Geological Repository Systems for Safe Disposal of Spent Nuclear Fuels and Radioactive Waste (Woodhead Publishing Series in Energy)

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Geological repository systems for safe disposal of spent nuclear fuels and radioactive waste

© Woodhead Publishing Limited, 2010

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Woodhead Publishing Series in Energy: Number 9

Geological repository systems for safe disposal of spent nuclear fuels and radioactive waste Edited by Joonhong Ahn and Michael J. Apted

CRC Press Boca Raton Boston New York Washington, DC

WOODHEAD

PUBLISHING LIMITED Oxford Cambridge New Delhi


Published by Woodhead Publishing Limited, Abington Hall, Granta Park, Great Abington, Cambridge CB21 6AH, UK www.woodheadpublishing.com Woodhead Publishing India Private Limited, G-2, Vardaan House, 7/28 Ansari Road, Daryaganj, New Delhi – 110002, India www.woodheadpublishingindia.com Published in North America by CRC Press LLC, 6000 Broken Sound Parkway, NW, Suite 300, Boca Raton, FL 33487, USA First published 2010, Woodhead Publishing Limited and CRC Press LLC # Woodhead Publishing Limited, 2010 The authors have asserted their moral rights. This book contains information obtained from authentic and highly regarded sources. Reprinted material is quoted with permission, and sources are indicated. Reasonable efforts have been made to publish reliable data and information, but the authors and the publishers cannot assume responsibility for the validity of all materials. Neither the authors nor the publishers, nor anyone else associated with this publication, shall be liable for any loss, damage or liability directly or indirectly caused or alleged to be caused by this book. Neither this book nor any part may be reproduced or transmitted in any form or by any means, electronic or mechanical, including photocopying, microfilming and recording, or by any information storage or retrieval system, without permission in writing from Woodhead Publishing Limited. The consent of Woodhead Publishing Limited does not extend to copying for general distribution, for promotion, for creating new works, or for resale. Specific permission must be obtained in writing from Woodhead Publishing Limited for such copying. Trademark notice: Product or corporate names may be trademarks or registered trademarks, and are used only for identification and explanation, without intent to infringe. British Library Cataloguing in Publication Data A catalogue record for this book is available from the British Library. Library of Congress Cataloging in Publication Data A catalog record for this book is available from the Library of Congress. Woodhead Publishing ISBN 978-1-84569-542-2 (book) Woodhead Publishing ISBN 978-1-84569-978-9 (e-book) CRC Press ISBN 978-1-4398-3109-0 CRC Press order number: N10189 The publishers’ policy is to use permanent paper from mills that operate a sustainable forestry policy, and which has been manufactured from pulp which is processed using acid-free and elemental chlorine-free practices. Furthermore, the publishers ensure that the text paper and cover board used have met acceptable environmental accreditation standards. Typeset by Data Standards Limited, Frome, Somerset, UK Printed by TJ International Limited, Padstow, Cornwall, UK


Contents

Contributor contact details

xv

Woodhead Publishing Series in Energy

xx

Preface

xxiii

Part I

Part I Introduction to geological disposal of spent nuclear fuels and radioactive waste

1

Multiple-barrier geological repository design and operation strategies for safe disposal of radioactive materials M. APTED, INTERA Inc., USA; and J. AHN, University of California, Berkeley, USA

3

1.1 1.2 1.3 1.4 1.5 1.6 1.7

Introduction 4 Multiple-barrier geological repository for radioactive materials 5 Basic disposal strategies for radioactive materials 7 Containment of radioactive materials 7 Constraints on concentration of radioactive materials 12 Summary 24 References 25

2

Spent nuclear fuel recycling practices, technologies and impact on geological repository systems 29 M. S. Y. CHU, M. S. Chu & Associates, LLC, USA

2.1 2.2 2.3 2.4 2.5 2.6

Background and introduction Current spent nuclear fuel reprocessing technologies Advanced spent nuclear fuel recycling technologies Impacts of spent nuclear fuel recycling on geological disposal Future trends References


29 32 33 36 41 42

vi

Contents

3

Near-surface, intermediate depth and borehole disposal of low-level and short-lived intermediate-level radioactive waste 43 I. G. CROSSLAND, Crossland Consulting, UK

3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 3.10

Introduction Outline of the sections Safety requirements for near-surface disposal Safety of disposal facilities Styles of near-surface disposal Designing for safety Future trends Sources of further information and advice IAEA requirements for geological disposal References

43 46 46 48 53 66 72 74 75 79

4

Underground research facilities and rock laboratories for the development of geological disposal concepts and repository systems 82 I. BLECHSCHMIDT and S. VOMVORIS, National Cooperative for the Disposal of Radioactive Waste–NAGRA, Switzerland

4.1 4.2 4.3 4.4 4.5 4.6 4.7

Introduction Types of URLs and their role in the staged development of geological repositories Planning and designing an underground research facility: basic considerations Public outreach and the role of URLs as training platforms Case studies Future trends References

82 85 98 101 102 115 117

Part II Geological repository systems: characterisation, site surveying and construction 5

Crystalline geological repository systems: characterisation, site surveying and construction technologies and techniques

121

A. J. HOOPER, Alan Hooper Consulting Ltd, UK 5.1 5.2 5.3 5.4 5.5

Introduction Lithologies Geological structure Rock mechanics and geotechnical properties Hydrogeology


121 123 125 127 128

Contents

vii

5.6 5.7 5.8 5.9 5.10 5.11 5.12 5.13

Geochemistry Radionuclide transport Disturbance by excavation or waste emplacement Stability Feasibility of construction Future trends Sources of further information References

133 138 140 144 146 148 150 150

6

Clay geological repository systems: characterisation and site surveying technologies and techniques 153 J. DELAY, National Radioactive Waste Management Agency, France

6.1 6.2 6.3 6.4 6.5 6.6 6.7 6.8 6.9 6.10 6.11 6.12 6.13 6.14 6.15

Foreword Specific features of a clay site survey Survey tools Survey strategy Technologies Geological mapping Geophysical seismic surveys Drilling Underground structures surveys Core lab analysis Integration of results Future trends Sources of further information Acknowledgements References

153 155 159 162 166 166 167 168 178 178 179 180 181 182 182

7

Assessing the long-term stability of geological environments for safe disposal of radioactive waste

188

K. J. WILSON and K. R. BERRYMAN, GNS Science, New Zealand 7.1 7.2 7.3 7.4 7.5

Introduction 188 Long-term volcano-tectonic stability issues for safe disposal of radioactive waste 189 Geochemical stability issues for safe disposal of radioactive waste 193 Potential climate change issues for safe disposal of radioactive waste 194 Using geological, geophysical and geochemical techniques for quantifying stability for safe disposal of radioactive waste 196


viii

Contents

7.6

Modelling long-term stability for safe disposal of radioactive waste 209 Future trends 210 Summary 212 Sources of further information and advice 213 Acknowledgements 213 References 213

7.7 7.8 7.9 7.10 7.11 8

Far-field process analysis and radionuclide transport modelling in geological repository systems

222

M. MAZUREK, University of Bern, Switzerland 8.1 8.2 8.3 8.4 8.5 8.6

Framework Transport and retardation in argillaceous sedimentary formations Transport and retardation in crystalline-basement environments Quantifying radionuclide transport: two case studies Emerging trends References

222 227 232 238 249 252

Part III Engineered barrier systems for geological repositories: containment materials and technology 9

Immobilisation of spent nuclear fuel and high-level radioactive waste for safe disposal in geological repository systems

261

E. R. VANCE and B. D. BEGG, Australian Nuclear Science and Technology Organisation, Australia 9.1 9.2 9.3 9.4 9.5 9.6 9.7 9.8

Generation of high-level waste from nuclear fuel Historical waste form development for processing Candidate waste forms and disposition schemes Inert matrix fuels Geological disposal Conclusions Acknowledgements References


261 265 271 278 280 283 284 284

Contents 10

Development and application of low-pH concretes for structural purposes in geological repository systems

ix

286

M. C. ALONSO , J. L. GARCIÁ CALVO and A. HIDALGO, Eduardo Torroja Institute for Construction Sciences, Spain; and L. FERNAŃDEZ LUCO, INTECIN–Universidad de Buenos Aires, Argentina 10.1 10.2 10.3 10.4 10.5 10.6 10.7 11

Introduction Functional cementitious material requirements for geological disposal Design and properties of low-pH cements Development and production of low-pH concretes: shotcrete plug application Long-term durability Sources of further information and advice References Development and application of smectitic buffer and backfill materials in geological repository systems

286 290 294 304 311 316 317 323

R. PUSCH, SWECO AB/Geodevelopment International AB, Sweden 11.1 11.2 11.3 11.4 11.5 11.6

Introduction Types, properties and fabrication of the buffer Design and performance of the buffer Types, properties and fabrication of backfill Long-term performance References

323 326 331 341 344 351

12

Near-field processes, evolution and performance assessment in geological repository systems

353

W. ZHOU, Rensselaer Polytechnic Institute, USA; and R. ARTHUR, INTERA Inc., USA 12.1 12.2 12.3 12.4 12.5 12.6 12.7

Introduction Near-field component: engineered barrier system (EBS) Near-field component: host rock Summary description of near-field containment and isolation Overview of near-field process modeling Future trends in near-field analysis References


353 355 359 365 366 374 376

x

Contents

13

Nuclear waste canister materials, corrosion behaviour and long-term performance in geological repository systems

379

F. KING, Integrity Corrosion Consulting Ltd, Canada; and D. W. SHOESMITH, University of Western Ontario, Canada 13.1 13.2 13.3 13.4 13.5 13.6 13.7 13.8 14

Introduction Environmental aspects important for nuclear waste canister performance Selection of nuclear waste canister materials Corrosion behaviour of candidate nuclear waste canister materials Long-term performance of nuclear waste canisters Future trends Sources of further information and advice References

379 380 389 391 410 412 413 414

Post-containment performance of geological repository systems: source-term release and radionuclide migration in the near- and far-field environments 421 Ch. POINSSOT and C. FILLET, CEA, Commissariat a` l’Energie Atomique et aux Energies Alternatives, France; and J.-M. GRAS, JMG Consulting, France

14.1 14.2 14.3 14.4 14.5 14.6 14.7 14.8 14.9 14.10 14.11 14.12 14.13

Introduction Waste form degradation Long-term nuclear glass performance Long-term behaviour of spent nuclear fuel (SNF) Brief overview of low- and intermediate-level waste (L/ILW) performance Conclusion on the relative performance of the different waste forms Radionuclide fate after release Role of aqueous chemical processes in defining the relevant aqueous chemical species Significance of retention processes as a net retardation effect Coupling with transport processes Summary of the behaviour of the main radionuclides Conclusion References


421 423 427 437 452 458 458 460 466 471 477 478 479

Contents

xi

Part IV Performance assessment, expert judgement and knowledge management for geological repository systems 15

Safety assessment for deep geological disposal of high-level radioactive waste in geological repository systems

497

P. N. SWIFT, Sandia National Laboratories, USA 15.1 15.2 15.3 15.4 15.5 15.6 15.7 15.8 15.9

Introduction Goals of a safety assessment Steps in a typical safety assessment Acknowledging uncertainty in a safety assessment Applications of safety assessment Future trends Sources of further information and advice Acknowledgments References

497 499 501 509 513 516 518 518 519

16

Safety assessment for near-surface disposal of low- and intermediate-level radioactive waste 522 M. W. KOZAK, INTERA Inc., USA

16.1 16.2 16.3

16.4 16.5 16.6 16.7 16.8 17

Introduction 522 Definition and performance measures for near-surface disposal of low- and intermediate-level radioactive waste 523 Key issues and development of safety assessment for nearsurface disposal of low- and intermediate-level radioactive waste 525 Safety assessment methodology for near-surface disposal of low- and intermediate-level radioactive waste 528 Application of safety assessment for near-surface disposal of low- and intermediate-level radioactive waste 531 Future trends 539 Sources of further information and advice 543 References 544 Treatment of uncertainty in performance assessments for the geological disposal of radioactive waste

547

J. C. HELTON, Arizona State University, USA; and C. J. SALLABERRY, Sandia National Laboratories, USA 17.1 17.2 17.3

Introduction Conceptual structure of a performance assessment Propagation of uncertainty


547 549 557

xii

Contents

17.4 17.5 17.6 17.7 17.8

Computational design of a performance assessment Sensitivity analysis Concluding discussion Acknowledgments References

565 569 572 574 575

18

Assessment of expert judgments for safety analyses and performance assessment of geological repository systems

580

K. E. JENNI, Insight Decisions, USA; and A. VAN LUIK, US Department of Energy, USA 18.1 18.2 18.3 18.4 18.5 18.6 18.7 18.8 18.9 18.10 19

Introduction 580 Quantifying uncertainties for decision analyses 584 Biases in judgments and the development of formal probability elicitation protocols 587 Assessments with multiple experts 593 Aggregating assessments from multiple experts 598 Degrees of rigor and formality in assessing expert judgments 600 Future trends 602 Sources of further information and advice 605 Acknowledgments 605 References 606 Application of knowledge management systems for safe geological disposal of radioactive waste 610 H. UMEKI, Japan Atomic Energy Agency, Japan; and H. TAKASE, Quintessa K.K, Japan

19.1 19.2 19.3 19.4 19.5 19.6 19.7 19.8 19.9

Introduction Knowledge management: definitions and nomenclature Disposal programme structures and knowledge flows Identification of critical problems and development of solutions Japan Atomic Energy Agency (JAEA) knowledge management system (KMS): the basic concept JAEA KMS: demonstration of application to safety case development Assessment of knowledge engineering and advanced information technology Constructing and visualising safety case arguments for geological disposal of radioactive waste Compiling, synthesising and organising knowledge


610 611 612 614 615 618 624 626 628

Contents 19.10 19.11 19.12 19.13 19.14

Facilitation of communication, multidisciplinary collaboration and efficient use of resources Future trends Sources of further information and advice Acknowledgements References

xiii

632 633 635 636 636

Part V Radiation protection, regulatory methodologies environmental monitoring and social engagement for geological repository systems 20

Radiation protection principles and development of standards for geological repository systems

641

M. JENSEN, Swedish Radiation Safety Authority, Sweden 20.1 20.2 20.3 20.4 20.5 20.6 20.7 20.8 20.9 21

Introduction Understanding safety of geological disposal Dose and/or risk in geological repository systems Probability and risk in geological repository systems Assessment of probability for scenarios Time scales in geological repository systems Optimization and best available technology (BAT) in geological repository systems Future trends References Development of risk-informed, performance-based regulations for geological repository systems

641 642 646 649 653 654 657 659 660 663

T. MCCARTIN and J. KOTRA, US Nuclear Regulatory Commission, USA; and G. WITTMEYER, Center for Nuclear Waste Regulatory Analyses, USA 21.1 21.2 21.3 21.4 21.5 21.6

Introduction Regulatory principles and methodologies for safe geological disposal Development and application of methodologies Future trends Disclaimer References


663 664 668 675 675 676

xiv

Contents

22

Environmental monitoring programs and public engagement for siting and operation of geological repository systems: experience at the Waste Isolation Pilot Plant (WIPP)

678

J. CONCA and T. KIRCHNER, New Mexico State University, USA 22.1 22.2 22.3 22.4

22.5 22.6 22.7 22.8 22.9 22.10 23

Introduction History of salt and site selection of the Waste Isolation Pilot Plant (WIPP) History and current status of CEMRC Survey of factors related to contaminant exposure and perceptions of environmental risks in the region around the Waste Isolation Pilot Plant (WIPP) Internal dosimetry and whole body monitoring of area citizens Air monitoring of geological repository systems Future trends Conclusions Acknowledgments References Methods for social dialogue in the establishment of radioactive waste management programmes

678 682 689

693 696 702 715 716 716 717 719

K. ANDERSSON, Karita Research AB, Sweden 23.1 23.2 23.3 23.4 23.5 23.6 23.7 23.8

Introduction The emergence of participation in nuclear waste management Rationales for participation in nuclear waste management programmes The Swedish dialogue and transparency process Public participation processes in nuclear waste management programmes The context of social dialogue in nuclear waste management programmes Conclusions References Index

719 721 723 725 727 730 735 737 741



(* = main contact)

Chapter 2

Editors

M. S. Y. Chu M. S. Chu & Associates, LLC 1333 Camino Cerrito SE Albuquerque, NM 87123 USA Email: [email protected]

J. Ahn* Department of Nuclear Engineering University of California, Berkeley Berkeley, CA 94720-1730 USA Email: [email protected] M. J. Apted INTERA Incorporated 3900 S. Wadsworth Blvd, Suite 555 Denver, CO 80235 USA Email: [email protected]

Chapter 3 I. G. Crossland Crossland Consulting Nympsfield Gloucestershire GL10 3UB UK Email: i.g.crossland@btinternet. com

Chapter 1 M. J. Apted* INTERA Incorporated 3900 S. Wadsworth Blvd, Suite 555 Denver, CO 80235 USA Email: [email protected] J. Ahn Department of Nuclear Engineering University of California, Berkeley Berkeley, CA 94720-1730 USA Email: [email protected]

Chapter 4 I. Blechschmidt* and S. Vomvoris Department of International Services and Projects National Cooperative for the Disposal of Radioactive Waste– NAGRA Hardstrasse 73 CH-5430 Wettingen Switzerland Email: ingo.blechschmidt@nagra. ch; [email protected]


xvi


Chapter 5

Chapter 9

A. J. Hooper Alan Hooper Consulting Ltd 5 Wickwar Road Kingswood Wotton-under-Edge Gloucestershire GL12 8RF UK Email: [email protected]

E. R. Vance* and B. D. Begg Australian Nuclear Science and Technology Organisation PMB 1 Menai NSW 2234 Australia Email: [email protected]; [email protected]

Chapter 6

Chapter 10

J. Delay Meuse/Haute-Marne Centre Andra, National Radioactive Waste Management Agency RD 960-55290 Bure France Email: [email protected]

Dr M. C. Alonso*, J. L. Garcı´ a Calvo and Dr A. Hidalgo Department of Construction Materials: Physics and Chemistry Eduardo Torroja Institute for Construction Sciences CSIC Serrano Galvache, 4 28033 Madrid Spain Email: [email protected]; [email protected]; [email protected]

Chapter 7 K. J. Wilson* and K. R. Berryman GNS Science PO Box 30368 Lower Hutt New Zealand Email: [email protected]; [email protected]

Chapter 8 M. Mazurek Rock–Water Interaction Institute of Geological Sciences University of Bern Baltzerstr. 3 CH-3012 Bern Switzerland Email: [email protected]

Dr L. Fernańdez Luco Facultad de Ingenierı´ a–INTECIN– Universidad de Buenos Aires Laboratorio de Materiales y Estructuras Av. Las Heras, 2214 1117 Cuidad Autońoma de Buenos Aires Argentina Email: lfernandez@fi.uba.ar

Chapter 11 R. Pusch SWECO AB/Geodevelopment International AB SE-22370 Lund Sweden


Contributor contact details Email: pusch@geodevelopment. ideon.se

Chapter 12 W. Zhou* Rensselaer Polytechnic Institute Department of Mechanical, Aerospace and Nuclear Engineering, 110 8th St Troy, NY 12180 USA Email: [email protected] R. Arthur INTERA Inc. 3900 S. Wadsworth Blvd Suite 555 Denver, CO 80228 USA Email: [email protected]

Chapter 13 F. King* Integrity Corrosion Consulting Ltd Nanaimo British Columbia Canada V9T 1K2 Email: [email protected] D. Shoesmith Chemistry Building University of Western Ontario London Ontario Canada N6A 5B7 Email: [email protected]

Chapter 14 C. Poinssot* Radio Chemistry and Processes Department Nuclear Energy Division

xvii

Commissariat a` l’Energie Atomique et aux Energies Alternatives Marcoule, BP 17171 30207 Bagnols-sur-Ce`ze France Email: [email protected] C. Fillet Nuclear Energy Division Commissariat a` l’Energie Atomique et aux Energies Alternatives Saclay 91191 Gif sur Yvette France Email: catherine.fi[email protected] J.-M. Gras JMG Consulting 77920 Samois-sur-Seine France Email: [email protected]

Chapter 15 P. N. Swift Sandia National Laboratories Mail Stop 1369 Albuquerque, NM 87185-1369 USA Email: [email protected]

Chapter 16 M. W. Kozak INTERA Inc. 3900 South Wadsworth Blvd Suite 555 Denver, CO 87110 USA Email: [email protected]


xviii


Chapter 17 J. C. Helton* Department 1545 Sandia National Laboratories Albuquerque, NM 87185-0748 USA Email: [email protected] C. J. Sallaberry Department 6783 Sandia National Laboratories Albuquerque, NM 87185-1370 USA Email: [email protected]

Chapter 20 Mikael Jensen Swedish Radiation Safety Authority Dept of Radioactive Materials SE-171 16 Stockholm Sweden Email: [email protected]

Chapter 21

Chapter 18 K. E. Jenni* Insight Decisions, LLC 1616 17th Street, Suite 268 Denver, CO 80202 USA Email: kjenni@insightdecisions. com A. van Luik US Department of Energy Office of Civilian Radioactive Waste Management 1551 Hillshire Dr. Las Vegas, NV 89134 USA Email: [email protected]

Chapter 19

T. McCartin* and J. Kotra Mail stop: E2-B2 US Nuclear Regulatory Commission Washington, DC 20555-0001 USA Email: Timothy.McCartin@nrc. gov; [email protected] G. Wittmeyer Center for Nuclear Waste Regulatory Analyses Southwest Research Institute 6220 Culebra Road San Antonio, TX 78238-5166 USA Email: [email protected]

Chapter 22

H. Umeki* Japan Atomic Energy Agency 2-1-8 Uchisaiwaicho, Chiyoda-ku, Tokyo 100-8577 Japan Email: [email protected] H. Takase Quintessa K.K.

A7-707 2-3-1 Minatomirai, Nishi-ku Yokohama 220-6007 Japan Email: [email protected]

J. Conca* and T. Kirchner Carlsbad Environmental Monitoring and Research Center New Mexico State University 1400 University Drive Carlsbad, NM 1-575-706-0214 USA Email: [email protected]



Chapter 23 K. Andersson Karita Research AB Box 6048 SE-187 06 TA¨BY Sweden Email: [email protected]


xix


1

Generating power at high efficiency: Combined cycle technology for sustainable energy production Eric Jeffs

2

Advanced separation techniques for nuclear fuel reprocessing and radioactive waste treatment Edited by Kenneth L. Nash and Gregg J. Lumetta

3

Bioalcohol production: Biochemical conversion of lignocellulosic biomass Edited by K. W. Waldron

4

Understanding and mitigating ageing in nuclear power plants: Materials and operational aspects of plant life management (PLiM) Edited by Philip G. Tipping

5

Advanced power plant materials, design and technology Edited by Dermot Roddy

6

Stand-alone and hybrid wind energy systems: Technology, energy storage and applications Edited by J. K. Kaldellis

7

Biodiesel science and technology: From soil to oil Jan C. J. Bart, Natale Palmeri and Stefano Cavallaro

8

Developments and innovation in carbon dioxide (CO2) capture and storage technology Volume 1: Carbon dioxide (CO2) capture, transport and industrial applications Edited by M. Mercedes Maroto-Valer

9

Geological repository systems for safe disposal of spent nuclear fuels and radioactive waste Edited by Joonhong Ahn and Michael J. Apted

10 Wind energy systems: Optimising design and construction for safe and reliable operation Edited by John D. Sørensen and Jens N. Sørensen



xxi

11 Solid oxide fuel cell technology: Principles, performance and operations Kevin Huang and John Bannister Goodenough 12 Handbook of advanced radioactive waste conditioning technologies Edited by Michael I. Ojovan 13 Nuclear reactor safety systems Edited by Dan Gabriel Cacuci 14 Materials for energy efficiency and thermal comfort in buildings Edited by Matthew R. Hall 15 Handbook of biofuels production: Processes and technology Edited by Rafael Luque, Juan Campelo and James Clark 16 Developments and innovation in carbon dioxide (CO2) capture and storage technology Volume 2: Carbon dioxide (CO2) storage and utilisation Edited by M. Mercedes Maroto-Valer 17 Oxy-fuel combustion for power generation and carbon dioxide (CO2) capture Edited by Ligang Zheng 18 Small and micro combined heat and power (CHP) systems: Advanced design, performance, materials and applications Edited by Robert Beith 19 Hydrocarbon fuel conversion technology: Advanced processes for clean fuel production Edited by M. Rashid Khan 20 Modern gas turbine systems: High efficiency, low emission, fuel flexible power generation Edited by Peter Jansohn 21 Concentrating solar power (CSP) technology: Developments and applications Edited by Keith Lovegrove and Wes Stein 22 Nuclear corrosion science and engineering Edited by Damien Fe´ron 23 Power plant life management and performance improvement Edited by John Oakey 24 Direct-drive wind and marine energy systems Edited by Markus Mueller 25 Advanced membrane science and technology for sustainable energy and environmental applications Edited by Angelo Basile and Suzana Nunes



Preface

Since the mid-20th century, nuclear power generation, radioactive materials utilization in medicine and industry, and nuclear weapons production have produced significant amounts of radioactive wastes at various different levels of radioactivity concentrations. Some radionuclides have half-lives much longer than the use of nuclear and radiation technologies. Therefore, a reliable technological and societal system is necessary for management of these radioactive wastes that assure safety, for both current and future generations. Geological disposal has been selected, developed, and implemented over the past half century by virtually all nuclear countries as the safest and most effective final disposition method for radioactive materials. The goal of geological disposal is simple – to keep hazardous radioactive wastes away from people. Thanks to spontaneous radioactive decay, the radiological hazard of such wastes decreases over time, unlike other types of hazardous wastes produced by modern societies. Thus, if we can successfully isolate the materials of concern for a ‘sufficiently’ long time, then we can achieve the goal of limiting the amount of radiation from radioactive materials to future populations. Achieving this seemingly straightforward goal, however, requires the consideration and integration of diverse technological, regulatory, and social factors. How can important, safety-related questions be collected from concerned stakeholders and addressed? Technologically, how can we find and characterize suitable geological disposal sites and then design repository systems that can limit any eventual radiological releases to acceptably safe low levels? How can we assure that the multiple engineered and natural barriers of such repository systems function as we expect far into the future? More fundamentally, how is long-term safety to be evaluated and independently confirmed? The concept of geological disposal started as a relatively simple measure, but the history of its development shows that the issue of geological disposal has triggered fundamental


xxiv

Preface

questions regarding the principles, standards, and approaches to successful implementation of final disposal concepts. While many countries that have generated nuclear waste are just beginning the process toward geological disposal, several important milestones were achieved in the last decade of the 20th century and the first decade of the 21st century. The Waste Isolation Pilot Plant (WIPP) was commissioned in 1999 in Carlsbad, New Mexico, as a deep geological repository for the disposal of US defense-generated transuranic wastes. Other disposal systems for low-level and intermediate-level wastes have also been implemented in Sweden, Finland, Japan, and elsewhere. License applications for deep disposal of used nuclear fuel and reprocessed highlevel waste are at, or near, the point of submittal in the US, Sweden, and Finland. Significant progress toward final disposal of nuclear wastes is also occurring in France, Switzerland, and Belgium. The last fifty years in waste disposal programs has also seen some setbacks and adjustments in response to scientific, regulatory, and societal concerns. The 1960s Project Salt Vault in Lyons, Kansas in the US was abandoned after local public objections. Early siting attempts for low- and intermediate-level wastes in Switzerland were rejected by popular local votes. In the mid-1990s the initial environmental impact statement for a deep geological repository in Canada for the disposal of used CANDU fuel was found to be technically sound but not socially supported. A public inquiry into potential disposal of intermediate-level waste in the Sellafield area of the United Kingdom also led to the concept being rejected. Decisions on progress toward nuclear waste disposal in Germany have fluctuated during decades of social and political debate centered on nuclear power. In the US, the Yucca Mountain Repository Program submitted a license application to the Nuclear Regulatory Commission in 2008, but at the time of writing the fate of this program remains uncertain. It is notable, however, that following on from these setbacks, national repository programs have learned and applied important lessons on the types of technological, social, and regulatory factors that require open consideration, transparent documentation, and close integration. Successful geological disposal program re-starts are being made in many countries, in part because nuclear wastes from power generation, medical applications, and industrial uses already exist, and have so far been held in temporary storage facilities. There is an acknowledged need for these wastes to be managed and disposed of safely to protect current and future generations. As the issue of long-term safety is universal, active, and mutually beneficial, international discussions and collaborations have played important roles since the 1970s, and have been particularly helpful and effective in developing national geological disposal programs. The results of this collaboration and information sharing include the core part of


Preface

xxv

conceptual design and philosophy, regulatory frameworks, and approaches to safety assessment by nuclear countries. Thus, it is now understood that the geological repository is not just a large-scale civil engineering project that constructs a bunch of tunnels containing radioactive waste canisters, but also a systematic process that is supported by a set of scientific principles, independent regulatory reviews, and methodologies on information exchange and decision-making with the general public in order to be successful. Based on this recent progress and development, the Editors of this book thought that it would be timely to compile a book about geological disposal that reviews the state-of-the-art of the technology, regulatory philosophy, and social-interaction framework that are contributing to sustained progress toward assuring long-term safety of geological disposal. Because geological disposal encompasses a wide range of ‘-ologies’, we asked experts from many diverse fields to contribute chapters; while standard topics such as materials science, design, and geoscience are addressed, we purposefully broadened our topics to include regulatory and societal areas as well. The book consists of five parts. The first part, containing Chapters 1 to 4, is an introduction to geological disposal, and explains design strategy, relation with fuel cycle, and conceptual designs for low-level and high-level waste repositories. The second part, including Chapters 5 to 8, discusses site characterization, site surveying, and construction for various types of host rocks, such as crystalline rock, clay, and salt. The third part, from Chapters 9 to 14, deals with engineered-barrier technologies, ranging from wastesolidification materials to buffer/backfill materials, and modeling of their interaction to evaluate radionuclide release rates from the barriers. The fourth part, from Chapters 15 to 19, covers the performance assessment for the geological disposal system, which is a methodology developed for evaluation of the long-term performance measure to be compared with the regulatory guidelines for safety judgment. The last part, from Chapters 20 to 23, discusses how we can interpret the results of performance assessment and communicate with the public for regulatory decision-making. The Editors wish to thank the authors for their excellent contributions of these chapters, which broadly cover the state-of-the-art in subjects pertinent to geological disposal. The reference list at the end of each chapter will also be useful for the reader who would like to learn more. The Editors also stress that continued international progress is being made worldwide on geological disposal and that interested readers can use the authors and their affiliations as an aid to keep pace with such future developments in this field. In developing this book the Editors have been blessed with assistance from Ian Borthwick, an excellent commissioning editor, and the team at Woodhead Publishing Limited, UK. Without Ian’s patience and stimulating comments and suggestions, this book could not have been realized.


xxvi

Preface

Last but not least, the Editors would like to give special thanks to Professor Thomas H. Pigford and Professor Paul L. Chambre´, both emeriti of the University of California at Berkeley. Their guidance, mentoring, and formal instruction have been a primary source of education and insights that have helped to guide our careers since the 1980s. Joonhong Ahn Michael J. Apted


1 Multiple-barrier geological repository design and operation strategies for safe disposal of radioactive materials M . A P T E D , INTERA Inc., USA; J . A H N , University California, Berkeley, USA

Abstract: This introductory chapter provides basic insights and guiding principles for establishing and evaluating the long-term safe isolation of radioactive wastes in geological repositories. The intended audience is new technical researchers and reviewers, interested in understanding how their specific expertise is integrated into a multi-discipline safety assessment. The focus is on deep geological disposal, appropriate for the disposal of spent nuclear fuel (SNF), reprocessed high-level waste (HLW), and long-lived, intermediate level waste (LL/ILW). Many of the principles discussed here, however, equally apply to near-surface disposal of lower activity wastes. Two basic types of processes affecting the long-term safe containment and isolation of radioactive waste in deep geological repositories are examined; (1) delay-and-decay processes and (2) concentration-attenuation processes. The robustness of different types of isolation processes, based on their effectiveness and reliability, are discussed. A ‘top-down’ safety assessment of an integrated, multiple-barrier repository system is vital in order to identify and prioritize safety-important barriers and processes, and to use such safety-importance insights to guide an efficient and effective research, development, and design program. Key words: containment, geological repository, high-level waste (HLW), mass transfer, multiple barriers, performance assessment, radioactive waste, repository design, safety assessment, safety importance, solubility, sorption, spent nuclear fuel (SNF), transport.

3 © Woodhead Publishing Limited, 2010

4

Geological repository systems for safe disposal

1.1

Introduction

Geological disposal has been the recommended approach for the permanent disposal of radioactive wastes since the seminal US National Academy of Sciences/National Research Council’s 1957 report The Disposal of Radioactive Wastes on Land (NAS, 1957). In the NAS report, disposal of liquid high-level wastes from weapons production was recommended. In this regard, the concept of the geologic disposal at the beginning was quite different from present ones, which are based on solid waste forms. The bases for the contemporary concept and principles of geologic disposal and the method for safety assessment were considered to be established in the 1970s, (e.g. see NEA, 1977, 1991; KBS, 1983). As of today, over 30 nations with nuclear power plants (NPPs) are engaged at various stages in exploring the feasibility, conducting site selection, characterization and design programs, and/or licensing of geological repositories (Witherspoon and Bodvarsson, 2006). There are also programs exploring the possibility of international repositories for a volunteer consortium of nations (e.g. IAEA, 2004). While interesting options involving advanced fuel cycle transmutation, reprocessing, and even extended storage are also being evaluated in some countries (e.g. NEA, 2002), it has become a well-accepted consensus that all of these options eventually require geological disposal as part of the overall solution for safe management and permanent disposal of radioactive wastes (NAS, 2001; Chapman and McCombie, 2003). While confidence has been well established in the engineering community, it is still recognized as a contentious regulatory, social, and scientific issue. Discussions include scientific points, such as long-term safety confirmation, so-called unknown ‘unknowns’ issue (MacFarlane and Ewing, 2006), as well as social points, such as equity between generations and equity between the repository-host community and the rest of the country (NAS, 2003). On the regulatory side, especially in the United States, there have been active discussions involving regulatory agencies, academia, and law makers in order to reach consensus for safety standards for long-term safety of geologic disposal (NAS, 1990, 1995). These discussions have affected, and have been affected by, the site selection processes and development of a repository concept in each country. Often, discussions have included various options of nuclear fuel cycles, hoping that some type of fuel cycles could reduce difficulties significantly. As mentioned in the previous paragraph, however, after three decades of discussion, we reached an understanding that geologic disposal of radioactive wastes is necessary for any type of fuel cycle. The purpose of this introductory chapter is to provide some basic insights and guiding principles for establishing and evaluating the long-term safe


Multiple-barrier geological repository design/operation strategies

5

isolation of radioactive wastes in geological repositories. The focus is on deep geological disposal, appropriate for the disposal of spent nuclear fuel (SNF), reprocessed high-level waste (HLW), and long-lived intermediate level waste (LL/ILW), as advocated by the International Atomic Energy Agency (IAEA, 2009). Many of the principles discussed here, however, equally apply to near-surface, trench-type disposal of lower activity wastes, as discussed in other chapters of this book. The intended audience for this chapter are workers just beginning in the field of geological disposal of radioactive wastes, experts from traditional engineering and scientific disciplines who may be called upon to review multi-disciplinary geological repository programs, and any interested reader with college-level mathematical and technical training. For more experienced workers, there are numerous sources on advanced research, design, and development studies that are being conducted internationally; indeed, the other chapters in this volume provide an excellent window into many of these important, on-going research, development, and deployment (RD&D) areas. By contrast, this chapter’s objective is to provide a short primer on the basis for repository concepts and a simplified context by which to understand better the relevance and safety importance of current RD&D studies.

1.2

Multiple-barrier geological repository for radioactive materials

To understand basic strategies for assuring long-term, safe disposal of radioactive waste, it is first necessary to introduce the concept of a system of multiple barriers. Figure 1.1 shows a representative deep-geological repository design with multiple barriers that include (moving from the inside to the outside): . . . .

a waste form (also called a ‘waste matrix’), a combination of metal canisters (also called ‘container,’ ‘package’, or ‘overpack’), an encompassing buffer and backfill, and the host rock (geosphere) of the repository site.

The main objective of this multiple-barrier system is to secure as long a time as possible for radionuclides contained in the waste form to reach the human environment, so that radiological impacts of disposed wastes would be reduced to an acceptable level. Placing high-level nuclear waste to stable rock formation at > 300 m depth by itself provides important radiationshield, anti-intrusion, and psychological barriers that greatly promote public safety and security compared to temporary surface storage.


6


1.1 Representative illustration of a multiple-barrier deep geological repository system.

When radioactive waste is emplaced within such a conceptual repository design, as shown in Fig. 1.1, two basic periods of analysis regarding the long-term isolation of that nuclear waste can be identified: . .

containment period, during which groundwater is initially prevented from contacting the waste form by the canister, and after failure of the canister due to corrosion; low-release period, during which groundwater can contact the waste form, allowing the release and transport of dissolved radionuclides through the multiple barriers of the repository eventually to reach the accessible surface environment.

Repository concepts are therefore developed around a combination of both engineered and natural barriers that can favorably affect the containment and low-release behavior of a repository. The number, type, and assigned safety functions of these various multiple barriers varies among concepts, depending on factors such as the type of waste form, radionuclide inventory in the waste form, the type of host rock, the geological, hydrological, and geochemical settings, the required regulatory time scale for safety analysis, etc. In order to develop insights into specific repository concepts, it is useful to identify and understand basic disposal strategies and how they are successful.



1.3

7

Basic disposal strategies for radioactive materials

Two basic strategies, or principles, for assuring the long-term safe disposal of nuclear waste by a geological repository can be broadly identified: . .

containment (sometimes called ‘delay and decay’) and constraining concentration (sometimes called ‘dilute and disperse’).

There are, as discussed below, many different ways in which these strategies may be implemented into a repository concept.

1.4

Containment of radioactive materials

A safety-important characteristic of radioactive waste is that radioactive decay acts to reduce the total radiological hazard over time. Each radionuclide has a characteristic half-life (t1/2), in which the initial mass inventory/activity of a radionuclide (A0) will decrease by one-half for every time-period (t) equal to its half-life. Mathematically, this is expressed as AðtÞ ¼ A0 2t=t1=2 . As a heuristic rule-of-thumb, it can be assumed that the initial activity A0 (initial mass inventory) of a given radionuclide may be considered to have decayed to insignificance (reduced by an arbitrary factor of 1024) over a period of 10 half-lives. Therefore, containment of a radionuclide with a halflife of t1/2 anywhere within the multiple-barrier repository system (from the engineered barrier system (EBS) to the geosphere) for a time period of 10t1/2, will effectively eliminate that specific radionuclide from safety considerations.

1.4.1 Canister containment To explore the effectiveness of ‘delay and decay’, the impact of the canister containment can first be considered. For certain repository concepts, the targeted containment time (tc) for a canister is 1000 years (e.g. JNC, 2000), whereas for other repository concepts the targeted containment time for the canister is on the order of 1 000 000 years (e.g. SKB, 2006; US DOE, 2008). Figure 1.2 shows the impact of a 1000-year canister for representative radionuclides occurring in nuclear waste.* The 1000-year canister effectively eliminates short-lived radionuclides with half-lives less than 100 years (i.e. *

For heat-producing waste forms such as SNF and HLW, there is an initial rise in temperature due to radioactive decay of relatively short-lived radionuclides, such as Co-60 (t1/2 = 5.27 years), Sr-90 (t1/2 = 29.1 years), Cs-137 (t1/2 = 30 years), and Am-241 (t1/2 = 432 years). A 1000-year (or greater) containment time effectively allows the dissipation of such radiogenic heating, so the need for analysis of the effects of elevated temperature and elevated temperature gradients on the post-containment release of radionuclides can be obviated.


8


1.2 Impact of canister containment on the reduction and elimination of the initial inventory as a function of the radionuclide half-life. A containment time (tc) of 1000 years is assumed for illustration purposes.

less than 0.1tc), such as Cs-137 and Sr-90, and partially attenuates the initial inventory of Am-241 with a 432-year half-life. If the canister lifetime is extended to 1 000 000 years, radionuclides with half-lives up to 100 000 years can be eliminated. Note, however, that there are numerous radionuclides present in SNF/HLW/ILW with half-lives greater than 100 000 years for which a 1 000 000-year container will have no significant ‘delay and decay’ impact on reducing the initial inventory. In addition to the canister containment time, we can consider additional residence time in the waste form until radionuclides are released by wastematrix dissolution. Furthermore, hardly soluble nuclides such as actinides would form precipitates in the vicinity of the waste-matrix dissolution location. Thus, in geochemical conditions, radionuclides would stay in the vicinity of the waste form for a substantially long time even after the canister failure. This effect, however, should be categorized as the concentration constraint, and is discussed in the following section.

1.4.2 Transport time Another example of ‘delay and decay’ by a repository system is the transport time of radionuclides to migrate through various barriers of the repository system. Following the same formulation as for canister impacts, if the transport time (tt) is equal to or greater than 10 times the half-life of a given



9

radionuclide, then that radionuclide will effectively decay to insignificance during that transport. Diffusion and advection are the two limiting processes for the aqueous transport of dissolved radionuclides through both natural and engineered barriers. The actual transport time of radionuclides is a function of the transport rate and solute–solid interaction processes that act to retard that transport (NEA, 1993). The most commonly cited of these retardation processes is ‘sorption’, for which a sorption coefficient Kd (in units of m3/ kg) can be measured.* Diffusive transport within a repository can be achieved either by selection of a host rock site with extremely low permeability, such as clay or mudstone (e.g. Verstricht and DeBruyn, 2000; Nagra, 2002; Andra, 2005), or the inclusion of an engineered ‘buffer’ barrier (Apted, 1995; JNC, 2000; SKB, 2006). Conversely, groundwater may advectively flow through porous or fractured repository host rock, and formulations for modeling such flow depend on a number of rock-specific and site-specific features, including degree fracturing, transmissivity of fractures, interconnected porosity, regional hydraulic head, hydraulic conductivity, and anisotropy/layering of rock (Domenico and Schwartz, 1998). As an illustration of the impact of transport time, Fig. 1.3 extends Fig. 1.2 and shows as a solid line where transport time (tt) is equal to half-life (t1/2), with the inclined dashed line showing where transport time equals 10 times half-life ðtt ¼ 10t1=2 Þ. Any radionuclides with transport times lying in the upper, left-hand region above this inclined dashed line would effectively decay to insignificance during the transport. As an example of the effectiveness of transport as a ‘containment’ or ‘delay and decay’ barrier, Fig. 1.3 shows as larger, darker-colored circles the calculated diffusive-transport time (tdt) for a representative set of radionuclides, assuming diffusion through a 0.7-m thick (b) buffer with a porosity (ε) of 0.4 and a density (r) of 1800 kg/m3, with an effective (i.e. unretarded) diffusion coefficient (De) of 3.15610–2 m2/s and using representative buffer sorption coefficients (JNC, 2000). A characteristic diffusive-transport time for a sorbing radionuclide can be calculated based on this simple formulation for Fick’s law: b2 ðe þ rKd tdt ¼ De For a non-sorbing radionuclide with Kd = 0, the diffusive-transport time through a 0.7-m thick buffer would be about 6 years, while for a strongly sorbing radionuclide with Kd = 1 m3/ kg, the diffusive-transport time *

‘Sorption’ can entail and combine a number of different types of solute–solid interactions, depending on how it is measured and normalized. See NEA (1993) for a basic review.


10


1.3 Impact of transport time on the reduction and elimination of the initial inventory as a function of the radionuclide half-life. The larger, darker-colored circles are calculated transport times for specific radionuclides diffusing across a 0.7-meter thick buffer. The smaller, lighter-colored cirles are the calculated combined transport times for radionuclides diffusing across the same buffer plus advective transport through 100 meters of rock at a flow rate of 1 meter per year.

through the same thickness buffer would be about 28 000 years. Figure 1.3 shows that diffusive transport through a 0.7-m thick buffer would eliminate many radionuclides with short- to intermediate-length half-lives, although longer-lived radionuclides with half-lives greater than 50 000 years would not experience a significant reduction in their initial inventory. If the diffusion coefficient can be reduced, the transport time can be increased. For example, with the 1-m thick Richard barrier buffer (Apted, 1995; US DOE, 1995), suitable for use in unsaturated host rock, with De = 3.15610–5 m2/ year, the diffusive-transport time for a non-sorbing radionuclide with Kd = 0, would be about 127 000 years, and over 200 million years for a sorbing radionuclide with Kd = 1 m3/ kg. Now consider the additional ‘delay and decay’ impacts attributable to advective transport through a host rock of a repository once radionuclides have been released from the buffer. Figure 1.3 shows as smaller, lightercolored circles the combination of calculated advective-transport time (tat) added to the previously calculated diffusive-transport times through the buffer (tdt) for selected radionuclides. A simplistic advective-transport model is used to calculate transport time, assuming retarded transport of the same set of representative radionuclides through a generic host rock (JNC,



11

2000), further assuming a 100 m path length (L) from the edge of the buffer to some subvertical major water-bearing fracture zone, a rock porosity (ε) of 0.01, a bulk density (r) of 2800 kg/m3 for the rock, and a flow rate (U) 1 m/ year: tat ¼

Lf1 þ ½ð1 eÞ=erKd g U

For a non-sorbing radionuclide with Kd = 0, the advective-transport time through 100 m of rock would be about 10 000 years, while for a strongly sorbing radionuclide with Kd = 1 m3/ kg, the advective-transport time through 100 m of rock would be about 38 000 years. The difference in position between the larger, darker-colored circles and the smaller, lighter colored circles in Fig. 1.3 represents the additional ‘delay and decay’ contribution attributable to retarded advective transport through this generic host rock (JNC, 2000). Note that while there is some contribution toward reducing the inventories of radionuclides with short and intermediate half-lives due to advective transport, it can be readily seen in Fig. 1.3 that even combined diffusive and advective transport cannot, by themselves, lead to effective delay and decay reductions in the inventory of all radionuclides, and especially radionuclides with half-lives greater than about 50 000 years. Of course, for different repository concepts and host rocks, input parameters affecting both diffusive- and advective-transport times can be different, possibly leading to more (or less) significant ‘containment’ capabilities for buffer and host rock. For fractured host rock there are additional processes (e.g. matrix diffusion (Neretnieks, 1980; Ahn, 1988)) that can retard even chemically non-sorbing radionuclides such as I-129 and act to further delay advective transport. For hydrologically saturated, lowpermeability clay and mudstone host rock in which only diffusive transport occurs, the entire thickness of the host rock formation could provide an extremely wide ‘diffusive transport barrier’. For example, the diffusivetransport time through a 100 m thick clay formation for a non-sorbing radionuclide with Kd = 0 would be about 127 000 years, with a diffusivetransport time of over 200 million years for a radionuclide with Kd = 1 m3/ kg.

1.4.3 Additional issues One concern with ‘delay and decay’ approaches is that the decay of activity (mass inventory) for a given radionuclide leads to an increase in the activity (mass inventory) for its decay-daughter radionuclide, and this daughter radionuclide (or other, subsequent, grand-daughter nuclides) may present a greater radiological hazard than the parent radionuclide. For example,


12


ORIGEN-2 calculations of radioactive decay in a 60 000 MWd/MTU pressurized water reactor spent fuel show that the initial activity of Ra226 increases approximately by a factor of 26105 after 1000 years of containment, and by a factor of about 76107 after 100 000 years of containment (Roddy et al., 1986). Recent safety assessments for the disposal of spent fuel in long-lived copper canisters emplaced in fractured granite (SKB, 2006; Posiva, 2007) show that uranium-series daughter radionuclides (Th-230 and Ra-226) and neptunium-series daughter radionuclides (Th-229) can actually become significant contributors to the calculated dose rate for such repository concepts, despite the fact of the trivial initial abundances of these daughter radionuclides in spent fuel (Roddy et al., 1986). There are additional mitigating factors to be considered in evaluating the safety impacts of ‘delay and decay’ barriers and processes. If, for example, the regulatory period for safety assessment is 100 000 years, then a mean canister lifetime of greater than 100 000 years would, by itself, provide assurance of regulatory compliance. Furthermore, there are physical processes, such as matrix diffusion, that can retard and delay the advective migration of even non-sorbing radionuclides such as I-129, as will be discussed in the next section. Also, an extended mean time of canister failure may imply a broadening in the overtemporal distribution of canister failure times, and extended path length of advective transport through the host rock may lead to significant lateral dispersion (hence, dilution) of transported radionuclides. These latter processes, however, can more correctly be placed into the second major strategy for waste isolation, constraints on concentration.

1.5

Constraints on concentration of radioactive materials

As argued in the previous section, containment (‘delay and decay’) strategies, via extended canister lifetimes or extended transport times through repository barriers, can be effective in reducing and even eliminating the initial inventory (hence, any potential release) of radionuclides with short and intermediate half-lives. The effectiveness of a containment-only strategy, however, is more problematic for longer-lived radionuclides. A second, post-containment/low-release period strategy based on additional processes is therefore needed to ensure successfully the safe disposal of nuclear waste containing long-lived radionuclides. During the low-release period, radionuclides will be released as dissolved species into groundwater contacting waste forms, followed by aqueous transport through the set of engineered and natural barriers of a repository system, potentially leading to radiological doses to future humans. Such



13

potential doses will be related to the concentration of dissolved radionuclides that may eventually reach the biosphere. Therefore, this second isolation strategy can be grouped under the collective term of ‘constraints on concentration’, also sometimes referred to as ‘dilute and disperse’. In the context of this overview, ‘constraints on concentration’ refers to any process or barrier that acts to reduce the concentrations of radionuclides as they migrate from one location to another location within the repository system.

1.5.1 Waste-form dissolution and radioelement solubility The most obvious and arguably the most important constraints on concentration arise when groundwater first contacts a waste form at the time of containment failure. Waste-form dissolution is driven by the fact that the groundwater is undersaturated* with respect to the waste-form components. The dissolution of the waste form leads to an increase in the concentration of radioactive and non-radioactive components in the groundwater contacting the waste-form surface, as schematically shown in Fig. 1.4. The dissolution rate (Rd) can be related to the time-dependent ion activity product† of dissolved components, Q(t), and the theoretical equilibrium constant for the dissolving waste form, Keq (Aagaard and Helgeson, 1982; NEA, 1993): QðtÞ Rd ¼ kþ 1 Keq where k+ is the initial dissolution rate of the waste form when Q(t) = 0. Note that both k+ and Rd are typically expressed in units of mass of waste form released per unit surface area of exposed waste form per unit time. In an open system, where dissolved components are immediately transported away from the surface of a dissolving waste form, the dissolution rate remains constant, Rd = k+ (point A in Fig. 1.4). In a closed system, the concentrations (i.e. thermodynamic activities) of dissolved components of the waste form would theoretically increase * It can be somewhat confusing in repository-related literature that both hydrologists and geochemists use the term ‘saturation’, although with completely different meanings. Hydrologically, ‘saturation’ refers to the degree in which open pores of the rock are filled with water, whereas geochemically, ‘saturation’ refers to the necessary concentration of dissolved chemical species to cause the precipitation of a specific solid phase containing those species. † In thermodynamic analyses, chemical reactions are written in terms of effective concentrations or activities of dissolved species (ai), rather than their concentrations (ci). By definition, the extent that activity differs from concentration is expressed by the activity coefficient (γi), such that ai ¼ gi ci . For example, see Aagaard and Helgeson (1982) and Langmuir (1997) for more detailed discussions.


14


1.4 Schematic diagram of the evoluation of concentration of a representative radioelement i at the surface of a dissolving nuclear waste form under expected repository conditions.

ðQðtÞ ? Keq ; as t ? ?Þ until an equilibrium saturation (or ‘solubility limit’) is reached between the contacting water and the waste form (point C in Fig. 1.4). Note that at equilibrium, while the net change in concentration over time is zero, there is actually a dynamic balance in which the forward (dissolution) and reverse (precipitation) rates are equal but non-zero (NEA, 1993). Thus, there are two limiting rate processes potentially imposing constraints on concentration of released radionuclides: . .

rate of dissolution (Rd) of species from the waste-form surface (surfacereaction control) and rate of transport (Rt) of dissolved species away from the waste-form surface (near-field transport control).

Such limiting rate processes have long been recognized and applied in natural geochemical and diagenetic systems (e.g. Berner, 1978), as well as in chemical engineering systems (Chambre´ et al., 1982). However, which of these two processes is the dominant constraint or control on radionuclide concentrations released from waste forms in a repository? Detailed mass-transfer analyses of the relative importance of



15

surface reaction or near-field transport with respect to conditions of repository systems have been conducted (Chambre´ et al., 1982, 1988; KBS, 1983; NAS, 1983). These analyses incorporate the time-dependent dissolution rate of a waste form under repository conditions as a complex function of the waste-package geometry, initial waste-form dissolution rate, diffusion in the buffer*, advection in the host rock, and environmental conditions of the near field. In such analyses, it was found that if the dissolution rate is relatively greater than the diffusive transport rate away from the waste-form surface, the concentration of radioelement i will continue to increase at the wasteform surface until the solubility limit Cs is reached. Thereafter, it will be Cs that constrains the concentration of i at the waste-form surface, even though there may be continued dissolution of the waste form. Conversely, if dissolution is relatively slower than the diffusive transport away from the waste-form surface, then the concentration of a dissolved radioelement at the waste-form surface approaches a value of αCs, where α (in this case < 1) is the factor determined by the mass transfer around the waste form, inventories of radionuclide i and the waste form, and the dissolution rate of the waste form (Chambre´ et al., 1988). For most radioelements in waste forms in most repository conditions, the former case is observed (NAS, 1983; Chambre´ et al., 1988). This means that solubility limits imposed by radioelement-bearing solids will constrain concentration, and the dissolution rate of the waste-form matrix is not a long-term constraint. Figure 1.4 schematically shows the time-dependent evolution of concentration for a representative radionuclide i at the surface of a waste form under repository conditions. Initially at time tA there is a high dissolution rate (Rd = k+) because the contacting water is assumed to be totally undersaturated with respect to radionuclide i, causing a rapid increase in concentration of radionuclide i. At the longer time of tB, the dissolution rate decreases because of the gradual increase in dissolved wasteform components (i.e. QðtÞ ? Keq ; as t ? ?). Eventually, the solubility limit for a phase C containing radionuclide i is reached at time tC, and phase C will precipitate. Thereafter, the solubilitylimit of that phase C will set a fixed boundary condition for the concentration of radionuclide i at the waste-form surface. Note, however, that even after the solid phase that set solubility-limited concentration for radioelement i precipitates, the initial waste form will continue to dissolve. Also note that the time-dependent decrease in the dissolution rate shown in Fig. 1.4 illustrates why it would be excessively and unreasonably * For most deep geological repository concepts (Witherspoon and Bodvarsson, 2006), a lowpermeability buffer is emplaced around the waste canister, to ensure diffusion-limited transport conditions within the engineered barriers system (EBS).


