A STUDY OF IN-PACKAGE NUCLEAR

Service de M´ etrologie Nucl´ eaire

A STUDY OF IN-PACKAGE NUCLEAR

CRITICALITY IN POSSIBLE BELGIAN SPENT NUCLEAR FUEL REPOSITORY DESIGNS

Th` ese de doctorat r´ ealis´ ee sous Olivier Wantz la direction de Messieurs les Professeurs

A. Dubus et R. Beauwens

en vue de l’obtention du grade de

Docteur en Sciences Appliqu´ ees

Belgian Spent Nuclear Fuel Repository Designs

Olivier Wantz

Service de M´ etrologie Nucl´ eaire Facult´ e des Sciences Appliqu´ ees Universit´ e Libre de Bruxelles

15th April 2005

Je voudrais tout d’abord remercier le Professeur Robert Beauwens de m’avoir accueilli dans son Service il y a cinq ans en me proposant de devenir chercheur sous contrat avec ONDRAF/NIRAS.

Je le remercie non seulement pour la confiance qu’il m’a accord´ee durant ces ann´ees mais

´egalement pour son ´ecoute attentive et ses conseils.

Olivier Smidts a ´et´e mon premier directeur de th`ese et je regrette qu’il ne travaille plus d´esormais dans le Service de M´etrologie Nucl´eaire. Je voudrais lui t´emoigner ma gratitude pour les deux ann´ees pass´ees en sa compagnie.

Je remercie le Professeur Alain Dubus qui a ´et´e mon second directeur pour ses remarques et critiques sur le travail que j’ai effectu´e.

Je veux remercier ONDRAF/NIRAS, et plus particuli`erement Messieurs Peter De Preter et Wim Cool qui m’ont fait confiance au cours de ces 5 ann´ees.

Je tiens aussi a remercier l’ensemble des membres du Service de M´etrologie Nucl´eaire, et plus particuli`erement Mme Jos´ee Immers qui a toujours ´et´e d’une grande gentillesse envers moi.

Plus personnellement, je veux exprimer ma reconnaissance vis-`a-vis de mes parents ainsi que ma soeur qui m’ont soutenu durant ces ann´ees de doutes et de joies. Je remercie ma compagne Lynn pour sa patience, son ´ecoute attentive, et ses petits conseils toujours judicieux.

Je souhaite d´edier ce travail `a Julie. Elle a toujours ´et´e la personne que j’admire le plus pour toutes ses qualit´es, son sens de l’humour, sa compr´ehension, sa tendresse, et son amiti´e a toute

´epreuve.

I INTRODUCTION 1

1 General Introduction 2

1.1 Context . . . . 2

1.2 Organization of the Work . . . . 3

1.3 Main Achievements . . . . 4

2 Radioactive Waste 5 2.1 Introduction . . . . 5

2.2 Definition . . . . 5

2.3 Origin . . . . 5

2.3.1 Introduction . . . . 5

2.3.2 Waste from the Nuclear Fuel Cycle . . . . 6

2.3.3 Waste from the Production and Use of Radionuclides . . . . 6

2.3.4 Waste from the Decommissioning of Nuclear Facilities . . . . 7

2.3.5 Waste Not Deriving from Nuclear Activities . . . . 7

2.4 International Classification . . . . 7

2.5 Belgian Situation . . . . 8

2.5.1 Classification . . . . 8

2.5.2 Description . . . . 10

2.5.3 Belgian Inventory . . . . 12

2.6 HLW Management . . . . 14

2.6.1 Introduction . . . . 14

2.6.2 HLW Management Options . . . . 14

2.6.3 HLW Repository Characterization . . . . 16

2.6.4 HLW-SFA Repository Example . . . . 17

2.7 Conclusions . . . . 20

3 Nuclear Criticality Risk 21 3.1 Introduction . . . . 21

3.2 Nuclear Fission Reactions . . . . 21

3.2.1 Isotopes . . . . 21

3.2.2 Examples . . . . 22

3.2.3 Fission Products . . . . 23

3.2.4 Multiplication Factor . . . . 24

3.2.5 Feedback Mechanisms . . . . 26

3.3 Risk . . . . 27

3.3.1 Definitions . . . . 27

3.3.2 Nuclear Risk . . . . 27

3.3.3 Radiological Consequences . . . . 28

3.3.4 Nuclear Criticality Risk . . . . 29

3.4 Conclusions . . . . 31

II SCENARIOS DEVELOPMENT 32 4 Scenarios Development Procedures 33 4.1 Introduction . . . . 33

