFR 2.390]

14.2 Evaluation Findings

In Section 14 of the revised CAR, DCS provided design basis information pertaining to

emergency management measures for the proposed facility. Based on the staff’s review of the

revised CAR and supporting information provided by the applicant relevant to emergency

management, the staff finds that DCS, at the construction authorization stage, has met the BDC

for emergency capability set forth in 10 CFR 70.64(a)(6). Accordingly, the staff concludes,

pursuant to 10 CFR Section 70.23(b), that the design bases of the PSSCs pertaining to

emergency management provide reasonable assurance of protection against natural

phenomena and the consequences of potential accidents.

14.3 References

(DCS, 2002b) Ihde, R., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Mixed Oxide Fuel Fabrication Facility Construction Authorization Request

Revision. October 31, 2002 (including page changes through February 9, 2005).
14-4

(NRC, 1998) U.S. Nuclear Regulatory Commission. NUREG/CR-6410,”Nuclear Fuel Cycle

Facility Accident Analysis Handbook.” Washington, DC, 1998.

(NRC, 2000) U.S. Nuclear Regulatory Commission. NUREG-1718, “Standard Review Plan for

the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility.” Washington,

DC, January 2000.

(NRC, 2001) Persinko, A., U.S. Nuclear Regulatory Commission. Memorandum to E. Leeds,

U.S. Nuclear Regulatory Commission, RE: January 4–5, 2001, Summary of Meeting with Duke

Cogema Stone & Webster to Discuss Design Basis for the Mixed Oxide Fuel Fabrication Facility. 

January 24, 2001. 
15-1

15. MANAGEMENT MEASURES

This chapter of the final safety evaluation report (FSER) contains the staff’s review of

management measures described by the Duke Cogema Stone & Webster (DCS or the

applicant) in Chapter 15 of the revised construction authorization request (CAR) (DCS, 2002b). 

Management measures are defined in Title 10, Section 70.4, “Definitions,” of the Code of

Federal Regulations (10 CFR 70.4) as functions, performed by a licensee, generally on a

continuing basis, that are applied to items relied on for safety (IROFS), to provide reasonable

assurance that the items are available and reliable to perform their functions, when needed. As

further stated in the 10 CFR 70.4 definition, management measures include configuration

management (CM), maintenance, training and qualifications, procedures, audits and

assessments, incident investigations, records management, and other quality assurance (QA)

elements. Management measures are made applicable to the revised CAR review by

10 CFR 70.64(a)(1). For purposes of establishing quality standards, 10 CFR 70.64(a)(1)

requires that the design of the proposed mixed oxide (MOX) fuel fabrication facility (MFFF, or the

facility) “must be developed and implemented in accordance with management measures, to

provide adequate assurance that IROFS will be available and reliable to perform their function

when needed.” 

In the MOX Project Quality Assurance Plan (MPQAP), Revision 2, the applicant had previously

addressed the QA program description requirements (which are referenced in FSER Section

15.1). The staff issued a safety evaluation report (SER) approving the MPQAP, including

additional Duke Cogema, Stone & Webster (DCS) clarification and commitments regarding two

QA issues, on October 1, 2001 (NRC, 2001, Section 15.1). Subsequently, DCS revised the

MPQAP to address the additional clarifications and commitments and submitted Revision 3 of

the MPQAP (DCS, 2002a, Section 15.1). The staff verified that the MPQAP, Revision 3,

adequately incorporated all of the DCS clarifications and commitments noted in the SER and

approved the MPQAP, Revision 3, for construction activities (NRC, 2003, Section 15.1). The

staff concluded, pursuant to 10 CFR 70.23(b), that the MPQAP, as applied to all structures,

systems, and components (SSCs) of the proposed facility, will provide reasonable assurance of

protection against natural phenomena and the consequences of potential accidents. The scope

of this 10 CFR 70.23(b) finding was described in the conclusion of the MPQAP SER as

pertaining to the construction of the SSCs for the proposed facility, including all related design

procurement and fabrication activities. As further stated in the conclusion to the MPQAP SER,

this finding did not pertain to any startup testing or operation of the proposed MFFF. 

Accordingly, since the staff in its October 1, 2001, SER, approved the applicant’s general QA

program as described above, this FSER focuses on the additional QA elements and other

management measures described in revised CAR Chapter 15 and evaluates these management

measures pursuant to 10 CFR 70.64(a)(1). The objective of this FSER review is to determine

whether the proposed management measures, together with the previously approved MPQAP,

establish a QA program which provides reasonable assurance of protection against natural

phenomena and the consequences of potential accidents, as required for revised CAR approval

by 10 CFR 70.23(b). For the reasons stated in the following FSER sections, the staff concludes,

pursuant to 10 CFR 70.23(b), that the proposed management measures set forth in the revised

CAR, together with the previously approved MPQAP, establish a QA program which provides

reasonable assurance of protection against natural phenomena and the consequences of

potential accidents.
15-2

The staff evaluated the information provided by the applicant for management measures by

reviewing Chapter 15 of the revised CAR, other sections of the revised CAR, supplementary

information provided by the applicant, and relevant documents available at the applicant’s offices

but not submitted by the applicant. The review of management measures was closely

coordinated with the review of accident sequences described in the safety assessment of the

design bases (see Chapter 5 of this FSER). Note that management measures will be applied to

IROFS, and that IROFS need not be specified until the applicant submits its integrated safety

analysis (ISA) summary (see 10 CFR 70.65(b)(6)). Thus, the staff will perform a more detailed

evaluation of management measures as part of the review of any later license application DCS

may submit. 

The review of the applicant’s description of management measures is addressed in the following

sections in the order that the applicant presented them in the revised CAR, beginning with QA. 

The staff used Chapter 15.0 in NUREG-1718, “Standard Review Plan for the Review of an

Application for a Mixed Oxide (MOX) Fuel Fabrication Facility,” issued 2000, as guidance in

performing the review.

15.1 Quality Assurance

15.1.1 Conduct of Review

The staff reviewed the QA descriptions and commitments of the applicant’s revised CAR (DCS,

2002b) for the MFFF in accordance with NUREG-1718. The purpose of this review is to

establish that the applicant has a plan to integrate its implementation of management measures

into its approved MPQAP, which is applicable to the design, fabrication, and construction of all

SSCs, including those designated as principal SSCs (PSSCs) (NRC, 2001). The following

regulations apply specifically to QA:

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

requirements for identifying, controlling, and reporting defects in a facility, activity, or basic

component supplied to a facility licensed under the Atomic Energy Act which could create

a substantial safety hazard.

• 10 CFR 70.64(a)(1) states that, for quality standards, the facility design must be

developed and implemented in accordance with management measures, to provide

adequate assurance that IROFS will be available and reliable to perform their function

when needed.

• 10 CFR 70.64(a)(8) states that the design of IROFS must provide for inspection and

testing to ensure that IROFS will be available and reliable to perform their functions when

needed.

• 10 CFR 70.23(b) states, as a prerequisite to revised CAR approval, that the QA program

must be found to provide reasonable assurance of protection against natural phenomena

and the consequences of potential accidents. This section further states that the criteria

in Appendix B, “Quality Assurance Criteria for Nuclear Power Plants and Fuel

Reprocessing Plants,” (referred to hereafter as Appendix B) to 10 CFR Part 50, “Domestic

Licensing of Production and Utilization Facilities,” will be used in determining whether the
15-3

QA program provides reasonable assurance of protection against natural phenomena and

the consequences of potential accidents.

In the revised CAR Section 15.1, the applicant stated that the application of management

measures will ensure that the PSSCs are available and reliable to perform their intended design

functions.