16


conservative to extrapolate short-term (days to years) laboratory dissolution, or ‘leach rate’, data to predict radioelement-release behavior over repository-relevant time scales of hundreds of years or more (NAS, 1983).

1.5.2 Additional waste-form considerations Because radioelement solubility and the waste-form dissolution rate provide such fundamental constraints on radioelement concentrations, it is important to note additional considerations and alternative bounding cases: Metastability In the previous discussion of Fig. 1.4, it was assumed that phase C was the only stable solid phase that could precipitate from solution. Waste forms, such as high-level waste glasses, are thermodynamically unstable in water, and cannot re-precipitate from solution. Indeed, glass is thermodynamically unstable with respect to the potential formation of many different, compositionally related crystalline (mineral) solids that could precipitate when glass dissolves under natural conditions (Dibble and Tiller, 1981). As an illustration, consider the dissolution of borosilicate glass in which numerous polymorphic phases of SiO2(±H2O) could precipitate, each phase with a different solubility limit for the dissolved species H4SiO4. Which of these many silica polymorphs will precipitate to impose a solubility limit for H4SiO4? In natural, low-temperature ( 50 meters) distances to the near-surface environment. Such high-isolation sites typically do not require or greatly rely upon engineered barriers to assure long-term isolation. Other geological formations can be called hydrological flow sites, in which groundwater (hence, dissolved radionuclides) at the repository depth moves through open pores and/or fractures at slow rates over long distances before reaching the near-surface environment (e.g. fractured granite, fractured tuff, sedimentary rock formations). For such hydrological flow sites, the relative roles and contributions of host rock and engineered barriers have been summarized (SKB, 1992): The . . . safety assessment shows that a repository constructed deep down in . . . crystalline basement with engineered barriers possessing long-term stability fulfills the safety requirements proposed by the [regulatory] authorities with ample margin. The safety of such a repository is only slightly dependent on the ability of the surrounding rock to retard and sorb leaking radioactive materials. The primary function of the rock is to provide stable mechanical and chemical conditions over a long time period so that the long-term performance of the engineered barriers is not jeopardized. As in all engineering enterprises, the attributes and isolation capabilities of the man-made and natural components of a repository system must be assessed and appropriately balanced. The design must also be feasible to construct, emplace waste, and monitor safety during the operational period. It is worth noting that mature repository programs in countries such as Sweden, Finland, and the US now entering into a licensing phase have experienced evolution in their design concepts. Such evolution has been an iterative process motivated by a number of factors, including the collection of new site characterization data, revisions to regulatory standards, adaptation of designs to pre-closure and post-closure issues, and continuing technical oversight and regulatory review regarding the likelihood and possible consequences of natural-event scenarios.

1.7

References

Aagaard P and Helgeson H (1982), ‘Thermodynamics and kinetic constraints on reaction rates among minerals and aqueous solutions: I. Theoretical considerations’, American Journal of Science, 282, 237–285. AECL (1994), ‘The disposal of Canada’s nuclear fuel waste: the Vault Model for postclosure assessment’, AECL-10714, COG-93-4, Atomic Energy of Canada


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Limited, Whiteshell Nuclear Research Establishment, Pinawa, Manitoba, Canada. Ahn J (1988), ‘Mass transfer and transport of radionuclides in fractured porous rock’, PhD dissertation, University of California, Berkeley. Ahn J (2008), ‘Effects of repository conditions on environmental impact reduction by recycling’, in 10th OECD/NEA Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation, 6–10 October 2008, Mito, Japan, OECD/NEA. Ahn J and Suzuki A (1993), ‘Diffusion of the 241Am → 237Np decay chain limited by their elemental solubilities in artificial barriers of high-level radioactive waste repositories’, Nuclear Technology, 101(1), 79–91. Ahn J, Kawasaki D and Chambre´. P (2002), ‘Relationship among performance of geological repositories, canister-array configuration, and radionuclide mass in waste’, Nuclear Technology, 140, 94–112. Andra (2005), ‘Synthesis of an evaluation of the feasibility of a geological repository in an Argillaceous Formation’, in Dossier 2005 Argile, National Agency for Radioactive Waste Management, Chaˆtenay-Malabry, France. Apted M (1995), ‘Robust EBS design and source-term analysis for the partially saturated Yucca Mountain Site’, in Proceedings of the 4th Annual International Conference on High-level Radioactive Waste Management, vol. 2, Las Vegas, Nevada, pp. 485–490. Apted M and Engel D (1988), ‘Analysis of congruent matrix release, precipitation and time-dependent containment failure on spent fuel performance,’ in Scientific Basis for Nuclear Waste Management XI, edited by M Apted and R Westerman, Material Research Society, Pittsburgh, Pennsylvania, pp. 303– 312. Berner R (1978), Early Diagenesis: A Theoretical Approach, Princeton University Press, Princeton, New Jersey. Bruno J and Sandino A (1987), ‘Radionuclide co-precipitation’, SKB-TR-87-23, Swedish Spent Fuel and Nuclear Waste Management Company, Stockholm, Sweden. Chambre´ P, Pigford T H, Fujita A, Kanki T, Kobayashi A, Lung H, Ting D, Sato Y and Zavoshy S J (1982), ‘Analytical performance models for geologic repositories’, LBL-14842, Lawrence Berkeley Laboratory, California. Chambre´ P, Kang C, Lee W and Pigford T (1988), ‘The role of chemical reaction in waste-form performance’, in Scientific Basis for Nuclear Waste Management XI, edited by M Apted and R Westerman, Material Research Society, Pittsburgh, Pennsylvania, pp. 285–291. Chapman N and McCombie C (2003), Principles and Standards for the Disposal of Long-lived Radioactive Wastes, Pergamon Press, London, UK. Dibble W and Tiller W (1981), ‘Kinetic model of zeolite paragenesis in tuffaceous sediments’, Clay and Clay Minerals, 29, 323–330. Domenico P and Schwartz F (1998), Physical and Chemical Hydrology, John Wiley & Sons, Inc., New York. EPRI (2006), ‘Program on technology innovation: EPRT Yucca Mountain spent fuel repository evaluation – 2006 Progress Report’, Report 1013445, Electric Power Research Institute, Palo Alto, California. Garisto N and Garisto F (1988), ‘Mass-transport precipitation coupling in finite



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systems’, AECL-9562, Atomic Energy of Canada Limited, Whiteshell Nuclear Research Establishment, Pinawa, Manitoba, Canada. IAEA (2005), ‘Multilateral approaches to the nuclear fuel cycle’, Expert Group Report to the Director General of the International Atomic Energy Agency, International Atomic Energy Agency, Vienna. IAEA (2009), ‘Classification of radioactive waste: general safety guide’, IAEA Safety Series GSG-1, International Atomic Energy Agency, Vienna. JNC (2000), ‘H12: Project to establish the scientific and technical basis for HLW disposal in Japan’, Japan Nuclear Cycle Development Institute, Tokai-Mura, Naka-Gun, Ibaraki, Japan. Johnson L and Tait J (1997), ‘Release of segregated nuclides from spent fuel’, SKBTR-97-18, Swedish Spent Fuel and Nuclear Waste Management Company, Stockholm, Sweden. Kawasaki D, Ahn J, Chambre´ P L and Halsey W G (2004), ‘Congruent release of long-lived radionuclides from multiple canister arrays’, Nuclear Technology, 148, 181–193. KBS (1983), ‘Final storage of spent nuclear fuel: KBS-3, SKBF/KBS Report KBS-3, Swedish Nuclear Fuel Supply Company, Stockholm, Sweden. Langmuir D (1997), Aqueous Environmental Geochemistry, Prentice Hall, Upper Saddle River, New Jersey. Langmuir D and Apted M (1992), ‘Backfill modifications using geochemical principles to optimize high-level nuclear waste isolation’, in Scientific Basis for Nuclear Waste Management XV, edited by C. Sombret, Material Research Society, Pittsburgh, Pennsylvania, pp. 13–24. MacFarlane A M and Ewing R C (eds) (2006), Uncertainty Underground: Yucca Mountain and the Nation’s High-Level Nuclear Waste, ISBN-10: 0-262-13462-4, ISBN-13: 978-0-262-13462-0. Murakami H and Ahn J (2008), ‘Development of compartment models for radionuclide transport in repository region’, in 12th International High-Level Radioactive Waste Management Conference (IHLRWM), 7–11 September 2008, Las Vegas, Nevada, American Nuclear Society. Nagra (1994), ‘Kristallin-I: conclusions from the Regional Investigation Programme for siting a HLW repository in crystalline basement of Northern Switzerland’, Nagra NTB 93-09E, Wettingen, Switzerland. Nagra (2002), Project Opalinus Clay: Safety Report: demonstration of disposal feasibility for spent fuel, vitrified high-level waste and long-lived intermediatelevel waste’, Nagra NTB 02-05, Wettingen, Switzerland. NAS (1957), The Disposal of Radioactive Wastes on Land, US National Academy of Sciences/National Research Council, National Academy Press, Washington, DC. NAS (1983), A Study of the Isolation System for Geologic Disposal of Radioactive Waste, US National Academy of Sciences/National Research Council, National Academy Press, Washington, DC. NAS (1990), Rethinking High-Level Radioactive Waste Disposal, US National Academy of Sciences/National Research Council, National Academy Press, Washington, DC. NAS (1995), Technical Bases for Yucca Mountain Standards, US National Academy


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of Sciences/National Research Council, National Academy Press, Washington, DC. NAS (2001), Disposition of High-Level Waste and Spent Nuclear Fuel, The Continuing Societal and Technical Challenges, US National Academy of Sciences/National Research Council, National Academy Press, Washington, DC. NAS (2003), One Step at a Time: The Staged Development of Geologic Repositories for High-Level Radioactive Waste, US National Academy of Sciences/National Research Council, National Academy Press, Washington, DC. NEA (1977), Objectives, Concepts and Strategies for the Management of Radioactive Waste Arising from Nuclear Power Programmes, Nuclear Energy Agency, Organization for Economic Co-operation and Development, Paris, France. NEA (1991), Disposal of Radioactive Waste: Review of Safety Assessment Methods, Nuclear Energy Agency, Organization for Economic Co-operation and Development, Paris, France. NEA (M Apted, ed.) (1993), The Status of Near-field Modelling: Proceedings of a Technical Workshop – Cadarache, France, 11–13 May 1993, Nuclear Energy Agency, Organization for Economic Co-operation and Development, Paris, France. NEA (2002), Accelerator-Driven Systems (ADS) and Fast Reactors (FR) in Advanced Nuclear Fuel Cycles, A Comparative Study, Nuclear Energy Agency, Organization for Economic Co-operation and Development, Paris, France. Neretnieks I (1980), ‘Diffusion in the rock matrix: an important factor in radionuclide retardation?’, Journal of Geophysical Research, 85, 4379. Posiva (2007), ‘Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto: Summary Report’, Posiva 2007-06, Posiva Oy, Eurajoki, Finland. Roddy J, Claiborne H, Ashline R, Johnson P and Rhyne B (1986), ‘Physical and decay characteristics of commercial LWR spent fuel’, ORNL/TM-9591, Oak Ridge National Laboratory, Oak Ridge, Tennessee. SKB (1992), ‘SKB’91: final disposal of spent nuclear fuel: importance of the bedrock for safety’, SKB-TR-92-20, Swedish Spent Fuel and Nuclear Waste Management Company, Stockholm, Sweden. SKB (2006), ‘Long-term safety for KBS-3 repositories at Forsmark and Laxemar – a first evaluation’, SKB-TR-06-09, Swedish Spent Fuel and Nuclear Waste Management Company, Stockholm, Sweden. Spahiu K, Cui D and Lundstro¨m M (2004), ‘The fate of radiolytic oxidants during spent fuel leaching in the presence of dissolved near-field hydrogen’, Radiochimica Acta, 92, 625–629. US DOE (1995), Total System Performance Assessment. US DOE (2008), License Application. Verstricht J and DeBruyn D (2000), ‘Belgian concept for HLW disposal: development and demonstration’, in Waste Management 2000, Tucson, Arizona. Witherspoon P and Bodvarsson G (2006), ‘Geological challenges in radioactive waste isolation: Fourth Worldwide Review’, LBNL-59808, Lawrence Berkeley National Laboratory, Berkeley, California.


2 Spent nuclear fuel recycling practices, technologies and impact on geological repository systems M . S . Y . C H U , M. S. Chu & Associates, LLC, USA

Abstract: This chapter discusses the potential impacts of spent nuclear fuel recycling on geological repositories. This chapter first provides background information on spent nuclear fuel management practices in the US and internationally, and then discusses the various advanced recycling technologies that are being used or considered and finally the potential impacts of these recycling technologies on the disposal of high-level wastes in geological repositories. Key words: spent nuclear fuel, waste management, recycling technologies, disposal, geologic repository.

2.1

Background and introduction

Essential to recycling of spent nuclear fuel is the separation of fissionable material from the spent nuclear fuel to reuse it in new reactor fuel. In the United States, spent nuclear fuel and high-level radioactive waste have been produced and accumulated since the 1950s. Nuclear fuel reprocessing technology was developed by the US government during the Manhattan project while building atomic bombs. Large-scale reprocessing plants began at nuclear weapons production sites in the Hanford site, Washington, and the Savannah River site in South Carolina. In 1956, the US Atomic Energy Commission (AEC), the predecessor agency of the US Department of Energy (DOE) and the US Nuclear Regulatory Commission (NRC) announced a program to encourage private industry to begin reprocessing spent nuclear fuel. Commercial reprocessing first started at the West Valley plant in New York in 1966. Around the same period of time, several other commercial entities attempted to reprocess waste; however, for a variety of reasons (economical as well as political) all 29 © Woodhead Publishing Limited, 2010

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commercial reprocessing activities were terminated by the early 1970s. In 1977, President Carter announced that the United States would defer commercial reprocessing indefinitely due to proliferation risks associated with separating plutonium from spent nuclear fuel. Although President Reagan lifted this indefinite ban on commercial reprocessing in 1981, no commercial reprocessing facilities have been developed since the 1970s due to economic as well as political reasons. Past reprocessing in the US generated millions of gallons of highly radioactive liquid waste. These liquid high-level wastes have been stored in tanks in the Hanford and Savannah sites, and efforts have been under way to vitrify the waste into glass logs in preparation for the final disposal in a geological repository. Currently, nuclear energy supplies about 20% of the United States’ electricity needs and spent nuclear fuel and high-level wastes are stored temporarily at 121 commercial and government sites in 39 states in the US. A commercial spent nuclear fuel inventory, which includes spent nuclear fuel generated by 104 operating reactors and 14 reactors that are no longer in operation, is currently estimated to be 58 000 metric tons of heavy metals (MTHM). This inventory is estimated to be increasing by approximately 2000 MTHM per year from the current operating reactors. As of January 2007, 47 license extensions have been granted to the existing nuclear reactors for 20-year license extensions. It is estimated that the cumulative inventory of spent nuclear fuel will be 109 300 MTHM at 2040. If all 104 operating reactors received license extensions, then the inventory at 2050 would be about 130 000 MTHM. Recently, the US Nuclear Regulatory Commission (NRC) has indicated that between 2007 and 2010, 23 new license applications (34 total numbers of units) from nuclear industries are expected. With this new wave of potential reactors, the inventory of spent nuclear fuel will grow significantly in the next few decades. The issues of the ultimate disposal of spent nuclear fuel and radioactive waste have been the subject of scientific studies and political debates for many decades around the world. In the US, the Nuclear Waste Policy Act (NWPA) was enacted by Congress in 1982 and establishes a process for the siting, construction, and operation of geological repository for the permanent disposal of spent nuclear fuel and high-level waste. The NWPA also designated the Federal government with the responsibility for the disposal of commercial spent nuclear fuel. Between 1982 and 1987, a suite of potential sites around the country was selected and studied. In 1987, Congress amended the NWPA and selected the Yucca Mountain site, Nevada, as the only site for further study and characterization. Since then, extensive site characterization, experimental and modeling investigations, and engineering designs have been performed at the Yucca Mountain site.


Spent nuclear fuel recycling

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In 2002, Congress passed a joint resolution that approved the DOE to prepare for a license application for Yucca Mountain as the disposal site. On 3 June 2008, the Department of Energy submitted the license application to the NRC, seeking construction authorization for the repository at Yucca Mountain (DOE, 2008a). The NWPA has a provision that sets a statutory limit on the amount of spent nuclear fuel and high-level waste that can be disposed of at Yucca Mountain to be 70 000 MTHM. DOE estimates that with the current and projected inventory of spent nuclear fuel, the limit for the Yucca Mountain repository will be reached by approximately 2010. It is apparent, given the future trend of nuclear energy, that the demand for disposal capacity will far exceed the 70 000 MTHM limit by law. Recently DOE submitted a report to Congress expressing the need to either develop a second repository or to remove the statuary disposal limit of 70 000 MTHM (DOE, 2008b). Although the US has not been reprocessing commercial spent nuclear fuel, the government has been conducting research activities in the recycling of spent nuclear fuel. In particular, DOE initiated the Advanced Fuel Cycle Initiative (AFCI) in 2002 to advance recycling technologies in separating and transmutating long-lived radionuclides to reduce radiotoxicity in spent nuclear fuel. Furthermore, in 2006, as part of President Bush’s Advanced Energy Initiative, DOE launched the Global Nuclear Energy Partnership (GNEP) program (DOE, 2007). The goal of the GNEP is to recycle used nuclear fuel to assure maximum energy recovery, to reduce proliferation concerns, to allow developing countries safely and securely to deploy nuclear power, and to reduce the volume and toxicity of wastes that require deep geological repository disposal. It is a long-standing US national security policy objective to reduce proliferation risks throughout the nuclear fuel cycle through comprehensive efforts to prevent the risks of nuclear weapons materials. An important element of the GNEP program is to create a safe, orderly system to support commercial nuclear power internationally without adding to the dangers of weapons proliferation. As a result, the GNEP emphasizes advanced reprocessing technologies that aim to reduce the proliferation potential associated with the weapons-usable materials inherent in the nuclear fuel cycle, while acquiring nuclear energy economically. In the studies of nuclear fuel cycle options, open or closed, it is recognized that residual nuclear waste will be generated that requires final disposal. The National Research Council in 1957 recommended that deep geological isolation would be a suitable approach for disposal (NRC, 1957). Other nations have adopted the same policy since then. However, no nation yet has a fully functioning geological disposal repository for high-level waste. In the US, the Waste Isolation Pilot Plant in New Mexico has an operating


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geological repository in salt for the disposal of defense transuranic (TRU) waste.

2.2

Current spent nuclear fuel reprocessing technologies

The PUREX (plutonium–uranium extraction) process has been the reprocessing technology for the past 40 years. It was first developed in the US for separating pure plutonium for military purposes. It is a solvent extraction process using the extractant tributylphosphate (TBP) mixed in a largely inert hydrocarbon solvent. Since the opening of the first PUREX plant at the Savannah River site in 1954, the PUREX process has been utilized in a variety of flowsheets and is still being used in all commercial reprocessing plants currently operating in the world. This technology has been improved rapidly to adapt continuously to the evolving characteristics of spent nuclear fuel and other constraints such as regulatory requirements. For example, national and international regulations are more stringent regarding safety and security, and waste released and public dose exposure limits have been lowered. Commercial PUREX reprocessing in general consists of four main technological operations: fuel handling and shearing, fuel dissolution, materials separation and purification, and waste treatment and conditioning (IAEA, 2005). At present PUREX reprocessing produces two types of waste: . .

waste from the process itself in the form of a liquid solution of fission products and actinides, and waste comprising hulls and end fittings from the structure of the fuel, insoluble fission products, and other operational residues.

The first type of liquid waste is usually concentrated and then vitrified, and is classified as high-level waste ready to be disposed of in a deep geological repository. The second type of waste can be compacted into containers and usually stored for further disposition. Several countries currently have operating reprocessing plants. The La Hague reprocessing plants UP2 and UP3 in France have been reprocessing spent nuclear fuel for decades using the PUREX technology. La Hague has a capacity of 1700 MTFM per year. The Thermal Oxide Reprocessing Plant (THORP) at Sellafield in the United Kingdom uses the PUREX technology with a capacity of 1200 MTHM. Japan has a small reprocessing plant at Tokai-mura and is beginning operation of a reprocessing plant at Rokkasho-mura with a capacity of 800 MTHM per year. Russia has a 400 MTHM/year commercial reprocessing plant at Mayak. India has been



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reprocessing spent nuclear fuel since the 1960s on smaller research scales. China has a reprocessing demonstration plant and is beginning to design a commercial reprocessing plant with a capacity of 800 MTHM/year in the near future. In France and the United Kingdom, improvement in the operations of the PUREX process continues to reduce the amount of waste generated. Improved sorting procedures and increased package concentrations have allowed the volume of waste generated to be reduced. However, volatile radionuclides from the processes are removed by caustic scrubbing and then released to the sea. This practice is not permitted in the US.

2.3

Advanced spent nuclear fuel recycling technologies

In the early years of reprocessing in the US, the goal was to separate pure plutonium for use in nuclear weapons. In recent years, as interests in commercial nuclear energy grew, the focus has shifted to an emphasis on non-proliferation, decreasing losses of fissile and fertile elements to waste, and optimizing waste production for final disposal. The main issues with the PUREX process are twofold. First, a pure plutonium stream is separated and increases the proliferation risk. Second, minor actinides and heat-generating radionuclides go into the waste stream, increasing the radiotoxicity and volume of waste, as well as subutilizing the emplacement of waste in a geological repository because of the thermal effects. Since 2002, the Advanced Fuel Cycle Initiative (AFCI) from the US DOE focuses on the R&D of fuel cycles that could have substantial environmental, non-proliferation, and economic advantages over the once-through fuel cycle. Specifically, the AFCI focuses on (1) separation technologies for spent nuclear fuel, (2) transmutation technologies for minor actinides and long-lived fission products from spent nuclear fuel, and (3) development of advanced proliferation resistant reactor fuels that will consume plutonium. The goals of the R&D are (1) to remove over 90% of the uranium, (2) to remove over 99% of cesium and strontium, and (3) to separate the transuranic elements (plutonium, neptunium, americium, and curium) for recycling. Although there are several emerging R&D technologies being investigated internationally (IAEA, 2008), there are two key technology areas, as discussed below.


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2.3.1 Advanced aqueous technologies A new generation of fuel reprocessing technologies is now in development under the DOE programs. In particular, a family of liquid reprocessing technologies derived from PUREX is being investigated. This family is called UREX+ technologies. UREX consists of a suite of processes where specific groups of radionuclides are removed to tailor products and compositions of the desired products and waste streams. All of these processes involve an initial extraction of uranium and technitium from the spent fuel dissolver solution. Other extraction steps are then added to meet the various separation requirements. These UREX+ processes use four main extraction steps to separate U, Tc, Cs/Sr, FP, and Pu/MA into six separate fractions. Additional extractions could also be used to partition Pu/ Np, Am, and Cm. In the UREX+ processes, U is denitrated from the aqueous stream and stored as an oxide powder for reuse. Tc is incorporated in a metallic alloy, Sr/Cs can be separated but may contain dilute nitric acid or some organic chemicals, lanthanides are separated and stabilized, and Pu and minor actinides (Np, Am, and Cm) are separated as a group for reuse. Many of the UREX+ processes under investigation are still at a laboratory scale and have not been optimized with respect to minimizing the number of separation cycles or optimizing separation efficiencies. Potential product and waste/storage materials are summarized below (see Figure 2.1). 1. 2. 3. 4. 5. 6.

Uranium: storage as U3O8. Cs/Sr aluminosilicate for storage. Tc in metallic alloy for disposal. Other fission products in vitrified high-level waste for disposal. Cladding hulls are compressed and disposed as HLW. Tritium, collected as tritiated water, may be grouted or encapsulated for disposal. 7. Xenon and krypton gases may be immobilized in zeolite and disposed as HLW. 8. Carbon-14 is recovered as CO2, converted to carbonate, and disposed as HLW. 9. Transuranics may be recycled for fuel fabrication. 10. Iodine may be trapped in silver-coated zeolite, converted to potassium iodate, and disposed as HLW. Continuing work is still under way to optimize the detailed flowsheets and steps.



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2.1 UREX+ process and waste/storage products.

2.3.2 Advanced pyroprocessing technology Pyroprocessing is a non-aqueous process conducted in molten chloride salts that make use of electrochemical dissolution, selective reduction, and adsorption to partition groups of elements. Pyroprocessing is currently not being used commercially, but has been the subject of much R&D. Pyroprocessing technology was originally proposed by the Argonne National Laboratory as a process with the potential to treat all DOE spent nuclear fuel. Later it was used as a potential treatment method for sodiumbonded fuel. As applied to SNF reprocessing, a demonstration project has been studied at the Idaho National Laboratory for reprocessing of Experimental Breeder Reactor II (EBR-II) since 1996. In this demonstration project, chopped fuel rod elements are placed in a steel anode basket in an electrorefiner that contains a KCl–LiCl molten salt eutectic system at upwards of 500 8C. Upon passage of a constant electrolysis current between the anode basket and the steel cathode, U, Pu, transuranic elements (TRU), the alkalis and alkaline earth metals, and rare earths are oxidized into the molten salt. The stainless steel from the cladding, most of the Zr, and the noble metals remain in the anode baskets. The oxidized uranium (U+3) is reduced to the metal and deposited on to the cathode. After a given period of time of electrolysis, the U cathode is removed and uranium is cast into


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metal ingot for future reuse or storage. The salt containing Pu and TRU elements, alkalis and alkaline earths, and some fission products is removed from the electrorefiner, mixed with zeolite, and then mixed with glass and hot pressed into a glass-bonded ceramic waste form. The material remaining in the anode basket is volatilized from the molten salt and cast into metal ingots as the metal waste form. An NAS committee evaluated this process and found no technical barriers in the reprocessing of EBR-II SNF using the pyroprocessing process (NRC, 2000). Specifically, the committee concluded that: . .

The physical and mechanical behavior of the ceramic waste form under repository conditions should be comparable to that of borosilicate glass. The corrosion rate of the metal waste form appears to be low and similar to the stainless steel and C22 alloy to be used in Yucca Mountain.

Pyroprocessing is most suitable for metallic spent fuel, but can also process oxide fuel after a reduction step. In an electrorefiner containing molten electrolytic salt, the anode will contain metals such as Tc, zirconium, iron, and molybdenum. One cathode will contain most of the uranium as metal, while the other cathode contains rare earth fission products plus TRU elements. Fission products such as Sr and Cs will remain in the molten salt. The uranium metal is converted to an appropriate form for reuse, and TRU metal is also reused for fuel fabrication. The metal left at the anode is heated in a metal waste furnace to produce a solid metallic waste form for final disposal. The fission-product-laden salt is circulated through a zeolite ion exchange bed and the zeolite matrix is consolidated into a monolithic ceramic waste form for disposal (Fig. 2.2). R&D is underway at the Idaho National Laboratory to optimize the pyroprocessing technology. Key R&D areas include separation of Sr/Cs from the salt so a separate waste stream can be produced, process modeling, and better actinide recovery. The Korea Atomic Research Institute is also conducting R&D on pyroprocessing technology.

2.4

Impacts of spent nuclear fuel recycling on geological disposal

The Yucca Mountain repository is used in this paper as an example for analyzing the potential impacts of recycling, since Yucca Mountain is a wellstudied and fully analyzed site. Current inventory destined for disposal at Yucca Mountain has a legislative limit of 70 000 MTHM. This inventory consists of 63 000 MTHM of spent nuclear fuel from commercial sites, and the remaining 70 000 MTHM would consist of about 2333 MTHM of DOE



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2.2 Pyroprocessing flow chart.

spent fuel and high-level waste. Between 11 000 and 17 000 waste packages will be emplaced in the repository. The DOE submitted the License Application (LA) to the Nuclear Regulatory Commission in June of 2008. In the LA, a Total Systems Performance Assessment (TSPA) was used to calculate potential releases of radionuclides from the repository under all plausible scenarios for up to a million years after repository closure. The TSPA assesses the capabilities of various barriers at the Yucca Mountain site. The performance of the repository is controlled by the natural and engineered features of the site, which act in concert to prevent or reduce the movement of radionuclides to the accessible environment. Three barriers are important to the waste isolation at Yucca Mountain: the upper natural barrier, the engineered barrier system, and the lower natural system. Details of calculations of infiltration of water into the mountain under various climatic and hydrological conditions are used to develop models for the seepage of groundwater into emplacement drifts underground. These models are incorporated into the TSPA model. Models of the engineered barrier system use these seepage rates along with the chemical and thermal conditions at the repository to calculate the degradation rates of the various engineered systems, i.e. drip shield, waste packages, and the


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waste forms. The release rates of radionuclides from the waste forms and waste packages into the unsaturated zone below the repository are then estimated. The flow of water movement within the unsaturated zone is then used to calculate the transport of radionuclides to the saturated zone 300 meters below the repository. Radionuclide concentrations at wells 20 km from the repository are then used as potential doses to individuals. Recently, the EPA finalized the environmental protection standards for Yucca Mountain (EPA 40 CFR 197, 2008). In this standard, compliance will be judged against a standard of 15 mrem/year dose at times up to 1000 years after disposal and against a standard of 100 mrem/year dose at times after 1000 years and up to 1 million years after disposal. The results of the TSPA in the LA shows that the radionuclides that dominate the annual doses typically have a combination of unique characteristics such as (1) large initial inventory in the waste, (2) moderate to high solubility, (3) long half-life (e.g. ≥ 105 years), and (4) low sorption in the transport paths. The radionuclides that are important to dose also depend on the time frame considered (i.e. 10 000 or 1 000 000 years) because of the effect of radionuclide decay and the effect of retardation from sorption along the flow path. Ingrowth of radionuclide through the decay chain can also be an important process that determines the role and importance of radionuclide in the actinium, uranium, neptunium, and thorium decay chains. In the License Application, results of compliance calculations show that the mean annual dose to an individual is 0.24 mrem for the first 10 000 years after disposal and the medium annual dose to an individual is 0.9 mrem for a time period up to one million years after disposal (peak dose at 720 000 years). The LA also shows that the main contributors to the mean annual dose for 10 000 years after closure are (ranked from highest to lowest): Tc99, C-14, Pu-239, I-129, Cl-36, Pu-230, Se-79, and Np-237. The single largest contributor is Tc-99, which accounts for 51% of the maximum mean dose. For the time period between 10 000 years and one million years, LA shows the main contributors at the peak mean annual dose are (from highest to lowest): Pu-242, Np-237, Ra-226, and I-129. These four radionuclides account for about 77% of the total mean dose. Although Ra-226 has a relatively short half-life (1600 years), it is sustained through chain decay of the longer-lived radionuclides of Th-230 (half-life of 7.546104 years) and U-234 (half-life of 2.466105 years). An important strategy for the Yucca Mountain repository concept is to manage temperatures within and between emplacement drifts to allow water always to drain freely in the rock between the emplacement drifts. At the Yucca Mountain design, there are severe temperature constraints at various locations underground. For example, the rock temperature midway between the drifts must always remain below 96 8C. This constraint ensures that any



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water flowing downward through the mountain will flow through the rocks, preventing the retention of large volumes of water above the repository that potentially could flood the repository as it cools. A temperature limit of 200 8C is imposed on rock surrounding the repository to prevent alteration of its crystalline structure. For the current inventory destined for Yucca Mountain, the temperature limits are met by setting the emplacement drifts 81 meters apart while specifying a linear heat load of 1.45 kW per meter and a maximum thermal load of 11.8 kW per waste package (DOE, 2008a). These constraints result in a repository footprint of 1150 acres for the disposal of 70 000 MTHM of waste. These thermal requirements are intimately linked to the characteristics of the waste inventory. The Yucca Mountain inventory consists of 90% commercial spent nuclear fuels and 10% government waste (vitrified HLW and DOE spent nuclear fuels). The decay heat characteristics are thus largely controlled by commercial spent nuclear fuels and the thermal requirements are mostly derived from commercial spent nuclear fuel characteristics. In general, in the first three hundred years after disposal, the heat is dominated by Sr-90 and Cs-137, while later in time it is dominated by long-lived isotopes of actinide elements plutonium and americium, specifically Pu-241 and its decay daughter Am-241. As spent nuclear fuel is recycled, there are three major changes to the inventory of residual wastes that may need disposal: 1. 2. 3.

Radionuclide composition and characteristics. Heat generation characteristics. Physical and chemical characteristics of waste forms.

These changes in inventory will result in profound changes to the performance of the repository in terms of both potential releases of radionuclides and the design of waste emplacement in the repository. Wigeland et al. (2006) performed scoping analyses on the potential impacts of separation and transmutation on the utilization (emplacement of wastes) of Yucca Mountain. They used thermal performance at Yucca Mountain to establish separation and transmutation criteria for commercial spent nuclear fuel. In this study, Wigeland used models that are simpler than those used by the Yucca Mountain project. Wigeland used thermal models that are decoupled from hydrological effects by assuming key hydrologic conditions (such as rock saturation states and surface water infiltration rates) as input. Results from the simplified models were compared with the more complicated coupled models and good agreements were found. It was found that if 99.9 % of the plutonium and americium are removed from spent PWR fuel, it is possible to increase the linear loading in the drift from the reference value of 1.1 MTHM/m to about 5.9 MTHM/m. This loading increase can be interpreted as an increase in area loading of the repository


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by a factor of 5.4. If the separation efficiency is lowered to 99 and 90%, the drift loading can be increased to factors of 5.3 and 4.3 respectively. Once plutonium and americium are mostly removed, the decay heat is due almost entirely to Cs and Sr. It is shown by Wigeland that if Cs and Sr can be removed at 99.9% efficiency after 99.9% of the plutonium and americium are removed, the drift loading can be further increased to 47.0 MTHM/m, a factor 42.7 higher than the reference case. Wigeland and Morris (2006) also studied the impact of reprocessing of SNF on potential peak dose rates from Yucca Mountain. A DOE simplified model was used to calculate the transport of radionuclides as they are released from waste packages. An analysis was performed to see the impact on peak dose if the entire SNF inventory was put in a glass waste form with no change in radionuclide contents. It is shown that the waste form change has a large impact on dose, where the peak dose is reduced to 50% of the reference case due to the reduced degradation rate of the glass waste form. It is shown by Wigeland and Morris that if 99.9% of the uranium is removed from the inventory, the peak dose is reduced to about 85% of the reference case. On other hand, if both plutonium and uranium can be removed from SNF at 99.9% efficiency, the peak dose rate can be reduced by a factor of 5.0. Furthermore, if all actinides can be removed at an efficiency rate of 99.9%, the peak dose rate can be reduced by a factor of about 90. The peak dose rate is then dominated by Tc-99 and I-129. It is concluded by Wigeland that it is possible to dispose of reprocessed waste from 7 000 000 MTHM of PWR spent fuel and have essentially the same peak dose rate as for 70 000 MTHM of PWR spent fuel. For this reprocessed inventory, the repository temperature would be slightly lower than the reference case. Overall, the removal of actinides at a separation efficiency of between 99 and 99.9% would offer a significant range of benefit for the utilization of a geological repository such as Yucca Mountain. Ahn (2007) analyzed the impacts of removal of actinides from commercial spent nuclear fuel on the environmental impact of Yucca Mountain. In this study, the indicator for environmental impact is measured by the ‘radiotoxicity’ of a radionuclide, which is defined as the volume of water needed to dilute the radionuclide to the permissible concentration for drinking. Simplifying assumptions are used in this study, such as all waste packages are assumed to fail at the same time and radionuclides are assumed to be released either congruently or with solubility limited. With this, the study shows that with a reduction of the TRU nuclide inventory by a factor of 100, the environmental impact (i.e. radiotoxicity) would be reduced by the same factor. The analyses by Wigeland and Ahn are based on generic reprocessing and the waste form was assumed to be similar to vitrified glass. As the advanced reprocessing technologies are being developed, more information on waste



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forms will be available for further analysis. For pyroprocessing technology, two waste forms are generated: the metal waste form and the ceramic waste form. The ceramic waste form is expected to be very similar to borosilicate glass; however, the characteristics of the metal waste form may be quite different from the glass waste form. One interesting aspect of the metal waste form is that Tc-99 will be incorporated into the metal waste form from pyroprocessing, and technetium will be in a reduced form. This reduced form of Tc-99 may have a significantly lower release rate than the reference case, potentially reducing the peak dose from the repository. The metal waste form has undergone extensive qualification examination and testing for acceptance to the Yucca Mountain repository (Frank et al., 2007). The metal waste form being tested is stainless steel and zirconium alloys. The empirical model that predicts the overall corrosion behavior of the metal waste form shows that uniform aqueous corrosion is the main degradation mechanism for radionuclide release. The study also shows that the metallic waste form can be very robust in the retention of Tc and other fission materials.

2.5

Future trends

The current Nuclear Waste Policy Act (NWPA) and the NRC regulations contain rules that govern the management and disposal of all classes of nuclear waste, both high-level and low-level waste. These rules were derived from waste generated from activities pertinent at the time, and are based on the ‘origin of generation of waste’. With new types of waste generated from potential reprocessing in the US, new approaches to managing nuclear waste may be considered. Gombert (2008) laid out an integrated waste management strategy for the Global Nuclear Energy Partnership program. In this strategy, logical risk-based considerations were used for potential disposition paths for the various waste generated by both UREX and pyroprocessing reprocessing technologies. Gombert found that all wastes generated by processes being developed under GNEP and AFCI programs can be disposed of under the current regulations; however, significant efficiencies can be realized if the separated wastes can be dispositioned based on their characteristics and risks rather than their origins. A recent decision was made by the incoming new US administration of Obama to halt or delay the development of the Yucca Mountain repository. This decision could have very significant impact on the whole nuclear energy policy. Current law does not provide an alternative repository site to Yucca Mountain and it does not authorize the DOE to open temporary centralized storage facilities without a permanent repository in operation. Without congressional action, therefore, the default alternative to Yucca Mountain would be indefinite on-site storage of spent nuclear fuel at reactor sites. The


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new administration also indicates that it is in favor of a closed fuel cycle policy and intends to convene a study to address reprocessing and the alternatives of nuclear waste issues. A National Academy of Sciences study of reprocessing technologies found that none of the separation and transmutation technologies would eliminate the need of a geological repository (NRC, 1995). With further R&D, however, if uranium, plutonium, and other long-lived radionuclides can be recycled back into a reactor in the future, the residual waste could contain only shorter-lived fission products and would be quite benign in terms of radiotoxicity. If this were the case, performance requirements for the final disposal at a repository could be much easier to achieve as the longterm uncertainties would be much eliminated. At the same time, alternative disposal options – e.g. alternative disposal sites – may be considered.

2.6

References

Ahn J (2007), ‘Deterministic assessment of environmental impact of Yucca Mountain Repository measured by radiotoxicity’, Journal of Nuclear Science and Technology, 44, 431. DOE (Department of Energy) (2007), ‘Global Nuclear Energy Partnership Strategic Plan’, GNEP-167312. DOE (Department of Energy) (2008a), ‘Yucca Mountain Repository License Application’, DOE/RW-0573. DOE (Department of Energy) (2008b), ‘The Report to the President and the Congress by the Secretary of Energy on the Need for a Second Repository’. Frank S M, Kaiser D D and Marsden K C (2007), ‘Immobilization of technetium in a metallic waste form’, Global. Gombert D (2008), ‘Global Nuclear Energy Partnership Integrated Waste Management Strategy’, GNEP-WAST-WAST-AI-RT-2008-000214, Department of Energy. IAEA (International Atomic Energy Agency) (2005), ‘Status and trends in spent fuel reprocessing’, IAEA-TECDOC-1467. IAEA (International Atomic Energy Agency) (2008), ‘Spent fuel reprocessing options’, IAEA-TECDOC-1587. NRC (National Research Council) (1957), The Disposal of Radioactive Waste on Land, National Academy Press, Washington, DC. NRC (National Research Council) (1995), Nuclear Waste: Technologies for Separations and Transmutation, National Academy Press, Washington, DC. NRC (National Research Council) (2000), Electrometallurgical Techniques for DOE Spent Fuel Treatment, National Academy Press, Washington, DC. Wigeland R A and Morris E E (2006), ‘Processing requirements for PWR spent fuel to reduce estimated peak dose rate associated with potential releases from a geologic repository’, Argonne National Laboratory, ANL-AFCL-166. Wigeland R A, Bauer T H, Fanning T H and Morris E E (2006), ‘Separations and transmutation criteria to improve utilization of a geological repository’, Nuclear Technology, 154 (April), 95–106.


3 Near-surface, intermediate depth and borehole disposal of low-level and short-lived intermediate-level radioactive waste I . G . C R O S S L A N D , Crossland Consulting, UK

Abstract: This chapter is concerned with the permanent disposal of radioactive waste in near-surface, intermediate-depth and borehole disposal facilities. The main styles of disposal are described and their advantages and disadvantages are explained in terms of their ability to meet the requisite safety standards and, hence, their suitability for housing the various categories of radioactive waste. Key words: radioactive waste, near-surface disposal, borehole disposal, intermediate depth disposal, long-term safety.

3.1

Introduction

3.1.1 Historical background to near-surface disposal Decades before deep geological disposal was conceived as an idea, radioactive waste was being generated by activities such as the production of radium for therapeutic uses, research into weapons and atomic power, and, beginning in the 1950s, power production itself. Many of these wastes were dumped at sea but, where the level of radioactivity was relatively low and a suitable site existed, disposal at the surface was also used. The Drigg low-level waste disposal site in the UK, for instance, was opened in 1959. For the most part these disposal facilities consisted of a simple trench into which the waste was tumble tipped. The prohibition of sea dumping by the London Dumping Convention, first as a two-year moratorium in 1982 and later as an outright ban, increased the volume of waste requiring disposal on land and created a need for more and larger disposal facilities. In some cases this led to closer scrutiny of existing near-surface facilities and a move away from simple trench systems to more engineered structures. Following a 43 © Woodhead Publishing Limited, 2010

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critical report from a parliamentary committee in 1986, for instance, disposal at the Drigg site changed from tumble tipping to a concept in which the wastes are encapsulated in concrete. This evolution in engineering practice has allowed near-surface disposal to continue to occupy an important niche in radioactive waste management. In Sweden, the Forsmark facility, opened in 1988, pioneered the disposal of short-lived low and intermediate waste at so-called ‘intermediate depths’: typically tens of metres below grade. A number of similar facilities, both existing and planned, have followed.

3.1.2 Current role of near-surface and borehole disposal in the overall context of radioactive waste management Longstanding practice in Europe classified radioactive waste according to the amount of radioactivity that it contained. Three basic categories were defined: low-, intermediate- and high-level wastes (LLW, ILW and HLW respectively). These categories were primarily designed to reflect the degree of hazard associated with each, especially when handling or storing them. It soon came to be recognised, however, that waste management is made more efficient if the waste categorisation anticipates the intended disposal route and, because the radionuclide half-life is an important determinant of the long-term safety of a disposal facility, this needs to be taken into account in the waste categorisation. The existing IAEA waste categorisation system (IAEA, 1994a) recognises this by subdividing ILW into long- and short-lived subcategories where ‘long half-life’ is taken to be greater than about 30 years; this allows strontium-90 and caesium-137 to be classified as short-lived. At the same time, the amount of long-lived radionuclides that are permitted in LLW is strictly limited. This leads to the possibility of disposing of LLW and short-lived ILW (collectively LILW) in an engineered near-surface facility. More recent developments (IAEA, 2009a) have recognised the usefulness of two new waste categories: the first is very low-level waste (VLLW), which should be sufficiently limited in both total and specific activity to allow disposal in a landfill (i.e. with minimal engineered barriers). The second is very short-lived waste, which is suitable for decay storage followed by release as non-radioactive waste. Low-level wastes have low specific activity and, normally, high volume: often the volume requiring disposal within a national programme will amount to hundreds of thousands or even millions of cubic metres. Highlevel wastes, on the other hand, are likely to amount to, at most, a few thousand cubic metres. While the low volumes of HLW and its very high radiation field provide strong motives for deep disposal, the large volumes of LLW demand more economic methods, and these can be justified on account


Disposal of intermediate radioactive waste

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of its lower activity. Typically, these more economic methods employ various methods of volume reduction coupled with near-surface disposal. By providing disposal facilities with a range of depths and engineering sophistication, wastes may be allocated to disposal facilities in a costeffective way. As discussed later, wastes containing radionuclides with halflives of a few thousand years (typically, radium-226 and carbon-14) may be allocated for intermediate-depth disposal. Particular difficulties can arise with respect to disused sealed radioactive sources. The problem is that these do not fit well into the existing waste categories because, even when their activity is modest, their small volume ensures that the specific activity is high. Small sources may, nevertheless, be disposed to a near-surface facility by simply mixing the sealed source with LLW and then immobilising the whole inside a standard container. At higher activities, however, the source activity may exceed the limit for a single waste package. Furthermore, even when the source size is below the package limit, the presence of a concentrated source may create a ‘hot spot’ in a near-surface disposal that can create difficulties for the safety case*. Consequently, the permitted activity of sealed sources in a near-surface disposal will usually be limited by regulation; normally there will be some sort of sliding scale that allows higher activities for shorter-lived radionuclides. Borehole disposal offers a potential solution for the disposal of long-lived sealed sources by providing the necessary levels of safety without going to the great expense of deep geological disposal.

3.1.3 Defining the ‘near surface’: limits to human intrusion Radioactive waste disposal aims to make the disposed material permanently safe in a manner that ensures that future generations are not required to perform any active management or, once the period of institutional control is over, even to know that the disposal facility exists. An important issue for safety, therefore, is that of human intrusion, i.e. the possibility that humans could unwittingly damage the facility and cause it to become a source of radiation exposure to themselves and others. Typical human intrusion scenarios include the construction of a dwelling on top of the facility, road construction, archaeological investigations, exploratory drilling, etc. In addressing these issues, reasonable intrusion depths are 3 m for house construction and 10 m for construction of a major road (NEA, 1987). In general, 30 m may be taken to be the limit of what is feasible in terms of excavations from the surface for road or railway construction in a competent rock (Ouzounian and Ozanam, 2009). Deeper excavations may *

The safety case is the collection of arguments and evidence that is intended to demonstrate the safety of a facility. A safety case will usually include one or more safety assessments.


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be associated with quarrying or open-cast mining, but one would expect to be able to eliminate these possibilities by siting a repository so as to avoid locations where there is potential for exploitation of minerals. Exploratory drilling can occur to any depth, of course, though the likelihood of such an intrusion diminishes significantly with the depth of disposal (Baker et al., 1997). A final consideration is that high-rise buildings could be compromised by the existence of underground voids (i.e. a repository) at depths of less than 50 m (Yamato, 2005). These figures suggest that, exploratory drilling apart, a reasonable value for the maximum depth of human intrusion is about 30 m. This value has been used by the BOSS (BOrehole disposal of Sealed Sources) borehole disposal system, which aims to avoid human intrusion by ensuring that the minimum depth of cover for buried disused sealed sources is 30 m. For an underground cavern this minimum depth may be increased to 50 m to allow for the possibility that high-rise buildings could be constructed immediately above it. Such figures are not to be regarded as absolute: where a facility is to be located in hilly terrain, for example, one may wish to admit the possibility of human intrusion by tunnelling (for a road, for example). Similarly, additional cover may be needed to offset the effect of erosion. While accepting these caveats, 30 m provides a useful discriminator between the near surface and intermediate depths, as we shall call them here.

3.2

Outline of the sections

Section 3.3 outlines the IAEA Safety Principles and requirements for radioactive waste disposal and then goes on to consider their main implications, while Section 3.4 investigates the safety of disposal facilities. Section 3.5 provides a description of the various styles of near-surface, intermediate-depth and borehole disposal facilities. Section 3.6 is concerned with the factors that influence the safety of such facilities. Section 3.7 briefly examines emerging and future trends in near-surface disposal. Finally, Section 3.8 provides a list of useful sources of information.

3.3

Safety requirements for near-surface disposal

3.3.1 IAEA safety principles and requirements With the exception of IAEA-sponsored international treaties, IAEA documents have no force of law. They do, nonetheless, form the basis for legislation and regulation for nuclear and radioactive waste issues in many countries. IAEA documents are arranged in a hierarchy. At the top are the fundamental safety principles (IAEA, 2006a) and the Basic Safety Standards (FAO et al., 1996).



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The IAEA fundamental safety principles (shown in the box below) apply to every kind of nuclear facility and supersede the earlier, sector-specific documents such as the waste safety principles. In many respects the operation Principle 1. Responsibility for safety The prime responsibility for safety must rest with the person or organization responsible for facilities and activities that give rise to radiation risks. Principle 2. Role of government An effective legal and governmental framework for safety, including an independent regulatory body, must be established and sustained. Principle 3. Leadership and management for safety Effective leadership and management for safety must be established and sustained in organizations concerned with, and facilities and activities that give rise to, radiation risks. Principle 4. Justification of facilities and activities Facilities and activities that give rise to radiation risks must yield an overall benefit. Principle 5. Optimization of protection Protection must be optimized to provide the highest level of safety that can reasonably be achieved. Principle 6. Limitation of risks to individuals Measures for controlling radiation risks must ensure that no individual bears an unacceptable risk of harm. Principle 7. Protection of present and future generations People and the environment, present and future, must be protected against radiation risks. Principle 8. Prevention of accidents All practical efforts must be made to prevent and mitigate nuclear or radiation accidents. Principle 9. Emergency preparedness and response Arrangements must be made for emergency preparedness and response for nuclear or radiation incidents. Principle 10. Protective actions to reduce existing or unregulated radiation risks Protective actions to reduce existing or unregulated radiation risks must be justified and optimized.


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(and the regulation of the operation) of a disposal facility is very similar to that of other nuclear facilities. The distinguishing feature of disposal is the recognition that at some stage the operator will relinquish or lose control over the facility. This may be a deliberate decision or it could be an involuntary act arising, for example, from a breakdown of organised society. For this reason, Principle 7 – protection of present and future generations – has special resonance for disposal and leads to a requirement to demonstrate, through post-closure safety assessment, that even in the absence of active control, the facility will remain adequately and permanently safe. The Basic Safety Standards (FAO et al., 1996) are concerned with the protection of human health from ionising radiation. These are expressed through the principles relating to justification, optimisation and risk limitation (Principles 4, 5 and 6). Coming below the fundamental safety principles and the Basic Safety Standards in the hierarchy of IAEA documents are the safety requirements. Requirements documents use the word ‘shall’ i.e. each requirement is obligatory. At the present time there are requirements documents that apply to near-surface disposal (IAEA, 1999) and to deep geological disposal (IAEA, 2009b). The document describing the requirements for near-surface disposal (IAEA, 1999) contains over 120 ‘shall’ statements but, in fact, all these can be more succinctly expressed through the 25 requirements that appear in the corresponding document for deep geological disposal (shown in Section 3.9) and, indeed, the IAEA has indicated its intention to develop a combined disposal requirements document. The implementation of these requirements is explained through three guidance documents that apply to near-surface disposal (IAEA, 2010a), deep geological disposal (IAEA, 2010b) and borehole disposal (IAEA, 2009b). Following standard regulatory practice (and in contrast to requirements documents), statements in guidance documents replace ‘shall’ with ‘should’. Without wishing to diminish the importance of the other requirements, it is probably true to say that the key safety issues for disposal are passive safety (Requirement 5), multiple safety functions (Requirement 7), containment of radionuclides (Requirement 8) and isolation of the waste so as to reduce the risk of human intrusion (Requirement 9).

3.4

Safety of disposal facilities

3.4.1 Operational safety It is convenient to separate safety issues into operational and post-closure phases where operational safety is concerned with the safety of workers and the public during the operational period. Here it is necessary to show that the facility is acceptably safe both in normal operation and in non-normal



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operation, where the latter includes incidents, accidents and emergencies. This demonstration is usually done using an operational safety assessment. Operationally, a radioactive waste disposal facility is much like any other nuclear plant; it has a radiological protection plan (including an emergency plan) and there is careful control and monitoring of both worker doses and radioactive contamination around and within the facility. It is normal for repository construction and repository operation to be ongoing simultaneously and, if this is the case, it will need to be recognised in the operational safety assessment. The operational safety assessment and any other relevant information such as security arrangements, financial guarantees, record keeping, quality management, etc., will be contained in the operational safety case, whose purpose is to describe the operation of the facility and demonstrate its safety. Worker doses will depend on the nature of the operations and the type of waste that is received. If, for instance, the facility is designed to accept shortlived intermediate-level wastes that emit significant levels of gamma radiation, there will probably be a need for remote handling of waste packages so that worker doses can be kept within the regulatory constraints. From this example we see that the way in which the facility is designed to be operated will be a factor in determining the waste acceptance criteria, i.e. the formal definition of the type of waste that can, and cannot, be accepted for disposal. Other operational factors that are likely to figure in the waste acceptance criteria are the weight, geometry and lifting arrangements of waste packages and the levels of removable contamination permitted on their surfaces. The end of the operational period is often defined to coincide with the start of the institutional control period (discussed in Section 3.4.4); alternatively, it may be defined to include the institutional control period – it does not really matter so long as there is no ambiguity.

3.4.2 Post-closure safety Post-closure safety needs to be addressed in a separate post-closure safety assessment. This aims to show that, even though future generations may not know of its existence, the design of the facility is such that the facility is permanently safe. To achieve this, all disposal facilities deploy multiple barriers to provide separate safety functions (Requirement 7) whereby engineered and natural barriers work together to provide the requisite level of passive safety, containment and isolation (Requirements 5, 8 and 9 respectively). The design aims to ensure that no undue reliance is placed on any one barrier. That said, a near-surface disposal is likely to rely more on the engineering barriers than the natural ones. The acceptability of this practice rests on the radionuclides in the waste being short-lived.