4.2 Definitions . . . . 33

4.2.1 Features, Events, and Processes . . . . 33

4.2.2 Scenarios . . . . 34

4.2.3 Scenarios Development . . . . 35

4.3 Aims of Scenarios Development . . . . 35

4.4 The Bottom-Up Procedure: The Sandia/NRC Procedure . . . . 37

4.4.1 Introduction . . . . 37

4.4.2 Identification of FEPs . . . . 37

4.4.3 Classification of FEPs . . . . 38

4.4.4 Screening of FEPs . . . . 38

4.4.5 Scenarios Construction . . . . 38

4.4.6 Screening of Scenarios . . . . 40

4.4.7 Use of Expert Judgment in Scenarios Selection Procedure . . . . 40

4.5 The Top-Down Procedure: The Joint SKI/SKB Investigations . . . . 41

4.5.1 Introduction . . . . 41

4.5.2 System Elements Identification . . . . 41

4.5.3 Element States Identification . . . . 41

4.5.4 Incompatible States Identification and Screening . . . . 41

4.5.5 Scenarios Formation and Final Screening . . . . 42

4.6 Conclusions . . . . 42

5 Criticality Scenarios Development: Examples 44 5.1 Introduction . . . . 44

5.2 Yucca Mountain Repository Project (USA) . . . . 44

5.2.1 Introduction . . . . 44

5.2.2 U.S. Design . . . . 44

5.2.3 U.S. Waste . . . . 46

5.2.4 U.S. Waste Packages . . . . 46

5.2.5 U.S. Inventory . . . . 49

5.2.6 First Approach to Criticality Scenarios Development . . . . 49

5.2.7 In-Package Critical Configurations . . . . 50

5.2.8 Scenarios Classification . . . . 53

5.2.9 Conclusions . . . . 54

5.3 KBS-3 Repository Project (Sweden) . . . . 55

5.3.1 Introduction . . . . 55

5.3.2 KBS-3 Repository Design . . . . 55

5.3.3 Swedish Spent Nuclear Fuel . . . . 55

5.3.4 Spent Fuel Canisters . . . . 57

5.3.5 Swedish Inventory . . . . 58

5.3.6 Plutonium Criticality Scenarios . . . . 58

5.3.7 Uranium Criticality Scenarios . . . . 58

5.3.8 Conclusions . . . . 59

5.4 Conclusions . . . . 59

6 In-Package Criticality Scenarios Development: SAFIR2 60

6.1 Introduction . . . . 60

6.2 SAFIR2 Repository Design . . . . 60

6.2.1 Introduction . . . . 60

6.2.2 ZAGALC Waste Repository Design . . . . 61

6.2.3 ZAGALS Waste Repository Designs . . . . 62

6.3 Comparison Between Yucca Mountain and SAFIR2 Designs . . . . 63

6.3.1 Introduction . . . . 63

6.3.2 Comparative Approach . . . . 63

6.3.3 Disruptive FEPs Identification . . . . 64

6.4 Bottom-Up Scenarios Development . . . . 65

6.4.1 Introduction . . . . 65

6.4.2 FEPs Listing and Screening . . . . 65

6.4.3 Scenarios Formation . . . . 66

6.4.4 Scenarios Screening . . . . 68

6.4.5 Bottom-Up Criticality Scenarios List . . . . 69

6.5 Top-Down Scenarios Development . . . . 71

6.5.1 Introduction . . . . 71

6.5.2 System and Element States . . . . 71

6.5.3 Combination of System States and First Screening . . . . 72

6.5.4 Second Screening and Top-Down Criticality Scenarios List . . . . 75

6.6 Conclusions . . . . 75

7 In-Package Criticality Scenarios Development: Supercontainer Designs 76 7.1 Introduction . . . . 76

7.2 Supercontainer 1 . . . . 76

7.2.1 Introduction . . . . 76

7.2.2 Design . . . . 76

7.2.3 Bottom-Up Scenarios . . . . 77

7.2.4 Top-Down Scenarios . . . . 79

7.2.5 Conclusions . . . . 80

7.3 Supercontainer 2 . . . . 81

7.3.1 Introduction . . . . 81

7.3.2 Design . . . . 81

7.3.3 Bottom-Up Scenarios . . . . 81

7.3.4 Top-Down Scenarios . . . . 83

7.3.5 Conclusions . . . . 84

7.4 Supercontainer 3 . . . . 84

7.4.1 Introduction . . . . 84

7.4.2 Design . . . . 84

7.4.3 Bottom-Up Scenarios . . . . 84

7.4.4 Top-Down Scenarios . . . . 85

7.4.5 Conclusions . . . . 86

7.5 Conclusions . . . . 86

III CRITICALITY CALCULATIONS 87 8 The MCNP Computer Code 88 8.1 Introduction . . . . 88

8.2 Monte Carlo Methods . . . . 88

8.3 Introduction to MCNP Features . . . . 89

8.3.1 Nuclear Data and Reactions . . . . 89

8.3.2 Neutron Thermal S(α, β) Tables . . . . 89

8.3.3 Source Specification . . . . 89

8.3.4 Output . . . . 89

8.3.5 Estimation of Monte Carlo Errors . . . . 90

8.3.6 Variance Reduction . . . . 90

8.3.7 Tally Choice . . . . 90

8.3.8 Nonanalog Monte Carlo . . . . 90

8.3.9 Variance Reduction Tools in MCNP . . . . 91

8.3.10 MCNP Geometry . . . . 91

8.3.11 MCNP Input . . . . 92

8.3.12 MCNP Plotter . . . . 92

8.4 Criticality Calculations . . . . 92

8.4.1 Introduction . . . . 92

8.4.2 Criticality Program Flow . . . . 92

8.4.3 Estimation of k eff Confidence Intervals . . . . 93

8.4.4 Recommendation for Making a Good Criticality Calculation . . . . 96

8.5 Conclusions . . . . 97

9 In-Package Criticality Calculations: SAFIR2 99 9.1 Introduction . . . . 99

9.2 Modelling and Hypotheses . . . 102

9.2.1 FAs . . . 102

9.2.2 Fuels . . . 102

9.2.3 Canisters . . . 102

9.2.4 Disposal Gallery and Host Rock . . . 103

9.3 Results . . . 103

9.3.1 Acceptation Criterium . . . 103

9.3.2 Comparison with Previous Results . . . 103

9.3.3 Geometry Simplification . . . 104

9.3.4 Water Moderation Effects . . . 106

9.3.5 FAs Geometry Alteration Effects . . . 106

9.3.6 Gallery Geometry Alteration Effects . . . 106

9.4 Conclusions . . . 107

10 In-Package Criticality Calculations: Supercontainer 1 112 10.1 Introduction . . . 112

10.2 Modelling and Hypotheses . . . 112

10.2.1 FAs . . . 112

10.2.2 Fuels . . . 113

10.2.3 Overpacks . . . 113

10.2.4 Supercontainer . . . 113

10.2.5 Disposal Gallery and Host Rock . . . 114

10.3 Results . . . 115

10.3.1 Water Moderation and Burn-up Effects . . . 115

10.3.2 SFAs Geometry Alteration Effects . . . 117

10.3.3 Supercontainer Internal Geometry Variations Effects . . . 121

10.4 Comparison with Scenarios . . . 125

10.5 Conclusions . . . 126

11 In-Package Criticality Calculations: Supercontainer 2 127

11.1 Introduction . . . 127

11.2 Modelling and Hypotheses . . . 127

11.2.1 FAs . . . 127

11.2.2 Fuel . . . 128

11.2.3 Overpack . . . 128

11.2.4 Supercontainer . . . 129

11.2.5 Disposal Gallery and Host Rock . . . 129

11.2.6 Additional Informations . . . 130

11.3 Results . . . 132

11.3.1 Introduction . . . 132

11.3.2 Overpack Water Density Variation . . . 132

11.3.3 Overpack Water Volume Fraction Variation . . . 135

11.3.4 Influence of the Overpack Material . . . 140

11.4 Comparison with Scenarios . . . 141

11.5 Conclusions . . . 142

12 In-Package Criticality Calculations: Supercontainer 3 143 12.1 Introduction . . . 143

12.2 Modelling and Hypotheses . . . 144

12.2.1 FAs . . . 144

12.2.2 Fuels . . . 144

12.2.3 Overpack . . . 144

12.2.4 Supercontainer . . . 145

12.2.5 Disposal Gallery and Host Rock . . . 145

12.2.6 Additional Informations . . . 145

12.3 Results . . . 148

12.3.1 Introduction . . . 148

12.3.2 SC3 Internal Geometry Variation . . . 148

12.3.3 FAs Geometry Variation . . . 151

12.3.4 Evolution in Time . . . 157

12.4 Comparison with Scenarios . . . 182

12.5 Conclusions . . . 182

IV CONCLUSIONS AND PERSPECTIVES 183 13 General Conclusions and Perspectives 184 13.1 Context . . . 184

13.2 Scenarios Development . . . 185

13.3 Criticality Calculations . . . 186

13.4 Consequences . . . 186

13.5 Perspectives . . . 187

APPENDICES 188 A Bowman and Venneri 189 A.1 Introduction . . . 189

A.2 Studied Configurations . . . 189

A.2.1 Introduction . . . 189

A.2.2 Case A: Water Ingress in TFMs and Rock Systems (Negative Feedback) . 190

A.2.3 Case B: TFMs Migration to Wet Rock Systems (Negative Feedback) . . . 190

A.2.4 Case C: TFMs Migration to wet rock Systems (Positive Feedback) . . . . 191

A.2.5 Case D: TFMs Migration to Wet Rock Systems (Positive Feedback) . . . 191

A.2.6 Case E: Drying-Out of TFMs Deposits in Rock (Positive Feedback) . . . 191

A.2.7 Case F: Small Volume Systems (Positive Feedback) . . . 191

A.3 Summary of the Results . . . 191

A.4 Conclusions . . . 192

B The Oklo Natural Fission Reactors 194 B.1 Introduction . . . 194

B.2 Situation . . . 194

B.3 The Oklo Phenomenon . . . 194

B.3.1 Uranium Accumulation at Oklo . . . 194

B.3.2 Criticality at Oklo . . . 195

B.3.3 Cyclic Time Evolution at Oklo . . . 195

B.4 Conclusions . . . 196

C Yucca Mountain Out-Package Criticality Scenarios 197 C.1 Introduction . . . 197

