Spent Fuel Transportation 

Risk Assessment

Final Report

Office of Nuclear Materials Safety and Safeguards

NUREG-2125
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Spent Fuel Transportation 

Risk Assessment 

Final Report

Manuscript Completed: September 2013

Date Published: January 2014

John R. Cook, NRC Project Manager

Office of Nuclear Materials Safety and Safeguards

NUREG-2125

iii NUREG-2125

ABSTRACT

The U.S. Nuclear Regulatory Commission (NRC) is responsible for issuing regulations for the 

packaging of spent fuel (and other large quantities of radioactive material) for transport that 

provide for public health and safety during transport (Title 10 of the Code of Federal Regulations

(10 CFR) Part 71, “Packaging and Transportation of Radioactive Waste,” dated 

January 26, 2004). In September 1977, the NRC published NUREG-0170, “Final Environmental 

Statement on the Transportation of Radioactive Material by Air and Other Modes,” which 

assessed the adequacy of those regulations to provide safety assurance. In that assessment, 

the measure of safety was the risk of radiation doses to the public under routine and accident 

transport conditions, and the risk was found to be acceptable. Since that time, there have been 

two affirmations of this conclusion for spent nuclear fuel (SNF) transportation, each using 

improved tools and information. This report presents the results of a fourth investigation into the 

safety of SNF transportation. The risks associated with SNF transportation come from the 

radiation that the spent fuel emits, which is attenuated—but not eliminated—by the 

transportation casks shielding and the possibility of the release of some quantity of radioactive 

material during a severe accident. This investigation shows that the risk from the radiation 

emitted from the casks is a small fraction of naturally occurring background radiation and the 

risk from accidental release of radioactive material is several orders of magnitude less. 

Because there have been only minor changes to the radioactive material transportation 

regulations between NUREG-0170 and this risk assessment, the calculated dose caused by the 

external radiation from the cask under routine transport conditions is similar to what was found 

in earlier studies. The improved analysis tools and techniques, improved data availability, and a 

reduction in uncertainty has made the estimate of accident risk from the release of radioactive 

material in this study approximately five orders of magnitude less than what was estimated in 

NUREG-0170. The results demonstrate that NRC regulations continue to provide adequate 

protection of public health and safety during the transportation of SNF.

v NUREG-2125

CONTENTS

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

LIST OF FIGURES ................................................................................................................... ix

LIST OF TABLES.................................................................................................................. xvii

EXECUTIVE SUMMARY .......................................................................................................xxiii

PUBLIC SUMMARY .............................................................................................................xxvii

ACKNOWLEDGEMENTS....................................................................................................xxxix

ACRONYMS AND ABBREVIATIONS..................................................................................... xli

1. INTRODUCTION................................................................................................................... 1

1.1 Organization of this Report ............................................................................................. 1

1.1.1 Chapter 1 and Appendix A.................................................................................... 1

1.1.2 Chapter 2 and Appendix B.................................................................................... 1

1.1.3 Chapter 3 and Appendix C.................................................................................... 1

1.1.4 Chapter 4 and Appendix D.................................................................................... 1

1.1.5 Chapter 5 and Appendix E.................................................................................... 2

1.1.6 Chapter 6 and Public Summary ............................................................................ 2

1.1.7 Bibliography .......................................................................................................... 2

1.1.8 Review Process .................................................................................................... 2

1.2 Historical Transportation Risk Studies and the Purpose of this Analysis ......................... 2

1.3 Risk ................................................................................................................................ 5

1.3.1 Accident Data ....................................................................................................... 5

1.3.2 Spent Nuclear Fuel Transportation Scenarios....................................................... 6

1.4 Regulation of Radioactive Materials Transportation........................................................ 7

1.5 Selection of Casks........................................................................................................ 10

2. RISK ANALYSIS OF ROUTINE TRANSPORTATION........................................................ 15

2.1 Introduction................................................................................................................... 15

2.2 Radiation Emitted during Routine Transportation.......................................................... 16

2.3 The RADTRAN Model of Routine, Incident-Free Transportation................................... 18

2.3.1 The Basic RADTRAN Model ............................................................................... 18

2.3.2 Individual and Collective Doses .......................................................................... 18

2.3.3 Doses to Members of the Public Occupying Vehicles that Share the Route........ 30

2.3.4 Doses at Truck and Train Stops.......................................................................... 33

2.4 Doses to Workers ......................................................................................................... 39

2.5 Chapter Summary......................................................................................................... 40

3. CASK RESPONSE TO IMPACT ACCIDENTS ................................................................... 45

3.1 Introduction................................................................................................................... 45

3.2 Finite Element Analyses of Casks................................................................................. 45

3.2.1 Rail-Steel Cask ................................................................................................... 46

3.2.2 Rail-Lead Cask ................................................................................................... 50

3.2.3 Truck-DU Cask ................................................................................................... 58

3.3 Impacts onto Yielding Targets ...................................................................................... 62

3.4 Effect of Impact Angle................................................................................................... 64

3.5 Impacts with Objects..................................................................................................... 66

3.6 Response of Spent Fuel Assemblies ............................................................................ 69

3.7 Chapter Summary......................................................................................................... 70
NUREG-2125 vi

4. CASK RESPONSE TO FIRE ACCIDENTS......................................................................... 73

4.1 Introduction................................................................................................................... 73

4.2 Description of Accident Scenarios ................................................................................ 74

4.2.1 Pool size ............................................................................................................. 74

4.2.2 Fire Duration ....................................................................................................... 74

4.2.3 Hypothetical Accident Configurations for the Rail Casks ..................................... 75

4.2.4 Hypothetical Accident Configuration for the Truck Cask...................................... 78

4.3 Analysis of Fire Scenarios Involving Rail Casks............................................................ 78

4.3.1 Simulations of the Fires....................................................................................... 79

4.3.2 Simulations of the Rail Casks.............................................................................. 79

4.3.3 Simulation of the Spent Nuclear Fuel Region...................................................... 81

4.3.4 Rail-Steel Cask Results ...................................................................................... 81

4.3.5 Rail-Lead Cask Results....................................................................................... 94

4.4 Truck Cask Analysis ................................................................................................... 104

4.4.1 Simulation of the Truck Cask ............................................................................ 104

4.4.2 Simulation of the Spent Nuclear Fuel Region.................................................... 105

4.4.3 Truck-DU Cask Results..................................................................................... 105

4.5 Chapter Summary....................................................................................................... 109

5. TRANSPORTATION ACCIDENTS ................................................................................... 111

5.1 Types of Accidents and Incidents ............................................................................... 111

5.2 Accident Probabilities ................................................................................................. 112

5.3 Accidents with Neither Loss of Lead Shielding nor Release of Radioactive Material... 114

5.4 Accidental Loss of Shielding ....................................................................................... 117

5.4.1 Loss of Lead Gamma Shielding ........................................................................ 117

5.4.2 Loss of Neutron Shielding ................................................................................. 121

5.5 Accidental Release of Radioactive Materials .............................................................. 122

5.5.1 Spent Fuel Inventory......................................................................................... 122

5.5.2 Conditional Probabilities and Release Fractions ............................................... 124

5.5.3 Dispersion......................................................................................................... 124

5.5.4 Consequences and Risks from Accidents Involving Release of Radioactive 

Material............................................................................................................. 126

5.6 Potential Criticality ...................................................................................................... 129

5.7 Chapter Summary....................................................................................................... 130

6. FINDINGS AND CONCLUSIONS ..................................................................................... 131

6.1 Routine Transportation ............................................................................................... 131

6.2 Transportation Accidents ............................................................................................ 135

6.3 Effect of Transportation of Higher Burnup Spent Nuclear Fuel.................................... 138

6.4 Findings and Conclusion............................................................................................. 139

APPENDIX A CASK DETAILS AND CERTIFICATES OF COMPLIANCE..........................A-1

A.1 Cask Descriptions ..................................................................................A-1

A.1.1 Truck Casks.................................................................................A-1

A.1.2 Rail Casks....................................................................................A-2

A.2 Certificates of Compliance .....................................................................A-4

APPENDIX B DETAILS OF RISK ANALYSIS OF ROUTINE, INCIDENT-FREE 

TRANSPORTATION.....................................................................................B-1

B.1 Introduction ............................................................................................B-1

B.2 The RADTRAN Model of Routine Transportation .................................B-1

B.2.1 Description of the RADTRAN Program ........................................B-1

B.2.2 The RADTRAN Software ............................................................B-5
vii NUREG-2125

B.3 RADTRAN Input Parameters .................................................................B-6

B.3.1 Vehicle-Specific Input Parameters................................................B-6

B.3.2 Route-Specific Input Parameters..................................................B-7

B.3.3 Other Parameters ........................................................................B-8

B.3.4 RADTRAN Input and Output Files................................................B-9

B.4 Routes ...................................................................................................B-12

B.5 Results...................................................................................................B-19

B.5.1 Maximally Exposed Resident In-Transit Dose ..............................B-19

B.5.2 Unit Risk—Rail Routes.................................................................B-20

B.5.3 Unit Risk—Truck Routes..............................................................B-21

B.5.4 Doses along Selected Routes......................................................B-24

B.6 Interpretation of Collective Dose ............................................................B-53

APPENDIX C DETAILS OF CASK RESPONSE TO IMPACT ACCIDENTS .......................C-1

C.1 Introduction ............................................................................................C-1

C.2 Finite Element Analysis of the Rail-Steel Cask .......................................C-1

C.2.1 Problem Statement ......................................................................C-1

C.2.2 Geometric Assumptions and Mesh ..............................................C-1

C.2.3 Material Properties.......................................................................C-6

C.2.4 Criteria for Element Death and Bolt Failure..................................C-10

C.2.5 Analysis Results ..........................................................................C-10

C.3 Finite Element Analysis of the Rail-Lead Cask .......................................C-21

C.3.1 Problem Statement ......................................................................C-21

C.3.2 Geometric Assumptions and Mesh ..............................................C-22

C.3.3 Material Properties.......................................................................C-26

C.3.4 Criteria for Element Death and Bolt Failure..................................C-31

C.3.5 Analysis Results ..........................................................................C-32

C.3.6 Determination of Lid Gaps ...........................................................C-40

C.4 Impacts against Yielding Targets ...........................................................C-54

C.4.1 Introduction..................................................................................C-54

C.4.2 Method.........................................................................................C-55

C.4.3 Soil Targets..................................................................................C-57

C.4.4 Concrete Targets .........................................................................C-58

C.4.5 Hard Rock Targets.......................................................................C-60

C.4.6 Results for Real Target Calculations ............................................C-61

C.4.7 Impacts against Water .................................................................C-61

C.5 Response of Spent Fuel Assemblies......................................................C-61

C.5.1 Introduction..................................................................................C-61

C.5.2 Description and Method ...............................................................C-62

C.5.3 Finite Element Models .................................................................C-64

C.5.4 Discussion and Conclusions ........................................................C-76

APPENDIX D DETAILS OF CASK RESPONSE TO FIRE ACCIDENTS.............................D-1

D.1 Introduction ............................................................................................D-1

D.2 Finite Volume Domain and Boundary Conditions for the Container 

Analysis Fire Environment (CAFE).........................................................D-2

D.3 The Rail-Steel Cask ............................................................................D-4

D.3.1 Geometric Consideration .............................................................D-5
NUREG-2125 viii

D.3.2 Rail-Steel Cask Thermal Behavior and Model Assumptions.D-13

D.3.3 Rail-Steel Cask Materials and Thermal Properties ................D-15

D.3.4 Rail-Steel Cask, Finite Element Model, and Boundary 

Conditions ..................................................................................D-23

D.3.5 Rail-Steel Cask Thermal Analysis Results..............................D-25

D.4 Rail Cask with Lead Shielding................................................................D-36

D.4.1 Geometry Considerations ............................................................D-36

D.4.2 Rail-Lead Cask Thermal Behavior and Model Assumptions .D-39

D.4.3 Rail-Lead Cask Materials and Thermal Properties.................D-42

D.4.4 Rail-Lead Cask Finite Element Model .....................................D-47

D.4.5 Rail-Lead Cask Thermal Analysis Results ..............................D-48

D.5 Truck Cask with Depleted Uranium ........................................................D-59

D.5.1 Geometric Considerations............................................................D-59

D.5.2 Truck-DU Thermal Behavior and Model Assumptions ...........D-61

D.5.3 Truck-DU Materials and Thermal Properties...........................D-62

D.5.4 Truck-DU P/Thermal Finite Element Model.............................D-63

D.5.5 Truck-DU Cask Thermal Analysis Results ..............................D-64

D.6 Fire Environment Benchmarks for Container Analysis............................D-67

D.6.1 Large Calorimeter Test and Benchmark Results..........................D-67

D.6.2 Small Calorimeter Test and Benchmark Results ..........................D-70

D.6.3 Summary of Benchmark Results..................................................D-71

D.7 Summary ...............................................................................................D-71

APPENDIX E DETAILS OF TRANSPORTATION ACCIDENTS .........................................E-1

E.1 Types of Accidents and Incidents...........................................................E-1

E.2 Accident Probabilities.............................................................................E-2

E.2.1 Historic Accident Frequencies......................................................E-2

E.2.2 Development of Conditional Accident Probabilities ......................E-2

E.3 Accident Risks and Consequences ........................................................E-6

E.3.1 Loss of Lead Gamma Shielding ...................................................E-6

E.3.2 Loss of Neutron Shielding............................................................E-17

E.4 Release of Radioactive Materials in Accidents .......................................E-20

E.4.1 Spent Fuel Inventory....................................................................E-20

E.4.2 Dispersion of Released Radionuclides.........................................E-21

E.4.3 Release Fractions........................................................................E-23

E.5 Summary ...............................................................................................E-28

BIBLIOGRAPHY ...............................................................................................Bibliography-1
ix NUREG-2125

LIST OF FIGURES

Figure PS-1 Model of a spent fuel cask in routine, incident-free transportation and 

radiation dose to a member of the public............................................................ xxx

Figure PS-2 Collective doses from background and from a truck shipment of spent 

nuclear fuel ....................................................................................................... xxxi

Figure PS-3 Collective doses (person-Sv) from routine truck transportation..........................xxxii

Figure PS-4 Corner impact onto a rigid target at a 97-kph (60-mph) accident scenario for 

a spent fuel cask and the deformations produced by the impact .......................xxxii

Figure PS-5 Engulfing fire scenario and the temperature contours in the rail cask 

following a 3-hour fire duration .........................................................................xxxiii

Figure PS-6 Average collective dose risk from release and loss-of-shielding (LOS) 

accidents..........................................................................................................xxxiv

Figure PS-7 Average collective dose risk from accidents that have no impact on the 

cargo................................................................................................................ xxxv

Figure PS-8 Calculated results for an illustrative rail shipment using the rail-lead cask ........xxxvi

Figure 1-1 The four tests for Type B packages....................................................................... 9

Figure 1-2 Photograph and cross-section of the NAC-STC cask ...........................................12

Figure 1-3 Basic layout and cross-section of the HI-STAR 100 rail transport cask ................13

Figure 1-4 GA-4 cask ............................................................................................................14

Figure 2-1 The upper figure is an exploded view of a generic spent fuel cask .......................17

Figure 2-2 RADTRAN model of the vehicle in routine, incident-free transportation................18

Figure 2-3 Diagram of a truck route as modeled in RADTRAN..............................................20

Figure 2-4 A segment of I-80 through Salt Lake City, UT.......................................................22

Figure 2-5 Highway and rail routes from Maine Yankee Nuclear Plant site............................23

Figure 2-6 Highway and rail routes from Kewaunee Nuclear Plant ........................................24

Figure 2-7 Highway and rail routes from Indian Point Nuclear Plant ......................................25

Figure 2-8 Highway and rail routes from Idaho National Laboratory ......................................26

Figure 2-9 Diagram used in RADTRAN for calculating radiation doses to occupants of 

other vehicles.......................................................................................................32

Figure 2-10 Diagram of truck stop model.................................................................................35

Figure 2-11 Collective doses from background and from Maine Yankee to ORNL truck 

shipments of spent nuclear fuel............................................................................44

Figure 3-1 Impact orientations analyzed................................................................................46

Figure 3-2 Finite element mesh of the Rail-Steel cask...........................................................47

Figure 3-3 Details of the finite element mesh for the impact limiters of the Rail-Steel 

cask .....................................................................................................................48

Figure 3-4 Finite element mesh of the Rail-Steel closure bolts and the closure end 

impact limiter attachment bolts .............................................................................49

Figure 3-5 Plastic strain in the welded canister of the Rail-Steel for the 193-kph

(120-mph) side impact case .................................................................................50

Figure 3-6 Finite element mesh of the Rail-Lead cask...........................................................51

Figure 3-7 Details of the finite element mesh for the impact limiters of the Rail-Lead 

cask .....................................................................................................................52
NUREG-2125 x

Figure 3-8 Finite element mesh of the Rail-Lead closure bolts for both the inner and 

outer lids ..............................................................................................................53

Figure 3-9 Deformed shape of the Rail-Lead cask following the 193-kph (120-mph) 

impact onto an unyielding target in the end-on orientation....................................54

Figure 3-10 Deformed shape of the Rail-Lead cask following the 193-kph (120-mph) 

impact onto an unyielding target in the corner orientation.....................................55

Figure 3-11 Deformed shape of the Rail-Lead cask following the 145-kph (90-mph) 

impact onto an unyielding target in the side orientation ........................................56

Figure 3-12 Measurement of closure gaps ..............................................................................57

Figure 3-13 Deformed shapes and plastic strains in the generic steel-DU-steel truck cask 

from NUREG/CR-6672 (impact limiter removed) following 193-kph 

(120-mph) impacts in the (clockwise from top left) end-on, center of gravity 

(CG)-over-corner, and side-on orientation............................................................61

Figure 3-14 Force generated by the Rail-Lead cask penetrating hard desert soil ....................63

Figure 3-15 Probability distribution for impact angles ..............................................................65

Figure 3-16 Influence of impact angle on effective velocity......................................................65

Figure 3-17 Deformations to the GA-4 truck cask after a 97-kph (60-mph) side impact 

onto a rigid semi-circular column..........................................................................66

Figure 3-18 Configuration of locomotive impact analysis.........................................................67

Figure 3-19 Sequential views of a 129-kph (80-mph) impact of a locomotive into a GA-4 

truck cask.............................................................................................................68

Figure 3-20 Results of a finite element simulation of an elevated freeway collapse onto a 

GA-4 spent fuel cask............................................................................................68

Figure 3-21 Finite element model of a PWR fuel assembly .....................................................69

Figure 3-22 Detailed finite element model of a single fuel rod .................................................70

Figure 3-23 Maximum strains in the rod with the highest loads ...............................................70

Figure 4-1 Cask lying on ground concentric with fuel pool.....................................................75

Figure 4-2 Cask lying on ground 3 meters (10 feet) from pool fire .........................................76

Figure 4-3 Cask lying on ground 18 meters (60 feet) from pool fire .......................................77

Figure 4-4 Regulatory pool fire configuration.........................................................................77

Figure 4-5 Truck-DU cask lying on ground concentric with fuel pool......................................78

Figure 4-6 Finite element models (cut views) of the two rail casks analyzed .........................80

Figure 4-7 Temperature distribution of the Rail-Steel cask at the end of the 30-minute 

800°C (1472°F) regulatory uniform heating ..........................................................81

Figure 4-8 Temperature of key cask regions, Rail-Steel cask undergoing regulatory 

uniform heating ....................................................................................................82

Figure 4-9 Temperature distribution of the Rail-Steel cask at the end of the 30-minute 

regulatory CAFE fire.............................................................................................83

Figure 4-10 Temperature of key cask regions, Rail-Steel cask undergoing regulatory 

CAFE fire .............................................................................................................84

Figure 4-11 Comparison of regulatory fire analysis for Rail-Steel cask: Uniform heating 

vs. CAFE fire........................................................................................................85

Figure 4-12 Gas temperature plots from the regulatory CAFE fire analysis .............................86

Figure 4-13 Fuel concentration plots from the regulatory CAFE fire analysis...........................87

Figure 4-14 Temperature distribution of the Rail-Steel cask at the end of the 3-hour 

concentric CAFE fire with cask on ground............................................................88
xi NUREG-2125

Figure 4-15 Temperature of key cask regions, Rail-Steel cask with cask on ground, 

concentric fire.......................................................................................................89

Figure 4-16 Gas temperature plots from the CAFE fire analysis of the cask on ground...........90

Figure 4-17 Fuel concentration plots from the CAFE fire analysis of the cask on ground ........91

Figure 4-18 Temperature distribution of the Rail-Steel cask at the end of the 3-hour, 

3-meter (10-foot) offset CAFE fire with cask on ground........................................92

Figure 4-19 Temperature of key cask regions, Rail-Steel cask with cask on ground, 

3-meter (10-foot) offset fire...................................................................................92

Figure 4-20 Temperature distribution of the Rail-Steel cask at the end of the 3-hour 

18-meter (60-foot) offset CAFE fire with cask on ground ......................................93

Figure 4-21 Temperature of key cask regions, Rail-Steel cask with cask on ground, 

18-meter (60-foot) offset fire.................................................................................93

Figure 4-22 Temperature distribution of the Rail-Lead cask at the end of the 30-minute 

800°C (1472°F) regulatory uniform heating ..........................................................94

Figure 4-23 Temperature of key cask regions, Rail-Lead cask undergoing regulatory 

uniform heating ....................................................................................................95

Figure 4-24 Temperature distribution of the Rail-Lead cask at the end of the 30-minute 

regulatory CAFE fire.............................................................................................96

Figure 4-25 Temperature of key cask regions, Rail-Lead cask in regulatory CAFE fire ...........97

Figure 4-26 Comparison of regulatory fire analysis for Rail-Lead cask: Uniform heating 

vs. CAFE fire........................................................................................................98

Figure 4-27 Temperature distribution of the Rail-Lead cask at the end of the 3-hour 

concentric CAFE fire with cask on ground............................................................99

Figure 4-28 Temperature of key cask regions, Rail-Lead cask with cask on ground, 

concentric fire.....................................................................................................100

Figure 4-29 Temperature distribution of the Rail-Lead cask at the end of the 3-hour 

3-meter (10-foot) offset CAFE fire with cask on ground......................................100

Figure 4-30 Temperature of key cask regions, Rail-Lead cask with Cask on ground, 

3-meter (10-foot) offset fire.................................................................................101

Figure 4-31 Temperature distribution of the Rail-Lead cask at the end of the 3-hour 

18-meter (60-foot) offset CAFE fire with cask on ground ....................................101

Figure 4-32 Temperature of key cask regions, Rail-Lead cask with cask on ground, 

18-meter (60-foot) offset fire...............................................................................102

Figure 4-33 Rail-Lead cask lead gamma shield region – maximum lead melt at the 

middle of the cask – Scenario: Cask on ground, 3-hour concentric pool fire......103

Figure 4-34 Rail-Lead cask lead gamma shield region – maximum lead melt at the 

middle of the cask – Scenario: Cask lying on ground, 3-hour 3-meter 

(10-foot) offset pool fire ......................................................................................103

Figure 4-35 Finite element model (cut view) of the Truck-DU cask........................................104

Figure 4-36 Temperature distribution of the Truck-DU cask at the end of the 1-hour 

concentric CAFE fire with cask on ground..........................................................105

Figure 4-37 Temperature of key cask regions, Truck-DU cask with cask on ground, 

concentric fire.....................................................................................................106

Figure 4-38 Gas temperature plots........................................................................................107

Figure 4-39 Fuel concentration plots .....................................................................................108

Figure 4-40 Peak temperature in the DU shielding during the thermal event .........................109
NUREG-2125 xii

Figure 5-1 Accident frequencies in the United States from 1991 until 2007 .........................113

Figure 5-2 Radiation dose rates to the maximally exposed individual (MEI) from loss of 

lead gamma shielding at distances from 1 to 5 meters from the cask carrying 

spent fuel ...........................................................................................................118

Figure 5-3 Radiation dose rates to the MEI from loss of lead gamma shielding at 

distances from 10 to 100 meters from the cask carrying spent fuel ....................119

Figure 5-4 Air and ground concentrations of radioactive material following a release..........125

Figure 6-1 Collective doses (person-Sv) from routine truck transportation...........................132

Figure 6-2 Collective doses (person-Sv) from routine rail transportation..............................133

Figure 6-3 Maximum individual dose (Sv) from routine transportation .................................135

Figure 6-4 Average collective dose risk from release and loss of lead shielding (LOS) 

accidents............................................................................................................136

Figure 6-5 Average collective dose risk from accidents that have no impact on the 

cargo..................................................................................................................138

Figure B-1 RADTRAN model of the vehicle in routine, incident-free transportation............B-2

Figure B-2 Diagram of a truck route as modeled in RADTRAN..........................................B-4

Figure B-3 Diagram of a truck-stop model .........................................................................B-5

Figure B-4 RadCat vehicle screen .....................................................................................B-6

Figure B-5 RADTRAN unit risk input file for the Truck-DU cask .........................................B-10

Figure B-6 RADTRAN unit risk input file for the Rail-Lead and Rail-Steel casks ................B-11

Figure B-7 Highway and rail routes from the Maine Yankee Nuclear Plant (NP) site..........B-16

Figure B-8 Highway and rail routes from the Kewaunee NP site........................................B-17

Figure B-9 Highway and rail routes from Indian Point NP site............................................B-18

Figure B-10 Highway and rail routes from Idaho National Laboratory ..................................B-19

Figure B-11 RADTRAN output for maximum individual truck doses.....................................B-20

Figure B-12 RADTRAN output for maximum individual rail doses........................................B-20

Figure B-13 RADTRAN output for Table B-8........................................................................B-21

Figure B-14 Parameters for calculating doses to occupants of highway vehicles sharing 

the route with the radioactive shipment.....................................................................B-41

Figure C-1 Half-symmetric mesh of Rail-Steel cask.......................................................C-2

Figure C-2 Impact-limiter mesh ...................................................................................C-3

Figure C-3 Impact-limiter mesh with honeycomb removed, showing the internal support 

structure ...................................................................................................C-3

Figure C-4 Mesh of lid closure bolts and impact-limiter attachment bolts .........................C-4

Figure C-5 Locations and magnitudes of internal gaps in the model................................C-5

Figure C-6 Orientation of cask model for material property definitions.............................C-5

Figure C-7 Rail-Steel cask end impact at 193 kph........................................................ C-11

Figure C-8 Rail-Steel cask corner impact at 48 kph...................................................... C-12

Figure C-9 Rail-Steel cask corner impact at 97 kph...................................................... C-13

Figure C-10 Rail-Steel cask corner impact at 145 kph .................................................... C-14

Figure C-11 Rail-Steel cask side impact at 193 kph ....................................................... C-14

Figure C-12 Plastic strain in the interior welded canister of the Rail-Steel cask from the 

end impact at 48 kph ................................................................................ C-15
xiii NUREG-2125

Figure C-13 Plastic strain in the interior welded canister of the Rail-Steel cask from the 

end impact at 97 kph ................................................................................ C-16

Figure C-14 Plastic strain in the interior welded canister of the Rail-Steel cask from the 

end impact at 145 kph............................................................................... C-16

Figure C-15 Plastic strain in the interior welded canister of the Rail-Steel cask from the

end impact at 193 kph............................................................................... C-17

Figure C-16 Plastic strain in the interior welded canister of the Rail-Steel cask from the 

corner impact at 48 kph............................................................................. C-17

Figure C-17 Plastic strain in the interior welded canister of the Rail-Steel cask from the 

corner impact at 97 kph............................................................................. C-18

Figure C-18 Plastic strain in the interior welded canister of the Rail-Steel cask from the 

corner impact at 145 kph........................................................................... C-18

Figure C-19 Plastic strain in the interior welded canister of the Rail-Steel cask from the 

corner impact at 193 kph........................................................................... C-19

Figure C-20 Plastic strain in the interior welded canister of the Rail-Steel cask from the 

side impact at 48 kph................................................................................ C-19

Figure C-21 Plastic strain in the interior welded canister of the Rail-Steel cask from the 

side impact at 97 kph................................................................................ C-20

Figure C-22 Plastic strain in the interior welded canister of the Rail-Steel cask from the

side impact at 145 kph .............................................................................. C-20

Figure C-23 Plastic strain in the interior welded canister of the Rail-Steel cask from the 

side impact at 193 kph .............................................................................. C-20

Figure C-24 Half-symmetric mesh of Rail-Lead cask...................................................... C-23

Figure C-25 Impact-limiter mesh .................................................................................. C-24

Figure C-26 Impact-limiter mesh with wood removed ..................................................... C-24

Figure C-27 Mesh of inner and outer lid closure bolts..................................................... C-25

Figure C-28 Orientation of cask model for material property definitions............................ C-26

Figure C-29 Rail-Lead cask end impact at 48 kph.......................................................... C-32

Figure C-30 Rail-Lead cask end impact at 97 kph.......................................................... C-33

Figure C-31 Rail-Lead cask end impact at 145 kph ........................................................ C-34

Figure C-32 Rail-Lead cask end impact at 193 kph ........................................................ C-35

Figure C-33 Rail-Lead cask corner impact at 48 kph ...................................................... C-35

Figure C-34 Rail-Lead cask corner impact at 97 kph ...................................................... C-36

Figure C-35 Rail-Lead cask corner impact at 145 kph .................................................... C-36

Figure C-36 Rail-Lead cask corner impact at 193 kph .................................................... C-37

Figure C-37 Rail-Lead cask side impact at 48 kph ......................................................... C-37

Figure C-38 Rail-Lead cask side impact at 97 kph ......................................................... C-38

Figure C-39 Rail-Lead cask side impact at 145 kph ....................................................... C-38

Figure C-40 Rail-Lead cask side impact at 193 kph ....................................................... C-39

Figure C-41 Gap opening locations for end impact orientation ........................................ C-41

Figure C-42 Gap opening locations for corner impact orientation..................................... C-41

Figure C-43 Gap opening locations for side impact orientation........................................ C-42
NUREG-2125 xiv

Figure C-44 Gaps in the inner and outer lids of the Rail-Lead cask from the end impact at 

48 kph..................................................................................................... C-43

Figure C-45 Gaps in the inner and outer lids of the Rail-Lead cask from the end impact at 

97 kph..................................................................................................... C-44

Figure C-46 Gaps in the inner and outer lids of the Rail-Lead cask from the end impact at 

145 kph ................................................................................................... C-45

Figure C-47 Gaps in the inner and outer lids of the Rail-Lead cask from the end impact at 

193 kph ................................................................................................... C-46

Figure C-48 Gaps in the inner and outer lids of the Rail-Lead cask from the corner impact 

at 48 kph ................................................................................................. C-47

Figure C-49 Gaps in the inner and outer lids of the Rail-Lead cask from the corner impact 

at 97 kph ................................................................................................. C-48

Figure C-50 Gaps in the inner and outer lids of the Rail-Lead cask from the corner impact 

at 145 kph................................................................................................ C-49

Figure C-51 Gaps in the inner and outer lids of the Rail-Lead cask from the corner impact 

at 193 kph................................................................................................ C-50

Figure C-52 Gaps in the inner and outer lids of the Rail-Lead cask from the side impact at 

48 kph..................................................................................................... C-51

Figure C-53 Gaps in the inner and outer lids of the Rail-Lead cask from the side impact at 

97 kph..................................................................................................... C-52

Figure C-54 Kinetic energy time-histories for the Rail-Lead cask from 193-kph (120-mph) 

impact analyses in the end, side, and corner orientations............................. C-56

Figure C-55 Force generated by the Rail-Lead cask penetrating hard desert soil.............. C-58

Figure C-56 Comparison of test force-deflection curves with those derived from the 

empirical equations................................................................................... C-60

Figure C-57 PWR fuel assembly .................................................................................. C-62

Figure C-58 Fuel rod schematic drawing....................................................................... C-63

Figure C-59 Beam fuel assembly finite element model ................................................... C-65

Figure C-60 Cross-section of a 17 × 17 fuel assembly (with guide tubes in blue) .............. C-65

Figure C-61 Acceleration curves applied to fuel assembly beam model ........................... C-66

Figure C-62 Plastic strain in a spacer grid from a 100 g impact................................. C-67

Figure C-63 Hexahedron test model for solid rod-to-rod contact in Abaqus/Explicit ........... C-68

Figure C-64 Test model for beam-to-beam contact in Abaqus/Explicit ............................. C-68

Figure C-65 Comparison of contact forces between solid rod and beam element rod........ C-69

Figure C-66 Comparison of contact forces as a function of impact velocity....................... C-69

Figure C-67 Continuum rod finite element model ........................................................... C-70

Figure C-68 Kinetic energy for Analysis Case 1............................................................. C-72

Figure C-69 Maximum equivalent plastic strain versus time for four inter-spacer-grid 

locations.................................................................................................. C-72

Figure C-70 Maximum equivalent plastic strain field in cladding for Analysis Case 1 ......... C-73

Figure C-71 Maximum equivalent plastic strain versus time for three spacer grid locations C-73

Figure C-72 Schematic showing maximum equivalent plastic strain for spacer grid and 

inter-spacer-grid locations ......................................................................... C-74
xv NUREG-2125

Figure C-73 Maximum equivalent plastic strain field in cladding for Analysis Case 2 ......... C-75

Figure C-74 Maximum equivalent plastic strain versus time for four inter-spacer-grid 

locations.................................................................................................. C-75

Figure C-75 Maximum equivalent plastic strain versus time for four spacer grid locations.. C-76

Figure C-76 Schematic showing maximum equivalent plastic strain for spacer grid and 

inter-spacer-grid locations, Analysis Case 2................................................ C-76

Figure D-1 CAFE three-dimensional domain: (a) CAFE regulatory fire, (b) cask on 

ground and at the center of the pool, (c) cask on the ground and 3 m 

(10 feet) from the edge of the pool, and (d) cask on the ground and 18.3 m 

(60 feet) from the edge of the pool ..............................................................D-4

Figure D-2 Rail-Steel cask transportation system..........................................................D-5

Figure D-3 Rail-Steel cask: (a) assembly of MPC and overpack and (b) cask with 

limiters .....................................................................................................D-6

Figure D-4 Rail-Steel cask overpack: (a) cross-sectional view through the center of the 

cask and (b) cross-sectional view through the midplane of the overpack.........D-7

Figure D-5 Rail-Steel cask MPC: (a) cross-sectional view through the axis of the cask 

and (b) cross-sectional view through the midplane of the overpack ................D-8

Figure D-6 (a) Fuel assembly (from Holtec International, 2000) and (b) fuel rod.............. D-10

Figure D-7 Fuel-basket region (left) and equivalent fuel-basket region (right).................. D-11

Figure D-8 (a) Upper and (b) lower impact limiters in a Rail-Steel cask .......................... D-12

Figure D-9 Rail-Steel cask finite element mesh............................................................ D-24

Figure D-10 Rail-Steel cask regulatory uniform-heating results (P/Thermal) ..................... D-26

Figure D-11 Rail-Steel cask CAFE regulatory fire .......................................................... D-28

Figure D-12 Rail-Steel cask CAFE fire with cask on ground and at the pool center ........... D-30

Figure D-13 Rail-Steel cask CAFE fire with cask on ground 3 m (10 feet) from the edge of 

the pool ................................................................................................... D-32

Figure D-14 Rail-Steel cask CAFE fire with cask on ground 18.3 m (60 feet) from the 

edge of the pool ....................................................................................... D-34

Figure D-15 Rail-Lead cask components with the directly loaded fuel basket shown to the 

right ........................................................................................................ D-37

Figure D-16 Cross-section view of the Rail-Lead cask with the directly loaded fuel basket. D-38

Figure D-17 Axial burnup profile for the directly loaded fuel basket.................................. D-40

Figure D-18 Three-dimensional quarter section of the directly loaded basket ................... D-41

Figure D-19 The Rail-Lead cask mesh.......................................................................... D-47

Figure D-20 Rail-Lead cask regulatory uniform-heating results ....................................... D-49

Figure D-21 Rail-Lead cask CAFE regulatory fire........................................................... D-51

Figure D-22 Rail-Lead cask on ground at the pool center ............................................... D-53

Figure D-23 Rail-Lead cask on ground 3.0 m (10 feet) from the edge of the pool.............. D-55

Figure D-24 Rail-Lead cask on ground 18.3 m (60 feet) from the edge of the pool ............ D-57

Figure D-25 Components of Truck-DU cask .................................................................. D-59

Figure D-26 Truck-DU cask mesh ................................................................................ D-64

Figure D-27 CAFE three-dimensional domain with the Truck-DU cask on the ground........ D-65

Figure D-28 Truck-DU cask on the ground at the pool center.......................................... D-66
NUREG-2125 xvi

Figure D-29 Large calorimeter fire test: (a) test setup and (b) fire fully engulfing the 

calorimeter............................................................................................... D-68

Figure D-30 Side view (looking from the north) of calorimeter and test setup.................... D-69

Figure D-31 CAFE benchmark results using fully engulfed large calorimeter: 

(a) temperatures averaged along the 0-, 90-, 180-, and 270-degree side 

looking at the calorimeter from the negative z-direction and (b) temperatures 

averaged over each ring starting from Ring 1 located on the positive side of 

the z-axis................................................................................................. D-69

Figure D-32 CAFE benchmark results using a small calorimeter 1.5 m (4.9 feet) from the 

edge of the fire......................................................................................... D-71

Figure E-1 Event tree for highway accidents......................................................................E-3

Figure E-2 Rail accident event tree ....................................................................................E-5

Figure E-3 The RADTRAN loss-of-shielding two-dimensional model .................................E-7

Figure E-4 Radiation dose rates to the MEI from loss of lead gamma shielding at 

distances from 1 to 5 meters from the cask carrying spent fuel ........................E-10

Figure E-5 Radiation dose rates to the MEI from loss of lead gamma shielding at 

distances from 10 to 100 meters from the cask carrying spent fuel...................E-10

Figure E-6 Event tree branch for a rail fire accident ...........................................................E-15

Figure E-7 RADTRAN input and output for Table E-11 ......................................................E-18
xvii NUREG-2125

LIST OF TABLES

Table PS-1 Collective Dose from Routine Transport for the Truck Route from Maine 

Yankee Nuclear Power Plant to Oak Ridge National Laboratory ......................... xxx

Table 1-1 NRC-Certified Commercial Light-Water Power Reactor Spent Fuel Casks............11

Table 1-2 Casks Chosen and Reasons for Selection ............................................................14

Table 2-1 External Radiation Doses from the Casks in this Study.........................................17

Table 2-2 Maximum Individual In-Transit Doses....................................................................19

Table 2-3 Characteristics of Rural, Suburban, and Urban Routes Used in RADTRAN ..........21

Table 2-4 Truck Route Segment Lengths and Population Densities, Kewaunee NP to 

Skull Valley ...........................................................................................................23

Table 2-5 Specific Routes Modeled.......................................................................................27

Table 2-6 Collective Doses to Residents near the Route (person-Sv) Per Shipment for 

Rail Transportation................................................................................................28

Table 2-7 Collective Doses to Residents near the Route (person-Sv) for Truck 

Transportation Per Shipment.................................................................................29

Table 2-8 Collective Doses (person-Sv) Per Shipment to Occupants of Trains Sharing 

Rail Routes ...........................................................................................................31

Table 2-9 Collective Doses (person-Sv) Per Shipment to Occupants of Vehicles 

Sharing Truck Routes ...........................................................................................33

Table 2-10 Input Data for Calculating Doses at Truck and Train Stops ...................................35

Table 2-11 Collective Doses at Rail Stops on the Maine Yankee-to-Hanford Route................36

Table 2-12 Collective Dose to Residents near Stops and Workers at Stops and on Board 

the Train................................................................................................................37

Table 2-13 Collective Doses to Residents near Truck Stops...................................................38

Table 2-14 Occupational Doses and Dose Rates from Routine Incident-Free 

Transportation.......................................................................................................39

Table 2-15 Total Collective Dose in Person-Sv from Routine Transportation for Each Rail 

Route for the Rail-Lead Cask ................................................................................41

Table 2-16 Total Collective Dose in Person-Sv from Routine Transportation for Each Rail 

Route for the Rail-Steel Cask................................................................................42

Table 2-17 Total Collective Dose in Person-Sv from Routine Transportation for Each 

Highway Route for the Truck Cask........................................................................43

Table 3-1 Maximum Lead Slump for the Rail-Lead Cask from Each Analysis Case ..............54

Table 3-2 Available Areas for Leakage from the Rail-Lead Cask ..........................................58

Table 3-3 Comparison of Analyses between This Study and NUREG/CR-6672....................60

Table 3-4 Deformation of the Closure Region of the Steel-DU-Steel Truck Cask from 

NUREG/CR-6672, mm..........................................................................................62

Table 3-5 Peak Contact Force for the Rail-Lead Cask Impacts onto an Unyielding 

Target ...................................................................................................................63

Table 3-6 Equivalent Velocities for Impacts onto Various Targets with the Rail-Lead 

Cask, kph..............................................................................................................64

Table 3-7 Accident Speeds That Result in the Same Damage as a Perpendicular 

Impact, kph ...........................................................................................................66
NUREG-2125 xviii

Table 5-1 Illustrations of Net Probability..............................................................................112

Table 5-2 Scenarios and Conditional Probabilities of Rail Accidents Involving the 

Rail-Lead Cask ...................................................................................................114

Table 5-3 Dose to an Emergency Respondera from a Cask in a No-Shielding Loss, 

No-Release Accident ..........................................................................................115

Table 5-4 Collective Dose Risks to the Public from a No Shielding Loss, No Release 

Accident Involving Rail Casks (person Sv) ..........................................................116

Table 5-5 Collective Dose Risks to the Public from a No Shielding Loss, No Release 

Accident Involving a Truck Cask (person Sv) ......................................................117

Table 5-6 Average Railcar Accident Frequencies and Accidents Per Shipment on the 

Routes Studied ...................................................................................................120

Table 5-7 Collective Dose Risks Per Shipment in Person-Sv for a Loss of Lead 

Shielding Accident Involving a Lead-Shielded Rail Cask .....................................120

Table 5-8 Doses to an Emergency Responder or Other Individual 5 Meters (16.4 feet) 

from the Cask for 10 Hours .................................................................................121

Table 5-9 Radionuclide Inventory for Accident Analysis of the Rail-Lead Cask ...................123

Table 5-10 Parameters for Determining Release Functions for the Accidents that Would 

Result in Release of Radioactive Material ...........................................................124

Table 5-11 Doses (Consequences) in Sv to the Maximally Exposed Individual from 

Accidents that Involve a Release ........................................................................126

Table 5-12 Total Collective Dose Risk (person-Sv) for Release Accidents Per Shipment 

for Each Route ....................................................................................................128

Table 5-13 Total Collective Dose Risk (person-Sv) for Each Route from a Loss of Lead 

Shielding Accident...............................................................................................128

Table 5-14 Total Collective Dose Risk (person-Sv) from Release and Loss of Lead 

Shielding Accidents.............................................................................................128

Table 5-15 Total Collective Dose Risk (person-Sv) from No-Release, No-Loss of 

Shielding Accidents Involving the Rail-Lead Cask ...............................................129

Table 5-16 Total Collective Dose Risk (person-Sv) from Loss of Neutron Shielding for 

Accidents Involving the Rail-Lead Cask ..............................................................129

Table B-1 Vehicle-Specific Parameters .............................................................................B-7

Table B-2 Route Parameters for Unit Risk Calculation ......................................................B-8

Table B-3 Parameter Values in the RADTRAN 6 Analysis.................................................B-9

Table B-4 Specific Routes Modeled...................................................................................B-13

Table B-5 Population Multipliers ........................................................................................B-14

Table B-6 Population Corrections for 2010 Census Data...................................................B-15

Table B-7 Maximum Individual Doses ...............................................................................B-19

Table B-8 Individual Doses (“Unit Doses” or “Unit Risks”) to Various Receptors for Rail 

Routes ..............................................................................................................B-21

Table B-9 Unit Risk Values for Various Receptors along Truck Routes .............................B-23

Table B-10 Collective Doses to Residents along the Route (Person-Sv) from Rail 

Transportation; Shipment Origin—INL...............................................................B-25

Table B-11 Collective Doses to Residents along the Route (Person-Sv) from Rail 

Transportation; Shipment Origin—Indian Point .................................................B-26

Table B-12 Collective Doses to Residents along the Route (Person-Sv) from Rail 

Transportation; Shipment Origin—Kewaunee ...................................................B-28
xix NUREG-2125

Table B-13 Collective Doses to Residents along the Route (Person-Sv) from Rail 

Transportation; Shipment Origin—Maine Yankee..............................................B-29

Table B-14 Collective Doses to Residents along the Route (Person-Sv) from Truck-DU; 

Shipment Origin—Maine Yankee ......................................................................B-31

Table B-15 Collective Doses to Residents along the Route (Person-Sv) from Truck-DU; 

Shipment Origin—Indian Point ..........................................................................B-33

Table B-16 Collective Doses to Residents along the Route (Person-Sv) from Truck-DU; 

Shipment Origin INL..........................................................................................B-35

Table B-17 Collective Doses to Residents along the Route (Person-Sv) from Truck-DU; 

Shipment Origin—Kewaunee ............................................................................B-36

Table B-18 Collective Doses (Person-Sv) to Occupants of Trains Sharing the Route..........B-39

Table B-19 States Comprising the 10 EPA Regions ............................................................B-40

Table B-20 Collective Doses to Persons Sharing the Route (Person-Sv) from Truck-DU; 

Shipment Origin—Maine Yankee ......................................................................B-42

Table B-21 Collective Doses to Persons Sharing the Route (Person-Sv) from Truck-DU; 

Shipment Origin—Indian Point ..........................................................................B-44

Table B-22 Collective Doses to Persons Sharing the Route (Person-Sv) from Truck-DU; 

Shipment Origin—INL .......................................................................................B-46

Table B-23 Collective Doses to Persons Sharing the Route (Person-Sv) from Truck-DU; 

Shipment Origin—Kewaunee ............................................................................B-47

Table B-24 Example of Rail Stops on the Maine Yankee-to-Hanford Rail Route .................B-48

Table B-25 Doses at Rail Stops on the Maine Yankee-to-Hanford Rail Route .....................B-49

Table B-26 Collective Doses to Residents near Truck Stops...............................................B-51

Table B-27 Occupational Doses Per Shipment from Routine, Incident-Free 

Transportation...................................................................................................B-52

Table C-1 Peak Acceleration from Each Analysis of the Rail-Steel Cask...........................C-21

Table C-2 Peak Acceleration from Each Analysis of the Rail-Lead Cask ...........................C-40

Table C-3 Available Areas for Leakage from the Rail-Lead Cask ......................................C-54

Table C-4 Peak Contact Force for the Rail-Lead Cask Impacts against an Unyielding 

Target ...............................................................................................................C-57

Table C-5 Equivalent Velocities for Rail-Lead Cask Impacts against Various Targets .......C-61

Table C-6 Properties of Fuel Assembly .............................................................................C-64

Table C-7 Zircaloy-4 Material Parameters .........................................................................C-66

Table C-8 Standard Material Properties.............................................................................C-71

Table C-9 Load Case Parameter Changes........................................................................C-71

Table C-10 Analysis Case Summary ...................................................................................C-77

Table D-1 Axial Burnup Profile in the Active Fuel Region of the Rail-Steel Cask ................D-13

Table D-2 Thermal Conductivities for the Rail-Steel Cask Materials...................................D-16

Table D-3 Specific Heat for the Rail-Steel Cask Materials ..................................................D-17

Table D-4 Densities for the Rail-Steel Cask Materials ........................................................D-18

Table D-5 Emissivity for Some of the Rail-Steel Cask Materials and Paints .......................D-18

Table D-6 Effective Thermal Conductivity for the Fuel-Basket Region................................D-20

Table D-7 Effective Thermal Conductivity of the Aluminum Heat Conduction Elements .....D-21

Table D-8 Effective Thermal Conductivity of the Intermediate Shells in the In-Plane 

Directions...........................................................................................................D-22
NUREG-2125 xx

Table D-9 Effective Conductivity of the Neutron Shield Region ..........................................D-23

Table D-10 Thermal Conductivities for the Rail-Lead Cask Materials ...................................D-43

Table D-11 Specific Heat for the Rail-Lead Cask Materials ..................................................D-43

Table D-12 Densities for the Rail-Lead Cask Materials ........................................................D-44

Table D-13 Emissivity for Some of the Rail-Lead Cask Materials .........................................D-44

Table D-14 Effective Thermal Properties of the Directly Loaded Fuel Basket.......................D-46

Table D-15 Effective Thermal Conductivities for the Neutron Shield Region of the Rail-Lead 

Cask ..................................................................................................................D-47

Table D-16 Axial Burnup Profile in the Active Fuel Region of the Truck-DU Cask ................D-61

Table D-17 Thermal Conductivities for the Truck-DU Cask Materials ...................................D-62

Table D-18 Volumetric Specific Heat for the Truck-DU Cask Materials.................................D-62

Table D-19 Effective Thermal Conductivities for the Truck-DU Cask Materials.....................D-63

Table D-20 Effective Volumetric Specific Heat for the Truck-DU Cask Materials ..................D-63

Table E-1 Truck and Railcar Accidents Per Kilometer, 1991 through 2007 ........................E-2

Table E-2 Parameters of Lead Shield Slumping from Impact .............................................E-8

Table E-3 Radiation Dose Rates (Sv/h) to the MEI at Various Distances for the Cask .......E-9

Table E-4 The “Conditional Dose Risk” in Sv to the MEI at Distances from the Cask from 

1 to 5 Meters for 1 Hour.....................................................................................E-11

Table E-5 The “Conditional Dose Risk” in Sv to the MEI at Distances 10 to 100 Meters 

from the Cask for 10 Hours................................................................................E-12

Table E-6 Average Railcar Accident Frequencies and Accidents Per Shipment on the 

Routes Studied ..................................................................................................E-13

Table E-7 Collective Dose Risks Per Shipment (Person-Sv) from Loss of Lead Shielding, 

Including Accident and Conditional Probabilities ................................................E-14

Table E-8 Radiation Dose (Sv) to the MEI at Various Distances from a Cask That Has 

Been in a Fire ....................................................................................................E-15

Table E-9 Events Leading to a Train Fire That Could Involve a Spent Fuel Cask...............E-16

Table E-10 Some Parameters Used in Calculating Loss of Neutron Shielding .....................E-17

Table E-11 Doses to an Emergency Responder 5 Meters from the Cask from an Accident 

with a Loss of Neutron Shielding........................................................................E-18

Table E-12 Collective Conditional Doses to Nearby Residents in Person-Sv from Loss of 

Neutron Shielding ..............................................................................................E-19

Table E-13 Collective Conditional Dose Risks in Person-Sv from Loss of Neutron 

Shielding............................................................................................................E-19

Table E-14 Collective Dose Risks in Person-Sv from Loss of Neutron Shielding..................E-20

Table E-15 Inventory for the Rail-Lead Cask........................................................................E-21

Table E-16 Parameters for Determining Release Functions for the Accidents That Would 

Result in Release of Radioactive Material..........................................................E-25

Table E-17 Sources of the Parameter Values in Table E-16 ................................................E-25

Table E-18 MEI Doses (Consequences) in Sv from Accidents That Involve a Release ........E-26

Table E-19 MEI Conditional Dose Risks in Sv from Accidents That Involve a Release ........E-26

Table E-20 Collective Inhalation and External Dose Risks for the Side-Impact, 

193-kph-Impact-Speed Accident for the 16 Rail Routes Analyzed .....................E-27
xxi NUREG-2125

Table E-21 Total Collective Dose Risks (Person-Sv) for Each Route from an Accident 

Involving Release, Per Shipment .......................................................................E-28

xxiii NUREG-2125

EXECUTIVE SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) has conducted several risk assessments and 

other analyses to evaluate the safety of transportation of spent power reactor nuclear fuel during 

the past 35 years. Regulations, shipping practices, and cask designs for transporting 

radioactive material have remained essentially unchanged during this time. Therefore, the 

actual per shipment risk over this time period also would have remained essentially the same. 

What has changed during this period is the calculated risks. This change was brought about by 

the improved ability to evaluate cask responses and their spent fuel contents to accident 

environments. The improvements include advancements in tools available to determine those 

responses and to calculate the consequences and risks that result from their response. This 

has resulted in a decrease in the calculated per shipment risk. The consequences and risks 

resulting from accidents calculated in this study are several orders of magnitude less than those 

calculated in previous risk assessments.

In this study, the risk associated with the transportation of spent nuclear fuel (SNF) was 

estimated by examining the behavior of three NRC-certified casks during routine transportation 

and in transportation accidents. Two casks are designed for transport by railroad: (1) a cask 

with steel gamma shielding and an inner welded canister for the spent fuel and (2) a cask with

lead gamma shielding that can transport spent fuel within an inner welded canister (referred to 

in this report as canistered fuel) or without an inner canister (referred to as directly loaded fuel).

A third cask with depleted uranium (DU) gamma shielding is designed to transport directly 

loaded spent fuel by highway. The response of these casks is typical of other cask designs. 

The use of certified cask designs means this risk assessment includes the factors of safety 

typically included in cask designs but not specifically considered in previous risk assessments.

The risks associated with routine shipments (incident-free) and shipments where an accident 

occurs are calculated separately. During routine transportation, the risk and the consequence 

are the same. In this case, the dose to residents living along a transportation route, to people 

sharing the highway or railway, people at stops, and transportation workers are all calculated. 

Regulations allow limited external radiation from the cask. The dose of radiation to members of 

the public during routine transportation is a small fraction of the naturally occurring background 

radiation that individuals experience.

If an accident occurs during shipment, most likely there is no damage to the cask. In this type of 

accident the shipping vehicle is stopped for a period of time, which exposes people in the 

vicinity of this stop (nearby residents, emergency response workers, etc.) to the allowed 

external radiation from the cask. If the accident is more severe, the shielding effectiveness of 

the cask could be reduced. If the cask is involved in a fire, the plastic neutron shielding material 

could melt, resulting in a slightly elevated amount of radiation emanating from the cask. If the 

lead shielded cask was involved in an exceptionally severe long-lasting fire, there could be a 

reduction in the effectiveness of the gamma shielding. The response of the cask to fire 

accidents was determined using detailed computer analyses. Even in the worst-case fires 

analyzed, no cask experienced a seal failure that could have led to a release of radioactive 

material from the spent fuel cask.

For impact accidents, the steel shielded cask with inner welded canister and the DU-shielded 

cask have no release and no loss of gamma shielding effectiveness even under the most 

severe impacts studied, which encompass all historic or even realistic accidents. The lead 
NUREG-2125 xxiv

shielded cask experiences some loss of gamma shielding effectiveness during severe impacts. 

Also, when spent fuel is transported without an inner welded canister some release of 

radioactive material could occur during exceptionally severe impacts.

If material were to be released, weather conditions at the accident location would affect the 

dispersal of that material. The risk assessment uses national average weather conditions 

because the time and location of an accident are unknown. The number of people exposed to 

the dispersed material is a function of the population density at the site of the accident, which is 

determined from census data. The amount of material released, the dispersion, and the 

population density are combined to determine the consequence (potential effects) of a release. 

The estimated dose from the most severe accident scenarios evaluated in this study is less than 

that required to produce an immediate injury or fatality.

Accident risk is the product of the consequence of the accident and its probability. The 

probability of an accident that has an effect on the cask is the product of the probability that the 

cask is involved in an accident and the conditional probability that the accident is severe enough 

to reduce the shielding or containment effectiveness of the cask. The conditional probability is 

based on State accident statistics for all types of heavy trucks and railcars. The accident 

probability is determined by multiplying these State-by-State accident rates by the distance 

traveled within each State. This was done for 16 representative truck routes and 

16 representative rail routes. The representative routes chosen are for illustrative purposes 

only, and no SNF shipments are planned from any of the points of origin to any of the 

destinations.

The study reached the findings listed below.

• The collective dose risks from routine transportation are very small. These doses are 

approximately four to five orders of magnitude less than the collective background 

radiation dose.

• The routes selected for this study adequately represent the routes for SNF transport, 

and there was relatively little variation in the risks per kilometer (km) over these routes.

• Radioactive material would not be released in an accident if the fuel is contained in an 

inner welded canister inside the cask.

• Only rail casks without inner welded canisters would release radioactive material, and 

only then in exceptionally severe accidents.

• If there were an accident during a spent fuel shipment, there is only about 

one-in-a-billion chance that the accident would result in a release of radioactive material.

• If there were a release of radioactive material in a spent fuel shipment accident, the dose 

to the maximally exposed individual (MEI) would be less than 2 sieverts (Sv) (200 rem)

and would not result in an acute lethality.

• The collective dose risks for the two types of extremely severe accidents (accidents 

involving a release of radioactive material and loss of lead shielding (LOS) accidents) 

are negligible compared to the risk from a no-release, no-loss of shielding accident.
xxv NUREG-2125

• The risk of gamma shielding loss from a fire is negligible.

• None of the fire accidents investigated in this study resulted in a release of radioactive 

material.

Based on these findings, this study reconfirms that radiological impacts from spent fuel 

transportation conducted in compliance with NRC regulations are low. In fact, they are

generally less than previous, already low, estimates. Accordingly, this study also reconfirms the 

NRC’s previous conclusion that regulations for transportation of radioactive material are 

adequate to protect the public against unreasonable risk.

A more complete plain-language summary of the report is given in the following section.

xxvii NUREG-2125

PUBLIC SUMMARY
NUREG-2125 xxviii

Nuclear fission in power reactors produces a large amount of energy, which has been 

harnessed to make electricity. Fission also creates radioactive products that are contained in 

fuel rod pins in nuclear fuel assemblies. Therefore, spent nuclear fuel is very radioactive when 

first removed from a reactor, but it decays and becomes less radioactive over time. Because of 

this radioactivity, people have some concerns when spent fuel is moved in trucks and by rail 

over public roads and railroads.

Thirty-five years ago, the U.S. Nuclear Regulatory Commission (NRC) responded to these

concerns by estimating the radiological impact of transporting radioactive materials, including 

spent fuel. This analysis resulted in NUREG-0170, “Final Environmental Statement on the 

Transportation of Radioactive Material by Air and Other Modes,” issued in 1977 (U.S. Nuclear 

Regulatory Commission, 1977). NUREG-0170 provided an environmental impact statement 

(EIS) for transportation of all types of radioactive material by road, rail, air, and water, and 

concluded the following:

• The average radiation dose to members of the public from routine transportation of 

radioactive materials is a fraction of their background dose.

1

• The radiological risk from accidents in transporting radioactive materials is very small 

compared to the nonradiological risk from accidents involving large trucks or freight 

trains.

On the basis of this EIS, NRC regulations in 1981 were considered “adequate to protect the 

public against unreasonable risk from the transport of radioactive materials.” However, the 

adequacy of these regulations continued to be questioned, in part, because the EIS was based 

on estimates of radiation dose and accident rates, for which not much data or information had 

been available. Among the questions not fully resolved: What constitutes “reasonable” risk and 

what are actual consequences should an accident happen?

The present work uses advanced models, risk assessment methods, and updated data to 

provide a current assessment of the risks and consequences of transporting spent nuclear fuel.

All commodities that are transported by truck or rail can be involved in accidents. Trucks and 

railcars carrying spent nuclear fuel transportation casks are no exception. The NRC recognizes 

this, and it requires that spent fuel casks be designed and built to withstand severe 

transportation accidents. NUREG-0170 and later studies of casks have considered accident 

conditions more severe than those the regulations require the cask to demonstrate their ability 

to withstand. A 1987 study applied actual accident statistics to projected spent fuel 

transportation (Fischer et al., 1987). This study, known as the “Modal Study,” also recognized 

that accidents could be described in terms of the strains they produced in the cask (for impacts) 

and the increase in cask temperature (for fires). Like NUREG-0170, the 1987 study based risk 

estimates on models because the limited number of accidents that had occurred involving spent 

fuel shipments was not sufficient to support projections or predictions. The Modal Study’s

refinement of modeling techniques and use of accident frequency data resulted in smaller 

assessed risks than had been projected by NUREG-0170.

 

1 The background dose is the average dose any individual will receive over the period of a year while 

conducting routine, everyday activities (3.6 millisieverts).
xxix NUREG-2125

In 2000, a study of two generic truck casks and two generic rail casks analyzed the cask 

structures and response to accidents by using computer modeling techniques (Sprung et al., 

2000). The study used semitrailer truck and rail accident statistics for general freight shipments 

because, even though more than 1,000 spent fuel shipments had been completed in the United 

States by 2000 and many thousands more had been completed safely internationally, there had 

been too few accidents involving spent fuel shipments to provide statistically valid accident 

rates.

Through a series of risk assessments, the release of radioactive material from a cask in an 

accident—and its subsequent dispersion—has been modeled with increasing refinement. 

NUREG-0170 assumed that most very severe accidents would result in release of all of the fuel 

particles created by the accident to the environment (the cask did not serve as a barrier to 

release). Although this engineering judgment overstated the release, it was nevertheless used 

because analytical capabilities at the time did not permit a more accurate assessment. The 

2000 study analyzed the physical properties of spent fuel rods in a severe accident and revised 

estimates of material released to 1 percent or less of the NUREG-0170 estimates. Accordingly, 

risk estimates were revised downward. The 2000 study also verified that an accidental release 

of radioactive material could only be through the seals at the end of the cask where the lid is 

attached. In other words, an accident could cause seal failure, but would not breach the cask 

body (Sprung et al., 2000).

The present study models certified cask designs (rather than generic casks) and the commercial 

spent nuclear fuel that these casks are certified to transport. It evaluated two rail casks and a 

truck cask.

Almost all spent fuel casks are shipped without incident. However, even this routine, 

incident-free transportation causes radiation exposures because all loaded spent fuel casks 

emit some external radiation. The radiation dose rates for spent fuel shipments are measured 

before each shipment and must be maintained within regulatory limits. The radiation dose from 

this external radiation to any member of the public during routine transportation, including stops, 

is barely discernible compared to the public’s natural background radiation. Figure PS-1

illustrates a rail cask and the way in which the radiation to a member of the public is modeled.

One hundred times the dose at 1 meter (3.3 feet) from the cask measured in millisieverts/hour 

(the dose measured in millirem/hour) is known as the Transport Index, which is used to 

represent the amount of radiation coming from the cask during routine transportation.

The external radiation from the spent fuel cask results in a very small dose to each member of 

the public along the route traveled by the cask. The collective dose from routine transportation 

is the sum of all of these doses. This study examined several example transportation routes

considered to be representative of possible cross-country transport. No actual spent fuel 

transport has occurred, or is planned to occur, on the routes studied. Table PS-1 and Figure 

PS-2 show the possible total dose in person-sieverts (person-Sv) to all of the workers and 

members of the public who would be exposed to radiation along one of these routes—the truck 

shipment from the Maine Yankee Nuclear Power Plant to Oak Ridge National Laboratory.

Table PS-1 and Figure PS-2 include the background radiation dose to exposed workers and 

members of the public during the time of the shipment.
NUREG-2125 xxx

Figure PS-1 Model of a spent fuel cask in routine, incident-free transportation and 

radiation dose to a member of the public

Relative sizes of the cask and member of the public are approximately to scale.

Table PS-1 Collective Dose from Routine Transport for the Truck Route from Maine 

Yankee Nuclear Power Plant to Oak Ridge National Laboratory (person-Sv)

Exposed Population Rural Suburban Urban

Urban 

Rush Hour Total

Residents near route 0.0000050 0.000089 0.0000020 0.00000045 0.000096

Traffic on the route 0.00013 0.00024 0.000054 0.0000050 0.00046

Residents near truck stops 0.00000056 0.000012 * * 0.000012

Truck crew 0.00059 0.000076 0.00067

Escort 0.000000047 0.0000000043 0.000000051

Inspectors (10 inspections) 0.0016

People at truck stops 0.00086

Truck stop workers 0.000013

Total dose from spent fuel shipment 0.0037

Background 7.56

* Most truck stops are located in rural or suburban areas.

Note: (1 Sv = 100,000 mrem)
xxxi NUREG-2125

Figure PS-2 Collective doses from background and from a truck shipment of spent 

nuclear fuel (person-Sv)

Figure note: (1 Sv = 100,000 mrem)

The collective doses calculated for routine transportation are higher for this study than for either 

NUREG/CR-6672, “Re-examination of Spent Fuel Shipment Risk Estimates,” (Sprung et al., 

2000) or NUREG-0170 (NRC, 1977), but still a very small fraction of background dose. Figure 

PS-3 compares the collective doses from truck transportation from the three studies. In 

NUREG-0170, the analysis was for a single route; in NUREG/CR-6672, the analysis was for 

200 representative routes (Sprung et al., 2000); and in this study, the analysis is for 16 truck 

routes (as well as 16 rail routes). The collective average dose in the present study is larger than 

the NUREG/CR-6672 result because present populations are generally larger, particularly along 

rural routes; the number of vehicles sharing the highways with the spent fuel transport is now 

much larger (see Chapter 2); and the number and length of refueling stops is much greater. 

These increases were somewhat offset by the greater vehicle speeds used in the present study.

This study uses current (1991 to 2007) truck and rail accident statistics to determine the 

probability of an accident and the severity of that accident. Researchers performed detailed 

analyses to evaluate how the casks would respond to the accident scenarios. Figure PS-4

shows a cask response to one impact scenario, a 97 kilometer per hour (kph) (60 mile per hour 

(mph)) corner impact onto a rigid target, and the resulting deformations. Almost all of the 

deformation is in the impact limiter, a device that is added to the cask to absorb energy, much 

like the bumper of a car. Similar analyses were performed for impacts at 48 kph (30 mph), 97

kph (60 mph), 145 kph (90 mph), and 193 kph (120 mph) in end-on (lid down), corner, and 

side-on orientations for two cask designs. These impact speeds encompass all accidents for 

truck and rail transportation.
NUREG-2125 xxxii

Figure PS-3 Collective doses (person-Sv) from routine truck transportation

Figure PS-4 Corner impact onto a rigid target at a 97-kph (60-mph) accident scenario for 

a spent fuel cask and the deformations produced by the impact
xxxiii NUREG-2125

Figure PS-5 shows one fire scenario, a 3-hour engulfing fire, and the resulting temperature 

distribution in the cask. Additional simulations were performed with the fire offset from the cask.

These fires include all fire-related accidents in rail transportation. The longest duration for an 

engulfing fire during truck transportation is 1 hour because of the amount of fuel that is carried 

onboard a tanker truck.

Detailed impact simulations were performed for two spent fuel casks—the NAC STC and the HI 

STAR 100—intended for transportation by railroad. In addition, the results for a third cask, the 

GA-4, which is intended for transportation by truck, were inferred from earlier analyses.

Detailed fire simulations were performed for all three casks.

Figure PS-5 Engulfing fire scenario and the temperature contours in the rail cask 

following a 3-hour fire duration

The transparency of the flames has been increased so the cask can be seen; in the actual fire 

simulation, and in a real fire, the flames are opaque.

The impact and thermal analysis results indicate that no accident involving the truck 

transportation cask would result in the release of radioactive material or reduction in the 

effectiveness of the gamma shielding. The only radiological consequence of an accident would 

be exposure to external radiation from the cask because of the long-duration stop associated 

with the accident. The stop needs to be long enough for responders to clear the accident scene 

and to arrange for shipment to resume. During this stop, emergency responders could be fairly 

close to the cask. Because there is no loss in effectiveness of the gamma shielding, the 

radiation dose to these responders would be a small fraction of the allowed occupational dose.

For rail transport of spent fuel that is in an inner welded canister, this study shows that there 

would be no release of radioactive material. For casks using lead gamma shielding, the most 

severe accidents evaluated led to a reduction in the effectiveness of that shielding, which results

in an elevated external radiation level. In addition, for rail transport of spent fuel that is not in an 
NUREG-2125 xxxiv

inner welded canister, some radioactive material is released following exceptionally severe and 

improbable accidents.

The calculated collective dose risk (the summation of dose to all exposed individuals times the 

probability of the accident) from accidents has decreased with each successive risk 

assessment. Figure PS-6 compares the average collective dose risks from releases and loss of 

lead shielding from the three studies (NUREG-0170 did not calculate loss of lead shielding).

This study also considered accident doses from a source that was not analyzed in the prior 

studies—the dose that results from accidents in which there is neither release nor loss of lead 

shielding, but there is increased exposure to a cask that is stopped for an extended period of 

time. Figure PS-7 shows the average collective dose risks for this scenario for the three casks 

studied. This scenario is important because more than 99.999999 percent of all accident 

scenarios do not lead to either release of radioactive material or loss of shielding. Figure PS-8 

provides a summary of all the accident probabilities and risks. The first pie chart shows that 

only about 1 in 1,000 trips would result in an accident. The second pie chart shows that if an 

accident occurs, only about 1 in 2,000 accidents is more severe than the regulatory accident 

conditions. The third pie chart shows that if an accident is more severe than the regulatory 

accident conditions, only about 3 in 1,000,000 will result in either loss of gamma shielding or 

release of radioactive material.

Figure PS-6 Average collective dose risk from release and loss-of-shielding (LOS) 

accidents

The LOS bar for NUREG/CR-6672 is not to scale.
xxxv NUREG-2125

Figure PS-7 Average collective dose risk from accidents that have no impact on the 

cargo
NUREG-2125 xxxvi

Most shipments are routine

99.86% of shipments occur without 

accidents

Collective Dose from Routine 

Transportation: 2.3 Person-mSv

Most accidents are less severe than 

the hypothetical accident casks are 

designed to withstand

99.95% of accidents would not exceed 

regulatory requirements

Collective Dose Risk from 10-hour 

Stop: 0.085 Person-mSv

Casks provide safety well beyond 

the regulatory requirements

99.99973% of accidents that are more 

severe than the regulatory hypothetical 

accident do not lead to release or loss 

of lead gamma shielding.

Collective Dose Risk from Loss of Lead 

Gamma Shielding: 2.5 x10-10

Person-mSv

Collective Dose Risk from Release:

3.5 x10-11 Person-mSv

Dose to a hypothetical maximally 

exposed individual (1.6 Sv) would not 

result in an acute fatality.

Figure PS-8 Calculated results for an illustrative rail shipment using the rail-lead cask
xxxvii NUREG-2125

A final point of comparison between the studies is the maximum consequence of an accident.

For NUREG-0170, this was about 110 person-Sv; for NUREG/CR-6672, it was about 

9,000 person-Sv; and for this study, it is 2.2 person-Sv. The reduction in consequence is the 

result of using the actual spent fuel being shipped, a smaller release fraction, and improvements 

in the RADTRAN model. This study estimated the effects of an accident on the maximally 

exposed individual (a theoretical person located at the point of highest concentration of 

potentially released radioactive material for 10 hours). The estimate for such an individual is 

calculated to be a dose of 1.6 Sv, and would not cause an acute fatality (fatality within two 

monthes of receiving the radiation dose).

As noted above, the purpose of this analysis was to reproduce (and, in some cases, extend) risk 

analyses previously considered in NUREG-0170, the Modal Study, and NUREG/CR-6672 using 

updated models and methods. The study reached the following findings:

• The collective doses from routine transportation are vanishingly small. These doses are 

about four to five orders of magnitude less than collective background radiation doses.

• The routes selected for this study adequately represent the routes for spent nuclear fuel 

transport, and there was relatively little variation in the risks per kilometer over these 

routes.

• Radioactive material would not be released in an accident if the fuel is contained in an 

inner welded canister inside the cask.

• Only rail casks without inner welded canisters would release radioactive material, and 

only then in exceptionally severe accidents.

• The regulatory hypothetical accident conditions are more severe than 99.995 percent of 

all accidents.

• The certification process not only assures that casks will survive the hypothetical 

accident conditions, but that they also survive 99.9999 percent of more severe

accidents. Therefore, if there was an accident during a spent fuel shipment, there is less 

than one-in-a-billion chance the accident would result in a release of radioactive 

material.

• If there was a release of radioactive material in a spent fuel shipment accident, the dose 

to the maximally exposed individual would be less than 2 Sv (200 rem), and would not 

cause an acute fatality.

• The collective dose risks for the two types of extra-regulatory accidents (accidents 

involving a release of radioactive material and loss of lead shielding) are negligible 

compared to the risk from a no-release, no-loss-of-shielding accident.

• The risk of loss of shielding from a fire is negligible.

• None of the fire accidents investigated in this study resulted in a release of radioactive 

material.
NUREG-2125 xxxviii

Based on these findings, this study reconfirms that radiological impacts from spent fuel 

transportation conducted in compliance with NRC regulations are low. In fact, this study’s 

radiological impact estimates are generally less than the already low estimates reported in 

earlier studies. Accordingly, with respect to spent fuel transportation, this study reconfirms the 

previous NRC conclusion that the regulations for transportation of radioactive material are 

adequate to protect the public against unreasonable risk.
xxxix NUREG-2125

ACKNOWLEDGEMENTS

The U.S. Nuclear Regulatory Commission (NRC) staff acknowledges the considerable 

assistance in the development of this report from the staff at Sandia National Laboratories, in 

particular: Douglas J. Ammerman, Ph.D. (Principal Researcher); Nicole L. Breivik, Ph.D.; Victor 

G. Figueroa; Carlos Lopez; Ruth F. Weiner, Ph.D.; and David R. Miller.

The NRC staff also acknowledges the valuable external peer review and comments provided on 

this report by the staff at Oak Ridge National Laboratory, in particular: Matt Feldman (Peer 

Review Team Lead); Cecil Parks, Ph.D.; Richard Hale; Bryan Broadhead, Ph.D.; Juan Carbajo, 

Ph.D.; Mike Muhlheim, Ph.D.; Phillip Maziasz, Ph.D.; and Allen Smith, Ph.D. (Spectrum 

subcontractor to ORNL).

This report was reviewed by staff from NRC’s Office of Nuclear Material Safety and Safeguards: 

Christopher Bajwa; Gordon Bjorkman, Ph.D.; Robert Einziger, Ph.D.; Anita Gray, Ph.D.; Joseph 

Borowsky; and John R. Cook (Project Manager and Review Team Lead).

xli NUREG-2125

ACRONYMS AND ABBREVIATIONS

ALARA as low as is reasonably achievable

AMAD activity median aerodynamic diameter

Btu British thermal unit

BWR boiling-water (nuclear) reactor

C Celsius

CAFE container analysis fire environment

CFD computational fluid dynamics

CFR Code of Federal Regulations

CG center of gravity

Ci curie

cm centimeter(s)

COC certificate of compliance

CRUD Chalk River unidentified deposit

DOE U.S. Department of Energy

DOT U.S. Department of Transportation

DU depleted uranium

EIS Environmental Impact Statement

EPA U.S. Environmental Protection Agency

EQPS equivalent plastic strain

F Fahrenheit

FE finite element

FSS fuel support structure

g acceleration due to gravity

gal gallon

GWD gigawatt days

HAC hypothetical accident condition

ILSS impact limiter support structure

in inch

INL Idaho National Laboratory

KE kinetic energy

km kilometer(s)

kph kilometers per hours

ksi 1000 pounds per square inch

lb pound(s)

lbm pound(s) mass

LOS loss of (lead) shielding

MCNP Monte Carlo N-Particle

MEI maximally exposed individual

MJ million Joules

MLEP multi-linear elastic/plastic

mm millimeter(s)

MN million Newtons

MPC multi-purpose canister

MPa million Pascals

mph miles per hour

mrem millirem

MTU metric ton of uranium

MWd megawatt-days
NUREG-2125 xlii

NP nuclear plant

NRC U.S. Nuclear Regulatory Commission

OFA Optimized Fuel Assembly

ORNL Oak Ridge National Laboratory

psi pounds per square inch

PWR pressurized-water reactor

rem roentgen equivalent man

SAR safety analysis report

SNF spent nuclear fuel

Sv sievert

TBq terabecquerels = 1012 becquerels

TC thermocouple

TEDE total effective dose equivalent

TI transport index

W watt

CHEMICAL SYMBOLS

Am americium

Cm curium

Co cobalt

Cs cesium

Eu europium

I iodine

Kr krypton

O oxygen

Pb lead

Pu plutonium

Ru ruthenium

Sb antimony

Sr strontium

Te tellurium

U uranium

Y yttrium

Zr zirconium
1 NUREG-2125

1. INTRODUCTION

1.1 Organization of this Report

The body of the report consists of an executive summary and six chapters. The chapters 

describe the risk analysis qualitatively. Each chapter in this study has an associated appendix 

that describes the analytical methods and calculations used to arrive at the results discussed in 

the chapters. Descriptions of programs, calculations, and codes used are located in the 

relevant appendices.

1.1.1 Chapter 1 and Appendix A

Chapter 1 gives an introduction to the study, a brief background, a discussion of risk as applied 

to the transportation of radioactive materials, a discussion of cask selection, and a review of the 

organization of the report. Appendix A contains details of certified spent fuel casks and the 

certificates of compliance for the casks used in this study.

1.1.2 Chapter 2 and Appendix B

Chapter 2 and Appendix B discuss RADTRAN2 analysis of incident-free transportation. During 

routine (incident-free) transportation, spent fuel transportation casks deliver an external dose to 

anyone in proximity to the shipment. This chapter describes the consequence of the external 

dose. In most previous transportation risk studies, the regulatory maximum dose rate of

0.1 millisieverts (mSv)/hour at 2 meters from the cask was assumed to be the external dose rate 

from every cask evaluated in the particular study. The present study uses the actual predicted 

external dose rate from U.S. Nuclear Regulatory Commission (NRC)-certified casks, as reported 

in the safety analysis reports (SARs) for those casks.

1.1.3 Chapter 3 and Appendix C

Chapter 3 and Appendix C address the structural analyses used to determine the cask 

response to accidents and the parameters that determine loss of lead gamma shielding and 

releases of radioactive material. The results of detailed analyses of the impact of the casks with 

impact limiters onto rigid targets at speeds of 48 kilometers per hour (kph), 97 kph, 145 kph, and 

193 kph (30 miles per hour (mph), 60 mph, 90 mph, and 120 mph) in end, corner, and side-on 

orientations are given. Results are supplied for impacts onto other surfaces or objects. The 

response of the fuel assemblies that the casks carry is also discussed.

1.1.4 Chapter 4 and Appendix D

Chapter 4 and Appendix D address the thermal analyses used to determine the cask response 

to accidents and the parameters that determine loss of lead gamma shielding and potential 

releases of radioactive material. The results from fire analyses that completely engulf the cask 

as well as those offset from the cask are given. The temperature response of the cask seals, 

the shielding material, and the spent fuel is provided.

 

2 RADTRAN is the radioactive material transportation risk assessment code originally developed for the NRC 

in the 1970s by Sandia National Laboratories.
NUREG-2125 2

1.1.5 Chapter 5 and Appendix E

Chapter 5 and Appendix E address RADTRAN analysis of transportation accidents, 

development of accident event trees and conditional probabilities, development of the 

radionuclide inventory and radioactive materials releases and dispersion of released material in 

the environment. The chapter also discusses accidents where no releases occur (the most 

likely accidents) and the radioactive cargo is not affected at all, but the vehicle is held for many 

hours at the accident location before it is permitted to continue.

1.1.6 Chapter 6 and Public Summary

Chapter 6 summarizes the results of the analyses. The Public Summary at the front of the report

contains a “plain language” summary of this study.

1.1.7 Bibliography

The bibliography is located after the Appendices. It contains all cited references and other 

bibliographic material. Citations in the text (e.g., Sprung et al., 2000, Figure 7.1) include specific 

page, figure, or table references where appropriate.

1.1.8 Review Process

The draft and final versions of this NUREG have undergone technical and editorial reviews. 

Before the draft NUREG was published for public comment, it went through an internal NRC 

technical review and an external technical peer review conducted by Oak Ridge National 

Laboratories. The draft NUREG had a 60-day public comment period, and written comments 

were received from four organizations. The resolution of these comments is included in the 

Public Comment Resolution Report (NRC, 2013a). In addition, the report was reviewed by the 

NRC Advisory Committee on Reactor Safeguards Subcommittee on Radiation Protection and 

Nuclear Materials and by the full Advisory Committee on Reactor Safeguards (ACRS). The 

responses to comments from these two committees are included in the ACRS Comment 

Resolution Report (NRC, 2013b). The final NUREG incorporates changes to address public 

and ACRS comments, and has been reviewed by the NRC prior to publication.

1.2 Historical Transportation Risk Studies and the Purpose of this Analysis

The purpose of this study was to analyze the radiological risks of transporting spent nuclear fuel 

(SNF) in routine transportation and transportation accidents, using the latest available data and 

modeling techniques. This study primarily analyzes cask behavior rather than the behavior of 

the spent fuel being transported. The study is the latest in a series of assessments of this type

that analyzes the behavior of NRC-certified casks carrying fuel of known isotopic composition 

and burnup. The studies preceding this one were based on conservative and generic 

assumptions.

This study is not intended to be a risk assessment for any particular transportation campaign 

and does not include the probabilities or consequences of malevolent acts. It does not address 

the acceptance of the risks associated with transportation of SNF but can be used to inform 

such discussions.

The NRC certifies casks used to transport SNF under Title 10 of the Code of Federal 

Regulations (10 CFR) Part 71, “Packaging and Transportation of Radioactive Material,” dated 
3 NUREG-2125

January 26, 2004. The adequacy of these regulations was confirmed in NUREG-0170, “Final 

Environmental Statement on the Transportation of Radioactive Material by Air and Other 

Modes” (NRC, 1977), an environmental impact statement (EIS) for transportation of all types of 

radioactive material by road, rail, air, and water. Several conclusions drawn from this EIS are 

listed below.

• The average radiation dose to members of the public from routine transportation of 

radioactive materials is a fraction of the existing background radiation dose.

• The radiological risk from accidents in transporting radioactive materials is very small 

compared to the nonradiological risk from accidents involving large trucks or freight 

trains.

• The regulations in force at the time of the EIS were determined by the Commission to be

“adequate to protect the public against unreasonable risk from the transport of 

radioactive materials” (46 FR 21629; April 13, 1981).

The risk assessment of NUREG-0170 was based on very conservative estimates of risk 

parameters and on models available at the time; these models would be considered imprecise 

today. The NRC concluded that the regulations were adequate because even very conservative 

estimates of risk parameters did not result in unacceptable risk. The NRC also recognized that 

the agency’s policies on radioactive materials transportation should be “subject to close and 

continuing review.” Two comprehensive contractor reports on spent fuel transportation have 

been issued since 1977: the Modal Study (Fischer et al., 1987) and NUREG/CR-6672, 

“Re-examination of Spent Fuel Shipment Risk Estimates,” (Sprung et al., 2000).3 The Modal 

Study was the first intensive examination of vehicle accident statistics and the first to categorize 

the frequency of severe accidents by structural and thermal response of a transportation cask. 

The Modal Study concluded that the frequency of accidents severe enough to produce 

significant cask damage was considerably less than NUREG-0170 estimated. The Modal Study 

was not a risk analysis because it did not consider the radiological consequence of accidents, 

but risks less than those estimated in NUREG-0170 could be inferred.

NUREG/CR-6672 refined the mechanical stress/thermal stress combinations of the Modal Study 

and recast them as a matrix of accident-related impact speeds and fire temperatures. In 

addition, NUREG/CR-6672 developed expressions for the behavior of spent fuel in accidents 

and potential release of this material, and analyzed the potential releases. The enhanced 

modeling capabilities available for NUREG/CR-6672 allowed analyses of the detailed structural 

and thermal response of transportation casks to accidents. NUREG/CR-6672 also used results 

of experiments by Lorenz et al. (1980), Sandoval et al. (1988), and Sanders et al. (1992) to 

estimate releases of radioactive material from the fuel rods to the cask interior and from the 

cask interior to the environment, following very severe accidents. The radionuclides available 

for release in the accidents studied in NUREG/CR-6672 are from relatively low burnup 

(30 gigawatt days per metric ton uranium (GWD/MTU)) and relatively high burnup (60 

GWD/MTU) pressurized-water reactor (PWR) and boiling-water reactor (BWR) fuel, although 

the transportability of the high burnup fuel was not considered. NUREG/CR-6672 studied the 

 

3 “Modal Study” and “NUREG/CR-6672” are the names by which these documents are referred to in the 

general transportation literature. The actual titles are in the bibliography of this document.
NUREG-2125 4

behavior of two generic truck casks and two generic rail casks; each generic cask encompassed 

design features of several NRC-certified casks.

The risks calculated in NUREG/CR-6672 were several orders of magnitude less than the 

estimates of NUREG-0170, concluding that no radioactive material would be released in more 

than 99.99 percent of accidents involving spent fuel shipments. These smaller risk estimates 

resulted from the use of refined and improved analytical and modeling techniques, exemplified 

by the finite element (FE) analyses of cask structure, and some experimental data substituted 

for the engineering judgments used in NUREG-0170.

In addition to the NRC-sponsored risk assessments cited above, there have been many other 

studies on the subject of spent fuel transportation. Perhaps one of the most independent, 

objective, authoritative, and recent analyses is the National Research Council report

(co-sponsored by the NRC), “Going the Distance?—The Safe Transport of Spent Nuclear Fuel 

and High-Level Radioactive Waste in the United States” (Committee on Transportation of 

Radioactive Waste, 2006). This reference is recommended to readers interested in further 

information on transportation package safety, transportation risk, and particularly for its 

coverage of societal topics beyond the scope of the technical risk assessment in the present 

study. One of the “Going the Distance” findings was: 

The radiological risks associated with the transportation of spent fuel and 

high-level waste are well understood and are generally low, with the possible 

exception of risks from releases in extreme accidents involving very long 

duration, fully engulfing fires.

In part because of that finding, the NRC sponsored several studies to investigate the potential

consequence from severe historical fire accidents if a spent fuel cask was involved. Two of 

these studies investigated tunnel fires (Adkins et al., 2006; Adkins et al., 2007) and one 

investigated the response of a spent fuel cask to an accident below a highway overpass 

(Bajwa et al., 2011). While these three studies examined environments where fire accidents 

actually occurred, they made assumptions about the placement of a cask within that 

environment that would cause the most damage to the cask without considering the probability 

of the placement. (The accident risk analyses in Chapter 5 discuss the probability of the 

scenarios analyzed and the resultant risk.) This study also evaluates severe fire accident 

consequences (but not modeling any particular historical accidents), as well as their associated 

probabilities, to provide a risk perspective.

The present study analyzes the behavior of three currently certified casks carrying 

Westinghouse 17×17 PWR fuel assemblies with 45 GWD/MTU burnup, the highest burnup that 

any of the three casks were certified to carry as of 2008 (the time of the analyses; some of the 

casks already have had changes to their allowed contents). In the future these casks may be 

certified to carry higher burnup fuel that has been cooled for a longer time and with a similar 

source term. A brief discussion on the effect of this change is provided in Section 6.3. For 

routine transportation, the risks are slightly larger than those estimated in NUREG/CR-6672 

because although the actual external dose rates are less than the regulatory maximum used in 

the other studies, populations along the routes have increased significantly. For accidents, the 

radiological risks calculated in the current study are at least an order of a magnitude less. The 

reduction in the estimates of risk from those in NUREG–0170 and NUREG/CR-6672 is the 

result of new data (such as event trees and accident probabilities) and observations and 

improved modeling techniques.
5 NUREG-2125

1.3 Risk

Understanding transportation risk is integral to understanding the environmental and related 

human health impact of radioactive materials transportation. A large amount of data exists for 

deaths, injuries, and damage from traffic accidents, but there are no data on health effects that 

radioactive materials transportation cause since no such effects have been observed. 

Therefore, regulators and the public rely on estimates of risk to gauge the potential effects of 

radioactive materials transportation. The risk estimates consider the potential accidents and 

events, where they could occur, and how severe they might be. Risk estimates include 

estimating the likelihood and severity of transportation accidents, as well as the calculation of 

exposure of workers and members of the public to ionizing radiation from routine transportation.

Risk is usually defined by answering the questions posed by the risk “triplet,” which is identified 

below:

• What can happen (the scenario)?

• How likely is it (the probability)?

• What is the outcome if it happens (i.e., how bad is it (the consequence))?

A risk number (quantitative risk) is calculated by multiplying the probability and consequence for 

a particular scenario. The probability of a scenario is always less than or equal to 1, because 

the maximum probability of an event is 1 (100 percent); an event with 100 percent probability 

(probability=1) of occurrence is an event that is certain to happen. In reality, very few events 

are certain to happen or certain not to happen (zero probability). The probability of most events 

is between these two extremes. Transportation accidents involving large trucks, for example, 

have a very low probability. The probability of a traffic accident for all highway vehicles is about

0.0000012 per km (or 1.2 in 1,000,000 km) (0.000002 per mile (or 2 in 1,000,000 miles)),

according to the U.S. Department of Transportation (DOT) Bureau of Transportation Statistics 

(DOT, 2007). The probability of a particular traffic accident scenario is even smaller, as shown 

in the event trees in Appendix E (Figures E-1 and E-2).

1.3.1 Accident Data

The only data available to estimate the future probability of a scenario are how often that 

scenario has occurred in the past. The probability of the scenario can be considered the same 

as its historical frequency. In the case of transportation accidents, enough accidents must have 

occurred in the past so that future accidents per kilometer can be predicted with reasonable 

accuracy. That is, the sample must be large enough to be sampled randomly. The most 

applicable frequency would be the frequency of accidents involving vehicles carrying SNF, but 

there have been too few of these for a statistically valid prediction.4 The sample size could have 

been increased by using international data, but regulations and practices in other countries are 

not consistent with those in the United States. In any case, there have not been enough 

accidents worldwide involving spent fuel transportation to provide an adequate statistical 

database. Even accidents involving all hazardous materials transportation do not provide a 

 

4 The U.S. Department of Transportation’s (DOT) Bureau of Transportation Statistics lists accidents per year 

for all classes of hazardous materials. The 2009 database lists 76 class 7 (radioactive materials) rail and 

highway incidents in the past 10 years; 

http://www.phmsa.dot.gov/staticfiles/PHMSA/DownloadableFiles/Files/tenyr_ram.pdf. These data did not 

specify the type of radioactive material involved. Not all of these incidents are accidents by DOT definition.
NUREG-2125 6

large enough database from which to generate statistics on a State-by-State basis. The 

database used in this study is the frequency of highway accidents involving large semitrailer

trucks and the frequency of freight rail accidents (DOT, 2007). Freight rail accident frequency is 

based on accidents per railcar-mile.

1.3.2 Spent Nuclear Fuel Transportation Scenarios

Several scenarios categorize transportation risk in this study. The most probable is routine 

transportation of SNF without incidents or accidents between the beginning and end of the trip.

Routine transportation is an example of the risk triplet identified previously.

• What can happen? The scenario is routine incident-free transportation.

• How likely is it? The probability is 100 percent (even if the shipment is involved in an 

accident, it still has an incident-free segment and dose).

• What if it happens? The consequence is a radiation dose less than 1 percent of 

background to individuals near the cask or along the route.

The doses and risks from routine transportation are analyzed in Chapter 2.

The accident scenarios discussed in this study are:

(1) Accidents in which the spent fuel cask is not damaged or affected.

• Minor traffic accidents (fender benders, flat tires) resulting in minor damage to 

the vehicle.

• Accidents in which damage to the vehicle is enough that it cannot move from the 

scene of the accident under its own power. There is no damage to the spent fuel 

cask that results in increased radiation in this type of accident.

• Accidents involving a traffic death, injury, or both, but no damage to the spent 

fuel cask that results in increased radiation in this scenario.

(2) Accidents in which the spent fuel cask is affected.

• Accidents involving loss of shielding (either neutron or gamma shielding) but no 

release of radioactive material.

• Accidents in which a release of radioactive material occurs.

In the first type of accidents, the only potential radiation dose to the public is from exposure of 

members of the public to external radiation emanating from the cask while the vehicle is 

stopped. In the current study, all of these accidents assume that the vehicle is stopped for 

10 hours. Only the second type of accidents involves release of radioactive material.

Traffic accident statistics (accident frequencies) are used in the analysis to calculate risks.

Average traffic accident frequencies since 1996 for large semitrailer trucks are about 

1.3 accidents per million highway kilometers (which is about the same as the accident rate for 
7 NUREG-2125

all highway vehicles). For freight rail, average frequencies since 1996 are about 1 accident per 

10 million railcar kilometers. The overall accident probability is the product of the probability that 

an accident will happen and the conditional probability that it will be a particular type of accident.

The consequence of an accident scenario could be a dose of ionizing radiation, either from 

external radiation from a stationary cask or from radioactive material released in an accident.

The risk associated with an accident is the product of the overall accident probability and the 

accident consequence. This risk is expressed in dose units (e.g., Sv), but because it includes 

accident probability, it is referred to as “dose risk.” Similarly, collective dose risk (person-Sv) 

includes the accident probability and the collective (population) dose.

1.4 Regulation of Radioactive Materials Transportation

DOT regulates the transportation of radioactive materials as part of hazardous materials 

transport regulations, primarily under Title 49, “Transportation,” to CFR Part 173, “Shippers—

General Requirements for Shipments and Packaging,” dated October 1, 2011. Mode specific 

regulations are given in Parts 174 to 177 and specifications for packagings are given in 

Part 178. In addition, 49 CFR 174.471 allows the use of packagings certified by the NRC under 

10 CFR Part 71. The regulations of 10 CFR Part 20, “Standards for Protections against

Radiation,” also are relevant. NRC transportation regulations primarily apply to the 

transportation of packages. DOT regulations include labeling, occupational and vehicle 

standards, registration requirements, reporting requirements, and packaging regulations.

Generally, DOT packaging regulations apply to industrial and Type A packaging whereas the 

NRC regulations apply to Type A fissile materials packaging and Type B packaging. Industrial 

and Type A nonfissile packages are designed to resist the stresses of routine transportation and 

are not certified to maintain their integrity in accidents, although many do. Type B packages are 

used to transport very hazardous quantities of radioactive materials. They are designed to 

maintain their integrity in severe accidents because the NRC recognizes that any transport 

package and vehicle may be in traffic accidents. This study addresses SNF transportation; 

therefore, it is only concerned with SNF for Type B packaging. (For the remainder of this report, 

the term “cask” will be used to refer to the contents plus the packaging.)

Nuclear fuel that has undergone fission in a reactor is extremely hot and radioactive when it is 

removed from the reactor. To cool the fuel thermally and allow the highly radioactive and 

short-lived fission products in the fuel to decay, the fuel is discharged from the reactor into a 

large pool of water. The fuel usually remains in the pool as long as there is space for it. After 

the fuel has cooled sufficiently, it can be moved to dry surface storage at the reactor or 

transported to a storage site or other destination. Currently, very little transportation of spent 

commercial power reactor fuel takes place in the United States and there are no plans to 

transport SNF before it has cooled for 5 years. The transportation casks are rated for heat load, 

which often determines the cooling time needed for the fuel to be transported. Shielding or 

other considerations may also drive the required cooling time.

10 CFR Part 71

The NRC recognizes that vehicles carrying radioactive materials are as likely as any vehicles of 

similar size traveling on similar routes to be in accidents. Therefore, transportation packages for 

very radioactive materials such as SNF are designed to maintain their integrity in severe 
NUREG-2125 8

accidents.

5 Packages meeting this requirement are Type B packages, which include the casks 

considered in this analysis—the NAC-STC (NAC, 2004) and Holtec HI-STAR 100 (Holtec 

International, 2000) rail casks, and the GA-4 (General Atomics, 1998) legal-weight truck casks.

Type B packages are designed to pass the sequential series of tests described in 

10 CFR 71.73, “Hypothetical Accident Conditions.” These tests are summarized below.

(1) A 9-meter (30-foot) drop onto an essentially unyielding horizontal surface. “Essentially 

unyielding” in this context means the target is hard and heavy enough that the package 

absorbs nearly all of the impact energy and the target absorbs very little energy. This 

test condition is more severe than most transportation accidents.

(2) A 1-meter (40-inch)6 drop onto a fixed 15-centimeter (cm) (6-inch) diameter steel 

cylinder to test the package’s resistance to punctures.

(3) An 800 degrees Celsius (C) (1,475 degrees Fahrenheit (F)) fire that fully engulfs the 

package for 30 minutes.

(4) Immersion under 0.9 meters (3 feet) of water. In addition, a nonsequential immersion in 

15 meters (50 feet) of water for 1 hour.

Figure 1-1 illustrates this sequence of tests.

In addition to the immersion test of 10 CFR 71.73, an undamaged cask carrying spent fuel is 

also required by 10 CFR 71.61, “Special Requirements for Type B Packages Containing More 

Than 105

A2,” to withstand an external pressure of 2 million Pascals (MPa) (290 pounds per 

square inch (psi)) for a period of not less than 1 hour without collapse, buckling, or inleakage of 

water. This pressure is equivalent to an immersion in 200 meters (660 feet) of water.

The package tests in 10 CFR 71.73 were developed to envelope real-life accidents. These 

tests are not intended to represent any specific transportation route, any specific historical 

transportation accident, or a “worst-case” accident. These tests are intended to simulate the 

damaging effects of a severe transportation accident in a manner that provides international 

acceptability, uniformity, and repeatability. All International Atomic Energy Agency Member 

States use these tests.

 

5 Although regulations allow the release of a specific quantity of each radionuclide, Type B casks typically are 

designed to remain leak-tight. 6 When discussing the regulations, the conversion between SI units and English units are those in the 

regulations. The actual arithmetic conversion factors are used in other areas of this report rather than the 

nominal conversions adopted by convention within the regulations.
9 NUREG-2125

Figure 1-1 The four tests for Type B packages

The tests are performed on a package design (either physically using a full-scale prototype or 

sub-scale test unit, or via computational modeling), but not on every package that will be used 

to transport SNF. A package designer may create computer models to evaluate the 

performance of a package design or components of the package design, build full-size or scale 

model packages for physical testing, or incorporate references to previous satisfactory 

demonstrations of a similar nature. In practice, the safety analysis performed for Type B 

packages often incorporates a combination of physical testing, computer modeling, and 

engineering evaluation. The SAR packaging contains information on the package design’s 

performance in the tests and an evaluation against the acceptance criteria in 10 CFR Part 71.

The SAR is used to apply for package certification. During the certification process, the NRC 

reviews the SAR to ensure that the package design meets all criteria specified in 

10 CFR Part 71.

NRC regulations specify that release of material from the package can be no more than the 

amount allowed to be shipped in a nonaccident resistant Type A package. The regulation also 

specifies a maximum post-test external radiation dose rate of 0.01 Sv per hour (1 rem/hr) at 

1 meter (40 inches) from the package surface.

10 CFR Part 20

Regulations in 10 CFR Part 20, “Standards for Protection against Radiation,” prescribe the 

largest allowable radiation dose that a member of the public may receive from NRC-licensed 

facilities, exclusive of background radiation, diagnostic or therapeutic radiation, or material 

discharged to the environment in accordance with NRC regulations. This section of the code 

does not apply to transport, but provides doses that can be compared to those calculated in this 

study. These doses are listed below.
NUREG-2125 10

• 1 mSv per year (100 mrem per year) total effective dose equivalent (TEDE), including 

both external and committed internal dose.

• 0.02 mSv per hour (2 mrem per hour) in any unrestricted area from external sources. As 

shown in Table 2-12, for example, doses from routine, incident-free transportation are 

considerably below these limits.

• 5 mSv per year (500 mrem per year) from a licensed facility if the licensee can show the 

need and expected duration of doses larger than 1 mSv (100 mrem) per year.

Although the regulations state clearly that these dose limits do not include background, it can

provide a useful comparison to other sources of radiation exposure since it affects everyone. 

The average background radiation dose in the United States is 0.0036 Sv (360 mrem) per year.

Part 20 also regulates occupational doses to 0.05 Sv per year (5 rem per year) TEDE.

1.5 Selection of Casks

Past risk assessments of spent fuel transportation have used generic cask designs with features 

similar to real casks but generally without all of the conservatisms that are part of real cask 

designs, such as assumptions on material strength and energy-absorbing capabilities of impact 

limiters. In the current study, the risk assessment was performed using actual cask designs 

with all of the design margins that contribute to their robustness. Because it is too costly and 

time-consuming to examine all casks, a subset of casks was selected for the risk assessment. 

Appendix A lists the various NRC-certified spent fuel casks at the time the study began, 

provides options for choosing the casks, describes some important features of the various cask 

designs, and finally concludes with the casks chosen.

Table 1-1 lists the casks that were NRC-certified as of 2006 (the date when the cask selections 

for this study were made) for the transportation of irradiated commercial light-water power 

reactor fuel assemblies. Those above the heavy line are older designs that were no longer 

used, but still had valid certificates. Those below the heavy line were more modern and 

additional units of these designs could be built. The casks chosen for this study came from the 

latter group. Appendix A includes brief descriptions of these casks.
11 NUREG-2125

Table 1-1 NRC-Certified Commercial Light-Water Power Reactor Spent Fuel Casks

Cask Package ID Canister

Contents (Number 

of assemblies) Type

IF-300 USA/9001/B( )F No 7 PWR, 17 BWR Rail

NLI-1/2 USA/9010/B( )F No 1 PWR, 2 BWR Truck

TN-8 USA/9015/B( )F No 3 PWR Overweighta

TN-9 USA/9016/B( )F No 7 BWR Overweighta

NLI-10/24 USA/9023/B( )F No 10 PWR, 24 BWR Rail

NAC-LWT USA/9225/B(U)F-96 No 1 PWR, 2 BWR Truck

GA-4 USA/9226/B(U)F-85 No 4 PWR Truck

NAC-STC USA/9235/B(U)F-85 Both 26 PWR Rail

NUHOMS®-MP187 USA/9255/B(U)F-85 Yes 24 PWR Rail

HI-STAR 100 USA/9261/B(U)F-85 Yes 24 PWR, 68 BWR Rail

NAC-UMS USA/9270/B(U)F-85 Yes 24 PWR, 56 BWR Rail

TS125 USA/9276/B(U)F-85 Yes 21 PWR, 64 BWR Rail

TN-68 USA/9293/B(U)F-85 No 68 BWR Rail

NUHOMS®-MP197 USA/9302/B(U)F-85 Yes 61 BWR Rail

a Overweight truck

Note: The casks in bold type are the ones selected for this study.

The casks chosen for detailed analysis were the NAC-STC (Figure 1-2) and the HI-STAR 100 

(Figure 1-3) rail casks. The GA-4 truck cask (Figure 1-4) was used to evaluate truck shipments, 

but detailed impact analyses of this cask were not performed because previous analyses of both 

truck and rail casks have shown that truck casks have significantly lower probability of release 

of radioactive material in impact accidents (Sprung et al., 2000). The impact analyses from 

Sprung et al. were used to assess the response of the GA-4 cask. Appendix A includes the 

complete certificate of compliance (COC) for each of these casks (as of April 12, 2010). The 

NAC-STC cask was chosen because it is certified for transport of spent fuel either with or 

without an internal welded canister. For transport of spent fuel without an internal canister, the 

NAC-STC’s COC allows the use of elastomeric or metallic O-rings. Although five casks in the 

group use lead for their gamma shielding, only the NAC-STC cask can transport fuel not 

contained within an inner welded canister. As noted in the analyses of Chapters 3, 4, and 5, the 

inclusion of spent fuel without an inner welded canister ensures that the potential pathway for 

radioactive material release into the environment was considered. The HI-STAR 100 rail cask 

was chosen because it was the only all-steel cask in the group certified for transport of fuel in an 

inner welded canister. The GA-4 truck cask was selected because it has a larger capacity than 

the NAC-LWT; therefore, it was more likely to be used in a large spent fuel transportation 

campaign. The chosen casks included all three of the most common shielding options: lead,

depleted uranium (DU), and steel.

Table 1-2 summarizes the casks chosen.

The choice of rail casks allowed for a comparison between directly loaded and canistered fuel, a 

comparison between a Steel-Lead-Steel cask and an All-Steel cask, and a comparison between 

elastomeric and metallic O-ring seals.
NUREG-2125 12

Figure 1-2 Photograph and cross-section of the NAC-STC cask 

Figure source: (courtesy of NAC International)
13 NUREG-2125

Figure 1-3 Basic layout and cross-section of the HI-STAR 100 rail transport cask 

Figure source: (from Haire and Swaney, 2005, and Holtec International, 2000)
NUREG-2125 14

Figure 1-4 GA-4 cask

Figure source: (courtesy of General Atomics)

Table 1-2 Casks Chosen and Reasons for Selection

Cask Chosen Type of Cask Reason for Consideration in this Study

HI-STAR 100 Rail Cask4 Rail-Steel Cask This was the only all-steel cask in the group that 

was certified for transport of fuel in an inner welded 

canister

NAC-STC Rail Cask7 Rail-Lead Cask Only the NAC-STC cask of this group can transport 

fuel that is not contained within an inner welded 

canister, thus ensuring the maximum potential for 

radioactive material released into the environment 

was considered.

GA-4 Truck Cask Truck-DU The GA-4 truck cask was chosen because its large

capacity made it more likely to be used in any large 

transportation campaign.

Detailed analyses in this report use the geometry and properties of the specific casks, but other 

similar casks are likely to respond in a similar manner. Therefore, the rest of this report refers to 

the HI-STAR 100 rail cask as Rail-Steel, the NAC-STC rail cask as Rail-Lead, and the GA-4 

truck cask as Truck-DU.

 

7 The choice of rail casks allowed comparison between directly loaded and canistered fuel, comparison 

between a Steel-Lead-Steel cask and an All-Steel cask, and comparison between elastomeric O-ring seals 

and metallic O-ring seals.
15 NUREG-2125

2. RISK ANALYSIS OF ROUTINE TRANSPORTATION

2.1 Introduction

NUREG–0170 (U.S. Nuclear Regulatory Commission (NRC), 1977), “Final Environmental 

Statement on the Transportation of Radioactive Material by Air and Other Modes,” was the first 

comprehensive assessment of the environmental and health impact of transporting radioactive 

materials. It documented estimates of the radiological consequences and risks associated with 

shipment by truck, train, plane, or barge of approximately 25 different radioactive materials, 

including power reactor spent fuel. However, little actual data on spent nuclear fuel (SNF)

transportation was available in 1977 and computational modeling of such transportation was in 

its infancy.

The RADTRAN computer code (Taylor and Daniel, 1977) is used in this chapter to estimate 

risks from routine8 transportation of SNF. Sandia National Laboratories initially developed 

RADTRAN for the NRC’s NUREG–0170 risk assessment. During the past several decades, the 

calculation method and RADTRAN code have improved to stay current with computer 

technology and supporting input data have been collected and organized. The basic RADTRAN 

analysis approach has not changed since the original development of the code, and the risk 

assessment method used in the RADTRAN code is accepted worldwide; about 25 percent of 

the 500 RADTRAN users are international.9

RADTRAN 6.0, integrated with the input file generator RADCAT (Neuhauser et al., 2000,

10

Weiner et al., 2009) is the version used in this study. The incident-free module of RADTRAN, 

the model used for the analysis in this chapter, was validated by measurement 

(Steinman et al., 2002), and verification and validation of RADTRAN 6.0 are documented in 

Dennis et al., 2008.

This chapter discusses risks to the public and workers when transportation of casks containing 

spent fuel takes place without incident and the transported casks are undamaged.

Nonradiological vehicular accident risk, which is orders of magnitude larger than the radiological 

transportation risk, is not analyzed in this study. The risks and consequences of accidents and 

incidents interfering with routine transportation are discussed in Chapter 5.

This chapter includes the following:

• A brief discussion of ionizing radiation emitted during transportation

• A description of the RADTRAN model of routine transportation

• Radiation doses from a single routine shipment to:

- Members of the public who live along the transportation route and near stops

- Occupants of vehicles that share the route with the radioactive shipment

 

8 The term “routine transportation” is used throughout this document to mean incident- or accident-free 

transportation. 9 The currently registered RADTRAN users are listed on a restricted-access Web site at Sandia National 

Laboratories. 10 Neuhauser et al. (2000) is the technical manual for RADTRAN 5 and is cited because the basic equations 

for the incident-free analyses in RADTRAN 6 are the same as those in RADTRAN 5. The technical manual 

for RADTRAN 6 is not yet available.
NUREG-2125 16

- Various groups of people at stops

- Workers

Appendix B includes detailed results of the RADTRAN calculations for this analysis. All 

references are listed in the bibliography. Weiner et al. (2009) provides a discussion of 

RADTRAN use and applications.