The applicant submitted MPQAP, Revision 2, as the required description of a QA program that

meets the general QA requirements of Appendix B. The applicant has committed to compliance

with the provisions of Parts I and II of American Society of Mechanical Engineers (ASME) NQA1-1994, “Quality Assurance Program Requirements for Nuclear Facilities,” as revised by the

ASME NQA-1a-1995 Addenda, issued 1994 (ASME, 1994), and U.S. Nuclear Regulatory

Commission (NRC) RG 1.28, Revision 3, “Quality Assurance Program Requirements (Design

and Construction), issued 1985 (NRC, 1985).” These ASME NQA-1 and G provisions are

hereafter referred to as NQA-1. The staff reviewed, compared, and evaluated the MPQAP

against the Appendix B requirements and the NQA-1 provisions, as well as NUREG-1718

guidance. The QA areas reviewed by the staff include the applicant’s descriptions in the

MPQAP for organization, QA function, design control, procurement document control,

instructions, procedures, drawings, document control, control of purchased items, identification

and control of items, control of special processes, inspection, test control, control of measuring

and test equipment, handling, storage, and shipping, inspection, test, and operating status,

nonconformances, corrective action, QA records, audits and assessments, and the applicant’s

provisions for continuing QA. The results of this review are documented in the MPQAP SER

dated October 1, 2001 (NRC, 2001). Additional discussions on the implementation and

application of the applicant’s QA program description in the MPQAP, Revision 2, as

supplemented by responses to requests for additional information (RAIs), were held during

telephone meetings and in-office reviews, particularly on the quality level categorization

commitments, including classification criteria and the categorization process. DCS incorporated

all additional clarifications and commitments referred to in the MPQAP SER into Revision 3 of

the MPQAP (DCS, 2002a). The staff reviewed the MPQAP, Revision 3, verified that it

appropriately incorporated all commitments and clarifications, and approved it for construction

application (NRC, 2003). In MPQAP, Revision 3, the applicant commits to invoking and

complying with the applicable requirements of 10 CFR Part 21 for all design, procurement,

fabrication, and construction activities.

15.1.2 Evaluation Findings

Based on the staff’s review of revised CAR Section 15.1 and the clarifications and commitments

made by the applicant in response to NRC RAIs relevant to the QA program, the staff finds that

DCS has appropriately committed to invoking the applicable 10 CFR Part 21 requirements for

design, procurement, and construction. Additionally, the staff finds, pursuant to

10 CFR 70.64(a)(1), that the PSSCs are being properly designed and developed, thus providing

adequate assurance that IROFS will be available and reliable to perform their intended design

functions when needed. However, as stated in the introduction to this chapter, the staff will

perform a more detailed evaluation of management measures as part of the review of the

license application, which the applicant plans to submit to the NRC if its revised CAR is

approved. The staff also finds, pursuant to 10 CFR 70.64(a)(8), that the inspection and testing

elements of QA are in place, so as to adequately ensure that IROFS will be available and

reliable to perform their functions when needed. 
15-4

Based on the above findings, together with its earlier approval of the MPQAP (NRC, 2001; DCS,

2002a), the staff concludes, pursuant to 10 CFR 70.23(b), that the QA program at the proposed

facility will provide reasonable assurance of protection against natural phenomena and the

consequences of potential accidents. The scope of this conclusion pertains to the construction of

the facility’s PSSCs and includes all related design, procurement, and fabrication activities but

does not include any startup testing or operation of the proposed facility. 

15.1.3 References

(ASME, 1994) American Society of Mechanical Engineers. ASME–NQA–1–1994, “Quality

Assurance Requirements for Nuclear Facility Applications” (as revised by the ASME NQA-1a1995 Addenda). New York, 1994/1995. 

(DCS, 2002a) Hastings, P., Duke Cogema Stone & Webster. Letter RE: Quality Assurance

Program for Construction of the Mixed Oxide Fuel Fabrication Facility. March 26, 2002.

(DCS, 2002b) Ihde, R., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Mixed Oxide Fuel Fabrication Facility Construction Authorization Request

Revision. October 31, 2002 (including page changes through February 9, 2005).

(NRC, 1985) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.28, Rev. 3, “Quality

Assurance Program Requirements (Design and Construction).” Washington, DC, August 1985.

(NRC, 2001) Persinko, A., U.S. Nuclear Regulatory Commission. Letter to P. Hastings, Duke

Cogema Stone & Webster, RE: Duke Cogema Stone & Webster Quality Assurance Program for

the Construction of the MFFF. October 1, 2001.

(NRC, 2003) Persinko, A., U.S. Nuclear Regulatory Commission. Letter to P. Hastings, Duke

Cogema, Stone & Webster, RE: DCS Quality Assurance Program for Construction of the Mixed

Oxide Fuel Fabrication Facility. January 10, 2003.

15.2 Configuration Management

15.2.1 Conduct of Review

This section of the FSER contains the staff’s review of the CM system committed to by the

applicant in Section 15.2 of the revised CAR (DCS, 2002). The objective of this review is to

verify that the applicant has adequately planned for the implementation of an acceptable CM

system which will provide reasonable assurance that the PSSCs identified by the applicant will

be available and reliable to perform their safety function when needed. The review, for

construction authorization, is to determine whether the applicant has adequately planned for CM

to be accomplished during design and construction and whether necessary policies, personnel,

procedures, and instructions will be in place to begin CM during the design and construction of

the PSSCs.

The following regulations apply specifically to CM:

• 10 CFR 70.4 defines the term “configuration management” as a management measure

“that provides oversight and control of design information, safety information, and records
15-5

of modifications (both temporary and permanent) that might impact the ability of items

relied on for safety to perform their functions when needed.”

• 10 CFR 70.64(a)(1) states that, for quality standards, the MFFF design must 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.

Section 15.2.3, “Areas of Review,” of NUREG-1718 defines the review areas and states that the

description of the CM system is to be reviewed with emphasis on the processes for documenting

an established baseline configuration and controlling changes to it to preclude inadvertent

degradation of safety. The review is to include the applicant’s descriptions of the organizational

structure responsible for CM activities and the process, procedures, and documentation required

for modifying SSCs, PSSCs, and IROFS and the supporting management measures. The

review is to focus on the applicant’s management level controls that ensure (1) the disciplined

documentation of engineering, installation, and operation of modifications, (2) the training and

qualification of affected staff; (3) revision and distribution of operating, test, calibration,

surveillance, and maintenance procedures and drawings, (4) postmodification testing, and (5)

operational readiness review. The review topics are to include CM policy, design requirements,

document control, change control, and assessments.

15.2.1.1 CM Policy

The applicant described its overall CM system in revised CAR Section 15.2 and presented its

CM policy in Section 15.2.1, which states that CM is provided for PSSCs throughout facility

design, construction, testing, operation and deactivation, to provide the means to establish and

maintain a technical baseline for the facility. The applicant’s CM system during design and

construction is the responsibility of the Deputy Project Manager—MFFF Engineering and

Construction. The CM system controls documents in accordance with QA procedures for design

control, document control, and records management. During the design and construction

phases, the applicant’s CM is, and will be, based on Section 3, “Design Control,” of the MPQAP

and associated MPQAP requirements and procedures for design and construction documents

and activities that establish and maintain the technical baseline. The staff reviewed these

MPQAP commitments and requirements for CM and determined that they were acceptable for

construction activities, including design, procurement, and fabrication. The staff review and

conclusion were documented in an SER, on October 1, 2001 (NRC, 2001). Design documents

and changes undergo interdisciplinary review and verification. Proper implementation is verified

and reflected in the design basis documentation. The applicant will update the CM description to

include details of the operational CM program in the license application. If the facility is licensed

to operate, any changes to the CM program would be governed by 10 CFR 70.72, “Facility

Changes and Change Process.”

Section 15.2 of the revised CAR states that CM is applied to SSCs. The applicant addressed its

commitment to CM application during design and construction and specifically described (1)

establishing and controlling the design bases to include all SSCs, not just PSSCs and IROFS,

(2) how the CM process functions for documenting the baseline configuration and controlling all

changes, and (3) how it provides for change control during construction (i.e., that the same

design control procedures are to be used for all SSCs, not just PSSCs and IROFS). During any
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construction of the proposed facility, all field changes, as-built configurations, and

nonconformances will be reviewed for impact to the design basis.