C.2 Near-Field Critical Configurations . . . 197

C.2.1 Failure of the Waste Package Bottom after Failure of the Top . . . 197

C.2.2 Failure of the Waste Package Bottom First . . . 198

C.3 Far-Field Critical Configurations . . . 201

C.3.1 Unsaturated Zone . . . 201

C.3.2 Saturated Zone . . . 202

C.4 Conclusions . . . 203

D NEA FEPs Database Screening 205 D.1 Introduction . . . 205

D.2 Assessment Basis (IFEP 0) . . . 206

D.3 External Factors (IFEP 1) . . . 208

D.3.1 Repository Issues (IFEP 1.1) . . . 208

D.3.2 Geological Processes and Effects (IFEP 1.2) . . . 210

D.3.3 Climatic Processes and Effects (IFEP 1.3) . . . 211

D.3.4 Future Human Actions (Active) (IFEP 1.4) . . . 213

D.3.5 Other (IFEP 1.5) . . . 215

D.4 Disposal System Domain: Environmental Factors (IFEP 2) . . . 215

D.4.1 Wastes and Engineered Features (IFEP 2.1) . . . 215

D.4.2 Geological Environment (IFEP 2.2) . . . 222

D.4.3 Surface Environment (IFEP 2.3) . . . 225

D.4.4 Human Behavior (IFEP 2.4) . . . 226

D.5 Radionuclide/Contaminant Factors (IFEP 3) . . . 228

D.5.1 Contaminant Characteristics (IFEP 3.1) . . . 228

D.5.2 Contaminant Release/Migration Factors (IFEP 3.2) . . . 229

D.5.3 Exposure Factors (IFEP 3.3) . . . 231

D.6 Conclusions . . . 232

E MCNP Input File Example 233 E.1 Introduction . . . 233

E.2 Listing of the Input File Example . . . 233

E.3 Description of the Input File Example . . . 237

E.4 Figures . . . 238

F Atomic Densities for MCNP Input 245

F.1 Introduction . . . 245

F.2 Fuel Assemblies . . . 245

F.2.1 UOX Fuels . . . 245

F.2.2 MOX Fuel . . . 246

F.2.3 Zircalloy Cladding . . . 246

F.3 Sand Filling . . . 246

F.4 Canister . . . 247

F.5 Supercontainer . . . 247

F.6 Gallery Walls . . . 247

F.7 Boom Clay . . . 248

Bibliography 249

2.1 A typical (Westinghouse) 17 × 17 FA. . . . 11

2.2 The KBS-3 multi-barrier design. . . . 18

2.3 The multi-barrier concept. . . . 19

3.1 Example of the thermal fission reaction of ²³⁵ U . . . . 22

3.2 Fission products yields for the thermal fission reaction of ²³⁵ U . . . . 23

3.3 Typical neutron fate in a reactor. . . . 24

4.1 Sandia scenarios selection procedure. . . . 37

4.2 Sandia scenarios formation (left) and screening (right). . . . 39

4.3 Identification of the incompatible states. . . . 42

5.1 General presentation of the Yucca Mountain site. . . . 45

5.2 Schematic view of different U.S. waste packages and the engineered barriers in a disposal gallery. . . . 47

5.3 Detailed waste package for 21 PWR uncanistered fuel assemblies. . . . 48

5.4 Top of the criticality FEPs tree. . . . 50

5.5 The criticality FEPs tree: water accumulates in waste package. . . . 51

5.6 The criticality FEPs tree: absorbers dissolution. . . . 52

5.7 The criticality FEPs tree: waste package internal structures degrade first. . . . . 53

5.8 The criticality FEPs tree: water flows through the waste package. . . . 54

5.9 The Svea 100 BWR assembly comprises four 5 × 5 arrays of pins with a 2.5 mm gap between adjacent arrays. . . . 56

5.10 F 17 × 17 PWR assembly containing no burnable poison. . . . 56

5.11 F 17 × 17 PWR assembly containing burnable poison. . . . 56

5.12 Canister design for 12 BWR SFAs. . . . 57

5.13 Canister design for 4 PWR SFAs. . . . 57

6.1 Schematic view of the SAFIR2 ZAGALC waste repository design. . . . 61

6.2 The SAFIR2 reference design for ZAGALS waste (gallery cross section). . . . 62

6.3 Bottom-up scenarios formation: initial criticality FEPs tree. . . . 67

6.4 Bottom-up scenarios first screening: incompatible combinations between FEPs are screened off. . . . 70

6.5 The 5 components of the system and their respective states. . . . 72

6.6 Incompatible system states are linked by dotted lines. . . . 73

7.1 The SC1 design (vertical cross section of a disposal gallery). . . . 77

7.2 Upper part of the SC1 in-package criticality FEPs Tree. FEPs 1, 2, and 3 are assumed to have occurred. . . . 78

7.3 The SC2 overpack holding four fuel assemblies and its cross-shaped separator. . . 81

7.4 Upper part of the SC2 in-package criticality FEPs Tree. FEPs 1, 2, and 7 are

assumed to have occurred. . . . 82

7.5 The SC3 full metal overpack of holding 4 fuel assemblies. . . . 85

9.1 The SAFIR2 reference design. . . 100

9.2 The SAFIR2 variant design . . . 101

9.3 Influence of moderation inside the canisters. . . 108

9.4 Influence of the fuel assemblies geometry (reference design) [nominal distance between the fuel rods centres = 1.26 cm]. . . 109

9.5 Influence of the fuel assemblies geometry (variant design) [nominal distance be- tween the fuel rods centres = 1.26 cm]. . . 110

9.6 Influence of the geometry of the disposal gallery (reference design) [nominal dis- tance between the disposal tubes and the gallery centres = 45 cm]. . . 111

10.1 The SC1 design (vertical cross section of a disposal gallery). . . 113

10.2 Influence of moderation inside the overpacks. . . 116

10.3 Influence of the FAs geometry (0 % vol. water inside the overpacks) [nominal distance between the fuel rods centres = 1.26 cm]. . . 118

10.4 Influence of the FAs geometry (40 % vol. water inside the overpacks) [nominal distance between the fuel rods centres = 1.26 cm]. . . 119

10.5 Influence of the FAs geometry (100 % vol. water inside the overpacks) [nominal distance between the fuel rods centres = 1.26 cm]. . . 120

10.6 Influence of the supercontainer geometry (0 % vol. water inside the overpacks) [nominal distance between the overpacks = 10 cm]. . . 122

10.7 Influence of the supercontainer geometry (40 % vol. water inside the overpacks) [nominal distance between the overpacks = 10 cm]. . . 123

10.8 Influence of the supercontainer geometry (100 % vol. water inside the overpacks) [nominal distance between the overpacks = 10 cm]. . . 124

11.1 The SC2 overpack holding four fuel assemblies and its cross-shaped separator. . . 129

11.2 MCNP cross section of a supercontainer in a disposal gallery. . . 131

12.1 The SC3 full metal overpack of holding 4 fuel assemblies. . . 143

12.2 Influence of the cross separator geometry (additional cooling time= 0 year / water density in overpack = 1.00) [nominal cross separator thickness = 8 cm]. . . 149

12.3 Influence of the burn-up (additional cooling time = 0 year / water density in overpack = 1.00 / distance between fuel rods = 1.26 cm). . . 150

12.4 Influence of the FAs geometry (additional cooling time = 0 year / water density in overpack = 0.00) [nominal distance between the fuel rods centres = 1.26 cm]. . 152

12.5 Influence of the FAs geometry (additional cooling time = 0 year / water density in overpack = 0.25) [nominal distance between the fuel rods centres = 1.26 cm]. . 153

12.6 Influence of the FAs geometry (additional cooling time = 0 year / water density in overpack = 0.50) [nominal distance between the fuel rods centres = 1.26 cm]. . 154

12.7 Influence of the FAs geometry (additional cooling time = 0 year / water density in overpack = 0.75) [nominal distance between the fuel rods centres = 1.26 cm]. . 155

12.8 Influence of the FAs geometry (additional cooling time = 0 year / water density in overpack = 1.00) [nominal distance between the fuel rods centres = 1.26 cm]. . 156

12.9 Influence of the FAs geometry (additional cooling time = 10 ³ years / water density in overpack = 0.00) [nominal distance between the fuel rods centres = 1.26 cm]. . 158

12.10Influence of the FAs geometry (additional cooling time = 10 ⁴ years / water density in overpack = 0.00) [nominal distance between the fuel rods centres = 1.26 cm]. . 159

12.11Influence of the FAs geometry (additional cooling time = 10 ⁵ years / water density

in overpack = 0.00) [nominal distance between the fuel rods centres = 1.26 cm]. . 160