2.2 Radiation Emitted during Routine Transportation 

The RADTRAN model for calculating radiation doses is based on the well-understood behavior 

of ionizing radiation, which is that it can be absorbed by various materials, including air. 

Absorption of ionizing radiation depends on the energy and type of radiation and the absorbing 

material.

Spent nuclear fuel is very radioactive, emitting ionizing radiation in the form of alpha, beta, 

gamma, and neutron radiation. Casks used to transport SNF have thick walls that absorb most 

of the emitted ionizing radiation, thereby shielding workers and the public.

Figure 2-1 shows two generic cask diagrams with the shielding identified. This generic cask 

does not show the cross-section of any of the three casks used in this study.

Alpha and beta radiation cannot penetrate the casks’ walls (a few millimeters of paper and 

plastic actually absorb both well). The steel and lead layers of the cask wall absorb most of the 

gamma and neutron radiation emitted by spent fuel, although adequate neutron shielding also 

requires a neutron absorber layer, such as a polymer or boron compound. In certifying spent 

fuel casks, the NRC allows very low external dose rates for gamma and neutron radiation. For 

spent uranium-based fuel, the gamma radiation typically dominates the external dose rate.

Absorbed radiation dose is measured in sieverts (Sv) in the International System of Units, rem 

or millirem in the historic English unit system (millirem is abbreviated as mrem in this 

document). Average U.S. background radiation from naturally occurring and some medical 

sources is 0.0036 Sv (360 mrem) per year (Shleien et al., 1998, Figure 1.1). The recent 

increase in diagnostic use of ionizing radiation, as in computerized tomography, has suggested 

increasing the average background to 0.0062 Sv (620 mrem). This background value is cited 

on the NRC Web site11. The present study, however, uses the older value of 0.0036 Sv per 

year. A single dental x ray delivers a dose of 4×10-5 Sv (4 mrem) and a single mammogram 

delivers 1.3×10-4 Sv (13 mrem) (Stabin, 2009). The maximum radiation dose rate from a spent 

fuel cask that regulation allows is 10-4 Sv per hour (10 mrem/hour), measured at 2 meters 

(about 6.6 feet) from the outside of the cask (10 CFR Part 71), or about 0.00014 Sv/hour (14 

mrem per hour) at 1 meter (40 inches) from a cask 4 to 5 meters (13 to 17 feet) long.

The external radiation doses from the casks in this study (Figures 1-2 to 1-4), determined from 

values reported in the cask SARs, are shown in Table 2-1 (Holtec, 2000; NAC, 2004; General 

Atomics, 1998).

 

11 http://www.nrc.gov/about-nrc/radiation/around-us/doses-daily-lives.html
17 NUREG-2125

Figure 2-1 The upper figure is an exploded view of a generic spent fuel cask

Figure source: (Sandia National Laboratories archive)

The lower figure is a cross-section of the layers of the cask wall.

Table 2-1 External Radiation Doses from the Casks in this Study

Truck-DU Rail-Lead Rail-Steel

Transportation mode Highway Rail Rail

Dose rate Sv/h (mrem/h) at 1 m (40 

inches)

0.00014 (14) 0.00014 (14) 0.000103 (10.3)

 Gamma fraction 0.77 0.89 0.90

 Neutron fraction 0.23 0.11 0.10

The calculated radiation dose to workers and members of the public from a routine shipment is 

based on the external dose rate at 1 meter from the spent fuel cask as shown in Figure 2-2.
NUREG-2125 18

This dose rate, when expressed in mrem per hour (or mSv per hour times 100), is numerically 

equal to the transport index (TI). Doses from the external radiation from the cask depend on the 

external dose rate, the distance of the receptor from the cask, the exposure time, and 

intervening shielding.

Figure 2-2 RADTRAN model of the vehicle in routine, incident-free transportation

The cask in this diagram is positioned horizontally and the critical dimension is the cask length.

Figure note: (TI = transport index, CD = critical dimension, r = radius)

2.3 The RADTRAN Model of Routine, Incident-Free Transportation

2.3.1 The Basic RADTRAN Model

For analysis of routine transportation, RADTRAN models the cask as a sphere with a radiation 

source at its center and assumes that the dimensions of the trailer or railcar carrying the cask 

are the same as the cask dimensions. The emission rate of the radiation source is based on the 

TI instead of a shielding calculation. The radiation source is modeled as a virtual source at the 

center of the sphere shown in Figure 2-2 that produces the same TI as the cask. The diameter 

of this spherical model, called the “critical dimension,” is the longest dimension of the actual 

spent fuel cask.

When the distance to the receptor (r in Figure 2-2) is much larger than the critical dimension, 

RADTRAN models the dose to the receptor as proportional to 1/r2

. When the distance to the 

receptor r is similar to or less than the critical dimension, as for crew or first responders, 

RADTRAN models the dose to the receptor as proportional to 1/r. The RADTRAN spherical 

model overestimates the measured dose by a few percent (Steinman et al., 2002).

2.3.2 Individual and Collective Doses

The dose to workers and the public from a cask during routine transportation depends on the 

amount of time workers or the public are exposed to the cask, the distance from the cask, the 
19 NUREG-2125

external radiation from the cask, and intervening shielding. When the vehicle carrying the cask 

is traveling along the route, the faster the vehicle goes, the less exposure there is to anyone 

along the vehicle’s route. Therefore, an individual member of the public residing near the 

transport route receives the largest dose from a moving vehicle when he or she is as close as 

possible to the vehicle and the vehicle is traveling as slowly as possible. For trucks and trains 

carrying spent fuel at a speed of 24 kilometers per hour (kph) (15 miles per hour (mph)) and a 

distance of 30 meters (approximately 100 feet) are assumed for maximum exposure.12

Table 2-2 shows the maximum dose to an individual member of the public under these 

conditions. The Rail-Lead cask has a higher dose than the Rail-Steel cask because it has a 

higher TI. The Truck-DU (depleted uranium) cask has a higher dose than the Rail-Lead cask 

(same TI) because it has a longer critical dimension; therefore, it takes more time to pass a 

receptor. The transit speed used for both rail and truck transport in the calculation of the 

maximum individual dose is 24 kph (15 mph). These doses are about the same as 1 minute of 

average background: 6.9×10-9 Sv (6.9×10-4 mrem).

Table 2-2 Maximum Individual In-Transit Doses

Cask (mode) Dose, Sv (mrem)

Rail-Lead (rail) 5.7x10-9 (5.7x10-4)

Rail-Steel (rail) 4.3x10-9 (4.3x10-4)

Truck-DU (truck) 6.7x10-9 (6.7x10-4)

When a vehicle carrying a spent fuel cask travels along a route, the people who live along that 

route and the people in vehicles that share the route are exposed to the external radiation from 

the cask. Doses to groups of people are collective doses; the units of a collective dose are 

person-Sv (person-rem). A collective dose, sometimes called a population dose, is essentially 

an average individual dose multiplied by the number of people exposed.13 RADTRAN 

calculates collective doses along transportation routes by integrating over the width of a band 

along the route where the population resides (the r in Figure 2-2) and then integrating along the 

route. Collective doses to people on both sides of the route are included. The exposed 

population is in a band 770 meters (approximately 0.5 miles) on either side of the route: from 

30 meters (100 feet) from the center of the route to 800 meters (0.5 miles).

Figure 2-3 shows how these bands are defined with examples of distances within the bands.

Occupants of vehicles that share the route with the radioactive shipment also receive a radiation 

dose from the spent fuel cask. The collective dose to occupants depends on the average 

number of occupants per vehicle and the number of vehicles per hour that pass the radioactive 

shipment in both directions.

 

12 Thirty meters is typically as close as a person on the side of the road can get to a vehicle traveling on an 

interstate highway. 13 Appendix B contains a detailed discussion on the collective dose.
NUREG-2125 20

Figure 2-3 Diagram of a truck route as modeled in RADTRAN (not to scale)

Any route can be divided into as many sections as desired for dose calculation (e.g., the dose to 

residents of a single house or city block). However, as a practical matter, routes are divided into 

rural, suburban, and urban segments according to the population per square mile (population 

density). Table 2-3 summarizes the characteristics of each population type that is part of the 

RADTRAN dose calculation. References for these parameter values can be found in the Table 

2-3 footnotes.
21 NUREG-2125

Table 2-3 Characteristics of Rural, Suburban, and Urban Routes Used in RADTRAN

Highway routes are Interstate or other limited-access highways.

Basis Highway Rail

Rural Suburban Urban Rural Suburban Urban

Population 

density per 

km2 (per mi2

)

a TRAGIS

0 to 54 

(0 to 

139)

54 to 1,284 

(139 to 

3,326)

>1,284 

(>3,326)

0 to 54 

(0 to 139)

54 to 1,284 

(139 to 

3,326)

>1,284 

(>3,326)

Sidewalk 

occupant/ 

resident 

ratiob,g

Urban 

Areas NA NA 6 NA NA NA

Shielding by 

buildingsb Historic 

RADTRAN 

use

0

(outside)

13%

(wood)

98.2%

(concrete, 

brick)

0

(outside)

13%

(wood)

98.2%

(concrete, 

brick)

U.S. average 

vehicle speed 

kph (mph)c,d DOT 108 (67) 108 (67) 102(63) 40 (25) 40 (25) 24 (15)

U.S. average 

vehicles per 

hourb,e DOT 1119 2,464 5,384 17 17 17

Occupants of 

other 

vehiclesb,f DOT 1.5 1.5 1.5 1 1 5

a Johnson and Michelhaugh, 2003 b Weiner et al., 2009 Appendix D c DOT, 2004a d DOT, 2004b, Appendix D e DOT, 2009 (these are average railcars per hour) f DOT, 2008, Table 1-11 g Applies only to sidewalks on secondary roads in urban areas

Each route clearly has a distribution of rural, urban, and suburban areas, as indicated in the 

example of the truck route in Figure 2-4, which shows a segment of Interstate 80 through Salt 

Lake City, UT. The broad stripe is the half-mile band on either side of the highway. The red 

areas are urban populations, the yellow areas are suburban, and the green areas are rural.

Instead of analyzing each separate, rural, urban, and suburban segment of this stretch of 

highway, the rural, suburban, and urban areas are each combined for RADTRAN dose 

calculations. The routing code WebTRAGIS (Johnson and Michelhaugh, 2003) provides these 

combinations for each State traversed by a particular route. WebTRAGIS (Transportation 

Routing Analysis Geographic Information System) software, developed by Oak Ridge National 

Laboratory (ORNL), determines routes from specified starting and ending points for highway, 

rail, or waterway transportation within the continental United States. Various criteria for the

route(s) to be determined may be specified including Highway Route Controlled Quantity 

(HRCQ) criteria, which are used for the 16 truck routes presented within this document. 

WebTRAGIS also uses data from the ORNL database Landscan USA to determine populations 

along the route and bin these populations into rural, suburban, and urban subsets for each State 

within the route. For population determinations, the user may specify the buffer zone from 
NUREG-2125 22

which the population is calculated. The default buffer zone is 800 meters (on either side of the 

route) and this default buffer is used in this document. TRAGIS can be used to create the 

route-specific population input data for the Sandia National Laboratory developed software 

RADTRAN as was done for this study.

Figure 2-4 A segment of I-80 through Salt Lake City, UT

Table 2-4 shows the WebTRAGIS output for a truck route from Kewaunee Nuclear Plant (NP), 

WI, to Skull Valley, UT.

The maps in Figures 2-5 through 2-8 show the 16 truck and 16 rail routes analyzed in this 

report. These illustrative routes were selected as representative of possible cross-country 

transport. No actual spent fuel transport has occurred or is planned from any of these points of 

origin to any of these destinations. The maps are adapted from the output of the routing code 

WebTRAGIS (Johnson and Michelhaugh, 2003).
23 NUREG-2125

Table 2-4 Truck Route Segment Lengths and Population Densities,

Kewaunee NP to Skull Valley

The route segment of Figure 2-4 is in bold.

State Kilometers (miles) Persons/km2 (persons/mi2

)

a

Rural Suburban Urban Rural Suburban Urban

Illinois 45 (28) 1.2 (0.7) 15.4 (40) 267 (691) 2,049 (5,301)

Iowa 394 (245) 95 (59.1) 5.1(3.2) 15.7 (41) 268 (693) 2,185 (5,653)

Nebraska 652 (405) 76 (47.2) 7 (4.4) 10 (26) 269 (696) 2,401 (6,212)

Utah 197 (123) 38 (23.6) 15 (9.3) 7.5 (19.4) 407 (1,053) 2,412 (6,240)

Wisconsin 191 (119) 85 (52.8) 19.9 (12.4) 21.4 (55) 337 (872) 2,660 (6,882)

Wyoming 607 (377) 34 (21.1) 3.4 (2.1) 4.9 (13) 399 (1,032) 1,967 (5,089)

a The populations density is a WebTRAGIS output, calculated by averaging the population density along the 

rural, suburban, or urban route length within each State.

Figure 2-5 Highway and rail routes from Maine Yankee Nuclear Plant site

Figure note: NP stands for Nuclear Plant and ORNL stands for Oak Ridge National Laboratory.
NUREG-2125 24

Figure 2-6 Highway and rail routes from Kewaunee Nuclear Plant

Figure note: NP stands for Nuclear Plant and ORNL stands for Oak Ridge National Laboratory.
25 NUREG-2125

Figure 2-7 Highway and rail routes from Indian Point Nuclear Plant

Figure note: NP stands for Nuclear Plant and ORNL stands for Oak Ridge National Laboratory.
NUREG-2125 26

Figure 2-8 Highway and rail routes from Idaho National Laboratory

Figure note: INL stands for Idaho National Laboratory and ORNL stands for Oak Ridge National Laboratory.

The route segment lengths and population densities are entered into RADTRAN, which then 

calculates the collective doses to residents along the route segments. Collective doses, which 

depend on route length and on the populations along the route, were calculated for 1 shipment 

over each of 16 truck and 16 rail routes. Collective doses are reported as person-Sv.

The sites where the shipments originated include two nuclear generating plants (Indian Point 

and Kewaunee), a storage site at a fully decommissioned nuclear plant (Maine Yankee), and 

INL. The routes modeled are shown in Table 2-5. Both truck and rail versions of each route are 

analyzed.
27 NUREG-2125

Table 2-5 Specific Routes Modeled

Origin Destination

Population within 

800 m (1/2 mile)

Total 

Kilometers

Urban 

Kilometers

Rail Truck Rail Truck Rail Truck

Maine 

Yankee 

Site, ME

Hanford, WA 1,647,190 1,129,685 5,084 5,013 355 116

Deaf Smith County, TX 1,321,024 1,427,973 3,362 3,596 211 165

Skull Valley, UT 1,451,325 1,068,032 4,068 4,174 207 115

Oak Ridge, TN 1,146,478 1,137,834 2,125 1,748 161 135

Kewaunee 

NP, WI

Hanford, WA 476,914 423,163 3,028 3,453 60 52

Deaf Smith County, TX 677,072 494,920 1,882 2,146 110 60

Skull Valley, UT 806,115 505,226 2,755 2,620 126 58

Oak Ridge, TN 779,613 646,034 1,395 1,273 126 92

Indian 

Point NP, 

NY

Hanford, WA 961,026 869,763 4,781 4,515 229 97

Deaf Smith County, TX 1,027,974 968,282 3,088 3,074 204 109

Skull Valley, UT 1,517,758 808,107 3,977 3,672 229 97

Oak Ridge, TN 1,146,245 561,723 1,264 1,254 207 60

Idaho 

National 

Lab, ID

Hanford, WA 164,399 132,662 1,062 959 20 15

Deaf Smith County, TX 298,590 384,912 1,913 2,291 40 52

Skull Valley, UT 169,707 132,939 455 466 26 19

Oak Ridge, TN 593,680 569,240 3,306 3,287 75 63

Note: Urban Kilometers are Included in Total Kilometers, (1 Kilometer = 0.6214 miles)

These routes represent a variety of route lengths and populations. The routes include the 

eastern United States, western United States, and cross-country routes. They vary in length 

and include a variety of urban areas. Two of the three nuclear plants chosen as origin sites

(Kewaunee, WI, and Maine Yankee, ME) and two of the destination sites (Hanford, WA, and 

Skull Valley, UT) are origins and destinations used in NUREG/CR-6672, “Re-examination of 

Spent Fuel Shipment Risk Estimates,” (Sprung et al., 2000). Indian Point Nuclear Plant, NY, 

involves a different set of cross-country and east coast routes than Maine Yankee. It also is an 

operating nuclear plant whereas Maine Yankee has been decommissioned and is now a surface 

storage facility. Since this study could be used for both commercial nuclear power plant and 

U.S. Department of Energy (DOE) spent fuel shipments, INL was included as an origin site.

The destination sites include two proposed repository sites (Deaf Smith County, TX, and 

Hanford, WA) (DOE, 1986), the site of the proposed private fuel storage facility (Skull Valley, 

UT), and ORNL. These routes were not intended to provide a “worst case” result, but were 

chosen to provide representative results over a broad range of conditions and large segments of 

the country.

The route segments and population densities were provided by WebTRAGIS. The WebTRAGIS 

output files are available from ADAMS (ML 13249A339)(NRC, 2013c). Population densities 

were updated from the 2000 census using the 2008 Statistical Abstract (U.S. Bureau of the 

Census, 2008, Tables 13 and 21), which includes data up through 2006. Updates were only 

made when the difference between the 2006 and 2000 population densities was 1 percent or 

more. The population data in Table 2-5 is directly from the WebTRAGIS outputs, and do not 

include the updates to 2006. All other tables in this NUREG use the updated 2006 populations.

It was anticipated that 2010 census data would be available for this final report, but at the time 
NUREG-2125 28

of this writing it still was not possible to obtain updated population data in WebTRAGIS. An 

estimation of the error introduced by not updating to the 2010 census can be made by using the 

same method that was used to adjust the populations to 2006 data. The correction factors for 

each state are given in Table B-6. The collective doses reported in Table 2-6 and Table 2-7 are 

in units of person-Sv. Table 2-6 and Table 2-7 present collective doses for rail and truck, 

respectively, for the 16 routes. State-by-State collective doses are tabulated in Appendix B.

Table 2-6 Collective Doses to Residents near the Route (person-Sv) Per Shipment for 

Rail Transportation (1 Sv = 105 mrem)

FROM/TO

Rail-Lead Rail-Steel

Rural Suburban Urban Total Rural Suburban Urban Total

MAINE YANKEE

ORNL 1.5x10-5 1.8x10-4 9.0x10-6 2.1x10-4 1.2x10-5 1.4x10-4 6.8x10-6 1.6x10-4

DEAF SMITH 1.9x10-5 2.2x10-4 1.1x10-5 2.5x10-4 1.4x10-5 1.7x10-4 8.7x10-6 1.9x10-4

HANFORD 2.4x10-5 2.6x10-4 1.3x10-5 2.9x10-4 1.8x10-5 2.0x10-4 9.9x10-6 2.3x10-4

SKULL VALLEY 2.6x10-5 2.7x10-4 1.0x10-5 2.9x10-4 2.0x10-5 2.0x10-4 7.6x10-6 2.2x10-4

KEWAUNEE

ORNL 1.0x10-5 1.1x10-4 6.7x10-6 1.3x10-4 7.9x10-6 8.3x10-5 5.1x10-6 9.6x10-5

DEAF SMITH 8.2x10-6 9.5x10-5 5.8x10-6 1.1x10-4 6.3x10-6 7.2x10-5 4.4x10-6 8.3x10-5

HANFORD 1.2x10-5 9.3x10-5 3.0x10-6 1.1x10-4 9.3x10-6 7.1x10-5 2.3x10-6 8.3x10-5

SKULL VALLEY 1.4x10-5 1.2x10-4 6.6x10-6 1.4x10-4 1.1x10-5 9.0x10-5 5.0x10-6 1.1x10-4

INDIAN POINT

ORNL 7.5x10-6 1.4x10-4 1.4x10-5 1.6x10-4 5.7x10-6 1.1x10-4 1.1x10-5 1.2x10-4

DEAF SMITH 1.7x10-5 1.8x10-4 1.2x10-5 2.0x10-4 1.3x10-5 1.3x10-4 8.9x10-6 1.5x10-4

HANFORD 2.2x10-5 2.1x10-4 1.3x10-5 2.5x10-4 1.7x10-5 1.6x10-4 9.9x10-6 1.9x10-4

SKULL VALLEY 2.3x10-5 2.0x10-4 1.3x10-5 2.4x10-4 1.7x10-5 1.5x10-4 1.0x10-5 1.8x10-4

IDAHO NATIONAL LAB

ORNL 1.8x10-5 1.1x10-4 3.7x10-6 1.3x10-4 1.4x10-5 8.6x10-5 2.8x10-6 1.0x10-4

DEAF SMITH 6.6x10-6 5.8x10-5 2.2x10-6 6.7x10-5 5.0x10-6 4.5x10-5 1.7x10-6 5.2x10-5

HANFORD 5.3x10-6 3.0x10-5 1.1x10-6 3.6x10-5 4.0x10-6 2.3x10-5 8.2x10-7 2.8x10-5

SKULL VALLEY 3.0x10-6 2.5x10-5 1.5x10-6 3.0x10-5 2.3x10-6 1.9x10-5 1.1x10-6 2.2x10-5
29 NUREG-2125

Table 2-7 Collective Doses to Residents near the Route (person-Sv) for Truck 

Transportation Per Shipment (1 Sv=105 mrem)

FROM TO

Truck-DU

Rural Suburban Urban Urban 

Rush 

Houra

Total

MAINE 

YANKEE

ORNL 5.0x10-6 8.9x10-5 2.0x10-6 4.5x10-7 9.6x10-5

DEAF SMITH 1.0x10-5 1.2x10-4 2.1x10-6 4.8x10-7 1.4x10-4

HANFORD 1.4x10-5 1.0x10-4 1.5x10-6 3.2x10-7 1.2x10-4

SKULL VALLEY 1.1x10-5 9.5x10-5 1.5x10-6 3.3x10-7 1.1x10-4

KEWAUNEE

ORNL 4.1x10-6 4.6x10-5 1.1x10-6 2.5x10-7 5.2x10-5

DEAF SMITH 6.6x10-6 3.9x10-5 7.6x10-7 1.7x10-7 4.7x10-5

HANFORD 9.1x10-6 4.1x10-5 7.0x10-7 1.5x10-7 5.1x10-5

SKULL VALLEY 7.3x10-6 3.1x10-5 6.7x10-7 1.5x10-7 3.9x10-5

INDIAN 

POINT

ORNL 4.1x10-6 6.4x10-5 1.6x10-7 1.6x10-7 6.9x10-5

DEAF SMITH 1.3x10-5 1.3x10-4 6.9x10-7 3.1x10-7 1.4x10-4

HANFORD 1.3x10-5 7.6x10-5 2.6x10-7 2.6x10-7 8.9x10-5

SKULL VALLEY 1.0x10-5 6.6x10-5 2.7x10-7 2.7x10-7 7.7x10-5

IDAHO 

NATIONAL 

LAB

ORNL 8.8x10-6 5.3x10-5 7.7x10-7 1.7x10-7 6.3x10-5

DEAF SMITH 4.6x10-6 3.0x10-5 6.9x10-7 1.5x10-7 3.7x10-5

HANFORD 5.5x10-6 8.8x10-6 1.1x10-7 4.2x10-8 1.4x10-5

SKULL VALLEY 1.2x10-6 1.0x10-5 2.7x10-7 5.9x10-8 1.2x10-5

a During rush hour RADTRAN halves the truck speed and doubles the vehicle density to take into account 

traffic jams and gridlock. Detailed data for the actual traffic speed and density on a city-by-city basis is 

not available. The rush-hour collective dose is in addition to the urban (non-rush-hour) collective dose; 

both are included in the total.

Collective dose is best used in making comparisons (e.g., in comparing the risks of routine 

transportation along different routes, by different modes (truck or rail), or in different casks).

Several comparisons can be made from the results shown in Table 2-6 and Table 2-7.

• Suburban residents sustain the largest dose for all routes and shipment modes. The 

urban dose is less than the suburban dose because urban residences are modeled as 

83 percent shielded, while suburban residences are modeled as13 percent shielded.

• Urban residents sustain a larger dose from a single rail shipment than a truck shipment 

on the same State route even though urban population densities are similar and the 

external dose rates from the cask are nearly the same. As shown in Table 2-5, most 

(though not all) rail routes have more urban miles than the analogous truck route. Train 

tracks go from city center to city center whereas trucks carrying spent fuel must use 

interstates and bypasses. In several cases shown in Table 2-5, the rail route had twice 

as many urban miles as the corresponding truck route. Also, train speeds in urban 

areas are only one-fourth of truck speeds.

• Overall, collective doses are larger for a single shipment on rail routes than truck routes 

because rail routes are often longer, especially in the western United States, where 

there is rarely a choice of railroads and train speeds are lower than truck speeds, 
NUREG-2125 30

especially in urban areas. However, rail casks hold about six times as much spent fuel 

as the truck cask. Therefore, to move a given amount of spent fuel would take six truck 

shipments for each rail shipment, making the total dose from shipping by truck higher.

• The collective doses shown in Table 2-6 and Table 2-7 are all very small. However, they 

are not the only doses people along the route receive. Background radiation is 

0.0036 Sv (360 mrem) per year in the United States, or 4.1×10-7 Sv/hour

(0.041 mrem/hr). The contribution of a single shipment to the population’s collective 

dose is illustrated in the following example of the Maine Yankee to ORNL truck route:

- From Table 2-7 the total collective dose to residents for this route is 9.6×10-5

person-Sv (9.6 person-mrem).

- From Table 2-5, there are 1,137,834 people within 800 meters (1/2 mile) of the 

route.

- Background is 4.1×10-7 Sv/hour (0.041 mrem/hr), which everyone is exposed to 

all the time, whether a shipment occurs or not.

- A truck traveling at an average of 108 km per hour (67 mph) travels the 1,748 km 

(1086 miles) in 16 hours.

- During those 16 hours, the 1,137,834 people will have received a collective 

background dose of 7.56 person-Sv, (756 person-rem) about 80,000 times the 

collective dose from the shipment.

- To illustrate, the total collective dose during a shipment to these 1,137,834 

people is not 9.6×10-5 person-Sv (9.6×10-3 person-rem), but 7.560096 person-Sv

(756.0096 person-rem).

- The NRC recommends that collective dose only be used for comparative 

purposes (NRC, 2008).

- The appropriate comparison between the collective dose from this shipment of 

spent fuel is not a comparison between 9.6×10-5 person-Sv (9.6×10-3

person-rem) from the shipment and zero dose if there is no shipment, but 

between 7.560096 person-Sv (756.0096 person-rem) if there is a shipment and 

7.560000 person-Sv (756.0000 person-rem) if there is no shipment.

Appendix B, Section B.6 contains a more complete discussion of collective dose.

2.3.3 Doses to Members of the Public Occupying Vehicles that Share the Route

Rail

Most U.S. rail is either double track or equipped with “passing tracks” that let one train pass 

another. When a train passes the train carrying the spent fuel cask, occupants of the passing 

train will receive some external radiation. Most trains in the United States carry freight, and the 

only occupants of the passing train are crew members. Only about 1 railcar in 60 has an 

occupant.
31 NUREG-2125

The dose to occupants of other trains in this situation depends on train speed and the external 

dose rate from the spent fuel casks. Table 2-8 shows the collective dose to public passengers 

of trains sharing the route, assuming for calculation purposes that train occupants are 

represented by one person in each passing railcar in rural and suburban areas, and five people 

in urban areas.14 The rural and suburban collective doses probably are unrealistically high, 

since most freight rail shipments going through rural and many suburban areas never encounter 

a passenger train. Data were not available to account for the occupancy of actual passenger 

trains, including commuter rail, that share rail routes with freight trains.

Table 2-8 Collective Doses (person-Sv) Per Shipment to Occupants of Trains Sharing 

Rail Routes (1 Sv=105 mrem)

SHIPMENT 

ORIGIN/ 

DESTINATION

Rail-Lead Cask Rail-Steel Cask

Rural Suburban Urban Total Rural Suburban Urban Total

MAINE YANKEE

ORNL 2.0x10-5 1.2x10-5 7.5x10-6 4.0x10-5 1.5x10-5 9.3x10-6 5.6x10-6 3.0x10-5

DEAF SMITH 3.8x10-5 1.3x10-5 9.7x10-6 6.1x10-5 2.9x10-5 1.0x10-5 7.4x10-6 4.6x10-5

HANFORD 6.2x10-5 1.7x10-5 1.6x10-5 9.0x10-5 4.7x10-5 1.3x10-5 1.2x10-5 6.8x10-5

SKULL VALLEY 4.8x10-5 1.6x10-5 9.6x10-6 7.4x10-5 3.6x10-5 1.2x10-5 7.3x10-6 5.5x10-5

KEWAUNEE

ORNL 1.4x10-5 7.0x10-6 5.8x10-6 2.7x10-5 1.0x10-5 5.3x10-6 4.4x10-6 2.0x10-5

DEAF SMITH 2.4x10-5 5.2x10-6 5.1x10-6 3.4x10-5 1.8x10-5 4.0x10-6 3.9x10-6 2.6x10-5

HANFORD 4.2x10-5 6.7x10-6 2.8x10-6 5.2x10-5 3.2x10-5 5.1x10-6 2.1x10-6 3.9x10-5

SKULL VALLEY 3.5x10-5 7.8x10-6 5.8x10-6 4.9x10-5 2.7x10-5 5.9x10-6 4.4x10-6 3.7x10-5

INDIAN POINT

ORNL 9.2x10-6 8.1x10-6 9.6x10-6 2.7x10-5 7.0x10-6 6.1x10-6 7.2x10-6 2.0x10-5

DEAF SMITH 3.6x10-5 1.1x10-5 9.4x10-6 5.6x10-5 2.8x10-5 8.2x10-6 7.1x10-6 4.3x10-5

HANFORD 6.0x10-5 1.4x10-5 1.1x10-5 8.5x10-5 4.6x10-5 1.1x10-5 8.0x10-6 6.5x10-5

SKULL VALLEY 4.8x10-5 1.3x10-5 1.1x10-5 6.5x10-5 3.6x10-5 1.0x10-5 8.0x10-6 4.9x10-5

INL

ORNL 4.6x10-5 7.1x10-6 3.4x10-6 5.7x10-5 3.5x10-5 5.4x10-6 2.6x10-6 4.3x10-5

DEAF SMITH 2.7x10-5 3.2x10-6 1.9x10-6 3.2x10-5 2.1x10-5 2.5x10-6 1.4x10-6 2.5x10-5

HANFORD 1.5x10-5 1.7x10-6 9.3x10-7 1.8x10-5 1.2x10-5 1.3x10-6 7.0x10-7 1.4x10-5

SKULL VALLEY 5.5x10-6 1.5x10-6 1.2x10-6 8.2x10-6 4.2x10-6 1.1x10-6 9.0x10-7 6.2x10-6

Truck

Unlike trains, trucks carrying spent fuel share the primary highway system with many cars, light 

trucks, and other vehicles. The occupants of any car or truck that passes the spent fuel cask in 

 

14 The five persons per railcar in urban areas are assumed to include occupants of passenger trains. 

Passenger trains carry more than five per car, but the majority of railcars even in urban areas carry freight 

only. This estimate is consistent with estimates made in past studies.
NUREG-2125 32

either direction will receive a small radiation dose. This dose is modeled in RADTRAN as 

shown in Figure 2-9. RADTRAN assumes there is always a vehicle in the adjacent lane.

Figure 2-9 Diagram used in RADTRAN for calculating radiation doses to occupants of 

other vehicles 

Figure source: (from Neuhauser et al., 2000)

The radiation dose to occupants of other vehicles depends on the exposure distance and time, 

the number of other vehicles on the road, and the number of people in the other vehicles. 

Occupants of the vehicles that share the route are closer to the cask than residents or others 

beside the route. Occupants of vehicles moving in the opposite direction from the cask are 

exposed to radiation from the cask for considerably less time because the vehicles involved are 

moving past each other. The exposure time for vehicles traveling in the same direction as the 

cask is assumed to be the time needed to travel the link at the average speed 

(Neuhauser et al., 2000). It is assumed that there is always a vehicle in the adjacent lane at the 

position of the cask and a vehicle in the same lane at the MIN distance from the cask. The 

number of other vehicles that share truck routes is very large; the average number of vehicles 

per hour on U.S. interstate and primary highways in 200415 (Weiner et al., 2009, Appendix D) 

were:

• 1,119 on rural segments, about 2.5 times the 1977 vehicle density

• 2,464 on suburban segments, almost four times the 1977 vehicle density

• 5,384 on urban segments, about twice the 1977 vehicle density

Each vehicle was assumed to have an average of 1.5 occupants since most cars and light 

trucks traveling on freeways have one or two occupants. State highway departments provide 

 

15 2004 is the most recent year for which data have been validated. 
33 NUREG-2125

traffic count data but do not provide vehicle occupancy data. If two occupants are assumed, the 

collective doses are one-third larger.

Detailed discussion and State-by-State results are presented in Appendix B. The collective 

doses for truck traffic are shown in Table 2-9.

Table 2-9 Collective Doses (person-Sv) Per Shipment to Occupants of Vehicles Sharing 

Truck Routes (1 Sv=105 mrem)

FROM TO

Truck-DU

Rural Suburban Urban

Urban 

Rush 

Houra Totalb

MAINE 

YANKEE

ORNL 1.3x10-4 2.4x10-4 5.2x10-5 4.8x10-5 4.6x10-4

DEAF SMITH 2.8x10-4 3.3x10-4 6.9x10-5 6.4x10-5 7.3x10-4

HANFORD 4.5x10-4 3.0x10-4 4.3x10-5 4.0x10-5 8.3x10-4

SKULL VALLEY 3.7x10-4 2.5x10-4 4.4x10-5 4.5x10-5 7.0x10-4

KEWAUNEE

ORNL 9.6x10-5 1.4x10-4 4.8x10-5 4.4x10-5 3.3x10-4

DEAF SMITH 1.8x10-4 8.9x10-5 2.2x10-5 2.0x10-5 3.1x10-4

HANFORD 3.4x10-4 1.4x10-4 3.3x10-5 3.0x10-5 5.4x10-4

SKULL VALLEY 2.4x10-4 8.6x10-5 2.5x10-5 2.3x10-5 3.8x10-4

INDIAN POINT

ORNL 1.8x10-4 2.1x10-4 3.3x10-5 3.0x10-5 4.6x10-4

DEAF SMITH 2.8x10-4 3.1x10-4 5.6x10-5 5.2x10-5 6.9x10-4

HANFORD 4.2x10-4 2.2x10-4 4.8x10-5 4.4x10-5 7.2x10-4

SKULL VALLEY 3.6x10-4 2.2x10-4 4.5x10-5 4.1x10-5 6.6x10-4

IDAHO 

NATIONAL 

LAB

ORNL 3.0x10-4 1.5x10-4 2.4x10-5 2.2x10-5 5.0x10-4

DEAF SMITH 2.2x10-4 7.3x10-5 2.7x10-5 2.5x10-5 3.4x10-4

HANFORD 1.0x10-4 8.5x10-5 9.5x10-6 8.7x10-6 2.0x10-4

SKULL VALLEY 3.7x10-5 3.2x10-5 8.5x10-6 7.8x10-6 8.5x10-5

a During rush hour the truck speed is halved and the vehicle density is doubled, for details see Section 

B-5.3 in Appendix B. b Total includes the sum of Rural, Suburban, Urban, and Urban Rush Hour.

Comparing Table 2-6 to Table 2-8, the collective dose to residents for rail transport is generally 

larger (except in rural areas) than the collective dose to people sharing the rail line. In contrast,

comparing Table 2-7 to Table 2-9 shows that for all routes and population densities the 

collective dose to those sharing the highway is greater than the collective dose to nearby 

residents.

2.3.4 Doses at Truck and Train Stops

Trucks and trains occasionally stop on long trips. Common carrier freight trains stop to 

exchange freight cars, change crews, and, when necessary, change railroads. The rail stops at 

the origin and destination of a trip are called “classification stops” and are 27 hours long. Spent 

fuel casks may be carried on both dedicated trains and regular freight trains; however, in 
NUREG-2125 34

practice, previous spent fuel shipments have been carried on dedicated trains. A dedicated 

train is a train that carries a single cargo from origin to destination. Coal unit trains are an

example of dedicated trains. The analyses conducted in this study assume that the casks are 

transported on dedicated trains, which eliminates the need for intermediate classification stops.

When a train is stopped, the dose to anyone nearby depends on the distance between that 

person and the cask and the time that the individual is exposed. People exposed at a rail stop 

include those listed below.

• railyard workers (including inspectors)

• train crew (passenger trains do not typically enter railyards)

• residents who live near the railyard

The semi-tractor trucks that carry Truck-DU casks each have two 300-liter (80-gallon) fuel 

tanks. They generally stop to refuel when half of the fuel is gone, approximately every 845 km 

(525 miles) (DOE, 2002). Trucks carrying spent fuel also are stopped at the origin and 

destination of each trip. Mandatory rest and crew changes are combined with refueling stops 

whenever possible.

The people likely to be exposed at a refueling truck stop are listed below.

• the truck crew of two; usually one crew member at a time fills the tanks

• other people using the truck stop (since these trucks stop at public truck stops)

• residents of areas near the stop

Some States inspect spent fuel cask shipments when the trucks enter the State. Inspection 

stations may be combined with truck weigh stations; therefore, inspectors of both the truck 

carrying the spent fuel and the trucks carrying other goods can be exposed in addition to crew 

from other trucks. When the vehicle is stopped, receptor doses depend only on distance from 

the source and exposure time, so that any situation in which the cask and the receptor stay at a 

fixed distance from each other can be modeled as a stop. These stop-like exposure situations 

include inspections, vehicle escorts, vehicle crew when the vehicle is in transit, and occupants 

of other vehicles near the stopped vehicle. Any of these situations can be modeled in 

RADTRAN. Appendix B provides details on the calculations performed for situations in this 

analysis.

Figure 2-10 is a diagram of the model used to calculate doses at truck stops. The inner circle 

defines the area occupied by people who share the stop with the spent fuel truck, who are 

between the truck and the building, and who are not shielded from the truck’s external radiation.

People in buildings at the stop are shielded.

Table 2-10 lists the input data used to calculate doses at truck and train stops.
35 NUREG-2125

Figure 2-10 Diagram of truck stop model (not to scale)

Table 2-10 Input Data for Calculating Doses at Truck and Train Stops

Data Interstate 

Highway Freight Rail

Minimum distance from nearby residents, m (ft) 30 (100) 200 (660)

Maximum distance from nearby residents, m (miles) 800 (1/2) 800 (1/2)

Stop time for rail classification (hours) NA 27

Stop time in transit for railroad change (hours) NA <<1 to 4

Stop time at truck stops (hours) 0.83 NA

Minimum distance to people sharing the stop, m (ft) 1 (3.3)a NA

Maximum distance to people sharing the stop, m (ft) 15 (50)a NA

a From Griego et al., 1996

Rail

Trains are stopped for classification for 27 hours at the beginning and end of a trip. The 

collective dose from the radioactive cargo to the railyard workers at these classification stops for 

the two rail casks studied is:

• 1.46×10-5 person-Sv (1.46 person-mrem) for the Rail-Lead

• 1.09×10-5 person-Sv (1.09 person-mrem) for the Rail-Steel
NUREG-2125 36

The average dose (calculated by dividing the collective dose by the number of exposed people) 

to an individual living between 200 and 800 meters from a classification yard is:

• 3.5×10-7 Sv (0.035 mrem) from the Rail-Lead cask

• 2.7×10-7 Sv (0.027 mrem) from the Rail-Steel cask

Table 2-11 shows the train stops doses to yard workers and residents near the stops for the 

Maine Yankee-to-Hanford rail route calculated using the input data from Table 2-10. The doses 

for all 16 rail routes were calculated in a similar fashion and are presented in Table 2-12. The 

difference in collective dose to residents near stops from route-to-route is primarily due to the 

different population densities at the classification stops, which may be either in rural or suburban 

areas.

Table 2-11 Collective Doses at Rail Stops on the Maine Yankee-to-Hanford Route 

(person-Sv) (1 Sv=105 mrem)

Stop

Route type 

(R, S, U) 

and State

Time 

(hours)

Railyard Worker Residents Near Stop

Rail-Lead Rail-Steel Rail-Lead Rail-Steel

Classification, 

origin S, ME 27 1.5x10-5 1.1x10-5 2.3x10-5 1.8x10-5

In route1 S, ME 4.0 2.2x10-6 1.6x10-6 3.4x10-6 2.6x10-6

In route 2 R, NY 4.0 2.2x10-6 1.6x10-6 9.2x10-7 6.9x10-7

In route 3 S, IL 2.0 1.1x10-6 8.1x10-7 1.2x10-5 9.4x10-6

Classification, 

destination S, WA 27 1.5x10-5 1.1x10-5 1.9x10-5 1.4x10-5
37 NUREG-2125

Table 2-12 Collective Dose to Residents near Stops and Workers at Stops and on Board 

the Train (person-Sv) (1 Sv=105 mrem)

ORIGIN DESTINATION

RESIDENTS NEAR STOPS

RAILYARD WORKERS, 

CREW, AND ESCORTS

RAIL LEAD RAIL STEEL RAIL LEAD RAIL STEEL

MAINE 

YANKEE

ORNL 1.1x10-4 8.5 x 10-5 3.4x10-4 2.3x10-4

DEAF SMITH 5.3 x10-5 5.0 x 10-5 5.1x10-4 3.7x10-4

HANFORD 1.1x10-4 8.8 x 10-5 7.6x10-4 5.6x10-4

SKULL VALLEY 5.4 x10-5 4.1 x 10-5 6.2x10-4 4.5x10-4

KEWAUNEE

ORNL 1.1x10-4 8.3 x 10-5 2.3x10-4 1.5x10-4

DEAF SMITH 6.8 x10-5 5.2 x 10-5 3.0x10-4 2.1x10-4

HANFORD 1.1x10-4 8.7 x 10-5 4.7x10-4 3.3x10-4

SKULL VALLEY 1.2 x10-4 9.1 x 10-5 4.3x10-4 3.0x10-4

INDIAN 

POINT

ORNL 1.3x10-4 1.0 x 10-4 2.1x10-4 1.4x10-4

DEAF SMITH 5.9 x10-5 4.5 x10-5 4.8x10-4 3.4x10-4

HANFORD 1.1x10-4 8.3 x 10-5 7.2x10-4 5.2x10-4

SKULL VALLEY 5.6 x10-5 4.3 x 10-5 6.0x10-4 4.4x10-4

INL

ORNL 9.5 x10-5 7.2 x 10-5 5.1x10-4 3.6x10-4

DEAF SMITH 7.7 x10-5 5.8 x10-5 3.1x10-4 2.1x10-4

HANFORD 5.6 x10-5 4.3x 10-5 1.8x10-4 1.2x10-4

SKULL VALLEY 3.1x10-6 2.4 x 10-6 9.5x10-5 5.0x10-5

Truck

Table 2-13 shows the collective doses to residents near stops for the rural and suburban 

segments of the 16 truck routes studied calculated using the input data from Table 2-10. Urban 

stops were not modeled because trucks carrying spent fuel casks are unlikely to stop in urban 

areas (this is because most truck stops are not within urban areas, those that are within

metropolitan areas are usually in industrial areas that do not have urban population density, and 

because the DOT routing rules require using urban bypass routes). Appendix B provides a 

detailed discussion and example of the calculations performed to derive this table.
NUREG-2125 38

Table 2-13 Collective Doses to Residents near Truck Stops (person-Sv) (1 Sv=105 mrem)

Origin Destination Type Persons/km2

(persons/mi2) Number of Stops Dose

MAINE 

YANKEE

ORNL Rural 19.9 (51.5) 1.1 7.4 x10-7

Suburban 395 (1023) 0.93 1.0 x10-5

Deaf Smith Rural 18.6 (48.2) 2.5 1.5 x10-6

Suburban 371 (961) 1.6 1.7 x10-5

Hanford Rural 15.4 (39.9) 4.3 2.2 x10-6

Suburban 325 (842) 1.5 1.4 x10-5

Skull Valley Rural 16.9 (43.8) 3.5 1.9 x10-6

Suburban 333 (861) 1.3 1.2 x10-5

KEWAUNEE

ORNL Rural 19.8 (51.3) 0.81 5.2 x10-7

Suburban 361 (935) 0.59 6.0 x10-6

Deaf Smith Rural 13.5 (35.0) 2.0 8.6 x10-7

Suburban 339 (878) 0.52 5.0 x10-6

Hanford Rural 10.5 (27.2) 3.4 1.2 x10-6

Suburban 316 (818) 0.60 5.4 x10-6

Skull Valley Rural 12.5 (32.4) 2.6 1.1 x10-6

Suburban 325 (840) 0.44 4.1 x10-6

INDIAN POINT

ORNL Rural 20.5 (53.1) 0.71 4.7 x10-7

Suburban 388 (1005) 0.71 7.8 x10-6

Deaf Smith Rural 17.1 (44.3) 2.3 1.3 x10-6

Suburban 370 (958) 1.2 1.3 x10-5

Hanford Rural 13.0 (33.7) 4.1 1.8 x10-6

Suburban 338 (875) 1.1 1.1 x10-5

Skull Valley Rural 14.2 (36.8) 3.3 1.5 x10-6

Suburban 351 (909) 0.93 9.3 x10-6

IDAHO 

NATIONAL 

LAB

ORNL Rural 12.4 (32.1) 3.1 1.3 x10-6

Suburban 304 (787) 0.72 6.3 x10-6

Deaf Smith Rural 7.8 (20.2) 2.3 5.8 x10-7

Suburban 339 (878) 0.35 3.4 x10-6

Hanford Rural 6.5 (16.8) 0.43 9.0x10-8

Suburban 200 (518) 0.57 3.2 x10-6

Skull Valley Rural 10.1 (26.2) 0.42 1.4 x10-7

Suburban 343 (888) 0.11 1.1 x10-6

The rural and suburban population densities in Table 2-13 are averages for the entire route. An 

analogous calculation can be made for each State traversed. However, in neither case can it be 

determined beforehand exactly where the truck will stop to refuel. In some cases (e.g., INL to 

Skull Valley) the truck may not stop at all since the total distance from INL to the Skull Valley 

site is only 466.2 km (290 miles). The route from Indian Point to ORNL illustrates another 

situation. This route is 1,028 km (639 miles) long and would include one truck stop. This stop 

could occur in a rural or suburban area. The results shown in Table 2-13 are general average 

doses at stops.
39 NUREG-2125

2.4 Doses to Workers

Radiation doses to workers are limited in accordance with the regulations in 10 CFR Part 20,

“Standards for Protection against Radiation,” which states maintaining worker exposure to 

ionizing radiation “as low as is reasonably achievable” (ALARA). ALARA applies to 

occupational doses since workers potentially are exposed to much larger doses than the 

general public. For example, the cab of a truck carrying a loaded Truck-DU cask is shielded so 

that 63 percent of the radiation from the end of the cask is blocked.

Occupational doses from routine, incident-free radioactive materials transportation include 

doses to truck and train crew, railyard workers, truck-stop workers, inspectors, and escorts.

Workers not included are those who handle spent fuel containers in storage, load and unload 

casks from vehicles or during intermodal transfer, and attendants who refuel trucks in areas 

where truck refueling stops in the United States no longer have such attendants.16

Table 2-14 summarizes the occupational doses. All doses are reported per hour except for the 

truck stop worker (reported for the maximum truck stop time) and the rail classification yard 

workers. All doses are individual doses (Sv) except for the railyard worker collective doses.

Table 2-14 Occupational Doses and Dose Rates from Routine Incident-Free 

Transportation (1 Sv=105 mrem)

Cask and route 

type

Train crew 

in transit: 

3 people; 

person-Sv/

km

Truck 

crew in 

transit 

2 people; 

personSv/kma

Escort: 

Sv/houra

Inspector: 

Average 

Sv per 8 

inspection

sc

Truck 

stop 

worker: 

Sv per 

stop

Rail 

classification 

yard workers: 

person-Sv 

/stop

Rail-Lead 

rural/suburban 4.3x10-7 5.8x10-6 1.5x10-5

Rail-Lead urban 7.2x10-7 5.8x10-6 b

Rail-Steel 

rural/suburban 3.3x10-7 4.4x10-6 1.1x10-5

Rail-Steel urban 5.5x10-7 4.4x10-6 b

Truck - DU 

rural/suburban 3.8x10-7 4.9x10-9 1.5x10-3 6.7x10-6

Truck - DU 

urban 3.6x10-7 4.9x10-9

a The truck crew is shielded while in transit to sustain a maximum dose of 0.02 mSv/hour b Even classification yards within metropolitan areas do not typically have urban population densities 

because of the large area the classification yard occupies. c The average number of State boundaries crossed for all 16 routes is eight. The average dose to an 

inspector from each of these inspections is 1.64 x 10-4 Sv (0.0164 rem).

Doses to rail crew and rail escorts are similar. Spent fuel may be transported in dedicated trains 

so that both escorts and train crew are assumed to be within a distance of one railcar length of 

 

16 The States of Oregon and New Jersey still require gas station attendants to refuel cars and light duty 

vehicles, but heavy truck crews do their own refueling.
NUREG-2125 40

the railcar carrying the spent fuel. Escorts in the escort car are not shielded because they must 

maintain line-of-sight to the railcar carrying spent fuel. Train crew members are in a crew 

compartment and were assumed to have some shielding, resulting in an estimated dose about 

25 percent less than the escort. The largest collective doses are to railyard workers. The 

number of workers in railyards is not constant and the number of activities that brings these 

workers into proximity with the shipment varies as well. This analysis assumes the dose to the 

worker doing an activity for each activity (e.g., inspection, coupling and decoupling the railcars, 

moving the railcar into position for coupling). The differences between doses in the Rail-Lead 

case and the Rail-Steel case reflect differences in cask dimensions and in external dose rate.

Truck crew members are shielded so that they receive a maximum dose of 2.0×10-5 Sv/hr

(2.0 mrem/hr). This regulatory maximum was imposed in the RADTRAN calculation. Truck 

inspectors generally spend about 1 hour within 1 meter (3.3 feet) of the cargo (Weiner and 

Neuhauser, 1992), resulting in a relatively large dose. An upper bound to the duration of a truck 

refueling stop is about 50 minutes (0.83 hours) (Griego et al., 1996). The truck stop worker 

whose dose is reflected in Table 2-14 is assumed to be outside (unshielded) at 15 meters (50 

feet) from the truck during the stop. Truck stop workers in concrete or brick buildings are

shielded from any radiation.

2.5 Chapter Summary

A summary of the results for the incident-free transport of spent fuel in the three casks analyzed 

in this study are presented in Table 2-15, Table 2-16, and Table 2-17.
41 NUREG-2125

Table 2-15 Total Collective Dose in Person-Sv from Routine Transportation for Each Rail 

Route for the Rail-Lead Cask (1 Sv=105 mrem)

Origin Destination

Residents 

Along 

Route

Occupants 

of Vehicles 

Sharing 

Route

Residents 

near Stop

Railyard 

Crew and 

Escorts

Total

MAINE 

YANKEE

ORNL 2.1x10-4 4.0x10-5 1.1x10-4 3.4x10-4 7.0x10-4

Deaf Smith 2.5x10-4 6.1x10-5 5.3 x10-5 5.1x10-4 8.7x10-4

Hanford 2.9x10-4 9.0x10-5 1.1x10-4 7.6x10-4 1.2x10-3

Skull Valley 2.9x10-4 7.4x10-5 5.4 x10-5

6.2x10-4 1.1x10-3

KEWAUNEE

ORNL 1.3x10-4 2.7x10-5 1.1x10-4 2.3x10-4 5.0x10-4

Deaf Smith 1.1x10-4 3.4x10-5 6.8 x10-5 3.0x10-4 5.1x10-4

Hanford 1.1x10-4 5.2x10-5 1.1x10-4 4.7x10-4 7.4x10-4

Skull Valley 1.4x10-4 4.9x10-5 1.2 x10-4 4.3x10-4 7.4x10-4

INDIAN 

POINT

ORNL 1.6x10-4 2.7x10-5 1.3x10-4

2.1x10-4 5.3x10-3

Deaf Smith 2.0x10-4 5.6x10-5 5.9 x10-5

4.8x10-4 8.0x10-3

Hanford 2.5x10-4 8.5x10-5 1.1x10-4

7.2x10-4 1.2x10-3

Skull Valley 2.4x10-4 6.5x10-5 5.6 x10-5

6.0x10-4 9.3x10-3

INL

ORNL 1.3x10-4 5.7x10-5 9.5 x10-5

5.1x10-4 8.0x10-4

Deaf Smith 6.7x10-5 3.2x10-5 7.7 x10-5

3.1x10-4 4.9x10-4

Hanford 3.6x10-5 1.8x10-5 5.6 x10-5

1.8x10-4 3.0x10-4

Skull Valley 3.0x10-5 8.2x10-6 3.1x10-6

9.5x10-5 1.4x10-4
NUREG-2125 42

Table 2-16 Total Collective Dose in Person-Sv from Routine Transportation for Each Rail 

Route for the Rail-Steel Cask (1 Sv=105 mrem)

Origin Destination

Residents 

Along 

Route

Occupants 

of Vehicles 

Sharing 

Route

Residents 

Near Stop

Railyard 

Crew and 

Escorts

Total

MAINE 

YANKEE

ORNL 1.6x10-4 3.0x10-5 8.5 x 10-5 2.3x10-4 5.1x10-4

Deaf Smith 1.9x10-4 4.6x10-5 5.0 x 10-53.7x10-4 6.7x10-4

Hanford 2.3x10-4 6.8x10-5 8.8 x 10-55.6x10-4 9.5x10-4

Skull Valley 2.2x10-4 5.5x10-5 4.1 x 10-54.5x10-4 7.7x10-4

KEWAUNEE

ORNL 9.6x10-5 2.0x10-5 8.3 x 10-5 1.5x10-4 3.5x10-4

Deaf Smith 8.3x10-5 2.6x10-5 5.2 x 10-5 2.1x10-4 3.7x10-4

Hanford 8.3x10-5 3.9x10-5 8.7 x 10-5 3.3x10-4 5.4x10-4

Skull Valley 1.1x10-4 3.7x10-5 9.1 x 10-5 3.0x10-4 5.4x10-4

INDIAN POINT

ORNL 1.2x10-4 2.0x10-5 1.0 x 10-4 1.4x10-4 3.8x10-4

Deaf Smith 1.5x10-4 4.3x10-5 4.5 x10-5 3.4x10-4 5.8x10-4

Hanford 1.9x10-4 6.5x10-5 8.3 x 10-55.2x10-4 8.6x10-4

Skull Valley 1.8x10-4 4.9x10-5 4.3 x 10-54.4x10-4 7.1x10-4

INL

ORNL 1.0x10-4 4.3x10-5 7.2 x 10-5 3.6x10-4 5.7x10-4

Deaf Smith 5.2x10-5 2.5x10-5 5.8 x10-5 2.1x10-4 3.4x10-4

Hanford 2.8x10-5 1.4x10-5 4.3x 10-5 1.2x10-4 2.0x10-4

Skull Valley 2.2x10-5 6.2x10-6 2.4 x 10-6 5.0x10-5 8.0x10-5

A code that estimates risk is never completely precise because the input data are estimates and 

projections. To account for this imprecision, RADTRAN uses assumptions and values that 

overestimate doses. Actual measurements confirm that RADTRAN overestimates doses by a 

small margin. Therefore, the doses calculated in this chapter should be regarded as 

overestimates.

The individual and collective doses calculated are for a single shipment and, even though 

overestimated, they are uniformly very small. Maximum individual doses are comparable to 

background doses and are less than doses from many medical diagnostic procedures.

Collective doses are orders of magnitude less than the collective background dose, as shown in 

Figure 2-11 for an example shipment from Maine Yankee to ORNL. This route assumes ten 

inspection stops at State boundaries. The NRC recommends that collective doses (average 

doses integrated over a population) only be used for comparisons (NRC, 2008). The proper 

comparison for collective doses is between the background collective dose plus the shipment 

dose and the background dose if there is no shipment. The collective dose, however, is never

zero in the absence of a shipment.
43 NUREG-2125

Table 2-17 Total Collective Dose in Person-Sv from Routine Transportation for Each 

Highway Route for the Truck Cask (1 Sv=105 mrem)

Origin Destination

Residents 

Along 

Route

Occupants 

of 

Vehicles 

Sharing 

Route

Residents 

Near Stop

Persons 

Sharing 

Stop

Crew/ 

Truck 

Stop 

Worker

Total

MAINE 

YANKEE

ORNL 9.6x10-5 4.6x10-4 1.2x10-5 8.6x10-4 6.8x10-4 2.1x10-3

Deaf Smith 1.4x10-4 7.3x10-4 1.8x10-5 9.2x10-4 1.4x10-3 3.2x10-3

Hanford 1.2x10-4 8.3x10-4 1.4x10-5 1.3x10-3 1.9x10-3 4.2x10-3

Skull Valley 1.1x10-4 7.0x10-4 1.4x10-5 1.1x10-3 1.6x10-3 3.5x10-3

KEWAUNEE

ORNL 5.2x10-5 3.3x10-4 6.6x10-6 3.2x10-4 4.9x10-4 1.2x10-3

Deaf Smith 4.7x10-5 3.1x10-4 5.8x10-6 5.7x10-4 8.3x10-4 1.8x10-3

Hanford 5.1x10-5 5.4x10-4 6.6x10-6 9.0x10-4 1.3x10-3 2.9x10-3

Skull Valley 3.9x10-5 3.8x10-4 5.1x10-6 6.8x10-4 1.0x10-3 2.2x10-3

INDIAN POINT

ORNL 6.9x10-5 4.6x10-4 8.3x10-6 3.2x10-4 4.9x10-4 1.3x10-3

Deaf Smith 1.4x10-4 6.9x10-4 1.4x10-5 7.9x10-4 1.2x10-3 2.9x10-3

Hanford 8.9x10-5 7.2x10-4 1.2x10-5 1.2x10-3 1.7x10-3 3.9x10-3

Skull Valley 7.7x10-5 6.6x10-4 1.1x10-5 9.5x10-4 1.4x10-3 3.1x10-3

INL

ORNL 6.3x10-5 5.0x10-4 7.5x10-6 8.6x10-4 1.3x10-3 2.7x10-3

Deaf Smith 3.7x10-5 3.4x10-4 4.0x10-6 6.0x10-4 8.8x10-4 1.9x10-3

Hanford 1.4x10-5 2.0x10-4 1.1x10-6 2.3x10-4 3.7x10-4 8.5x10-4

Skull Valley 1.2x10-5 8.5x10-5 1.2x10-6 1.2x10-4 1.8x10-4 1.6x10-3
NUREG-2125 44

Figure 2-11 Collective doses from background and from Maine Yankee to ORNL truck 

shipments of spent nuclear fuel (person-Sv) (1 Sv=105 mrem)
45 NUREG-2125

3. CASK RESPONSE TO IMPACT ACCIDENTS

3.1 Introduction

Spent fuel casks are required to be accident resistant. During the U.S. Nuclear Regulatory 

Commission (NRC) certification process the cask designer must demonstrate, among other 

things, that the cask would survive a free fall from a height of 9 meters (30 feet) falling onto a 

flat, essentially unyielding, target in the orientation most likely to damage the cask 

(10 CFR 71.73, “Hypothetical Accident Conditions”). The NRC’s required high standards and 

conservative approaches for this demonstration include the use of conservative (usually 

minimum) material properties in analyses, allowing only small amounts of yielding, and the use 

of materials with high ductility. These approaches ensure that the casks not only will survive 

impacts at the speed created because of the 9-meter drop but will also survive much higher 

speed impacts.

In addition to the conservative designs that the certification process ensures, two additional 

requirements of the 9-meter drop provide safety when compared to actual accidents. The first 

requirement is that the impact must be onto an essentially unyielding target. This implies that 

the cask will absorb all of the kinetic energy of the impact and the target will absorb none. For 

impacts onto real surfaces, both the cask and the target absorb the kinetic energy. The second 

requirement is that the vertical impact must be onto a horizontal target. This requirement 

ensures that at some point during the impact, the velocity of the cask will be zero, and all of the 

kinetic energy is converted into strain energy (i.e., absorbed by the cask). Most real accidents 

occur at an angle, and the kinetic energy of the cask is absorbed by multiple impacts instead of 

one impact. This chapter discusses these three aspects.

3.2 Finite Element Analyses of Casks

Previous risk studies have used generic casks. The Modal Study (Fischer et al., 1987) 

assumed that any accident more severe than the regulatory hypothetical impact accident would 

lead to a cask release. In NUREG/CR-6672, “Re-examination of Spent Fuel Shipment Risk 

Estimates,” (Sprung et al., 2000), the impact limiters of the generic casks were assumed to be 

unable to absorb more energy than the amount from the regulatory hypothetical impact accident 

(i.e., a 9-meter (30-foot) free fall onto an essentially rigid target). Modeling limitations at the 

time of the studies required both of these assumptions. In reality, casks and impact limiters 

have excess capacity to resist impacts. In the current study, three NRC-certified casks were 

used instead of generic casks, and the actual impact resistance capability of those cask designs 

was included in the analyses. However, for the truck cask no new finite element (FE) analyses

were performed. The current study relied upon analyses performed for other studies, some of 

which used a generic truck cask.

The response to impacts of 48 kph, 97 kph,145 kph, and 193 kph (equal to 30 mph, 60 mph, 

90 mph, and 120 mph) onto an unyielding target in the end, corner, and side orientations for the 

Rail-Steel and Rail-Lead spent fuel transportation casks were determined using the nonlinear 

transient dynamics explicit FE code PRESTO (SIERRA, 2009). PRESTO is a Lagrangian code, 

using a mesh that follows the deformation to analyze solids subjected to large, suddenly applied 

loads. The code is designed for a massively parallel computing environment and for problems 

with large deformations, nonlinear material behavior, and contact. PRESTO has a versatile 

element library that incorporates both continuum (3D) and structural elements, such as beams 

and shells. The results from analyses using this type of code have been compared to results 
NUREG-2125 46

from both regulatory and high-speed impact tests. A recent Safety Analysis Report Addendum 

for the PAT-1 air transport package compared the very large deformations seen in full-scale 

testing of this package to those calculated using nonlinear explicit dynamics (Yoshimura et al., 

2010). There have also been comparisons between full-scale regulatory drop tests of two spent 

fuel casks in Germany with explicit dynamic finite element analyses (Kishimoto et al., 2007, 

Musolff et al., 2007).

In addition to the detailed analyses of rail casks performed for this study, the response of the 

Truck-DU (depleted uranium) spent fuel transportation cask was inferred based on the FE

analyses performed for the generic casks in NUREG/CR-6672. The direction of the cask travel 

was perpendicular to the surface of the unyielding target in all of the analyses performed.

Figure 3-1 is a pictorial representation of the three impact orientations analyzed. In all of the 

analyses, the spent fuel basket and fuel elements were treated as a uniform homogenous 

material. The density of this material was adjusted to achieve the correct weight of the loaded 

basket. The overall behavior of the material was conservative (i.e., because it acts as a single 

entity that affects the cask all at once instead of many smaller parts that affect the cask over a 

longer period of time) for assessing the effect the cask contents had on the behavior of the 

cask. A sub-model of a single assembly was used to calculate the detailed response of the fuel 

assemblies.

End Corner Side

Figure 3-1 Impact orientations analyzed

3.2.1 Rail-Steel Cask

Finite Element Model

Figure 3-2 shows the overall FE model of the Rail-Steel cask depicted in Figure 1-3. This cask 

has steel gamma-shielding material and transports 24 pressurized-water reactor (PWR)

assemblies in a welded multipurpose canister (MPC). The impact limiters on each end of the 

cask are designed to absorb the kinetic energy of the cask during the regulatory hypothetical 

impact accident. They are made of an interior stainless steel support structure, an aluminum 

honeycomb energy absorber, and a stainless steel skin. Figure 3-3 shows the FE mesh of the 

closure end impact limiter. The one on the other end of the cask differs only in how it is 

attached to the cask. The aluminum honeycomb has direction-dependent properties. The 
47 NUREG-2125

strong direction of the honeycomb is oriented in the primary crush direction, requiring the FE

model to include the individual blocks of honeycomb material, rather than a single material for 

the entire impact limiter. The cask has a single solid steel lid attached with fifty-four 1-5⁄8-inch 

diameter bolts and sealed with dual metallic O-rings. Figure 3-4 shows the FE mesh of the 

closure bolts (bolts used to attach the closure end impact limiter are also shown) and the level 

of mesh refinement included in these important parts. Appendix C provides details of the FE

models, including material properties, contact surfaces, gaps, and material failure.

Figure 3-2 Finite element mesh of the Rail-Steel cask
NUREG-2125 48

Impact limiter showing the various blocks of honeycomb

Impact limiter with the honeycomb removed to reveal the inner support structure

Figure 3-3 Details of the finite element mesh for the impact limiters of the Rail-Steel cask
49 NUREG-2125

Figure 3-4 Finite element mesh of the Rail-Steel closure bolts and the closure end impact 

limiter attachment bolts

The highly refined mesh in these critical parts ensures an accurate assessment of the closure 

response.

Analysis results

As expected, for all end, corner, and side impacts of the 48 kph (30 mph) impact analyses—the 

impact velocity from the regulatory hypothetical impact accident—the impact limiter absorbed 

almost all of the cask’s kinetic energy and there was no damage (i.e., permanent deformation) 

to the cask body or canister. As the impact velocity increases, additional damage to the impact 

limiter occurs for all orientations because it absorbs more kinetic energy. This shows the 

margin of safety in the impact limiter design. At 97 kph (60 mph) there is still no significant 

damage to the cask body or canister. At an impact speed of 145 kph (90 mph), damage to the 

cask and canister begins. The impact limiter has absorbed all the kinetic energy it can, and any 

additional kinetic energy must be absorbed by plastic deformation in the cask body.

For the side impact at 145 kph (90 mph), several lid bolts fail in shear but the lid remains 

attached. At this point, the metallic seal no longer maintains the leak-tightness of the cask, but 

the spent fuel remains contained within the welded canister. Even at the highest impact speed

of 193 kph (120 mph), the welded canister remains intact for all orientations. Figure 3-5 shows 

the deformed shape and plastic strain in the canister for the 193 kph (120 mph) impact in a side 

orientation. This case has the most plastic strain in the canister. The peak value of plastic 

strain in this case is 0.7. This value is specified by the equivalent plastic strain (EQPS), which is

a representation of the magnitude of local permanent deformation. The canister’s stainless 

steel material can easily withstand plastic strains greater than 1 (Blandford et al., 2007). These 

results demonstrate that no impact accident will lead to release of material from the Rail-Steel 

canister. Appendix C includes similar figures for the other orientations and speeds and criteria 

for the failure model.
NUREG-2125 50

Figure 3-5 Plastic strain in the welded canister of the Rail-Steel for the 193-kph 

(120-mph) side impact case

3.2.2 Rail-Lead Cask

Finite Element Model

Figure 3-6 shows the overall FE model of the Rail-Lead cask depicted in Figure 1-2. This cask 

has lead gamma-shielding material and transports either 26 directly-loaded PWR assemblies or 

24 PWR assemblies in a welded MPC. The impact limiters at each end of the cask are 

designed to absorb the cask’s kinetic energy during the regulatory hypothetical impact accident.

The impact limiters are made of redwood and balsa wood energy-absorbing material and a 

stainless steel skin. Figure 3-7 shows the FE mesh of the closure end impact limiter (the impact 

limiter on the other end of the cask is identical). The cask has a dual lid system. The inner lid is 

attached with 42 38-millimeter (mm) (1.5-inch) diameter bolts and sealed with dual elastomeric 

O-rings if the cask is only used for transportation and metallic O-rings if the cask is used for 

storage before transportation. The outer lid is attached with 36 25-mm (1-inch) diameter bolts 

and sealed with a single elastomeric O-ring if the cask is only used for transportation and a 

metallic O-ring if the cask is used for storage before transportation. Figure 3-8 shows the FE

mesh of the closure bolts and the level of mesh refinement included in these important parts.

Appendix C includes details of the FE models.
51 NUREG-2125

Figure 3-6 Finite element mesh of the Rail-Lead cask
NUREG-2125 52

Impact limiter showing the two different types of wood. The yellow is balsa and the red 

is redwood.

Impact limiter with the wood removed to reveal the inner attachment bolts

Figure 3-7 Details of the finite element mesh for the impact limiters of the Rail-Lead cask
53 NUREG-2125

Figure 3-8 Finite element mesh of the Rail-Lead closure bolts for both the inner and 

outer lids

The longer bolts are for the inner lid and the shorter ones for the outer lid.

Analysis Results

The impact limiter absorbed almost all of the kinetic energy of the cask for the 48 kph impact 

analyses—the impact velocity from the regulatory hypothetical impact accident—and no 

damage to the cask body occurred. The response of the Rail-Lead cask was more complicated

than that of the Rail-Steel cask. For the end orientation, as the impact velocity increased,

initially there was additional damage to the impact limiter because it was absorbing more kinetic 

energy, which shows the margin of safety in the impact limiter design. There is no significant 

damage to the cask body at 97 kph (60 mph). At an impact speed of 145 kph (90 mph),

damage to the cask begins. The impact limiter has absorbed all the kinetic energy it can and 

any additional kinetic energy is absorbed by plastic deformation in the cask body. At this speed 

there is significant slumping of the lead gamma shielding material, resulting in a loss of lead 

shielding near the end of the cask away from the impact point. As the impact velocity is 

increased to 193 kph (120 mph), the lead slump becomes more pronounced and there is 

enough plasticity in the lids and closure bolts to result in a possible loss of sealing capability.

For the directly loaded cask (without a welded multipurpose canister (MPC)) there could be 

some loss of radioactive contents if the cask has metallic seals. This would not be the case if 

the cask has elastomeric seals. A more detailed discussion of leakage is provided later in this 

section. Figure 3-9 shows the deformed shape of the Rail-Lead cask following the 193 kph (120

mph) impact in the end-on orientation. The amount of lead slump from this impact is 35.5 cm

(14.0 in) and the area without lead shielding is visible in Figure 3-9. Table 3-1 gives the amount 

of lead slump in each of the analysis cases.
NUREG-2125 54

Figure 3-9 Deformed shape of the Rail-Lead cask following the 193-kph (120-mph) impact 

onto an unyielding target in the end-on orientation

Table 3-1 Maximum Lead Slump for the Rail-Lead Cask from Each Analysis Casea

Speed,

kph (mph)

Max. Slump End,

cm (in)

Max. Slump Corner,

cm (in)

Max. Slump Side,

cm (in)

48 (30) 0.64 (0.25) 0.17 (0.065) 0.01 (0.004)

97 (60) 1.83 (0.72) 2.51 (0.99) 0.14 (0.054)

145 (90) 8.32 (3.28) 11.45 (4.51) 2.09 (0.82)

193 (120) 35.55 (14.00) 31.05 (12.22) 1.55 (0.61)

a The measurement locations for each impact orientation are given in Appendix C.

For corner impacts at 97 kph (60 mph) and 145 kph (90 mph), there is some damage to the 

cask body and deformation of the impact limiter, which results in lead slump and closure bolt 

deformation. The amount of closure deformation in these two cases is not sufficient to cause a 

leak if the cask is sealed with elastomeric O-rings, but it is enough to cause a leak if the cask is 

sealed with metallic O-rings. For a corner impact at 193 kph (120 mph), there is more 

significant deformation to the cask, more lead slump, and a larger gap between the lid and the 

cask body. Figure 3-10 shows the deformed shape of the cask for this impact analysis. The 

Bulging of Lead
55 NUREG-2125

deformation in the seal region is sufficient to cause a leak if the cask has metallic O-rings but 

not if it has elastomeric O-rings. The maximum amount of lead slump is 31 cm (12 inches).

Figure 3-10 Deformed shape of the Rail-Lead cask following the 193-kph (120-mph)

impact onto an unyielding target in the corner orientation

In the side impact, as the impact velocity increases from 48 kph (30 mph) to 97 kph (60 mph), 

the impact limiter ceases to absorb additional energy and there is permanent deformation of the 

cask and closure bolts. The resulting gap in between the lids and the cask body is sufficient to 

allow leakage if there is a metallic seal, but not if there is an elastomeric seal. This gap 

calculation between the cask body and lid is conservative because the clamping force applied 

by bolt preload was neglected in the analysis (i.e., the clamping force acts to keep the lid and 

cask body together). When the impact speed is increased to 145 kph (90 mph), the amount of 

damage to the cask increases significantly. In this case, many bolts from the inner and outer lid 

fail in shear and there is a gap between each of the lids and the cask. This gap is sufficient to 

allow leakage if the cask is sealed with either elastomeric or metallic O-rings.

Figure 3-11 shows the deformed shape of the cask following this impact. The response of the 

cask to the 193 kph (120 mph) impact is similar to that from the 145 kph (90 mph) impact, 
NUREG-2125 56

except that the gaps between the lids and the cask are larger. Appendix C shows the deformed 

shapes for all of the cases analyzed.

Figure 3-11 Deformed shape of the Rail-Lead cask following the 145-kph (90-mph) impact 

onto an unyielding target in the side orientation

Leak Area

The certificate of compliance (COC) for the Rail-Lead cask allows transportation of spent fuel in 

three different configurations. The analyses conducted for this study were all direct-loaded fuel 

cases, but the results can be applied to cases with an internal canister. The impact limiter and 

cask body are the same for that case. The addition of the internal canister adds strength and 

stiffness to the cask in the closure region because it has a 203-mm (8-in) thick lid that will inhibit 

the rotation of the cask wall and reduce any gaps between the closure lids and the cask.

Figure 3-12 shows the deformation of the closure region for the 193 kph (120 mph) end impact.

Gaps for the outer lid were measured as the shortest distance from Node A to the surface 

opposite it and gaps for the inner lid were measured as the shortest distance from Node B to the 

surface opposite it. None of the analyses show sufficient deformation into the interior volume of 

the cask to cause a failure of the internal welded canister. Therefore, as with the Rail-Steel

cask, if the spent fuel is transported in an inner welded canister, there would be no release from 

any of the impacts.

In cases without an inner canister, the cask can be used for dry spent-fuel storage before 

shipment or to transport fuel removed from pool storage and immediately shipped. In the first of 

these two cases, metallic O-rings provide the seal between the lids and the cask body. This 

type of seal is less tolerant to movement between the lids and the cask and a closure opening 

greater than 0.25 mm will cause a leak. If the cask is used for direct shipment of spent fuel, 

elastomeric O-rings provide the seal between the lids and the cask body. While no tests of the 

effect of gap on leak rates for the lids of this cask have been performed, it is assumed that this 

type of seal can withstand closure openings of 2.5 mm (0.10 in) without leaking (Sprung et al.,

2000).

Note the gaps between the 

lids and the cask body
57 NUREG-2125

Figure 3-12 Measurement of closure gaps

Table 3-2 gives the calculated axial gap in each analysis and the corresponding leak area for 

both metallic and elastomeric seals. The leak areas are calculated for the lid with the smaller 

gap because if any leakage from the cask is to occur, both lids must leak.
NUREG-2125 58

Table 3-2 Available Areas for Leakage from the Rail-Lead Cask

Orientation Speed,

kph

(mph)

Location Lid Gap,

mm (in)

Seal Type Hole Size,

mm2 (in2)

End

48

(30)

Inner 0.226 (0.0089) Metalb none

Outer 0 Elastomer none

97

(60)

Inner 0.056 (0.0022) Metal none

Outer 0.003 (0.00012) Elastomer none

145

(90)

Inner 2.311 (0.091) Metal none

Outer 0.047 (0.00185) Elastomer none

193

(120)

Inner 5.588 (0.220) Metal 8796 (13.63)

Outer 1.829 (0.072) Elastomer none

Corner

48

(30)

Inner 0.094 (0.0037) Metal none

Outer 0.089 (0.0035) Elastomer none

97

(60)

Inner 0.559 (0.022) Metal 65 (0.10)

Outer 0.381 (0.015) Elastomer none

145

(90)

Inner 0.980 (0.0386) Metal 599 (0.928)

Outer 1.448 (0.057) Elastomer none

193

(120)

Inner 2.464 (0.097) Metal 1716 (2.660)

Outer 1.803 (0.071) Elastomer none

Side

48

(30)

Inner 0.245 (0.0096) Metal none

Outer 0.191 (0.0075) Elastomer none

97

(60)

Inner 0.914 (0.036) Metal 799 (1.24)

Outer 1.600 (0.063) Elastomer none

145

(90)

Inner 8a (0.3) Metal >10000 (>16)

Outer 25a (1) Elastomer >10000 (>16)

193

(120)

Inner 15a (0.6) Metal >10000 (>16)

Outer 50a (2) Elastomer >10000 (>16)

a Estimated. The method used to calculate the gaps for the other cases is explained in Appendix C. For 

these cases, there was bolt failure and the gap was too large to measure using the standard method, but 

the resultant leak area is sufficiently large that any change to it would not change the cask-release 

fraction. b The metal seal for the Rail-Lead cask is installed only when the cask has been used for dry storage prior 

to transportation. Currently, none of these casks are used for dry storage and there are no plans for 

using them that way in the future.

3.2.3 Truck-DU Cask

Detailed FE analyses of the Truck-DU cask were not performed for this study because the 

response of the truck casks in NUREG/CR–6672 indicated there were no gaps between the lid 

and the cask body at any impact speed. Therefore, the results discussed here are based on the 

FE analysis of the generic steel-DU-steel truck cask performed for NUREG/CR–6672. In 

general, results from the analyses performed for this study confirm that the analyses performed 
59 NUREG-2125

for NUREG/CR–6672 were conservative (see Table 3-3); therefore, the results discussed below 

are likely to be an overestimate of the damage to the Truck-DU cask from severe impacts.

Figure 3-13 shows the deformed shape and plastic strain contours for the generic 

steel-DU-truck cask from Appendix A to NUREG/CR–6672 (Figures A-15, A-19, and A-22).

None of the impacts caused strains great enough to fail the cask wall, and in all cases the 

deformation in the closure region was insufficient to cause seal failure.

Table 3-4 (extracted from Table 5.6 of NUREG/CR–6672) provides the deformation in the seal 

region for each case. There would be no release of radioactive contents in any of these cases.
NUREG-2125 60

Table 3-3 Comparison of Analyses between This Study and NUREG/CR-6672

Item/Cask Rail-Steel 6672 Monolithic Steel

Deformed 

Shape 

145 kph

(90 mph)

(Figure A-35 of NUREG/CR-6672)

Failed 

Bolts

No Yes

Item/Cask Rail-Lead 6672 SLS Rail

Deformed 

Shape 

145 kph

(90 mph)

(Figure A-24 of NUREG/CR-6672)

Gap Size Inner Lid - 0.980 mm (0.039 in)

Outer Lid – 1.448 mm (0.057 in)

6.096 mm (0.240 in)

Failed 

Bolts

No Yes
61 NUREG-2125

 

Figure 3-13 Deformed shapes and plastic strains in the generic steel-DU-steel truck cask 

from NUREG/CR-6672 (impact limiter removed) following 193-kph (120-mph) impacts in 

the (clockwise from top left) end-on, center of gravity (CG)-over-corner, and side-on 

orientation
NUREG-2125 62

Table 3-4 Deformation of the Closure Region of the Steel-DU-Steel Truck Cask from 

NUREG/CR-6672, mm (in)

Cask

Analysis

Velocity

Corner Impact End Impact Side Impact

Opening Sliding Opening Sliding Opening Sliding

Steel-DU-Steel 

Truck

48 kph

30 mph

0.508

(0.02)

1.778

(0.07)

0.127-0.305

(0.005-0.012)

0.025-0.127

(0.001-0.005)

0.254

(0.01)

0.508

(0.002)

97 kph

60 mph

2.032

(0.08)

1.778

(0.07)

0.254-0.508

(0.01-0.02)

0.076-0.152

(0.003-0.006)

0.254

(0.01)

0.254

(0.01)

145 kph

90 mph

0.508

(0.02)

2.540

(0.1)

- - 0.254

(0.01)

0.508

(0.02)

193 kph

120 mph

0.762

(0.03)

3.810

(0.15)

0.330

(0.013)

0.762

(0.03)

0.102

(0.004)

0.508

(0.02)

3.3 Impacts onto Yielding Targets

The analysis results discussed in Section 3.2 were for impacts onto an unyielding, essentially 

rigid, target. All real-impact accidents involve targets that yield to some extent. When a cask 

impacts a real target, the amount of impact energy the target and cask absorb depends on the 

relative strength and stiffness of the two objects. For an impact onto a real target to produce the 

same amount of damage as the impact onto an unyielding target, the force applied to the cask 

has to be the same. If the target is not capable of sustaining that level of force, it cannot 

produce the corresponding level of cask damage.

For the Rail-Lead cask (the only one of the three investigated in this study with any release), the 

peak force associated with each impact analysis performed is supplied in Table 3-5. In this 

table, the cases with non-zero hole sizes from Table 3-4 have bold text. It can be seen that in 

order to produce sufficient damage for the cask to release any material, the yielding target has 

to be able to apply a force to the cask greater than 146 million Newtons (MN), or 33 million 

pounds. Very few real targets are capable of applying this amount of force. A hard rock is the 

closest thing to an unyielding target. In this study, hard rock is defined as rock that requires 

blasting operations to remove. While not all classes of this type of rock are equally strong, all of 

them are assumed to absorb negligible energy during an impact; therefore, they are treated as 

rigid.

If the cask hits a flat target, such as the ground, roadway, or railway, it will penetrate into the 

surface. The greater the contact force between the cask and the ground, the greater the 

penetration depth. Figure 3-14 shows the relationship between penetration depth and force for 

the Rail-Lead cask impacting onto hard desert soil in an end-on orientation. As the cask 

penetrates the surface, some of its kinetic energy is absorbed by the surface. The amount of 

energy the target absorbs is equal to the area underneath the force versus the penetration 

curve seen in Figure 3-14. For example, the end impact at 97 kph (60 mph) onto an unyielding 

target requires a contact force of 124 MN (27.9 x 106 pounds). A penetration depth of 

approximately 2.2 meters (7.2 feet) will cause the soil to exert this amount of force. The soil 

absorbs 142 million Joules (MJ) (105 x 106 foot pounds) of energy when penetrated to this 

depth. Adding the energy absorbed by the soil to the 41 MJ (30 x 106 foot pounds) of energy 

absorbed by the cask yields a total absorbed energy of 183 MJ (135 x 106 foot pounds). For the 

cask to have this amount of kinetic energy, it would have to be traveling at 205 kph (127 mph). 

Therefore, a 205 kph (127 mph) impact onto hard desert soil causes the same amount of 
63 NUREG-2125

damage as a 97 kph (60 mph) impact onto an unyielding target. A similar calculation can be 

performed for other impact speeds, orientations, and target types.

Figure 3-14 Force generated by the Rail-Lead cask penetrating hard desert soil

Table 3-6 provides the resulting equivalent velocities. Similar to Table 3-5, the cases resulting

in non-zero hole sizes are identified in bold text. Where the calculated velocity is more than 250 

kph (155 mph), the value in the table is listed as “>250 (>155).” No accident velocities are more 

than this. The concrete target used is a 23-cm-thick slab on engineered fill, which is typical of 

many concrete roadways and concrete retaining walls adjacent to highways. Appendix C

contains details on the calculation of equivalent velocities.

Table 3-5 Peak Contact Force for the Rail-Lead Cask Impacts onto an Unyielding Target 

Orientation Speed,

kph (mph)

Accel.

(g)

Contact Force 

(Millions of 

Pounds)

Contact Force

(MN)

End

48 (30) 58.5 14.6 65.0

97 (60) 111.6 27.9 123.9

145 (90) 357.6 89.3 397.1

193 (120) 555.5 138.7 616.8

Corner

48 (30) 36.8 9.2 40.9

97 (60) 132.2 33.0 146.8

145 (90) 256.7 64.1 285.1

193 (120) 375.7 93.8 417.2

Side

48 (30) 76.1 19.0 84.5

97 (60) 178.1 44.5 197.8

145 (90) 411.3 102.7 456.7

193 (120) 601.1 150.0 667.4

Note: bold numbers are for the cases where there may be seal leaks
NUREG-2125 64

Table 3-6 Equivalent Velocities for Impacts onto Various Targets with the Rail-Lead

Cask, kph (mph)

Orientation Rigid (or hard rock) Soil Concrete

End

48 (30) 102 (63) 71 (44)

97 (60) 205 (127) 136 (85)

145 (90) >250 (>155) >250 (>155)

193 (120) >250 (>155) >250 (>155)

Corner

48 (30) 73 (45) 70 (43)

97 (60) 236 (147) 161 (100)

145 (90) >250 (>155) >250 (>155)

193 (120) >250 (>155) >250 (>155)

Side

48 (30) 103 (64) 79 (49)

97 (60) 246 (153) 185 (115)

145 (90) >250 (>155) >250 (>155)

193 (120) >250 (>155) >250 (>155)

3.4 Effect of Impact Angle

The regulatory hypothetical impact accident requires the cask’s velocity to be perpendicular to 

the impact target. All of the analyses were conducted with this type of impact. During transport, 

the usual scenario is that the velocity is parallel to the nearby surfaces, and therefore, most 

accidents that involve impact with surfaces occur at a shallow angle. This is not necessarily 

true, however, for impacts with structures or other vehicles.

Accident databases do not include impact angle as one of their parameters, so there is no 

information on the relative frequency of impacts at various angles. Given that vehicles usually 

travel parallel to the nearby surfaces, for this study a triangular distribution of impact angles was 

used. Figure 3-15 shows the assumed step-wise distribution of impact angle probabilities. For 

impacts onto hard targets, which are necessary to damage the cask, the component of the 

velocity that is parallel to the impact surface has very little effect on the amount of damage to 

the cask. This requires the accident speed to be higher for a shallow angle impact than a 

perpendicular one to achieve the same amount of damage. Figure 3-16 depicts an example of 

an impact at a shallow angle and the component of the velocity perpendicular to the surface.

Table 3-7 provides the cumulative probability of exceeding an impact angle range and the 

accident speeds required to have the velocity component in the direction perpendicular to the 

target.

Using the information from Table 3-6 and Table 3-7 along with the event trees in Appendix E 

and the assumptions that half of the impacts into tunnels are hard rock surfaces and half are 

concrete leads to the result that 99.95 percent of all rail impact accidents are less severe than 

the regulatory hypothetical accident of 10 CFR 71.73, “Hypothetical Accident Conditions”.
65 NUREG-2125

Figure 3-15 Probability distribution for impact angles

Figure 3-16 Influence of impact angle on effective velocity
NUREG-2125 66

Table 3-7 Accident Speeds That Result in the Same Damage as a Perpendicular Impact, 

kph (mph)

Angle Prob.

Cum. 

Prob.

VAcc so 

VPerp = 48 kph

(30mph)

VAcc so 

VPerp = 97 kph

(60 mph)

VAcc so 

VPerp = 145 kph

(90 mph)

VAcc so 

VPerp = 193 kph

(120 mph)

0 - 10 0.2000 1.0000 278 (173) 556 (345) 834 (518) 1112 (691)

10 - 20 0.1778 0.8000 141 (88) 282 (175) 423 (263) 565 (351)

20 - 30 0.1556 0.6222 97 (60) 193 (120) 290 (180) 386 (240)

30 - 40 0.1333 0.4667 75 (47) 150 (93) 225 (140) 300 (186)

40 - 50 0.1111 0.3333 63 (39) 126 (78) 189 (117) 252 (157)

50 - 60 0.0889 0.2222 56 (35) 111 (69) 167 (104) 223 (139)

60 - 70 0.0667 0.1333 51 (32) 103 (64) 154 (96) 206 (128)

70 - 80 0.0444 0.0667 49 (30.4) 98 (61) 147 (91) 196 (122)

80 - 90 0.0222 0.0222 48 (30) 97 (60) 145 (90) 193 (120)

3.5 Impacts with Objects

The preceding sections dealt with impacts onto flat surfaces, but a large number of impacts 

occur on surfaces that are not flat. These include impacts into columns and other structures, 

impacts by other vehicles, and, more rarely, impacts by collapsing structures. These types of 

impacts were not explicitly included in this study, but recent work by Sandia National 

Laboratories (NRC, 2003a; Ammerman and Gwinn, 2004; Ammerman et al., 2005) has shown 

the GA-4 cask response to some of these impacts. The result of an impact into a large, 

semi-circular, rigid column is shown in Figure 3-17 (NRC, 2003a). While this impact led to 

significant permanent deformation of the cask, the level of strain was not high enough to cause 

tearing of the containment boundary and there was no permanent deformation in the closure 

region and no loss of containment.

Figure 3-17 Deformations to the GA-4 truck cask after a 97-kph (60-mph) side impact 

onto a rigid semi-circular column

Figure source: (from NRC, 2003b)

Collision by a railroad locomotive could potentially cause cask damage and is probably the most 

severe type of collision with another vehicle that could occur. Ammerman et al. (2005) 

investigated several different scenarios of this type of collision. The overall configuration of the 

general analysis case is shown in Figure 3-18. Most trains involve more locomotives and 

trailing cars than used in this analysis, but additional train mass has little effect on the force 

acting on the cask. The impact duration is short and the coupling between the cars is flexible, 

so the impact is over before the inertia of more cars can influence it. Variations on the general 

configuration included the most common locomotive scenarios: (1) having a level crossing
67 NUREG-2125

where the truck tires and locomotive wheels are at the same elevation, (2) having a raised 

crossing where the bottom of the trailer’s main beams are at the same elevation as the top of 

the tracks, and (3) having a skewed crossing so the impact is at 67 degrees instead of at 

90 degrees. For all analyses, the truck was assumed to be stopped and train velocities were 

considered to be 113 kph (70 mph) and 129 kph (80 mph).

Figure 3-18 Configuration of locomotive impact analysis 

Figure source: (Ammerman et al., 2005)

None of the analyses led to deformations that would cause a release of radioactive material 

from the cask or resulted in cask accelerations high enough for the fuel rod cladding to fail. 

Figure 3-19 shows a sequence of the impact. The front of the locomotive is severely damaged 

and the trailer is totally destroyed, but there is very little deformation of the cask—only minor 

denting where the collision posts of the locomotive hit the cask.

The collapse of a bridge onto a cask occurs less frequently, but it also has the potential to 

damage a cask. This type of accident occurred when an elevated portion of the Nimitz Freeway 

collapsed during the Loma Prieta earthquake near San Francisco on October 17, 1989. This 

scenario was analyzed to determine if it would cause a release of spent fuel from the GA-4 truck 

cask (Ammerman and Gwinn, 2004). The analysis assumed that the cask was lying directly on 

the roadway (negating the cushioning effect of the trailer and impact limiters) and a main beam 

of the elevated freeway fell and hit the middle of the cask. Stresses in the cask and damage to 

the beam are shown in Figure 3-20. As in the other analyses for impacts with objects, no loss of 

containment would occur from this accident.
NUREG-2125 68

Figure 3-19 Sequential views of a 129-kph (80-mph) impact of a locomotive into a GA-4 

truck cask 

Figure source: (Ammerman et al., 2005)

Figure 3-20 Results of a finite element simulation of an elevated freeway collapse onto a 

GA-4 spent fuel cask 

Figure source: (Ammerman and Gwinn, 2004), 270 ips = 15.3 mph = 24.7 kph
69 NUREG-2125

3.6 Response of Spent Fuel Assemblies

The FE analyses of the casks in this study did not include the individual components of the 

spent fuel assemblies. Instead, the total mass of the fuel and its support structure were 

combined into an average material. A detailed model of a spent fuel assembly was developed 

to determine the response of individual components (Kalan et al., 2005). Figure 3-21 shows this 

model. In the figure, the fuel rods are shown in yellow, the guide tubes in green, the spacer 

grids in red, the end plates in light blue, and the impact surface in dark blue. The loads 

associated with a 100 g17 cask impact in a side orientation were then applied to this detailed 

model. Kalan et al., 2005, only analyzed the side impact of spent fuel assemblies because the 

strains associated with the rods buckling during an end impact are limited by the constrained 

lateral deformations the basket provides. The side impact results in forces in each fuel rod at 

their supports and in many of the fuel rods midway between the supports where they impact on 

the rods above or below them. A detailed FE model used to determine the response of the rod 

with the highest loads is shown in Figure 3-22. There is slight yielding of the rod at each 

support location and slightly more yielding where the rods impact each other.

Figure 3-21 Finite element model of a PWR fuel assembly

Figure 3-23 shows the maximum plastic strain at each location. The largest of these strains is 

slightly below 2 percent, which is half the plastic strain capacity of irradiated zircaloy at the 

maximum burnup allowed in the Rail-Lead cask (45,000 MWd/MTU) (Sanders et al., 1992); 

therefore, the fuel rods will not crack. The peak acceleration of the cask would have to be 

above 200 g for the cladding to fail. The only impacts severe enough to crack the rods are 

those with impact speeds onto an essentially unyielding target of 145 kph (90 mph) or higher.

Appendix C includes a detailed description of the fuel assembly modeling.

 

17 g refers to the acceleration due to gravity. A 100 g impact results in a deceleration of the cask equal to 100 

times the acceleration due to gravity.
NUREG-2125 70

Figure 3-22 Detailed finite element model of a single fuel rod

Figure 3-23 Maximum strains in the rod with the highest loads

3.7 Chapter Summary

Detailed FE analyses performed for two spent fuel transportation rail casks indicate that casks 

are very robust structures capable of withstanding almost all impact accidents without release of 

radioactive material. In fact, when spent fuel is transported within an inner welded canister or in 

a truck cask, no impacts result in release. Even the rail cask without an inner welded canister 

can withstand impacts much more severe than the regulatory impact without releasing any 

material.

The analyses in this chapter and the event trees in Chapter 5 combine to show that 99.95

percent of the impact accidents are less severe than the regulatory hypothetical accident of 

10 CFR 71.73.
71 NUREG-2125

In the worst orientation (i.e., side impact), an impact speed onto a rigid target at more than 

97 kph (60 mph) is required to cause seal failure in a rail cask. If the cask has an inner welded 

canister, even this impact will not lead to a release of radioactive material. A 97 kph side 

impact onto a rigid target produces a force of approximately 200 MN (45 million pounds) and is 

equivalent to a 185 kph (115 mph) impact onto a concrete roadway or abutment, or a 246 kph 

(153 mph) impact onto hard soil. For impacts onto hard rock, which may be able to resist these 

large forces, impacts at angles less than 30 degrees require a speed of more than 193 kph

(120 mph) to be equivalent.

Assessment of previous analyses performed for spent fuel truck transportation casks, including 

impacts onto flat rigid targets, into cylindrical rigid targets, by locomotives, and by falling bridge 

structures, indicate that truck casks will not release their contents in any impact accidents.

In summary, the sequence of events necessary for there to be the possibility of any release is a 

rail transport cask with no welded canister travelling at an impact velocity greater than 97 kph

(60 mph). This cask would have to impact in a side orientation and the surface would have to 

be hard rock with an impact angle greater than 30 degrees.

73 NUREG-2125

4. CASK RESPONSE TO FIRE ACCIDENTS

4.1 Introduction

Certified Type B casks are designed to withstand a fully-engulfing fire for 30 minutes while 

maintaining critical functions, including protecting the public from doses of radiation exceeding

regulatory limits. Certification analyses of the hypothetical accident condition (HAC) fire 

specified in 10 CFR 71.73, “Hypothetical Accident Conditions,” generally impose a thermal 

environment on the cask similar to or more severe than most thermal environments a cask may 

be exposed to in actual transportation accidents involving a fire (Fischer et al., 1987). Large 

open-pool fires can burn at temperatures higher than the average temperature of 800 degrees 

Celsius (C) (1,475 degrees Fahrenheit (F)) specified in HAC fire regulations. Actual fire plumes 

have location- and time-varying temperature distributions that vary from about 600 degrees C 

(1,112 degrees F) to more than 1,200 degrees C (2,192 degrees F) (Koski, 2000; 

Lopez et al., 1998). Therefore, an evenly-applied 800 degrees C (1,475 degrees F) fire 

environment used in a certification analyses applies a similar heating as an actual fire.

This risk study used computer codes capable of modeling both fire behavior and the thermal 

responses of objects engulfed in those fires in a realistic way18 to analyze the response of the 

Rail-Steel and the Rail-Lead casks to three different fire configurations. This chapter describes 

these configurations and discusses the casks’ temperature responses. An analysis of the 

thermal performance of the Truck-DU (depleted uranium) cask when exposed to a severe fire 

scenario is also presented.

The thermal response of each cask is compared to two characteristic temperature limits: the 

rated seal temperature (350 degrees C (662 degrees F) for elastomeric seals used in the 

Rail-Lead cask and the Truck-DU cask and 649 degrees C (1,200 degrees F) for the metallic 

seal used in the Rail-Steel cask) and the fuel rod burst rupture temperature (750 degrees C 

(1,382 degrees F) for all casks (Lorenz, 1980)). These temperature limit values are the same 

as those used in NUREG/CR-6672, “Re-examination of Spent Fuel Shipment Risk Estimates,”

for the elastomeric seal and fuel rod burst temperature. The Rail-Steel cask seal temperature 

limit is obtained from Table 2.1.2 and Table 4.1.1 in the HI-STAR 100 safety analysis report 

(SAR) (Holtec International, 2000). Section 7.2.5.2 in NUREG/CR-6672 explains that 350

degrees C (662 degrees F) is a conservative temperature limit the spent nuclear fuel (SNF)

transportation industry typically uses for elastomeric seals. Section 7.2.5.2 of NUREG/CR-6672 

also provides the rationale for the use of 750 degrees C (1,382 degrees F) as the fuel rod burst 

rupture temperature. These temperature limits are used in this study to determine if the cask 

seals or fuel rods would be compromised under any of the accident scenarios analyzed. If only 

the seals are compromised, a CRUD-only release ensues. If the fuel rods and seals are both 

compromised, a release of CRUD and spent fuel constituents would ensue. In either case, the 

consequences of the release would have to be evaluated. In addition, other thermally-induced 

phenomena that could cause a degradation of the package are considered. These include the 

melting of lead in the Rail-Lead cask, solid/solid phase changes in the uranium in the DU-Truck 

casks, and rapid oxidation of the stainless steel in all of the casks.

 

18 Computational fluid dynamics fire codes are capable of modeling flame behavior, soot formation, flow of hot 

gasses, and other physical phenomena found in actual fires.
NUREG-2125 74

To make sure that all relevant failure mechanisms were considered, staff requested Oak Ridge 

National Laboratory (ORNL) to perform a phenomena identification and ranking review of Draft 

NUREG-2125 (Smith et al., 2013). This review identified two additional phenomena, but 

concluded that the temperatures obtained by the casks were not sufficient for these phenomena 

to result in cask failure.

4.2 Description of Accident Scenarios

4.2.1 Pool size 

Three fire accident scenarios are analyzed for each rail cask and one for the truck cask. A 

hydrocarbon fuel pool that conforms to the HAC fire described in 10 CFR 71.73 is used as the 

basis for each scenario. This regulation specifies a hydrocarbon fuel pool that extends between 

1 and 3 meters (3.3 and 10 feet) horizontally beyond the external surface of a cask. To ensure 

that the fire fully engulfed the large casks analyzed in this study, all fuel pools extended 

3 meters (10 feet) from the sides of the cask.

4.2.2 Fire Duration

The fire duration postulated for the rail cask analyses is based on the capacity of a large rail 

tank car. Typical large rail tank cars can carry about 113,562 liters (30,000 gallons) of 

flammable or combustible liquids (i.e., hydrocarbon-based liquids). To estimate the duration of 

the fires, all of the fuel in the tank car is released and assumed to form a pool with the 

dimensions of a regulatory pool fire for the rail casks analyzed. That is, fuel pools extending

horizontally 3 meters (10 feet) beyond the surfaces of the casks are used in the fire models.

Provided that relatively small differences exist between the overall dimensions of the Rail-Steel 

cask and the Rail-Lead cask, these fuel pools are similar in size and are nominally 14 m×9 m 

(46 feet×29.5 feet). A pool of this size would have to be 0.9 meters (3 feet) deep to pool 

113,562 liters (30,000 gallons) of liquid fuel, a condition extremely unlikely to occur in any 

accident scenario. If all of the fuel in this pool were to ignite and burn (i.e., none of the fuel runs 

off or soaks into the ground), the pool fire would burn for approximately 3 hours. This fire 

duration is estimated using a nominal hydrocarbon fuel recession (evaporation) rate of 5 mm

(0.2 inches) per minute, which is typical of large pool fires (SFPE, 2002; Lopez et al., 1998; 

Quintiere, 1998). This large pool area could also burn for up to 3 hours—although it would be 

even less likely—if the liquid fuel flows at exactly the right rate to feed and maintain the pool 

area for the duration of the fire. Since these pooling conditions are very difficult to obtain, the 

fire duration presented here is considered conservative. NUREG/CR-7034, “Analysis of Severe 

Railway Accidents Involving Long Duration Fires,” corroborates that it is very difficult for a rail 

cask to be subjected to long-duration, large fires (Adams et al., 2011). Nonetheless, a 3-hour 

fire that does not move over time, and is capable of engulfing a rail cask over the duration of the 

fire, is conservatively used for the analysis of the two rail casks in this study.

In the case of the Truck-DU cask, fire duration is based on the fuel capacity of a typical 

petroleum tank truck. One of these tank trucks can transport approximately 34,070 liters 

(9,000 gallons) of gasoline on the road. Provided that the overall dimensions of the Truck-DU 

cask are 2.3 meters×6 meters (7.5 feet×19.7 feet), a regulatory pool that extends horizontally 

3 meters (10 feet) beyond the outer surface of the cask would be 8.3 meters×12 meters

(27.2 feet×39.4 feet). To pool 34,070 liters (9,000 gallons) of gasoline in this area, the pool 

would have to be 0.3 meter (1 foot) deep, a configuration difficult to obtain in an accident 

scenario and therefore unlikely to occur. This type of pool fire would burn for a little more than 

1 hour. As discussed for the rail cask pool fire, the other possibility of maintaining an engulfing
75 NUREG-2125

fire which can burn for that duration is if, for example, gasoline flowed at the right rate to 

maintain the necessary fuel pool conditions. This scenario is also very unlikely. 

NUREG/CR-7035, “Analysis of Severe Roadway Accidents Involving Long Duration Fires,”

corroborates the assertion that it is very difficult for a truck cask to be subjected to long duration, 

large fires (Adams and Mintz, 2011). Nevertheless, 1 hour is used as the duration of a fire not 

moving over time for the conservative analysis of the Truck-DU cask.

4.2.3 Hypothetical Accident Configurations for the Rail Casks

Three fire accident scenarios that differ from the regulatory HAC fire configuration are analyzed 

in this study for the rail casks. These are:

(1) Cask lying on the ground in the middle of (concentric with) a pool of flammable liquid 

(such as gasoline) as depicted in Figure 4-1. This scenario represents the case in which 

the liquid fuel spilled because of an accident flows to the location where the cask comes 

to rest following the accident and forms a large pool under (and concentric with) the 

cask.

Figure 4-1 Cask lying on ground concentric with fuel pool

(2) Cask lying on the ground 3 meters (10 feet) away from the pool of flammable liquid (with 

the side of the cask aligned with the long side of the fuel pool) as depicted in Figure 4-2.

This scenario represents the hypothetical case in which the fuel pool and the cask are 

separated by the width of one rail car. This could be the case in an accident in which the 

rail cars derail in an “accordion” fashion.

Cask in the middle of flammable liquid 

fuel pool region(shown in orange)

Fire engulfing the cask
NUREG-2125 76

Figure 4-2 Cask lying on ground 3 meters (10 feet) from pool fire

(3) Cask lying on the ground 18 meters (60 feet) from the pool of flammable liquid (with the 

side of the package aligned with the long side of the fuel pool) as depicted in Figure 4-3.

This scenario represents the hypothetical case in which the pool of flammable liquid and 

the cask are separated by the length of one rail car. This represents an accident in 

which the separation between a tank car carrying flammable liquid and the railcar 

carrying the SNF package is maintained (the distance of a buffer rail car, which is always 

required when radioactive and flammable/hazardous liquids are transported on the same 

train19) after the accident. For this scenario, the most damaging cask position is 

assumed (i.e., the side of the cask is assumed to face the fire).

For each scenario, calm wind conditions leading to a vertical fire are assumed. Only the cask 

and the fuel pool are represented for the analysis. For conservatism, objects that would be 

present and could shield or protect the cask from the fire (i.e., such as the conveyance or other 

rail cars) are not included. All analyses include decay heat from the cask content.

Before the accident scenarios were analyzed, two additional 30-minute regulatory HAC fire 

analyses were performed for each rail cask based on conditions described in 10 CFR 71.73. In 

the first analysis, a commercially-available FE heat transfer code is used to apply an 

800 degrees C (1,475 degrees F) uniform-heating fire condition to the casks. In the second 

analysis, a benchmarked computational fluid dynamics (CFD) computer model with radiation 

heat transfer is used.

 

19 49 CFR 174.85, “Position in Train of Placarded Cars, Transport Vehicles, Freight Containers, and Bulk 

Packagings”
77 NUREG-2125

Figure 4-3 Cask lying on ground 18 meters (60 feet) from pool fire

In the computer model, each cask is positioned 1 meter above the fuel pool (as described in 

10 CFR 71.73) and a realistic fire fully engulfs the cask as shown in Figure 4-4. The FE uniform 

heating analyses results were compared to those in the SARs to ensure that the cask models 

used in these analyses were representative. The CFD fire analyses results are compared to the 

results obtained from the uniform-heating FE analyses to demonstrate that the realistic CFD fire 

imposes conditions similar to uniform heating.

Cask elevated 1 meter (3.3 feet) above 

flammable liquid fuel pool region (shown in 

orange)

Regulatory fire engulfing the cask

Figure 4-4 Regulatory pool fire configuration
NUREG-2125 78

4.2.4 Hypothetical Accident Configuration for the Truck Cask

In the case of the truck cask, only the most severe hypothetical accident configuration (i.e., the 

cask is assumed to be concentric with a flammable fuel pool and is fully engulfed by fire) is 

analyzed because none of the temperature limits were reached and the offset fire scenarios 

would be less severe.

Figure 4-5 presents this hypothetical accident configuration.

Cask in the middle of flammable liquid fuel pool 

region (shown in orange)

Fire engulfing the cask

Figure 4-5 Truck-DU cask lying on ground concentric with fuel pool

4.3 Analysis of Fire Scenarios Involving Rail Casks

Advanced computational tools generated the data necessary for this risk study. Heat transfer 

from the fire to the cask body was simulated for hypothetical fire accidents. Two computer 

codes, including all the relevant heat transfer and fire physics, were used in a coupled manner.

This allows for the simultaneous detailed modeling of realistic external fire environments and 

heat transfer within the cask’s complex geometry. This section contains brief descriptions of the 

models and detailed information on the computer models, including material properties, 

geometry, and boundary conditions. Appendix D presents the assumptions used for model 

generation and subsequent analyses.

This section presents the results from the fire and heat transfer analyses performed on the 

Rail-Steel and Rail-Lead casks. The scale in the temperature distribution plots of all the 

Rail-Steel cask analysis results are the same to make comparisons easier. The same is done 

for the Rail-Lead cask plots.