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The post-closure safety assessment will generally include a ‘design’ or ‘normal’ scenario, which estimates the expected radiological outcomes assuming that the facility performs as it is designed to do. This requires the post-closure safety assessment to anticipate the degradation of the repository structures (e.g. waste package corrosion, erosion of the repository cap) and to demonstrate that the repository will perform satisfactorily during and following events, such as storms or seismicity, that might reasonably be expected to occur during the assessment period, i.e. with a probability of, say, greater than 5%. To perform such a safety assessment it will be necessary to determine the storm and seismic hazards at the chosen site. For a near-surface disposal, key issues for post-closure safety are the integrity of the waste packages, the mechanical stability of the waste stack and the overlying cap, and the hydraulic properties of the surrounding environment. The main purpose of the repository cap, for instance, is to avoid or reduce infiltration of meteoric water into the waste. If the waste stack settles, say because of excess voidage, this may lead to cracking of the cap, increased water infiltration and leaching of radionuclides from the waste. Where a near-surface repository is located in the unsaturated zone, a key parameter is the distance between the base of the waste stack and the groundwater table. The greater this distance, the more radionuclides will be retarded as they migrate downwards towards the aquifer. Such effects need to be considered as part of the post-closure safety assessment. As another example, a below-grade near-surface repository sited in impermeable clay may be subject to ‘bathtubbing’ whereby the excavation fills up with meteoric water. Radionuclides will leach into this water and, as the water spills over on to the adjacent land, will contaminate the land and any crops that are growing there. Such examples illustrate the situations that might be analysed as part of the normal scenario. The post-closure safety assessment should also examine the consequences of less likely events such as severe storms and seismicity and demonstrate that, allowing for the lower probability of such events, the calculated risk is acceptably low. These ‘less likely scenarios’ usually include a series of ‘what if’ calculations that would probably include assessments of the consequences of premature failure of the engineered barriers or increased rates of erosion and flooding. Another standard type of scenario is that of human intrusion. This is usually considered to occur immediately after the end of the institutional control period and will normally include an assessment of the radiological consequences of someone building a house on top of the facility, eating food grown in contaminated soil and extracting drinking water (and, possibly, irrigation water) from a borehole that passes through or close to the facility into an underlying aquifer. In this context the main purpose of the



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institutional control period is to prevent human intrusion until the radionuclides have decayed sufficiently that the radiological consequences of human intrusion are acceptable. The post-closure safety assessment is an important document for the derivation of the waste acceptance criteria. The assessment should provide, in particular, (1) a limit on the total radioactive content of the repository and (2) a limit on the specific activity of individual waste packages. These limits are often expressed in terms (separately) of beta/gamma emitting radionuclides and alpha emitting radionuclides. This has the advantage of simplicity and may also facilitate the link between waste categorisation and the disposal route. In this way, all waste classified as LLW (for example) would be eligible for disposal in the facility. Alternatively, limits on total activity and specific activity may be expressed as separate values for individual radionuclides. A problem with this second method is that the radionuclides that are of greatest interest for disposal (e.g. iodine-129) are generally difficult to measure, so that it is usually necessary to establish a correlation between each radionuclide of interest and a different ‘indicator’ radionuclide (usually a gamma emitter) that is easier to measure. Typical examples are correlations between iodine-129 and caesium-137 (both fission products) or between carbon-14 and cobalt-60 (both neutron activation products). Consideration of the plume of contaminated water that could flow from the repository once the engineered barriers have degraded leads to the establishment of limits on the total activity. Other calculations, such as (1) exposure of the waste following removal of the cap due to erosion and (2) use of contaminated soil in a human intrusion scenario, lead to limits on specific activity. A particularly useful exposition of the derivation of activity limits for near-surface disposal is provided by IAEA (2003a). The waste acceptance criteria may also include limits on other parameters such as container strength, composition of the immobilisation grout, free water content and voidage (e.g. IAEA, 2000a; Morales, 2005). In some cases, there may also be a limitation on cellulose content on the grounds that, under cementitious conditions, cellulose can degrade to an organic acid that forms a complex with actinide ions and so makes them soluble and mobile. In most cases it is the repository operator that puts the wastes into the disposal container and grouts them into place and, where this is the case, it will be the disposer, rather than the producer, that takes on some of the burden of ensuring that the packages comply with the waste acceptance criteria.

3.4.3 Safety of mining and milling wastes The mining, milling and processing of uranium, thorium and rare earth minerals produces a range of so-called NORM (naturally occurring


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radioactive material) wastes that are high in volume. In general, mining wastes consist of excavated rock while tailings are the more finely divided residue that remains after the target element(s) is removed from the finely ground (i.e. milled) ore. In general, the proportion of the target element that is removed from the extracted ore is no more than a few tens of percents. Consequently, the concentration of (say) uranium in the waste rock and tailings materials is not so different from that of the original ore body. Ore processing may also produce wastes whose specific activity exceeds that of the original ore: an example is the radium-containing scales formed on the inside of pipes. For wastes like this, the term TENORM (technically enhanced NORM) may sometimes be used, especially in the USA. Invariably, NORM and TENORM wastes contain long-lived radioactive elements. Standard practice is to pile up the waste rock and the tailings on the surface. Sometimes a proportion of the waste rock may be returned to the mine, especially if it is needed for mechanically supporting the excavated cavities. However, the increase in volume that occurs when the rock is broken up ensures that the mine will never accommodate all the waste rock that it produces. Consequently, unless another use such as road construction can be found for the waste rock and tailings, it is inevitable that waste management facilities – rock piles for instance – will be left in situ around the mine. It is usual for the safety of these facilities to be governed by mining regulations rather than regulations for radiological protection; because these wastes are long-lived, there is usually an expectation that the facilities will be subject to perpetual care. The acceptability of this rests on the argument that the radionuclides are naturally occurring and that their specific activity and the associated hazard are low.

3.4.4 Significance of the institutional control period In terms of the IAEA requirements for disposal (Section 3.9), two requirements concern passive safety. Requirement 6 states that safety shall be ensured by passive means so as to minimise the need for action after the facility has been closed. Requirement 10 complements this, stating that in the post-closure period the site shall be appropriately supervised in order to protect and preserve the passive safety barriers. For the most part, the thing that this post-closure supervision is intended to protect against is human intrusion, and we may regard post-closure supervision as an engineered safety barrier whose main safety function is the prevention of human intrusion. Clearly, this safety function will have particular importance for near-surface disposal because of the greater risk of human intrusion there. The institutional control period is the time span over which post-closure supervision may be assumed to apply. A commonly used institutional



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control period is 300 years. This is convenient because it represents ten halflives and, therefore, a factor of about 1000 reduction in the activity of two of the most commonly encountered fission products, strontium-90 and caesium-137. It also happens to represent a period of time over which, it may be credibly argued, institutional control could be continued. Evidence for this comes from many organisations such as faith bodies, legislative assemblies and long-established private institutions. Reduced to very simple terms, the institutional control period may be seen as the period of time over which radionuclides in a repository should decay to near-exemption levels. This has implications for the waste acceptance criteria, as explained in Section 3.6.2. For mining and milling wastes the situation is rather different. The long-lived (albeit naturally occurring) radionuclides demand that surface facilities should be subject to perpetual control to avoid human intrusion. In low population locations, regular inspections of fences and signs to warn off potential intruders may be sufficient given that the low specific activity of the wastes requires longterm exposure to accumulate a significant dose. In facilities that are located near urban areas, more innovative solutions may be sought. At Port Hope in Canada, for instance, large volumes of historical radium-containing wastes were dumped in various locations around urban areas. The community decided to deal with the potential hazard by moving the wastes to one site, covering them with a thick layer of earth and designating the flattened, covered piles as a recreational area for common use. Institutional control measures are sometimes described as either ‘active’ or ‘passive’ (IAEA, 2003b). Active measures include activities such as a security presence and environmental monitoring. Passive measures usually include archiving of repository documents, control of ownership and, perhaps, markers designed to indicate the presence of a hazard. The main advantage of categorising the measures in this way is that it recognises that some forms of institutional control are likely to be more durable than others. The approach may lead to the institutional control period being divided into active and passive periods.

3.5

Styles of near-surface disposal

3.5.1 General Radioactive waste disposal facilities are expensive to develop and this, together with the general issue of sustainability, provides a strong motivation for waste minimisation. This important subject falls outside the remit of this document apart from noting that simple housekeeping measures on nuclear plant can lead to very significant reductions in waste production. Especially useful in this respect are strict segregation of waste


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streams and the avoidance of unnecessary waste creation by, for example, forbidding staff from taking disposable items (e.g. newspapers) into controlled working areas. Waste minimisation is concerned with waste prevention (indeed, this would be a better name for it) and should not be confused with volume reduction, which is a process that is applied after the waste has been created. Pre-disposal management of radioactive waste forms the basis for a suite of IAEA requirements and guidance documents (IAEA, 2000b, 2003c). While this subject largely falls beyond the scope of the present discussion, one aspect – volume reduction – has a strong influence on disposal costs and is so commonly used that it will be briefly described. Given that LLW may often include miscellaneous contaminated trash, volume reduction is often achieved by compaction or incineration. In waste compaction, waste is typically placed into 200 litre steel drums that are subsequently crushed in a high-force compactor to produce ‘pucks’. These are then placed inside a second (often concrete or concrete-lined) container and backfilled with a concrete grout (Fig. 3.1). In general, the larger this second container is, the greater will be the degree of volume reduction achieved. Where waste is capable of being incinerated, this may also be a useful volume reduction option, remembering that the resulting ash will have a considerably higher specific activity than the starting material and that it will itself need to be immobilised and packaged in some way. Incineration is also a useful option for incinerable liquids such as contaminated oils, which would otherwise be difficult to dispose. In the case of large-volume decommissioning wastes, worker doses will be minimised if these are placed in the vault intact and grouted into place. Because of the volume increase that occurs when fragmented components (e.g. heat exchangers) are placed into waste packages, it will often be found that emplacement of the intact component will also be most efficient in terms of volume utilisation.

3.5.2 Surface facilities: trenches and engineered vaults The needed level of sophistication of a repository design – for which one may read the number of engineered barriers – primarily depends on the nature of both the waste and the site. In desert areas with low rainfall, high evapotranspiration and a deep water table, safe disposal may be possible with a covered over mound or, more probably, simple trenches backfilled with excavated material. Nevertheless, safety assessments for such sites should still consider the full range of features, events and processes (FEPs), supplying justification for the exclusion of those that are not applicable. It will be necessary to go beyond the consideration of average rainfall and evapotranspiration values to consider the rainfall pattern, which may be



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3.1 Sectioned waste package consisting of a series of ‘pucks’ (supercompacted 200 litre drums) cemented into a larger drum (courtesy Ondraf Niras).

associated with infrequent but intense rainfall events. Such facilities may be found in operation in the USA and Australia and are planned in Egypt, Iran and other countries. At sites in temperate and tropical regions the groundwater table will often be close to the surface and here engineering measures will usually be necessary (1) to reduce the rate of leaching of radionuclides from the waste (usually by encapsulating them in concrete) and (2) to divert meteoric water away from the wastes through the installation of a low-permeability cap on


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top of the facility. Typically, wastes are cemented into steel or concrete containers and then placed inside concrete-lined vaults. The repository cap, which is put in place when the vaults are full, usually consists of at least a metre-thick layer of low-permeability clay, which is compacted after being put in place. This may be covered by an anti-bio-intrusion layer of loose rock of, maybe, half a metre, followed by soil that supports a layer of vegetation that provides some protection against erosion. Water management is an important issue when operating these facilities. During the operational phase when the wastes are still uncovered, meteoric water cannot be allowed to accumulate inside the waste cells and, if water does enter, measures must be taken to remove it. This may require the water to be decontaminated before it can be discharged to the environment. To reduce infiltration of meteoric water it is common to use a temporary ‘roof’ over a cell while it is in the process of being filled. This type of facility may be constructed on the ground surface or just below it. An above grade facility, i.e. one located directly on the ground surface, maximises the distance between the base of the facility and the groundwater table but needs some degree of lateral support. Placing the facility in a shallow excavation requires less lateral support and has some advantages with respect to access for waste placement and installation of a stable, erosion-resistant cap. Facilities of this type are found in France, Japan, Spain, the UK and Ukraine (Figs 3.2 and 3.3).

3.2 Aerial view of the closed near-surface repository at Centre de la Manche, France. The ‘corrugated’ surface is used to drain water off the cap (courtesy Andra).



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3.3 Operations at the Drigg, UK, near-surface repository. Half-height ISO containers form the outer layer of the waste package (courtesy BNFL).

3.5.3 Subsurface facilities: silos, caverns and tunnels Intermediate depth disposal Specially created subsurface facilities for disposal of low- and short-lived intermediate-level wastes at intermediate depths are in operation in Finland and Sweden; others are planned in France, Japan and the Republic of Korea. In Sweden the Fo¨rsmark facility has been in operation since 1988. LLW is deposited in tunnels that are located about 60 m below the bottom of the Baltic Sea in crystalline bedrock, as shown schematically in Fig. 3.4. Each tunnel is subdivided into a series of concrete cells where the waste is placed without further cementation. Short-lived ILW is lowered into a 50-m deep underground concrete silo that is divided into individual shafts and surrounded by a clay buffer. Wastes are lowered into the shafts and then backfilled with a concrete grout (Skogsberg and Ingvarsson, 2006). Radioactivity was first discovered in France so the country has a long history of utilisation of radioactive materials and, consequently, many historic wastes, of which radium-containing wastes form a significant part. France also has large volumes of graphite waste following the decom-


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3.4 Schematic view of the Fo¨rsmark LILW repository showing the silo and four disposal tunnels (copyright SKB, artist Jan M. Rojmar).

missioning of its first-generation, gas-cooled, graphite moderated reactors. These graphite wastes contain a range of radionuclides and, of those that occur in significant quantities, carbon-14 is the most long-lived. Both carbon-14 and radium-226 have half-lives of a few thousand years (to be precise, 5730 and 1600 years respectively). The radium wastes have much lower activity than the graphite wastes and a key objective of any disposal scheme will be to contain emitted radon. One option that is being considered is disposal in a specially created facility in a low-permeability clay or marl formation that is at least 50 m thick. The depth of the repository could be up to 200 m. Current plans (Ouzounian and Ozanam, 2009) envisage a simple tunnel with radium and graphite wastes housed in separate vaults. The waste package details have yet to be decided but the space-filling characteristics of rectilinear packages gives them a clear advantage in terms of efficiency of volume utilisation and avoidance of large voids that require backfilling. A similar facility, shown diagrammatically in Fig. 3.5, is proposed for use in Japan at Rokkasho (Yamato, 2005). In the Czech Republic a former limestone mine and munitions store has been converted to a repository for low- and intermediate-level waste. Under normal conditions the mine is dry though there is a possibility of flooding



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3.5 Investigatory tunnel for the proposed intermediate depth repository at Rokkasho in Japan (after Yamato, 2005) (courtesy JNFL).

during very severe storm events. To accommodate such rare events, the design incorporates a novel hydraulic cage design that diverts water away from the wastes (Biurren et al., 2005). In the Republic of Korea a subsurface repository for LILW is being constructed close to the Wolsung Nuclear Power Plant (NPP) on the southeastern coast of the country. Current plans envisage a drift running from the surface to a depth of 80 m where, in the first phase, six underground silos, 27 m in diameter and 50 m deep, will provide disposal capacity for 100 000 drums, each of 200 litres. Within the drums, wastes are encapsulated in cement or polymer. Drums are loaded into the silo inside a rectilinear concrete overpack that holds 16 drums in a single layer 464 array. Deep disposal of long and intermediate-level waste (LILW) While it may be convenient to distinguish between ‘intermediate-depth’ and ‘deep geological’ disposal there is, in fact, little difference between them. The choice of depth primarily rests on one, or a combination, of the following: the regulatory requirements in the specific country, the chosen location and stakeholders’ perceptions of the potential hazard. Deep disposal of LILW may result when former mines are adapted for the


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purpose. This strategy has been successfully pursued in Germany where three former mines, all of them unsaturated, serve as repositories for LILW. Two, those at Morsleben and Asse, are disused salt mines, where large cavities resulting from salt excavation have been partly filled with LLW. The third is the former iron mine at Konrad, where LILW is to be disposed in the old mine galleries, which remain unsaturated, at depths between 800 and 1300 m. More unusually, new deep facilities may be deployed for LILW disposal, as in Canada*, which envisages a disposal facility at Kincardine (near the Bruce power plant), where LILW from the power plants would be disposed in a limestone formation at a depth of 660 m. A second facility could be constructed at a depth of 500 m in granite underlying the Chalk River Laboratories site for LILW that is currently held there.

3.5.4 Mining and milling wastes The regulatory regime for mining and milling wastes may differ significantly from those applicable to radioactive waste from the nuclear industry. It is normal for these naturally occurring wastes to be excluded from the regulations governing radioactive material and for them to be controlled in a similar manner to other mildly toxic (but non-radioactive) wastes. Issues arising from their disposition may not be limited to the presence of toxic or radioactive materials. A common problem, for instance, is that oxidation of iron sulphide minerals in the wastes may lead to the formation of acidic compounds and acid run-off (IAEA, 2004a), just as it does in the similar phenomenon known as acid mine drainage. This can have a very negative effect on the surrounding environment, both from the acid itself and from the increased solubility of toxic metal ions in meteoric water. Surface accumulations of mining wastes can be ancient – rock piles are known that go back to the Bronze Age. These wastes frequently represent a minor hazard due to the presence of toxic and/or radioactive elements in low concentrations; there may also be a physical hazard from landslides when, as is very often the case, these wastes are deposited on sloping ground. A great deal of effort has been (and is still to be) expended on remediation of such facilities. Necessary actions usually include stabilising the rock piles and covering them with a metre or so of uncontaminated soil. Stabilisation may entail slope reduction, compaction, placing of supporting structures at the foot of the pile and water management. These measures reduce the physical hazard and prevent dispersion of the wastes. The cover has a number of purposes: it prevents wind-blown contamination, reduces radon *

Ken Nash, Chair, Nuclear Waste Management Organization, speech to Canadian Nuclear Association Seminar, 23 February 2006, http://www.nwmo.ca/.



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exhalation (in the case of radium-containing wastes), mitigates external radiation levels on the surface of the pile and, by encouraging vegetation to grow, controls erosion and promotes stability. The properties of the mining and milling wastes will, clearly, depend on the nature of the operation, e.g. whether operations are limited to mining or whether they include ore concentration and smelting. Mining produces rock wastes, while ore processing produces powdered material that is mixed with water in a slurry form. For all these types of waste we find that rich ore bodies tend to produce wastes that are relatively small in volume but high in metal (e.g. uranium) concentration. The reverse is the case for low-grade ore bodies. Smelting produces molten wastes that usually solidify to a glassy slag; if radioactivity levels are low, this may be a useful pozzolanic (cementproducing) material when crushed to powder. Other wastes are likely to include contaminated items of plant, especially pipework that has become coated with scale and may be sufficiently radioactive to be classified as ILW. Wastes arising from milling and ore concentration are water-borne and need to be allowed to settle to allow the solid and liquid phases to separate. This is usually done through the creation of a tailings dam and lake. There are two main forms: valley dam impoundments and ring dyke impoundments (IAEA, 2004a). The mill tailings are pumped in the form of slurry on to the lake. Water agitation is kept to a minimum and the solids fall to the bottom so that the supernatant water can be removed to be reused. In some earlier designs (especially in desert locations) water removal may occur mostly by evaporation (IAEA, 2004b). Key safety issues during operation are the stability of the dam (keeping in mind seismicity, for instance) and management of contaminated water. It is possible, for instance, that contaminated water may seep through the deposited tailings into the underlying ground where potable water may be located. Measures to prevent this include specially created dewatering drifts below the tailings facility, use of wicks laid within the tailings as conduits for water removal and groundwater abstraction to lower the surrounding groundwater table. Water that is removed in this way may be used as process water; surplus water will need to be treated before discharge to the environment (IAEA, 2004b). When the operations cease, it is usual for tailings management facilities to be converted to disposal facilities. This will require a number of engineering measures that must be designed to meet the needs of the site. The tailings, for instance, will no longer be covered by water, which allows the possibility of wind-borne releases. There may also be some radon exhalation. This will necessitate the capping of the tailings pond with clean soil and a layer of vegetation. Other measures might include improvements with respect to mechanical stability and use of an anti-bio-intrusion layer and/or a low


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permeability layer within the cap, which could include a geotextile membrane (IAEA, 2004a). Regardless of what engineering measures are implemented, the existence of an above-grade disposal facility containing long-lived or chemically toxic wastes will usually require some form of permanent management in the form of surveillance, environmental monitoring, repairs to the cap, etc. Where mining is an open-cast operation, it is more probable that the mine and tailings wastes will be used to backfill the excavated hole. It is even possible for this to be done during the course of operation. It is usual for the coarser mine wastes to be placed in the bottom of the hole so that the finer, more easily leachable tailing material may be placed above the level of the groundwater. If the hole is lined with clay to reduce interaction with the surrounding groundwater and if another use can be found for the surplus rock created by the increase in volume on excavation, such a design may provide a walk-away solution, i.e. it may require no long-term management.

3.5.5 Borehole facilities for large- and small-volume waste packages Surprisingly perhaps, the use of borehole facilities for radioactive waste storage or disposal is a relatively common practice throughout the world. Those that are most frequently found are the RADON borehole facilities developed in the former Soviet times and still in use, especially in Russia. These RADON boreholes may be conveniently subdivided into two types: small and large diameters. The small-diameter RADON boreholes consist of a roughly 10 cm diameter borehole leading to a larger stainless steel container placed at a depth of about 3 m (Sobolev et al., 2001). Disused sealed sources are dropped down the borehole so that they fall into the container. An unusual feature of these facilities is that the total activity of sources can be so great (PBq quantities) that it has been found useful to pour molten lead into the container to facilitate heat dissipation. While these facilities were originally intended for disposal, it seems that they are now regarded as storages. A similar type of facility has also been deployed (without the addition of lead) for disused sealed sources at Pelindaba in South Africa. The larger-diameter RADON boreholes (Fig. 3.6) are a more recent development (Prozorov et al., 2001). At the Moscow RADON site, boreholes are drilled to 40 to 50 m depth in clayey sediments above the saturated zone. Hole diameters can vary but the smallest is 1.9 m cased to a usable diameter of 1.5 m with the soil-casing annulus filled with bentonite– cement mortar. Wastes are packaged in 200-litre drums and emplaced in the borehole to within 2 m of the surface. Wastes are generally ungrouted in the



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3.6 Large-diameter borehole at the MosRadon site near Moscow (source: IAEA TECDOC-1368).

borehole to aid retrieval should this be needed. At the present time the emplacement is classed as storage but this could, with regulatory approval, be simply converted to disposal by filling the voidage with a free-flowing bentonite–cement mortar. Even larger diameter boreholes have been used for radioactive waste


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disposal in Western Australia at the Mount Walton Intractable Waste facility (Hartley et al., 1994) and in the USA where there are facilities at the Nevada test site (Cochrane and Hasan, 2005) and Savannah River (Fig. 3.7). At the Nevada test site a 3 m diameter hole was drilled to a depth of 37 m, this being about 200 m above the saturated zone. Drummed wastes are loaded into the hole to within 21 m of the surface and backfilled with excavated soil. PBq quantities of caesium-137 and transuranic waste have been disposed in this way. The safety case (Cochran et al., 2001a, 2001b) largely rests on the very low availability of water at the site.

3.7 Schematic diagram of the Greater Confinement Test facility at Savannah River (source: US NRC NUREG/CR-3774 V.5, Alternative methods for disposal of low-level radioactive waste. Task 2E: technical requirements for shaft disposal of low level radioactive waste).



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3.5.6 The International Atomic Energy Agency (IAEA) borehole disposal of sealed sources (BOSS) concept for disused sealed sources A relatively new development in borehole disposal facilities is the BOSS (BOrehole disposal of Sealed Sources) concept that stems from IAEA’s efforts to retrieve, ‘condition’ and safely store disused radium sources through its African Regional Cooperative Agreement for Research, Development and Training related to Nuclear Science and Technology – the so-called ‘AFRA’ project. In this context ‘conditioning’ entails the sealing of radium sources into 3-mm thick stainless steel capsules. Later developments, drawing on concerns over radiological security, have widened the remit of the conditioning campaign to encompass sources other than radium. This has included the development of a mobile hot cell that allows high activity sources to be remotely handled and conditioned (Crossland et al., 2011). Recognising that storage is not a permanent solution, the AFRA project then turned its attention to devising a safe yet economic means of disposal. Much of the development work was performed for IAEA by the Nuclear Energy Corporation of South Africa (Necsa), which produced a borehole disposal concept that largely relies on the containment/corrosion resistance properties of stainless steel by placing the conditioned sources inside a concrete-filled 6-mm thick stainless steel container. The disposal containers are then placed at a suitable depth (but always greater than 30 m to avoid

3.8 Schematic diagram of the BOSS borehole disposal concept (copyright Crossland Consulting).


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human intrusion) within a specially drilled borehole and concreted into place using a fluid cement grout (Fig. 3.8). In the absence of localised corrosion – something that can be assured quite simply under these cementitious conditions – corrosion rates are extremely low, so low, in fact, that most radioactive sources can be expected to decay to exemption levels before the disposal container is breached (IAEA, 2010c).

3.6

Designing for safety

3.6.1 Stakeholder views An early approach to site selection for near-surface disposal (IAEA, 1994b) adopted a strong technical focus, ranking the suitability of sites in terms of their physical characteristics, i.e. their geology, hydrology, geomorphology, etc. The danger with this approach is that it implies a search for the ‘best’ site when, in reality, experience has shown that, while the best is unattainable, virtually any site, with few exceptions, has the capability to host a disposal facility safely. This is not to say that all sites are equally good but, rather, that it is usually possible to overcome a site’s shortcomings through the use of more robust engineered barriers or, alternatively, by choosing a more appropriate geological horizon. A further difficulty with a site selection process that is focused on technical criteria is that it overlooks the issue of social acceptability. In many countries, the willingness (or not) of the local community to host a near-surface disposal facility is the single most important determinant in site selection. In most cases, the only communities willing to host such a facility are those that already have a nuclear facility in their locality. The argument that is usually deployed is that these communities have a higher level of understanding of nuclear issues and therefore appreciate the benefits of a new disposal facility in terms of jobs and inward investment. An argument that is less often heard is that such a ‘nuclear community’ may already hold significant quantities of waste in storage. If other communities are unwilling to receive this, the nuclear community has two choices: a continuation of the status quo or an agreement to host a new disposal facility. Where a community agrees to perform this role for the benefit of the country as a whole, it will, clearly, seek to accrue other benefits such as infrastructure improvements. For the owner of the wastes there may also be significant savings in transport costs, of course. Stakeholder involvement may not be limited to the issue of site selection; there are at least two cases where the local community has had a significant influence on the facility design. In the Port Hope area of Ontario, Canada, the local community wished to use relocated mining and milling wastes to create an area that could be used for local leisure activities, and this had



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repercussions with respect to design. In Belgium, in 1999, three communities were initially competing to host a near-surface repository for LILW and, in July 2006, the government decided to locate the repository at Dessel where the STOLA local partnership (STudie- en Overleggroep Laagactief Afval– Dessel or, in English, the study and consultative group on low-level waste– Dessel) had been established as a ‘consultation platform’ to investigate the ‘technical feasibility and social acceptability of the final disposal of Belgian low-level and short-lived waste within municipal boundaries’ (STOLA, 2004). The proposed repository design has been strongly influenced by STOLA. In particular, the floor of the disposal modules is supported by pillars, creating a basement whose purpose is to make it possible to perform continuous visual checks and intervene in the case of cracks or leaks.

3.6.2 Waste acceptance criteria The operational and post-closure safety assessments for a near-surface disposal facility should be used to define the waste acceptance criteria that will determine the wastes that can, and cannot, be disposed to the facility. However, since we may also choose to categorise waste according to its disposal route, it follows that there should also be a strong link between waste categorisation and safety assessment. In practice, mostly for historical reasons, this link is not always evident. Broadly speaking, near-surface disposal is reserved for LLW and shortlived ILW (collectively, ‘LILW’). Typically, LLW consists of contaminated trash and other wastes and may include contaminated clothing, tissues and wipes, and disused items of plant such as NPP secondary circuit pipework and pumps. Short-lived ILW usually consists of NPP operational wastes, especially wastes generated by water cleanup plant. These may include spent ion exchange resins and sludges, both of which must be conditioned into a suitable waste form before disposal. Operationally, the main distinction between LLW and short-lived ILW is that the former has a sufficiently low specific activity to allow manual handling whereas the latter requires remote handling and, therefore, an investment in remote handling equipment, which may be considerable. The upper limit on specific activity for beta/gamma emitters in LLW is usually around 10 kBq/g, which, in theory, could produce a contact dose rate of up to 5 mSv/h. In practice, limits on the total activity that may be contained within a single package and the fact that most of the radionuclides are present as surface contaminants mean that packages will be self-shielding so that dose rates will usually be much lower than this. Where remote handling equipment is available, this will allow much higher activity wastes to be emplaced. In this case the limiting specific activity for disposable short-lived radionuclides is more likely to be


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determined by post-closure safety rather than operational safety. The broad aim is that, after the 300-year institutional control period, the specific activity of the wastes should have fallen to an acceptable level. There are many ways of estimating this level and here we illustrate a simple method that might be applied. At the end of the institutional control period we may suppose that the repository cap is still in place, but an intrusion occurs by drilling through the cap into the waste and the drill cuttings are spread over the surface. This contaminated ground is then cultivated. The area of agricultural land needed to support a critical group of, say, five adults using modern, highinput agricultural practice is about 2500 m2 and a reasonable cultivation depth is 10 cm, which produces a total soil volume of 250 m3. If we assume that contaminated drill cuttings have a volume of about 0.5 m3 and are distributed through the soil volume, we derive a dilution factor of 500. Note that it is not necessary to assume uniform mixing because the higher doses and risks arising from more heavily contaminated areas will be offset by lower doses and risks from the less heavily contaminated areas. A reasonable expectation is that radiation exposure from consumption of contaminated food and water will be acceptable if the specific activity of the cultivatable soil falls below exemption levels. By way of example, the Basic Safety Standards document (FAO et al., 1996) has an exemption level for specific activity of caesium-137 of 10 Bq/g. For this radionuclide, a 300-year institutional control period represents around ten half-lives or about 1000 times reduction in activity. Combining this with the factor of 500 dilution in soil we may therefore expect the limiting specific activity for near-surface disposal of short-lived ILW to be about 5 MBq/g, which is about a factor of 500 greater than the beta/gamma limit for LLW (for this radionuclide). LILW will, of course, contain some long-lived radionuclides and it is necessary to set activity limits for these also. In this case we can largely ignore decay during the institutional control period so that, following the argument in the previous paragraph, the limiting values on specific activity will be around 100 (the dilution factor) times the exemption values. For uranium-238 and radium-226, this produces a limiting specific activity of 5 kBq/g. This estimate is close to those usually deployed in practice: France (Dutzer, 2002), for example, currently has a total alpha-specific activity limit of 4 kBq/g (calculated at 300 years) for the Centre de l’Aube near-surface disposal facility and the limit for the UK Drigg facility is similar. As a general rule, disposal at greater depths (as with the Fo¨rsmark facility) will permit higher specific activities to be disposed.



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3.6.3 Disposal environment With few exceptions, the groundwater pathway is the most important pathway for radionuclide migration from a repository. Consequently, the hydrology and hydrogeology of a site will have a profound effect on the repository design. As a general rule, making a safety case will be simpler in environments that have little or no water present (e.g. desert environments) than in those that do. Deserts have the further advantages of low agricultural productivity and low population density, which reduce the risk of human intrusion. Consequently, for disposal in a very dry environment (as at the Nevada test site) it will usually be possible to achieve adequate levels of safety with relatively few engineered barriers. In contrast, where a repository is sited below or close to the water table, it will usually be necessary to deploy engineered barriers such as a repository cap to reduce infiltration of meteoric water and encapsulation of the wastes in concrete to reduce leaching of radionuclides. There are a number of characteristics that may reduce the technical suitability of a site to host a disposal facility. These include the presence of an active geological fault, periodic flooding, rapid erosion or landslip. Even in these cases, however, it would often be possible to engineer the site so that these effects were insignificant for safety. Waste repositories are usually fairly simple structures so that, in the case of a nearby active fault, the repository structures could be designed to have sufficient mechanical strength to resist the disruptive effect of fault movements. Similarly, periodic flooding could be prevented through flood defences or by raising the ground level at the facility. It may be possible to overcome erosion and landslip through quite simple civil engineering measures such as slope stabilisation. All these measures would, of course, increase costs and this itself may be a sufficient disincentive to cause the developer to look for an alternative site or to place the facility at greater depth. There are a few site characteristics that have the capability to eliminate a site from further consideration altogether, and these are generally nontechnical. Sites that have special religious, social or wildlife significance would fall into this category and might include, for example, cemeteries, battle sites and national parks.

3.6.4 Engineered barriers It is normal for developers to design repositories following the multi-barrier approach. This means that the design deploys a series of engineered and natural barriers that, so far as possible, act independently of each other so that, if one or two were to fail, safety would still be assured. The first engineered barrier usually consists of the waste form, which is


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often created in two stages: conditioning and immobilisation. Conditioning is required if the waste has an unsuitable physical or chemical form; liquid and gaseous wastes, for instance, will not permitted to be disposed in this form and must be converted into solid substances. Similarly, chemically active wastes (e.g. alkali metals) must be conditioned by chemical reaction to form a more passive material. Immobilisation (sometimes called encapsulation) means that the conditioned waste is embedded in a solid matrix; this is usually concrete but could be a polymeric material or, more unusually these days, bitumen. When the waste is directly mixed with an immobilising material the resulting waste form is usually described as monolithic. The immobilisation matrix is likely to have a number of safety functions, such as radiation shielding during handling operations, reduction of the rate at which radionuclides may be leached from the waste by water, and consolidation of finely divided material. The conditioned and immobilised waste, i.e. the waste form, will invariably be held within a container and the whole is called a waste package. Containers are usually made from mild steel, cast iron, stainless steel or concrete. As already explained, the repository structure may take a number of forms. In some cases, the geometry of a facility may be simply determined by the footprint of the site. For near-surface disposal, semi-buried concrete vaults are commonly used. Where the vaults take the form of concrete-lined trenches, these may be subdivided into cells through the use of cross walls. Waste packages are emplaced into the cells or vaults and may be backfilled with concrete or previously excavated earth so as to fill up the gaps between the waste packages. The purpose of backfilling is to reduce voidage and, therefore, settlement of the waste so that subsequent shrinkage of the waste stack does not cause the repository cap to fail. The backfill may also act as a sorbent that reduces the rate of migration of radionuclides in groundwater. A repository cap may then be placed over the cells for the purpose of reducing the infiltration of meteoric water. Finally, institutional control, perhaps in the form of a security presence, fences, warning notices, etc., is implemented for the purpose of avoiding damage due to human intrusion. Depending primarily upon the nature of the site and the wastes, different kinds and combinations of engineered barriers will be deployed so as to achieve the requisite levels of operational and post-closure safety. For mining and milling wastes, for example, engineered barriers may consist of little more than an earth cover, measures to promote mechanical stability and regular monitoring of the condition of the mounded wastes. For LILW, on the other hand, we might expect to see an impermeable cap, monolithic waste forms, and an active security presence that is intended to prevent unauthorised access for hundreds of years.



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3.6.5 Natural barriers Natural barriers are less important for near-surface disposal than for deep geological disposal primarily because, if an unwitting human intruder were to take up residence on top of the facility, there would be no natural barriers, only engineered barriers, between him and the waste. Nonetheless, the surrounding environment (and therefore the potential natural barriers) will be important in determining the repository geometry – especially its depth, which may be chosen to keep the base of the repository above the water table. Natural barriers are likely to have raised importance in fault scenarios (e.g. premature failure of the cap) or at very long timescales when, although radionuclides or toxic materials are present in only small quantities, all the engineered barriers will have become degraded. In a near-surface facility where infiltrating meteoric water leaches out radionuclides and carries them down to an underlying aquifer, the natural barriers are represented by (a) the unsaturated rock that lies between the base of the repository and the underlying aquifer and (b) the saturated rock of the underlying aquifer that lies between the base of the repository and the so-called discharge point, i.e. the location at which this contaminated water might be utilised for drinking and other uses. Often the discharge point is assumed to be a water abstraction well (e.g. IAEA, 2003b). In these circumstances, the effectiveness of the natural barriers is largely expressed through the following parameters: the net rate of infiltration of water through the wastes, the radionuclide retention coefficients for the saturated and unsaturated rocks, the distance between the base of the repository and the groundwater table, and the hydraulic conductivity, porosity and hydraulic gradient in the aquifer. The geochemical properties of the surrounding environment are also important, but usually in the sense that certain geochemical regimes are to be avoided, especially where natural groundwater comes into contact with repository materials. In near-surface disposal the most usual example is the avoidance of unwanted cement–groundwater reactions, such as sulphateinduced degradation of concrete. At greater depths – in borehole disposal, for example – high chloride levels in groundwater could lead to rapid steel corrosion.

3.6.6 Safety functions A very useful way of defining and describing the roles of the engineered and natural barriers in providing adequate levels of operational and post-closure safety is through an explanation of their safety functions. It has been found that this helps the project team to adopt a consistent approach to meeting the design aims; it may also be an important input to the regulatory review. In the case of a near-surface disposal, for example, we would expect to see


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every major engineered component listed together with an explanation of its role in providing operational and/or post-closure. This may be conveniently presented as a table that should also show the timeframe over which each barrier is expected to operate.

3.7

Future trends

3.7.1 Remediation of historical near-surface disposal facilities In a few countries there are a number of disposals that were performed under (now) outdated regulation. Such disposals were legally made but they would not be permitted under modern-day regulation. The question then arises of whether these disposals should be remediated. This could be a relatively minor action such as improvements to the repository cap or, at the other extreme, it could include retrieval of the wastes followed by repackaging and a fresh disposal. ICRP (2000) has provided useful guidance on how to evaluate such a situation based on the principle that such a remediation should do more good than harm. In making such an evaluation, the risks to future generations are balanced against the risks to workers who would be required to perform the remediation. On the assumption that older regulations will not be so very different from those pertaining to the present day, there is a general expectation that if the disposals were compliant with the then-prevailing regulations, remediation will not extend so far as retrieval. There exist, however, some sites where such old disposals sit alongside ongoing disposals that are being made in compliance with current regulation. When assessing the post-closure safety of such sites, it will usually be necessary to consider the site as a whole, i.e. including all radionuclide releases for all facilities on the site. Given that new near-surface disposal sites can be extremely difficult to create, it is this aspect, rather than the approach recommended by ICRP (2000), that is likely to drive the need (or not) to remediate. Where such a situation is discovered, and pending a definitive resolution (which may require a government decision), the regulator may require future waste emplacements at the site to be regarded as storage, which entails certain requirements with respect to waste retrieval. Such situations have arisen at the Drigg site in the UK (Fig. 3.9) and at RADON sites in the Russian Federation. Given that the availability of radioactive waste disposal sites seems unlikely to keep pace with the increasing volume of waste requiring disposal, we may anticipate that this situation will be repeated elsewhere.

3.7.2 Intermediate depth disposal Intermediate depth disposal is an emerging trend for wastes that are unsuitable for near-surface disposal but not so radioactive or long-lived that



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3.9 The Drigg LLW repository in north-west England showing the location of past, present and possible future disposals. Since the photograph was taken, Vault 8 has been nearly filled (courtesy Sellafield Ltd).

they require deep geological disposal. The archetypal examples of this are the intermediate depth facilities proposed in Japan and France (Section 3.5.3). In some ways, however, the term ‘intermediate depth’ is unfortunate because it implies that the needed depth of disposal directly correlates with the activity of the wastes – what might be simply (if crudely) paraphrased as ‘intermediate depth disposal for intermediate-level wastes’. In fact, depth of disposal is as much a function of the chosen site and stakeholder perceptions as the nature of the wastes. Compare, for instance, the different approaches towards LILW disposal in Canada and Belgium that were described in Sections 3.5.2 and 3.6.1. Nevertheless, proposals for intermediate depth disposal are indicative of a more optimised approach in which the economic benefits of disposal at an appropriate depth are brought to the fore.

3.7.3 Borehole disposal As described in Section 3.5, at the present time borehole disposal facilities may be primarily seen as a variant of near-surface disposal. For disposals at greater depths, borehole facilities offer two significant advantages. The first is flexibility in terms of disposal depth and choice of host rocks. The second is cost, which should be lower because of the reduced volume of rock that needs to be excavated per unit volume of waste. This is especially important for the many countries that, while they have no nuclear programme, do have significant numbers of long-lived disused radioactive sources requiring


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disposal. For these countries the IAEA BOSS concept, or something similar, may offer a permanent solution at reasonable cost. The two great advantages of borehole disposal – flexibility and cost – are not limited to disused sealed sources; the use of very deep boreholes for disposal of spent fuel and HLW has been mooted by, for instance, SKB (1989). It seems to us, nonetheless, that development beyond the conceptual stage is unlikely for the foreseeable future given the investments already made in ‘conventional’ repository technology.

3.8

Sources of further information and advice

Almost without exception, national waste management agencies have policies promoting openness and transparency that require them to provide virtually free access to their reports and research findings. Many of these reports are available in English and most of them can be freely downloaded on the Internet. Hard copies can also be ordered over the Internet or requested by mail. Where hard copies are supplied, there may be a charge to cover administrative costs and postage. The British, Finnish, French, Swedish and US agencies (DOE) are particularly useful in this respect. The relevant organisations, addresses and websites are shown below. Organisation Full name and address Agence Nationale pour la Gestion des Déchets Andra Radioactifs 1/7, rue Jean Monnet, Parc de la Croix-Blanche, 92298 Châtenay-Malabry, Cedex, France International Atomic Energy Agency, Wagramer IAEA Strasse 5, PO Box 100, Vienna, Austria Nuclear Decommissioning Authority, Waste NDA Management Division (formerly UK Nirex Ltd) Curie Ave, Harwell, Didcot, Oxon, UK Nuclear Energy Agency of the Organisation for NEA-OECD Economic Development and Cooperation Issy-les-Moulineaux, Paris, France

Website www.andra.fr

www.ieae.org www.nda.gov.uk

www.nea.fr

Posiva

Posiva Oy, Olkiluoto, FI-27160 Eurajoki, Finland

SCK-CEN

Studiecentrum voor Kernenergie/Centre d'Etude www.sckcen.be de l'Energie Nucléaire Boeretang 200, BE-2400 MOL, Belgium

SKB

Svensk Kärnbränslehantering, Blekholmstorget www.skb.se 30, Box 250, SE-101 24 Stockholm

US DOE

US Department of Energy www.energy.gov 1000 Independence Ave, SW, Washington, DC 20585, USA


www.posiva.fi


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The website of the Belgium research organisation SCK/CEN is also a useful source. Work performed within the European Community Research Framework programmes tends to be listed on project-specific websites and also on the websites of the organisations involved. Two other useful sources, both for international regulatory standards and technical information, are the IAEA and OECD-NEA. Virtually all IAEA reports (including all those listed in the references) and even draft safety standards are available in downloadable form.

3.9

IAEA requirements for geological disposal

IAEA requirements for geological disposal (IAEA, 2006b) are listed below. Requirement 1. Government responsibilities The government is required to provide an appropriate national legal and organisational framework within which disposal facilities for radioactive waste can be sited, designed, constructed, operated and closed. This shall include: confirmation at a national level of the need for different types of disposal facilities; the definition of the steps in the development and licensing for different types of facility; the clear allocation of responsibilities, the securing of financial and other resources, and the provision of independent regulatory functions related to each planned disposal facility. Requirement 2. Regulatory body responsibilities The regulatory body shall establish the regulatory requirements for the development of each type of disposal facility and shall set out the procedures for meeting the requirements for the various stages of the licensing process. It shall also set conditions for the development, operation and closure of each individual disposal facility and shall carry out such activities as are necessary to ensure that the conditions are met. Requirement 3. The responsibilities of the operator The operator of a disposal facility shall be responsible for its safety. The operator shall carry out safety assessments and develop a safety case, and shall carry out all the necessary activities for siting, design, construction, operation and closure, in compliance with the regulatory requirements and within the national legal infrastructure. Requirement 4. Importance of safety in the development process Throughout the development of a disposal facility, an appropriate understanding of the relevance and implications for safety of the available


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options shall be developed by the operator, for achieving the ultimate goal of providing an optimised level of operational and post-closure safety. Requirement 5. Passive safety The operator shall site, design, construct, operate and close the disposal facility in such a way that safety is ensured by passive means to the extent possible and the need for actions to be taken after the closure of the facility is minimised. Requirement 6. Understanding and confidence in safety The operator of a disposal facility shall develop an adequate understanding of the facility and its host environment and the factors that influence its postclosure safety over suitably long time periods, so that a sufficient level of confidence in safety is achieved. Requirement 7. Multiple safety functions The host environment shall be selected and the engineered components of the facility designed so as to ensure that post-closure safety is provided by means of multiple safety functions; i.e. containment and isolation of the waste shall be provided by means of a number of barriers whose performance is achieved by diverse physical and chemical processes and whose individual adequacy shall be demonstrated. The overall performance of the disposal system shall not be unduly dependent on a single barrier or safety function. Requirement 8. Containment The engineered barriers, including the waste form and packaging, shall be so designed, and the host environment shall be so selected, as to provide containment of the waste during the period when radioactive decay has not yet significantly reduced the hazard posed by the waste and in the case of heat-generating waste when the waste produces heat energy in amounts that could adversely affect the containment. Requirement 9. Isolation The disposal facility shall be sited in a host environment depth that provides isolation of the waste from the biosphere and from humans over time periods of several hundreds of years for short-lived waste and at least several thousand years for high-level waste, with account taken of both the natural evolution of the disposal system and disturbing events. Requirement 10. Post-closure supervision An appropriate level of supervision shall be applied in order to protect and preserve the passive safety barriers to the extent that this is needed in order to fulfil the functions that they are assigned in the post-closure safety case. Requirement 11. Step-by-step development and evaluation



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Disposal facilities shall be developed, operated and closed in a series of steps, each supported, as necessary, by iterative evaluations of the site, of the options for design, construction, operation and management, and of the performance and safety of the disposal system. Requirement 12. Preparation of the safety case and safety assessment A safety case and supporting safety assessment shall be prepared and updated by the operator, as necessary, at each step in the development, operation and closure of a disposal facility. The safety case and safety assessment shall be sufficiently detailed and comprehensive to provide the necessary technical input for informing the regulatory and other decisions needed at each step. Requirement 13. Scope of the safety case and safety assessment The safety case for a disposal facility shall describe all the safety relevant aspects of the site, the design of the facility, and the managerial and regulatory controls. The safety case and its supporting assessments shall illustrate the level of protection provided and shall provide assurance that safety requirements will be met. Requirement 14. Documentation of the safety case and safety assessments The safety case and its supporting safety assessments shall be documented to a level of detail and quality sufficient to inform and support the decision to be made at each step and to allow for their independent review. Requirement 15. Site characterisation The site for a disposal facility shall be characterised at a level of detail sufficient to support both a general understanding of the characteristics of the site, including its present condition, its probable natural evolution, possible natural events and also human plans and actions that may affect the facility or its vicinity over the period of interest with regard to safety, and a specific understanding of the impact on safety of features, events and processes associated with the site and the facility. Requirement 16. Design The disposal facility and its engineered barriers shall be designed to contain the waste with its associated hazard, to be physically and chemically compatible with the host geological and/or surface environment and to provide post-closure safety features that complement those afforded by the host environment. The facility and its engineered barriers shall also be designed to ensure safety during the operational period. Requirement 17. Construction A disposal facility shall be constructed in accordance with the design as described in the approved safety case and safety assessments. It shall be


78


constructed in such a way as to preserve the post-closure safety functions of the host environment that have been shown to be important by the safety case. The construction activities shall be carried out to ensure safety during the operational period. Requirement 18. Operation A disposal facility shall be operated in accordance with the conditions of the licence and the relevant regulatory requirements to maintain safety during the operational period, and in such a manner as to preserve the post-closure safety functions assumed in the safety case. Requirement 19. Closure A disposal facility shall be brought to closure in such a way that the safety functions shown by the safety case to be important for the post-closure period are provided for. Plans for closure, including the transition from active management of the facility, shall be well defined and practicable, so that closure can be carried out safely at an appropriate time. Requirement 20. Waste acceptance Waste packages and unpackaged waste accepted for emplacement in a disposal facility shall conform to criteria fully consistent with and derived from the safety case for the operational and post-closure safety of the disposal facility. Requirement 21. Monitoring programmes A programme of monitoring shall be carried out prior to and during the construction and operation of a disposal facility, and after closure if this is part of the safety case. This programme shall be designed to collect and update the information needed to confirm the conditions necessary for the safety of workers and members of the public and the protection of the environment during the operation of the facility and to confirm the absence of any conditions that could reduce the post-closure safety of the facility. Requirement 22. Post-closure and institutional controls Plans shall be prepared for the post-closure period to address institutional control and the arrangements for maintaining the availability of information on the disposal facility. These plans shall be consistent with passive safety and shall form part of the safety case on which authorisation to close the facility is granted. Requirement 23. Consideration of nuclear safeguards Nuclear safeguard requirements shall be considered in the design and operation of any disposal facility that may accept materials or wastes to which nuclear safeguards apply. Nuclear safeguards shall be implemented in such a way as not to compromise the safety of the disposal facility.



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Requirement 24. Management systems Management systems* to provide for assurance of quality shall be applied to all safety-related activities, systems and components throughout all the steps of the development and operation of a disposal facility. The level of assurance for each element shall be commensurate with its importance to safety. Requirement 25. Existing disposal facilities The safety of existing disposal facilities shall be assessed periodically, when a modification is planned, when changes with respect to the authorised conditions happen and at least every ten years. In the event that any safety requirements are not met, measures shall be put in place to upgrade the safety of the facility.

3.10

References

Baker A J, Chambers A V, Jackson C P, Porter J D, Sinclair J E, Sumner P J, Thorne M C and Watson S P (1997), ‘Nirex 97: An assessment of the post-closure performance of a deep waste repository at Sellafield; Volume 3: The groundwater pathway’, Nirex Report S/97/012-3. Biurren E, Haverkamp B and Kucérka M (2005), ‘Closure of a chamber in the Richard Underground Repository’, in WM’05 Conference, 27 February–3 March 2005, Tucson, Arizona. Cochran J R and Hasan A (2005), ‘Greater confinement disposal of radioactive waste in borehole facilities’, in Disposal of Low Activity Radioactive Waste: Proceedings of an International Symposium on Disposal of Low Activity Radioactive Waste, 13–17 December 2004, Cordoba, Spain, Session 5, p. 393, IAEA, Vienna. Cochran J R, Crowe B and Colarusso A (2001a), ‘Results of the performance assessment for the classified transuranic waste disposed at the Nevada Test Site (1)’, in WM’01 Conference, 25 February–1 March 2001, Tucson, Arizona. Cochran, J R et al. (2001b), ‘Compliance assessment document for the transuranic wastes in the greater confinement disposal boreholes at the Nevada Test Site, Volume 2: Performance assessment’, SAND2001-2977, Sandia National Laboratories, Albuquerque, New Mexico. Crossland I G, Jova-Sed L A and Nachmilner L (2011), ‘A comprehensive system for the disposal of disused sealed sources’, in Radioactive Waste Management: Science and Technology, A special edition of International Journal of Environmental Engineering Science (IJEES), 2011, 3(1), in press. Dutzer M (2002), ‘From waste packages acceptance criteria to waste packages acceptance process at the Centre de l’Aube Disposal Facility’, in Radioactive Waste Products (RADWAP 2002) Proceedings, edited by R D Odoj, P Baier, P * The term ‘management system’ reflects and includes all the initial concepts of ‘quality control’ (controlling the quality of products) and its evolution through ‘quality assurance’ (the system to ensure the quality of products) and ‘quality management’ (the system to manage quality).


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Brennecke and K Kuehn, 23–26 September 2002, Forschungzentrum Juelich GmbH, Germany. FAO et al., Food and Agriculture Organization of the United Nations, International Atomic Energy Agency, International Labour Organisation, Nuclear Energy Agency of the Organisation for Economic Co-Operation and Development, Pan American Health Organization and World Health Organization (1996) International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources, Safety Series 115, IAEA, Vienna. Hartley B M, Wall B, Munslow-Davies L, Toussaint L F, Hirschberg K-J, Terry K W and Shepherd M (1994) ‘The establishment of a radioactive waste disposal facility in Western Australia for low level waste’, in Proceedings of the 9th Pacific Basin Nuclear Conference, Nuclear Energy, Science and Technology Pacific Partnership, 1–6 May 1994, Sydney, Australia. IAEA (1994a), ‘Classification of radioactive waste’, Safety Series, Safety Guide G1.1, International Atomic Energy Agency, Vienna. IAEA (1994b), ‘Siting of near surface disposal facilities’, Safety Series, Safety Guide No. 111-G-3.1, International Atomic Energy Agency, Vienna. IAEA (1999), ‘Near surface disposal of radioactive waste’, Safety Standards Series No. WS-R-1, International Atomic Energy Agency, Vienna. IAEA (2000a), ‘Inspection and verification of waste packages for near surface disposal’, TECDOC-1129, International Atomic Energy Agency, Vienna. IAEA (2000b), ‘Predisposal management of radioactive waste, including decommissioning’, Safety Standards Series WS-R-2, International Atomic Energy Agency, Vienna. IAEA (2003a), ‘Derivation of activity limits for the disposal of radioactive waste in near surface disposal facilities’, TECDOC-1380, International Atomic Energy Agency, Vienna. IAEA (2003b), ‘Radioactive waste management glossary’, International Atomic Energy Agency, Vienna. IAEA (2003c), ‘Predisposal management of low and intermediate level radioactive waste’, Safety Standards Series WS-G-2.5, International Atomic Energy Agency, Vienna. IAEA (2004a), ‘The long term stabilization of uranium mill tailings. Final Report of a Co-ordinated Research Project 2000–2004’, TECDOC-1403, International Atomic Energy Agency, Vienna. IAEA (2004b), ‘Treatment of liquid effluent from uranium mines and mills. Report of a Co-ordinated Research Project 1996–2000’, TECDOC-1419, International Atomic Energy Agency, Vienna. IAEA (2006a), ‘Fundamental safety principles’, Safety Series SF1, International Atomic Energy Agency, Vienna. IAEA (2006b), ‘Geological disposal of radioactive waste safety requirements’, Safety Standards Series WS-R-4, International Atomic Energy Agency, Vienna. IAEA (2009a), ‘Classification of radioactive waste, general safety guide,’ Safety Standards Series GSG-1, International Atomic Energy Agency, Vienna. IAEA (2009b), ‘Borehole disposal facilities for radioactive waste’, Safety Standards Series SSG-1, International Atomic Energy Agency, Vienna. IAEA (2010a), ‘Near surface disposal of radioactive waste’, Draft Safety Guidance DS356, International Atomic Energy Agency, Vienna, in preparation.