12.12Influence of the FAs geometry (additional cooling time = 10 ⁶ years / water density

in overpack = 0.00) [nominal distance between the fuel rods centres = 1.26 cm]. . 161

12.13Influence of the FAs geometry (additional cooling time = 10 ³ years / water density in overpack = 0.25) [nominal distance between the fuel rods centres = 1.26 cm]. . 164

12.14Influence of the FAs geometry (additional cooling time = 10 ³ years / water density in overpack = 0.50) [nominal distance between the fuel rods centres = 1.26 cm]. . 165

12.15Influence of the FAs geometry (additional cooling time = 10 ³ years / water density in overpack = 0.75) [nominal distance between the fuel rods centres = 1.26 cm]. . 166

12.16Influence of the FAs geometry (additional cooling time = 10 ³ years / water density in overpack = 1.00) [nominal distance between the fuel rods centres = 1.26 cm]. . 167

12.17Influence of the FAs geometry (additional cooling time = 10 ⁴ years / water density in overpack = 0.25) [nominal distance between the fuel rods centres = 1.26 cm]. . 168

12.18Influence of the FAs geometry (additional cooling time = 10 ⁴ years / water density in overpack = 0.50) [nominal distance between the fuel rods centres = 1.26 cm]. . 169

12.19Influence of the FAs geometry (additional cooling time = 10 ⁴ years / water density in overpack = 0.75) [nominal distance between the fuel rods centres = 1.26 cm]. . 170

12.20Influence of the FAs geometry (additional cooling time = 10 ⁴ years / water density in overpack = 1.00) [nominal distance between the fuel rods centres = 1.26 cm]. . 171

12.21Influence of the FAs geometry (additional cooling time = 10 ⁵ years / water density in overpack = 0.25) [nominal distance between the fuel rods centres = 1.26 cm]. . 172

12.22Influence of the FAs geometry (additional cooling time = 10 ⁵ years / water density in overpack = 0.50) [nominal distance between the fuel rods centres = 1.26 cm]. . 173

12.23Influence of the FAs geometry (additional cooling time = 10 ⁵ years / water density in overpack = 0.75) [nominal distance between the fuel rods centres = 1.26 cm]. . 174

12.24Influence of the FAs geometry (additional cooling time = 10 ⁵ years / water density in overpack = 1.00) [nominal distance between the fuel rods centres = 1.26 cm]. . 175

12.25Influence of the FAs geometry (additional cooling time = 10 ⁶ years / water density in overpack = 0.25) [nominal distance between the fuel rods centres = 1.26 cm]. . 176

12.26Influence of the FAs geometry (additional cooling time = 10 ⁶ years / water density in overpack = 0.50) [nominal distance between the fuel rods centres = 1.26 cm]. . 177

12.27Influence of the FAs geometry (additional cooling time = 10 ⁶ years / water density in overpack = 0.75) [nominal distance between the fuel rods centres = 1.26 cm]. . 178

12.28Influence of the FAs geometry (additional cooling time = 10 ⁶ years / water density in overpack = 1.00) [nominal distance between the fuel rods centres = 1.26 cm]. . 179

12.29Variation of k eff ± 2σ with time (water density in overpack = 1.00 / distance between the fuel rods centres = 1.26 cm). . . 180

12.30Evolution in time of the k eff (Swedish case. . . 181

A.1 Criticality curves for different radii in spherical geometries. . . 190

B.1 Schematic geological profile across the western margin of the Franceville Basin. . 195

B.2 Idealised cross section of a fossil reactor zone at Oklo. . . 196

C.1 The criticality tree: near-field scenarios. . . 198

C.2 The criticality tree: near-field scenarios. . . 199

C.3 The criticality tree: near-field scenarios. . . 200

C.4 The criticality tree: near-field scenarios. . . 200

C.5 The criticality tree: far-field scenarios. . . 202

C.6 The criticality tree: near-field scenarios. . . 203

C.7 The criticality tree: near-field scenarios. . . 204

E.1 Fuel, fuel rods, control and instrument tubes (cells). . . 239

E.2 Fuel, fuel rods, control and instrument tubes (surfaces). . . 240

E.3 Fuel assembly (cells). . . 241

E.4 Canisters inside the supercontainer (cells). . . 242

E.5 Gallery lateral cross section (cells). . . 243

E.6 Gallery vertical cross section (cells). . . 244

2.1 UOX fresh fuel assemblies characteristics. . . . 10

2.2 Typical 1450 days irradiated (45 GWd/tHM) UOX SFA radionuclide inventory. . 12

2.3 Typical 1450 days irradiated (45 GWd/tHM) MOX SFA radionuclide inventory. . 13

2.4 Belgian radioactive waste inventory. . . . 13

2.5 Number of SFAs foreseen for the Belgian repository. . . . 13

2.6 Composition of residues from reprocessing of 1,000 kg of SFAs. . . . 15

4.1 Illustration of consequences on the barrier states caused by individual FEPs. . . 42

5.1 Approximate inventories of U.S. nuclear waste destined for geological disposal and quantities planned for Yucca Mountain. . . . 49

6.1 First scenarios screening: incompatible system elements states are screened off. . 74

9.1 Atomic densities of the UOX and MOX fuels. . . 102

9.2 Reference design: UOX modelling validation. . . 104

9.3 Variant design: UOX modelling validation. . . . 104

9.4 Reference design: UOX geometry simplification. . . 105

9.5 Variant design: UOX geometry simplification. . . 105

9.6 Reference design: MOX geometry simplification. . . 105

9.7 Variant design: MOX geometry simplification. . . 105

10.1 Atomic densities (in 10 ³⁰ atom/m ³ ) of the UOX and MOX fuels as a function of the burn-up. . . 114

11.1 Zircalloy-4 mass composition. . . 128

11.2 UOX fuel isotopic composition. . . 128

11.3 Overpack materials compositions and densities. . . 129

11.4 Void overpack. . . 133

11.5 Water density inside the overpack = 0.25 g/cm ³ . . . 133

11.6 Water density inside the overpack = 0.50 g/cm ³ . . . 133

11.7 Water density inside the overpack = 0.75 g/cm ³ . . . 134

11.8 Water density inside the overpack = 1.00 g/cm ³ . . . 134

11.9 Pure sand filled overpack (no water). . . 135

11.10Water volume fraction in overpack = 10 %. . . 136

11.11Water volume fraction in overpack = 20 %. . . 136

11.12Water volume fraction in overpack = 30 %. . . 137

11.13Water volume fraction in overpack = 40 %. . . 137

11.14Water volume fraction in overpack = 50 %. . . 137

11.15Water volume fraction in overpack = 60 %. . . 138

11.16Water volume fraction in overpack = 70 %. . . 138

11.17Water volume fraction in overpack = 80 %. . . 138

11.18Water volume fraction in overpack = 90 %. . . 139

11.19Overpack material influence on k eff ± 2σ. . . 140

12.1 UOX fuel isotopic composition (additional cooling time = 0 years). . . 145

12.2 UOX fuel isotopic composition (additional cooling time = 10 ³ years). . . 146

12.3 UOX fuel isotopic composition (additional cooling time = 10 ⁴ years). . . 146

12.4 UOX fuel isotopic composition (additional cooling time = 10 ⁵ years). . . 147

12.5 UOX fuel isotopic composition (additional cooling time = 10 ⁶ years). . . 147

12.6 Isotopic contribution to k eff (additional cooling time = 0 year). . . 163

12.7 Isotopic contribution to k eff (additional cooling time = 10 ⁴ years). . . 163

F.1 UOX fuels atomic densities composition in 10 ^{− 2} atom barn ^{− 1} cm ^{− 1} . . . 246

F.2 MOX fuel atomic densities. . . 247

F.3 Stainless steel weight fractions and atomic densities. . . 247

F.4 Supercontainer concrete composition. . . 247

F.5 Supercontainer concrete atomic densities. . . 248

F.6 Gallery walls atomic densities. . . 248

F.7 Boom Clay weight composition. . . 248

F.8 Boom Clay atomic densities. . . 248

INTRODUCTION

General Introduction

1.1 Context

About 60 percent of the electricity production in Belgium originates from nuclear power plants.