Results are presented in the following order:

(1) 800 degrees C (1,475 degrees F) uniform heating exposure for 30 minutes (based on 

10 CFR 71.73)

(2) 30-minute CFD pool fire using the container analysis fire environment (CAFE) code 

(based on 10 CFR 71.73)
79 NUREG-2125

(3) 3-hour container analysis fire analysis (CAFE) pool fire (cask on ground concentric with 

pool)

(4) 3-hour CAFE pool fire (cask on ground 3 meters from pool)

(5) 3-hour CAFE pool fire (cask on ground 18 meters from pool)

4.3.1 Simulations of the Fires

Fire simulations are performed with the CAFE code (Suo-Anttila et al., 2005). CAFE is a CFD 

and radiation heat transfer computer code capable of realistically modeling fires that is coupled 

to a commercially-available FE analysis computer code to examine the effects of fires on 

objects. CAFE has been benchmarked against large-scale fire tests specifically designed to 

obtain data for calibration of fire codes (del Valle, 2008; del Valle et al., 2007; Are et al., 2005; 

Lopez et al., 2003). Appendix D contains details on the benchmark exercises performed to 

ensure that proper input parameters are used to realistically represent the engulfing and offset 

fires assumed in this study.

4.3.2 Simulations of the Rail Casks

The heat transfer within the Rail-Steel and the Rail-Lead casks is modeled with the computer 

code MSC PATRAN-Thermal (P-Thermal) (MSC, 2008). This code is commercially available 

and may be used to solve a variety of heat transfer problems. P-Thermal has been coupled 

with CAFE, allowing for a refined heat transfer calculation within complex objects, such as spent 

fuel casks, with realistic external fire boundary conditions.

Both the Rail-Steel and Rail-Lead casks have a polymeric neutron shield that is assumed to 

melt completely and be replaced by air at its operational temperature limit (see Appendix D).

The Rail-Lead cask has a lead gamma shield that can change phase upon reaching its melting 

temperature. Unlike the neutron shield, the thermal energy absorbed in the process of melting 

the gamma shield is included in the analyses. The thermal expansion effects of the lead are not 

included in the heat transfer calculations but are considered in the estimation of the gamma 

shielding reduction. Thick multilayered carbon steel walls provide the gamma shielding in the 

Rail-Steel cask. Therefore, melting is not a consideration for this cask under any condition to 

which it is exposed.

Impact limiters are modeled as undamaged (not deformed). The Rail-Steel cask has aluminum 

honeycomb impact limiters and the Rail-Lead cask has wood impact limiters. Prior analyses of 

casks comparing the thermal results for regulatory fires with impact limiters damaged from the 

regulatory impact test vs. undamaged limiters has shown very little effect from including the 

damage (Lopez and Ammerman, 2004). Damage to impact limiters from an extra-regulatory 

impact is not considered because of the extremely low probability that an extra-regulatory 

impact will be followed by a long-duration fire. Spaces between components are explicitly 

modeled in both casks because they could have a significant effect on the cask’s thermal 

response. FE models of the two casks are shown in Figure 4-6 and Appendix D presents 

details on cask modeling.
NUREG-2125 80

Rail-Steel cask

Rail-Lead cask

Figure 4-6 Finite element models (cut views) of the two rail casks analyzed
81 NUREG-2125

4.3.3 Simulation of the Spent Nuclear Fuel Region

The interior of the package comprising the fuel basket and the fuel assemblies is not modeled 

explicitly. A homogenized SNF region, comprised of all materials and geometric features of the 

fuel basket and fuel assemblies, is represented as a solid cylinder inside the cask. The thermal 

response of the homogenized SNF region is similar to the overall response for the more detailed 

fuel basket model and assemblies reported in NUREG/CR-6886, “Spent Fuel Transportation 

Package Response to the Baltimore Tunnel Fire Scenario,” (Adkins et al., 2006) and provides 

enough resolution for the purposes of this study. Appendix D presents details on how the 

effective properties of the homogenized SNF region are determined and applied to the models. 

The condition of the fuel is not important for this analysis, but the fuel response model in 

Appendix C suggests that following a regulatory impact accident there would be no failure of the 

fuel cladding or collapse of the fuel spacer grids.

4.3.4 Rail-Steel Cask Results

Results for the Rail-Steel cask are presented in the order specified at the beginning of 

Section 4.3 in Figure 4-7 through Figure 4-21. Figure 4-7 through Figure 4-10 contain the 

temperature distribution and transient temperature response of key cask regions for the 

regulatory 800 degrees C uniform heating and regulatory CAFE fire.

°C

External surface 3⁄8-cut view

Figure 4-7 Temperature distribution of the Rail-Steel cask at the end of the 30-minute 

800°C (1472°F) regulatory uniform heating
NUREG-2125 82

Figure 4-8 Temperature of key cask regions, Rail-Steel cask undergoing regulatory 

uniform heating

As modeled using FE, the uniform external heating produces an even temperature response 

around the circumference of the cask. However, the realistic uneven fire heating of the exterior, 

as modeled using CAFE, produces temperatures that vary around the circumference. For 

comparison, results from the uniform (FE) regulatory fire simulation are plotted against the 

hottest regional temperatures obtained from the regulatory CAFE (nonuniform) fire simulation.
83 NUREG-2125

°C 

External surface External surface - Bottom

1⁄2-cut view 3⁄8-cut view

Figure 4-9 Temperature distribution of the Rail-Steel cask at the end of the 30-minute 

regulatory CAFE fire
NUREG-2125 84

Figure 4-10 Temperature of key cask regions, Rail-Steel cask undergoing regulatory

CAFE fire

Figure 4-11 presents this thermal response comparison and illustrates that the uniform heating 

thermal environment described in 10 CFR 71.73 heats up the seal region of the Rail-Steel cask 

more than a real fire, even though a real fire can heat the cask to a temporary and localized 

thermal environment greater than 800 degrees C. A real fire applies a time- and space-varying 

thermal load to an object that it engulfs. In particular, large fires have an internal region where 

fuel exists in the form of gas, but not enough oxygen is available for that fuel to burn. That 

region is typically called the “vapor dome.” The lack of oxygen in the vapor dome is attributed to 

poor air entrainment in larger diameter pool fires, where most of the oxygen is consumed in the 

plume region’s perimeter. Since combustion is inefficient inside the vapor dome, this region 

remains cooler than the rest of the fire envelope. Thus, the presence of regions cooler than 

800 degrees C within a real fire makes it possible for fires with peak flame temperatures above 

800 degrees C to have an overall effect on internal temperatures of a thermally massive object 

that is similar to those obtained by applying a simpler heating condition, such as the one 

specified in 10 CFR 71.73. The non-uniform heating of a real fire leads to locations on the cask 

exterior, and especially on the impact limiter exterior, whose temperature is higher than 

800 degrees C (1472 degrees F). The way the impact limiters were modeled in these analyses 

and conservatism in the CAFE code increase the maximum temperatures reported on the outer

shell of the impact limiters. The peak temperature of this region shown in Figure 4-9 is above 

1300 degrees C (2370 degrees F), a temperature where rapid oxidation of the thin stainless 

steel shell might be of concern. However, this extremely high surface temperature is an artifact 

of the conservative modeling process. That is, due to the boundary conditions specified for this 

simulation, the model over-predicts the temperature on this surface. In actual fire tests 

performed at Sandia and elsewhere throughout the world, rapid oxidation of thin stainless steel 

shells covering energy absorbing material has never been observed [Andersen et al., 1978, 

Pierce et al., 2003, Gelder and May, 2006]. The boundary conditions do have the intended 

effect of rendering the model conservative in that the high surface temperature results in 

additional energy being imparted to the cask, thereby resulting in conservatively high peak 

temperatures throughout the cask.
85 NUREG-2125

Figure 4-11 Comparison of regulatory fire analysis for Rail-Steel cask: Uniform heating 

vs. CAFE fire

The “Outer Wall” CAFE curve is the average of the two “Outer Surface” CAFE curves for the 

sides of the cask as presented in Appendix D, Figure D-11.

The vapor dome effects on the temperature distribution within a fire and the concentration of 

unburned fuel available in the vapor dome for the CAFE regulatory analysis is illustrated in

Figure 4-12 and Figure 4-13.

Note that the plots in Figure 4-12 and Figure 4-13 are snapshots of the distributions at an 

arbitrary time during fire simulation. In reality, the fire moves slightly throughout the simulation, 

causing these distributions to vary over time. Nevertheless, these plots show representative 

distributions for the cask and fire configuration shown.

Appendix D provides additional plots with more information about temperature distributions at 

different locations in the cask.
NUREG-2125 86

Longitudinal plane (through the middle of the cask)

Transverse plane (at mid-length of the cask)

Figure 4-12 Gas temperature plots from the regulatory CAFE fire analysis
87 NUREG-2125

Longitudinal plane (through the middle of the cask)

Transverse plane (at mid-length of the cask)

Figure 4-13 Fuel concentration plots from the regulatory CAFE fire analysis

Results from the analysis of the cask lying on the ground and concentric with a pool fire that 

burns for 3 hours are presented in Figure 4-14 and Figure 4-15. As in the regulatory 

configuration, in which the cask is elevated 1 meter above the hydrocarbon fuel pool, the vapor 

dome affected the temperature distribution of the cask. This is evident by the cooler 

temperatures observed at the bottom of the cask. In this scenario, even after 3 hours in the fire, 
NUREG-2125 88

temperatures at the bottom of the package are cooler than temperatures observed in the 

regulatory configuration. However, the top of the cask in this configuration heats up more than 

the rest of the cask. This differs from what is observed in the regulatory configuration, in which 

the hotter regions are found on the sides of the cask.

°C 

External surface External surface - Bottom

1⁄2-cut view 3⁄8-cut view

Figure 4-14 Temperature distribution of the Rail-Steel cask at the end of the 3-hour 

concentric CAFE fire with cask on ground
89 NUREG-2125

Figure 4-15 Temperature of key cask regions, Rail-Steel cask with cask on ground, 

concentric fire

Figure 4-16 and Figure 4-17 are the fire temperature distribution and fuel concentration plots at

an arbitrary time during the CAFE fire simulation. In this case, the concentration of unburned 

fuel under the cask is high; therefore, the fire temperature under the cask is lower than what is 

observed in the regulatory configuration.
NUREG-2125 90

Longitudinal plane (through the middle of the cask)

Transverse plane (at mid-length of the cask)

Figure 4-16 Gas temperature plots from the CAFE fire analysis of the cask on ground
91 NUREG-2125

Longitudinal plane (through the middle of the cask)

Transverse plane (at mid-length of the cask)

Figure 4-17 Fuel concentration plots from the CAFE fire analysis of the cask on ground

Results of the offset fire analyses are summarized in Figure 4-18 through Figure 4-21. In the 

case of the 3-meter offset, the side of the cask facing the fire received heat by thermal radiation.

The heat that the cask absorbed during the 3-hour exposure caused the cask temperature to 

rise, as depicted in Figure 4-18 and Figure 4-19. Similarly, the 18-meter offset fire caused the 

cask temperature to rise as illustrated in Figure 4-20 and Figure 4-21. These results show that 

offset fires, even as close to the cask as 3 meters, do not represent a threat to this thermally 

massive SNF transportation cask. The maximum temperatures observed in the seal and in the 
NUREG-2125 92

SNF region did not reach their temperature limits. Therefore, offset fire scenarios will not cause 

this package to release radioactive material.

°C 

External surface 3⁄8-cut view

Figure 4-18 Temperature distribution of the Rail-Steel cask at the end of the 3-hour, 

3-meter (10-foot) offset CAFE fire with cask on ground

Figure 4-19 Temperature of key cask regions, Rail-Steel cask with cask on ground, 

3-meter (10-foot) offset fire
93 NUREG-2125

°C 

External surface 3⁄8-cut view

Figure 4-20 Temperature distribution of the Rail-Steel cask at the end of the 3-hour 

18-meter (60-foot) offset CAFE fire with cask on ground

Figure 4-21 Temperature of key cask regions, Rail-Steel cask with cask on ground, 

18-meter (60-foot) offset fire
NUREG-2125 94

Summary of Rail-Steel Cask Analysis Results

The results show that the Rail-Steel cask is capable of protecting fuel rods from burst rupture 

and of maintaining containment when exposed to the severe fire environments analyzed as part 

of this study. That is, while the neutron shield material is conservatively assumed to be absent 

during the fire accident, the SNF region stays below 750 degrees C (1,382 degrees F) and the 

seal region stayed under 649 degrees C (1,200 degrees F) for all the scenarios considered.

Furthermore, this cask uses a welded canister that will not be compromised under these thermal 

loads. This cask will not experience loss of gamma shielding because the shielding is a thick 

multilayered carbon steel wall, which is not affected in a way that could reduce its ability to 

provide shielding.

4.3.5 Rail-Lead Cask Results

The thermal response of the Rail-Lead cask to the same fire environments discussed for the 

Rail-Steel cask is presented in this section. The 30-minute regulatory fire results are 

summarized in Figure 4-22 through Figure 4-26.

°C 

External surface 3⁄8-cut view

Figure 4-22 Temperature distribution of the Rail-Lead cask at the end of the 30-minute 

800°C (1472°F) regulatory uniform heating
95 NUREG-2125

Figure 4-23 Temperature of key cask regions, Rail-Lead cask undergoing regulatory 

uniform heating
NUREG-2125 96

°C 

External surface External surface - Bottom

1⁄2-cut view 3⁄8-cut view

Figure 4-24 Temperature distribution of the Rail-Lead cask at the end of the 30-minute 

regulatory CAFE fire
97 NUREG-2125

Figure 4-25 Temperature of key cask regions, Rail-Lead cask in regulatory CAFE fire

Results from the uniform regulatory fire simulation are plotted against the hottest regional 

temperatures from the CAFE (nonuniform) regulatory fire simulation. This plot is shown in 

Figure 4-26. As with the Rail-Steel cask, this figure illustrates that the uniform heating thermal 

environment described in 10 CFR 71.73 heats the seal region of the Rail-Lead cask more than a 

nonuniform real fire may, even though a real fire may impart to the cask a localized thermal 

environment greater than 800 degrees C (1,472 degrees F).
NUREG-2125 98

Figure 4-26 Comparison of regulatory fire analysis for Rail-Lead cask: Uniform heating 

vs. CAFE fire

The “Outer Wall” CAFE curve is the average of the two “Outer Surface” CAFE curves for the 

sides of the cask as presented in Appendix D, Figure D-21.

Analyses results of the cask lying on the ground heated by the concentric and offset fires are 

summarized in Figure 4-27 through Figure 4-32. These plots show similar trends to those 

observed in the Rail-Steel cask for the same configurations.

Two of the scenarios analyzed show melting of the lead gamma shield in the Rail-Lead cask.

Lead melts at 328 degrees C (622 degrees F). During that process, it absorbs (stores) heat 

while maintaining its temperature relatively constant at 328 degrees C. As a result, the heatup 

rate of parts of the cask slows down while the lead melts. This is why the curve of the region 

inward from the gamma shield region (i.e., the edge of the SNF region) in Figure 4-28 and 

Figure 4-30 shows a change in slope at about 328 degrees C. This effect is seen more clearly 

in the slower heating case shown in Figure 4-30. Once the lead melting process is complete, 

the cask resumes heating up if the external source is still at a higher temperature. Note that a 

similar effect is observed when the lead solidifies at 328 degrees C during the postfire cooling 

period. In this case, the cooling rate of portions of the cask slows down while the lead solidifies.

This can also be clearly seen in Figure 4-30.

Appendix D contains plots with additional information on temperature distributions at more cask 

locations. The gradual thermal expansion and contraction of the gamma shield region during 

cask heating and cooling is another effect considered in cases where lead melted. This effect is 

discussed in the next subsection.
99 NUREG-2125

°C 

External surface External surface - Bottom

1⁄2-cut view 3⁄8-cut view

Figure 4-27 Temperature distribution of the Rail-Lead cask at the end of the 3-hour 

concentric CAFE fire with cask on ground
NUREG-2125 100

Figure 4-28 Temperature of key cask regions, Rail-Lead cask with cask on ground, 

concentric fire

°C 

External surface 3⁄8-cut view

Figure 4-29 Temperature distribution of the Rail-Lead cask at the end of the 3-hour 

3-meter (10-foot) offset CAFE fire with cask on ground
101 NUREG-2125

Figure 4-30 Temperature of key cask regions, Rail-Lead cask with Cask on ground, 

3-meter (10-foot) offset fire

°C 

External surface 3⁄8-cut view

Figure 4-31 Temperature distribution of the Rail-Lead cask at the end of the 3-hour 

18-meter (60-foot) offset CAFE fire with cask on ground
NUREG-2125 102

Figure 4-32 Temperature of key cask regions, Rail-Lead cask with cask on ground, 

18-meter (60-foot) offset fire

Melting of the Lead Gamma Shield

There are two cases in which a portion of the lead gamma shield melts. These are the 3-hour 

concentric fire and the 3-hour, 3-meter (10-foot) offset fire. The lead gamma shield region that 

melted for each case is shown in red in Figure 4-33 and Figure 4-34. These two figures only 

show the lead portion of the cask wall. As these figures show, approximately 88 percent of the 

lead melts in the case of the 3-hour concentric fire, whereas only about 30 percent of the lead 

melts in the 3-hour, 3-meter (10-foot) offset fire. Because of melting and thermal expansion of 

some of the lead gamma shield, some loss of lead shielding is observed, which translates to an 

increase in gamma radiation exposure. The width of the streaming path (i.e., the gap created 

because of lead melt, expansion, and subsequent contraction as it solidifies) is estimated. For 

this estimate, it is assumed that the thermal expansion of the lead permanently deforms 

(buckles) the interior wall of the cask, enabling calculation of the gap in the lead gamma shield.

The lead region gap that the concentric fire case causes is assumed to appear on the top 

portion of the cask. That is, after the lead melts and buckles the interior wall of the cask 

because of its thermal expansion, molten lead is assumed to flow to the lower portions of the 

cask’s gamma shield region, which allows a gap to form on the top portion. From a geometric 

analysis that considered the expansion and contraction of the lead and a conservative cask wall 

deformation, this gap is estimated to be about 0.5 meters (20 inches), which translates to an 

8.1 percent loss of lead shielding. In the 3-meter offset fire, the gap is assumed to form on the 

top portion of the molten lead region shown in Figure 4-34. In this case, the gap is estimated to 

be about 0.127 m (5 inches), which translates to a 2 percent loss of lead shielding. These gaps 

are estimated using geometric information and temperature-dependent density values of lead 

(i.e., 11.35 g/cm3 (0.41 lb/in3

) for solid lead and 10.6 g/cm3 and 10.3 g/cm3 (0.38 lb/in3 and 

0.37 lb/in3

) for molten lead at temperatures of 384 degrees C and 577 degrees C (723 degrees 
103 NUREG-2125

F and 1071 degrees F), respectively). The loss-of-shielding fractions reported in this section are 

used in Chapter 5 to estimate the consequences.

Figure 4-33 Rail-Lead cask lead gamma shield region – maximum lead melt at the middle 

of the cask – Scenario: Cask on ground, 3-hour concentric pool fire

Figure 4-34 Rail-Lead cask lead gamma shield region – maximum lead melt at the middle 

of the cask – Scenario: Cask lying on ground, 3-hour 3-meter (10-foot) offset pool fire
NUREG-2125 104

Summary of Rail-Lead Cask Analysis Results

The results presented here show that the Rail-Lead cask is also capable of protecting the fuel 

rods from burst rupture and of maintaining containment when exposed to the severe fire 

environments analyzed, even when the neutron shield material is conservatively assumed to be 

absent during the fire accident. However, some reduction of gamma shielding is estimated to 

occur in two cases. Partial loss of lead shielding is expected when the cask is exposed to a

concentric fire that burns for longer than 65 minutes and for casks that receive heat from a fire 

offset by 3 meters (10 feet) and that burns for longer than 2 hours and 15 minutes. 

Nevertheless, no release of radioactive material is expected if this cask was exposed to any of 

these severe thermal environments because the elastomeric seals did not reach their 

temperature limit. This ensures the cask is capable of maintaining containment (i.e., preventing 

any radioactive material from getting out of the package) under any of the fire environments 

analyzed.

4.4 Truck Cask Analysis

A three-dimensional (3D) analysis of the Truck-DU cask engulfed in a large fire is performed for 

this study. The cask is assumed to lie on the ground concentric with the hydrocarbon fuel pool 

fire. As explained in Section 4.2.2, the fire is assumed to last 1 hour. Results from the fire and 

heat transfer analyses performed on the Truck-DU cask are presented in this section.

4.4.1 Simulation of the Truck Cask

The heat transfer to and within the Truck-DU cask is modeled using P-Thermal/CAFE. The 

cask has a hydrogenous neutron shield that is assumed to disappear completely and replaced 

by air at its operational temperature limit (see Appendix D). In this cask, a layer of DU within the 

cask wall provides the gamma shielding. Melting of the DU is not a concern for this cask under 

any of the conditions to which it is exposed. The aluminum honeycomb impact limiters are 

modeled as undamaged (not deformed). Decay heat was included in the analysis. The FE

model of the cask is shown in Figure 4-35. Appendix D presents cask modeling details.

Figure 4-35 Finite element model (cut view) of the Truck-DU cask
105 NUREG-2125

4.4.2 Simulation of the Spent Nuclear Fuel Region

As with the rail casks, the SNF region comprising the fuel basket and the fuel assemblies is not 

modeled explicitly for the Truck-DU cask. Instead, a homogenized SNF region is used. All 

materials and geometric features of the fuel basket and fuel assemblies are represented as a 

single solid inside the cask. Appendix D presents the effective properties of the homogenized 

SNF region.

4.4.3 Truck-DU Cask Results

The results from the analysis of the cask lying on the ground and concentric with a pool fire that 

burns for 1 hour are presented in Figure 4-36 and Figure 4-37.

°C 

External surface External surface - Bottom

1⁄2-cut view 3⁄8-cut view

Figure 4-36 Temperature distribution of the Truck-DU cask at the end of the 1-hour 

concentric CAFE fire with cask on ground

As observed with the rail casks, the vapor dome affected the temperature distribution of the 

truck cask. This is evident by the cooler temperatures observed at the bottom of the cask. 
NUREG-2125 106

Even after 1 hour in the fire, the temperatures at the bottom of the cask are lowest and the 

temperatures at the top are highest.

Figure 4-37 Temperature of key cask regions, Truck-DU cask with cask on ground, 

concentric fire

Figure 4-38 and Figure 4-39 are the fire temperature distribution and fuel concentration plots at 

an arbitrary time during the CAFE fire simulation. Note that the concentration of unburned fuel 

under the cask is high. This means that poor combustion is occurring in that zone, leading to 

cooler temperatures of the cask’s lower region.
107 NUREG-2125

Longitudinal plane (through the middle of the cask)

Transverse plane (at mid-length of the cask)

Figure 4-38 Gas temperature plots

CAFE fire analysis of the truck cask on ground
NUREG-2125 108

Longitudinal plane (through the middle of the cask)

Transverse plane (at mid-length of the cask)

Figure 4-39 Fuel concentration plots

CAFE fire analysis of the Truck-DU cask lying on ground

Summary of Truck-DU Cask Analysis Results

The results presented here show that the Truck-DU cask is capable of protecting the SNF rods 

from burst rupture and of maintaining containment when exposed to the severe fire environment 

analyzed in this study. That is, while the neutron shield material is conservatively assumed to 
109 NUREG-2125

be absent during the fire accident, the SNF region temperature is less than 750 degrees C 

(1,382 degrees F) and the seal region temperature is less than 350 degrees C (662 degrees F).

This cask will not experience gamma shielding loss because a thick steel-DU wall provides the 

shielding, which is not affected in a way that could reduce its ability to provide shielding. The 

peak temperature in the DU gamma shielding is 406 degrees C (763 degrees F) as shown in 

Figure 4-40, well below the temperature where uranium goes through a crystal lattice phase 

change or the temperature where it can undergo intermetallic reactions (eutectic formation) with 

the stainless steel cask walls.

Figure 4-40 Peak temperature in the DU shielding during the thermal event

4.5 Chapter Summary

This chapter presents the realistic analyses of four fire accident scenarios. These accident 

scenarios are identified below: 

• the HAC fire described in 10 CFR 71.73 (c) (4)

• a cask on the ground concentric with a fuel pool sufficiently large to engulf the cask

• a cask on the ground with a pool fire offset by the width of a rail car (3 meters) 

• a cask on the ground with a pool fire offset by the length of a rail car (18 meters) 

518

479

441

402

363

325

286

248

209

170

132

93

°C
NUREG-2125 110

Analyses are performed for the Rail-Steel and the Rail-Lead casks for these four fire accident 

scenarios. An analysis of a Truck-DU cask on the ground concentric with a hydrocarbon fuel 

pool sufficiently large enough to engulf the cask is also performed. Probable worst-case fire 

accident scenarios for a rail cask transported by railway and for a truck cask transported by 

roadway were represented within the cases analyzed. The neutron shield material of each cask 

analyzed was assumed to melt and flow out of the cask instantly at the beginning of the fire.

Results show that neither the Rail-Steel cask nor the Rail-Lead cask would lose the containment 

boundary seal in any of the accidents considered in this study. In addition, the SNF rods did not 

reach burst rupture temperature. Because the seals did not fail, if the rods had been failed by 

an impact event that occurred before the fire, there still would have been no release from either 

cask. Also, if the actual rod burst temperature is lower than 750 degrees C, as is suggested by 

the data of Chung and Kassner (1978), and some rod burst occurs, there still would have been 

no release from either of the casks. However, some loss of gamma shielding is expected with

the Rail-Lead cask in the event of a 3-hour engulfing fire and a 3-hour, 3-meter (10-foot) offset 

fire. Nevertheless, no release of radioactive material is expected to occur as a result of these 

hypothetical fire accidents because containment is not lost in any of the cases studied. In the 

case of the Truck-DU cask, containment would be maintained in the 1-hour fire accident. These 

results demonstrate the adequacy of current regulations to ensure the safe transport of SNF. 

Furthermore, the results demonstrate that SNF casks designed to meet current regulations will 

prevent the loss of radioactive material in realistic severe fire accidents.
111 NUREG-2125

5. TRANSPORTATION ACCIDENTS

5.1 Types of Accidents and Incidents

The different types of accidents that can interfere with routine transportation of spent nuclear 

fuel (SNF) are listed below.

• Accidents in which the spent fuel cask is not damaged or affected.

- Minor traffic accidents (“fender-benders,” flat tires) resulting in minor damage to 

the vehicle. These usually are called “incidents.”20

- Accidents that damage the vehicle or trailer enough so that the vehicle cannot 

move from the scene of the accident under its own power, but do not result in 

damage to the spent fuel cask.

- Accidents involving a death or injury, or both, but no damage to the spent fuel 

cask.

• Accidents in which the spent fuel cask is affected.

- Accidents resulting in the loss of lead gamma shielding or neutron shielding (or 

both), but no radioactive material is released.

- Accidents in which radioactive material is released.

Accident risk is expressed as “dose risk,” which is the product of the radiation dose resulting 

from the accident and the probability of that accident. The units used for dose risk are sieverts

(Sv). When the consequence to an entire population is considered, the accident risk is 

expressed as “collective dose risk,” and the units are person-Sv.

When an accident happens at a particular spot along the route, the vehicle carrying the spent 

fuel cask stops. Therefore, there can only be one accident for a shipment; resumption of the 

shipment essentially is a new shipment. Accidents can result in damage to spent fuel in the 

cask even if no radioactive material is released. While this would not result in additional 

exposure to members of the public, workers engaged in accident recovery operations—

including unloading or subsequently opening the cask at a facility—would be affected. 

Accidents damaging the fuel but not damaging the cask and potential consequent risk to 

workers are not included in this study.

 

20 In U.S. Department of Transportation terminology, an “accident” is an event that results in a death, an injury, 

or enough damage to the vehicle that it cannot move under its own power. All other events that occur in 

nonroutine transportation are “incidents.” This document uses the term “accident” for both accidents and 

incidents.
NUREG-2125 112

5.2 Accident Probabilities

Risk is the product of probability and consequence of a particular accident scenario. The 

probability, or likelihood, that a spent fuel cask will be in a specific type of accident is a 

combination of two factors—

• The probability that the vehicle carrying the spent fuel cask will be in an accident.

• The conditional probability that the accident will be a certain type of accident. This is a 

conditional probability because it depends on the vehicle being in an accident.

The net accident probability is the product of the probability of an accident and the conditional 

probability of a particular type of accident. A few hypothetical examples are given in Table 5-1

to illustrate the probability calculation.

Table 5-1 Illustrations of Net Probability

Accident 

Probability for a 

5,000 km 

(3,107-mile)

Cross-Country 

Tripa Accident Scenario

Conditional 

Probabilityb

Net Probability of 

Accident

0.0099 Truck collision with a 

gasoline tank truck 0.82×0.003 = 0.00246 0.82×0.003×0.0099 = 2.44×10-5

0.00066

Derailment into slope

>80 kph (>50 mph), no 

fire

0.7355×0.9846×

(0.06048+0.00005)×

0.9887×0.0011 = 

0.0000476

0.0000476×0.00066

= 3.14×10-8

0.00066

Railcar accident on a

bridge at 48-80 kph 

(30-50 mph), no fire

0.7355×0.9846×

0.2665×0.0113 = 

0.00218

0.00218×0.00066 

= 1.44×10-6

a Calculated from DOT, 2005, Table 1-32. b From event trees in Appendix E.

Accident probability is calculated from the number of accidents per kilometer (accident 

frequency) for a particular type of vehicle as recorded by the U.S. Department of Transportation 

(DOT) and reported by the Bureau of Transportation Statistics. Large truck accidents and 

freight rail accidents are the two data sets used in this analysis. The DOT has compiled and 

validated national accident data for truck and rail from 1971 through 2007 (DOT, 2008), but the 

accident rates declined definitively between 1971 and the 1990s. For this analysis, rates from 

1996 through 2007 are used: 0.0019 accidents per thousand large truck-km (0.0031 accidents 

per thousand large truck miles) and 0.00011 accidents per thousand railcar-km

(0.00018 accidents per thousand railcar miles).

Figure 5-1 shows the accidents per truck-km and per railcar-km for this period. The logarithmic 

scale is used on the vertical axis to show the entire range.
113 NUREG-2125

Figure 5-1 Accident frequencies in the United States from 1991 until 2007

Figure note: (1 km = 0.62137 miles)

As Chapters 3 and 4 show, however, the only accidents that could result in either loss of 

radiation shielding or release of radioactive material are rail accidents involving the Rail-Lead

cask when fuel is directly loaded inside the cask (i.e., the fuel is not contained in a welded 

canister inside the cask). These accidents are listed below.

• Collisions with hard rock or equivalent at impact speeds greater than 97 kph (60 mph) 

that result in some loss of lead gamma radiation shielding or damage to the cask seals.

Hard rock is not necessarily an unyielding target; however, collision of a cask with hard 

rock is the only type of collision along a transportation route that could damage the cask 

sufficiently to result in the release of radioactive material or loss of lead shielding.

• Fires of long-enough duration to compromise the lead shielding.

Whether these accidents happen depends on the likelihood (conditional probability) of the 

accident scenario as well as on accident frequency. The event trees for truck and rail, 

Figures E-1 and E-2 of Appendix E, show some elements of accident scenarios in each branch 

of the respective event tree. The dependence on probability is illustrated in Figure E-6, which 

shows the sequence of events necessary for a pool fire that can burn long enough to 

compromise the seals and lead shielding.

Table 5-2 shows the conditional probabilities of accidents that could result in a radiation dose to 

a member of the public. Sections E.3 to E.5 of Appendix E provide the analysis resulting in 

these conditional probabilities. The calculation of these probabilities is done using the typical 

method for risk assessments, but because of the large degree of safety that spent fuel casks

provide, only extremely low probability events have the possibility of leading to a radiation dose 

to the public. For these extremely low probability events, the results are reported to the 

precision of the calculation (to aid understanding of derivation of results), but they should be 

considered accurate only to the order of magnitude. 

1.E-08

1.E-07

1.E-06

19911992199319941995199619971998199920002001200220032004200520062007

Accidents/km

Annual Vehicle Accidents 1991-2007

Railcar Truck
NUREG-2125 114

Table 5-2 Scenarios and Conditional Probabilities of Rail Accidents Involving the 

Rail-Lead Cask

Accident Scenario for the Rail-Lead Cask

Conditional probability of gamma shield loss or 

radioactive material content release exceeding 

10 CFR 71.51 quantitiesa

Loss of lead shielding from impactb 8.3×10-10

Loss of lead shielding from firec 10-14 to 10-10

Radioactive materials release from impactd 5.1×10-10

Radioactive materials release from fire 0

a More than 99.999999 percent of potential accidents would result in neither loss of lead shielding nor a 

release of radioactive material. b From the cases in Table E-2 of Appendix E with lead slump greater than 1 percent. c From the fire event tree, Figure E-6 in Appendix E. d From the sum of probabilities in the last row of Table 5-10 for the casks with metal seals. The probability 

of release would be less if the cask is shipped with elastomer seals.

5.3 Accidents with Neither Loss of Lead Shielding nor Release of Radioactive 

Material

The conditional probability that an accident involving a lead-shielded cask will be the type with 

no release and no lead shielding loss, as the footnote to Table 5-2 states, is 99.999999 percent.

The only type of cask that could lose gamma shielding is a lead-shielded cask such as the 

Rail-Lead cask. The only type of cask that could release radioactive material in an accident is a 

cask carrying uncanistered spent fuel. Although the Truck-DU cask carries uncanistered fuel, it

would not release any radioactive material under any scenario postulated in this report. The 

Rail-Steel cask carries only canistered fuel and would not release any radioactive material.

Neither Truck-DU casks nor Rail-Steel casks are lead-shielded; therefore shielding loss would 

not occur.

Doses to emergency responders from an accident in which no material is released and no loss 

of lead gamma shielding are shown in Table 5-3, and collective dose risks to the public from this 

type of accident are shown in Table 5-4 and Table 5-5. These radiation doses depend on the 

following—

• The external dose rate from the cask (Table 2-1).

• A10-hour stop (DOE, 2002) at the scene of the accident, until the vehicle and cask, or 

both, can be moved safely. Ten hours may overstate the stop time for most accidents.

• An average distance of 5 meters (16.4 feet) between the cask and the first responders 

and others who remain with the cask.

• For collective dose risks, the average rural, urban, and suburban population densities for 

each route.

The radiation doses in Table 5-3 and dose risks in Table 5-4 and Table 5-5 are the 

consequences of all Truck-DU accidents, all Rail-Steel accidents, and 99.999999 percent of the 

Rail-Lead accidents.
115 NUREG-2125

Table 5-3 Dose to an Emergency Respondera from a Cask in a No-Shielding Loss, 

No-Release Accident

Cask Dose in Sv (mrem)

10-hour allowed dose in Sv (mrem) derived 

from the 1-hour dose in 10 CFR 71.51

Truck-DU 1.0 x10-3 (100) 0.1 (10,000)

Rail-Lead 9.2x10-4 (92) 0.1 (10,000)

Rail-Steel 6.9x10-4 (69) 0.1 (10,000)

a Includes police, incident command, firefighters, EMTs, and any other emergency responders.

Table 5-4 and Table 5-5 show collective dose risks in person-Sv for the 10-hour stop following

the accident. The routes chosen are for illustrative purposes only, and no SNF shipments are 

planned from any of these points of origination to any of these destinations. Doses are shown 

for rural, suburban, and urban segments of each route, but an accident only happens once on 

any route. Therefore, each listed dose risk is the collective dose residents on that route 

segment would receive if the accident happened at any spot on that type of route segment.

The average individual U.S. background dose for 10 hours is 4.1×10-6 Sv (0.41mrem). Average 

background doses during the 10-hour stop for the 16 truck routes analyzed are—

• rural: (4.1 10-6 Sv)×(16.8 persons/km2

)×π×(0.8 km)2 = 0.000138 person-Sv (13.8 

person-mrem)

• suburban: (4.1 10-6 Sv)×(463 persons/km2

)×π×(0.8 km)2 = 0.00382 person-Sv (382 

person-mrem)

• urban: (4.1 10-6 Sv)×(2,682 persons/km2

)×π×(0.8 km)2 = 0.0221 person-Sv (2,210 

person-mrem)

If the Truck-DU cask, for example, is in a no-shielding loss, no-release accident, the average 

collective dose (the sum of the background dose and the dose because of the accident) to 

residents for the 10 hours following the accident would be—

• rural: 0.000141 person-Sv (14.1 person-mrem)

• suburban: 0.003881 person-Sv (388.1 person-mrem)

• urban: 0.022108 person-Sv (2,210.8 person-mrem)

The background and accident collective doses would be indistinguishable from the collective 

background dose. Any dose to an individual is well below the dose that 10 CFR 71.51, 

“Additional Requirements for Type B Packages,” allows, which is to be expected.
NUREG-2125 116

Table 5-4 Collective Dose Risks to the Public from a No-Shielding Loss, No-Release 

Accident Involving Rail Casks (person-Sv) (1 Sv=105 mrem)

FROM/TO

Rail-Lead Rail-Steel

Rural Suburban Urbana Total Rural Suburban Urbana Total

MAINE YANKEE

ORNL 3.1x10-6 5.3x10-5 6.6x10-6 6.3x10-5 2.3x10-6 4.0x10-5 5.0x10-6 4.8x10-5

DEAF SMITH 2.3x10-6 5.7x10-5 6.8x10-6 6.6x10-5 1.7x10-6 4.3x10-5 5.2x10-6 5.0x10-5

HANFORD 5.7x10-6 5.2x10-5 6.3x10-6 6.4x10-5 4.3x10-6 3.9x10-5 4.8x10-6 4.8x10-5

SKULL 

VALLEY 2.8x10-6 5.1x10-5 5.3x10-6 6.0x10-5 2.1x10-6 3.9x10-5 4.0x10-6 4.5x10-5

KEWAUNEE

ORNL 3.1x10-6 5.7x10-5 7.2x10-6 6.8x10-5 2.3x10-6 4.3x10-5 5.4x10-6 5.1x10-5

DEAF SMITH 1.5x10-6 6.1x10-5 7.2x10-6 6.9x10-5 1.2x10-6 4.6x10-5 5.4x10-6 5.2x10-5

HANFORD 1.5x10-6 5.3x10-5 6.6x10-6 6.1x10-5 1.2x10-6 4.0x10-5 5.0x10-6 4.6x10-5

SKULL 

VALLEY 2.0x10-6 6.2x10-5 6.0x10-6 7.0x10-5 1.5x10-6 4.7x10-5 4.5x10-6 5.3x10-5

INDIAN POINT

ORNL 2.6x10-6 7.2x10-5 8.7x10-6 8.3x10-5 2.0x10-6 5.4x10-5 6.6x10-6 6.3x10-5

DEAF SMITH 1.9x10-6 5.9x10-5 7.5x10-6 6.9x10-5 1.4x10-6 4.5x10-5 5.7x10-6 5.2x10-5

HANFORD 1.9x10-6 5.6x10-5 7.2x10-6 6.5x10-5 1.4x10-6 4.3x10-5 5.5x10-6 5.0x10-5

SKULL 

VALLEY 2.2x10-6 6.0x10-5 6.6x10-6 6.9x10-5 1.7x10-6 4.6x10-5 5.0x10-6 5.2x10-5

IDAHO NATIONAL LAB

ORNL 1.9x10-6 6.0x10-5 5.8x10-6 6.8x10-5 1.4x10-6 4.6x10-5 4.4x10-6 5.2x10-5

DEAF SMITH 8.0x10-7 6.0x10-5 5.3x10-6 6.6x10-5 6.0x10-7 4.6x10-5 4.0x10-6 5.0x10-5

HANFORD 1.0x10-6 6.0x10-5 6.7x10-6 6.8x10-5 7.5x10-7 4.6x10-5 5.1x10-6 5.2x10-5

SKULL 

VALLEY 2.0x10-6 5.9x10-5 7.1x10-6 6.8x10-5 1.5x10-6 4.4x10-5 5.4x10-6 5.1x10-5

AVERAGE 2.3x10-6 5.8x10-5 6.7x10-6 6.7x10-5 1.7x10-6 4.4x10-5 5.1x10-6 5.1x10-5

a The urban dose is less than the suburban dose because urban residences are 83 percent shielded, while 

suburban residences are 13 percent shielded.
117 NUREG-2125

Table 5-5 Collective Dose Risks to the Public from a No-Shielding Loss, No-Release 

Accident Involving a Truck Cask (person-Sv) (1 Sv=105 mrem)

FROM TO

Truck-DU

Rural Suburban Urbana Total

MAINE 

YANKEE

ORNL 4.2x10-6 7.2x10-5 9.1x10-6 8.5x10-5

DEAF SMITH 3.9x10-6 6.7x10-5 8.4x10-6 7.9x10-5

HANFORD 3.2x10-6 5.9x10-5 8.4x10-6 7.1x10-5

SKULL VALLEY 3.5x10-6 6.1x10-5 8.6x10-6 7.3x10-5

KEWAUNEE

ORNL 4.1x10-6 6.6x10-5 8.3x10-6 7.8x10-5

DEAF SMITH 2.8x10-6 6.2x10-5 8.4x10-6 7.3x10-5

HANFORD 2.2x10-6 5.8x10-5 8.4x10-6 6.9x10-5

SKULL VALLEY 2.6x10-6 5.9x10-5 8.6x10-6 7.0x10-5

INDIAN POINT

ORNL 3.6x10-6 6.7x10-5 8.2x10-6 7.9x10-5

DEAF SMITH 3.6x10-6 6.7x10-5 8.2x10-6 7.9x10-5

HANFORD 2.7x10-6 6.2x10-5 8.4x10-6 7.3x10-5

SKULL VALLEY 3.0x10-6 6.4x10-5 8.5x10-6 7.6x10-5

IDAHO 

NATIONAL 

LAB

ORNL 2.6x10-6 5.5x10-5 7.9x10-6 6.6x10-5

DEAF SMITH 1.6x10-6 6.2x10-5 6.8x10-6 7.0x10-5

HANFORD 1.4x10-6 3.6x10-5 5.2x10-6 4.3x10-5

SKULL VALLEY 2.1x10-6 6.2x10-5 8.4x10-6 7.3x10-5

AVERAGE 2.9x10-6 6.1x10-5 8.1x10-6 7.2x10-5

a The urban dose risk is less than the suburban dose risk because urban residences are 83 percent shielded, 

while suburban residences are 13 percent shielded

5.4 Accidental Loss of Shielding

Section E.3.1 to Appendix E (loss of gamma shielding) and Section E.3.2 (loss of neutron 

shielding) provide details on dose calculations from shielding losses.

5.4.1 Loss of Lead Gamma Shielding

Type B transportation packages are designed to safely carry radioactive material and require 

shielding adequate to meet the external dose regulation of 10 CFR Part 71, “Packaging and 

Transportation of Radioactive Material.” SNF is extremely radioactive and requires shielding 

that absorbs gamma radiation and neutrons. The sum of the external radiation doses from 

gamma radiation and neutrons should not exceed 0.0001 Sv (10 mrem) per hour at 2 meters 

(6.7 feet) from the cask, as 10 CFR 71.47, “External Radiation Standards for All Packages,”

stipulates.

Each SNF transportation cask analyzed uses a different material to serve as gamma shielding.

They also may use different neutron shielding, but it is not usually part of the accident analysis.

The Rail-Steel cask has a steel wall thick enough to attenuate gamma radiation to acceptable 

levels. The Truck-DU cask uses metallic DU. Neither of these shields would lose their 
NUREG-2125 118

effectiveness in an accident. The Rail-Lead cask has a lead gamma shield that could have its 

effectiveness reduced in an accident. Lead is relatively soft as compared to DU or steel and

melts at a considerably lower temperature (330 degrees C, 626 degrees F).

In a hard impact, the lead shield will slump, and a small section of the spent fuel in the cask will 

be shielded only by the steel shells. Figure 5-2 and Figure 5-3 show the maximum individual 

radiation dose at various distances from the damaged cask for a range of gaps in the lead 

shield. In the figures, the dose estimates for the large gaps are depicted on the left side of the 

graph and the fraction of lead shield lost (gap size) increases from left to right. Figure 5-2

shows that doses larger than the external dose that 10 CFR 71.51 allows (0.01 Sv/hour (1 

rem/hour) at 1 meter (3.3 feet) from the cask) occur when the lead shielding gap is more than 

2 percent of the shield.

Figure 5-2 Radiation dose rates to the maximally exposed individual (MEI) from loss of 

lead gamma shielding at distances from 1 to 5 meters from the cask carrying spent fuel

The horizontal axis represents the fraction of shielding lost (the shielding gap).

Figure note: (1 m = 3.3 feet, 1 Sv = 105 mrem)
119 NUREG-2125

Figure 5-3 Radiation dose rates to the MEI from loss of lead gamma shielding at 

distances from 10 to 100 meters from the cask carrying spent fuel

The vertical axis is logarithmic so that all of the doses can be shown on the same graph. The 

horizontal axis represents the fraction of shielding lost (the shielding gap) (1 m = 3.3 feet, 1 Sv = 

105 mrem).

One-in-a-billion accidents (from the first row of Table 5-2) could cause loss of lead shielding that 

results in a dose rate exceeding the regulatory dose rate specified in 10 CFR 71.51. The “one 

in a billion” is a conditional probability, conditional on an accident happening. The total 

probability of such an accident includes both this conditional probability and the probability that 

there will be an accident. The probability of an accident is shown in the right-hand column of

Table 5-6. For example, the probability that an accident resulting in lead shielding loss leading 

to a dose rate greater than 0.01 Sv/hr (1 rem/hr) will happen on the rail route from Maine 

Yankee Nuclear Plant site to Hanford is:

(8.3×10-10)*(0.00214) = 1.74×10-12

or about twice-in-a-trillion Maine Yankee to Hanford shipments.

This very small probability indicates that severe accidents, which are more traumatic to the cask 

than the tests shown in Figure 1-1, are unlikely to happen. Conditions that can cause enough 

lead shielding loss to result in radiation doses to the public above those that 10 CFR 71.51 

allows are extreme conditions.

The overall collective dose risks to the resident population from a lead shielding loss accident 

on the 16 rail routes studied are shown in Table 5-7. These include accidents in which resulting 

dose rates would be within regulatory limits. The doses are the total of rural, suburban, and 

urban doses from Table E-7 in Appendix E. The expected dose to any member of the 

population along the routes, at least 10 meters (33 feet) from the cask, is within the limits of 

10 CFR 71.51. The Indian Point-to-ORNL collective dose risk is comparatively large because 

the suburban and urban populations along this route are about 20 percent higher than along the 

other routes, and the rail accident rate per kilometer is an order of magnitude larger.
NUREG-2125 120

Table 5-6 Average Railcar Accident Frequencies and Accidents Per Shipment on the

Routes Studied

ORIGIN DESTINATION

AVERAGE 

ACCIDENTS 

PER KM

ROUTE 

LENGTH 

(KM)

PROBABILITY OF AN 

ACCIDENT FOR THE 

TOTAL ROUTE

MAINE YANKEE

ORNL 6.5 × 10-7 2125 0.00139

DEAF SMITH 5.8 × 10-7 3362 0.00194

HANFORD 4.2 × 10-7 5084 0.00214

SKULL VALLEY 5.1 × 10-7 4086 0.00208

KEWAUNEE

ORNL 4.3 × 10-7 1395 0.00060

DEAF SMITH 3.3 × 10-7 1882 0.00062

HANFORD 2.4 × 10-7 3028 0.00073

SKULL VALLEY 3.7 × 10-7 2755 0.00103

INDIAN POINT

ORNL 8.8 × 10-6 1264 0.0112

DEAF SMITH 6.2 × 10-7 3088 0.00192

HANFORD 4.4 × 10-7 4781 0.00212

SKULL VALLEY 5.5 × 10-7 3977 0.00217

INL

ORNL 3.6 × 10-7 3306 0.00120

DEAF SMITH 3.5 × 10-7 1913 0.00067

HANFORD 3.2 × 10-7 1062 0.00034

SKULL VALLEY 2.8 × 10-7 455 0.00013

Table 5-7 Collective Dose Risks Per Shipment in Person-Sv for a Loss of Lead Shielding 

Accident Involving a Lead-Shielded Rail Cask (1 Sv=105 mrem)

SHIPMENT ORIGIN ORNL

DEAF 

SMITH HANFORD SKULL VALLEY

MAINE YANKEE 2.5x10-13 2.7x10-13 2.7x10-13 2.6x10-13

KEWAUNEE 1.0x10-13 6.3x10-14 5.4x10-14 1.1x10-13

INDIAN POINT 3.5x10-12 2.4x10-13 2.5x10-13 2.7x10-13

IDAHO NATIONAL LAB 9.9x10-14 4.1x10-14 2.1x10-14 1.5x10-14

Table 5-7 is a summary of Table E-7 in Appendix E. The collective dose (consequence) for 

each route is calculated by dividing the dose risks in Table E-7 by the appropriate probabilities. 

The resulting total consequence for all routes is about 800 person-Sv (80,000 person-rem).

The conditional probability that a lead shielding gap will occur after a fire involving the cask is 

about 10-19. The conditional probability is so small because the following has to occur before a 

fire is close enough to the cask—and burns hot enough and long enough—to do any damage to 

the lead shield:

• The train must be in an accident resulting in a major derailment or the location of the fire 

will be too far removed from the cask to damage the lead shielding.
121 NUREG-2125

• There must be at least one tank car of flammable material involved in the accident 

(either on the train carrying the spent fuel cask or on another train involved in the 

accident).

• The derailment must result in a pileup. By regulation, railcars carrying spent fuel casks 

are required to have buffer cars and are never located directly adjacent to a railcar 

carrying hazardous or flammable material.

• The flammable material must leak out so that it can ignite.

• The pileup must be such that the resulting fire is no further from the cask than a railcar 

length.

The probability of a pileup and the probability that the cask is within a railcar length from the fire

are very small. Assessing the conditional probability without these two events, and considering 

only the more likely events, results in a conditional probability of about 10-10, or approximately 

1 in 10 billion.

Appendix E discusses in detail the event trees and probabilities for fire accidents.

5.4.2 Loss of Neutron Shielding

The type of fuel that can be transported in the three casks considered has relatively low neutron 

emission but does require neutron shielding, usually a hydrocarbon or carbohydrate polymer 

that often contains a boron compound. All three of the casks studied have polymer neutron 

shields. Table 5-8 shows the total radiation dose resulting from a loss of neutron shielding to 

individuals who are approximately 5 meters from a fire-damaged cask for 10 hours. The dose 

allowed by 10 CFR 71.51 is provided for comparison. Neutrons are absorbed by air much 

better than gamma radiation; therefore, external neutron radiation would have an impact on

receptors close to the cask but not on the general public.

Table 5-8 Doses to an Emergency Responder or Other Individual 5 Meters (16.4 feet) 

from the Cask for 10 Hours

Cask

Total Dose in Sv

(mrem)

10-hour allowed total dose in Sv 

(mrem) from 10 CFR 71.51

Truck-DU 0.0073 (730) 0.1(10,000)

Rail-Lead 0.0076 (760) 0.1(10,000)

Rail-Steel 0.0076 (760) 0.1(10,000)

Impacts caused by severe accidents, even those that cause breaches in the seals, will not 

significantly damage the neutron shield. However, the neutron shielding on any of the three 

casks is flammable and could be damaged or destroyed in a fire.

The neutron doses do not exceed the allowable dose cited in the regulation. These doses could 

result from a regulatory fire accident. The conditional probability of this neutron dose is 0.0063 

for a truck fire accident and 0.0000001 for a rail fire accident. The conditional probability of a 

fire for the Truck-DU cask is much higher than that for the two rail casks. These occur, in part,

because truck accidents always include a potential source of fuel (the gas tanks of the truck) 
NUREG-2125 122

whereas many railcar accidents do not involve the locomotive. They also occur, in part,

because of the way the event trees were constructed. The truck event tree does not distinguish 

between minor fires and those severe enough to damage the neutron shielding, while the rail 

event tree only considers severe fires. Therefore the conditional probability of a truck fire is 

quite conservative (overstated). Details are discussed in Section E.3.2 of Appendix E.

The loss of neutron shielding produces a much smaller dose to an emergency responder than 

would happen if there was a loss of gamma shielding of 7 percent. The 10 hour dose to an 

emergency responder at 5 meters (16.4 feet) for the rail lead cask after a loss of neutron 

shielding accident from Table 5-8 is 0.0076 Sv (760 mrem), while multiplying the 5-meter 

(16.4-foot) dose rate in Figure 5-2, 0.007 Sv/hr (700 mrem/hr) by the assumed ten-hour 

exposure time results in a dose of 0.07 Sv (7,000 mrem) after a loss of 7 percent of lead 

shielding accident. Both of these doses are probably overestimates of what would actually 

happen in either of these types of accidents because loss of shielding is relatively easy to 

mitigate, and such actions would likely take place before any extended emergency response 

activities close to the cask were carried out.

5.5 Accidental Release of Radioactive Materials

Radioactive materials released into the environment are dispersed in the air and some deposit 

on the ground. If a spent fuel cask is in a severe enough accident, spent fuel rods can tear or 

be otherwise damaged, releasing fission products and very small particles of spent fuel into the 

cask. If the cask seals are damaged, these radioactive substances can be swept from the 

interior of the cask through the seals into the environment. Release to the environment requires 

the accident be severe enough to damage the fuel rods and release the pressure in the rods or 

there will be no positive pressure to sweep material from the cask into the environment.

Chapters 3 and 4 discuss the potential accidents that could result in such a release. This 

chapter discusses the probability of such accidents and the consequences of the release of 

these radionuclides.

5.5.1 Spent Fuel Inventory

Spent nuclear fuel contains many different radionuclides. The amount of each fission product 

nuclide in the SNF depends on the type of reactor fuel and how much 235U (uranium-235) was in 

the fuel (the enrichment) when it was loaded into the reactor. The amount of each fission 

product in the spent fuel also depends on how much nuclear fission has taken place in the 

reactor (the burnup). Finally, the amount of each radionuclide depends on the time that has 

passed between removal of the fuel from the reactor and transportation in a cask (the cooling 

time) because the fission products undergo radioactive decay during this time. Plutonium, 

americium, curium, thorium, and other actinides produced in the reactor decay to a sequence of 

radioactive elements that are the progeny of the actinide. These progeny increase in 

concentration as the original actinide decays. However, there is never more radioactive 

material as a result of decay than there was initially.
123 NUREG-2125

The fuel studied in this analysis is pressurized-water reactor (PWR) fuel that has “burned” 

45,000 MWD/MTU and cooled for 9 years.

21 The Rail-Lead cask, the only cask studied that 

could release radioactive material in an accident, is certified to carry 26 PWR assemblies.

The spent fuel inventory for accident analysis was selected by normalizing the radionuclide 

concentrations in the spent fuel by radiotoxicity (see Section E.4.1 to Appendix E). The 

resulting inventory is shown in Table 5-9.

Table 5-9 Radionuclide Inventory for Accident Analysis of the Rail-Lead Cask

Radionuclide Name Form

Terabecquerels (TBq) Curies (Ci)

26 Assemblies 26 Assemblies

241Am americium particle 193 5,210

240Pu plutonium particle 184 4,970

238Pu plutonium particle 180 4,850

241Pu plutonium particle 10,440 282,000

90Y yttrium particle 40,400 1,090,000

90Sr strontium particle 40,400 1,090,000

137Cs cesium volatile 50,400 1,360,000

239Pu plutonium particle 71.9 1,940

244Cm curium particle 31.5 852

134Cs cesium volatile 3030 81,800

154Eu europium particle 146 3,950

106Ru ruthenium particle 467 12,600

243Cm curium particle 1.16 31.3

243Am americium particle 0.995 26.9

144Ce cerium particle 180 4,850

242Pu plutonium particle 0.614 16.6

125Sb antimony particle 431 11,600

155Eu europium particle 607 16,400

242mAm americium particle 0.163 4.40

242Am americium particle 0.162 4.38

60Co cobalt CRUD 55.6 1,500

125mTe tellurium particle 105 2,840

234U uranium particle 0.572 15.5

85Kr krypton gas 3,340 90,100

The 60Co (cobalt-60) inventory listed is not part of the nuclear fuel. It is the main constituent of a 

corrosion product, Chalk River unidentified deposit (CRUD), which accumulates on the outside 

of the rods and is formed by corrosion of hardware in the reactor. It is listed here with the 

inventory because it is released to the environment under the same conditions that spent fuel 

particles are released.

 

21 This was approximately the shortest time needed for the fuel to cool sufficiently to meet thermal 

requirements for cask certification. Although relatively short-term, this time was considered somewhat 

typical when this study began. Considerably longer-term spent fuel storage scenarios are now being 

considered, but these longer-term scenarios were not considered in this study.
NUREG-2125 124

5.5.2 Conditional Probabilities and Release Fractions

Seven accident scenarios involving the Rail-Lead cask, described in Chapter 3, could result in 

material releases to the environment. Table 5-10 provides details of these scenarios pertinent

to calculating the resulting doses. Section E.4.3 to Appendix E provides a detailed description 

of the movement of radionuclide particles from fuel rods to the cask interior and from the cask 

interior to the environment. The last row in the table provides the conditional probabilities of 

each of these releases. The total conditional probability that an accident will lead to a release 

for the cask using metal seals is 1.08x10-9 (or one-in-a-billion accidents) and for the cask using 

elastomer seals it is 3.57x10-10.

Table 5-10 Parameters for Determining Release Functions for the Accidents that Would 

Result in Release of Radioactive Materiala

Cask 

Orientation End Corner Side Side Side Side Corner

Rigid Target 

Impact Speed,

kph (mph)

193 (120) 193 (120) 193 (120) 193 (120) 145 (90) 145 (90) 145 (90)

Seal metal metal elastomer metal elastomer metal metal

Cask to 

Environment 

Release 

Fraction

Gas 0.800 0.800 0.800 0.800 0.800 0.800 0.800

Particles 0.70 0.70 0.70 0.70 0.70 0.70 0.64

Volatiles 0.50 0.50 0.50 0.50 0.50 0.50 0.45

CRUD 0.001 0.001 0.001 0.001 0.001 0.001 0.001

Rod to 

Cask 

Release 

Fraction

Gas 0.12 0.12 0.12 0.12 0.12 0.12 0.12

Particles 4.80x10-6 4.80x10-6 4.80x10-6 4.80x10-6 4.80x10-6 4.80x10-6 2.40x10-6

Volatiles 3.00x10-5 3.00x10-5 3.00x10-5 3.00x10-5 3.00x10-5 3.00x10-5 1.50x10-5

CRUD 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Conditional 

Probability 5.96x10-12 3.57x10-11 1.79x10-11 1.79x10-11 3.40x10-10 3.40x10-10 6.79x10-10

a Discussion of the values in this table is given in Section E.4.3 of Appendix E.

5.5.3 Dispersion

Material swept from the cask and released into the environment is dispersed by wind and 

weather. The dispersion is modeled using the accident model in RADTRAN 6, which is a 

Gaussian dispersion model. The release would be at about 1.5 meters above ground level 

since the cask is sitting on a railcar. The gas sweeping from the cask is warmer than ambient; 

therefore, the release is elevated. Under these conditions, the maximum ground level air 

concentration and deposition are 21 meters downwind from the release. The dispersion was 

modeled using neutral weather conditions (Pasquill: stability D, wind speed 4.7 m/sec

(10.5 mph)). It was repeated using very stable meteorology (Pasquill: stability F, wind speed 

0.5 m/sec (1.1 mph)), but the difference was negligible because of the relatively low elevation of 

the release. The maximally exposed individual (MEI) would be located directly downwind from 

the accident, 21 meters (69 feet) from the cask.

Figure 5-4 shows air and ground concentrations of released material as a function of downwind 

distance. The upwind side of the maximum concentration is short because the plume rise is 

very fast. Therefore the x-axis (downwind distance) is foreshortened so that the plume rise and 

gradual decay can be shown in the same graph. The concentrations shown are along the 
125 NUREG-2125

plume centerline and are the maximum concentrations in the plume. The figure shows the 

exponential decrease of airborne concentrations as the downwind distance increases. The 

ground (deposited) concentration also decreases in the downwind direction.

a. Airborne concentration of radioactive material released from the cask in an 

accident

Figure note: 1 meter = 3.3 feet

b. Concentration of radioactive material deposited after release from the cask in an 

accident

Figure 5-4 Air and ground concentrations of radioactive material following a release

Figure note: 1 meter = 3.3 feet

1.E-08

1.E-07

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

0.E+00 5.E+04 1.E+05 2.E+05

Air Concentration

Downwind Distance (m)

Air Concentration (TBq-sec/cu.m.-TBq release) 

1.E-14

1.E-13

1.E-12

1.E-11

1.E-10

1.E-09

1.E-08

1.E-07

0.E+00 5.E+04 1.E+05 2.E+05

Ground deposition

Downwind Distance (m)

Ground Deposition (TBq/sq.m.-TBq release) 
NUREG-2125 126

5.5.4 Consequences and Risks from Accidents Involving Release of Radioactive Material

The dose from accidents that would involve a release is shown in Table 5-11. Section E.4.3 to 

Appendix E provides a detailed discussion on how these values were obtained.

Table 5-11 Doses (Consequences) in Sv to the Maximally Exposed Individual from 

Accidents that Involve a Release (1 Sv=105 mrem)

Cask 

Orientation

Impact Speed,

kph (mph)

Seal

Material Inhalation Resuspension

Cloud-shi

ne

Groundshine Total

End 193 (120) metal 1.6 0.014 8.8x10-5 9.4x10-4 1.6

Corner 193 (120) metal 1.6 0.014 8.8x10-5 9.4x10-4 1.6

Side 193 (120) elastomer 1.6 0.014 8.8x10-5 9.4x10-4 1.6

Side 193 (120) metal 1.6 0.014 8.8x10-5 9.4x10-4 1.6

Side 145 (90) elastomer 1.6 0.014 4.5x10-6 3.6x10-5 1.6

Side 145 (90) metal 1.6 0.014 8.8x10-5 9.4x10-4 1.6

Corner 145 (90) metal 0.73 0.0063 5.1x10-5 9.0x10-4 0.74

The doses listed in Table 5-11 are consequences, not risks. The dose to the MEI is not the sum 

of the total doses. Each cask orientation is a different accident scenario and results in a set of 

internal (includes inhalation and resuspension) and external (includes cloudshine and 

groundshine) doses. Only one accident scenario can happen at a time. These doses would not 

result in either acute illness or death (Shleien et al., 1998). The internal and external doses are 

listed separately because they have different physiological effects. In external doses, the 

receptor would receive a dose only as long as he or she is exposed to the deposited or airborne 

material. If people near the accident are evacuated—and evacuation can take as much as 1

day—then they would only receive an external dose for 1 day. The most significant dose is the 

inhalation dose. All exposures to the dispersed material last until the end of the evacuation 

time, which for this analysis was 24 hours.

Inhaled radioactive particles lodge in the body and are eliminated slowly through physiological 

processes that depend on the chemical form of the radionuclide. The inhaled dose is called a 

“committed” dose because the exposure is for as long as the radionuclide is in the body. The 

activity of the nuclide, however, decreases exponentially as it decays (as shown in the 

Inhalation column of Table 5-11). The resuspension dose is also an inhaled dose. The 

U.S. Nuclear Regulatory Commission (NRC) considers the total effective dose equivalent: the 

sum of the internal and external doses, which allows the doses to be added (the total is shown 

in the last column of Table 5-11).

A pool fire co-located with the cask and burning for a long enough time could severely damage 

the seals. None of the fires analyzed in this report caused sufficient seal damage to result in a 

release of radioactive material. The conditional probability of the series of events required to 

produce the most severe fire scenario analyzed is about 10-19 (discussed in detail in Section 

E.3.1.2 to Appendix E), so analysis of a more severe fire is meaningless. Even a fire offset from 

the cask but close enough to damage lead shielding has a conditional probability of between

10-14 and 10-10.
127 NUREG-2125

The NRC has conducted several analyses of historic fire accidents that include conservative 

assumptions regarding the placement of a cask within those fires (Adkins et al., 2006; Adkins et 

al., 2007, and Bajwa et al., 2012). Consider first the railroad tunnel fire event similar to the 

Baltimore Tunnel Fire (Adkins et al., 2006) (the consequence of the truck tunnel fire was three 

hundred times lower). Based on the rail event tree and the fire branch from Appendix E, the 

conditional probability that a pool fire occurs in a tunnel is 7x10-9

. For this event to be as severe 

as that analyzed, the car carrying flammable liquid would need to be only one car away from the 

car carrying the spent fuel cask (DOT regulations require a buffer car between a spent fuel car 

and other freight). If we assume the train consist is formed randomly, the probability that the 

closest car to the cask car is carrying flammable hazardous material is 0.055 (from DOT, 2010).

Combining these two probabilities gives a net conditional probability of a pool fire in a tunnel as 

close as possible to a cask of 4x10-10

. This probability does not include any information about 

the duration of that pool fire, but if it is assumed that all of these types of fires are as severe as 

the Baltimore Tunnel Fire, this number can be used to estimate the effect on the transportation 

risk assessment. Adkins et al. conservatively estimated this fire could cause a release of 0.3 A2

of material from a rail cask without inner welded canister. This compares to the impact release 

of 8.4 A2 with the same probability. Therefore, even with the conservative assumptions about 

the amount of release and the severity of the fire, including tunnel fires will only increase the 

accident collective dose risk by 4 percent.

The MacArthur Maze highway fire (Bajwa et al., 2012) may lead to a release of radioactive 

material. The truck event tree in Appendix E does not provide sufficient data to determine the 

probability of this event, so investigation of the historical accident record is required. In the past 

20 years there have been two fires similar to this one, so that provides a starting point. There 

are about 400,000 large truck accidents each year (DOT, 2008, Table 2-23), so the probability 

that a severe tanker truck fire occurs below a bridge is approximately 3x10-7

. Neither of these 

two accidents involved another truck, which would be necessary for a spent fuel cask to be 

involved in the accident. From the truck event tree in Appendix E, the conditional probability of

a collision with a gasoline tanker is 2.5x10-3

. Combining these two probabilities gives the 

conditional probability that a truck carrying a spent fuel cask is involved in a MacArthur Maze 

type event is 6x10-10. For this event to cause a release, the spent fuel cask must also be 

co-located with the fire and not protected by intervening structure (such as the tractor, the truck 

bed, or the gasoline tanker). There is no statistical data to provide an estimate for this 

probability, but it is likely to be less than 0.05. Therefore, the probability of a fire such as that 

analyzed in Bajwa et al. (2012) is less than 3x10-11, a factor of 17 less probable than the impact 

accident that results in an 8.4 A2 release. Therefore, this type of accident would not significantly 

change the results of this study unless it resulted in more than 140 A2 of release.

Table 5-12 shows the total collective dose risk from the universe of release accidents. The 

accident with the most severe consequence could result in a release of 8.4 times the amount of 

radioactive material that can be transported in a container that is not accident resistant (8.4

A2s). Such an accident would result in a collective dose of 6.8 person-Sv to an exposed 

population of 58,000, calculated by multiplying RADTRAN output for dose and plume footprint 

area by a population density of 41.46 persons/km2 (107.4 persons/mi2

) (the U.S. average minus 

Alaska). Of the three casks in this study, only the Rail-Lead cask could result in a release in 

each type of accident considered.

These dose risks are negligible by any standard.
NUREG-2125 128

Table 5-12 Total Collective Dose Risk (person-Sv) for Release Accidents Per Shipment 

for Each Route (1 Sv=105 mrem)

ORNL DEAF SMITH HANFORD

SKULL 

VALLEY

MAINE YANKEE 3.5x10-14 4.1x10-14 3.2x10-14 3.0x10-14

KEWAUNEE 1.8x10-14 1.2x10-14 5.4x10-15 1.4x10-14

INDIAN POINT 1.5x10-11 5.9x10-13 5.3x10-13 1.9x10-13

IDAHO NATIONAL LAB 9.4x10-14 1.5x10-13 4.1x10-14 2.7x10-13

Table 5-13, which is the same as Table 5-7, shows total collective dose risks from loss-of-lead 

shielding accidents. Table 5-7 is repeated here for ease of comparison. The sum of the two

tables is shown in Table 5-14.

Table 5-13 Total Collective Dose Risk (person-Sv) for Each Route from a Loss of Lead 

Shielding Accident (1 Sv=105 mrem)

ORNL DEAF SMITH HANFORD

SKULL 

VALLEY

MAINE YANKEE 2.5x10-13 2.7x10-13 2.7x10-13 2.6x10-13

KEWAUNEE 1.0x10-13 6.3x10-14 5.4x10-14 1.1x10-13

INDIAN POINT 3.5x10-12 2.4x10-13 2.5x10-13 2.7x10-13

IDAHO NATIONAL LAB 9.9x10-14 4.1x10-14 2.1x10-14 1.5x10-14

Table 5-14 Total Collective Dose Risk (person-Sv) from Release and Loss of Lead 

Shielding Accidents (1 Sv=105 mrem)

ORNL DEAF SMITH HANFORD

SKULL 

VALLEY

MAINE YANKEE 2.8x10-13 3.1x10-13 3.0x10-13 2.9x10-13

KEWAUNEE 1.2x10-13 7.6x10-14 5.9x10-14 1.2x10-13

INDIAN POINT 1.9x10-11 8.3x10-13 7.9x10-13 4.6x10-13

IDAHO NATIONAL LAB 1.9x10-13 1.9x10-13 6.1x10-14 2.9x10-13

Table 5-15 shows the total collective dose risk for an accident involving the Rail-Lead shielded 

cask in which there is no loss of lead shielding or release. Since the collective dose risk for this 

type of accident depends on the transport index (TI), the collective dose risk from an accident 

involving the truck cask would be the same. For the Rail-Steel cask carrying canistered fuel, the 

collective dose risk would be slightly less because the TI is smaller. For this analysis, the cask 

was assumed to be immobilized for 10 hours.
129 NUREG-2125

Table 5-15 Total Collective Dose Risk (person-Sv) from No-Release, No-Loss of 

Shielding Accidents Involving the Rail-Lead Cask (1 Sv=105 mrem)

ORNL

DEAF 

SMITH HANFORD

SKULL 

VALLEY

MAINE YANKEE 6.3x10-5 6.6x10-5 6.4x10-5 6.0x10-5

KEWAUNEE 6.8x10-5 6.9x10-5 6.1x10-5 7.0x10-5

INDIAN POINT 8.3x10-5 6.9x10-5 6.5x10-5 6.9x10-5

IDAHO NATIONAL LAB 6.8x10-5 6.6x10-5 6.8x10-5 6.8x10-5

Note: (See Table 5-4)

Table 5-16 shows the collective accident dose risk for the 16 rail routes from loss of neutron 

shielding for the Rail-Lead cask. This table is extracted from Table E-14 in Appendix E.

Table 5-16 Total Collective Dose Risk (person-Sv) from Loss of Neutron Shielding for 

Accidents Involving the Rail-Lead Cask (1 Sv=105 mrem)

ORNL DEAF SMITH HANFORD

SKULL 

VALLEY

MAINE YANKEE 8.90x10-14 1.16x10-13 1.13x10-13 1.12x10-13

KEWAUNEE 3.48x10-14 3.41x10-14 3.72x10-14 5.46x10-14

INDIAN POINT 6.94x10-13 1.13x10-13 1.14x10-13 1.22x10-13

IDAHO NATIONAL LAB 5.88x10-14 3.48x10-14 1.09x10-14 7.15x10-15

5.6 Potential Criticality

Spent fuel casks are required to demonstrate that they will remain subcritical following the 

hypothetical accident sequence of 10 CFR 71.73, “Hypothetical Accident Conditions.” In a 

transportation risk assessment, it must also be determined if the cask remains subcritical 

following more severe accidents. Because spent fuel casks are under moderated (Elam et al., 

2003) a criticality event requires the addition of moderator (water) into the cask. For water to 

get into the cask, there must be a failure in the seals. In the accidents investigated in this study, 

only impacts into hard rock surfaces at speeds greater than 93 kph (60 mph) have the potential 

for failing the seals. Impacts into water at speeds up to the maximum recorded accident speed

cannot cause a seal failure due to the lack of shear strength of the water. Therefore, for 

addition of moderator to be possible the cask would have to first impact a hard rock surface and 

then fall into a body of water. Even if the cask fell into a body of water after an impact caused 

the seal to fail, it would have to be in the right configuration for sufficient water to enter the cask 

that moderation is possible. The starting conditional probability for this is 4x10-10 accidents that 

produce a seal failure. The rail event tree does not provide any information about the probability 

of water, but the truck event tree gives 0.009 as the probability that there is water under a 

bridge. This is likely an overestimation of the chance that there is water near hard rock surface.

Even if water is present, the cask must rebound from the hard rock surface in such a way that it 

lands in the water. Then, if it lands in the water, the water must be deep enough to submerge 

the cask. Combined, the conditional probability that the cask gets flooded if there is a seal 

failure has to be less than 10-5

. Even this is not a sufficient condition for there to be a criticality 

event. The fuel rubble must still be arranged in a manner that supports criticality. Given these 

extremely low probabilities, it can be deduced that a criticality event is not credible.
NUREG-2125 130

5.7 Chapter Summary

The conclusions that can be drawn from the risk assessment that apply to the three types of 

casks studied as presented in this chapter are listed below.

• The 16 truck and 16 rail routes selected for study are an adequate representation of 

U.S. routes for SNF transportation, and there was relatively little variation in the risks per 

kilometer over these routes.

• The overall collective dose risks are very small.

• The collective dose risks for the two types of extra-regulatory accidents (accidents 

involving a release of radioactive material and loss-of-lead-shielding accidents) are 

negligible compared to the risk from a no-release, no-loss-of-shielding accident. There 

is no expectation of any release from spent fuel shipped in inner welded canisters from 

any impact or fire accident analyzed.

• The collective dose risk from loss of lead shielding is comparable to the collective dose 

risk from a release, though both are very small. The doses and collective dose risks 

from loss of lead shielding are smaller than those calculated in NUREG/CR-6672, 

“Re-examination of Spent Fuel Shipment Risk Estimates,” because of better precision in 

the finite element (FE) modeling and a more accurate model of the dose from a gap in 

the lead shield.

• The conditional risk of either a release or loss of lead shielding from a fire is negligible.

• The consequences (doses) of some releases and some loss of lead shielding scenarios 

that occur with extremely low probability are larger than those cited in 10 CFR 71.51

“Additional Requirements for Type B Packages”; but would not result in an acute 

lethality. Only one in a billion accidents would result in these doses.
131 NUREG-2125

6. FINDINGS AND CONCLUSIONS

The U.S. Nuclear Regulatory Commission (NRC) first assessed the health and safety impacts of 

spent fuel transportation in NUREG-0170, “Final Environmental Statement on the 

Transportation of Radioactive Material by Air and Other Modes,” published in 1977. Based on 

NUREG-0170, the Commissioners concluded that the regulations in force at the time of the 

environmental impact statement were “adequate to protect the public against unreasonable risk 

from the transport of radioactive materials” (46 FR 21629; April 13, 1981). The present 

document presents the most recent NRC assessment of the risks of transporting commercial 

spent nuclear fuel (SNF). Both NUREG-0170 and this document estimate the radiological 

impact for spent fuel transport conducted in compliance with Title 10 of the Code of Federal 

Regulations (10 CFR) Part 71 regulations, “Packaging and Transportation of Radioactive 

Material.” Other NRC studies, including the Modal Study (Fischer et al., 1987) and 

NUREG/CR-6672, “Re-examination of Spent Fuel Shipment Risk Estimates,” (Sprung et al., 

2000), also provided spent fuel shipment risk assessments.

Regulations and regulatory compliance analyses are different from risk assessments. A 

regulation must be conservative because its purpose is to ensure safety, and 10 CFR Part 71, 

which regulates transportation, requires a conservative estimate (i.e., overestimate) of the 

damage to a cask in an accident and the radiation emitted from the cask during routine 

transportation. The original environmental assessment for 10 CFR Part 71, NUREG-0170, was 

also conservative, but for a different reason: only limited data were available to perform the 

assessment. Therefore, NUREG-0170 deliberately used conservative parameter estimates. 

The NRC’s conclusion was that NUREG-0170 showed that even with conservative assumptions 

transportation of radioactive materials provide adequate public safety.

When an assessment is used to inform regulation, it should be as realistic as possible to provide 

information necessary to confirm or revise the regulations it informs. Realistic assessment 

depends on data availability and accurate and precise modeling techniques, which have 

become increasingly available since 1977. Consequently, the Modal Study and 

NUREG/CR-6672 made progress in assessing transportation risks more realistically. As a 

result, both the calculated consequences and risks of radioactive materials transportation 

decreased. The decrease in risk means that the regulations provide for a greater level of safety 

than previously recognized.

The present study is more accurate than previous analyses. Certified spent fuel casks are 

analyzed, rather than generic designs. Recent (2005 or later) accident frequency and 

population data are used in the analyses and the modeling techniques also were upgraded. 

This study, the Spent Fuel Transportation Risk Assessment, is another step toward building a 

complete picture of SNF transportation radiological safety. It also presents the current state of 

art for such analyses. The results of this study are compared with preceding risk assessments 

in the figures that follow.

6.1 Routine Transportation

Figure 6-1 and Figure 6-2 show results of routine truck and rail transportation of a single 

shipment of SNF using the single example route from NUREG-0170, the average of the 

200 routes from NUREG/CR-6672, and the average of the 16 truck or rail routes from this study.
NUREG-2125 132

Figure 6-1 Collective doses (person-Sv) from routine truck transportation

Figure 6-1 plots average collective radiation dose (person-sieverts (Sv)) from truck 

transportation, and Figure 6-2 plots average collective radiation dose from rail transportation.

These average doses include doses to the population along the route, doses to occupants of 

vehicles sharing the route, doses at stops, and doses to vehicle crew and other workers. Doses 

without the crew and worker dose (labeled public only) are also shown.

Collective doses from routine transportation directly depend on the population along the 

route and the number of other vehicles that share the route, and, inversely, on vehicle 

speed. Doses to occupants of vehicles that share the route depend inversely on the 

square of the vehicle speed.

NUREG-0170 results for truck transportation were based on a single long route; constant values 

of rural, suburban, and urban population densities; different and conservative vehicle speeds on 

rural, urban, and suburban roads; a fixed rate of vehicle stops; and 1975 estimates of vehicle 

density (vehicles per hour), all of which led to conservative results. NUREG/CR-6672 used 

more realistic distributed route lengths, population densities, vehicle occupancy and density, 

vehicle dose rate and stop time, and the means of the distributions as parameters. 

Figure 6-1 shows that the conservatism was decreased by more than a factor of three.

The collective average dose in the present study is larger than the NUREG/CR-6672 result 

because present populations are generally larger, particularly along rural routes, and vehicle 

densities are much greater (see Chapter 2). The higher vehicle speeds used in the present 

study offset these increases. The largest contributor to higher doses in this study is the 

parameters used for stops. In this study, stops were assumed to occur every 845 kilometers

(525 miles) versus 1,290 kilometers (802 miles) and last for 50 minutes versus 30 minutes. The 

combination of these two factors results in a 2.5 times increase in the stop dose. This is 

especially significant because the greatest contributor to the public collective dose is from 

2.4x10-3

1.4x10-3

1.6x10-3

4.4x10-4
133 NUREG-2125

people sharing truck stops with the cask (56 percent of the collective dose). The second largest 

contributor is from people sharing the highway with the cask (38 percent of the collective dose).

Residents along the route only receive 6 percent of the collective dose and residents near truck 

stops only receive 1 percent.

Figure 6-2 shows the differences between NUREG 0170, NUREG/CR-6672, and the present 

study for calculating average doses to the public for routine rail transportation.

Figure 6-2 Collective doses (person-Sv) from routine rail transportation

The difference in dose between the Rail-Lead cask and the Rail-Steel cask occurs because the 

latter cask has a smaller external dose rate (Chapter 2). The differences in crew doses 

between the studies reflect the considerable difference between the methods the different 

studies used.

Differences in the collective doses from routine transportation between the cited studies are not 

the result of differences in external radiation from the spent fuel casks. The 1975 version of 

10 CFR Part 7122 specified the same limit on external radiation (the transport index (TI)) as Part 

71 specifies today. Instead, these differences reflect improvements to modeling methods and 

the increase in population and traffic levels. Also, the groups of people exposed that various 

studies considered has changed. For example, this study includes inspector doses not included 

in the other two studies.

 

22 A copy is provided in NUREG-0170.

4.9x10-4

2.6x10-4

2.0x10-4

3.2x10-4

1.6x10-4

6.7x10-4
NUREG-2125 134

The differences in results are primarily due to vehicle speed, population and vehicle densities, 

and differences in calculating train crew and railyard worker doses. These differences are 

summarized below.

•

• Differences in vehicle speed. The faster the cask moves past a receptor, the less that 

receptor is exposed. NUREG-0170 and NUREG/CR 6672 used 80 kph (50 mph) for all 

truck routes and 64 kph (40 mph) on rural rail routes, 40 kph (25 mph) on suburban rail 

routes, and 24 kph (15 mph) on urban rail routes. The truck speeds used in this study 

are 108 kph (67 mph) on rural routes, 102 kph (63 mph) on suburban routes, and 97 kph 

(60 mph) on urban routes. The rail speeds are 40 kph (25 mph) on rural and suburban 

routes and 24 kph (15 mph) on urban routes. The present speeds are based on data 

instead of the estimated values previous studies used.

• Differences in populations along the routes. NUREG-0170 used 6 persons per km2

(15.5 persons per mi2

) for rural populations, 719 per km2 (1862 per mi2) for suburban 

routes, and 3,861 per km2 (10,000 per mi2

) for urban routes. NUREG/CR-6672 used 

1990 census data provided by the codes HIGHWAY and INTERLINE and used the 

mean values of Gaussian distributions of population densities on 200 routes in the 

United States. This study uses 2000 census data provided by WebTRAGIS (Johnson 

and Michelhaugh, 2002), with some updates based on 2008 Bureau of Census data

(U.S. Bureau of the Census, 2008), for the rural, suburban, and urban truck and rail 

route segments in each State traversed for each of the 16 origin/destination pairs 

studied. The variation from the NUREG-0170 values is considerable.

• Differences in vehicles per hour on highways. NUREG-0170 and NUREG/CR-6672 both 

used the 1975 values of 470 vehicles per hour on rural routes, 780 on suburban routes, 

and 2,800 on urban routes. This study used 2002 state vehicle density data for each 

State traversed. The national average vehicle density is 1,119 vehicles per hour on rural 

routes, 2,464 on suburban routes, and 5,384 on urban routes. This large difference in 

vehicle density contributes to the difference in collective doses for routine truck 

transportation between NUREG/CR-6672 and this study.

• Differences in calculating doses to rail crew. NUREG-0170 estimated the distance 

between the container carrying radioactive material and the crew member to calculate 

doses to rail and railyard crew. NUREG/CR-6672 used the Wooden (1980) calculation 

of doses to railyard workers and did not calculate a dose to the train crew. This study 

calculated all doses using the formulations in RADTRAN 6, calculated an in-transit crew 

dose, used an updated value for the time of a classification stop (27 hours instead of

30 hours), and used in-transit stop times from WebTRAGIS instead of the stop dose 

formula, which is pegged to total trip length and used in NUREG/CR-6672. The 

in-transit crew dose calculated in this study was small enough that it contributed a 

negligible amount to these doses.

Dose to the maximally exposed individual (MEI) is a better indication than collective dose of the 

radiological effect of routine transportation. The same event results in different collective doses 

depending on the population affected, which varies by location and the consideration of rush 

hour. The MEI dose is shown in Figure 6-3 for NUREG-0170 and for the three cask types of 

this study. NUREG/CR-6672 did not calculate this dose for routine transportation. The 

reduction is because of the higher speeds this study used.
135 NUREG-2125

Figure 6-3 Maximum individual dose (Sv) from routine transportation

6.2 Transportation Accidents

Radiological accident risk is expressed in units of “dose risk” that include the probability of an 

accident and the conditional probability of certain types of accidents. Dose units (Sv) are used 

because probability is a unitless number. The dose risk to a population (as distinct from the 

dose risk to an individual) is collective dose risk, which has units of person-Sv. NUREG-0170, 

NUREG/CR-6672, and this study all used the RADTRAN version available at the time of the 

study to calculate dose risk, but the input parameters differed significantly. These parameters 

were based primarily on the detail and precision of the assessment of package performance,

modeling improvements, and the availability of accident and population data. In addition, 

improvements in RADTRAN and other modeling codes described in earlier chapters resulted in 

a more accurate analysis of cask behavior in an accident.

The results shown in Figure 6-4 and Figure 6-5 for this study are averages over the 16 rail

routes studied. As discussed in Chapters 3, 4, and 5, a lead-shielded rail cask, the Rail-Lead 

cask in this study, is the only cask type of the three studied that indicated either release of 

radioactive material or loss of lead gamma shielding in an accident.

The results in Figure 6-4 reflect the different amounts of radioactive material released and the 

different amounts of lead shielding lost as estimated in the respective studies. NUREG-0170 

used a scheme of 8 different accident scenarios; 4 postulated release of the entire releasable 

contents of the cask, 2 postulated no release, 1 postulated a 10 percent release, and 

1 postulated a 1 percent release. The range of conditional probabilities varied from 1×10-5 for 

the most severe (100 percent release) accident to 80 percent for the 2 no-release scenarios. 

The NUREG-0170 “universe” of accidents and their consequences was primarily based on 

engineering judgment, which was clearly conservative.

6.7x10-9

4.3x10-9

5.7x10-9

1.0x10-8
NUREG-2125 136

Figure 6-4 Average collective dose risk from release and loss of lead shielding (LOS)

accidents

Figure note: The LOS bars are not to scale.

NUREG/CR-6672 analyzed the structural and thermal behavior of four generic cask designs—

two truck and two rail casks—in great detail, and analyzed the behavior of the five groups that 

best describe the physical and chemical nature of the radioactive materials potentially released 

from SNF through the casks. These five groups are particulate matter, semi-volatile 

substances, ruthenium, gas, and CRUD. The spent fuels considered were high burnup and low 

burnup pressurized-water reactor (PWR) and boiling-water reactor (BWR) fuel. This analysis 

resulted in 19 truck accident scenarios and 21 rail accident scenarios, each with an attendant 

possibility, including a no-release scenario, which had better than 99.99 percent probability.

The present study followed the analytical outline of the NUREG/CR 6672 analysis, but analyzed 

the structural and thermal behavior of a certified lead-shielded cask design loaded with fuel that 

the cask is certified to transport. Instead of the 19 truck scenarios and 21 rail scenarios that 

included potential releases of radioactive material, the current study resulted in only 7 rail 

scenarios that included releases, as described in Chapters 3 and 5. The seals are the only 

parts of the cask structure that could be damaged enough to allow a release. Release could 

take place through the seals only if the seals fail and if the cask is carrying uncanistered fuel. 

No potential truck accident scenario resulted in seal failure, nor did any fire scenario. In the 

present study, only the Rail-Lead cask response to extremely severe accident conditions 

resulted in a release. A comparison of the collective dose risks from potential releases in this 

study to both NUREG-0170 and NUREG/CR-6672 is appropriate, since the latter two studies 

considered only potential releases. The collective dose risks decrease with each succeeding 

study as expected, since the overall conditional probability of release and the quantity of 

material potentially released decreases with each successive study. The decrease in release is 

primarily because of the replacement of conservative estimates of cask performance in an 

1.1x10-12

9.4x10 4.9x10 -6 -4

4.7x10-13

9.1x10-13
137 NUREG-2125

accident with FE analyses of cask performance in an accident. Basically, in succeeding studies, 

the calculated performance of the cask is better (it releases less) than estimated previously.

The collective dose risk from a release depends on dispersion of the released material, which 

either remains suspended in the air, producing cloudshine, or is deposited on the ground, 

producing groundshine, or is inhaled. All three studies used the same basic Gaussian 

dispersion RADTRAN model, although the RADTRAN 6 model is much more flexible than the 

previous versions and can model elevated releases. NUREG-0170 only calculated doses from 

inhaled and resuspended material. NUREG/CR-6672 included groundshine and cloudshine as 

well as inhaled material, but overestimated the dose from inhaled resuspended material. The 

combination of improved assessment of cask damage and dispersion modeling has resulted in 

the decrease in collective dose risk from releases shown in Figure 6-4.

Frequently, public interest in the transportation of SNF focuses solely on the consequences of 

possible accidents without regard to the likelihood that an accident will occur. The maximum 

estimated consequence, based on average population density, from the accident with the 

largest release is 2.18 person-Sv (218 person-rem). This consequence is orders of magnitude 

less than the 110 person-Sv (11,000 person-rem) in NUREG-0170 and the 9,000 person-Sv 

(900,000 person-rem) estimated in NUREG/CR-6672 Figure 8.27. The reduction in 

consequence is the result of using the actual spent fuel being shipped, a smaller release 

fraction, and improvements in the RADTRAN model. The maximum estimated dose to any 

person from this accident is 1.6 Sv (160 rem), and would not cause an acute fatality (fatality 

within two monthes of receiving the radiation dose).

NUREG-0170 did not consider a loss of spent fuel cask lead shielding, which can result in a 

significant dose increase from gamma radiation emitted by the cask contents.

NUREG/CR-6672 analyzed 10 accident scenarios in which the lead gamma shield could be 

compromised and then calculated a fractional shield loss for each. An accident dose risk was 

calculated for each potential fractional shield loss.

The present study followed the same general calculation scheme, but with a more sophisticated 

model of gamma radiation from the cask due to the damaged shield and using 18 potential 

accident scenarios instead of 10. Most of the difference between the NUREG/CR-6672 dose 

risks from shielding loss and this study is the inclusion of accident scenarios that have a higher 

conditional probability (i.e., accidents that are more likely to happen) than any scenarios in 

NUREG/CR-6672. The consequence of a loss of lead shielding estimated in NUREG/CR-6672 

Table 8.13 is 41,200 person-Sv (4,120,000 person-rem), about 100 times the 690 person-Sv 

(6,900 person-rem) estimated in this study because of the more conservative loss of lead 

shielding model used in NUREG/CR-6672 and the overestimation of the amount of lead slump 

in that study. Loss of lead shielding clearly affects only casks with a lead gamma shield; casks 

using DU or thicker steel shielding would not be affected.

More than 99.999999 percent of potential accident scenarios do not affect the cask at all and 

would not result in a release of radioactive material or an increased dose from loss of lead 

shielding. However, these accidents would result in an increased external radiation dose from 

the cask to the population near the accident because the cask would remain at the accident 

location until it could be moved. A nominal 10-hour delay in moving the cask was assumed for 

this study. The resulting collective dose risk is shown in Figure 6-5 for all three cask types 

studied. Even including this additional consequence type, the accident collective dose risk from 

this study is less than that reported in either NUREG-0170 or NUREG/CR-6672.
NUREG-2125 138

Figure 6-5 Average collective dose risk from accidents that have no impact on the cargo

For the most probable accident, one that does not involve either loss of shielding or release of 

radioactive material, the most significant consequence, in addition to any nonradiological 

consequence of the accident itself, is the external dose from a cask immobilized at the accident 

site.

Figure 6-5 shows the average collective dose risks from this type of accident for the 16 truck 

routes and 16 rail routes studied. The most significant parameters contributing to this dose risk 

are the accident frequency and the length of time that the cask sits at the accident location. 

Even in this case, the significant parameter in the radiological effect of the accident is not the 

amount or rate of radiation released, but the exposure time.

Each of the three transportation risk assessments conducted for the NRC show that the NRC 

regulation of transportation casks ensures safety and health. The use of data in place of 

engineering judgment shows that accidents severe enough to cause a loss of shielding or 

release of radioactive material are improbable and the consequences of such unlikely accidents 

would require mitigation, but would not result in large radiation doses to even the MEI. 

Moreover, these consequences depend on the size of the population exposed rather than on 

the radiation or radioactive material released.

6.3 Effect of Transportation of Higher Burnup Spent Nuclear Fuel

At the time the analyses for this report were completed, the maximum burnup for the spent fuel 

transported in any of the casks was 45 gigawatt days per metric ton uranium (GWD/MTU). 

Current reactor operations result in spent fuel with burnup levels higher than this. A detailed 

examination of the effect of the higher burnup levels is outside the scope of this document, but 

this section provides some general insights on expected changes resulting from transporting 

these higher burnup spent fuels.

2.2x10-5

1.7x10-5

2.4x10-5
139 NUREG-2125

The regulatory external dose rates must still be met, so there is no effect on incident-free 

transport or on the results from accidents that do not result in cask damage. The higher burnup 

fuel will have to be cooled longer before it is transported to meet the cask’s decay heat and 

dose rate limits and the expected radiation emanating from the fuel should not change

substantially (it cannot increase above the regulatory surface dose rates, and the casks studied 

here are either at that limit or very near to it). Therefore, results from loss of shielding accidents 

will not change significantly. In all of the accidents that are severe enough to have a release 

path from the cask, the acceleration level is high enough to fail the cladding of all of the fuel, 

whether it is high burnup or not. Higher burnup fuel has a rim layer with a higher concentration 

of radionuclides. This will lead to the rod-to-cask release fraction being higher but will not affect 

the cask-to-environment release fraction (Table 5-10 gives the release fractions used in this 

study.). In addition, the isotopic mixture of the higher burnup fuel cooled for a longer period of 

time will have more transuranic isotopes and less fission product. For example, the inventory of 

241Am (Americium-241) goes up from 193 Tetrabecquerels (TBq) at 45 GWD burnup to 1,980 

TBq at 60 GWD burnup (5,210 curies (Ci) to 53,400 Ci) and the inventory of 90Sr (strontium-90) 

drops from 40,400 TBq to 30,600 TBq (1,090,000 Ci to 826,000 Ci). Insufficient data exists to 

accurately estimate the rod-to-cask release fractions for higher burnup fuel. If the release 

fractions remain the same, the effect of the change in radionuclide inventory increases the 

number of A2s released by a factor of 5.9. This increase does not alter the conclusions of this 

study.

6.4 Findings and Conclusion

The following findings are reached from this study:

• The collective doses from routine transportation are very small. Theses doses are about 

four to five orders of magnitude less than collective background radiation doses.

• The routes selected for this study adequately represent the routes for SNF transport, 

and there was relatively little variation in the risks per kilometer over these routes.

• Radioactive material would not be released in an accident if the fuel is contained in an 

inner welded canister inside the cask.

• Only rail casks without inner welded canisters would release radioactive material and 

only then in exceptionally severe accidents.

• If there were an accident during a spent fuel shipment, there is only about a 

one-in-a-billion chance that the accident would result in a release of radioactive material.

• If there were a release of radioactive material in a spent fuel shipment accident, the dose 

to the MEI would be less than 2 Sv (200 rem) and would not result in an acute lethality.

• The collective dose risks for the two types of extra-regulatory accidents (accidents 

involving a release of radioactive material and loss of lead shielding accidents) are 

negligible compared to the risk from a no-release, no-loss of shielding accident.

• The risk of loss of lead shielding from a fire is negligible.
NUREG-2125 140

• None of the fire accidents investigated in this study resulted in a release of radioactive 

material.

Based on these findings, this study reconfirms that radiological impacts from spent fuel 

transportation conducted in compliance with NRC regulations are low. They are, in fact,

generally less than previous, already low, estimates. Accordingly, with respect to spent fuel 

transportation, this study reconfirms the previous NRC conclusion that regulations for 

transportation of radioactive material are adequate to protect the public against unreasonable 

risk.
APPENDIX A

CASK DETAILS AND CERTIFICATES OF COMPLIANCE

A-1

A.1 Cask Descriptions

This appendix provides a listing and brief description of the spent fuel transport casks that were 

considered for evaluation in this risk analysis. It also provides the certificates of compliance for 

those casks selected for evaluation.

A.1.1 Truck Casks

GA-4

The Steel-DU-Steel cask design is stiffer than lead casks and has 

smaller deformations.

The 4 PWR assembly capacity of this cask makes it the likely 

workhorse truck cask for any large transportation campaign.

Elastomeric seals (ethylene propylene) allow larger closure 

deformations before leakage.

Truck casks have hydrogenous neutron shielding.

Larger capacity allows for larger radioactive material inventory and 

possible larger consequences from an accident.

The design is from the late 1980s; General Atomics used finite 

element analyses and model test results in certification.

The depleted uranium (DU) shielding is made from five segments, 

which have been shown to not result in gaps during the regulatory 

accident sequence, but which could possibly result in gaps during 

extra-regulatory accidents.

The cask body has a square cross-section, which provides more 

possible orientations.

The cask has an aluminum honeycomb impact limiter.

NAC-LWT

The steel-lead-steel design is relatively flexible, which should result 

in plastic deformation of the body before seal failure.

The NAC-LWT cask contains either a single pressurized-water 

reactor (PWR) assembly or two boiling-water reactor (BWR) 

assemblies.

The cask has both elastomeric and metallic seals. The low 

compression of the elastomeric seal (metallic is primary) allows little 

closure movement before leakage but may perform better in a fire.

The lead shielding could melt during severe fires, leading to loss of 

shielding.

With liquid neutron shielding, the tank is likely to fail in 

extra-regulatory impacts.

The bottom end impact limiter is attached to the neutron shielding 

tank, making side drop analysis more difficult.

The NAC-LWT has an aluminum honeycomb impact limiter.

The cask is very similar to the generic steel-lead-steel cask from 

NUREG/CR-6672, “Reexamination of Spent Fuel Shipment Risk 

Estimates.”

The cask is being used for foreign research reactor shipments.
A-2

A.1.2 Rail Casks

NAC-STC

The cask has a steel-lead-steel design, which is relatively flexible 

and should result in plastic deformation of the body before seal 

failure.

The NAC-STC cask is certified for both direct loaded fuel and fuel in 

a welded canister.

The cask can contain either 26 directly loaded PWR assemblies or 

one transportable storage container (three configurations, all for 

PWR fuel).

The cask can have either elastomeric or metallic seals. A 

configuration must be chosen for analysis.

The lead shielding used could melt during severe fires, leading to 

loss of shielding.

The NAC-STC has polymer neutron shielding.

The cask has a wood impact limiter (redwood and balsa).

This cask is similar to the steel-lead-steel rail cask from 

NUREG/CR-6672.

Two casks have been built and are being used outside of the United 

States.

NAC-UMS

The NAC–UMS cask has a steel-lead-steel design, which is 

relatively flexible and should result in plastic deformation of the body 

before seal failure.

The fuel is in a welded canister.

Baskets for 24 PWR assemblies or 56 BWR assemblies are 

available.

Elastomeric seals allow larger closure deformations before leakage.

The lead shielding could melt during severe fires, leading to loss of 

shielding.

The cask has polymer neutron shielding.

The cask has a wood impact limiter (redwood and balsa).

The cask is similar to the steel-lead-steel rail cask from 

NUREG/CR-6672.

The NAC-UMS cask has never been built.

HI-STAR 100

The HI-STAR 100 cask has a layered all-steel design.

The fuel is in a welded canister.

Baskets for 24 PWR assemblies or 68 BWR assemblies are 

available.

The cask has metallic seals, resulting in smaller closure 

deformations before leakage.

The cask has polymer neutron shielding.

The cask has aluminum honeycomb impact limiters.

At least seven of these casks have been built and are being used for 

dry storage; no impact limiters have been built.

The HI-STAR 100 is proposed as the transportation cask for the 

Private Fuel Storage facility.(PFS)
A-3

TN-68

The TN-68 cask has a layered all-steel design.

Directly loaded fuel is used in the cask.

The TN-68 has 68 BWR assemblies.

Metallic seals result in smaller closure deformations before leakage.

The cask has polymer neutron shielding.

The cask has a wood impact limiter (redwood and balsa).

At least 24 TN-68 casks have been built and are being used for dry 

storage; no impact limiters have been built.

MP-187

The MP-187 cask has a steel-lead-steel design, which is relatively 

flexible and should result in plastic deformation of the body before 

seal failure.

The fuel is in a welded canister.

There are 24 PWR assemblies.

Metallic seals result in smaller closure deformations before leakage.

The MP-187 has hydrogenous neutron shielding.

The cask has aluminum honeycomb and polyurethane foam impact 

limiters (chamfered rectangular parallelepiped).

This cask has never been built.

MP-197

The MP-197 cask has a steel-lead-steel design, which is relatively 

flexible and should result in plastic deformation of the body before 

seal failure.

The fuel is in a welded canister.

There are 61 BWR assemblies.

Elastomeric seals allow larger closure deformations before leakage.

The MP-197 has hydrogenous neutron shielding.

The cask has a wood impact limiter (redwood and balsa).

This cask has never been built.

TS125

The TS125 cask has a steel-lead-steel design, which is relatively 

flexible and should result in plastic deformation of the body before 

seal failure.

The fuel is in a welded canister.

There are basket designs for 21 PWR assemblies or 64 BWR 

assemblies.

Metallic seals result in smaller closure deformations before leakage.

The TS125 has polymer neutron shielding.

The cask has aluminum honeycomb impact limiters.

This cask has never been built.
A-4

A.2 Certificates of Compliance

Certificates of Compliance follow on pages A-5 to A-76.
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION 

NUMBER PAGE PAGES 

9261 7 71-9261 USA/9261/B(U)F-96 1 OF 7 

2. PREAMBLE 

a. This certificate is issued to certify that the package (packaging and contents) described in Item 5 below meets the applicable safety standards 

set forth in Title 10, Code of Federal Regulations, Part 71, “Packaging and Transportation of Radioactive Material.” 

b. This certificate does not relieve the consignor from compliance with any requirement of the regulations of the U.S. Department of 

Transportation or other applicable regulatory agencies, including the government of any country through or into which the package will be 

transported. 

3. THIS CERTIFICATE IS ISSUED ON THE BASIS OF A SAFETY ANALYSIS REPORT OF THE PACKAGE DESIGN OR APPLICATION 

a. ISSUED TO (Name and Address) b. TITLE AND IDENTIFICATION OF REPORT OR APPLICATION 

Holtec International 

Holtec Center 

555 Lincoln Drive West 

Marlton, NJ 08053 

Holtec International Report No. HI-951251.Safety 

Analysis Report for the Holtec International Storage, 

Transport, And Repository Cask System (HI-STAR 

100 Cask System) Revision 12, dated October 9, 

2006, as supplemented.

4. CONDITIONS 

 This certificate is conditional upon fulfilling the requirements of 10 CFR Part 71, as applicable, and the conditions specified below. 

5

(a) Packaging 

 (1) Model No.: HI-STAR 100 System 

 (2) Description 

The HI-STAR 100 System is a canister system comprising a Multi-Purpose Canister (MPC) 

inside of an overpack designed for both storage and transportation (with impact limiters) of 

irradiated nuclear fuel. The HI-STAR 100 System consists of interchangeable MPCs that 

house the spent nuclear fuel and an overpack that provides the containment boundary, helium 

retention boundary, gamma and neutron radiation shielding, and heat rejection capability. The 

outer diameter of the overpack of the HI-STAR 100 is approximately 96 inches without impact 

limiters and approximately 128 inches with impact limiters. Maximum gross weight for 

transportation (including overpack, MPC, fuel, and impact limiters) is 282,000 pounds. 

Specific tolerances germane to the safety analyses are called out in the drawings listed below. 

The HI-STAR 100 System includes the HI-STAR 100 Version HB (also referred to as the HISTAR HB). 

 Multi-Purpose Canister

There are seven Multi-Purpose Canister (MPC) models designated as the MPC-24, MPC-24E, 

MPC-24EF, MPC-32,MPC-68, MPC-68F, and the MPC-HB. All MPCs are designed to have 

identical exterior dimensions, except 1) MPC-24E/EFs custom-designed for the Trojan plant, 

which are approximately nine inches shorter than the generic Holtec MPC design; and 2) 

MPC-HBs custom-designed for the Humboldt Bay plant, which are approximately 6.3 feet 

5.(a)(2) Description (continued) 

A-5
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION 

NUMBER PAGE PAGES 

9261 7 71-9261 USA/9261/B(U)F-96 2 OF 7 

shorter than the generic Holtec MPC designs. The two digits after the MPC designate the 

number of reactor fuel assemblies for which the respective MPCs are designed. The MPC-24 

series is designed to contain up to 24 Pressurized Water Reactor (PWR) fuel assemblies; the 

MPC-32 is designed to contain up to 32 intact PWR assemblies; and the MPC-68 and MPC68F are designed to contain up to 68 Boiling Water Reactor (BWR) fuel assemblies. The 

MPC-HB is designed to contain up to 80 Humboldt Bay BWR fuel assemblies. 

The HI-STAR 100 MPC is a welded cylindrical structure with flat ends. Each MPC is an 

assembly consisting of a honeycombed fuel basket, baseplate, canister shell, lid, and closure 

ring. The outer diameter and cylindrical height of each generic MPC is fixed. The outer 

diameter of the Trojan MPCs is the same as the generic MPC, but the height is approximately 

nine inches shorter than the generic MPC design. A steel spacer is used with the Trojan plant 

MPCs to ensure the MPC-overpack interface is bounded by the generic design. The outer 

diameter of the Humboldt Bay MPCs is the same as the generic MPC, but the height is 

approximately 6.3 feet shorter than the generic MPC design. The Humboldt Bay MPCs are 

transported in a shorter version of the HI-STAR overpack, designated as the HI-STAR HB. 

The fuel basket designs vary based on the MPC model. 

 Overpack

The HI-STAR 100 overpack is a multi-layer steel cylinder with a welded baseplate and bolted 

lid (closure plate). The inner shell of the overpack forms an internal cylindrical cavity for 

housing the MPC. The outer surface of the overpack inner shell is buttressed with 

intermediate steel shells for radiation shielding. The overpack closure plate incorporates a 

dual O-ring design to ensure its containment function. The containment system consists of 

the overpack inner shell, bottom plate, top flange, top closure plate, top closure inner O-ring 

seal, vent port plug and seal, and drain port plug and seal. 

 Impact Limiters

The HI-STAR 100 overpack is fitted with two impact limiters fabricated of aluminum 

honeycomb completely enclosed by an all-welded austenitic stainless steel skin. The two 

impact limiters are attached to the overpack with 20 and 16 bolts at the top and bottom, 

respectively. 

 (3) Drawings 

The package shall be constructed and assembled in accordance with the following drawings 

or figures in Holtec International Report No. HI-951251, Safety Analysis Report for the Holtec 

International Storage, Transport, And Repository Cask System (HI-STAR 100 Cask System), 

Revision 12, as supplemented: 

A-6
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION 

NUMBER PAGE PAGES 

9261 7 71-9261 USA/9261/B(U)F-96 3 OF 7 

5.(a)(3) Drawings (continued) 

 (a) HI-STAR 100 Overpack Drawing 3913, Sheets 1-9, Rev. 9 

 (b) MPC Enclosure Vessel Drawing 3923, Sheets 1-5, Rev. 16 

 (c) MPC-24E/EF Fuel Basket Drawing 3925, Sheets 1-4, Rev. 5 

 (d) MPC-24 Fuel Basket Assembly Drawing 3926, Sheets 1-4, Rev. 5 

 

 (e) MPC-68/68F/68FF Fuel Basket Drawing 3928, Sheets 1-4, Rev. 5 

(f) HI-STAR 100 Impact Limiter Drawing C1765, Sheet 1, Rev. 4; Sheet 2, Rev. 3; 

 Sheet 3, Rev. 4, Sheet 4, Rev. 4; Sheet 5, Rev. 2; 

Sheet 6, Rev. 3; and Sheet 7, Rev. 1. 