In Section 15.2 of the revised CAR, the applicant described the design documents under CM,

which include calculations, safety analyses, design criteria, engineering drawings, system

descriptions, technical documents, and specifications that establish design requirements. The

scope of CM expands throughout the design process. During any construction, startup, and

operations, the scope of documents under CM would expand to include vendor, test, and

inspection data; startup, test, operating, and administrative procedures; and nonconformance

reports. These documents will include those generated through functional interfaces with QA,

maintenance, and training and qualifications of personnel. The applicant described how the CM

system is implemented through or related to other management measures and described these

interfaces and relationships.

The approved MPQAP establishes the framework for the applicant’s CM system and other

management measures for PSSCs (NRC, 2001; NRC, 2003). The CM system records are

generated and processed in accordance with the requirements of the MPQAP. Maintenance

requirements are established as part of the design basis which is controlled under CM, and

records provide evidence of compliance with preventive and corrective maintenance schedules. 

Training and qualification of personnel are controlled in accordance with the MPQAP provisions

and will be considered part of the design basis controlled under CM. Corrective actions and

changes resulting from audits, assessments, and incident investigations will be evaluated and

controlled in accordance with provisions of the MPQAP and QA procedures. Plant procedures

will be controlled in accordance with the MPQAP and QA procedures and will be reviewed for

impact on the design basis.

The applicant’s description of CM includes the designation of PSSCs under the QA classification

and grading provisions of Section 2 of the MPQAP. The grading approach to SSCs includes

applying the most stringent QA controls to SSCs with the highest safety significance. All

controls for all PSSCs are controlled under CM and documented in the same manner. The

applicant’s commitments for SSC classification and the grading process are discussed in the

SER approving the MPQAP (NRC, 2001), and are incorporated in MPQAP, Revision 3 (NRC,

2003).

15.2.1.2 Design Requirements

The applicant stated that the organization structure and staffing interfaces for the CM system,

for design and construction, will be administered by the MFFF engineering organization. The

lead discipline engineers will have primary technical responsibility for the work performed by

their disciplines and will be responsible for the conduct of interdisciplinary reviews. Reviews will

also be conducted by construction management, operations, QA, and procurement personnel. 

The design control process will also interface with document control and records management

processes controlled by QA procedures. The PSSCs are designated as Quality Level 1 (QL-1)

using the approved MPQAP classification process, and their associated design documents will

be subject to review and verification. Analyses constituting the safety assessment of the design

bases are subject to these same requirements, and changes are evaluated to ensure

consistency with the design bases. Design bases documented in CAR Chapters 5 through 11

will be consistent with those in, and flowed down from, the design requirements and basis of

design documents, analyses, specifications, and drawings. The CM system will capture these
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requirements and resulting design bases in accordance with design control, document control,

and records management procedures.

The SSCs will be classified based on their safety significance and role in preventing or mitigating

design basis accidents in accordance with the categories of QA classification described in

Section 2 of the MPQAP and approved by the staff (NRC, 2001; NRC, 2003).

15.2.1.3 Document Control

Document control will be implemented in accordance with Section 6 of the MPQAP. These

provisions were reviewed by the staff and found to be adequate for design and construction of

the proposed facility (NRC, 2001). 

15.2.1.4 Change Control

The applicant’s description of control of changes to the technical baseline identifies that these

changes are controlled under Section 3, “Design Control,” of the MPQAP and associated

procedures. The change control process includes technical, management, and safety reviews

before implementation. The review process includes reviews to ensure consistency with the

approved safety assessment of the design bases of PSSCs and includes provisions for

appropriate reviews at the design, construction, and operations phases. 

15.2.1.5 Assessments

The applicant confirms that assessments, including initial and periodic examinations of the CM

system, will be conducted to determine the system’s effectiveness and to correct deficiencies. 

The applicant committed that such assessments will be systematically planned and conducted in

accordance with an overall facility audit and assessment program as described by DCS in

revised CAR Section 15.6. 

The applicant also committed to updating the CM system to reflect any changes to the proposed

facility made between submittal of the revised CAR and the expected submittal of its later

license application.

15.2.2 Evaluation Findings

In Chapter 15.2 of the revised CAR, DCS committed to implement and update its CM system at

the proposed facility. Management-level policies and procedures, including a safety review of

any proposed activity involving SSCs, are described that will ensure that the relationship

between design requirements, construction, and facility documentation is maintained as part of a

new design or change in an existing design. The administrative control will ensure that the

organizational structure, procedures, and responsibilities necessary to implement CM are in

place or committed to; that the design requirements and bases are documented and supported

by analyses and the documentation is maintained current; that documents, including drawings,

are appropriately stored and accessible; that drawings and related documents adequately

describe SSCs; that procedures adequately describe how the applicant will achieve and maintain

strict consistency among the design requirements, facility construction, and facility

documentation; and that methods are in place for suitable analysis, review, approval, and

implementation of identified changes to SSCs. The applicant described its approach to QL
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categorization and grading of controls for SSCs and identified the process, criteria, and control

to be applied. Pursuant to 10 CFR 70.64(a)(1), the staff finds that the applicant’s proposed CM

system is a management measure ensuring that the facility design is being properly developed

and implemented, so as to provide adequate assurance that IROFS will be available and reliable

to perform their function when needed. The applicant will describe its operation phase change

process, for 10 CFR 70.72 provisions, in more detail in its license application. Accordingly, the

staff concludes, pursuant to 10 CFR 70.23(b), that the CM system set forth in the MPQAP and

the revised CAR is part of a QA program which will provide reasonable assurance of protection

against natural phenomena and the consequences of potential accidents.

15.2.3 References

(DCS, 2001) Hastings, P., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Response to Request for Additional Information (DCS-NRC-000059). August

31, 2001.

(DCS, 2002) Ihde, R., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Mixed Oxide Fuel Fabrication Facility Construction Authorization Request

Revision. October 31, 2002 (including page changes through February 9, 2005).

(DOE, 1993) U.S. Department of Energy. DOE–STD–1073–93–Pt.1 and –Pt.2, “DOE Standard

Guide for Operational Configuration Management Program.” Washington, DC, 1993. 

(NRC, 2001) Persinko, A., U.S. Nuclear Regulatory Commission. Letter to P. Hastings, Duke

Cogema Stone & Webster, RE: DCS Quality Assurance Program for Construction of the MFFF. 

October 1, 2001.

(NRC, 2003) Persinko, A., U.S. Nuclear Regulatory Commission. Letter to P. Hastings, Duke

Cogema Stone & Webster, RE: DCS Quality Assurance Program for Construction of the MFFF. 

January 10, 2003.

15.3 Maintenance

15.3.1 Conduct of Review

This section of the FSER contains the staff’s review of the Maintenance Program committed to

by the applicant in Section 15.3 of the revised CAR (DCS, 2002). The objective of this review is

to determine whether the proposed facility will have a maintenance program which will provide

adequate assurance that IROFS (other than personnel activities) will be available and reliable to

perform their safety function when needed. 

The following regulations apply specifically to the staff’s review of the proposed Maintenance

Program:

• 10 CFR 70.4 defines the term “management measures” as including a maintenance

program.

• 10 CFR Section 70.64(a)(1) states that the MFFF design must be developed and

implemented in accordance with management measures to provide adequate assurance
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that items relied on for safety will be available and reliable to perform their function when

needed.

• 10 CFR Section 70.64(a)(8) states that the design of IROFS must provide for

maintenance to ensure that IROFS will be available and reliable to perform their functions

when needed.

15.3.1.1 Maintenance Program

In revised CAR Section 15.3, DCS described and committed to implementation of a

maintenance program, including safety controls, surveillance/monitoring, corrective

maintenance, preventive maintenance, functional testing, and work control methods and

describes the relationship of the maintenance elements to other management measures. The

applicant will describe its maintenance program in more detail in the license application. 

Preventive maintenance activities, surveillance, and performance trending will be done to

provide reasonable and continuing assurance that IROFS will be available and reliable to

perform their safety functions.

15.3.1.2 Safety Controls

The applicant committed to providing safety controls by specifying maintenance requirements for

calibration frequency, functional testing requirements, and replacement of specified components

for IROFS.