Belgium owns 7 nuclear pressurized water reactors, which are located in two sites: 4 reactors in Doel and 3 reactors in Tihange. Together they have a capacity of approximately 5, 900 MWe.

All these reactors use classical uranium oxide fuel assemblies. Two of them (Doel3, Tihange2) have also accepted a limited number of mixed (uranium and plutonium) oxide fuel assemblies.

These mixed fuel assemblies came from the reprocessing of spent uranium oxide fuel assemblies in La Hague (France). The reprocessing of spent fuel gives birth to vitrified high-level waste, and to different isotopes of uranium and plutonium, which can be used in the manufacture of mixed oxide fuel assemblies.

Each country producing radioactive waste must find a solution to dispose them safely. The internationally accepted solution is to dispose high-level radioactive waste in a deep and stable geological layer. This seems to be the most secure and environment-friendly way to get rid off the high-level radioactive waste. One of the few stable geological layers, which could accept radioactive waste in Belgium, is the Boom clay layer. Another possible layer is the Ypresian clay layer, but it is not the reference option for the moment. The Boom clay layer is quite thin (about 100 m thick) and is not at a large depth (about 240 m below the ground surface) at the proposed disposal site, beneath the SCK • CEN Nuclear Research Centre in Mol. A large number of studies have already been performed on the Boom clay layer, and on the possibility of building a high-level radioactive waste repository in this geological medium.

Since 1993, the Belgian government has promulgated a moratorium on the reprocessing of spent uranium oxide fuels in La Hague. Since then, spent fuel assemblies are considered as waste, and ONDRAF/NIRAS (the Belgium Agency for Radioactive Waste and Enriched Fissile Materials) has thus to deal with them as waste. This rises a number of questions on how to deal with this new kind of waste. A solution is to directly dispose these spent fuel assemblies in containers in a repository, just like the other high-level radioactive waste. This repository would be build in the Boom clay layer at a depth of about 240 m beneath the SCK • CEN.

One of the questions raised by this new kind of waste is “could the direct disposal of the spent nuclear fuel assemblies lead to nuclear criticality risks in the future?”. Nuclear criticality is the ability of a system to sustain a nuclear fission chain reaction. This question was not a key issue with vitrified high-level waste because these do not include fissile uranium and plutonium isotopes, which could lead to a criticality event. The spent fuel repository will be designed in order to totally avoid the occurrence of a criticality event at the closure time. But in the future history of the repository, external events could possibly affect this. These events could maybe lead to criticality inside the repository, and this has also to be avoided. This work tries to answer this question, and to determine how to avoid a long-term criticality event inside the repository.

The only complete research work answering this question has been performed in the U.S. for the Yucca Mountain repository but this design is fully different from the Belgian one studied here:

for example, the waste are not only spent fuel waste, and the geological layer is volcanic tuff.

1.2 Organization of the Work

This manuscript is mainly based on an applied research work for ONDRAF/NIRAS on the risks of a criticality in different high-level radioactive waste repositories designs.

The present work is divided in four parts. Part I “Introduction” gives basic knowledge about radioactive waste (Chapter 2), and nuclear criticality risks (Chapter 3). Chapter 2 “Radioactive Waste” gives the definition of radioactive waste, their different origins, and the international and Belgian classifications of radioactive waste. It also describes the management options for the high-level radioactive waste, and the expected characteristics of a high-level radioactive waste repository. Chapter 3 “Nuclear Criticality Risks” defines the notions of fission reaction, multiplication factor (which gives a measure of the potential of criticality event for a given system), and risk (particularly nuclear risk). Nuclear criticality risks in a repository can be studied by two ways:

• Determination of the different ways a criticality event can occur in the future history of the repository (developed in Part II), and subsequent calculation of the multiplication factor in different situations given by the scenarios (developed in Part III).

• Determination of the consequences of postulated criticality events in a repository design (not developed in this work).

Part II “Scenarios Development” concerns the future evolution of geological repositories and the ways a criticality event could occur in such a repository. Scenarios are used to assess the long- term safety of high-level radioactive waste repository designs. The goal here is to assess long-term in-package criticality risks in different repository designs proposed by ONDRAF/NIRAS: the SAFIR2 designs and three supercontainers designs, which were recently proposed. Chapter 4

“Scenarios Development Procedures” is devoted to the different methodologies used to develop scenarios for assessing the future evolution of a high-level radioactive waste repository. Two sys- tematic approaches are detailed with examples. Chapter 5 “Criticality Scenarios Development:

Examples” details the scenarios developed in the U.S. and in Sweden to determine how a criti- cality event could occur in specific repositories. The application of these systematic approaches for the study of in-package criticality scenarios in different Belgian spent nuclear fuel repository designs is developed in Chapter 6 “In-Package Criticality Scenarios Development: SAFIR2”, and Chapter 7 “In-Package Criticality Scenarios Development: Supercontainer Designs”.

Part III “Criticality Calculations” concerns the determination of the multiplication factor in a series of nuclear systems. Chapter 8 “The MCNP Computer Code” describes the Monte Carlo computer code used to perform the calculations. Chapter 9 “Criticality Calculations: SAFIR2”

shows the results of the calculations of the multiplication factor in the two SAFIR2 repository designs. The main goal of this chapter was to compare the results obtained with the MCNP code with calculations performed with another code to check the correct implementation and use of MCNP. Chapter 10 “Criticality Calculations: SC1”, Chapter 11 “Criticality Calculations:

SC2”, and Chapter 12 “Criticality Calculations: SC3” give the results of the calculation of the multiplication factor for different repository designs based on three supercontainers designs in a number of possible future repository evolutions. Different parameters are varied and the multi- plication factor is computed in all these cases. A comparison between the criticality calculations results and the scenarios developed in Part II is also provided for the different supercontainer designs.

Part IV “Conclusions and Perspectives” gives the final conclusions and perspectives of this

research work.

1.3 Main Achievements

The main achievements of this work are:

• A first set of in-package criticality scenarios for different design options for a Belgian spent fuel repository in the Boom clay layer.

• A large number of criticality calculations with different parameters (fuel type, fuel burn- up, fuel enrichment, distance between the fuel assemblies, distance between the fuel rods, water fraction inside the overpack) for the different design options.

• A preliminary study of the effects of the spent fuel assemblies isotopic evolution with time on the multiplication factor.

• For the first time, a coupling between the in-package criticality scenarios and the criticality

calculations has been performed.

Radioactive Waste

2.1 Introduction

The first section of this chapter gives the international definition of radioactive waste. Then, the different possible origins of these radioactive waste are listed, showing that not only the nuclear industry produces radioactive waste. The three categories of the International Atomic Energy Agency classification of radioactive waste are also detailed.

The Belgian situation regarding these waste is then presented. The particular Belgian classifi- cation distinguishes conditioned radioactive waste in groups, categories, classes, and fluxes and can be related to the international classification. The waste of interest in this research work (spent nuclear fuel assemblies from the 7 Belgian nuclear power reactors) are detailed along with a first estimation of their inventory.

The last sections of this chapter are devoted to different high-level radioactive waste manage- ment options. These options are detailed putting the emphasis on the internationally accepted option: the geological disposal, whose aim is to isolate the waste from the environment and the human beings in stable geological formations. An overview of the desired safety properties is presented in an example of a high-level radioactive waste repository design in Sweden.

These notions have been presented in [1].