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IAEA (2010b), ‘Geological disposal of radioactive waste’, Draft Safety Guidance DS334, International Atomic Energy Agency, Vienna, in preparation. IAEA (2010c), ‘BOSS technical manual’, Waste Technology Section, International Atomic Energy Agency, Vienna, in preparation. ICRP (2000), ‘Protection of the public in situations of prolonged radiation exposure. The application of the Commission’s system of radiological protection to controllable radiation exposure due to natural sources and long-lived radioactive residues’, ICRP Publication 82, Pergamon. Morales A (2005), ‘El Cabril low and intermediate level waste disposal facility (Spain): new acceptance criteria’, in WM’05 Conference, 27 February–3 March 2005, Tucson, Arizona. NEA (1987), ‘Shallow land disposal of radioactive waste: reference levels for the acceptance of long lived radionuclides. A Report by an NEA Expert Group’, Nuclear Energy Agency of the Organisation for Economic Co-Operation and Development, Paris. Ouzounian G and Ozanam O (2009), ‘Disposal project for low-level long-lived radioactive waste in France’, in WM2009 Conference, 1–5 March 2009, Phoenix, Arizona. Prozorov L, Tkatchenko A, Titkov V and Korneva S (2001), ‘Prospects of large diameter well construction at ‘‘RADON’’ sites’, in WM’01 Conference, 25 February–1 March 2001, Tucson, Arizona. SKB (1989), ‘Storage of nuclear waste in very deep boreholes: feasibility study and assessment of economic potential’, SKB Technical Report 89-39, SKB, Stockholm. Skogsberg M and Ingvarsson R (2006), ‘Operational experience from SFR – Final repository for low- and intermediate-level waste in Sweden’, in Topseal Conference Session III, September 2006, http://www.euronuclear.org/events/ topseal/transactions/ . Sobolev I A, Ojovan M I and Karlina O K (2001), ‘Management of spent sealed radioactive sources at regional facilities ‘‘RADON’’ in Russia’, in ICEM’01 8th International Conference on Environmental Management, 30 September–4 October 2001, Bruges, Belgium. STOLA, 2004, ‘Belgian low-level and short-lived waste: does it belong in Dessel? Choosing a sustainable solution’, STOLA, Dessel, November 2004. Note that STOLA ceased to exist in July 2005 and was immediately succeeded by STORA, which is tasked with monitoring all nuclear affairs in Dessel. STOLA documents are downloadable from http://www.stora.be/. Yamato A (2005), ‘The current status of the JNFL subsurface disposal plan for relatively high low level radioactive waste’, in Proceedings of an International Conference: Safety of Radioactive Waste Disposal, Tokyo, 3–7 October 2005, p. 303, IAEA, Vienna.


4 Underground research facilities and rock laboratories for the development of geological disposal concepts and repository systems I. BLECHSCHMIDT and S. VOMVORIS, National Cooperative for the Disposal of Radioactive Waste–NAGRA, Switzerland

Abstract: Underground research laboratories (URLs) play an important and multi-faceted role in the development of deep geological repository systems for the disposal of radioactive waste, both from a scientific and technological point of view as well as for building public confidence. For more than three decades several countries have been conducting extended experimental and demonstration programmes in such facilities. The results produced from these investigations in URLs have proven to be valuable both in generic terms, i.e. developing and assessing the disposal concept and improving general acceptance, and in specific terms, as an essential means for detailed characterisation, design and assessment of potential repository systems. Key words: underground research laboratories, role in radioactive waste disposal programmes, case studies, Grimsel Test Site, Mont Terri Project, in situ experiments, large-scale experiments, bentonite, FEBEX, CRR, LTD, GMT, ESDRED, TEM, HG-A.

4.1

Introduction

4.1.1 Definition and roles of underground rock laboratories (URLs) The acronym ‘URL’ stands for underground research laboratory or underground rock laboratories. It has become the accepted generic term for underground facilities in which activities are carried out in support of radioactive waste repository development programmes (NEA, 2001a). A URL is any underground facility in which characterisation, testing, 82 © Woodhead Publishing Limited, 2010

Underground research facilities and rock laboratories

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Table 4.1 Relationship between laboratory studies, in situ experiments in URLs and natural analogues (modified after Kickmaier, 2002) Method

Characteristics

Laboratory experiments (‘classic’ lab studies)

Well-defined boundary Weeks to years conditions, unrealistic environment Defined but complex Several years to decades boundary conditions, realistic environment (‘repository relevant’) Boundary conditions less Up to millions of years well defined, realistic environment

Underground research laboratory (in situ experiments) Natural analogues

Duration of the ‘experiments’ (observation period)

technology development and/or demonstration activities are carried out in support of the development of deep geological repositories for radioactive waste. It may be an elaborate, purpose-built facility in which large research programmes can be carried out over many years or a quite simple facility, for example one attached to existing underground excavations, in which quite specific investigations may be made. URLs are located in rocks that are considered to be suitable for the construction of radioactive waste repositories, such as granite, salt, clay/shale, volcanic tuff, etc. Experiments in underground rock laboratories represent one of the complementary approaches that are needed for building the scientific understanding and for gathering the diversity of data to analyse the potential behaviour of repository systems over long time spans. URLs provide the link between investigations in ‘classic’ laboratories, where spatial scale is quite small, and natural analogues, where spatial and temporal scale are much larger (Table 4.1). The work in URLs allows an in-depth investigation of the selected geological environment and the engineered components of the repository, providing the opportunity to test models at more appropriate (repositoryrelevant) scales and boundary conditions than can be achieved from the surface only. Consequently, the RD&D programmes in the URLs form an important part of the overall radioactive waste management strategies (Fig. 4.1). In national radioactive waste programmes the general objectives of underground research facilities can be summarised as follows (modified after McCombie and Kickmaier, 2000): . .

Developing the technology and methodology needed for underground experimentation (especially important for countries with little tradition of tunnelling, mining or resource exploitation) Providing data that improve the basic understanding of the behaviour of


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4.1 URLs as part of the overall waste management strategy (modified after Kickmaier, 2002).

. .

repository system components – and of their interactions – as input to qualitative and quantitative assessments Demonstration and, at an advanced stage, optimisation of engineering components and system processes Building confidence within the scientific and technical community as well as within the public that the understanding of the important processes governing repository performance is adequate

The value of URLs in communicating with the public and with ‘decision makers’ has been recognised as a valuable non-technical asset. Although the nature of the facilities and their often remote locations preclude site visits by vast numbers of the public, many thousands of members of the general public are received each year in the operating URLs. In addition, illustrative material from experiments in these facilities is frequently used in public documents on disposal produced from within and outside the waste management community. A more recent trend has been towards the use of URLs to demonstrate the feasibility of the technology proposed for implementation in underground repositories. Currently many experimental programmes in URLs are focusing on engineering practicability and demonstrating the behaviour of large multi-component systems. The high costs and long timescales that must be considered for such experiments make them an appropriate focus for international cooperation.



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4.1.2 Chapter set-up Section 4.1 gives an introduction to underground research laboratories (URLs). The different types of URLs and their role in the development of deep geologic repositories are described in Section 4.2. Section 4.3 summarises the basics for planning and designing an underground research facility taking into account most recent examples. Public outreach and the role of URLs as training platforms are discussed in Section 4.4. A few case studies of in situ characterisation and testing, engineered barriers and demonstration experiments are highlighted in Section 4.5. The conclusions and future needs are summarised in Section 4.6.

4.2

Types of URLs and their role in the staged development of geological repositories

4.2.1 Different types of URLs Many different definitions have been used in the past thirty years to describe the work which is performed in URLs such as ‘first-generation URLs’, ‘rock characterisation facilities – RCFs’, ‘off-site URLs’, ‘second-generation URLs’, ‘site-specific URLs’, ‘on-site URLs’ or ‘performance confirmation facilities’, are some of the most frequently used names. As the objectives between the various types of URLs can differ a lot, leading to a certain degree of confusion, for the remainder of this chapter we are using the most accepted definitions, as is shown below. We will distinguish URLs in two major categories: . .

Generic URLs Site-specific URLs

Generic URLs (equivalent to off-site URLs or first-generation URLs) are independent of final disposal sites and comprise facilities that are developed for research and testing purposes at a site that will not be used for waste disposal; they provide, however, information that is expected to support the disposal of radioactive waste elsewhere. Both non-destructive as well as destructive experiments are performed (experiments often culminate with a post-mortem phase) and a flexible approach, in which modifications are possible, is warranted. The major roles of generic URLs are: . .

Development of methodology (preparations for in situ testing – surface/ underground) and testing of the transferability of data obtained in the laboratory to in situ tests (e.g. sorption processes, rock mechanics) Collection of information (host rock – barrier properties, engineering feasibility, interactions between host rock and engineered barriers, work


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. .

Geological repository systems for safe disposal process, empirical observations, data sets/observations for model testing, independent evidence, data for direct use, e.g. by use of transfer factors, modelling, etc.) Concept testing and demonstration (e.g. engineered barriers – emplacement, evolution, retrieval) Platform for interaction (professional community – different organisations, different disciplines; broad public)

Site-specific URLs (on-site URLs, second-generation URLs or rock characterisation facilities) as the name suggests, are located in the host rock in an area that is considered as a potential location for a repository. They include facilities that are developed for specific investigations at the given site and may, indeed, be a forerunner to the development of a repository at the site. Because they may become part of a future repository, activities performed as well as the URL itself should not unduly affect the potential future repository; for example, they should not have a negative impact on the host rock performance and their planning and execution is part of the development of the specific repository project with more formal requirements. The major differences between these two URL categories are summarised in Fig. 4.2. A special case within the site-specific URLs is the ‘performance confirmation facility’, which is continuously gaining more relevance as the

4.2 Major differences between ‘generic URLs’ and ‘site-specific URLs’ (based on Kickmaier, 2002).



87

Table 4.2 Overview of generic Underground Research Laboratories (modified after IAEA, 2001; NEA, 2001a; NEA 2001b) Generic URL Operation

Host rock, depth

Comments

Whiteshell Canada Underground AECL Research Laboratory (URL) 1984–2003

Granite 240–420 m

Purpose built; generic Shaft sealing on-going

Olkiluoto Research Tunnel 1992–

Finland Posiva

Granite (tonalite) 60–100 m

Purpose built; parallel to repository facilities

Amelie 1986–1992

France ANDRA

Bedded salt

Pre-existing tunnels

Fanay-Auge`res 1980–1990

IRSN

Granite Uranium mine


Sediments (shale) 250 m


Permian rock salt anticline Mining levels 490– 800 m, cavern 950 m


Tournemire (Tournemire Research Tunnel) 1990– Asse Mine 1965–1997

Country Organisation

IRSN Germany GSF

Tono 1986–2006

Japan JNC/JAEA

Sediments, uranium Pre-existing tunnels mine 130 m

Kamaishi 1988–1998

Granite, Fe mine


Mizunami Underground Research Laboratory (MIU) 2004–(shaft sinking initiation)

Granite 1000 m (shaft)

Purpose built; generic Surface investigations since 1996

Horonobe Underground Research Laboratory 2005–(shaft sinking initiation)

Sedimentary rock 500 m (shaft)

Purpose built; generic Surface investigations since 2000

KURT– Korean Underground Research Tunnel 2006–

South Korea KAERI

Granite 90 m

Purpose built; generic

Stripa Mine 1976–1992

Sweden SKB

Granite, Fe mine 360–410 m



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Table 4.2 (cont.) Generic URL Operation


¨ spo¨ Hard Rock A Laboratory (HRL) 1995–

Host rock, depth

Comments

Granite 200–460 m (ramp/ spiral)


Grimsel Test Site (GTS) 1984–

Switzerland Nagra

Granite 450 m

Purpose built; parallel to existing tunnels

Mt Terri (FMT) 1995–

SNHGS

Opalinus Clay 400 m

Purpose built; parallel to existing tunnels

Climax, Nevada 1978–1983

USA US-DOE

Granite, mine 420 m


G-Tunnel, Nevada 1979–1990

Tuff 300 m


Busted Butte, Yucca Mountain, Nevada 1998–

Bedded tuff 100 m


development of deep disposal of radioactive waste moves forward in several countries. It is expected that in the near future they will warrant the designation of ‘third-generation URLs’ and they will comprise facilities designed for confirming key phenomena of importance during and after waste emplacement in an existing repository, e.g. provide input to the decision-making for the final closure of the repository.*

4.2.2 Past and present URLs Tables 4.2 and 4.3 summarise past and present URLs in a succinct form. Consistent with the definitions above, the underground facilities are broadly categorised as in generic and site-specific URLs. The first category is further divided into the ones using pre-existing underground facilities, i.e. tunnels originally developed for different purposes, and the purpose-built laboratories. Examples of generic URLs are presented in Figs 4.3 to 4.7. During the last 40 years more than 20 facilities have been utilised for underground research and testing (Table 4.2); about one-third of these facilities are currently in active operation or under construction. The number of site-specific facilities, as expected, is increasing, e.g. the ONKALO facility in Finland (Fig. 4.11), the Bure URL (Meuse/Haute *

Excerpt from P. Zuidema’s presentation at the celebration ceremony for the 25th anniversary of the Grimsel Test Site (GTS) in Switzerland (September 2009).



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Table 4.3 Overview of site-specific Underground Research Laboratories (modified after IAEA 2001; NEA 2001a; NEA 2001b) Site-specific URL Operation


Host rock, depth

Comments

High-Activity Disposal Experiment Site URL, Mol (HADES-URF) 1984–

Belgium Boom clay (plastic Purpose built; sitespecific GIE EURDICE clay) 230 m

ONKALO, Olkiluoto 2003–

Finland Posiva

Granite (tonalite) 500 m (ramp)

Purpose built; sitespecific

Meuse/Haute-Marne (Bure URL) 2000–

France ANDRA

Shale (indurated clays) 450–500 m


Gorleben 1985–1990

Germany BfS, DBE

Salt dome below 900 m


Konrad 1980–

Limestone, Fe mine Facility in former 800–1300 m iron mine

Morsleben (ERAM) 1981–1998

Salt dome, K/salt mine below 525 m

Facility in decommissioning stage

Salt (bedded) 655 m

Operating repository since 1999*

Welded tuff 300 m (ramp)


Waste Isolation Pilot Plant (WIPP) 1982–(1999) Exploratory Studies Facility (ESF), Yucca Mountain, Nevada 1996–

USA US-DOE

* A part of the underground facility continues as a performance confirmation facility or third-generation URL.

Marne) in France (Fig. 4.10) and the Exploratory Studies Facility (ESF) at the proposed Yucca Mountain site in the USA. The ESF (Fig. 4.9) has already completed a broad in situ experimentation programme at the rock formation (tuff) of the proposed high-level waste repository. At the time of writing this text, the Bure URL has completed the first phase of tests at the formation of interest (Callovo-Oxfordian) and is developing the next phase of underground testing. In Spring 2009, ONKALO passed the mid-point of the target depth and had already characterised important geological features crossed by the access ramp. Both in Finland and in the USA the intention has been to integrate part of the URL in the infrastructure of the repository facility. For ANDRA’s programme in the Meuse/Haute Marne region this is still an open question and depends on the final location of the repository area, which is to be proposed around 2015. Examples of site-specific URLs are presented in Figs 4.8 to 4.11.


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4.3 Grimsel Test Site (GTS), generic URL in granite, purpose-built parallel to existing tunnels, 400 m, horizontal ‘ramp type’ access (1983–1984: construction/1984–1986: basic site characterisation/1986–: scientific programme organised in GTS project phases).

4.4 A¨spo¨ Hard Rock Laboratory, purpose-built generic URL in granite, 450 m, ramp and shaft access (1986: investigations/1988: selection of site/1990–1995: construction/1995– : scientific programme) (# SKB 2009). © Woodhead Publishing Limited, 2010


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4.5 Mont Terri rock laboratory (FMT), generic URL in clay formation, purpose-built parallel to existing motorway tunnels, horizontal ‘ramp type’ access, operated by swisstopo (modified after Bossart and Thury, 2008).

The A¨spo¨ underground facility in Sweden deserves a special mention here, since it is a purpose-built laboratory but will not be part of the repository facility (Fig. 4.4). This addresses an important issue, namely whether or not an underground research laboratory should always be constructed as the first step of a repository facility. In the Swedish case, it is argued that the answer depends on many different aspects, but most importantly on the characteristics of the rock formation at hand; for example, in widespread rocks with properties varying homogeneously in space, fundamental research could be carried out in a representative generic URL. The two key assumptions for this statement to be valid are homogeneity and representativeness. Homogeneity is used here in a statistical sense; for example, the spacing between two large-scale fracture zones or the thickness of fracture zones follows a statistical distribution that has characteristics that are homogeneous in space (a constant mean value and standard deviation for a second order approximation). Representativeness, the second key assumption, is associated with scale; the location and geometric characteristics of the URL should be such that they can capture the smallscale variability of the key properties of the formation and can allow upscaling to make an inference about the parameters at large scales. Under these two conditions it can be confidently argued that a generic purpose-built URL at depths similar to those of the planned repository can suffice for performing most of the engineering and scientific tests, and


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4.6 Mizunami URL (status of excavation in December 2009: main shaft: 413.0 m; ventilation shaft: 436.3 m), purpose-built generic URL in granite with sedimentary cover, shaft access (1996: project initiated/2003: construction started) (# JAEA 2009).

definitely for carrying out a lot of the illustration or demonstration tests. It should be emphasised that this statement does not imply that during construction of the repository there will be no test facility or in situ testing. On the contrary, it is expected that a number of test niches and galleries will be constructed to gain specific data for the rock formations of interest and feed into the finalisation of the designs and safety assessments. These data would be (a) of a confirmatory nature and (b) a prerequisite for the engineering design. The need, however, to perform fundamental tests for



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4.7 KURT–KAERI underground research tunnel, purpose-built generic URL in granite, ramp access (2003: project initiation and site characterisation/2004: basic and detailed design and achieved construction license/2005: start construction/2006: completion of the tunnel/2007: start of operation) (# KAERI 2009).

4.8 HADES (Mol) URL, site-specific URL in plastic (boom) clay, shaft access (1980: construction start/1985: URL operational/1987: first extension / 1997–2000: second shaft /2001–2002: connecting gallery/ 2006-2007: PRACLAY gallery) (# SCK-CEN 2009).


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4.9 YUCCA Mountain Exploratory Studies Facility (ESF), site-specific URL in welded tuff, 300 m, ramp access (1987: selection of site/1994–: construction/1997–: drift scale thermal tests started/1999–2008: demonstration experiments/2009: programme stopped/re-evaluation of options) (courtesy of the US Department of Energy).

4.10 Meuse/Haute-Marne URL at Bure, site-specific URL in CallovoianOxfordian Argillites, shaft access (1999: authorisation for installation/ 2002: construction suspended/2003: construction restarted) (# Andra 2009).



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4.11 ONKALO Underground Rock Characterisation Facility, sitespecific URL (URL in granite), ramp access (2004–2011: construction/ 2004–: investigation/2010: characterisation depth /2011: excavation complete/2015–: construction of repository (# POSIVA 2009).

concept development or model confirmation, e.g. a long-term migration test to verify the performance assessment models, could have already been satisfied with the generic URL.

4.2.3 The two bounding approaches for developing a URL programme As the title suggests the two approaches described here are the two extreme cases, presented to more clearly illustrate the points. In reality, it is expected that a hybrid approach will be followed that lies somewhere in the middle. The top-down approach starts with a very systematic evaluation of the needs for the URL in the context of the repository programme and derives the specific experiments from these needs. The bottom-up approach starts from an evaluation of the possible experiments that can be performed and builds up experiments or activities that would serve the repository programme needs. From a technical and scientific point of view the first approach would seem the preferable one. However, as mentioned in Section 4.1, the roles of a URL extend beyond technical ones, e.g. focal points for training, know-how build-up, etc. It is in this context, and for programmes that are at an early stage, that the bottom-up approach is usually followed. Thus, the welljustified literature survey of what has been or is being done in other URLs occasionally tends to become the proposed URL programme rather than providing input for the programme of the new URL. As the repository


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programme in a country develops, e.g. as the first set of provisional safety analyses is completed or as engineering designs are developed, the planning can focus on new experiments that should serve the specific needs of and be derived from the high-level requirements of the respective repository programme.

4.2.4 Evolution of URL programmes over time The research, development and demonstration (RD&D) programme to be followed in a URL depends on the stage of the national programme, the knowledge and expertise available internationally for similar geologic formations or repository designs and the long-term plans for the repository development. The RD&D programme and the resulting activities in the URL evolve dynamically over time and should be targeted to fulfilling the evolving needs of its various users. ‘Users’ here is used broadly to define the various technical disciplines involved in a geological repository realisation, e.g. those represented in the site and field characterisation group, in the repository design and engineering group or in the performance assessment group, as well as the non-technical disciplines, e.g. the group responsible for communication with the various stakeholders, policy- and decision-makers and the public. It suffices to mention that a URL, both an existing one as well as a newly built one, can benefit from experience accumulated to date elsewhere. The strategy adopted most of the time in selecting medium- to long-term experimental programmes is a trade-off between the following two aspects: (a) a set of experiments that have enough overlap with existing programmes to strengthen the confidence, through verification or benchmarking, that the developments and results are confirmed by international comparison and (b) a set of experiments or projects that focus either on activities specific to the repository system selected for the particular programme or on novel topics for which a significant contribution to the state-of-knowledge can be accomplished. The initial experimental programmes are usually determined by shortterm needs, partly because during the short term one has to perform a certain type of fundamental characterisation tests to develop the basic understanding and databases for the design of more complex experiments. The basic evolution of user needs and respective programmes in a URL is shown in Fig. 4.12 for the case of the Grimsel Test Site (GTS) in Switzerland (www.grimsel.com); the GTS has been in operation since 1984 and for each of the three past decades of operation the main focus is as summarised in this figure. At the beginning – in the early 1980s – emphasis was placed on developing the appropriate tools and techniques that would be needed for carrying out more complicated experiments. This was a very important



4.12

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Evolution of needs → example of the Grimsel Test Site (GTS).

objective at that time because existing technologies from other industries (e.g. oil and gas industry, mining, tunnel engineering) had to be adapted to the requirements of radioactive waste disposal: the low-permeability host rock formations to be evaluated, the longevity of equipment and instrumentation and the desire for non-intrusive or non-destructive testing techniques are just a few examples of the additional requirements posed in the context of radioactive waste disposal. A fundamental toolkit was obtained during that period, which allowed designing and performance of more complex experiments but also had a significant contribution to the site exploration techniques employed by Nagra at that time. Three examples of such techniques are (a) the multi-packer hydraulic systems developed, which were employed in the crystalline rock characterisation from a tunnel at the Piz Pian Grand potential repository site (Nagra, 1988), (b) the evaluation of techniques and materials for sealing of boreholes drilled from underground exploration tunnels, as preparation for the underground characterisation of the Wellenberg potential repository site (Coons et al., 1987; see also Blu¨mling and Adams, 2008) and (c) the cross-hole tomography studies, as


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preparation for the second phase of the crystalline characterisation programme in Switzerland (Niva et al., 1988). Similar examples can be found in other first-generation URLs, e.g. at the A¨spo¨ Hard Rock laboratory in Sweden (SKB, 1996). New technologies are of course being tested and evaluated continuously, but compared to the first decade this is done rather on an as-needed or ad hoc basis; such an example is wireless data transmission techniques that have been tested at GTS since 2006. The second decade shows an additional emphasis on performance assessment questions, in particular testing and verification of models for radionuclide migration, study of the effects of special materials to be used in repository construction (e.g. concrete) on the natural system or other issues such as migration of repository-generated gas. A unique characteristic of the GTS among rock laboratories worldwide is the existence of a radiation controlled zone (IAEA Level B/C) in one of the investigation tunnels, which allows experiments to be carried out with radioactive tracers in the geosphere under realistic conditions. With this set-up, it has been possible to check the results of small-scale laboratory experiments in the field and to test directly model calculations of the migration of radioactive substances. During this decade, the emphasis expanded to include large-scale engineering demonstration experiments, such as FEBEX (full-scale engineered barriers experiment) and GMT (gas migration test). These types of experiments, associated with a certain maturity of the URL and associated repository development programmes, also take place elsewhere, as shown in Table 4.3. Since the end of the 1990s, further steps have been taken towards more integrated (and more complex) projects with: (a) field experiments under repository-relevant boundary conditions, as far as possible (large-scale, long-term, realistic hydrogeological conditions, etc.), and (b) projects addressing the implementation of a geological repository (engineering feasibility, operational aspects, closure, monitoring and possible effects of repository construction on the surrounding rock). Example case studies of detailed programmes are presented in Section 4.5.

4.3

Planning and designing an underground research facility: basic considerations

This section is meant as a short primer mainly for programmes that want to develop their own URL facility. Our starting point is that national strategies for URL development strongly depend on the following: 1.

What is the timescale for the radioactive waste disposal and how do activities feature in this timescale?


Underground research facilities and rock laboratories 2.

3.

4. 5. 6.

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Has the potential host rock been selected and what requirements does this decision (or lack thereof, if the host rock is still open) have on the URL programme? Is the implementation of an URL needed to develop and test a disposal concept and is going underground the most efficient way to satisfy research and testing needs? Can desired information be obtained by cooperating in work performed in the URL of another country? Can an existing underground facility be adapted for generic URL work in a cost-effective manner? Is the overall waste-disposal programme sufficiently advanced to provide continuity when the URL work under consideration is completed?

In the following two sections the requirements, timing and resources needed for planning a URL are discussed in the context of the six points listed above.

4.3.1 General requirements for implementation of a sitespecific URL Key issues to consider in this context are discussed using the ONKALO Underground Rock Characterisation Facility in Finland, which represents an example of a site-specific URL currently under construction. A fundamental requirement is that a site-specific URL must be constructed without jeopardising the long-term safety of the selected repository site. Fulfilling this requirement also renders it possible later to link the facility to the repository as a part of it (Posiva, 2003a). The URL buildings and structures should comply with the land use plan of the region, ensuring the needs of present operations and future projects in the area (Posiva, 2003a). The URL shall be based on proven technology that should help to minimise technical risks and to improve working conditions. It shall enable the operation of the final disposal facility according to the defined time schedule and the collection of sufficient information and knowledge of the repository host rock and other underground conditions to ensure the long-term safety of the final disposal and to meet the requirements of the construction licence for the repository. Consequently, the investigation programme for the URL shall be drawn up to allow the collection of sufficient information and knowledge required for the application of the construction licence. The occupational and operational safety level must correspond to the safety level of the planned repository. Furthermore, visits to the URL shall be possible during and after its construction. The URL shall comply with the plans of the final


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disposal facility, e.g. compliance of URL underground structures with repository design to allow disposal tunnel characterisation, use of the horizontal and vertical deposition concepts, use of one-level and two-level repository layouts and the possibility of extension of the repository. The URL construction shall keep the natural conditions of the bedrock, and conditions otherwise favourable with respect to long-term safety, sufficiently unchanged in the repository area. It shall keep the environmental impacts small in order not to affect the natural surroundings considerably or change the local habitants’ living conditions. Different stages of investigation phases can be defined (Posiva, 2003b): .

.

.

.

First stage. Surface-based investigations before the construction of the access ramp/shaft starts. The main results are an improved description of the target rock volume and the access locations of the URL and the establishment of baseline conditions. Second stage. Construction of an access ramp and/or a shaft down to the planned repository depth accompanied by parallel investigations on the surface. The main products are monitoring responses of the geosphere to the construction activities, the improved characterisation of the target rock volumes and the completion of the detailed design of the URL. Third stage. Completion of the URL construction at the target depth, together with underground investigations, including site-specific tests of repository technology and experiments related to the long-term performance of the multi-barrier system. Fourth stage. The repository excavation stage. The underground characterisation of the site will continue to be an important activity during the operational stage, especially when designing and constructing the deposition tunnels.

4.3.2 Timing of URL development and the resources required Having a URL offers a lot of advantages, as has been discussed above, but it also carries a substantial commitment in terms of resources, both human and financial. Thus, the decision to develop one’s own URL has to be evaluated within the whole repository development programme. Construction costs for a URL may easily be on the order of 100 million euros and, once a URL is started, a significant portion of a disposal programme’s budget may be used to support it. The entire life cycle costs of the URL, including decommissioning, are significantly higher than these initial costs. Andersson (1999) reports that four European URLs spend between 5 and 11 million euros annually on research and development.


Underground research facilities and rock laboratories Table 4.4

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Participation and financing schemes

The entire URL – a joint programme

This has been the case with the Stripa project, where each of the partners had a ‘share’ of the whole project. The financial responsibility was according to the shares held and access was guaranteed to all results.

Joint programme This is a variation of the above scheme, whereby each partner – no fixed sharing has the same ‘voting’ rights but there is no fixed key of contribution. URL run by one organisation – participation fee

Under this scheme each partner organisation contributes directly to the operational costs of the URL and then additionally participates with personnel or resources in the experiments of interest.

URL run by one organisation – participation in a project

This is a variation of the participation scheme mentioned above. Here a partner organisation participates in or conducts its own project. The projects are the constituent elements of the URL and it is each project that contributes to the URL costs.

The timing of site-specific URL development strongly depends on the following points: . . . .

Are specific data needed that can only be obtained in a site-specific URL? Have all necessary data been collected before the system is disturbed? Have all technical, logistical and regulatory preconditions been met? Is the programme ready to demonstrate full capability to build a repository?

Table 4.4 shows the different financing models that have been followed so far for setting up cooperation projects in URLs. Note that a cooperative programme would, in most cases, be associated with generic URLs.

4.4

Public outreach and the role of URLs as training platforms

4.4.1 Public outreach In parallel with their main role as research facilities, URLs have also become established as important platforms for active interaction with the interested public and thus contribute significantly to the acceptance of the scientific and engineering work being performed in the area of geological disposal. By supplying direct information to the public from (and in) a ‘realistic’ environment, the feasibility of safe disposal can be illustrated and conveyed in a convincing manner. As part of the public outreach programmes, thousands of visitors are guided through the underground facilities every


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year and can form their own impression of the studies being carried out. They can look over the shoulders of the researchers and experience the engineering and scientific work at close hand.

4.4.2 URLs as training platforms URLs also play an important role in the dissemination of accumulated knowledge and experience to a broad spectrum of young scientists worldwide. The GTS, for example, serves as an internationally recognised training platform for future generations of experts in the management of radioactive waste. Most importantly, it offers its infrastructure and databases for educational purposes. The educational activities take into account the needs of implementers, regulators and research organisations. As members of IAEA Network Centres of Excellence, ‘Training in and Demonstration of Waste Disposal Technologies in Underground Research Facilities’, many URLs provide their facilities and experts with knowledge for theoretical and practical training in all aspects of underground research, engineering and related issues.

4.5

Case studies

4.5.1 In situ characterisation and testing (near-field and farfield processes) Tests with radionuclides colloid and radionuclide retardation (CRR) experiment Studies of natural colloids in deep groundwaters from widely ranging environments (e.g. Yucca Mountain, Wellenberg, Schwarzwald, Oklo, Cigar Lake, etc.) have been ongoing for several decades. In addition, laboratory programmes on colloid generation, stability and classical batch experiments studying radionuclide uptake on colloids (and the possibilities for subsequent release) have been reported and a significant database already exists. It is generally accepted that five requirements must be fulfilled to prove that colloid-facilitated transport of radionuclides in a potential repository host rock may be of significance to the long-term performance of a waste repository: . . . . .

Colloids must be present. Colloids must be mobile. Colloids must be stable under the given groundwater conditions (geochemical and hydrogeological environment). Radionuclide association with the colloids must take place. The association must be irreversible.



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4.13 Three-dimensional view of the CRR test site. The view direction is perpendicular to the experimental shear zone.

The colloid and radionuclide retardation (CRR, www.grimsel.com) experiment was dedicated to improving the understanding of the in situ retardation of safety-relevant actinides and fission products associated with bentonite colloids in the vicinity of the engineered barrier system (EBS)/host rock interface. In addition to a series of in situ dipole experiments that were carried out at the Grimsel Test Site (GTS), the project partners, namely ANDRA (F), ENRESA (E), FZK-INE (D), JNC (J), USDOE/Sandia (USA) and Nagra (CH), funded an extensive programme of laboratory and modelling investigations. The aims of CRR were: examination of the in situ migration of bentonite colloids in fractured rocks, investigation of the interactions between safety-relevant radionuclides and bentonite colloids in the laboratory and in situ and, in addition, the testing of the applicability of numerical codes for representing colloidmediated radionuclide transport. The starting point of the CRR experiment was at the interface between the bentonite buffer (EBS) and the geosphere. A radionuclide and bentonite colloid cocktail was injected into a dipole flow field of 2.2 m length and the breakthrough of radionuclides and bentonite colloids was monitored and compared (Fig. 4.13). The results of the CRR experiment can be summarised as follows: . .

Bentonite colloids were generated at the bentonite–host rock interface under laboratory conditions. High bentonite colloid concentrations (up to 1000 mg/l) remain stable


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. . .

Geological repository systems for safe disposal even over long timescales (months), showing very slow aggregation rates in the Grimsel groundwater. A high recovery of up to 90% of the injected bentonite colloids in the dipole flow field indicated high colloid mobility under the applied (short-term and fast-flow) experimental conditions. Batch experiments revealed a strong decrease in sorption to fracture infill and granite in the presence of colloids. The recovery of tri- and tetravalent actinides in the in situ experiment was increased from about 30% in the absence of to 60% in the presence of bentonite colloids. Colloidally transported radionuclides arrived slightly earlier at the extraction borehole than did conservative tracers.

The following conclusions can be drawn from the CRR project: .

.

.

The bentonite buffer is a potential source of artificial colloids within the EBS of a potential repository. To date, there is no in situ evidence for the formation and release of such colloids into the geosphere. Ongoing in situ experiments might give a first indication but, as they were not specifically designed for this purpose, these data might be biased by artefacts of the experimental layout. Bentonite colloids were found to be mobile within the tested shear zone. However, the migration of particles strongly depends on the internal structure of the shear zone and on the (applied) flow field. To date, there are no in situ experiments that were performed under realistic (i.e. repository-relevant) groundwater flow conditions. The radionuclide uptake of bentonite colloids has been proven for at least the tri- and tetravalent actinides, both with classical batch experiments in the laboratory and with in situ experiments. Within the experimental timescales, the colloidal association of these radionuclides appears to be irreversible.

Tests with radionuclides: long-term diffusion (LTD) It is well known that matrix diffusion is an important process in retarding radionuclides in fractured rocks. Rock matrix diffusion is particularly important when determining dose and risk calculations for weakly and nonsorbing radionuclides such as 129I and 14C. However, there have been few long-term field scale experiments to evaluate matrix diffusion of radionuclides in fractured rock with a minimal disturbance to the in situ condition. The long-term diffusion (LTD, www.grimsel.com) project is a series of experiments that aims to obtain quantitative information on matrix diffusion under in situ conditions. Partners and contributors are JAEA and AIST (2006–2008) of Japan, NRI and RAWRA of the Czech Republic,



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HYRL of Finland and Nagra and PSI of Switzerland. The project is divided into several work packages in which are: .

.

. .

An in situ monopole diffusion experiment where radionuclide tracers diffuse into undisturbed rock matrix with subsequent geochemical analysis of matrix samples combined with predictive and post-mortem modelling exercises to increase confidence in the modelling of long-term and large-scale diffusion processes. Characterisation of the pore space geometry (including determination of in situ porosity for comparison with laboratory-derived data) using 14Cdoped PMMA resin injection and NHC-9 chemical porosimetry techniques. A study of natural tracers in the rock matrix to elucidate evidence for long-term diffusion processes. Detailed characterisation of the flow paths in a water-conducting fracture and investigation of the in situ matrix diffusion paths in core material from earlier GTS experiments.

Circulation of a cocktail of sorbing, weakly sorbing and non-sorbing radionuclides (3H, 22Na, 131I and 134Cs) in the monopole was started in June 2007 and continued until August 2009 (Fig. 4.14). Water samples retrieved on a regular basis have been analysed at the Paul Scherrer Institute (PSI).

4.14 In situ monopole diffusion experiment where radionuclide tracers diffuse into the undisturbed rock matrix (borehole 8 m long, diameter 56 mm).


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Compared with predictive models carried out by four modelling teams, a much higher decrease in Cs was seen in the analysis of water samples than was predicted in the models. This was due to sorption on the borehole wall, and was confirmed after the borehole was overcored in November 2009. Analysis of overcored subsamples showed that the diffusion profiles of 22Na, stable I and 3H were more or less as predicted by the models. Overall the main conclusions so far from Phase 1 of the LTD project can be summarised as follows: . . . .

A stronger than expected Cs sorption was found on the borehole wall in the in situ diffusion experiment. A 10–20% decrease in porosity was found in samples measured in situ compared with samples measured in the laboratory. Evidence of unlimited connected porosity was obtained from natural tracer studies. Evidence of matrix diffusion retarding 237Np and 235U was found after only 60 days circulation in an advective fracture.

In the next phase of the LTD project, the monopole will be overcored in order to investigate the Cs sorption sites and concentrations of the radionuclides that have diffused into the rock matrix from the borehole. Future in situ experiments include a novel di-monopole experiment that allows in situ monitoring of matrix diffusion to be carried out and one to study diffusion from a shear zone using soluble, non-sorbing radionucide tracers is planned.

4.5.2 Engineered barriers Large-scale process testing full-scale engineered barrier system experiment (FEBEX) The FEBEX experiment at the Grimsel Test Site (www.grimsel.com) consists of an in situ full-scale engineered barrier system (EBS) test performed under natural conditions (Fig. 4.15). A ‘mock-up’ test, at almost full scale, runs in parallel in the laboratories at Ciemat in Madrid (a full description of the ‘in situ’ and the ‘mock-up’ test, covering the period 1994– 2004 is included in ENRESA, 2006, and references therein). The FEBEX experiment has been the research subject in three subsequent, European research projects: FEBEX, FEBEX-II and NF-PRO, under the leadership of ENRESA. Starting in 2008, a consortium of four partners (SKB, Ciemat, Posiva and Nagra) was brought together, which continues running the in situ experiment and the mock-up as part of Grimsel Phase VI under the name FEBEXe (e = extension) until 2012, when excavation of the second heater is planned.


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4.15

The current layout of the FEBEXe experiment at the GTS.



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The experiment is based on the Spanish reference concept for the disposal of radioactive waste in crystalline rock in which the canisters enclosing the conditioned waste are placed horizontally in drifts and surrounded by a clay barrier constructed of highly compacted bentonite blocks. With the start of the heating in 1997, the FEBEX experiment (heater 2 and EBS; see Fig. 4.1) is the longest running full-scale experiment. A constant temperature of 100 8C has been maintained at the heaters/bentonite contact during this time, while the bentonite buffer has been slowly hydrating with the water naturally coming from the rock. A total of 632 sensors of diverse types were installed in the clay barrier, the rock mass, the heaters and the service zone to measure the following variables: temperature, humidity, total pressure, displacement, water pressure, etc. Partial dismantling of the in situ test was carried out during 2002. After five years of heating, heater 1 was removed and the materials recovered (bentonite, metals, instruments, etc.) have been analysed to investigate the different types of processes undergone, while the second heater continued. The aim of the FEBEX experiment is to study the behaviour of the nearfield components (EBS, host rock) for a high-level radioactive waste repository in crystalline rock. The three main objectives are: . . .

To demonstrate the feasibility of handling and constructing an engineered barriers system. To study the thermohydromechanical (THM) processes in the near field. To study the thermohydrogeochemical (THG) processes in the near field.

As the experiment has now been running over more than 11 years, a unique dataset including crucial long-term observations in terms of sealing, saturation and corrosion has been generated. The current status is that the outer buffer is practically saturated and hydration is in progress at the inner buffer. This was predicted by THM models, but occurs at a slower pace than expected. Especially, one can observe that: . . .

Temperatures in the bottom part continue being higher than those in the sides and upper part for each section. Temperature trends are increasing very slowly in the buffer and in the rock close to the heater. The total pressure in general continues increasing into the buffer and also in the contact bentonite/rock and bentonite/plug. The water content in the buffer continues increasing at all points. The biggest differences can be observed in the inner part but it is clearly rising elsewhere.

Modelling efforts also continue within FEBEXe with the objective to:


Underground research facilities and rock laboratories . . .

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further elucidate the observed behaviour and the processes involved, assess how the results can be extrapolated to the near field in the long term and assess the transferability of the results to the national disposal concepts of the FEBEXe partners.

Large-scale process testing: gas migration test (GMT) The gas migration test (GMT, www.grimsel.com) (under the leadership of RWMC/Japan) has been a world premiere project, focusing on the migration of waste-generated gas – in the GMT case simulated by nitrogen – through the engineered barrier system (vent and bentonite/sand barrier) in a large-scale, in situ, realistic set-up (Fig. 4.16). The results of the field test are reported elsewhere (see, for example, Vomvoris et al., 2003, Shimura et al., 2006). Important conclusions, e.g. on the robustness of the engineered barriers to gas migration, have been reached and, more importantly, the experience gained with GMT has been used to design similar large-scale experiments elsewhere (e.g. the HG-A test at the FMT).

4.16 Gas migration test (GMT) at the GTS.


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4.5.3 Demonstration experiments Engineering studies and demonstration of repository design (ESDRED) The European research project ESDRED (www.esdred.info) comprises various experiments aiming at the demonstration of technological concepts for handling and disposal of radioactive waste. The long low-pH shotcrete plug in GTS is part of ESDRED Module 4 whose object is the construction and test of the low-pH shotcrete plug at full scale (Fig. 4.17). The aim of the test is to demonstrate the support capacity of such a plug under realistic conditions (Spillmann et al., 2009). This involves the swelling pressure of a bentonite buffer applied to one side of the plug. In order to demonstrate the applicability of the experimental concept a full-scale and fully monitored test of a low-pH shotcrete plug was carried out at the Grimsel Test Site (www.grimsel.com). To simulate real conditions of a filled repository, the plug is loaded with the swelling pressure of a bentonite buffer. The bentonite was provided with an artificial hydration system (vertical mats) to accelerate the saturation process and to impose a pore water pressure in the bentonite buffer. The cementitious materials used in repositories plugs must be designed to support either the rock of the underground works or mechanically a swelling material seal. The concrete used for these two applications must comply with two main requirements:

4.17 Layout of the ESDRED shotcrete plug including the monitoring components of the TEM project at the GTS.


Underground research facilities and rock laboratories . .

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The pH has to be as low as possible (below 11) in order to minimise chemical interaction with the bentonite or with the surrounding geological medium. The concrete has to be emplaced by the shotcreting technique.

Suitable concrete formulations have been developed and a short low-pH shotcrete plug was tested at the A¨spo¨ URL (Sweden). The full-sized bentonite and shotcrete plug in the Grimsel Test Site was completed in March 2007. The plug is monitored by a network of hard-wired sensors that measure total pressure, water pressure and water content. Test and evaluation of monitoring (TEM) techniques The ESDRED plug experiment also provides the opportunity to test and evaluate different monitoring methods (TEM project, www.grimsel.com) in a situation that would have similarities with a repository vault-end seal. Six additional sensors constitute the wireless monitoring network. The data from the wireless monitoring sensors (two pore pressure, two total pressure and two water content sensors) are temporarily stored on a battery powered data logger. A magneto-inductive transmitter at the tunnel face and a receiver in the access tunnel provide the wireless data transmission. To protect the standard data logger and magneto-inductive system from the expected high swelling pressures of the bentonite (up to 5 MPa), they were installed in small boreholes. The data are then transmitted by MI antennas to the data acquisition unit located in the tunnel. Non-intrusive monitoring techniques have the advantage that the technical barrier and surrounding excavation damage zone are not disturbed by the monitoring system. A regular fan of six monitoring boreholes was drilled to evaluate the seismic tomography as a non-intrusive monitoring technique. A seismic source and receivers explore the area between the boreholes. Repeat measurements allow investigating changes in the water content or pressure during the saturation process.

4.5.4 Hydromechanical evolution of backfill and host rock Gas path through host rock and along seal sections: HG-A experiment The excavation damaged zones (EDZs) around cylindrical excavations, such as sealed sections in tunnels or shafts, have been the subject of extensive experimental and theoretical investigations in the field of underground waste disposal. Conceptual and numerical models have been developed to describe the hydromechanical processes associated with the creation and evolution of damaged zones during the operational phase and after backfilling of the underground structures. However, only limited data


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from tunnel-scale experiments are available for a comprehensive validation of the EDZ models. The HG-A experiment is a large-scale experiment focusing on hydromechanical processes as part of a long-term geoscientific research programme at the Mont Terri Underground Rock Laboratory located in the Opalinus Clay formation in the Jura Mountains of Switzerland and operated by swisstopo/Swiss Geological Survey (www. mont-terri.ch). A horizontal microtunnel with a diameter of 1 m and a length of 13 m was drilled in a claystone formation. Monitoring was carried out before, during and after tunnel excavation. After installing monitoring instruments in the open tunnel, the end of the tunnel was backfilled with sand (test interval) and a large hydraulic packer was emplaced in the seal section. The packer was inflated and subsequently the test interval was saturated with synthetic pore water (a key reference for this case study is Marschall et al., 2008). Figure 4.5 shows the layout of the Mont Terri URL with the HG-A experimental site located in a niche of Gallery 04. The specific aims of the HG-A experiment are to: . . .

provide evidence for barrier function of the Opalinus Clay on the tunnel scale, investigate mechanical self-sealing of EDZ after tunnel closure (packer inflation) and provide evidence for a gas transport capacity of Opalinus Clay for both the intact host rock and EDZ.

The microtunnel was excavated during February 2005 using a steel augur from a niche in Gallery 04. The microtunnel was excavated parallel to a bedding strike, so that bedding runs along the tunnel. This was done to imitate the expected relationship between bedding and emplacement tunnel orientation in a deep repository in Northern Switzerland, where bedding is expected to be almost horizontal and emplacement tunnels to be subhorizontal (Nagra, 2002). The experimental layout is shown in Fig. 4.18. Excavation was monitored using piezometers and deformation gauges installed in boreholes. This was subsequently supplemented with additional piezometer boreholes and borehole stressmeters (Fig. 4.19). The first 6 m of the microtunnel was lined with a steel casing immediately after excavation to stabilize the opening. The gap behind the liner was then cement-grouted, but not sealed. The next 3 m from 6–9 m along the tunnel forms the seal section where a purpose-built hydraulic megapacker (940 mm in diameter and 3000 mm in length) was installed. The final 4 m of the microtunnel is the test section. This section was instrumented with piezometers, extensometers, strain gauges and time domain reflectometers (TDRs) to measure pressure, deformation and water content. After instrumentation the test section was filled with sand behind a retaining wall. The seal section was instrumented with piezometers, total pressure cells



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4.18 Gas path through the host rock and along seal sections: the HG-A experiment layout.

4.19 Schematic drawing of the microtunnel and the site instrumentation. Grey zone coding refers to the steel liner (black), the seal section (grey in the middle part) and the backfilled test section (grey part at the end).

and TDRs prior to the installation of the megapacker. Following the installation of the megapacker the volume between the retaining wall and the megapacker was filled with a cement grout. The post-excavation experimental sequence is summarised in Fig. 4.20. The test and sealing sections were instrumented during 2006 and the megapacker was emplaced and first inflated in November 2006 (Phase 2).


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4.20

Post-excavation experimental sequence.

During 2007 a series of water injection tests were performed to test the performance of the flow control units and the long-term monitoring system. A year-long multi-rate water injection was started in January 2008 and is still ongoing. At this point, the experiment is approaching the end of Phase 3. It is anticipated that Phase 4, gas injection, will be started during 2010 and Phase 5, the post-gas hydraulic testing, will start in 2012. Following Phase 5 it is planned to excavate and perform post-mortem characterisation of the test and sealing sections. The following preliminary conclusions can be drawn from the HG-A project (Marschall et al., 2008): 1.

Hydromechanical processes. The responses to injection and changes in megapacker pressure indicate a tightly coupled system where imposed test section pressure changes cause deformation of the test and sealing section, while stress changes in the seal section lead to deformation of the test section and consequent pressure changes. In addition to these short-term responses the measurements indicate ongoing deformation of the test section, presumably resulting from the response to far-field stresses and possible softening of the rock due to resaturation. The coupled nature of the responses together with the ongoing transient



2.

3.

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responses to excavation and subsequent resaturation make interpretation complex. Flowpaths and self-sealing. The hydraulic response suggests that flow paths are dominated by the EDZ around the tunnel and that the intact host rock represents a significant low permeability barrier. Initial interpretations and models suggest that the total conductance of the EDZ is reducing with time. It is assumed that this reduction relates to reconsolidation and swelling of the rock, gel creation and sealing of fractures within the EDZ. The influence of the megapacker pressure on EDZ properties is harder to determine and will require more detailed consideration of the coupled responses seen during the low and high effective stress periods. Barrier function of the undisturbed rock. The low diffusivity responses to saturation and pressurisation of the test section that have been observed in the monitoring boreholes support the current estimates of the diffusivity of the intact Opalinus Clay. To date, the uncertainty in the storativity prevents any better estimate of metre-scale hydraulic conductivity.

Future trends

Experience from more than 30 years has shown that the investigation and testing programmes in Underground Research Laboratories form an important part of the overall waste management strategies of the various national programmes. URL projects contribute in many different ways to the geological disposal programme and their role evolves with time, stage of the programme, state of know-how and technology. Broad international experience exists to design and implement a URL programme in a fast and economic way. In the last few decades, many of the established URLs emerged as internationally well-known research laboratories in the area of safe disposal of radioactive waste in deep geological repositories. They are characterised by a broad spectrum of scientific and engineering projects, close international collaboration and know-how exchange and active, transparent dialogue with the public and their elected representatives.

4.6.1 What has been obtained from URLs . .

Development of methods and equipment for underground characterisation and testing of the reliability of the different methods. Determination of reliability of surface-based methods of site characterisation.


116 . . . . . . .

Geological repository systems for safe disposal Application of site-exploration strategies and strategies to adapt underground systems as more information is acquired. Testing and development of conceptual and numerical models of processes potentially relevant to radionuclide transport through rock. Quantification of impacts of excavation on the local system. Further development and assessment of excavation techniques. Simulation of effects caused by emplacement of radioactive waste. Demonstration of engineered-barrier systems (feasibility). Demonstration of repository engineering on a large scale.

4.6.2 Future needs Within a waste disposal programme, URLs have a direct technical function and also a confidence-building role. In advanced programmes, there is a trend towards large-scale, realistic, integrated experiments in which a number of interacting components and/or processes are simultaneously studied. As fundamental scientific problems move towards solution and the implementation of repositories draws closer, the confidence building role grows in importance. Large-scale, long-term, integrated experiments play a key role in raising technical and public confidence. Full- or large-scale experiments performed to date have highlighted potential optimisation areas (feedback to Nagra’s programme, for example, leading to selection of a combination of bentonite blocks and granular bentonite as buffer). The timing of these experiments should be determined considering the overall geological disposal programme and the steps required leading up to the submission of the licence application. In emerging programmes, the need for know-how development in some of the fundamental issues will exist, but whether this need is satisfied with a new URL or being involved in cooperative projects in existing URLs is an issue that has to be addressed by each programme individually. Visits to URLs represent not only an occasion for gathering visual impressions but also an opportunity for in-depth discussions between technical experts and decision makers. URLs also provide an excellent focus for mutually beneficial international cooperative efforts. In any cost–benefit analysis of the value of a URL in a national programme, the purely scientific/technical contributions of the facility must not be considered in isolation; the less direct or intangible benefits associated with raising of technical and public confidence in the competence of the implementer must also be weighed. Finally, URLs provide a training platform for the transfer of knowledge and experience, a role gaining increasingly in importance.



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References

Andersson J (1999), ‘A study of the co-operation of research in underground facilities within the EU on aspects of disposal of radioactive waste’, European Commission, Brussels. Blu¨mling P and Adams J (2008), ‘Grimsel Test Site – Investigation Phase IV: borehole sealing, Nagra Technical Report NTB 07-01, Nagra, Wettingen, Switzerland. Bossart P and Thury M (2008), ‘Mont Terri Rock Laboratory. Project, Programme 1996 to 2007 and Results’, Report Swiss Geological Survey 3, swisstopo, 3084 Wabern, Switzerland. Coons A, Bergstro¨m P, Gnirk et al. (1987), ‘State-of-the-art report on potentially useful materials for sealing nuclear waste repositories’, Nagra Technical Report NTB 87-33, Nagra, Wettingen, Switzerland. ENRESA (2006), ‘Full-scale engineered barriers experiment: updated final report (1994–2004)’, Transedit, ISSN1134-380X. IAEA (2001), ‘The use of scientific and technical results obtained from underground research laboratory investigations for the geological disposal of radioactive waste’, IAEATECDOC-1243. Kickmaier W (2002), ‘The role of rock laboratories’, Bulletin 34, pp. 4–9, Nagra, Wettingen, Switzerland. McCombie C and Kickmaier W (2000), ‘Underground research laboratories: their roles in demonstrating repository concepts and communicating with the public’, in Euradwaste 1999: Radioactive Waste Management Strategies and Issues: Fifth European Commission Conference on Radioactive Waste Management and Disposal and Decommissioning, edited by C Davies, Luxembourg, 15–18 November 1999, EUR 19143 EN, pp. 274–281. Marschall P, Trick T, Lanyon G W, Delay J and Shao H (2008), ‘Hydro-mechanical evolution of damaged zones around a microtunnel in a claystone formation of the Swiss Jura Mountains’, ARMA, San Francisco, California. Nagra (1988), ‘Untersuchungen zur Standorteignung im Hinblick auf die Endlagerung schwach- und mittelaktiver Abfa¨lle: Berichterstattung u¨ber die Untersuchungen der Phase 1 am potentiellen Standort Piz Pian Grand (Gemeinden Mesocco und Rossa, GR)’, Nagra Technical Report NTB 88-19. Nagra, Wettingen, Switzerland. Nagra, (2002), ‘Projekt Opalinuston – Synthese der geo-wissenschaftlichen Untersuchungsergebnisse. Entsor-gungsnachweis fu¨r abgebrannte Brennelemente, verglaste Hochaktive sowie Langlebige mittelaktive Abfa¨lle’, Nagra Technical Report NTB 02-03, Nagra, Wettingen, Switzerland. NEA (2001a), ‘The role of underground laboratories in nuclear waste disposal programme’, OECD/Nuclear Energy Agency, Paris. NEA (2001b), ‘Radioactive Waste Management Committee: going underground for testing, characterisation and demonstration (a technical position paper)’, 6/ REV, OECD/Nuclear Energy Agency (OECD/NEA), Paris. Niva B, Olsson O and Blu¨mling P (1988), ‘Grimsel Test Site – radar crosshole tomography with application to migration of saline tracer through fracture zones’, Nagra Technical Report NTB 88-31, Nagra, Wettingen, Switzerland.