 (g) HI-STAR 100 Assembly for Transport Drawing 3930, Sheets 1-3, Rev. 2 

 (h) Trojan MPC-24E/EF Spacer Ring Drawing 4111, Sheets 1-2, Rev. 0 

 (i) Damaged Fuel Container Drawing 4119, Sheet 1-4, Rev. 1 

 for Trojan Plant SNF 

 (j) Spacer for Trojan Failed Fuel Can Drawing 4122, Sheets 1-2, Rev. 0 

 (k) Failed Fuel Can for Trojan SNC Drawings PFFC-001, Rev. 8 and 

 PFFC-002, Sheets 1 and 2, Rev. 7 

 (l) MPC-32 Fuel Basket Assembly Drawing 3927, Sheets 1-4, Rev. 6 

 (m) HI-STAR HB Overpack Drawing 4082, Sheets 1-7, Rev. 3 

 (n) MPC-HB Enclosure Vessel Drawing 4102, Sheets 1-4, Rev. 1 

 (o) MPC-HB Fuel Basket Drawing 4103, Sheets 1-3, Rev. 5 

 (p) Damaged Fuel Container HB Drawing 4113, Sheets 1-2, Rev. 1 

5.(b) Contents 

 (1) Type, Form, and Quantity of Material 

 (a) Fuel assemblies meeting the specifications and quantities provided in Appendix A to 

this Certificate of Compliance and meeting the requirements provided in Conditions 

5.b(1)(b) through 5.b(1)(i) below are authorized for transportation. 

A-7
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION 

NUMBER PAGE PAGES 

9261 7 71-9261 USA/9261/B(U)F-96 4 OF 7 

5.(b)(1) Type, Form, and Quantity of Material (continued) 

 (b) The following definitions apply: 

Damaged Fuel Assemblies are fuel assemblies with known or suspected 

cladding defects, as determined by review of records, greater than pinhole 

leaks or hairline cracks, empty fuel rod locations that are not filled with dummy 

fuel rods, missing structural components such as grid spacers, whose structural 

integrity has been impaired such that geometric rearrangement of fuel or gross 

failure of the cladding is expected based on engineering evaluations, or that 

cannot be handled by normal means. Fuel assemblies that cannot be handled 

by normal means due to fuel cladding damage are considered FUEL DEBRIS. 

Damaged Fuel Containers (or Canisters) (DFCs) are specially designed fuel 

containers for damaged fuel assemblies or fuel debris that permit gaseous and 

liquid media to escape while minimizing dispersal of gross particulates. 

The DFC designs authorized for use in the HI-STAR 100 are shown in Figures 

1.2.10, 1.2.11, and 1.I.1 of the HI-STAR 100 System SAR, Rev. 12, as 

supplemented. 

Fuel Debris is ruptured fuel rods, severed rods, loose fuel pellets, and fuel 

assemblies with known or suspected defects which cannot be handled by 

normal means due to fuel cladding damage, including containers and 

structures supporting these parts. Fuel debris also includes certain Trojan 

plant-specific fuel material contained in Trojan Failed Fuel Cans. 

Incore Grid Spacers are fuel assembly grid spacers located within the active 

fuel region (i.e., not including top and bottom spacers). 

Intact Fuel Assemblies are fuel assemblies without known or suspected 

cladding defects greater than pinhole leaks or hairline cracks and which can be 

handled by normal means. Fuel assemblies without fuel rods in fuel rod 

locations shall not be classified as intact fuel assemblies unless dummy fuel 

rods are used to displace an amount of water greater than or equal to that 

displaced by the original fuel rod(s). Trojan fuel assemblies not loaded into 

DFCs or FFCs are classified as intact assemblies. 

Minimum Enrichment is the minimum assembly average enrichment. Natural 

uranium blankets are not considered in determining minimum enrichment. 

Non-Fuel Hardware is defined as Burnable Poison Rod Assemblies (BPRA), 

Thimble Plug Devices (TPDs), and Rod Cluster Control Assemblies (RCCAs). 

Planar-Average Initial Enrichment is the average of the distributed fuel rod 

initial enrichments within a given axial plane of the assembly lattice. 

A-8
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION 

NUMBER PAGE PAGES 

9261 7 71-9261 USA/9261/B(U)F-96 5 OF 7 

5.(b)(1)(b) Definitions (continued) 

Trojan Damaged Fuel Containers (or Canisters) are Holtec damaged fuel 

containers custom-designed for Trojan plant damaged fuel and fuel debris as 

depicted in Drawing 4119, Rev. 1. 

Trojan Failed Fuel Cans are non-Holtec designed Trojan plant-specific 

damaged fuel containers that may be loaded with Trojan plant damaged fuel 

assemblies, Trojan fuel assembly metal fragments (e.g., portions of fuel rods 

and grid assemblies, bottom nozzles, etc.), a Trojan fuel rod storage container, 

a Trojan Fuel Debris Process Can Capsule, or a Trojan Fuel Debris Process 

Can. The Trojan Failed Fuel Can is depicted in Drawings PFFC-001, 

Rev. 8 and PFFC-002, Rev. 7. 

Trojan Fuel Debris Process Cans are Trojan plant-specific canisters 

containing fuel debris (metal fragments) and were used to process organic 

media removed from the Trojan plant spent fuel pool during cleanup 

operations in preparation for spent fuel pool decommissioning. Trojan Fuel 

Debris Process Cans are loaded into Trojan Fuel Debris Process Can 

Capsules or directly into Trojan Failed Fuel Cans. The Trojan Fuel Debris 

Process Can is depicted in Figure 1.2.10B of the HI-STAR100 System SAR, 

Rev. 12, as supplemented. 

Trojan Fuel Debris Process Can Capsules are Trojan plant-specific canisters 

that contain up to five Trojan Fuel Debris Process Cans and are vacuumed, 

purged, backfilled with helium and then seal-welded closed. The Trojan Fuel 

Debris Process Can Capsule is depicted in Figure 1.2.10C of the HI-STAR 100 

System SAR, Rev. 12, as supplemented. 

Undamaged Fuel Assemblies are fuel assemblies where all the exterior rods 

in the assembly are visually inspected and shown to be intact. The interior 

rods of the assembly are in place; however, the cladding of these rods is of 

unknown condition. This definition only applies to Humboldt Bay fuel assembly 

array/class 6x6D and 7x7C. 

ZR means any zirconium-based fuel cladding materials authorized for use in a 

commercial nuclear power plant reactor. 

 (c) For MPCs partially loaded with stainless steel clad fuel assemblies, all 

remaining fuel assemblies in the MPC shall meet the more restrictive of the 

decay heat limits for the stainless steel clad fuel assemblies or the applicable 

ZR clad fuel assemblies.

(d) For MPCs partially loaded with damaged fuel assemblies or fuel debris, all 

remaining ZR clad intact fuel assemblies in the MPC shall meet the more 

A-9
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION 

NUMBER PAGE PAGES 

9261 7 71-9261 USA/9261/B(U)F-96 6 OF 7 

 5.(b)(1)(b) Definitions (continued) 

restrictive of the decay heat limits for the damaged fuel assemblies or the intact 

fuel assemblies. 

(e) For MPC-68s partially loaded with array/class 6x6A, 6x6B, 6x6C, or 8x8A fuel 

assemblies, all remaining ZR clad intact fuel assemblies in the MPC shall meet 

the more restrictive of the decay heat limits for the 6x6A, 6x6B, 6x6C, and 

8x8A fuel assemblies or the applicable Zircaloy clad fuel assemblies. 

 (f) PWR non-fuel hardware and neutron sources are not authorized for 

transportation except as specifically provided for in Appendix A to this CoC. 

 (g) BWR stainless-steel channels and control blades are not authorized for 

transportation. 

 (h) For spent fuel assemblies to be loaded into MPC-32s, core average soluble 

boron, assembly average specific power, and assembly average moderator 

 temperature in which the fuel assemblies were irradiated, shall be determined 

according to Section 1.2.3.7.1 of the SAR, and the values shall be compared 

against the limits specified in Part VI of Table A.1 in Appendix A of this 

Certificate of Compliance. 

 (i) For spent fuel assemblies to be loaded into MPC-32s, the reactor records on 

spent fuel assemblies average burnup shall be confirmed through physical 

burnup measurements as described in Section 1.2.3.7.2 of the SAR. 

5.(c) Criticality Safety Index (CSI)= 0.0 

6. In addition to the requirements of Subpart G of 10 CFR Part 71: 

(a) Each package shall be both prepared for shipment and operated in accordance with detailed 

written operating procedures. Procedures for both preparation and operation shall be 

developed. At a minimum, those procedures shall include the provisions provided in Chapter 

7 of the HI-STAR SAR. 

 (b) All acceptance tests and maintenance shall be performed in accordance with detailed written 

procedures. Procedures for acceptance testing and maintenance shall be developed and 

shall include the provisions provided in Chapter 8 of the HI-STAR SAR. 

7. The maximum gross weight of the package as presented for shipment shall not exceed 282,000 

pounds, except for the HI-STAR HB, where the gross weight shall not exceed 187,200 pounds. 

8. The package shall be located on the transport vehicle such that the bottom surface of the bottom 

impact limiter is at least 9 feet (along the axis of the overpack) from the edge of the vehicle. 

A-10
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION 

NUMBER PAGE PAGES 

9261 7 71-9261 USA/9261/B(U)F-96 7 OF 7 

9. The personnel barrier shall be installed at all times while transporting a loaded overpack. 

10. The package authorized by this certificate is hereby approved for use under the general license 

provisions of 10 CFR 71.17. 

11. Transport by air of fissile material is not authorized. 

12. Revision No. 6 of this certificate may be used until May 31, 2010. 

13. Expiration Date: March 31, 2014 

Attachment: Appendix A 

REFERENCES: 

Holtec International Report No. HI-951251, Safety Analysis Report for the Holtec International Storage, 

Transport, And Repository Cask System (HI-STAR 100 Cask System), Revision 12, dated October 9, 2006. 

Holtec International supplements dated June 29, July 27, August 3, September 27, October 5, and 

December 18, 2007; January 9, March 19, and September 30, 2008; and February 27, 2009. 

 FOR THE U.S. NUCLEAR REGULATORY COMMISSION 

/RA/ 

 Eric J. Benner, Chief 

 Licensing Branch 

 Division of Spent Fuel Storage and Transportation 

 Office of Nuclear Material Safety 

 and Safeguards 

 

 Date: May 8, 2009 

A-11
 APPENDIX A 

CERTIFICATE OF COMPLIANCE NO. 9261, REVISION 7 

MODEL NO. HI-STAR 100 SYSTEM

A-12
Appendix A - Certificate of Compliance 9261, Revision 7

INDEX TO APPENDIX A 

Page: Table: Description: 

Page A-1 to A-23 Table A.1 Fuel Assembly Limits 

Page A-1 MPC-24: Uranium oxide, PWR intact fuel assemblies 

listed in Table A.2. 

A-2 MPC-68: Uranium oxide, BWR intact fuel assemblies 

listed in Table A.3 with or without Zircaloy channels. 

A-3 MPC-68: Uranium oxide, BWR damaged fuel assemblies, 

with or without Zircaloy channels, placed in damaged fuel 

containers. Uranium oxide BWR damaged fuel assemblies 

shall meet the criteria specified in Table A.3 for fuel 

assembly array/class 6x6A, 6X6C, 7x7A, or 8x8A. 

A-4 MPC-68: Mixed oxide (MOX), BWR intact fuel 

assemblies, with or without Zircaloy channels. MOX BWR 

intact fuel assemblies shall meet the criteria specified in 

Table A.3 for fuel assembly array/class 6x6B. 

A-5 MPC-68: Mixed oxide (MOX), BWR damaged fuel 

assemblies, with or without Zircaloy channels, placed in 

damaged fuel containers. MOX BWR damaged fuel 

assemblies shall meet the criteria specified in Table A.3 

for fuel assembly array/class 6x6B. 

A-6 MPC-68: Thoria rods (ThO2 and UO2) placed in Dresden 

Unit 1 Thoria Rod Canisters 

A-7 MPC-68F: Uranium oxide, BWR intact fuel assemblies, 

with or without Zircaloy channels. Uranium oxide BWR 

intact fuel assemblies shall meet the criteria specified in 

Table A.3 for fuel assembly array/class 6x6A, 6x6C, 7x7A, 

or 8x8A. 

A-8 MPC-68F: Uranium oxide, BWR damaged fuel 

assemblies, with or without Zircaloy channels, placed in 

damaged fuel containers. Uranium oxide BWR damaged 

fuel assemblies shall meet the criteria specified in Table 

A.3 for fuel assembly array/class 6x6A, 6x6C, 7x7A, or 

8x8A. 

A-9 MPC-68F: Uranium oxide, BWR fuel debris, with or 

without Zircaloy channels, placed in damaged fuel 

containers. The original fuel assemblies for the uranium 

oxide BWR fuel debris shall meet the criteria specified in 

Table A.3 for fuel assembly array/class 6x6A, 6x6C, 

7x7A, or 8x8A. 

A-13
Appendix A - Certificate of Compliance 9261, Revision 7

A-10 Table A. 1 

(Cont'd) 

MPC-68F: Mixed oxide (MOX), BWR intact fuel 

assemblies, with or without Zircaloy channels. MOX BWR 

intact fuel assemblies shall meet the criteria specified in 

Table A.3 for fuel assembly array/class 6x6B. 

A-11 MPC-68F: Mixed oxide (MOX), BWR damaged fuel 

assemblies, with or without Zircaloy channels, placed in 

damaged fuel containers. MOX BWR damaged fuel 

assemblies shall meet the criteria specified in Table A.3 

for fuel assembly array/class 6x6B. 

A-12 MPC-68F: Mixed Oxide (MOX), BWR fuel debris, with or 

without Zircaloy channels, placed in damaged fuel 

containers. The original fuel assemblies for the MOX BWR 

fuel debris shall meet the criteria specified in Table A.3 for 

fuel assembly array/class 6x6B. 

A-13 MPC-68F: Thoria rods (ThO2 and UO2) placed in Dresden 

Unit 1 Thoria Rod Canisters. 

A-15 MPC-24E: Uranium oxide, PWR intact fuel assemblies 

listed in Table A.2. 

A-16 MPC-24E: Trojan plant damaged fuel assemblies. 

A-17 MPC-24EF: Uranium oxide, PWR intact fuel assemblies 

listed in Table A.2. 

A-18 MPC-24EF: Trojan plant damaged fuel assemblies. 

A-19 MPC-24EF: Trojan plant Fuel Debris Process Can 

Capsules and/or Trojan plant fuel assemblies classified as 

fuel debris. 

A-20 to A-21 MPC-32: Uranium oxide, PWR intact fuel assemblies in 

array classes 15X15D, E, F, and H and 17X17A, B, and C 

as listed in Table A.2. 

A-22 to A-23 MPC-HB: Uranium oxide, intact and/or undamaged fuel 

assemblies and damaged fuel assemblies, with or without 

channels, meeting the criteria specified in Table A.3 for 

fuel assembly array/class 6x6D or 7x7C. 

A-24 to A-27 Table A.2 PWR Fuel Assembly Characteristics 

A-28 to A-33 Table A.3 BWR Fuel Assembly Characteristics 

A-34 Table A.4 Fuel Assembly Cooling, Average Burnup, and Initial 

Enrichment MPC-24/24E/24EF PWR Fuel with Zircaloy 

Clad and with Non-Zircaloy In-Core Grid Spacers. 

A-14
Appendix A - Certificate of Compliance 9261, Revision 7

INDEX TO APPENDIX A 

Page: Table: Description: 

A-34 Table A.5 Fuel Assembly Cooling, Average Burnup, and Initial 

Enrichment MPC-24/24E/24EF PWR Fuel with Zircaloy 

clad and with Zircaloy In-Core Grid Spacers. 

A-35 Table A.6 Fuel Assembly Cooling, Average Burnup, and Initial 

Enrichment MPC-24/24E/24EF PWR Fuel with Stainless 

Steel Clad. 

A-35 Table A.7 Fuel Assembly Cooling, Average Burnup, and Initial 

Enrichment-MPC-68. 

A-36 Table A.8 Trojan Plant Fuel Assembly Cooling, Average Burnup, and 

Initial Enrichment Limits. 

A-36 Table A.9 Trojan Plant Non-Fuel Hardware and Neutron Source 

Cooling and Burnup Limits. 

A-37 Table A.10 Fuel Assembly Cooling, Average Burnup, and Minimum 

Enrichment MPC-32 PWR Fuel with Zircaloy Clad and 

with Non-Zircaloy In-Core Grid Spacers. 

A-37 Table A.11 Fuel Assembly Cooling, Average Burnup, and Minimum 

Enrichment MPC-32 PWR Fuel with Zircaloy Clad and 

with Zircaloy In-Core Grid Spacers. 

A-38 Table A.12 Fuel Assembly Maximum Enrichment and Minimum 

Burnup Requirement for Transportation in MPC-32. 

A-39 Table A.13 Loading Configurations for the MPC-32. 

A-40 References. 

A-15
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.1 (Page 1 of 23) 

Fuel Assembly Limits 

I. MPC MODEL: MPC-24 

 A. Allowable Contents 

 1. Uranium oxide, PWR intact fuel assemblies listed in Table A.2 and meeting the following 

specifications: 

a. Cladding type: ZR or stainless steel (SS) as specified in Table A.2 

for the applicable fuel assembly array/class 

b. Maximum initial enrichment: As specified in Table A.2 for the applicable fuel 

assembly array/class. 

c. Post-irradiation cooling time, average 

burnup, and minimum initial 

enrichment per assembly 

i. ZR clad: 

ii. SS clad: 

An assembly post-irradiation cooling time, average 

burnup, and minimum initial enrichment as specified 

in Table A.4 or A.5, as applicable. 

An assembly post-irradiation cooling time, average 

burnup, and minimum initial enrichment as specified 

in Table A.6, as applicable. 

d. Decay heat per assembly: 

i. ZR Clad: 

ii. SS Clad: 

≤833 Watts 

≤488 Watts 

e. Fuel assembly length: < 176.8 inches (nominal design) 

f. Fuel assembly width: < 8.54 inches (nominal design) 

g. Fuel assembly weight: < 1,680 lbs 

 B. Quantity per MPC: Up to 24 PWR fuel assemblies. 

 C. Fuel assemblies shall not contain non-fuel hardware or neutron sources. 

 D. Damaged fuel assemblies and fuel debris are not authorized for transport in the MPC-24. 

 E. Trojan plant fuel is not permitted to be transported in the MPC-24. 

A-16
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.1 (Page 2 of 23) 

Fuel Assembly Limits 

II. MPC MODEL: MPC-68 

 A. Allowable Contents 

 1. Uranium oxide, BWR intact fuel assemblies listed in Table A.3, except assembly classes 6x6D and 

7x7C, with or without Zircaloy channels, and meeting the following specifications: 

a. Cladding type: ZR or stainless steel (SS) as specified in Table A.3 

for the applicable fuel assembly array/class. 

b. Maximum planar-average initial 

enrichment: 

As specified in Table A.3 for the applicable fuel 

assembly array/class. 

c. Initial maximum rod enrichment: As specified in Table A.3 for the applicable fuel 

assembly array/class. 

d. Post-irradiation cooling time, average 

burnup, and minimum initial 

enrichment per assembly: 

 i. ZR clad: 

 ii. SS clad: 

An assembly post-irradiation cooling time, average 

burnup, and minimum initial enrichment as specified 

in Table A.7, except for (1) array/class 6x6A, 6x6C, 

7x7A, and 8x8A fuel assemblies, which shall have a 

cooling time > 18 years, an average burnup <

30,000 MWD/MTU, and a minimum initial 

enrichment > 1.45 wt% 235U, and (2) array/class 

8x8F fuel assemblies, which shall have a cooling 

time > 10 years, an average burnup < 27,500 

MWD/MTU, and a minimum initial enrichment > 2.4 

wt% 235U. 

An assembly cooling time after discharge > 16 

years, an average burnup < 22,500 MWD/MTU, and 

a minimum initial enrichment > 3.5 wt% 235U. 

e.Decay heat per assembly: 

i. ZR Clad: 

a. SS Clad: 

≤272 Watts, except for array/class 8X8F fuel 

assemblies, which shall have a decay heat ≤183.5 

Watts. 

≤83 Watts 

f. Fuel assembly length: < 176.2 inches (nominal design) 

g. Fuel assembly width: < 5.85 inches (nominal design) 

h. Fuel assembly weight: < 700 lbs, including channels 

A-17
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.1 (Page 3 of 23) 

Fuel Assembly Limits 

II. MPC MODEL: MPC-68 (continued) 

 A. Allowable Contents (continued) 

 2. Uranium oxide, BWR damaged fuel assemblies, with or without Zircaloy channels, placed in 

damaged fuel containers. Uranium oxide BWR damaged fuel assemblies shall meet the criteria 

specified in Table A.3 for fuel assembly array/class 6x6A, 6x6C, 7x7A, or 8x8A, and meet the 

following specifications: 

a. Cladding type: ZR 

b. Maximum planar-average initial 

enrichment: 

As specified in Table A.3 for the applicable fuel 

assembly array/class. 

c. Initial maximum rod enrichment: As specified in Table A.3 for the applicable fuel 

assembly array/class. 

d. Post-irradiation cooling time, average 

burnup, and minimum initial 

enrichment per assembly: 

An assembly post-irradiation cooling time > 18 

years, an average burnup < 30,000 MWD/MTU, and 

a minimum initial enrichment > 1.45 wt% 235U. 

e. Fuel assembly length: < 135.0 inches (nominal design) 

f. Fuel assembly width: < 4.70 inches (nominal design) 

g. Fuel assembly weight: < 550 lbs, including channels and damaged fuel 

containers 

A-18
Appendix A - Certificate of Compliance 9261, Revision 7

 Table A.1 (Page 4 of 23) 

 Fuel Assembly Limits 

II. MPC MODEL: MPC-68 (continued) 

 A. Allowable Contents (continued) 

 3. Mixed oxide (MOX), BWR intact fuel assemblies, with or without Zircaloy channels. MOX BWR 

intact fuel assemblies shall meet the criteria specified in Table A.3 for fuel assembly array/class 

6x6B and meet the following specifications: 

a. Cladding type: ZR 

b. Maximum planar-average initial 

enrichment: 

As specified in Table A.3 for fuel assembly 

array/class 6x6B. 

c. Initial maximum rod enrichment: As specified in Table A.3 for fuel assembly 

array/class 6x6B. 

d. Post-irradiation cooling time, average 

burnup, and minimum initial 

enrichment per assembly: 

An assembly post-irradiation cooling time > 18 

years, an average burnup < 30,000 MWD/MTIHM, 

and a minimum initial enrichment > 1.8 wt% 235U for 

the UO2 rods. 

e. Fuel assembly length: < 135.0 inches (nominal design) 

f. Fuel assembly width: < 4.70 inches (nominal design) 

g. Fuel assembly weight: < 400 lbs, including channels 

A-19
Appendix A - Certificate of Compliance 9261, Revision 7

 Table A.1 (Page 5 of 23) 

 Fuel Assembly Limits 

II. MPC MODEL: MPC-68 (continued) 

 A. Allowable Contents (continued) 

 4. Mixed oxide (MOX), BWR damaged fuel assemblies, with or without Zircaloy channels, placed in 

damaged fuel containers. MOX BWR damaged fuel assemblies shall meet the criteria specified in 

Table A.3 for fuel assembly array/class 6x6B and meet the following specifications: 

a. Cladding type: ZR 

b. Maximum planar-average initial 

enrichment: 

As specified in Table A.3 for array/class 6x6B. 

c. Initial maximum rod enrichment: As specified in Table A.3 for array/class 6x6B. 

d. Post-irradiation cooling time, average 

burnup, and minimum initial 

enrichment per assembly: 

An assembly post-irradiation cooling time > 18 

years, an average burnup < 30,000 MWD/MTIHM, 

and a minimum initial enrichment > 1.8 wt% 235U for 

the UO2 rods. 

e. Fuel assembly length: < 135.0 inches (nominal design) 

f. Fuel assembly width: < 4.70 inches (nominal design) 

g. Fuel assembly weight: < 550 lbs, including channels and damaged fuel 

containers. 

A-20
Appendix A - Certificate of Compliance 9261, Revision 7

 Table A.1 (Page 6 of 23) 

 Fuel Assembly Limits 

II. MPC MODEL: MPC-68 (continued) 

 A. Allowable Contents (continued) 

 5. Thoria rods (ThO2 and UO2) placed in Dresden Unit 1 Thoria Rod Canisters (as shown in Figure 

1.2.11A of the HI-STAR 100 System SAR, Revision 12) and meeting the following specifications: 

a. Cladding type: ZR 

b. Composition: 98.2 wt.% ThO2, 1.8 wt. % UO2 with an enrichment 

of 93.5 wt. % 235U. 

c. Number of rods per Thoria Rod 

Canister: 

< 18 

d. Decay heat per Thoria Rod Canister: < 115 Watts 

e. Post-irradiation fuel cooling time and 

average burnup per Thoria Rod 

Canister: 

A fuel post-irradiation cooling time > 18 years and an 

average burnup < 16,000 MWD/MTIHM. 

f. Initial heavy metal weight: < 27 kg/canister 

g. Fuel cladding O.D.: > 0.412 inches 

h. Fuel cladding I.D.: < 0.362 inches 

i. Fuel pellet O.D.: < 0.358 inches 

j. Active fuel length: < 111 inches 

k. Canister weight: < 550 lbs, including fuel 

 B. Quantity per MPC: Up to one (1) Dresden Unit 1 Thoria Rod Canister plus any combination of 

damaged fuel assemblies in damaged fuel containers and intact fuel assemblies, up to a total of 68. 

 C. Fuel assemblies with stainless steel channels are not authorized for loading in the MPC-68. 

 D. Dresden Unit 1 fuel assemblies (fuel assembly array/class 6x6A, 6x6B, 6x6C, or 8x8A) with one 

Antimony-Beryllium neutron source are authorized for loading in the MPC-68. The Antimony-Beryllium 

source material shall be in a water rod location. 

A-21
Appendix A - Certificate of Compliance 9261, Revision 7

 Table A.1 (Page 7 of 23) 

 Fuel Assembly Limits 

III. MPC MODEL: MPC-68F 

 A. Allowable Contents 

 1. Uranium oxide, BWR intact fuel assemblies, with or without Zircaloy channels. Uranium oxide BWR 

intact fuel assemblies shall meet the criteria specified in Table A.3 for fuel assembly array/class 

6x6A, 6x6C, 7x7A, or 8x8A and meet the following specifications: 

a. Cladding type: ZR 

b. Maximum planar-average initial 

enrichment: 

As specified in Table A.3 for the applicable fuel 

assembly array/class. 

c. Initial maximum rod enrichment: As specified in Table A.3 for the applicable fuel 

assembly array/class. 

d. Post-irradiation cooling time, average 

burnup, and minimum initial 

enrichment per assembly: 

An assembly post-irradiation cooling time > 18 

years, an average burnup < 30,000 MWD/MTU, and 

a minimum initial enrichment > 1.45 wt% 235U. 

 e. Fuel assembly length: < 176.2 inches (nominal design) 

f. Fuel assembly width: < 5.85 inches (nominal design) 

g. Fuel assembly weight: < 400 lbs, including channels 

A-22
Appendix A - Certificate of Compliance 9261, Revision 7

 Table A.1 (Page 8 of 23) 

 Fuel Assembly Limits 

III. MPC MODEL: MPC-68F (continued) 

 A. Allowable Contents (continued) 

2. Uranium oxide, BWR damaged fuel assemblies, with or without Zircaloy channels, placed in 

damaged fuel containers. Uranium oxide BWR damaged fuel assemblies shall meet the criteria 

specified in Table A.3 for fuel assembly array/class 6x6A, 6x6C, 7x7A, or 8x8A, and meet the 

following specifications: 

 a. Cladding type: ZR 

b. Maximum planar-average initial 

enrichment: 

As specified in Table A.3 for the applicable fuel 

assembly array/class. 

c. Initial maximum rod enrichment: As specified in Table A.3 for the applicable fuel 

assembly array/class. 

d. Post-irradiation cooling time, average 

burnup, and minimum initial 

enrichment per assembly: 

An assembly post-irradiation cooling time > 18 

years, an average burnup < 30,000 MWD/MTU, and 

a minimum initial enrichment > 1.45 wt% 235U. 

e. Fuel assembly length: < 135.0 inches (nominal design) 

f. Fuel assembly width: < 4.70 inches (nominal design) 

g. Fuel assembly weight: < 550 lbs, including channels and damaged fuel 

containers 

A-23
Appendix A - Certificate of Compliance 9261, Revision 7

 Table A.1 (Page 9 of 23) 

Fuel Assembly Limits 

III. MPC MODEL: MPC-68F (continued) 

 A. Allowable Contents (continued) 

 3. Uranium oxide, BWR fuel debris, with or without Zircaloy channels, placed in damaged fuel 

containers. The original fuel assemblies for the uranium oxide BWR fuel debris shall meet the 

criteria specified in Table A.3 for fuel assembly array/class 6x6A, 6x6C, 7x7A, or 8x8A, and meet 

the following specifications: 

 a. Cladding type: ZR 

b. Maximum planar-average initial 

enrichment: 

As specified in Table A.3 for the applicable original 

fuel assembly array/class. 

c. Initial maximum rod enrichment: As specified in Table A.3 for the applicable original 

fuel assembly array/class. 

d. Post-irradiation cooling time, average 

burnup, and minimum initial 

enrichment per assembly: 

An assembly post-irradiation cooling time > 18 

years, an average burnup < 30,000 MWD/MTU, and 

a minimum initial enrichment > 1.45 wt% 235U for the 

original fuel assembly. 

e. Fuel assembly length: < 135.0 inches (nominal design) 

f. Fuel assembly width: < 4.70 inches (nominal design) 

g. Fuel assembly weight: < 550 lbs, including channels and damaged fuel 

containers 

A-24
Appendix A - Certificate of Compliance 9261, Revision 7

 Table A.1 (Page 10 of 23) 

 Fuel Assembly Limits 

III. MPC MODEL: MPC-68F (continued) 

 A. Allowable Contents (continued) 

 4. Mixed oxide (MOX), BWR intact fuel assemblies, with or without Zircaloy channels. MOX BWR 

intact fuel assemblies shall meet the criteria specified in Table A.3 for fuel assembly array/class 

6x6B and meet the following specifications: 

a. Cladding type: ZR 

b. Maximum planar-average initial 

enrichment: 

As specified in Table A.3 for fuel assembly 

array/class 6x6B. 

c. Initial maximum rod enrichment: As specified in Table A.3 for fuel assembly 

array/class 6x6B. 

d. Post-irradiation cooling time, average 

burnup, and minimum initial 

enrichment per assembly: 

An assembly post-irradiation cooling time > 18 

years, an average burnup < 30,000 MWD/MTIHM, 

and a minimum initial enrichment > 1.8 wt% 235U for 

the UO2 rods. 

e. Fuel assembly length: < 135.0 inches (nominal design) 

f. Fuel assembly width: < 4.70 inches (nominal design) 

g. Fuel assembly weight: < 400 lbs, including channels 

A-25
Appendix A - Certificate of Compliance 9261, Revision 7

 Table A.1 (Page 11 of 23) 

 Fuel Assembly Limits 

III. MPC MODEL: MPC-68F (continued) 

 A. Allowable Contents (continued) 

 5. Mixed oxide (MOX), BWR damaged fuel assemblies, with or without Zircaloy channels, placed 

in damaged fuel containers. MOX BWR intact fuel assemblies shall meet the criteria specified in 

Table A.3 for fuel assembly array/class 6x6B and meet the following specifications: 

a. Cladding type: ZR 

b. Maximum planar-average initial 

enrichment: 

As specified in Table A.3 for array/class 6x6B. 

c. Initial maximum rod enrichment: As specified in Table A.3 for array/class 6x6B. 

d. Post-irradiation cooling time, average 

burnup, and minimum initial 

enrichment per assembly: 

An assembly post-irradiation cooling time > 18 

years, an average burnup < 30,000 MWD/MTIHM, 

and a minimum initial enrichment > 1.8 wt% 235U for 

the UO2 rods. 

e. Fuel assembly length: < 135.0 inches (nominal design) 

f. Fuel assembly width: < 4.70 inches (nominal design) 

g. Fuel assembly weight: < 550 lbs, including channels and damaged fuel 

containers 

A-26
Appendix A - Certificate of Compliance 9261, Revision 7

 Table A.1 (Page 12 of 23) 

 Fuel Assembly Limits 

III. MPC MODEL: MPC-68F (continued) 

 A. Allowable Contents (continued) 

 6. Mixed oxide (MOX), BWR fuel debris, with or without Zircaloy channels, placed in damaged fuel 

containers. The original fuel assemblies for the MOX BWR fuel debris shall meet the criteria 

specified in Table A.3 for fuel assembly array/class 6x6B and meet the following specifications: 

a. Cladding type: ZR 

b. Maximum planar-average initial 

enrichment: 

As specified in Table A.3 for original fuel assembly 

array/class 6x6B. 

c. Initial maximum rod enrichment: As specified in Table A.3 for original fuel assembly 

array/class 6x6B. 

d. Post-irradiation cooling time, average 

burnup, and minimum initial 

enrichment per assembly: 

An assembly post-irradiation cooling time > 18 

years, an average burnup < 30,000 MWD/MTIHM, 

and a minimum initial enrichment > 1.8 wt% 235U for 

the UO2 rods in the original fuel assembly. 

 e. Fuel assembly length: < 135.0 inches (nominal design) 

 f. Fuel assembly width: < 4.70 inches (nominal design) 

 g. Fuel assembly weight: < 550 lbs, including channels and damaged fuel 

containers 

A-27
Appendix A - Certificate of Compliance 9261, Revision 7

 Table A.1 (Page 13 of 23) 

 Fuel Assembly Limits 

III. MPC MODEL: MPC-68F (continued) 

 A. Allowable Contents (continued) 

 7. Thoria rods (ThO2 and UO2) placed in Dresden Unit 1 Thoria Rod Canisters (as shown in Figure 

1.2.11A of the HI-STAR 100 System SAR, Revision 12) and meeting the following specifications: 

a. Cladding Type: ZR 

b. Composition: 98.2 wt.% ThO2, 1.8 wt. % UO2 with an enrichment 

of 93.5 wt. % 235U. 

c. Number of rods per Thoria Rod 

Canister: 

< 18 

d. Decay heat per Thoria Rod Canister: < 115 Watts 

e. Post-irradiation fuel cooling time and 

average burnup per Thoria Rod 

Canister: 

A fuel post-irradiation cooling time > 18 years and an 

average burnup < 16,000 MWD/MTIHM. 

f. Initial heavy metal weight: < 27 kg/canister 

g. Fuel cladding O.D.: > 0.412 inches 

h. Fuel cladding I.D.: < 0.362 inches 

i. Fuel pellet O.D.: < 0.358 inches 

j. Active fuel length: < 111 inches 

k. Canister weight: < 550 lbs, including fuel 

A-28
Appendix A - Certificate of Compliance 9261, Revision 7

 Table A.1 (Page 14 of 23) 

 Fuel Assembly Limits 

III. MPC MODEL: MPC-68F (continued) 

 B. Quantity per MPC: 

Up to four (4) damaged fuel containers containing uranium oxide or MOX BWR fuel debris. The 

remaining MPC-68F fuel storage locations may be filled with array/class 6x6A, 6x6B, 6x6C, 7x7A, and 

8x8A fuel assemblies of the following type, as applicable: 

 1. Uranium oxide BWR intact fuel assemblies; 

 2. MOX BWR intact fuel assemblies; 

 3. Uranium oxide BWR damaged fuel assemblies placed in damaged fuel containers; 

 4. MOX BWR damaged fuel assemblies placed in damaged fuel containers; or 

 5. Up to one (1) Dresden Unit 1 Thoria Rod Canister. 

 C. Fuel assemblies with stainless steel channels are not authorized for loading in the MPC-68F. 

 D. Dresden Unit 1 fuel assemblies (fuel assembly array/class 6x6A, 6x6B, 6x6C or 8x8A) with one 

Antimony-Beryllium neutron source are authorized for loading in the MPC-68F. The AntimonyBeryllium neutron source material shall be in a water rod location. 

A-29
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.1 (Page 15 of 23) 

Fuel Assembly Limits 

IV. MPC MODEL: MPC-24E 

 A. Allowable Contents 

 1. Uranium oxide, PWR intact fuel assemblies listed in Table A.2 and meeting the following 

specifications: 

a. Cladding type: ZR or stainless steel (SS) as specified in Table A.2 

for the applicable fuel assembly array/class 

b. Maximum initial enrichment: As specified in Table A.2 for the applicable fuel 

assembly array/class. 

c. Post-irradiation cooling time, average 

burnup, and minimum initial enrichment 

per assembly 

i. ZR clad: 

ii. SS clad: 

iii. Trojan plant fuel 

iv Trojan plant non-fuel hardware and 

neutron sources 

Except for Trojan plant fuel, an assembly postirradiation cooling time, average burnup, and 

minimum initial enrichment as specified in Table A.4 

or A.5, as applicable. 

An assembly post-irradiation cooling time, average 

burnup, and minimum initial enrichment as specified 

in Table A.6, as applicable. 

An assembly post-irradiation cooling time, average 

burnup, and minimum initial enrichment as specified 

in Table A.8. 

Post-irradiation cooling time, and average burnup as 

specified in Table A.9 

d. Decay heat per assembly 

i. ZR Clad: 

ii. SS Clad: 

Except for Trojan plant fuel, decay heat ≤ 833 Watts. 

Trojan plant fuel decay heat: ≤ 725 Watts 

≤ 488 Watts 

e. Fuel assembly length: < 176.8 inches (nominal design) 

f. Fuel assembly width: < 8.54 inches (nominal design) 

g. Fuel assembly weight: < 1,680 lbs, including non-fuel hardware and neutron 

sources 

A-30
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.1 (Page 16 of 23) 

Fuel Assembly Limits 

IV. MPC MODEL: MPC-24E 

 A. Allowable Contents (continued) 

 2. Trojan plant damaged fuel assemblies meeting the applicable criteria listed in Table A.2 and 

meeting the following specifications: 

a. Cladding type: ZR 

b. Maximum initial enrichment: 3.7% 235U 

c. Fuel assembly post-irradiation cooling 

time, average burnup, decay heat, and 

minimum initial enrichment per 

assembly 

An assembly post-irradiation cooling time, average 

burnup, and initial enrichment as specified in Table 

A.8 

Decay Heat: ≤ 725 Watts 

d. Fuel assembly length: < 169.3 inches (nominal design) 

e. Fuel assembly width: < 8.43 inches (nominal design) 

f. Fuel assembly weight: < 1,680 lbs, including DFC or Failed Fuel Can 

B. Quantity per MPC: Up to 24 PWR intact fuel assemblies. For Trojan plant fuel only, up to four (4) 

damaged fuel assemblies may be stored in fuel storage locations 3, 6, 19, and/or 22. The remaining 

MPC-24E fuel storage locations may be filled with Trojan plant intact fuel assemblies. 

C. Trojan plant fuel must be transported in the custom-designed Trojan MPCs with the MPC spacer 

installed. Fuel from other plants is not permitted to be transported in the Trojan MPCs. 

D. Except for Trojan plant fuel, the fuel assemblies shall not contain non-fuel hardware or neutron sources. 

Trojan intact fuel assemblies containing non-fuel hardware may be transported in any fuel storage 

location. 

E. Trojan plant damaged fuel assemblies must be transported in a Trojan Failed Fuel Can or a Holtec 

damaged fuel container designed for Trojan Plant fuel. 

F. One (1) Trojan plant Sb-Be and /or up to two (2) Cf neutron sources in a Trojan plant intact fuel 

assembly (one source per fuel assembly) may be transported in any one MPC. Each fuel assembly 

neutron source may be transported in any fuel storage location. 

G. Fuel debris is not authorized for transport in the MPC-24E. 

H. Trojan plant non-fuel hardware and neutron sources may not be transported in the same fuel storage 

location as a damaged fuel assembly. 

A-31
Appendix A - Certificate of Compliance 9261, Revision 7

 Table A.1 (Page 17 of 23) 

 Fuel Assembly Limits 

V. MPC MODEL: MPC-24EF 

 A. Allowable Contents 

 1. Uranium oxide, PWR intact fuel assemblies listed in Table A.2 and meeting the following 

specifications: 

a. Cladding type: ZR or stainless steel (SS) as specified in Table A.2 

for the applicable fuel assembly array/class. 

b. Maximum initial enrichment: As specified in Table A.2 for the applicable fuel 

assembly array/class. 

c. Post-irradiation cooling time, average 

burnup, and minimum initial enrichment 

per assembly 

i. ZR clad: 

ii. SS clad: 

iii Trojan plant fuel: 

iv Trojan plant non-fuel hardware and 

neutron sources: 

Except for Trojan plant fuel, an assembly postirradiation cooling time, average burnup, and 

minimum initial enrichment as specified in Table A.4 

or A.5, as applicable. 

An assembly post-irradiation cooling time, average 

burnup, and minimum initial enrichment as specified 

in Table A.6, as applicable. 

An assembly post-irradiation cooling time, average 

burnup, and minimum initial enrichment as specified 

in Table A.8. 

Post-irradiation cooling time, and average burnup as 

specified in Table A.9. 

d. Decay heat per assembly: 

a. ZR Clad: 

b. SS Clad: 

Except for Trojan plant fuel, decay heat ≤ 833 Watts. 

Trojan plant fuel decay heat: ≤ 725 Watts. 

≤ 488 Watts 

e. Fuel assembly length: < 176.8 inches (nominal design) 

f. Fuel assembly width: < 8.54 inches (nominal design) 

g. Fuel assembly weight: < 1,680 lbs, including non-fuel hardware and neutron 

sources. 

A-32
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.1 (Page 18 of 23) 

Fuel Assembly Limits 

V. MPC MODEL: MPC-24EF 

 A. Allowable Contents (continued) 

 2. Trojan plant damaged fuel assemblies meeting the applicable criteria listed in Table A.2 and 

meeting the following specifications: 

a. Cladding type: ZR 

b. Maximum initial enrichment: 3.7% 235U 

c. Fuel assembly post-irradiation cooling 

time, average burnup, decay heat, and 

minimum initial enrichment per assembly: 

An assembly post-irradiation cooling time, average 

burnup, and initial enrichment as specified in Table 

A.8. 

Decay Heat: ≤ 725 Watts 

d. Fuel assembly length: < 169.3 inches (nominal design) 

e. Fuel assembly width: < 8.43 inches (nominal design) 

f. Fuel assembly weight: < 1,680 lbs, including DFC or Failed Fuel Can. 

A-33
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.1 (Page 19 of 23) 

Fuel Assembly Limits 

V. MPC MODEL: MPC-24EF 

 A. Allowable Contents (continued) 

 3. Trojan Fuel Debris Process Can Capsules and/or Trojan plant fuel assemblies classified as fuel 

debris, for which the original fuel assemblies meet the applicable criteria listed in Table A.2 and 

meet the following specifications: 

a. Cladding type: ZR 

b. Maximum initial enrichment: 3.7% 235U 

c. Fuel debris post-irradiation cooling time, 

average burnup, decay heat, and 

minimum initial enrichment per 

assembly: 

Post-irradiation cooling time, average burnup, and 

initial enrichment as specified in Table A.8. 

Decay Heat: ≤ 725 Watts 

d. Fuel assembly length: < 169.3 inches (nominal design) 

e. Fuel assembly width: < 8.43 inches (nominal design) 

f. Fuel assembly weight: < 1,680 lbs, including DFC or Failed Fuel Can. 

B. Quantity per MPC: Up to 24 PWR intact fuel assemblies. For Trojan plant fuel only, up to four (4) 

damaged fuel assemblies, fuel assemblies classified as fuel debris, and/or Trojan Fuel Debris Process 

Can Capsules may be stored in fuel storage locations 3, 6, 19, and/or 22. The remaining MPC-24EF 

fuel storage locations may be filled with Trojan plant intact fuel assemblies. 

C. Trojan plant fuel must be transported in the custom-designed Trojan MPCs with the MPC spacer 

installed. Fuel from other plants is not permitted to be transported in the Trojan MPCs. 

D. Except for Trojan plant fuel, the fuel assemblies shall not contain non-fuel hardware or neutron sources. 

Trojan intact fuel assemblies containing non-fuel hardware may be transported in any fuel storage 

location. 

E. Trojan plant damaged fuel assemblies, fuel assemblies classified as fuel debris, and Fuel Debris 

Process Can Capsules must be transported in a Trojan Failed Fuel Can or a Holtec damaged fuel 

container designed for Trojan Plant fuel. 

F. One (1) Trojan plant Sb-Be and /or up to two (2) Cf neutron sources in a Trojan plant intact fuel 

assembly (one source per fuel assembly) may be transported in any one MPC. Each fuel assembly 

neutron source may be transported in any fuel storage location. 

G. Trojan plant non-fuel hardware and neutron sources may not be transported in the same fuel storage 

location as a damaged fuel assembly. 

A-34
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.1 (Page 20 of 23) 

Fuel Assembly Limits 

VI. MPC MODEL: MPC-32 

 A. Allowable Contents 

 1. Uranium oxide, PWR intact fuel assemblies in array/classes 15x15D, E, F, and H and 17x17A, B, 

and C listed in Table A.2 and meeting the following specifications: 

a. Cladding type: 

ZR 

b. Maximum initial enrichment: As specified in Table A.2 for the applicable fuel 

assembly array/class. 

c. Post-irradiation cooling time, maximum 

average burnup, and minimum initial 

enrichment per assembly: 

An assembly post-irradiation cooling time, average 

burnup, and minimum initial enrichment as specified 

in Table A.10 or A.11, as applicable. 

 d. Minimum average burnup per assembly 

(Assembly Burnup shall be confirmed per 

Subsection 1.2.3.7.2 of the SAR, which is 

hereby included by reference) 

Calculated value as a function of initial enrichment. 

See Table A.12. 

 e. Decay heat per assembly: ≤ 625 Watts 

 f. Fuel assembly length: < 176.8 inches (nominal design) 

 g. Fuel assembly width: < 8.54 inches (nominal design) 

 h. Fuel assembly weight: < 1,680 lbs 

 i. Operating parameters during irradiation of the assembly (Assembly operating parameters shall 

be determined per Subsection 1.2.3.7.1 of the SAR, which is hereby included by reference) 

Core ave. soluble boron concentration: < 1,000 ppmb 

 Assembly ave. moderator temperature: < 601 K for array/classes 15x15D, E, F, and H 

< 610 K for array/classes 17x17A, B, and C 

 Assembly ave. specific power: < 47.36 kW/kg-U for array/classes 15x15D, E, F, and 

H 

< 61.61 kW/kg-U for array/classes 17x17A, B, and C 

A-35
Appendix A - Certificate of Compliance 9261, Revision 7

 

Table A.1 (Page 21 of 23) 

Fuel Assembly Limits 

VI. MP C MODEL: MPC-32 (continued) 

 B. Quantity per MPC: Up to 32 PWR intact fuel assemblies. 

C. Fuel assemblies shall not contain non-fuel hardware. 

 

D. Damaged fuel assemblies and fuel debris are not authorized for transport in MPC-32. 

 

E. Trojan plant fuel is not permitted to be transported in the MPC-32. 

A-36
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.1 (Page 22 of 23) 

Fuel Assembly Limits 

VII. MPC MODEL: MPC-HB 

 A. Allowable Contents 

 1. Uranium oxide, INTACT and/or UNDAMAGED FUEL ASSEMBLIES, DAMAGED 

FUEL ASSEMBLIES, and FUEL DEBRIS, with or without channels, meeting the 

criteria specified in Table A.3 for fuel assembly array/class 6x6D or 7x7C and the 

following specifications: 

a. Cladding type: ZR 

b. Maximum planar-average enrichment: As specified in Table A.3 for the applicable 

fuel assembly array/class. 

c. Initial maximum rod enrichment: As specified in Table A.3 for the applicable 

fuel assembly array/class. 

d. Post-irradiation cooling time, average 

burnup, and minimum initial enrichment 

per assembly: 

An assembly post-irradiation cooling time ≥

29 years, an average burnup ≤ 23,000 

MWD/MTU, and a minimum initial 

enrichment ≥ 2.09 wt% 235U. 

 e. Fuel assembly length: < 96.91 inches (nominal design) 

 f. Fuel assembly width: < 4.70 inches (nominal design) 

 g. Fuel assembly weight: < 400 lbs, including channels and DFC 

 h. Decay heat per assembly: ≤ 50 W 

 h. Decay heat per MPC: ≤ 2000 W 

A-37
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.1 (Page 23 of 23) 

Fuel Assembly Limits 

VII. MPC MODEL: MPC-HB (continued) 

B. Quantity per MPC-HB: Up to 80 fuel assemblies 

C. Damaged fuel assemblies and fuel debris must be stored in a damaged fuel container. 

Allowable Loading Configurations: Up to 28 damaged fuel assemblies/fuel debris, in 

damaged fuel containers, may be placed into the peripheral fuel storage locations as 

shown in SAR Figure 6.I.3, or up to 40 damaged fuel assemblies/fuel debris, in damaged 

fuel containers, can be placed in a checkerboard pattern as shown in SAR Figure 6.I.4. 

The remaining fuel locations may be filled with intact and/or undamaged fuel assemblies 

meeting the above applicable specifications, or with intact and/or undamaged fuel 

assemblies placed in damaged fuel containers. 

 NOTE 1: The total quantity of damaged fuel or fuel debris permitted in a single damaged 

fuel container is limited to the equivalent weight and special nuclear material quantity of one 

intact assembly. 

NOTE 2: Fuel debris includes material in the form of loose debris consisting of zirconium 

clad pellets, stainless steel clad pellets, unclad pellets, or rod segments up to a maximum of 

one equivalent fuel assembly. A maximum of 1.5 kg of stainless steel clad is allowed per 

cask. 

 

A-38
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.2 (Page 1 of 4) 

PWR FUEL ASSEMBLY CHARACTERISTICS (Note 1) 

Fuel Assembly 

Array/Class 14x14A 14x14B 14x14C 14x14D 14x14E

Clad Material 

(Note 2) ZR ZR ZR SS Zr 

Design Initial U 

(kg/assy.) (Note 3) < 407 < 407 < 425 < 400 < 206 

Initial Enrichment 

(MPC-24, 24E, and 

24EF) 

(wt % 235U) 

< 4.6 (24) 

< 5.0 (24E/EF) 

< 4.6 (24) 

< 5.0 (24E/EF) 

< 4.6 (24) 

< 5.0 (24E/EF) 

< 4.0 (24) 

< 5.0 (24E/EF) 

< 5.0 

No. of Fuel Rod 

Locations 179 179 176 180 173 

Fuel Clad O.D. (in.) > 0.400 > 0.417 > 0.440 > 0.422 > 0.3415 

Fuel Clad I.D. (in.) < 0.3514 < 0.3734 < 0.3880 < 0.3890 < 0.3175 

Fuel Pellet Dia. (in.) < 0.3444 < 0.3659 < 0.3805 < 0.3835 < 0.3130 

Fuel Rod Pitch (in.) < 0.556 < 0.556 < 0.580 < 0.556 Note 6 

Active Fuel 

Length (in.) < 150 < 150 < 150 < 144 < 102 

No. of Guide Tubes 17 17 5 (Note 4) 16 0 

Guide Tube 

Thickness (in.) > 0.017 > 0.017 > 0.038 > 0.0145 N/A

A-39
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.2 (Page 2 of 4) 

PWR FUEL ASSEMBLY CHARACTERISTICS (Note 1) 

Fuel Assembly 

Array/Class 15x15A 15x15B 15x15C 15x15D 15x15E 15x15F

Clad Material 

(Note 2) ZR ZR ZR ZR ZR ZR 

Design Initial U 

(kg/assy.) (Note 3) < 464 < 464 < 464 < 475 < 475 < 475 

Initial Enrichment 

(MPC-24, 24E, and 

24EF) 

 (wt % 235U) 

< 4.1 (24) 

< 4.5 

(24E/EF) 

< 4.1 (24) 

< 4.5 

(24E/EF) 

< 4.1 (24) 

< 4.5 

(24E/EF) 

< 4.1 (24) 

< 4.5 

(24E/EF) 

< 4.1 (24) 

< 4.5 

(24E/EF) 

< 4.1 (24) 

< 4.5 

(24E/EF) 

Initial Enrichment 

(MPC-32) 

(wt. % 235U) 

(Note 5) 

N/A N/A N/A (Note 5) (Note 5) (Note 5) 

No. of Fuel Rod 

Locations 204 204 204 208 208 208 

Fuel Clad O.D. (in.) > 0.418 > 0.420 > 0.417 > 0.430 > 0.428 > 0.428 

Fuel Clad I.D. (in.) < 0.3660 < 0.3736 < 0.3640 < 0.3800 < 0.3790 < 0.3820 

Fuel Pellet Dia. 

(in.) < 0.3580 < 0.3671 < 0.3570 < 0.3735 < 0.3707 < 0.3742 

Fuel Rod Pitch (in.) < 0.550 < 0.563 < 0.563 < 0.568 < 0.568 < 0.568 

Active Fuel Length 

(in.) < 150 < 150 < 150 < 150 < 150 < 150 

No. of Guide 

and/or Instrument 

Tubes 

21 21 21 17 17 17 

Guide/Instrument 

Tube Thickness 

(in.) 

> 0.015 > 0.015 > 0.0165 > 0.0150 > 0.0140 > 0.0140

A-40
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.2 (Page 3 of 4) 

PWR FUEL ASSEMBLY CHARACTERISTICS (Note 1) 

Fuel Assembly 

Array/ Class 15x15G 15x15H 16x16A 17x17A 17x17B 17x17C

Clad Material 

(Note 2) SS ZR ZR ZR ZR ZR 

Design Initial U 

(kg/assy.) (Note 3) < 420 < 475 < 443 < 467 < 467 < 474 

Initial Enrichment 

(MPC-24, 24E, and 

24EF) 

(wt % 235U) 

< 4.0 (24) 

< 4.5 

(24E/EF) 

< 3.8 (24) 

< 4.2 

(24E/EF) 

< 4.6 (24) 

< 5.0 

(24E/EF) 

< 4.0 (24) 

< 4.4 

(24E/EF) 

< 4.0 (24) 

< 4.4 

(24E/EF) 

(Note 7) 

< 4.0 (24) 

< 4.4 

(24E/EF) 

Initial Enrichment 

(MPC-32) 

(wt. % 235U) 

(Note 5) 

N/A (Note 5) N/A (Note 5) (Note 5) (Note 5) 

No. of Fuel Rod 

Locations 204 208 236 264 264 264 

Fuel Clad O.D. (in.) > 0.422 > 0.414 > 0.382 > 0.360 > 0.372 > 0.377 

Fuel Clad I.D. (in.) < 0.3890 < 0.3700 < 0.3320 < 0.3150 < 0.3310 < 0.3330 

Fuel Pellet Dia. (in.) < 0.3825 < 0.3622 < 0.3255 < 0.3088 < 0.3232 < 0.3252 

Fuel Rod Pitch (in.) < 0.563 < 0.568 < 0.506 < 0.496 < 0.496 < 0.502 

Active Fuel 

Length (in.) < 144 < 150 < 150 < 150 < 150 < 150 

No. of Guide and/or 

Instrument Tubes 21 17 5 (Note 4) 25 25 25 

Guide/Instrument 

Tube 

Thickness (in.) 

> 0.0145 > 0.0140 > 0.0400 > 0.016 > 0.014 > 0.020

A-41
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.2 (Page 4 of 4) 

PWR FUEL ASSEMBLY CHARACTERISTICS (Note 1) 

Notes: 

1. All dimensions are design nominal values. Maximum and minimum dimensions are 

specified to bound variations in design nominal values among fuel assemblies within a given 

array/class. 

2. ZR Designates cladding material made of Zirconium or Zirconium alloys. 

3. Design initial uranium weight is the nominal uranium weight specified for each assembly by 

the fuel manufacturer or reactor user. For each PWR fuel assembly, the total uranium 

weight limit specified in this table may be increased up to 2.0 percent for comparison with 

users’ fuel records to account for manufacturer tolerances. 

4. Each guide tube replaces four fuel rods. 

5. Minimum burnup and maximum initial enrichment as specified in Table A.12. 

6. This fuel assembly array/class includes only the Indian Point Unit 1 fuel assembly. This fuel 

assembly has two pitches in different sectors of the assembly. These pitches are 0.441 

inches and 0.453 inches 

7. Trojan plant-specific fuel is governed by the limits specified for array/class 17x17B and will 

be transported in the custom-designed Trojan MPC-24E/EF canisters. The Trojan MPC24E/EF design is authorized to transport only Trojan plant fuel with a maximum initial 

enrichment of 3.7 wt.% 235U. 

A-42
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.3 (Page 1 of 6) 

BWR FUEL ASSEMBLY CHARACTERISTICS (Note 1) 

Fuel Assembly 

Array/Class 6x6A 6x6B 6x6C 7x7A 7x7B 8x8A

Clad Material 

(Note 2) ZR ZR ZR ZR ZR ZR 

Design Initial U 

(kg/assy.) (Note 3) < 110 < 110 < 110 < 100 < 195 < 120 

Maximum planaraverage initial 

enrichment 

(wt.% 235U) 

< 2.7 

< 2.7 for the 

UO2 rods. 

See Note 4 

for MOX 

rods 

< 2.7 < 2.7 < 4.2 < 2.7 

Initial Maximum Rod 

Enrichment 

(wt.% 235U) 

< 4.0 < 4.0 < 4.0 < 5.5 < 5.0 < 4.0 

No. of Fuel Rod 

Locations 35 or 36 

35 or 36 

(up to 9 

MOX rods) 

36 49 49 63 or 64 

Fuel Clad O.D. (in.) > 0.5550 > 0.5625 > 0.5630 > 0.4860 > 0.5630 > 0.4120 

Fuel Clad I.D. (in.) < 0.5105 < 0.4945 < 0.4990 < 0.4204 < 0.4990 < 0.3620 

Fuel Pellet Dia. (in.) < 0.4980 < 0.4820 < 0.4880 < 0.4110 < 0.4910 < 0.3580 

Fuel Rod Pitch (in.) < 0.710 < 0.710 < 0.740 < 0.631 < 0.738 < 0.523 

Active Fuel 

Length (in.) < 120 < 120 < 77.5 < 80 < 150 < 120 

No. of Water Rods 

(Note 11) 1 or 0 1 or 0 0 0 0 1 or 0 

Water Rod 

Thickness (in.) > 0 > 0 N/A N/A N/A > 0 

Channel 

Thickness (in.) < 0.060 < 0.060 < 0.060 < 0.060 < 0.120 < 0.100

A-43
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.3 (Page 2 of 6) 

BWR FUEL ASSEMBLY CHARACTERISTICS (Note 1) 

Fuel Assembly 

Array/Class 8x8B 8x8C 8x8D 8x8E 8x8F 9x9A

Clad Material 

(Note 2) ZR ZR ZR ZR ZR ZR 

Design Initial U 

(kg/assy.) (Note 3) < 185 < 185 < 185 < 185 < 185 < 177 

Maximum planaraverage initial 

enrichment 

(wt.% 235U) 

< 4.2 < 4.2 < 4.2 < 4.2 < 4.0 < 4.2 

Initial Maximum Rod 

Enrichment 

(wt.% 235U) 

< 5.0 < 5.0 < 5.0 < 5.0 < 5.0 < 5.0 

No. of Fuel Rod 

Locations 63 or 64 62 60 or 61 59 64 74/66 

(Note 5) 

Fuel Clad O.D. (in.) > 0.4840 > 0.4830 > 0.4830 > 0.4930 > 0.4576 > 0.4400 

Fuel Clad I.D. (in.) < 0.4295 < 0.4250 0.4230 < 0.4250 < 0.3996 < 0.3840 

Fuel Pellet Dia. (in.) < 0.4195 < 0.4160 < 0.4140 < 0.4160 < 0.3913 < 0.3760 

Fuel Rod Pitch (in.) < 0.642 < 0.641 < 0.640 < 0.640 < 0.609 < 0.566 

Design Active Fuel 

Length (in.) < 150 < 150 < 150 < 150 < 150 < 150 

No. of Water Rods 

(Note 11) 1 or 0 2 1 - 4 

(Note 7) 5 N/A 

(Note 12) 2 

Water Rod 

Thickness (in.) > 0.034 > 0.00 > 0.00 > 0.034 > 0.0315 > 0.00 

Channel 

Thickness (in.) < 0.120 < 0.120 < 0.120 < 0.100 < 0.055 < 0.120

A-44
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.3 (Page 3 of 6) 

BWR FUEL ASSEMBLY CHARACTERISTICS (Note 1) 

Fuel Assembly 

Array/Class 9x9B 9x9C 9x9D 9x9E 

(Note 13)

9x9F 

(Note 13) 9x9G

Clad Material 

(Note 2) ZR ZR ZR ZR ZR ZR 

Design Initial U 

(kg/assy.) (Note 3) < 177 < 177 < 177 < 177 < 177 < 177 

Maximum planaraverage initial 

enrichment 

(wt.% 235U) 

< 4.2 < 4.2 < 4.2 < 4.0 < 4.0 < 4.2 

Initial Maximum 

Rod Enrichment 

(wt.% 235U) 

< 5.0 < 5.0 < 5.0 < 5.0 < 5.0 < 5.0 

No. of Fuel Rods 72 80 79 76 76 72 

Fuel Clad O.D. (in.) > 0.4330 > 0.4230 > 0.4240 > 0.4170 > 0.4430 > 0.4240 

Fuel Clad I.D. (in.) < 0.3810 < 0.3640 < 0.3640 < 0.3640 < 0.3860 < 0.3640 

Fuel Pellet Dia. 

(in.) < 0.3740 < 0.3565 < 0.3565 < 0.3530 < 0.3745 < 0.3565 

Fuel Rod Pitch (in.) < 0.572 < 0.572 < 0.572 < 0.572 < 0.572 < 0.572 

Design Active Fuel 

Length (in.) < 150 < 150 < 150 < 150 < 150 < 150 

No. of Water Rods 

(Note 11) 

1 

(Note 6) 1 2 5 5 1 

(Note 6) 

Water Rod 

Thickness (in.) > 0.00 > 0.020 > 0.0300 > 0.0120 > 0.0120 > 0.0320 

Channel 

Thickness (in.) < 0.120 < 0.100 < 0.100 < 0.120 < 0.120 < 0.120

A-45
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.3 (Page 4 of 6) 

BWR FUEL ASSEMBLY CHARACTERISTICS (Note 1) 

Fuel Assembly 

Array/Class 10x10A 10x10B 10x10C 10x10D 10x10E

Clad Material (Note 2) ZR ZR ZR SS SS 

Design Initial U 

(kg/assy.) (Note 3) < 186 < 186 < 186 < 125 < 125 

Maximum planaraverage initial 

enrichment (wt.% 235U) 

< 4.2 < 4.2 < 4.2 < 4.0 < 4.0 

Initial Maximum Rod 

Enrichment (wt.% 235U) < 5.0 < 5.0 < 5.0 < 5.0 < 5.0 

No. of Fuel Rod 

Locations 

92/78 

(Note 8) 91/83 (Note 9) 96 100 96 

Fuel Clad O.D. (in.) > 0.4040 > 0.3957 > 0.3780 > 0.3960 > 0.3940 

Fuel Clad I.D. (in.) < 0.3520 < 0.3480 < 0.3294 < 0.3560 < 0.3500 

Fuel Pellet Dia. (in.) < 0.3455 < 0.3420 < 0.3224 < 0.3500 < 0.3430 

Fuel Rod Pitch (in.) < 0.510 < 0.510 < 0.488 < 0.565 < 0.557 

Design Active Fuel 

Length (in.) < 150 < 150 < 150 < 83 < 83 

No. of Water Rods (Note 

11) 2 1 (Note 6) 5 (Note 10) 0 4 

Water Rod Thickness 

(in.) > 0.0300 > 0.00 > 0.031 N/A > 0.022 

Channel Thickness (in.) < 0.120 < 0.120 < 0.055 < 0.080 < 0.080

A-46
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.3 (Page 5 of 6) 

BWR FUEL ASSEMBLY CHARACTERISTICS (Note 1) 

Fuel Assembly Array/Class 6x6D 7x7C 

Clad Material 

(Note 2) 

Zr Zr 

Design Initial U 

(kg/assy.)(Note 3) 

≤ 78 ≤ 78 

Maximum planar-average 

initial enrichment (wt.% 235U) 

≤ 2.6 ≤ 2.6 

Initial Maximum Rod 

Enrichment (wt.% 235 U) 

≤ 4.0 

(Note 14) 

≤ 4.0 

No. of Fuel Rod Locations 36 49 

Fuel Clad O.D. (in.) ≥ 0.5585 ≥ 0.486 

Fuel Clad I.D. (in.) ≤ 0.505 ≤ 0.426 

Fuel Pellet Dia. (in.) ≤ 0.488 ≤ 0.411 

Fuel Rod Pitch (in.) ≤ 0.740 ≤ 0.631 

Active Fuel Length (in.) ≤ 80 ≤ 80 

No. of Water Rods (Note 11) 0 0 

Water Rod Thickness (in.) N/A N/A 

Channel Thickness (in.) ≤ 0.060 ≤ 0.060 

A-47
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.3 (Page 6 of 6) 

BWR FUEL ASSEMBLY CHARACTERISTICS (Note 1) 

Notes: 

1. All dimensions are design nominal values. Maximum and minimum dimensions are 

specified to bound variations in design nominal values among fuel assemblies within a given 

array/class. 

2. ZR designates cladding material made from Zirconium or Zirconium alloys. 

3. Design initial uranium weight is the uranium weight specified for each assembly by the fuel 

manufacturer or reactor user. For each BWR fuel assembly, the total uranium weight limit 

specified in this table may be increased up to 1.5% for comparison with users’ fuel records 

to account for manufacturer’s tolerances. 

4. < 0.635 wt. % 235U and < 1.578 wt. % total fissile plutonium (239Pu and 241Pu), (wt. % of total 

fuel weight, i.e., UO2 plus PuO2). 

5. This assembly class contains 74 total fuel rods; 66 full length rods and 8 partial length rods. 

6. Square, replacing nine fuel rods. 

7. Variable 

8. This assembly class contains 92 total fuel rods; 78 full length rods and 14 partial length 

rods. 

9. This assembly class contains 91 total fuel rods, 83 full length rods and 8 partial length rods. 

10. One diamond-shaped water rod replacing the four center fuel rods and four rectangular 

water rods dividing the assembly into four quadrants. 

11. These rods may be sealed at both ends and contain Zr material in lieu of water. 

12. This assembly is known as “QUAD+” and has four rectangular water cross segments 

dividing the assembly into four quadrants. 

13. For the SPC 9x9-5 fuel assembly, each fuel rod must meet either the 9x9E or 9x9F set of 

limits for clad O.D., clad I.D., and pellet diameter. 

14. Only two assemblies may contain one rod each with an initial maximum enrichment up to 

5.5 wt%. 

A-48
Appendix A - Certificate of Compliance 9261, Revision 7

 Table A.4 

 FUEL ASSEMBLY COOLING, AVERAGE BURNUP, AND INITIAL ENRICHMENT 

MPC-24/24E/24/EF PWR FUEL WITH ZIRCALOY CLAD AND 

WITH NON-ZIRCALOY IN-CORE GRID SPACERS 

Post-irradiation 

Cooling Time 

(years)

Assembly Burnup 

(MWD/MTU)

Assembly Initial 

Enrichment

(wt. % U-235)

> 9 < 24,500 > 2.3 

> 11 < 29,500 > 2.6 

> 13 < 34,500 > 2.9 

> 15 < 39,500 > 3.2 

> 18 < 44,500 > 3.4 

 Table A.5 

 FUEL ASSEMBLY COOLING, AVERAGE BURNUP, AND INITIAL ENRICHMENT 

MPC-24/24E/24EF PWR FUEL WITH ZIRCALOY CLAD AND 

WITH ZIRCALOY IN-CORE GRID SPACERS 

Post-irradiation 

Cooling Time 

(years)

Assembly Burnup 

(MWD/MTU)

Assembly Initial 

Enrichment

(wt. % U-235)

> 6 < 24,500 > 2.3 

> 7 < 29,500 > 2.6 

> 9 < 34,500 > 2.9 

> 11 < 39,500 > 3.2 

> 14 < 44,500 > 3.4 

A-49
Appendix A - Certificate of Compliance 9261, Revision 7

 Table A.6 

 FUEL ASSEMBLY COOLING, AVERAGE BURNUP, AND INITIAL ENRICHMENT 

 MPC-24/24E/24EF PWR FUEL WITH STAINLESS STEEL CLAD 

Post-irradiation 

Cooling Time 

(years)

Assembly Burnup 

 (MWD/MTU)

Assembly Initial 

Enrichment

(wt. % U-235)

> 19 < 30,000 > 3.1 

> 24 < 40,000 > 3.1 

 Table A.7 

 FUEL ASSEMBLY COOLING, AVERAGE BURNUP, AND INITIAL ENRICHMENT 

 MPC-68 

Post-irradiation 

Cooling Time 

(years)

Assembly Burnup 

(MWD/MTU)

Assembly Initial 

Enrichment 

(wt. % U-235)

> 5 < 10,000 > 0.7 

> 7 < 20,000 > 1.35 

> 8 < 24,500 > 2.1 

> 9 < 29,500 > 2.4 

> 11 < 34,500 > 2.6 

> 14 < 39,500 > 2.9 

> 19 < 44,500 > 3.0 

A-50
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.8 

TROJAN PLANT FUEL ASSEMBLY COOLING, AVERAGE BURNUP, 

 AND INITIAL ENRICHMENT LIMITS (Note 1) 

Post-irradiation Cooling 

Time (years)

Assembly Burnup

(MWD/MTU)

Assembly Initial 

Enrichment 

(wt.% 235U)

≥16 ≤42,000 ≥3.09 

≥16 ≤37,500 ≥2.6 

≥16 ≤30,000 ≥2.1 

NOTES: 

1. Each fuel assembly must only meet one set of limits (i.e., one row) 

Table A.9 

TROJAN PLANT NON-FUEL HARDWARE AND NEUTRON SOURCES 

COOLING AND BURNUP LIMITS 

Type of Hardware or 

Neutron Source

Burnup

(MWD/MTU)

Post-irradiation Cooling 

Time 

(Years)

BPRAs ≤15,998 ≥24 

TPDs ≤118,674 ≥11 

RCCAs ≤125,515 ≥9 

Cf neutron source ≤15,998 ≥24 

Sb-Be neutron source with 4 

source rods, 16 burnable 

poison rods, and 4 thimble 

plug rods 

≤45,361 ≥19 

Sb-Be neutron source with 4 

source rods, 20 thimble plug 

rods 

≤88,547 ≥9 

A-51
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.10 

FUEL ASSEMBLY COOLING, AVERAGE BURNUP, AND MINIMUM ENRICHMENT MPC-32 

PWR FUEL WITH ZIRCALOY CLAD AND WITH NON-ZIRCALOY IN-CORE GRID SPACERS 

Post-irradiation

cooling time 

(years)

Assembly burnup

(MWD/MTU)

Assembly Initial 

Enrichment 

(wt. % U-235)

≥12 ≤24,500 ≥2.3 

≥14 ≤29,500 ≥2.6 

≥16 ≤34,500 ≥2.9 

≥19 ≤39,500 ≥3.2 

≥20 ≤42,500 ≥3.4 

Table A.11 

FUEL ASSEMBLY COOLING, AVER AGE BURNUP, AND MINIMUM ENRICHMENT MPC-32 

PWR FUEL WITH ZIRCALOY CLAD AND WITH ZIRCALOY IN-CORE GRID SPACERS 

Post-irradiation

cooling time 

(years)

Assembly burnup

(MWD/MTU)

Assembly Initial 

Enrichment 

(wt.% U-235)

≥8 ≤24,500 ≥2.3 

≥9 ≤29,500 ≥2.6 

≥12 ≤34,500 ≥2.9 

≥14 ≤39,500 ≥3.2 

≥19 ≤44,500 ≥3.4 

A-52
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.12 

FUEL ASSEMBLY MAXIMUM ENRICHMENT AND MINIMUM BURNUP REQUIREMENTS 

FOR TRANSPORTATION IN MPC-32 

Fuel Assembly 

Array/Class

Configur

ation 

(Note 2)

Maximum 

Enrichment 

(wt.% U235)

Minimum Burnup (B) as a Function of 

Initial Enrichment (E) (Note 1) (GWD/MTU)

15x15D, E, F, H A 4.65 B = (1.6733)*E3-(18.72)*E2+(80.5967)*E-88.3 

 B 4.38 B = (2.175)*E3-(23.355)*E2+(94.77)*E-99.95 

 C 4.48 B = (1.9517)*E3-(21.45)*E2+(89.1783)*E-94.6 

 D 4.45 B = (1.93)*E3-(21.095)*E2+(87.785)*E-93.06 

17x17A,B,C A 4.49 B = (1.08)*E3-(12.25)*E2+(60.13)*E-70.86 

 B 4.04 B = (1.1)*E3-(11.56)*E2+(56.6)*E-62.59 

 C 4.28 B = (1.36)*E3-(14.83)*E2+(67.27)*E-72.93 

 D 4.16 B = (1.4917)*E3-(16.26)*E2+(72.9883)*E-79.7 

NOTES: 

1. E = Initial enrichment (e.g., for 4.05 wt.% , E = 4.05). 

2. See Table A.13. 

3. Fuel Assemblies must be cooled 5 years or more. 

A-53
Appendix A - Certificate of Compliance 9261, Revision 7

Table A.13 

LO ADING CONFIGURATIONS FOR THE MPC-32 

CONFIGURATION ASSEMBLY SPECIFICATIONS

A $ Assemblies that have not been located in any cycle under a 

control rod bank that was permitted to be inserted during full 

power operation (per plant operating procedures); or 

$ Assemblies that have been located under a control rod bank 

that was permitted to be inserted during full power operation 

(per plant operating procedures), but where it can be 

demonstrated, based on operating records, that the insertion 

never exceeded 8 inches from the top of the active length 

during full power operation. 

B $ Of the 32 assemblies in a basket, up to 8 assemblies can be 

from core locations where they were located under a control 

rod bank, that was permitted to be inserted more than 8 

inches during full power operation. There is no limit on the 

duration (in terms of burnup) under this bank. 

$ The remaining assemblies in the basket must satisfy the 

same conditions as specified for configuration A. 

C $ Of the 32 assemblies in a basket, up to 8 assemblies can be 

from core locations where they were located under a control 

rod bank, that was permitted to be inserted more than 8 

inches during full power operation. Location under such a 

control rod bank is limited to 20 GWD/MTU of the assembly. 

$ The remaining assemblies in the basket must satisfy the 

same conditions as specified for configuration A. 

D $ Of the 32 assemblies in a basket, up to 8 assemblies can be 

from core locations where they were located under a control 

rod bank, that was permitted to be inserted more than 8 

inches during full power operation. Location under such a 

control rod bank is limited to 30 GWD/MTU of the assembly. 

$ The remaining assemblies in the basket must satisfy the 

same conditions as specified for configuration A. 

A-54
Appendix A - Certificate of Compliance 9261, Revision 7

REFERENCES:

Holtec International Report No. HI-951251, Safety Analysis Report for the Holtec International 

Storage, Transport, And Repository Cask System (HI-STAR 100 Cask System), Revision 12, 

dated October 6, 2006, as supplemented.

A-55
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES 1 a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 

9235 11 71-9235 USA/9235/B(U)F-96 1 OF 12 

2. PREAMBLE 

a. This certificate is issued to certify that the package (packaging and contents) described in Item 5 below meets the applicable safety standards set 

forth in Title 10, Code of Federal Regulations, Part 71, “Packaging and Transportation of Radioactive Material.” 

b. This certificate does not relieve the consignor from compliance with any requirement of the regulations of the U.S. Department of Transportation or 

other applicable regulatory agencies, including the government of any country through or into which the package will be transported. 

3. THIS CERTIFICATE IS ISSUED ON THE BASIS OF A SAFETY ANALYSIS REPORT OF THE PACKAGE DESIGN OR APPLICATION 

a. ISSUED TO (Name and Address) b. TITLE AND IDENTIFICATION OF REPORT OR APPLICATION 

NAC International 

3930 East Jones Bridge Road, Suite 200 

Norcross, Georgia 30092 

NAC International, Inc., application dated 

February 19, 2009. 

4. CONDITIONS 

 This certificate is conditional upon fulfilling the requirements of 10 CFR Part 71, as applicable, and the conditions specified below. 

5 (a) Packaging

 (1) Model No.: NAC-STC 

 (2) Description: For descriptive purposes, all dimensions are approximate nominal values. 

Actual dimensions with tolerances are as indicated on the Drawings. 

A steel, lead and polymer (NS4FR) shielded shipping cask for (a) directly loaded irradiated 

PWR fuel assemblies, (b) intact, damaged and/or the fuel debris of Yankee Class or 

Connecticut Yankee irradiated PWR fuel assemblies in a canister, and (c) non-fissile, solid 

radioactive materials (referred to hereafter as Greater Than Class C (GTCC) as defined in 

10 CFR Part 61) waste in a canister. The cask body is a right circular cylinder with an impact 

limiter at each end. The package has approximate dimensions as follows: 

 Cavity diameter 71 inches 

 Cavity length 165 inches 

 Cask body outer diameter 87 inches 

 Neutron shield outer diameter 99 inches 

 Lead shield thickness 3.7 inches 

 Neutron shield thickness 5.5 inches 

 Impact limiter diameter 124 inches 

 Package length: 

 without impact limiters 193 inches 

 with impact limiters 257 inches 

The maximum gross weight of the package is about 260,000 lbs. 

The cask body is made of two concentric stainless steel shells. The inner shell is 1.5 inches 

thick and has an inside diameter of 71 inches. The outer shell is 2.65 inches thick and has 

A-56
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES 1 a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 

9235 11 71-9235 USA/9235/B(U)F-96 2 OF 12 

5.(a)(2) Description (Continued) 

an outside diameter of 86.7 inches. The annulus between the inner and outer shells is filled 

with lead. 

The inner and outer shells are welded to steel forgings at the top and bottom ends of the 

cask. The bottom end of the cask consists of two stainless steel circular plates which are 

welded to the bottom end forging. The inner bottom plate is 6.2 inches thick and the outer 

bottom plate is 5.45 inches thick. The space between the two bottom plates is filled with a 

2-inch thick disk of a synthetic polymer (NS4FR) neutron shielding material. 

The cask is closed by two steel lids which are bolted to the upper end forging. The inner lid 

(containment boundary) is 9 inches thick and is made of Type 304 stainless steel. The outer 

lid is 5.25 inches thick and is made of SA-705 Type 630, H1150 or 17-4PH stainless steel. 

The inner lid is fastened by 42, 1-1/2-inch diameter bolts and the outer lid is fastened by 36, 

1-inch diameter bolts. The inner lid is sealed by two O-ring seals. The outer lid is equipped 

with a single O-ring seal. The inner lid is fitted with a vent and drain port which are sealed by 

O-rings and cover plates. The containment system seals may be metallic or Viton. Viton 

seals are used only for directly-loaded fuel that is to be shipped without long-term interim 

storage. 

The cask body is surrounded by a 1/4-inch thick jacket shell constructed of 24 stainless steel 

plates. The jacket shell is 99 inches in diameter and is supported by 24 longitudinal stainless 

steel fins which are connected to the outer shell of the cask body. Copper plates are bonded 

to the fins. The space between the fins is filled with NS4FR shielding material. 

Four lifting trunnions are welded to the top end forging. The package is shipped in a 

horizontal orientation and is supported by a cradle under the top forging and by two trunnion 

sockets located near the bottom end of the cask. 

The package is equipped at each end with an impact limiter made of redwood and balsa. 

Two impact limiter designs consisting of a combination of redwood and balsa wood, encased 

in Type 304 stainless steel are provided to limit the g-loads acting on the cask during an 

accident. The predominantly balsa wood impact limiter is designed for use with all the 

proposed contents. The predominately redwood impact limiters may only be used with 

directly loaded fuel or the Yankee-MPC configuration. 

The contents are transported either directly loaded (uncanistered) into a stainless steel fuel 

basket or within a stainless steel transportable storage canister (TSC). 

The directly loaded fuel basket within the cask cavity can accommodate up to 26 PWR fuel 

assemblies. The fuel assemblies are positioned within square sleeves made of stainless 

steel. Boral or TalBor sheets are encased outside the walls of the sleeves. The sleeves are 

laterally supported by 31, 1⁄2-inch thick, 71-inch diameter stainless steel disks. The basket 

also has 20 heat transfer disks made of Type 6061-T651 aluminum alloy. The support disks 

A-57
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 

9235 11 71-9235 USA/9235/B(U)F-96 3 OF 12 

5.(a)(2) Description (Continued) 

and heat transfer disks are connected by six, 1-5/8-inch diameter by 161-inch long threaded 

rods made of Type 17-4 PH stainless steel. 

The TSC shell, bottom plate, and welded shield and structural lids are fabricated from 

stainless steel. The bottom is a 1-inch thick steel plate for the Yankee-MPC and 1.75-inch 

thick steel plate for the CY-MPC. The shell is constructed of 5/8-inch thick rolled steel plate 

and is 70 inches in diameter. The shield lid is a 5-inch thick steel plate and contains drain 

and fill penetrations for the canister. The structural lid is a 3-inch thick steel plate. The 

canister contains a stainless steel fuel basket that can accommodate up to 36 intact Yankee 

Class fuel assemblies and Reconfigured Fuel Assemblies (RFAs), or up to 26 intact 

Connecticut Yankee fuel assemblies with RFAs, with a maximum weight limit of 35,100 lbs. 

Alternatively, a stainless steel GTCC waste basket is used for up to 24 containers of waste. 

One TSC fuel basket configuration can store up to 36 intact Yankee Class fuel assemblies or 

up to 36 RFAs within square sleeves made of stainless steel. Boral sheets are encased 

outside the walls of the sleeves. The sleeves are laterally supported by 22 1⁄2-inch thick , 

69-inch diameter stainless steel disks, which are spaced about 4 inches apart. The support 

disks are retained by split spacers on eight 1.125-inch diameter stainless steel tie rods. The 

basket also has 14 heat transfer disks made of Type 6061-T651 aluminum alloy. 

The second fuel basket is designed to store up to 26 Connecticut Yankee Zirc-clad 

assemblies enriched to 3.93 wt. percent, stainless steel clad assemblies enriched up to 4.03 

wt. percent, RFAs, or damaged fuel in CY-MPC damaged fuel cans (DFCs). Zirc-clad fuel 

enriched to between 3.93 and 4.61 wt. percent, such as Westinghouse Vantage 5H fuel, 

must be stored in the 24-assembly basket. Assemblies approved for transport in the 26-

assembly configuration may also be shipped in the 24-assembly configuration. The 

construction of the two basket configurations is identical except that two fuel loading 

positions of the 26-assembly basket are blocked to form the 24-assembly basket. 

RFAs can accommodate up to 64 Yankee Class fuel rods or up to 100 Connecticut Yankee 

fuel rods, as intact or damaged fuel or fuel debris, in an 8x8 or 10x10 array of stainless steel 

tubes, respectively. Intact and damaged Yankee Class or Connecticut Yankee fuel rods, as 

well as fuel debris, are held in the fuel tubes. The RFAs have the same external dimensions 

as a standard intact Yankee Class, or Connecticut Yankee fuel assembly. 

The TSC GTCC basket positions up to 24 Yankee Class or Connecticut Yankee waste 

containers within square stainless steel sleeves. The Yankee Class basket is supported 

laterally by eight 1-inch thick, 69-inch diameter stainless steel disks. The Yankee Class 

basket sleeves are supported full-length by 2.5-inch thick stainless steel support walls. The 

support disks are welded into position at the support walls. The Connecticut Yankee GTCC 

basket is a right-circular cylinder formed by a series of 1.75-inch thick Type 304 stainless 

steel plates, laterally supported by 12 equally spaced welded 1.25-inch thick Type 304 

stainless steel outer ribs. The GTCC waste containers accommodate radiation activated and 

surface contaminated steel, cutting debris (dross) or filter media, and have the same external 

dimensions of Yankee Class or Connecticut Yankee fuel assemblies. 

A-58
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 

9235 11 71-9235 USA/9235/B(U)F-96 4 OF 12 

5.(a)(2) Description (Continued) 

The Yankee Class TSC is axially positioned in the cask cavity by two aluminum honeycomb 

spacers. The spacers, which are enclosed in a Type 6061-T651 aluminum alloy shell, 

position the canister within the cask during normal conditions of transport. The bottom 

spacer is 14-inches high and 70-inches in diameter, and the top spacer is 28-inches high and 

also 70-inches in diameter. 

The Connecticut Yankee TSC is axially positioned in the cask cavity by one stainless steel 

spacer located in the bottom of the cask cavity. 

5.(a)(3) Drawings 

(i) The cask is constructed and assembled in accordance with the following Nuclear Assurance 

Corporation (now NAC International) Drawing Nos.: 

 

 423-800, sheets 1-3, Rev. 14 

423-802, sheets 1-7, Rev. 20 

423-803, sheets 1-2, Rev. 8 

 423-804, sheets 1-3, Rev. 8 

423-805, sheets 1-2, Rev. 6 

423-806, Rev. 7 

423-807, sheets 1-3, Rev. 3 

 423-811, sheets 1-2, Rev. 11 

 423-812, Rev. 6 

 423-900, Rev. 6 

 423-209, Rev. 0 

 423-210, Rev. 0 

 423-901, Rev. 2 

 

(ii) For the directly loaded configuration, the basket is constructed and assembled in accordance 

with the following Nuclear Assurance Corporation (now NAC International) Drawing Nos.: 

 

 423-870, Rev. 5 

423-871, Rev. 5 

423-872, Rev. 6 

 

 423-873, Rev. 2 

 423-874, Rev. 2 

 423-875, sheets 1-2, Rev. 7 

(iii) For the Yankee Class TSC configuration, the canister, and the fuel and GTCC waste baskets 

are constructed and assembled in accordance with the following NAC International Drawing Nos.: 

 

 455-800, sheets 1-2, Rev. 2 

 455-801, sheets 1-2, Rev. 3 

 455-820, sheets 1-2, Rev. 2 

 455-870, Rev. 5 

 455-871, sheets 1-2, Rev. 8 

 455-871, sheets 1-3, Rev. 7P21

 455-872, sheets 1-2, Rev. 12 

 455-872, sheets 1-2, Rev. 11P11

 455-873, Rev. 4 

 455-881, sheets 1-3, Rev. 8 

 455-887, sheets 1-3, Rev. 4 

 

 455-888, sheets 1-2, Rev. 8 

 455-891, sheets 1-2, Rev. 1 

455-891, sheets 1-3, Rev. 2PO1

455-892, sheets 1-2, Rev. 3 

 455-892, sheets 1-3, Rev. 3P01

 455-893, Rev. 3 

 455-894, Rev. 2 

 455-895, sheets 1-2, Rev. 5 

 455-895, sheets 1-2, Rev. 5P01

 455-901, Rev. 0P01

 455-902, sheets 1-5, Rev. 0P41

 455-919, Rev. 2 

 

1

Drawing defines the alternate configuration that accommodates the Yankee-MPC damaged fuel can. 

A-59
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES

1 a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 

9235 11 71-9235 USA/9235/B(U)F-96 5 OF 12 

5.(a)(3) Drawings (Continued) 

(iv) For the Yankee Class TSC configuration, RFAs are constructed and assembled in accordance 

with the following Yankee Atomic Electric Company Drawing Nos.: 

 

 YR-00-060, Rev. D3 

 YR-00-061, Rev. D4 

 YR-00-062, sheet 1, Rev. D4 

 YR-00-062, sheet 2, Rev. D2 

 YR-00-062, sheet 3, Rev. D1 

 YR-00-063, Rev. D4 

 YR-00-064, Rev. D4 

 YR-00-065, Rev. D2 

 YR-00-066, sheet 1, Rev. D5 

 YR-00-066, sheet 2, Rev. D3 

(v) The Balsa Impact Limiters are constructed and assembled in accordance with the following NAC 

International Drawing Nos.: 

 423-257, Rev. 2 

 423-258, Rev. 2 

 423-843, Rev. 2 

 423-859, Rev. 0 

(vi) For the Connecticut Yankee TSC configuration, the canister and the fuel and GTCC waste 

baskets are constructed and assembled in accordance with the following NAC International Drawing 

Nos.: 

 414-801, sheets, 1-2 Rev. 1 

414-820, Rev. 0 

 414-870, Rev. 3 

 414-871, sheets 1-2, Rev. 6 

 414-872, sheets 1-3, Rev. 6 

 414-873, Rev. 2 

 414-874, Rev. 0 

 414-875, Rev. 0 

 414-881, sheets 1-2, Rev. 4 

 414-882, sheets 1-2, Rev. 4 

 414-887, sheets 1-4, Rev. 4 

 414-888, sheets 1-2, Rev. 4 

 414-889, sheets 1-3, Rev. 7 

 414-891, Rev. 3 

 414-892, sheets 1-3, Rev. 3 

 414-893, sheets,1-2, Rev. 2 

 414-894, Rev. 0 

 414-895, sheets 1-2, Rev. 4 

(vii) For the Connecticut Yankee TSC configuration, DFCs and RFAs are constructed and 

assembled in accordance with the following NAC International Drawing Nos.: 

 414-901, Rev. 1 

 414-902, sheets 1-3, Rev. 3 

 414-903, sheets 1-2, Rev. 1 

 414-904, sheets 1-3, Rev. 0 

A-60
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES 1 a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 

9235 11 71-9235 USA/9235/B(U)F-96 6 OF 12 

5.(b) Contents 

(1) Type and form of material 

(i) Irradiated PWR fuel assemblies with uranium oxide pellets. Each fuel assembly may have 

a maximum burnup of 45 GWD/MTU. The minimum fuel cool time is defined in the Fuel Cool 

Time Table, below. The maximum heat load per assembly is 850 watts. Prior to irradiation, 

the fuel assemblies must be within the following dimensions and specifications: 

Assembly Type 14x14 15x15 16x16 17x17 17x17 

(OFA)

FramatomeCogema

17x17

Cladding Material Zirc-4 Zirc-4 Zirc-4 Zirc-4 Zirc-4 Zirconium 

Alloy

Maximum Initial 

Uranium Content 

(kg/assembly) 

407 469 402.5 464 426 464

Maximum Initial 

Enrichment (wt% 235U) 4.2 4.2 4.2 4.2 4.2 4.5

Minimum Initial 

Enrichment (wt% 235U) 1.7 1.7 1.7 1.7 1.7 1.7

Assembly CrossSection (inches) 7.76 

to 8.11 

8.20

to 8.54

8.10

 to 8.14

8.43

to 8.54 8.43 8.425to 8.518

Number of Fuel 

Rods per Assembly 

176 

to 179 

204

to 216 236 264 264 264(1)

Fuel Rod OD (inch) 0.422 

to 0.440 

0.418

to 0.430 0.382 0.374to 0.379 0.360 0.3714to 0.3740

Minimum Cladding 

Thickness (inch) 0.023 0.024 0.025 0.023 0.023 0.0204

Pellet Diameter (inch) 0.344 

to 0.377 

0.358

to 0.390 0.325 0.3225to 0.3232 0.3088 0.3224to 0.3230

Maximum Active Fuel 

Length (inches) 146 144 137 144 144 144.25

Notes: 

(1) - Fuel rod positions may also be occupied by solid poison shim rods or solid zirconium alloy or 

stainless steel fill rods. 

A-61
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 

9235 11 71-9235 USA/9235/B(U)F-96 7 OF 12

5.(b)(1)(i) Contents - Type and Form of Material - Irradiated PWR fuel assemblies (Continued) 

 

FUEL COOL TIME TABLE 

Minimum Fuel Cool Time in Years

Fuel Assembly Burnup (BU)

Uranium 

Enrichment 

(wt% U-235) 

BU < 30 

GWD/MTU 

30 < BU < 35 

GWD/MTU 

35 < BU < 40 

GWD/MTU 

40 < BU < 45 

GWD/MTU

Fuel Type 14x14 15x15 16x16 17x17 14x14 15x15 16x16 17x17 14x14 15x15 16x16 17x17 14x14 15x15 16x16 17x17

1.7<E<1.9 8 7 6 7 10 10 7 9 -- -- -- -- -- -- -- -- 

1.9<E<2.1 7 7 5 7 9 9 7 8 12 13 9 11 -- -- -- -- 

2.1<E<2.3 7 7 5 6 9 8 6 8 11 11 8 10 -- -- -- -- 

2.3<E<2.5 6 6 5 6 8 8 6 7 10 10 8 9 14 15 12 14 

2.5<E<2.7 6 6 5 6 8 7 6 7 10 9 7 9 13 14 10 12 

2.7<E<2.9 6 6 5 5 7 7 5 6 9 9 7 8 12 12 9 11 

2.9<E<3.1 6 5 5 5 7 7 5 6 9 8 6 8 11 11 8 10 

3.1<E<3.3 5 5 5 5 7 6 5 6 8 8 6 7 10 10 8 9 

3.3<E<3.5 5 5 5 5 6 6 5 6 8 7 6 7 10 10 7 9 

3.5<E<3.7 5 5 5 5 6 6 5 6 7 7 6 7 9 9 7 9 

3.7<E<3.9 5 5 5 5 6 6 5 6 7 7 6 7 9 9 7 9 

3.9<E<4.1 5 5 5 5 6 6 5 6 7 7 6 7 8 9 7 9 

4.1<E<4.2 5 5 5 5 5 6 5 6 6 7 6 7 8 8 7 9 

4.2<E<4.3 -- -- -- 5(1) -- -- -- 6(1) -- -- -- 7(1) -- -- -- 9(1)

4.3<E<4.5 -- -- -- 5(1) -- -- -- 6(1) -- -- -- 7(1) -- -- -- 8(1)

Notes: 

(1) - Framatome-Cogema 17x17 fuel only. 

A-62
NRC FORM 

618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES 1 a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 

9235 11 71-9235 USA/9235/B(U)F-96 8 OF 12 

5.(b)(1) Contents - Type and Form of Material (Continued) 

(ii) Irradiated intact Yankee Class PWR fuel assemblies or RFAs within the TSC. The 

maximum initial fuel pin pressure is 315 psig. The fuel assemblies consist of uranium 

oxide pellets with the specifications, based on design nominal or operating history record 

values, listed below: 

Assembly 

Manufacturer/Type 

UN 

 16x16 

CE 1

 16x16 

West. 

18x18 

Exxon 2

16x16 

Yankee 

RFA 

Yankee 

DFC 

Cladding Material Zircaloy Zircaloy SS Zircaloy Zirc/SS Zirc/SS

Maximum Number of Rods 

per Assembly 237 231 305 231 64 305

Maximum Initial 

Uranium Content 

(kg/assembly) 

246 240 287 240 70 287

Maximum Initial 

Enrichment 

(wt% 235U) 

4.0 3.9 4.94 4.0 4.94 4.97 3

Minimum Initial Enrichment 

(wt% 235U) 4.0 3.7 4.94 3.5 3.5 3.5 3

Maximum Assembly 

Weight (lbs) ≤ 950 ≤ 950 ≤ 950 ≤ 950 ≤ 950 ≤ 950

Maximum Burnup 

(MWD/MTU) 32,000 36,000 32,000 36,000 36,000 36,000

Maximum Decay Heat per 

Assembly (kW) 0.28 0.347 0.28 0.34 0.11 0.347

Minimum Cool Time 

(yrs) 11.0 8.1 22.0 10.0 8.0 8.0

Maximum Active Fuel 

Length (in) 91 91 92 91 92 N/A

 Notes: 

1. Combustion Engineering (CE) fuel with a maximum burnup of 32,000 MWD/MTU, a minimum enrichment of 

3.5 wt. percent 235U, a minimum cool time of 8.0 years, and a maximum decay heat per assembly of 0.304 

kW is authorized.

2. Exxon assemblies with stainless steel in-core hardware shall be cooled a minimum of 16.0 years with a 

maximum decay heat per assembly of 0.269 kW. 

3. Stated enrichments are nominal values (fabrication tolerances are not included).

A-63
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 

9235 11 71-9235 USA/9235/B(U)F-96 9 OF 12 

5.(b)(1) Contents - Type and Form of Material (Continued) 

(iii) Solid, irradiated, and contaminated hardware and solid, particulate debris (dross) or filter 

media placed in a GTCC waste container, provided the quantity of fissile material does not 

exceed a Type A quantity, and does not exceed the mass limits of 10 CFR 71.15. 

(iv) Irradiated intact and damaged Connecticut Yankee (CY) Class PWR fuel assemblies 

(including optional stainless steel rods inserted into the CY intact and damaged fuel assembly 

reactor control cluster assembly (RCCA) guide tubes that do not contain RCCAs), RFAs, or 

DFCs within the TSC. The maximum initial fuel pin pressure is 475 psig. The fuel assemblies 

consist of uranium oxide pellets with the specifications, based on design nominal or operating 

history record values, listed below: 

Assembly Manufacturer/Type PWR 1

15x15 

PWR 2

15x15 PWR 3 CY-MPC 

RFA4

CY-MPC 

DFC5

Cladding Material SS Zircaloy Zircaloy Zirc/SS Zirc/SS

Maximum Number of Assemblies 26 26 24 4 4

Maximum Initial Uranium 

Content (kg/assembly) 433.7 397.1 390 212 433.7

Maximum Initial Enrichment 

(wt% 235U) 4.03 3.93 4.61 4.616

 4.616

Minimum Initial Enrichment (wt% 

235U) 3.0 2.95 2.95 2.95 2.95

Maximum Assembly Weight (lbs) ≤ 1,500 ≤ 1,500 ≤ 1,500 ≤ 1,600 ≤ 1,600

Maximum Burnup (MWD/MTU) 38,000 43,000 43,000 43,000 43,000

Maximum Decay Heat per 

Assembly (kW) 0.654 0.654 0.654 0.321 0.654

Minimum Cool Time (yrs) 10.0 10.0 10.0 10.0 10.0

Maximum Active Fuel Length (in) 121.8 121.35 120.6 121.8 121.8

Notes:

1. Stainless steel assemblies manufactured by Westinghouse Electric Co., Babcock & Wilcox Fuel Co., Gulf Gen. Atomics, Gulf 

Nuclear Fuel, & Nuclear Materials & Man. Co.

2. Zircaloy spent fuel assemblies manufactured by Gulf Gen. Atomics, Gulf Nuclear Fuel, & Nuclear Materials & Man. Co., and 

Babcock & Wilcox Fuel Co.

3. Westinghouse Vantage 5H zircaloy clad spent fuel assemblies have an initial uranium enrichment > 3.93 % wt. U235. 4. Reconfigured Fuel Assemblies (RFA) must be loaded in one of the 4 oversize fuel loading positions.

5. Damaged Fuel Cans (DFC) must be loaded in one of the 4 oversize fuel loading positions. 

6. Enrichment of the fuel within each DFC or RFA is limited to that of the basket configuration in which it is loaded.

A-64
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 

9235 9 71-9235 USA/9235/B(U)F-96 10 OF 12 

5.(b) Contents (Continued) 

(2) Maximum quantity of material per package 

(i) For the contents described in Item 5.(b)(1)(i): 26 PWR fuel assemblies with a 

maximum total weight of 39,650 lbs. and a maximum decay heat not to exceed 22.1 

kW per package. 

(ii) For the contents described in Item 5.(b)(1)(ii): Up to 36 intact fuel assemblies to the 

maximum content weight limit of 30,600 lbs. with a maximum decay heat of 12.5 kW 

per package. Intact fuel assemblies shall not contain empty fuel rod positions and any 

missing rods shall be replaced by a solid Zircaloy or stainless steel rod that displaces 

an equal amount of water as the original fuel rod. Mixing of intact fuel assembly types 

is authorized. 

(iii) For intact fuel rods, damaged fuel rods and fuel debris of the type described in Item 

5.(b)(1)(ii): up to 36 RFAs, each with a maximum equivalent of 64 full length Yankee 

Class fuel rods and within fuel tubes. Mixing of directly loaded intact assemblies and 

damaged fuel (within RFAs) is authorized. The total weight of damaged fuel within 

RFAs or mixed damaged RFA and intact assemblies shall not exceed 30,600 lbs. with 

a maximum decay heat of 12.5 kW per package. 

(iv) For the contents described in Item 5.(b)(1)(iii): for Connecticut Yankee GTCC waste up 

to 24 containers of GTCC waste. The total cobalt-60 activity shall not exceed 196,000 

curies. The total weight of the waste containers shall not exceed 18,743 lbs. with a 

maximum decay heat of 5.0 kW. For all others, up to 24 containers of GTCC waste. 

The total cobalt-60 activity shall not exceed 125,000 curies. The total weight of the 

waste and containers shall not exceed 12,340 lbs. with a maximum decay heat of 2.9 

kW. 

(i) For the contents described in Item 5.(b)(1)(iv): up to 26 Connecticut Yankee fuel 

assemblies, RFAs or damaged fuel in CY-MPC DFCs for stainless steel clad 

assemblies enriched up to 4.03 wt. percent and Zirc-clad assemblies enriched up to 

3.93 wt. percent. Westinghouse Vantage 5H fuel and other Zirc-clad assemblies 

enriched up to 4.61 wt. percent must be installed in the 24-assembly basket, which 

may also hold other Connecticut Yankee fuel types. The construction of the two 

basket configurations is identical except that two fuel loading positions of the 26 

assembly basket are blocked to form the 24 assembly basket. The total weight of 

damaged fuel within RFAs or mixed damaged RFAs and intact assemblies shall not 

exceed 35,100 lbs. with a maximum decay heat of 0.654 kW per assembly for a 

canister of 26 assemblies. A maximum decay heat of 0.321 kW per assembly for 

Connecticut Yankee RFAs and of 0.654 kW per canister for the Connecticut Yankee 

DFCs is authorized. 

5.(c) Criticality Safety Index: 0.0 

A-65
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 

9235 11 71-9235 USA/9235/B(U)F-96 11 OF 12 

6. Known or suspected damaged fuel assemblies or rods (fuel with cladding defects greater than pin 

holes and hairline cracks) are not authorized, except as described in Item 5.(b)(2)(iii). 

7. For contents placed in a GTCC waste container and described in Item 5.(b)(1)(iii): and which contain 

organic substances which could radiolytically generate combustible gases, a determination must be 

made by tests and measurements or by analysis that the following criteria are met over a period of 

time that is twice the expected shipment time: 

The hydrogen generated must be limited to a molar quantity that would be no more than 4% by 

volume (or equivalent limits for other inflammable gases) of the TSC gas void if present at STP (i.e., 

no more than 0.063 g-moles/ft3 at 14.7 psia and 70°F). For determinations performed by analysis, the 

amount of hydrogen generated since the time that the TSC was sealed shall be considered. 

8. For damaged fuel rods and fuel debris of the quantity described in Item 5.(b)(2)(iii) and 5.(b)(2)(v): if 

the total damaged fuel plutonium content of a package is greater than 20 Ci, all damaged fuel shall be 

enclosed in a TSC which has been leak tested at the time of closure. For the Yankee Class TSC the 

leak test shall have a test sensitivity of at least 4.0 X 10-8 cm3

/sec (helium) and shown to have a leak 

rate no greater than 8.0 X 10-8 cm3

/sec (helium). For the Connecticut Class TSC the leak test shall 

have a test sensitivity of at least 1.0 X 10-7 cm3

/sec (helium) and shown to have a leak rate no greater 

than 2.0 X 10-7 cm3

/sec (helium). 

9. In addition to the requirements of Subpart G of 10 CFR Part 71: 

 (a) The package must be prepared for shipment and operated in accordance with the Operating 

Procedures in Chapter 7 of the application, as supplemented. 

 (b) Each packaging must be acceptance tested and maintained in accordance with the 

Acceptance Tests and Maintenance Program in Chapter 8 of the application, as 

supplemented, except that the thermal testing of the package (including the thermal 

acceptance test and periodic thermal tests) must be performed as described in NAC-STC 

Safety Analysis Report. 

 (c) For packaging Serial Numbers STC-1 and STC-2, only one of these two packagings must be 

subjected to the thermal acceptance test as described in Section 8.1.6 of the NAC-STC Safety 

Analysis Report. 

10. Prior to transport by rail, the Association of American Railroads must have evaluated and approved 

the railcar and the system used to support and secure the package during transport. 

11. Prior to marine or barge transport, the National Cargo Bureau, Inc., must have evaluated and 

approved the system used to support and secure the package to the barge or vessel, and must have 

certified that package stowage is in accordance with the regulations of the Commandant, United 

States Coast Guard. 

A-66
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 

9235 11 71-9235 USA/9235/B(U)F-96 12 OF 12

12. Transport by air is not authorized. 

13. Packagings must be marked with Package Identification Number USA/9235/B(U)F-96. 

14. The package authorized by this certificate is hereby approved for use under the general license 

provisions of 10 CFR 71.17. 

15. Revision No. 9 of this certificate may be used until May 31, 2010. 

16. Expiration date: May 31, 2014. 

 

 REFERENCES

NAC International, Inc., application dated: February 19, 2009. 

As supplemented June 3, 2009. 

 

 FOR THE U.S. NUCLEAR REGULATORY COMMISSION 

 

 /RA/ 

 

 Eric J. Benner, Chief 

 Licensing Branch 

 Division of Spent Fuel Storage and Transportation 

 Office of Nuclear Material Safety 

 and Safeguards 

Date: June 12, 2009 

A-67
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES 1 a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 

9226 3 71-9226 USA/9226/B(U)F-85 1 OF 9 

2. PREAMBLE 

a. This certificate is issued to certify that the package (packaging and contents) described in Item 5 below meets the applicable safety standards set 

forth in Title 10, Code of Federal Regulations, Part 71, “Packaging and Transportation of Radioactive Material.” 

b. This certificate does not relieve the consignor from compliance with any requirement of the regulations of the U.S. Department of Transportation or 

other applicable regulatory agencies, including the government of any country through or into which the package will be transported. 

3. THIS CERTIFICATE IS ISSUED ON THE BASIS OF A SAFETY ANALYSIS REPORT OF THE PACKAGE DESIGN OR APPLICATION 

a. ISSUED TO (Name and Address) b. TITLE AND IDENTIFICATION OF REPORT OR APPLICATION 

General Atomics 

3550 General Atomics Court 

San Diego, California 92121-1122 

General Atomics application dated 

January 6, 2009 

4. CONDITIONS 

 This certificate is conditional upon fulfilling the requirements of 10 CFR Part 71, as applicable, and the conditions specified below. 

5. 

a. Packaging 

 (1) Model No.: GA-4 

 (2) Description 

The GA-4 Legal Weight Truck Spent Fuel Shipping Cask consists of the packaging (cask and 

impact limiters) and the radioactive contents. The packaging is designed to transport up to 

four intact pressurized-water reactor (PWR) irradiated spent fuel assemblies as authorized 

contents. The packaging includes the cask assembly and two impact limiters, each of which 

is attached to the cask with eight bolts. The overall dimensions of the packaging are 

approximately 90 inches in diameter and 234 inches long. 

The containment system includes the cask body (cask body wall, flange, and bottom plate); 

cask closure; closure bolts; gas sample valve body; drain valve; and primary O-ring seals for 

the closure, gas sample valve, and drain valve. 

 Cask Assembly

The cask assembly includes the cask, the closure, and the closure bolts. Fuel spacers are 

also provided when shipping specified short fuel assemblies to limit the movement of the 

fuel. The cask is constructed of stainless steel, depleted uranium, and a hydrogenous 

neutron shield. The cask external dimensions are approximately 188 inches long and 40 

inches in diameter. A fixed fuel support structure divides the cask cavity into four spent fuel 

compartments, each approximately 8.8 inches square and 167 inches long. The closure is 

recessed into the cask body and is attached to the cask flange with 12 1-inch diameter bolts. 

The closure is approximately 26 inches square, 11 inches thick, and weighs about 1510 lbs. 