15.3.1.3 Maintenance Elements

The applicant’s description requires that surveillance and monitoring of IROFS, including

instrument calibration and testing, be performed at specified intervals to measure the degree to

which IROFS meet performance specifications. The results of surveillances would be trended

and, when indicated by potential performance degradation, preventive frequencies adjusted or

other corrective action taken. Incident investigations may also identify root causes of failures

related to maintenance type or frequency, and lessons learned from these investigations would

be factored into the maintenance program. Procedures would prescribe compensatory

measures for surveillance tests that could be performed only while equipment was out of

service.

Preventive maintenance measures described by the applicant include preplanned and scheduled

periodic refurbishment, overhaul, or replacement of IROFS to ensure their continued safety

function. Planning would include results of surveillance and monitoring and instrument

calibration and testing.

Corrective maintenance would be performed for repair or replacement of equipment that has

unexpectedly degraded or failed. Corrective maintenance would restore IROFS to acceptable

performance through a planned, systematic, controlled, and documented approach for the

activities.

Following initial installation, functional testing of IROFS would be performed as part of the

applicant’s periodic surveillance testing and also after corrective or preventive maintenance or

calibration to ensure that the item is capable of performing its safety function when required. 
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The functional testing would be conducted using approved procedures, which would include

compensatory measures that may be necessary while the test of equipment or systems is being

conducted. 

15.3.1.4 Work Control Methods

As stated in CAR Section 15.3.3, the applicant committed to maintenance-related work control

methods including maintenance management and tracking, which would involve integration of

maintenance activities with ongoing operations activities. Work control methods would also

include appropriate interfaces with radiation protection and associated work permits,

lockout/tagout requirements, and procedures.

15.3.1.5 Maintenance Relationship to Other Management Measures

The applicant’s committed maintenance function will interface with the configuration

management and procedure systems by obtaining the approved and controlled drawings,

specifications, and procedures. Personnel would be trained in the maintenance of IROFS

through the training program, and records of performance trends and maintenance history would

be maintained. 

15.3.2 Evaluation Findings

In Chapter 15.3 of the revised CAR, DCS described and committed to implement its

Maintenance Program to be used on IROFS and associated activities at the proposed facility. 

Pursuant to 10 CFR 70.64(a)(1), the staff finds that the applicant’s proposed Maintenance

Program is a management measure providing adequate assurance that IROFS will be available

and reliable to perform their function when needed. Additionally, pursuant to

10 CFR 70.64(a)(8), the staff finds that the maintenance elements of QA are in place, so as to

adequately ensure that IROFS will be available and reliable to perform their functions when

needed. The applicant will describe its maintenance program in more detail in its license

application. Accordingly, the staff concludes, pursuant to 10 CFR 70.23(b), that the

Maintenance Program set forth in the revised CAR is part of a QA program which will provide

reasonable assurance of protection against natural phenomena and the consequences of

potential accidents.

15.3.3 References

(DCS, 2002) Ihde, R., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Mixed Oxide Fuel Fabrication Facility Construction Authorization Request

Revision. October 31, 2002 (including page changes through February 9, 2005).

(NRC, 1984) U.S. Nuclear Regulatory Commission. Inspection Manual, Procedure 88025,

“Maintenance and Surveillance Testing.” Washington, DC, May 23, 1984.

(NRC, 1989) U.S. Nuclear Regulatory Commission. “Guidance on Management

Controls/Quality Assurance, Requirements for Operation, Chemical Safety, and Fire Protection

for Fuel Cycle Facilities.” Federal Register: Vol. 54, No. 53., pp. 11590–11598. March 21,

1989.
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(NRC, 1996) U.S. Nuclear Regulatory Commission. Inspection Manual, Procedure 88062,

“Maintenance and Inspection. Washington, DC, January 1996.

(NRC, 1997) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.160, Rev. 2,

“Monitoring the Effectiveness of Maintenance at Nuclear Power Plants. Washington, DC, March

1997.

(NRC, 2003) Persinko, A., U.S. Nuclear Regulatory Commission. Letter to P. Hastings, Duke

Cogema, Stone & Webster, RE: DCS Quality Assurance Program for Construction of the MOX

Fuel Fabrication Facility. January 10, 2003. 

15.4 Training and Qualification

15.4.1 Conduct of Review

This section of the FSER contains the staff’s review of the training and qualification information

provided by the applicant in Section 15.4 of the revised CAR (DCS, 2002). The objective of this

review is to determine whether personnel who would perform activities relied on for safety will

understand, recognize the importance of, and be qualified to perform these activities in a

manner that adequately protects the public, worker health and safety, and the environment. The

staff evaluated the applicant’s provisions for training and qualification by reviewing the revised

CAR, the applicant’s QA program description, the MPQAP, responses to NRC staff RAIs, and

relevant documents available at the applicant’s offices but not submitted by the applicant.

The following regulations apply specifically to the staff’s review of the proposed training and

qualification program:

• 10 CFR 70.4 defines the term “management measures” as including a training and

qualification program.

• 10 CFR 70.64(a)(1) states that the MFFF design must 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.

Pursuant to Section 15.4.3 of NUREG-1718, the staff’s review of the proposed training and

qualification program included the following areas: 

• organization and management of training

• analysis and identification of functional areas requiring training

• position training requirements

• development of the basis for training, including objectives

• organization of instruction using lesson plans and other training guides

• evaluation of trainee learning
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• conduct of on-the-job training

• evaluation of training effectiveness

• personnel qualification

• applicant’s provisions for continuing assurance, including the needs for retraining or

reevaluation of qualification 

15.4.1.1 Organization and Management of Training

The applicant described its training program for the operations phase of the facility in revised

CAR Section 15.4 and stated that training program requirements will apply to plant personnel

who perform activities relied on for safety. The applicant’s approved MPQAP addresses training

and qualification requirements during the facility design and construction phases, including QA

training of personnel for nondestructive examination, for inspection and test personnel, and for

auditors. The staff reviewed these MPQAP commitments and requirements for training and

qualification and determined that they were acceptable for construction activities, including

design, procurement, and fabrication (NRC, 2001). DCS specifically committed to the

establishment of an operational training program in accordance with its description in revised

CAR Section 15.4 and will update the program information in any later license application it may

submit.

The applicant’s training program description requires that line managers are responsible for the

content and effective conduct of training for their personnel. Line managers will be given the

authority to implement training, and their responsibilities are included in position descriptions. 

The applicant’s training organization will provide support with planning, directing, analyzing,

developing, conducting, evaluating, and controlling a systematic performance-based training

process. Plant procedures will establish the requirement for indoctrination and training of

personnel performing activities relied on for safety. Lesson plans, which will be used for

classroom and on-the-job training, will be included in the CM system and will be updated based

on design changes or plant modifications. Auditable training records will be maintained to

support management information needs for personnel training, job performance, and

qualifications.

15.4.1.2 Analysis and Identification of Functional Areas Requiring Training

The applicant will perform a needs/job analysis and identify tasks to ensure that appropriate

training is provided to personnel. A task list will be developed and updated as needed and will

be reviewed as part of the systematic valuation of training effectiveness. 

15.4.1.3 Position Training Requirements

The applicant will develop minimum training requirements for positions whose activities are

relied on for safety. Entry-level criteria will be defined for the positions, including minimum

educational, technical, and experience requirements. Initial identification of job-specific training

requirements will be based on experience from MELOX and La Hague operations in France and

from U.S. experience.
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15.4.1.4 Development of the Basis for Training

The applicant will establish learning objectives that identify the training content, based on the

needs/job analyses and position-specific requirements. The task list will be used to develop the

desired posttraining performance objectives, including the knowledge, skills, and abilities that the

trainee should demonstrate; the conditions under which required actions will take place; and the

standards of performance the trainee should achieve on completion of the training activity.

15.4.1.5 Organization of Instruction Using Lesson Plans and Other Training Guides

The applicant will use the learning objectives, derived from specific job performance

requirements and the needs/job analysis, to develop lesson plans and other training guides. 

Lesson plans are approved before use and are used for classroom and on-the-job training.