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Posiva Oy (2003a), ‘ONKALO: underground rock characterisation facility – main drawing stage’, Posiva Working Report 2003-26, Posiva Oy, Helsinki. Posiva Oy (2003b), ‘ONKALO: underground characterisation and research programme (UCRP)’, Posiva Report 2003-03, Posiva Oy, Helsinki. Shimura T, Fujiwara A, Vomvoris S, Marschall P, Lanyon B, Ando K and Yamamoto S ‘Large scale gas migration test at Grimsel Test Site’, in Proceedings of the 11th International High-Level Radioactive Waste Management Conference, 30 April–4 May 2006, Las Vegas, Nevada, ANS/ HLRWM, p. 4. SKB (1996) A¨spo¨ Hard Rock Laboratory, 10 Years of Research, SKB, Stockholm, Sweden.. Spillmann T, Fries T, Wetzig V, Koch AS, Garcia-Sin˜eriz J-L, Ba´rcena I (2009), Beton mit tiefer Alkalita¨t: Demonstrationsversuche im Felslabor Grimsel: Stollenverschluss aus niedrig-pH Spritzbeton, Tec21 SIA/21, pp. 17–19. Vomvoris S, Lanyon B, Marschall P, Ando K, Adachi T, Fujiwara A and Yamamoto S (2003), ‘Sand/bentonite barriers and gas migration: the GMT large-scale in-situ test in the Grimsel Test Site’, in Materials Research Society Proceedings on Scientific Basis for Nuclear Waste Management XXVI, Vol. 757, Materials Research Society, Warrendale, Pennsylvania. Websites of the various URL projects and http://www.ha.nea.fr/html/rwm/docs/ 2001/index.html.


5 Crystalline geological repository systems: characterisation, site surveying and construction technologies and techniques A . J . H O O P E R , Alan Hooper Consulting Ltd, UK

Abstract: This chapter describes the technologies and techniques that are available or are being developed for the characterisation of crystalline sites, where the repository host rock is a strong igneous or metamorphic rock that is water saturated. Individual sections are presented on each of the categories of geoscientific information, namely lithologies, geological structure, rock mechanics and geotechnical properties, hydrogeology, geochemistry and radionuclide transport properties, which typically are sought and then must be integrated in a site characterisation programme. Further sections consider respectively: disturbance by excavation and waste emplacement, the stability of crystalline rocks in the context of waste disposal and the feasibility of construction in such rocks. Key words: deterministic and stochastic treatment as a function of length scale, spatial variability of flow, geochemical and transport properties, fractures and fracture networks, borehole drilling and testing, development and testing of conceptual models.

5.1

Introduction

This chapter describes the technologies and techniques that are available or are being developed for the characterisation of crystalline sites. In the context of geological disposal of radioactive wastes, the term crystalline is used to describe strong igneous or metamorphic rocks such as granites or consolidated tuffaceous rocks. Such rocks typically have low porosities and groundwater is present predominantly in discontinuities, or fractures, within the rock matrix. Much attention is devoted to the characterisation of the fracture systems in crystalline rocks since it is these that will provide the

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principal pathway for any groundwater flow and radionuclide transport away from the repository. Crystalline rocks have many properties that make them suitable host rocks for an underground disposal system located hundreds of metres below ground. In some countries, such as Sweden and Finland, the crystalline rocks that have been investigated persist from the prospective repository depth all the way to the ground surface, whereas in other countries they are typically overlain by sedimentary rocks whose characteristics must also be understood in order to develop the necessary designs and safety assessments, as discussed in Chapter 6. This provides a clear-cut example of the necessity to ensure that the site characterisation techniques and technologies that are selected for use are appropriate to the geological and hydrogeological conditions at a site. Sweden and Finland have two of the most advanced programmes in the world dealing with the geological disposal of spent nuclear fuel and both have identified prospective sites for locating a repository in the crystalline bedrock (SKB, 2008; Andersson et al., 2007). Significant contributions have been, and continue to be, made to the development of technologies and techniques for characterising crystalline sites in these countries. The spent fuel and/or high-level waste disposal programmes of other countries, including Canada (AECL, 1994), France (Andra, 2005) and Switzerland (Nagra, 1994), have made notable contributions also, even when at the stage of testing the feasibility of disposal in crystalline rock, as did investigations in the UK of a potential crystalline host rock, overlain by sedimentary rocks, for the disposal of predominantly long-lived intermediate-level wastes arising from the reprocessing of spent fuel (Nirex, 1997). The bulk of the chapter is taken up with a description of the technologies and techniques that can be used to characterise key aspects of crystalline sites to provide the sort of information required to develop repository designs, safety assessments and the necessary geological understanding. This analysis is strongly informed by the published experiences of national radioactive waste management programmes such as those listed above, which in turn have used the experience of other industries such as mineral extraction, geotechnical engineering, geothermal energy, water resource exploration and, in some cases, hydrocarbon exploration. Individual sections are presented on each of the categories of geoscientific information, namely lithologies, geological structure, rock mechanics and geotechnical properties, hydrogeology, geochemistry and radionuclide transport properties, which typically are sought in a site characterisation programme. However, an important point to note is that each category of information must be related to others so that a consistent picture of the site emerges rather than a collection of separately obtained findings that will not give the necessary overall understanding of a site. These sections on techniques and


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technologies are followed by shorter sections that consider, respectively, disturbance by excavation and waste emplacement, the stability of crystalline rocks in the context of waste disposal and the feasibility of construction in such rocks. This is followed by a short discussion of possible future trends in the field of characterising prospective repository sites in crystalline rocks.

5.2

Lithologies

The classical starting point for the description of a site is the distribution of different rock types, involving the identification of rock units on the basis of properties that are potentially relevant to aspects of the repository design and/or safety assessment.

5.2.1 Geophysical surveys The location of the site itself may be defined as a result of regional geophysical and mapping surveys that enable the identification of a suitably sized volume of the potential host rock. The relevant geophysical surveys at this stage are often conducted as low-level aerial surveys and include the following: . . .

Magnetic surveys that determine the distinctive magnetic properties of different rock types. Radiometric surveys that determine the distinctive radionuclide content of different rock types. Electromagnetic surveys that determine the distinctive conductivities of different rock types.

These surveys are very useful in providing information on the larger-scale structural elements, such as lineaments, within the rock mass. Such large structural elements are often used to define the boundaries within which the prospective repository site will be located. Gravity surveys are also useful at the early, reconnaissance, stage of site characterisation to identify the different rock types on the basis of distinctive gravitational properties. However, this geophysical technique is not well suited to aerial surveying and is most often used in manual surveys conducted on the ground surface.

5.2.2 Surface mapping Geophysical surveys are especially valuable when there is no outcrop of the crystalline rock because it is overlain by sedimentary cover, or perhaps relatively thick Quaternary deposits (soils and gravels deposited over the last 1.5 million years or so), or when the area of interest is below the sea or some


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other large body of water. However, geological mapping of outcrops at the surface or of exposures in trenches, when available, provides much useful information, particularly when combined with the current state of development of electron microscopy and analytical techniques. The identification of the mineralogical and chemical compositions and the microstructural grain size in outcropping rocks provides an important input to the subdivision of the crystalline rock mass into different rock units.

5.2.3 Borehole drilling Boreholes are drilled in order to obtain similarly detailed information on the rocks occurring at depth. Two distinct types of drilling are used in repository site investigation: core drilling and percussion drilling. In core drilling, intact drill cores are taken out of the borehole. In suitable geological conditions this technique can be used throughout the length of a borehole and there are many examples of a continuous core record being obtained over depth intervals of more than one thousand metres. The drill cores are examined in detail, much as in the case of outcrop mapping, and typically samples are also taken for other sorts of analyses. In percussion drilling the rock is crushed in the borehole and the investigations to identify the lithological properties are done on the crushed material when it is removed from the borehole. A number of geophysical tools have been developed for use in boreholes and some of these can complement the sampling afforded by abstraction of drill cores or crushed rock. In relatively simple rocks, the density/photoelectric factor (PEF) and neutron tools that are used to measure rock porosity will also resolve the dominant rock type. Other lithology-sensitive measurement techniques, such as the elemental capture spectroscopy (primarily measuring silicon, calcium, iron and sulphur) and natural gamma-ray tool (measuring thorium, uranium and potassium), can be used if the rock is more complex.

5.2.4 Role of rock units The identification of different rock units as subdivisions in the crystalline host rock is valuable in establishing the spatial distribution of key physical and chemical properties within the rock volume. For example, in the work carried out by SKB, the Swedish Nuclear Fuel and Waste Management Company, to characterise the candidate repository sites at Forsmark and Laxemar, rock units are identified to be associated with certain fracture network characteristics, such as fracture density and permeability, or thermal conductivities, resulting from the precise mineral content (Stephens et al., 2007; Wahlgren et al., 2008). Such properties are important in



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evaluating the mechanical and/or hydrogeological properties of the rock mass and in developing repository designs.

5.3

Geological structure

There are strong relationships between the rock types and the aspects of geological structure that are important to repository development, so many of the techniques and technologies discussed above are valuable in characterising the structure. The structure of a prospective repository site, placed within a regional context, can be seen as the basis for developing ideas on the location of the repository and for influencing other elements of the site characterisation programme. The focus of attention in respect of a crystalline site is a description of the structure in terms of the discontinuities in the rock mass, since these will determine the mechanical and hydrogeological properties that are important in determining the safety of the repository and should control decisions on design in order to achieve safety.

5.3.1 Length scales A fairly standard approach has developed whereby discontinuities in the rock mass are classified in terms of their different length scales and are then subject to different means of characterisation according to that classification. The term deformation zone is used to describe extensive discontinuities that are likely to include large numbers of individual fractures. These deformation zones are subdivided in terms of the likely implication that their size will have for repository safety and design. At the top end of the length scale regional deformation zones typically extend over tens of kilometres, and, if present, are likely to represent an important boundary for the site investigations. Major deformation zones more local to the prospective repository site may extend over a length of the order of one kilometre. At a potentially suitable repository site there will be relatively few major deformation zones such that these are typically characterised deterministically in site characterisation programmes conducted to date. The classification of minor deformation zones refers to a length scale of a few hundred metres where the lengths of individual fractures also can be of this magnitude. In most crystalline rocks that have been investigated to date, the number of deformation zones in this category within the prospective repository volume has been such that they cannot be investigated individually and a geostatistical method is required to characterise their properties. Finally, there are individual fractures that may have length scales on the order of ten metres.


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5.3.2 Characterisation of deformation zones As already noted, surface mapping and aeromagnetic surveying provide good information on the regional deformation zones and these zones may also have clear signatures in the topography of the region. Given their likely significance for civil engineering projects and resource exploitation, e.g. mineral extraction, there will often be good pre-existing geoscientific information. Major deformation zones more local to the prospective repository site are also readily identified by the same methods, but greater precision is required in defining their location at depth so borehole drilling is often used in conjunction with surface-based information to locate these deformation zones, at depth. Particularly for steeply dipping deformation zones, the use of inclined or deviated boreholes can prove invaluable. More detailed investigations are required in order to characterise the frequency and spatial distribution of minor deformation zones. Recent work in Sweden has shown the value of magnetic surveys conducted manually from the ground surface, which are more sensitive than the previously conducted aerial surveys (Triumf, 2007). The identification of magnetic anomalies is used in conjunction with detailed surface mapping of fracture sets to identify the distribution of the minor deformation zones within the surveyed area. Boreholes can then be drilled to intersect the deformation zone indicated from the survey, as in the case for the major deformation zones. For gently dipping deformation zones, advances in the technology of surface-based seismic reflection surveys have proved successful in identifying minor deformation zones in recent years. This represents a potentially powerful technique when a vertical seismic profiling (VSP) tool can be used to pick up the seismic signature in a borehole drilled subsequently to intersect the deformation zone (Juhlin, 2007).

5.3.3 Characterisation of fractures and fracture networks Information on frequency, orientation and dimensions is required to understand the rock mechanical and hydrogeological significance of fractures and fracture networks. This information is obtained by examination of fractures retrieved in drillcore and by obtaining images of the borehole wall. Special techniques have been developed to correctly orientate the drill core in three-dimensional space and hence to correlate the fracture being examined in the retrieved core with the image of the fracture at depth in the borehole wall. Geophysical and hydrogeological data obtained from the borehole are also used to characterise the fractures and valuable evidence on the fracture type can be obtained from the mineralogical signature of any fracture filling that is observed. This information on the fractures and fracture networks can then be used to identify domains



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occurring between the deterministically defined deformation zones within the rock mass on the basis of the predominating fracture characteristics.

5.3.4 Discrete fracture network models The geostatistical information on fractures and fracture networks required by other disciplines is typically presented as a discrete fracture network (DFN) model (e.g. Munier, 2004). Typically the DFN will comprise, at a minimum, the following components for each fracture set identified on the basis of its orientation: . . .

5.4

The orientation of the set. The size distribution of fractures in the set (often expressed in terms of areal dimensions). The amount, or intensity, of fracturing (e.g. areal density per unit volume of rock).

Rock mechanics and geotechnical properties

Information is required on the mechanical properties of the rock mass in which the waste deposition will occur and through which access tunnels and/ or shafts will be constructed. In particular, it is required to understand the strength of the rock, its deformation properties under loading and the in situ stress regime. Given the occurrence of deformation zones and fractures in many crystalline rocks, this information is required on each of the main structural elements within the rock mass and has to be in a form that allows analysis of the response of the overall rock mass at a range of different length scales according to the element of repository design that is of interest.

5.4.1 Mechanical properties The strength properties, such as uniaxial compressive strength and tensile strength, of the intact rock not intersected by fractures can be determined by a suite of well-known laboratory tests on drillcore samples. Similarly, the mechanical properties of fractures in the rock can be determined by carrying out tilt tests and direct shear tests on drillcore samples containing a fracture. The deformation properties of the intact rock, classically expressed in terms of Young’s modulus and Poisson’s ratio, can be obtained from the results of uniaxial and triaxial compression tests on drillcore samples. The results from the laboratory tests are typically used in empirically based rock mass classification systems to determine the rock mass deformation and strength properties that are relevant to repository design and to determining the response to mechanical or thermal loading. The


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empirically based rock mass classification systems draw on a large database generated from geotechnical engineering projects and are widely used in underground tunnelling and civil engineering works in other industries. There is an increasing trend to use theoretically based numerical modelling to determine the relevant rock mass properties and such an approach has been used alongside the empirical approach in SKB’s site characterisation programme, for example (Glamheden et al., 2007).

5.4.2 In situ stress The in situ stress can be determined by using various techniques. It is resolved into maximum horizontal stress, minimum horizontal stress and vertical stress, so that both the magnitudes of these and the orientation of the maximum and minimum stress direction are required. Commonly applied methods include overcoring, where a small diameter borehole is drilled at the base of a cored borehole. Strain gauges are fixed to the wall of this borehole and the instrumented rock is then retrieved to the surface by overcoring, allowing the direct measurement of the stress. Techniques involving hydraulic pressure are also commonly used to measure in situ stress where the hydraulic pressure on an identified fracture in a borehole wall is increased and decreased to determine the pressure at which it just opens and closes. Hydraulic fracturing is also used where the hydraulic pressure in a section of a borehole is increased until fracturing of the rock is induced.

5.5

Hydrogeology

Information on the hydrogeology of a potential repository site is required to compile a hydrogeological description on a regional and local scale that is sufficiently detailed to support judgements on the suitability of the site and to meet the information needs of repository design, safety assessment and environmental impact assessment. The hydrogeological description is universally presented as a conceptual model and the hydrogeological information obtained in site characterisation is aimed at supporting the definitions of boundary conditions for the model and of the initial conditions and justifying the hydraulic parameter values that are assigned to various components of the model. Figure 5.1 provides a typical schematic illustration of the major features of a groundwater flow system, in this example showing the SW–NE cross-section of the regional groundwater flow system affecting the Borrowdale Volcanic Group rocks at Sellafield as determined by Nirex (1997).



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5.1 A schematic illustration of the major features of the groundwater flow system in the region around the former potential repository zone (PRZ) at Sellafield (copyright United Kingdom Nuclear Decommissioning Authority; reproduced by kind permission).

5.5.1 Hydraulic parameters Much of the information on the hydrogeology of a site comes from the programme of testing that is carried out in the various boreholes drilled at the site. Tests can be carried out in single boreholes or can involve cross-hole or interference tests between two or more boreholes. The characteristics of interest primarily concern the derivation of hydraulic parameters and include measures of environmental pressure, environmental and freshwater heads, hydraulic conductivity, permeability, transmissivity, storativity, specific storage, hydraulic diffusivity and measures of connectivity. Various methods are used for the interpretation of each identified parameter, e.g. by direct measurement (associated with the respective testing or data acquisition technique), through preliminary interpretation or through modelling. A range of hydrogeological testing methods are available which themselves yield different types of information. The methods identified in a recently conducted review (Golder Associates, 2006) are: . . . . . . . .

pumping tests conducted in open boreholes without packers; interference measurements conducted in open observation boreholes; single-hole packer tests; interference tests conducted in observation boreholes equipped with multi-packer systems; fluid logging; flow logging; differential flow logging; and long-term monitoring of pressure in boreholes equipped with multi-level piezometers or multi-packer systems.


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5.5.2 Hydrogeological conceptual model It is recognised as of great importance to relate the targets and design of the hydrogeological testing programme to the development of the hydrogeological conceptual model of the site. Particularly at the early stages of a site characterisation programme there will be the potential for viable alternative conceptual models. An important role of the testing programme is to enable discrimination, where possible, between such alternatives and to identify the conceptual and parametric uncertainties associated with the model that emerges from this process as having the best justification. The structural building blocks of the hydrogeological conceptual model(s) are the hydrogeological units, which are recognised on the basis of their hydraulic properties and initial condition. An important element in the design of a hydrogeological testing programme is a consideration of how hydrogeological units will be defined and which part of the system will be treated deterministically and which stochastically. As noted in the discussion on structure, a fractured crystalline rock formation will probably be assumed to have stochastically distributed properties whereas major deformation zones are likely to be represented deterministically in terms of location and size. However, within such deterministically represented features, the hydraulic parameters may be represented as constant or stochastically distributed. In many cases, it may be uncertain at the outset how the various hydrogeological units in the system are to be treated for modelling purposes and the initial programme of investigations and

5.2 Discrimination between deterministic and stochastic treatment of hydrogeological features of different length scales as applied by United Kingdom Nirex Limited in its investigation of the Borrowdale Volcanic Group basement rocks (copyright United Kingdom Nuclear Decommissioning Authority; reproduced by kind permission).



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hydrogeological testing will be designed to resolve this. Figure 5.2 shows how discrimination between the deterministic and stochastic treatment of hydrogeological features was applied by United Kingdom Nirex Limited in its investigation of the Borrowdale Volcanic Group basement rocks.

5.5.3 Hydrogeological measurements Single-hole tests are mainly conducted to derive the hydraulic parameter transmissivity. In rock formations where the flow geometry is less complex the derived transmissivity could be used to derive the hydraulic conductivity of the tested formation. In fractured rock formations the derived transmissivity is representative of only a small proportion of the test interval and in most cases it would not be clear how many fractures and what totally hydraulically effective aperture could be attributed to the transmissivity. In fractured rock masses, it is essential to recognise the importance of scale in relating the results of hydrogeological testing to the large-scale behaviour of hydrogeological units. The further inputs required to construct a hydrogeological model in fractured crystalline rock can be established from a combination of tests, typically as follows: . .

.

Interference tests give information on the connectivity of the system and its relation to scale and direction. Fluid and flow logging measurements use fluid conductivity, temperature or flow in open holes under static or pumping conditions to seek to identify discrete features (fractures) that are hydrogeologically active and thus enable derivation of the flowing fracture frequency. Long-term discrete pressure measurements in boreholes are used for deriving the static pressure and equivalent freshwater head of the isolated section and their natural (or induced) fluctuations. The monitoring installations are also used as observation points during interference tests.

Characterising fractures and fracture networks The upscaled hydrogeological properties for the units that are modelled stochastically are usually determined through the use of hydrogeological discrete fracture network (DFN) models, where similar principles apply as described in the discussion on the use of DFN models in the characterisation of geological structure. Figure 5.3 shows a realisation of the fracture network within a 2-km-sided cube of the Borrowdale Volcanic Group produced from a discrete fracture network model. In this figure flowing features are represented as planar fractures and each fracture is tessellated


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5.3 A realisation of the flowing features within a 2-km-sided cube of the Borrowdale Volcanic Group rocks produced from a discrete fracture network model (copyright United Kingdom Nuclear Decommissioning Authority; reproduced by kind permission).

into domains with sides of approximately 40 m, within which the transmissivity is taken to be constant. It is particularly important in constructing a hydrogeological DFN to be able to characterise the fracture population that is capable of carrying groundwater flow. The borehole flow logging tool developed by Posiva in Finland has proved very effective in characterising individual fractures that are conducting flow at the point where they are intersected in a monitored borehole (O¨hberg and Rouhiainen, 2000). However, other fractures may also be capable of conducting flow; the concept of potentially flowing fractures has commonly been used whereby fractures that are identified from drillcore logging and/or borehole imaging and geophysical measurements as having the characteristics of a flowing fracture (e.g. open aperture, mineralogical evidence) are also taken into consideration. Figure 5.4 shows how a model domain can be constructed to simulate groundwater flow in the fracture network of the crystalline host rock of a repository. The figure is taken from a simulation of the volume of the Borrowdale Volcanic Group of rocks bounded by major



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5.4 The model domain used for the calculations of groundwater flow through the fracture network in the Borrowdale Volcanic Group rocks at Sellafield (copyright United Kingdom Nuclear Decommissioning Authority; reproduced by kind permission).

deformation zones having approximate dimensions of 3 km62 km6750 m (Nirex, 1997).

5.5.4 Testing the conceptual model Good scientific practice has been developed over a number of years involving the testing of the hydrogeological model that results from the process outlined here against other information obtained from site investigations that has not been used in the development of the model. Various types of information have been used depending on the nature of the prospective repository site, and in particular whether the crystalline rock is overlain by sedimentary formations. In general the types of information used have included the following: . . . .

5.6

conservative tracer tests; groundwater level responses in boreholes to large-scale drawdowns, induced in interference tests, for example; surface and near-surface groundwater levels and discharges; and hydrochemistry of water sampled from fractures and from the rock matrix.

Geochemistry

Geochemistry is taken here to mean the chemical characteristics of: .

the groundwater in the fracture network in the crystalline rock;


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Geological repository systems for safe disposal the pore water in the crystalline rock matrix; and the minerals formed on the surface of fractures (including secondary minerals such as calcite).

It includes the associated media such as colloids, microbes and gases that, with the groundwater and rocks, make up the present subsurface environment at the prospective repository site. It is important to understand the geochemistry for a number of reasons, and these dictate the sampling and analysis that is undertaken. Obtaining good-quality groundwater samples that are representative of the undisturbed conditions at depth is notoriously difficult, given the necessity to use borehole technology to obtain the samples combined with the paucity of groundwater and low hydraulic conductivities that characterise a potentially suitable rock mass for the siting of a repository. Similarly, it has not been possible until recently to characterise reliably the hydrochemistry of the pore water in a crystalline rock matrix for the purposes of a repository project. Improvements in the techniques used to obtain and analyse samples have been made on the basis of international experience in the field such that the difficulties can be overcome sufficiently to obtain enough good quality samples to characterise the hydrochemistry of a prospective repository site. An important contribution has been made by the development of protocols to qualify representative samples that are not significantly contaminated by materials introduced by drilling or by the sampling method (e.g. Laaksoharju et al., 2008a). Some geochemical measurements, such as pH or redox potential, Eh, can be made by specially designed down-hole instruments.

5.6.1 Multi-disciplinary approach To various degrees, the sampled groundwater compositions reflect processes relating to ancient through to modern water/rock interactions and mixing of groundwaters of different origins. Such complex data require a multidisciplinary approach to their interpretation in order to produce a hydrogeochemical model that can be used, for example, in hydrogeological, radionuclide transport and safety assessment modelling. The hydrogeochemical model should represent a site-scale hydrogeochemical interpretation that is fully integrated with the corresponding geological and hydrogeological models. Laaksoharju et al. (2008b) propose that the objective for such a model in the Swedish geological repository programme is that it should clearly show the following: . .

the major lithological and structural units comprising the site; knowledge of the major groundwater flow directions from hydrogeological modelling;


Crystalline geological repository systems . . . .

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the relationship of chemistry (i.e. mixing of end members, chemical reactions, etc.) to these major hydraulically conducting pathways; a clear indication of the groundwater types representative of the main hydraulic rock mass; units characterised by fractures (fracture zones) of lower transmissivities; and an indication of the chemistry of the rock matrix pore space fluid/ groundwater.

5.6.2 Application of geochemical information The specific uses to which such a model can be put are highly dependent upon the location of the site and the repository concept under consideration. However, a general scheme can be defined for the application of geochemical information and hence the targets for geochemical characterisation, as, for example, by Bath (2008): 1. 2.

3.

4.

5.

Performance of the engineered barrier system: groundwater compositions at repository depth and up-gradient of repository location. Groundwater flow directions and rates: locations of groundwater volumes with distinct compositions; hydrochemistry of transmissive deformation zones and fractures; hydrochemical properties at groundwater flow model boundaries; and depth of active modern groundwater circulation. Pathways for radionuclide transport and retention in the geosphere: groundwater compositions at repository depth; hydrochemistry of transmissive deformation zones and fractures down-gradient of repository location; and geochemical description of radionuclide transport pathways including fracture minerals and adjacent rock matrix. Palaeohydrogeology (i.e. groundwater flows and compositions in the past): locations of groundwater volumes with distinct compositions and hydrochemical properties at groundwater flow model boundaries. Biosphere pathways and processes, including the geosphere–biosphere interface zone: hydrochemistry of shallow groundwaters, soil waters and surface waters; geochemical composition of soils; and baseline geochemical conditions of the surface environment.

Identification of objectives The approach that has been developed for determining the samples and analyses that are required involves the identification of the specific objectives that characterisation of the geochemistry has to achieve in


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support of such a scheme. Bath (2008) identifies the following objectives that are generally recognisable in the crystalline rock programmes that have been documented to date. Develop a redox model This can be used in assessing the performance of engineered barriers and inform the speciation of redox-sensitive radionuclides. The following information would be used: . . . .

Use geochemical modelling to interpret Eh measurements and redoxsensitive solute data with particular attention to sulphate reduction, FeII/FeIII equilibria and methane/carbon dioxide equilibria. Support the interpretation of the sulphate/sulphide system with data on stable sulphur-34/sulphur-32 isotopic ratios. Use microbial population data to develop an understanding of microbial mediation of chemical conditions in general and redox in particular. Use data on redox parameters and microbial populations as a function of depth from the surface to understand the development of a redox front with infiltration.

Understand controls on groundwater composition . .

Use graphical and statistical numerical analyses of data for non-reactive solutes (e.g. chloride) and stable isotopes (e.g. oxygen-18/oxygen-16 ratios) to determine solute sources and groundwater mixing ratios. Determine the capacity of rocks and minerals in the prospective repository location to control or buffer the key hydrochemical conditions such as Eh, pH and total inorganic carbon.

Evaluate groundwater ages and solute travel times .

. . .

Use non-reactive solutes and stable isotopes to: support and test the interpretation of groundwater flow paths in the hydrogeological model; constrain groundwater sources and mixing (as above); and calibrate a palaeohydrogeological model. Use chloride and oxygen-18/oxygen-16 data from extracted rock matrix waters to characterise matrix diffusion and the palaeohydrochemical equilibrium between matrix porewater and mobile groundwater. Use tritium data to identify where modern groundwater recharge occurs. Interpret carbon-14 and carbon-13/carbon-12 data to obtain groundwater ages.


Crystalline geological repository systems . .

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Interpret chlorine-36 data in terms of chlorine residence times and groundwater ages. Interpret helium-4 data in terms of groundwater mixing and residence times, using estimates of its in situ production from the uranium and thorium present in the rocks.

Support development of the surface compartments of the hydrogeological model . .

Use tritium data to identify the depth of penetration of recently recharged water. Use chloride and other geochemical and isotope data to identify areas of groundwater discharge.

Evaluate palaeohydrogeological stability . . .

Use observations of calcite (secondary mineral) morphology to interpret past variations of groundwater salinity. Use iron, manganese and rare earth element contents of calcite and other secondary minerals to interpret past stability or variability of redox conditions. Use oxygen-18/oxygen-16 and carbon-13/carbon-12 data to identify calcite sources and qualitative ages, and also to interpret carbon biogeochemistry.

Quantify concentrations and fluxes of analogue solutes and mobile species .

.

Interpret dissolved concentrations of naturally occurring uranium, thorium, radium and other solutes that are analogues for repository radionuclides to test consistency with the outputs of radionuclide transport models. Interpret variations of concentrations of colloids, microbes and gases in terms of understanding their sources, compositions, reactions and transport/retention mechanisms.

Establish baseline hydrochemical conditions . .

Determine the chemical compositions of undisturbed groundwaters and surface waters in the siting area. Interpret dissolved gas compositions in shallow groundwaters and in soil gases.


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Identify any hazardous or anomalous geochemical conditions or scenarios . .

5.7

Interpret the production, concentrations and outgassing of radon, methane and any other hazardous dissolved gases at the prospective repository depth. Identify any other geochemical conditions that pose potential environmetal hazards (e.g. the presence of arsenic or fluoride) that could affect the performance of the engineered barrier system (e.g. sulphide mineral oxidation to give sulphuric acid, penetration of low ionic strength waters that would degrade swelling clays) or that could influence radionuclide transport (e.g. naturally occurring organic complexants).

Radionuclide transport

It is considered that on the macroscopic scale radionuclide transport in crystalline rock will occur along advective flow channels that are within deformation zones and fractures. The retardation processes that are most often considered to act to limit the rate of radionuclide transport in these flow channels are rock-matrix diffusion and sorption. The characterisation of the relevant processes uses a combination of in situ and laboratory techniques and technologies. It relies heavily on information on lithologies, geological structure, hydrogeology and geochemistry, both to provide the framework for the transport model and to provide data on the compositions and mineralogies of rocks and fracture infills, and of groundwater and matrix porewater.

5.7.1 Transport data As exemplified by Crawford (2008), the transport data required include effective diffusivities for radionuclide transport in the rock matrix, rockmatrix porosities, the surface area of the porosity, cation exchange capacities of the minerals in the rocks and fracture infills, and the sorption properties for radionuclides on to rocks and fracture infills in contact with groundwater of various compositions. All of this information can be obtained from what are now relatively standard laboratory measurements, albeit that they are demanding to perform under the carefully controlled conditions required to ensure that they provide representative data. Other disciplines in the overall site characterisation are important in ensuring that the rocks and fracture infills used in making the measurements are representative of those that will be contacted along the advective flow path and that appropriate synthetic groundwaters are used in the radionuclide sorption measurements.



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The effective diffusivity is obtained through the formation factor that is defined as the ratio of effective diffusivity of a solute to the free diffusivity of that solute at infinite dilution in water. The formation factor is a geometrical parameter dependent upon the porosity, tortuosity and constrictivity of the pore space available for diffusion into the rock matrix. Formation factors can be measured both in the laboratory, using through-diffusion experiments involving tritiated water and by electrical resistivity measurements, and in situ, using a high-resolution, downhole electrical resistivity tool. The in situ measurements are important in ensuring that the formation factor is representative of conditions at depth where the rock is subjected to stress and to identify where artificially high formation factors may have resulted from opening of pores in the laboratory samples in the absence of the stress field.

5.7.2 Flow data The extent of the interaction of radionuclides dissolved in groundwater in the advective flow channels with the fracture surface and the rock matrix will be controlled by the surface area in contact with the flowing water for a given water flow rate. The ratio of the flow wetted surface to flow rate is termed the hydrodynamic transport resistance or, more conveniently, the Ffactor (Andersson et al., 1998). The necessary information on flow and its distribution in the crystalline rock mass comes from the hydrogeological model and will follow the approach to characterising the hydrogeological properties of deterministically defined deformation zones and stochastically modelled fracture networks as described above. Typically a two-dimensional model will be constructed of the deterministic deformation zones such that it provides the lateral and depth-dependent variation of the transmissivity, calibrated against borehole measurements in the deformation zone. A hydrogeological discrete fracture network (DFN) model is used to provide F-factors for representative flow paths in the stochastically modelled fracture network and the frequency of the occurrence of the modelled flow paths. Information on representative flow path lengths and hydraulic gradients is required to support this technique. It is recognised that there is the potential to underestimate the effect of flow channelling in individual borehole measurements either because the flow channel is not intersected or, in the case that its properties are inferred, the flow rate that the channel can support is underestimated because of the bulk properties of the structure in which it occurs. Flow-channelling effects in the connected fracture network should be captured in the hydrogeological DFN modelling, but other, more-localised, channelling is usually accounted for by scoping calculations until the stage of undertaking underground


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investigations. Once underground, mapping of open fractures and determination of flowing fracture frequencies, coupled with information on the relationship between rock mechanical properties and fracture transmissivities, would lead to an improved understanding of the extent and nature of flow-channelling effects.

5.7.3 Description of transport characteristics The information on radionuclide retardation by rock-matrix diffusion and sorption can be combined with F-factors for the various sections of flow paths from the repository depth to the surface in safety assessment models, which in this way capture the relevant site characterisation information in evaluating the transport of radionuclides along these flow paths. The transport properties derived by the approach described here can be tested by various designs of tracer tests where the movement of tracer solutes in flow paths between sections of injection boreholes and receptor boreholes is measured under controlled conditions. Single well injection and withdrawal tests have also been used in some characterisation programmes.

5.8

Disturbance by excavation or waste emplacement

The excavation of the repository underground works, comprising the deposition areas, access tunnels, drifts and shafts will perturb the natural system in a number of ways. The rock surrounding the excavations can be subject to mechanical, hydrogeological and chemical disturbance, as will be outlined below. The main impact of waste emplacement considered in respect of geological repositories is the thermal perturbation caused by the disposal of heat-generating waste. There are also chemical perturbations to be considered but these are more concept-specific. These processes of disturbance will not occur separately and independently of one another. The coupling between the different processes requires to be understood sufficiently to assess the implications for repository design and safety. This typically involves the construction of thermal–hydraulic– mechanical–chemical (THMC) coupled process models and considerable international effort has been devoted to develop capabilities in this area, using carefully controlled experiments, often conducted in underground laboratories in validation testing of the outputs of such models (NEA, 2001).



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5.8.1 Mechanical disturbance Deformation of the rock Deformation of the rock mass surrounding the excavated opening can result from the response of the rock to the in situ stress field. This is most significant when there is a high, anisotropic stress field such as found at the underground research laboratory at Lac do Bonnet (Canada) that was operated in the past in support of the Canadian spent fuel waste management programme (Martin, 1990). Using information obtained from field experiments on such deformation phenomena, the orientation of tunnels and their cross-sectional geometry and dimensions can be optimised to minimise the extent of such deformations. Disturbance to fracture system Opening of new and pre-existing fractures in the rock wall surrounding the excavations and movement along fracture planes can occur in response to stress relief. This type of disturbance is often split into subcategories. ‘Excavation damage’ is used to describe perturbations that are not reversible, in extreme cases involving the movement of rock blocks or mechanical spalling of smaller rock fragments. ‘Excavation disturbance’ describes perturbations that may be transient in nature where the relevant aspects of the natural system are expected to return towards their unperturbed state following the closure of the repository, e.g. changes in aperture of fractures in response to stress relief. Characterisation of the rock mechanics and geotechnical properties as described above is designed to understand the quality of the rock mass and hence its response to excavation, among other things. Thus the extent and nature of any excavation damage will be assessed in advance and appropriate measures taken in terms of design of the underground openings. Limited damage of the rock surface resulting from the excavation process itself is unavoidable, but a range of excavation methods have been developed for use in crystalline rock and the extent of surface damage can be controlled by the choice of method if necessary. In recent years large-scale underground experiments have been conducted to measure the nature and extent of excavation disturbance in the rock wall surrounding an excavated tunnel in terms of fracture opening, movement on fractures and any resulting changes in hydraulic properties, e.g. the ZEDEX experiment conducted in the Aspo Hard Rock Laboratory in Sweden (Emsley et al., 1997). Such experiments use similar techniques to those applied in surface-based site characterisation, although it is worth mentioning the value of acoustic emission tests to detect the extent of


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movement of fractures. The information obtained on the characteristics of the damaged/disturbed zones can be used to construct appropriate structural and hydrogeological models of these features for use in conjunction with the equivalent model for the undisturbed system.

5.8.2 Hydrogeochemical disturbance The excavation of the underground works will create a void in the rock mass at essentially atmospheric pressure and result in a hydraulic pressure gradient towards the excavations. This means that the pre-existing, natural groundwater system will be perturbed by flows towards the excavations. The amount of such perturbing flow will depend on the pressure gradient and flow path characteristics, and also on the groundwater control measures that are implemented in constructing the underground works. This so-called drawdown has the potential to be used to confirm the understanding of the hydrogeology. Once the repository has been filled with wastes and sealed and closed, it would be expected that the pre-existing pressure field would be restored. However, one of the potentially important long-term impacts of the drawdown would be to disturb the groundwater chemistry at the repository depth and in the surrounding rocks. This requires careful consideration, particularly when, as is often found, there are marked variations in groundwater salinity as a function of depth, and water of different salinity would then be introduced into the repository system from the overlying rocks. The hydrogeological and geochemical models described earlier can be used to evaluate the extent and spatial distribution of the resulting hydrochemical changes. The capability to carry out such evaluations can be tested beforehand by comparing the predictions of equivalent models in relation to the sites of underground research laboratories against information on the hydrochemical perturbation that occurs as a consequence of their excavation (e.g. Vieno et al., 2003). The introduction of air will also perturb the hydrochemistry and the mineralogical chemistry, particularly when one bears in mind that chemically reducing conditions are usually found at depth. However, the fact that the groundwater flow is towards the repository excavations when they contain air means that the depth of such disturbance into the rock wall will be limited. Methods exist from other industries such as mining and civil engineering to assess the extent of phenomena, such as acid mine drainage (where sulphide minerals such as pyrite are oxidised in air to produce sulphuric acid), that must be taken into account in describing the perturbed condition of the repository.



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5.8.3 Thermal disturbance In repository designs for the disposal of heat-generating wastes in crystalline rocks it is typically found that the temperature of the rock surrounding the emplaced waste will be increased by a few tens of degrees Centigrade for a few hundred years and by just a few degrees over timescales on the order of one thousand years or more. As noted above, this will cause effects that interact with the effects of other forms of disturbance, e.g. localised changes in rock stresses and hydraulic pore pressures in the immediate vicinity of the waste. Combined modelling and experimental testing, particularly using heater tests in underground research laboratories, is used to gain an adequate understanding of the coupling of the thermal processes with others. Thermal rock spalling, where the thermally induced stresses could cause detachment of fragments of rock, currently receives a great deal of attention since it has the potential to degrade the performance of some engineered barrier systems that would adjoin the rock surface. This provides a good example of the interaction between site characteristics and repository design since the engineered barrier design or its method of emplacement would probably need to be adapted to overcome this difficulty.

5.8.4 Chemical disturbance by the wastes or engineered barrier materials The emplacement of wastes and engineered barriers will introduce a number of materials that are foreign in respect of the pre-existing natural system. The disturbance created will depend strongly upon the chemical nature of the materials in relation to the natural hydrochemistry and mineral chemistry. Many engineered barrier materials are selected on the basis that they will not react significantly with the groundwater and consequentially their impact on the chemistry of the natural system will be minimal. The materials that have been identified as having the greatest potential impact in this respect are cement-based materials, variously used in grouts within certain waste packages and as repository construction materials, and metallic wastes or waste containers that will corrode to generate gas under the prevailing hydrochemical conditions. Groundwater that contacts cement-based materials will be conditioned to high alkalinity and will leach the chemical constituents of the cement, typically leading to increased concentrations of sodium and potassium (in the short term) and of calcium. Extensive integrated modelling and experimental work, including the use of natural analogue systems, has been conducted to develop models for the evaluation of the nature and extent of the change to groundwater chemistry and mineral composition along flow paths leading from the repository (Nirex, 2002a).


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The generation of gas has the potential to create an overpressure and, if it exceeds the solubility limit, a separate gas phase that will interact with the groundwater flow system. Again, extensive integrated modelling and experimental work has been conducted to develop models for the evaluation of any physical disruption of the surrounding rock mass and of two-phase gas-water flow in fractures (Nirex, 2002b).

5.9

Stability

The scientific and technical factors taken into account in selecting a candidate repository site for characterisation include, in particular, its longterm stability and its buffering capacity with respect to external and internal perturbations, along with its ability to accommodate the waste disposal facility and its ability to prevent or attenuate the potential release of radionuclides. In building a safety case, it is therefore important to assess: . . .

the features, events and processes (FEPs) that could influence the evolution of the geosphere; the long-term stability of the favourable conditions displayed by the host formation; and the buffering capacity of the formation with respect to perturbations.

The key issue is to evaluate the resilience of the main safety functions of the geosphere (including its flow and transport properties) to natural perturbations. The relevance of various naturally occurring processes and events will depend upon the timeframe to be considered, but timeframes on the order of about one million years are typically considered. Stability, in this sense, does not imply that steady-state conditions exist; the geosphere is constantly evolving, although in many cases rather slowly, and such evolution is perfectly acceptable for safe geological disposal. What is important is that this evolution is properly understood.

5.9.1 Geoscientific understanding In a general sense, crystalline rock formations are regarded as very stable: many (in Scandinavia and Canada in particular) have ages on the order of billions of years and are associated with very low surface erosion rates. This has been an important reason in many countries for considering such formations for radioactive waste repositories. Therefore many crystalline rocks provide intrinsically stable environments, particularly from a mechanical standpoint, and provide good buffering against external events and processes. Considerably younger crystalline rocks have also been considered, e.g. in Japan, but their relative youth does not imply that they are, necessarily, any



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less stable, or less suitable for disposal purposes. There may, however, be intrinsic differences between very old and relatively recent rocks, especially with regard to the types of deformation zones present and the variability of fracture orientations, which could have implications with regard to stability. The natural phenomena (processes and events) identified to have the potential to perturb the stability of a crystalline rock have remained constant over a number of years and are as follows: . . . .

seismicity (in respect of the reactivation of existing fractures, faults and deformation zones); climate change (leading to either colder or warmer surface temperatures, and hence, variously, to ice sheets, permafrost or sea level change); uplift and erosion; and volcanism.

The geoscientific understanding of these phenomena was evaluated at a workshop of the OECD Nuclear Energy Agency Geosphere Stability Project (NEA, 2009). Regarding the understanding and handling of perturbations that affect the stability, the findings of the workshop, following a review of the results of site characterisation and research programmes, can be broadly summarised as follows: . . . . . .

There is, in general, good confidence in the understanding of the magnitude, cause, characteristics and frequency of perturbing phenomena. There is more limited confidence as to where and when a perturbation will occur. Confidence is also more limited regarding the volume of rock affected by a perturbation, but there are methods available to study such issues. The extent to which a repository is affected by a perturbation can often be conservatively addressed by using bounding and/or pessimistic approaches. Conclusions regarding many of these phenomena are supported by the results of natural analogue studies. Such evidence is useful in supporting a safety case, but often relevant site-specific observations provide stronger evidence than those from a general natural analogue study.

5.9.2 Geosphere stability in safety cases There are assessment tools (deterministic, probabilistic and bounding) for taking account of uncertainties in relation to geological stability in a safety case and also a wealth of examples of how these tools have been applied.


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The bounding analyses often carried out as part of a safety case are likely to provide pessimistic bounds, which are also likely to be broader than the bounds produced by the use of so-called proxy data, obtained from investigations of analogous geological systems. Ultimately the evolution of the geosphere and the associated stability issues must be evaluated in a repository and site-specific context. For aspects of stability to which the repository design or safety assessment is particularly sensitive, there may be a need to constrain the assessment with more realistic site-specific information.

5.9.3 Implications for repository design As an example, analyses of the effects of seismic perturbations on the stability of the geosphere have shown that they have important implications for repository design, especially in relation to selecting the location of the repository within the rock mass to provide a suitable separation from active or potentially active deformation zones or faults. At the repository scale, the effects of stress magnitude and stress orientation, combined with consideration of the potential reactivation of deformation zones and the larger fractures in future seismic events, provide important constraints on the location and orientation of disposal tunnels and on the positioning of waste containers. In some geological environments (perhaps, for example, in Japan) the orientation of the maximum horizontal stress may be expected to remain approximately constant for a considerable time in the future, so that it is the effects of future seismicity that are of greatest interest. This appears also to be the case in other, seismically quieter countries, such as Sweden and Finland; here also it is considered that the greatest perturbation to the geological stability would be caused by a future seismic event during rapid glacial retreat as a result of a future transition from a cold climate state to a warmer one. An understanding of the long-term stability of the hydrogeochemistry may be very important in cases where the intrusion of altered groundwaters at repository depth could degrade the performance of engineered barriers, for example. The construction of a palaeohydrogeochemical model to understand how the system has responded to equivalent perturbations in the past is noted above as an important objective of geochemical characterisation of a site.

5.10

Feasibility of construction

The wealth of experience obtained from tunnelling and civil engineering projects in crystalline rocks has enabled the construction of empirical rock mass classification systems to inform the design and choice of construction



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and support methods in such environments. Provided that the rock mechanics and geotechnical properties have been characterised appropriately, the orientation, geometry and dimensions of underground openings can be optimised in terms of creating a suitably stable environment in which the necessary repository operations of emplacing waste containers and the associated engineered barriers can be conducted safely. Following this approach, tunnels or vaults having diameters or spans on the order of 20 metres or more can be excavated at prospective repository depths on the order of 500 metres in good-quality crystalline rock and with minimal rockbolting will give openings that will remain stable for about 100 years or more.

5.10.1 Construction methods Well-established construction methods are available to excavate such openings, involving the use of drilling and blasting techniques or of tunnel-boring machines. The comparative attributes of these two general methods for excavating tunnels at depth in crystalline rock have been evaluated at the Aspo Hard Rock Laboratory (HRL) in Sweden (Ba¨ckblom et al., 2004). Also, in connection with evaluations of the potential mechanical disturbance caused by excavation, various controlled blasting methods have been developed to enable the damage to the rock surface to be limited, if required. If the excavation has to pass through difficult features such as fracture zones carrying high water flows, a suite of methods has been developed to deal with these situations both from experience in underground research laboratories and in other industries. In the example of high water flows in fracture zones, grouting is likely to be used, as exemplified by the experience gained at the Aspo HRL.

5.10.2 Stabilising the excavations As already noted, the excavated opening is likely to be stabilised by rock bolting with the frequency and length of the rock bolts being determined by the local rock quality. Consideration would have to be given to the choice of materials used in connection with this control method to avoid introducing materials that may be detrimental to repository performance, e.g. organic resins. The spalling of small fragments of rock can be prevented by the use of shotcrete, a surface coating of a specially formulated concrete. Consideration would have to be given to the chemical perturbation that this would cause in the long term, which might necessitate its removal prior to backfilling and sealing the relevant section of the repository. Similar considerations will apply to any use of concrete in structural components of the underground workings, such as floor slabs below waste handling


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machines, and some programmes are studying the potential development of specially formulated concretes that will minimise the chemical perturbation caused by their use (e.g. Vuorinen et al., 2004). Groundwater inflows into the excavations can be controlled to some extent by grouting the flowing fractures or more simply by piping the water into a sump if the flow rate does not warrant grouting. The construction of sealing systems to isolate various compartments of the underground system is covered as part of the discussion of engineered barriers in another chapter.

5.10.3 Design considerations As noted in the discussion on stability, the location and layout of the underground works will be optimised relative to key site characteristics and their evolution in response to potential natural perturbations that may occur in the future. The example of avoiding deformation zones and fractures that may be reactivated has been discussed already and other features, such as fractures carrying excessive flow of groundwater, are also typically identified as requiring to be avoided. A formal process of identifying requirements that should be met to ensure the safety functions of the engineered barriers and geosphere, as now used in the most advanced programmes, provides a framework for making these design decisions (NEA, 2004).

5.11

Future trends

Existing techniques and technologies already provide a high capability to characterise crystalline sites, as evidenced by the interpreted results of site characterisation programmes conducted in Finland and Sweden and the associated identification of remaining issues. In addition, good experience continues to be gained from underground research laboratories such as the Aspo HRL and from the ONKALO rock characterisation facility that is being developed at the selected repository site at Olkiluoto in Finland. Nevertheless, possible improvements have been identified, not least because waste management programmes have an obligation to take account of all relevant scientific and technological advances. An obvious development will be in the use of information obtained from investigations undertaken underground to test the understanding gained from surfacebased investigations, in particular the various models used to represent that understanding. Although research facilities such as the Aspo HRL and the Canadian URL have provided a dress rehearsal of these techniques, this way of testing and calibration of the data and modelling that are to be used in repository design and safety assessment will provide new insights. Aspects that are considered important in these advanced programmes are as follows:


Crystalline geological repository systems . . .

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calibrating the size distribution and size intensity models for fractures and fracture systems at repository depth from underground investigations; obtaining information on in-plane flow channelling from investigations of coupling of rock mechanics properties and fracture transmissivities in tunnels; and characterising the nature and extent of any changes in hydraulic conductivity that are induced by excavation disturbance as a function of local conditions.

Technological improvements that may be made in connection with underground investigations include: . . .

methods for stabilising investigation boreholes, particularly when intersecting features carrying high water flows; laser scanning to produce three-dimensional surface mapping of rock walls; and methods for measuring water inflows that reflect the previously unperturbed natural system more reliably.

Various technological improvements have been identified for possible future use in surface-based investigations, although in some cases they may also have application underground. A number of these relate to geophysical tools for use in boreholes, as follows: . . . .

a nuclear magnetic resonance tool to relate measured petrophysical properties to fluid flow; an ultrasonic tool (acoustic televiewer) to determine fracture orientation; a downhole tool for in situ measurement of rock thermal conductivity as a function of localised rock characteristics; and ‘walk-around’ borehole seismic techniques to image the key features in the rock mass between boreholes.

In addition, there are recognised benefits in conducting logging while drilling, rather than after the borehole has been drilled to target depth, and this technique has been trialled successfully. It has been recognised that there is value in understanding the occurrence and spatial distribution of different fracture minerals on the site scale, so attempts will be made to develop such models in relation to the role that fracture minerals play both in determining radionuclide transport properties and in buffering the hydrochemistry (e.g. with respect to pH and Eh). In the field of modelling in relation to site characterisation, there are likely to be improved techniques for treating spatial variability in relation to geological structure and hydrogeology and for modelling coupled THMC


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processes over length scales of interest in repository design and safety assessment. Several initiatives are underway in the area of evaluating the stability of crystalline sites (NEA, 2009). An example is the International Tectonics Meeting (ITM) methodology of the Nuclear Waste Management Organisation (NUMO) of Japan, which considers the likelihood of tectonic activity affecting the stability of a potential repository in Japan. It is aimed at assessing the likelihood of a site being significantly affected by volcanic events and rock deformation processes within 104 years and the potential scale and nature of any such impacts as a function of their type and geographic distribution.

5.12

Sources of further information

The site descriptive modelling reports reports published by the Swedish Nuclear Fuel and Waste Management Company (SKB, 2008) and the Finnish waste management company, Posiva Oy (Andersson et al., 2007), respectively on the Forsmark and Olkiluoto sites, and the references cited therein, provide a comprehensive, state-of-the-art view of the characterisation of crystalline sites. These overall site description reports are supported by tiers of reports that provide successively more details on the acquisition and use of information from the relevant geoscientific disciplines that have been discussed in this chapter. All such reports are readily accessible (in May 2009) through the respective websites of these two organisations, respectively www.skb.se and www.posiva.fi. An assessment of geosphere characterisation techniques conducted in the context of the United Kingdom radioactive waste management programme (Nirex, 2007, and references cited therein) provides a good summary of the capabilities of a comprehensive range of techniques used in waste management programmes and other relevant industrial applications.

5.13

References

AECL (1994), ‘Environmental impact statement on the concept for disposal of Canada’s nuclear fuel waste’, Atomic Energy of Canada Limited Report AECL-10711, COG-93-1, Atomic Energy of Canada Ltd, Ottawa. Andersson J, Ahokas H, Hudson J, Koskinen L, Luukonen A, Pitka¨nen P, Mattila J, Ikonen A and Yla¨-Mella M (2007), ‘Olkiluoto Site Description 2006’, Posiva Oy Report POSIVA 2007-03, Posiva Oy, Olkiluouto, Finland. Andersson J, Hermanson J, Elert M, Gylling B, Moreno L and Selroos J-O (1998), ‘Derivation and treatment of the flow-wetted surface and other geosphere parameters in the transport models FARF31 and COMP23 for use in safety assessment’, SKB Report R-98-60, Svensk Ka¨rnbra¨nslehantering AB, Stockholm, Sweden.