2.2 Definition

The International Atomic Energy Agency (IAEA) defines radioactive waste as [2]: “For legal and regulatory purposes, radioactive waste may be defined as materials that contain or are contaminated with radionuclides at concentrations or activities greater than clearance levels as established by the regulatory body, and for which no use is foreseen. (It should be recognised that this definition is purely for regulatory purposes, and that material with activity concentra- tions equal or less than clearance levels is radioactive from a physical viewpoint, although the associated radiological hazards are negligible.)”

2.3 Origin

This section is mainly based on a IAEA Safety Series report [2].

2.3.1 Introduction

Many activities involving the use of radionuclides and nuclear power generation result in gen-

eration of radioactive waste. Such activities include all steps of the nuclear fuel cycle (i.e. the

activities associated with the generation of nuclear power) as well as other non-fuel-cycle ac-

tivities. Radioactive waste may also be generated outside the nuclear activities by the (mostly

large scale) processing of raw materials containing naturally occurring radionuclides.

The radioactive waste, which are generated, are as varied in form, activity and type of contam- ination as they are in type of generating activity. They may be solid, liquid or gaseous. Within these groups are a variety of waste types such as trash, spent radioactive sources, pumps, pipes, ion exchange resins, sludges, and Spent Nuclear Fuel (SNF). Activity levels range from ex- tremely high levels associated with spent fuel and residues from fuel reprocessing to very low levels associated with radioisotope applications in laboratories, hospitals, etc. Equally broad is the spectrum of half-lives of the radionuclides contained in the radioactive waste. Which radionuclides are present will depend on the generating process; they may include uranium and other naturally occurring, transuranic and specific man-made radionuclides.

2.3.2 Waste from the Nuclear Fuel Cycle

The nuclear fuel cycle refers to activities associated with the supply of fuel and the management of radioactive materials involved with the production of nuclear power. Although several nuclear fuel cycles are possible, the following description is limited to radioactive waste generated in the uranium fuel cycle. The major steps generating radioactive waste in the uranium fuel cycle are:

• Mining and milling: these waste result from the production of uranium. It contains low concentrations of uranium and is contaminated principally by its daughter products, e.g.

thorium, radium and radon.

• Fuel supply: these waste may result from purification, conversion and enrichment of ura- nium and the fabrication of fuel elements. These include contaminated trapping materials from off-gas systems, lightly contaminated trash, and residues from recycle or recovery operations. These radioactive waste generally contain uranium and, in the case of mixed oxide fuel, also plutonium.

• Reactor operations/power generation: these waste result from treatment of cooling water and storage ponds, equipment decontamination, and routine facility maintenance. Re- actor waste are normally contaminated with fission products and activation products.

Radioactive waste generated from routine operations include contaminated clothing, floor sweepings, paper and concrete. Radioactive waste from treatment of the primary coolant systems and off-gas system include spent resins and filters as well as some contaminated equipment. Radioactive waste may also be generated from replacement of activated core components such as control rods or neutron sources.

• Management of spent fuel: in addition to the radioactive waste described above, reactor operations generate SNF. This material contains uranium, fission products and actinides.

It generates significant heat when freshly removed from the reactor. SNF is either consid- ered as waste or waste are generated from reprocessing operations. Reprocessing operations generate solid and liquid radioactive waste streams. Solid radioactive waste such as fuel element cladding hulls, hardware, and other insoluble residues are generated during fuel dissolution. They may contain activation products, as well as some undissolved fission products, uranium and plutonium. The principal liquid radioactive waste stream, how- ever, is the nitric acid solution which contains both high activity fission products and actinides in high concentrations.

2.3.3 Waste from the Production and Use of Radionuclides

The production and use of radionuclides are not directly related to nuclear power production.

These activities generate smaller quantities of radioactive waste than do fuel cycle activities:

• Research activities: these include a variety of activities and facilities such as research

reactors, accelerators, and laboratory activities. All may generate radioactive waste, with

the type and volume of waste dependent on the research conducted.

• Radioisotope production: the type and volume of radioactive waste produced depend on the radioisotope and its production method. Generally, the volume of radioactive waste generated from these activities is small, but specific activities might be significant.

• Radioisotope applications: the use of radioisotopes may generate small volumes of ra- dioactive waste. The type and volume of radioactive waste produced will depend on the application.

2.3.4 Waste from the Decommissioning of Nuclear Facilities

At the end of the useful life of a nuclear facility, adequate actions have to be taken to retire it from service, finally leading to unrestricted release or use of the site. The activities in decontam- ination and dismantling of a nuclear facility and the cleanup of the site will lead to radioactive waste, which may vary greatly in type, activity, size and volume, and may be activated or con- taminated. These waste may consist of solid materials such as process equipment, construction materials and tools. To reduce the amount of radioactive waste, decontamination of materials is widely applied. Radioactive liquid waste streams may originate from decontamination pro- cesses. Decommissioning waste may contain the radionuclide spectrum which has been used or generated in the respective nuclear facility.

2.3.5 Waste Not Deriving from Nuclear Activities

From several years, it has been recognized that industrial activities other than those mentioned above may also generate waste, which in some cases may be considered as radioactive. This is the case in industrial activities where raw materials containing naturally occurring radionuclides are processed on a large scale, for example the production of artificial fertilizers and the extraction of oil and gas.

In these cases, the natural radionuclides present at mostly low concentrations in the raw material (phosphate ore, oil, gas, etc.), are concentrated during the processing. They are found either in the products or in the different gaseous, liquid or solid waste streams. The concentration of the radionuclides in the waste streams may exceed the levels for exempt waste (see section 2.4).

2.4 International Classification

IAEA classifies the radioactive waste into three categories [2]:

• Exempt Waste (EW) contain so little radioactive material that they cannot be consid- ered as radioactive and might be exempted from nuclear regulatory control. Although radioactive from a physical point of view, these waste may be safely disposed of without specifically considering their radioactive properties.

• Low and Intermediate Level Waste (LILW): radioactive waste in which the concentration of or quantity of radionuclides is above clearance levels established by the regulatory body, but with a radionuclide content and thermal power below those of high-level waste. LILW are often separated into short-lived and long-lived waste:

– LILW Short-Lived (LILW-SL) contain low concentrations of long-lived radionuclides.

Short-lived waste may be disposed of in near surface disposal facilities.

– LILW Long-Lived (LILW-LL) contain long-lived radionuclides in quantities that need a high degree of isolation from the biosphere. Plans call for the disposal of long-lived waste in geological repositories.

• High-Level Waste (HLW) contain large concentrations of both short- and long-lived ra-

dionuclides, so that a high degree of isolation from the biosphere (usually via geological

disposal) is needed to ensure safety. These waste generate significant quantities of heat from radioactive decay, and normally continue to generate heat for several centuries. These waste include:

– The radioactive liquid containing most of the fission products and actinides originally present in SNF and forming the residue from the first solvent extraction cycle in reprocessing and some of the associated waste streams.

– Solidified high-level waste from the radioactive liquid discussed above, and SNF itself if it is declared as waste. SNF can either be reprocessed or be directly considered as waste.

– Any other waste with a comparable activity level.

2.5 Belgian Situation

2.5.1 Classification

The Belgian classification of radioactive waste differs from the conventional international one defined by IAEA. This classification, which only concerns conditioned waste, is based on two

“groups”, divided in “categories”, “classes”, and “fluxes” [3]. Although the Belgian and IAEA classifications differ, they are fully consistent.

2.5.1.1 Groups The two groups are:

• The open group: the radioactive waste that belong to this group do not require geological disposal thanks to their radiological properties. The activity concentrations and the half- live period of the elements forming these waste are small.

• The geological group: this group contains all the waste containing elements that need to be isolated in order to protect population and environment. This isolation is now thought to be achievable by disposal in stable geological formations.

2.5.1.2 Categories

The four categories are defined as follows:

• A: This category of waste belongs to the open group. The activities and the half-lives of the radionuclides in these waste are relatively “small”. A list of 20 radionuclides is the basis of a calculation leading to a criterion. If the sum of the 20 quotients of each radionuclide concentration to a maximum concentration is less than unity, then the “criterion X” is reached and the waste belong to the category A.