A-68
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES 1 a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 

9226 3 71-9226 USA/9226/B(U)F-85 2 OF 9 

5.a. (2) (continued) 

The cask has two ports allowing access to the cask cavity. The closure lid has an integral 

half-inch diameter port (hereafter referred to as the gas sample valve) for gas sampling, 

 venting, pressurizing, vacuum drying, leakage testing, or inerting. A 1-inch diameter port in 

the bottom plate allows draining, leakage testing, or filling the cavity with water. A separate 

drain valve opens and closes the port. The primary seals for the gas sample valve and drain 

valve are recessed from the outside cask surface as protection from punctures. The gas 

sample valve and the drain valve also have covers to protect them during transport. 

 Cask

The cask includes the containment (flange, cask body, bottom plate and drain valve seals); 

the cavity liner and fuel support structure; the impact limiter support structure; the trunnions 

and redundant lift sockets; the depleted uranium gamma shield; and the neutron shield and 

its outer shell. The cask body is square, with rounded corners and a transition to a round 

outer shell for the neutron shield. The cask has approximately a 1.5 inch thick stainless steel 

body wall, 2.6 inch thick depleted uranium shield (reduced at the corners), and 0.4 inch thick 

stainless steel fuel cavity liner. 

The cruciform fuel support structure consists of stainless steel panels with boron-carbide 

(B4C) pellets for criticality control. A continuous series of holes in each panel, at right angles 

with the fuel support structure axis, provides cavities for the B4C pellets. The fuel support 

structure is welded to the cavity liner and is approximately 18 inches square by 166 inches 

long and weighs about 750 lbs. 

The flange connects the cask body wall and fuel cavity liner at the top of the cask, and the 

bottom plate connects them at the bottom. The gamma shield is made up of five rings, which 

are assembled with zero axial tolerance clearance within the depleted uranium cavity, to 

minimize gaps. The impact limiter support structure is a slightly tapered 0.4 inch thick shell 

on each end of the cask. The shell mates with the impact limiter’s cavity and is connected to 

the cask body by 36 ribs. 

The neutron shield is located between the cask body and the outer shell. The neutron shield 

design maintains continuous shielding immediately adjacent to the cask body under normal 

conditions of transport. The details of the design are proprietary. The design, in conjunction 

with the operating procedures, ensures the availability of the neutron shield to perform its 

function under normal conditions of transport. 

Two lifting and tie-down trunnions are located about 34 inches from the top of the cask body, 

and another pair is located about the same distance from the bottom. The trunnion outside 

diameter is 10 inches, increasing to 11.5 inches at the cask interface. Two redundant lift 

sockets are located about 26 inches from the top of the cask body and are flush with the 

outer skin. 

A-69
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES 1 a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 

9226 3 71-9226 USA/9226/B(U)F-85 3 OF 9 

5.a. (2) (continued) 

Materials

All major cask components are stainless steel, except the neutron shield, the depleted 

uranium gamma shield, and the B4C pellets contained in the fuel support structure. All O-ring 

seals are fabricated of ethylene propylene. 

Impact Limiters

The impact limiters are fabricated of aluminum honeycomb, completely enclosed by an allwelded austenitic stainless steel skin. Each of the two identical impact limiters is attached to 

the cask with eight bolts. Each impact limiter weighs approximately 2,000 lbs. 

 (3) Drawings 

The packaging is constructed and assembled in accordance with the following GA Drawing 

Number: 

 Drawing No. 031348, 

 sheets 1 through 19, Revision D (Proprietary Version) 

 GA-4 Spent Fuel Shipping Cask Packaging Assembly 

5.(b) Contents 

 

 (1) Type and Form of Material: 

 (a) Intact fuel assemblies. Fuel with known or suspected cladding defects greater than hairline 

cracks or pinhole leaks is not authorized for shipment. 

 (b) The fuel authorized for shipment in the GA-4 package is irradiated 14x14 and 15x15 PWR fuel 

assemblies with uranium oxide fuel pellets. Before irradiation, the maximum enrichment of any 

assembly to be transported is 3.15 percent by weight of uranium-235 ( 235U). The total initial 

uranium content is not to exceed 407 Kg per assembly for 14x14 arrays and 469 Kg per 

assembly for 15x15 arrays. 

c) Fuel assemblies are authorized to be transported with or without control rods or other non-fuel 

assembly hardware (NFAH). Spacers shall be used for the specific fuel types, as shown on 

sheet 17 of the Drawings. 

(d) The maximum burnup for each fuel assembly is 35,000 MWd/MTU with a minimum cooling 

time of 10 years and a minimum enrichment of 3.0 percent by weight of 235U or 45,000 

MWd/MTU with a minimum cooling time of 15 years (no minimum enrichment). 

 

(e) The maximum assembly decay heat of an individual assembly is 0.617 kW. The maximum 

total allowable cask heat load is 2.468 kW (including control components and other NFAH 

when present). 

 

A-70
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES 1 a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 

9226 3 71-9226 USA/9226/B(U)F-85 4 OF 9 

5.b. (1) (continued) 

(f) The PWR fuel assembly types authorized for transport are listed in Table 1. All parameters are 

design nominal values. 

 (2) Maximum Quantity of Material per Package 

 (a) For material described in 5.b.(1): four (4) PWR fuel assemblies. 

 (b) For material described in 5.b.(1): the maximum assembly weight (including control 

components or other NFAH when present) is 1,662 lbs. The maximum weight of the cask 

contents (including control components or other NFAH when present) is 6,648 lbs., and the 

maximum gross weight of the package is 55,000 lbs. 

Table 1 - PWR Fuel Assembly Characteristics 

Fuel Type

Mfr.-Array 

(Versions) 

Design 

Initial U 

(kg/assy.) 

No. of 

Fuel 

Rods 

Fuel Rod 

Pitch (in.)

Pellet 

Diameter 

(in.) 

Zr Clad 

Thickness 

(in.) 

Active Fuel 

Length 

(in.) 

W-15x15 

(Std/ZC) 

469 204 0.563 0.3659 0.0242 144 

W-15x15 

(OFA) 

463 204 0.563 0.3659 0.0242 144 

BW-15x15 

(Mk.B,BZ,BGD) 

464 208 0.568 0.3686 0.0265 142 

Exx/A-15x15 

(WE) 

432 204 0.563 0.3565 0.030 144 

CE-15x15 

(Palisades) 

413 204 0.550 0.358 0.026 144 

CE-14x14 

(Ft.Calhoun) 

376 176 0.580 0.3765 0.028 128 

W-14x14 

(Model C) 

397 176 0.580 0.3805 0.026 137 

CE-14x14 

(Std/Gen.) 

386 176 0.580 0.3765 0.028 137 

Exx/A-14x14 

(CE) 

381 176 0.580 0.370 0.031 137 

A-71
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES 1 a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 

9226 3 71-9226 USA/9226/B(U)F-85 5 OF 9 

5.b(2)(b)(continued) 

Fuel Type

Mfr.-Array 

(Versions) 

Design 

Initial U 

(kg/assy.) 

No. of 

Fuel 

Rods 

Fuel Rod 

Pitch (in.)

Pellet 

Diameter 

(in.) 

Zr Clad 

Thickness 

(in.) 

Active Fuel 

Length 

(in.) 

W-14x14 

(OFA) 

358 179 0.556 0.3444 0.0243 144 

W-14x14 

(Std/ZCA,/ZCB) 

407 179 0.556 0.3674 0.0225 145.5 

Exx/A-14x14 

(WE) 

379 179 0.556 0.3505 0.030 142 

5.c. Criticality Safety Index (CSI): 100 

6. Fuel assemblies with missing fuel pins shall not be shipped unless dummy fuel pins that displace an 

equal amount of water have been installed in the fuel assembly. 

7. In addition to the requirements of Subpart G of 10 CFR 71: 

 a. Each package shall be both prepared for shipment and operated in accordance with detailed 

written operating procedures. Procedures for both preparation and operation shall be developed 

using the specifications contained within the application. At a minimum, those procedures shall 

require the following provisions: 

(1) Identification of the fuel to be loaded and independent verification that the fuel meets the 

specifications of Condition 5.b of the CoC. 

 (2) That before shipment the licensee shall: 

 (a) Perform a measured radiation survey to assure compliance with 49 CFR 173.441 

and 10 CFR 71.47 and assure that the neutron and gamma measurement 

instruments are calibrated for the energy spectrums being emitted from the 

package. 

 (b) Verify that measured dose rates meet the following correlation to demonstrate 

compliance with the design bases calculated hypothetical accident dose rates: 

 3.4 x (peak neutron dose rate at any point on cask surface at its midlength) + 

 1.0 x (gamma dose rate at that location) ≤ 1000 mR/hr. 

 (c) Verify that the surface removable contamination levels meet the requirements of 49 

CFR 173.443 and 10 CFR 71.87. 

 

A-72
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES 1 a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 

9226 3 71-9226 USA/9226/B(U)F-85 6 OF 9 

7.a.(2) (continued) 

 (d) Inspect all containment seals and closure sealing surfaces for damage. Leak test 

all containment seals with a gas pressure rise test after final closure of the package. 

The leak test shall have a test sensitivity of at least 1 x10-3 standard cubic 

centimeters per second of air (std-cm3

/sec) and there shall be no detectable 

pressure rise. A higher sensitivity acceptance and maintenance test may be 

required as discussed in Condition 7.b.(5), below. 

 

 (3) Before leak testing, the following closure bolt and valve torque specifications: 

 (a) The cask lid bolts shall be torqued to 235 ± 15 ft-lbs. 

(b) The gas sample valve and drain valve shall be torqued to 20 ± 2 ft-lbs. 

 (4) During wet loading operations and prior to leak testing, the removal of water and residual 

moisture from the containment vessel in accordance with the following specifications: 

(a) Cask evacuation to a pressure of 0.2 psia (10 mm Hg) or less for a minimum of 1 hour. 

(b) Verifying that the cask pressure rise is less than 0.1 psi in 10 minutes. 

 (5) Before shipment, independent verification of the material condition of the neutron shield as 

described in SAR Section 7.1.1.4 or 7.1.2.4. 

 b. All fabrication acceptance tests and maintenance shall be performed in accordance with detailed 

written procedures. Procedures for fabrication, acceptance testing, and maintenance shall be 

developed using the specifications contained within the application and shall include the 

following provisions: 

(1) All containment boundary welds, except the final fabrication weld joint connecting the cask body 

wall to the bottom plate, shall be radiographed and liquid-penetrant examined in accordance with 

ASME Code Section III, Division 1, Subsection NB. Examination of the final fabrication weld joint 

connecting the cask body wall to the bottom plate may be ultrasonic and progressive liquid 

penetrant examined in lieu of radiographic and liquid penetrant examination. 

(2) The upper lifting trunnions and redundant lifting sockets shall be load tested, in the cask axial 

direction, to 300 percent of their maximum working load (79,500 lbs. minimum) per trunnion and 

per lifting socket, in accordance with the requirements of ANSI N14.6. The upper and lower 

lifting trunnions shall be load tested, in the cask transverse direction, to 150 percent of their 

maximum working load (20,625 lbs. minimum) per trunnion, in accordance with the requirements 

of ANSI N14.6. 

A-73
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES 1 a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 

9226 3 71-9226 USA/9226/B(U)F-85 7 OF 9 

7.b.(continued) 

(3) The cask containment boundary shall be pressure tested to 1.5 times the Maximum Normal 

Operating Pressure of 80 psig. The minimum test pressure shall be 120 psig. 

(4) All containment seals shall be replaced within the 12-month period prior to each shipment. 

 (5) A fabrication leakage test shall be performed on all containment components including the O-ring 

seals prior to first use. Additionally, all containment seals shall be leak tested after the third use 

of each package and within the 12-month period prior to each shipment. Any replaced or 

repaired containment system component shall be leak tested. The leakage tests shall verify that 

the containment boundary leakage rate does not exceed the design leakage rate of 1 x10-7 stdcm3

/sec. The leak tests shall have a test sensitivity of at least 5 x 10-8 std-cm3/sec. 

(6) The depleted uranium shield shall be gamma scanned with 100 percent inspection coverage 

during fabrication to ensure that there are no shielding discontinuities. The neutron shield 

supplier shall certify that the shield material meets the minimum specified requirements 

(proprietary) used in the applicant’s shielding analysis. 

(7) Qualification and verification tests to demonstrate the crush strength of each aluminum 

honeycomb type and lot to be utilized in the impact limiters shall be performed. 

(8) The boron carbide pellets, fuel support structure and fuel cavity dimensions, and 235U content in 

the depleted uranium shall be fabricated and verified to be within the specifications of Table 2 to 

ensure criticality safety. 

A-74
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES 1 a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 

9226 3 71-9226 USA/9226/B(U)F-85 8 OF 9 

Table 2 

Specified Parameter Minimum Maximum 

B4C boron enrichment 96 wt% 10B N/A 

Diameter of each B4C pellet 0.426 in 0.430 in 

Height of each B4C pellet stack 7.986 in 8.046 in 

Mass of 10B in each B4C pellet stack 31.5 g N/A 

Mass of each B4C pellet stack 43.0 g 45.0 g 

Diameter of each fuel support 

structure hole 0.432 in 0.44 in 

Fuel support structure nominal hole pitch N/A 0.55 in 

Fuel support structure hole depth minus 

B4C pellet-stack height 

 (at room temperature) 

0.009 in 0.129 in 

Thickness of each fuel support 

structure panel 0.600 in 0.620 in 

Fuel cavity width N/A 9.135 in 

235U content in depleted uranium 

shielding material N/A 0.2 wt% 

8. Transport of fissile material by air is not authorized. 

9. The package authorized by this certificate is hereby approved for use under the general license 

provisions of 10 CFR 71.17. 

10. Expiration Date: October 31, 2013. 

A-75
NRC FORM 618

(8-2000) 

10 CFR 71 

U.S. NUCLEAR REGULATORY COMMISSION

CERTIFICATE OF COMPLIANCE

FOR RADIOACTIVE MATERIAL PACKAGES 1 a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 

9226 3 71-9226 USA/9226/B(U)F-85 9 OF 9 

REFERENCES

General Atomics Application for the GA-4 Legal Weight Truck Spent Fuel Shipping Cask, January 6, 2009. 

 FOR THE U.S. NUCLEAR REGULATORY COMMISSION 

/RA/ 

 Eric J. Benner, Chief 

 Licensing Branch 

 Division of Spent Fuel Storage and Transportation 

 Office of Nuclear Material Safety 

 and Safeguards 

Date February 5, 2009

 

A-76
APPENDIX B

DETAILS OF RISK ANALYSIS OF ROUTINE, 

INCIDENT-FREE TRANSPORTATION

B-1

B.1 Introduction

In NUREG-0170, “Final Environmental Statement on the Transportation of Radioactive Material 

by Air and Other Modes,” issued December 1977 (NRC, 1977), the U.S. Nuclear Regulatory 

Commission (NRC) documented estimates of the radiological consequences and risks 

associated with shipping by truck, train, plane, or barge. This report covered about 25 different 

radioactive materials, including power reactor spent fuel. These estimates were calculated 

using Version 1 of the RADTRAN code (Taylor and Daniel, 1977), which was developed for the 

NRC by Sandia National Laboratories specifically to support the NUREG-0170 study. In this 

new updated study, researchers used the computational tool RADTRAN Version 6.0, integrated 

with the input file generator RadCat (Neuhauser et al., 2000,

1 Weiner et al., 2009).

Researchers widely accept the basic risk-assessment method employed in the RADTRAN 

code.

2 A software quality assurance plan, consistent with American National Standards Institute 

guidelines, tracks changes to the code. The incident-free module of an earlier version of 

RADTRAN—RADTRAN 5.25—was validated by measurement (Steinman et al., 2002); 

RADTRAN 6.0, the version used in the current study, employs this same module. 

Dennis et al. (2008) documents the verification and validation of RADTRAN 6.0.

B.2 The RADTRAN Model of Routine Transportation

B.2.1 Description of the RADTRAN Program

RADTRAN calculates the radiological consequences and risks associated with the shipment of 

a specific radioactive material in a specific package along a specific route. Shipments that take 

place without the occurrence of accidents are routine, incident-free shipments, and the radiation 

doses to various receptors (exposed persons) are called “incident-free doses.” Because the 

probability of routine, incident-free shipment is essentially equal to one, RADTRAN calculates a 

dose rather than a risk for such shipments.3 The dose from a routine shipment is based on the 

external dose from the part of the vehicle carrying the radioactive cargo, referred to as the 

“vehicle” in this discussion of RADTRAN. Doses to receptors from the external radiation from 

the vehicle depend on the distance between the receptor and the radioactive cargo being 

transported and the exposure time. Exposure time is the length of time the receptor is exposed 

to external emissions from the radioactive cargo. The doses in routine transportation depend 

only on the external dose rate from the cargo and not on the radioactive inventory of the cargo.

 

1 Neuhauser et al., 2000, is the technical manual for RADTRAN 5, and is cited because the basic equations 

for the incident-free analyses in RADTRAN 6 are the same as those in RADTRAN 5 and the technical 

manual for RADTRAN 6 is not yet available.

2 RADTRAN was used to calculate risks for NUREG-0170 (NRC, 1977), the Yucca Mountain Final 

Environmental Impact Statement (U.S. Department of Energy, 2002), the recertification of the Waste 

Isolation Pilot Plant, and other studies. RADTRAN today has 600 registered users, about 25 percent of 

whom are U.S. Government contractors and about 25 percent of whom are international users. The list of 

users is proprietary.

3 The probability of a transportation incident or accident depends on the trip length and is about 10-3 for a 

cross-country trip. The probability of routine transportation on such a trip is 1 - 0.001, or 0.999, or essentially 

one. For a shorter trip, the probability of routine transportation is even closer to one.
B-2

RADTRAN models the vehicle as a spherical radiation source traveling along the route. The 

source strength is the transport index (TI), 100 times the dose rate in millisieverts per hour 

(mSv/h)

4 at 1 meter (40 inches) from the cask, which is treated as an isotropically radiating 

virtual source at the center of the sphere, as shown in Figure B-1 (see Neuhauser et al. (2000) 

for a detailed explanation).

Figure B-1 RADTRAN model of the vehicle in routine, incident-free transportation

When the distance to the receptor r is much larger than the critical dimension, RADTRAN 

models the dose to the receptor as proportional to 1/r2

. When the distance to the receptor r is 

similar to or less than the critical dimension, as for crew or first responders, RADTRAN models 

the dose to the receptor as proportional to 1/r. The TIs for the Rail-Lead and the Rail-Steel 

casks were calculated from the dose rates at 2 meters, as reported in the safety analysis reports 

of these casks (Holtec International, 2004, NAC international, 2004) and are shown in 

Table B-1.

Equation B-1 serves as the basic equation for calculating incident-free doses to a population 

along a transportation route:

(B-1) dxdr

r r x

r B r

V

Qk DR

D x

x

x

v

∫ ∫ 











−

− =

∞

−∞

max

min

2 2

0 (exp( ))( ( )) ( ) μ

 

4 One mSv = 100 millirem (mrem). Thus, 100 times the dose rate in mSv/h at 1 meter (40 inches) from the 

package is equivalent to the dose rate in mrem/h.
B-3

Where:

x is the distance between the receptor and the source, perpendicular to the route

Q includes factors that correct for unit differences

k0 is the package shape factor5

DRv is the vehicle external dose rate: the TI

V is the vehicle speed

μ is the radiation attenuation factor 

B is the radiation buildup factor

r is the distance between the receptor and the source along the route

Neuhauser et al. (2000) provides additional details of the application of this and similar 

equations.

External radiation from casks carrying used nuclear fuel includes both gamma and neutron 

radiation. For calculating doses from gamma radiation, RADTRAN uses Equation B-2 for 

conservatism.

(B-2) (e-μr) * B(r) = 1

For calculating doses from neutron radiation, on the other hand, RADTRAN uses Equation B-3, 

in which the coefficients are characteristics of the material.

(B-3) (e-μr) * B(r) = (e-μr) * (1 + a1r + a2r

2 + a3r3 + a4r4

)

Equation B-2 can be rewritten (Neuhauser et al., 2000) as Equation B-4.

(B-4) [ ] n n

v f I f I

V

Qk DR D(x) * * 0 = γ γ +

Where:

fγ is the gamma fraction of the external radiation

fn is the neutron fraction of the external radiation

Iγ is the double integral in Equation B-1 using Equation B-2

In is the double integral in Equation B-1 using Equation B-3

Collective (population) doses are calculated by integrating over the band along the route where 

the population resides (the x integration in Equation B-1) and then integrating along the route 

from minus to plus infinity (the r integration in Equation B-1). Figure B-2 illustrates this 

calculation method for a truck route. The x integration limits in Figure B-2 are not to scale: xmin

is usually 30 meters (98 feet) (200 meters (656 feet) near a rail classification stop) and xmax is 

usually 800 meters (1/2 mile). Integration of x to distances greater than 800 meters (1/2 mile) 

results in risks not significantly different from integration to 800 meters (1/2 mile), because the 

decrease in dose with distance is exponential.

 

5 For details on the package shape factor, please see Equations B-4 and B-5 and accompanying text of 

Neuhauser et al. (2000).
B-4

Figure B-2 Diagram of a truck route as modeled in RADTRAN (from U.S. Department of 

Energy, 2002)

The 845-km value is the average distance a very large truck travels on half of its fuel capacity. 

The 161-km (100-mile) value is the distance between spent fuel shipment inspections required 

by regulation.

Variants of Equation B-1 are used to calculate doses to members of the public (at stops), 

vehicle crew members and other workers, occupants of vehicles that share the route with the 

vehicle carrying the radioactive cargo, and any other receptor identified. Figure B-3 is a 

diagram of the model used to calculate doses at truck stops. The inner circle defines the area 

occupied by people who are between the spent fuel truck and the building and who are not 

shielded from the truck’s external radiation. Griego et al. (1996) provides the dimensions of this 

circle and the average number of people who occupy it, along with the method used to 

determine these values.
B-5

Figure B-3 Diagram of the truck-stop model

B.2.2 The RADTRAN Software

This section briefly describes the RADTRAN software program. The RadCat 3.0 User Guide 

(Weiner et al., 2009) provides a full description of the software and how to use it. The equations 

that RADTRAN uses, variants of Equation B-1, are programmed in FORTRAN 95. RADTRAN 

uses the following information:

• an input text file that contains the input parameters, as defined by the RADTRAN user 

• a text file that contains an internal library of 148 radionuclides, with their associated dose 

conversion factors and half-lives

• a binary file that contains the societal ingestion doses for one curie of each radionuclide 

in the internal radionuclide library

• dilution factors and isopleth areas for several weather patterns

Only the first of these is used in calculating doses from incident-free transportation; the other 

three are used in accident analysis and will be discussed in Appendix E.

The input text file can be written directly using a text editor or can be constructed using the input 

file generator RadCat. RadCat, programmed in XML and running under Java Web Start, 

provides a series of screens that guide the user in entering values for RADTRAN input 

parameters. Figure B-4 shows a RadCat screen.
B-6

Figure B-4 RadCat vehicle screen

RADTRAN output is a text file that can be saved as text or as a spreadsheet.

B.3 RADTRAN Input Parameters

B.3.1 Vehicle-Specific Input Parameters

RADTRAN does not allow for the offset of the package from the trailer edge, so the physical 

dimensions of the package are considered the physical dimensions of the vehicle. Table B-1

shows the vehicle-specific input parameters to RADTRAN and includes the parameter values 

used in this analysis. The Rail-Steel model is calculated as if transporting canistered fuel; the 

Rail-Lead model is based on transporting uncanistered fuel. The analysis includes a third

model—the Truck-DU model—a truck cask with depleted uranium (DU) gamma shielding. 

While the Truck-DU is a truck cask, the other two are rail casks. This analysis assumes that the 

Truck-DU cask is transported by truck and the Rail-Lead and Rail-Steel casks are transported

by rail.
B-7

Table B-1 Vehicle-Specific Parameters

Truck-DU Rail-Lead Rail-Steel

Transportation mode Highway Rail Rail

Length (critical dimension) 5.94 m 4.90 m 5.08 m

Diameter (“crew view”) 2.29 m 2.5 m 3.2 m

Distance from cargo to crew cab 3.5 m 150 m minimum 150 m minimum

TI 14 14.02 10.34

Gamma fraction 0.77 0.89 0.90

Neutron fraction 0.23 0.11 0.10

Number of packages per vehicle 1 per truck 1 per railcar 1 per railcar

Number of crew 2 3 3

Exclusive use? yes N/A N/A

Dedicated rail N/A yes yes

17 × 17 PWR assemblies 4 26 24

B.3.2 Route-Specific Input Parameters

Table B-2 shows the route parameters for a unit risk calculation. These route parameters are 

the common input parameters for the 16 specific rail routes and 16 specific truck routes 

analyzed.
B-8

Table B-2 Route Parameters for Unit Risk Calculation

(from U.S. Department of Transportation, 2004, 2006)

Parameter Interstate Highway Freight Rail

Rural vehicle speed (U.S. average kph) 108 40.4

Suburban vehicle speed (U.S. average kph) 102 40.4

Urban vehicle speed (U.S. average kph) 97 24

Rural vehicle density (U.S. average vehicles/h) 1,119 17a

Suburban vehicle density (U.S. average vehicles/h) 2,464 17

Urban vehicle density (U.S. average vehicles/h) 5,384 17

Persons per vehicle 1.5 2

Farm fraction 0.5 0.5

Minimum distance of stop from nearby residents (m) 30 200

Maximum distance of stop from nearby residents (m) 800 800

Stop time for classification (hours) N/A 27

Stop time in transit for railroad change (hours) N/A variable

Stop time for truck inspections (hours) 0.75 N/A

Stop time at truck stops (hours) 0.83 N/A

Average number of people sharing the stop 6.9b N/A

Minimum distance to people sharing the stop (m) 1b N/A

Maximum distance to people sharing the stop (m) 15b N/A

Truck stop worker distance from cask (m) 15 N/A

Truck stop worker shielding factor 0.018 N/A

Truck crew shielding factor 0.377 N/A

Escort distance from cask (m) 4 16 a Railcars per hour b Griego et al., 1996

B.3.3 Other Parameters

RADTRAN includes a set of parameters whose values are not generally known by the user and 

which have been used routinely in transportation risk assessments. RADTRAN contains default 

values for these parameters, but all default values can be changed by the user. Table B-3 lists 

the parameter values used in the incident-free analysis.

Additional input parameters are rural, suburban, and urban route lengths and population 

densities; characteristics of stops along a route; and the TI of the package.
B-9

Table B-3 Parameter Values in the RADTRAN 6 Analysis

Parameter Value

Shielding factor for residents (fraction of energy impacting the receptor):

R = rural, S = suburban, U = urban

R = 1.0

S = 0.87

U = 0.018

Fraction of outside air in urban buildings 0.25

Fraction of urban population on sidewalk 0.48

Fraction of urban population in buildings 0.52

Ratio of sidewalk occupants to residents (applies only to secondary roads in urban areas; 

does not apply to rail) 6

Distance from in-transit shipment for maximum exposure (m) 

(maximally exposed individual (MEI) exposure) 30

Vehicle speed for maximum exposure (km/hr) (MEI exposure) 24

Distance from in-transit shipment to nearest resident in rural and suburban areas (m) 30

Distance from in-transit shipment to nearest resident in urban areas (m) 27

Width of the population band on each side of the route (m) 800

Distance between vehicles or trains (m) 3.0

Minimum number of rail classification stops 1

B.3.4 RADTRAN Input and Output Files

Figure B-5 shows the incident-free unit risk input file for the Truck-DU cask. Figure B-6 shows

the incident-free unit risk input file for the Rail-Lead and Rail-Steel casks. In the interests of 

space, only the portion of the input files relevant to routine incident-free transportation are 

shown.
B-10

OUTPUT CI_REM

FORM UNIT

DIMEN 1 0 18

PARM 0 1 3 0

PACKAGE GA4 14.0 0.77 0.22999999999999998 5.94

VEHICLE -1 GA_4 1.400000E01 0.77 0.23 5.94 1.0 2.0 3.5 0.38 2.29

GA4 1.0

FLAGS

IACC 2

IUOPT 2

REGCHECK 0

MODSTD

DISTOFF FREEWAY 3.000000E01 3.000000E01 8.000000E02

DISTOFF SECONDARY 2.700000E01 3.000000E01 8.000000E02

DISTOFF STREET 5.000000E00 8.000000E00 8.000000E02

DISTON

FREEWAY 1.500000E01

SECONDARY 3.000000E00

STREET 3.000000E00

ADJACENT 4.000000E00

MITDDIST 3.000000E01

MITDVEL 2.400000E01

RR 1.000000E00

RU 1.800000E-02

RS 8.700000E-01

LINK R GA_4 1.0 108.0 1.5 1.0 1119.0 1.0 0.0 R 1 0.5

LINK S GA_4 1.0 108.0 1.5 1.0 2464.0 1.0 0.0 S 1 0.0

LINK U GA_4 0.9 102.0 1.5 1.0 5384.0 1.0 0.0 U 1 0.0

LINK U_RUSH GA_4 0.1 51.0 1.5 1.0 10760.0 1.0 0.0 U 1 0.0

STOP STOP_1 GA_4 9180.0 1.0 15.0 1.0 0.83

STOP RURAL GA_4 1.0 30.0 800.0 1.0 0.83

STOP SUBURBAN GA_4 1.0 30.0 800.0 0.87 0.83

STOP URBAN GA_4 1.0 30.0 800.0 0.018 0.83

EOF

Figure B-5 RADTRAN unit risk input file for the Truck-DU cask
B-11

RADTRAN 6 July 2008

&& SEE APPENDIX A.2 FOR DETAILS

&& UNIT RISK FACTOR

&& INCIDENT-FREE TRANSPORT

&& AVERGE TI

&& PWR 35GHWD/MTHM BURNUP 10 YEAR COOLED 24 ASSEMBLIES

&& REMARK

TITLE NAC-STC 

OUTPUT CI_REM

FORM UNIT

DIMEN 1 0 18

PARM 0 1 3 0

PACKAGE NAC-STC 14.02 0.89 0.10999999999999999 4.9

PACKAGE HI-STAR_100 10.034 0.9 0.09999999999999998 5.08

VEHICLE -2 NAC-STC 1.400000E01 0.89 0.11 4.9 1.0 3.0 150.0 1.0 2.5

 NAC-STC 1.0

 HI-STAR 0.0

VEHICLE -2 HI-STAR 1.030000E01 0.9 0.1 5.08 1.0 3.0 150.0 1.0 3.2

 NAC-STC 0.0

 HI-STAR 1.0

FLAGS

 IACC 2

 ITRAIN 2

 IUOPT 2

 REGCHECK 0

MODSTD

 DDRWEF 1.800000E-03

 FMINCL 1.000000E00

 DISTOFF RAIL 3.000000E01 3.000000E01 8.000000E02

 DISTON

RAIL 3.000000E00

 MITDDIST 3.000000E01

 MITDVEL 2.400000E01

 RPD 6.000000E00

 RR 1.000000E00

 RU 1.800000E-02

 RS 8.700000E-01

LINK NAC_R NAC-STC 1.0 40.4 3.0 1.0 8.0 1.0 0.0 R 3 0.5

LINK NAC_S NAC-STC 1.0 40.4 3.0 1.0 8.0 1.0 0.0 S 3 0.0

LINK NAC_U NAC-STC 1.0 24.0 3.0 1.0 17.0 1.0 0.0 U 3 0.0

LINK HISTAR_R HI-STAR 1.0 40.4 3.0 1.0 8.0 1.0 0.0 R 3 0.5

LINK HISTAR_S HI-STAR 1.0 40.4 3.0 1.0 8.0 1.0 0.0 S 3 0.0

LINK HISTAR_U HI-STAR 1.0 24.0 3.0 1.0 17.0 1.0 0.0 U 3 0.0

STOP CLASSIFICATION NAC-STC 1.0 200.0 800.0 1.0 27.0

STOP CLASSIFICATION HI-STAR 1.0 200.0 800.0 1.0 27.0

EOF

Figure B-6 RADTRAN unit risk input file for the Rail-Lead and Rail-Steel casks
B-12

B.4 Routes

This study analyzes both the per-kilometer doses from a single shipment on rural, suburban, 

and urban route segments and doses to receptors from a single shipment between 

16 representative pairs of origins and destinations, chosen to represent a range of route lengths 

and a variety of populations. The actual truck and rail routes were selected for a number of 

reasons. The combination of four origins and four destinations represent a variety of route 

lengths and population densities and both private and government facilities, as well as a large 

number of States. The selected routes also include the origins and destinations analyzed in 

NUREG/CR-6672, “Reexamination of Spent Fuel Shipment Risk Estimates,” issued March 2000

(Vol. 1 available as Agencywide Documents Access and Management System (ADAMS) 

Accession No. ML003698324), thereby permitting the results of the studies to be compared.

Spent nuclear fuel and high-level radioactive waste from power reactors are currently stored at 

77 locations in the United States (67 nuclear generating plants, five storage facilities at sites of 

decommissioned nuclear plants, and five U.S. Department of Energy defense facilities). The 

origin sites (Table B-4) include two nuclear generating plants (Indian Point and Kewaunee), a 

storage site (Maine Yankee), and a national laboratory (Idaho National Laboratory). The 

destination sites include the two proposed repository sites not characterized (Deaf Smith 

County, TX, and Hanford, WA) (U.S. Department of Energy, 1986), the site of the proposed 

private fuel storage facility (Skull Valley, UT), and a national laboratory site (Oak Ridge, TN). 

Table B-4 shows the routes modeled. The populations within 800 meters (1/2 mile) of the route 

were determined from output of the WebTRAGIS (Johnson and Michelhaugh, 2003) routing 

code, modified by the increase in population between 2000 and 2006 (see Table B-5). Both 

truck and rail versions of each route are analyzed. These routes are used for illustrative 

purposes only. No actual spent fuel shipments on these routes are occurring or planned.
B-13

Table B-4 Specific Routes Modeled

(Urban Kilometers Are Included in Total Kilometers)

Origin Destination

Population within 

800 m (1/2 mile)

Total 

Kilometers

Urban 

Kilometers

Rail Truck Rail Truck Rail Truck

Maine 

Yankee 

Site, ME

Hanford, WA 1,647,190 1,129,685 5,084 5,013 355 116

Deaf Smith County, TX 1,321,024 1,427,973 3,362 3,596 211 165

Skull Valley, UT 1,451,325 1,068,032 4,068 4,174 207 115

Oak Ridge, TN 1,146,478 1,137,834 2,125 1,748 161 135

Kewaunee 

Nuclear 

Plant 

(NP), WI

Hanford, WA 476,914 423,163 3,028 3,453 60 52

Deaf Smith County, TX 677,072 494,920 1,882 2,146 110 60

Skull Valley, UT 806,115 505,226 2,755 2,620 126 58

Oak Ridge, TN 779,613 646,034 1,395 1,273 126 92

Indian 

Point

NP, NY

Hanford, WA 961,026 869,763 4,781 4,515 229 97

Deaf Smith County, TX 1,027,974 968,282 3,088 3,074 204 109

Skull Valley, UT 1,517,758 808,107 3,977 3,672 229 97

Oak Ridge, TN 1,146,245 561,723 1,264 1,254 207 60

Idaho 

National 

Lab, ID

Hanford, WA 164,399 132,662 1,062 959 20 15

Deaf Smith County, TX 298,590 384,912 1,913 2,291 40 52

Skull Valley, UT 169,707 132,939 455 466 26 19

Oak Ridge, TN 593,680 569,240 3,306 3,287 75 63

WebTRAGIS, which uses census data from the 2000 census, provided the route segments and 

population densities. The 2008 Statistical Abstract (U.S. Census Bureau, 2008) provided 

updated population data through 2006. Table 13 of the 2008 Statistical Abstract shows the 

percent increase in population for each of the 50 United States, as well as for the United States 

as a whole. Table 20 of the Abstract shows the percent increase in population for all 

metropolitan areas within the United States with more than 250,000 people. Data from these 

two tables were combined to give population multipliers for States along routes for which the 

collective dose and the population increase were significant enough to make a correction.

Table B-5 shows the population multipliers used. “Significant” was defined as a population 

difference of more than 1 percent (i.e., multipliers between 0.99 and 1.01 were not considered 

significant). The State-specific multiplier was applied to rural and suburban routes through the 

State (even though some of these routes would be within the largest metropolitan area), and the 

multiplier for the largest metropolitan area in that State was applied to the urban route segments 

(even though some of the urban segments may not be within the largest metropolitan area). For 

States without a metropolitan area with more than 250,000 people (Delaware, Montana, 

North Dakota, South Dakota, Vermont, and Wyoming), the Statewide increase was used. It was 

anticipated that 2010 census data would be available for this final report, but at the time of this 

writing it still was not possible to obtain updated population data in WebTRAGIS. An estimation 

of the error introduced by not updating to the 2010 census can be made by using the same 

method that was used to adjust the populations to 2006 data. The correction factors for each 

state are given in Table B-6.
B-14

Table B-5 Population Multipliers

State

Rural, 

Suburban, 

Urban 

Designation

Population 

Multiplier State

Rural, 

Suburban, 

Urban 

Designation

Population 

Multiplier

Arkansas Rural, Suburban 1.051 New 

Hampshire

Rural, Suburban 1.064

Urban 1.069 Urban 1.058

Colorado Rural, Suburban 1.105New Jersey Rural, Suburban 1.037

Urban 1.105 Urban 1.027

Connecticut Rural, Suburban 1.029New Mexico Rural, Suburban 1.075

Urban 1.020 Urban 1.119

Delaware Rural, Suburban 1.089New York Rural, Suburban 1.017

Urban 1.089 Urban 1.027

District of 

Columbia

Rural, Suburban 1.017

Ohio Rural, Suburban 1.011

Urbana 1.017 Urban 0.984

Idaho Rural, Suburban 1.133Oklahoma Rural, Suburban 1.037

Urban 1.221 Urban 1.070

Illinois Rural, Suburban 1.033Oregon Rural, Suburban 1.082

Urban 1.045 Urban 1.109

Indiana Rural, Suburban 1.038Pennsylvania Rural, Suburban 1.013

Urban 1.092 Urban 1.025

Iowa Rural, Suburban 1.019South Dakota Rural, Suburban 1.036

Urban 1.110 Urban 1.036

Kansas Rural, Suburban 1.028Tennessee Rural, Suburban 1.061

Urban 1.037 Urban 1.109

Kentucky Rural, Suburban 1.041Texas Rural, Suburban 1.127

Urban 1.051 Urban 1.163

Maine Rural, Suburban 1.037Utah Rural, Suburban 1.142

Urban 1.054 Urban 1.102

Maryland Rural, Suburban 1.060Vermont Rural, Suburban 1.025

Urban 1.103 Urban 1.025

Massachusetts Rural, Suburban 1.014Virginia Rural, Suburban 1.080

Urban 1.014 Urban 1.103

Minnesota Rural, Suburban 1.050Washington Rural, Suburban 1.085

Urban 1.069 Urban 1.072

Missouri Rural, Suburban 1.044Wisconsin Rural, Suburban 1.036

Urban 1.036 Urban 1.006

Montana Rural, Suburban 1.047Wyoming Rural, Suburban 1.043

Urban 1.047 Urban 1.043

Nebraska Rural, Suburban 1.033

Urban 1.072

a For urban areas within the District of Columbia, the growth rate of the District was used rather than the 

growth rate of metropolitan Washington. The growth rate of metropolitan Washington was used for urban 

areas within Maryland and Virginia.
B-15

Table B-6 Population Corrections for 2010 Census Data

State Rural/Suburban 

Correction

Urban 

Correction

Arkansas 1.038 1.072

Colorado 1.058 1.056

Connecticut 1.019 1.019

Delaware 1.052 0.895

District of Columbia 1.034 1.055

Idaho 1.069 1.086

Illinois 1.000 0.995

Indiana 1.027 1.055

Iowa 1.022 1.066

Kansas 1.032 1.052

Kentucky 1.032 1.051

Maine 1.005 1.000

Maryland 1.028 1.055

Massachusetts 1.017 1.023

Minnesota 1.027 1.034

Missouri 1.025 1.006

Montana 1.048 1.107

Nebraska 1.033 1.052

New Hampshire 1.001 0.994

New Jersey 1.008 1.004

New Mexico 1.053 1.087

New York 1.004 1.004

North Dakota 1.058 1.177

Ohio 1.005 0.983

Oklahoma 1.048 1.069

Oregon 1.035 1.041

Pennsylvania 1.021 1.023

South Dakota 1.042 1.198

Tennessee 1.051 1.093

Texas 1.070 1.061

Utah 1.084 1.053

Vermont 1.003 1.064

Virginia 1.046 1.055

Washington 1.052 1.054

West Virginia 1.019 0.996

Wisconsin 1.023 1.031

Wyoming 1.094 1.076

Parameters such as population density and route segment lengths, which are specific to each 

route, were developed using WebTRAGIS. Figures B-7 through B-10 are WebTRAGIS maps 

showing the routes.
B-16

Figure B-7 Highway and rail routes from the Maine Yankee Nuclear Plant (NP) site
B-17

Figure B-8 Highway and rail routes from the Kewaunee NP site
B-18

Figure B-9 Highway and rail routes from Indian Point NP site
B-19

Figure B-10 Highway and rail routes from Idaho National Laboratory (INL)

B.5 Results

B.5.1 Maximally Exposed Resident In-Transit Dose

The largest dose from a moving vehicle to an individual member of the public is sustained when 

that individual is 30 meters (a conservative estimate of the Interstate right-of-way) from the 

moving vehicle and the vehicle is moving at the slowest speed it would be likely to maintain. 

This speed is 24 kilometers per hour (kph) (16 miles per hour (mph)) for both rail and truck. 

Table B-7 shows the maximum individual dose, in Sv, for each package. These doses are 

directly proportional to the external dose rate (TI) of each package. For comparison, a single 

dental x-ray delivers a dose of 4×10-5 Sv (Stabin, 2009), about 7,000 times the doses shown in 

Table B-7.

Table B-7 Maximum Individual Doses

Package (mode) Dose in Sv

Truck-DU (truck) 6.7 × 10-9

Rail-Lead (rail) 5.7 × 10-9

Rail-Steel (rail) 4.3 × 10-9
B-20

Figure B-11 and Figure B-12 show the portion of the RADTRAN output file that reflects these 

doses.

 RUN DATE: [ 03-02-2010 AT 18:07 ] PAGE 11

 TRUCK URF -- PUBLIC

 MAXIMUM INDIVIDUAL IN-TRANSIT DOSE

 GA_4 6.70 x 10-7 REM

Figure B-11 RADTRAN output for maximum individual truck doses

 RUN DATE: [ 02-19-2010 AT 10:55 ] PAGE 10

 NAC-STC

 MAXIMUM INDIVIDUAL IN-TRANSIT DOSE

 NAC-STC 5.67 x 10-7 REM

 HI-STAR 4.30 x 10-7 REM

Figure B-12 RADTRAN output for maximum individual rail doses

B.5.2 Unit Risk—Rail Routes

Table B-8 shows the doses to railyard workers along the route, to residents and others along 

the route, and to occupants of vehicles that share the route from a single shipment (one rail 

cask) traveling 1 kilometer past a population density of one person per square kilometer (km2

). 

The dose units are person-Sv. The doses are calculated assuming one cask on a train because 

railcar-km is the unit usually used to describe freight rail transport. The data in this table may be 

used to calculate collective doses along routes as follows:

• This is a conservatively calculated dose that assumes that the railyard crew receives a 

fraction of the classification yard dose when the train stops. The railyard crew dose is 

multiplied by the length of each type of route traveled. The classification yard 

occupational collective dose (Wooden, 1986), assuming a 30-hour classification stop, is 

integrated into RADTRAN. This integrated dose was adjusted to reflect the 27-hour stop 

(Table B-3) (U.S. Department of Transportation, 2006).

• The area of the band occupied by the population along the route is equal to the product 

of the kilometers traveled and the band width (usually 800 meters (1/2 mile) on each 

side of the route). RADTRAN calculates the doses to residents along the route by 

integrating over this area. This “unit dose to a resident along the route” is multiplied by 

the area of the band and the appropriate population density (obtained from a routing 

code like WebTRAGIS).
B-21

Table B-8 Individual Doses (“Unit Doses” or “Unit Risks”) to Various Receptors for Rail 

Routesa

Cask and route type Resident along 

route, Sv-kmb

Resident near 

railyard Sv-km2/hc

Occupants of 

vehicles sharing the 

route,

d person-Sv/kmc

Rail-Lead rural 7.3 × 10-10 3.5 × 10-7 1.6 × 10-8

Rail-Lead suburban 6.3 × 10-10 3.5 × 10-7 1.6 × 10-8

Rail-Lead urban 2.2 × 10-11 3.5 × 10-7 4.6 × 10-8

Rail-Steel rural 5.6 × 10-10 2.7 × 10-7 1.2 × 10-8

Rail-Steel suburban 4.8 × 10-10 2.7 × 10-7 1.2 × 10-8

Rail-Steel urban 1.7 × 10-11 2.7 × 10-7 3.5 × 10-8

a The units of the dose to the residents near a railyard where a train has stopped, Sv-km2

/h, reflect the 

output of the RADTRAN stop model, which incorporates the area occupied. b To obtain the collective dose to residents along the route, multiply this number by the route length and the 

population density. c To obtain the collective dose to residents near a railyard, multiply this number by the population density 

and the stop duration. d To obtain the collective dose to occupants of vehicles sharing the route, multiply this number by the route 

length (the vehicle density and occupants/vehicle used are the national average).

Figure B- 13 shows the RADTRAN output for Table B-8 (in rem). The relevant data in the 

output are in bold print.

 IN-TRANSIT POPULATION EXPOSURE IN PERSON-REM

 LINK CREW OFF LINK ON LINK

 NAC_R 4.32 x 10-5

 7.29 x 10-8 1.63 x 10-6

 NAC_S 4.32 x 10-5

 6.34 x 10-8 1.63 x 10-6

 NAC_U 7.28 x 10-5

 2.21 x 10-9 4.63 x 10-6

 HISTAR_R 3.27 x 10-5

 5.55 x 10-8 1.24 x 10-6

 HISTAR_S 3.27 x 10-5

 4.83 x 10-8 1.24 x 10-6

 HISTAR_U 5.51 x 10-5

 1.68 x 10-9 3.50 x 10-6

 STOP EXPOSURE IN PERSON-REM

 ANNULAR AREA CLASSIFICA 3.51 x 10-5

 ANNULAR AREA CLASSIFICA 2.66 x 10-5

Figure B-13 RADTRAN output for Table B-8

B.5.3 Unit Risk—Truck Routes

Table B-9 shows the doses to truck crew, residents and others along the route and to occupants 

of vehicles that share the route from a single shipment (one truck cask) traveling 1 kilometer 

past a population density of one person/km2

. The dose units are person-Sv. Rural, suburban, 

and urban doses to residents living near stops are calculated by multiplying the appropriate stop 

dose (truck stops are not typically located in urban areas) by the appropriate population density 

(obtained from a routing code like WebTRAGIS). The number of stops on each route segment 

is calculated by dividing the length of the route segment by 845 kilometers (average distance 
B-22

between refueling stops for a large semidetached trailer truck (U.S. Department of 

Energy, 2002, Appendix J). The area of the band occupied by the population along the route is 

equal to the kilometers traveled multiplied by, for example, 1.6 for a band 800 meters wide on 

each side of the route.
B-23

Table B-9 Unit Risk Values for Various Receptors along Truck Routes

Resident 

along 

route,

a

Sv-km

Resident 

near 

stops,

b

Sv-km2/h

Occupants of vehicles sharing route,

c person-Sv/km

EPA Regions (see Table B-19)

1 2 3 4 5 6 7 8 9 10

Truck-DU 

rural 3.1 × 10-10 3.26 × 10-8 4.8 × 10-8 1.1 × 10-7 2.3 × 10-7 1.6 × 10-7 1.3 × 10-7 9.9 × 10-8 1.0 × 10-7 8.8 × 10-8 1.6 × 10-7 1.2 × 10-7

Truck-DU 

suburban 2.7 × 10-10 2.84 × 10-8 8.0 × 10-8 2.3 × 10-7 4.0 × 10-7 3.1 × 10-7 2.7 × 10-7 1.7 × 10-7 1.8 × 10-7 2.2 × 10-7 4.1 × 10-7 2.9 × 10-7

Truck-DU 

urban 5.2 × 10-12 2.4 × 10-7 4.6 × 10-7 6.4 × 10-7 6.2 × 10-7 4.9 × 10-7 3.3 × 10-7 2.7 × 10-7 4.1 × 10-7 8.4 × 10-7 6.3 × 10-7

Truck-DU 

urban 

rush 

hourd

1.2 × 10-12 2.2 × 10-7 4.3 × 10-7 5.9 × 10-7 5.8 × 10-7 4.5 × 10-7 3.1 × 10-7 2.5 × 10-7 3.8 × 10-7 7.7 × 10-7 5.8 × 10-7

6.9 people 

sharing 

stop 

(personSv) e

2.3 × 10-4

a To obtain the collective dose to residents along the route, multiply this number by the route length and the population density. b To obtain the collective dose to residents near a truck stop, multiply this number by the population density and the stop duration. c To obtain the collective dose to occupants of vehicles sharing the route, multiply this number by the route length (the vehicle 

density and occupants/vehicle used are the regional average). d One-tenth of the urban route segment is considered “rush-hour km,” which is equivalent to the truck spending 10 percent of the 

urban transit distance during rush hour. RADTRAN has historically assumed that the vehicle density doubles during rush hour 

and the vehicle speed is halved. The slower vehicle speed impacts the dose to urban residents along the route, but the vehicle 

density does not. Both factors influence the dose to occupants of vehicles sharing the route. The unit risk factors in the table 

incorporate 9/10 of the distance not during rush hour and 1/10 of the distance during rush hour, so both factors should be 

multiplied by the total urban distance. e Average number of people outside at truck stops, from Griego et al., 1996
B-24

B.5.4 Doses along Selected Routes

Doses to receptors along the routes shown in Table B-4 are presented below.

B.5.4.1 Collective Doses to Receptors along the Route

Using route data from WebTRAGIS, researchers calculated collective doses from incident-free 

transportation. For all route segments, collective doses were calculated for the resident 

population. Tables B-10 to B-13 show collective doses along rail routes; Tables B-14 to B-17

show collective doses along highway routes. Blank cells in the tables indicate that no route 

miles for that population type were present along that route (e.g., not all routes transit urban 

areas in all States).

Note: In each of these tables, one value is bolded. This value is used in a sample calculation 

that follows the table.
B-25

Table B-10 Collective Doses to Residents along the Route (Person-Sv) from Rail 

Transportation; Shipment Origin—INL

DEST. AND 

ROUTE

Rail-Lead Rail-Steel

Rural Suburban Urban Total Rural Suburban Urban Total

ORNL

Colorado 1.3 × 10-7 5.8 × 10-7 0 7.1 × 10-7 1.0 × 10-7 4.4 × 10-7 0 5.4 × 10-7

Idaho 1.7 × 10-6 7.6 × 10-6 3.4 × 10-7 9.7 × 10-6 1.3 × 10-6 5.8 × 10-6 2.6 × 10-7 7.4 × 10-6

Illinois 1.8 × 10-6 1.7 × 10-5 4.5 × 10-7 1.9 × 10-5 1.3 × 10-6 1.3 × 10-5 3.4 × 10-7 1.4 × 10-5

Indiana 1.7 × 10-6 8.2 × 10-6 1.8 × 10-7 1.0 × 10-5 1.3 × 10-6 6.3 × 10-6 1.4 × 10-7 7.7 × 10-6

Kansas 1.3 × 10-6 6.6 × 10-6 1.6 × 10-7 8.1 × 10-6 9.7 × 10-7 5.0 × 10-6 1.2 × 10-7 6.1 × 10-6

Kentucky 2.6 × 10-6 2.1 × 10-5 8.6 × 10-7 2.5 × 10-5 2.0 × 10-6 1.6 × 10-5 6.6 × 10-7 1.9 × 10-5

Missouri 2.4 × 10-6 2.2 × 10-5 1.1 × 10-6 2.6 × 10-5 1.8 × 10-6 1.7 × 10-5 8.7 × 10-7 2.0 × 10-5

Nebraska 3.5 × 10-6 1.2 × 10-5 3.5 × 10-7 1.6 × 10-5 2.7 × 10-6 9.5 × 10-6 2.6 × 10-7 1.2 × 10-5

Tennessee 1.2 × 10-6 7.8 × 10-6 4.2 × 10-8 9.1 × 10-6 9.4 × 10-7 6.0 × 10-6 3.2 × 10-8 6.9 × 10-6

Wyoming 1.4 × 10-6 8.8 × 10-6 2.1 × 10-7 1.0 × 10-5 1.1 × 10-6 6.7 × 10-6 1.6 × 10-7 7.9 × 10-6

DEAF SMITH

Colorado 3.3 × 10-6 4.1 × 10-5 1.7 × 10-6 4.6 × 10-5 2.5 × 10-6 3.2 × 10-5 1.3 × 10-6 3.5 × 10-5

Idaho 1.7 × 10-6 7.6 × 10-6 3.4 × 10-7 9.7 × 10-6 1.3 × 10-6 5.8 × 10-6 2.6 × 10-7 7.4 × 10-6

Oklahoma 1.1 × 10-7 1.8 × 10-7 0 2.9 × 10-7 8.3 × 10-8 1.4 × 10-7 0 2.2 × 10-7

Texas 4.1 × 10-7 3.4 × 10-6 5.9 × 10-8 3.8 × 10-6 3.1 × 10-7 2.6 × 10-6 4.4 × 10-8 2.9 × 10-6

Wyoming 1.1 × 10-6 6.0 × 10-6 1.5 × 10-7 7.3 × 10-6 8.5 × 10-7 4.6 × 10-6 1.2 × 10-7 5.6 × 10-6

HANFORD

Idaho 3.7 × 10-6 1.6 × 10-5 6.0 × 10-7 2.0 × 10-5 2.8 × 10-6 1.2 × 10-5 4.6 × 10-7 1.5 × 10-5

Oregon 1.4 × 10-6 9.2 × 10-6 2.2 × 10-7 1.1 × 10-5 1.1 × 10-6 7.0 × 10-6 1.7 × 10-7 8.3 × 10-6

Washington 1.2 × 10-7 4.4 × 10-6 2.6 × 10-7 4.7 × 10-6 8.9 × 10-8 3.3 × 10-6 2.0 × 10-7 3.6 × 10-6

SKULL VALLEY

Idaho 1.4 × 10-6 6.6 × 10-6 3.3 × 10-7 8.3 × 10-6 1.1 × 10-6 5.0 × 10-6 2.5 × 10-7 6.3 × 10-6

Utah 1.6 × 10-6 1.9 × 10-5 1.1 × 10-6 2.2 × 10-5 1.2 × 10-6 1.4 × 10-5 8.5 × 10-7 1.6 × 10-5

Sample calculation: Urban route from INL to Hanford through Idaho, Rail-Lead cask

RADTRAN output (unit risk): 2.21 × 10-9 person-rem (from Figure B-13)

Population density: 2,281 persons/km2

Population multiplier: 1.133

Route segment length: 10.5 km

Population (collective) dose = 2.21 × 10-9 * 2,281 * 1.133 * 10.5 = 6.00 × 10-5 person-rem

Convert to SI units: 6.00 × 10-5 person-rem * 0.01 person-Sv/person-rem = 6.00 × 10-7 person-Sv

Cells with zero values indicate no route miles of this population density.
B-26

Table B-11 Collective Doses to Residents along the Route (Person-Sv) from Rail 

Transportation; Shipment Origin—Indian Point

DEST. AND 

ROUTES

Rail-Lead Rail-Steel

Rural Suburban Urban Total Rural Suburban Urban Total

ORNL

Delaware 1.2 × 10-8 7.3 × 10-6 8.2 × 10-7 8.2 × 10-6 9.5 × 10-9 5.6 × 10-6 6.3 × 10-7 6.2 × 10-6

DC 3.2 × 10-9 8.9 × 10-7 4.5 × 10-7 1.3 × 10-6 2.4 × 10-9 6.8 × 10-7 3.5 × 10-7 1.0 × 10-6

Maryland 6.9 × 10-7 2.2 × 10-5 2.0 × 10-6 2.5 × 10-5 5.3 × 10-7 1.7 × 10-5 1.5 × 10-6 1.9 × 10-5

New Jersey 4.0 × 10-7 1.2 × 10-5 1.4 × 10-6 1.4 × 10-5 3.1 × 10-7 9.1 × 10-6 1.1 × 10-6 1.0 × 10-5

New York 3.0 × 10-8 1.6 × 10-6 3.4 × 10-6 5.0 × 10-6 2.3 × 10-8 1.3 × 10-6 2.5 × 10-6 3.8 × 10-6

Pennsylvania 4.9 × 10-8 8.6 × 10-6 3.2 × 10-6 1.2 × 10-5 3.8 × 10-8 6.5 × 10-6 2.4 × 10-6 9.0 × 10-6

Tennessee 2.2 × 10-6 3.0 × 10-5 6.6 × 10-7 3.3 × 10-5 1.7 × 10-6 2.3 × 10-5 5.0 × 10-7 2.5 × 10-5

Virginia 4.1 × 10-6 5.9 × 10-5 2.3 × 10-6 6.5 × 10-5 3.1 × 10-6 4.5 × 10-5 1.7 × 10-6 5.0 × 10-5

DEAF SMITH

Illinois 1.5 × 10-6 2.7 × 10-5 2.4 × 10-6 3.1 × 10-5 1.1 × 10-6 2.0 × 10-5 1.8 × 10-6 2.3 × 10-5

Indiana 2.1 × 10-6 1.1 × 10-5 5.4 × 10-7 1.4 × 10-5 1.6 × 10-6 8.6 × 10-6 4.1 × 10-7 1.1 × 10-5

Iowa 3.0 × 10-7 6.2 × 10-7 3.1 × 10-8 9.5 × 10-7 2.3 × 10-7 4.8 × 10-7 2.4 × 10-8 7.2 × 10-7

Kansas 2.0 × 10-6 1.8 × 10-5 7.9 × 10-7 2.1 × 10-5 1.5 × 10-6 1.4 × 10-5 6.0 × 10-7 1.6 × 10-5

Missouri 1.2 × 10-6 7.0 × 10-6 2.4 × 10-7 8.5 × 10-6 9.2 × 10-7 5.4 × 10-6 1.8 × 10-7 6.5 × 10-6

New York 5.5 × 10-6 6.1 × 10-5 4.9 × 10-6 7.1 × 10-5 4.2 × 10-6 4.6 × 10-5 3.7 × 10-6 5.4 × 10-5

Ohio 2.5 × 10-6 3.2 × 10-5 2.3 × 10-6 3.6 × 10-5 1.9 × 10-6 2.4 × 10-5 1.7 × 10-6 2.8 × 10-5

Oklahoma 4.5 × 10-7 4.0 × 10-6 5.2 × 10-8 4.5 × 10-6 3.4 × 10-7 3.1 × 10-6 3.9 × 10-8 3.4 × 10-6

Pennsylvania 4.1 × 10-7 9.3 × 10-6 4.9 × 10-7 1.0 × 10-5 3.1 × 10-7 7.1 × 10-6 3.7 × 10-7 7.8 × 10-6

Texas 7.3 × 10-7 5.1 × 10-6 1.2 × 10-7 6.0 × 10-6 5.6 × 10-7 3.9 × 10-6 9.4 × 10-8 4.5 × 10-6

HANFORD

Idaho 9.8 × 10-7 6.7 × 10-6 9.6 × 10-8 7.8 × 10-6 7.5 × 10-7 5.1 × 10-6 7.3 × 10-8 5.9 × 10-6

Illinois 1.4 × 10-6 2.1 × 10-5 2.2 × 10-6 2.4 × 10-5 1.0 × 10-6 1.6 × 10-5 1.7 × 10-6 1.9 × 10-5

Indiana 2.1 × 10-6 1.1 × 10-5 5.4 × 10-7 1.4 × 10-5 1.6 × 10-6 8.6 × 10-6 4.1 × 10-7 1.1 × 10-5

Minnesota 3.2 × 10-6 2.9 × 10-5 1.2 × 10-6 3.4 × 10-5 2.4 × 10-6 2.2 × 10-5 8.9 × 10-7 2.6 × 10-5

Montana 2.2 × 10-6 1.3 × 10-5 1.4 × 10-7 1.6 × 10-5 1.7 × 10-6 1.0 × 10-5 1.1 × 10-7 1.2 × 10-5

New York 5.5 × 10-6 6.1 × 10-5 4.9 × 10-6 7.1 × 10-5 4.2 × 10-6 4.6 × 10-5 3.7 × 10-6 5.4 × 10-5

North Dakota 1.0 × 10-6 8.2 × 10-6 2.6 × 10-7 9.5 × 10-6 7.7 × 10-7 6.3 × 10-6 2.0 × 10-7 7.2 × 10-6

Ohio 2.5 × 10-6 3.2 × 10-5 2.3 × 10-6 3.6 × 10-5 1.9 × 10-6 2.4 × 10-5 1.7 × 10-6 2.8 × 10-5

Pennsylvania 4.1 × 10-7 9.3 × 10-6 4.9 × 10-7 1.0 × 10-5 3.1 × 10-7 7.1 × 10-6 3.7 × 10-7 7.8 × 10-6

Washington 1.1 × 10-6 1.3 × 10-5 6.5 × 10-7 1.5 × 10-5 8.5 × 10-7 1.0 × 10-5 5.0 × 10-7 1.2 × 10-5

Wisconsin 1.7 × 10-6 8.2 × 10-6 3.7 × 10-7 1.0 × 10-5 1.3 × 10-6 6.2 × 10-6 2.8 × 10-7 7.8 × 10-6
Table B-11 Collective Doses to Residents along the Route (Person-Sv) from Rail 

Transportation; Shipment Origin—Indian Point (continued)

B-27

DEST. AND 

ROUTES

Rail-Lead Rail-Steel

Rural Suburban Urban Total Rural Suburban Urban Total

SKULL VALLEY

Colorado 1.3 × 10-7 5.8 × 10-7 0 7.1 × 10-7 1.0 × 10-7 4.4 × 10-7 0 5.4 × 10-7

Illinois 1.3 × 10-6 2.1 × 10-5 2.7 × 10-6 2.5 × 10-5 9.9 × 10-7 1.6 × 10-5 2.1 × 10-6 1.9 × 10-5

Indiana 2.1 × 10-6 1.1 × 10-5 5.4 × 10-7 1.4 × 10-5 1.6 × 10-6 8.6 × 10-6 4.1 × 10-7 1.1 × 10-5

Iowa 4.0 × 10-6 1.8 × 10-5 3.4 × 10-7 2.2 × 10-5 3.1 × 10-6 1.4 × 10-5 2.6 × 10-7 1.7 × 10-5

Nebraska 4.2 × 10-6 2.0 × 10-5 6.2 × 10-7 2.5 × 10-5 3.2 × 10-6 1.5 × 10-5 4.7 × 10-7 1.9 × 10-5

New York 5.5 × 10-6 6.1 × 10-5 4.9 × 10-6 7.1 × 10-5 4.2 × 10-6 4.6 × 10-5 3.7 × 10-6 5.4 × 10-5

Ohio 2.5 × 10-6 3.2 × 10-5 2.3 × 10-6 3.6 × 10-5 1.9 × 10-6 2.4 × 10-5 1.7 × 10-6 2.8 × 10-5

Pennsylvania 4.1 × 10-7 9.3 × 10-6 4.9 × 10-7 1.0 × 10-5 3.1 × 10-7 7.1 × 10-6 3.7 × 10-7 7.8 × 10-6

Utah 1.3 × 10-6 1.8 × 10-5 1.1 × 10-6 2.0 × 10-5 9.7 × 10-7 1.4 × 10-5 8.6 × 10-7 1.6 × 10-5

Wyoming 1.4 × 10-6 9.5 × 10-6 2.3 × 10-7 1.1 × 10-5 1.0 × 10-6 7.2 × 10-6 1.8 × 10-7 8.4 × 10-6

Sample calculation: Urban route from Indian Point to Deaf Smith through Indiana, Rail-Lead cask

RADTRAN output (unit risk): 2.21 × 10-9 person-rem (from Figure B-13)

Population density: 2,305.9 persons/km2

Population multiplier: 1.0

Route segment length: 10.6 km

Population (collective) dose = 2.21 × 10-9 * 2,305.9 * 1.0 * 10.6 = 5.40 x 10-5 person-rem

Convert to SI units: 5.40 × 10-5 person-rem * 0.01 person-Sv/person-rem = 5.40 × 10-7 person-Sv

Cells with zero values indicate no route miles of this population density.
B-28

Table B-12 Collective Doses to Residents along the Route (Person-Sv) from Rail 

Transportation; Shipment Origin—Kewaunee

DEST. AND 

ROUTES

Rail-Lead Rail-Steel

Rural Suburban Urban Total Rural Suburban Urban Total

ORNL

Illinois 2.4 × 10-7 2.1 × 10-5 2.5 × 10-6 2.3 × 10-5 1.8 × 10-7 1.6 × 10-5 1.9 × 10-6 1.8 × 10-5

Indiana 2.1 × 10-6 1.1 × 10-5 5.4 × 10-7 1.4 × 10-5 1.6 × 10-6 8.6 × 10-6 4.1 × 10-7 1.1 × 10-5

Kentucky 3.2 × 10-6 1.6 × 10-5 7.0 × 10-7 2.0 × 10-5 2.4 × 10-6 1.3 × 10-5 5.3 × 10-7 1.5 × 10-5

Ohio 2.2 × 10-6 3.0 × 10-5 1.4 × 10-6 3.3 × 10-5 1.6 × 10-6 2.3 × 10-5 1.1 × 10-6 2.5 × 10-5

Tennessee 7.4 × 10-7 5.0 × 10-6 4.1 × 10-8 5.7 × 10-6 5.6 × 10-7 3.8 × 10-6 3.1 × 10-8 4.4 × 10-6

Wisconsin 1.9 × 10-6 2.5 × 10-5 1.5 × 10-6 2.9 × 10-5 1.5 × 10-6 1.9 × 10-5 1.1 × 10-6 2.2 × 10-5

DEAF SMITH

Illinois 1.6 × 10-6 3.5 × 10-5 3.0 × 10-6 4.0 × 10-5 1.2 × 10-6 2.7 × 10-5 2.3 × 10-6 3.0 × 10-5

Iowa 3.0 × 10-7 6.2 × 10-7 3.1 × 10-8 9.5 × 10-7 2.3 × 10-7 4.8 × 10-7 2.4 × 10-8 7.2 × 10-7

Kansas 2.0 × 10-6 1.8 × 10-5 7.9 × 10-7 2.1 × 10-5 1.5 × 10-6 1.4 × 10-5 6.0 × 10-7 1.6 × 10-5

Missouri 1.2 × 10-6 7.0 × 10-6 2.8 × 10-7 8.5 × 10-6 9.2 × 10-7 5.4 × 10-6 2.2 × 10-7 6.5 × 10-6

Oklahoma 4.5 × 10-7 4.0 × 10-6 5.2 × 10-8 4.5 × 10-6 3.4 × 10-7 3.1 × 10-6 3.9 × 10-8 3.4 × 10-6

Texas 7.3 × 10-7 5.1 × 10-6 1.2 × 10-7 6.0 × 10-6 5.6 × 10-7 3.9 × 10-6 9.4 × 10-8 4.6 × 10-6

Wisconsin 1.9 × 10-6 2.5 × 10-5 1.5 × 10-6 2.9 × 10-5 1.5 × 10-6 1.9 × 10-5 1.1 × 10-6 2.2 × 10-5

HANFORD

Idaho 9.8 × 10-7 6.7 × 10-6 9.6 × 10-8 7.8 × 10-6 7.5 × 10-7 5.1 × 10-6 7.3 × 10-8 5.9 × 10-6

Minnesota 3.3 × 10-6 3.0 × 10-5 9.2 × 10-7 3.5 × 10-5 2.5 × 10-6 2.3 × 10-5 7.0 × 10-7 2.6 × 10-5

Montana 2.2 × 10-6 1.3 × 10-5 1.4 × 10-7 1.3 × 10-5 1.7 × 10-6 1.0 × 10-5 1.1 × 10-7 1.0 × 10-5

North Dakota 1.0 × 10-6 8.2 × 10-6 2.6 × 10-7 9.5 × 10-6 7.7 × 10-7 6.3 × 10-6 2.0 × 10-7 7.2 × 10-6

Washington 1.1 × 10-6 1.3 × 10-5 6.5 × 10-7 1.5 × 10-5 8.5 × 10-7 1.0 × 10-5 5.0 × 10-7 1.2 × 10-5

Wisconsin 3.5 × 10-6 2.2 × 10-5 8.9 × 10-7 2.6 × 10-5 2.7 × 10-6 1.7 × 10-5 6.8 × 10-7 2.0 × 10-5

SKULL VALLEY

Colorado 1.3 × 10-7 5.8 × 10-7 0 7.1 × 10-7 1.0 × 10-7 4.4 × 10-7 0 5.4 × 10-7

Illinois 1.4 × 10-6 2.7 × 10-5 2.8 × 10-6 3.1 × 10-5 1.1 × 10-6 2.1 × 10-5 2.1 × 10-6 2.4 × 10-5

Iowa 4.0 × 10-6 1.8 × 10-5 3.4 × 10-7 2.2 × 10-5 3.1 × 10-6 1.4 × 10-5 2.6 × 10-7 1.7 × 10-5

Nebraska 4.2 × 10-6 2.0 × 10-5 6.2 × 10-7 2.5 × 10-5 3.2 × 10-6 1.5 × 10-5 4.7 × 10-7 1.9 × 10-5

Utah 1.3 × 10-6 1.8 × 10-5 1.1 × 10-6 2.0 × 10-5 9.7 × 10-7 1.4 × 10-5 8.6 × 10-7 1.6 × 10-5

Wisconsin 1.9 × 10-6 2.5 × 10-5 1.5 × 10-6 2.9 × 10-5 1.5 × 10-6 1.9 × 10-5 1.1 × 10-6 2.2 × 10-5

Wyoming 1.4 × 10-6 9.5 × 10-6 2.3 × 10-7 1.1 × 10-5 1.0 × 10-6 7.2 × 10-6 1.8 × 10-7 8.4 × 10-6

Sample calculation: Rural route from Kewaunee to ORNL through Ohio, Rail-Steel cask

RADTRAN output (unit risk): 5.55 × 10-8 person-rem (from Figure B- 13)

Population density: 14.8 persons/km2

Population multiplier: 1.0

Route segment length: 200.6 km

Population (collective) dose = 5.55 × 10-8 * 14.8 * 1.0 * 200.6 = 1.65 × 10-4 person-rem

Convert to SI units: 1.65 × 10-4 person-rem * 0.01 person-Sv/person-rem = 1.65 × 10-6 person-Sv

Cells with zero values indicate no route miles of this population density.
B-29

Table B-13 Collective Doses to Residents along the Route (Person-Sv) from Rail 

Transportation; Shipment Origin—Maine Yankee

DEST. AND 

ROUTES

Rail-Lead Rail-Steel

Rural Suburban Urban Total Rural Suburban Urban Total

ORNL

Kentucky 3.2 × 10-6 1.6 × 10-5 7.0 × 10-7 2.0 × 10-5 2.4 × 10-6 1.3 × 10-5 5.3 × 10-7 1.5 × 10-5

Maine 9.3 × 10-7 1.6 × 10-5 6.2 × 10-7 1.7 × 10-5 7.1 × 10-7 1.2 × 10-5 4.7 × 10-7 1.3 × 10-5

Massachusetts 1.3 × 10-6 2.9 × 10-5 1.8 × 10-6 3.2 × 10-5 1.0 × 10-6 2.2 × 10-5 1.4 × 10-6 2.4 × 10-5

New 

Hampshire 3.8 × 10-7 7.5 × 10-6 2.5 × 10-7 8.2 × 10-6 2.9 × 10-7 5.7 × 10-6 1.9 × 10-7 6.2 × 10-6

New York 4.8 × 10-6 5.2 × 10-5 1.8 × 10-6 5.9 × 10-5 3.7 × 10-6 4.0 × 10-5 1.4 × 10-6 4.5 × 10-5

Ohio 3.6 × 10-6 4.9 × 10-5 3.2 × 10-6 5.6 × 10-5 2.7 × 10-6 3.7 × 10-5 2.5 × 10-6 4.3 × 10-5

Pennsylvania 4.2 × 10-7 9.4 × 10-6 5.1 × 10-7 1.0 × 10-5 3.2 × 10-7 7.2 × 10-6 3.9 × 10-7 7.9 × 10-6

Tennessee 7.4 × 10-7 5.0 × 10-6 4.1 × 10-8 5.7 × 10-6 5.6 × 10-7 3.8 × 10-6 3.1 × 10-8 4.4 × 10-6

Vermont 6.7 × 10-8 5.2 × 10-7 0 5.9 × 10-7 5.1 × 10-8 4.0 × 10-7 0 4.5 × 10-7

DEAF SMITH

Illinois 1.5 × 10-6 2.7 × 10-5 2.4 × 10-6 3.1 × 10-5 1.1 × 10-6 2.0 × 10-5 1.8 × 10-6 2.3 × 10-5

Indiana 2.1 × 10-6 1.1 × 10-5 5.4 × 10-7 1.4 × 10-5 1.6 × 10-6 8.6 × 10-6 4.1 × 10-7 1.1 × 10-5

Iowa 3.0 × 10-7 6.2 × 10-7 3.1 × 10-8 9.5 × 10-7 2.3 × 10-7 4.8 × 10-7 2.4 × 10-8 7.2 × 10-7

Kansas 2.0 × 10-6 1.8 × 10-5 7.9 × 10-7 2.1 × 10-5 1.5 × 10-6 1.4 × 10-5 6.0 × 10-7 1.6 × 10-5

Maine 9.3 × 10-7 1.6 × 10-5 6.2 × 10-7 1.7 × 10-5 7.1 × 10-7 1.2 × 10-5 4.7 × 10-7 1.3 × 10-5

Massachusetts 1.3 × 10-6 2.9 × 10-5 1.8 × 10-6 3.2 × 10-5 1.0 × 10-6 2.2 × 10-5 1.4 × 10-6 2.4 × 10-5

Missouri 1.2 × 10-6 7.0 × 10-6 2.4 × 10-7 8.5 × 10-6 9.2 × 10-7 5.4 × 10-6 1.8 × 10-7 6.5 × 10-6

New 

Hampshire 3.8 × 10-7 7.5 × 10-6 2.5 × 10-7 8.2 × 10-6 2.9 × 10-7 5.7 × 10-6 1.9 × 10-7 6.2 × 10-6

New York 4.8 × 10-6 5.2 × 10-5 1.8 × 10-6 5.9 × 10-5 3.7 × 10-6 4.0 × 10-5 1.4 × 10-6 4.5 × 10-5

Ohio 2.5 × 10-6 3.2 × 10-5 2.3 × 10-6 3.6 × 10-5 1.9 × 10-6 2.4 × 10-5 1.7 × 10-6 2.8 × 10-5

Oklahoma 4.3 × 10-7 3.9 × 10-6 4.9 × 10-8 4.5 × 10-6 3.3 × 10-7 3.0 × 10-6 3.7 × 10-8 3.4 × 10-6

Pennsylvania 4.1 × 10-7 9.3 × 10-6 4.9 × 10-7 1.0 × 10-5 3.1 × 10-7 7.1 × 10-6 3.7 × 10-7 7.8 × 10-6

Texas 7.3 × 10-7 5.1 × 10-6 1.2 × 10-7 6.0 × 10-6 5.6 × 10-7 3.9 × 10-6 9.4 × 10-8 4.5 × 10-6

Vermont 6.7 × 10-8 5.2 × 10-7 0 5.9 × 10-7 5.1 × 10-8 4.0 × 10-7 0 4.5 × 10-7

HANFORD

Idaho 9.8 × 10-7 6.7 × 10-6 9.6 × 10-8 7.8 × 10-6 7.5 × 10-7 5.1 × 10-6 7.3 × 10-8 5.9 × 10-6

Illinois 1.4 × 10-6 2.1 × 10-5 2.2 × 10-6 2.4 × 10-5 1.0 × 10-6 1.6 × 10-5 1.7 × 10-6 1.9 × 10-5

Indiana 2.1 × 10-6 1.1 × 10-5 5.4 × 10-7 1.4 × 10-5 1.6 × 10-6 8.6 × 10-6 4.1 × 10-7 1.1 × 10-5

Maine 9.3 × 10-7 1.6 × 10-5 6.2 × 10-7 1.7 × 10-5 7.1 × 10-7 1.2 × 10-5 4.7 × 10-7 1.3 × 10-5

Massachusetts 1.3 × 10-6 2.9 × 10-5 1.8 × 10-6 3.2 × 10-5 1.0 × 10-6 2.2 × 10-5 1.4 × 10-6 2.4 × 10-5

Minnesota 3.2 × 10-6 2.9 × 10-5 1.2 × 10-6 3.4 × 10-5 2.4 × 10-6 2.2 × 10-5 8.9 × 10-7 2.6 × 10-5

Montana 2.2 × 10-6 1.3 × 10-5 1.4 × 10-7 1.6 × 10-5 1.7 × 10-6 1.0 × 10-5 1.1 × 10-7 1.2 × 10-5

New 

Hampshire 3.8 × 10-7 7.5 × 10-6 2.5 × 10-7 8.2 × 10-6 2.9 × 10-7 5.7 × 10-6 1.9 × 10-7 6.2 × 10-6

New York 4.8 × 10-6 5.2 × 10-5 2.2 × 10-6 5.9 × 10-5 3.7 × 10-6 4.0 × 10-5 1.7 × 10-6 4.5 × 10-5

North Dakota 1.0 × 10-6 8.2 × 10-6 2.6 × 10-7 9.5 × 10-6 7.7 × 10-7 6.3 × 10-6 2.0 × 10-7 7.2 × 10-6
B-30

Table B-13 Collective Doses to Residents along the Route (Person-Sv) from Rail 

Transportation; Shipment Origin Maine Yankee (continued)

DEST. AND 

ROUTES

Rail-Lead Rail-Steel

Rural Suburban Urban Total Rural Suburban Urban Total

HANFORD (cont.)

Ohio 2.5 × 10-6 3.2 × 10-5 2.3 × 10-6 3.6 × 10-5 1.9 × 10-6 2.4 × 10-5 1.7 × 10-6 2.8 × 10-5

Pennsylvania 4.1 × 10-7 9.3 × 10-6 4.9 × 10-7 1.0 × 10-5 3.1 × 10-7 7.1 × 10-6 3.7 × 10-7 7.8 × 10-6

Vermont 6.7 × 10-8 5.2 × 10-7 0 5.9 × 10-7 5.1 × 10-8 4.0 × 10-7 0 4.5 × 10-7

Washington 1.1 × 10-6 1.3 × 10-5 6.5 × 10-7 1.5 × 10-5 8.5 × 10-7 1.0 × 10-5 5.0 × 10-7 1.2 × 10-5

Wisconsin 1.7 × 10-6 8.2 × 10-6 3.7 × 10-7 1.0 × 10-5 1.3 × 10-6 6.2 × 10-6 2.8 × 10-7 7.8 × 10-6

SKULL VALLEY

Colorado 1.3 × 10-7 5.8 × 10-7 0 7.1 × 10-7 1.0 × 10-7 4.4 × 10-7 0 5.4 × 10-7

Illinois 1.3 × 10-6 2.2 × 10-5 2.7 × 10-6 2.6 × 10-5 1.0 × 10-6 1.7 × 10-5 2.0 × 10-6 2.0 × 10-5

Indiana 2.1 × 10-6 1.1 × 10-5 5.4 × 10-7 1.4 × 10-5 1.6 × 10-6 8.6 × 10-6 4.1 × 10-7 1.1 × 10-5

Iowa 4.0 × 10-6 1.8 × 10-5 3.4 × 10-7 2.2 × 10-5 3.1 × 10-6 1.4 × 10-5 2.6 × 10-7 1.7 × 10-5

Maine 9.6 × 10-7 1.6 × 10-5 1.6 × 10-7 1.7 × 10-5 7.3 × 10-7 1.2 × 10-5 1.2 × 10-7 1.3 × 10-5

Massachusett

s

6.5 × 10-7 2.8 × 10-5 8.5 × 10-7 3.0 × 10-5 4.9 × 10-7 2.1 × 10-5 6.5 × 10-7 2.3 × 10-5

Nebraska 4.2 × 10-6 2.0 × 10-5 6.2 × 10-7 2.5 × 10-5 3.2 × 10-6 1.5 × 10-5 4.7 × 10-7 1.9 × 10-5

New 

Hampshire 1.1 × 10-7 3.7 × 10-6 4.9 × 10-8 3.8 × 10-6 8.6 × 10-8 2.8 × 10-6 3.7 × 10-8 2.9 × 10-6

New York 4.8 × 10-6 5.2 × 10-5 1.8 × 10-6 5.9 × 10-5 3.7 × 10-6 4.0 × 10-5 1.4 × 10-6 4.5 × 10-5

Ohio 2.5 × 10-6 3.2 × 10-5 2.3 × 10-6 3.6 × 10-5 1.9 × 10-6 2.4 × 10-5 1.7 × 10-6 2.8 × 10-5

Pennsylvania 4.1 × 10-7 9.3 × 10-6 4.9 × 10-7 1.0 × 10-5 3.1 × 10-7 7.1 × 10-6 3.7 × 10-7 7.8 × 10-6

Utah 1.3 × 10-6 1.8 × 10-5 1.1 × 10-6 2.0 × 10-5 9.7 × 10-7 1.4 × 10-5 8.6 × 10-7 1.6 × 10-5

Vermont 6.7 × 10-8 5.2 × 10-7 0 5.9 × 10-7 5.1 × 10-8 4.0 × 10-7 0 4.5 × 10-7

Wyoming 1.4 × 10-6 9.5 × 10-6 2.3 × 10-7 1.1 × 10-5 1.0 × 10-6 7.2 × 10-6 1.8 × 10-7 8.4 × 10-6

Sample calculation: Rural route from Maine Yankee to Skull Valley through Nebraska, Rail-Steel cask

RADTRAN output (unit risk): 5.55 × 10-8 person-rem (from Figure B- 13) 

Population density: 9.3 persons/km2

Population multiplier: 1.0

Route segment length: 621.7 km

Population (collective) dose = 5.55 × 10-8 * 9.3 * 1.0 * 621.7 = 3.21 × 10-4 person-rem

Convert to SI units: 3.17 × 10-4 person-rem * 0.01 person-Sv/person-rem = 3.21 × 10-6 person-Sv

Cells with zero values indicate no route miles of this population density.
B-31

Table B-14 Collective Doses to Residents along the Route (Person-Sv) from Truck-DU; 

Shipment Origin—Maine Yankee

DESTINATION ROUTES Rural Suburban Urban

Urban 

Rush 

Hour

Total

ORNL

Connecticut 2.8 × 10-7 1.5 × 10-5 5.6 × 10-7 1.2 × 10-7 1.58 × 10-5

Maine 4.0 × 10-7 7.3 × 10-6 6.6 × 10-8 1.5 × 10-8 7.74 × 10-6

Maryland 3.4 × 10-8 1.3 × 10-6 1.3 × 10-8 3.0 × 10-9 1.33 × 10-6

Massachusetts 2.7 × 10-7 1.2 × 10-5 2.0 × 10-7 4.5 × 10-8 1.23 × 10-5

New Hampshire 4.7 × 10-8 1.5 × 10-6 7.4 × 10-9 1.6 × 10-9 1.59 × 10-6

New Jersey 1.8 × 10-7 6.4 × 10-6 3.1 × 10-7 6.9 × 10-8 6.92 × 10-6

New York 2.1 × 10-9 1.7 × 10-6 5.1 × 10-7 1.1 × 10-7 2.28 × 10-6

Pennsylvania 1.1 × 10-6 1.3 × 10-5 1.6 × 10-7 3.6 × 10-8 1.42 × 10-5

Tennessee 7.6 × 10-7 9.4 × 10-6 7.9 × 10-8 1.8 × 10-8 1.02 × 10-5

Virginia 1.8 × 10-6 1.9 × 10-5 1.3 × 10-7 2.8 × 10-8 2.13 × 10-5

West Virginia 1.1 × 10-7 2.6 × 10-6 7.2 × 10-9 1.6 × 10-9 2.73 × 10-6

DEAF SMITH

Connecticut 1.5 × 10-6 9.9 × 10-6 9.5 × 10-8 2.1 × 10-8 1.15 × 10-5

Maine 2.9 × 10-7 1.5 × 10-5 5.7 × 10-7 1.3 × 10-7 1.60 × 10-5

Maryland 4.0 × 10-7 6.8 × 10-6 3.7 × 10-8 8.2 × 10-9 7.28 × 10-6

Massachusetts 3.4 × 10-8 1.3 × 10-6 1.3 × 10-8 3.0 × 10-9 1.34 × 10-6

New Hampshire 2.7 × 10-7 1.2 × 10-5 2.0 × 10-7 4.5 × 10-8 1.23 × 10-5

New Jersey 4.7 × 10-8 1.5 × 10-6 7.4 × 10-9 1.6 × 10-9 1.59 × 10-6

New York 2.4 × 10-7 8.6 × 10-6 1.7 × 10-7 3.9 × 10-8 9.03 × 10-6

Oklahoma 2.4 × 10-8 4.3 × 10-6 2.0 × 10-7 4.5 × 10-8 4.53 × 10-6

Pennsylvania 1.6 × 10-6 8.4 × 10-6 1.1 × 10-7 2.4 × 10-8 1.02 × 10-5

Tennessee 9.4 × 10-7 1.1 × 10-5 1.3 × 10-7 3.0 × 10-8 1.19 × 10-5

Texas 2.9 × 10-6 2.5 × 10-5 3.9 × 10-7 8.7 × 10-8 2.83 × 10-5

Virginia 3.9 × 10-7 2.2 × 10-6 9.5 × 10-8 2.1 × 10-8 2.76 × 10-6

West Virginia 1.7 × 10-6 1.8 × 10-5 1.2 × 10-7 2.7 × 10-8 2.03 × 10-5
B-32

Table B-14 Collective Doses to Residents Along the Route (Person-Sv) from Truck-DU; 

Shipment Origin Maine Yankee (continued)

DESTINATION ROUTES Rural Suburban Urban Urban 

Rush Hour Total

HANFORD

Connecticut 2.7 × 10-7 1.1 × 10-5 3.0 × 10-7 6.6 × 10-9 1.18 × 10-5

Idaho 1.4 × 10-6 6.6 × 10-6 9.6 × 10-8 2.1 × 10-9 8.15 × 10-6

Illinois 8.1 × 10-7 6.5 × 10-6 1.2 × 10-7 2.7 × 10-9 7.42 × 10-6

Indiana 8.3 × 10-7 7.1 × 10-6 1.2 × 10-7 2.6 × 10-9 8.10 × 10-6

Iowa 1.9 × 10-6 6.8 × 10-6 5.8 × 10-8 1.3 × 10-9 8.73 × 10-6

Maine 4.0 × 10-7 6.8 × 10-6 3.7 × 10-8 8.2 × 10-10 7.28 × 10-6

Massachusetts 2.8 × 10-7 1.2 × 10-5 2.1 × 10-7 4.6 × 10-9 1.28 × 10-5

Nebraska 2.0 × 10-6 5.4 × 10-6 8.8 × 10-8 2.0 × 10-9 7.48 × 10-6

New Hampshire 4.7 × 10-8 1.5 × 10-6 7.4 × 10-9 1.6 × 10-10 1.59 × 10-6

New York 2.8 × 10-7 5.7 × 10-6 5.0 × 10-8 1.1 × 10-9 6.05 × 10-6

Ohio 1.3 × 10-6 1.2 × 10-5 1.7 × 10-7 3.8 × 10-9 1.39 × 10-5

Oregon 8.1 × 10-7 2.9 × 10-6 2.6 × 10-8 5.8 × 10-10 3.74 × 10-6

Pennsylvania 2.0 × 10-6 1.1 × 10-5 8.2 × 10-8 1.8 × 10-9 1.33 × 10-5

Utah 6.3 × 10-7 4.0 × 10-6 1.8 × 10-8 4.1 × 10-10 4.68 × 10-6

Washington 8.9 × 10-8 8.4 × 10-7 5.0 × 10-8 1.1 × 10-9 9.92 × 10-7

Wyoming 9.1 × 10-7 3.6 × 10-6 3.5 × 10-8 7.8 × 10-10 4.54 × 10-6

SKULL VALLEY

Connecticut 2.7 × 10-7 1.1 × 10-5 3.0 × 10-7 6.6 × 10-9 1.18 × 10-5

Illinois 8.1 × 10-7 6.5 × 10-6 1.2 × 10-7 2.7 × 10-9 7.42 × 10-6

Indiana 8.3 × 10-7 7.1 × 10-6 1.2 × 10-7 2.6 × 10-9 8.10 × 10-6

Iowa 1.9 × 10-6 6.8 × 10-6 5.8 × 10-8 1.3 × 10-9 8.73 × 10-6

Maine 4.0 × 10-7 6.8 × 10-6 3.7 × 10-8 8.2 × 10-10 7.28 × 10-6

Massachusetts 2.7 × 10-7 1.2 × 10-5 2.0 × 10-7 4.5 × 10-9 1.23 × 10-5

Nebraska 2.0 × 10-6 5.4 × 10-6 8.8 × 10-8 2.0 × 10-9 7.48 × 10-6

New Hampshire 4.7 × 10-8 1.5 × 10-6 7.4 × 10-9 1.6 × 10-10 1.59 × 10-6

New York 2.8 × 10-7 5.7 × 10-6 5.0 × 10-8 1.1 × 10-9 6.05 × 10-6

Ohio 1.3 × 10-6 1.2 × 10-5 1.7 × 10-7 3.8 × 10-9 1.39 × 10-5

Pennsylvania 2.0 × 10-6 1.1 × 10-5 8.2 × 10-8 1.8 × 10-9 1.33 × 10-5

Utah 5.2 × 10-7 4.6 × 10-6 2.0 × 10-7 4.5 × 10-9 5.40 × 10-6

Wyoming 9.1 × 10-7 3.6 × 10-6 3.5 × 10-8 7.8 × 10-10 4.54 × 10-6

Sample calculation: Rural route from Maine Yankee to ORNL through New York, Truck-DU cask

RADTRAN output (unit risk): 3.1 × 10-10 Sv-km (from Table B-9)

Population density: 4.4 persons/km2

Population multiplier: 1.0

Route segment length: 1.6 km

Population (collective) dose = 3.1 × 10-10 * 4.4 * 1.0 * 1.6 = 2.15 × 10-9 person-Sv
B-33

Table B-15 Collective Doses to Residents along the Route (Person-Sv) from Truck-DU; 

Shipment Origin—Indian Point

DESTINATION ROUTES Rural Suburban Urban U. Rush Hour Total

ORNL

Maryland 5.4 × 10-8 2.1 × 10-6 3.0 × 10-9 3.0 × 10-9 1.3 × 10-6

New Jersey 3.9 × 10-7 1.4 × 10-5 3.9 × 10-8 3.9 × 10-8 9.0 × 10-6

New York 7.5 × 10-8 7.0 × 10-6 4.9 × 10-8 4.9 × 10-8 4.7 × 10-6

Pennsylvania 9.4 × 10-7 1.1 × 10-5 3.0 × 10-8 3.0 × 10-8 1.2 × 10-5

Tennessee 7.9 × 10-7 9.7 × 10-6 1.5 × 10-8 1.5 × 10-8 1.1 × 10-5

Virginia 1.7 × 10-6 1.8 × 10-5 2.7 × 10-8 2.7 × 10-8 2.0 × 10-5

West Virginia 1.1 × 10-7 2.6 × 10-6 1.6 × 10-9 1.6 × 10-9 2.7 × 10-6

DEAF SMITH

Arkansas 2.3 × 10-6 1.6 × 10-5 2.2 × 10-8 2.2 × 10-8 1.1 × 10-5

Maryland 5.4 × 10-8 2.1 × 10-6 3.0 × 10-9 3.0 × 10-9 1.3 × 10-6

New Jersey 3.9 × 10-7 1.4 × 10-5 3.9 × 10-8 3.9 × 10-8 9.0 × 10-6

New York 4.7 × 10-8 4.3 × 10-6 4.9 × 10-8 4.9 × 10-8 4.7 × 10-6

Oklahoma 1.7 × 10-6 8.7 × 10-6 2.6 × 10-8 2.6 × 10-8 1.0 × 10-5

Pennsylvania 9.4 × 10-7 1.1 × 10-5 3.0 × 10-8 3.0 × 10-8 1.2 × 10-5

Tennessee 2.9 × 10-6 2.5 × 10-5 3.9 × 10-7 8.7 × 10-8 2.8 × 10-5

Texas 2.9 × 10-6 2.5 × 10-5 8.7 × 10-8 2.1 × 10-8 2.8 × 10-6

Virginia 3.9 × 10-7 2.2 × 10-6 2.1 × 10-8 2.7 × 10-8 2.0 × 10-5

West Virginia 1.7 × 10-6 1.8 × 10-5 2.7 × 10-8 1.6 × 10-9 2.7 × 10-6

HANFORD

Idaho 1.4 × 10-6 6.6 × 10-6 1.6 × 10-9 2.1 × 10-8 8.1 × 10-6

Illinois 7.8 × 10-7 6.3 × 10-6 2.1 × 10-8 2.7 × 10-8 7.4 × 10-6

Indiana 8.6 × 10-7 7.1 × 10-6 2.7 × 10-8 2.6 × 10-8 8.1 × 10-6

Iowa 1.9 × 10-6 6.8 × 10-6 2.6 × 10-8 1.3 × 10-8 8.7 × 10-6

Nebraska 2.0 × 10-6 5.4 × 10-6 1.3 × 10-8 2.0 × 10-8 7.5 × 10-6

New Jersey 2.6 × 10-7 6.7 × 10-6 2.0 × 10-8 3.2 × 10-8 7.1 × 10-6

New York 4.7 × 10-8 4.3 × 10-6 3.2 × 10-8 4.9 × 10-8 4.7 × 10-6

Ohio 1.3 × 10-6 1.2 × 10-5 4.9 × 10-8 3.8 × 10-8 1.4 × 10-5

Oregon 8.9 × 10-7 3.2 × 10-6 3.8 × 10-8 5.8 × 10-9 4.1 × 10-6

Pennsylvania 1.8 × 10-6 8.7 × 10-6 5.8 × 10-9 8.1 × 10-9 1.1 × 10-5

Utah 6.3 × 10-7 4.0 × 10-6 8.1 × 10-9 4.1 × 10-9 4.7 × 10-6

Washington 8.9 × 10-8 8.4 × 10-7 4.1 × 10-9 1.1 × 10-8 9.9 × 10-7

Wyoming 9.1 × 10-7 3.6 × 10-6 1.1 × 10-8 7.8 × 10-9 4.5 × 10-6
Table B-15 Collective Doses to Residents along the Route (Person-Sv) from Truck-DU; 

Shipment Origin—Indian Point (continued)

B-34

DESTINATION ROUTES Rural Suburban Urban U. Rush Hour Total

SKULL VALLEY

Illinois 8.1 × 10-7 6.5 × 10-6 7.8 × 10-9 2.7 × 10-8 7.4 × 10-6

Indiana 8.3 × 10-7 7.1 × 10-6 2.7 × 10-8 2.6 × 10-8 8.1 × 10-6

Iowa 1.9 × 10-6 6.8 × 10-6 2.6 × 10-8 1.3 × 10-8 8.7 × 10-6

Nebraska 2.0 × 10-6 5.4 × 10-6 1.3 × 10-8 2.0 × 10-8 7.5 × 10-6

New Jersey 2.6 × 10-7 6.7 × 10-6 2.0 × 10-8 3.2 × 10-8 7.1 × 10-6

New York 4.7 × 10-8 4.3 × 10-6 3.2 × 10-8 4.9 × 10-8 4.7 × 10-6

Ohio 1.3 × 10-6 1.2 × 10-5 4.9 × 10-8 3.8 × 10-8 1.4 × 10-5

Pennsylvania 1.8 × 10-6 8.7 × 10-6 3.8 × 10-8 8.1 × 10-9 1.1 × 10-5

Utah 5.1 × 10-7 4.6 × 10-6 8.1 × 10-9 4.5 × 10-8 5.4 × 10-6

Wyoming 9.1 × 10-7 3.6 × 10-6 4.5 × 10-8 7.8 × 10-9 4.5 × 10-6

Sample calculation: Rural route from Indian Point to Hanford through Idaho, truck cask

RADTRAN output (unit risk): 3.05 × 10-8 person-rem (from Table B-9)

Population density: 11.3 persons/km2

Population multiplier: 1.133

Route segment length: 357 km

Population (collective) dose = 3.05 × 10-8

*11.3*1.133*357 = 1.39 × 10-4 person-rem

Convert to SI units: 1.39 × 10-4 person-rem*0.01 person-Sv/person-rem = 1.39 × 10-6 person-Sv
B-35

Table B-16 Collective Doses to Residents along the Route (Person-Sv) from Truck-DU; 

Shipment Origin INL

DESTINATION ROUTES Rural Suburban Urban Urban 

Rush Hour Total

ORNL

Colorado 1.0 × 10-6 4.7 × 10-6 1.4 × 10-7 3.1 × 10-8 5.86 × 10-6

Idaho 6.3 × 10-7 2.7 × 10-6 2.4 × 10-8 5.4 × 10-9 3.32 × 10-6

Illinois 9.1 × 10-7 6.0 × 10-6 1.9 × 10-8 4.3 × 10-9 6.98 × 10-6

Kansas 1.6 × 10-6 6.7 × 10-6 1.1 × 10-7 2.5 × 10-8 8.50 × 10-6

Kentucky 5.8 × 10-7 2.3 × 10-6 1.8 × 10-9 4.1 × 10-10 2.86 × 10-6

Missouri 1.2 × 10-6 1.5 × 10-5 3.1 × 10-7 6.9 × 10-8 1.65 × 10-5

Tennessee 1.4 × 10-6 8.8 × 10-6 1.1 × 10-7 2.5 × 10-8 1.03 × 10-5

Utah 6.6 × 10-7 4.2 × 10-6 1.8 × 10-8 4.1 × 10-9 4.89 × 10-6

Wyoming 7.9 × 10-7 2.5 × 10-6 2.5 × 10-8 5.5 × 10-9 3.30 × 10-6

DEAF SMITH

Colorado 1.3 × 10-6 1.6 × 10-5 4.4 × 10-7 9.8 × 10-8 1.74 × 10-5

Idaho 6.3 × 10-7 2.7 × 10-6 2.4 × 10-8 5.4 × 10-9 3.32 × 10-6

New Mexico 1.2 × 10-6 5.6 × 10-6 1.8 × 10-7 4.1 × 10-8 7.02 × 10-6

Texas 5.3 × 10-8 9.4 × 10-8 0 0 1.47 × 10-7

Utah 6.6 × 10-7 4.2 × 10-6 1.8 × 10-8 4.1 × 10-9 4.89 × 10-6

Wyoming 7.9 × 10-7 2.5 × 10-6 2.5 × 10-8 5.5 × 10-9 3.30 × 10-6

HANFORD

Idaho 1.7 × 10-6 8.7 × 10-6 1.1 × 10-7 2.5 × 10-8 1.05 × 10-5

Oregon 3.0 × 10-6 2.5 × 10-8 5.6 × 10-10 5.6 × 10-9 3.83 × 10-6

Washington 8.4 × 10-7 5.0 × 10-8 1.1 × 10-9 1.1 × 10-8 9.92 × 10-7

SKULL VALLEY Idaho 6.3 × 10-7 2.7 × 10-6 2.4 × 10-8 5.4 × 10-9 3.32 × 10-6

Utah 6.0 × 10-7 7.4 × 10-6 2.4 × 10-7 5.4 × 10-8 8.33 × 10-6

Sample calculation: Suburban route from INL to Deaf Smith through Utah, Truck-DU cask

RADTRAN output (unit risk): 2.73 × 10-8 person-rem (from Table B-9)

Population density: 260.1 persons/km2

Population multiplier: 1.102

Route segment length: 53.4 km

Population (collective) dose = 2.73 × 10-8 * 260.1 * 1.102 * 53.4= 4.18 × 10-4 person-rem

Convert to SI units: 4.18 × 10-4 person-rem * 0.01 person-Sv/person-rem = 4.18 × 10-6 person-Sv

Cells with zero values indicate no route miles of this population density.
B-36

Table B-17 Collective Doses to Residents along the Route (Person-Sv) from Truck-DU; 

Shipment Origin—Kewaunee

DESTINATION ROUTES Rural Suburban Urban Urban 

Rush Hour Total

ORNL

Illinois 2.1 × 10-7 9.8 × 10-6 3.6 × 10-7 7.9 × 10-8 1.05 × 10-5

Indiana 1.3 × 10-6 1.2 × 10-5 2.0 × 10-7 4.5 × 10-8 1.31 × 10-5

Kentucky 1.2 × 10-6 1.0 × 10-5 1.3 × 10-7 2.9 × 10-8 1.17 × 10-5

Ohio 5.9 × 10-8 8.3 × 10-7 1.1 × 10-8 2.4 × 10-9 9.04 × 10-7

Tennessee 3.6 × 10-7 6.2 × 10-6 7.8 × 10-8 1.7 × 10-8 6.67 × 10-6

Wisconsin 9.8 × 10-7 7.6 × 10-6 3.5 × 10-7 7.8 × 10-8 9.00 × 10-6

DEAF SMITH

Illinois 7.5 × 10-7 3.3 × 10-6 1.2 × 10-8 2.7 × 10-9 4.02 × 10-6

Iowa 1.4 × 10-6 7.4 × 10-6 6.3 × 10-8 1.4 × 10-8 8.90 × 10-6

Kansas 1.0 × 10-6 6.9 × 10-6 1.6 × 10-7 3.6 × 10-8 8.16 × 10-6

Missouri 6.5 × 10-7 6.1 × 10-6 6.2 × 10-8 1.2 × 10-8 6.81 × 10-6

Oklahoma 1.1 × 10-6 6.0 × 10-6 1.8 × 10-9 2.1 × 10-8 7.18 × 10-6

Texas 3.9 × 10-7 2.2 × 10-6 9.5 × 10-8 2.1 × 10-8 2.76 × 10-6

Wisconsin 1.2 × 10-6 7.6 × 10-6 2.8 × 10-7 6.1 × 10-8 9.20 × 10-6

HANFORD

Idaho 2.1 × 10-7 4.1 × 10-6 6.0 × 10-8 1.1 × 10-8 4.03 × 10-6

Minnesota 1.7 × 10-6 2.7 × 10-6 1.3 × 10-8 2.9 × 10-9 4.46 × 10-6

Montana 2.1 × 10-6 8.2 × 10-6 1.0 × 10-7 2.2 × 10-8 1.03 × 10-5

South Dakota 1.5 × 10-6 3.9 × 10-6 3.1 × 10-8 6.9 × 10-9 5.35 × 10-6

Washington 1.0 × 10-6 9.2 × 10-6 2.0 × 10-7 4.5 × 10-8 1.05 × 10-5

Wisconsin 2.1 × 10-6 1.1 × 10-5 2.7 × 10-7 6.0 × 10-8 1.38 × 10-5

Wyoming 5.6 × 10-7 1.6 × 10-6 2.2 × 10-8 4.9 × 10-9 2.21 × 10-6

SKULL VALLEY

Illinois 7.5 × 10-7 3.3 × 10-6 1.2 × 10-8 2.7 × 10-9 4.02 × 10-6

Iowa 1.9 × 10-6 6.8 × 10-6 5.8 × 10-8 1.3 × 10-8 8.73 × 10-6

Nebraska 2.0 × 10-6 5.4 × 10-6 8.8 × 10-8 2.0 × 10-8 7.48 × 10-6

Utah 5.1 × 10-7 4.6 × 10-6 2.0 × 10-7 4.5 × 10-8 5.39 × 10-6

Wisconsin 1.2 × 10-6 7.6 × 10-6 2.8 × 10-7 6.1 × 10-8 9.20 × 10-6

Wyoming 9.1 × 10-7 3.6 × 10-6 3.5 × 10-8 7.8 × 10-9 4.54 × 10-6

Sample calculation: Urban route from Kewaunee to Skull Valley through Wisconsin, Truck-DU cask, not 

during rush hour

RADTRAN output (unit risk): 5.22 × 10-10 person rem (from Table B-9)

Population density: 2,660 persons/km2

Population multiplier: 1.0

Route segment length: 19.9 km

Population (collective) dose = 5.22 × 10-10 * 2,660 * 1.0 * 19.9 = 2.76 × 10-5 person-rem

Convert to SI units: 2.76 × 10-5 person-rem * 0.01 person-Sv/person-rem = 2.76 × 10-7 person-Sv

Collective dose is best used in making comparisons, for example, in comparing the risks of 

routine transportation along different routes. All collective doses modeled are of the order of 

10-5 person-Sv (1 person-mrem) or less. The tables show that, in general, urban residents 

sustain a slightly larger dose from rail transportation than from truck transportation on the same 

State route, even though urban population densities are similar. For example, for the Maine 

urban segment of the Maine-Yankee-to-ORNL route, the collective dose differs depending on 

the transportation type used:
B-37

• The truck route urban population density is 2,706 persons/km2 (7009 persons/mi2

) and 

the collective dose is 6.6×10-8 person-Sv (6.6×10-3 person-mrem).

• The rail route urban population density is 2,527 persons/km2

, (6,545 persons/mi2), but 

the collective dose, 6.2×10-7 person-Sv (6.2×10-2 person-mrem) from the Rail-Lead cask, 

is almost 10 times larger than the dose from the Truck-DU cask, even though the 

external dose rates from the two casks are nearly the same.

Doses from rail transportation through urban areas are larger than those from truck 

transportation because train transportation was designed, and train tracks were laid, to go from 

city center to city center. Trucks carrying spent fuel, on the other hand, are required to use the 

Interstate highway system, and to use bypasses around cities where such bypasses exist. In 

the example presented, the truck traverses 5 kilometers of urban route while the train traverses 

13 urban kilometers. In addition, the average urban train speed is 24 kph (15 mph) while the 

average urban truck speed is 102 kph (63.4 mph). A truck carrying a cask through an urban 

area at about four times the speed of a train carrying a similar cask will deliver one-quarter the 

dose of the rail cask.

B.5.4.2 Doses to Occupants of Vehicles Sharing the Route

The dose to occupants of vehicles sharing a highway route (typically referred to as the on-link 

dose) consists of the sum of three components:

(1) dose to persons in vehicles traveling in the opposite direction to the shipment

(2) dose to persons in vehicles traveling in the same direction as the shipment

(3) dose to persons in passing vehicles

In the case of rail, there is a dose only to occupants of railcars (the rail analog to highway 

vehicles) traveling in the opposite direction, because passing on parallel track is rarely the case. 

RADTRAN uses Equation B-4 to calculate the dose to occupants of vehicles traveling in the 

opposite direction. The result is as follows:

(B-5) [ ] n n

v

opp f I f I

V

Qk DR

V

N PPV D * * **

2* 0  + 

 



 = γ γ

Where:

Dopp is the dose to occupants of railcars traveling in the opposite direction

N is the number of railcars sharing the route

PPV is the number of passengers per railcar

The other terms are defined as in preceding equations. The factor of 2 is included to account 

for the vehicle moving toward the radioactive cargo and then away from it. An additional factor 

of (N*PPV/V) accounts for the dose to people in the oncoming vehicle, which is assumed to be 

traveling at the same speed as the cargo. N is the number of oncoming vehicles per hour and 

P is the number of persons per vehicle.
B-38

Rail

Table B-18 provides the dose to occupants of railcars other than the railcar carrying the 

radioactive cargo and moving in the opposite direction. The vehicle occupancies used to 

calculate the table, one person on rural and suburban segments, and five people on urban 

segments, have been used historically in RADTRAN since 1988. The occupancy is consistent 

with the following considerations:

• Freight trains carry a crew of three, but all but one or two of the 60 to 120 cars on a 

freight train are unoccupied.