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Andra (2005), Dossier 2005 Granite, Agence Nationale pour la Gestion des Dećhets Radioactifs, Paris, France. Ba¨ckblom G, Christiansson R and Lagerstedt L (2004), ‘Choice of rock excavation methods for the Swedish deep repository for spent nuclear fuel’, SKB Report R-04-62, Svensk Ka¨rnbra¨nslehantering AB, Stockholm, Sweden. Bath A (2008), ‘Geochemistry strategy for GeoCORE’, Intellisci Ltd, Technical Note 0607.19 prepared for Nirex, Nuclear Decommissioning Authority, Harwell. Crawford J (ed.) (2008), ‘Bedrock transport properties Forsmark, Site descriptive modelling SDM-Site Forsmark’, SKB Report R-08-48, Svensk Ka¨rnbra¨nslehantering AB, Stockholm, Sweden. Emsley S, Olsson O, Stenberg L, Alheid H-J and Falls S (1997), ‘ZEDEX – a study of damage and disturbance from tunnel excavation by blasting and tunnel boring’, SKB Report TR-97-30, Svensk Ka¨rnbra¨nslehantering AB, Stockholm, Sweden. Glamheden R, Fredriksson A, Ro¨shoff K, Karlsson J, Hakami H and Christiansson R (2007), ‘Rock mechanics Forsmark, site descriptive modelling Forsmark stage 2.2’, SKB Report R-07-31, Svensk Ka¨rnbra¨nslehantering AB, Stockholm, Sweden. Golder Associates (2006), ‘Geosphere Characterisation Project – state of the art in hydrogeological testing and interpretation’, Report 05527634.500/A.0 prepared for Nirex, United Kingdom Nirex Limited, Harwell. Juhlin C (2007), ‘Integrated interpretation of surface and borehole (VSP) seismic data along profiles 2 and 5, Forsmark, Sweden’, in Geology – Background Complementary Studies, Forsmark Modelling Stage 2.2, edited by M Stephens and K Skagius, SKB Report R-07-56, Svensk Ka¨rnbra¨nslehantering AB, Stockholm, Sweden. Laaksoharju M, Smellie J, Tullborg E-L, Gimeno M, Hallbeck L, Molinero J and Waber N (2008a), ‘Bedrock hydrogeochemistry Forsmark, site descriptive modelling SDM-site Forsmark’, SKB Report R-08-47, Svensk Ka¨rnbra¨nslehantering AB, Stockholm, Sweden. Laaksoharju M, Smellie J, Tullborg E-L, Gimeno M, Go´mez J, Auque´ L, Molinero J, Gurban I, Hallbeck L, Buckau G, Gascoyne M and Wallin B (2008b), ‘The hydrogeochemical modelling approach used within the Swedish site investigation programme’, in Abstracts, 33rd International Geological Congress, Oslo, compiled by K Ahlbom and M Stephens, SKB Report R-0897, Svensk Ka¨rnbra¨nslehantering AB, Stockholm, Sweden. Martin C (1990), ‘Characterising in-situ stress domains at the AECL underground research laboratory’, Canadian Geotechnical Journal, 27(5), 631–646. Munier R (200), ‘Statistical analysis of fracture data, adapted for modelling discrete fracture networks – version 2’, SKB Report R-04-66, Svensk Ka¨rnbra¨nslehantering AB, Stockholm, Sweden. Nagra (1994), ‘Kristallin-1 safety assessment report’, Nagra Technical Report 93-22, Nationale Genossenschaft fu¨r die Lagerung radioaktiver Abfa¨lle, Baden, Switzerland. NEA (2001), ‘Going underground for testing, characterisation and demonstration (a technical position paper)’, Nuclear Energy Agency Report NEA/RWM(2001) 6/ REV, OECD Nuclear Energy Agency, Paris.


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NEA (2004), ‘Engineered barrier systems (EBS) design requirements and constraints’, Nuclear Energy Agency Report 4548, OECD, Paris. NEA (2009), ‘Stability and buffering capacity of the geosphere for long-term isolation of radioactive waste – application to crystalline rock’, Nuclear Energy Agency Report 6362, OECD, Paris. Nirex (1997), ‘Nirex 97 – an assessment of the post-closure performance of a deep waste repository at Sellafield’, Nirex Science Report S/97/012, United Kingdom Nirex Limited, Harwell. Nirex (2002a), ‘Research on the alkaline disturbed zone resulting from cement– water–rock reactions around a cementitious repository’, Nirex Report N/054, United Kingdom Nirex Limited, Harwell. Nirex (2002b), ‘Review of the work undertaken on gas migration in the geosphere’, Nirex Report N/023, United Kingdom Nirex Limited, Harwell. Nirex (2007), ‘Geosphere characterisation project-status report October 2006’, Nirex Report N/136, United Kingdom Nirex Limited, Harwell. O¨hberg A and Rouhiainen P (2000), ‘Posiva groundwater flow measuring techniques’, Posiva Oy Report POSIVA 2000-12, Posiva Oy, Olkiluoto, Finland. SKB (2008), ‘Site description of Forsmark at completion of the site investigation phase-SDM-site Forsmark’, SKB Report TR-08-05, Svensk Ka¨rnbra¨nslehantering AB, Stockholm, Sweden. Stephens M, Fox A, La Pointe P, Simeonov A, Isaksson H, Hermanson J and O¨hman J (2007), ‘Geology Forsmark-site descriptive modelling Forsmark stage 2.2’, SKB Report R-07-45, Svensk Ka¨rnbra¨nslehantering AB, Stockholm, Sweden. Triumf C-A (2007), ‘Oskarshamn site investigation – assessment of probable and possible dolerite dykes in the Laxemar sub-area from magnetic total field data and digital elevation models’, SKB Report P-07-223, Svensk Ka¨rnbra¨nslehantering AB, Stockholm, Sweden. Vieno T, Lehikoinen J, Lo¨fman J, Nordman H and Me´sza´ros F (2003), ‘Assessment of disturbances caused by construction and operation of ONKALO’, Posiva Oy Report POSIVA 2003-06, Posiva Oy, Olkiluoto, Finland. Vuorinen U, Lehikoinen J, Imoto H, Yamamoto T and Alonso M (2004), ‘Injection grout for deep repositories, subproject 1 – low-pH cementitious grout for larger fractures, leach testing of grout mixes and evaluation of the long-term safety’, Posiva Oy Working Report 2004-46, Posiva Oy, Olkiluoto, Finland. Wahlgren C-H, Curtis P, Hermanson J, Forssberg O, O¨hman J, Fox A, La Pointe P, Drake H, Triumf C-A, Mattsson H, Thunehed H and Juhlin C (2008), ‘Geology Laxemar-site descriptive modelling SDM-site Laxemar’, SKB Report R-08-54, Svensk Ka¨rnbra¨nslehantering AB, Stockholm, Sweden.


6 Clay geological repository systems: characterisation and site surveying technologies and techniques J . D E L A Y , National Radioactive Waste Management Agency, France

Abstract: The concept of a disposal facility in a clay site relies on the idea that long-term safety will be assured by the argillaceous rock and the repository layout. Therefore, the main objective of the studies carried out on the host formation is to determine its confinement abilities and its stability over a long period of time. Since the physical properties of the argillaceous rocks constrain the layout of the underground structures, the natural properties of the host formation are of concern in the survey strategy. This chapter sets out the technologies and techniques that will be implemented in the framework of such a geological survey. Key words: clay site, argillaceous rock, survey strategy, geological survey, physical properties of the rocks.

6.1

Foreword

The argillaceous sedimentary formations capable of hosting a disposal facility must (Andra, 2005): . .

.

Protect the waste packages from superficial phenomena, mainly climatic (erosion, glaciation) and human activities. Ensure the confinement of radionuclides, i.e. possess intrinsic properties restricting water migration and promoting the retardation of radionuclides. These properties must be ensured over long periods of time (from thousands to several hundred thousand years). Maintain these favourable properties in spite of the perturbations induced on the rock by the construction of the facility (excavation damaged zone, i.e. fractures caused by stress redistribution around the underground structures after excavation). In addition, the barrier properties of the rock must remain intact despite the effects from heat153 © Woodhead Publishing Limited, 2010

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Geological repository systems for safe disposal generating waste packages and the degradation of the packages and the engineered barriers.

In Europe, three argillaceous sedimentary formations are the subjects of intensive research carried out through both surface surveys and underground laboratory experiments. They are: (1) the Boom Clay at Mol in Belgium (ONDRAF: Organisme National des Dećhets Radioactifs et des Matie`res Fissiles, 2001), (2) the Callovo-Oxfordian at Bure, in France (Andra: Agence Nationale pour la Gestion des Dećhets Radioactifs, 2005) and (3) the Opalinus Clay at Benken (Nagra: Nationale Genossenschaft fu¨r die Lagerung radioaktiver Abfa¨lle, 2002a) and Mont Terri (Bossart and Thury, 2008) in Switzerland. These rock formations consist mostly of clay minerals, carbonates, quartz and other silicates. In a natural environment, these components may exist in greatly varying proportions and all intermediary facies may be found, whether pure clay, limestone or siltstone. In addition to these major components, organic matter and accessory minerals, such as pyrite, can also be found (Boisson, 2005; Yven et al., 2007). Siliceous shale (porcelanite) and diatomaceous shale studied for nuclear waste disposal in Japan (Horonobe site) may also be considered as equivalents of argillaceous rocks (Hama et al., 2007). However, due to poor compaction and fractures, their hydraulic conductivity is high and a disposal concept cannot rely only on the confining properties of this rock. In Canada a Deep Geologic Repository for low- and intermediate-level radioactive waste has been proposed in a massive Ordovician argillaceous limestone overlain by 200 m of Ordovician shale (Jensen et al., 2009; Mazurek, 2004). In this particular case, the hydraulic conductivity of the argillaceous limestone and the shale is very low and the rock formation is able to ensure the confinement of radionuclides. Techniques for the survey of the host formation are identical to those implemented in the case of more argillaceous host formations. The major aims of a survey carried out on an argillaceous sedimentary site are (1) to identify the geometry of the host formation, i.e. its thickness and sedimentary or tectonic boundaries, (2) to reconstruct the depositional environment and (3) to assess the variability in space of the physical characteristics of the rock. The size of the studied site varies according to the geographic and geological context and the concept and waste inventory of the disposal facility. This site covers the footprint of the facility and a zone extending from several hundred metres to several kilometres around the footprint. This chapter illustrates the site surveys carried out in the framework of deep geological disposal projects in clay formation. It is essentially focused on surface techniques and technologies. It mainly relies on the French, Swiss


Clay geological repository systems

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and Belgium experiences. The chapter shows examples of lessons learned in following the practical applications of safety requirements (IAEA, 2007) or safety guides (IAEA, 1994) internationally acknowledged.

6.2

Specific features of a clay site survey

6.2.1 Characteristics of the lithology and consequences for the confining properties The confining properties of argillaceous rock formations come from their texture, the nature and relative proportion of their minerals, their low permeability and their deformation properties under stress. Regarding migration and retention properties, the differences encountered between various rock formations of this type depend on their porosities, i.e. on the degree of compaction and diagenetic cementation, and on the relative proportions of the minerals (Samper et al., 2008; Wersin et al., 2008). Clay minerals consist of alumina-silicate crystals in the form of platelets, which trap water molecules. These platelets are electrically charged at their surface with a negative charge. Thus, they are able to retain positively charged ions (sorption process). The negatively charged ions are pushed back from the platelets, reducing thereby the space between the clusters of platelets in which they move (Sammartino et al., 2003) and thus delaying the migration. The pore waters of such rocks are generally reducing and have circumneutral pH values. These environmental conditions are efficiently buffered by the mineral assemblage (Gaucher et al., 2009). Carbonates are in equilibrium with the carbon dioxide dissolved in water. These minerals regulate the pH of the medium, ensuring thereby the chemical stability of the pore water (Vinsot et al., 2008a). The pore water composition also controls the solubility limits of the radionuclides. One of the difficulties met in the characterisation of argillaceous sites is to collect in situ interstitial water unperturbed by sampling conditions. In the survey phases carried out in boreholes from the surface, this characterisation is made on interstitial water extracted from solid samples. The analysis of this water, associated with modelling taking into account the fluid/rock equilibria, provides a good estimation of the pore water composition (Gaucher et al., 2009). Recent results obtained through direct sampling in dedicated boreholes have significantly enhanced the knowledge of the phenomena constraining the chemical composition of pore water (Vinsot et al., 2008b). The porosity of argillaceous rocks greatly varies accordingly, namely to the degree of compaction and diagenesis of the sediments. As an example, porosities of the Opalinus Clay at Mont Terri and Benken, in Switzerland,


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and of the Callovo-Oxfordian clays at Bure, in France, are 12 to 18% in volume (Mazurek et al., 2008). In more compacted clay rocks, it can be much less. In the Callovo-Oxfordian formation the pore size is about 10 to 100 nanometres. This particular geometry accounts for the very low hydraulic conductivity values of these rocks (1012–1014 m/s; Delay, 2007). Consequently, in clay rocks, the chemical elements dissolved in water migrate mainly through diffusion, i.e. through the effect of their own movement, rather than advection, i.e. carried by the flowing water. In a natural environment located at a depth of a few hundred metres, this porosity is usually water saturated. However, a gaseous phase may exist and the size of the pores act as a capillary barrier. In the case of a nuclear waste disposal facility, gas migration mechanisms are also studied since the excavation of the drifts will lead to partial dewatering of the rock, and some components of the disposal facility will release gases. These gases, mostly hydrogen produced by the corrosion of metals, may increase the pore pressure and may also lead to dilatant micro-fracturing in the vicinity of the repository (Marschall et al., 2005; Boulin et al., 2008).

6.2.2 Mechanical and thermomechanical properties of clay rocks Owing mainly to their high content in clay minerals, argillaceous rocks display a complex rheological behaviour. Indeed, as a rule, argillaceous rocks present a low mechanical resistance and a poor ability to conduct heat. However, their creep and swelling properties lead to the self-sealing of fissures or fractures (Bastiaens et al., 2007; Bock et al., 2010). This selfsealing may lead to nearly complete sealing of the excavation-damaged zone of the underground structures after repository closure. The proportions of carbonates or silts, the water content or other specific features may significantly modify each of these mechanical properties. Thus, the Opalinus Clay studied at the Benken site and the Callovo-Oxfordian clays studied at Bure behave differently according to their mechanical strength (uniaxial strength). The strength of Opalinus Clay decreases with an increasing carbonate content, whereas it increases in the case of the Callovo-Oxfordian clays. These different behaviours can be accounted for by micro-structural variations. Results of carbonate micro-fabric investigations confirm that Opalinus Clay contains mainly coarse-grained shell fragments, while the Callovo-Oxfordian clays show a more homogeneous distribution of fine-grained fragments, mainly non-biogenic carbonates. In conclusion, not only their carbonate content but also their grain size distribution, their shape and spatial distribution control the mechanical



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behaviour of clays (Klinkenberg et al., 2009). Besides, an increase in quartz content increases the thermal conductivity of the rock. However, the mechanical behaviour of an underground structure depends not only on the intrinsic properties of the rock but also on specific factors such as the depth of the structure and its orientation with respect to the natural stress field. Excavation conditions, support and lining also play a role (Blu¨mling et al., 2007).

6.2.3 Characteristics required from the geological environment Besides the intrinsic characteristics of the argillaceous rock, the host formation and overlying and underlying formations must display certain favourable characteristics in order to be considered as a possible site for hosting a nuclear waste disposal facility. A conceptual geological model has therefore to be developed to establish the favourable nature of the geological environment. This conceptual geological model describes the evolution of the host formation from the phase of sedimentary deposit up to its present state. This model must also be able to describe its future evolution over the time required for the demonstration of repository safety. Firstly, this model is geometrical since it determines the dimensional characteristics of the argillaceous host formation and overlying and underlying formations. This model is also genetic since it helps to reconstruct the burial history of the rock and its micro-structural and geochemical evolution. Finally, this model is dynamic since it attempts to describe the evolution of the pore water migrating through diffusion in the host formation and through advection in more permeable formations. Over the area covering the disposal footprint, the argillaceous host formation must be: .

Thick enough to ensure the long-term confinement of radionuclides. Large-scale hydraulic conductivity must therefore be assessed through tests or through a set of pressure measurements carried out over a long period of time. Since the concentration profiles of natural solutes in the clay pore water (ions, isotopes, gases, i.e. so-called natural tracers) provide information on the vertical transport (Fig. 6.1) (Nagra, 2002b; Bossart and Thury, 2008), the transport mechanisms can be assessed by analysing their concentrations in pore waters extracted from cores. The CLAYTRAC project (Mazurek et al., 2009) on natural tracer profiles across argillaceous formations evaluates the relevance of natural tracer data in the understanding of past geological evolution. Data from nine sites have been analysed to support the scientific studies of the geological barrier and the feasibility of a safe geological repository.


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6.1 Chloride concentration profile in the Opalinus Clay layer at Mont Terri (Switzerland). This profile is accounted for by the diffusion mechanisms of ions in the layer and the difference in concentrations with the Trias and Dogger pore waters at the boundaries of this layer (Mont Terri Project – swisstopo Reports, from Bossart and Thury, 2008; reproduced by permission of swisstopo).

6.2 Burial history of the sedimentary formations in the Zu¨rcher Weinland (Switzerland) (from Nagra, 2002a, 2002b; reproduced by permission of swisstopo).

.

.

Deep enough to ensure long-term protection against erosion and climatic surface phenomena. The recent and past evolution of a potential siting area must therefore be analysed with respect to glacial and fluvial erosion, sedimentation and seismicity arising from crustal movement seismicity (Nagra, 2002a). Homogeneous enough with limited tectonic discontinuities and litholo-



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gical variability. This analysis relies on the paleogeographic reconstruction of the depositional environment and the geological history over millions of years (Fig. 6.2). The tectonic setting of the site must also be analysed in order to identify more specifically the past tectonic evolution and the different tectonic stress fields that may have existed or may still exist in the formation. The analysis of the past evolution helps us to understand the sequence of occurrence of faults or folds, and their amplitude. This analysis leads to an assessment of the future tectonic evolution of the site and the hydrogeological role of the tectonic structures and the consequences on future radionuclide release into the geosphere. The simpler the structure of the geological formations, i.e. plane, homogeneous and distant from active tectonic zones, the easier it is to apprehend their characteristics as a whole. The tectonic intra-plate zones of this type of formation are considered as the most favourable for a disposal facility. This comprehensive characterisation of a site survey must ensure that nothing will interfere with the integrity of the repository. The surface area of the studied sector may cover scales of up to several hundred square kilometres (Fig. 6.3). More broadly, all types of impact, not only radiological, are of interest. For instance, the construction and operation of a disposal facility must have no impact on existing natural resources (ground water, geothermal resources, iron ore, oil and gas exploitation). This involves identifying natural resources and assessing their potential.

6.3

Survey tools

Geological characterisation methods fall under three categories: (1) surface methods, (2) survey and observation boreholes, and (3) survey drifts. Specific techniques and technologies are associated with each method to describe or measure the characteristics of the underground. These methods have been developed over time by oil (Downey, 2004) or mining exploration companies (Petit-Dominguez et al., 2008), by civil engineering companies in the framework of the construction of tunnels and by organisations in charge of the management and protection of hydrogeological or geothermal resources. More recently, drift excavation technologies have also made headway, thanks to the refinement of techniques applied in the framework of large infrastructural works such as hydroelectrical installations (Sapigni et al., 2003).


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6.3 Potential implantation of a disposal site in the Benken region (Switzerland) and boundaries of the sector surveyed through threedimensional seismic and investigation boreholes (from Nagra, 2002a, 2002b; reproduced by permission of swisstopo (BA091588)).

6.3.1 Surface methods Surface-based geological methods, such as outcrop mapping and microtectonic analyses, are used to characterise the sedimentary geologic setting with particular attention given to interpreting stratigraphic relationships, geometry and structure (i.e. faults and fracture networks). These methods can be complemented by near-surface ‘light’ geophysics acquisition (e.g. electrical recording, radar, etc.) in order to determine specific layer geometry



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and localise discontinuities. The geologic models supported and derived from these data are used, in part, for the static correction of two- and threedimensional seismic data necessary to compensate for velocity variations in overburden. Two-dimensional and three-dimensional seismic methods are used primarily to reconstruct the geometric framework of the geological media (bedrock stratigraphy, presence of faults or fractures) (Lavergne, 1986). Results are available in the form of line data (two-dimensional) or blocs of data (three-dimensional). At depths commonly considered for a geological waste disposal facility (a few hundred metres) these methods can reveal the presence of faults a few metres in vertical throw. Three-dimensional technical resolution is better than two-dimensional as it relies on the spatial coherence of data sets and emphasises the identification of low-amplitude discontinuities (Birkhaüser et al., 2001; Andra, 2005). Stratigraphic inversion or analysis of seismic data has been used to infer petrophysical attributes such as images of porosity variation in carbonates and siltstones. However, in argillaceous formations, only a qualitative estimate of variability can currently be made on the basis of such data. A seismic survey requires the completion of a well calibration to identify the depth of seismic reflectors and the related geological layers, i.e. to convert the time information into depth information by a reliable velocity model (Boyer and Mari, 1997; Birkhaüser et al., 2001). Finally, in order to complete this inventory of surface methods, the installation of a seismological measurement network should be included in this category. This network provides recordings of the natural seismicity and contributes to the general survey of the site.

6.3.2 Drillings and borehole measurements The design of a drilling programme starts with the ranking of the desired objectives and information. Ideally, one should be able to collect simultaneously in the same borehole the maximum information on hydrogeology, rock mechanics, geochemistry, geophysics, etc. However, the technical requirements specific to each of these domains are sometimes incompatible and this set of information often cannot be obtained in a single borehole. Thus, the inability to carry out all the measurements in the same borehole involves drilling several dedicated boreholes with specific technical construction characteristics (cored boreholes, destructive boreholes), as well as designing a measurement programme and a completion adapted to the known or assumed geological and hydrogeological conditions. In practical terms, one of the issues in clay drilling is borehole stability. This stability depends, in part, on the techniques used for drilling and the associated mud.


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Polymer muds are commonly used in sedimentary rocks, but clay drilling is often more efficient with oil-based mud or silicate polymer mud.

6.3.3 Underground structures The construction of a survey shaft or drift is a common means to survey a geological site considered for the construction project of a tunnel in difficult conditions. In the case of studies dedicated to the construction of a radioactive waste disposal facility, pilot laboratories are also built with the view to acquire complementary data and clearing uncertainties regarding the behaviour of the formation and the stability of the underground structures. This type of survey has been conducted in methodological laboratories at Mol, in Belgium (Mertens et al., 2004), and Mont Terri, in Switzerland (Bossart and Thury, 2008), as well as in the Andra laboratory at Bure (Delay et al., 2007a). The construction follow-up of shafts, drifts and ramps greatly enhances the knowledge of the geological environment by providing a change of scale in the sedimentological and structural observations (Langer, 1991). In addition, a good assessment of the ground water productivity and analysis of the producing horizons (porous or fractured horizons) can be made on the basis of water inflows.

6.4

Survey strategy

6.4.1 Analysis of available data Preliminary studies for setting up a geological survey programme include the analysis of all available data (Stewart, 2002). Geophysical, climatic, hydrological, hydrogeological, seismic and other data are collected in addition to geological maps established with more or less accuracy according to the region. This may include information derived from historic oil drilling or seismic campaigns conducted within the sedimentary basins.

6.4.2 Definition of the survey zone boundaries The boundaries of the areas and depths to be surveyed are determined by the predictable extension of radionuclide migration over very long times and the occurrence and location of potential discharge zones; i.e. the potential future changes of the hydrogeological system have also to be taken into account. These potential discharge zones may be controlled by large extension (or regional scale) hydrogeological systems (McEwen, 2007). These constraints require study areas of several thousand square kilometres.



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6.4 Extent of the Boom Clay and locations of the investigation boreholes (Belgium) (from Wemaere et al., 2008; reproduced by permission of CEN SCK).

The data necessary for carrying out such large-scale analyses can be found in scientific or technical papers published by national geological surveys. In practical terms, on-site surveys are restricted to a sector limited by faults, outcrops of layers and hydrological limits (river or coastal boundary), which constitute the local limits of regional hydrogeological models. Thus, in France (Delay et al., 2007b), regional hydrogeological studies covered the Paris Basin as a whole but surveys carried out by Andra in 2003 in a preliminary survey phase only covered a sector of about 700 km2. During this campaign, a total of eight boreholes were drilled from five platforms distributed throughout this sector. Figure 6.4 shows, for the Mol site (Belgium), the zone covered by the survey and locations of the boreholes used to build the hydrogeological model (Wemaere et al., 2008). If the sector is large enough, the repository location can be optimised within a less extended zone considered as favourable. In the case of the French project, this zone is called the ‘transposition zone’, i.e. a zone where the results obtained at the underground research laboratory may be considered transferable, both with respect to the formation-specific confining properties and to the nature


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6.5 Map covering the sector studied in France. Location of the Bure underground research laboratory (URL) and boundaries of the transposition zone. Survey grid (seismic lines) used at Bure to determine the level of knowledge on the transposition zone and define the locations of new boreholes. E and F drilling pads are outside the boundaries of the transposition zone (source: Andra).

and occurrence of perturbations induced by the construction and operation of the repository (Fig. 6.5).

6.4.3 Selection of the survey grid and key parameters of the tools adapted to this grid The amount of information to be acquired on the defined perimeter is determined according to several criteria. Most important is the knowledge



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about the variability scales of the geological environment. This variability is sedimentary (characteristic length of the sedimentary bodies, extension of the various facies, etc.), but it also comprises structural features (density, length of the structures, etc.). Based on this information, and taking into account available data, it is possible to determine the means and different associated survey grids necessary to fulfil the characterisation objectives. The survey campaign conducted in France in 2007 and 2008 on the transposition zone illustrates this approach. It relies on a 3 km63 km grid of the zone. The length of 3 km was retained not only because it corresponds to a characteristic frequency of the variability of the Callovo-Oxfordian layer (second-order sedimentary variations) but also of the underlying and overlying formations. This survey area is also consistent with the area in which surface and subsurface investigations were conducted during the development and operation of the underground laboratory at Bure. The approach was implemented based on an uncertainty analysis following previous campaigns. Each square was assigned an uncertainty with respect to the others according to the quantity and quality of available data. The objective was to develop a survey approach that would yield data as consistent as possible throughout the transposition zone. The differences between the squares of the grid were minimised by adding complementary ‘survey points’ corresponding to the boreholes. The result converged towards the implantation of four drilling platforms (A to D, Fig. 6.5) within the transposition zone. Two other platforms (E and F, Fig. 6.5) were established according to specific geological criteria related to the fractures of the underlying and overlying formations outside the transposition zone. Since vertical steep structures are under-represented in vertical boreholes, a two-dimensional seismic grid was designed to investigate such a structure with the knowledge of typical fault characteristics and geometry. In order to calibrate the seismic records according to borehole data, these seismic lines pass through the locations of the drilling platforms.

6.4.4 Specific features Detailed studies may be conducted on specific objects if they can play a role in the construction, operation or long-term safety of the disposal facility. They may be structural objects, such as faults or fractures located within, or at a distance from, the footprint of the disposal facility, or they may be current phenomena such as seismic activity and topographic modifications linked, for instance, to glacial rebound (Boulton et al., 2004). The scope of the characterisation studies will be adapted according to the specific features of each disposal project. However, certain recurrent themes of study must be taken into account in the design programmes from the beginning. For instance, specific to clays, the anisotropy of the miner-


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alogical structure linked to the orientation of the minerals and the conditions of their sedimentation may lead to anisotropies of the confining properties, such as the hydraulic conductivity and diffusion coefficient (Wenk et al., 2008). The issue of ‘abnormal’ pressures must also be addressed at the design stage of the measurement and analysis programmes in settings with clay rocks. ‘Abnormal’ pressures are pressures measured in the ground that are not induced by hydraulic head or salinity gradients. Indeed, it is quite common to find ‘abnormal’ pressures (over- or underpressures) in clays. This phenomenon, identified and well documented, thanks to deep oil exploration boreholes (Hunt, 1996 pp. 290–320), is often assigned to compaction or gas production phenomena encountered at great depths. Other in situ and laboratory studies conducted on these subjects (Neuzil, 2000; Rousseau-Gueutin et al., 2009) have led to consider osmotic components, at least partially, for overpressures or underpressures. However, it is likely that in most cases these ‘abnormal’ pressures are caused by coupled phenomena. These phenomena may be hydrodynamic with biphasic components (capillarity), hydromechanical, thermohydromechanical or thermo-osmotic. Modifications of the hydrodynamic limits induced by climatic (glaciation/deglaciation) or topographical changes (erosion) could partially account for the occurrence and/or persistence of these situations. Whatever the reason, the existence of ‘abnormal’ pressure heads reveals the occurrence of low formation scale permeabilities (Fig. 6.6).

6.5

Technologies

Geoscientific technologies and measurement tools are under continuous development and numerous specialised journals and professional associations provide information on the latest enhancements. The sections below describe the main technologies used in this field but only elaborate on those that have undergone a specific development relevant to the survey of a radioactive waste disposal site.

6.6

Geological mapping

Establishing or, more broadly, updating geological maps is an essential stage for any disposal project. If observation conditions are favourable, tectonic and sedimentary surveys using light surface geophysics (electrical, VLF, light seismic, radar) are commonly undertaken. A combination of these measurements determines the stratigraphy of the layers and may reveal structures such as buried valleys infilled to depths of tens to hundreds of metres by drift (Jørgensen et al., 2003). When the surveyed sector is too vast,



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6.6 Pressure and permeability profiles obtained within the Paleozoic sedimentary sequence beneath the Bruce site, Ontario, Canada. Environmental head profiles have been corrected for salinity. Anomalous under- and overpressure conditions, indicative of low formation scale bedrock permeabilities, are observed in the Silurian, Ordovician and Cambrian sediments (reproduced by permission of NWMO).

the aim of the geological mapping is limited to identifying the zones of fractures or sedimentary features.

6.7

Geophysical seismic surveys

These technologies are most commonly used by oil companies for the survey of deep levels (over 1000 metres). They must be adapted to the characterisation of argillaceous sites at depths of several hundred metres. Attempts have been made at preserving the high frequencies of the signal in order to increase the resolution power. In the same way, the corrections of surface velocity anomalies (static corrections) must be the object of particular care since, when they are correctly applied, the quality of the seismic imagery in the first few hundred metres can be significantly improved. Usually, acquisition parameters are defined according to the objective characteristics (nature, depth, continuity, desired resolution, etc.). Most often, they rely on the feedback of previous seismic campaigns conducted within the proposed area. One or two days should therefore be dedicated to


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on-site tests, prior to acquisition, in order to verify and optimise these parameters. As regards acquisition parameters, the data processing sequence and interpretation must be adapted to the specific features of the site. Besides the static corrections mentioned above, the velocity analyses must be the object of a strict follow-up.

6.8

Drilling

6.8.1 Drilling programme Once the geometry of the studied formations and the survey grid are defined, drilling is a rapid means for collecting information on the physical and chemical nature of the rock and interstitial fluids. Depending on the characterisation objectives, the boreholes can be vertical, slanted or deviated; vertical boreholes are suited to the study of tabular structures, deviated or inclined boreholes to the study of subvertical fractures and subhorizontal or directional boreholes to the survey of future drifts axes. Cored boreholes Cored boreholes provide a detailed knowledge of the geology and direct observation of the petrophysical, sedimentological and structural features. In addition, thanks to an adapted sampling, a detailed mineralogical characterisation of each horizon can be achieved in the laboratory as well as various direct measurements such as density, porosity and permeability. In clay formations, coring is carried out with a drilling mud that ensures the stability of the borehole walls, the lifting of the cuttings and the cooling of the tool. However, drilling mud creates a ‘cake’ along the walls, which induces a skin effect that conceals the hydraulic properties of the geological formation. The fluid used for hydrogeological measurements is, if possible, a polymer-based mud. For stress measurements, non-hydrated rock core samples and a very high-quality borehole wall surface can be obtained in most argillaceous levels using oil-based mud. Percussion-air drilling In general, percussion drilling is implemented in aquifers overlying and underlying the host formation, and the boreholes are often dedicated to hydrogeochemical characterisation. Thus, fluids are collected in these formations to provide data for the regional hydrogeological and hydrogeochemical models.



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The main constraints of boreholes dedicated to hydrogeochemical sampling are as follows: .

.

The drilling technique must be as little contaminating as possible with respect to the interstitial fluids extracted through pumping. Overpressures must be avoided to prevent in-depth invasion of the formation by a possibly contaminating fluid. All fluids and their additives used at the drilling stage (grease, oil, water, etc.) must be sampled eventually to trace the source of chemical pollutants. All the equipment used for drilling or for hydrogeological tests must be clean (physically and chemically). This equipment must release no product or substance liable to alter the chemical quality of the water.

Percussion-air drilling is recommended for short boreholes (less than 200 metres deep) in usually productive formations. This type of drilling is rapid and only requires light equipment and tools commonly used by drillers in the exploitation of water resources for human and agricultural needs. For deep boreholes, i.e. depths ranging between 200 and 800 metres, the most effective technique is reverse-air circulation percussion drilling. With reverse-air circulation the cuttings are lifted inside the rods instead of the hole/tube annular. Thus, since they have not been in contact with the overlaying formations liable to contaminate them, the rock cuttings and formation fluid cuttings reaching the surface are more representative of the formation (Fig. 6.7). Geological programme The geological programme comprises the analysis of the cuttings or cores and the follow-up of the drilling parameters. The monitoring of mechanical and hydraulic parameters while drilling provides information on the characteristics of the drilled layers. When drilling mud, it is standard practice to monitor the levels of the mud tanks (total losses, inflows) and the presence of gas. There are also devices for monitoring the chemical composition of the mud. In cored boreholes the tagging of the mud with chemical elements or fluorescent tracers allows a precise management of the mud flows during the drilling and pumping phases. It enables quality control of the samples dedicated to geochemical analysis. These parameters are recorded and displayed in real time using data acquisition and processing systems provided by ‘mud-logging’ companies. This real-time analysis is necessary in the case of an exploration borehole as monitored parameters can vary rapidly. A constant rate of drilling and maintaining other borehole parameters as steady as possible will ensure the best quality measurements.


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6.7 Principle of the reverse-air percussion drilling technique (source: Andra).

6.8.2 Geophysical wireline logging Well logging allows a detailed record of the geologic formations penetrated by a borehole. The log is based on physical measurements made by geophysical tools lowered into the hole. Historically, the first logging measurements consisted of basic electrical logs (resistivity) and spontaneous potential (SP) logs. They were used to identify oil bearing formations. In the case of an investigation survey they are also used to evaluate the zone around the borehole invaded with the drilling mud.



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Electric, sonic and porosity tools, as well as tools for measuring natural gamma radiations and gamma–gamma density radiations, are used to estimate the physical properties of the rock formations. These indirect measurements are also made accurately to determine the depth at which data are obtained through coring. Thus, sonic and acoustic tools combined with density measurements are used to calibrate the reflection seismics. High-resolution electrical or acoustical imaging logs are used to visualise the formation, compute the formation dip and analyse thinly bedded and fractured geological layers. With a high-quality resolution, they can display sedimentary features such as porous horizon distribution or clay layer orientation and organisation. Two main types of tools are available to measure the porosity of the drilled formations: neutron source tools and nuclear magnetic resonance (NMR) tools. NMR tools must be retained for borehole porosity measurements since neutron source tools alter the natural carbon-14 content of the formation fluids. Besides a total bulk porosity value, the NMR tool also provides free fluid and bound water porosity estimates, as well as a qualitative estimation of the permeability (Coates et al., 1999). The analysis of these measurements made in different types of borehole (Fig. 6.8) brings an understanding of the formation lateral and horizontal variability (Lefranc et al., 2008).

6.8 Interpretation of natural gamma-ray logs. Example of a lithostratigraphic correlation in the Callovo-Oxfordian clay formation at Bure (Andra, 2005).


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6.9 Example of an interpretation of hydrogeological logs during a pumping test in a Dogger borehole at Bure. It is to be noted that an important inlet is identified through thermal logging and spinner and two secondary inlets are identified through fluid logging (source: Andra).

6.8.3 Hydrogeological logging tools Two types of wireline tools are used in the hydrogeological logging programmes: (1) flow logging tools, i.e. mechanical, thermal and chemical tools for locating and quantifying the flow zones and (2) fluid sampling probes. Flow logging tools can be used without pumping to highlight the differences in heads and flows between permeable levels. Generally, flow loggings are associated with a pumping phase to characterise water inflows with transmissivities ranging between 109 and 105 m2/s. Measurement of the water level in the borehole during the pumping phase also provides data to estimate a global transmissivity value. This value is compared with the individual measurements obtained through wireline flow loggings. Figure 6.9 shows loggings recorded in the pumping phase associating various techniques in the same borehole. Flow logging Flow logging is used in boreholes when vertical water fluxes generate velocities higher than ca. 5 mm/s. Each of the inlets can then be identified and, assuming a constant head throughout the tested aquifer, the



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transmissivity value of each zone can be deduced. This method is considered well suited to boreholes with transmissivitiy values ranging between 107 and 105 m2/s. Different types of borehole flow measuring tools are available on the market. The most widely used are hydrogeological micro-spinners. However, hydrogeological tools are more accurate than oil tools because the spinner is replaced by a turbine with a much better sensitivity and resolution. Electric conductivity fluid logging Fluid conductivity logging (Tsang et al., 1990) has been developed in Switzerland in the framework of regional siting programmes for high-level waste. After the drilling phase, the mud in the borehole is replaced with water with a contrasting electrical conductivity with respect to the pore water. During a pumping phase, electrical conductivity logs show spikes at the inflow points. These peaks broaden and move along the borehole according to the inlets flow rates. In the analysis a numerical inversion simulation is used to obtain the locations and rates of the inlets. Fluid conductivity flow logging has been successfully applied in the Bure area boreholes with transmissivity values ranging between 109 and 105 m2/s (Delay and Distinguin, 2004). Heat-pulse flowmeters Heat-pulse flowmeters are used to measure flows using either the travel time of a thermal pulse or the dilution of heat released by a continuous heat source. A Finnish adaptation of the method has been implemented to characterise the detailed heterogeneity of flow rates ranging between about 0.2 ml/min and over 1 l/min (O¨hberg and Rouhiainen, 2000). Thus, for 25cm diameter boreholes (suitable for a 4-inch pumping device) the transmissivity covered by this type of tool ranges between 108 and 106 m2/s. Samplers When sampling cannot be made from the surface at the outflow of the pumping device, samplers are handled with a cable. The most sophisticated samplers can bring out fluids at the bottom-hole pressure without modifying the fluid/gas ratio. Sampling with a cable consists in lowering into the borehole, at a given depth, a fluid sampling device. The sampling is triggered off either by remote


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control from the surface or by an integrated time switch. There are numerous types of samplers: standard, high pressure (which keeps the sample under pressure), large size and inert (with a special lining of the sampling chamber, e.g. Teflon).

6.8.4 Packer tests For borehole measurements of low hydraulic conductivities in clay rocks, owing to low flows, the borehole sections to be characterised must be isolated with inflatable packers. In these measuring chambers, different types of equipment are used to apply a hydraulic pulse in underpressure or overpressure conditions. Isolation of the test interval is achieved by inflating the packers. Following the isolation of the interval from the lower and upper parts of the borehole, a variation in pressure is applied to the interval (slug or pulse test). The recovery of the pressure or water level to its initial state is recorded. The interpretation of the pressure variations during the recovery of the initial state, according to time, helps to determine the permeability and storativity of the rock formation (Hsieh et al., 1981). The tubing which supports the down-hole equipment constitutes the ‘test string’. The packer consists of a set of sensors (pressure, temperature) and devices (shutting tools, etc.). A surface control and acquisition system is used for recording and displaying in real time all the parameters collected at the bottom of the borehole (pressure, temperature, flow rate, conductivity) and for controlling the tools (opening and shutting of the valves, inflating and deflating the packers). Usually, the test string is lowered into the borehole to carry out a set of tests, starting with the deepest one. When a test is completed the packers are deflated and the system is moved up to the next test interval, and so on. Borehole slug tests A slug test is carried out to assess the permeability of the formation in the immediate vicinity of the borehole for values ranging between 107 and 1010 m/s. A slug test consists in monitoring, in the tubing supporting the down-hole equipment, the return to equilibrium of a water column initially in equilibrium with the formation. In the case of an overpressure choc, the test fluid in the borehole penetrates into the formation, whereas in the case of a depression choc the flow is reversed and the formation fluid enters the borehole. The latter option must be retained whenever borehole conditions are favourable so that injecting the test formation with a foreign fluid can be avoided.



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Borehole pulse tests A pulse test is carried out to assess the permeability of the formation in the immediate vicinity of the borehole for values ranging between 109 and 1014 m/s. Pulse tests require a device that can apply a choc and equipment to measure the pressure in the test interval when the flow is insignificant. The reliability of a pulse test requires assessing as accurately as possible the volume of the fluid involved in the test. This volume depends on the compressibility of the down-hole equipment and the physical properties of the fluid. The equivalent compressibility of the system therefore depends on the compressibility of the fluid (test fluid) and equipment (metal parts) in the test chamber, on the packers (rubber + inflating fluid), as well as on the elasticity of the borehole walls in the test zone.

6.8.5 Geomechanical tests: acquisition of a stress profile Several technologies are available to determine the stress state throughout a sedimentary sequence. The variations in horizontal and vertical natural stresses are the consequences of the rheological properties and geometrical characteristics of the stratigraphic layers as well as the loading history of the rock mass. Technologies typically used in deep boreholes in the framework of characterisation surveys are described in the following paragraphs (Wileveau et al., 2007) (Fig. 6.10). Other methods, such as overcoring, used in mines or underground laboratories are not well adapted to deep boreholes. Traditional hydraulic fracturing test and the hydraulic test on pre-existing fracture (HTPF) method The hydraulic fracturing technique is used to estimate the value of the minimum principal stress (Haimson, 1993). It relies on the fact that a hydraulic fracture will be created and will propagate normal to the minimum principal stress. The pressure at which the fracture closes, i.e. closure pressure, is an estimate of the minimum principal stress. When the borehole is parallel to a principal stress, the normal to the fracture plane is the orientation of the minimum principal stress. Acoustic and electrical wireline logging tools are used to detect the fracture dip and orientation. Borehole images are acquired before the tests and compared with images acquired after the test. Instead of creating a new fracture, the same methodology may be used to determine the stress component acting normal to a pre-existing plane of weakness, either a bedding plane or a pre-existing fracture. This is the so-


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6.10 In situ stress profile established at Bure (France). It is to be noted that when the content in clay is high (Callovo-Oxfordian), the minor and vertical constraint values are very similar (source: Andra, from Wileveau et al., 2007).

called HTPF (hydraulic test on pre-existing fracture) method (Cornet, 1986). Combined with hydraulic fracturing information, such tests help to determine additional components of the stress tensor. Sleeve fracturing test The sleeve fracturing test consists in the strictly mechanical reopening of a hydraulic fracture. This test evaluates the stress concentration at the borehole wall in a given orientation. Desroches and Kurkjian (1999) propose to interpret this test with the same assumptions as those proposed by Haimson (1993) to interpret reopening fractures in low-permeability rocks. Systematic analysis of breakouts in boreholes Another method consists in analysing the position and orientation of breakouts in deviated boreholes of different orientations (Brudy and



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Zoback, 1999). In deviated boreholes, the orientation of breakouts along the hole provides information on the shape of the stress ellipsoid (Peska and Zoback, 1995; Haimson, 2007). This information provides a ranking of the stress components and an estimate of the maximum horizontal principal stress, when knowledge about the minimum horizontal stress σh (sigmah) and the vertical stress σv (sigmav) is available.

6.8.6 Borehole long-term monitoring equipments Once the borehole and associated in situ measurements are completed borehole instrumentation can be installed. Generally, such completions are installed to observe the stabilisation of formation pressures, which in lowpermeability sediments may occur over periods of 1 to 2 years. It takes, for instance, one or two years for a hydraulic head to stabilise in a deep borehole in a formation with a 1013 m/s average permeability. Hydrogeological equipment commonly installed in a deep borehole falls under three main types: .

.

.

Pressure sensors (sometimes associated with temperature and conductivity sensors) installed below the water level, or in depth in front of the inlets. This type of equipment is used in the most permeable geological formations. Pressure and temperature sensors installed in a concrete backfilled measuring chamber at the bottom of the borehole. These types of sensor are put in place permanently without the intent to retrieve them (Delay et al., 2007c). Their advantage is that they are perfectly isolated from the rest of the borehole, but their drawback is that they cannot be replaced, or recalibrated, in case there seems to be a problem with their functioning. Wireless or wireline data transmission has been used successfully. Multi-packer completions are used to follow up simultaneously up to twenty static levels in the same borehole. This type of device is relatively costly, but it is very effective even with low permeability environments.

Sometimes, other types of completion are installed with aims other than strictly hydrogeological ones. In France, a sophisticated completion was installed in a borehole from the surface at a depth of 537 metres to follow up the diffusion of radioactive tracers (Delay et al., 2007c). This equipment, installed in 2004, is still functioning in 2009. Finally, in certain geothermal contexts, innovative optic fibre temperature measuring devices have been installed in boreholes to detect transmissive formations or zones over the long term (Yamano and Goto, 2005).


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6.8.7 Workover operations When it is technically possible, it may be sensible to perform additional wireline logs or surveys. Such measurements can be particularly useful in quantifying the evolution of the geometry and temperature in the borehole, or changes in formation permeabilities. Workover operations are commonly scheduled in survey programmes to avoid costs associated with drill rig stand-by.

6.9

Underground structures surveys

Underground structures, whether built for a laboratory or a disposal facility, provide a unique opportunity to observe geological layers in large size and to carry out in-depth geological analysis. In addition, the geomechanical behaviour of the structures can be assessed in situ following the excavation of the front face. In the case of the construction of a deep disposal facility, a scientific follow up will be conducted to confirm the previous observations provided by surface investigations and will include the development of experimental niches, drifts or experimental cells or vaults. Any operation linked to the excavation or safety of the excavation worksite can be adopted by the investigation programme. For example, boreholes advanced beyond the excavation face to detect pressurised water or gas zones can be used to collect samples of uncontaminated groundwater in the more permeable formations.

6.10

Core lab analysis

When a borehole is cored, the samples are packaged in a specific way according to the type of analysis. Four main types of sampling and analysis are carried out in clays: .

.

Sampling for geological studies: mostly mineralogical and microstructural analysis. The identification of fossils and microfossils is also carried out to interpret the depositional environment (Thierry et al., 2006). Sampling for geomechanical studies. In clays, it is important that a sample will not undergo structural modifications due to decompression to maintain it as close as possible to its in situ state. Thus, following its extraction from the core barrel, the sample is stored in a special cell where it is cemented under load. The main elastic, poroelastic and thermal properties can then be determined in the laboratory (Homand et al., 2006).


Clay geological repository systems .

.

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Sampling for permeability measurements (permeability to water or gas). The samples, also packaged in a confining cell, are the subject of measurements carried out parallel and perpendicular to the stratigraphy (Delay et al., 2006). Sampling for geochemical studies. The samples are packaged in specific cells or gas-tight bags as soon as they are extracted from the core barrel. In these geochemical cells air is replaced either by vacuum or by a gas, which restricts the reactions with the rock minerals. Sophisticated leaching or squeezing extraction techniques are implemented in specialised laboratories to determine the geochemistry of the pore water (Gaucher et al., 2006). To extract the gases from the cores, the samples are usually packaged directly on site in air-tight cells and gas extraction is carried out on the spot.

6.11

Integration of results

The common goal of the characterisation studies, sample analysis and modelling is to provide documents in support of the licensing for the construction and operation of a nuclear waste disposal facility. It usually consists of three types of technical documents using data acquired on the sites: (1) a geosynthesis, (2) a disposal facility design and (3) a safety analysis. The current practice is to gather the analysis and field measurements in a geosynthesis that provides a global understanding of past and current functioning of the geological system. In parallel, the scientific results are used in the design of the disposal facility, i.e. in the description of its general architecture and the geometry of the modules and disposal cells. In this respect, the rock mechanics results are of particular significance for clay rocks. The geosynthesis is used to build the conceptual model integrating the confinement properties of the geological environment. The conceptual model serves as a frame to describe the evolution of the disposal facility through scenarios. These scenarios take into account the sensitivity of the geological environment to perturbations induced by the construction and operation of the disposal facility, including its closure and long-term monitoring stages, and finally the long-term geological evolution of the whole system during the post-closure phase. In addition, the safety functions, i.e. the objectives to be achieved by a disposal facility, are analysed in a safety analysis according to various scenarios.


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6.12

Future trends

In recent years, significant breakthroughs have been made by the various national and international agencies (NEA, IAEA, etc.) thanks to the surveys and research studies carried out in the framework of international collaborations. They are taken into account when updating safety requirements (IAEA, 2006, 2007). Numerous research programmes and partnerships carried out in the framework of European projects have also provided opportunities to share human and technical means as well as experimentation sites. In this respect, projects carried out at the Mol and Mont Terri sites have played a pioneering role in international cooperation by making available to the scientific community the first underground structures specifically dedicated to this type of research. Currently, the surveys of sedimentary basins and argillaceous environments are making rapid headway, thanks, in part, to the efforts of oil companies interested in CO2 sequestration in deep geological layers. In some respects, this subject is very close to activities carried out in the past ten years in the framework of nuclear waste disposal. As regards subjects more specifically related to nuclear waste disposal, in spite of successful attempts made in boreholes from the surface, the fields of experimentation dealing with the interactions between the geological environment (clays and shales) and disposal materials (glass, iron, cement), as well as subjects dealing with the sealing of shafts and drifts, will continue to be studied, for the greater part, in URLs and on samples in surface laboratories. The same applies to studies dedicated to transport and microbiological phenomena. Two other major scientific and technical axes must be developed from boreholes. The first one deals with the geomechanical characterisation of argillaceous formations. At the present, it is difficult to predict reliably over the long term, the behaviour of a shaft or a drift, and the extent of the damaged zone around these structures. The presence of hydromechanical couplings also raises numerous theoretical and experimental questions, regarding in particular the dewatering and rehydratation phenomena of clays. The second axis is the analysis of gas migration in argillaceous rocks. Whether in the frame of the analysis of natural gas migration or radionuclide migration induced by the generation of gas in the disposal facility, experiments carried out from the surface should be considered to acquire a better understanding of the elementary phenomena and hydraulic and geomechanical couplings.



6.13

181

Sources of further information

An exhaustive review of websites on existing or potential radioactive disposal facilities has been published by Rempe (2007). The reader will find there all the sources of information on all ongoing projects worldwide. The status of these projects is regularly updated by the DOE in a document called Geological Challenges in Radioactive Waste Isolation, the fourth edition of which came out in April 2006 (Witherspoon and Bodvarsson, 2006). The reader will find in this issue a synthesis made by each country involved in a disposal facility project. The agencies in charge of radioactive waste management bring out, according to the legislations or administrative processes, synthesis reports presenting the progress of their projects to their safety authority and their government. These reports are detailed syntheses of sites and projects. They are the objects of international peer reviews and are made available to the public. The most recent ones are: . . .

For Belgium, the SAFIR 2 (ONDRAF, 2001) report. For Switzerland, reports on the ‘Opalinus Clay’ project (Nagra, 2002a, 2002b, 2002c). For France, Dossier 2005 (Andra, 2005).

Two files are scheduled in France in the coming years, one in 2013 for Public Debate and the other in 2015 for the Licensing of the Bure site, and one in Canada in 2012 for Public Hearing regarding the Bruce site project. Information regarding the Canadian project on the proposed Bruce site Deep Geologic Repository may be found at http://www.nwmo.ca/. Numerous congresses, symposiums and journals regularly present stateof-the-art progress in the various fields of earth sciences. Scientific journals compile the approaches of the academic community, research agencies, waste producers and reviewers of research programmes. Specialised publications also take stock of the progress of scientific works. Some of the most recent are as follows: . .

The document published by the Mont Terri consortium: Mont Terri Rock Laboratory Project, Programme 1996 to 2007 and Results (Bossart and Thury, 2008). Papers published following the congress ‘Clays in natural and engineered barriers for radioactive waste confinement – Lille 2007’, organised by Andra (France), SKB (Sweden), Nagra (Switzerland) and ONDRAF (Belgium) (Special Issue 2008, Physics and Chemistry of the Earth; see the papers mentioned in the reference).


182 .

Geological repository systems for safe disposal Reviews in Engineering Geology XIX – Deep Geologic Repositories (Rempe, 2008) published by the Geological Society of America.

Regarding geophysics, and more specifically the two- and three-dimensional seismic reflection, the websites of the main contractors present their activities and major research works that are under way. Most of them offer integrated services, from acquisition to interpretation. Among others are: http://www.cgg.com/ http://www.westerngeco.com/ http://www.dmt.de/en/home.html http://www.beicip.com/

6.14

Acknowledgements

The bulk of this review is based on information provided by Andra, SCKCEN, Nagra and the Mont Terri Project. I warmly thank Andreas Gautschi (Nagra) and Mark Jensen (NWMO) for reviewing this chapter. I would also thank Isabelle Wermaere (SCKCEN), Paul Bossart (Mont Terri Project – Swisstopo), Ce´line Righini (Andra), Agne`s Vinsot (Andra), Lise Feuillaˆtre (Andra), Georges Vigneron (Andra) and Anne de Henning for their contribution to this work.

6.15

References

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7 Assessing the long-term stability of geological environments for safe disposal of radioactive waste K.J. WILSON and K.R. BERRYMAN, GNS Science, New Zealand

Abstract: The future stability of the earth’s crust is a critical factor in the site selection process for geological repositories of hazardous radioactive waste. The chapter compares the timeframes of geological repository operation with the timeframes of geological processes. The potential effects of tectonic instability and climate change are discussed and common techniques of assessing and modelling crustal stability are outlined. Key words: volcano-tectonic stability, geological repository, climate change and nuclear waste, long-term tectonic deformation, modelling tectonic stability.

7.1

Introduction

The stability of the earth’s crust over the lifetime of a nuclear waste repository is a critical factor in the site selection process for geological repositories. Events such as earthquakes and volcanic eruptions can have varying effects on the safe operation of the geological repository, depending on the size and proximity of the event. The effects may range from mild shaking to catastrophic rupture or exposure of the waste repository. Fortunately the principles and processes driving geologic instability, namely plate tectonics and climate change, are well understood. There are a number of techniques used by geoscientists to characterise the level of risk and to predict the long-term behaviour of the earth’s crust. In this chapter we review the timeframes relevant to nuclear waste repositories and compare these with the timeframes over which various geological processes operate. In general, geological repositories are expected to be effective (i.e. keep the hazardous waste isolated from the environment) 188 © Woodhead Publishing Limited, 2010

Assessing the long-term stability of geological environments

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for 100 000 to 1 000 000 years. Many methods have been developed for geoscientists to assess the present and short-term geological stability (e.g. 100 years) of a site, but assessing the long-term stability requires an understanding of how volcano-tectonic processes change and migrate over time. Processes to consider include the migration of existing fault zones into new areas, the cumulative impact of slow uplift and erosion, and the possibility of new volcanoes or geothermal zones developing. At timescales greater than several thousands of years the potential effects of climate change, such as sea level change and ice sheet loading, also need to be considered. We describe the geological and geophysical techniques that can be used to quantify past and present deformation, and also describe how models of long-term crustal stability are developed.