• B: This category, which belongs to the geological group, is defined by an exclusion princi- ple. The waste that belong to this category:

– Do not respect the criterion X.

– Do generate low quantities of heat (less than 20 W/m ³ ).

• C: These waste, in the geological group, hold high quantities of alpha and beta emitters and generate significant quantities of heat (more than 20 W/m ³ ). They need to be cooled during a determined period of time.

• R: This category, which belongs to the open group, gathers radium-low-contaminated

waste.

Although the Belgian and IAEA classifications differ, they are fully consistent. Category A corresponds to LILW-SL, category B to LILW-LL, and category C to HLW. Category R has no IAEA equivalent, since the industrial production of radium is a historical feature peculiar to a small number of countries such as Belgium and Canada.

2.5.1.3 Classes

The twenty classes are defined according to the type of interim storage and final disposal options for the conditioned waste.

The codes assigned to the classes are alphanumeric codes with four or more characters:

• The first character (letter) indicates the type of storage. As far as possible, the letter reflects the activity level: L (low), M (medium), H (high), Z (very high), R (radium).

• The next three characters are the letters AGA (Actief Geconditionneerd Afval, which is the Dutch acronym for Active Conditioned Waste), are common to all classes and so have no discriminating function. They have a phonetic function while maintaining a certain likeness with old codes.

• The fifth character indicates the type of disposal. The absence of a fifth character indicates a technically definite surface disposal class. T (temporary) indicates a technically possible surface disposal class (temporarily, the decision to opt for a type of disposal is not taken).

L (long-lived) indicates a class intended for deep disposal.

• The next character, if any, indicates an origin:

– C: waste originating from the reprocessing of SNF in COGEMA-La Hague.

– P: PAMELA, which is the plant that vitrified high-level liquid and solid waste. This plant, which is on the Belgoprocess site at Dessel, is now used solely for cementing high-level solid waste.

– E: EUROBITUM, which is an installation for bituminizing medium-active liquid waste, and is located on the Belgoprocess site at Dessel.

– S: waste which belong for the moment to SYNATOM.

• A sequential digit again if any indicates one or more variants of waste, matrix or package.

The waste intended for deep disposal are classified in 11 classes:

• ZAGALC, ZAGALS: Waste with a very high level of activity and which generate a large quantity of heat come from nuclear fuel used to generate electricity, and are the direct product of that fuel. Depending on whether the SNF is reprocessed (ZAGALC) or not (ZAGALS), very high-level waste will mainly consist of glass enclosing unusable material removed from the fuel, or the spent fuel itself. These waste belong to category C.

• HAGALC2, HAGALP1, HAGALP2, HAGALP3: HLW that generate moderate amounts of heat are also the result of the reprocessing of SNF. Class H waste belong to category C.

• MAGALC, MAGALE, MAGAL: Medium-level waste with low or negligible heat gen-

eration and containing radionuclides with long half-lives are the outcome either of the

processing/conditioning of liquid process effluent and decontamination operations, with

possible preconcentration, or, of the conditioning of solid waste that are highly contami-

nated by alpha and/or beta emitters. The class M waste belong to category B.

• LAGAL: The low-level waste designated for deep disposal have a concentration of ra- dionuclides with long half-lives that exceeds the levels viewed as acceptable for surface disposal. They are also low irradiating, with a contact dose rate that is below, or at maxi- mum of the order of, 5 mSv/h, with the concentrations of activity in beta-gamma emitters having moderate or low values. LAGAL waste belong to category B.

• RAGAL: This class contains radium-bearing waste and belongs to category B waste.

2.5.1.4 Fluxes

More than one hundred fluxes distinguish between the conditioned waste packages according to their physical, chemical, and radiological characteristics.

2.5.2 Description

The waste considered in this study are the Spent Fuel Assemblies (SFAs) originating from the seven Belgian Pressurized Water Reactors (PWRs) nuclear power plants in Doel and in Ti- hange. These waste are labelled ZAGALS (Zeer hoog Actief Geconditionneerd Afval Langlevend Synatom, which is the Dutch acronym for “high-level and long-lived conditioned radioactive waste originating from Synatom”) [3].

2.5.2.1 Uranium ZAGALS Waste

These waste are the SFAs originating from all the Belgian nuclear power plants. These fuel assemblies are made of fuel pellets, which are composed of Uranium OXide (UOX), contained inside Zircalloy fuel rods. These fuel rods are assembled in a square lattice with spacing grids and with top and bottom nozzles. The main characteristics of these fuel assemblies, before irradiation, are given in Table 2.1 (the dimensions are supposed not significantly altered by irradiation) for the seven Belgian nuclear power plants. Figure 2.1 presents a typical 17 × 17 Fuel Assembly (FA).

After irradiation, these SFAs are stored (either in pools or in dry conditions inside canisters) for a still to be determined cooling period (of about 60 years) before being possibly placed in disposal canisters.

Reactor Doel 1 Tihange 1 Doel 3 Doel 4 Doel 2 Tihange 2 Tihange 3

Assembly 14 × 14 15 × 15 17 × 17 17 × 17

Width [mm] 198 214 214 214

Length [mm] 2940 4110 4110 4885

Mass [kg] 385 650 670 780

U mass [kg] 265 431 461 528

Structure mass [kg] 120 219 209 242

Table 2.1: UOX fresh fuel assemblies characteristics [3].

2.5.2.2 MOX ZAGALS Waste

These fuel assemblies are conceptually similar to the UOX ones, except for the composition of

the fuel pellets, which are made of Mixed (uranium and plutonium) OXide (MOX). They have

been used in the Doel 3 and Tihange 2 nuclear power reactors.

Figure 2.1: A typical (Westinghouse) 17 × 17 FA [4].

2.5.2.3 ZAGALS Radionuclides Inventory

Table 2.2 lists the main radionuclides which are present in a typical 1450 days irradiated UOX SFA. This typical UOX SFA has a 45 GWd/tHM (Giga Watt Day per ton Heavy Metal) burn- up, a 4.0 % ²³⁵ U initial enrichment, and has a 1450 days reactor core irradiation period. The radionuclides present in a typical 1450 days irradiated MOX SFA are listed in Table 2.3 . This typical MOX SFA has a 45 GWd/tHM burn up, a 4.93 % ²³⁹ P u + ²⁴¹ P u initial enrichment, and has a 1450 days reactor core irradiation period.

Isotope [Bq] Isotope [Bq]

14 C 8.90E+09 ²³⁵ U 2.60E+08

79 Se 9.40E+09 ²³⁶ U 5.60E+09

90 Sr 4.90E+14 ²³⁷ N p 9.10E+09

93 Zr 4.50E+10 ²³⁸ P u 5.60E+13

99 T c 2.90E+11 ²³⁸ U 5.40E+09

107 P d 2.90E+09 ²³⁹ P u 6.20E+12

126 Sn 1.80E+10 ²⁴⁰ P u 1.00E+13

129 I 7.40E+08 ²⁴¹ Am 9.00E+13

135 Cs 1.00E+10 ²⁴¹ P u 2.70E+14

137 Cs 7.50E+14 ²⁴² P u 5.10E+10

226 Ra 4.90E+05 ²⁴³ Am 6.30E+11

229 T h 3.10E+03 ²⁴⁴ Cm 1.50E+13

230 T h 2.50E+06 ²⁴⁴ P u 1.20E+04

231 P a 8.40E+05 ²⁴⁵ Cm 9.70E+09

232 T h 1.40E+01 ²⁴⁶ Cm 1.70E+09

233 U 1.90E+06 ²⁴⁷ Cm 5.80E+03

234 U 1.00E+10 ²⁴⁸ Cm 1.50E+04

Table 2.2: Typical 1450 days irradiated (45 GWd/tHM) UOX SFA radionuclide inventory [3].