• Urban track carries almost all passenger rail traffic.

• Dose is calculated for one cask on a train, and rail statistics are per railcar, not per train.

The dose to occupants of other trains depends on train speed and the external dose rate from 

the spent fuel cask. Train speeds are available only for the entire United States, and not for 

each State. Therefore, Table B-18 shows the doses to occupants of trains that share the route 

with either a loaded Rail-Lead cask or a loaded Rail-Steel cask for the rural, suburban, and 

urban segments of each entire route, rather than State by State.
B-39

Table B-18 Collective Doses (Person-Sv) to Occupants of Trains Sharing the Route

SHIPMENT 

ORIGIN/

DESTINATION

Rail-Lead Cask Rail-Steel Cask

Rural Suburban Urban Total Rural Suburban Urban Total

MAINE YANKEE

ORNL 2.0 × 10-5 1.2 × 10-5 7.5 × 10-6 4.0 × 10-5 1.5 × 10-5 9.3 × 10-6 5.6 × 10-6 3.0 × 10-5

DEAF SMITH 3.8 × 10-5 1.3 × 10-5 9.7 × 10-6 6.1 × 10-5 2.9 × 10-5 1.0 × 10-5 7.4 × 10-6 4.6 × 10-5

HANFORD 6.2 × 10-5 1.7 × 10-5 1.6 × 10-5 9.0 × 10-5 4.7 × 10-5 1.3 × 10-5 1.2 × 10-5 6.8 × 10-5

SKULL VALLEY 4.8 × 10-5 1.6 × 10-5 9.6 × 10-6 7.4 × 10-5 3.6 × 10-5 1.2 × 10-5 7.3 × 10-6 5.5 × 10-5

KEWAUNEE

ORNL 1.4 × 10-5 7.0 × 10-6 5.8 × 10-6 2.7 × 10-5 1.0 × 10-5 5.3 × 10-6 4.4 × 10-6 2.0 × 10-5

DEAF SMITH 2.4 × 10-5 5.2 × 10-6 5.1 × 10-6 3.4 × 10-5 1.8 × 10-5 4.0 × 10-6 3.9 × 10-6 2.6 × 10-5

HANFORD 4.2 × 10-5 6.7 × 10-6 2.8 × 10-6 5.2 × 10-5 3.2 × 10-5 5.1 × 10-6 2.1 × 10-6 3.9 × 10-5

SKULL VALLEY 3.5 × 10-5 7.8 × 10-6 5.8 × 10-6 4.9 × 10-5 2.7 × 10-5 5.9 × 10-6 4.4 × 10-6 3.7 × 10-5

INDIAN POINT

ORNL 9.2 × 10-6 8.1 × 10-6 9.6 × 10-6 2.7 × 10-5 7.0 × 10-6 6.1 × 10-6 7.2 × 10-6 2.0 × 10-5

DEAF SMITH 3.6 × 10-5 1.1 × 10-5 9.4 × 10-6 5.6 × 10-5 2.8 × 10-5 8.2 × 10-6 7.1 × 10-6 4.3 × 10-5

HANFORD 6.0 × 10-5 1.4 × 10-5 1.1 × 10-5 8.5 × 10-5 4.6 × 10-5 1.1 × 10-5 8.0 × 10-6 6.5 × 10-5

SKULL VALLEY 4.8 × 10-5 1.3 × 10-5 1.1 × 10-5 6.5 × 10-5 3.6 × 10-5 1.0 × 10-5 8.0 × 10-6 4.9 × 10-5

INL

ORNL 4.6 × 10-5 7.1 × 10-6 3.4 × 10-6 5.7 × 10-5 3.5 × 10-5 5.4 × 10-6 2.6 × 10-6 4.3 × 10-5

DEAF SMITH 2.7 × 10-5 3.2 × 10-6 1.9 × 10-6 3.2 × 10-5 2.1 × 10-5 2.5 × 10-6 1.4 × 10-6 2.5 × 10-5

HANFORD 1.5 × 10-5 1.7 × 10-6 9.3 × 10-7 1.8 × 10-5 1.2 × 10-5 1.3 × 10-6 7.0 × 10-7 1.4 × 10-5

SKULL VALLEY 5.5 × 10-6 1.5 × 10-6 1.2 × 10-6 8.2 × 10-6 4.2 × 10-6 1.1 × 10-6 9.0 × 10-7 6.2 × 10-6

Sample calculation: Urban segment from Maine Yankee to Skull Valley, Rail-Lead cask

RADTRAN output (unit risk): 4.63 × 10-6 person-rem (from Figure B- 13)

Route urban length: 207 km (from Table B-4)

Population (collective) dose = 4.63 × 10-6

* 207= 9.58 × 10-4 person-rem

Convert to SI units: 9.58 × 10-4 person-rem * 0.01 person-Sv/person-rem = 9.58 × 10-6 person-Sv

The RADTRAN calculation incorporates the number of occupants of other trains, the train 

speed, and the railcars per hour. This value is then multiplied by the total rural, suburban, and 

urban kilometers, respectively, of the route.

Truck

Vehicle density data for large semidetached trailer trucks traveling U.S. Interstates and primary 

highways is available and well qualified. Every State records traffic counts on major (and most 

minor) highways and publishes these routinely. Researchers have used average vehicle 

density data from each of the 10 U.S. Environmental Protection Agency (EPA) regions (Weiner 

et al., 2009, Appendix D). This study used the EPA regions because they include all of the 

“lower 48” United States (Alaska and Hawaii are included in EPA Region 10 but are not 

considered in this risk assessment because no spent fuel will be shipped to or from those 

States). Table B-19 shows the 10 EPA regions and the average vehicle density data for the 

region, except for region 10, where the average vehicle density is from the three states listed.
B-40

Table B-19 States Comprising the 10 EPA Regions

Region States Included in Region

Vehicles per Hour

Rural Suburban Urban

1 Connecticut, Massachusetts, Maine, New 

Hampshire, Rhode Island, Vermont 439 726 2,129

2 New Jersey, New York 1,015 2,094 4,163

3 Delaware, Maryland, Pennsylvania, Virginia, 

West Virginia 2,056 3,655 5,748

4

Alabama, Florida, Georgia, Kentucky, 

Mississippi, North Carolina, South Carolina, 

Tennessee

1,427 2,776 5,611

5 Illinois, Indiana, Michigan, Minnesota, Ohio, 

Wisconsin 1,200 2,466 4,408

6 Arkansas, Louisiana, New Mexico, Oklahoma, 

Texas 897 1,498 3,003

7 Iowa, Kansas, Missouri, Nebraska 926 1,610 2,463

8 Colorado, Montana, North Dakota, South 

Dakota, Utah, Wyoming 795 1,958 3,708

9 Arizona, California, Nevada 1,421 3,732 7,517

10 Idaho, Oregon, Washington 1,123 2,670 5,624

The calculation of doses to occupants sharing the highway route with the radioactive materials 

truck includes the dose to vehicles passing the radioactive cargo and vehicles in an adjoining 

lane, as well as vehicles traveling in the opposite direction. Equations 28 and 34 in Neuhauser 

et al. (2000) describe this calculation.

Figure B-14 is the diagram accompanying these equations and shows the parameters used in 

the calculation. Table B-1 gives the parameter values. RADTRAN assumes there is always a 

vehicle in the position next to the truck and always a vehicle behind the truck at a distance MIN 

from the cask.
B-41

Figure B-14 Parameters for calculating doses to occupants of highway vehicles sharing 

the route with the radioactive shipment

(from Figure 3-2 of Neuhauser et al., 2000)

Tables B-20 to B-23 show the doses to individuals in vehicles sharing the highway route with 

the truck carrying a loaded Truck-DU cask. The RADTRAN calculation incorporates the number 

of occupants of other vehicles, the vehicle speed, and the vehicles per hour. This value is then 

multiplied by the rural, suburban, and urban kilometers, respectively, of each State transited.
B-42

Table B-20 Collective Doses to Persons Sharing the Route (Person-Sv) from Truck-DU; 

Shipment Origin—Maine Yankee 

DESTINATION ROUTES Rural Suburban Urban Urban 

Rush Hour Total

ORNL

Connecticut 2.0 × 10-6 9.2 × 10-6 9.2 × 10-6 8.5 × 10-6 2.9 × 10-5

Maine 2.9 × 10-6 6.7 × 10-6 1.1 × 10-6 1.0 × 10-6 1.2 × 10-5

Maryland 1.3 × 10-6 4.9 × 10-6 9.0 × 10-7 8.3 × 10-7 8.0 × 10-6

Massachusetts 1.7 × 10-6 8.7 × 10-6 3.4 × 10-6 3.2 × 10-6 1.7 × 10-5

New Hampshire 3.7 × 10-7 1.4 × 10-6 1.9 × 10-7 1.8 × 10-7 2.1 × 10-6

New Jersey 4.5 × 10-6 1.6 × 10-5 6.6 × 10-6 6.1 × 10-6 3.3 × 10-5

New York 7.5 × 10-7 2.1 × 10-6 1.3 × 10-5 1.2 × 10-5 2.7 × 10-5

Pennsylvania 3.0 × 10-5 4.8 × 10-5 7.0 × 10-6 6.5 × 10-6 9.2 × 10-5

Tennessee 1.7 × 10-5 3.2 × 10-5 4.2 × 10-6 3.9 × 10-6 5.6 × 10-5

Virginia 6.4 × 10-5 9.3 × 10-5 6.2 × 10-6 5.7 × 10-6 1.7 × 10-4

West Virginia 2.8 × 10-6 1.2 × 10-5 4.5 × 10-7 4.1 × 10-7 1.5 × 10-5

DEAF SMITH

Arkansas 3.1 × 10-5 2.1 × 10-5 2.8 × 10-6 2.6 × 10-6 5.8 × 10-5

Connecticut 2.0 × 10-6 9.2 × 10-6 9.2 × 10-6 8.5 × 10-6 2.9 × 10-5

Maine 2.9 × 10-6 6.8 × 10-6 7.3 × 10-7 6.8 × 10-7 1.1 × 10-5

Maryland 1.3 × 10-6 4.9 × 10-6 9.0 × 10-7 8.3 × 10-7 8.0 × 10-6

Massachusetts 1.7 × 10-6 8.7 × 10-6 3.4 × 10-6 3.2 × 10-6 1.7 × 10-5

New Hampshire 3.7 × 10-7 1.4 × 10-6 1.9 × 10-7 1.8 × 10-7 2.1 × 10-6

New Jersey 4.5 × 10-6 1.6 × 10-5 6.6 × 10-6 6.1 × 10-6 3.3 × 10-5

New York 7.5 × 10-7 6.8 × 10-6 6.9 × 10-6 6.4 × 10-6 2.1 × 10-5

Oklahoma 4.2 × 10-5 1.6 × 10-5 2.8 × 10-6 2.6 × 10-6 6.4 × 10-5

Pennsylvania 3.0 × 10-5 4.8 × 10-5 7.0 × 10-6 6.5 × 10-6 9.2 × 10-5

Tennessee 7.8 × 10-5 8.6 × 10-5 2.0 × 10-5 1.8 × 10-5 2.0 × 10-4

Texas 2.2 × 10-5 3.1 × 10-6 2.4 × 10-6 2.2 × 10-6 2.9 × 10-5

Virginia 6.4 × 10-5 9.3 × 10-5 6.2 × 10-6 5.7 × 10-6 1.7 × 10-4

West Virginia 2.8 × 10-6 1.2 × 10-5 4.5 × 10-7 4.1 × 10-7 1.5 × 10-5
B-43

Table B-20 Collective Doses to Persons Sharing the Route (Person-Sv) from Truck-DU; 

Shipment Origin—Maine Yankee (continued)

DESTINATION ROUTES Rural Suburban Urban Urban 

Rush Hour Total

HANFORD

Connecticut 1.7 × 10-6 8.0 × 10-6 5.1 × 10-6 4.7 × 10-6 2.0 × 10-5

Idaho 4.4 × 10-5 2.3 × 10-5 4.6 × 10-6 4.2 × 10-6 7.6 × 10-5

Illinois 2.4 × 10-5 2.0 × 10-5 5.0 × 10-6 4.6 × 10-6 5.4 × 10-5

Indiana 1.8 × 10-5 2.6 × 10-5 4.6 × 10-6 4.3 × 10-6 5.3 × 10-5

Iowa 4.0 × 10-5 1.7 × 10-5 1.4 × 10-6 1.3 × 10-6 6.0 × 10-5

Maine 2.9 × 10-6 6.8 × 10-6 7.3 × 10-7 6.8 × 10-7 1.1 × 10-5

Massachusetts 1.7 × 10-6 8.7 × 10-6 3.4 × 10-6 3.2 × 10-6 1.7 × 10-5

Nebraska 6.7 × 10-5 1.3 × 10-5 1.9 × 10-6 1.8 × 10-6 8.4 × 10-5

New Hampshire 3.7 × 10-7 1.4 × 10-6 1.9 × 10-7 1.8 × 10-7 2.1 × 10-6

New York 2.5 × 10-6 4.6 × 10-6 1.1 × 10-6 9.9 × 10-7 9.2 × 10-6

Ohio 2.8 × 10-5 6.9 × 10-5 4.0 × 10-6 3.7 × 10-6 1.0 × 10-4

Oregon 3.7 × 10-5 9.5 × 10-6 1.4 × 10-6 1.3 × 10-6 4.9 × 10-5

Pennsylvania 8.7 × 10-5 6.9 × 10-5 4.0 × 10-6 3.7 × 10-6 1.6 × 10-4

Utah 1.6 × 10-5 1.1 × 10-5 6.2 × 10-7 5.7 × 10-7 2.8 × 10-5

Washington 7.6 × 10-6 2.1 × 10-6 2.6 × 10-6 2.4 × 10-6 1.5 × 10-5

Wyoming 7.5 × 10-5 1.0 × 10-5 2.1 × 10-6 2.0 × 10-6 8.9 × 10-5

SKULL VALLEY

Connecticut 1.7 × 10-6 8.0 × 10-6 5.1 × 10-6 4.7 × 10-6 2.0 × 10-5

Illinois 2.4 × 10-5 2.0 × 10-5 5.0 × 10-6 4.6 × 10-6 5.4 × 10-5

Indiana 1.8 × 10-5 2.6 × 10-5 4.6 × 10-6 4.3 × 10-6 5.3 × 10-5

Iowa 4.0 × 10-5 1.7 × 10-5 1.4 × 10-6 1.3 × 10-6 6.0 × 10-5

Maine 2.9 × 10-6 6.8 × 10-6 7.3 × 10-7 6.8 × 10-7 1.1 × 10-5

Massachusetts 1.7 × 10-6 8.7 × 10-6 3.4 × 10-6 3.2 × 10-6 1.7 × 10-5

Nebraska 6.7 × 10-5 1.3 × 10-5 1.9 × 10-6 1.8 × 10-6 8.4 × 10-5

New Hampshire 3.7 × 10-7 4.8 × 10-6 4.8 × 10-7 4.4 × 10-6 1.0 × 10-5

New York 5.8 × 10-6 1.3 × 10-5 2.1 × 10-6 1.9 × 10-6 2.3 × 10-5

Ohio 2.8 × 10-5 4.1 × 10-5 7.3 × 10-6 6.7 × 10-6 8.3 × 10-5

Pennsylvania 8.7 × 10-5 6.9 × 10-5 4.0 × 10-6 3.7 × 10-6 1.6 × 10-4

Utah 1.8 × 10-5 8.1 × 10-6 6.1 × 10-6 5.6 × 10-6 3.8 × 10-5

Wyoming 7.5 × 10-5 1.0 × 10-5 2.1 × 10-6 2.0 × 10-6 8.9 × 10-5

Sample Calculation: Rural segment from Maine Yankee to ORNL through Connecticut (EPA Region 1)

Unit risk (From Table B-9): 4.80 × 10-8 person-Sv/km

Rural route segment length: 40.7 km

Dose to occupants of vehicles sharing the route: 4.80 × 10-8 * 40.7 = 1.95 × 10-6 person-Sv
B-44

Table B-21 Collective Doses to Persons Sharing the Route (Person-Sv) from Truck-DU; 

Shipment Origin—Indian Point

DESTINATION ROUTES Rural Suburban Urban Urban RH Total

ORNL

Maryland 1.3 × 10-6 4.9 × 10-6 9.0 × 10-7 8.3 × 10-7 7.9 × 10-6

New Jersey 4.5 × 10-6 1.6 × 10-5 6.6 × 10-6 6.1 × 10-6 3.3 × 10-5

New York 1.3 × 10-6 6.5 × 10-6 7.6 × 10-6 7.0 × 10-6 2.2 × 10-5

Pennsylvania 3.0 × 10-5 4.8 × 10-5 7.0 × 10-6 6.5 × 10-6 9.2 × 10-5

Tennessee 1.7 × 10-5 3.4 × 10-5 3.8 × 10-6 3.5 × 10-6 5.8 × 10-5

Virginia 6.4 × 10-5 9.3 × 10-5 6.2 × 10-6 5.7 × 10-6 1.7 × 10-4

West Virginia 6.4 × 10-5 1.2 × 10-5 4.5 × 10-7 4.1 × 10-7 7.7 × 10-5

DEAF SMITH

Arkansas 3.1 × 10-5 2.1 × 10-5 2.8 × 10-6 2.6 × 10-6 5.7 × 10-5

Maryland 1.3 × 10-6 4.9 × 10-6 9.0 × 10-7 8.3 × 10-7 7.9 × 10-6

New Jersey 4.5 × 10-6 1.6 × 10-5 6.6 × 10-6 6.1 × 10-6 3.3 × 10-5

New York 1.3 × 10-6 6.5 × 10-6 7.6 × 10-6 7.0 × 10-6 2.2 × 10-5

Oklahoma 4.2 × 10-5 1.6 × 10-5 2.8 × 10-6 2.6 × 10-6 6.3 × 10-5

Pennsylvania 3.0 × 10-5 4.8 × 10-5 7.0 × 10-6 6.5 × 10-6 9.2 × 10-5

Tennessee 7.8 × 10-5 8.6 × 10-5 2.0 × 10-5 1.8 × 10-5 2.0 × 10-4

Texas 2.2 × 10-5 3.1 × 10-6 2.4 × 10-6 2.2 × 10-6 3.0 × 10-5

Virginia 6.4 × 10-5 9.3 × 10-5 6.2 × 10-6 5.7 × 10-6 1.7 × 10-4

West Virginia 2.8 × 10-6 1.2 × 10-5 4.5 × 10-7 4.1 × 10-7 1.6 × 10-5

HANFORD

Idaho 4.4 × 10-5 2.3 × 10-5 4.6 × 10-6 4.2 × 10-6 7.6 × 10-5

Illinois 2.4 × 10-5 2.0 × 10-5 5.0 × 10-6 4.6 × 10-6 5.4 × 10-5

Indiana 1.8 × 10-5 2.6 × 10-5 4.6 × 10-6 4.3 × 10-6 5.3 × 10-5

Iowa 4.0 × 10-5 1.7 × 10-5 1.4 × 10-6 1.3 × 10-6 6.0 × 10-5

Nebraska 6.7 × 10-5 1.3 × 10-5 1.9 × 10-6 1.8 × 10-6 8.4 × 10-5

New Jersey 4.8 × 10-6 1.3 × 10-5 5.6 × 10-6 5.2 × 10-6 2.9 × 10-5

New York 1.3 × 10-6 6.5 × 10-6 7.6 × 10-6 7.0 × 10-6 2.2 × 10-5

Ohio 1.5 × 10-6 7.6 × 10-6 8.1 × 10-6 7.4 × 10-6 2.5 × 10-5

Oregon 3.7 × 10-5 9.5 × 10-6 1.4 × 10-6 1.3 × 10-6 4.9 × 10-5

Pennsylvania 8.0 × 10-5 5.7 × 10-5 2.2 × 10-6 2.0 × 10-6 1.4 × 10-4

Utah 1.6 × 10-5 1.1 × 10-5 6.2 × 10-7 5.7 × 10-7 2.8 × 10-5

Washington 7.6 × 10-6 2.1 × 10-6 2.6 × 10-6 2.4 × 10-6 1.5 × 10-5

Wyoming 7.5 × 10-5 1.0 × 10-5 2.1 × 10-6 2.0 × 10-6 8.9 × 10-5
Table B-21 Collective Doses to Persons Sharing the Route (Person-Sv) from Truck-DU; 

Shipment Origin—Indian Point (continued)

B-45

DESTINATION ROUTES Rural Suburban Urban Urban RH Total

SKULL VALLEY

Illinois 2.4 × 10-5 2.0 × 10-5 5.0 × 10-6 4.6 × 10-6 5.4 × 10-5

Indiana 1.8 × 10-5 2.6 × 10-5 4.6 × 10-6 4.3 × 10-6 5.3 × 10-5

Iowa 4.0 × 10-5 1.7 × 10-5 1.4 × 10-6 1.3 × 10-6 6.0 × 10-5

Nebraska 6.7 × 10-5 1.3 × 10-5 1.9 × 10-6 1.8 × 10-6 8.4 × 10-5

New Jersey 5.6 × 10-6 1.5 × 10-5 5.9 × 10-6 5.5 × 10-6 3.2 × 10-5

New York 1.5 × 10-6 7.6 × 10-6 8.1 × 10-6 7.4 × 10-6 2.5 × 10-5

Ohio 2.8 × 10-5 4.1 × 10-5 7.3 × 10-6 6.7 × 10-6 8.3 × 10-5

Pennsylvania 8.0 × 10-5 5.7 × 10-5 2.2 × 10-6 2.0 × 10-6 1.4 × 10-4

Utah 1.8 × 10-5 8.1 × 10-6 6.1 × 10-6 5.6 × 10-6 3.8 × 10-5

Wyoming 7.5 × 10-5 1.0 × 10-5 2.1 × 10-6 2.0 × 10-6 8.9 × 10-5

Sample Calculation: Urban rush hour segment from Indian Point to Hanford through Idaho (EPA Region 10)

Unit risk (From Table B-9): 5.80 × 10-7 Sv

Urban route segment length: 7.3 km

Dose to occupants of vehicles sharing the route: 5.80 × 10-7 * 7.3 = 4.23 × 10-6
B-46

Table B-22 Collective Doses to Persons Sharing the Route (Person-Sv) from Truck-DU; 

Shipment Origin—INL

DESTINATION ROUTES Rural Suburban Urban Urban Rush 

Hour Total

ORNL

Colorado 3.1 × 10-5 1.1 × 10-5 4.0 × 10-6 3.7 × 10-6 5.0 × 10-5

Idaho 2.2 × 10-5 8.0 × 10-6 1.3 × 10-6 1.2 × 10-6 3.3 × 10-5

Illinois 2.5 × 10-5 2.4 × 10-5 1.1 × 10-6 1.0 × 10-6 5.1 × 10-5

Kansas 6.2 × 10-5 1.4 × 10-5 2.7 × 10-6 2.5 × 10-6 8.1 × 10-5

Kentucky 1.8 × 10-5 1.1 × 10-5 1.2 × 10-7 1.2 × 10-7 2.9 × 10-5

Missouri 2.5 × 10-5 2.3 × 10-5 7.2 × 10-6 6.7 × 10-6 6.2 × 10-5

Tennessee 3.3 × 10-5 3.5 × 10-5 5.2 × 10-6 4.8 × 10-6 7.8 × 10-5

Utah 1.3 × 10-5 1.1 × 10-5 6.2 × 10-7 5.7 × 10-7 2.5 × 10-5

Wyoming 7.5 × 10-5 1.0 × 10-5 2.1 × 10-6 2.0 × 10-6 8.9 × 10-5

DEAF SMITH

Colorado 3.9 × 10-5 3.6 × 10-5 1.9 × 10-5 1.8 × 10-5 1.1 × 10-4

Idaho 2.2 × 10-5 8.0 × 10-6 1.3 × 10-6 1.2 × 10-6 3.3 × 10-5

New Mexico 6.4 × 10-5 9.8 × 10-6 4.8 × 10-6 4.4 × 10-6 8.3 × 10-5

Texas 7.7 × 10-6 1.7 × 10-7 0 0 7.9 × 10-6

Utah 1.3 × 10-5 1.1 × 10-5 6.2 × 10-7 5.7 × 10-7 2.5 × 10-5

Wyoming 7.0 × 10-5 7.6 × 10-6 1.5 × 10-6 1.4 × 10-6 8.1 × 10-5

HANFORD

Idaho 5.5 × 10-5 6.3 × 10-5 5.4 × 10-6 5.0 × 10-6 1.3 × 10-4

Oregon 3.7 × 10-5 2.0 × 10-5 1.4 × 10-6 1.3 × 10-6 6.0 × 10-5

Washington 7.6 × 10-6 2.1 × 10-6 2.6 × 10-6 2.4 × 10-6 1.5 × 10-5

SKULL VALLEY Idaho 2.2 × 10-5 8.0 × 10-6 1.3 × 10-6 1.2 × 10-6 4.2 × 10-5

Utah 1.5 × 10-5 1.5 × 10-5 7.2 × 10-6 6.6 × 10-6 4.4 × 10-5

Sample Calculation: Suburban segment from INL to Deaf Smith through Utah (EPA Region 8)

Unit risk (From Table B-9): 2.20 × 10-7 Sv

Suburban route segment length: 53.4 km

Dose to occupants of vehicles sharing the route: 2.15 × 10-7 * 53.4 = 1.148 × 10-5

Cells with zero values indicate no route miles of this population density.
B-47

Table B-23 Collective Doses to Persons Sharing the Route (Person-Sv) from Truck-DU; 

Shipment Origin—Kewaunee

DESTINATION ROUTES Rural Suburban Urban Urban 

Rush Hour Total

ORNL

Illinois 3.7 × 10-6 2.0 × 10-5 1.4 × 10-5 1.3 × 10-5 5.1 × 10-5

Indiana 3.3 × 10-5 3.8 × 10-5 8.3 × 10-6 7.7 × 10-6 8.7 × 10-5

Kentucky 2.7 × 10-5 4.3 × 10-5 7.2 × 10-6 6.7 × 10-6 8.4 × 10-5

Ohio 1.4 × 10-6 2.5 × 10-6 5.4 × 10-7 5.0 × 10-7 4.9 × 10-6

Tennessee 1.1 × 10-5 1.8 × 10-5 4.4 × 10-6 4.1 × 10-6 3.8 × 10-5

Wisconsin 2.0 × 10-5 2.1 × 10-5 1.3 × 10-5 1.2 × 10-5 6.6 × 10-5

DEAF SMITH

Illinois 2.0 × 10-5 1.2 × 10-5 5.9 × 10-7 5.4 × 10-7 3.3 × 10-5

Iowa 3.2 × 10-5 1.6 × 10-5 1.6 × 10-6 1.4 × 10-6 5.1 × 10-5

Kansas 2.9 × 10-5 1.2 × 10-5 3.5 × 10-6 3.2 × 10-6 4.8 × 10-5

Missouri 1.4 × 10-5 1.1 × 10-5 1.3 × 10-6 1.2 × 10-6 2.8 × 10-5

Oklahoma 3.4 × 10-5 1.1 × 10-5 2.8 × 10-6 2.6 × 10-6 5.0 × 10-5

Texas 2.2 × 10-5 3.1 × 10-6 2.4 × 10-6 2.2 × 10-6 3.0 × 10-5

Wisconsin 2.5 × 10-5 2.3 × 10-5 9.8 × 10-6 9.0 × 10-6 6.7 × 10-5

HANFORD

Idaho 9.3 × 10-6 1.1 × 10-5 3.0 × 10-6 2.8 × 10-6 2.6 × 10-5

Minnesota 5.2 × 10-5 1.3 × 10-5 5.4 × 10-7 5.0 × 10-7 6.6 × 10-5

Montana 9.6 × 10-5 3.0 × 10-5 5.4 × 10-6 5.0 × 10-6 1.4 × 10-4

South Dakota 5.3 × 10-5 1.2 × 10-5 1.0 × 10-6 9.5 × 10-7 6.7 × 10-5

Washington 4.6 × 10-5 3.0 × 10-5 1.1 × 10-5 1.0 × 10-5 9.7 × 10-5

Wisconsin 4.6 × 10-5 4.0 × 10-5 9.9 × 10-6 9.2 × 10-6 1.1 × 10-4

Wyoming 4.0 × 10-5 4.1 × 10-6 1.4 × 10-6 1.3 × 10-6 4.7 × 10-5

SKULL 

VALLEY

Illinois 2.0 × 10-5 1.2 × 10-5 5.9 × 10-7 5.4 × 10-7 3.3 × 10-5

Iowa 4.0 × 10-5 1.7 × 10-5 1.4 × 10-6 1.3 × 10-6 6.0 × 10-5

Nebraska 6.7 × 10-5 1.3 × 10-5 1.9 × 10-6 1.8 × 10-6 8.4 × 10-5

Utah 2.4 × 10-5 1.0 × 10-5 8.8 × 10-6 8.1 × 10-6 4.4 × 10-5

Wisconsin 2.5 × 10-5 2.3 × 10-5 9.8 × 10-6 9.0 × 10-6 6.7 × 10-5

Wyoming 7.5 × 10-5 1.0 × 10-5 2.1 × 10-6 2.0 × 10-6 8.9 × 10-5

Sample Calculation: Urban segment from Kewaunee to Skull Valley through Wisconsin (EPA Region 5), not 

during rush hour

Unit risk (From Table B-9): 4.90 × 10-7 Sv

Urban route segment length: 19.9 km

Dose to occupants of vehicles sharing the route: 4.90 × 10-7 * 19.9 = 9.75 × 10-6

B.5.4.3 Doses from Stopped Vehicles

Rail

Trains are stopped in classification yards at the origin and destination of the trip. The usual 

length of these classification stops is 27 hours. The collective dose to the railyard workers at 

these classification stops from the radioactive cargo is calculated internally by RADTRAN and is 

based on calculations of Wooden (1986), which the authors of this document have verified. 

This “classification yard dose” for the two rail casks studied is as follows:
B-48

• For the Rail-Lead: 1.5 × 10-5 person-Sv (1.5 person-mrem)

• For the Rail-Steel: 1.1 × 10-5 person-Sv (1.1 person-mrem)

These collective doses include doses to the train crew while the train is in the yard.

The collective dose to people living near a classification yard is calculated by multiplying the 

average dose from the rail cask to an individual living near a classification yard, as shown in 

Table B-8, by the population density between 200 and 800 meters (656 feet and 1⁄2 mile) from 

the railyard. The population density is obtained from WebTRAGIS, and the integration from 200 

to 800 meters (656 feet and 1⁄2 mile) (Table B-2) is performed by RADTRAN.

Most train stops along any route are shown in the WebTRAGIS output for that route. 

Table B-24 shows the stops on the rail route from Maine Yankee to Hanford as an example.

Table B-24 Example of Rail Stops on the Maine Yankee-to-Hanford Rail Route

Stop Reason Route type (R, S, U)a and 

State Time (hours)

Classification Initial classification S, ME 27

1 Railroad transfer (short line to ST) S, ME 4.0

2 Railroad transfer (ST to CSXT) R, NY 4.0

3 Railroad transfer (CSXT to IHB) S, IL 2.0

4 Railroad transfer (IHB to BNSF) S, IL <<1

5 Railroad transfer (BNSF to UP) S, WA <<1

Classification Final classification S, WA 27 a Determined from the WebTRAGIS output

Railyard worker collective doses can then be calculated for Stops 1 and 2 in Table B-24. 

Parameter values are from Table B-24 and the classification yard dose above:

Dose: (4/27)*(1.5 × 10-5

) = 2.2 × 10-6 person-Sv (0.22 person-mrem) for the Rail-Lead cask

Dose: (4/27)*(1.1 × 10-5

) = 1.6 × 10-6 person-Sv (0.16 person-mrem) for the Rail-Steel cask

The doses for stop 3 would be 1⁄2 of these values.

The above equations include a factor of 4/27 because the classification stop doses are 

calculated by RADTRAN for activities lasting a total of 27 hours, and the in-transit stops are for 

only 4 hours.

The average dose to an individual living 200 to 800 meters (656 feet and 1⁄2 mile) from a 

classification yard, as calculated by RADTRAN, is as follows:

• 3.5×10-7 Sv (0.035 mrem) from the Rail-Lead cask

• 2.7×10-7 Sv (0.027 mrem) from the Rail-Steel cask

Collective doses to residents near a yard (a classification yard or railroad stop) are then 

calculated from the following general expression:
B-49

Dose (person-Sv) = (population density) * (dose/h to resident near yard) * (stop time) *

(shielding factor)

Thus, for a suburban population density of 373.8 persons/km2 (968 persons/mi2

) (the suburban 

population density through Maine along the Maine Yankee-to-Hanford route) living near Stop 1 

in Table B-24, the dose can be calculated as follows:

Dose = (373.8 persons/km2

) * (3.5 × 10-7 Sv-km2/h) * (4 h) * 0.87 = 4.6×10-4 person-Sv

Table B-25 gives results for the stops.

Table B-25 Doses at Rail Stops on the Maine Yankee-to-Hanford Rail Route

Stop

Route 

type 

(R, S, U)a

and State

Time 

(hours)

Railyard worker dose 

(person-Sv)b

Residents near stop 

(person-Sv)

Rail-Lead Rail-Steel Rail-Lead Rail-Steel

Classification, 

origin S, ME 27 1.5 × 10-5 1.1 × 10-5 2.3 × 10-5 1.8 × 10-5

1 S, ME 4.0 2.16 × 10-6 1.61 × 10-6 4.6 × 10-4 3.5 × 10-4

2 R, NY 4.0 2.16 × 10-6 1.61 × 10-6 2.5 × 10-5 1.9 × 10-5

3 S, IL 2.0 1.08 × 10-6 8.05 × 10-7 2.9 × 10-4 2.2 × 10-4

Classification, 

destination S, WA 27 1.5 × 10-5 1.1 × 10-5 1.9 × 10-5 1.4 × 10-5

a Determined from the WebTRAGIS output b The yard worker dose depends only on the length of time the railcar is stopped in the yard; it is

independent of population density and shielding factor.

Truck

Doses at truck stops are calculated differently. There are two types of receptors at a truck stop, 

in addition to the truck crew—residents who live near the stop and people who share the stop 

with the refueling truck. Griego et al. (1996) conducted some time and motion studies at a 

number of truck stops. They found that the average number of people at a stop between the 

gas pumps and the nearest building was 6.9, the average distance from the fuel pump to the 

nearest building was 15 meters, and the longest refueling time for a large semidetached trailer 

truck was 0.83 hour (50 minutes). With these parameters, the collective dose to the people 

sharing the stop would be 2.3 × 10-4 person-Sv (23 person-mrem) (Table B-9). The relationship 

between the collective dose and the number of receptors is not linear in this case.

The collective dose to residents near the stop is calculated in the same way as for rail 

transportation, using data from Table B-9 plus the population density of the region around the 

stop and the stop time:

Dose (person-Sv) = (population density) * (dose/h to resident near stop) * (stop time)

Thus, for a rural population density of 15.4 persons/km2 (40 persons/mi2

) (the average along the 

Maine Yankee-to-Hanford truck route), the following dose can be calculated:

Dose/stop = (15.4 persons/km2

) * (3.3×10-8 Sv-km2/h) * (0.83 h) = 4.2×10-5 person-Sv
B-50

The population density used in the calculation is the density around the truck stop; appropriate 

residential shielding factors are used in the calculation. Unlike a train, the truck will stop several 

times on any truck route to fill the fuel tanks. Very large trucks generally carry two 80-gallon 

tanks each and stop for fuel when the tanks are half empty. A semidetached trailer truck 

carrying a Truck-DU cask can travel an average of 845 kilometers (525 miles) (U.S. Department 

of Energy, 2002) before needing to refuel. The number of refueling (and rest) stops depends on 

the length of each type of route segment. This calculation uses the following equations:

Route segment length (km)/(845 km/stop) = stops/route segment

Dose (person-Sv) = (population/km2

) * (dose to resident near stop (Sv-km2/h)) *

(stops/route segment) * (hours/stop)

Table B-26 shows the collective doses to residents near stops for the rural and suburban 

segments of the 16 truck routes in Table B-4. Trucks carrying Truck-DU casks of spent fuel are 

unlikely to stop in urban areas.

The rural and suburban population densities in Table B-26 are the averages for the entire route. 

An analogous calculation can be made for each State traversed. However, in neither case can 

one determine beforehand exactly where the truck will stop to refuel. In some cases (e.g., INL 

to Skull Valley) the truck might not stop at all because the total distance from INL to the Skull 

Valley site is only 466.2 kilometers (289.7 miles). The route from Indian Point to ORNL 

illustrates another situation. This route is 1,028 kilometers (638.8 miles) long and would thus 

include one truck stop, which could be in either a rural or a suburban area.
B-51

Table B-26 Collective Doses to Residents near Truck Stops

Origin Destination Type Persons/

km2

Average 

number 

of stopsa

Person-Sv

Residents 

near 

Stops

Persons 

Sharing 

Stops

Total

Maine 

Yankee

ORNL

Rural 19.9 1.1 7.4 × 10-7 3.9 × 10-4 3.9 × 10-4

Suburban 395 0.93 1.0 × 10-5 4.7 × 10-4 4.8 × 10-4

Deaf Smith

Rural 18.6 2.5 1.5 × 10-6 5.6 × 10-4 5.6 × 10-4

Suburban 371 1.6 1.7 × 10-5 3.6 × 10-4 3.8 × 10-4

Hanford

Rural 15.4 4.3 2.2 × 10-6 9.7 × 10-4 9.8 × 10-4

Suburban 325 1.5 1.4 × 10-5 3.4 × 10-4 3.5 × 10-4

Skull Valley

Rural 16.9 3.5 1.9 × 10-6 7.9 × 10-4 7.9 × 10-4

Suburban 332.5 1.3 1.2 × 10-5 2.9 × 10-4 3.0 × 10-4

Kewaunee

ORNL

Rural 19.8 0.81 5.2 × 10-7 1.8 × 10-4 1.8 × 10-4

Suburban 361 0.59 6.0 × 10-6 1.3 × 10-4 1.4 × 10-4

Deaf Smith

Rural 13.5 2.0 8.6 × 10-7 4.5 × 10-4 4.5 × 10-4

Suburban 339 0.52 5.0 × 10-6 1.2 × 10-4 1.2 × 10-4

Hanford

Rural 10.5 3.4 1.2 × 10-6 7.7 × 10-4 7.7 × 10-4

Suburban 316 0.60 5.4 × 10-6 1.4 × 10-4 1.4 × 10-4

Skull Valley

Rural 12.5 2.6 1.1 × 10-6 5.9 × 10-4 5.9 × 10-4

Suburban 324.5 0.44 4.1 × 10-6 9.9 × 10-5 1.0 × 10-4

Indian Point

ORNL

Rural 20.5 0.71 4.7 × 10-7 1.6 × 10-4 1.6 × 10-4

Suburban 388 0.71 7.8 × 10-6 1.6 × 10-4 1.7 × 10-4

Deaf Smith

Rural 17.1 2.3 1.3 × 10-6 5.2 × 10-4 5.2 × 10-4

Suburban 370 1.2 1.3 × 10-5 2.7 × 10-4 2.8 × 10-4

Hanford

Rural 13.0 4.1 1.8 × 10-6 9.2 × 10-4 9.2 × 10-4

Suburban 338 1.1 1.1 × 10-5 2.5 × 10-4 2.6 × 10-4

Skull Valley

Rural 14.2 3.3 1.5 × 10-6 7.4 × 10-4 7.4 × 10-4

Suburban 351 0.93 9.3 × 10-6 2.1 × 10-4 2.2 × 10-4

INL

ORNL Rural 12.4 3.1 1.3 × 10-6 7.0 × 10-4 7.0 × 10-4

Suburban 304 0.72 6.3 × 10-6 1.6 × 10-4 1.7 × 10-4

Deaf Smith Rural 7.8 2.3 5.8 × 10-7 5.2 × 10-4 5.2 × 10-4

Suburban 339 0.35 3.4 × 10-6 7.9 × 10-5 8.2 × 10-5

Hanford Rural 6.5 0.43 2.0 × 10-7 9.7 × 10-5 9.7 × 10-5

Suburban 200 0.57 9.4 × 10-7 1.3 × 10-4 1.3 × 10-4

Skull Valley Rural 10.1 0.42 1.4 × 10-7 9.5 × 10-5 9.5 × 10-5

Suburban 343 0.11 1.1 × 10-6 2.5 × 10-5 2.6 × 10-5

The number of stops is the kilometers of the route segment divided by 845 kilometers, the distance between 

stops, so it may be a fraction. Retaining the fraction allows the calculation to be repeated.
B-52

Sample Calculation: Rural Stop from Maine Yankee to ORNL

Stop dose from RADTRAN output: 3.26 × 10-6 rem = 3.26 × 10-8 Sv. This takes into account the 30- to 800-meter

width of the population band on either side of the route.

Average rural population density: 19.9 persons/km2

Total rural km = 731

Distance between truck stops: 845 km (U.S. Department of Energy, 2002)

Number of truck stops: 731/845 = 0.865

Collective dose: 19.9 * 0.865 * 3.26 × 10-8 = 5.6 × 10-7

B.5.4.4 Occupational Doses

Occupational doses from routine, incident-free radioactive materials transportation include 

doses to truck and train crew, railyard workers, inspectors, and escorts. Not included are 

workers who handle spent fuel containers in storage, loading and unloading casks from vehicles 

or during intermodal transfer, and attendants who would refuel trucks, because truck refueling 

stops in the United States no longer have such attendants.6

Table B-27 summarizes the occupational doses.

Table B-27 Occupational Doses Per Shipment from Routine, Incident-Free 

Transportation

Cask and route 

type

Train crew 

in transit:

a

3 people; 

person-Sv

Truck 

crew in 

transit:

2 people; 

person-Sv

Escort:

Sv/houra

Inspector: 

Sv per 

inspection

Truck 

stop 

worker: 

Sv per 

stop

Rail 

classification 

yard 

workers: 

person-Sv

Rail-Lead

rural/suburban 5.4 × 10-9 5.8 × 10-6 1.5 × 10-5

Rail-Lead urban 9.1 × 10-8 5.8 × 10-6

Rail-Steel 

rural/suburban 4.1 × 10-9 4.4 × 10-6 1.1 × 10-5

Rail-Steel urban 6.8 × 10-9 4.4 × 10-6

Truck-DU 

rural/suburban 3.8 × 10-7 4.9 × 10-9 1.6 × 10-4 6.7 × 10-6

Truck-DU urban 3.6 × 10-7 4.9 × 10-9

a The truck crew is shielded while in transit to sustain a maximum dose of 0.02 mSv/h

The doses to rail crew and rail escorts are similar. Spent fuel may be transported in dedicated 

trains so that both escorts and train crew are assumed to be within a railcar carrying the spent 

fuel. Escorts in the escort car are not shielded because they must maintain a line of sight to the 

railcar carrying spent fuel. Train crew members are in a crew compartment and are assumed to 

have some shielding, resulting in an estimated dose about 25 percent less than the escort. The 

largest collective doses are to railyard workers. The number of workers in railyards is not a 

 

6 The States of Oregon and New Jersey still require gas station attendants to refuel cars and light-duty 

vehicles, but heavy truck crew members do their own refueling.
B-53

constant, and the number of activities that brings these workers into proximity with the shipment 

varies as well. This analysis assumes the dose to the worker doing an activity for each activity

he or she does (e.g., inspection, coupling and decoupling the railcars, moving the railcar into 

position for coupling). The differences between doses in the Rail-Lead case and the Rail-Steel 

case reflect the differences in cask dimensions and in external dose rate.

Truck crew members are shielded so that they receive a maximum dose of 2.0×10-5 Sv/h

(2 mrem/hr). This regulatory maximum was imposed in the RADTRAN calculation. Truck 

inspectors generally spend about 1 hour within 1 meter (40 inches) of the cargo (Weiner and 

Neuhauser, 1992), resulting in a relatively large dose. An upper bound to the duration of a truck 

refueling stop is about 50 minutes (0.83 hours) (Griego et al,, 1996). The truck stop worker 

whose dose is reflected in Table B-27 is assumed to be outside (unshielded) at 15 meters 

(49 feet) from the truck during the stop. Truck stop workers who are in concrete or brick 

buildings would be shielded from any radiation.

B.6 Interpretation of Collective Dose

Collective dose is essentially the product of an average radiation dose and the number of 

people who receive that average dose. The following example—a suburban segment on a 

particular route—is typical of all routes in all States; only the specific numbers change.

The following parameters characterize a representative segment of the 

Maine-Yankee-to-Hanford truck route; the suburban segment through Illinois, shown below, is a 

representative example:

• Route segment length: 73 km (45 miles)

• Suburban population density: 324 persons/km2 (839 persons/mi2

)

• Area occupied by that population: 0.800 km * 2 * 73 = 116.8 km2 (45 mi2

)

• Total suburban population exposed to the shipment = 37,800 people

• From Table B-14, the collective radiation dose to that population, from routine, 

incident-free transportation = 6.5×10-6 person-Sv (0.65 person-mrem)

• U.S. background = 0.0036 Sv per year (4.1×10-7 Sv/h) (360 mrem/year = 0.041 mrem/hr)

• At an average speed of 108 kph (67 mph), time of population exposure = 0.675 h

The background dose sustained by each member of this population is 2.8×10-7 Sv

(0.028 mrem), for a total collective dose of 0.0105 person-Sv (1,050 person-mrem). The total 

collective dose is thus 0.0105065 person-Sv (1,050.65 person-mrem) with the shipment, and 

0.0105000 person-Sv (1,050 person-mrem) without the shipment. The collective dose from 

routine, incident-free transport is a very small increase in the collective dose the population 

continually receives from natural sources.

APPENDIX C

DETAILS OF CASK RESPONSE TO IMPACT ACCIDENTS

C-1

C.1 Introduction

For this study, the researchers performed explicit dynamic finite element calculations of the two 

spent fuel rail transportation casks described in Chapter 1 and shown in Figures 1-2 and 1-3 to 

assess their response to impact analyses. Information below provides the details of these 

analyses. In addition, the researchers summarized past explicit dynamic finite element 

analyses of the spent fuel truck transportation cask described in Chapter 1 and shown in 

Figure 1-4, and deduced the response to the same impact events and to other events based on

those analyses.

C.2 Finite Element Analysis of the Rail-Steel Cask

C.2.1 Problem Statement

Simulate the impact of a loaded Rail-Steel cask onto an unyielding surface. Consider the 

impact velocities of 48 kilometers per hour (kph) (30 miles per hour (mph)), 97 kph (60 mph), 

145 kph (90 mph), and 193 kph (120 mph). Include end, side, and center-of-gravity (CG) 

over-corner impact orientations. Based on the results, assess the integrity of the containment 

boundary and estimate the extent of any possible breach. Although the deformation and failure 

of the lid-closure bolts is of interest, the ultimate question of containment breach can be 

determined by assessing the integrity of the inner container. Predict the possible breach of the 

cask using plastic strains in the stainless steel inner container.

C.2.2 Geometric Assumptions and Mesh

A finite element model of the Rail-Steel cask was developed for use with the Sierra Mechanics 

code PRESTO. PRESTO is a nonlinear, transient dynamics finite element code developed at 

Sandia National Laboratories (Sandia) and is used extensively for weapons qualification work. 

The Rail-Steel cask model was developed and modified over several years to improve the initial 

limitations of PRESTO. Regulations required the model to include the most important geometric 

features without becoming so large that it could not be run on the available computational 

platforms. The final half-symmetric model consisted of 1.4 million solid hexahedral (hex) 

elements. The drop event lasted approximately 0.5 seconds. The simulation of this drop event 

required approximately 6 to 8 days of run time on 256 processors of a high-performance 

computer at Sandia.

An earlier version of the model used shell elements in areas of thin-walled components. The 

code had difficulty with contact between hexes and embedded shells, and the boundary 

conditions between the shells and hexes required careful and complicated consideration. 

Ultimately, the shell elements were replaced by hex elements with two or three elements 

through the thickness. Although two elements through the thickness are considered insufficient 

to correctly predict bending response, these instances were limited to components for which

bending responses were not considered important. For example, the outer shell of the impact 

limiters was modeled with two hex elements through the thickness. The purpose of this outer 

layer is to provide constraint to the aluminum honeycomb that comprises the impact limiter. The 

details of how it bends and folds away from the honeycomb are not important and not accurate 

with two elements through the thickness. Figures C-1 through C-4 show the model details. To 

allow for internal impacts, gaps were included between the fuel region and the canister and 

between the canister and the cask interior. Figure C-5 shows the location and magnitude of 

these gaps.
C-2

Figure C-1 Half-symmetric mesh of Rail-Steel cask
C-3

Figure C-2 Impact-limiter mesh

Figure C-3 Impact-limiter mesh with honeycomb removed, showing the internal support 

structure

Closure bolts were modeled with hex elements, with a minimum of four elements across the 

diameter of the bolt, as shown in Figure C-4. Any preload that would normally exist in these 

bolts because of tightening the bolts when they are installed was neglected. This assumption is 

conservative because it increases the amount of movement the closure lid would experience in 
C-4

any of the impact cases considered and maximizes any gaps that form between the closure lid 

and the cask body.

The total mass of the cask was 165,000 kilograms (kg) (weight of the cask was 

364,700 pounds (lb)). This weight is high because of an incorrect density value for the 

aluminum honeycomb that was not discovered until after the runs were completed. The 

overweight of the impact limiters results in a more severe loading environment because it 

increases the amount of kinetic energy that must be absorbed. The consequence of this 

increase is that all results are slightly conservative.

Figure C-4 Mesh of lid closure bolts and impact-limiter attachment bolts
C-5

Figure C-5 Locations and magnitudes of internal gaps in the model

The orientation of the model is important to the definition of orthotropic material properties. The 

cask model is oriented as shown in Figure C-6 and the impact direction is changed for the three 

impact conditions. For an end drop, the initial velocity is in the +z direction. For a side drop, the 

initial velocity is in the –x direction. And for a center-of-gravity (CG) over-corner drop, the initial 

velocity is in 0.38269x + 0.92388z direction.

Figure C-6 Orientation of cask model for material property definitions
C-6

C.2.3 Material Properties

These analyses placed primary importance on the response of the closure bolts. The threaded 

ends of the bolts were modeled as fixed into their mating parts using equivalent nodes. The 

remainder of the bolt was allowed to slide into its through-hole. Bolt failure was predicted by 

considering the equivalent plastic strain (EQPS) required for failure. Researchers assessed the 

value of EQPS that constitutes failure using tensile test data and references. Section C.2.4 

provides details.

The analyses assumed that the aluminum honeycomb in the impact limiters was equally strong 

in the axial and radial directions and weaker in the circumferential direction. Properties were not 

varied at 15-degree increments, as specified by the design. Instead, properties were defined in 

the global x-y-z directions and aligned with the loading direction at the point of impact. The 

honeycomb was modeled with an orthotropic crush material model. The model has been used 

for many years in PRESTO and in the commercially available finite element method code 

LS-DYNA (LSTC, 1999). It is known to behave poorly at the transition to a fully compacted 

state when the material transitions from a unidirectional compaction to an isotropic compression 

with Poisson’s expansion. For lower impact velocities (48 and 97 kph (30 and 60 mph)), this 

was not an issue. However, for the higher impact velocities, the model became unstable at 

material lockup. To allow the code to continue running, elements that were not correlating 

correctly were deleted. Because such elements had already absorbed the energy of the impact 

and were now just maintaining volume, their deletion was not considered important to the 

overall cask response.

Material properties are listed below, along with the parameters required by PRESTO (SIERRA

Solids Mechanics Team, 2009): All inputs were in English units, so those are the values listed 

first with SI units in parentheses.

C.2.3.1 Material SA350-LF3

Material SA350-LF3 low-alloy steel (Holtec, 2004) is used for top lid and cask bottom.

Density = 0.00074 lb-s2

/in4 (7.9 g/cm3)

Material model ep_power_hard

Young’s Modulus = 28.0 × 106 psi (193 × 103 MPa)

Poisson’s Ratio = 0.27

Yield Stress = 37.0 × 103 psi (255 MPa)

Hardening Constant = 192746.0 psi (1,329 MPa)

Hardening Exponent = 0.748190

Lüders Strain = 0.0

C.2.3.2 Material SA203E

Material SA203-E nickel alloy (Klamerus et al., 1996) is used for the overpack inner wall.

Density = 0.00074 lb-s2

/in4 (7.9 g/cm3)

Material model ep_power_hard

Young’s Modulus = 28.0 × 106 psi (193 × 103 MPa)

Poisson’s Ratio = 0.27

Yield Stress = 40.0 × 103 psi (276 MPa)

Hardening Constant = 192746 psi (1,329 MPa)
C-7

Hardening Exponent = 0.748190

Lüders Strain = 0.0

C.2.3.3 Material SA-516, GR70

Material SA-516, Grade 70 (Klamerus et al., 1996) is used for overpack external wall, buttress 

plates, and impact-limiter gusset plates.

Density = 0.00074 lb-s2

/in4 (7.9 g/cm3)

Material model ep_power_hard

Young’s Modulus = 29.0 × 106 psi (200 × 103 MPa)

Poisson’s Ratio = 0.3

Yield Stress = 53.097 × 103 psi (366 MPa)

Hardening Constant = 0.131331 × 106 psi (90.55 MPa)

Hardening Exponent = 0.479290

Lüders Strain = 0.00781

C.2.3.4 Material Testfoam

Material properties were taken from typical aluminum honeycomb data, as measured at Sandia 

(Hinnerichs et al., 2006). Properties used were for holtite and impact-limiter aluminum cross-ply 

honeycomb.

Density = 0.0003002 lb-s2

/in4 (3.2 g/cm3)

Material model orthotropic_crush

Young’s Modulus = 4 × 106 psi (27.6 × 103 MPa)

Poisson’s Ratio = 0.3

Yield Stress = 40000 psi (27.6 MPa)

Ex = 5.00 × 104 psi (345 MPa)

Ey = 5.00 × 104 psi (345 MPa)

Ez = 5.00 × 104 psi (345 MPa)

Gxy = 2.50 × 104 psi (172 MPa)

Gyz = 2.50 × 104 psi (172 MPa)

Gzx = 2.50 × 104 psi (172 MPa)

Vmin = 0.70

Crush xx = 2300_T

Crush yy = 2300_T

Crush zz = 2300_L

Crush xy = 2300_T

Crush yz = 2300_T 

Crush zx = 2300_T 

Function 2300_L

0 1415.384615

0.05 2123.076923 

0.1 2300

0.4 2300

0.5 1592.307692

0.6 3737.5 

0.7 20000

0.9 20000

Function 2300_T

0 1415.384615

0.05 2123.076923
C-8

0.1 2300

0.4 2300

0.5 1592.307692

0.6 3737.5

0.7 20000

0.9 20000

C.2.3.5 Material Internals

This material is used for cask contents inside the inner container.

Density = 0.00029 lb-s2

/in4 (3.1 g/cm3)

Material model orthotropic_crush

Young’s Modulus = 0.5 × 106 psi (3,450 MPa)

Poisson’s Ratio = 0.3

Yield Stress = 20,000.0 psi (138 MPa)

Ex = 0.5 × 106 psi (3,450 MPa)

Ey = 0.5 × 106 psi (3,450 MPa)

Ez = 2.2 × 106 psi (15.2 × 103 MPa)

Gxy = 0.25 × 106 psi (1,720 MPa)

Gyz = 1.1 × 106 psi (7,580 MPa

Gzx = 1.1 × 106 psi (7,580 MPa)

Vmin = 0.70

Crush xx = 2300_T

Crush yy = 700_W

Crush zz = 2300_L

Crush xy = foam_cross_1

Crush yz = foam_cross_2

Crush zx = foam_cross_1

Function foam_cross_1

0 1000

0.6 1000

0.7 10000

0.8 10000

Function foam_cross_2

0 500

0.6 500

0.7 5000

0.8 5000

C.2.3.6 Material SB637 

Material SB637-N07718 (U.S. Department of Defense, 1993) is used for lid closure bolts.

Density = 0.00074 lb-s2

/in4 (7.9 g/cm3)

Material model ml_ep_fail

Young’s Modulus = 28.6 × 106 (197 × 103 MPa)

Poisson’s Ratio = 0.3

Yield Stress = 160000 psi (1,100 MPa)

Beta = 1.0

Hardening Function = MLEP_Hardening

Young’s Modulus Function = constant_one

Poisson’s Ratio Function = constant_one
C-9

Yield Stress Function = constant_one

Critical Tearing Parameter = 0.13

Critical Crack Opening Strain = 0.01

C.2.3.7 Material 304SS

Material 304SS is used for the inner welded container, bottom impact-limiter bolts, top 

impact-limiter bolts, and shell surrounding the impact limiters (Hucek, 1986).

Density = 0.00074 lb-s2

/in4 (7.9 g/cm3)

Material model ep_power_hard

Young’s Modulus = 53.3 × 106 psi1 (367 × 103 MPa)

Poisson’s Ratio = 0.3

Yield Stress = 46.246 × 103 psi (319 MPa)

Hardening Constant = 319.05 × 103 psi (2,200 MPa)

Hardening Exponent = 0.68

Lüders Strain = 0.0

 1 The modulus of this material was artificially increased to resolve a contact chatter problem within the finite 

element model. This has very little effect on the response of the cask because this material is only used for 

thin shells that have relatively low strength and stiffness.
C-10

C.2.4 Criteria for Element Death and Bolt Failure

For all attachment bolts, element failure is defined according to PRESTO (SIERRA Solid 

Mechanics Team, 2009) convention.

Criterion is element value of EQPS > 1.12

Death on inversion = on

To account for instability in the orthotropic crush material model, elements are removed from the 

mesh if the following condition occurs, stated in the PRESTO element death convention:

• criterion is element value of solid_angle <= 0.05

• criterion is max nodal value of velocity(1) > 20,000

• criterion is max nodal value of velocity(2) > 20,000

• criterion is max nodal value of velocity(3) > 20,000

• criterion is max nodal value of velocity(1) < -20,000

• criterion is max nodal value of velocity(2) < -20,000

• criterion is max nodal value of velocity(3) < -20,000

• death on inversion = on 

The impact-limiter gusset plates and aluminum impact-limiter honeycomb are in contact within 

the impact limiter. The honeycomb would likely fail before the gusset plates in an experiment. 

Because of the homogenized material modeling of the honeycomb and the relatively coarse 

mesh, the gusset plates are significantly deformed by the honeycomb. The failure of the gusset 

plates is defined according to PRESTO convention and includes the following conditions:

• criterion is element value of time-step < -0.01

• criterion is element value of volume <= 0.0

• death on inversion = on

C.2.5 Analysis Results

Figures C-7 through C-11 depict the deformed shape of the cask following each impact

analysis.
C-11

Figure C-7 Rail-Steel cask end impact at 193 kph (120 mph)
C-12

Figure C-8 Rail-Steel cask corner impact at 48 kph (30 mph)
C-13

Figure C-9 Rail-Steel cask corner impact at 97 kph (60 mph)
C-14

Figure C-10 Rail-Steel cask corner impact at 145 kph (90 mph)

Figure C-11 Rail-Steel cask side impact at 193 kph (120 mph)
C-15

In Figures C-12 through C-23, the EQPS in the welded inner canister is shown for each analysis 

case. The same contour interval is used for each figure and was chosen so that areas that 

were near failure would show up as red and could be clearly seen. All areas that are dark blue 

have plastic strains that are much lower than the failure strain and are not of concern.

Figure C-12 Plastic strain in the interior welded canister of the Rail-Steel cask from the 

end impact at 48 kph (30 mph)
C-16

Figure C-13 Plastic strain in the interior welded canister of the Rail-Steel cask from the 

end impact at 97 kph (60 mph)

Figure C-14 Plastic strain in the interior welded canister of the Rail-Steel cask from the 

end impact at 145 kph (90 mph)
C-17

Figure C-15 Plastic strain in the interior welded canister of the Rail-Steel cask from the 

end impact at 193 kph (120 mph)

Figure C-16 Plastic strain in the interior welded canister of the Rail-Steel cask from the 

corner impact at 48 kph (30 mph)
C-18

Figure C-17 Plastic strain in the interior welded canister of the Rail-Steel cask from the 

corner impact at 97 kph (60 mph)

Figure C-18 Plastic strain in the interior welded canister of the Rail-Steel cask from the 

corner impact at 145 kph (90 mph)
C-19

Figure C-19 Plastic strain in the interior welded canister of the Rail-Steel cask from the 

corner impact at 193 kph (120 mph)

Figure C-20 Plastic strain in the interior welded canister of the Rail-Steel cask from the 

side impact at 48 kph (30 mph)
C-20

Figure C-21 Plastic strain in the interior welded canister of the Rail-Steel cask from the 

side impact at 97 kph (60 mph)

Figure C-22 Plastic strain in the interior welded canister of the Rail-Steel cask from the 

side impact at 145 kph (90 mph)

Figure C-23 Plastic strain in the interior welded canister of the Rail-Steel cask from the 

side impact at 193 kph (120 mph)
C-21

C.2.5.1 Analysis Summary

As expected, for all end, corner, and side impacts of the 48-kph (30-mph) impact analyses (the 

impact velocity from the regulatory hypothetical impact accident), the impact limiter absorbed 

almost all of the kinetic energy of the cask. No damage (permanent deformation) occurred to

the cask body or canister. As the impact velocity increases, the first effect is additional damage 

to the impact limiter (for all orientations) because it is absorbing more kinetic energy (this shows 

the margin of safety in the impact-limiter design). At 97 kph (60 mph), no significant damage to 

the cask body or canister occurred. At an impact speed of 145 kph (90 mph), damage to the 

cask and canister appears to begin. The impact limiter has absorbed all of the kinetic energy it 

can, and any additional kinetic energy is absorbed by plastic deformation in the cask body. 

Table C-1 gives the peak acceleration for each impact case. As expected, the accelerations for 

the side impacts are the highest and those for the corner impacts are the lowest.

Table C-1 Peak Acceleration from Each Analysis of the Rail-Steel Cask

Orientation Speed, kph

(mph) Peak Accel., g

End

48 (30) 71

97 (60) 115

145 (90) 212

193 (120) 276

Corner

48 (30) 66

97 (60) 86

145 (90) *

193 (120) 233

Side

48 (30) *

97 (60) *

145 (90) 355

193 (120) 472

* Data from the finite element output file for these cases was not available.

For the side impact at 145 kph (90 mph), several of the lid bolts fail in shear (criteria for the 

failure model are included in Section C.2.4 above), but the lid remains attached. At this point, 

the metallic seal no longer maintains the leaktightness of the cask, but the spent fuel remains 

contained within the welded canister. Even at the highest impact speed, 193 kph (120 mph), 

the welded canister remains intact for all orientations, so the response of the closure is of less 

importance.

C.3 Finite Element Analysis of the Rail-Lead Cask

C.3.1 Problem Statement

Simulate impact of a loaded Rail-Lead cask onto an unyielding surface. Consider impact 

velocities of 48 kph (30 mph), 97 kph (60 mph), 145 kph (90 mph), and 193 kph (120 mph). 

Include end, side, and CG over-corner impact orientations. Based on the results, assess the 

integrity of the containment boundary and estimate the extent of any possible breach. Estimate 

the deformation and failure of the lid closure bolts and any resulting gap between the lids and 

the cask. Estimate the maximum lead slump distance.
C-22

C.3.2 Geometric Assumptions and Mesh

Researchers developed a finite element model of the Rail-Lead cask for use with the Sierra 

Mechanics code PRESTO (SIERRA Solid Mechanics Team, 2009). PRESTO is a nonlinear, 

transient dynamics finite element code developed at Sandia. The finite element model was built 

primarily of hex elements. Shell elements were used for the thin stainless steel skin that wraps 

around the impact limiters. The final half-symmetric model consisted of 750,000 elements. The 

drop event lasted approximately 0.5 seconds. The simulation of this drop event required 

approximately 36 to 60 hours of run time on 64 processors of the RedSky high-performance 

computer at Sandia.

The model details are shown in Figures C-24 through C-27. Unlike the Rail-Steel cask, the 

basket in the Rail-Lead storage/transport cask completely fills the internal space of the cask. 

Gaps between the individual fuel elements and the cask lid are possible, but the probability of 

each of these fuel elements contacting the lid at the same time is very small. Thus, no gap was 

included in the model. Also, the presence of a gap could increase the force acting on the fuel 

elements, but for the severe impacts of concern in this study, such a scenario is unlikely to 

influence the overall deformation of the cask lid region because the fuel impact onto the lid 

would occur while the lid is being pushed onto the cask by the impact limiter, regardless of any 

initial gaps between the fuel and the lid.

Closure bolts were modeled with hex elements, with a minimum of four elements across the 

diameter of the bolt, as shown in Figure C-27. The model neglected any preload that would 

normally exist in these bolts as a result of the bolts being tightened during installation. This 

assumption is conservative because it increases the amount of movement the closure lid will 

experience in any of the impact cases considered and maximizes any gaps that form between 

the closure lid and the cask body.

The total mass of the cask was 112,000 kg (total weight of the cask was 247,300 lb).
C-23

Figure C-24 Half-symmetric mesh of Rail-Lead cask
C-24

Figure C-25 Impact-limiter mesh

Figure C-26 Impact-limiter mesh with wood removed
C-25

Figure C-27 Mesh of inner and outer lid closure bolts

The orientation of the model is important to the definition of orthotropic material properties. The 

cask model is oriented as shown in Figure C-28, and the impact direction is changed for the 

three impact conditions. For an end drop, the initial velocity is in the -y direction. For a side 

drop, the initial velocity is in the –x direction. And for a CG over-corner drop, the initial velocity 

is in a 0.169912x - 0.98546y direction.
C-26

Figure C-28 Orientation of cask model for material property definitions

C.3.3 Material Properties

Material properties are listed below, along with the parameters required by PRESTO (SIERRA

Solids Mechanics Team, 2009): All inputs were in English units, so those are the values listed 

first with SI units in parentheses.

C.3.3.1 Material Redwood

This material is used for top and bottom impact limiter.

Density = 5.682 × 10-5 lb-s2

/in4 (0.61 g/cm3)

Material model orthotropic_crush

Young’s Modulus = 1.5 × 106 psi (10.3 × 103 MPa)

Poisson’s Ratio = 0.3

Yield Stress = 20000 psi (138 MPa)

Vmin=0.9

Ex = 1.5 × 106 psi (10.3 × 103 MPa)

Ey = 0.3 × 106 psi (2.1 × 103 MPa)

Ez = 1.5 × 106 psi (10.3 × 103 MPa)

Gxy = 0.2 × 106 psi (1.4 × 103 MPa)

Gyz = 0.2 × 106 psi (1.4 × 103 MPa)

Gzx = 0.2 × 106 psi (1.4 × 103 MPa)

Crush xx = redwood_strong

Crush yy = redwood_weak

Crush zz = redwood_strong 

Crush xy = redwood_shear
C-27

Crush yz = redwood_shear

Crush zx = redwood_shear

Function redwood_strong

strain stress, psi (MPa)

0. 2000 (13.8)

0.14 4200 (29.0)

0.28 5100 (35.2)

0.42 5430 (37.4)

0.57 6100 (42.1)

0.71 10100 (69.6)

0.80 15000 (103)

0.90 20000 (138)

Function redwood_weak

strain stress, psi (MPa)

0. 400 (2.76)

0.14 986 (6.80)

0.28 1200 (8.27)

0.42 1275 (8.79)

0.57 1432 (9.87)

0.71 2371 (16.3)

0.80 3521 (24.3)

0.90 4690 (32.3)

Function redwood)_shear

strain stress, psi (MPa)

0.0 1000 (6.9)

0.60 1000 (6.9)

0.70 10000 (69)

0.90 10000 (69)

C.3.3.2 Material Balsa

This material is used for outer corner of top and bottom impact limiters.

Density = 1.5 × 10-5 lb-s2

/in4 (0.16 g/cm3)

Material model orthotropic_crush

Young’s Modulus = 1.5 × 106 psi (10.3 × 103 MPa)

Poisson’s ratio = 0.3

Yield Stress = 20000 psi

Vmin = 0.9

Ex = 1.5 × 106 psi (10.3 × 103 MPa)

Ey = 0.3 × 106 psi (2.1 × 103 MPa)

Ez = 1.5 × 106 psi (10.3 × 103 MPa)

Gxy = 0.2 × 106 psi (1.4 × 103 MPa)

Gyz = 0.2 × 106 psi (1.4 × 103 MPa)

Gzx = 0.2 × 106 psi (1.4 × 103 MPa)

Crush xx = balsa_strong

Crush yy = balsa_weak

Crush zz = balsa_strong

Crush xy = balsa_shear

Crush yz = balsa_shear

Crush zx = balsa_shear
C-28

Function balsa_strong

strain stress, psi (MPa)

0. 2000 (13.8)

0.14 4200 (29.0)

0.28 5100 (35.2)

0.42 5430 (37.4)

0.57 6100 (42.1)

0.71 10100 (69.6)

0.80 15000 (103)

0.90 20000 (138)

Function balsa_weak

strain stress, psi (MPa)

0. 400 (2.76)

0.14 986 (6.80)

0.28 1200 (8.27)

0.42 1275 (8.79)

0.57 1432 (9.87)

0.71 2371 (16.3)

0.80 3521 (24.3)

0.90 4690 (32.3)

Function balsa_shear

strain stress, psi (MPa)

0.0 1000 (6.9)

0.60 1000 (6.9)

0.70 10000 (69)

0.90 10000 (69)

C.3.3.3 Material 304 SS

Properties for 304 stainless steel were obtained from tensile tests conducted at Sandia.

Elastic values match the Rail-Lead safety analysis report (SAR) (NAC, 2004), but the complete 

response curve is used for plasticity.

This material is used for inner and outer cask wall, shell surrounding impact limiters, and 

impact-limiter attachment bolts.

Density = 7.48e-4 lb-s2

/in4 (8.0 g/cm3)

Material model ml_ep_fail

Young’s Modulus = 28.0 × 106 psi (193 × 103 MPa)

Poisson’s Ratio = 0.27

Yield Stress = 33.0 × 103 psi2 (228 MPa)

Beta = 1.0

Young’s Modulus Function = 304_SS_YM

Poisson’s Ratio Function = 304_SS_PR

Yield Stress Function = 304_SS_YS

Hardening Function = 304_SS_H

 2 The yield strength for this material is generally much higher than 33 kilopounds per square inch (ksi), but this 

value was used to be consistent with the value from the SAR. The actual yield strength for this material is 

generally closer to the 46 ksi used for the Rail-Steel cask analyses.
C-29

Critical Tearing Parameter = 7.779

Critical Crack Opening Strain = 0.20

Function 304_SS_H

strain stress, psi (MPa)

0.0 0. (0)

0.0395 23.4 × 103 (161 MPa)

0.0782 34.9 × 103 (241 MPa)

0.1151 45.1 × 103 (311 MPa)

0.1509 54.0 × 103 (372 MPa)

0.1857 61.7 × 103 (425 MPa)

0.2197 68.5 × 103 (472 MPa)

0.2527 74.7 × 103 (515 MPa)

0.2848 80.5 × 103 (555 MPa)

0.3165 86.0 × 103 (593 MPa)

0.3470 91.2 × 103 (629 MPa)

0.3767 96.4 × 103 (665 MPa)

0.4077 101.5 × 103 (700 MPa)

0.4378 106.4 × 103 (734 MPa)

0.4690 111.4 × 103 (768 MPa)

0.5209 119.1 × 103 (821 MPa)

0.5797 128.4 × 103 (885 MPa)

0.6595 140.6 × 103 (969 MPa)

0.7520 156.5 × 103 (1,080 MPa)

0.8639 176.3 × 103 (1,220 MPa)

1.0129 204.2 × 103 (1,410 MPa)

1.2049 242.9 × 103 (1,680 MPa)

1.4476 298.5 × 103 (2,060 MPa)

1.7499 382.8 × 103 (2,640 MPa)

2.1246 519.1 × 103 (3,580 MPa)

2.5960 754.3 × 103 (5,200 MPa)

3.1689 1161.6 × 103 (8,010 MPa)

3.7371 1624.0 × 103 (11,200 MPa)

6.0 3465.5 × 103 (23,900 MPa)

C.3.3.4 Material Filler

This material is used for internals.

Density = 2.92 × 10-4 lb-s2

/in4 (3.1 g/cm3)

Material model elastic

Young’s Modulus = 122.0 × 103 psi (841 MPa)

Poisson’s Ratio = 0.30

C.3.3.5 Material 17-4 SS

Properties for 17-4 stainless steel were obtained from tensile tests conducted at Sandia. Elastic 

values match Rail-Lead SAR (NAC, 2004), but the complete response curve is used for 

plasticity.

This material is used for outer lid and outer lid bolts.

Density = 7.48 × 10-4 lb-s2

/in4 (8.0 g/cm3)

Material model ml_ep_fail
C-30

Young’s Modulus = 28.0 × 106 psi (193 × 103 MPa)

Poisson’s Ratio = 0.28

Yield Stress = 100000. psi (689 MPa)

Beta = 1.0

Young’s Modulus Function = 304_SS_YM

Poisson’s Ratio Function = 304_SS_PR

Yield Stress Function = 304_SS_YS

Hardening Function = 17_4_SS_H

Critical Tearing Parameter = 10.0

Critical Crack Opening Strain = 0.20

Function 17_4_SS_H

strain stress psi (MPa)

0 100000.0 (689)

0.00407825 136477.69 (941)

0.00879119 153992.02 (1,060)

0.01402863 161193.41 (1,110)

0.01969711 164727.25 (1,140)

0.02677325 166808.60 (1,150)

0.03772328 168627.66 (1,160)

0.12541256 176332.05 (1,220)

0.24107482 183114.13 (1,260)

0.37338829 196318.29 (1.350)

0.51621765 212319.68 (1,460)

0.67105461 234527.78 (1,620)

0.84082846 261327.83 (1,800)

1.03088417 297249.64 (2,050)

1.24626188 344040.44 (2,370)

1.49347177 408459.72 (2,820)

1.78071924 499087.83 (3,440)

2.13871929 625460.64 (4,310)

C.3.3.6 Material SB-637

Material SB-637 Grade N07718 nickel alloy steel (NAC, 2004) is used for inner lid bolts.

Density = 7.324 × 10-4 lb-s2

/in4 (7.8 g/cm3)

Material model elastic_plastic

Young’s Modulus = 29.0 × 106 psi (200 × 103 MPa)

Poisson’s Ratio = 0.32

Yield Stress = 150.8 × 103 psi (1,040 MPa)

Hardening Modulus = 531.4 × 103 psi (3,664 MPa)

Beta = 1.0

C.3.3.7 Material Pb

Lead (Hoffman and Attaway, 1991) is used for the midcask wall.

Density = 1.06 × 10-3 lb-s2

/in4 (11.3 g/cm3)

Material model elastic_plastic

Young’s Modulus = 2.0 × 106 psi 13.8 × 103 MPa)

Poisson’s Ratio = 0.3

Yield Stress = 1700. psi (11.7 MPa)
C-31

Hardening Modulus = 2000 psi (13.8 MPa)

Beta = 1

C.3.3.8 Material NS-4-FR

A solid synthetic polymer, NS-4-FR is used for neutron shielding inserts in top and bottom lids.

The neutron-shielding material was developed by BISCO Industries, Inc., and is now supplied 

by Genden Engineering Services & Construction Company.

NS-4-FR is an epoxy resin that contains boron.

Density = 1.571 × 10-4 lb-s2

/in4 (1.7 g/cm3)

Material model elastic

Young’s Modulus = 0.561 × 105 psi (387 MPa)

Poisson’s Ratio = 0.3

C.3.4 Criteria for Element Death and Bolt Failure

To account for instability in the orthotropic crush material model, elements are removed from the 

mesh if the following condition occurs, stated in the PRESTO (SIERRA Solid Mechanics Team, 

2009) element death convention:

• criterion is max nodal value of velocity(1) > 20,000

• criterion is max nodal value of velocity(2) > 20,000

• criterion is max nodal value of velocity(3) > 20,000

• criterion is max nodal value of velocity(1) < -20,000

• criterion is max nodal value of velocity(2) < -20,000

• criterion is max nodal value of velocity(3) < -20,000

• death on inversion = on

For the impact-limiter attachment bolts, failure of elements is defined according to the PRESTO 

convention. This means that failure occurs when the critical tearing parameter (Wellman and 

Salzbrenner, 1992) is reached, as defined for 304 stainless steel:

Material criterion = ml_ep_fail

Failure of the outer lid and outer lid bolts was defined according to the PRESTO convention 

when a maximum value of EQPS was reached in 17-4 stainless steel. The PRESTO 

convention established this value of EQPS using an analysis of a tensile test specimen, and it 

defined failure at the true strain that corresponds to the true stress approximately midway 

between the true stress at maximum load and the final true stress. It chose the conservative 

value to compensate for the relatively coarse mesh in the bolt:

Criterion is element value of EQPS > 1.5

Failure of the inner lid bolts was defined according to the PRESTO convention when a 

maximum value of EQPS was reached in SB-637 Grade N07718 nickel alloy steel:

Criterion is element value of EQPS > 0.1
C-32

C.3.5 Analysis Results

Figures C-29 through C-40 depict the deformed shape of the cask following each impact 

analysis.

Figure C-29 Rail-Lead cask end impact at 48 kph (30 mph)
C-33

Figure C-30 Rail-Lead cask end impact at 97 kph (60 mph)
C-34

Figure C-31 Rail-Lead cask end impact at 145 kph (90 mph)
C-35

Figure C-32 Rail-Lead cask end impact at 193 kph (120 mph)

Figure C-33 Rail-Lead cask corner impact at 48 kph (30 mph)
C-36

Figure C-34 Rail-Lead cask corner impact at 97 kph (60 mph)

Figure C-35 Rail-Lead cask corner impact at 145 kph (90 mph)
C-37

Figure C-36 Rail-Lead cask corner impact at 193 kph (120 mph)

Figure C-37 Rail-Lead cask side impact at 48 kph (30 mph)
C-38

Figure C-38 Rail-Lead cask side impact at 97 kph (60 mph)

Figure C-39 Rail-Lead cask side impact at 145 kph (90 mph)
C-39

Figure C-40 Rail-Lead cask side impact at 193 kph (120 mph)

C.3.5.1 Analysis Summary

For the impact analyses at 48 kph (30 mph), the impact velocity from the regulatory hypothetical 

impact accident), the impact limiter absorbed almost all of the kinetic energy of the cask. No 

damage to the cask body occurred. The response of the Rail-Lead cask is more complicated 

than that of the Rail-Steel cask. Table C-2 gives the peak acceleration for each impact case. 

As expected, the accelerations for the side impacts are the highest and those for the corner 

impacts are the lowest. For the end orientation, as the impact velocity increases, there is 

initially additional damage to the impact limiter because it is absorbing more kinetic energy (this 

shows the margin of safety in the impact-limiter design). At 97 kph (60 mph), there is no 

significant damage to the cask body or canister. At an impact speed of 145 kph (90 mph), 

damage to the cask and canister appears to begin. The impact limiter has absorbed all the 

kinetic energy it can and any additional kinetic energy is absorbed by plastic deformation in the 

cask body. At this speed, significant slumping of the lead gamma shielding material occurs, 

resulting in a loss of shielding near the end of the cask away from the impact point (Chapter 5 

and Appendix E discuss this further). As the impact velocity is increased to 193 kph (120 mph), 

the lead slump becomes more pronounced and there is enough plasticity in the lids and closure 

bolts to result in a loss of sealing capability. For the directly loaded cask (without a welded 

multipurpose canister), some loss of radioactive contents could occur if the cask has metallic 

seals but not if the cask has elastomeric seals. A more detailed discussion of leakage is 

provided later in this section.

For the corner impacts at 97 and 145 kph (60 and 90 mph), there is some damage to the cask 

body, in addition to deformation of the impact limiter, which results in lead slump and closure 

bolt deformation. The amount of deformation to the closure in these two cases is not sufficient 

to cause a leak if the cask is sealed with elastomeric O-rings, but it is enough to cause a leak if 

the cask is sealed with metallic O-rings. For a corner impact at 193 kph (120 mph), there is 

more significant deformation to the cask, more lead slump, and a larger gap between the lid and 

the cask body. Figure C-36 shows the deformed shape of the cask for this impact analysis. 

The deformation in the seal region is sufficient to cause a leak if the cask has metallic O-rings 

but not if it has elastomeric O-rings. The maximum amount of lead slump is 31 centimeters

(12 inches).
C-40

Table C-2 Peak Acceleration from Each Analysis of the Rail-Lead Cask

Orientation Speed, kph (mph) Peak Accel., g

End 48 (30) 58.5

97 (60) 111.6

145 (90) 357.6

193 (120) 555.5

Corner 48 (30) 36.8

97 (60) 132.2

145 (90) 256.7

193 (120) 375.7

Side 48 (30) 76.1

97 (60) 178.1

145 (90) 411.3

193 (120) 601.1

In the side impact, as the impact velocity increases from 48 kph (30 mph) to 97 kph (60 mph), 

the impact limiter ceases to absorb additional energy and permanent deformation of the cask 

and closure bolts occurs. The resulting gap between the lids and the cask body is sufficient to 

allow leakage if there is a metallic seal but not if there is an elastomeric seal. When the impact 

speed is increased to 145 kph (90 mph), the amount of damage to the cask increases 

significantly. In this case, many of the bolts from both the inner and outer lid fail in shear, and 

there is a gap between each of the lids and the cask. This gap is sufficient to allow leakage if 

the cask is sealed with either elastomeric or metallic O-rings. Figure C-39 shows the deformed 

shape of the cask following this impact. The response of the cask to the 193-kph (120-mph) 

impact is similar to that from the 145-kph (90-mph) impact, except that the gaps between the 

lids and the cask are larger.

C.3.6 Determination of Lid Gaps

Possible gaps between the lids and the cask were extracted from the final drop results. The 

longitudinal orientation of the cask was along the y-direction, so the difference in y-direction 

displacement between the lid and the cask gave a measure of the gap. Researchers paired a 

node on the cask with the nearest node on the lid for this gap calculation. The nodes did not 

align exactly in the x-z plane. Researchers then calculated two gap values for the end drop 

orientation because the deformations were axisymmetric. For side down and CG over-corner 

orientations, researchers calculated gap values at five equally spaced locations around the 

half-circumference of the cask, as shown in Figures C-41 through C-43.
C-41

Figure C-41 Gap opening locations for end impact orientation

Figure C-42 Gap opening locations for corner impact orientation

delta 1

delta 2

delta 1

(impact corner)

delta 

2

delta 3

delta 

4

delta 

5
C-42

Figure C-43 Gap opening locations for side impact orientation.

The next set of Figures (Figures C-44 through C-53) show plots of the gap sizes as a function of 

time for the inner and outer lid for each analysis case except the side impacts at 145 kph (90 

mph) and 193 kph (120 mph). These two analyses resulted in very large gaps and the method 

used to calculate the gap was no longer accurate (at this large of a gap, changes in the hole 

size do not change the amount of material released, so accuracy is not needed). All of the gaps 

calculated are somewhat conservative because the bolts did not include any preload. Preload 

decreases the gap size because the bolts do not start to elongate until the preload is overcome. 

As an example, if the 18-cm-long (7.1-inch-long) inner lid bolts are preloaded to 50 percent of 

their yield strength (0.5 * 1,040 = 520 MPa (75.4 ksi)), the elastic elongation is 0.46 mm 

(0.018 inches). This indicates that the calculated gap for the inner lid is probably overestimated 

by this amount.

delta 1 (impact location)

delta 2

delta 3

delta 4

delta 5
C-43

Figure C-44 Gaps in the inner and outer lids of the Rail-Lead cask from the end impact at 

48 kph (30 mph)

0

0.001

0.002

0.003

0.004

0.005

0.006

0.007

0.008

0.009

0.01

0 0.005 0.01 0.015 0.02 0.025 0.03

Opening (inches)

Time (sec)

STC End 30 MPH Inner Lid Gaps

Z1

Z2

0

0.00002

0.00004

0.00006

0.00008

0.0001

0.00012

0 0.005 0.01 0.015 0.02 0.025 0.03 0.035

Opening (inches)

Time (sec)

STC End 30 MPH Outer Lid Gaps

Z1

Z2
C-44

Figure C-45 Gaps in the inner and outer lids of the Rail-Lead cask from the end impact at

97 kph (60 mph)

0

0.002

0.004

0.006

0.008

0.01

0.012

0.014

0 0.005 0.01 0.015 0.02 0.025 0.03 0.035 0.04

Opening (inches)

Time (sec)

STC End 60 MPH Inner Lid Gaps

Z1

Z2

0

0.00002

0.00004

0.00006

0.00008

0.0001

0.00012

0.00014

0 0.005 0.01 0.015 0.02 0.025 0.03 0.035 0.04

Opening (inches)

Time (sec)

STC End 60 Outer Lid Gaps

Z1

Z2
C-45

Figure C-46 Gaps in the inner and outer lids of the Rail-Lead cask from the end impact at 

145 kph (90 mph)

0

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0.08

0.09

0.1

0 0.005 0.01 0.015 0.02 0.025 0.03 0.035 0.04

Opening (inches)

Time (sec)

STC End 90 MPH Inner Lid Gaps

Z1

Z2

0

0.0002

0.0004

0.0006

0.0008

0.001

0.0012

0.0014

0.0016

0.0018

0.002

0 0.005 0.01 0.015 0.02 0.025 0.03

Opening (inches)

Time (sec)

STC End 90 MPH Outer Lid Gaps
C-46

Figure C-47 Gaps in the inner and outer lids of the Rail-Lead cask from the end impact at 

193 kph (120 mph)

0

0.05

0.1

0.15

0.2

0.25

0.3

0 0.005 0.01 0.015 0.02 0.025 0.03 0.035 0.04

Opening (inches)

Time (sec)

STC End 120 MPH Inner Lid Gaps

Z1

Z2

0

0.02

0.04

0.06

0.08

0.1

0.12

0 0.005 0.01 0.015 0.02 0.025 0.03 0.035 0.04

Opening (inches)

Time (sec)

STC End 120 MPH Outer Lid Gaps

Z1

Z2
C-47

Figure C-48 Gaps in the inner and outer lids of the Rail-Lead cask from the corner impact 

at 48 kph (30 mph)

0

0.002

0.004

0.006

0.008

0.01

0.012

0.014

0.016

0 0.01 0.02 0.03 0.04 0.05 0.06 0.07 0.08

Opening (inches)

Time (sec)

STC CG-over-Corner 30 MPH Inner Lid Gaps

Z1

Z2

Z3

Z4

Z5

0

0.001

0.002

0.003

0.004

0.005

0.006

0.007

0 0.01 0.02 0.03 0.04 0.05 0.06 0.07 0.08

Opening (inches)

Time (sec)

STC CG-over-Corner 30 MPH Outer Lid Gaps

Z1

Z2

Z3

Z4

Z5
C-48

Figure C-49 Gaps in the inner and outer lids of the Rail-Lead cask from the corner impact 

at 97 kph (60 mph)

0

0.005

0.01

0.015

0.02

0.025

0 0.005 0.01 0.015 0.02 0.025 0.03 0.035 0.04 0.045

Opening (inches)

Time (sec)

STC CG-over-Corner 60 MPH Inner Lid Gaps

Z1

Z2

Z3

Z4

Z5

0

0.002

0.004

0.006

0.008

0.01

0.012

0.014

0.016

0 0.005 0.01 0.015 0.02 0.025 0.03 0.035 0.04 0.045 0.05

Opening (inches)

Time (sec)

STC CG-over-Corner 60 MPH Outer Lid Gaps

Z1

Z2

Z3

Z4

Z5
C-49

Figure C-50 Gaps in the inner and outer lids of the Rail-Lead cask from the corner impact 

at 145 kph (90 mph)

0

0.005

0.01

0.015

0.02

0.025

0.03

0.035

0.04

0.045

0 0.005 0.01 0.015 0.02 0.025 0.03 0.035 0.04

Opening (inches)

Time (sec)

STC CG-over-Corner 90 MPH Inner Lid Gaps

Z1

Z2

Z3

Z4

Z5

0

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0 0.005 0.01 0.015 0.02 0.025 0.03 0.035 0.04

Opening (inches)

Time (sec)

STC CG-over-Corner 90 MPH Outer Lid Gaps

Z1

Z2

Z3

Z4

Z5
C-50

Figure C-51 Gaps in the inner and outer lids of the Rail-Lead cask from the corner impact 

at 193 kph (120 mph)

0

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0.08

0.09

0.1

0 0.005 0.01 0.015 0.02 0.025 0.03

Opening (inches)

Time (sec)

STC CG-over-Corner 120 MPH Inner Lid Gaps

Z1

Z2

Z3

Z4

Z5

0

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0.08

0.09

0 0.005 0.01 0.015 0.02 0.025 0.03

Opening (inches)

Time (sec)

STC CG-over-Corner 120 MPH Outer Lid Gaps

Z1

Z2

Z3

Z4

Z5
C-51

Figure C-52 Gaps in the inner and outer lids of the Rail-Lead cask from the side impact 

at 48 kph (30 mph)

0

0.002

0.004

0.006

0.008

0.01

0.012

0 0.005 0.01 0.015 0.02 0.025 0.03 0.035 0.04

Opening (inches)

Time (sec)

STC Side 30 MPH Inner Lid Gaps

Z1

Z2

Z3

Z4

Z5

0

0.001

0.002

0.003

0.004

0.005

0.006

0.007

0.008

0 0.005 0.01 0.015 0.02 0.025 0.03 0.035 0.04

Opening (inches)

Time (sec)

STC Side 30 MPH Outer Lid Gaps

Z1

Z2

Z3

Z4

Z5
C-52

Figure C-53 Gaps in the inner and outer lids of the Rail-Lead cask from the side impact 

at 97 kph (60 mph)

0

0.005

0.01

0.015

0.02

0.025

0.03

0.035

0.04

0.045

0.05

0 0.005 0.01 0.015 0.02 0.025 0.03 0.035 0.04

Opening (inches)

Time (sec)

STC Side 60 MPH Inner Lid Gaps

Z1

Z2

Z3

Z4

Z5

0

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0 0.005 0.01 0.015 0.02 0.025 0.03 0.035 0.04

Opening (inches)

Time (sec)

STC Side 60 MPH Outer Lid Gaps

Z1

Z2

Z3

Z4

Z5
C-53

To calculate any leak size based on the gaps, researchers had to take the compliance of the 

O-rings into account. The Rail-Lead cask can be sealed with either elastomeric O-rings or 

metallic O-rings. Elastomeric O-rings of the types used in transportation casks can typically 

maintain a seal when the opening between the mating surfaces opens by 2.5 millimeters

(0.10 inch). This number is used as the compliance for the cases with elastomeric O-rings. 

Unfortunately, no data are available for the specific O-rings used in this cask, and the actual 

compliance may be more or less than 2.5 millimeters. In any case, the hole sizes from the case 

with elastomeric O-rings are less than those for the cases with metallic O-rings. Metallic O-rings 

are much less tolerant to gaps, and a value of 0.25 millimeters (0.010 inch) is used as the 

compliance for the cases with metallic O-rings. For the end impact analyses, the gap size is 

uniform for the entire circumference of the seal, and the hole size is calculated by subtracting 

the compliance of the o-ring from the gap and multiplying by the circumference. If either the 

inner seal or the outer seal has a gap less than the compliance, there is no leak area. For end 

impacts, the only case in which any leakage occurs is the 193-kph (120-mph) impact with 

metallic O-rings.

For the corner and side impacts, the amount of gap varies around the circumference of the seal, 

and a more complicated algorithm is needed to calculate the hole size. As in the end impact, 

the compliance of the seal is subtracted from the gap and a trapezoidal area between 

measurement locations is assumed. In the corner impact, none of the gaps are large enough to 

overcome the compliance of elastomeric O-rings. But some leakage would occur at impacts of

97 kph (60 mph), 145 kph (90 mph), and 193 kph (120 mph) for the case in which the cask is 

sealed with metallic O-rings. The calculated hole sizes for these three cases are 65, 599, and 

1,716 square millimeters (mm2

) (0.10, 0.928. and 2.66 in2), respectively. In the side impact at 

97 kph (60 mph), the gaps are not sufficient to cause a leakage with elastomeric seals. But with 

metallic seals, a hole size of 799 mm2 (1.24 in2

) is calculated. In the 145-kph (90-mph) and 

193-kph (120-mph) analyses, a number of bolts fail and very large openings appear between 

the lids and the cask body. In these cases, both the elastomeric and metallic seals fail and the 

resulting hole size is more than 10,000 mm2

(16 in2). Table C-3 gives the final gap and hole 

sizes for each of the analyses.
C-54

Table C-3 Available Areas for Leakage from the Rail-Lead Cask

Orientation Speed

(kph)

Location Lid Gap

(mm)

Seal 

Type

Hole 

Size

(mm2)

End

48 Inner 0.226 Metal none

Outer 0 Elastomer none

97 Inner 0.056 Metal none

Outer 0.003 Elastomer none

145 Inner 2.311 Metal none

Outer 0.047 Elastomer none

193 Inner 5.588 Metal 8796

Outer 1.829 Elastomer none

Corner

48 Inner 0.094 Metal none

Outer 0.089 Elastomer none

97 Inner 0.559 Metal 65

Outer 0.381 Elastomer none

145 Inner 0.980 Metal 599

Outer 1.448 Elastomer none

193 Inner 2.464 Metal 1716

Outer 1.803 Elastomer none

Side

48 Inner 0.245 Metal none

Outer 0.191 Elastomer none

97 Inner 0.914 Metal 799

Outer 1.600 Elastomer none

145 Inner 8 Metal > 10000

Outer 25 Elastomer > 10000

193 Inner 15 Metal > 10000

Outer 50 Elastomer > 10000

C.4 Impacts against Yielding Targets

C.4.1 Introduction

The finite element results discussed in the previous section apply only to impacts against a rigid 

target. For this type of impact, the cask absorbs the entire kinetic energy of the impact. For 

finite element analyses, a rigid target is easily implemented by enforcing a no-displacement 

boundary condition at the target surface. In real life, the construction of a rigid target is 

impossible, but it is possible to construct a target that is sufficiently rigid that increasing its 

rigidity does not increase the amount of damage to the cask. This is because in real impacts 

energy absorption is shared between the cask and the target. If the target is much weaker than 

the cask, the target will absorb most of the energy. If the target is much stronger than the cask, 

the cask will absorb most of the energy. In this section, partitioning of the drop energy between 

the four generic casks and several “real-world” targets will be developed to obtain impact 

speeds onto real surfaces that give the same damage as impacts onto rigid targets. 

Researchers considered impacts onto hard desert soil, concrete highways, and hard rock. They 
C-55

did not specifically test impacts onto water surfaces, but this scenario is also discussed. In 

addition, the probability of puncture of the cask caused by impact against a nonflat surface (or 

impact by a puncture probe) is developed.

C.4.2 Method

For each finite element calculation for impact against a rigid target, the total kinetic energy of the 

finite element model is output at 100 time-steps through the analysis. The total kinetic energy is 

one-half of the sum of the mass associated with each node times the velocity of that node 

squared. Figure C-54 shows kinetic energy time-histories for the steel-lead-steel truck cask for 

each orientation from the 197-kph (120-mph) impact analyses. From the time-history of kinetic 

energy, researchers derived a velocity time-history. They calculated the rigid-body velocity for 

each time-step by assuming that all of the kinetic energy of the model is caused by velocity in 

the direction of the impact. Equation C-1 shows this mathematically:

(C-1)

Where vt is the velocity at time t, KEt is the kinetic energy at time t, mi is the mass associated 

with node i, and the summation is over all of the nodes in the finite element model.

∑ = i

t

m

2KE

t v
C-56

Figure C-54 Kinetic energy time-histories for the Rail-Lead cask from 193-kph (120-mph) 

impact analyses in the end, side, and corner orientations

For each analysis, the peak contact force is determined. Table C-4 lists these forces. For an 

impact against a real target to be as damaging to the cask as the impact against the rigid target, 

the target must be able to impart a force equal to this peak force to the cask. Because casks 

are complex structures and the rate of load application for impacts against yielding targets is 

slower than for a rigid target, it is likely the actual damage to the cask for the yielding target 

impacts would be less than that calculated using this method. The wide variety of yielding 

target types makes it impossible to quantify the amount of conservatism that results.

0.E+00

1.E+08

2.E+08

3.E+08

4.E+08

5.E+08

6.E+08

7.E+08

8.E+08

0 0.005 0.01 0.015 0.02 0.025 0.03

Kinetic Energy (in-lb)

Time (sec)

NAC-STC 193 KPH (120 MPH) Impacts

Side

End

Corner
C-57

Table C-4 Peak Contact Force for the Rail-Lead Cask Impacts against an Unyielding 

Target*

Orientation Speed

(kph)

Accel.

(G)

Contact 

Force 

(Millions of 

Pounds))

Contact 

Force

(MN)

End 48 58.5 14.6 65.0

97 111.6 27.9 123.9

145 357.6 89.3 397.1

193 555.5 138.7 616.8

Corner 48 36.8 9.2 40.9

97 132.2 33.0 146.8

145 256.7 64.1 285.1

193 375.7 93.8 417.2

Side 48 76.1 19.0 84.5

97 178.1 44.5 197.8

145 411.3 102.7 456.7

193 601.1 150.0 667.4

* Bold numbers correspond to cases in which seals might leak.

The energy absorbed by the target in developing this force is added to the initial kinetic energy 

of the cask. This total absorbed energy is used to calculate an equivalent velocity by replacing 

KEt in Equation C-1 with the total energy.

C.4.3 Soil Targets

The force that hard desert soil imparts to a cask following an impact was derived from results of 

impact tests performed by Gonzales (1987), Waddoups (1975), and Bonzon and 

Schamaun (1976). The tests by Gonzales and Waddoups used casks that were comparable to 

Rail-Lead casks, but much smaller. The tests by Bonzon and Schamaun were with casks that 

were less stiff than the Rail-Lead cask. This large amount of test data was used to develop an 

empirical soil target force-deflection equation that is a function of impactor area. Figure C-55

shows the force-deflection curves for impact of the Rail-Lead cask against a soil target. Corner 

impacts were assumed to have the same contact area on the soil target as the end impacts, so 

only two curves are shown. Similar curves were developed for each of the other casks. 

Comparison of Figure C-55 with the forces in Table C-4 shows that many of the impacts will 

result in very large soil penetrations. This is consistent with the results seen in the tests 

performed by Waddoups, where casks were dropped 610 meters (2,000 feet) from a helicopter. 

Penetration depths for these impacts were up to 2.4 meters (8 feet), and the equivalent rigid 

target impact velocity was less than 48 kph (30 mph). Integration of the force-deflection curve 

up to the peak contact force determines the amount of energy absorbed by the target.
C-58

Figure C-55 Force generated by the Rail-Lead cask penetrating hard desert soil

C.4.4 Concrete Targets

The force imparted to a cask by impact onto a concrete target is derived from test results by 

Gonzales (1987). In his series of tests, a cask-like test unit struck two types of concrete targets, 

one 12 inches thick and one 18 inches thick, at velocities from 48 to 97 kph (30 to 60 mph). All 

of the impacts were in an end-on orientation. Based on the results of these tests and 

engineering mechanics, researchers derived an empirical relationship between the force and 

energy absorbed. For impacts against concrete slab targets, two mechanisms exert large 

forces on the cask. The first is the generation of a shear plug in the concrete. The force 

required to produce this shear plug is linearly related to the impact velocity, the diameter of the 

impacting body, and the thickness of the concrete. Equation C-2 gives the empirical equation 

for the force required to produce the shear plug:

(C-2)

Where Fs is the force required to produce the shear plug, Cs is an empirical constant (16.84)

derived by curve fitting to experimental results, ve is the equivalent impact velocity, di is the 

diameter of the impacting object, and tc is the thickness of the concrete slab.

The energy absorbed in producing this shear plug is linearly related to the cask diameter, the 

square of the impact velocity, and the fourth root of the slab thickness. Equation C-3 gives the 

empirical equation for the energy required to produce the shear plug:

(C-3)

Where Es is the energy required to produce the shear plug and Ce is an empirical 

constant (0.00676) derived by curve fitting to experimental results.

After the shear plug is formed, further resistance to penetration is achieved by the behavior of 

the subgrade and soil beneath the concrete. This material is being penetrated by the cask and 

0

20

40

60

80

100

120

140

0

100

200

300

400

500

600

700

0 5 10 15

Force (Millions of Pounds)

Force (MN)

Penetration Distance (meters)

Fs = Csvedi

tc

Es = Cedive

2

tc

0.25
C-59

the shear plug. Generally, the shear plug forms with 45-degree slopes on the side. Therefore, 

the diameter of the soil being penetrated is equal to the cask diameter plus twice the slab 

thickness. The behavior of the subgrade and soil is assumed to be the same as that of the hard 

desert soil used for the soil target impacts. Figure C-56 compares force-deflection curves 

derived from these equations with those resulting from the tests performed by Gonzales.

For corner and side impacts, an equivalent diameter is calculated to fit the empirical equations. 

For each case, the diameter is calculated by assuming the shear plug forms when the concrete 

target has been penetrated 5 cm (2 inches). The area of the equivalent diameter is equal to the 

area of the concrete in contact with the cask when the penetration depth is 5 cm (2 inches). To 

calculate the equivalent velocity for concrete targets, the force required to generate the shear 

plug must be compared to the peak contact force for the impact onto the rigid target. The 

velocity required to produce this force can be calculated from Equation C-2. The kinetic energy 

associated with this velocity is absorbed by the combined effects of producing the shear plug, 

penetrating the subgrade and soil beneath the concrete, and deforming the cask. The energy 

absorbed in producing the shear plug is calculated by Equation C-3; the energy absorbed by the 

cask is equal to the kinetic energy of the rigid target impact; and the energy absorbed by the 

subgrade and soil is calculated in a manner similar to that for the soil impact discussed above. 

If the amount of energy to be absorbed by the soil is sufficiently high, the force in the soil will be 

higher than the force required to produce the shear plug. In this case, an iterative approach is 

necessary to derive an equivalent velocity so that the maximum force generated in penetrating 

the subgrade and soil beneath the concrete is equal to the peak contact force for the rigid target 

impact.

The only test data available on the orientation of impacts onto concrete targets is for end 

impacts. In this orientation, the contact area between the cask and the concrete does not 

increase with increasing penetration distance. To use the empirical relationships developed for 

end impacts with other impact orientations, an equivalent diameter must be determined. For 

both the side and corner impacts, the equivalent diameter was calculated to have an area equal 

to the area of the cask 5 mm (2 inches) above the contact point. For side impact orientations, 

this area is a rectangle. For corner impact orientations, this area is a truncated parabola. The 

shape of the contact area recognizes that the impact limiter will partially deform before it 

generates sufficient force to cause the concrete to fail and that the failure mode of the concrete 

is not a simple plug formation as it is for the end impact case.
C-60

Figure C-56 Comparison of test force-deflection curves 

with those derived from the empirical equations

C.4.5 Hard Rock Targets

For impacts against hard rock targets, the target is assumed to be a semi-infinite half plane. 

The force and energy absorbed by the target is determined by the volumetric behavior of the 

rock. For hard rock surfaces, this behavior is sufficiently stiff that the target absorbs very little 

energy. For this reason, these impacts are treated as rigid target impacts.
C-61

C.4.6 Results for Real Target Calculations

Table C-5 gives the results for impacts against soil and concrete targets.

Table C-5 Equivalent Velocities for Rail-Lead Cask Impacts against Various Targets

(in kph)

Orientation Rigid Soil Concrete

End

48 102 71

97 205 136

145 >250 >250

193 >250 >250

Corner

48 73 70

97 236 161

145 >250 >250

193 >250 >250

Side

48 103 79

97 246 185

145 >250 >250

193 >250 >250

C.4.7 Impacts against Water

Equivalent velocities for impacts against water targets for velocities greater than the regulatory 

impact are assumed to be above the range of possible impact velocities (240 kph = 150 mph). 

The incompressible nature of water makes perfectly flat impacts quite severe. As the impact 

velocity increases, smaller deviations from the perfectly flat orientation are sufficient to cause 

the lack of shear strength in water to dominate the response. Because perfectly flat impacts are 

very improbable, this approach is justified.

C.5 Response of Spent Fuel Assemblies

C.5.1 Introduction

The response of spent power reactor fuel assemblies to impact accidents is not well 

understood. While this area has been investigated in the past (Sanders et al., 1992), those 

models tended to be relatively crude and imprecise. In addition, utility companies have renewed 

their interest in shipping higher-burnup spent fuel. Therefore, it is essential to determine a more 

accurate response of spent fuel assemblies to impact loads that might be caused by 

transportation or handling accidents or malevolent acts. Sandia has performed a series of 

computational analyses to predict the structural response of a spent nuclear fuel assembly that 

is subjected to a hypothetical regulatory impact accident, as defined in Title 10 of the Code of 

Federal Regulations (10 CFR) 71.73, “Hypothetical Accident Conditions.” This study performs a 

structural analysis of a typical pressurized-water reactor (PWR) fuel assembly using the 

Abaqus/Explicit finite element analysis code. The configuration of the pellet and cladding 

interface and the material properties of the pellet have been varied in the model to account for 

possible variations in actual spent fuel assemblies.
C-62

C.5.2 Description and Method

Figure C-57 shows a typical PWR fuel assembly, which consists of a series of fuel pins, or rods, 

grouped together in a square array. The fuels rods are held in place by a series of equally 

spaced grids. Within the array of fuel tubes are a series of guide tubes in which control rods are 

placed for controlling the fission reaction during operation. The guide tubes are attached to 

endplates, nozzles, or end fittings, which provide rigidity for handling.

Figure C-57 PWR fuel assembly

Figure C-58 is a schematic representation of an individual fuel rod. This rod is constructed by 

stacking a series of uranium dioxide (UO2) pellets inside a zirconium tube, placing a spring on 

the top of the pellet stack, and welding on end caps. A plenum is added at the top of the 

assembly to provide a sufficient volume to collect released fission gases.

Spacer Grids

Control Rod

Guide Tube
C-63

Figure C-58 Fuel rod schematic drawing

The working environment of a reactor is extremely harsh. The fuel rods are subjected to 

neutron radiation, large thermal gradients, large stress caused by external water pressure, and 

large local stress from contact between the pellet and the cladding. On the first power cycle, the 

uranium pellet cracks into pie-shaped pieces caused by the large radial temperature gradients 

across the pellet. Over a short period of time (months), the pellets shrink as fine porosity in the 

fuel is removed by radiation densifications. The cladding slowly creeps down onto the pellet 

because of its high operating temperature and the external pressure of the coolant. The pellet 

also begins to expand because of fission product swelling. Over a period of 1 to 2 years, the 

initial gap between the fuel rod and the pellet is eliminated. However, the contact between the 

cladding and the fuel pellet is not necessarily circular and uniform. This leads to local increases 

in the cladding stress. In addition, zirconium is one of the few elements that react with both 

oxygen and hydrogen. This can lead to a reaction between the zirconium dioxide (ZrO2) layers 

on the inner cladding surface and the fuel pellet to form a bonding interface of (U,Zr)O2 between 

the fuel pellet and the cladding, which in essence bonds the pellet to the cladding wall. In 

addition, hydride precipitants can also form in the Zircaloy cladding wall.