15.4.1.6 Evaluation of Trainee Learning

The applicant will evaluate trainee mastery of learning objectives through observation and

demonstration or oral or written tests as appropriate. Evaluations will measure the trainee’s

skills and knowledge of job performance requirements.

15.4.1.7 Conduct of On-the-Job Training

The applicant’s description includes requirements for on-the-job training to be performed for

IROFS activities using current performance-based training materials, conducted by designated

personnel who are competent in the program standards and methods of conducting the training. 

Completion of on-the-job training will be demonstrated by actual task performance where

feasible and appropriate. When the actual task cannot be performed by the trainee, a simulation

of the task is performed using the conditions encountered with task performance, including

references, tools, and equipment reflecting the actual task to the extent practical.

15.4.1.8 Evaluation of Training Effectiveness

The applicant will systematically and periodically evaluate the training program’s effectiveness in

producing competent employees. These evaluations will include feedback from trainees and will

identify program strengths and weaknesses, determine whether program content matches

current job needs, and determine if corrective actions are needed to improve the program’s

effectiveness. Evaluation objectives will be developed, results will be documented, and changes

made to procedures, practices, or training materials as necessary.

15.4.1.9 Personnel Qualification 

The applicant discussed its commitments for personnel qualification in revised CAR Section

15.4.9. Qualification requirements for key management positions, addressed in revised CAR

Chapter 4, are in accordance with the applicable guidance in Section 15.4.4.3 of NUREG-1718. 

Revised CAR Chapter 4 currently stresses the organization for design and construction of the

proposed facility. The applicant will supplement the information in Chapter 4 when it submits its

license application to address operations and the qualifications of key plant management

positions. 
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15.4.1.10 Applicant’s Provisions for Continuing Assurance

The applicant’s description of its provisions for continuing assurance of training and qualification

of plant personnel includes an evaluation of personnel performing activities relied on for safety to

determine that they continue to understand, recognize the importance of, and have the

appropriate qualifications needed to perform their activities. The evaluation may be by written or

oral test or by performance evaluation. The evaluation results will be documented, and

retraining or other appropriate action taken when indicated. Retraining will also be required for

plant modifications, procedure changes, and QA program changes when needed.

15.4.2 Evaluations Findings

In Chapter 15.4 of the revised CAR, DCS described its management measures for training and

qualification to be used during the operations phase of the proposed facility. Based on that

information and the discussion provided in the sections above for training and qualification, the

staff finds that the personnel who perform activities relied on for safety will understand,

recognize the importance of, and be qualified to perform these activities in a manner that will

adequately protect the public, worker health and safety, and the environment. The applicant will

establish an operational training program in accordance with its description in revised CAR

Section 15.4. Pursuant to 10 CFR 70.64(a)(1), the staff finds that the applicant’s proposed

training and qualification program is a management measure providing adequate assurance that

IROFS will be available and reliable to perform their function when needed. The applicant will

update the training program description for operations activities in its license application. 

Accordingly, the staff concludes, pursuant to 10 CFR 70.23(b), that the management measures

for training and qualification set forth in the MPQAP and the revised CAR are part of a QA

program which will provide reasonable assurance of protection against natural phenomena and

the consequences of potential accidents.

15.4.3 References

(ASME, 1994) American Society of Mechanical Engineers. ASME–NQA–1–1994, “Quality

Assurance Requirements for Nuclear Facility Applications.” New York, 1994.

(DCS, 2002) Ihde, R., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Mixed Oxide Fuel Fabrication Facility Construction Authorization Request

Revision. October 31, 2002 (including page changes through February 9, 2005).

(NRC, 2001) Persinko, A., U.S. Nuclear Regulatory Commission. Letter to P. Hastings, Duke

Cogema Stone & Webster, RE: Duke Cogema Stone & Webster Quality Assurance Program for

the Construction of the MFFF. October 1, 2001.

(NRC, 1993) U.S. Nuclear Regulatory Commission. NUREG-1220, Rev. 1, “Training Review

Criteria and Procedures.” Washington, DC, January 1993. 

(NRC, 2003) Persinko, A., U.S. Nuclear Regulatory Commission. Letter to P. Hastings, Duke

Cogema Stone & Webster, RE: Duke Cogema Stone & Webster Quality Assurance Program for

the Construction of the MFFF. January 10, 2003.
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15.5 Plant Procedures

15.5.1 Conduct of Review

This section of the FSER contains the staff’s review of procedure information submitted by the

applicant in Section 15.5 of the revised CAR (DCS, 2002). The objective of this review is to

determine whether the applicant can adequately control potential facility operations in areas to

be identified as IROFS, by developing, reviewing, approving, and controlling the implementation

of written plant procedures that would protect the workers, the public, and the environment

during any potential testing, startup, and operation of the facility. The staff evaluated the

applicant's provisions for plant procedures by reviewing Section 15.5 of the revised CAR, other

sections of the revised CAR, responses to NRC staff RAIs, and relevant documents available at

the applicant’s offices but not submitted by the applicant.

The following regulations apply specifically to the staff’s review of the proposed plant

procedures:

• 10 CFR 70.4 defines the term “management measures” as including plant procedures.

• 10 CFR 70.64(a)(1) states that the MFFF design must be developed and implemented in

accordance with management measures, to provide adequate assurance that IROFS will

be available and reliable to perform their function when needed.

For the construction approval, in accordance with guidance in Section 15.5.4.3, NUREG-1718,

the staff review is limited to verifying whether the applicant has adequately committed to

establish a process for the production, use, and management control of written plant

procedures.

15.5.1.1 Plant Procedures Commitment Description

In revised CAR Section 15.5, the applicant discussed general commitments for plant procedures

applicable to the startup, testing, and operations phases of the proposed facility and stated that

the MPQAP contains the applicant’s description of procedures for the design and construction

phases. The staff reviewed the MPQAP requirements, in particular Section 5.0, “Instructions,

Drawings and Procedures,” and determined that they were acceptable for construction activities,

including design, procurement, and fabrication (NRC, 2001). The applicant will describe its plant

procedures for the facility’s proposed startup, testing, and operation in more detail in any later

license application it may submit. 

Revised CAR Section 15.5 provides additional description of the provisions for procedures and

identifies four types of plant procedures that would be used to control activities during any facility

operations—operating procedures, administrative procedures, maintenance procedures, and

emergency procedures. All such procedures would be prepared, issued, used, and controlled

under the CM system and the MPQAP requirements.

Operating procedures would be used to directly control process operations by workstation and

control room operators. They would include directions for normal operations, including startup

and some testing, operation and shutdown, as well as off-normal conditions, including alarm

response. They would also include operating limits and controls, controls to ensure operational
15-16

safety and hold- or checkpoints. Administrative procedures would be used to perform

management control activities that support operations, including CM, safety, human-system

interface, QA, design control, training and qualification, audits and assessment, incident

investigations, records and document control, and reporting. Maintenance procedures would

address preventive and corrective maintenance, surveillance for calibration, inspection and

testing, and functional testing following maintenance. Emergency procedures would address the

preplanned actions of plant personnel in the event of an emergency.

The applicant described its commitments during any facility operations to review all plant

procedures at least every 5 years to ensure continued accuracy and usefulness. Emergency

procedures would be initially reviewed annually and subsequently every 2 years. Additional

reviews and modifications of procedures would be based on facility operating experience,

incidents, and identified inadequacies.

The applicant will develop procedures for test control for the preoperational testing program.

These procedures will provide testing criteria for determining when tests will be required and

how the activities will be performed. Tests will simulate the most adverse design conditions

feasible, and results will be documented and evaluated and acceptability determined by

responsible personnel. The applicant committed to developing all required plant procedures

before receiving any special nuclear material (SNM) at the site of the proposed facility and to

validating the procedures during startup testing of the proposed facility (CAR Section 15.5.5).