7.2

Long-term volcano-tectonic stability issues for safe disposal of radioactive waste

There are various timeframes over which the stability of geological environment needs to be assessed when considering potential geological repository sites and assessing their future performance. As geological repositories are constructed, filled and interned, and as the radioactive waste decays, there are changing levels of acceptable risk from geological instability. We will discuss the timescales over which the geological repositories are operational and contrast this with the timescales over which geological processes change or evolve. Construction and filling of the repositories is expected to take decades. During this initial period the waste is at its highest level of radiation and it is not fully sealed from the biosphere; therefore the hazard from the waste is at its maximum. Some geological repository projects require that the repository remain open for a certain period after filling for monitoring or retrieval of the waste. For example, the US Nuclear Regulatory Commission requires nuclear waste be fully retrievable for 50 years after emplacement and it is planned to keep the proposed Yucca Mountain repository open for 100 years after it is filled. Once a repository is sealed the time period over which it must provide protection depends upon the rate of radioactive decay and release of radionuclides. A common benchmark when assessing the hazard of the waste, if it were to be released into the environment, is the comparison of the waste radioactivity to the radioactivity of natural uranium ore, the reasoning being that once the waste has decayed to levels equal to that found in nature then it is no longer a hazard to the biosphere. The time over which nuclear waste decays to ‘natural levels’ depends on the nature of the waste; processed high-level waste (HLW) will decay to natural levels within a few


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thousand years, while spent fuel will take one to several hundreds of thousands of years (Chapman, 2006). Ten thousand years is generally considered a minimum time period that a geological repository will operate for and most organisations assess the stability of geological repositories over timescales of > 100 000 years. The USA has radiation standards for geological repositories based on millirem levels; a millirem is a measure of radiation dose absorbed by humans. New standards introduced in 2008 require repositories to have a radiation limit of 15 millirem per year for the first 10 000 years after disposal and 100 millirem per year between 10 000 to 1 million years (EPA, 2008). The Nuclear Waste Management Organisation of Japan (NUMO) focuses on a time period of 100 000 years for assessing tectonic stability (Apted et al., 2004). Geological processes affecting the stability of the earth’s crust are primarily related to earthquakes, volcanoes and climate change. The relevance of each geological hazard in relation to deep repositories varies over different timescales. For example, the probability of an ice sheet covering large parts of the earth in the next few thousand years is extremely low but the probability of earthquakes along plate boundaries is very high. Geological processes likely to operate in the short term ( 1 000 000 years. For example, the Snake River Plain/Yellowstone hotspot has migrated to the northeast at approximately 40 km/million years over the past 15 Ma (Nash et al., 2006). At timescales > 10 000 years climate changes in the form of natural glacial–interglacial cycles. These cycles are generally related to orbital cycles that operate on periodicities of 23 000, 41 000 and 100 000 years (Hays et al., 1976). Due to strains associated with the loading and unloading of ice sheets at high latitudes and global sea level changes on the order of 150 m, climate cycles can have significant effects on the stability of the earth’s crust. On timescales greater than several million years the location and nature of plate boundaries can change. For example, the Himalayas have formed since the Indian plate collided with the Eurasian plate approximately 45 million years ago (Fig. 7.2). A new divergent plate boundary, manifest as the East African Rift, has developed through Ethiopia in the past 20 Ma (Pik et al., 2008). Long-term climate fluctuations also occur on timescales of tens of millions of years. For example, the early Eocene period (55–40 Ma) and late Cretaceous period (100–65 Ma) are recognised as times of warmer temperatures, or ‘thermal maximums’ (Poulsen et al., 2003; Zachos et al.,


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7.2 Schematic diagram showing the northward migration of the Indian continental plate and its eventual collision with the Eurasian plate. This first-order plate boundary change occurred over timescales greater than several millions of years (source: adapted from USGS).

2008). Causes of these long-term climate variations may be related to factors such as plate tectonics (e.g. continent positions changing with respect to the poles and influencing oceanic circulation), mega-scale volcanism and methane release from continental shelves.



7.3

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Geochemical stability issues for safe disposal of radioactive waste

Knowledge of the geochemistry and flow patterns of groundwater are of critical importance to the effective operation of geological repositories. Both of these factors can influence the corrosion rate of the engineered barriers encasing the waste, plus the entrainment rate and flow path of mobile radionuclides that are eventually released from the repository. While highly impermeable host rocks are typically selected for geological repositories, the host rock will still contain some groundwater (the exception being evaporite formations, which are impermeable). Ideally, the groundwater will be anoxic (reducing) with very low rates and volumes of flow (Chapman, 2006). Geological processes that can affect groundwater geochemistry over time are primarily concerned with volcanic and geothermal activity, climate change causing salinity and oxygenation variations, and faulting affecting the rock permeability. In general nuclear waste repositories will not be sited in or near active volcanic zones. However, there can be the possibility of future volcanic intrusions into previously non-volcanic areas (Mahoney et al., 2009). A magmatic intrusion close to the surface of the earth’s crust (e.g. within 8 km) will cause significant changes to groundwater. The flow patterns will be affected by the convective circulation of heated water and geochemistry will be altered, particularly by mixing of near-neutral pH water and steam to produce oxidising ClSO4 waters (Henley and Ellis, 1983). Reducing conditions are favourable in geological repositories because many radioelements such as uranium, neptunium, technetium and selenium are more soluble in fresh oxidising groundwater. Many radionuclides vary in solubility according to pH; for example, 239Pu and 238U are less soluble under alkaline conditions (Nirex, 2001). Magmatic intrusions and geothermal heating in the vicinity of nuclear waste repositories therefore have the potential to alter the flow, oxidation and pH of the groundwater and this can affect the solubility and distribution of radionuclides. Climatic variations can affect the geochemistry of groundwater in several ways. One example is the changing salinity of groundwater caused by sea level changes. With changing salinity the performance of bentonite buffers is altered; compacted bentonite is a commonly proposed engineered barrier for nuclear waste (Arcos et al., 2003; Tanaka et al., 2007). On a short timescale anthropogenic-induced climate warming in the next several thousand years will probably cause a rise in sea level with associated salinisation of groundwater in coastal areas. Over longer timescales continental areas at high latitudes, such as on the Canadian and Fennoscandian shields, show evidence of hypersaline brines forming during glacial periods of the Pleistocene. The brines migrate through the crystalline basement in response


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to ice sheet loading and unloading (Starinsky and Katz, 2003). Strains associated with ice sheet loading have also been proposed as a mechanism for injecting fresh, oxygenated water into deep groundwater (Nirex, 2005a; Talbot, 1999). Fault rupture has also been suggested as a mechanism by which permeability pathways may be opened into deep repositories, potentially leading to a ‘freshening’ of the groundwater. Fault rupture will not necessarily create a permeable pathway through rock because the fault plane can be filled with a ‘fault gouge’ material of pulverised, very fine grained rock, which may not be permeable. However, there have been reports of water well levels changing in response to earthquakes at distances of > 1000 km from the earthquake epicentre, showing that groundwater circulation can be affected by earthquakes (Montgomery and Manga, 2003).

7.4

Potential climate change issues for safe disposal of radioactive waste

Climate and the geological environment are interrelated in many complex ways and therefore climate change can potentially affect the stability of the geological environment. In recent times, the consideration of future climate change is mainly concerned with global warming. There is evidence that the global average temperature is rising due to anthropogenic increases in carbon dioxide in the atmosphere (IPCC, 2007). There are also natural cycles of climate change related to orbital variations of the earth (also called Milankovich cycles); at present the earth is in an interglacial period but less than 20 000 years before the present it was in a glacial period. The timescales and effects of both anthropogenic-induced global warming and natural Milankovich cycles on geological repositories will be discussed. Evidence of global warming comes from a variety of datasets including rising average surface temperatures, rising sea levels and decreasing snow and ice cover (IPCC, 2007). These trends have been documented over the past several decades (with some records extending back to 1850 AD). Therefore the timescale of global warming is on the order of tens to hundreds of years, although predicting future trends remains controversial and typically has a large degree of uncertainty. There are many known and potential impacts of global warming but the phenomena most likely to impact geological repositories are those related to the coastal zone, namely rising sea levels and increased storminess. Predictions of sea level rise for the period up to 2099 AD range from 0.18 to 0.59 m (IPCC, 2007). Beyond the next century some scenarios incorporating ice sheet melting predict up to 7–10 m of sea level rise in the next millennium (Gregory et al., 2004; Kerr, 2006; Overpeck et al.,



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2006). Geological repositories located near the coastline and near sea level will probably see an increased frequency of seawater inundation due to sea level rise, e.g. the inundation of repository access tunnels or ventilation shafts. Accompanying sea level rise will be an increase in the inland extent of storm surge inundation. On relatively flat coastlines a small rise in sea level ( 250 m depth, but such surficial disturbance could greatly alter the rock weathering processes, groundwater levels and strain patterns in the bedrock, while also affecting the surface infrastructure and access tunnels. Shallow landslides are more frequent and widespread than deep-seated landslides but will generally not greatly affect a geological repository unless such an event occurs during the initial construction phases. Landslide hazard can be assessed by mapping the active landslides and the landforms of previous landslides and studying the correlations between landslides and other landscape characteristics such as rock type, bedding direction, slope angle and aspect and rainfall (e.g. Yamagishi et al., 2002; Van Den Eeckhaut et al., 2009). Other important geological features that will typically be identified by geological mapping are fault lines and volcanoes. Fault lines are usually recognised by the juxtaposition of two rock types that are not in chronological or stratigraphic order. For example, when Quaternary sediments (past 1.8 Ma) are overlain by Cretaceous (145–65 Ma) units; they cannot have been deposited in such an order and therefore must be separated by a fault line. Further details on how active fault lines are mapped and analysed will be discussed in Section 7.5.3. Active volcanoes are very obvious features in the landscape and simple to map at a coarse scale, but most volcanoes represent many eruptive events and at a detailed scale the ‘volcano’ may be represented as several edifices, vents and deposits. The classification of volcanoes and determination of frequency and magnitude to underpin probabilistic hazard can be a complex exercise (see, for example, Mahoney et al., 2009; Wetmore et al., 2009). Dormant or inactive volcanoes can be less obvious in the landscape but are usually identified by morphology (e.g. a cone shape or caldera lake) and geology (i.e. basaltic, andesitic or other such volcanic rocks). There are various dating methods


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used, such as radiocarbon, luminescence, 40K–40Ar and 39Ar –40Ar, to attain the age of volcanic features so that the history and hazard of the volcano can be assessed.

7.5.2 Measuring current crustal deformation using global positioning system (GPS) When subjected to tectonic stresses, the earth’s crust will deform. Global positioning system (GPS) techniques measure mm-scale movement of the earth’s surface related to active tectonic deformation. GPS measurements are taken at survey points permanently attached to the ground either by intermittent (survey-style) or continuous (daily, automated) collection of phase and pseudorange data from the constellation of GPS satellites that orbit the earth. GPS measurements enable the quantification of tectonic strain at different timescales (e.g. interseismic tectonic deformation over periods of decades, to coseismic movement over seconds during an earthquake) and at different spatial scales (e.g. regional to continental scale deformation). GPS techniques are useful in assessing tectonic stability of potential nuclear waste repositories as they enable the determination of regional tectonic strain rates, the location of possible new blind faults with no obvious surface expression (e.g. ‘hidden faults’) and the detection of aseismic slip. Elevated strains around a geological repository could disrupt the repository in numerous ways, including through the reactivation or formation of new faults, creep deformation of the engineered barriers and flexural folding of the host rock. An example of GPS techniques being used to measure tectonic strain is at Yucca Mountain, where a dense GPS network was installed to monitor the potential geological repository site. Prior to GPS installation it was thought that the Yucca Mountain area was under relatively little strain because nearby Quaternary faults have extremely low slip rates ( 0.7 mm/year of deformation currently accumulating on unidentified structures in the Yucca Mountain region (Wernicke et al., 1998; Hill and Blewitt, 2006). At a larger scale GPS has been used to map the tectonic strain across most plate boundaries around the world; it can also help identify blocks within tectonically active areas that are undergoing smaller amounts of deformation, relative to the surrounding region. Faults with no previously identified surface expression (hidden or blind faults) can be detected in some cases with the aid of GPS measurements (e.g. Donnellan et al., 1993; Stevens et al., 2002), and zones of very high strain where faults may develop in the future can also be identified. For example,



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in southern Kyushu, Japan, a zone of high left-lateral shear not identified by active faulting studies was defined from GPS studies (Wallace et al., 2009). The shear zone is interpreted to be associated with the subduction of a ridge, but because the ridge subduction point is continually migrating along the margin, the zone of high shear also migrates and does not localize in one place long enough to allow a fault zone to develop that is identifiable at the ground surface (Wallace et al., 2009). Based on the migration rate of the subducting ridge, the future location of a high left lateral shear strain can be predicted. GPS techniques can also be used to detect aseismic slip, which is when a fault slips without an accompanying earthquake. Aseismic fault creep has been detected on a few crustal faults (e.g. Azzaro et al., 2001; Lyons et al., 2002), but this typically occurs along faults with high slip rates that are identifiable at the surface. GPS techniques are most useful when used to identify aseismic slip at subduction zones. These are called slow slip events in which decimetres of movement occur on the interface between two plates over timescales of days to months (e.g. Dragert et al., 2001; Obara et al., 2004; Douglas et al., 2005). For example, the Manawatu slow slip event on the Hikurangi subduction interface, New Zealand, produced 35 cm of slip over 18 months. If this movement had been released suddenly it would likely have generated a MW 7.0 earthquake (Wallace and Beavan, 2006). Slow slip events along subduction interfaces are unlikely to be a direct hazard to nuclear repositories as there is no ground shaking associated with the events and the events occur at depths over several kilometres. However, slow slip events may alter the tectonic strain pattern in the upper plate of subduction margins and the identification of such events allows a better understanding of the seismic behaviour of subduction zones.

7.5.3 Active fault mapping and paleoseismology The presence of active faults is one of the soundest pieces evidence that an area is seismically active and will continue to be so in the future. The common definition of an active fault is one that has ruptured in the Holocene period (past 11 700 years). However, when considering the location of nuclear waste repositories it is necessary to use a longer timeframe and consider all faults that have been active within the Quaternary period (past 1.8 Ma). Active faults can pose a direct hazard to a repository by fault rupture through the repository (hence disruption of the engineered barrier and alteration of groundwater flow paths). Active faults near to the repository can also pose a hazard of severe ground shaking, landsliding, changing groundwater levels, uplift and possible oxidation, as well as altering the local tectonic strain field. A common approach to locating active faults is first to use aerial photos


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to identify geomorphic scarps and lineaments, followed by mapping and trenching the fault to constrain the frequency, timing and size of past events. This field of research is called paleoseismology. Active faults can often be recognised in the landscape by a fault scarp; on a flat surface this will be recognisable as a linear step cross-cutting all features of the surface or, at the foot of mountain ranges, fault scarps can be identified by a distinct change in the surface slope (Fig. 7.3). The bedrock around active faults can be sheared and weakened by fault movements, so often fault lines produce alignments of linear valleys as streams take advantage of the weakened bedrock. Active faults can also be recognised by offset landscape features; e.g. streams, ridgelines and terraces can be displaced by faults (Fig. 7.3).

7.3 Photograph of the Alpine Fault, a major strike-slip fault in South Island, New Zealand. Prominent geomorphic markers that help to identify the active fault are shown (source: K. Wilson).



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Once a fault line has been identified in the landscape it is important to understand several parameters. When did it last rupture? How frequently does it rupture? What size are the earthquakes? Paleoseismic investigations typically involve mapping the fault, observing local geology, geomorphology and rock outcrops, and trenching the fault. When mapping fault lines, offset features of the landscape will be recorded. Some of the offset features such as river terraces will have a known age or be able to be dated; therefore a fault slip rate can be calculated. For example, if a fault offsets a last glacial age river terrace (c. 20 000 years old) by 10 m the fault will have a slip rate of 0.5 m/1000 years (0.5 mm/year). The timing of the last rupture can be determined by dating the youngest offset feature. The frequency of fault rupture can be determined from dating successively offset landscape features (e.g. a sequence of abandoned channels) or by trenching. Fault trenching involves digging a large pit across the fault scarp, typically > 20 m long and > 2 m deep. Within fault trenches successive sedimentary units offset by the fault can be dated to determine the timing of past earthquakes (e.g. Lienkaemper et al., 2002; Kelson et al., 2006; Villamor et al., 2007). For some well-studied faults the size of the past earthquakes can be estimated using scaling relationships developed from historical earthquakes. These commonly relate earthquake magnitude and measurable geomorphic features such as the surface rupture length, the maximum surface displacement and the average surface displacement (Wells and Coppersmith, 1994; Hanks and Bakun, 2002). For example, Field et al. (2008) assembled active fault data for California and converted fault data to equivalent earthquake magnitude and frequency so that the historical record of earthquakes could be supplemented by earthquakes identified in the geological record. The paleoseismology methods discussed above can be applied to active faults with an identifiable scarp. However, many active faults, particularly those with very long recurrence intervals, do not display an obvious surface expression. Alternative methods of identifying active fault locations are required. For example, in Australia, active faults with very low slip rates (~0.02 mm/year) bounding the Flinders Range, south-central Australia, have been identified (Quigley et al., 2006; Hillis et al., 2008). The scarps of the faults were covered by Holocene sediments but exposures of Neoproterozoic schist (500–1000 Ma) thrust over Pleistocene (< 1.8 Ma) river gravels within stream exposures allowed identification of fault activity within the Quaternary period. Other techniques to recognise zones of active faulting include the identification of drainage pattern asymmetries (where active faults have altered drainage patterns, e.g. Cox, 1994, Meghraoui et al., 2001), identification of liquefaction features (where ground shaking causes liquefaction of soft sediments such as alluvial sands, e.g. Guccione, 2005; Singh and Jain, 2007), identification of seismites and seismically


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triggered clastic dikes (sedimentary layers shaken by earthquakes, typically recognised by the reorganisation of magnetic signals in the sediment, e.g. Marco and Agnon, 2005; Levi et al., 2006) and identification of earthquaketriggered speleothem breakage events in caves (e.g. Lemeille et al., 1999; Pons-Branchu et al., 2004). Active fault mapping identifies areas in which nuclear repositories should not be sited, while the results of paleoseismology studies provide important information for seismic hazard analysis. Seismic hazard analyses use earthquake catalogues compiled from instrumental and historic data. However, in areas with short historic records or with seismogenic sources with long recurrence intervals, paleoseismology is needed to characterise low-frequency but large-magnitude events.

7.5.4 Historical seismological record The historical seismological record is derived from two sources: the instrumentally recorded earthquake catalogue and written historical records of earthquakes in the pre-instrumental period. The timescales over which the two sources record earthquakes varies. For example, Italy and Japan have written historical records of earthquakes dating back > 1000 years whereas written historical records for New Zealand date back to only 1840 AD due to its short history of European occupation. Historic records of earthquakes come from a variety of sources, usually written records but also oral histories. For example, the 1700 MW 9 Cascadia subduction earthquake is recorded in Native American oral histories and there is also a written account of the resulting tsunami in Japanese literature (Satake et al., 1996). The instrumentally recorded earthquake catalogue varies by region and magnitude. For example, there is considered to be a complete record of global shallow earthquakes ≥ M 7 since the early 1900s and a complete record of global shallow ≥ M 6 events only since the early 1960s (Pe´rez and Scholz, 1984). The instrumental record for smaller events depends upon the seismic network of each country. For example, New Zealand has had a seismic network since the 1930s, whereas the Ecuador national seismic network was only installed in 1990. Mapping the distribution of historical seismicity provides a clear indication of tectonically active and inactive regions (Fig. 7.4). When considering the seismological issues for location of a nuclear waste repository a probabilistic seismic hazard analysis (PSHA) will usually be undertaken. The aim of this is to understand the hazards of earthquake shaking and abrupt fault displacement at any site. The first step in a PSHA is to locate all known earthquake sources in the area (usually within 100–200 km of the site). One input for the earthquake sources is the mapped active faults, which indicate the sources for large earthquakes (described in the



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7.4 An extract from the New Zealand earthquake catalogue showing the size and depth of earthquakes in New Zealand of ≥ M 4.0 from 1964 to 1997. This shows large areas of New Zealand that are seismically quiescent. The earthquake catalogue data are used in probabilistic seismic hazard analysis (see Stirling et al., 2001).

previous section). The other input is the historic seismicity record, which records the moderate-to-large background earthquakes. Background earthquakes are those events less than or equal in size to the ground-rupturing fault events, and can occur both on and away from the faults. They are wellillustrated by maps of seismicity (Fig. 7.4). The background earthquakes that do not occur on mapped faults may represent tectonic strain accumulation in between major faults or events on unmapped or blind faults (Stirling et al., 2009). Because earthquakes of <M 6.5 generally do not rupture the ground surface they will not be recorded by paleoseismic methods; therefore the frequency of such events needs to be derived from the historic siesmicity record.


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An important use of historical seismicity catalogues is in the understanding of earthquake magnitude–frequency relationships. Large earthquakes occur less frequently, while small earthquakes occur with great frequency (specifically, the Gutenberg–Richter relationship shows that the logarithm of the frequency of the earthquake occurrence is a decreasing function of magnitude; Gutenburg and Richter, 1944). These relationships mean that the historic seismicity record can use the frequency of the smallto-moderate sized earthquakes to predict the frequency of large earthquakes. The magnitude–frequency relationships are key components of probabilistic seismic hazard analyses.

7.5.5 Indicators or tectonic uplift or subsidence Uplift and subsidence of the earth’s surface occurs in response to a variety of processes, e.g. glacial loading, volcanic intrusion or tectonic forcing. Vertical deformation of the earth’s surface should be considered in the process of siting nuclear waste repositories for two main reasons. Firstly, the initial burial depth of a repository could increase or decrease over time if the earth’s surface is raised or lowered. For example, if a repository is buried at a depth of 300 m at a site undergoing uplift at 3 mm/year, it will only take 100 000 years for the repository to reach the surface (if surface erosion keeps pace with the rock uplift). Secondly, vertical deformation at the earth’s surface may reflect an active ‘hidden’ structure at depth. Uplift and subsidence can be measured over a range of time and spatial scales (e.g. Litchfield et al., 2009). For short-term (minutes to decades) measurements of vertical deformation, common methods include: . . . . .

Continuous GPS (continuous measurement of a point to cm-scale accuracy for vertical motion) Geodetic surveying (levelling to mm-scale precision between widely spaced locations and comparing survey results over years to decades) Tide gauges (these record relative sea-level changes and can record land movement over years to decades) InSAR (using radar satellite images from different times to detect changes in the surface elevation to centimetres of precision) Biologic or geomorphic markers such as shorelines and coral atolls can record land-level changes over years to decades in rapidly deforming areas

Over the medium term (1000–100 000 years) geomorphic surfaces such as river terraces and marine terraces can record uplift. For example, river terraces that formed by aggradation during the last glacial have been uplifted above modern river levels along the parts of the Hikurangi margin, New Zealand. The terraces record differential uplift related to tectonic



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processes along the subduction margin (Litchfield and Berryman, 2006). Marine terraces record former shorelines; if the sea level history is known and the terrace can be dated, then the comparison between the terrace elevation and modern sea level provides a measure of tectonic uplift (e.g. Berryman, 1993; Hsieh et al., 2004; Ota and Yamaguchi, 2004). Marine terraces are typically used to record uplift in the Holocene period or since the late Pleistocene (last 125 000 years). Buried shorelines or marginal marine sediments can also record tectonic subsidence (e.g. Cochran et al., 2006). Techniques to measure tectonic upift and subsidence over the long term (100 000’s to millions of years) usually apply to large areas and reflect plate tectonic processes such as continental collision or rifting. Marine sedimentary rocks high in the landscape can record long-term uplift, as can mudstone-porosity measurements and low-temperature thermochronology. Peneplains, large areas of formerly flat relief that are uplifted and incised, also record long-term tectonic uplift. Similarly, marine planation surfaces, formed by erosion when the sea level was at a low stand, can record tectonic subsidence when identified in marine seismic lines. Tectonic processes causing uplift and subsidence can vary from local faulting on a small scale to plate boundary collisional and extensional processes at the large scale. Marine and fluvial terraces can be particularly important features in nuclear waste repository investigations as they can represent the surface expression of uplift on blind faults. Blind faults are those that do not rupture the ground surface; hence no surface scarps are produced. Where fluvial or marine terraces, which are assumed to have initially formed flat (or slightly downstream-sloping for fluvial terraces), are warped, it may indicate tectonic deformation at depth. For example, blind thrust faults have been mapped using the deformation patterns recorded by mid-late Pleistocene fluvial and marine terraces on the eastern edge of the Apennine mountains in central Italy; these blind faults are thought to be the source of several historical M 5.2 earthquakes (Vannoli et al., 2004). Marine terraces can often also record uplift on unmapped offshore faults; e.g. at the Pakarae River mouth area, New Zealand, a suite of seven Holocene terraces records uplift that is associated with an offshore thrust fault (Ota et al., 1991; Wilson et al., 2006). On a larger scale, the distribution of uplift and subsidence at collisional plate boundaries can indicate areas of greater or lesser tectonic strain. For example, along the northern coast of California, in the vicinity of the Mendecino Triple Junction, there are many marine terraces. The uplift rates are generally very high (up to 4 mm/year) and variable due to complex interactions between the San Andreas fault, the southern end of the Cascadia subduction zone and the Mendecino fracture zone (Merritts, 1996). Large-scale subsidence can occur due to crustal thinning or by


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tectonic loading. Crustal thinning occurs at divergent plate boundaries where continental plates are rifted apart, or within transtensional basins along transform plate boundaries. Tectonic loading occurs adjacent to large mountain ranges or volcanoes where the weight of the massif depresses the crust. Subsidence can also occur at subduction margins characterised by tectonic erosion; this is when the subducting plate scrapes material off the underside of the upper plate thus thinning the upper plate. However, this process generally occurs within 100 km of the subduction trench.

7.5.6 Geophysical techniques for detecting crustal structure and volcanic intrusions There are several geophysical methods that can be applied when seeking to understand the structure and stability of the earth’s crust. Techniques such as gravity surveys and seismic tomography can be used for investigating crustal structure, while aeromagnetics, magnetotellurics and heat flow can be used for detecting volcanic intrusions, such as magma chambers, as well as crustal structure. These techniques and their application to assessing geological stability are outlined below. Aeromagnetic measurements characterise the earth’s magnetic characteristics as measured from the air. The data are collected using a magnetometer aboard an aircraft; the magnetometer records variations in the intensity of the ambient magnetic field. After corrections have been made for atmospheric and regional affects a map is compiled that generally shows the distribution and relative abundance of magnetic minerals in the upper parts of the earth’s crust. As different rocks will contain different amounts of magnetic minerals this technique can be used relatively quickly, but coarsely, to map the major structures of the upper crust. Features such as major fault lines, folds and basins can be identified (but no estimate can be made of the age of the structures). Of relevance to geological repository investigations is that aeromagnetics is a useful tool for identifying magnetic anomalies that may be associated with subsurface volcanic rocks or magma bodies at shallow depth. In currently active volcanic regions the surface geology is often dominated by recent volcanic sediments such as pyroclastic flows; similarly, ancient volcanic features can be covered by non-volcanic Quaternary sediments. For example, a high-resolution aeromagnetic survey of the Amargosa Valley identified 20 anomalies that may represent basaltic volcanoes buried beneath 100–400 m of alluvium (O’Leary et al., 2002). This site is located only 25 km from Yucca Mountain and it has been estimated that the 20 newly identified volcanic centres would raise the probability of Yucca Mountain site disruption by 40% (Perry et al., 2004; Smith and Keenan, 2005).



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Just as different rock units vary in magnetism, they also vary in density. Gravity surveys are a method of detecting lateral changes in gravity related to subsurface rock density. The surveys can be carried out at a variety of scales, from airborne surveys of large areas to higher resolution ground surveys over several kilometres. Raw gravity measurements are corrected for elevation, latitude and topography and the regional gravity field is subtracted, which yields a map of gravity anomalies. Positive gravity anomalies are often associated with the presence of anomalous high-density bodies in the shallow crust compared with a ‘standard’ earth model, whereas gravity lows are associated with low-density bodies, which may be located anywhere in the crust or the upper mantle. If there is some knowledge of the density of rocks in the area, crustal structure models can be produced. The models try to fit the gravity data as closely as possible but there is often more than one solution; therefore gravity surveys are best used in conjunction with other techniques such as seismic tomography (see below) and aeromagnetics. Gravity techniques have often been used successfully in volcano-tectonic regions where basement faults bounding the volcanoes are typically buried beneath young volcanic deposits. For example, basement faults beneath the Tongariro Volcanic Centre, New Zealand, are covered by up to 800 m of volcanic deposits but vertical displacements of 250–300 m on basement faults have been modelled using gravity data; this allows minimum rates of volcanic rift subsidence to be resolved (Cassidy et al., 2009). Gravity models of the Summa–Vesuvius volcano, Italy, allow identification of lineations related to normal faults in the carbonate basement up to 2 km beneath the surface (Cella et al., 2007). Beneath the volcanic island of Pico, Azores, gravity anomalies corresponding to lowdensity bodies are thought to represent shallow, small intrusions of magma at the intersections of tectonic lineaments, highlighting how gravity can help understand both crustal structure and volcanic hazards (Nunes et al., 2006). Seismic tomography is another geophysical method that is used for understanding deep crustal structure. Tomography uses the velocity of seismic waves to build three-dimensional models of the earth’s crust. Areas through which waves move quickly tend to be cooler or consist of denser rock. Conversely, areas where waves move slowly indicate warmer or less dense rock. Seismic tomography can be carried out at a variety of scales – from using global networks of seismometers to detect regions of upwelling and downwelling deep in the earth’s mantle, to using regional seismic networks to study tomography at a plate boundary or around an individual volcano. It can be particularly useful for detecting areas of partial melt and fluids in the crust. For example, at the scale of plate boundary, seismic tomography of the central North Island, New Zealand, shows the depth of the subducted Pacific Plate at up to 65 km depth and significant volumes of partial melt beneath the backarc volcanic zone (Reyners et al., 2006). At a


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smaller scale, seismic tomographic studies of Mount St Helens, USA, have shown velocity anomalies associated with a linear zone of high seismicity, probably a fault, and with a magma chamber at 2–3.5 km depth (Waite and Moran, 2009). Another geophysical method for understanding crustal structure is magnetotelluric measurements. Magnetotelluric data maps the spatial variation of the earth’s electrical resistivity by measuring naturally occurring electric and magnetic fields at the earth’s surface. In general, rocks containing fluids, such as water, and rocks at high temperature will have a low resistivity. Conversely, dry and cold rocks will have high resistivities. Magnetotelluric measurements can produce images of the earth’s electrical resistivity structure from a few hundred metres to several hundred kilometres depth depending on the frequency of the electromagnetic signals. This method is particularly sensitive to fluids in the crust because fluids dramatically lower the resistivity of the host rocks. For example, magnetotelluric surveys across the San Andreas Fault show that the fault zone is characterised by a vertical zone of low electrical resistivity; this low resistivity is due to the presence of fluids on the highly fractured fault zone to a depth of approximately 4 km (Unsworth et al., 1997). Magnetotellurics is frequently used in volcanic and geothermal regions to image magma chambers, volcanic vents and regions of hydrothermal alteration. Beneath the Aso volcano, Japan, magnetotelluric surveys could differentiate which craters were underlain by active hydrothermal systems and which were underlain by consolidated magma (Kanda et al., 2008). Heat flow measurements are a further geophysical method used primarily in monitoring volcanoes, as well as for detecting melt intrusions and hydrothermal heating at some distances from volcanic centres. This is important for nuclear waste repository tectonic stability as anomalously high rates of crustal heat flow may be an early precursor to the onset of volcanism. The heat flow within the continental crust depends on the natural radioactivity in the crust, the tectonic setting and heat flux from the mantle below (Stein, 1995). An example of the effects of high heat flow due to magma intrusion is at Yellowstone, USA, where the average heat flow is 30 to 40 times greater than that of an average continental crust. Crustal heat flow is usually measured using data from deep boreholes and wells. In volcanically active zones infrared remote sensing is used to monitor surface heat flow. Groundwater temperature and chemistry is also monitored as this is affected by shallow magma intrusions. Precursor increases in heat flow have been detected using satellite data on several active volcanoes. For example, a temperature anomaly of several hundred degrees above average was detected in the months leading up to the Lascar volcano eruption in Chile in 1986 (Rothery et al., 1988).



7.6

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Modelling long-term stability for safe disposal of radioactive waste

Extrapolating current levels of knowledge and confidence estimates far into the future has been recognised by nuclear safety commissions throughout the world as a challenging problem. The recent regulatory guide from the Canada Nuclear Safety Commission (CNSC, 2006) typifies how the longterm safety issue (and by inference the contributing hazard and risk assessment) is being addressed. CNSC note that ‘the demonstration of safety will rely less on quantitative predictions and more on qualitative arguments as the timescale increases’, and that ‘long term quantitative predictions should therefore not be considered as guaranteed impacts, but rather as safety indicators’. These statements represent a pragmatic view in the absence of mature procedures. Inevitably, quantitative assessment, including uncertainty analysis, will become the expectation in future, even if a qualitative approach is acceptable for the moment. At present the extent of hazard data and quantitative modelling of tectonic processes is limited, but with increasing computing capacity, new dating techniques and determination, a longer time series of information and modelling is achievable. The challenge before the scientific and risk community is to develop robust quantitative approaches to bound uncertainties and to entertain the appropriate alternative tectonic and volcanic evolution patterns in an area of interest over extended timeframes. Instrumental records exist for perhaps one or two centuries at best, historical documentation of events exists for a few millennia at most and the geological record loses its resolution quite quickly as one looks back in time. In many environments a reasonable level of preservation of events and processes in the geological archive may exist for 10–100 thousand years, but over longer periods event preservation will be more patchy, and the ability to date events accurately and calculate rates of process requires considerable resource. Additionally, it becomes increasingly important to evaluate whether large-scale boundary conditions, such as motions of the tectonic plates, can be assured to be constant. For example, the assumption that the volcanic front in a subduction margin setting is essentially fixed in space and time because it coincides with melting at the c. 100 km depth contour on the subduction interface may not be valid over long periods of interest such as 1 million years, because subduction processes can evolve on this timeframe. Thus, extrapolation of hazard estimates, based on well-constrained datasets for perhaps the past 100 thousand years at most, to 1 million years, brings great uncertainty. Traditional statistical approaches to uncertainty treatment at long return periods mean that the extreme tails of probability density functions (PDFs) of hazard parameters govern the hazard estimates (e.g. Stepp et al., 2001;


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Hanks et al., 2006). Until recently, few data or models have existed to constrain the upper reasonable bound on statistical relations, and thus there is a tendency for hazard estimates to be extremely conservative. Studies that specifically address upper bound constraints on hazard parameters are important in the context of long-term geological stability assessment, but thus far have been largely restricted to the earthquake ground motion field. Approaches include modelling (e.g. Andrews et al., 2007) and statistical studies (e.g. Rhoades et al., 2008; Strasser et al., 2009). Similar approaches to bounding other key parameters of the hazard model or to limit the range of evolutionary tectonic paths that are reasonable over certain time periods in a region offer hope for more certain long-term assessments of site stability.

7.7

Future trends

A range of innovative and alternate approaches to site suitability assessment over and above traditional deterministic and latterly probabilistic methodologies are being developed (Fig. 7.5). The deterministic and probabilistic approaches are well developed and indeed expected for nuclear power plants

7.5 Steps in the process of evaluating the long-term stability of the geological environment for the purposes of radioactive and spent nuclear fuel repositories.



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and increasing to geological repository siting and safety assessment (e.g. Cornell, 1968; Reiter, 1990; Stepp et al., 2001; IAEA 2002; Gu¨rpinar, 2005). To a large extent the innovations extend existing procedures or bring techniques from other disciplines to the hazard area. Many approaches are designed to capture a reasonable representation of uncertainty in hazard and risk calculations and to identify and separate epistemic (uncertainty because of lack of knowledge) from aleatory (variability that inherently exists in natural earth systems) components (SSHAC, 1997; Bommer and Abrahamson, 2006). Because quantitative data are often lacking for all aspects of tectonic hazard assessment at critical facilities, judgement is required to assess many aspects of the hazard and risk assessment. An experienced group of experts may be used to parameterise the quantitative models or assess the appropriate range of alternate geologic models that may characterise the site, or, in terms of very long-term assessment, how a region may evolve over periods of 100 000– 1 000 000 years. Considerable effort has been placed on developing a structured environment in which expert judgement can be used (e.g. SSHAC, 1997; Coppersmith et al., 2009). Coppersmith et al. (2009) list the following key requirements in achieving successful expert elicitation: . . . . . . .

Experts should be trained in probability theory, uncertainty quantification and ways to avoid common cognitive biases. Comprehensive and user-friendly databases should be provided to the expert. Experts should be required to evaluate all potentially credible hypotheses. Workshops and other interactions among the experts and proponents of alternative viewpoints should be encouraged. Assessment interviews should start with conceptual models and develop towards parameter assessments. Feedback should be provided to the experts to give them insight into the significance of their assessments to the hazard results. Complete documentation should include the technical basis for all assessments.

There remains considerable debate about the relative merits of probabilistic and deterministic hazard procedures (e.g. Bommer, 2002; Klu¨gel, 2005). In our view future developments will see a narrowing of these conflicts and in fact a recognition that deterministic scenarios should often be accepted into probabilistic methodologies as alternate viable models or parameters. These are readily structured in decision trees (often called logic trees), which are increasingly being formulated logically to structure probability estimates of components of hazard and risk assessment (e.g. SSHAC, 1997; Kessler and McGuire, 1999; Coppersmith et al., 2009). Decision trees often become very


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large for assessment of complex systems, and increasingly a Monte Carlo sampling approach, paying heed to the weighting of alternate parameters or branches at each level of the decision tree, is used to generate the hazard curves (e.g. Jaquet et al., 2000; Perry et al., 2001). Particularly within the carbon sequestration field there has been an emphasis on deterministic scenario modelling using ‘features, events and processes’ or FEPs (e.g. Maul et al., 2004). Such an approach has also been developed occasionally in the nuclear waste industry (e.g. Swift et al., 1999; NEA/OECD, 2000) and particularly in the Swedish waste repository programme (e.g. Last et al., 2004; Kozak et al., 2006). These individual FEPs may be considered viable alternate geologic models, processes or event chains, which can readily be absorbed into probabilistic hazard methodology. Fuzzy logic theory has also been applied to probabilistic hazard estimation as a means of capturing uncertainty in data inputs (e.g. Elshayeb, 2004; Goldsworthy et al., 2004) Another modelling technique that may be of substantial value in forecasting long-term evolution of a geological repository is the use of the finite element approach. This technique is now used extensively to model the evolution of geological structure, groundwater flow, performance of radioactive waste containment and nuclear facility engineering (e.g. Rautman et al., 1995; Chen et al., 2004; Vespa et al., 2005). As computational capacity increases it appears feasible that sufficiently large or detailed meshes may be able to be defined for scales ranging from repository site area geology to a regional context such that process-driven evolutionary models of long-term site stability could be developed. Such models would be a valuable approach to compare alongside statistical methods embedded in deterministic probabilistic calculations.

7.8

Summary

Assessment of the stability of the geologic environment for the purposes of nuclear waste repositories requires a long-term (100 000–1 000 000 years) understanding of tectonic and climatic processes. This is significantly longer than comparable stability assessments for other critical facilities, such as nuclear power plants and large dams, which typically have a lifetime of 1000 a are shown. Data sources: Nykyri et al. (2008, Tables 6-8 and 6-17), Nagra (2002b, Tables A2.1.2, A2.1.4, A2.1.6 and A2.8), Nagra (2002c, Table A3.4-1). Activities are reported per ton of initial heavy metal (tIHM) in the Swiss case, where IHM = original mass of U or U + Pu (Pu originates from reprocessed mixed-oxide (MOX) fuel). As Finland does not reprocess spent fuel and pursues direct disposal only, there is no MOX fuel (i.e. IHM = U) and the waste categories related to reprocessing (vitrified high-level waste, intermediatelevel waste) do not exist. Activities relate to the time at which disposal is foreseen, i.e. after 30 (Finland) or 40 (Switzerland) years of decay on the surface

Radionuclide

Half-life (a)

Activity of Finnish spent fuel (OL1 and OL2 reactors) (GBq/tU)

10

Be C

1.60E+06 5.70E+03

2.78E+01

Cl Ca Ni 79 Se 93 Zr 93 Mo 94 Nb 99 Tc 107 Pd 126 Sn 129 I 135 Cs 137 Cs 166m Ho 226 Ra 229 Th 230 Th 231 Pa 232 Th 233 U 234 U 235 U 236 U 237 Np 238 U 239 Pu 240 Pu 242 Pu 243 Am 245 Cm 246 Cm

3.00E+05 1.00E+05 7.50E+04 1.10E+06 1.50E+06 4.00E+03 2.00E+04 2.10E+05 6.50E+06 2.30E+05 1.60E+07 2.30E+06 1.37E+03 1.20E+03 1.60E+03 7.90E+03 7.50E+04 3.30E+04 1.40E+10 1.60E+05 2.50E+05 7.00E+08 2.30E+07 2.10E+06 4.50E+09 2.40E+04 6.60E+03 3.80E+05 7.40E+03 8.50E+03 4.70E+03

14

36 41 59

1.04E+00 1.32E+02 3.30E+00 9.08E+01 1.13E+01 3.00E+01 6.12E+02 4.29E+00 2.18E+01 1.14E+00 2.15E+01 2.36E+06

2.41E-03 5.59E+01 7.45E-01 1.30E+01 1.30E+01 1.16E+01 1.05E+04 1.98E+04 7.61E+01 7.62E+02 6.16E+00 1.19E+00

Activity of Swiss intermediatelevel waste (ILW-1 type) (GBq/tIHM)

Sorption Kd in crystalline bedrock at Olkiluoto (dilute/ brackish water) (m3/kg) 0

Activity of Swiss spent fuel (GBq/tIHM)

Activity of Swiss vitrified high-level waste (BNFL glass) (GBq/tIHM)

1.2E-01 8.3E+01

4.34E-02

4.15E+00 1.43E-02

0

4.09E-01 9.77E-01 7.94E+01 3.05E-02

2.72E+01 2.95E-04 1.34E+00 3.17E-01 7.25E+00 3.63E-01

0.1 0.0005 0.2 0.0005 0.02 0.05 0.001 0.001 0 0.05 0.05

1.5E+00 1.9E-01 8.8E+01 1.1E+00 1.1E+02 2.8E-01 8.8E+00 7.4E+02 6.7E+00 2.0E+01 1.8E+00 2.5E+01 2.3E+06 1.4E+00 1.9E-04 9.4E-06 2.1E-02 5.1E-04 2.7E-08 4.5E-03 6.5E+01 6.1E-01 1.4E+01 2.1E+01 1.3E+01 1.5E+04 2.7E+04 1.3E+02 1.5E+03 3.7E+01 8.0E+00

5.74E+02 4.58E+00 3.12E+01 1.28E-03 2.38E+01 1.41E+06 9.16E-02 1.16E-04 3.85E-06 6.72E-03 1.22E-03

3.40E-08 7.93E-10 4.08E-06 1.88E-07

2.14E-03 6.05E-02 1.83E-04 2.26E-03 1.22E+01 3.67E-03 4.09E+01 1.41E+02 2.14E-01 5.38E+02 3.97E+00 7.33E-01

4.08E-07 1.25E-02 2.27E-04 1.72E-03 1.68E-03 2.49E-03 3.85E+00 5.44E+00 1.72E-02 8.16E-02 3.40E-04 7.03E-05

3.63E-03 9.75E-03 2.72E-02 9.75E+02


0.2 0.2 0.2 0.05 0.2 0.1 0.1 0.1 0.1 0.2 0.1 0.5 0.5 0.5 0.04 0.04 0.04

Sorption Kd in Opalinus Clay (reference case) (m3/kg) 0.9 0.001 (Cinorg), 0 (Corg) 0 0.001 0.9 0 10 0.01 4 50 5 100 0.00003 0.5 0.5 50 0.0007 50 50 5 50 20 20 20 20 50 20 20 20 20 10 10 10

Far-field process analysis and radionuclide transport modelling

227

barriers is expected to evolve very slowly, but the conversion to dose (in mSv/a) becomes very uncertain. On a national level, the regulatory frameworks define the time scale and degree of rigorousness over which repository safety needs to be demonstrated. For example, in Finland, the constraint of a maximum annual dose of 0.1 mSv applies in the first thousands of years after repository closure (STUK, 2001). Over this period, the biosphere can be assumed not to differ substantially from the present situation. At later stages, predictions of the evolution of the biosphere become more uncertain, and therefore the regulation is based on specifying maximum radionuclide fluxes, while no conversion to dose is required from the implementer. In addition, the regulation acknowledges the fact that the reliability of the model calculations becomes limited at more distant times. The Swiss regulation (ENSI, 2009) defines a maximum dose of 0.1 mSv/a for the most strongly exposed individuals on the basis of the expected evolution of the repository system. Unlikely disruptive scenarios with a higher radiological risk must be shown to have a probability of less than 106 per year. In contrast to the Finnish regulation, the Swiss regulation requires the calculation of an individual dose up to a time scale of 1 Ma. At the same time, it acknowledges the fact that the evolution of the biosphere cannot be predicted accurately and suggests that the calculations be based on the current state of the biosphere and human habits. For times beyond 1 Ma, the radiological consequences of surface effects such as the possible erosive exhumation of the repository must be assessed and compared to those of natural radiation.

8.2

Transport and retardation in argillaceous sedimentary formations

The majority of the formations studied in the framework of the disposal of radioactive waste (whether at potential sites or in generic rock laboratories) have a limited thickness, typically 100–200 m (Table 8.2). However, the clayrich unit is frequently embedded within a thicker but lithologically more heterogeneous low-permeability sequence, also shown in Table 8.2. Given the limited thickness, it needs to be shown for a safety case that retardation of radionuclide transport is sufficient over such short transport paths. In the Swiss safety case (Entsorgungsnachweis; Nagra, 2002a, 2002b), the geosphere barrier is represented by a vertical migration path of 40 m through Opalinus Clay. This value is less than the half formation thickness and excludes the near-repository geosphere. Similarly, the additional retardation along the rest of the migration path (upper part of the low-permeability sequence, then aquifers) is neglected in the reference case.



{

*

113.2

160.0

102.6

257.4

312.1

219.0

102.6

257.4

7.9E-11

4

5

2.1E-11–2.8E-11

4.2E-11–5.6E-11

2

1.0E-15–1.0E-14

2E-12–4.3E-11*

4E-14 (4E-15–4E-13)

6E-15–3E-14

5E-14–5E-13

10 cm, i.e. well beyond the zone that was potentially affected by mechanical damage due to drilling or stress redistribution, indicating that a connected porosity exists (see Fig. 8.2). Porosity was then measured by a suite of laboratory methods (Mo¨ri, 2009). The mass proportion of resin in subsamples (representing the in situ porosity) was quantified by measuring the carbon content and a porosity of 0.2–0.5% was obtained to depths of up to 20 cm from the borehole. Note that matrix porosity did not vary systematically as a function of distance from natural fractures, and even several metres away a connected porosity was identified.

8.3.2 Matrix fluid studies at Laxemar and Forsmark, Sweden Methods to study the chemical characteristics of pore water in crystalline rocks from Sweden (A¨spo¨, Laxemar, Forsmark) have been developed by Waber and Smellie (2008). In this framework, Waber et al. (2007) studied the Cl contents and the δ18O and δ2H values of pore water residing in the



235

matrix of crystalline rock at Laxemar (Sweden) in the vicinity of a waterconducting fracture at a depth of 112 m. The Cl content of the immediately adjacent matrix water corresponds to that measured in the fracture water (around 80–90 mg/L). Over a distance of 6 m away from the fracture, the Cl content increases to reach a constant value around 280 mg/L beyond that distance. Waber et al. (2009) interpreted the profile as due to diffusion in response to the evolving salinity of the water circulating in the fracture. Transport modelling considering the Pleistocene and Holocene evolution of ground waters reproduced reasonably well the observed profile considering plausible boundary conditions and diffusion as the only transport process. This evidence is taken as a supporting argument for the existence of a connected matrix porosity to a distance of at least 6 m from the fracture. Analogous results were also obtained from other Swedish study sites (see Waber and Smellie, 2008, for an overview) and from the Grimsel test site (Eichinger, 2009).

8.3.3 Constraining the formation factor by electrical conductivity logging in boreholes in Sweden Electrical conductivity (or resistivity) logging is a standard tool of geophysical borehole characterisation. Depending on the tool, the penetration depth into the formation is in the order of decimetres to metres, i.e. well beyond the zone potentially mechanically disturbed by the drilling process. Under certain assumptions (see, for example, Lo¨fgren, 2004), the measured conductivity can be used to constrain the formation factor of the rock matrix according to Ff ¼

De d kr ¼ nt 2 ¼ t Dw kw

with Ff = formation factor (–) De = effective diffusion coefficient of a species (m2/s) Dw = free-water diffusion coefficient of a species (m2/s) nt = transport porosity (–) d = constrictivity (–) t2 = tortuosity (–) kw = electrical conductivity of free water (S/m) kr = electrical conductivity of the rock matrix saturated with pore water (S/m) The assumptions underlying the applicability of electrical-conductivity logs for estimating the formation factor include: (1) the electrical conductivity of


236


the pore water within the tested volume can be constrained, either based on measurements on ground water circulating in adjacent fractures or on the basis of matrix-leaching data from the drill core (Waber and Smellie, 2008), (2) test intervals can be identified in which no fractures are present, such that the measured conductivity characterises the matrix and (3) the measured conductivity reflects ion movement in water, whereas conduction through minerals or along their surfaces is negligible or can be properly accounted for. Mean values of the formation factor for the rock matrix of various Swedish crystalline-basement rocks are in the range of 105 to 104 for insitu measurements (e.g. Lo¨fgren and Neretnieks, 2003; Liu et al., 2006; Lo¨fgren, 2007). Relating these values to the free-water diffusion coefficient of anions such as Cl or I (ca. 26109 m2/s at room temperature; Li and Gregory, 1974; CRC, 2004) yields values for De of 261014 to 261013 m2/ s. This is about a factor of 20–40 less than the range obtained for indurated clay-rich sedimentary formations (Mazurek et al., 2009), reflecting the lower porosity and different pore structure. Laboratory measurements of the formation factor yield values that are up to an order of magnitude higher when compared to in situ determinations, and Lo¨fgren (2004) attributes this difference to stress release and sample damage in the case of the laboratory samples. Overall, the in situ measurements of the formation factor indicate that a connected porosity exists in the studied formations, even though the derivation of specific parameters is subject to uncertainties, in particular due to the generally unknown chemical composition of the pore water and the distinction of ion movement in the rock matrix from that in fractures.

8.3.4 In situ diffusion experiments in the Canadian underground research laboratory (Pinawa, Manitoba) A suite of test boreholes were drilled in the URL into granitic rock to a depth of 10 m to reach rock not affected by tunnel excavation (Vilks et al., 2003, 2004). Tracer cocktails were injected into packed-off intervals, and care was taken to prevent hydraulic gradients. After 15 to 28 months, some of the test volumes were overcored and studied in the laboratory. In a first test stage, four vertical boreholes were emplaced at the 420 m level. Iodide tracer penetration to a matrix depth of 2.5 to 8 cm was identified for diffusion times of 1.3 to 2.3 years (Vilks et al., 2003). The shape of the diffusion profile could not be modelled when assuming a homogeneous medium and radial geometry, and a composite model with a 1-cm thick skin considering a lower diffusion coefficient was needed to explain the data. The skin was thought to be due to stress redistribution along the borehole, given the exceptionally high natural stresses at the 420 m level (s1 = ca. 60 MPa).



237

In a second test stage (Vilks et al., 2004), subhorizontal instead of vertical boreholes were drilled at the 240, 300 and 420 m levels, mostly subparallel to s1 or s2, which minimised problems with core discing. The tracer distribution was used to calculate effective diffusion coefficients for iodide of 1.461013 to 1.161012 m2/s in the distal zone unaffected by stress redistribution. While these in situ values do not vary with depth, ancillary laboratory tests suggest increasing trends for laboratory-derived diffusion coefficients and porosity with depth, which is interpreted to be due to a progressively stronger effect of stress release with increasing depth. To summarise, the experiments demonstrated (1) the existence of diffusion (and therefore a connected porosity) in situ and (2) the fact that laboratory measurements may be affected by stress release and therefore overestimate in situ transport parameters in the rock matrix.

8.3.5 Treatment of matrix diffusion in geosphere-transport calculations The in situ measurement of matrix porosity and diffusion coefficients in crystalline rocks is a non-trivial undertaking. A number of pertinent studies have been conducted in recent times, all indicating that the rock matrix contains an interconnected, diffusion-accessible pore network throughout the rock body. Adjacent to fractures, porosity and diffusion coefficients may be enhanced if hydrothermal or low-temperature alteration affected the rocks. However, diffusion has been shown not to be limited to alteration rims or a specific zone around fractures. Nevertheless, the amount of site-specific experimental data pertinent to in situ conditions is limited, and therefore geosphere-transport models in general do not consider unlimited matrix diffusion. In the Swedish longterm safety evaluation for the Forsmark and Laxemar sites, the penetration depth of matrix diffusion is in the range 0.03–10 m in the probabilistic geosphere-transport model (SKB, 2006a, 2006c). The upper limit is in the range of the typical half spacing between water-conducting fractures and therefore essentially represents full matrix connectivity, whereas the lower limit corresponds to the length over which connectedness has been undisputedly demonstrated. In the radionuclide transport model used by Posiva for Olkiluoto, the potential repository site for spent fuel in Finland, Nykyri et al. (2008) applied a simplified model considering one single pathway to quantify radionuclide transport in the geosphere, and a constant matrix-penetration depth of 10 cm was assumed. To conclude, the migration of sorbing radionuclides from defective canisters will be retarded in the bentonite and in the host rock. On the other hand, non-sorbing radionuclides released into the bentonite will diffuse out


238


into the host rock, and this occurs rapidly in comparison to the respective half-lives (Table 8.1). In the geosphere, matrix diffusion is essentially the only retardation process for long-lived radionuclides that occur mainly as anions and so interact weakly or not at all with mineral surfaces, such as 129 36 I, Cl or 79Se.