On removal from the reactor, the state of the spent fuel assembly at any future time depends on 

the spent fuel’s environmental history, as well as on its condition when it was removed from the 

reactor. The internal gas pressure in a fuel rod having been removed from the reactor now 

provides tensile hoop and axial stresses on the cladding. This stress, along with changes in 

cladding temperature, may allow hydrogen to precipitate out and possibly reform along the 

circumferential directions (direction of highest stress). Plastic creep in the cladding may cause 

a gap to develop between the cladding and the fuel pellet and void spaces to develop in the 

cracked pellets. The current material conditions and stress state of any particular rod at the 

time of an accident is complex and unknown. Therefore, the current material properties and 

geometric configuration will be varied over a small range to attempt to account for the actual 

unknown material and geometric variations.

Table C-6 lists the nominal dimensions of a 17 × 17 PWR fuel assembly. Because of the large 

number of rods and the large ratio between the fuel assembly length and the fuel rod diameter, 

modeling a complete assembly using the finite element method is challenging. To build the 

entire model using continuum and structural shell elements with a high enough resolution in 

Cladding

Plenum

Fuel Pellet

Fuel - Cladding Gap

Bottom Cap

Top Cap
C-64

each fuel rod would produce a model with so many degrees of freedom as to be computationally 

intractable. Therefore, the current analysis is broken down into three steps. In the first step, the 

entire assembly is modeled using structural beam and shell elements. In the second step, the 

loads from the most highly loaded rod in the full assembly model are transferred to a single rod 

model constructed of continuum and structural shell elements. This model provides the detailed 

stress field necessary to determine the integrity of the fuel rod. Because of the severe nature of 

the reactor environments, significant material and geometric changes occur in the fuel rods. 

Very little, if any, test data is available for the Zircaloy-4 material under high irradiation 

conditions; therefore, as a third step, a series of parametric analyses were conducted with the 

continuum model to determine the sensitivity of the model to changes in the rod geometry and 

the pellet and cladding material properties.

Table C-6 Properties of Fuel Assembly

Assembly Type 17 × 17

Cladding Material Zircaloy-4

Assembly Cross-section, mm (in) 214.1-216.9 (8.43-8.54)

Number of Fuel Rods per Assembly 264

Fuel Rod OD, mm (in) 9.50 to 9.63 (0.374 to 0.379)

Minimum Cladding Thickness, mm (in) 0.58 (0.023)

Pellet Diameter, mm (in) 8.191 to 8.209 (0.3225 to 0.3232)

Maximum Active Fuel Length, m (in) 3.66 (144)

C.5.3 Finite Element Models

As described above, this analysis developed two major models. The first of these, the beam 

fuel assembly model, consists of beam and shell elements. This structural model determines 

the overall response of the fuel assembly. Using data from this model, researchers have 

developed a detailed continuum model of a single rod to determine a more detailed response of 

the most highly loaded rod. Several parametric analyses have been conducted, with the latter 

model to determine the effect of variations of rod material properties and geometry. In addition 

to these models, several smaller models have been developed to aid in the overall analysis. 

Initial models were developed to test the capabilities of the finite element codes. Researchers 

also developed small models when problems arose in the analyses. The following sections 

discuss all of these models, along with the final rod analysis.

C.5.3.1 Fuel Assembly Finite Element Model

Using the latest version of the Abaqus/Explicit finite element code, researchers constructed and 

analyzed a complete fuel assembly model (shown in Figure C-59), which incorporates 

three-dimensional beam elements for the fuel pins and control rods, shell elements for the 

spacer grid assemblies, and the support plates representing the basket walls. The endplates 

are modeled as solid plates using hexahedron elements so that the support rod beam elements 

can be attached. The model contains 265 fuel pins and 24 tie rods. There are a total of 

129,440 elements, with 41,616 beam elements. The length of each fuel rod and support rod 

has 144 beam elements. Figure C-60 presents the location of the guide tubes in the 

cross-section of the fuel assembly.
C-65

Figure C-59 Beam fuel assembly finite element model

Figure C-60 Cross-section of a 17 × 17 fuel assembly (with guide tubes in blue)

The fuel assembly model was loaded using acceleration curves developed from experimental 

data of a side impact drop test. Scientists used side loading because the fuel assemblies are 

much weaker in this loading direction and a previous study (Sanders et al., 1992) showed that 

the casks were more likely to fail from side loading than from other loading conditions. The 

full-scale data for the analysis was calculated from the 1⁄4-scale test data. Figure C-61 presents 
C-66

a plot of the full-scale data. Researchers generated an additional curve from the full-scale data 

to yield a maximum acceleration of 100 g while maintaining the same total impulse. The fuel 

rods are given an initial velocity of 13.4 meters per second (528 inches per second, 30 mph), 

which corresponds to a 9-meter (30-foot) drop test. The acceleration is applied to the lower 

plate, which represents the side of the fuel basket.

Figure C-61 Acceleration curves applied to fuel assembly beam model

The fuel rod material is modeled as unirradiated Zircaloy-4, using a power law hardening 

constitutive model fit to test data from the literature (Pierron et al., 2003). Table C-7 shows the 

calculated material parameters. These material properties are used for the fuel pins, the tie 

rods, and the support grid. This analysis models the fuel pins and tie rods as solid beams with a 

circular cross-section.

Table C-7 Zircaloy-4 Material Parameters

Elastic Modulus 89,600 MPa (13.0X103 ksi)

Yield Stress 448 MPa (65 ksi)

Lüders Strain 0.00

Hardening Constant 714 MPa (103.5 ksi)

Hardening Exponent 0.845

C.5.3.2 Fuel Assembly Model Results

For the lower acceleration curve given in Figure C-61, which represents a rail cask, there is no 

plastic deformation in the fuel rods or the spacer grids. The entire model remains elastic. For 

the analysis with the higher acceleration curve, there is no plastic deformation in the fuel rods 

and some plastic deformation in the spacer grids. Figure C-62 shows the most highly strained 

spacer grid. The lower three sections of the spacer grid buckle and a maximum plastic strain of 

28 percent is calculated.
C-67

Figure C-62 Plastic strain in a spacer grid from a 100 g impact

The contact forces from the beam fuel assembly model will be used as input to a single rod 

continuum model. Because these forces occur over very short durations during the analysis, it 

was necessary to obtain data points at each time-step in the fuel assembly model. Therefore, 

contact forces at a total of 20,349 time-steps were obtained from the fuel assembly analysis.

C.5.3.3 Beam Element Versus Solid Element Contact

In processing the contact forces from the beam fuel assembly model, researchers observed that 

the forces calculated during beam-to-beam contact were very large and acted over very short 

durations. These forces were much larger than those calculated in the model for the 

beam-to-shell contact. To investigate this difference in the magnitude and duration of the 

contact forces, researchers developed two additional models. The first, shown in Figure C-63, 

is a model of two impacting rods modeled with hexahedron elements. The second, shown in 

Figure C-64, is a model of two impacting rods modeled using beam elements. Because the 

beam elements in the beam fuel assembly model remain elastic, researchers evaluated these 

models for impact using elastic material properties.
C-68

Figure C-63 Hexahedron test model for solid rod-to-rod contact in Abaqus/Explicit

Figure C-64 Test model for beam-to-beam contact in Abaqus/Explicit

Figures C-65 and C-66 show the results from the two finite element rod models. For the same 

mass, impact velocity, and cross-sectional geometry, the two models generate two different sets 

of contact forces. As shown in Figure C-65, the beam element impact forces are much larger 

and shorter in duration than those generated from the hex rod model. The magnitudes of the 

forces differ by about a factor of 7. Researchers made an additional check comparing the 

hexahedron Abaqus/Explicit model to a similar model run in the Sandia code PRONTO 3D. 

Both codes generated similar contact and reaction forces. Continued evaluation of the two 

models generated the curves shown in Figure C-66. For the velocity range of interest, there is a 

good linear fit for each curve. Therefore, in transferring the loads between the beam fuel 

assembly model and the continuum beam model, the magnitude of the forces were scaled in 

accordance with the curves in Figure C-66. The length of each beam element impulse was 

increased to keep the integral of the curve the same. That is, the total impulse was maintained 

to conserve the change in momentum. The ratio of contact forces from this simple crossed-rod 

problem was then applied to the more complex fuel assembly analyses using the model of 

Figure C-59.
C-69

Figure C-65 Comparison of contact forces between solid rod and beam element rod

Figure C-66 Comparison of contact forces as a function of impact velocity

C.5.3.4 Continuum Rod Model

A continuum model was constructed using shell and hexahedral elements. The mesh is shown 

in Figure C-67, with a blowup of the end region showing the mesh density. The 

magenta-colored regions represent the locations of the spacer grids. A plane of symmetry 

occurs along the longitudinal axis of the beam. The symmetric model contains 

162,000 elements, with 139,000 hexahedron elements used to model the UO2 core and 

23,000 shell elements used to model the Zircaloy-4 cladding. The hexahedron core has 

16 elements across the diameter, and the semicircular arc of the cladding has 16 shell 

elements.

0

1000

2000

3000

4000

5000

6000

0 0.0005 0.001 0.0015 0.002 0.0025 0.003

Force (lbs)

Time (sec)

Contact Forces

Beam elem solid Rod

y = 17.387x + 63.234

R2 = 0.9927

y = 2.4994x + 8.7556

R2 = 0.9998

0

1000

2000

3000

4000

5000

6000

0 100 200 300 400

Force (lbs)

Velocity (in/sec)

Rod Contact Force

Beam Elements

Solid Beam

Linear (Beam Elements)

Linear (Solid Beam)
C-70

Figure C-67 Continuum rod finite element model

Researchers applied the contact forces obtained from the beam fuel assembly model for the 

100-g loading to a set of shell nodes running along the top and bottom of the symmetry plane. 

There are 1,446 nodes along each surface. Positive contact forces are applied to the bottom 

set of nodes and negative forces are applied to the upper nodes. As noted in the previous 

section, the forces from the beam fuel assembly model that result from beam-to-beam contact 

are scaled according to the curves in Figure C-66, and the duration of the load is then increased 

to conserve the change in momentum. In the region of the spacer grid, where there is 

beam-to-shell contact, the loads are not scaled. The new load curves are then interpolated from 

the element nodes in the beam fuel assembly model to a larger number of element nodes in the 

continuum model. Researchers give the rod model the same initial velocity as the beam fuel 

assembly model, 13.4 meters per second (528 inches per second).

The rod materials are also modeled using a power-law hardening model. The parameters are 

presented in Table C-8. The model was run for two different load cases, as shown in Table C-9. 

In the first case, the outside diameter of the UO2 core and the inside diameter of the cladding 

are the same; the Zircaloy-4 material is modeled as unirradiated fuel and the UO2 is also 

assumed to be pristine. In the second load case, the cladding material is assumed to be 

unirradiated, while the modulus of the UO2 is decreased by an order of magnitude to simulate a 

softer, crumbled material that has been irradiated. The results from both of these analyses are 

presented in the following section.
C-71

Table C-8 Standard Material Properties

Zircaloy Uranium Oxide

Elastic Modulus 89.6 × 103 MPa

(13.0 × 103 ksi)

193 × 103 MPa

(28.0 × 103 ksi)

Yield Stress 448 MPa (65 ksi) 149 MPa (21.6 ksi)

Lüders Strain 0.00 0.00

Hardening Constant 714 MPa (103.5 ksi) 714 MPa (103.5 ksi)

Hardening Exponent 0.845 0.845

Table C-9 Load Case Parameter Changes

Load Case parameters

Case Cladding Yield 

Strength, MPa

(psi)

UO2

Modulus, MPa

(psi)

Cladding Gap

(inches)

Case 1 448 (65,250) 193 × 103

(28 × 106)

None

Case 2 448 (65,250) 193 × 102

(28 × 105)

None

C.5.3.5 Continuum Rod Results

(a) Analysis Case 1

The first analysis case models unirradiated Zircaloy-4 material with no gap between the UO2 rod 

and the cladding. Figure C-68 presents the resulting kinetic energy plot for this analysis. 

Almost all of the kinetic energy is lost from the rod, which indicates that the load impulse applied 

in the continuum model matches the impulse generated in the beam fuel assembly model. 

There is a large decrease in the kinetic energy at approximate 5.2 milliseconds. This 

corresponds to the large loads applied to the rod caused by beam-to-beam contact forces at 

locations between the spacer grids. Figure C-69 illustrates these impacts, which show the 

maximum EQPS in the rod cladding as a function of time for three intergrid locations. 

A maximum plastic strain of 1.5 percent is observed between spacer grid locations G and H. 

Figure C-70 presents a detailed contour plot of this region.
C-72

Figure C-68 Kinetic energy for Analysis Case 1

Figure C-69 Maximum equivalent plastic strain versus time for four inter-spacer-grid 

locations

The spacer grids are specified by the letters in the legend (cf. Figure C-72).

0

200

400

600

800

1000

1200

0 0.005 0.01 0.015 0.02

Kinetic Energy (in lb)

Time (sec)

Continuum Rod

Analysis Case 1

0

0.002

0.004

0.006

0.008

0.01

0.012

0.014

0.016

0 0.01 0.02

EQPS

Time (sec)

Maximum EQPS Plots Among Interspacer 

Locations

Max EQPS, AB

Max EQPS, CD

Max EQPS, EF

Max EQPS, GH
C-73

Figure C-70 Maximum equivalent plastic strain field in cladding for Analysis Case 1

Figure C-71 presents the plastic strain in the rods at several spacer grid locations. These 

strains are approximately an order of magnitude smaller than intergrid strains. This indicates 

that the spacer grid contact is much softer than beam-to-beam contact.

Figure C-71 Maximum equivalent plastic strain versus time for three spacer grid 

locations

The spacer grids are specified by the letters in the legend (cf. Figure C-72).

EQPS

Maximum EQPS for Three Spacer Grid Locations

0

0.0005

0.001

0.0015

0.002

0.0025

0 0.005 0.01 0.015 0.02

Time (sec)

EQPS

Max EQPS, F

Max EQPS, A

Max EQPS, C
C-74

Figure C-72 shows the distribution of plastic strains along the length of the rod. The peak 

equivalent plastic strains are at the inter-rod locations between spacer grids G and H and 

between grids D and E. Strains at most spacer grid locations along the rod remains elastic. 

The maximum plastic strain in the rod at a spacer grid is 0.06 percent at spacer grid C.

Figure C-72 Schematic showing maximum equivalent plastic strain for 

spacer grid and inter-spacer-grid locations

A close examination of the strain distribution in Figure C-72 shows that they are not symmetric 

about the center of the beam, although the initial beam fuel assembly finite element model and 

its loading were symmetric. This artifice is a result of the beam contact algorithm in 

Abaqus/Explicit. As shown in Figure C-65, the impulses calculated for beam-to-beam contact

are only a few microseconds long—or roughly equal to three analysis time increments. 

Because the resolution of the impulse and the analysis time-step are of the same order of 

magnitude, any accumulative numerical error on the position of the beam element nodes may 

result in a change in the time of contact and therefore the magnitude of the contact force and 

the subsequent position and velocity of the nodes. This results in a slight asymmetry in the 

calculated beam forces in the beam fuel assembly model. These forces are subsequently 

applied to the continuum model, and the result is the asymmetry of the strain fields shown in 

Figure C-72.

(b) Analysis Case 2

For the second analysis case, the Zircaloy material properties remain the same, but the 

modulus of the UO2 is decreased by an order of magnitude to provide a probable overestimation 

of the softness in the postreactor UO2. The large cracks that develop in the fuel pellets during 

its in-core lifetime engender this softness. The largest plastic strains for this configuration are 

about one-third higher than those in the previous case of an unirradiated (pristine) UO2 core. 

The maximum EQPS is reached between spacer grids A and B and has a value of 1.98 percent. 

A contour plot of this region is presented in Figure C-73, which shows an axial region about 

2 inches long, with strain between 1 percent and 2 percent. Figure C-74 shows the maximum 

EQPS at four inter-spacer-grid locations as a function of time; Figure C-75 shows the maximum 

EQPS for four spacer grid locations. These curves are similar in shape to those in Analysis 

Case 1, in which large strains occur at 5.2 milliseconds. For this configuration, plastic strains 

appear in the rod at all but one of the spacer grid locations, and the maximum value of plastic 
C-75

strain for a spacer grid location is 0.67 percent at spacer grid C. Figure C-76 depicts a 

distribution of plastic strain over the entire rod.

Figure C-73 Maximum equivalent plastic strain field in cladding for Analysis Case 2

Figure C-74 Maximum equivalent plastic strain versus time for four inter-spacer-grid 

locations

The spacer grids are specified by the letters in the legend (cf. Figure C-76).

EQPS

0

0.005

0.01

0.015

0.02

0.025

0 0.01 0.02

EQPS

Time (sec)

Maximum EQPS at Four Locations Between Spacer 

Grids

Max EQPS, AB

Max EQPS, BC

Max EQPS, CD

Max EQPS, EF
C-76

Figure C-75 Maximum equivalent plastic strain versus time for four spacer grid locations

The spacer grids are specified by the letters in the legend (cf. Figure C-76).

Figure C-76 Schematic showing maximum equivalent plastic strain for spacer grid 

and inter-spacer-grid locations, Analysis Case 2

C.5.4 Discussion and Conclusions

In this study, the researchers conducted explicit dynamic finite element analyses of a PWR fuel 

assembly using two separate finite element models. The first model consisted of structural 

beam and shell elements and was used to determine the overall response of the complete fuel 

assembly to a regulatory side impact. Researchers applied loading data from this analysis to a 

continuum model of a single fuel pin to determine the localized stress and strain fields. They 

observed that during impact, the largest loads on the rods were generated from beam-to-beam 

contact.

Because of the lack of experimental data and the variability in properties of stored spent fuel 

rods, researchers conducted a series of analyses with variations in the stiffness of the UO2 core 

material. Table C-10 summarizes the parameters used in each analysis and the maximum 

plastic strain calculated in the cladding wall. From Table C-10, it can be concluded that an 

0

0.001

0.002

0.003

0.004

0.005

0.006

0.007

0.008

0 0.01 0.02

EQPS

Time (sec)

Maximum EQPS at Four Locations on Spacer Grids

Max EQPS, A

Max EQPS, B

Max EQPS, C

Max EQPS, D
C-77

order-of-magnitude change in the stiffness of the pellet material results in a 30-percent increase 

in the maximum plastic strain in the rod. The Case 2 maximum plastic strain is about half of the 

plastic strain to failure for the cladding of the fuel considered in this study. Thus, an 

acceleration pulse of about 200 g would be required to cause cladding failure. From Tables C-1

and C-2 it can be seen that only the impacts at 145 and 193 kph (90 and 120 mph) onto rigid 

targets generate accelerations greater than 200 g.

Table C-10 Analysis Case Summary

Case Cladding Yield 

Strength (psi)

UO2

Modulus (psi)

Cladding Gap

(inches)

Max EQPS

(%)

Case 1 65,250 28 × 106 None 1.5

Case 2 65,250 28 × 105 None 1.96

The materials in this study were modeled as isotropic and homogenous using an elastic plastic 

power-law hardening model. It is not clear that this approximation accurately models the 

response of the UO2 pellets. It is more likely that the initial response would not be a steep linear 

response as modeled, but would be nonlinear, with a soft initial reaction that would increase in 

stiffness as the pellet is squeezed. Researchers concluded that any attempt to estimate the 

nonlinear response of the pellet at this point would be pure conjecture.

APPENDIX D

DETAILS OF CASK RESPONSE TO FIRE ACCIDENTS

D-1

D.1 Introduction

For this study, researchers performed thermal analyses of Rail-Steel, Rail-Lead, and Truck-DU 

cask types to obtain the thermal response of these casks to the fire accident scenarios 

described in Chapter 4. The approach used to model these casks is similar to the ones used in 

the following:

• HI-STAR 100, NAC-STC, and GA-4 safety analysis reports (SARs) (Holtec 

International, 2000; NAC International, 2004, General Atomics, 1998)

• a combination of one-dimensional thermal resistance analysis (Incropera and 

Dewitt, 1996) 

• two- and three-dimensional finite element modeling

Thermal resistances are used to obtain effective thermal properties for several geometrically 

complex regions of the casks. These homogenized regions are then added back to the finite 

element model with the equivalent effective properties. This process eliminates some of the 

complexities of finite element discretization inherent in the models while preserving the essential 

thermal response of the casks.

For the Rail-Steel and Truck-DU casks, thermal resistance analysis reported in the Rail-Steel 

cask SAR (Holtec International, 2000) and in the Truck-DU cask SAR (General Atomic, 1998), 

respectively, are used but modified where necessary to reflect the current study. This appendix 

discusses these modifications in more detail in later sections. The approach used to model the 

Rail-Lead cask is similar to the approach used in the Rail-Lead cask SAR (NAC

International, 2004). The only exception is in how the contents of the cask are modeled. In the 

Rail-Lead cask SAR, the fuel-basket region and the rest of the overpack are modeled explicitly 

using a three-dimensional quarter section of the cask to obtain a steady-state solution. The 

maximum temperature difference between the center of the fuel-basket region and the inner 

wall of the overpack obtained in the steady-state solution is then used to calculate the 

fuel-basket cladding temperature for the regulatory uniform heating flux (see Title 10 of the

Code of Federal Regulations (10 CFR) 71.73, “Hypothetical Accident Conditions”), which did not 

include a fuel-basket region.

In this study, a three-dimensional quarter section of the fuel basket is used to obtain effective 

thermal properties for the Rail-Lead cask and fuel basket. The fuel-basket region is replaced in 

the full-scale three-dimensional finite element model using effective properties for the 

homogenized basket region. With the exception of the fuel basket region, results in the 

Rail-Lead cask SAR are used to obtain the thermal response of this cask, with minor changes to 

reflect the current study. Results taken from the Rail-Steel and Rail-Lead cask SARs and from 

the Truck-DU cask SAR are checked where possible using formulas taken directly from these 

reports or using formulas derived from independent analysis.

Some boundary conditions and material properties differ slightly from those used in the 

Rail-Steel and Rail-Lead cask SARs and in the Truck-DU cask SAR. The intent of this thermal 

analysis is to determine the temperature of critical components during and after a hypothetical 

fire accident using material properties and boundary conditions that closely resemble the 

conditions in a real fire accident. Because actual boundary conditions related to a severe fire 
D-2

are sometimes difficult to implement using available data or current analysis tools, researchers 

had to make some simplifications. For example, they assumed the insulation material used in 

the neutron shields of both casks decomposes completely when its operational temperature limit 

is reached. In such cases, researchers made conservative assumptions to maximize heat input 

to the casks, as is done in both SARs cited above. In the case of material properties, those 

presented in the SARs are preferred, followed by those in standard thermal textbooks and 

journals. For some materials, properties are available, but only over a limited temperature 

range. In such cases, the value available at the highest temperature is used for higher 

temperatures.

As mentioned in Chapter 4, MSC Software Corporation’s (MSC’s) Patran/Thermal (P/Thermal) 

(MSC Software Corporation, 2008) is the finite element heat transfer code used to solve the 

internal thermal response of the Rail-Steel and Rail-Lead casks in the regulatory uniform 

heating scenario. This scenario effectively simulates fire conditions using a spatially uniform 

radiation flux over the external surfaces of the casks as established in 10 CFR 71.73. The 

container analysis fire environment (CAFE) is the computational fluid dynamics code used to 

generate the fire environment for the CAFE regulatory and CAFE nonregulatory scenarios 

described in Chapter 4. For these scenarios, CAFE and P/Thermal are coupled together to 

obtain the thermal response of the Rail-Steel and Rail-Lead casks. CAFE generates more 

realistic fire conditions on the external surfaces of the casks, as opposed to spatially uniform 

heating conditions. P/Thermal uses CAFE-predicted external conditions to calculate the internal 

thermal response of the casks.

Researchers analyzed three fire accident scenarios different from the hypothetical accident 

condition (HAC) regulatory fire configuration for the rail casks and one fire accident scenario—

the most severe configuration of the rail cask analysis—for the truck cask with a fire lasting 

1 hour, as described in Chapter 4. These scenarios represent the accident case in which the 

fuel pool and the cask are concentric with each other (fully engulfing) or separated by one 

railcar width or one railcar length.

In the following sections, the geometry, material properties, and boundary conditions used to 

model the Rail-Steel, Rail-Lead, and Truck-DU casks are described and results are shown that 

supplement discussions in Chapter 4. The three-dimensional domain and the boundary 

conditions used in the CAFE runs are described first, followed by the geometry and boundary 

conditions used in the Rail-Steel and Rail-Lead cask analyses using the P/Thermal finite 

element models. Finally, this report presents and discusses the results from two CAFE runs 

used to benchmark the code in order to demonstrate the validity of the CAFE code for these 

types of analyses.

D.2 Finite Volume Domain and Boundary Conditions for the Container Analysis 

Fire Environment (CAFE)

CAFE (Suo-Antilla et al., 2005) uses the finite volume approach with orthogonal Cartesian 

discretization to solve (1) the three momentum equations for predicting the velocity and 

momentum field, (2) the mass continuity equation, (3) the energy equation for predicting the 

temperature field, (4) the equation of state, (5) a number of scalar transport equations for 

tracking the flow of species, and (6) two transport equations to solve thermal radiation within 

and outside the fire. CAFE uses a variable density algorithm (pressure implicit, split operator) to 

obtain a velocity field that satisfies both the momentum and continuity equations. CAFE has a 
D-3

number of turbulence models, but for this study a large eddy simulation formulation is used. 

Thermal radiation transport within and near the fires is split into two types: diffusive radiation 

inside the flame zone and clear air radiation outside the flame zone. Diffusive thermal radiation 

transport is modeled with the Rosseland approximation. Clear air radiation outside the flame 

zone is modeled using view factor methods.

CAFE is coupled to P/Thermal through a set of user-supplied subroutines that pass temperature 

and thermal heat flux data between both codes. CAFE uses a specialized scheme to map the 

temperature and heat fluxes to the exterior surfaces of the finite element model (Suo-Antilla 

et al., 2005). MSC Patran is the front-end code employed to generate the material database, 

the finite element discretization, and the boundary conditions used by P/Thermal. CAFE and 

P/Thermal are able to exchange data through a special boundary condition set up in Patran.

Figure D-1 illustrates the domain configurations used in the CAFE fire scenarios discussed in 

Chapter 4. Figure D-1(a) shows the computational fluid dynamics domain used for the CAFE 

regulatory run, and Figures D-1(b) through D-1(d) show the domain for the CAFE nonregulatory 

runs. As explained in Chapter 4, all nonregulatory CAFE runs for rail casks were determined 

based on a 113,562-liter (30,000-gallon (gal)) fuel spill. A rectangular pool is used to ensure 

that the specifications from 10 CFR 71.73 are met in the case of regulatory fires. For 

consistency, the pool remained rectangular in all other cases. The pool edges remained 

3 meters (m) (9.8 feet) away from the surface of the cask in all configurations.

The pool area is 9.25 × 13.80 m (30.35 × 45.28 feet) in the Rail-Steel cask configurations, and 

9.14 × 12.42 m (29.99 × 40.75 feet) in the Rail-Lead cask configurations. These pool areas 

correspond to a fully loaded rail tank car burning over a period of 3 hours, the maximum burn 

time based on 113.6 cubic meters (m3

) (30,000 gal) of fuel.

The pool area is 8.3 × 12 m (27.2 × 39.3 feet) in the Truck-DU cask configuration. This area 

corresponds to a fully loaded fuel tanker truck burning over a period of 1 hour, the maximum 

burn time based on 34.1 m3 (9,000 gal) of fuel. Only the scenario depicted in Figure D-1(b), the 

most severe fire scenario in the analysis of the rail casks, was analyzed for the Truck-DU cask.

An appropriate domain size is determined from del Valle et al. (2007) and del Valle (2009), in 

which thermal analyses were conducted with CAFE using a calorimeter the size of a rail cask. 

In these studies, results of CAFE runs were compared to experiments and showed good 

agreement. In the current study, the ground dimensions varied between cases because a larger 

domain is required for the cask offset cases, but they were at least 25 × 15 × 25 m 

(82 × 49 × 82 feet), about the size of the domain used in del Valle et al. (2007) and del 

Valle (2009). A mesh refinement study was conducted to assess the sensitivity of the cask 

external temperatures to mesh size and to determine an appropriate mesh size. Based on this 

study, a mesh with approximately 90,000 finite volumes was deemed acceptable for both casks. 

As observed in Figure D-1, the mesh is finer in the region near the pool. All CAFE scenarios 

used calm wind conditions; the velocity at the boundaries and inside the domain are originally 

set to zero, but are allowed to float as the fire develops.
D-4

(a) (b)

(c) (d)

Figure D-1 CAFE three-dimensional domain: (a) CAFE regulatory fire, (b) cask on 

ground and at the center of the pool, (c) cask on the ground and 3 m (10 feet) from the 

edge of the pool, and (d) cask on the ground and 18.3 m (60 feet) from the edge of the 

pool

D.3 The Rail-Steel Cask

The Rail-Steel cask is designed for transportation of a variety of spent fuel assemblies and is 

intended to fit horizontally on a railcar bed (see Figure D-2). Therefore, the Rail-Steel cask 

system is assumed to be in the horizontal position in all CAFE runs (see Figure D-1), as it would 

be after derailment if the flatbed railcar overturned or if the cask detached from a railcar after an 

accident. This thermal analysis only considers the thermally relevant components of the 

Rail-Steel cask. As stated in the introduction, this analysis uses some results reported in the 

Rail-Steel cask SAR (Holtec International, 2000). Values taken from the SAR were checked 

where possible to assess the validity of assumptions and to verify results.
D-5

Figure D-2 Rail-Steel cask transportation system

D.3.1 Geometric Consideration

The Rail-Steel cask consists of an overpack, a multipurpose canister (MPC), and two impact 

limiters; these components fit together as shown in Figure D-3. The MPC stores the nuclear 

spent fuel material, and the lid is seal welded to prevent the contents from leaking into the 

overpack cavity. The MPC is the first containment barrier in the Rail-Steel cask. The overpack 

is designed to temper both the heat and the neutron and gamma rays generated inside the 

MPC. The overpack is secured with a seal to prevent the contents from a breached canister 

from further leaking into the external environment. Thus, the overpack forms the second 

containment barrier in the Rail-Steel cask. During transportation, the overpack ends are fitted 

with impact limiters that, besides absorbing most of the impact energy during an impact, add 

another thermal insulation layer to the extreme ends of the overpack.
D-6

(a) (b)

Figure D-3 Rail-Steel cask: (a) assembly of MPC and overpack and (b) cask with limiters 

(from Holtec International, 2000)

D.3.1.1 The Overpack

The Rail-Steel overpack is a multilayered cylindrical vessel approximately 2.11 m 

(83.3 inches (in.)) in diameter and 5.16 m (203.1 in.) in length. The inner cavity of the overpack 

is approximately 1.75 m (68.8 in.) in diameter and 4.85 m (191.1 in.) in length. The inner cavity 

is formed by (1) welding a thick-walled cylinder, called the inner shell, to a metal base cup at the 

bottom and to a large diameter flange at the top and (2) bolting a closure plate onto the flange 

as shown in Figure D-4.

Five thin-walled cylinders, tightly fitted to one another and to the inner shell, form the next 

structural layer of the overpack, strengthening the overpack against puncture or penetration. 

These cylinders are jointly referred to as the intermediate shells and act as the gamma shield. 

Channels welded to the outermost intermediate shell extend radially outward and delimit the last 

layer of the overpack. These channels act as fins, enhancing conduction to the periphery of the 

overpack. Plates welded between the ends of each successive channel complete the outer 

enclosure shell of the overpack. The cavities formed between the channel walls and between 

the outermost intermediate shell and the outer enclosure plates are filled with a neutron shield 

material that provides thermal insulation, in addition to neutron attenuation. The outermost 

intermediate shell, the neutron shield region, and the outer enclosure shell effectively extend the 

diameter of the overpack an additional 32.3 centimeters (cm) (12.7 in.) beyond the perimeter of 

the flange and the metal base cup.
D-7

(a) (b)

Figure D-4 Rail-Steel cask overpack: (a) cross-sectional view through the center of the 

cask and (b) cross-sectional view through the midplane of the overpack

(from Holtec International, 2000)

The overpack shells, metal base cup, flange, closure plate, and neutron shield region are the 

major components of the overpack; together they comprise most of its volume. The thermal 

modeling explicitly represents the overpack shells, metal base cup, flange, and closure plate, 

with minor alterations to simplify the solid modeling and meshing process. The most significant 

change extends the length of the overpack shells to the length of the outer enclosure shell. This 

change impacts the regions outlined in red in Figure D-4(a). As observed, the regions affected 

are where the overpack shells meet the metal plate cup and flange. Note that these length 

changes are more pronounced near the inner shell and gradually diminish radially outward. At 

most, the total length in question is less than 10 percent of the total length of the outer shell. 

The materials used for the metal plate cup and flange (cryogenic steel) and the overpack shell 

(carbon steel) have nearly the same thermal properties. There are contact gap regions between 

shells that are not present in the metal plate cup and flange. Therefore, these changes are 

expected to have some effect on the overall thermal response of the overpack, but only in the 

radial direction and limited to the region in question. The intermediate shells and the neutron 

shield region are each represented as a single volume to minimize geometric complexity; 

however, their thermal properties are accounted for in the thermal model using the techniques 

described in Sections D.3.3.3 and D.3.3.4.

The overpack contains additional components used to service the overpack during normal 

operations or designed to function only during abnormal ambient conditions, such as fires. 

These features include seals, gas ports, rupture disks, and lifting and pocket trunnions, as 
D-8

shown in Figure D-4. The model does not include these components because their effects are 

assumed to be either (1) negligible because of their small volume and mass relative to the other 

components in the overpack; (2) highly localized, with no effect on the overall thermal 

performance of the cask at locations of interest; or (3) both.

D.3.1.2 Multipurpose Canister

The MPC is a cylindrical vessel approximately 1.73 m (68.3 in.) in diameter (outside) and 

4.83 m (190.3 in.) in length. The MPC is made from a cylindrical shell 1.2 cm (0.5 in.) thick and 

4.76 m (187.4 in.) in length, a circular baseplate 6.35 cm (2.5 in.) thick, and a circular plate lid 

24.1 cm (9.5 in.) thick (see Figure D-5a). The baseplate is welded to the bottom of the MPC 

shell, and this shell is in turn welded to the exterior surface of the lid. At the top, the MPC shell 

is flush against a large groove on the end perimeter of the circular plate lid. An annular closure 

ring welded to both the groove and the top of the shell seals the contents of the MPC. In the 

horizontal position, the shell and the base plate rest on the inner shell of the overpack. Drain 

and vent ports on the MPC lid are used to evacuate and fill the MPC with an inert gas (generally 

helium). With the exception of the closure ring and drain ports, all of these components are 

modeled explicitly. The closure ring is assumed to be part of the lid.

(a) (b)

Figure D-5 Rail-Steel cask MPC: (a) cross-sectional view through the axis of the cask

and (b) cross-sectional view through the midplane of the overpack

(from Holtec International, 2000)

The spent nuclear fuel (SNF) is stored (or SNF assemblies are stored) in a fuel basket inside 

the MPC (see Figure D-5b). The fuel basket is made by welding a series of perpendicular and 

parallel plates to form an array of storage cells. Each storage cell contains a single fuel 

assembly. The Rail-Steel cask is designed to carry four general types of MPCs: (1) the 

MPC-24/-24E/-24EF, which contains a maximum of 24 pressurized-water reactor (PWR) fuel 

assemblies, (2) the MPC-32, which contains a maximum of 32 PWR fuel assemblies, (3) the 
D-9

MPC-68/-68F, which contains a maximum of 68 boiling-water reactor (BWR) assemblies, and 

(4) the MPC-HB, which contains a maximum of 80 Humboldt Bay BWR assemblies. These 

MPC types are similar in design; however, the MPC-24 is designed to carry a greater specific 

heat load and the highest total heat load. For this reason, attention is focused on the MPC-24. 

In the MPC-24, the fuel cells are physically separated from one another by a gas pocket called 

the flux trap. The length of the fuel basket is approximately 4.48 m (176.5 in.). The fuel 

assembly might not reach this length; in such cases, spacers are installed on the baseplate and 

on the MPC lid to hold the fuel assemblies in place (see Figure D-5a).

A single fuel assembly consists of an array of fuel rods, each separated by a gas space (when 

the MPC is backfilled), as shown in Figure D-6a. The total number of rods per assembly varies 

with fuel assembly design. Each fuel rod consists of a number of cylindrical fuel pellets fitted 

into a thin-walled pipe, called the fuel cladding. The fuel cladding’s inner diameter is slightly 

larger than the diameter of the pellets, as shown in Figure D-6b. The fuel pellets are held tightly 

against each other using the force of a spring. The radial dimensions of the rod components 

vary between fuel rod designs. In general, the length of the fuel column is only a fraction of the 

total length of the fuel rod and marks the active fuel region. The total length of the fuel rod is 

approximately the same as the length of the fuel assembly. Additional supports are added to 

the ends of the fuel assembly and at regular intervals along the length of the assembly for 

structural integrity to maintain spacing between the rods, as well as for handling purposes.

The Rail-Steel cask system is designed to carry a number of PWR fuel rods. It is impractical to 

analyze the Rail-Steel cask system with all of these fuel rod designs. Similarly, it is impractical 

to model the MPC contents with all the components described above because (1) the wide 

range of component length scales creates additional meshing complexities and (2) alternative 

methods have been employed in the SAR literature and in this study to obtain equivalent 

thermal properties for the MPC internal contents with good results (see Section D.3.4). Hence, 

the Rail-Steel cask model does not explicitly represent the fuel-basket region, which includes 

the fuel assembly, basket walls, and flux trap gaps.

The MPC shell contains support structures that help keep the fuel basket laterally in place, as 

well as lift lugs used during loading and unloading operations. Some slots between the 

periphery of the fuel basket and the MPC shell wall contain thin-walled heat conduction 

elements. These conduction elements extend the full length of the basket and provide an 

effective heat conduction path between the MPC basket and MPC shell. With the exception of 

the heat conduction elements, all other structural elements in the fuel-basket periphery region 

are ignored for the same reason cited in Section D.3.1.1. The fuel heat conduction elements 

are not represented explicitly, but their thermal effect is included through the use of a simplified 

analytical model explained in the Rail-Steel cask SAR.
D-10

To simplify the modeled geometry, the fuel-basket region and fuel-basket periphery region are 

modeled as two concentric cylindrical regions extending the length of the fuel assembly (see 

Figure D-7).

(a)

(b)

Figure D-6 (a) Fuel assembly (from Holtec International, 2000) and (b) fuel rod

Cladding

Plenum

Fuel Pellet

Fuel - Cladding Gap

Bottom Cap

Top Cap
D-11

(a) (b)

Figure D-7 Fuel-basket region (left) and equivalent fuel-basket region (right)

(from Holtec International, 2000)

The diameter of the equivalent fuel-basket region (Zone 1) is calculated by using the hydraulic 

diameter of the fuel-basket periphery region (Zone 2). The hydraulic diameter takes into 

account the perimeter of the fuel basket, MPC support structures, MPC inner shell wall, and the 

area of the basket-to-shell gap—the total surface area between the perimeter of the fuel basket, 

MPC support structures, and MPC inner wall—through which heat transfer occurs. For the 

MPC-24 basket, the hydraulic diameter calculated using this method is approximately 12.7 cm 

(5 in.). The hydraulic diameter is also equal to the inner diameter of the MPC shell minus the 

inner diameter of the equivalent fuel-basket cylinder region. This provides a way to obtain the 

equivalent fuel-basket cylinder diameter and periphery annulus gap length (Zone 2). The 

periphery annulus gap length obtained from the hydraulic diameter calculation approximates the 

effective gap length through which heat is transferred between the irregular fuel-basket 

perimeter, the MPC support structures, and the MPC inner shell wall in the actual cask.

D.3.1.3 Rail-Steel Cask Impact Limiters

The impact limiters are relatively low-density cylindrical components that are not only designed 

to absorb energy during impact but also to serve as insulators during fires in the uncrushed 

state. The main body of the impact limiter has a maximum diameter of 3.25 m (128 in.) and a 

maximum length of 1.52 m (60 in.) (see Figure D-8).
D-12

(a) (b)

Figure D-8 (a) Upper and (b) lower impact limiters in a Rail-Steel cask

(from Holtec International, 2000)

Most of the impact limiter is honeycomb material enclosed in a thin-shelled metal wall. The 

honeycomb material and outer shell walls are supported in the interior of the limiter by one

small- and one large-diameter cylinder. Both of these cylinders are concentric but offset axially 

from each other. As observed in Figure D-8, the small cylinder has uniform thickness, but the 

large cylinder has several wall thicknesses to accommodate the ends of the overpack. The 

small cylinder has a cover plate at one end and is welded to a large circular plate on the other 

end. The large plate is in turn welded to the large cylinder at its perimeter. The space inside 

this small cylinder is filled with air. Triangular plates welded to the large- and small-diameter 

cylinders complete the support structure inside the impact limiter.

Each impact limiter contains a circular segment of neutron shielding, 6.35 cm (2.5 in.) thick, 

attached to the large plate on one side. This neutron shield provides axial neutron attenuation 

and forms a heat barrier between the impact limiter body and the overpack ends.

The impact limiters are bolted to the ends of the overpack. The upper limiter bolts to the sides 

of the flange through the protruding section of the large cylinder, and the lower limiter bolts to 

the overpack metal base cup through the cover plate, neutron shield, and large plate, as shown 

in Figure D-8(b).

The impact limiters are assumed to stay intact after the hypothetical accident scenarios 

described in Chapter 4. This assumption is reasonable because the height of the flatbed railcar 

is approximately that of the diameter of the overpack. This height precludes any significant 

damage to the impact limiter during an accident scenario involving, for example, the overturn of 

the railcar flatbed. Because the limiters are assumed to stay intact, they are modeled in their 

original shape. The neutron shield material in the limiters is retained, but only up to the surface 

temperature at which the material starts to degrade (see Section D.3.3.4). Only the 
D-13

large-diameter thick-walled cylinder is explicitly modeled because it serves as a direct 

conduction path from the outside to the inside of the limiter.

D.3.2 Rail-Steel Cask Thermal Behavior and Model Assumptions

The MPC-24 is designed to carry a maximum heat load of 20 kilowatts (kW) (0.833 kW per fuel 

assembly). This heat generation rate is nonuniform along the length of the active fuel region. 

Table D-1 shows the normalized, axial heat generation rate distribution for a typical Rail-Steel 

cask PWR assembly (Holtec International, 2000). This table is used to calculate the heat 

generation rate through the active length of the basket (i.e., in the axial direction). The 

Rail-Steel cask system is designed to reject heat passively to the environment under normal 

conditions of transport. Thus, heat is dissipated from the fuel rods to the exterior surfaces of the 

cask by a combination of conduction, convection, and radiation heat-transfer modes.

Table D-1 Axial Burnup Profile in the Active Fuel Region of the Rail-Steel Cask

Axial Distance from Bottom of Active Fuel 

(Percent of Active Fuel Length)

Normalized 

Value

0.000% to 4.167% 0.548

4.167% to 8.333% 0.847

8.333% to 16.667% 1.077

16.667% to 33.333% 1.105

33.333% to 50.000% 1.098

50.000% to 66.667% 1.079

66.667% to 83.333% 1.050

83.333% to 91.667% 0.960

91.667% to 95.833% 0.734

95.833% to 100.000% 0.467

For normal transportation conditions, the internal temperature is higher than the external 

temperature of the cask; therefore, heat will be dissipated outwardly starting from the fuel rods. 

Inside the fuel rods, heat is transferred outward by (1) conduction through the gas space 

between rods and (2) radiation exchange between different fuel rods and between the fuel rods 

and the walls of the basket. Convection is assumed to be negligible in this region because

radiation effects dominate at high temperatures. Heat is then dissipated by conduction through 

the gas space in the flux traps and by radiation between the basket walls. Convection is also 

assumed to be negligible in the flux trap region. In the fuel-basket periphery, heat is dissipated 

to the MPC shell by (1) conduction through the heat conduction elements and the gas and 

(2) radiation between the walls of the fuel basket and the MPC and between the inner walls of 

the heat conduction elements. In this region, natural convection loops enhance heat transfer 

between the inner walls of the heat conduction elements. The results of the Rail-Steel cask 

SAR take this effect into account. A two-dimensional finite element model is used to determine 

the heat convection coefficient for this region of the basket.
D-14

Heat transfer from the MPC shell to the overpack inner shell occurs through an MPC-overpack 

nonconcentric gap. In the horizontal position, the MPC makes contact with the overpack at the 

bottom. This contact gap is approximately 0.5 mm (0.02 in.) across. In this region, heat is also 

dissipated by conduction through the variable gas-filled gap and by radiation between the outer 

and inner walls of the MPC and overpack respectively. The Rail-Steel cask SAR used a

two-dimensional analytical model to obtain an effective conductivity through the variable-length 

gap, as discussed in Section D.3.3.2. Heat transfer through the inner and intermediate shells 

occurs by conduction through the shell material and through the contact gaps between the 

shells. These contact gaps are assumed to be five microns (2 × 10-4 in.) across as in the 

Rail-Steel cask SAR. Conduction in the neutron shield region occurs in parallel through the 

radial connectors and the neutron shield material. A simple thermal resistor network is used to 

calculate the effective thermal conductivity through intermediate shells and through the neutron 

shield region (see Sections D.3.3.3 and D.3.3.4).

The Rail-Steel cask system is designed to keep the temperature of components below their 

operational temperature limits1 for normal conditions of transport and for a 30-minute, fully

engulfing regulatory fire and subsequent cooldown period (10 CFR 71.73). For longer fully

engulfing fires, such as the ones depicted in this study, a significant amount of heat might be 

transferred to the interior of the Rail-Steel cask, raising the temperature of some of its 

components above their operational temperature limits. This is expected to occur in the neutron 

shield region. The operational temperature limit of the neutron shield insulation is 

149 degrees Celsius (C) (300 degrees Fahrenheit (F)). In the model for this study, the neutron 

shield material is assumed to decompose completely shortly after it reaches this temperature 

limit, immediately triggering thermal radiation exchange between the overpack enclosure shell 

and the outermost intermediate shell. The assumption used in this study is a significant 

departure from the assumption made in the SAR. Section D.3.3 will discuss this point further. 

As with the neutron shield, the aluminum honeycomb is expected to reach temperatures beyond 

the operational temperature limits. However, the honeycomb material is not expected to 

completely melt. Given the results in Pierce et al. (2003), the regression rate of the honeycomb 

material is expected to be minimal over a 3-hour period and have only a local effect.

Heat dissipation through the cross-section (i.e., in the axial direction) of the MPC and overpack 

and through the limiters is assumed to occur mostly by conduction. Heat conduction occurs in 

parallel through each of the materials that comprise this cross-section. Thermal radiation in the 

axial direction is possible; however, because view factors tend to diminish with the square of the 

distance and angle of view and the temperature gradients are weak along the axis compared to 

the radial direction (as observed in contour results presented in Chapter 4), these effects are 

neglected in the basket region. Thus, radiation effects are assumed to be mostly in the radial 

direction, except near the lateral ends of the MPC. Thermal radiation exchange occurs between 

the MPC outer surface and the overpack inner lid and between the MPC outer surface and the 

overpack bottom plate. In the limiters, the thin metal shell covering the neutron shield radiates 

to the small-diameter plate located directly across the air gap that fills the small-diameter 

cylinder (see Figure D-8 and the description in Section D.3.1.3).

 

1 The term “operational temperature limit” does not necessarily mean “melting point.” The Rail-Steel SAR 

provides operational temperature limits.
D-15

With the exception of the contact gaps already mentioned (e.g., between intermediate shell 

layers and between the MPC and the overpack), all contact gaps in the Rail-Steel cask are 

assumed to be perfect.

D.3.3 Rail-Steel Cask Materials and Thermal Properties

The Rail-Steel cask system is made from a variety of steel and aluminum alloys. The overpack 

inner shell is made from SA203-E cryogenic steel and the metal base cup, flange, and closure 

plate are made from SA350-LF3 cryogenic steel. The intermediate shells are made from 

SA516-70 carbon steel and the radial channels and enclosure plates are made from 

SA515-70 carbon steel. The neutron shield material is Holtite-A, a synthetic neutron-absorbing 

polymer with 1 percent boron carbide, sold commercially under the trade name NS-4-FR (Holtec 

International, 2000). The variable-length gap between the MPC and overpack is filled with 

helium.

The MPC shell, lid, and baseplate; the basket; and the fuel-cell walls are made from alloy X, a 

generic term used in various SARs that usually stands for one of the following stainless steel 

metals: SA304, SA304LN, SA316, or SA316LN (Holtec International, 2000). The thermal 

properties of SA304 are assumed for these components. Very little difference in thermal 

properties is found between SA304 and the other stainless steel materials already mentioned. 

On one side of each fuel cell wall is a thin layer of Boral sandwiched between the fuel cell wall 

and thin stainless steel sheathing. Boral is a neutron absorber made of boron carbide and 

aluminum alloy 1100 (Holtec International, 2000). The Boral layer and stainless steel sheathing 

extend the length of the active fuel region. The MPC-24 is designed to carry intact Zircaloy and 

stainless steel clad fuel assemblies. In this study, the fuel rods are assumed to be made from 

Zircaloy cladding, as in the Rail-Steel cask, for conservative results. The fuel pellets are 

uranium dioxide (UO2). The MPC heat conduction elements are made from aluminum 

alloy 1100. All void spaces inside the MPC are filled with helium (Holtec International, 2000).

The honeycomb in the impact limiter is made from aluminum 5052 and the large-diameter 

cylinder is made from carbon steel (SA516). The neutron shield segments are also made from 

Holtite-A.

Table D-2 provides the thermal conductivity for materials used in the Rail-Steel cask at several 

temperatures. For aluminum 1100 and the various carbon steels, data from the Rail-Steel cask 

SAR are available over a limited temperature range. The analysis in that report showed only a 

limited range of temperatures because the fire exposure was limited to 30 minutes with a 

subsequent cooldown. For these materials, the data trend is decreasing; therefore, the thermal 

conductivity value at the highest temperature is used at higher temperatures. This reflects a 

conservative assumption because the thermal conductivity values used are higher than they 

should be. Note also that Holtite-A is replaced with air once the temperature of the neutron 

shield region reached the operational temperature limit of that material. In reality, only a fraction 

of the Holtite-A decomposes. Some of the gases generated in the shield region outgas through 

the neutron shield rupture disks at high pressures. Up to 90 percent of these gases come from 

moisture in the Holtite-A (NRC, 2000). Experiments show that up to 50 percent (by weight) of 

the NS-4-FR eventually degrades by the time the temperature of the material reaches 

800 degrees C, leaving behind charred remains (Soo-Haeng et al., 1996), and these are not 

expected to combust (Soo-Haeng et al., 1996; NRC, 2000). The thermal conductivity of helium 

varies with pressure in addition to temperature; however, the pressure dependency is much 
D-16

weaker over the range of 101 to 689 kilopascals (14 to 100 pounds per square inch) 

(Petersen, 1970).

Table D-2 Thermal Conductivities for the Rail-Steel Cask Materials

Material

Thermal Conductivity W/m-°C (Btu/ft-hr-°F)

92°C

(200°F)

226°C

(450°F)

377°C

(700°F)

477°C

(900°F)

726°C

(1,340°F)

Air§ 0.026 

(0.015)

0.040 

(0.023)

0.050 

(0.028)

0.055 

(0.031)

0.067 

(0.038)

Stainless Steel§ 14.5 (8.3) 18.3 (10.5) 20.4 (11.8) 21.9 (12.6) 25.4 (14.6)

Aluminum Alloy 1100∗ 228 (131) 212 (122) ― ― ―

Aluminum-Honeycomb‡ 3.5 (2.0) 4.1 (2.4) 4.8 (2.8) 5.2 (3.0) ―

Boral (B4C)∗ 83.3 (48.2) 83.1 (48.0) 81.3 (47.0) 80.5 (46.5) ―

Carbon Steel - Int. Shells∗ 42.3 (24.5) 41.7 (24.1) 38.8 (22.4) ― ―

Carbon Steel - N. Shield∗ 50.7 (29.3) 49.1 (28.4) 42.6 (24.6) ― ―

Cryogenic Steel∗ 41.1 (23.8) 41.0 (23.7) 38.5 (22.3) ― ―

Helium§ 0.17 (0.098) 0.22 (0.12) 0.26 (0.15) 0.29 (0.16) 0.35 (0.20)

Holtite-A∗ 0.65 (0.37) ― ― ― ―

UO2

∗ 6.0 (3.4) 6.0 (3.4) 5.1 (2.9) ― ―

Zircaloy∗ 13.5 (7.8) 14.6 (8.4) 16.2 (9.3) 17.8 (10.2) ―

§ Incropera and Dewitt, 1996 ∗ Holtec International, 2000

‡ Thermophysical Properties Research Laboratory Inc., 2001

Dashes in the table indicate that properties at these temperatures were not available, or the material 

never reaches these temperatures. Where data is not available, it is assumed the values do not change 

with increasing temperature.

Table D-3 provides the specific heat for these same materials at several temperatures. 

Temperature-dependent values are given only for those materials that exhibited large variation 

in temperature. With the exception of stainless steel, aluminum 5052, and carbon steel, the 

specific heat of most materials used in the Rail-Steel cask is fairly constant. Of interest are the 

properties of carbon steel; the specific heat increases abruptly above 700 degrees C (1,292 F) 

and reaches a peak at around 768 degrees C (1,414 degrees F), the curie temperature. This 

behavior is associated with changes in the magnetic state of these materials and has been 

observed for a great number of carbon steel materials (Yafei, 2009). For Holtite-A, limited data 

are available above its operational temperature limit. Air properties are used beyond this limit. 

In addition, radiation exchange between the inner and outer surface of the neutron shield region 

is also allowed above this operational temperature limit to maximize heat input.
D-17

Table D-3 Specific Heat for the Rail-Steel Cask Materials

Material

Specific Heat J/kg-°C (Btu/lbm-°F)

92°C 

(200°F)

226°C 

(450°F)

377°C 

(700°F)

477°C 

(900°F)

726°C 

(1,340°F)

Air§ 1,010 (0.24) ― ― ― ―

Stainless Steel§ 482 (0.11) 535 (0.12) 563 (0.13) 575 (0.13) 611 (0.14)

Aluminum§ 903 (0.21) ― ― ― ―

Aluminum-Honeycomb‡ 890 (0.21) 976 (0.23) 1,057 (0.25) 1,100 (0.26) ―

Carbon Steel§ 434 (0.10) 505 (0.12) 590 (0.14) 653 (0.15) 1,169 (0.27)

Boral (B4C)* 2,478 (0.59) ― ― ― ―

Helium§ 5,193 (1.2) ― ― ― ―

Holtite-A* 1,632 (0.39) ― ― ― ―

UO2

* 234 (0.056) ― ― ― ―

Zircaloy* 304 (0.073) ― ― ― ―

§ Incropera and Dewitt, 1996

 

∗ Holtec International, 2000

‡ Thermophysical Properties Research Laboratory Inc., 2001

Dashes in the table indicate that properties at these temperatures were not available, or the material 

never reaches these temperatures. Where data is not available, it is assumed the values do not change 

with increasing temperature.

Table D-4 provides densities for stainless steel, carbon steel, Zircaloy, and UO2 at 92 degrees C 

(200 degrees F) and for air and helium at various temperatures. Because the density of most 

metals changes very little with temperature, only the values at 92 degrees C (200 degrees F) 

are used. The density of Holtite-A is assumed not to vary significantly from 92 degrees C 

(200 degrees F) to its operational temperature limit. Recall that air properties are used above 

this limit to replace Holtite-A.

Table D-5 shows the emissivity values obtained from the Rail-Steel cask SAR. The exterior 

surface of the Rail-Steel cask is coated with Carboline 890 paint, and the overpack inner 

surfaces are coated with Thermaline 450 paint. However, these coatings are only good up to 

216 degrees C (422 degrees F) and 262 degrees C (505 degrees F), respectively (Holtec 

International, 2000). Note also the internal surfaces of the heat conduction elements are 

sandblasted to increase radiation between opposite sides of the heat conduction elements.
D-18

Table D-4 Densities for the Rail-Steel Cask Materials

Material

Density kg/m3 (lbm/ft3)

92°C 

(200°F)

226°C 

(450°F)

377°C 

(700°F)

477°C 

(900°F)

726°C 

(1,340°F)

Air§ 0.98 (0.061) 0.69 (0.043) 0.54 (0.034) 0.46 (0.029) 0.35 (0.022)

Stainless Steel§ 7,900 (493) ― ― ― ―

Aluminum§ 2,702 (168) ― ― ― ―

Aluminum-Honeycomb‡ 526 (32) ― ― ― ―

Carbon Steel§ 7,854 (490) ― ― ― ―

Boral (B4C)∗ 544 (34) ― ― ― ―

Helium§ 0.14 (0.008) 0.10 (0.006) 0.077 

(0.0048)

0.065 

(0.0041)

0.048 

(0.003)

Holtite-A∗ 1,681 (105) ― ― ― ―

UO2

∗ 10,956 

(684) ― ― ― ―

Zircaloy∗ 6,551 (409) ― ― ― ―

§ Incropera and Dewitt, 1996

 

∗ Holtec International, 2000

‡ Thermophysical Properties Research Laboratory Inc., 2001

Dashes in the table indicate that properties at these temperatures were not available, or the material 

never reaches these temperatures. Where data is not available, it is assumed the values do not change 

with increasing temperature.

Table D-5 Emissivity for Some of the Rail-Steel Cask Materials and Paints

Material Emissivity

Zircaloy 0.80

Painted Surface 0.85

Rolled Carbon Steel 0.66

Stainless Steel 0.36

Sandblasted Aluminum 0.40

Only materials on the outside of the cask are expected to experience temperatures above 

726 degrees C (1,340 degrees F). For all of the casks evaluated, the exterior layer is stainless 

steel. The important parameter for heat transfer through this stainless steel is thermal 

diffusivity, which is equal to the thermal conductivity divided by the product of density and 

specific heat. At high temperatures, this value does not significantly change with increasing 

temperature. The value for thermal diffusivity of stainless steel used at 726 degrees C and 

above in these analyses was 5.26x10-6 m2

/s. Recent measurements of thermal diffusivity of 

stainless steel at high temperature (Rempe and Knudson, 2008) indicate a best fit for the 

thermal diffusivity at 1300 degrees C (2370 degrees F) is 5.8x10-6 m2

/s, only 10 percent higher 

than the value used. Some of the data points given by Rempe and Knudson at high 
D-19

temperature are actually below 5.26x10-6 m2

/s. Because of the large degree of scatter in the 

data for thermal diffusivity, a single analysis using a higher diffusivity (150 percent) was 

performed and resulted in only a couple of degrees of difference in predicted temperature. 

Because the predicted temperatures are not near any of their respective limits, it was 

determined that the results were appropriate.

D.3.3.1 Effective Thermal Properties of Fuel Basket and Fuel-Basket Periphery

Thermal properties for the fuel-basket region and fuel-basket periphery are obtained from the 

Rail-Steel cask SAR. In that report, the fuel basket and the fuel-basket periphery cross-sections 

were replaced with two concentric cylinders, each with equivalent effective thermal properties, 

as described in Section D.3.1.2. The procedure used to obtain the in-plane thermal 

conductivities of the fuel basket and fuel-basket periphery as a function of temperature is 

described in the Rail-Steel cask SAR, but is summarized here for completeness.

First, the cross-section of the fuel assembly is modeled using a detailed two-dimensional finite 

element model of the cross-section of a 17 × 17 Optimized Fuel Assembly (OFA) rod 

arrangement (see Figure D-7a), a uniform heat generation rate over each fuel rod, and a 

uniform temperature applied to the periphery of the fuel assembly. The 17 × 17 OFA used was 

determined to be the most resistive assembly design (Holtec International, 2000). The finite 

element model takes into account radiation between the rods and conduction across the helium 

gap. The effective thermal conductivity is obtained from the following equation:

(D-1) keff = 0.29468�qga2�

∆T

Where qg is the heat generation rate per fuel cell per unit length, a is half the length of one side 

of the fuel cell, and ∆T is the maximum temperature difference in the fuel assembly 

(Sanders et al., 1992). Because radiation is not linearly dependent on temperature, the model 

is run several times, each time with increasing uniform temperature near the edge of the fuel 

assembly to obtain effective properties at various temperatures. The detailed fuel assembly is 

thus replaced with a homogenized fuel cell region (see Figure D-7).

Second, the in-plane thermal conductivity of the basket storage wall, Boral, and stainless steel 

sheathing are replaced with an equivalent thermal conductivity using the thermal resistor 

network described in the Rail-Steel cask SAR. The representative network takes into account 

the thermal resistances perpendicular to the wall and along the wall.

Third, the cross-section of the MPC is modeled using a two-dimensional finite element 

representation of the homogenized fuel-basket walls, with a uniform heat generation rate 

applied over each homogenized fuel assembly and a uniform temperature applied over the 

perimeter of the MPC shell. The model in the Rail-Steel cask SAR took into account 

(1) conduction through the homogenized fuel assemblies, the helium gas in the flux traps, and 

the basket periphery; (2) radiation between homogenized basket walls; and (3) natural 

convection loops in the basket periphery. The effective conductivities of the basket region (kb) 

and periphery region (kp) are given by the following equations:

(D-2) (kb)eff = q

4π∆Tbm
D-20

(D-3) �kp�

eff = qW

As∆Tpb

Where

(D-4) ∆Tpb = ∆Tpm − ∆Tbm.

Here, q is the MPC heat generation per unit length, ∆Tbm is the maximum temperature 

difference in the basket, ∆Tpm is the maximum temperature difference in the MPC cross-section, 

As is the surface area per unit length, and W is the basket periphery annular gap length. 

Table D-6 give the equivalent fuel-basket thermal conductivities. The Rail-Steel cask SAR 

obtained the effective axial thermal conductivities of the fuel basket by using the resistor 

method, which reduces to an area-weighted average because the basket length (L) in the 

resistance (L/kA) is equal across all materials. The specific heat and density are obtained using 

a mass- and volume-weighted average respectively. Near the ends of the basket, the fuel rods 

are filled with gas, decreasing the in-plane and axial thermal conductivity of the basket slightly 

because the thermal conductivity of helium is smaller than that of the UO2 pellets. Note that the 

thermal conductivities did not vary much by temperature.

The properties in Table D-6 are used over the length of the basket. For consistency, 

temperature-varying properties are implemented in the thermal model.

Table D-6 Effective Thermal Conductivity for the Fuel-Basket Region

Effective Thermal Properties 92°C 

(200°F)

226°C 

(450°F)

377°C 

(700°F)

477°C 

(900°F)

726°C 

(1,340°F)

In-Plane Thermal Conductivity in 

W/m-°C (Btu/ft-hr-°F) 1.9 (1.1) 2.6 (1.5) 3.4 (1.9) ― ―

Axial Thermal Conductivity in 

W/m-°C (Btu/ft-hr-°F) 3.4 (1.9) 3.8 (2.2) 4.3 (2.5) 4.6 (2.6) ―

Specific Heat in J/kg-°C 

(Btu/lbm-°F) 305 (0.073)

Density in kg/m3 (lbm/ft3) 2,688 (168)

Dashes in the table indicate that properties at these temperatures were not available, or the material 

never reaches these temperatures. Where data is not available, it is assumed the values do not change 

with increasing temperature.

Fuel spacers separate the ends of the fuel assembly from the MPC lid and MPC bottom plate. 

In these regions, conduction is predominately through the helium gas and through the fuel 

spacer and fuel-basket walls. Thermal radiation also occurs between the walls of the basket 

and the fuel spacers.

The homogenized material properties used in the fuel-spacer region are estimated by taking into 

account the properties of the fuel region, fuel spacer, helium, fuel-basket ends, and thermal 

radiation. A sensitivity study using theoretical bounds indicated that the temperatures obtained 

in the regions of interest were barely influenced by the properties used.
D-21

Fourth, the thermal conductivity in the basket periphery is further enhanced to account for heat 

dissipation through heat conduction elements. The equivalent resistor network through the heat 

conduction elements is obtained using a two-dimensional analytical model explained in the 

Rail-Steel cask SAR. This resistance is added in parallel with the resistance obtained from the 

two-dimensional finite element model for the basket periphery region. Table D-7 provides the 

fuel-basket periphery’s in-plane conductivity.

Table D-7 Effective Thermal Conductivity of the Aluminum Heat Conduction Elements

Effective Thermal Properties 92°C 

(200°F)

226°C 

(450°F)

377°C 

(700°F)

477°C 

(900°F)

726°C 

(1,340°F)

In-Plane Thermal Conductivity in

W/m-°C (Btu/ft-hr-°F) 0.43 (0.25)

Axial Thermal Conductivity in W/m-°C 

(Btu/ft-hr-°F) 10 (5.8)

Specific Heat in J/kg-°C (Btu/lbm-°F) 964 (0.23)

Density in kg/m3 (lbm/ft3) 132 (8.25)

Researchers determined the axial effective thermal conductivity from an area-weighted average 

using aluminum 1100 and helium properties. Holtec International (2000) gives the area of the 

periphery region. The area of the heat conduction elements is estimated at 3.5 times the fuel 

basket’s cell pitch (27.3 cm (10.7 in.)) multiplied by the thickness of the elements (3.175 mm 

(0.125 in.)) and the total number of aluminum inserts (i.e., eight) (Holtec International, 2000). 

The specific heat and density of the fuel-basket periphery is obtained from an area- and 

mass-weighted average, respectively, again considering only aluminum 1100 and helium.

Heat transfer through the periphery region is further enhanced by radiation between the inner 

walls of the heat conduction elements and the walls of the MPC and fuel basket. As 

demonstrated in Table D-6, the emissivity of stainless steel and sandblasted aluminum are not 

very different.

D.3.3.2 Effective Thermal Properties of Multipurpose Canister-Overpack Helium Gap

In the horizontal position, the MPC rests on the overpack, forming a nonconcentric, 

variable-length helium gap. This gap is not modeled explicitly. Instead, the study used a 

two-dimensional analytical model derived in Holtec International (2000) to obtain an effective 

conductivity through the variable-length gap. This model included the effects of the contact 

region as explained below.

To account for radial heat dissipation through the variable-length helium gap and through the 

metal-to-metal contact area, equations for the overall heat conducted through these regions are 

summed and then equated to the overall heat conducted through a concentric gap to obtain an 

effective thermal conductivity for a constant-length helium gap (i.e., concentric gap). The 

following equation, taken from the Rail-Steel cask SAR, was used to obtain the effective thermal 

conductivity across the gap (kgap):
D-22

(D-5)

�kgap�

eff

kgas

= t

π ∫ 1t (1−cos θ)+ε cos θ dθ π

0

Where kgas is the conductivity of the gas, t is the thickness of the concentric gap, and ε (0.5 mm 

(0.02 in.)) is the metal-to-metal contact area width. Results reported in the SAR show that the 

effective conductivity through the equivalent concentric gap is twice the conductivity of helium.

D.3.3.3 Effective Thermal Properties of Overpack Intermediate Shells

The Rail-Steel cask consists of a series of shell-gas layers between the inner shell wall and the 

outermost intermediate shell of the overpack. The contact gaps are assumed to be 0.05 mm 

(0.002 in.) across (Holtec International, 2000). No radiation is assumed through these gaps 

because radiation accounts for less than 5 percent of the effective conductivity for gaps of this 

size. Researchers obtained the in-plane thermal conductivity by adding the resistances across 

each shell and gap in series (see Table D-8). The axial and circumferential conductivities are 

assumed to be that of the shell layer material because the thermal conductivity of air and the 

gap area of air contribute very little. Similarly, the specific heat and density of the intermediate 

shell layers are assumed to be equal to the intermediate shell material.

Table D-8 Effective Thermal Conductivity of the Intermediate Shells in the In-Plane 

Directions

Effective Thermal Properties 92 °C 

(200 °F)

226 °C 

(450 °F)

377 °C 

(700 °F)

477 °C (900

°F)

726 °C 

(1,340 °F)

In-Plane Thermal Conductivity 

W/m-°C (Btu/ft-hr-°F) 13.2 (7.6) 15.6 (9.0) 17.0 (9.8) 18.6 (10.7) 22.1 (12.7)

D.3.3.4 Effective Thermal Properties of Neutron Shield Region

The neutron shield region consists of the Holtite-A inside the cavities formed between the 

outermost intermediate shell and the outer enclosure shell and between the radial channels. 

Note that the outer enclosure shell is not included here because it is modeled explicitly. The 

neutron shield region includes the Holtite-A material and the radial sections of the channel 

(2 per channel for a total of 40). This region is also modeled as a single volume with 

homogenized thermal properties.

Table D-9 shows the effective properties in the neutron shield region. The effective thermal 

conductivity in the in-plane and axial direction are obtained by summing the resistance through 

the radial channels and through the neutron shield material in parallel. Because both the 

Holtite-A and radial channels extend the same length in the axial direction, the resistance 

equation in the axial direction reduces to an area-weighted average of the individual material 

conductivities. Researchers used air in place of Holtite-A to calculate the effective properties 

above 149 degrees C (300 degrees F), taking into account the radial channels.

The thermal conductivity in the circumferential direction is assumed to be that of Holtite-A 

because the total thickness of the radial channels in this direction is small compared to the total 
D-23

circumferential length of the Holtite-A. Note that this is a conservative assumption in the sense 

that heat dissipated through the neutron shield region is preferentially in the in-plane and axial 

directions as a result of the latter assumption. This assumption does not have an impact in the 

uniform-heating run, but it does have an impact on the CAFE fire runs, where heat input around 

the circumference of the cask varies. In this case, heat will be dissipated more readily through 

the in-plane direction, thus giving higher temperatures in the interior of the cask.

Table D-9 Effective Conductivity of the Neutron Shield Region

Effective Thermal Properties 92°C 

(200°F)

226°C 

(450°F)

377°C 

(700°F)

477°C 

(900°F)

726°C 

(1,340°F)

In-Plane Thermal Conductivity 

W/m-°C (Btu/ft-hr-°F) 4.3 (2.4) 3.5 (2.0) 3.2 (1.8) 3.1 (1.8) 2.7 (1.5)

Axial Thermal Conductivity 

W/m-°C (Btu/ft-hr-°F) 3.6 (2.0) 3.3 (1.9) 3.0 (1.7) 3.0 (1.7) 2.6 (1.5)

Specific Heat J/kg-°C 

(Btu/lbm-°F) 1,315 (0.31) 505 (0.12) 590 (0.14) 653 (0.15) 1,130 

(0.28)

Density kg/m3 (lbm/ft3) 2,113 (132) 552 (34)

Researchers obtained the specific heat and density of the neutron shield region using a 

mass- and area-weighted average, respectively. Holtite-A is expected to reach its temperature 

limit during the early transient period of a fire. When this happens, Holtite-A partially 

decomposes, leaving char residue behind. Most of the excess gas generated in Holtite-A 

outgases through the rupture disks when the pressure inside the neutron shield region reaches 

the disks’ design limits. In the thermal model, when Holtite-A’s temperature limit is reached, 

Holtite-A is replaced with air, and radiation is activated by setting the emissivity to an 

appropriate value. Note that air effectively lowers the specific heat and density of the neutron 

shield region. The effective specific heat of the neutron shield region is greatly influenced by 

the specific heat values of carbon steel because the density of air in the mass-weighted average 

is very small compared to carbon steel.

D.3.4 Rail-Steel Cask, Finite Element Model, and Boundary Conditions

A steady-state case was run to obtain the initial conditions of the Rail-Steel cask and to 

compare results against those provided in the Rail-Steel SAR and Adkins et al. (2006). The 

steady-state model consisted of exposing the Rail-Steel cask to a 37.8-degree C (100-degree F) 

ambient-temperature, radiation boundary condition. This boundary condition is applied over the 

entire outer surface of the cask using an emissivity value of 0.85. In addition, insolation is 

applied over the outer curved surfaces of the cask (193.8 watts per square meter (W/m2

)

(61.4 British thermal units per square foot per hour (Btu/ft2

-hr)) and over the flat ends of the cask 

(96.9 W/m2 (30.7 Btu/ft2

-hr)), as specified in American Society of Testing Materials E2230, 

“Standard Practice for Thermal Qualification of Type B Packages for Radioactive Material”

(ASTM, 2008). A convection boundary condition is also applied to the outer surface of the cask 

using a heat transfer coefficient of 3 W/m2

-°C (0.53 Btu/ft2-hr-°F). This value is obtained from a 

set of correlations described in the Rail-Steel cask SAR—assuming turbulent flow—and is within 

the same order of magnitude as values obtained from correlations in Incropera and 

Dewitt, 1996.
D-24

In general, steady-state results are slightly higher than those presented in the Rail-Steel cask 

SAR, but lower than those reported in Adkins et al. (2006). For example, the current study 

found a maximum fuel cladding temperature of 376 degrees C (710 degrees F), compared to 

372 degrees C (701 degrees F) in the Rail-Steel cask SAR and 392 degrees C (738 degrees F) 

in Adkins et al. (2006). The largest differences are observed in the extreme ends of the 

overpack, where temperatures in the Rail-Steel cask are lower (by approximately 25 degrees C

(45 degrees F)) than reported here and significantly lower (approximately 50 degrees C

(90 degrees F)) than what is reported in Adkins et al. (2006). These differences are attributed to 

dissimilarities in modeling assumptions and approaches and boundary conditions. For example, 

Adkins et al. (2006) assumed a gap between the overpack and the limiters. Overall, however, 

the temperatures obtained from these three studies showed similar spatial trends and good 

agreement given the differences cited above.

The steady-state case is used to assess the suitability of the mesh. The mesh is initially 

169,600 elements; this corresponds to a nominal element size of 10.2 cm (4 in.). This value is 

decreased to 5.1 cm (2 in.) and then increased to 15.2 cm (6 in.) to study the effects of element 

size on temperatures at locations of interest (as shown in the results of Chapter 4 and later in 

this appendix). Results of the 15.2-cm (6-in.) element-size mesh showed some difference in the 

temperatures in the interior of the cask when compared to those of the 10.2-cm (4-in.) 

element-size mesh. This is expected because large cells are created in the interior of the cask. 

Near the exterior of the overpack, small geometric features resulted in small size elements. 

Results of the 5.1-cm (2-in.) element-size mesh showed very little difference when compared to 

the 10.2-cm (4-in.) element-size mesh. The 5.1-cm (2-in.) element-size mesh had smaller 

elements in the interior and about the same near the exterior of the overpack. Therefore, a third 

case was run, this time using the 10.2-cm (4-in.) mesh, with a refined mesh near the exterior of 

the overpack. Results from this mesh showed some difference (less than 5 degrees in the 

neutron shield region), but not enough to justify the extra computational time needed to run this 

mesh. Figure D-9 shows the final mesh used to run the five scenarios described in Chapter 4.

Figure D-9 Rail-Steel cask finite element mesh
D-25

The uniform-heating case described in Chapter 4 was run initially to verify the Rail-Steel cask 

finite element model. This exercise gave an additional measure of confidence in the Rail-Steel 

cask model. The boundary conditions for this case consisted of exposing the Rail-Steel cask to 

an 800-degree C (1,472-degree F) ambient-temperature, radiation boundary condition. This 

boundary condition is applied over the entire outer surface of the cask using an emissivity value 

of 0.9. A convection boundary condition is also applied to the outer surface of the cask using a 

heat transfer coefficient of 85 W/m2

-°C (15.2 Btu/ft2-hr-°F). The Rail-Steel cask SAR obtains 

this value from a set of correlations described within that report. The Rail-Steel cask SAR also

assumes a vertical flame speed of 15 meters per second (m/s) (49 feet per second (ft/s)), a 

value significantly higher than that specified in Nakos (2005) but nevertheless conservative 

because it will result in a higher heat input to the cask. Convection accounts for about 10 to 

20 percent of the total heat input for large objects inside a fire; the rest is through thermal 

radiation (Nicolette and Larson, 1989).

The uniform-heating case is run for 30 minutes, followed by an 11.5-hour transient cooldown 

with insolation. During the cooldown period, the boundary conditions are set back to their 

steady-state case values, except for the emissivity of the outer cask, which remains the same to 

simulate what happens in actual fires—a blanket of soot covers the cask. Also, unlike the 

Rail-Steel cask SAR, the neutron shield region is assumed to contain air with radiation 

interaction between the outer enclosure shell and the outermost intermediate shell.

Overall, maximum temperatures obtained using the model developed in this study and in the 

Rail-Steel cask SAR are similar. The difference in purpose of the two analyses leads to some 

different assumptions, which in turn leads to slightly different results.

For the remaining cases, the external boundary conditions are obtained from CAFE, the 

computational fluid dynamics code coupled to P/Thermal. As mentioned in Section D.2, a 

boundary condition was set up in Patran that allowed CAFE results to be communicated to 

P/Thermal and vice-versa. The cooldown period for these cases also used the steady-state 

case boundary conditions (from 10 CFR 71.71, “Normal Conditions of Transport”).

D.3.5 Rail-Steel Cask Thermal Analysis Results

Figures D-10 through D-14 show results for the five scenarios already described in Chapter 4. 

These results are not discussed here, but are presented to supplement results discussed in 

Chapter 4. Figure D-10 shows results for the regulatory uniform-heating case cited in the 

previous section. This is the P/Thermal-only run. Figure D-11 shows results for the regulatory 

CAFE fire and, together with Figure D-10, might be useful in determining the differences 

between uniform and nonuniform fire conditions.

Nicolette and Larson (1989) discuss the effect that large objects have on fires and their 

implications to modeling large casks in fires. Figure D-12 shows results for the fully engulfing 

CAFE fire with the cask on the ground, and Figure D-13 and Figure D-14 show results for the 

cask on the ground but outside the fire. The last three cases are for a 3-hour fire and 

subsequent cooldown period.
D-26

Figure D-10 Rail-Steel cask regulatory uniform-heating results (P/Thermal)

0

20

40

60

80

100

120

140

160

180

200

0 2 4 6 8 10 12 14

Temperature (°C)

Time (hrs)

Seal Temperature

Seal-Near_Side

Seal-Top

Seal-Far_Side

Seal-Bottom

0

50

100

150

200

250

300

350

400

450

0 2 4 6 8 10 12 14

Temperature (°C)

Time (hrs)

Fuel Temperature

Fuel_Region-Center

Fuel_Edge-Near_Side

Fuel_Edge-Top

Fuel_Edge-Far_Side

Fuel_Edge-Bottom
D-27

Figure D-10 Rail-Steel cask regulatory uniform-heating results (P/Thermal) (continued)

0

50

100

150

200

250

300

350

0 5 10 15

Temperature (°C)

Time (hrs)

Cask Inner & Neutron Shield Temperature

Cask_Inner_Surface-Near_Side

Cask_Inner_Surface-Top

Cask_Inner_Surface-Far_Side

Cask_Inner_Surface-Bottom

Neutron_Shield_Inner_SurfaceNear_Side

Neutron_Shield_Inner_SurfaceTop

Neutron_Shield_Inner_SurfaceFar_Side

Neutron_Shield_Inner_SurfaceBottom

0

100

200

300

400

500

600

700

800

0 5 10 15

Temperature (°C)

Time (hrs)

Outer Surface Temperature

Cask_Outer_Surface-Near_Side

Cask_Outer_Surface-Top

Cask_Outer_Surface-Far_Side

Cask_Outer_Surface-Bottom
D-28

Figure D-11 Rail-Steel cask CAFE regulatory fire

0

20

40

60

80

100

120

140

160

180

0 2 4 6 8 10 12 14

Temperature (°C)

Time (hrs)

Seal Temperature

Seal-Near_Side

Seal-Top

Seal-Far_Side

Seal-Bottom

0

50

100

150

200

250

300

350

400

450

0 2 4 6 8 10 12 14

Temperature (°C)

Time (hrs)

Fuel Temperature

Fuel_Region-Center

Fuel_Edge-Near_Side

Fuel_Edge-Top

Fuel_Edge-Far_Side

Fuel_Edge-Bottom
D-29

Figure D-11 Rail-Steel cask CAFE regulatory fire (continued)

0

50

100

150

200

250

300

350

400

0 5 10 15

Temperature (°C)

Time (hrs)

Cask Inner & Neutron Shield Temperature

Cask_Inner_Surface-Near_Side

Cask_Inner_Surface-Top

Cask_Inner_Surface-Far_Side

Cask_Inner_Surface-Bottom

Neutron_Shield_Inner_SurfaceNear_Side

Neutron_Shield_Inner_SurfaceTop

Neutron_Shield_Inner_SurfaceFar_Side

Neutron_Shield_Inner_SurfaceBottom

0

100

200

300

400

500

600

700

800

900

1000

0 5 10 15

Temperature (°C)

Time (hrs)

Outer Surface Temperature

Cask_Outer_Surface-Near_Side

Cask_Outer_Surface-Top

Cask_Outer_Surface-Far_Side

Cask_Outer_Surface-Bottom
D-30

Figure D-12 Rail-Steel cask CAFE fire with cask on ground and at the pool center

0

50

100

150

200

250

300

350

400

0 2 4 6 8 101214161820222426

Temperature (°C)

Time (hrs)

Seal Temperature

Seal-Near_Side

Seal-Top

Seal-Far_Side

Seal-Bottom

0

100

200

300

400

500

600

0 2 4 6 8 101214161820222426

Temperature (°C)

Time (hrs)

Fuel Temperature

Fuel_Region-Center

Fuel_Edge-Near_Side

Fuel_Edge-Top

Fuel_Edge-Far_Side

Fuel_Edge-Bottom
D-31

Figure D-12 Rail-Steel cask CAFE fire with cask on ground and at the pool center

(continued)

0

100

200

300

400

500

600

700

800

0 5 10 15 20 25

Temperature (°C)

Time (hrs)

Cask Inner & Neutron Shield Temperature

Cask_Inner_Surface-Near_Side

Cask_Inner_Surface-Top

Cask_Inner_Surface-Far_Side

Cask_Inner_Surface-Bottom

Neutron_Shield_Inner_SurfaceNear_Side

Neutron_Shield_Inner_SurfaceTop

Neutron_Shield_Inner_SurfaceFar_Side

Neutron_Shield_Inner_SurfaceBottom

0

200

400

600

800

1000

1200

0 5 10 15 20 25

Temperature (°C)

Time (hrs)

Outer Surface Temperature

Cask_Outer_Surface-Near_Side

Cask_Outer_Surface-Top

Cask_Outer_Surface-Far_Side

Cask_Outer_Surface-Bottom
D-32

Figure D-13 Rail-Steel cask CAFE fire with cask on ground 3 m (10 feet) from the edge of 

the pool

0

50

100

150

200

250

300

350

0 2 4 6 8 10 12 14 16

Temperature (°C)

Time (hrs)

Seal Temperature

Seal-Near_Side

Seal-Top

Seal-Far_Side

Seal-Bottom

0

50

100

150

200

250

300

350

400

450

0 2 4 6 8 10 12 14 16

Temperature (°C)

Time (hrs)

Fuel Temperature

Fuel_Region-Center

Fuel_Edge-Near_Side

Fuel_Edge-Top

Fuel_Edge-Far_Side

Fuel_Edge-Bottom
D-33

Figure D-13 Rail-Steel cask CAFE fire with cask on ground 3 m (10 feet) from the edge of 

the pool (continued)

0

50

100

150

200

250

300

350

400

450

0 5 10 15 20

Temperature (°C)

Time (hrs)

Cask Inner & Neutron Shield Temperature

Cask_Inner_Surface-Near_Side

Cask_Inner_Surface-Top

Cask_Inner_Surface-Far_Side

Cask_Inner_Surface-Bottom

Neutron_Shield_Inner_SurfaceNear_Side

Neutron_Shield_Inner_SurfaceTop

Neutron_Shield_Inner_SurfaceFar_Side

Neutron_Shield_Inner_SurfaceBottom

0

100

200

300

400

500

600

700

0 5 10 15 20

Temperature (°C)

Time (hrs)

Outer Surface Temperature

Cask_Outer_Surface-Near_Side

Cask_Outer_Surface-Top

Cask_Outer_Surface-Far_Side

Cask_Outer_Surface-Bottom
D-34

Figure D-14 Rail-Steel cask CAFE fire with cask on ground 18.3 m (60 feet) from the edge 

of the pool

0

20

40

60

80

100

120

140

160

0 2 4 6 8 10 12 14 16

Temperature (°C)

Time (hrs)

Seal Temperature

Seal-Near_Side

Seal-Top

Seal-Far_Side

Seal-Bottom

0

50

100

150

200

250

300

350

400

450

0 2 4 6 8 10 12 14 16

Temperature (°C)

Time (hrs)

Fuel Temperature

Fuel_Region-Center

Fuel_Edge-Near_Side

Fuel_Edge-Top

Fuel_Edge-Far_Side

Fuel_Edge-Bottom
D-35

Figure D-14 Rail-Steel cask CAFE fire with cask on ground 18.3 m (60 feet) from the edge 

of the pool (continued)

0

50

100

150

200

250

0 5 10 15 20

Temperature (°C)

Time (hrs)

Cask Inner & Neutron Shield Temperature

Cask_Inner_Surface-Near_Side

Cask_Inner_Surface-Top

Cask_Inner_Surface-Far_Side

Cask_Inner_Surface-Bottom

Neutron_Shield_Inner_SurfaceNear_Side

Neutron_Shield_Inner_SurfaceTop

Neutron_Shield_Inner_SurfaceFar_Side

Neutron_Shield_Inner_SurfaceBottom

0

50

100

150

200

250

300

350

0 5 10 15 20

Temperature (°C)

Time (hrs)

Outer Surface Temperature

Cask_Outer_Surface-Near_Side

Cask_Outer_Surface-Top

Cask_Outer_Surface-Far_Side

Cask_Outer_Surface-Bottom
D-36

D.4 Rail Cask with Lead Shielding

The Rail-Lead cask (NAC International, 2004) is also certified to transport SNF material on 

railcars. This cask is chosen because it presents quite a different design philosophy from the 

Rail-Steel cask. The Rail-Lead cask uses lead for the gamma shield. Moreover, the Rail-Lead 

cask is certified to carry SNF without a separate canister. As in the Rail-Steel cask analysis, the 

Rail-Lead cask is assumed to be in the horizontal configuration, as it would be during 

transportation, and most likely after an accident scenario. Only the thermally relevant 

components of the Rail-Lead cask are considered to estimate the thermal response of this cask.

The Rail-Lead cask uses a single lead gamma shield, as opposed to a multilayer carbon steel 

gamma shield like the one used in the Rail-Steel cask. This lead shield melts at relatively low 

temperatures, but remains in the overpack in molten form until the temperature is low enough to 

change back to the solid state. This process impacts the ability of the cask to attenuate gamma 

rays, as described in Chapter 5 and Appendix E. One unique feature of the Rail-Lead cask is 

that it can transport the SNF in a directly loaded fuel basket, in addition to inside a canister, as 

used in the Rail-Steel cask. The directly loaded configuration is a significant design departure 

from the MPC configuration because there is no barrier between the fuel assemblies and the 

inner walls of the overpack. For this reason, this analysis focuses on the directly loaded 

configuration. Finally, the Rail-Lead cask uses wood-filled impact limiters, as opposed to an 

aluminum honeycomb, a minor difference from the thermal analysis point of view, but 

nevertheless important to note.

In most cases, results reported in the Rail-Lead cask SAR (NAC International, 2004) are used 

but modified where necessary, as done in the Rail-Steel cask analysis. The only significant 

departure is how the interior of the overpack is treated in the Rail-Lead cask SAR, as explained 

in the introduction to this appendix. Unlike the method used in that SAR, the directly loaded 

basket is replaced with a cylinder having equivalent effective thermal properties using a simple, 

three-dimensional finite element model and the thermal resistor network method. As done in 

the Rail-Steel cask analysis, the neutron shield region is replaced with an equivalent thermal 

region. The impact limiters are also modeled in the uncrushed state for the same reasons cited 

in Section D.3.1.3.

D.4.1 Geometry Considerations

The directly loaded Rail-Lead cask consists of an overpack, a fuel basket, and limiters at each 

end of the basket, as shown in Figure D-15. The directly loaded fuel basket is an open fuel 

container designed to fit snugly within the overpack interior cavity. The overpack is designed to 

attenuate both the heat and the neutron and gamma rays generated inside the fuel basket. The 

overpack contains two lids, each fitted with seals that completely seal the contents inside the 

overpack from the outside environment. The total length of the Rail-Lead cask, including the 

limiters, is approximately 6.5 m (256 in.).
D-37

Figure D-15 Rail-Lead cask components with the directly loaded fuel basket shown to 

the right (from NAC International, 2004)

D.4.1.1 Overpack

The Rail-Lead cask overpack is also a multilayer cylindrical vessel approximately 2.20 m 

(86.7 in.) in diameter and 4.90 m (193 in.) in length (see Figure D-15). The inner cavity of the 

overpack is approximately 1.80 m (71 in.) in diameter and 4.19 m (165 in.) in length. The 

cross-section of the overpack vessel is made of three shell layers arranged in the following 

order starting from the center of the overpack: an inner shell, a lead shell, and an outer shell 

(see Figure D-16).

As in the Rail-Steel cask, these shells are tightly coupled to each other and are welded to the 

overpack bottom plate and top flange. The lead shell acts as the gamma shield in this design. 

The thickness of the inner shell wall is not constant throughout, but tapers in slightly through 

most of the overpack side wall. This configuration allows the thickness of the lead shell to 

increase slightly through the same section of the overpack, where the gamma shielding is most 

needed. Radial channels are also welded to the outer shell to enhance heat transfer through 

the neutron shield region. The outer enclosure shell is formed the same way as in the 

Rail-Steel cask. Similarly, the cavities formed by the outer enclosure shell, the radial channels, 

and the outer enclosure shell are filled with a neutron shield material. The neutron shield region 

increases the diameter of the overpack an additional 29.2 cm (11.5 in.). Unlike the Rail-Steel 

cask, the overpack contains inner and outer lids that fit into the flange. Both the inner lid and 

bottom plate contain a cylindrical layer of neutron shield that is 5 cm (2 in.) thick.
D-38

Figure D-16 Cross-section view of the Rail-Lead cask with the directly loaded fuel basket

The thermal model explicitly represents the inner, outer, and lead shells; the flange; the inner 

and outer lids; and the bottom plate, with minor alterations to simplify the model. The most 

significant change is in making the thickness of the inner shell and lead shell constant 

throughout. Their thickness is kept equal to the corresponding thicknesses in the middle section 

of the overpack. As in the Rail-Steel cask model, the neutron shield region is represented as a 

single volume to minimize geometric complexity. As with the Rail-Steel cask, the Rail-Lead 

cask overpack contains a number of features that serve a special purpose. These features are 

omitted from the model for the same reasons they were omitted in the Rail-Steel cask model—

their effects are assumed to be either (1) negligible because of their small volume and mass 

relative to the other components in the overpack or (2) highly localized with no effect to the 

overall thermal performance of the cask at locations of interest or (3) both.

D.4.1.2 Directly Loaded Fuel Basket

In the Rail-Lead cask, the nuclear spent fuel is stored in a directly loaded basket (see 

Figure D-15). In this configuration, the fuel basket can store up to 26 PWR fuel assemblies. 

The total length of the fuel basket is 4.18 m (164.5 in.) and the diameter is a little less than the 

inner diameter of the overpack. The fuel basket consists of 31 support disks and 20 heat 

transfer disks, aligned parallel to each other, and each is precisely separated using six threaded 

rods and spacer nuts. The heat transfer disks are placed between the support disks in the 

region where the heat decay rate is at a maximum. Except for the end support disks, all support 

disks have the same thickness; the end support disks are twice as thick. Except for the end 

support disks, all heat transfer disks are slightly thicker than the support disks. Both disk types 

contain 26 square holes spaced at regular intervals and aligned between disks. Each square 

hole fits a thin-walled square fuel tube that extends almost the length of the basket. These 

tubes are welded to the disks and accommodate the fuel assemblies. The fuel assemblies 

extend almost the entire length of the fuel basket. The basket’s active fuel region is assumed to 

be 3.66 m (144 in.) in length, as suggested in the Rail-Lead cask SAR. Additional plates and a 

short cylinder are welded to the end support disks for extra support and to complete the 

fuel-basket design. The fuel basket fits within the inner cavity of the overpack, but small gaps
D-39

exist between the basket disks and the inner wall of the overpack and between the ends of the 

basket and the lid and bottom plate walls.

As in the Rail-Steel cask, each fuel assembly consists of an array of fuel rods, each separated 

by a helium gas space. The total number of rods in the fuel assembly, the dimensions of each 

rod, and the type of fuel cladding vary between assembly designs. Section D.3.1.2 more fully 

describes the fuel assembly and fuel rods.

The model does not explicitly include the fuel basket and fuel assemblies. Instead, a separate 

three-dimensional model was generated to obtain the effective properties of the basket in the 

in-plane and axial directions. Because the basket support disks, gas regions, and heat transfer 

disks repeat at regular intervals in the active fuel region, a three-dimensional quarter solid 

model of a section comprising two support disks and a heat transfer disk, as well as the gas and 

fuel tubes between them, was generated to obtain the effective properties of the basket in the 

in-plane and axial directions. The diameter of the support and heat transfer disks is assumed to 

be the same to simplify the solid modeling and mesh process. The same model is used for the 

portion of the fuel basket without the heat transfer disk. In this case, the material properties and 

boundary conditions for the heat transfer disk are replaced with those of the gas region.

D.4.1.3 Impact Limiters

The impact limiters in the Rail-Lead cask are cylindrical wood-filled structures, also encased in a 

thin metal shell. Each impact limiter is 3.15 m (124 in.) in diameter and 1.12 m (44 in.) in length 

(see Figure D-15). The depth of the cap where the overpack fits is 30.5 cm (12 in.). These 

limiters serve the same purpose as the impact limiters in the Rail-Steel cask (see 

Section D.3.1.3). Because the impact limiters are mostly wood and have very little metal in their 

structures, they are modeled as two coupled all-wood structures of the same volume and shape.

D.4.2 Rail-Lead Cask Thermal Behavior and Model Assumptions

The Rail-Lead cask is also designed to release heat passively under normal conditions of 

transport. In the directly loaded configuration, the basket is designed to accommodate a 

maximum heat load of 22.1 kW (0.85 kW per fuel assembly). Figure D-17 shows the 

normalized, axial heat generation rate distribution for a 0.85-kW PWR assembly. As with the 

Rail-Steel cask, heat is dissipated from the fuel rods to the exterior surfaces of the Rail-Lead 

cask by a combination of conduction, convection, and radiation heat transfer.
D-40

Figure D-17 Axial burnup profile for the directly loaded fuel basket

(from NAC International, 2004)

The fuel assembly design in the Rail-Lead cask is conceptually the same as in the Rail-Steel 

cask model; therefore, the same heat transfer mechanisms are present as described in 

Section D.3.2. The approach described in Section D.3.3.1 is also used in the Rail-Lead cask 

SAR to obtain the effective thermal conductivity of the fuel assembly in the radial direction. 

Values presented in the Rail-Lead cask SAR are used in this study and are not much different 

from those used in the Rail-Steel cask SAR, as expected. Heat generated in the assembly is 

dissipated by conduction through the fuel tube walls. From the tubes, heat is then radially 

dissipated by conduction through the support and heat transfer disks and through the gas in the 

void formed between the tubes and the inner wall of the overpack. Radiation to the adjacent 

tubes and disks, and to the inner wall of the overpack, also distributes heat. As in the Rail-Steel 

cask fuel basket, convection is limited to a few regions around the basket perimeter. However, 

unlike the HI-STAR configuration, the convective cells in the Rail-Lead cask fuel basket are 

confined to the gas void between adjacent disks. Moreover, heat dissipated from the adjacent 

disks through this void tends to decrease the temperature gradient across this void region, 

reducing temperature-gradient-induced flow. In the Rail-Lead cask model, convection is 

neglected in this region because it is not expected to be significant given the Nusselt values 

presented in the Rail-Steel cask SAR for a similar void configuration.

Heat is dissipated radially by conduction and radiation through the gap between the disks and 

the overpack inner wall. This gap is assumed to be 1.65 mm (0.065 in.) across, as stated in the 

Rail-Lead cask SAR. As mentioned before, a three-dimensional quarter section of the fuel 

basket is generated to obtain effective thermal conductivities in the in-plane and axial directions 

(see Figure D-18). The small gap between the disks and the inner wall of the overpack is 

included (not visible in Figure D-18). Except for convection, this model accounts for all modes 
D-41

of heat transfer, including radiation between the tubes, between the tubes and the disks, 

between the tubes and the inner shell (also not shown), between the disks, and between the 

disks and the inner shell. In the horizontal position, the disks make contact with the inner shell 

wall. To account for conduction through the contact area between the disks and the inner shell 

wall, the same method developed for the Rail-Steel cask is employed to enhance conductivity 

through the equivalent concentric gap (see Section D.3.3.2). Note that both the support and 

heat transfer disk diameters are assumed to be the same after thermal expansion.

Figure D-18 Three-dimensional quarter section of the directly loaded basket

The helium material is not shown.

Heat transfer through the inner shell, lead, and outer shell of the overpack occurs by conduction 

through the shell materials. These are modeled explicitly. As in the Rail-Steel cask, conduction 

in the neutron shield region occurs in parallel through the radial connectors and the neutron 

shield material.

Heat transfer from the cross-section of the directly loaded basket and overpack to the axial ends 

of the overpack is assumed to occur by conduction and radiation. Heat conduction occurs in 

parallel through each of the connecting materials that comprise the basket and overpack. The 

effective thermal properties are obtained in the same manner as in the Rail-Steel cask model. 

Radiation is assumed to occur from the end disks of the basket to the interior wall of the inner lid 

and bottom plate of the overpack.

The Rail-Lead cask system is also designed to keep the temperature of critical components 

below their design limits during and after a 30-minute, fully engulfing HAC scenario. For fire 

accident scenarios lasting longer than the HAC fire described in 10 CFR 71.73, a significant 

amount of heat may be transferred to the interior of the cask. As in the Rail-Steel cask, the 

neutron shield is expected to reach temperatures beyond its material temperature limits. Heat is 

then dissipated by conduction through the gas layer in the neutron shield and by radiation 

between the outer wall of the intermediate shell layer and the outer enclosure shell. Similarly, 

the lead shell is expected to melt because its melting point is around 321 degrees C 

Heat Transfer Disks

Support Disks

Fuel Assembly Fuel Tubes
D-42

(611 degrees F). The impact limiters are made of wood encased in a thin metal layer and are 

sealed to prevent moisture from deteriorating the wood over long periods of time. Because the 

impact limiters are assumed to stay intact (i.e., with the content sealed) after the initial accident 

event (e.g., derailment), the wood is not expected to char significantly. Therefore, this model 

does not take charring into account.

D.4.3 Rail-Lead Cask Materials and Thermal Properties

The Rail-Lead cask is made of stainless steel, lead, copper, aluminum, NS-4-FR, and Boral 

neutron absorber and is backfilled with helium. The inner and outer shell, the outer enclosure 

shell, the bottom plate, the top flange, and the inner lid of the overpack are made from type 304 

stainless steel. The gamma shield is made from copper-lead and the outer lid from type 630

stainless steel. The radial channels are made from a combination of type 304 stainless steel, 

copper, and a small section of carbon steel. The stainless steel in the channel is the main 

support component, while the copper enhances conduction through the channels. The 

overpack neutron shield is made from NS-4-FR. The impact limiters are redwood and balsa 

layers encased in a thin stainless steel shell.

In the fuel basket, the support disks, threaded rods, and spacer nuts are made from type 630 

stainless steel, and the top and bottom support plates, short cylinder, and fuel tubes are made 

from type 304 stainless steel. The heat transfer support disks are made from aluminum 

alloy 6061. As with the Rail-Steel cask, adjacent to each fuel tube wall is a layer of Boral 

sandwiched between the tube wall and a thin layer of stainless steel sheathing. The fuel rods 

are assumed to be Zircaloy as in the Rail-Steel cask analysis. The pellets are made from UO2. 

The empty gas space, which encompasses most of the volume inside the overpack cavity, is 

filled with helium.

Tables D-10 through D-13 provide the thermal conductivity, specific heat, density, and 

emissivity for those materials used in the Rail-Lead cask which differ from the Rail-Steel cask or 

for which the properties are significantly different (see Tables D-2 through D-5 for additional 

properties). The properties of NS-4-FR reported in the Rail-Lead cask SAR are marginally 

different from those reported for Holtite-A, as expected. The thermal conductivity of redwood 

and balsa vary depending on the direction of the grain. For balsa, values from the MSC Patran 

material database were used and compared well with values in Incropera and Dewitt (1996). 

Table D-10 gives the MSC Patran database references. NUREG-0361, “Safety Analysis Report 

for the Plutonium Air Transportable Package, Model PAT-1” (U.S. Nuclear Regulatory 

Commission, 1978), gives values through and along the grain for redwood; however, because

the Rail-Lead cask SAR does not specifically describe the arrangement of the wood layers in 

the limiters, average properties (along and through the grain) are assumed. The specific heat 

and density of copper-lead provided in the Rail-Lead cask SAR are slightly lower and higher, 

respectively, than for plain lead (Incropera and Dewitt, 1996); properties from the MSC Patran 

database are used instead because data are readily available above the melting temperature 

and include the specific latent heat of fusion (23.9 kilojoules per kilogram (kJ/kg) (10.3 Btu per 

pound of mass (Btu/lbm)). The specific heat of lead increases up to the melting point (by a 

factor of 1.07), but then remains approximately the same. Because these changes are small, 

the value at 92 degrees C (200 degrees F) is used throughout the rest of the temperature range.
D-43

Table D-10 Thermal Conductivities for the Rail-Lead Cask Materials

Material

Thermal Conductivity in W/m-°C (Btu/ft-hr-°F)

92°C 

(200°F)

226°C 

(450°F)

377°C 

(700°F)

477°C 

(900°F) 726°C (1,340°F)

Aluminum 6061∗ 171 (98.8) 176 (101.7) 176 (101.7)

Copper§ 402 (232.4) 386 (223.1) 376 (217.3) 369 (213.3) 352 (203.5)

Balsa¥ 0.050 

(0.029) ― ― ― ―

Lead‡ 33.9 (19.6) 29.3 (16.9) 16.7 (9.7) 15.3 (8.8) 14.7 (8.5)

Redwood€ 3.6 (2.0) 5.5 (3.1) ― ― ―

Stainless Steel Type 630∗ 17.5 (9.9) 18.3 (10.6) 20.7 (12.0) 24.6 (14.2) ―

∗ NAC International, 2004

§ Incropera and Dewitt, 1996 ¥ McAdams, 1954; Perry, 1963; Weast, 1966

‡ Kelley, 1960; Schorsch, 1966; Weast, 1966 € U.S. Nuclear Regulatory Commission, 1978

Dashes in the table indicate that properties at these temperatures were not available, or the material never 

reaches these temperatures. Where data is not available, it is assumed the values do not change with 

increasing temperature.

Table D-11 Specific Heat for the Rail-Lead Cask Materials

Material

Specific Heat in J/kg-K (Btu/lbm-F)

92°C (200°F) 226°C 

(450°F)

377°C (700°F) 477°C (900°F) 726°C 

(1,340°F)

Copper§ 390 (0.093) 406 (0.097) 422 (0.101) 431 (0.103) 451 (0.108)

Balsa¥ 2,302 (0.55) ― ― ― ―

Lead‡ 131 (0.031) ― ― ― ―

Redwood€ 2,386 (0.57) 3,898 (0.93) ― ― ―

§ Incropera and Dewitt, 1996 ¥ McAdams, 1954; Perry, 1963; Weast, 1966

‡ Kelley, 1960; Schorsch, 1966; Weast, 1966 € U.S. Nuclear Regulatory Commission, 1978

Dashes in the table indicate that properties at these temperatures were not available, or the material never 

reaches these temperatures. Where data is not available, it is assumed the values do not change with 

increasing temperature.
D-44

Table D-12 Densities for the Rail-Lead Cask Materials

Density in kg/m3 (lbm/ft3)

92°C (200°F) 226°C 

(450°F)

377°C 

(700°F)

477°C 

(900°F)

726°C 

(1,340°F)

Aluminum 6061* 2,823 (176) ― ― ― ―

Copper§ 8,933 (558) ― ― ― ―

Balsa¥ 130 (8.1) ― ― ― ―

Lead‡ 11,350 (709) ― ― ― ―

Redwood€ 352 (22) ― ― ― ―

∗ NAC International, 2004

§ Incropera and Dewitt, 1996 ¥ McAdams, 1954; Perry, 1963; Weast, 1966

‡ Kelley, 1960; Schorsch, 1966; Weast, 1966 € U.S. Nuclear Regulatory Commission, 1978

Dashes in the table indicate that properties at these temperatures were not available, or the material never 

reaches these temperatures. Where data is not available, it is assumed the values do not change with 

increasing temperature.

Table D-13 Emissivity for Some of the Rail-Lead Cask Materials

Material Emissivity

Aluminum 6061 0.22

Stainless Steel Type 630 0.58

With the exception of the basket and neutron region, all components are modeled explicitly. 

The impact limiters are modeled in their intact state, with properties of redwood and balsa, 

because the outer shell volume is significantly smaller than the total wood volume. Contact gap 

effects are assumed to be negligible. As in the Rail-Steel cask model, NS-4-FR is replaced with 

air when the former reaches its temperature limit, but only in the neutron shield region of the 

overpack. Radiation is activated in this region by setting the emissivity to the appropriate value.

D.4.3.1 Directly Loaded Fuel Basket

In the Rail-Lead cask SAR, fuel rods are evaluated to determine a representative fuel rod 

configuration. The fuel assembly is then modeled explicitly to obtain an equivalent in-plane 

thermal conductivity for the homogenized fuel assembly, as described in Section D.3.3.1. The 

fuel assembly’s axial conductivity is next obtained with an area-weighted average using the 

thermal conductivities of the individual components of the fuel rods and helium. The rest of the 

directly loaded basket with the homogenized fuel assembly is then included explicitly in the 

normal condition run, but is not included in the subsequent regulatory fire accident run. Instead, 

the maximum temperature difference between the fuel basket and the inner wall of the overpack 
D-45

calculated in the normal condition run is added to the inner wall temperature of the overpack 

calculated in the regulatory fire run to obtain an estimate of the temperature of the center of the 

fuel basket for the regulatory run. Because the Rail-Lead cask SAR did not provide 

homogenized properties for the fuel-basket region, this study used a different approach to 

obtain these properties. This alternate approach (1) reduces geometric modeling complexities 

while maintaining the overall response of the cask and (2) is consistent with the approach 

employed to model the Rail-Steel cask.

The directly loaded fuel basket is replaced with a homogenized cylinder having equivalent 

effective thermal conductivities in the in-plane and axial directions. As described in 

Section D.4.2, two variations of the same three-dimensional quarter section, finite element 

model are generated. The first model included two support disks, a heat transfer disk, and the 

fuel tubes and helium space between the disks (see Figure D-18). The second model did not 

include the heat transfer disk; instead, it is replaced with helium and the boundary conditions 

are modified to reflect this change.

Because the Rail-Lead cask SAR did not explicitly give the effective in-plane and axial 

conductivities for the fuel basket (i.e., the fuel basket was modeled explicitly in that SAR), these 

effective conductivities are obtained using the following four-step procedure.

First, the detailed cross-section of the fuel assembly is replaced with a homogenized fuel region 

having equivalent thermal properties. This analysis is done in the Rail-Lead cask SAR, as 

explained above, and this study includes the analysis results. As expected, the thermal 

conductivities reported in the Rail-Lead cask SAR are close to those reported in the Rail-Steel 

cask SAR for similar fuel assemblies, which serves as a check.

Second, the fuel tube, Boral, and stainless steel sheathing are replaced with a homogenized 

wall having an equivalent thermal conductivity, as described in the Rail-Lead cask SAR.