15.5.2 Evaluation Findings

In Chapter 15.5 of the revised CAR, DCS described its management measures for establishing

plant procedures to be used at the proposed facility. Based on that information and the

discussion provided in the sections above for plant procedures, the staff finds that the applicant

is committed to and would be capable of providing the management control necessary to

support operation of the proposed facility. Pursuant to 10 CFR 70.64(a)(1), the staff finds that

the applicant’s proposed plant procedures are a management measure providing adequate

assurance that IROFS will be available and reliable to perform their function when needed. The

applicant will describe its plant procedures for startup, testing, and operation in more detail in its

license application. Accordingly, the staff concludes, pursuant to 10 CFR 70.23(b), that the

management measures for establishing plant procedures set forth in the MPQAP and the

revised CAR are part of a QA program which will provide reasonable assurance of protection

against natural phenomena and the consequences of potential accidents.

15.5.3 References

(ASME, 1994) American Society of Mechanical Engineers. NQA–1–1994, “Quality Assurance

Requirements for Nuclear Facility Applications.” New York, 1994.

(DCS, 2002) Ihde, R., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Mixed Oxide Fuel Fabrication Facility Construction Authorization Request

Revision. October 31, 2002 (including page changes through February 9, 2005).

(NRC, 1978) U.S. Nuclear Regulatory Commission. Regulatory Guide 1.33, Rev. 2, “Quality

Assurance Program Requirements (Operation).” Washington, DC, February 1978.
15-17

(NRC, 1989) U.S. Nuclear Regulatory Commission. “Guidance on Management

Controls/Quality Assurance, Requirements for Operation, Chemical Safety, and Fire Protection

for Fuel Cycle Facilities.” Federal Register: Vol. 54, No. 53, pp. 11590–11598. March 21, 1989.

(NRC, 2001) Persinko, A., U.S. Nuclear Regulatory Commission. Letter to P. Hastings, Duke

Cogema Stone & Webster, RE: Duke Cogema Stone & Webster Quality Assurance Program for

the Construction of the MFFF. October 1, 2001.

(NRC, 2003) Persinko, A., U.S. Nuclear Regulatory Commission. Letter to P. Hastings, Duke

Cogema Stone & Webster, RE: Duke Cogema Stone & Webster Quality Assurance Program for

the Construction of the MFFF. January 10, 2003.

15.6 Audits and Assessments

15.6.1 Conduct of Review

This section of the FSER contains the staff’s review of the audits and assessment information

provided by the applicant in Section 15.6 of the revised CAR (DCS, 2002). The objective of this

review is to determine whether the applicant has developed and adequately described a system

of audits and assessments that provides reasonable assurance that the PSSCs identified by the

applicant at the MFFF construction and design stage, and the IROFS to be identified by the

applicant in its license application, will be available and reliable to perform their safety function

when needed. The staff evaluated the applicant’s system of audits and assessments by

reviewing Section 15.6 of the revised CAR, other sections of the revised CAR, responses to

NRC staff RAIs, and relevant documents available at the applicant’s offices but not submitted by

the applicant.

The following regulations apply specifically to the staff’s review of the proposed audits and

assessments program:

• 10 CFR 70.4 defines the term “management measures” as including an audits and

assessments program.

• 10 CFR 70.64(a)(1) states that the facility design must be developed and implemented in

accordance with management measures, to provide adequate assurance that IROFS will

be available and reliable to perform their function when needed.

Section 15.6.3 of NUREG-1718 defines the scope of the construction approval review pertaining

to the applicant’s planned system of audits and assessments. In reviewing the revised CAR, the

staff has evaluated this system and the applicant’s provisions for continued adherence to the system.

15.6.1.1 Audits and Assessments—General

In revised CAR Section 15.6.1, the applicant discussed general commitments to perform audits

and assessments and stated that they are to be performed in accordance with the applicant’s 

MPQAP requirements for SSCs and associated activities using a graded approach

commensurate with their safety significance. The staff reviewed the commitments for audits and

assessments in MPQAP Section 18.0, “Audits,” and Section 2.4, “Management Assessments,”

and determined that they are acceptable for construction activities (NRC, 2001). The safety
15-18

significance of SSCs and their associated activities will be used to determine the frequency and

rigor by which they are audited and assessed. The audits and assessments are to provide DCS

management with feedback on the technical adequacy of SSCs and activities by evaluating how

well the QA program is being implemented and feedback on the program’s effectiveness in

ensuring that SSCs are properly designed and constructed.

Audits and assessments of SSCs will be scheduled to provide coverage, consistency, and

coordination with ongoing work and at a frequency commensurate with the project status and

importance of the work. All functional areas performing work controlled by the MPQAP will be

audited at least once a year. Results of audits, assessments, surveillances, deficiencies, and

corrective action reports will be used to determine the scope and frequency of functional area

audits. Audits will be scheduled to begin as early in the life of the work as practical and will be

continued at intervals consistent with the work schedule. External audits of PSSC suppliers will

be performed before contract placement, with annual supplier evaluations and full audits

required every 3 years. Annual project assessments to determine the overall effectiveness of

the QA program will be conducted by the project manager, and each functional area performing

work on PSSCs will perform an internal management assessment annually. Additional audits

and assessments of specific functions will be conducted as directed by management to provide

an adequate assessment of compliance and effectiveness. 

The applicant committed to conducting its internal and external audits and assessments using

procedures in accordance with the approved MPQAP requirements (NRC, 2001). These

procedures will include requirements for scheduling and planning, certification of

audit/assessment personnel, development of audit plans and checklists, audit/assessment

performance, reporting and tracking of findings to closure, and closure of the audit/assessment. 

They will emphasize timely reporting and correction of findings to prevent recurrence.

The applicant identified the qualifications and responsibilities of the QA manager, who will have

overall responsibility for managing the QA program including audits and assessments of qualityaffecting activities. The QA verification manager is directly responsible for ensuring that audits

and assessments are conducted in accordance with the MPQAP requirements, including the

lead auditor/auditor certification program, audit program management, reporting findings to

management, evaluating the effectiveness of QA program implementation, approving audit

checklists and reports, maintaining the approved suppliers list, and providing input for

continuous program improvements.

The applicant described the training and qualification requirements and responsibilities for audit

and assessment personnel. These include lead auditor and auditor training and certification and

assessment personnel training appropriate to their activities. Audit and assessment procedures

will require that personnel be independent of the activity being audited or assessed, and that

they have appropriate authority, freedom, and access to make the audit process meaningful and

effective and to properly audit or assess the assigned areas or activities. Checklists will

establish acceptance criteria to determine acceptable performance and audit/assessment team

determinations, results, and reports will be reviewed and approved by appropriate management.

Audits and assessments will be conducted using written procedures/checklists and will include

detailed walkdowns of plant areas, including out-of-the-way and limited-access areas. If findings

result, the deficiencies will be accurately documented for accurate evaluation and timely

corrective actions, including immediate correction, where feasible and appropriate. Audit and
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assessment results will be reviewed by management having responsibility in the area

audited/assessed. Audit and assessment findings and recommendations will be documented

and distributed to appropriate management for review, and response will be required from

responsible managers. Audit and assessment results will be tracked by the applicant’s QA

organization. The data will be analyzed and trended and resultant reports, which indicate quality

trends and the effectiveness of management measures, will go to appropriate management for

review, response, corrective action, and followup.

15.6.1.2 Audits

The applicant described its requirement for audit team personnel, audit conduct, and reporting of

results in revised CAR Section 15.6.2. Audit team personnel will be independent of the areas

and activities being audited and have no direct responsibility for the items they audit. Technical

and programmatic audits will be performed to evaluate internal project activities using applicable

procedures. External suppliers will be evaluated using applicable supplier evaluation

procedures. Audit reports will be issued to appropriate management on a timely basis, followup

reviews will be performed to verify effective completion of corrective actions for audit findings,

and status of open findings will be routinely reported to project management. During

construction, internal audits as-built conditions against controlled drawings, specifications, and

procedures based on committed construction codes and standards.

15.6.1.3 Assessments

The applicant’s description and commitments for assessments include appropriate requirements

as discussed in Section 15.6.1.1, above. In particular, annual project assessments to determine

the overall effectiveness of the QA program will be conducted by the project manager, and each

functional area performing work on PSSCs will perform an internal management assessment

annually. Additional audits and assessments of specific functions will be conducted as directed

by management to provide an adequate assessment of compliance and effectiveness. 