2.5.3 Belgian Inventory

For a short-term approach, one can have a quite precise approximation of the waste inventory [5]. This approximation is based on the known inventory, and the estimates of the electricity producers.

From a long-term point of view, the estimations of the waste inventory are based on a reference scenario, which supposes the complete bringing down of the nuclear installations that are working now in the country. A good prediction also has to take into account that medical and non- nuclear industries will continue to produce radioactive waste. Table 2.4 gives an estimation of the Belgian conditioned radioactive waste inventory, which will be produced until year 2060 (the foreseen final date for the dismantling operations of the existing nuclear installations). For the B and C waste, the lower approximation corresponds to the hypothesis that all the SFAs will be reprocessed, the upper approximation corresponds to the hypothesis where no more SFAs will be reprocessed. The column “destination” gives the actual type of disposal foreseen for the different waste categories.

Currently, and based on the Belgian nuclear electricity generating programme (7 power reactors

operating during 40 years), the assumptions for the amounts of SFAs which have been and which

will be consumed, leads to an estimation of the total number of SFAs to be disposed of, which

is given in Table 2.5.

Isotope [Bq] Isotope [Bq]

14 C 4.30E+09 ²³⁵ U 4.10E+07

79 Se 8.00E+09 ²³⁶ U 3.90E+08

90 Sr 2.50E+14 ²³⁷ N p 7.30E+09

93 Zr 3.30E+10 ²³⁸ P u 2.40E+14

99 T c 2.80E+11 ²³⁸ U 5.20E+09

107 P d 5.90E+09 ²³⁹ P u 2.20E+13

126 Sn 2.50E+10 ²⁴⁰ P u 6.70E+13

129 I 9.20E+08 ²⁴¹ Am 4.50E+14

135 Cs 9.40E+09 ²⁴¹ P u 1.30E+15

137 Cs 7.50E+14 ²⁴² P u 2.40E+11

226 Ra 1.00E+06 ²⁴³ Am 5.20E+12

229 T h 1.80E+04 ²⁴⁴ Cm 2.00E+14

230 T h 1.10E+07 ²⁴⁴ P u 7.80E+05

231 P a 2.20E+05 ²⁴⁵ Cm 3.80E+11

232 T h 9.30E+05 ²⁴⁶ Cm 1.30E+10

233 U 4.40E+10 ²⁴⁷ Cm 4.90E+04

234 U 4.40E+10 ²⁴⁸ Cm 1.10E+05

Table 2.3: Typical 1450 days irradiated (45 GWd/tHM) MOX SFA radionuclide inventory [3].

Categories of Expected Destination nuclear waste volume [m ³ ]

A 60,000 surface or deep disposal

B & C 10,000 to 13,000 deep geological disposal

R 30,000 not yet decided

Table 2.4: Belgian radioactive waste inventory [5].

Reactor Type Quantity

Doel 1-2 UOX 1669

Tihange 1-2, Doel 3 UOX 4780

Tihange 3, Doel 4 UOX 3266

Tihange 2, Doel 3 MOX 144

Table 2.5: Number of SFAs foreseen for the Belgian repository [3].

2.6 HLW Management

2.6.1 Introduction

HLW are not dangerous for ever, because their composing radionuclides decay over time. How- ever, management must take into account long-term safety since some radionuclides take a very long time to decay. This long time decay, which can take several millions of years, focused attention for the first time on the need for some waste to be managed for a very long time beyond the lifetime of those who generated the waste. Long-lived radioactive waste are the first type of dangerous waste for which this need, and the questions they raise about equity between generations, were considered, leading to measures being taken to safeguard future generations.

Final disposal is only the last step in the waste management process, after collection, treatment, conditioning, storage and transport. For HLW, since heat output and radioactivity decrease with time, storage of the waste for some decades before final disposal greatly simplifies the tech- nical requirements for it, and reduces radiological risks.

All parts of the nuclear fuel cycle, from uranium mining and preparing fuel for use, through use of the fuel to generate electricity, to management of SNF and decommissioning, produce some radioactive waste. Radioactive waste include a diversity of materials with different physical, chemical and radioactive characteristics, requiring different types of management.

What to do with SFAs has been and is still an important challenge. To answer this last question, several other ones must be raised [6]:

• What are technically possible options?

• What is the presently economical strategy?

• What long-term alternatives are there?

One can readily think of many ways to manage SFAs and other HLW [6]:

• Dilute and disperse in the air and/or in large bodies of water. Nowadays, this is considered as an unacceptable solution for obvious environmental reasons.

• Delay action. This approach is favored by those who believe that the present political cli- mate makes disposal anywhere impossible and who expect public opinion about radioactive waste to become less negative in the future.

• Store for future use. Many people in the nuclear community feel that as conventional energy sources will be depleted in the 21st century, materials in SFAs will be needed for breeder reactors.

• Reprocess, retrieve, and recycle. Fissionable elements are especially useful, while some of the radioactive fission products have economic value.

• Isolate. This is the current preferred method to prevent radioactivity and radiation from affecting living beings and the environment.

One now explores the various processes involved by the last two points (recycling and isolation of SFAs and/or HLW) [6].

2.6.2 HLW Management Options 2.6.2.1 Reprocessing

SFAs can be reprocessed in specialized factories such as those implanted in La Hague (France)

and Sellafield (United Kingdom). The average composition of the residues from reprocessing

of 1, 000 kg of SFAs are given in Table 2.6. The weight of the reprocessed waste material is

one-tenth of the initial SFAs. The rest is mostly re-usable uranium and plutonium. Of course, since the radioactivity is from fission products, the activity per gram of waste is considerably higher.

The main objection to the reprocessing of SFAs lies in the fact that it might make plutonium more accessible for illegal purposes, including the manufacture of nuclear weapons or its use in terrorism. Another reason for not treating SFAs is the increasing radiation exposure for workers required to operate, maintain, and repair a reprocessing plant. The risk linked to the transport of SFAs and reprocessed materials is also often used as an argument against reprocessing.

Material Weight [kg]

Fission products 28.8

Uranium 4.8

Plutonium 0.04

Neptunium 0.48

Americium 0.14

Curium 0.04

Reprocessing chemicals 68.5

Total 102.8

Table 2.6: Composition of residues from reprocessing of 1,000 kg of SFAs [6].

2.6.2.2 Transmutation

The transmutation process consists in converting elements into other ones by nuclear reaction.

Such a process is possible through neutron bombardment, in an Accelerator Driven System (ADS). Applied to nuclear waste, transmutation would involve irradiation of waste: the neutrons are absorbed to produce new isotopes, which may have a short half-life or be stable. The process aids natural decay by effectively shortening half-lives. A large number of problems are still to be solved before this option could be applied at an industrial level.

2.6.2.3 Unrealistic Options

The disposal options presented in the next paragraphs have been studied and found to be unrealistic.

2.6.2.3.1 Space Disposal This option would seem to be an ideal solution for getting rid of the waste permanently, but there are some drawbacks such as the possibility of an aborted mission, with the spacecraft burning up on re-entry in the atmosphere, leading to a catastrophic atmospheric contamination. The high cost of the vehicle due to the extra shielding weight is also a severe disadvantage.

2.6.2.3.2 Ice-Sheet Disposal Intuition tells us the farther the waste are removed from

the habitations, the safer people will be. One of the most remote site on Earth is the polar

ice cap in Antarctica. Several methods of waste disposing have been proposed for this special

location. For example, containers would be allowed to melt down through the ice by means of

their own decay heat, with melting ice freezing above them as they descend. The containers

would eventually sink to the solid rock base one or two km down. This would appear to be a

very secure disposal, but there is possibly a water layer between the ice and the rock. Containers

would thus be exposed to water, which is connected to the sea. The ice disposal method is also

unacceptable because of the complexities of international ownership of Antarctica, the expense