Third, both three-dimensional quarter section models described above (and shown in 

Figure D-18) are used to obtain the in-plane and axial effective thermal conductivities. Each 

model is evaluated with two sets of boundary conditions:

(1) a uniform temperature applied over the outer circumference of the inner shell, adiabatic 

conditions over the in-plane ends, and uniform heat generation in the homogenized fuel 

assemblies

(2) adiabatic conditions applied over the outer circumference of the inner shell, a uniform 

temperature over one of the in-plane ends, a uniform heat flux over the other in-plane 

end, and no uniform heat generation in the homogenized fuel assemblies

In the first case, the in-plane thermal conductivity is obtained using the same procedure 

described in Section D.3.3.1. In the second case, the axial thermal conductivity is obtained 

using the standard relationship:

(D-6) keff = qA

L(Tq−Tt)
D-46

Here A is the cross-sectional area of the basket; L is the thickness across the modeled section; 

q is the uniform heat flux applied over one of the cross-sectional area’s axial ends; and Tt is the 

uniform temperature applied over the other cross-sectional area’s axial end. Tq is the average 

temperature where uniform heat flux is applied and is calculated using the simulation results. 

A second option is to apply constant (but different) temperatures at both axial ends of the 

basket, then calculate the total heat flow (qA) through the basket using the simulation results, 

and lastly calculate the effective axial conductivity using the above equation. To obtain 

temperature-dependent thermal conductivities, this third step is repeated a number of times 

using a wide range of uniform circumferential temperatures and applied heat fluxes.

Fourth, the thermal conductivities obtained in the third step are added using an equivalent 

thermal resistor network model to obtain in-plane and axial thermal conductivities, respectively, 

over the entire fuel basket.

Table D-14 shows the thermal properties used for the basket. These properties are applied to 

the homogenized fuel-basket cylinder. The equivalent specific heat and density are obtained 

using a mass- and volume-weighted average, respectively, over the individual component 

properties. The Rail-Lead cask SAR gives the volume of each component in the fuel basket 

(i.e., support disks, heat transfer disks, and fuel tubes).

Table D-14 Effective Thermal Properties of the Directly Loaded Fuel Basket

Effective Thermal Properties 92°C 

(200°F)

226°C 

(450°F)

377°C 

(700°F)

477°C 

(900°F)

726°C 

(1,340°F)

Radial Thermal Conductivity in 

W/m-°C (Btu/ft-hr-°F) 3.2 (1.8) 3.8 (2.1) 4.3 (2.4) 5.0 (2.8) 5.9 (3.4)

Axial Thermal Conductivity in 

W/m-°C (Btu/ft-hr-°F) 2.4 (1.4) 3.2 (1.8) 3.8 (2.1) 4.5 (2.6) 5.8 (3.3)

Specific Heat in

J/kg-°C (Btu/lbm-°F) 332 (0.079)

Density in kg/m3 (lbm/ft3) 2,450 (153)

D.4.3.2 Neutron Shield Region

The neutron shield region is modeled using the same approach as that used in the SARs (NAC

International, 2004; Holtec International, 2000). Both reports used the thermal resistor network 

method to obtain the in-plane and axial effective thermal conductivities (see Section D.3.3.4). In 

the case of the Rail-Lead cask, there are fewer radial channels than in the Rail-Steel cask; 

however, as will be demonstrated shortly, this shortcoming is compensated for by adding 

copper in the neutron shield region. Table D-15 shows the thermal properties used for the 

neutron shield region in the Rail-Lead cask. The circumferential thermal conductivity is 

assumed to be that of NS-4-FR. As before, the specific heat and density are obtained from a 

mass- and area-weighted average. Note that the thermal conductivity is slightly higher than in 

the Rail-Steel cask even though the Rail-Lead cask contains fewer channels. This is expected 
D-47

because the neutron shield in the Rail-Lead cask contains copper, which has a much higher 

thermal conductivity than carbon steel.

Table D-15 Effective Thermal Conductivities for the Neutron Shield Region of the 

Rail-Lead Cask

92°C 

(200°F)

226°C 

(450°F)

377°C 

(700°F)

477°C 

(900°F)

726°C 

(1,340°F)

In-Plane Thermal Conductivity in 

W/m-°C (Btu/ft-hr-°F) 8.1 (4.6) 7.9 (4.5) 7.7 (4.4) 7.7 (4.4) 7.4 (4.2)

Axial Thermal Conductivity in 

W/m-°C (Btu/ft-hr-°F) 7.6 (4.3) 7.3 (4.2) 7.3 (4.2) 7.2 (4.1) 6.9 (3.9)

Specific Heat in

J/kg-°C (Btu/lbm-°F) 1,406 (0.33) 535 (0.12) 563 (0.13) 575 (0.13) 611 (0.14)

Density in kg/m3 (lbm/ft3) 1,983 (123) 380 (23)

D.4.4 Rail-Lead Cask Finite Element Model

The following description is short because most of the details are similar to those of the 

Rail-Steel cask analysis described in Section D.3.4. In the Rail-Lead cask runs, the cask model 

had 109,662 elements (see Figure D-19); this corresponds to a nominal element size of 10.2 cm 

(4 in.). The element count is less than in the Rail-Steel cask because the Rail-Lead cask is 

smaller and has fewer features which add to the element count. A mesh refinement study was 

also conducted with the Rail-Lead cask model with a similar outcome. The boundary conditions 

for the normal condition, steady-state run, the regulatory uniform-heating run, and the CAFE fire 

runs are the same as those discussed in Sections D.2 and D.3.4. They are not repeated here.

Figure D-19 The Rail-Lead cask mesh
D-48

Overall, maximum temperatures obtained using the model developed here and in the Rail-Lead 

cask SAR are also similar. Again, the difference in purpose of the two analyses leads to some 

different assumptions, which in turn lead to slightly different results.

D.4.5 Rail-Lead Cask Thermal Analysis Results

The following figures (Figure D-20 through Figure D-24) show additional results for the 

Rail-Lead cask not provided in Chapter 4. Figure D-20 shows results for the regulatory 

uniform-heating case. (Recall this is a P/Thermal-only run.) Figure D-21 shows results for the 

regulatory CAFE fire; Figure D-22 shows results for the fully engulfing CAFE fire run with the 

cask on the ground; and Figures D-23 and D-24 show results for the CAFE fire runs with the 

cask on the ground and outside the pool area. As with the Rail-Steel cask, the last three cases 

are run for a total of 3 hours. Chapter 4 discusses these results and their implications.
D-49

Figure D-20 Rail-Lead cask regulatory uniform-heating results

0

20

40

60

80

100

120

140

160

180

200

0 2 4 6 8 10 12 14

Temperature (°C)

Time (hrs)

Seal Temperature

Seal_Inner-Near_Side

Seal_Inner-Top

Seal_Inner-Far_Side

Seal_Inner-Bottom

Seal_Outer-Near_Side

Seal_Outer-Top

Seal_Outer-Far_Side

Seal_Outer-Bottom

0

50

100

150

200

250

300

350

0 2 4 6 8 10 12 14

Temperature (°C)

Time (hrs)

Fuel Temperature

Fuel_Region-Center

Fuel_Edge-Near_Side

Fuel_Edge-Top

Fuel_Edge-Far_Side

Fuel_Edge-Bottom
D-50

Figure D-20 Rail-Lead cask regulatory uniform-heating results (continued)

0

50

100

150

200

250

300

0 5 10 15

Temperature (°C)

Time (hrs)

Lead Temperature

Lead_Inner_Surface-Near_Side

Lead_Inner_Surface-Top

Lead_Inner_Surface-Far_Side

Lead_Inner_Surface-Bottom

Lead_Outer_Surface-Near_Side

Lead_Outer_Surface-Top

Lead_Outer_Surface-Far_Side

Lead_Outer_Surface-Bottom

0

100

200

300

400

500

600

700

800

0 5 10 15

Temperature (°C)

Time (hrs)

Outer Surface Temperature

Cask_Outer_Surface-Near_Side

Cask_Outer_Surface-Top

Cask_Outer_Surface-Far_Side

Cask_Outer_Surface-Bottom
D-51

Figure D-21 Rail-Lead cask CAFE regulatory fire

0

20

40

60

80

100

120

140

160

180

0 2 4 6 8 10 12 14

Temperature (°C)

Time (hrs)

Seal Temperature

Seal_Inner-Near_Side

Seal_Inner-Top

Seal_Inner-Far_Side

Seal_Inner-Bottom

Seal_Outer-Near_Side

Seal_Outer-Top

Seal_Outer-Far_Side

Seal_Outer-Bottom

0

50

100

150

200

250

300

350

400

0 2 4 6 8 10 12 14

Temperature (°C)

Time (hrs)

Fuel Temperature

Fuel_Region-Center

Fuel_Edge-Near_Side

Fuel_Edge-Top

Fuel_Edge-Far_Side

Fuel_Edge-Bottom
D-52

Figure D-21 Rail-Lead cask CAFE regulatory fire (continued)

0

50

100

150

200

250

300

350

0 5 10 15

Temperature (°C)

Time (hrs)

Lead Temperature

Lead_Inner_Surface-Near_Side

Lead_Inner_Surface-Top

Lead_Inner_Surface-Far_Side

Lead_Inner_Surface-Bottom

Lead_Outer_Surface-Near_Side

Lead_Outer_Surface-Top

Lead_Outer_Surface-Far_Side

Lead_Outer_Surface-Bottom

0

100

200

300

400

500

600

700

800

900

1000

0 5 10 15

Temperature (°C)

Time (hrs)

Outer Surface Temperature

Cask_Outer_Surface-Near_Side

Cask_Outer_Surface-Top

Cask_Outer_Surface-Far_Side

Cask_Outer_Surface-Bottom
D-53

Figure D-22 Rail-Lead cask on ground at the pool center

0

50

100

150

200

250

300

350

0 2 4 6 8 10 12 14 16

Temperature (°C)

Time (hrs)

Seal Temperature

Seal_Inner-Near_Side

Seal_Inner-Top

Seal_Inner-Far_Side

Seal_Inner-Bottom

Seal_Outer-Near_Side

Seal_Outer-Top

Seal_Outer-Far_Side

Seal_Outer-Bottom

0

100

200

300

400

500

600

700

800

0 2 4 6 8 10 12 14 16

Temperature (°C)

Time (hrs)

Fuel Temperature

Fuel_Region-Center

Fuel_Edge-Near_Side

Fuel_Edge-Top

Fuel_Edge-Far_Side

Fuel_Edge-Bottom
D-54

Figure D-22 Rail-Lead cask on ground at the pool center (continued)

0

100

200

300

400

500

600

700

800

900

0 5 10 15 20

Temperature (°C)

Time (hrs)

Lead Temperature

Lead_Inner_Surface-Near_Side

Lead_Inner_Surface-Top

Lead_Inner_Surface-Far_Side

Lead_Inner_Surface-Bottom

Lead_Outer_Surface-Near_Side

Lead_Outer_Surface-Top

Lead_Outer_Surface-Far_Side

Lead_Outer_Surface-Bottom

0

200

400

600

800

1000

1200

0 5 10 15 20

Temperature (°C)

Time (hrs)

Outer Surface Temperature

Cask_Outer_Surface-Near_Side

Cask_Outer_Surface-Top

Cask_Outer_Surface-Far_Side

Cask_Outer_Surface-Bottom
D-55

Figure D-23 Rail-Lead cask on ground 3.0 m (10 feet) from the edge of the pool

0

50

100

150

200

250

300

350

400

450

0 2 4 6 8 10 12 14 16

Temperature (°C)

Time (hrs)

Fuel Temperature

Fuel_Region-Center

Fuel_Edge-Near_Side

Fuel_Edge-Top

Fuel_Edge-Far_Side

Fuel_Edge-Bottom

0

50

100

150

200

250

300

350

0 2 4 6 8 10 12 14 16

Temperature (°C)

Time (hrs)

Seal Temperature

Seal_Inner-Near_Side

Seal_Inner-Top

Seal_Inner-Far_Side

Seal_Inner-Bottom

Seal_Outer-Near_Side

Seal_Outer-Top

Seal_Outer-Far_Side

Seal_Outer-Bottom
D-56

Figure D-23 Rail-Lead cask on ground 3.0 m (10 feet) from the edge of the pool (continued)

0

50

100

150

200

250

300

350

400

450

500

0 5 10 15 20

Temperature (°C)

Time (hrs)

Lead Temperature

Lead_Inner_Surface-Near_Side

Lead_Inner_Surface-Top

Lead_Inner_Surface-Far_Side

Lead_Inner_Surface-Bottom

Lead_Outer_Surface-Near_Side

Lead_Outer_Surface-Top

Lead_Outer_Surface-Far_Side

Lead_Outer_Surface-Bottom

0

100

200

300

400

500

600

700

0 5 10 15 20

Temperature (°C)

Time (hrs)

Outer Surface Temperature

Cask_Outer_Surface-Near_Side

Cask_Outer_Surface-Top

Cask_Outer_Surface-Far_Side

Cask_Outer_Surface-Bottom
D-57

Figure D-24 Rail-Lead cask on ground 18.3 m (60 feet) from the edge of the pool

0

20

40

60

80

100

120

140

160

180

0 2 4 6 8 10 12 14 16

Temperature (°C)

Time (hrs)

Seal Temperature

Seal_Inner-Near_Side

Seal_Inner-Top

Seal_Inner-Far_Side

Seal_Inner-Bottom

Seal_Outer-Near_Side

Seal_Outer-Top

Seal_Outer-Far_Side

Seal_Outer-Bottom

0

50

100

150

200

250

300

350

0 2 4 6 8 10 12 14 16

Temperature (°C)

Time (hrs)

Fuel Temperature

Fuel_Region-Center

Fuel_Edge-Near_Side

Fuel_Edge-Top

Fuel_Edge-Far_Side

Fuel_Edge-Bottom
D-58

Figure D-24 Rail-Lead cask on ground 18.3 m (60 feet) from the edge of the pool (continued)

0

50

100

150

200

250

0 5 10 15 20

Temperature (°C)

Time (hrs)

Lead Temperature

Lead_Inner_Surface-Near_Side

Lead_Inner_Surface-Top

Lead_Inner_Surface-Far_Side

Lead_Inner_Surface-Bottom

Lead_Outer_Surface-Near_Side

Lead_Outer_Surface-Top

Lead_Outer_Surface-Far_Side

Lead_Outer_Surface-Bottom

0

50

100

150

200

250

300

350

0 5 10 15 20

Temperature (°C)

Time (hrs)

Outer Surface Temperature

Cask_Outer_Surface-Near_Side

Cask_Outer_Surface-Top

Cask_Outer_Surface-Far_Side

Cask_Outer_Surface-Bottom
D-59

D.5 Truck Cask with Depleted Uranium

The Truck-DU cask is slightly different from the two previously analyzed casks. This cask is 

certified to transport up to four PWR spent fuel assemblies on a truck flat bed and uses depleted 

uranium (DU) for the gamma shield. In this analysis, the cask is assumed to be in the horizontal 

configuration, as it would most likely be after an accident scenario. For the Truck-DU cask, 

results reported in the Truck-DU SAR (General Atomics, 1998) are used but modified where 

necessary to reflect the current study.

D.5.1 Geometric Considerations

The Truck-DU consists of an overpack, a fuel basket, and limiters at each end. Like the 

Rail-Lead cask, the Truck-DU is a single containment cask with no MPC. Compared to the 

Rail-Steel and Rail-Lead casks, however, this cask is smaller in size (1.00 m (39.8 in.) in 

diameter at the center, 2.3 m (90 in.) in diameter at the impact limiters, and 5.94 m (234 in.) in 

length) because it carries only four spent fuel assemblies. Figure D-25 shows the layout of the 

Truck-DU cask.

Figure D-25 Components of Truck-DU cask (from General Atomics, 1998)

D.5.1.1 Fuel Assembly and Interior Cavity of the Overpack

The inner cavity of the Truck-DU is a rectangular box measuring 0.46 × 0.46 m (18 × 18 in.) in 

the cross-section and 4.25 m (167 in.) long. Inside this cavity, the fuel assemblies are stored 

within four slots formed by a steel fuel-support structure (FSS). Section D.3.1.2 discusses the 

details of the fuel assembly. The FSS is made from four panels that are 0.016 m (0.61 in.) thick 

and are arranged in a perpendicular cross pattern. This section of this document refers to the 

fuel assembly and FSS together as the fuel basket. Fuel spacers and other support structures 

complete the remaining space at the ends of the fuel basket. This section of this document 

refers to these regions as the fuel basket ends.

INNER HONEYCOMB 

OUTER HONEYCOMB 

END HONEYCOMB 

CORNER HONEYCOMB 
D-60

In this analysis, the fuel basket and the fuel-basket end regions are each represented as single 

volumes to minimize geometric complexity, but their thermal response is accounted for using 

effective properties.

D.5.1.2 Overpack

The overpack center cross-section is made from a five-layer cross-section. The first three inner 

layers are square with rounded corners. The first layer, the cavity liner, is a thin steel wall 

(9.5 mm (0.376 in.) thick) that separates the contents of the cask from the gamma shield. The 

second layer is a thick wall (6.7 cm (2.6 in.) at the center of the cask) of depleted uranium (DU)

which serves as the gamma shield. The third and last square layer is a thick wall (7.6 cm 

(3 in.)) of steel. In the axial direction, the DU layer tapers off and extends just past the ends of 

the inner cavity of the overpack. The cavity liner and the thick steel wall extend almost to the 

axial ends of the gamma shield. The cavity liner and the thick steel wall mate with a 

square-shaped steel flange at the top of the overpack and a square-shaped metal base cup at 

the bottom. The inner cavity of the overpack is sealed off from the environment using a steel lid 

(0.28 m (11 in.) thick at the center) which fits on the flange, as shown in Figure D-25. The metal 

base cup is 0.24 m (9.5 in.) thick.

The last two layers, the neutron shield and the thin steel outer skin wall (1 cm (0.4 in.) thick), 

form the rest of the center cross-section of the overpack. The outer surface of the neutron 

shield layer and the outer skin wall are circular in shape. In the axial direction, the neutron 

shield and the outer skin wall extend the interior plane wall of the impact limiters. Both layers 

mate with an impact limiter support structure (ILSS) at these extreme ends. The ILSS is design 

to support the impact limiters using a series of ribs, 1.9 cm (0.75 in.) thick, that extend radially 

outward from the exterior surface of the thick steel wall to the interior surface of the outer shell 

wall of the ILSS. The spaces between these ribs and between the exterior surface of the thick 

steel wall and the interior surface of the outer shell wall of the ILSS are filled with a neutron 

shield material. In the axial direction, the ILSS extends to the end of the lid at the top and to the 

metal base cup at the bottom.

The thermal model explicitly represents the cavity liner, gamma shield, thick steel wall, flange, 

base metal cup, lid, neutron shield region, and outer skin wall, with minor alterations to simplify 

the model. The ILSSs are represented as single volumes to minimize geometric complexity, but 

their thermal response is accounted for appropriately using effective properties. As with the rail 

casks, the Truck-DU cask overpack contains a number of features that serve a special purpose 

(e.g., valves, seals, and trunnions). The model omits these features for the same reasons they 

were omitted from the rail cask models—the effects of these features are assumed to be either 

(1) negligible because of their small volume and mass relative to the other components in the 

overpack, (2) highly localized with no effect to the overall thermal performance of the cask at 

locations of interest, or (3) both.

D.5.1.3 Impact Limiters

For the Truck-DU cask, impact limiters bolt to the top and bottom of the overpack. These 

impact limiters are similar to the Rail-Steel limiters in that they are made of aluminum 

honeycomb material encased in a thin steel shell. Figure D-25 illustrates the arrangement of 

the honeycomb material.
D-61

In this model, the impact limiters were assumed undamaged; hence, they are modeled using the 

geometry shown in Figure D-25. The encasing steel shell is neglected because the total volume 

of the shell is small compared to that of the rest of the honeycomb material.

D.5.2 Truck-DU Thermal Behavior and Model Assumptions

Like the rail casks, the Truck-DU cask is designed to release heat passively under normal 

conditions of transport. The Truck-DU fuel basket is designed to accommodate a maximum 

heat load of 2,468 W (a maximum of four fuel assemblies at 617 W per assembly). Table D-16

shows the normalized, axial heat generation rate distribution for a 617-W PWR assembly. This 

axial heat generation profile is applied over the active fuel region which encompasses only 

about 3.66 m (144 in.) of the total fuel assembly length.

Table D-16 Axial Burnup Profile in the Active Fuel Region of the Truck-DU Cask

Axial Distance from Bottom of Active 

Fuel (Percent of Active Fuel Length) Normalized Value

0 to 1.4 0.432

1.4 to 4.2 0.630

4.2 to 7.6 0.847

7.6 to 11.1 0.964

11.1 to 15.3 1.09

15.3 to 24.3 1.22

24.3 to 38.9 1.22

38.9 to 66.0 1.09

66.0 to 77.9 0.964

77.9 to 84.7 0.847

84.7 to 91.7 0.630

91.7 to 96.3 0.432

96.3 to 100 0.252

As with the rail casks, heat is dissipated from the fuel rods to the exterior surfaces of the 

Truck-DU cask by a combination of conduction, convection, and radiation heat transfer. Heat 

transfer from the fuel assemblies to the outer surface of the overpack and the limiters is similar 

to the other rail casks. The only exception is that there are fewer large voids through the 

cross-section of this cask. Heat dissipation from the center cross-section of the cask is 

predominately by conduction and radiation through the fuel assembly and the FSS. Conduction 

dominates through the overpack cross-section. In the axial direction, radiation occurs between 

the ends of the fuel assembly and inner cavity wall. Conduction through the honeycomb 

material is complex; however, effective properties found in the Truck-DU cask SAR are used to 

obtain the thermal response of the impact limiters.

The Truck-DU cask is also designed to keep the temperature of critical components below their 

design limits during and after a 30-minute, fully engulfing, HAC scenario. For fire accident 

scenarios lasting longer than the HAC fire described in 10 CFR 71.73, a significant amount of 

heat may be transferred to the interior of the cask. As in the rail casks, the temperature of the 
D-62

neutron shield material is expected to reach temperatures beyond its operational temperature 

limit. Heat then is assumed to be dissipated by conduction through a gas layer in the neutron 

shield region and by radiation between the outer surface of the thick steel wall layer and the 

inner surface of the outer skin wall.

D.5.3 Truck-DU Materials and Thermal Properties

The Truck-DU cask is made of stainless steel, DU, copper, aluminum, hydrogenous neutron 

absorber, Boral neutron absorber (B4C), and helium. With the exception of spacers, bolts, and 

the lifting trunnions, which this analysis ignores, all major components of the overpack are made 

from type XM-19 stainless steel. The outer skin wall of the overpack is made from a 

combination of XM-19 stainless steel and copper. The stainless steel serves as the main 

support component, while the copper enhances conduction in the axial direction. The neutron 

shield material is made from a hydrogenous material which continues to function above 

149 degrees C (300 degrees F). The impact limiters are made from aluminum alloy materials of 

various densities. This study ignores the stainless steel shell (XM-11 and XM-19) encasing the 

honeycomb material. As with previous fuel-basket wall materials, the FSS is made from 

stainless steel and B4C.

With the exception of the XM-19 stainless steel and the honeycomb material, all material 

properties can be found in Sections D.3.3 and D.4.3. Tables D-17 and D-18 show the material 

properties used for the XM-19 stainless steel and the honeycomb material. The honeycomb 

material is classified by location in the limiter (see Figure D-25).

Table D-17 Thermal Conductivities for the Truck-DU Cask Materials

Material

Thermal Conductivity in W/m-°C (Btu/hr-ft-°F)

92°C 

(200°F)

226°C 

(450°F)

377°C 

(700°F)

477°C 

(900°F)

726°C 

(1,340°F)

XM-19 12.3 (7.1) 15.2 (8.8) 17.0 (9.8) 18.7 (10.7) 22.8 (13.2)

Inner Honeycomb kr/kz 8.7/2.6 (5.0/1.5)

Outer Honeycomb kr/kz 6.5/2.0 (3.8/1.2)

Corner Honeycomb kr/kz 1.7/2.9 (0.98/1.8)

End Honeycomb kr/kz 2.6/8.6 (1.5/5.0)

Table D-18 Volumetric Specific Heat for the Truck-DU Cask Materials

Material Volumetric Specific Heat (ρCp)

in J/m3-°C (Btu/ft3-°F)

XM-19 4,287,264 (63.9)

Inner Honeycomb 155,300 (2.3)

Outer Honeycomb 117,000 (1.7)

Corner Honeycomb 39,290 (0.58)

End Honeycomb 155,300 (2.3)
D-63

D.5.3.1 Effective Thermal Properties

Effective properties were used for the active fuel-basket region, the ends of the fuel basket, the 

neutron shield region, the ILSS region, and the outer skin wall (see Tables D-19 and 

Table D-20). These properties were obtained from the Truck-DU cask SAR (General 

Atomics, 1998). For the HAC scenarios, the hydrogenous neutron shield material was replaced 

with air above 149 degrees C (300 degrees F). Recall that radiation heat transfer was added 

between the outer surface of the thick steel wall layer and the inner surface of the outer skin 

wall to increase heat transfer to the interior of the cask during the fire (as was done in the rail 

cask analysis).

Table D-19 Effective Thermal Conductivities for the Truck-DU Cask Materials

Material

Thermal Conductivity in W/m-°C (Btu/hr-ft-°F)

92°C (200°F) 226°C 

(450°F)

377°C 

(700°F)

477°C 

(900°F)

726°C 

(1,340°F)

Active Fuel Region kr/kz

1.2/4.5 

(1.0/3.8)

1.8/4.9 

(1.5/4.2)

2.3/5.2 

(2.0/4.4)

2.8/5.7 

(2.4/4.9)

4.9/7.3 

(4.2/6.2)

Fuel Region Ends kr/kz

0.28/3.3 

(0.24/2.8)

0.31/3.8 

(0.26/3.2)

0.33/4.1 

(0.28/3.5)

0.35/4.6 

(0.30/3.9)

0.40/6.3 

(0.34/5.3)

Neutron Shield Region 

kr/kz

1.7/0.15 

(1.5/0.12) N/A N/A N/A N/A

ILSS kr_bottom/kr_top/kz

2.2/3.9/0.85 

(1.8/3.3/0.72

)

2.8/4.9/1.0 

(2.3/4.1/0.85

)

3.2/5.6/1.2 

(2.7/4.8/1.0)

3.5/6.1/1.3 

(3.0/5.2/1.1)

4.2/7.5/1.5 

(3.6/6.4/1.3)

Outer Skin Wall kr/kz

12.2/41.5 

(10.4/35.4)

15.2/44.0 

(13.0/37.5)

17.0/45.5 

(14.5/38.8)

18.6/47.0 

(15.9/40.0)

22.8/50.3 

(19.4/42.9)

Table D-20 Effective Volumetric Specific Heat for the Truck-DU Cask Materials

Material Volumetric Specific Heat (ρCp)

in J/m3-°C (Btu/ft3-°F)

Active Fuel Region 938,700 (25.2)

Fuel Region Ends 1,263,000 (33.9)

Neutron Region 1,715,000 (46.0)

ILSS 1,225,000 (32.8)

Outer Skin Wall 3,882,000 (104.2)

D.5.4 Truck-DU P/Thermal Finite Element Model

In the Truck-DU runs, the cask model had 241,700 elements (see Figure D-26). The element 

count is higher than in the rail cask analysis because the Truck-DU has a number of smaller 

features that add to the element count. The boundary conditions for the normal-condition 

steady-state run; the regulatory uniform-heating run; and the CAFE fire run are the same as 
D-64

discussed in Sections D.2 and D.3.4. In this analysis, the fire is run for only 1 hour. This 

timeframe corresponds to the total fuel burning time for the maximum-capacity, fully loaded fuel 

tanker truck.

Overall, maximum temperatures obtained in the normal condition, steady-state run and in the 

regulatory uniform-heating case using the model developed for this study are similar to the 

results presented in the Truck-DU cask SAR. Again, the difference in purpose of the two 

analyses leads to some different assumptions, which in turn lead to slightly different results.

Figure D-26 Truck-DU cask mesh

D.5.5 Truck-DU Cask Thermal Analysis Results

For the Truck-DU cask, only one CAFE nonregulatory fire was run—simulating the cask being 

on the ground and at the center of the pool (see Figure D-27). This is the most severe case, as 

demonstrated in the Rail-Steel and Rail-Lead cask analyses. Figure D-28 shows additional 

results for this case not provided in Chapter 4. A discussion of these results and their 

implications is provided in Chapter 4.
D-65

Figure D-27 CAFE three-dimensional domain with the Truck-DU cask on the ground
D-66

Figure D-28 Truck-DU cask on the ground at the pool center

0

50

100

150

200

250

0 5 10 15 20 25 30 35 40

Temperature (°C)

Time (hrs)

Seal Temperature

Seal-Flat-Side

Seal-Upper-Corner

0

50

100

150

200

250

300

350

0 5 10 15 20 25 30 35 40

Temperature (°C)

Time (hrs)

Fuel Temperature

Fuel_Region-Center

Fuel_Edge-Flat-Side

Fuel_Edge-Corner
D-67

Figure D-28 Truck-DU cask on ground at the pool center (continued)

D.6 Fire Environment Benchmarks for Container Analysis

Large, fully engulfed objects have a great impact on the surrounding fire environment. To 

adequately predict incident heat flux to casks, computational fluid dynamics models must be 

employed with appropriate boundary conditions. Also, because of the impact that massive 

objects have on fires, computational fluid dynamics models must be validated against 

experimental data from tests that have similarly sized objects (Nicolette and Larson, 1989).

Since the development of the CAFE code (del Valle, et al., 2009; del Valle, 2007; 

Are, et al., 2005; Lopez, et al., 2003), there has been a continuing effort to benchmark and 

fine-tune this fire model by making use of relevant empirical data from experiments. Continuing 

with this effort, before running the cases described in Chapter 4, CAFE is benchmarked against 

experimental data obtained from two fire test series conducted at Sandia National Laboratory’s

Lurance Canyon Burn Site: (1) one using a large calorimeter in the center of the pool 

(Greiner, 2009; Kramer, 2008) and (2) the other using a smaller diameter calorimeter adjacent 

to the fire (Lopez et al., 2003). The large calorimeter is close to the size of the casks analyzed 

in this study and had a test setup and conditions that closely matched the regulatory 

hypothetical fire accident scenario outlined in 10 CFR 71.73 for certification of SNF 

transportation casks. The smaller-diameter calorimeter test is used to benchmark CAFE's 

ability to predict heat flux to objects outside the fire plume. This section briefly describes these 

experiments and shows benchmark results.

D.6.1 Large Calorimeter Test and Benchmark Results

The large calorimeter is a carbon steel cylindrical pipe approximately 2.43 m (96 in.) in diameter 

and 4.6 m (180 in.) in length, with nominal walls 2.54 cm (1 in.) thick, and had bolted lids on 

0

200

400

600

800

1000

1200

0 1 2 3 4 5

Temperature (°C)

Time (hrs)

Outer Surface Temperature

Cask_Outer_Surface-Side

Cask_Outer_Surface-45_Deg
D-68

each end (see Figure D-29(a)). The calorimeter is placed on two stands at the center of a fuel 

pool 7.93 m (26 feet) in diameter. The stands kept the calorimeter 1 m (39.4 in.) above the fuel 

surface. Approximately 7,500 liters (2,000 gallons) of JP8 jet fuel are used for each test. Total 

burn time varies with each test, but is at least 25 minutes. All tests are conducted in relatively 

low wind conditions (less than 5 m/s (11 mph)) to ensure that the calorimeter is fully or partially 

engulfed (see Figure D-29(b)).

(a) (b)

Figure D-29 Large calorimeter fire test: (a) test setup and (b) fire fully engulfing the 

calorimeter

Thermocouples (TCs) are installed on the interior walls of the calorimeter to measure interior 

surface temperatures. All TCs are installed in a ring configuration as shown in Figure D-30. 

Heat flux gages are placed just outside the round walls of the calorimeter in a ring configuration 

and outside the lids to measure incident heat fluxes close to the outer walls of the calorimeter. 

Fuel burn rates are measured using a TC rake—a linear array of TCs traversing the depth of the 

fuel layer at known distance intervals. Directional flow probes are installed just outside of the 

calorimeter walls to measure the flow speed of hot gases near the calorimeter walls. Finally, 

ultrasonic sensors placed on four towers—two sensor towers aligned with the calorimeter lids 

and two sensor towers perpendicular to the cylindrical section of the calorimeter, but on 

opposite sides—are used to measure windspeed and wind direction. Each tower is 

approximately 24.4 m (80 feet) from the center of the pool and has three ultrasonic sensors

placed 2, 8 and 10 m (6.5, 26.2, and 32.8 feet) from the ground.
D-69

Fuel Pool Instrumentation

Heat Flux 

Gages

Fuel Surface

Insulated

Wires

39"

72"

Heat Flux 

Gages

Thermocouple 

Rings

Figure D-30 Side view (looking from the north) of calorimeter and test setup

Note: The calorimeter is centered within the pool. This drawing is not to scale.

Figure D-31a shows average temperatures along the four circumferential sides of the 

calorimeter obtained from Test 1 and from the CAFE benchmark run. Data from Test 1 were

chosen because the wind conditions and fire characteristics of this test best matched the 

regulatory conditions specified in 10 CFR 71.73 and the fire scenarios analyzed in this study. 

The test readings were taken from TCs located at 0 degrees (north side; that is, pointing out of 

the page), 90 degrees (top side), 180 degrees (south side; that is, pointing into the page), and 

270 degrees (underneath). This plot illustrates that average temperature predictions obtained 

from CAFE envelop the average temperatures readings from the test.

(a) (b)

Figure D-31 CAFE benchmark results using fully engulfed large calorimeter: 

(a) temperatures averaged along the 0-, 90-, 180-, and 270-degree side looking at the 

calorimeter from the negative z-direction and (b) temperatures averaged over each ring 

starting from Ring 1 located on the positive side of the z-axis

From this perspective, CAFE overpredicts temperatures underneath and on the south side of 

the calorimeter and underpredicts temperatures on the top of the calorimeter. Figure D-31b 

shows a plot of average temperatures over each TC ring starting from the left side of the 

0

100

200

300

400

500

600

700

800

900

1000

0 5 10 15 20 25 30 35

Temperature (°C)

Time (min)

Test 1: 0 degrees

Test 1: 90 degrees

Test 1: 180 degrees

Test 1: 270 degrees

CAFE: 0 degrees

CAFE: 90 degrees

CAFE: 180 degrees

CAFE: 270 degrees

0

100

200

300

400

500

600

700

800

900

1000

0 5 10 15 20 25 30 35

Temperature (°C)

Time (min)

Test 1: Ring 1

Test 1: Ring 2

Test 1: Ring 3

Test 1 Ring 4

Test 1: Ring 5

CAFE: Ring 1

CAFE: Ring 2

CAFE: Ring 3

CAFE: Ring 4

CAFE: Ring 5

y

z
D-70

calorimeter and moving along the negative z-axis, as shown in Figure D-30. From this 

perspective, CAFE predicts the average temperatures over the rings reasonably well.

Closer inspection of the temperature histories obtained from CAFE at each of the nodes 

corresponding to TC locations revealed excellent agreement with test data over most of the 

cask, except at locations where the wind effects are strongest—the last two rings to the right of 

Figure D-30 at 90 degrees (top side), 180 degrees (south side) and 270 degrees (underneath). 

Temperatures at 180 and 270 degrees are higher than expected, while temperatures at 

90 degrees are underpredicted. Differences rapidly diminished going from the rings on the right 

side of the calorimeter to the rings on the left side, as shown in Figure D-30. Part of the reason 

for these discrepancies is the way in which the computational fluid dynamics model applies the 

wind boundary conditions. In the large calorimeter test series, windspeeds are obtained only at 

four locations around the pool and at three heights. These height-dependent data are applied 

uniformly over the corresponding cross-sections of the domain, which does not necessarily 

reflect the actual conditions in the test. This leads to windspeeds being higher than expected in 

some locations around the casks, such as the south side of the cask near Ring 5 (the rightmost 

ring in Figure D-30).

D.6.2 Small Calorimeter Test and Benchmark Results

Experimental data from a smaller pipe calorimeter is used to benchmark the view factor method 

used in CAFE (Lopez et al., 2003). The CAFE model for thermal radiation transport within and 

near large hydrocarbon fires is divided into two types—diffusive radiation inside the flame zone 

and clear air or view factor radiation outside the flame zone. Outside the flame zone, thermal 

radiation transport is modeled by the clear air or view factor method. The calculation of the view 

factor between the fire and an adjacent object is complicated by the fact that the outer surface 

of a fire (or smoky region) is dynamically changing as a result of the puffing and turbulent nature 

of flames (Lopez et al., 2003).

In the experiments, a calorimeter is positioned such that its axis is 1.5 m (4.9 feet) away from 

the center of the fuel pool. The wind blew the fire away from the calorimeter leaving a 

significantly larger gap between the pipe calorimeter and the plume. Figure D-32 presents the 

results from tests and CAFE. The temperatures shown are at the center ring of this calorimeter. 

The blue lines are obtained from experimental data and the black lines are obtained from CAFE. 

By looking at the temperature distribution of this very long pipe, it can be clearly seen how the 

external radiation algorithm worked on the far field object.
D-71

Figure D-32 CAFE benchmark results using a small calorimeter 1.5 m (4.9 feet) from the 

edge of the fire

D.6.3 Summary of Benchmark Results

The fully engulfing benchmark results show that CAFE bounds the experimental calorimeter 

temperatures. Inside the fire, CAFE underestimates temperatures near the top of the 

calorimeter, while it overestimates temperatures on all other sides of the calorimeter. Taken as 

a whole, these results show that CAFE slightly overpredicts the average temperature of the 

surface of the calorimeter. Therefore, it is expected that for the fully engulfing cases examined 

in this study, the cask surface temperatures predicted by CAFE will be close to or slightly higher 

than expected.

Outside the fire zone, CAFE is expected to predict reasonably accurate temperatures for objects 

near the fire. For objects further from the pool, results are expected to be less accurate given 

the method employed by the code. This is not a concern because the heat flux to objects 

outside the plume decreases with the distance squared, suggesting that the fire threat is also 

less severe with distance from the fuel pool as observed in the results for the 18.3-m (60-foot) 

standoff case.

D.7 Summary

This appendix discusses the method employed to obtain the thermal response of the Rail-Steel, 

Rail-Lead, and Truck-DU casks to several fires lasting longer than the HAC fire described in 

10 CFR 71.73.

The approach used to model internals of these casks is similar to that presented in the 

respective cask SARs. This appendix describes some mathematical models and results 

reported in these documents and used in this study. In addition, modifications made to the cask 

models to simplify the complexities inherent in the cask design are noted. In general, boundary 

conditions and material properties differ slightly from those used in the SARs. For consistency, 

the same properties were used in these casks when the same or similar type materials were

used. Because realistic boundary conditions are sometimes difficult to implement using 

available data and current analysis tools, some simplifications were also necessary.
D-72

MSC Patran is the front-end code employed to generate the material database, the finite 

element discretization, and the boundary conditions for the internals of the casks. P/Thermal is 

the finite element heat transfer code used to solve the internal thermal response of the casks. 

CAFE is the computational fluid dynamics code used to generate the fire environment for the 

hypothetical fires lasting longer than the hypothetical fire described in 10 CFR 71.73. For these 

scenarios, CAFE and P/Thermal are coupled together to obtain the thermal response of the 

casks. P/Thermal is also used to generate the regulatory fire environments used for model 

verification. Results from these P/Thermal regulatory fires were compared against results 

presented in the SARs for the same regulatory environments. This served as a check to the 

current models.

Four fire accident scenarios are analyzed for the rail casks and one hypothetical fire accident 

scenario—the worst case in the rail cask analyses—is analyzed for the truck cask. These

scenarios include the regulatory fire described in 10 CFR 71.73, a cask on the ground 

concentric with a fuel pool sufficiently large to engulf the cask, a cask on the ground with a pool 

fire offset by the width of a railcar (3 m (10 feet)), and a cask on the ground with a pool fire 

offset by the length of a railcar (18 m (60 feet)). These nonregulatory scenarios represent an 

accident in which a pool of flammable liquid and the cask are separated by one railcar width or 

by one railcar length following an accident. The results shown in this section demonstrate that 

the Rail-Steel, Rail-Lead, and Truck-DU casks maintain containment for the cases analyzed in 

this study.
APPENDIX E

DETAILS OF TRANSPORTATION ACCIDENTS

E-1

E.1 Types of Accidents and Incidents

The following types of accidents can interfere with routine transportation of spent nuclear fuel:

• accidents in which the spent fuel cask is not damaged or affected

– minor traffic accidents (e.g., fender-benders, flat tires) that result in minor 

damage to the vehicle—usually called “incidents”1

– accidents that damage the vehicle or trailer enough that the vehicle cannot move 

from the scene of the accident under its own power, but which do not result in 

damage to the spent fuel cask

– accidents that involve a death or injury, but no damage to the spent fuel cask

• accidents in which the spent fuel cask is affected

– accidents that result in loss of gamma shielding effectiveness but no release of 

radioactive material

– accidents in which radioactive material is released

Neutron shielding is always assumed to be lost in an accident because it is not designed to be 

accident-resistant.

This analysis considers the first three accident types, in which the spent fuel cask is not 

damaged or affected, together. Section 5.3 of Chapter 5 discusses the radiation doses and 

risks from these types of accidents. This appendix evaluates the last three types of accidents in 

detail. Only very severe accidents (those resulting in conditions much more severe than the 

regulatory accident) have the possibility of causing a loss of lead gamma shielding or release of 

radioactive material.

The analyses from Chapters 3 and 4 and Appendices C and D indicate that none of the 

accidents studied for this report lead to loss of radioactive material or gamma shielding 

effectiveness for the Rail-Steel or Truck-DU casks. These casks can only suffer a loss of 

neutron shielding. Some of the accident environments studied did lead to a loss of 

effectiveness of the gamma shielding for the Rail-Lead cask. When spent fuel is transported in 

this cask without an inner welded canister, some of the accident environments studied could 

result in a release of radioactive material. This appendix evaluates the probability and 

consequence of accidents that lead to a loss of shielding or release of radioactive material.

 

1 In U.S. Department of Transportation parlance, an “accident” is an event that results in a death, an injury, or 

enough damage to a vehicle that it cannot move under its own power. All other events that result in 

nonroutine transportation are termed “incidents.” This document uses the term “accident” for both accidents 

and incidents.
E-2

E.2 Accident Probabilities

E.2.1 Historic Accident Frequencies

The probability that a traffic accident occurs is based on historic accident frequencies. These 

have been developed and the statistics validated by the U.S. Department of Transportation 

(DOT). Table E-1 shows truck and railcar accidents from 1991 through 2007 (U.S. Department 

of Transportation, 2008). The following are the average accident frequencies for this period:

• 1.98 × 10-6 per kilometer (km) (3.19 × 10-6 per mile) for large trucks on Interstates and 

primary highways

• 1.32 × 10-7 per railcar-km (2.12 × 10-7 per railcar-mile) for freight rail

Table E-1 Truck and Railcar Accidents Per Kilometer, 1991 through 2007

YEAR TRUCK ACCIDENTS PER KM RAILCAR ACCIDENTS PER RAILCAR-KM

1991 2.39 × 10-6 2.08 × 10-7

1992 1.99 × 10-6 1.91 × 10-7

1993 2.19 × 10-6 1.68 × 10-7

1994 2.19 × 10-6 1.64 × 10-7

1995 2.39 × 10-6 1.53 × 10-7

1996 1.90 × 10-6 1.39 × 10-7

1997 1.89 × 10-6 1.32 × 10-7

1998 2.04 × 10-6 1.19 × 10-7

1999 1.84 × 10-6 1.12 × 10-7

2000 2.08 × 10-6 1.12 × 10-7

2001 1.99 × 10-6 1.18 × 10-7

2002 1.83 × 10-6 1.12 × 10-7

2003 1.85 × 10-6 1.02 × 10-7

2004 1.90 × 10-6 1.00 × 10-7

2005 1.73 × 10-6 1.06 × 10-7

2006 1.83 × 10-6 1.04 × 10-7

2007 1.59 × 10-6 9.60 × 10-8

Accident frequencies decreased 33.5 percent for trucks and 53.8 percent for railcars between 

1991 and 2007. This document uses the average because there are annual fluctuations. 

Figure 5.1 in Chapter 5 shows the accident frequency trends.

E.2.2 Development of Conditional Accident Probabilities

Each specific accident scenario is described by a conditional probability (“conditional” on an 

accident occurring). The total probability of a specific accident scenario is the product of the 

accident frequency and the conditional probability for that scenario. Conditional probabilities are 

derived from event trees, as described below.

E.2.2.1 Conditional Probabilities of Truck Accidents

A transportation accident scenario can be disaggregated into a series of events. The 

conditional probability of a particular event in the scenario is best illustrated with an event tree, 
E-3

which is a diagram that includes all possible accident scenarios. Each branch of the tree is the 

series of events that comprise a particular accident scenario. The conditional probability is the 

product of the probabilities along a particular branch.

Figure E-1 is an event tree for truck accidents (Mills et al., 2006). An illustrative example would 

be the calculation of the conditional probability of a truck colliding with another vehicle on a 

bridge and then falling from the bridge onto a rocky embankment.

Pconditional = Pcollision*Pbridge accident*Pfall off bridge*Psoft rock

Pconditional = (0.054)*(0.064)*(0.02)*(0.046) = 3.18 × 10-6

Figure E-1 Event tree for highway accidents (from Mills et al., 2006)

The far right column of Figure E-1 lists the conditional probabilities.

Mills et al. (2006) describe in detail the construction of the event tree in Figure E-1. Appendix C

discusses details of collision accidents and Appendix D provides additional information on fire 

accidents.

0.064
E-4

E.2.2.2 Conditional Probabilities of Rail Accidents

This study uses the event tree for rail found in Volpe, 2006, shown in Figure E-2.
E-5

Figure E-2 Rail accident event tree (from Volpe, 2006)

2

 

2 “Off bridge” means the accident does not occur on a bridge.
E-6

E.2.2.3 Uncertainty in Event Trees

Event trees are excellent tools for dividing the universe of accidents into categories. The 

resultant probabilities are only as precise as the data that were available to develop the event 

tree. This becomes especially problematic for events that occur very rarely, such as the most 

severe accidents. For these events, the branch point distributions are assumed to be the same 

as those distributions for events that occur more frequently. NUREG/CR-6672, 

“Re-Examination of Spent Fuel Risk Estimates,” issued in 2000 (Sprung et al., 2000),

investigated uncertainties associated with event trees and found the total effect of all 

uncertainties to be approximately an order of magnitude. No further attempt was made to 

quantify the uncertainty associated with the event trees in this study.

E.3 Accident Risks and Consequences

E.3.1 Loss of Lead Gamma Shielding

The Rail-Lead cask is the only cask studied that uses lead as a gamma shield, so loss of the 

shielding would occur only in rail accidents. The Rail-Lead gamma shield is an annular lead 

cylinder about 0.127 meter (5 inches) thick. The lead shell can slump in a sufficiently severe 

impact, leaving a gap in the lead shield that results in increased external gamma radiation. The 

RADTRAN computer code models a gap in the shield from an impact and translates this to an 

increase in the dose from the virtual radiation source (O’Donnell et al., 2004; 

Dennis et al., 2009) that is the basis for the incident-free transportation model (Figure B-1, 

Appendix B). Figure E-3 is a diagram of the loss-of-shielding model, which recognizes the 

two-dimensional symmetry of the lead-shielded cask.

This study used the Monte Carlo N-Particle (MCNP) transport code to calculate the photon 

density along the line of receptor points (Figure E-3), both with and without a void (gap) in the 

lead shield. The difference in photon density, which is a function of the gap size, was 

expressed as a multiplier of the external dose rate at 1 meter (i.e., the transport index (TI)) from

the fully shielded cask. Different gap sizes are modeled using different values of this multiplier.

The product of the multiplier and the TI of the fully shielded cask provides a new value of the TI, 

which RADTRAN then uses to calculate doses to receptors. Thus, the results of MCNP 

calculations provide a RADTRAN source for various gap sizes. The RADTRAN model 

overestimates the dose to a receptor because it models all loss of lead shielding as a gap in the 

lead shield, rather than thinning of the lead shield, for example. Therefore, the doses calculated 

using RADTRAN are larger than would be calculated using MCNP.

Figure E-3 shows only one side of the model because the model is symmetric and can apply to 

either side of the cask. All of the calculations for loss of lead shielding in this study assumed the 

void was uniform around the cask.
E-7

Figure E-3 The RADTRAN loss-of-shielding two-dimensional model

(from O’Donnell et al., 2004)

E.3.1.1 Loss of Lead Shielding from Impact

Section 3.2.2 described the various amounts of lead slump that result from impact speed and 

orientation. Table E-2 shows the conditional probabilities of each combination of impact speed 

and orientation. The rail event tree does not include information on the hardness of the 

surfaces struck for any of its branches and only impacts into hard rock surfaces are severe 

enough to cause lead slump (see Section 3.3 for a discussion on impacts into yielding targets). 

To account for this, researchers used the impact surface distributions from the truck event tree 

(Figure E-1). This event tree indicated that only 5.5 percent of impacts are into hard rock 

surfaces. Only impacts that are at angles greater than 45 degrees are severe enough to cause 

lead slump (see Section 3.4).

Using the triangular distribution of impact angles, 33.3 percent of accidents have an impact 

angle greater than 45 degrees. The rail event tree also does not include any information about 

what happens after an accident that occurs on a bridge. Again, the detail for this is taken from 

the truck event tree, which indicates that 2 percent of the accidents that occur on a bridge result 

in the vehicle falling off the bridge, and 5 percent of those cases result in an impact onto hard 

rock. For falling off a bridge, the impact angle distribution is uniform rather than triangular, so 

the probability that the angle is greater than 45 degrees is 50 percent. Using this information, 

researchers calculated the conditional probabilities for accident scenarios in which an impact 

could result in lead slump by adding the probabilities at a particular impact speed for impacts 

into a slope and impacts into an embankment multiplied by 0.055 for hard rock and multiplied 

by 0.333 for impact angle. Added to this is the probability that the impact occurs on a bridge,

multiplied by 0.02 for falling off the bridge, multiplied by 0.05 for hard rock, and multiplied by 0.5 
E-8

for impact angle. An example calculation for calculating the conditional probability of an impact 

at a speed greater than 145 kilometers per hour (kph) (90 mph) is shown below:

Rail event tree impact onto slope > 113 kph (70 mph) = 3.95 × 10-8 (from the derailment with no 

fire branch)

Rail event tree impact onto embankment > 113 kph (70 mph) = 1.43 × 10-8

Rail event tree accident on bridge > 113 kph (70 mph) = 4.10 × 10-7

Fraction of accidents at >113 kph (70 mph) that are at > 145 kph (90 mph) = 0.05

Conditional probability = {[(3.95 × 10-8 + 1.43 × 10-8

) × 0.055 × 0.333] + 

[(4.10 × 10-7

) × 0.02 × 0.05 × 0.5]} × 0.05 = 5.96 × 10-11

Table E-2 Parameters of Lead Shield Slumping from Impacta

Orientation

Impact 

Speed 

(kph)

Event Tree 

Impact Speed 

(kph)b

Maximum 

Slump 

(mm)c

Slumped 

fraction

Conditional 

Probability 

from Rail 

Event Treed

Conditional 

Probability 

Including 

Orientation

End 

(Probability =

0.1)

193 >113 (5%)e 355.48 0.0725 5.96 × 10-11 5.96 × 10-12

145 >113 (95%) 83.2 0.0170 1.13 × 10-9 1.13 × 10-10

97 80 to 113 18.28 0.00373 1.44 × 10-6 1.44 × 10-7

48 48 to 80 6.43 0.00131 6.34 × 10-6 6.34 × 10-7

Corner 

(Probability =

0.6)

193 >113 (5%)d 310.48 0.0634 5.96 × 10-11 3.57 × 10-11

145 >113 (95%) 114.52 0.0234 1.13 × 10-9 6.79 × 10-10

97 80 to 113 25.11 0.00512 1.44 × 10-6 8.62 × 10-7

48 48 to 80 1.65 0.000337 6.34 × 10-6 3.80 × 10-6

Side 

(Probability =

0.3)

193 >113 (5%)d 15.47 0.00316 5.96 × 10-11 1.79 × 10-11

145 >113 (95%) 20.88 0.00426 1.13 × 10-9 3.40 × 10-10

97 80 to 113 1.37 0.000280 1.44 × 10-6 4.31 × 10-7

48 48 to 80 0.09 0.0000184 6.34 × 10-6 1.90 × 10-6

a These are input parameters to the RADTRAN calculation. b Event tree impact speeds are binned. This column relates the binned speeds to the modeled impact speed. c These values are derived from the finite element analysis (Chapter 3 and Appendix C). d These values are the sum of probabilities of collision scenarios at a particular impact speed into slope and

embankment multiplied by 0.055, because only impacts onto hard surfaces can cause lead slump, multiplied 

by 0.333, because only impacts that are at angles greater than 45 degrees can cause lead slump, plus the 

probability the accident occurs on a bridge, multiplied by 0.02 for accidents that fall off the bridge, multiplied 

by 0.055 for hard targets, multiplied by 0.5 for impact angles greater than 45 degrees. e The event tree did not distinguish between impacts from 113 kph to 145 kph and from 145 kph to 193 kph. It 

is assumed that 5 percent of the impacts are greater than 145 kph and 95 percent are between 113 kph and 

145 kph.

Table E-2 also shows the amount of lead slump, both as an absolute amount and as a fraction 

of the longest dimension of the lead shield.
E-9

Table E-3 shows dose rates to the maximally exposed individual (MEI) at various distances from 

the cask. The populations within 800 meters (1/2 mile) of the rail routes, as shown in Table 2-5, 

are the populations that could be exposed for each of the 16 rail routes modeled.

Table E-3 Radiation Dose Rates (Sv/h) to the MEI at Various Distances for the Cask

Fraction of 

slumped 

lead

1 m 2 m 3 m 4 m 5 m 10 m 20 m 50 m

7.25 × 10-2 4.6 × 10-2 2.1 × 10-2 1.3 × 10-2 9.2 × 10-3 7.1 × 10-3 1.3 × 10-4 3.1 × 10-5 4.4 × 10-6

6.34 × 10-2 3.9 × 10-2 1.8 × 10-2 1.1 × 10-2 7.9 × 10-3 6.1 × 10-3 1.0 × 10-4 2.3 × 10-5 3.4 × 10-6

2.34 × 10-2 1.2 × 10-2 5.6 × 10-3 3.5 × 10-3 2.5 × 10-3 1.9 × 10-3 1.9 × 10-5 4.6 × 10-6 6.9 × 10-7

1.70 × 10-2 8.6 × 10-3 3.9 × 10-3 2.4 × 10-3 1.7 × 10-3 1.3 × 10-3 1.4 × 10-5 3.3 × 10-6 5.1 × 10-7

5.12 × 10-3 2.3 × 10-3 1.0 × 10-3 6.4 × 10-4 4.6 × 10-4 3.5 × 10-4 8.5 × 10-6 2.1 × 10-6 3.4 × 10-7

4.26 × 10-3 1.9 × 10-3 8.3 × 10-4 5.2 × 10-4 3.8 × 10-4 2.9 × 10-4 8.4 × 10-6 2.1 × 10-6 3.3 × 10-7

3.73 × 10-3 1.6 × 10-3 7.2 × 10-4 4.5 × 10-4 3.3 × 10-4 2.5 × 10-4 8.3 × 10-6 2.1 × 10-6 3.3 × 10-7

3.16 × 10-3 1.3 × 10-3 6.1 × 10-4 3.8 × 10-4 2.8 × 10-4 2.1 × 10-4 8.2 × 10-6 2.1 × 10-6 3.3 × 10-7

1.31 × 10-3 5.7 × 10-4 2.6 × 10-4 1.7 × 10-4 1.2 × 10-4 9.5 × 10-5 8.1 × 10-6 2.0 × 10-6 3.2 × 10-7

3.37 × 10-4 2.3 × 10-4 1.1 × 10-4 7.2 × 10-5 5.3 × 10-5 4.2 × 10-5 8.1 × 10-6 2.0 × 10-6 3.2 × 10-7

2.80 × 10-4 2.1 × 10-4 1.0 × 10-4 6.7 × 10-5 5.0 × 10-5 4.0 × 10-5 8.1 × 10-6 2.0 × 10-6 3.2 × 10-7

1.84 × 10-5 1.4 × 10-4 7.2 × 10-5 4.8 × 10-5 3.6 × 10-5 2.9 × 10-5 8.1 × 10-6 2.0 × 10-6 3.2 × 10-7

a The numbers in bold italics exceed the exceed the external dose rate of 10 CFR 71.51. The dose rates are 

the direct output of the RADTRAN loss-of-shielding model.3

Figures E-4 and E-5 show dose rates to the MEI as a function of the fraction of shielding lost 

and as a function of distance from the cask. Exposure to the highest dose rate for 30 minutes 

would lead to a dose similar to a head computerized tomography (CT) scan 

(Shleien et al., 1998).

 

3 A rigorous calculation was performed at Oak Ridge National Laboratory (ORNL) using a three-dimensional

radiation shielding model. The comparison of results with those in Table E-3 indicate that the results are 

acceptable. Results of the more rigorous calculation show generally good agreement; however, the ORNL 

1-m values are 33 to 67 percent higher than the Table E-3 values. The differences may be the result of the 

manner in which the loss-of-shielding model was incorporated into RADTRAN—the additional dose caused 

by the shielding gap was expressed as a multiplier to the external dose rate.
E-10

Figure E-4 Radiation dose rates to the MEI from loss of lead gamma shielding at 

distances from 1 to 5 meters from the cask carrying spent fuel

Figure E-5 Radiation dose rates to the MEI from loss of lead gamma shielding at 

distances from 10 to 100 meters from the cask carrying spent fuel4

Table E-3 shows how the dose rate to the MEI depends on the fraction of the lead shield lost 

and the distance from the cask. The far left column of Table E-3 shows lead slump of the cask 

as computed from the finite element analyses discussed in Chapter 3 and Appendix C. 

RADTRAN computed the dose rates shown in the Table using the model discussed in 

Section E.3.1. Only a few of the dose rates in the Table exceed the regulatory dose rate of 

0.01 sieverts per hour (Sv/h) (100 mrem/hr) from Title 10 of the Code of Federal Regulations

 

4 The vertical axis is logarithmic so that all of the dose rates can be shown on the same graph.
E-11

(10 CFR) 71.51, “Additional Requirements for Type B Packages.” These dose rates, which are 

shown in bold italics in the table, are the result of accidents that are much more severe than the 

regulatory accident and occur with a very small probability (less than one in a billion accidents 

would produce the amount of lead slump required to result in these dose rates).

Emergency responders would sustain the large dose rates that occur near the cask (1 to 

5 meters (3.3 to 16.4 feet) from the cask). The 1-to-5-meter dose rates can be considered 

occupational rather than public doses. Exposure at the maximum dose rate (maximum slump at 

1 meter (3.3 feet)) for 1 hour would approximate the annual occupational dose limit from 

10 CFR Part 20, “Standards for Protection against Radiation.” If a loss-of-shielding accident 

occurred on a public right of way (a railroad track in this case), no member of the public would 

be closer than 10 meters (33 feet). The public MEI dose rate (from the largest gap in the lead 

shield) would be 0.13 mSv/h (13 mrem/hr) (the 10-meter curve in Figure E-5).

The “dose risk” combines the probability of a particular accident with the consequence (the 

dose). Table E-3 estimates accident consequences (dose rates) only; it does not consider the 

probability of an accident severe enough to produce those consequences. The dose risk is 

much smaller than the dose because of the very low probability of an accident that produces a 

loss of shielding. Tables E-4 and E-5 show the conditional dose risk (i.e., the product of the 

conditional probability with the dose consequence) for each fractional loss of lead shielding. For 

distances less than 10 meters (33 feet) from the cask, these doses are for a 1-hour exposure 

and represent an occupational dose to emergency responders. For distances of 10 meters 

(33 feet) and greater, these doses are for a 10-hour exposure and represent the dose to the 

public. These dose risks are the risk of a particular accident scenario if there is an accident; 

they do not include the probability of an accident.

Table E-4 The “Conditional Dose Risk” in Sv to the MEI at Distances from the Cask from 

1 to 5 Meters for 1 Houra

Fraction of 

Slumped 

Lead

Conditional 

Probability

Distance from the Cask (m)

1 2 3 4 5

7.25 × 10-2 5.96 × 10-12 2.74 × 10-13 1.25 × 10-13 7.74 × 10-14 5.48 × 10-14 4.23 × 10-14

6.34 × 10-2 3.57 × 10-11 1.39 × 10-12 6.43 × 10-13 3.93 × 10-13 2.82 × 10-13 2.18 × 10-13

2.34 × 10-2 6.79 × 10-10 8.15 × 10-12 3.80 × 10-12 2.38 × 10-12 1.70 × 10-12 1.29 × 10-12

1.70 × 10-2 1.13 × 10-10 9.73 × 10-13 4.41 × 10-13 2.72 × 10-13 1.92 × 10-13 1.47 × 10-13

5.12 × 10-3 8.62 × 10-7 1.98 × 10-9 8.62 × 10-10 5.52 × 10-10 3.97 × 10-10 3.02 × 10-10

4.26 × 10-3 3.40 × 10-10 6.34 × 10-13 2.81 × 10-13 1.77 × 10-13 1.29 × 10-13 9.78 × 10-14

3.73 × 10-3 1.44 × 10-7 2.30 × 10-10 1.03 × 10-10 6.47 × 10-11 4.74 × 10-11 3.59 × 10-11

3.16 × 10-3 1.79 × 10-11 2.32 × 10-14 1.09 × 10-14 6.79 × 10-15 5.00 × 10-15 3.75 × 10-15

1.31 × 10-3 6.34 × 10-7 3.61 × 10-10 1.65 × 10-10 1.08 × 10-10 7.61 × 10-11 6.02 × 10-11

3.37 × 10-4 3.80 × 10-6 8.75 × 10-10 4.18 × 10-10 2.74 × 10-10 2.02 × 10-10 1.60 × 10-10

2.80 × 10-4 4.31 × 10-7 9.19 × 10-11 4.44 × 10-11 2.91 × 10-11 2.15 × 10-11 1.71 × 10-11

1.84 × 10-5 1.90 × 10-6 2.66 × 10-10 1.37 × 10-10 9.13 × 10-11 6.84 × 10-11 5.51 × 10-11

a The “conditional dose risk” is the product of conditional probabilities from Table E-2 and the 1- to 5-meter 

doses from Table E-3.
E-12

Table E-5 The “Conditional Dose Risk” in Sv to the MEI at Distances 10 to 100 Meters

from the Cask for 10 Hoursa

Fraction of 

Slumped 

Lead

Conditional 

Probability

Distance from the Cask (m)

10 20 50 100

7.25 × 10-2 5.96 × 10-12 7.74 × 10-15 1.85 × 10-15 2.62 × 10-16 5.93 × 10-17

6.34 × 10-2 3.57 × 10-11 3.57 × 10-14 8.22 × 10-15 1.22 × 10-15 2.82 × 10-16

2.34 × 10-2 6.79 × 10-10 1.29 × 10-13 3.12 × 10-14 4.69 × 10-15 1.16 × 10-15

1.70 × 10-2 1.13 × 10-10 1.58 × 10-14 3.74 × 10-15 5.77 × 10-16 1.37 × 10-16

5.12 × 10-3 8.62 × 10-7 7.33 × 10-11 1.81 × 10-11 2.93 × 10-12 7.53 × 10-13

4.26 × 10-3 3.40 × 10-10 2.84 × 10-14 7.13 × 10-15 1.13 × 10-15 2.87 × 10-16

3.73 × 10-3 1.44 × 10-7 1.19 × 10-11 3.02 × 10-12 4.74 × 10-13 1.19 × 10-13

3.16 × 10-3 1.79 × 10-11 1.47 × 10-15 3.75 × 10-16 5.90 × 10-17 1.47 × 10-17

1.31 × 10-3 6.34 × 10-7 5.13 × 10-11 1.27 × 10-11 2.03 × 10-12 5.15 × 10-13

3.37 × 10-4 3.80 × 10-6 3.08 × 10-10 7.61 × 10-11 1.22 × 10-11 3.07 × 10-12

2.80 × 10-4 4.31 × 10-7 3.49 × 10-11 8.62 × 10-12 1.38 × 10-12 3.47 × 10-13

1.84 × 10-5 1.90 × 10-6 1.54 × 10-10 3.80 × 10-11 6.08 × 10-12 1.50 × 10-12

a The “conditional dose risk” is the product of conditional probabilities from Table E-2; the 

10-, 20-, and 50-meter doses from Table E-3; and the 100-meter dose from Figure E-5.

The collective dose risk to an exposed population within a radius r of the cask may be 

calculated with Equation E-1:

(E-1) Collective Dose Risk = A ∗ L ∗ PD ∗ S ∗ ∑ Pci ∗ 10−6 ∗ ∫ Dir ∗ 2πrdr 800

20

n

i

= A ∗ L ∗ PD ∗ S ∗ 1.46 ∗ 10−12

Where:

A is the accident frequency per kilometer (km) on the route segment under consideration

L is the length of the route segment in km

PD is the population density per square kilometer (km2

)

S is the shielding factor caused by residence type from Table II-3 

Pci is the conditional probability of the ith fractional loss of shielding

10-6 is used to convert the integrated area from m2 to km2

r is the distance from the cask: 20 to 800 meters

Dir is the dose from the i

th fractional loss of shielding at a distance r from the cask (these 

values are tabulated in Table E-3)

2πrdr is the incremental area of the band at distance r from the cask (in square 

meters (m2

))
E-13

The index i indicates a particular fractional shielding loss, which Table E-3 summarizes. The 

summation in Equation E-1 is the conditional dose risk of all of the accidents considered (the 

“universe” of accidents). Only one accident is modeled, which is assumed to occur on any route 

segment, but not on more than one route segment. The summation in Equation E-1 is 

independent of accident location.

Population dose risk ultimately depends on the accident frequency and on the population along 

the route where the accident happens. The accident frequency, in accidents per kilometer, is 

equated to the accident probability. U.S. Department of Transportation (2008) provides the rail 

accident frequencies used in this analysis. Table E-6 shows the average railcar accident 

frequencies for each of the 16 rail routes analyzed. The routes shown are for illustrative 

purposes only, and no SNF shipments are planned from any of these points of origin to any of 

these destinations. The accident frequencies from Table E-6 are combined with the average 

dose risk integrated over the potentially exposed population.

Table E-6 Average Railcar Accident Frequencies and Accidents Per Shipment on the 

Routes Studied

ORIGIN DESTINATION

AVERAGE 

ACCIDENTS PER KM

ROUTE 

LENGTH 

(KM)

PROBABILITY OF AN 

ACCIDENT FOR THE 

TOTAL ROUTE

MAINE YANKEE

ORNL 6.5 × 10-7 2,125 0.00139

DEAF SMITH 5.8 × 10-7 3,362 0.00194

HANFORD 4.2 × 10-7 5,084 0.00214

SKULL VALLEY 5.1 × 10-7 4,086 0.00208

KEWAUNEE

ORNL 4.3 × 10-7 1,395 0.00060

DEAF SMITH 3.3 × 10-7 1,882 0.00062

HANFORD 2.4 × 10-7 3,028 0.00073

SKULL VALLEY 3.7 × 10-7 2,755 0.00103

INDIAN POINT

ORNL 8.8 × 10-6 1,264 0.0112

DEAF SMITH 6.2 × 10-7 3,088 0.00192

HANFORD 4.4 × 10-7 4,781 0.00212

SKULL VALLEY 5.5 × 10-7 3,977 0.00217

INL

ORNL 3.6 × 10-7 3,306 0.0012

DEAF SMITH 3.5 × 10-7 1,913 0.00067

HANFORD 3.2 × 10-7 1,062 0.00034

SKULL VALLEY 2.8 × 10-7 455 0.00013

Table E-7 shows the collective dose risks to populations on each side of the rail cask that has 

lost lead shielding on impact. These estimates include both the conditional probabilities and the 

accident frequencies on each route, as in Equation E-1.
E-14

Table E-7 Collective Dose Risks Per Shipment (Person-Sv) from Loss of Lead Shielding, 

Including Accident and Conditional Probabilities

ORIGIN TYPE

Destination

ORNL DEAF SMITH HANFORD SKULL 

VALLEY

MAINE 

YANKEE

Rural 2.11 × 10-14 2.65 × 10-14 4.74 × 10-14 3.59 × 10-14

Suburban 2.23 × 10-13 2.34 × 10-13 2.16 × 10-13 2.19 × 10-13

Urban 5.95 × 10-15 7.17 × 10-15 8.19 × 10-15 4.83 × 10-15

KEWAUNEE

Rural 9.64 × 10-15 6.29 × 10-15 8.06 × 10-15 1.34 × 10-14

Suburban 9.07 × 10-14 5.49 × 10-14 4.50 × 10-14 9.41 × 10-14

Urban 3.35 × 10-15 2.25 × 10-15 8.24 × 10-16 2.38 × 10-15

INDIAN POINT

Rural 1.12 × 10-13 2.25 × 10-14 3.02 × 10-14 3.04 × 10-14

Suburban 3.04 × 10-12 2.07 × 10-13 2.16 × 10-13 2.33 × 10-13

Urban 3.72 × 10-13 8.15 × 10-15 7.27 × 10-15 7.21 × 10-15

IDAHO 

NATIONAL 

LAB 

Rural 1.66 × 10-14 4.00 × 10-15 2.56 × 10-15 1.62 × 10-15

Suburban 8.09 × 10-14 3.60 × 10-14 1.76 × 10-14 1.26 × 10-14

Urban 1.34 × 10-15 6.49 × 10-16 3.71 × 10-16 4.38 × 10-16

Example: For the suburban route segment of the Maine-Yankee-to-Hanford route:

Accident rate: 4.2 × 10-7

/km (from Table E-6)

Segment Length: 1,135 km

Population density: 357 persons/km2

Suburban shielding factor: 0.87

Dose risk = 4.2 × 10-7 accident/km * 1135 km * 357 persons/km2 * 0.87 *

1.46 × 10-12 Sv-km2

/accident = 2.16 × 10-13 person-Sv (2.16 × 10-8 person-mrem)

E.3.1.2 Loss of Lead Shielding with Fire

The loss of lead shielding because of a fire occurs after the end of the fire when the cask cools. 

Lead expands as it melts and can buckle the innermost cask shell. When the melted lead cools 

and solidifies, it occupies the same volume as before expansion, but the volume available 

between the steel cask shells is larger because of the buckling of the inner shell, leaving a gap. 

Appendix D describes in full detail the melting of lead and the formation of a gap. Briefly, if the 

cask is offset from the fire, the gap would be in the section of lead shield facing the fire. In an 

engulfing fire, the gap would be at the upper surface of the cask. For conservatism, this study 

assumes that the loss of lead shielding is uniform around the cask and people in all directions 

are equally exposed. Therefore, in both cases, anyone facing the side of the cask with the 

shielding gap could sustain an increased radiation dose.

In an accident, two fire scenarios can result in a loss of lead shielding:

• Fire Scenario 1: A sufficiently hot pool fire engulfs a cask on the ground and can melt 

enough lead in 3 hours to create a fractional shield loss of 8.14 percent.
E-15

• Fire Scenario 2: A sufficiently hot pool fire offset from the cask burns for more than 

3 hours and can create a fractional shield loss of 2.01 percent.

Appendix D fully describes these scenarios. Table E-8 shows the doses sustained by the MEI, 

exposed for 1 hour, at various distances from the cask.

Table E-8 Radiation Dose (Sv) to the MEI at Various Distances from a Cask That Has 

Been in a Fire (Direct RADTRAN Output)

Reduction 

of Lead 

Shieldinga 1 m 2 m 5 m 10 m 20 m 50 m 100 m

0.0201 7.0 × 10-3 3.1 × 10-3 1.1 × 10-3 1.1 × 10-5 2.6 × 10-6 3.9 × 10-7 9.4 × 10-8

0.0814 3.5 × 10-2 1.6 × 10-2 5.4 × 10-3 1.1 × 10-4 2.6 × 10-5 3.7 × 10-6 8.5 × 10-7

a From the thermal analyses in Chapter 4 and Appendix D.

No lead shielding would be lost until after the fire was out and the cask had cooled enough for 

the lead to solidify; only then would there be a gap in the lead shield. Differential heating of the 

lead shield could result in geometry and associated volume changes that could impact the 

shielding effectiveness locally. Such effects would have minimal impact on the dose. Thus,

no one would be exposed for many hours after the accident, and with a fire this severe, nearby 

residents and the public would probably have been evacuated. The MEI in this case would be 

an emergency responder. Under these circumstances, measures could be taken to mitigate 

emergency responder exposures.

Volpe (2006), Figure 16, postulates a chain of events leading to a fire, from which the probability 

of these scenarios can be calculated. Figure E-6 shows the relevant portion of the Volpe figure.

Figure E-6 Event tree branch for a rail fire accident (from Volpe, 2006, Figure 16; “HM” 

stands for “hazardous material”)

To get to the starting point of the fire event tree, certain other events must occur. The first of 

these events is a major derailment, as shown in Table E-9. Volpe estimates that the speed at 

the time of the accident for such a derailment is at least 80 kph (50 miles per hour). If a pileup 

could occur in any kind of derailment other than in a tunnel, from Figure E-2, the probability of 

such a major collision/derailment/fire can be calculated as follows:
E-16

Pderailment * Pfire * (P80-to-113-kph collision + P>113-kph collision)) * [Poffbridge * (Pslope + Pembankment + Pstructure) +

Pon_bridge] = P

(0.7355) * (0.0155) * (0.06614 + 0.00061) * [0.9887 * (0.0011 + 0.0004 + 0.0077) + 0.0113] =

1.55 × 10-5

The summation of surfaces struck does not include the “other” branch of the event tree because 

this branch represents the most common case for derailments—the case in which the derailed 

cars remain upright on the side of the track. Table E-9 lists the other events in the scenario, 

together with the probability of each event. These events are a pileup, a flammable hazardous 

cargo within 10 meters (33 feet) (about half a railcar length), leaking of that hazardous 

substance, and ignition of a pool fire. The net probability of the sequence of events shown in 

Table E-9 following a major pileup is 1.35 × 10-14. The net probability depends on the very small 

pileup probability of 2.4 × 10-5

. Thus, it is instructive to estimate the probability without the 

assumption of a pileup. Using the “no pileup” branch, the net probability for the events of 

Table E-9 is 5.6 × 10-10, still an exceptionally small number. The uncertainties in this analysis 

are exceedingly large compared to the probabilities. Essentially, the sequence of events in 

Figure E-6, with or without the pileup, is unlikely to happen.

The conditional probability of Fire Scenario 1, a major derailment, with or without a pileup, that

leads to a 3-hour pool fire that surrounds the cask is:

(1.55 × 10-5

) * (5.6 × 10-10) = 8.7 × 10-15

The conditional probability of Fire Scenario 2, a major derailment that does not involve a pileup 

but leads to a 3-hour fire offset from the cask by more than 10 meters (33 feet), is:

(1.55 × 10-5

) * (2.36 × 10-5) = 3.7 × 10-10

Table E-9 Events Leading to a Train Fire That Could Involve a Spent Fuel Cask

Event Probability Alternative Event Probability

Major derailment (> 48 kph) 8 × 10-5 No major derailment 0.99992

Pileup 2.4 × 10-5 No pileup 0.99998

Offset < 10 m 2.38 × 10-5 Offset ≥10 m 0.99998

Flammable hazardous material 

in another railcar 0.6 No flammable 

material available 0.4

Release of flammable material 0.0394 No release of 

flammable material 0.9606

Pool fire 0.001 No pool fire 0.999
E-17

The average accident frequency for the 16 rail routes studied is 1.9 × 10-3 (the range is from 

1.3 × 10-4 to 1.1 × 10-2

).5 Thus, the average probability of an accidental fire that could cause 

loss of lead shielding in a rail cask is 1.9 × 10-3 * 8.7 × 10-15 = 1.7 × 10-17 if the cask is concentric 

with the fire and 1.9 × 10-3 * 3.7 × 10-10 = 7.0 × 10-13 if the cask and fire are offset by 10 meters 

(33 feet) or more. The largest dose risk to a person 1 meter (3.3 feet) away from the cask with 

2-percent lead loss for 1 hour would be 7.0 × 10-13 * 7.0 × 10-3 = 4.9 × 10-15 Sv (this dose risk is 

larger than that from the cask in the engulfing fire).

E.3.2 Loss of Neutron Shielding

The neutron shield is usually a hydrocarbon or carbohydrate polymer, sometimes borated, 

because boron and organic polymers are good neutron absorbers. Neutron shielding burns, 

and it could be destroyed in a fire. The neutron dose from loss of shielding in a fire is estimated 

using the parameters listed in Table E-10. The conditional probability of a truck fire is from 

Figure E-1. The conditional probability of a rail fire is a combination of the fire probability in 

Figure E-2 and the following steps from Table E-9:

• a pileup

• flammable cargo on the train

• release of the flammable cargo

Table E-10 Some Parameters Used in Calculating Loss of Neutron Shielding

Parameter Truck-DU Rail-Lead Rail-Steel

Conditional probability of a firea 0.0063b 8.9 × 10-8 8.9 × 10-8

Neutron dose rate at one meter from the 

cask in mSv/h (mrem/h)c 1.78 (178) 1.81 (181) 1.82 (182)

Shielding of residents none none none

Time until the cask is removed (h) 10 10 10

a The truck fire probability comes from Figure E-1 and the rail fire probability comes from Volpe (2006). b The conditional probability of a fire for the Truck-DU cask is much higher than that for the two rail casks, in 

part (1) because truck accidents always involve a source of fuel (the gas tanks of the truck) while many 

railcar accidents do not involve the locomotive and (2) because of the way that the event trees were 

constructed. The truck event tree does not distinguish between minor fires and ones that are severe enough 

to damage the neutron shielding, while the rail event tree only considers severe fires. Therefore, the 

conditional probability of a truck fire is quite conservative. c The neutron dose at 2 meters from the Truck-DU cask is from General Atomics (1998); the corresponding 

figure for the Rail-Lead cask is from Nuclear Acceptance Corporation International (2002); and the 

corresponding figure for the Rail-Steel cask is from Holtec International (2004). The respective TI values 

were calculated from these numbers using RADTRAN, which slightly overestimates results (Steinman et 

al., 2002). The RADTRAN external dose rate is then modeled as entirely neutron emission.

The other parameters are the same as those used in calculating doses from an accident in 

which there is no release of radioactive material and no loss of lead shielding (Section 5.4 of 

Chapter 5).

 

5 This value is obtained by multiplying the appropriate accident frequency per kilometer from Table E-7 by the 

appropriate route length in kilometers from Table 2-5.
E-18

Table E-11 shows the neutron doses to an emergency responder (5 meters from the cask) and 

Figure E-7 shows the RADTRAN output that was used to generate the doses.

Table E-11 Doses to an Emergency Responder 5 Meters from the Cask from an Accident 

with a Loss of Neutron Shielding

Cask 10-hour 10 CFR 71.51 dose in Sv

Truck-DU 0.0073

Rail-Lead 0.0076

Rail-Steel 0.0077

RADTRAN STOP INPUT FILE

STOP RESPONDERT GA_4 1.0 5.0 5.0 1.0 10.0

STOP RESPONDERNAC NAC-STC 1.0 5.0 5.0 1.0 10.0

STOP RESPONDERHI HISTAR 1.0 5.0 5.0 1.0 10.0

RADTRAN OUTPUT FILE

STOP EXPOSURE IN PERSON-REM

LINE-SOURCE RESPONDERT 7.29E-01

LINE-SOURCE RESPONDERNAC 7.61E-01

LINE-SOURCE RESPONDERHI 7.73E-01

Figure E-7 RADTRAN input and output for Table E-11

Table E-12 shows the collective doses to nearby residents on the 16 truck and 16 rail routes 

analyzed and Table E-13 shows the total collective dose risks, including accident frequency. 

For the Rail-Lead cask, the neutron doses would add to the gamma dose from the loss of lead 

shielding in the cases in which the fire was severe enough to result in a loss of lead shielding.
E-19

Table E-12 Collective Conditional Doses to Nearby Residents in Person-Sv

from Loss of Neutron Shielding

FROM TO Truck-DU Rail-Lead Rail-Steel

MAINE YANKEE

ORNL 7.49 × 10-4 7.17 × 10-4 7.40 × 10-4

DEAF SMITH 7.01 × 10-4 6.71 × 10-4 6.93 × 10-4

HANFORD 6.23 × 10-4 5.96 × 10-4 6.15 × 10-4

SKULL VALLEY 6.38 × 10-4 6.11 × 10-4 6.31 × 10-4

KEWAUNEE

ORNL 6.87 × 10-4 6.57 × 10-4 6.78 × 10-4

DEAF SMITH 6.41 × 10-4 6.13 × 10-4 6.33 × 10-4

HANFORD 5.98 × 10-4 5.72 × 10-4 5.91 × 10-4

SKULL VALLEY 6.17 × 10-4 5.91 × 10-4 6.10 × 10-4

INDIAN POINT

ORNL 7.28 × 10-4 6.97 × 10-4 7.20 × 10-4

DEAF SMITH 6.95 × 10-4 6.65 × 10-4 6.87 × 10-4

HANFORD 6.38 × 10-4 6.11 × 10-4 6.31 × 10-4

SKULL VALLEY 6.63 × 10-4 6.34 × 10-4 6.55 × 10-4

INL

ORNL 5.78 × 10-4 5.53 × 10-4 5.71 × 10-4

DEAF SMITH 6.16 × 10-4 5.89 × 10-4 6.08 × 10-4

HANFORD 3.78 × 10-4 3.62 × 10-4 3.73 × 10-4

SKULL VALLEY 6.41 × 10-4 6.13 × 10-4 6.33 × 10-4

Table E-13 Collective Conditional Dose Risks in Person-Sv from

Loss of Neutron Shielding

FROM TO Truck-DU Rail-Lead Rail-Steel

MAINE 

YANKEE

ORNL 4.7 × 10-6 6.4 × 10-11 6.6 × 10-11

DEAF SMITH 4.4 × 10-6 6.0 × 10-11 6.2 × 10-11

HANFORD 3.9 × 10-6 5.3 × 10-11 5.5 × 10-11

SKULL VALLEY 4.0 × 10-6 5.4 × 10-11 5.6 × 10-11

KEWAUNEE

ORNL 4.3 × 10-6 5.8 × 10-11 6.0 × 10-11

DEAF SMITH 4.0 × 10-6 5.5 × 10-11 5.6 × 10-11

HANFORD 3.8 × 10-6 5.1 × 10-11 5.3 × 10-11

SKULL VALLEY 3.9 × 10-6 5.3 × 10-11 5.4 × 10-11

INDIAN POINT

ORNL 4.6 × 10-6 6.2 × 10-11 6.4 × 10-11

DEAF SMITH 4.4 × 10-6 5.9 × 10-11 6.1 × 10-11

HANFORD 4.0 × 10-6 5.4 × 10-11 5.6 × 10-11

SKULL VALLEY 4.2 × 10-6 5.6 × 10-11 5.8 × 10-11

INL

ORNL 3.6 × 10-6 4.9 × 10-11 5.1 × 10-11

DEAF SMITH 3.9 × 10-6 5.2 × 10-11 5.4 × 10-11

HANFORD 2.4 × 10-6 3.2 × 10-11 3.3 × 10-11

SKULL VALLEY 4.0 × 10-6 5.5 × 10-11 5.6 × 10-11

To get the collective dose risk for loss of neutron shielding, the collective conditional dose risks 

from Table E-13 must be multiplied by the probability that an accident occurs. Table E-6 gives 

the accident probability for the rail routes. For the truck routes, the route length (from Table 2-5) 
E-20

is multiplied by the national average accident rate, 1.98 × 10-6

/km (3.19 × 10-6/mile). Table E-14

shows the resulting collective dose risks. Comparing these results to those from the loss of lead 

shielding in Table E-7, it can be seen that loss of lead shielding produces a higher collective 

dose risk, even though it is a much less probable event.

Table E-14 Collective Dose Risks in Person-Sv from Loss of Neutron Shielding

FROM TO Truck-DU Rail-Lead Rail-Steel

MAINE YANKEE

ORNL 4.67 × 10-8 8.90 × 10-14 9.17 × 10-14

DEAF SMITH 3.13 × 10-8 1.16 × 10-13 1.20 × 10-13

HANFORD 3.22 × 10-8 1.13 × 10-13 1.18 × 10-13

SKULL VALLEY 1.41 × 10-8 1.12 × 10-13 1.16 × 10-13

KEWAUNEE

ORNL 2.94 × 10-8 3.48 × 10-14 3.60 × 10-14

DEAF SMITH 1.70 × 10-8 3.41 × 10-14 3.47 × 10-14

HANFORD 1.97 × 10-8 3.72 × 10-14 3.87 × 10-14

SKULL VALLEY 9.83 × 10-9 5.46 × 10-14 5.56 × 10-14

INDIAN POINT

ORNL 4.11 × 10-8 6.94 × 10-13 7.17 × 10-13

DEAF SMITH 2.68 × 10-8 1.13 × 10-13 1.17 × 10-13

HANFORD 2.91 × 10-8 1.14 × 10-13 1.19 × 10-13

SKULL VALLEY 1.04 × 10-8 1.22 × 10-13 1.26 × 10-13

INL

ORNL 6.85 × 10-9 5.88 × 10-14 6.12 × 10-14

DEAF SMITH 1.77 × 10-8 3.48 × 10-14 3.62 × 10-14

HANFORD 2.24 × 10-9 1.09 × 10-14 1.12 × 10-14

SKULL VALLEY 2.61 × 10-8 7.15 × 10-15 7.28 × 10-15

E.4 Release of Radioactive Materials in Accidents

E.4.1 Spent Fuel Inventory

A Rail-Lead cask is the only cask studied that would release any radioactive material in an 

accident. Since there is no traffic accident that would result in a release from the Truck-DU or 

Rail-Steel cask, the inventory of those casks is not relevant to this analysis. The fuel used in 

this analysis is pressurized-water reactor (PWR) fuel, with a burnup of 45,000 megawatt-days

per metric ton of uranium (MTU) (the maximum burnup that a Rail-Lead cask would transport),

which has cooled for 9 years before transport (Appendix A). The radionuclide inventory of this 

fuel was determined using ORIGEN (Croff, 1980). The radionuclide activities in the inventory 

were “normalized” by dividing each activity by the A2 value for that radionuclide. The A2 value, 

the amount of the radionuclide that could be transported in a Type A container, is an indication 

of the radiotoxicity; the larger the A2 value, the smaller the radiotoxicity of that nuclide. 

Researchers then sorted and added the normalized radioactivities until they reached 

99.99 percent of the total normalized radioactivity. The “total normalized activity” referred to 

here is not the total A2 value as calculated by the formula in Appendix A, “Determination of A1

and A2,” to 10 CFR Part 71, “Packaging and Transportation of Radioactive Material.” Table E-15

lists the radionuclides selected this way, together with their actual radioactivities (not the 

normalized radioactivities). Normalized radioactivities are used only to identify 99.9 percent of 

the radiotoxicity. The actual activity is the basis for the release fraction of each radionuclide.
E-21

Table E-15 Inventory for the Rail-Lead Cask

Radionuclide Name Form

Terabecquerels (TBq) 

from 26 Assemblies

Curies (Ci) from 

26 Assemblies

241Am americium particle 193 5,210

240Pu plutonium particle 184 4,970

238Pu plutonium particle 180 4,850

241Pu plutonium particle 10,440 282,000

90Y yttrium particle 40,400 1,090,000

90Sr strontium particle 40,400 1,090,000

137Cs cesium volatile 50,400 1,360,000

239Pu plutonium particle 71.9 1,940

244Cm curium particle 31.5 852

134Cs cesium volatile 3030 81,800

154Eu europium particle 146 3,950

106Ru ruthenium particle 467 12,600

243Cm curium particle 1.16 31.3

243Am americium particle 0.995 26.9

144Ce cerium particle 180 4,850

242Pu plutonium particle 0.614 16.6

125Sb antimony particle 431 11,600

155Eu europium particle 607 16,400

242mAm americium particle 0.163 4.40

242Am americium particle 0.162 4.38

60Co cobalt CRUD1 55.6 1,500

125mTe tellurium particle 105 2,840

234U uranium particle 0.572 15.5

85Kr krypton gas 3,340 90,100

1

 CRUD” is an acronym for “Chalk River unidentified deposit,” a type of corrosion deposit that contains radioactive 

species.

E.4.2 Dispersion of Released Radionuclides

If a spent fuel cask transportation accident did result in the release of radioactive material, the 

public could be exposed if the material were to be dispersed through the air. Experimental work 

reviewed by Sprung et al. (2000) indicates that only very small particles with an activity mean 

aerodynamic diameter (AMAD)6 of 10 microns or less would be released from a cask in an 

accident because the only release path is through the seals at the ends of the cask. Ten 

 

6 The AMAD is the diameter of a sphere of density one gram per cubic centimeter that has the same inertial 

properties as the actual particle. 
E-22

microns is generally considered the upper limit of respirability. Thus, particles accidentally 

released from a cask would be released as a respirable aerosol.

The discussion below is an abbreviated discussion of air dispersion, a subject that is treated 

extensively and in detail in textbooks such as Wark and Warner (1981).

The basic equation for atmospheric dispersion of an aerosol is the Gaussian dispersion 

equation (Turner, 1994, Chapter 2):

(E-2)

Where:

CHI7 = the concentration of particles in the air

Q = the radioactivity or mass of airborne particles

u = the windspeed

σy and σz = meteorological constants; functions of the downwind distance x and the 

meteorological conditions

The wind direction is traditionally along the x-axis of a Cartesian coordinate system, the 

crosswind direction is y, and z represents the altitude above ground. When the plume of 

released material rises buoyantly to a height H, the Gaussian equation becomes:

(E-3)

Where H is the height to which the plume rises before being blown downwind. For a 

ground-level release along the plume centerline, Equations E-2 and E-3 reduce to the following:

(E-4)

Radioactive gases released in an accident will disperse in the air according to Equations E-1 

and E-3. Particles, however, have mass and will settle on the ground. Equation E-5 gives the 

settling velocity Vt—the terminal velocity of a particle in the indicated size range:

 

7 The Greek letter Χ is traditionally used to represent air concentration, but is so easily confused in typescript 

with the 24th letter of the alphabet that it is often written phonetically (“chi”).





 



−















− = 2

2

2

2

2 exp 2 exp 2

1

y z y z

y z

Q u

CHI

π σ σ σ σ

y z Q u

CHI

2π σ σ

1 =





















 



− +  +



 



− −















− = 2

2

2

2

2

2

2

( ) exp 2( ) exp 2 exp 21

y z y z z

y z H z H

Q u

CHI

π σ σ σ σ σ
E-23

(E-5) Vt = gρd2

18μ

Where:

g = gravitational acceleration

d = particle aerodynamic diameter

ρ = particle density

μ = air viscosity at ambient temperature

Equation E-6 (Wark and Warner, 1981, Chapter 5) describes the ground deposition rate:

(E-6)

Where wp is the particle deposition rate. These equations are programmed in RADTRAN.

Both wind and air-temperature profiles affect the dispersion of airborne material. The 

predominant motion of airborne material is downwind, while crosswind motion is diffusive. Light 

winds, stable air, and temperature inversions result in less dispersion and higher airborne and 

ground concentrations of radionuclides. Strong winds and turbulent air are good conditions for 

dispersion and result in lower airborne and deposited radionuclide concentrations and 

consequently result in lower radiation doses to the public, even though the plume of radioactive 

material may spread over a large area.

RADTRAN calculates external doses from deposited material (“groundshine”) and from material 

that remains suspended in the air (“cloudshine”). The code also calculates internal committed 

doses from airborne material that is inhaled, and from material that becomes resuspended in 

the air. The doses reported are the sums of the groundshine, cloudshine, inhaled, and 

resuspended inhaled doses, unless otherwise indicated. To determine public exposure, as 

discussed in 10 CFR 20.1301, “Dose Limits for Individual Members of the Public,” the 

U.S. Nuclear Regulatory Commission adds these doses to sum to a “total effective dose 

equivalent.” RADTRAN accommodates a number of atmospheric dispersion conditions.

E.4.3 Release Fractions

Release of radionuclides into the environment from a cask depends on releases from the fuel 

rods into the cask and from the cask to the environment. If the cask contains canistered fuel, 

the cask structural and thermal analyses in Chapters 3 and 4 show that the canister does not 

rupture, even under the most severe accidents analyzed, so no radioactive material can exit the 

cask. In the present study, therefore, only the Rail-Lead cask transporting uncanistered fuel 

could release any radioactive material or Chalk River unidentified deposit (CRUD) as a 

consequence of a traffic accident. This section only considers PWR spent fuel.























 



 



 − −















− = 2

2

2

2

2 exp 2 exp 2 z

t

y z y

p t u

xV H

y

u

V

Q

w

π σ σ σ σ
E-24

When fuel rods are fractured in an impact, they depressurize, and the consequent overpressure 

sweeps fuel particles out of the cask if there is a breach in the seal. The depressurization and 

release of material from the rod is described very clearly by Hanson et al. (2008):

When commercial spent nuclear fuel (CSNF) is handled in a dry environment, 

whether as fuel assemblies, canned, or within a container, one possible 

mechanism for radionuclide release is a drop accident scenario, [in which] it is 

possible that the cladding could fracture, and cans or containers could breach... 

(Sprung et al. 2000). Upon clad breach, it is expected that the rod would rapidly 

depressurize, releasing its fill gas (e.g., He) and fission gases (e.g., Kr, Xe) that 

have been released from the fuel matrix, depending on the size of the cladding 

defect and fuel burnup characteristics (Einziger and Beyer 2007).... It is also 

possible for fuel fines to be ejected as the high-pressure fill and fission gases 

rapidly escape through the defect.... (Hanson et al., 2008, Section 1)

The release fractions from the rods to the cask, under the described conditions, are developed 

from the data in Hanson et al. (2008) for 45,000 megawatt-days per MTU spent fuel.

Hanson et al. (2008) suggest that volatile fission products, like the cesium isotopes, exhibit 

release behavior similar to fission gases. However, any cesium isotope would be released as 

the oxide or chloride and would therefore behave more like volatile compounds than like gases. 

Because the volatile compounds tend to migrate to the fuel rim, and Einziger (2007) 

recommends 3 × 10-5 as an appropriate release fraction for rim material, the present analysis 

uses this release fraction for volatiles, including ruthenium.

Hanson et al. (2008) describe a number of mechanical tests performed on unoxidized fuel of 

varying burnup. Page 4.12 of Hanson et al. (2008) summarizes release fractions from these 

tests for the fuel that appears to be the most appropriate. This analysis uses a release fraction 

of 4.8 × 10-6

, based on the information in Hanson et al. (2008), for release of fine particles from 

the rod to the cask. An analysis that recognizes and accommodates the uncertainty in 

estimating release fractions would be appropriate but is beyond the scope of this study.

Figure 7.11 in Sprung et al. (2000) presents release fractions of several compounds as 

functions of the available leak area. The compounds studied represent the physical and 

chemical groups present in spent nuclear fuel—gas, volatiles, and particulate matter. This 

figure served as the basis for estimating the cask-to-environment release fractions of the 

physical and chemical groups studied.

Table E-16 summarizes the parameters from which release fractions were developed.
E-25

Table E-16 Parameters for Determining Release Functions for the Accidents That Would 

Result in Release of Radioactive Material

Cask 

Orientation End Corner Side Side Side Side Corner

Impact 

Speed (kph) 193 193 193 193 145 145 145

Seal metal metal elastomer metal elastomer metal metal

Cask-toEnvironment 

Release 

Fraction

Gas 0.800 0.800 0.800 0.800 0.800 0.800 0.800

Particles 0.70 0.70 0.70 0.70 0.70 0.70 0.64

Volatiles 0.50 0.50 0.50 0.50 0.50 0.50 0.45

CRUD 0.001 0.001 0.001 0.001 0.001 0.001 0.001

Rod-to-Cask 

Release 

Fraction

Gas 0.12 0.12 0.12 0.12 0.12 0.12 0.12

Particles 4.80 × 10-6 4.80 × 10-6 4.80 × 10-6 4.80 × 10-6 4.80 × 10-6 4.80 × 10-6 2.40 × 10-6

Volatiles 3.00 × 10-5 3.00 × 10-5 3.00 × 10-5 3.00 × 10-5 3.00 × 10-5 3.00 × 10-5 1.50 × 10-5

CRUD 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Conditional Probability for 

combined Rod-CaskEnvironment release

5.96 × 10-12 3.57 × 10-11 1.79 × 10-11 1.79 × 10-11 3.40 × 10-10 3.40 × 10-10 1.13 × 10-10

Table E-17 shows sources of the parameter values in Table E-16. The parameter values are 

consistent with those in Sanders et al. (1992).

Table E-17 Sources of the Parameter Values in Table E-16

Release 

fraction Comment

Cask-toEnvironment 

Release 

Fraction

Gas 0.800

The basis of each release fraction is the size of the 

gap in the seal—the leak area—provided for each 

combination of impact speed and orientation in 

Table C-1 in Appendix C. Release fractions were 

obtained from the graph in Figure 7.11 (pp. 7–53) in 

Sprung et al. (2000).

Particles 0.70

Particles—

Corner Impact 0.64

Volatiles 0.50

Volatiles—

Corner Impact 0.45

CRUD 0.001 This release fraction is based on Einziger and 

Beyer (2007) and discussed in Section E.5.4.1.

Rod-to-Cask 

Release 

Fraction

Gas 0.12 From Einziger, personal communication.

Particles 4.80 × 10-6

From the release fraction in Hanson et al. (2008), 

Table 4.10. Particles—

Corner Impact 2.4 × 10-6

Volatiles 3.00 × 10-5

Average of values in Hanson et al. (2008), 

Section 4.3, p. 4.12. Volatiles—

Corner Impact 1.5 × 10-5

CRUD 1.00 CRUD is on the outside of the rod.

The release from these potential accidents is not at ground level but at about 2 meters (6.6 feet) 

above ground, taking into account the height of the flatcar and the diameter of the horizontally 

mounted cask. The factor H in Equation E-4 is the release height, 2 meters (6.6 feet) in this 
E-26

case. The gas flowing from the cask is warmer than ambient and the heat rate is about 

660 watts per assembly,

8 so that the plume of material will be lofted slightly. Results of the 

RADTRAN model of Equation E-4 indicate a maximum air concentration and ground deposition 

at about 21 meters (69 feet) downwind from the cask. Because the release is slightly elevated 

above ground level and the maximum air concentration at the ground and the maximum 

deposition are downwind from the release point,

9 based on the postulated meteorological 

conditions, the MEI would be located at this point. Figures 5-4a and 5-4b in Chapter 5 present a

graph of the plume. Results of the RADTRAN calculation, the radiation dose (consequence) 

that could result if radioactive material were released in a spent fuel cask accident, are shown in 

Table E-18.

Table E-18 MEI Doses (Consequences) in Sv from Accidents That Involve a Release

Cask 

Orientation

Impact 

Speed 

(kph)

Seal Inhalation ReSuspension

CloudShineGroundShine Total

End 193 metal 1.6 0.014 8.8 × 10-5 9.4 × 10-4 1.6

Corner 193 metal 1.6 0.014 8.8 × 10-5 9.4 × 10-4 1.6

Side 193 elastomer 1.6 0.014 8.8 × 10-5 9.4 × 10-4 1.6

Side 193 metal 1.6 0.014 8.8 × 10-5 9.4 × 10-4 1.6

Side 145 elastomer 1.6 0.014 4.5 × 10-6 3.6 × 10-5 1.6

Side 145 metal 1.6 0.014 8.8 × 10-5 9.4 × 10-4 1.6

Corner 145 metal 0.73 0.0063 5.1 × 10-5 9.2 × 10-4 0.74

When the doses in Table E-18 are multiplied by the probabilities in Table E-16, the “conditional 

dose risks” of Table E-19 result.

Table E-19 MEI Conditional Dose Risks in Sv from Accidents That Involve a Release

Cask 

Orientation

Impact 

Speed 

(kph)

Seal Inhalation ReSuspension

CloudShineGroundShine Total

End 193 metal 9.5 × 10-12 8.3 × 10-14 5.2 × 10-16 5.6 × 10-15 9.5 × 10-12

Corner 193 metal 5.7 × 10-11 5.0 × 10-13 3.1 × 10-15 3.4 × 10-14 5.7 × 10-11

Side 193 elastomer 2.9 × 10-11 2.5 × 10-13 1.6 × 10-15 1.7 × 10-14 2.9 × 10-11

Side 193 metal 2.9 × 10-11 2.5 × 10-13 1.6 × 10-15 1.7 × 10-14 2.9 × 10-11

Side 145 elastomer 5.4 × 10-10 4.8 × 10-12 3.0 × 10-14 3.2 × 10-13 5.4 × 10-10

Side 145 metal 5.4 × 10-10 4.8 × 10-12 3.0 × 10-14 3.2 × 10-13 5.4 × 10-10

Corner 145 metal 8.3 × 10-11 7.1 × 10-13 5.8 × 10-15 1.0 × 10-13 8.3 × 10-11

Population doses are calculated by integrating the rural, suburban, and urban population 

densities, respectively, over the largest plume footprint in the dispersion calculation: 1,420 km2

(548 mi2

) for average meteorological stability (Pasquill Class D, windspeed of 4.7 meters per 

second (m/s) (10.5 mph). The calculation was repeated using very stable meteorology 

(Pasquill: stability F, windspeed 0.5 m/s (1.1 mph)), but the difference was negligible because 

 

8 For 9-year-cooled PWR fuel from the ORIGEN analysis. 660 watts per assembly = 17,160 watts per cask =

4.1 kcal/sec. 9 Earlier versions of RADTRAN (before RADTRAN 5.6) could not model elevated releases.
E-27

of the relatively low elevation of the release. Collective dose risks are calculated by multiplying 

each population dose by the appropriate conditional probability. As an example, Table E-20

presents the collective dose risks for the side-impact accident at an impact speed of 193 kph 

(120 mph).

Table E-20 Collective Inhalation and External Dose Risks for the Side-Impact, 

193-kph-Impact-Speed Accident for the 16 Rail Routes Analyzed

Collective Internal Dose Risk

(person-Sv)

Collective External Dose Risk

(person-Sv)

ORNL DEAF 

SMITH HANFORD SKULL VALLEY ORNL DEAF SMITH HANFORD SKULL VALLEY

MAINE YANKEE

Rural 5.9 × 10-13 7.5 × 10-13 1.4 × 10-12 1.0 × 10-12 1.1 × 10-13 1.4 × 10-13 2.5 × 10-13 1.9 × 10-13

Suburban 7.2 × 10-12 7.6 × 10-12 6.4 × 10-12 7.2 × 10-12 1.3 × 10-12 1.4 × 10-12 1.2 × 10-12 1.3 × 10-12

Urban 2.7 × 10-11 3.2 × 10-11 2.5 × 10-11 2.2 × 10-11 4.9 × 10-12 5.9 × 10-12 4.4 × 10-12 4.0 × 10-12

KEWAUNEE

Rural 2.7 × 10-13 1.8 × 10-13 2.3 × 10-13 3.8 × 10-13 4.9 × 10-14 3.2 × 10-14 4.1 × 10-14 6.8 × 10-14

Suburban 2.9 × 10-12 1.8 × 10-12 1.5 × 10-12 3.0 × 10-12 5.3 × 10-13 3.2 × 10-13 2.6 × 10-13 5.5 × 10-13

Urban 1.5 × 10-11 1.0 × 10-11 3.7 × 10-12 1.1 × 10-11 2.7 × 10-12 1.8 × 10-12 6.8 × 10-13 2.0 × 10-12

INDIAN POINT

Rural 3.2 × 10-12 6.3 × 10-13 7.3 × 10-13 8.5 × 10-13 5.7 × 10-13 1.1 × 10-13 1.3 × 10-13 1.5 × 10-13

Suburban 8.6 × 10-11 6.7 × 10-12 6.0 × 10-12 7.5 × 10-12 1.6 × 10-11 1.2 × 10-12 1.1 × 10-12 1.4 × 10-12

Urban 6.2 × 10-10 3.7 × 10-11 2.8 × 10-11 3.3 × 10-11 1.1× 10-10 6.7 × 10-12 5.1 × 10-12 5.9 × 10-12

IDAHO NATIONAL LAB

Rural 4.7 × 10-13 1.1 × 10-13 7.2 × 10-14 4.6 × 10-14 8.5 × 10-14 2.0 × 10-14 1.3 × 10-14 8.3 × 10-15

Suburban 2.6 × 10-12 1.2 × 10-12 5.7 × 10-13 4.1 × 10-13 4.7 × 10-13 2.1 × 10-13 1.0 × 10-13 7.4 × 10-14

Urban 6.1 × 10-12 2.9 × 10-12 1.7 × 10-12 2.0 × 10-12 1.1 × 10-12 5.3 × 10-13 3.0 × 10-13 3.6 × 10-13

The values are calculated as in the following example. This example is for one accident 

scenario i: a side-impact accident with an impact speed of 193-kph (120 mph).

Collective internal dose risk = (accident rate) * (route segment length) * Pcond_i * (doseinhalation_i + 

doseresuspension_i) * ∫[(population density) ∗ (plume footprint area)]. Internal dose includes the

doses from both direct inhalation and inhaled material resuspended in air. External doses 

include cloudshine and groundshine. The NRC cites the total effective dose equivalent, which 

includes both inhalation (internal) doses and external doses. The complete collective dose risk 

is the sum of the collective dose risks shown in Table E-20 over all accident scenarios, summed 

over the entire route:

Total collective dose risks (person-Sv) for each route =

 ∑ ∑ [dose riskinternali + dose riskexternal_i

7

i=1 ] r,s,u .

Table E-21 shows the total dose risk for each route.
E-28

Table E-21 Total Collective Dose Risks (Person-Sv) for Each Route from an Accident 

Involving Release, Per Shipment

ORNL DEAF SMITH HANFORD SKULL VALLEY

MAINE YANKEE 3.5 × 10-14 4.1 × 10-14 3.2 × 10-14 3.0 × 10-14

KEWAUNEE 1.8 × 10-14 1.2 × 10-14 5.4 × 10-15 1.4 × 10-14

INDIAN POINT 1.5 × 10-11 5.9 × 10-13 5.3 × 10-13 1.9 × 10-13

INL 9.4 × 10-14 1.5 × 10-13 4.1 × 10-14 2.7 × 10-13

E.5 Summary

The technical observations for the analysis of accidents are as follows:

• Event trees based on current accident statistics show that the conditional probability of a 

severe accident for either truck or rail is one in 100,000 or less. The probability of a fire 

that would damage a cask on a railcar enough to cause loss of gamma shielding or 

release of radioactive material is negligible.

• The analyses in Appendices C and D demonstrate that there could be no releases of 

radioactive material from a cask carrying canistered fuel, and that the only cask that 

could suffer a loss of lead shielding or release of radioactive material is the Rail-Lead

cask. Most accidents involving spent fuel casks—99.9999999 percent—do not lead to 

either a release of radioactive material or a loss of lead gamma shielding.

• A dose larger than the 10 CFR 71.51 limit would be sustained only for the 

extra-regulatory impacts in which more than 2 percent of the lead shielding is lost and 

only if the receptor is within 4 meters (13 feet) of the cask.
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ML103560271.




UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

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OFFICIAL BUSINESS
NUREG

-2125

Final

Spent Fuel Transportation Risk Assessment