15.6.1.4 Provisions for Continuing Assurance

The applicant’s provisions for adhering to its planned system of audits and assessments are

described in revised CAR Section 15.6.4 and include maintaining the applicant’s QA program

current through deactivation of the proposed facility. Appropriate changes to the QA program

and procedures for audits and assessments will be made as a result of reorganizations, revised

activities, lessons learned, changes to applicable regulations, and program improvements. The

applicant also committed to update the system of audits and assessments to reflect any

changes to the proposed facility made between submittal of the revised CAR and the expected

submittal of its later license application.

15.6.2 Evaluation Findings

In Chapter 15.6 of the revised CAR, DCS described its planned system for conducting audits

and assessments on PSSCs and IROFS at the proposed facility. Based on that information and

the discussion provided in the sections above for audits and assessments, the staff finds that the

applicant has adequately described its system for audits and assessments. Pursuant to 10 CFR

70.64(a)(1), the staff finds that the applicant’s proposed audits and assessments system is a

management measure providing adequate assurance that IROFS will be available and reliable to
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perform their function when needed. Accordingly, the staff concludes, pursuant to 10 CFR

70.23(b), that the system for audits and assessments set forth in the MPQAP and the CAR is

part of a QA program which will provide reasonable assurance of protection against natural

phenomena and the consequences of potential accidents.

15.6.3 References

(DCS, 2002) Ihde, R., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Mixed Oxide Fuel Fabrication Facility Construction Authorization Request

Revision. October 31, 2002 (including page changes through February 9, 2005).

(NRC, 2001) Persinko, A., U.S. Nuclear Regulatory Commission. Letter to P. Hastings, Duke

Cogema Stone & Webster, RE: Duke Cogema Stone & Webster Quality Assurance Program for

the Construction of the MFFF. October 1, 2001.

(NRC, 2003) Persinko, A., U.S. Nuclear Regulatory Commission. Letter to P. Hastings, Duke

Cogema Stone & Webster, RE: Duke Cogema Stone & Webster Quality Assurance Program for

the Construction of the MFFF. January 10, 2003.

15.7 Incident Investigations

15.7.1 Conduct of Review

This section of the FSER contains the staff’s review of the system for incident investigations

referenced by the applicant in Section 15.7.1 of the revised CAR (DCS, 2002). The objective of

this review is to determine whether DCS has developed and adequately described a process for

the systematic investigation of incidents, assignment and acceptance of corrective actions, and

followup to ensure completion of the actions. The staff evaluated the applicant's proposed

system by reviewing Section 15.7 of the revised CAR, MPQAP Section 16, responses to NRC

staff RAIs, and relevant documents available at the applicant's offices but not submitted by the applicant.

The following regulations apply specifically to the staff’s review of the proposed system for

conducting incident investigations:

• 10 CFR 70.4 defines the term “management measures” as including incident

investigations.

• 10 CFR 70.64(a)(1) states that the MFFF design must be developed and implemented in

accordance with management measures, to provide adequate assurance that IROFS will

be available and reliable to perform their function when needed.

With respect to the applicant’s proposed system for conducting incident investigations, Section

15.7.4.3 of NUREG-1718 limits the staff’s CAR review to verifying whether the applicant has

committed to establishing a system to adequately investigate incidents. Such a system includes

provisions for the assignment and acceptance of corrective actions and followup measures to

ensure completion of corrective actions.
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15.7.1.1 Incident Investigation and Corrective Action Process and Administration

In revised CAR Sections 15.7.1 and 15.7.2, the applicant discussed its commitments and

process to perform incident investigation and corrective action and stated that the process

currently in use during design and construction phases of the proposed facility is as described in

the applicant’s MPQAP Section 16, “Corrective Action.” Section 16 of the MPQAP was

previously reviewed and approved by the staff (NRC, 2001). It contains the applicant’s system

for identifying, classifying, following up, closing and trending conditions adverse to quality, and

preparing significant event reports. The staff has verified that this system is adequate for

ensuring that proper corrective action would be taken during any construction activities, including

design, procurement, and fabrication (NRC, 2001). In revised CAR Section 15.7.2, the applicant

committed to modifying this process before any facility startup testing to include the additional

specific actions that would be needed to support an operating facility. 

The applicant’s description in revised CAR Section 15.7.1 of the incident investigation and

corrective action process for design, construction, and operations includes management controls

to promptly identify incidents/findings, evaluate the need to stop work, and assign investigation

teams. It also provides for significance and root cause evaluations and corrective action

planning, management approval, implementation, completion and tracking, as well as tracking

and evaluation for adverse trends.

Corrective action process administration is discussed in revised CAR Section 15.7.2. The

incident investigation and deficiencies corrective action process will be administered by the

applicant’s QA organization during the design and construction phases of the proposed facility. 

The applicant committed to providing in the license application a detailed description of how the

incident investigations process would work during any operation of the proposed facility. Such a

description will be expected to address the prompt investigation of incidents, the use of qualified

investigative teams, monitoring and documenting corrective actions, investigating team plans,

methodologies, personnel qualifications and independence, and appropriate documentation and

records requirements.

15.7.2 Evaluation Findings

In Chapter 15.7 of the revised CAR, DCS described its planned system for performing incident

investigations relevant to PSSCs and associated activities during the construction of the

proposed facility, including design, procurement, and fabrication activities. Based on that

information and the discussion provided in the sections above for audits and assessments, the

staff finds that the applicant has adequately described its system for performing incident

investigations. Pursuant to 10 CFR 70.64(a)(1), the staff finds that the applicant’s proposed

incident investigations system is a management measure providing adequate assurance that

IROFS will be available and reliable to perform their function when needed. Accordingly, the

staff concludes, pursuant to 10 CFR 70.23(b), that the system for incident investigations set forth

in the MPQAP and the revised CAR is part of a QA program which will provide reasonable

assurance of protection against natural phenomena and the consequences of potential

accidents.
15-22

15.7.3 References

(DCS, 2002) Ihde, R., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Mixed Oxide Fuel Fabrication Facility Construction Authorization Request

Revision. October 31, 2002 (including page changes through February 9, 2005).

(DOE, 1992a) U.S. Department of Energy. DOE–STD–1010–92, “Guide to Good Practices for

Incorporating Operating Experiences.” Washington, DC, July 1992. 

(DOE, 1992b) U.S. Department of Energy. DOE–NE–STD–1004–92, "“Root Cause Analysis

Guidance Document.” Washington, DC, February 1992.

(NRC, 1986) U.S. Nuclear Regulatory Commission. NUREG/CR–4616, “Root Causes of

Component Failures Program: Methods and Applications.” Washington, DC, December 1986. 

(NRC, 1991) U.S. Nuclear Regulatory Commission. NUREG/CR–5665, “A Systematic

Approach to Repetitive Failures.” Washington, DC, February 1991.

(NRC, 1996) U.S. Nuclear Regulatory Commission. Information Notice 96–28, “Suggested

Guidance Relating to Development and Implementation of Corrective Action.” Washington, DC,

May 1996.

(NRC, 1999) U.S. Nuclear Regulatory Commission. “Domestic Licensing of Special Nuclear

Material (10 CFR Part 70).” Federal Register: Vol. 64, No. 146, pp. 41338–41357. July 30,

1999.

(NRC, 2001) Persinko, A., U.S. Nuclear Regulatory Commission. Letter to P. Hastings, Duke

Cogema Stone & Webster, RE: Duke Cogema Stone & Webster Quality Assurance Program for

the Construction of the MFFF. October 1, 2001.

(NRC, 2003) Persinko, A., U.S. Nuclear Regulatory Commission. Letter to P. Hastings, Duke

Cogema Stone & Webster, RE: Duke Cogema Stone & Webster Quality Assurance Program for

the Construction of the MFFF. January 10, 2003.

15.8 Records Management

15.8.1 Conduct of Review

This section of the FSER contains the staff’s review of the facility records management system

provided by the applicant in Section 15.8 of the revised CAR (DCS, 2002). The objective of this

review is to verify that the applicant has developed and adequately described a facility records

management system that complies with NRC requirements. The staff evaluated the applicant’s

facility records management system by reviewing Section 15.8 of the revised CAR, other

sections of the revised CAR, responses to NRC staff RAIs and relevant documents available at

the applicant's offices but not submitted by the applicant.

The following regulations apply specifically to the staff’s review of the proposed records

management system:
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• 10 CFR 70.4 defines the term “management measures” as including a records

management system.

• 10 CFR 70.64(a)(1) states that the MFFF design must be developed and implemented in

accordance with management measures, to provide adequate assurance that IROFS 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.

Section 15.8.3 of NUREG-1718 defines the scope of the construction approval review of the

applicant’s facility records management system to include (1) the process whereby records are

specified, created, verified, categorized, indexed, inventoried, protected, stored, maintained,

distributed, and deleted or preserved, (2) the handling and control of various kinds of records

and the methods of recording media that comprise the records, including contaminated and

classified records, and (3) the physical characteristics of the record storage facilities with respect

to the preservation and protection of the records for their designated lifetimes. 

The applicant’s description of its facility records management system, in revised CAR Section

15.8, addresses commitments for the records management program, including record

generation, receipt, storage, preservation, safekeeping, correction, retrieval, and disposition, for

program changes and provisions for continuing adequate records management.

15.8.1.1 Records Management Program

In revised CAR Section 15.8.1, the applicant described its facility records management system,

and stated that the system for controlling records management responsibilities and the

generation, review, approval, classification, verification, indexing, storage, protection,

maintenance, correction, retrieval and disposition of QA records will be in accordance with the

MPQAP requirements. The staff reviewed the MPQAP Section 17.0, “Quality Assurance

Records,” commitments for audits and assessments and determined that they were acceptable

for construction activities at the proposed facility (NRC, 2001). Section 17.0 of the MPQAP

commits the applicant to adhere to the requirements of Criterion 17, “Quality Assurance

Records,” in Appendix B to 10 CFR Part 50 and Basic Requirement 17 and Supplement 17S-1

of NQA-1-1994 Part I, as revised by RG 1.28, Revision 3. The applicant did not request to be

excepted from any of these requirements. The staff reviewed the applicant’s commitments and

the description of the QA program for records in accordance with NUREG-1718 and compared

them to the applicable requirements of Appendix B and the NQA-1 provisions. The staff

reviewed the MPQAP description of the facility QA records management system and verified

that MPQAP Section 17.0 meets the requirements of Appendix B and NQA-1.

The applicant further described the records management system in revised CAR Section 15.8.1,

including applicable project procedures, dual facility storage and fireproof backup tape storage

for electronic data management system, and use of fireproof storage for other documents such

as radiographs and microfilm. The procedures control the generation of records, their review

and approval as records, receipt process, storage, preservation, and safekeeping. Records

requiring correction or revision will be retrieved by authorized individuals in accordance with the

applicable procedures. Original records will be retained and the revision processed per the

applicable project procedures. Records file folders, and interfacing links to associated
15-24

documents/records, will be structured to ensure timely retrievability of records. All “lifetime” QA

records will be stored for the period of time in which the proposed facility may operate.

15.8.1.2 Records Management Program Changes

The applicant’s discussion includes the requirements that routine audits, surveillances, and

assessments of document control and records management will be performed to evaluate the

implementation of the program. Findings and observations from such oversight functions and

other monitoring activities may result in program improvements. 

15.8.1.3 Continuing Records Management Provisions

The applicant’s provisions for ensuring the continuing adequacy of its records management

system include commitments to keep the program procedures current. The records

management procedures will be revised based on lessons learned during implementation;

corrective actions from audits, surveillances, or assessments; improvements based on trend

analysis; and changes that result from regulations, commitments, reorganizations, revised

schedules, or program improvements. The applicant committed to update the facility records

management system to reflect any changes to the proposed facility made between submittal of

the revised CAR and the expected submittal of its later license application.

15.8.2 Evaluation Findings

In Chapter 15.8 of the revised CAR, DCS described its planned records management system to

be used at the proposed facility. Based on that information and the discussion provided in the

sections above for the records management system, the staff finds that the applicant has

adequately described its system for records management. Pursuant to 10 CFR 70.64(a)(1), the

staff finds that the DCS proposed records management system is a management measure

providing adequate assurance that IROFS will be available and reliable to perform their function

when needed, and that appropriate records will be maintained by or under the control of DCS

throughout the operating life of the proposed facility, if the NRC authorizes its operation. 

Accordingly, the staff concludes, pursuant to 10 CFR 70.23(b), that the records management

system set forth in the MPQAP and the revised CAR is part of a QA program which will provide

reasonable assurance of protection against natural phenomena and the consequences of

potential accidents. 

15.8.3 References

(DCS, 2002) Ihde, R., Duke Cogema Stone & Webster. Letter to U.S. Nuclear Regulatory

Commission, RE: Mixed Oxide Fuel Fabrication Facility Construction Authorization Request

Revision. October 31, 2002 (including page changes through February 9, 2005).

(NRC, 2001) Persinko, A., U.S. Nuclear Regulatory Commission. Letter to P. Hastings, Duke

Cogema Stone & Webster, RE: Duke Cogema Stone & Webster Quality Assurance Program for

the Construction of the MFFF. October 1, 2001.

(NRC, 2003) Persinko, A., U.S. Nuclear Regulatory Commission. Letter to P. Hastings, Duke

Cogema Stone & Webster, RE: Duke Cogema Stone & Webster Quality Assurance Program for

the Construction of the MFFF. January 10, 2003.
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.1821

2. TITLE AND SUBTITLE 3. DATE REPORT PUBLISHED

Final Safety Evaluation Report on the Construction Authorization Request for the Mixed Oxide MONTH YEAR

Fuel Fabrication Facility at the Savannah River Site, South Carolina

March 2005

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 NRCM 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 & Safeguards

Office of Nuclear Material Safety & Safeguards

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

9. SPONSORING ORGANIZATION - NAME AND ADDRESS QafNRC, type 'Same as above'; if contractor, provide NRCDfvision, Office or Region, U.S. NudearRegualotoy Commission,

and mailing address)

Same as above

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words or less)

The U.S. Department of Energy (DOE), National Nuclear Security Administration (NNSA), has contracted with Duke Cogema

Stone & Webster (DCS or applicant) to design, construct, and operate a proposed Mixed Oxide (MOX) Fuel Fabrication Facility

(MFFF) that would convert depleted uranium and weapon-grade plutonium into MOX fuel. The proposed MOX facility would be

located on the DOE's Savannah River Site in South Carolina. Use of the proposed facility to produce MOX fuel would be part of

the DOE's'surplus plutonium disposition program. The purpose of the DOE program is to ensure that plutonium produced for

nuclear weapons and declared exceess to national security needs is converted to proliferation-resistant forms.

Under the applicable 10 CFR Part 70 requirements, before a license to possess and use special nuclear material may be

issued for the proposed MFFF, the U.S. Nuclear Regulatory Commission (NRC) must first authorize construction of the

proposed MFFF. On February 28,2001, DCS submitted to the U.S. Nuclear Regulatory Commission (NRC) a Construction

Authorization Request (CAR). This Final Safety Evaluation Report (FSER) documents the NRC staffs review of the CAR, as

amended, and supplemental supporting information provided by the applicant. This FSER only addresses regulatory

requirements for approval of construction, and does not address operational aspects of the proposed facility.

12. KEY WORDSIDESCRIPTORS (List words orphrases that wvl assist researchers in locating the report.) 13. AVAILABILITY STATEMENT

Mixed Oxide Fuel Fabrication Facility unlimited

MOX 14. SECURITY CLASSIFICATION

fuel cycle (This Page)

MOX fuel unclassified

plutonium (This Report)

safety evaluation unclassified

15. NUMBER OF PAGES

16. PRICE

NR FOR .3(-04 PRNE NRCCE AE

NRIC FORM 335 (9-2004) PRINTED ON RECYCLED PAPER
Printed 

on recycled 

paper

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