8.4

Quantifying radionuclide transport: two case studies

The role of the host rock and of the geosphere as a whole in a multi-barrier disposal system is illustrated here using the Swiss spent fuel/vitrified highlevel waste programme targeted at Opalinus Clay in northeasten Switzerland and the Finnish spent fuel programme focused on the crystalline basement at Olkiluoto (southwestern Finland) as examples. Both programmes consider the disposal of spent fuel, but in the Swiss case, vitrified high-level waste and intermediate-level waste are also part of the programme. The choice of these case studies is motivated by the very different characteristics of the geosphere at both sites, and this also affects the design of the engineered barriers. The purpose is to present and compare the approaches and conclusions as presented in the literature. The disposal concepts for the two sites are illustrated in a highly simplified form in Fig. 8.3. The Swiss concept considers a repository hosted by Opalinus Clay, a flat-lying, indurated, about 100–120 m thick shale formation. The waste will be encapsulated in steel canisters and will be embedded in a bentonite buffer (note that copper canisters are considered as an alternative option; see Nagra, 2002b). Due to the scarcity of fractures and their high self-sealing potential, there is sufficient scientific argument to exclude fracture flow from the expected evolution scenario, and diffusion is considered as the main transport process (Nagra, 2002a; Gimmi et al., 2007). In the Finnish concept, waste contained in copper canisters and surrounded by bentonite will be emplaced into Precambrian crystalline bedrock where transport is dominated by advection in fractures. In the KBS-3V design, each vertical deposition hole will be located in a rock volume devoid of major fractures and faults, even though lowertransmissivity fractures may be intersected by deposition holes. (Note that the KBS-3H design with horizontal in-tunnel emplacement is also being considered.) Along the flow path from the repository to the surface, radionuclides may be transported in progressively higher-hierarchy fractures and faults, so the first part of the flow path is thought to have the largest transport resistance. The properties of the host rock in the two disposal concepts are contrasting, and these differences affect the site-specific safety concepts.



239

8.3 Simplified illustrations of the disposal concepts for spent fuel in Opalinus Clay (Switzerland) and in crystalline basement at Olkiluoto (Finland). Sketch is not to scale. Arrows indicate possible migration paths of radionuclides towards the surface. For Olkiluoto, the KBS-3V concept with vertical deposition holes is shown. An alternative concept with in-tunnel deposition (KBS-3H) is currently also under consideration. The reference depths for the repositories are 650 m below the surface in Switzerland and 420 m in Finland.

8.4.1 Features, events and processes In order to understand and quantify the performance of the geosphere safety functions (see above), the relevant phenomena need to be explored. A structured way to deal with phenomena and to minimise the risk of omitting important processes and events is to compile and evaluate lists of FEPs (features, events and processes). In this context, events can be seen as discontinuous phenomena that occur over short periods of time (such as earthquakes) and features relate to qualitative attributes of the system and to parameter values needed to quantify the processes and events. Lists of geosphere-related FEPs, together with a documentation of their understanding and relevance for repository safety, are available on an international basis (e.g. general documentation in NEA, 2000, and FEP evaluation for argillaceous rocks in Mazurek et al., 2003). However, due to the specific features of national disposal concepts, of the site-specific properties of the repository system and of the safety functions of the geosphere, an understanding of features, events and processes must be obtained on a site-specific level for each disposal programme (e.g. Nagra, 2002d, for Switzerland; SKB, 2006b, for Sweden; Miller and Marcos, 2007, for Finland). It is convenient to structure geosphere-related FEPs into the following classes:


240 1.

2.

3.

Geological repository systems for safe disposal Undisturbed system: characterisation of the current state of the far-field geosphere. This is a snapshot of an evolving system but useful as a starting point. Repository-induced perturbations: effects on the host rock induced by the construction and operation of the repository and effects due to the presence and evolution of the waste and of the engineered barrier system over long periods of time (thermal, hydraulic, chemical, mechanical, gas-related, microbiological, coupled). Long-term evolution: effects on the repository system due to the natural evolution of the geosphere and of surface conditions over the time scale of interest (e.g. deformation events, vertical movements, climate change, human activities).

The total number of FEPs considered to be relevant within the repository system (waste, canister, bentonite buffer, host rock, geosphere) is 55 in the Swiss case (Nagra, 2002b, 2002d) and 70 in the Finnish case (Miller and Marcos, 2007). In Table 8.3, subsets of these integral lists are presented, in particular FEPs related to the host rock and to the geosphere as a whole, including the natural long-term evolution. FEPs related to the waste and to the engineered barrier system, to interactions between the engineered barriers and the host rock, as well as those considering future human actions, are not listed. Table 8.3 includes 9 (Swiss case) and 12 (Finnish case) FEPs. These represent the relevant aspects of FEP classes 1 and 3 above (with the exception of future human activities) but exclude FEP class 2. The FEPs from both programmes are thematically grouped in Table 8.3. It becomes evident that a large number of FEPs are addressed in both programmes, in spite of the large differences of the host rocks and of the disposal concepts. In particular, the main commonalties include the issue of fluid movement in the host rock (advection, diffusion, gas migration), the retardation of solute transport in the host rock by physico-chemical processes, future tectonic activity and fault displacements, and finally effects from the surface (erosion, glaciations, permafrost). However, there are also differences, and a number of FEPs are considered to be relevant in one programme but not in the other: .

.

The most direct flow path across the host rock is short (40 m) in the case of Opalinus Clay, so flow-path length is an issue, whereas this is less critical in the crystalline basement at Olkiluoto where the flow path will be at least 10 times longer. Opalinus Clay acts as a colloid filter due to its nanometric pore-space structure, so colloidal transport of radionuclides is not an issue. On the other hand, it may play a role in the fractured crystalline basement at Olkiluoto. In particular, bentonite colloids may be released from the buffer by water flowing in fractures. Various laboratory and in situ



.

.

.

241

studies have shown that such colloids are stable at low water salinity (e.g. Degueldre et al., 1996) and they have the potential to transport sorbing radionuclides through the geosphere (e.g. Geckeis et al., 2004; Missana et al., 2003, 2004). However, because sorbing radionuclides will be strongly retarded (if not fully contained) by sorption within the bentonite buffer and so radionuclide releases from the buffer are expected to be dominated by non-sorbing species, colloidal transport in the geosphere is not thought to be of prime importance, at least not in the expected evolution scenario of the repository system. Retardation in aquifers and in confining units embedding Opalinus Clay has no analogue in the crystalline basement at Olkiluoto because no distinction can be made between host rock and embedding units of the geosphere (the two are essentially the same). Geomorphological evolution (such as denudation and linear erosion) are important in the Swiss situation where land uplift is balanced by erosion. The crystalline rocks at Olkiluoto are more resistant to erosion, which is considered to be limited in spite of ongoing uplift and future glaciations. The possible effects of deep permafrost and deep penetration of glacial waters are studied in the Finnish case but are considered to be irrelevant for Opalinus Clay in Switzerland. In central Europe, deep permafrost is not expected to occur in future climates. Deep penetration of surficial waters is not of concern for Opalinus Clay due to its setting in a sedimentary basin with several low-permeability units that limit vertical flow.

8.4.2 Treatment of the geosphere in safety-assessment calculations Expected evolution Calculations quantifying radionuclide transport across the whole multibarrier system in the Swiss and Finnish disposal concepts were performed, and these calculations highlight the differences among the geological settings, host-rock types and the characteristics of the engineered barriers. In the reference case for Opalinus Clay, the lifetime of the steel canister containing spent fuel is assumed to be 10 ka, and thereafter all canisters are assumed to be defective (Nagra, 2002b). This means that after this time radionuclides can be released from the whole inventory of the repository. This is a gross conservative simplification of a more complex evolution over which the canisters may start to leak. In contrast, the Finnish concept considers copper canisters with corrosion rates that are much lower than those of steel. In the main scenario, the copper canisters are assumed to



Rock–water interaction; radionuclide solubility, sorption and precipitation; matrix diffusion

Geochemical retardation of radionuclide transport

Gas migration Two-phase flow

Colloidal trans- An unfractured argillaceous sedimentary rock acts as a Possible colloids originate mainly from erosion of the port colloid filter due to the pore structure with apertures in the bentonite buffer. Because sorbing radionuclides are not expected to leave the bentonite (unless it is defective), nanometre range (Voegelin and Kretschmar, 2002) sorption on colloids in the host rock may not be important

—

A free gas phase may develop because natural methane The generation of gas is a concern for intermediate-level waste (anaerobic corrosion of metals, degradation of contents in ground water are near saturation at certain levels. organics), to a lesser degree for spent fuel and vitrified high- Gas is also produced by anaerobic corrosion of metals in the

Fracture minerals such as carbonates and sulphides buffer ground-water composition and redox state. These in turn affect the speciation, solubility and sorption characteristics of radionuclides. Matrix diffusion is a retardation process operating only in fractured, advection-dominated systems. It results in a dilution of radionuclide concentrations due to mixing with pore waters in the matrix, and it provides access to sorption sites in the porous rock matrix. Co-precipitation is another process providing retardation, but it is conservatively neglected

Advective/dispersive transport along fractures dominates, and the issue is to understand the heterogeneous spatial distribution and the rate of flow. Hydraulic gradients and flow rates are expected to change in future (glaciations, land uplift, sea-level changes, water salinity)

Multiple lines of evidence suggest that flow along discrete fractures is insignificant (Nagra, 2002a). Fractures are infrequent and hydraulically not active (self-sealing). Flow through the matrix occurs but is small when compared to diffusive transport

Ground water flow (advection); dispersion

Advective flow in Opalinus Clay; heterogeneous flow

Retardation occurs mainly via sorption on mineral surfaces. Stable reducing conditions prevail and are buffered by solid phases (sulphides, organic matter). Solubility limits of radionuclides (precipitation and co-precipitation of solids) are conservatively neglected

Minimum transport path is several hundreds of metres. The first part provides most transport resistance. At larger distances from the deposition hole, solutes may be transported in more transmissive faults

Crystalline basement at Olkiluoto, Finland

Critical parameter in the case of sedimentary units of limited thickness and disposal systems in which retention in the geosphere plays an important role. The minimum vertical path length in Opalinus Clay is 40 m

Opalinus Clay in northeastern Switzerland

—


Assessment of key phenomena

Length of vertical transport path in Opalinus Clay


Key phenomena (FEPs)

Table 8.3 Results of the FEP (features, events and processes) evaluation for repositories in Opalinus Clay of northeastern Switzerland and in the crystalline basement of Olkiluoto, southern Finland. Data sources: Nagra (2002d, Table A4.1.1) and Miller and Marcos (2007, Chapter 8). While these documents include a FEP analysis for the whole repository system, only FEPs that pertain to the far-field host rock, its wider frame (i.e. the geosphere) and natural effects on the geosphere originating from the surface are included here. FEPs related to all other system components (waste, canister, bentonite buffer), interactions between the engineered barriers and the host rock (thermal, chemical, hydraulic, mechanical, coupled) and the effects of future human actions (such as drilling or exploitation of resources) are excluded

System component

Host rock

System component

Geosphere

Surface environment



Deep permafrost is not expected to occur during future cold periods. The main issue relates to the nature and rate of erosion (in particular linear deep glacial erosion) and the evolution of the biosphere

The ongoing uplift is not balanced by erosion and results in a The expected amount of total erosion (denudation in response to uplift and linear erosion by rivers) is max. 200 m net gain in elevation. Glacial erosion is estimated at 1–4 m per in the coming 1 Ma. Deep glacial erosion will also occur but glacial cycle (Posiva, 2006) will be limited to the known overdeepened valleys where this type of erosion occurred in past glacial cycles (Nagra, 2002a)

Climatic evolu- Freezing and tion permafrost; methane hydrate formation; salt exclusion

Geomorpholo- — gical evolution

Deep permafrost is expected to occur during future cold periods and will affect the hydrogeological system. The penetration depth of permafrost is a matter of current research. Salt exclusion due to partial freezing of ground and pore water, as well as the formation of clath rates, are further potential consequences of cold climates

Not well characterised, therefore conservatively neglected in Retardation is considered along the whole flow path from the reference scenario canister to surface. However, the more distal parts of the flow path do not contribute much to retardation because flow is expected to be through major, higher-transmissivity structures

Fault activity is a concern because future glaciations may result in differential movements. Breaching of canisters is unlikely because the disposal holes will not be cross-cut by major fractures

Retardation in — local and regional aquifers

Reactivation The siting regions are in tectonically quiet areas (low displacements seismicity, no active faults). A low uplift rate is essential along existing because uplift is expected to be balanced by erosion fractures

Neotectonic activity

Microbes may be present throughout the geosphere and catalyse chemical reactions. Because mostly non-sorbing radionuclides are expected to reach the geosphere, microbial effects on their transport through the host rock are thought to be limited

level waste (anaerobic corrosion of metals). The build-up of repository. Gas transport will occur in fractures, with possible gas pressure is an issue in Opalinus Clay due to its very low effects on water flow permeability. Gas migration is expected to occur by diffusion in aqueous solution, two-phase flow and pathway dilation mainly in the horizontal direction, whereas macroscopic gas fracs are not expected (Nagra, 2004; Marschall et al., 2005) The amount of water displaced by gas migration is likely to be negligible (Rodwell, 2000).


Microbes cannot metabolise in the generally nanometric Microbial populations pore network of argillaceous sedimentary rocks (Stroesand processes Gascoyne, 2002)


Assessment of key phenomena

—


Key phenomena (FEPs)

244


remain intact over the entire period of interest. Radionuclides are only released if a canister is defective (initial welding defects, corrosion defects, defects due to rock shear), and this was quantified in the ‘defective canister scenarios’ (Nykyri et al., 2008). Here, scenario Sh1 is used for illustrative purposes. This scenario assumes that one single canister has an initial defect with a size of 1 mm. Figure 8.4 illustrates the calculated release rates of radionuclides from the bentonite buffer and from the host rock at both sites. Release is given in units of dose (mSv/a), using the site-specific dose-conversion factors. The following observations can be made: .

Given the earlier onset of release from the canister in the chosen scenario (initial canister defect), short-lived radionuclides such as 90Sr enter the geosphere at Olkiluoto but not in Opalinus Clay, where

8.4 Release of radionuclides from the bentonite buffer and from the host rock as a function of time. Left: reference case for a spent-fuel repository hosted by Opalinus Clay in northeastern Switzerland (based on Nagra, 2002b; all steel canisters are assumed to be defective). Right: case Sh1 (with assumed small defect in one single copper canister) for a KBS-3V spent-fuel repository in crystalline rock at Olkiluoto, Finland (based on Nykyri et al., 2008). Radionuclide fluxes are recalculated to units of dose using conversion factors for the reference biosphere. The recalculation is hypothetical for the releases from bentonite but facilitates the comparison with releases from the host rock. Note that the ranges of both axes are different between the two sites. Not shown: 137 Cs release from bentonite buffer at Olkiluoto (peak of 9.46107 mSv/a at 77 a; negligible release from host rock).



.

.

.

.

.

245

complete containment over the first 10 ka is assumed. Because Sr is sorbing, it decays completely within the host rock. In the case of Opalinus Clay, a substantial number of radionuclides migrate across the bentonite buffer and enter the host rock, including some nuclides of the U and Th decay chains. This is not the case at Olkiluoto, where all nuclides of the decay chains are contained within the bentonite. This is likely due to the fact that only one and not all canisters are assumed to have failed and due to the slow fuel dissolution and release rate of radionuclides through the small defect. Overall, the calculated release of radionuclides from the bentonite buffer into the host rock is much lower at Olkiluoto, reflecting the fact that the copper canisters are assumed to contain completely most of the activity over the entire period of interest. For Opalinus Clay, the highest release from the bentonite is predicted to occur shortly after the assumed complete failure of all steel canisters at 10 ka after the present. This peak represents the ‘instant release fraction’, i.e. the release of radionuclides residing along grain boundaries in the spent fuel and in the gap region. These radionuclides are assumed to be transferred completely into aqueous solution as soon as water penetrates the spent fuel; i.e. their release is independent of the slow fuel-dissolution rate. A corresponding peak is not seen in the Finnish case, most probably reflecting different assumptions made regarding the parameterisation of the instant release fraction and the transport from the fuel into the bentonite. Opalinus Clay is a very efficient transport barrier and contains most radionuclides. Only weakly or non-sorbing radionuclides (14C, 79Se, 36 Cl, 129I) are released into the embedding rock units. The reason for the good host-rock performance is the fact that transport occurs mainly by diffusion and because of the high sorption coefficients for many radionuclides, in particular for actinides (Table 8.1). The maximum doses are reduced by orders of magnitude even for the non-sorbing radionuclides, and the releases are shifted to substantially more distant times. The elevated releases of these radionuclides from the bentonite at early stages are attenuated. This finding contrasts with that for Olkiluoto, where retardation due to transport through the geosphere is weak for non-sorbing radionuclides such as 14C, 36Cl and 129I. The calculated release rates from the host rock into the biosphere are lower for the crystalline basement at Olkiluoto than for Opalinus Clay in Switzerland, essentially reflecting the assumption that only one copper canister (out of about 3000) fails in the former situation but all steel canisters fail in the latter. Nevertheless, both assessments yield doses that are orders of magnitude below regulatory limits. In conclusion, the weaker performance of the steel canisters in the Swiss


246

Geological repository systems for safe disposal concept is compensated by the high transport resistance of Opalinus Clay. In the Finnish concept, the canister, together with the surrounding bentonite buffer, is expected to retain most radionuclides over the whole period of interest (with the exception of non-sorbing nuclides, such as 14 C, 36Cl and 129I). The role of the crystalline bedrock as a transport barrier is limited in the radionuclide transport calculations, in part due to the comparatively low transport resistance (relatively high hydraulic conductivity), in part due to the difficulty to predict fully the dynamic evolution in the future (land uplift, glaciations, spatial heterogeneity), and due to the simplified and conservative approach used to represent retardation in the rock matrix.

Alternative and ‘what if’ calculations The expected evolution (or the reference case) represents the situation in which the barrier system functions according to the best current understanding. Alternative calculations explore uncertainties in scenarios, conceptualisations and parameter values. ‘What if’ calculations are meant to test the robustness of the disposal system, and they include scenarios and parameter values that are not supported by (or even in contradiction to) current scientific understanding. In the Finnish safety evaluation for a KBS-3V repository (Nykyri et al., 2008), a number of alternative and ‘what if’ calculations were performed. All underlying scenarios assume the failure of a canister in consequence of one of the following events and processes: . .

. .

An initial defect due to inadequate encapsulation of the spent fuel (hole up to 10 cm in size). A delayed defect due to corrosion (10 cm at times of 10–100 ka after the present). This scenario addresses the uncertainties regarding the stability of the geochemical environment (such as deep infiltration of oxygen-rich glacial water or corrosion due to sulphide-rich ground water). A defect growing with time (0–100 ka after the present) due to a disrupted bentonite buffer, resulting in access of sulphide-rich water and high flow rate. A delayed defect due to displacement along a fault cross-cutting the deposition hole (1–70 ka after the present). This scenario covers the issue of geomechanical (in)stability of the environment in response to future glacially triggered crustal movements.

Under these assumptions, the following cases were explored: .

Suite of ground-water compositions from glacial to saline, with effects



. . .

247

on anion-accessible porosity, diffusion coefficients, radionuclide solubilities and sorption characteristics. Increased water flow rates, enhancing the transfer of radionuclides to the fractured host rock and transport through the geosphere. Increased fuel-degradation rate (in particular in situations when oxygenated water penetrates to the repository depth). Formation of gas in the canister, leading to displacement of water that contains radionuclides out of the bentonite and to radionuclide migration in the gas phase.

The calculations show that, as in the base case, 129I, 135Cs and 14C contribute most to the total dose rates in the majority of the calculation cases. In ‘what if’ cases combining an initial canister defect with high flow rates and saline ground water, calculated peak doses would be reached in the range of only decades to a few centuries after emplacement, and the shortlived nuclides 90Sr and 226Ra would dominate the dose. This reflects the rapid flow and the reduced calculated retardation of these nuclides in the geosphere (transit times through the geosphere become short with respect to the decay of 90Sr and 226Ra, and matrix diffusion and sorption in the matrix are even more limited), and the higher salinity (resulting in higher assumed solubility limits and lower sorption Kd values). In the Swiss programme for deep disposal in Opalinus Clay (Nagra, 2002b), a number of parameter variations were conducted around the reference case. Variations of host-rock properties included higher/lower flow rates, higher diffusion coefficients and lower sorption coefficients, with only a limited effect (factor ≤4) on calculated dose rates from spent fuel. In the framework of alternative conceptualisations of processes in various system components (e.g. fuel dissolution, limited performance of the bentonite, recurrent flow events in the host rock during glacial cycles, radionuclide release along shaft and ramp, gas-induced displacement of pore water), the calculated dose rates originating from spent fuel are never substantially higher than in the reference case. Further, alternative scenarios explored the consequences of gas pathways on the transport of volatile radionuclides and of various kinds of human actions. Lastly, a substantial number of ‘what if’ calculations were performed. Some of these are presented here, all relating to spent fuel: .

Upward Darcy flow in Opalinus Clay is assumed to be 261012 m/s, i.e. 100 times the value in the reference case, in which case transport is advection-dominated. The resulting maximum doses are a factor 20 to 40 higher compared to the reference case and are reached at earlier times (30–80 ka after the present). The same radionuclides (129I, 36Cl, 14C and 79 Se) dominate. The increase of the maximum dose rate is lower than the


248

.

.

.

.

Geological repository systems for safe disposal increase of the Darcy flow because diffusion also contributes to transport in the reference case. Assuming the existence of a vertical fracture with a transmissivity of 16109 m2/s, cross-cutting all emplacement tunnels of the repository results in a maximum dose rate of 6.56104 mSv/a at a hydraulic gradient of 1 m/m. This is slightly more than one order of magnitude higher than in the reference case. The limited increase of the dose rate is partly due to the retardation by matrix diffusion that was considered to occur along the fracture. The peak dose rate occurs shortly after the assumed canister failure at 10 ka, corresponding to the contribution of the instant release fraction. In addition to the radionuclides that dominate in the reference case (Fig. 8.4, bottom left), 226Ra becomes important over the entire period of interest (i.e. up to 1 Ma). As a nuclide of the 4N+2 chain, it is continuously produced by ingrowth from 238U, so it is present in the fuel matrix at all times, in spite of its short half-life of 1.6 ka. The combination of the low sorption coefficient of 0.0007 m3/kg for Ra (Table 8.1) and the high advective flow rate assumed in this ‘what if’ case result in a transport time through Opalinus Clay that is on the order of the half-life of 226Ra, resulting in an incomplete containment of 226Ra in the host rock. Nevertheless, the peak dose rate is due to 129I and not 226Ra. An increase of the fuel-dissolution rates by factors of 10 and 100 translates into dose-rate increases by factors of about 4 and 10, indicating the non-linear relationship of these parameters (due to solubility limits and other factors). 129I,36Cl, 14C and 79Se dominate the dose rates. Assuming that a discrete gas phase is generated in the near field and then breaks through to the overlying aquifer, thereby bypassing the retardation capacity of the host rock, leads to a maximum additional dose rate from volatile 14C of about 56105 mSv/a. This value is similar to the total dose rate in the reference case, and is reached at around 10 ka after the present. Setting the sorption coefficient for 129I to zero in the bentonite and in the host rock doubles the maximum dose rate, and the maximum is reached at an earlier time. This illustrates that even the low sorption coefficients of 56104m3/kg (bentonite) and 36105 m3/kg (host rock) as used in the reference case have a marked retarding effect on iodine transport.

In conclusion, the alternative calculations in both disposal programmes were meant to quantify the consequences of various FEPs and related uncertainties on system performance, as listed in Table 8.3. The robustness of the disposal system was further explored by ‘what if’ calculations. In the Finnish analysis, the main objective of these calculations was to explore the



249

(in)sensitivity of the engineered barrier system to internal and external effects influencing geochemical stability and mechanical integrity. In the Swiss safety assessment, both the functionality of the engineered barriers and the performance of the host rock as a transport barrier were considered. All calculations at both sites resulted in maximum dose rates well below the regulatory limit of 0.1 mSv/a. Dose rates from spent fuel are dominated by radionuclides occurring in anionic form (129I, 36Cl, 79Se, 14C), by 14C potentially occurring in a gas phase and by the weakly sorbing cation 135Cs. With the exception of 14C, all these nuclides have half-lives > 100 ka. Actinides do not contribute to calculated doses because of their low solubilities and favourable sorption characteristics (Table 8.1), and are contained in the engineered barriers and/or in the geosphere. The only exception is 226Ra, which (together with 90Sr in the Finnish analysis) may become important in ‘what if’ calculations where unrealistically high flow rates through the geosphere are considered.

8.5

Emerging trends

8.5.1 In situ studies of porosity and transport parameters of the rock matrix In argillaceous systems, there is confidence that porosities, diffusion coefficients and sorption characteristics measured in the laboratory do not deviate substantially from the in situ values (e.g. Van Loon et al., 2004; Wersin et al., 2004, 2008). Van Loon et al. (2003) conducted through- and out-diffusion experiments with Opalinus Clay at pressures up to 15 MPa. No measurable effect on porosity was identified between experiments at 4 and 15 MPa, and the diffusion coefficient decreased only by about 20%. Diffusion coefficients measured in the laboratory and in situ were also used to model natural-tracer profiles across numerous argillaceous formations, and the resulting diffusion times were always within the range that is plausible and consistent with independent palaeo-hydro-geological arguments (Mazurek et al., 2009). In conclusion, stress release and other artefacts appear not to affect the pore structure of clays and shales substantially, and laboratory measurements can be transferred to in-situ conditions. Similarly, sorption Kd values based on laboratory measurements on powdered material are generally in agreement with in-situ values derived from modelling in situ diffusion experiments (e.g. Van Loon et al., 2005; Wersin et al., 2008). These conclusions are likely not to be true for crystalline-basement environments, where porosity is much smaller and even minor disturbances can have an impact on porosity and the tortuosity of the pore network. In recent years, efforts have been made in rock laboratories to characterise


250


matrix porosity and diffusion coefficients under in situ conditions. These efforts are continuing, with a trend towards long-term experiments, in order to achieve penetration distances of diffusing tracers well beyond the zone around the borehole where some degree of stress redistribution occurred. Examples of completed experiments include those by Vilks et al. (2003, 2004) in the Canadian underground research laboratory in Manitoba, the CP experiment at the Grimsel test site in Switzerland (Frieg et al., 1998; Schild et al., 2001; Mo¨ri et al., 2003; Ota et al., 2003). Long-term experiments are ongoing in the Swedish underground rock laboratory at A¨spo¨ (LTDE-SD; interim report in SKB, 2008b, Section 3.3) and in the Grimsel test site in Switzerland (LTD; preliminary information in Mo¨ri et al., 2007, and Eichinger et al., 2008). In the framework of these projects, diffusion of non-sorbing as well as sorbing tracers is studied, so the experiments are expected also to provide in situ information on sorption behaviour. While today Kd values for sorption are usually based on batch laboratory experiments using rock powders, there is evidence that the in situ mineral-surface area and sorption Kd values are smaller (e.g. Andre´ et al., 2008). The transferability of laboratory-derived Kd measurements to in situ conditions was first addressed by Frick (1994) and Heer and Hadermann (1996) in the framework of in situ migration experiments in granodiorite of the Grimsel test site (Central Switzerland). For Na, the Kd derived from modelling of the in situ experiments was about one order of magnitude smaller than the one obtained from batch sorption experiments on rock powders crushed to a grain size 100 m 15–20 cm 35% or 35% > SF > 15% is employed. Then the alkaline content in pores, Na+ and K+, remains around 100 mmol/l. BFS does not reduce the pH as do FA and SF.

Therefore, the low-pH cementitious materials based on OPC must include higher contents of mineral additions, over 40%, to ensure that free portlandite is significantly reduced or even avoided. To eliminate portlandite, mineral additions are recommended, particularly those that consume portlandite during hydration via pozzolanic reaction (Alonso et al., 2006; Cau Dit Coumes et al., 2006; Lagerblad et al., 2006). Silicabased materials (silica fume or fly ash, etc.) allow the production of low-pH cementious material, either by using isolated material (binary mixes) or by mixing several mineral blends (ternary or more blended cements). In the case of low-pH cementitious material production, it has been determined that if OPC is used as the basic cement of the low-pH blend, at least 55% of SiO2 must be in the binder, or pozzolan blended additions above 40% have to be employed. Indeed, the composition of the blended agent has a marked effect on the pH, with SF being the most effective (due to its higher silica content), as shown in Fig. 10.4. Formulations with a sufficient percentage of mineral addition, > 40%, and a total SiO2 content > 55% show a pH below 11 after 90 days of curing. Evolution of the pore fluid pH of several low-pH cement formulations is shown in Fig. 10.5. Table 10.4 compiles the pore fluid composition of lowpH cement pastes from fluid expression tests after 90 days of hydration (Garcı´ a Calvo et al., 2008). The higher the increase in the silica content of the binder, the higher the decrease in the pore fluid pH of the paste. This effect is more relevant at longer curing times (> 90 days) because over short times the pH can be even higher in the low-pH cement pastes, as already described by Larbi et al. (1990). These results indicate that the percentage of silica in the cement formulation has to be above 45% to obtain a pH below 12.5 at 90 days, indicating that the portlandite has been removed from the cement paste as will be shown below, when solid phase evolution is evaluated.


298


10.4 Silicate content of cement formulation influence on pore fluid pH at 90 days.

10.5 Evolution of the pore fluid pH of several cement pastes based on OPC.

Observing Table 10.4, a decrease in the alkaline content in pore solutions is evident in binder formulations with high mineral admixtures contents. With regards to Ca2+, in up to 90 days it seems to increase in the pore fluid with respect to OPC without blending addition. The increase of the SO42


Development and application of low-pH concretes

299

Table 10.4 Composition of the pore fluid of low-pH pastes based on OPC at 90 days of hydration Chemical composition (ppm) +

+

Sample

pH

Na

K

OPC 40% SF 50% SF 10%SF + 10%FA 35%SF + 30%FA 50%SF + 30%FA

12.9 12.2 11.2 12.6 11.2 10.5

603.27 119.7 165.1 118.50 175.09 220.75

2129.5 321.8 368.6 459.40 387.92 219.02

Ca2+

Si4+

Al3+

SO42

474.41 612.1 650.9 689.86 486.10 1208.0

1.23 9.82 60.27 ND 47.98 72.28

ND ND ND ND ND ND

12.78 84.40 2181 31.63 1819 3105

ND: non-detected.

10.6 Evolution of Na+, K+ and Ca2+ contents in a low-pH cement paste (60%OPC + 40%SF).

content in the pastes with lower pore fluid pHs is also very significant as well as the increase in the Si4+ amount, which is supposed to come mainly from the mineral admixtures used. Figure 10.6 shows the evolution with curing time of the Na+, K+ and Ca2+ contents in the pore solution of cement paste with 40% of SF. This figure clearly shows that over short periods of time low-pH cements have a decrease in the alkali content of the pore solution but an increase in the calcium one. Concerning the evolution of the low-pH cement paste microstructures, the portlandite content decreases with curing time and has fully disappeared at 90 days in those with a percentage of SiO2 ≥ 55%, which can be confirmed from thermal analysis (DTA/TG), such as those results shown in Fig. 10.7, where the portladinte content is graphed for six cement pastes at 90 days of curing. The introduction of a high content of mineral admixtures also caused relevant changes in the CSH gel compositions of the cement pastes, which is the main binding phase in all Portland cement-based systems. The Ca/Si


300


10.7 Portlandite content of the cement pastes studied at 90 days (TG analyses).

10.8

Influence of Na2Oeq on pH.

ratio of CSH in conventional Portland cement pastes varies from 1.2 to 2.3, with a mean of 1.75 (Klur et al., 1998). On the contrary, in the low-pH cements this ratio (measured using backscattered electron microscopy with EDX analyses) varies from 1.2 to 0.8. CSH gels with these CaO/SiO2 ratios are typically considered to be formed by longer length chains of tetrahedral silica (Cong and Kirkpatrick, 1996). Indeed, Stronach and Glasser (1997) noted that to obtain a low pH in the cement materials the CaO/SiO2 ratio must be lower than 1.1. Furthermore, although in conventional OPC pastes the increase in the content of alkalis, Na+ and K+, contributes to increase the pH in the pore water, if high contents of mineral admixtures are used this general rule cannot be applied. This can be seen in Fig. 10.8, which shows the influence of alkalis in the pH of the pore fluid.



301

There is another important issue regarding the alkaline content in the pore fluid of low-pH OPC pastes. A binding process during hydration that removes the alkalis from the pore solution of low-pH pastes could be taking place inside the new CSH gel. Glasser and Marr (1985) published findings that SF is very effective in removing K+ from the pore fluid, due to the formation of smaller CaO/SiO2 ratios in the cement pastes. CSH gels with high CaO/SiO2 ratios have a positively charged surface, but as the CaO/SiO2 ratio decreases the charge becomes neutral, leaving van der Waals forces to operate, and at the lowest CaO/SiO2 ratios the surface charges of the pores become negative. In fact, Hong and Glasser (1999) demonstrated that the alkaline binding process is more effective when the Ca/Si ratios of CSH gels are between 1.2 and 0.85, as occurs in low-pH cements. Therefore, to formulate a low-pH cement, adding pozzolanic materials to OPC should be advantageous in at least three respects: (1) portlandite formed by the hydration of OPC is converted into CSH by pozzolanic reaction, (2) OPC is diluted and (3) the CaO/SiO2 ratio of CSH is lowered, which enhances their sorption capacity of alkalis and reduces their equilibrium pH. Furthermore, these additions are known to improve some properties of the cementitious materials: bleeding is limited (Yogendran et al., 1987) and porosity is refined (Huang and Feldman, 1985; Durekovic, 1995), which in turn reduces permeability and diffusivity (Bentz et al., 2000; Oh et al., 2002). However, some difficulties might be expected with high contents of these mineral additions for the low-pH concrete applications, as shown in Section 10.4.

10.3.2 Low-pH cements based on calcium aluminate cement (CAC) Calcium aluminate cements (CACs) represent an interesting alternative because their pore water pH, ranging from 11.4 to 12.5, is reduced as compared to OPC (Gon˜i et al., 1991). The main difference between Portland and calcium aluminate cements lies in the nature of the active phase that leads to setting and hardening. Monocalcium aluminate (CA) is the principal active phase in the CAC, which reacts with water to give calcium aluminate hydrates. However, direct use of these binders comes up against one main difficulty. The hydration reaction of CAC at environmental temperatures produces hexagonal hydrated calcium aluminates, CaAl2O14H2O (CAH10) and Ca2Al2O13H16 (C2AH8). The stable phases are Ca3Al2(OH)12 (C3AH6) and Al(OH)3 (AH3), and the other phases will inevitably convert to these, decreasing the mechanical properties of the material (Midgley and Midgley, 1975; Fentiman et al., 1990; Xiandong and Kirkpatrick, 1993; Capmas and


302


George, 1994; Scrivener et al., 1999). The conversion reactions of CAC are shown schematically below: 3CAH10 ?C3 AH6 þ 2AH3 þ 18H

½10:1

3C2 AH8 ?2C3 AH6 þ AH3 þ 9H

½10:2

where CAH10 is CaAl2O14H2O, C2AH8 is Ca2Al2O13H16, C3AH6 is Ca3Al2(OH)12, AH3 is Al(OH)3 and H is H2O. Some authors have found that an interesting way to reduce the hydrate conversion and the decreasing strength is to replace some of the CAC by blast furnace slag (BFS) or a pozzolan such as microsilica and metakaolin (Majumdar et al., 1990a, 1990b; Collepardi et al., 1995; Ding et al., 1995; Quillin et al., 2001; Hidalgo et al., 2008, 2009). The reaction that avoids the conversion of CAC hexagonal hydrated phases could take place in the following way (Majumdar et al., 1990a, 1990b; Collepardi et al., 1995; Rivas Mercury et al., 2006): the silica content of the mineral addition would react with the calcium aluminates, initially avoiding the formation of the hexagonal form C2AH8 and, subsequently the conversion in the cubic form C3AH6. Therefore, instead this cubic phase, a hexagonal aluminate hydrate containing silica, called gehlenite (Ca2Al2SiO7·8H2O or C2ASH8), is proposed to be formed. Majumdar et al. (1990a, 1990b) considered that the amount of C2ASH8 is dependent on the capacity of a mineral addition to release silica. It is important to mention that in the literature there is not agreement between authors on the precipitation of the C2ASH8 phase (gehlenite). Although studies about stabilisation of CAC with mineral additions are scarce, this solid phase has been observed in CAC mixes with BFS, SF, metakaolin and FA. Majumdar and Singh (1992) and Fu et al. (1995) concluded that, in mixes based on CAC with a silica fume (SF) content between 30 and 50%, the gehlenite is the main hydration product before a week (T < 40 8C). Some authors (Rivas Mercury et al., 2006; Hidalgo et al., 2008, 2009) also reported that in the system CAC-SF and CAC-FA, silica reacts with the calcium aluminate phases in the cement and water to form different crystalline hydrates (with variable proportions of Ca, Al and Si) such as Ca2Al2SiO7·8H2O (C2ASH8, called gehlenite), Ca3Al2(SiO4)3-x (OH)4x (0 < x < 3) (katoite) and not very well defined and complex zeolite-type phases. The Ca(OH)2, Al(OH)3, CaAl2O4·10H2O, Ca2Al2O5·8H2O, Ca2Al2SiO7·8H2O and a series of solid solutions, with the general formula of Ca3Al2(SiO4)3-x(OH)4x, can be located in the CaO–Al2O3–SiO2–H2O system for temperatures below 100 8C. Indeed, the whole series exists with the end members, known as grossular when x = 0 and hydrogarnet if x = 3. Since these solid solutions are also present in nature, mineralogists classify



303

10.9 Pore fluid pH of several cement pastes based on CAC at 90 days of hydration.

them as hibschite for the minerals with 0.2 < x < 1.5 and as katoite for the ones with 1.5 < x < 3 (Rivas Mercury et al., 2003a, 2003b, 2007). The identification of the calcium carboaluminate hydrate C4AcH11 as a precursor of hydrogarnet–garnet solid solution members with silica in their composition, C3AS3-xH2x (CASH), when SF and/or FA are used, has been also proposed in the literature (Hidalgo et al., 2008, 2009). Therefore, it seems clear that the inclusion of mineral additions in mixes based on CAC increases their microstructure stability and even slightly decreases the pore fluid pH value over long time periods, as shown in Fig. 10.9. However, in this case there is neither a minimum nor a maximum percentage of mineral addition to employ for optimum low-pH cement developing, but shows that the percentages are related to avoidance of the conversion process. It is also remarkable that, as well as occurring in low-pH cements based on OPC, the percentage of alkalis does not alter significantly the pore fluid pH of these materials, and again the alkali binding process is observed, although it is less clear than in the case of OPC low-pH mixes.

10.3.3 Low-pH cements based on other cement types Calcium sulphoaluminate cements Raw mixes for calcium sulphoaluminate (CSA) clinkers differ from those for OPC in that they contain significant amounts of sulphates. The CSA clinker is mainly composed of yeelimite (4CaO·3Al2O3·SO3), belite (C2S) and an Al-rich ferrite (Andac and Glasser, 1995). CSA cement pastes harden through the formation of an initial ettringite skeleton, and its subsequent infilling by mixtures of ettringite, calcium monosulphoaluminate hydrate,


304


alumina and ferrite gel (Kasselouri et al., 1995; Zhang and Glasser, 2002). The reported pore solution pH values of CSA mortars are highly variable, ranging from 8.5 to 13 (Andac and Glasser, 1999; Kalogridis et al., 2000; Janotka et al., 2003), but they must be studied in depth before using in HLW underground repositories. Phosphocalcic cements These cements could seem attractive at first sight due to hydroxyapatite, which is the thermodynamically stable resulting product, has a very low solubility in neutral or basic medium and leads to an equilibrium pH within the range 7–9 (El Jazairi, 1987; Popovics et al., 1987; Driessens, 1995; Pera and Ambroise, 1998; Chow, 2000; Lemaıˆ tre, 2001). However, up to now, very few materials have been designed to be workable using contemporary engineering practices (Cau Dit Coumes et al., 2006). Magnesia phosphate cements Their pore solution usually ranges between 7 and 8 (Pera and Ambroise, 1998). These cements are based on an acid–base reaction between dead burnt magnesia and a phosphate salt (El Jazairi, 1987; Popovics et al., 1987; Seehra et al., 1993; Wagh et al., 1997; Soudee and Pera, 2000).

10.4

Development and production of low-pH concretes: shotcrete plug application

This section shows the design of a concrete mix to be shotcreted at a concrete plug where a low pH is mandatory. The meaning of ‘concrete design’ is to select suitable constituents (among the available ones) and to determine the proportions to achieve the desired performance both in the fresh and hardened states. As the design of the concrete plug involves different characteristics of the concrete (mechanical strength, deformation, hydraulic permeability, heat of hydration, shrinkage, potential durability, etc.) and other technological properties, such as consistency, pumpability and a low loss in workability with time, the design and optimisation of the concrete must be carried out on a performance basis. Thus, the first step is to determine the requirements imposed by the specific application to the fresh and hardened state, usually known as ‘functional requirements’, the main one being its low pH, which must be around 11, to prevent the bentonite in contact from deteriorating. As defined in Section 10.2, functional requirements for the concrete of the shotcreted plug made of low-pH cement can be summarised as follows: at the fresh state, it has to be workable, pumpable and projectable while


Development and application of low-pH concretes Table 10.5

305

Functional requirements for the shotcreted concrete

Requirement

Target

Complementary considerations

Maximum pH

11

Hydraulic conductivity

K < 1010 m/s

Measurement technique Age of measurement

Final mechanical properties Young modulus Poisson ratio Tensile strength Friction angle Cohesion Compressive strength

< 20 GPa 0.2–0.3 1 MPa > 378 2 MPa 10 MPa

Workability

> 2h

Pumpability Peak of hydration

500 m < 40 8C

Final means: long-term The age of testing to assess final properties will be established from test results

Negligible workability loss for at least 2 hours A long pumping distance can be expected Temperature rise is influenced by the accelerating admixture added at the shotcrete nozzle

compressive strength, elasticity modulus, water permeability and pH can be highlighted as the most relevant requirements for the hardened state. Moreover, if the concrete maintains a relative humidity in the pores above 80%, the properties of concrete change over time due to progressive hydration of the cement and the pozzolanic reactions. The result is an increase in strength and modulus of elasticity in conjunction with a decrease in permeability and the pH has to be taken into consideration as well. On the other hand, the projection of the concrete requires the mix to be pumpable in real working conditions (equipment available, distances and heights when pumping, temperature, etc.) and to adhere to a vertical surface (the bottom of the plug). This issue requires the use of an accelerator admixture at the nozzle. Since delays must be expected during work, it would be advantageous if the mixture could preserve its characteristics in a fresh state (workability and pumpability) for at least two hours after mixing. Last, but not least, the adiabatic temperature rise must be below 40 8C. Table 10.5 summarises the main functional requirement for the low-pH shotcrete of the concrete plug.

10.4.1 Mix design procedure Concrete can be seen as a composite material composed of an aggregate skeleton bound by a paste matrix. The paste itself is composed of the low-


306


10.10

Flowchart to select the constituents of the paste.

pH cement formulation, water and the chemical admixtures. As most of the physicochemical reactions occur at the paste phase, the compatibility among different constituents can be assessed in paste evaluations, the aggregate being almost inert. Thus, the selection of the concrete components is divided into two stages: paste components and aggregates proportioning. The flowchart in Fig. 10.10 shows the selection process of the paste constituents. As can be seen from the compatibility evaluation chart shown in Fig. 10.10, different tests have to be performed successively to determine suitable



307

combinations of low-pH cement formulation and the admixtures to be used. It must be kept in mind that chemical admixtures are developed and optimised for ‘standard cements’ and thus their efficiency with special cements (low-pH) cannot be neglected. Aggregates selection is a key issue in concrete design and production as about 70% of the concrete is made of aggregates and they strongly influence water demand, workability, pumpability and projectability of the concrete. To select suitable aggregates, two main considerations arise: the suitability of aggregates in terms of strength, surface hardness, dimensional stability, intrinsic durability, among others, and the aggregate grading, i.e. the particle size distribution. To produce low-rebound shotcrete, it is advisable not to use a maximum size over 16 mm and preferably not over 12 mm. Different available fractions can be mixed in order to compose an aggregate skeleton with a grading able to obtain a fresh concrete that fulfils the requirements for shotcrete. The easiest way to adjust the proportions of the aggregate fractions is to consider the limits suggested by different organisations, such as ACI (American Concrete Institute), AENOR (Spanish Standards Organisation), SCA (Shotcrete Concrete Association of the United Kingdom), etc. If the percentage passing through different sieves is plotted against the log of the sieve size, aggregate size distribution can be graphed. If the limits suggested are also indicated in the same plot, it is easy to verify the compliance with these limits, as shown in Fig. 10.11. Other aggregate characteristics, such as absorption (saturated in a dry-surface condition) and density, are also necessary to adjust concrete proportions.

10.11 Aggregate grading and bandwidth limits.


308


Chemical admixtures The importance of selecting suitable and compatible chemical admixtures when selecting the concrete constituents and designing the concrete mixture cannot be overemphasised. Once the compatibility between the low-pH cement formulation and the admixtures is verified, it is necessary to adjust the dosage of the chemical admixture in the concrete. The admixtures are used to modify the properties of the concrete, principally in the fresh state, and to accelerate hardening. The chemical admixtures used in the shotcrete plug discussed in this section were: air-entraining admixtures (used to improve the workability of the concrete mixes, which have been affected by crushed fine aggregate), superplasticisers or high-range water-reducing admixtures (they are essential when utilising ultra-fine mineral additions, such as silica fume, to properly disperse the particles; moreover they also allow for a reduction in the water content without compromising the fluidity of the mixtures) and accelerator admixtures (they are used directly at the nozzle; their effect is almost immediate and the goal is not only to achieve early strength but also to ensure that the concrete will adhere to the surface without slipping or collapsing. As their use can cause a significant reduction in the final mechanical strength of the concrete, they must be evaluated exhaustively together with the range of materials in question).

10.4.2 Concrete design Concrete mix design is based on two main parameters: water demand (for a given workability) and the water/binder ratio to attain compressive strength. For conventional concrete, using conventional cement types, there are design charts or design methods that provide preliminary data on these figures, such as Abrams law, sketched in Fig. 10.12. Each of the three curves corresponds to a nominal compressive strength of cement, according to European standards, but the nominal compressive strength of the low-pH formulation has to be determined, as well as the water demand. A rather simple procedure is to prepare mortars (no coarse aggregate in the formulation) of equivalent fluidity and to test their corresponding compressive strengths at a given age, because strength evolution with time strongly depends on the cement type. The comparison between samples cast with different water/cement (w/c) ratios and different cement types can be made from the percentage of the compressive strength at any age versus the 28-day strength. Thus, no matter what the actual strength level is, all samples exhibit the same relative value at 28 days: 100%. Figure 10.13 represents the relative strength (at 28 days) versus log age. From the figure, it can clearly be seen that the strength evolution of calcium aluminate based low-pH cement is negligible over 7



309

10.12 Relationship between compressive strength (28 days) and the w/c ratio (Abrams law).

10.13 Relative strength evolution versus log time for different cement types.

day curing while Portland cement based low-pH cements exhibit an increase of about 40% at 90 days. Therefore, long-term CAC based low-pH formulation properties can be assessed after a few days of curing (typically 7) but OPC based low-pH formulations should be tested after longer curing periods (typically 90 days). Indeed, the evolution of strength (and other properties) differs considerably, depending on whether the concrete is CAC or OPC based. The composition of the mixes can be set via the absolute volume method (the sum of the volumes equals 1000 dm3) using the aggregates and the paste components already defined. A cement content of approximately 300 kg/m3 was determined and the water adjusted to achieve suitable fluidity.


310


Table 10.6 Concrete proportions and properties in the fresh and hardened state Component

Weight (dry)

Properties in the fresh state

Water Cement Water/binder Gravel Fine gravel Sand Water reducer (1.8%)

277 307 0.90 616 200 818 5.5

Unit weight Slump Cohesion Aspect

2.23 t/m3 12 cm Good Good

Compressive strength 7d 28 d

16.2 MPa 29.0 MPa

Adjustments are usually necessary when preparing the trial mix, but in just a few trials it is possible to achieve reasonably good results. Nevertheless, the properties at the hardened state have to be determined at different ages (typically 7, 28 and 90 days) and the pumpability and projectability evaluated in real scale testing. A decrease in mechanical properties of about 35 to 40% of the sprayed concrete as a result of the use of an accelerator and the spraying technique have to be considered when setting the target for strength for conventional concrete.

10.4.3 Trial mix and properties of concrete in the fresh and hardened state Once initial proportions are set, a trial mix has to be tested in order to adjust the water content according to the required consistency (slump), checking the mix for cohesion, aspect (not too sandy, nor harsh) and unit weight. Once the requirements at the fresh state are fulfilled, cylindrical or cubic samples have to be cast to determine the compressive strength at different ages and the static elasticity modulus. Values have to comply with the requirement for compressive strength plus 40% (about) to compensate for further losses when the concrete is projected. Table 10.6 shows low-pH concrete proportions and properties at the fresh and hardened states.

10.4.4 Shotcrete trials and sampling Once the basic concrete is checked against compliance of the requirements, it is necessary to make trials with shotcrete, in real scale. The trials have to be done according to the following basic work process: . .

Conditioning or preparation of the trial zone Preparation of basic concrete in the facility or in a truck mixer


Development and application of low-pH concretes Table 10.7

Shotcrete properties and functional requirements

Compressive strength (MPa) Static elasticity modulus (GPa) pH Water permeability K (m/s) Pumpability Slum loss (2 h) Adiabatic temperature rise (8C)

. . .

311

Shotcrete properties

Requirements for shotcrete plug

23–25 14–16 10.4 2–561011 Verified Verified < 20

> 10 < 20 < 11 < 161010 — — < 40

Test of quality of fresh and adjusted concrete Samples of basic concrete cast (cylinders) Pump and/or spray trial in the trial zone

After 28 days of curing, samples were taken from reference panels and tested. Table 10.7 shows the results obtained and the compliance with functional requirements. Therefore, according to all the results shown in this section, it could be said that it is quite possible to construct shotcrete plugs that fulfil the functional requirements needed in an underground repository of high radioactive waste using low-pH cement formulations.

10.5

Long-term durability

Due to the location and the long service life of this type of product, their durability properties must also be guaranteed. Therefore, in this section two parameters closely related to the durability of low-pH cementitious materials are evaluated: resistance to long-term groundwater aggression and the use of low-pH cements in reinforced concretes (susceptibility to corrosion). Although this chapter focuses on the degradation processes expected in low-pH cementitious materials, some ideas are now given concerning the processes expected in the region including cementitious materials and bentonite. However, it must be said that in this subject most of the studies published in the literature deal with conventional Portland concretes instead of low-pH ones (which are more recent). As already stated in previous sections, concrete pore waters originating from a Portland cement have high alkalinity and are able to react with and modify the bentonite barrier. These early cement pore waters have alkali ions that will be transported by diffusion and possibly by advection due to bentonite suction if this material is emplaced under unsaturated conditions. During a relatively short-term stage that is characterised by the leaching of dissolved alkali hydroxides, the


312


alteration of bentonite is characterised by montmorillonite dissolution and sodium and potassium zeolites (e.g. analcime, phillipsite) are formed (Bauer and Velde, 1999; Vigil de la Villa et al., 2001; Ramı´ rez et al., 2005). In a second alkaline long-term stage, the system is buffered by the portlandite dissolution and Ca2+ is the predominant cationic species in the leaching solution of the concrete. In this case, zeolites are replaced presumably by cement phases such as calcium silicate hydrates (CSHs) in the altered bentonite region. Some authors (Taylor, 1990; Berner, 1992) include a third stage where the incongruent dissolution of CSH gels and other cement phases determine the pH buffering capacity from the concrete side. Modelling simulations (Fernańdez et al, 2009) have shown that the mineralogical transformations in the host rock–concrete–bentonite system become significant only after 103 years. At this time, the precipitation of secondary minerals (mainly analcime, Mg-saponite, ettringite and brucite) cover most of the volume over the first centimetres of the bentonite layer on both sides of the concrete barrier and this reduces the porosity to ~5%. Total clogging is observed after 56103 years, and thereafter the concrete phases dissolve gradually and still release Ca2+, which is incorporated in the exchanger complex of bentonite, displacing the rest of the exchangeable cations. Analcime and Mg-saponite precipitate along the bentonite section over the long term (105 years).

10.5.1 Interaction of groundwaters with low-pH concretes Although concrete is stable in high humid environments, direct contact with water that is stagnant, percolating or flowing produces a diffusion of the pore solution and also alteration of the solid phases. Degradation of concrete due to leaching occurs when the hydrates in cementitious materials dissolve into the surrounding water and the precipitation of new phases takes place. This may cause a loss of strength. Although the expected degradation rate is slow, its evaluation is very important for structures near field HLW repositories, where extremely long-term stability is needed. Due to the difficulties for long timescales involved in laboratory experiments using groundwaters, accelerated tests have to be used to qualify the different concretes for deep repositories. The leaching processes of materials that contain cement pastes are a combination of chemical reactions and diffusion transport and have to be studied thermodynamically and kinetically. Cement degradation depends on physical factors including the effect of porosity, compressive strength and density, as well as on leachant characteristics including the effect of pH, flow rate, temperature and water chemical composition. In realistic conditions of HLW repositories, concrete durability is based on its interaction with clays and granite groundwaters. In general,



313

groundwaters are mineralised solutions; however, their exact composition on site must be known in order to evaluate their interaction with low-pH concretes. For example, salinity of groundwater is very variable depending on the emplacement of the repository; in the south of Europe the more common groundwaters have low salinity, while in the north water with a high saline content of around 50 g/l is more probable. The presence of cementitious material may greatly alter the chemistry of water in the proposed repositories. These materials also provide a large reservoir of unstable Ca-silicate phases that will dissolve and reprecipitate at the rock–water or concrete–bentonite interfaces. Chemical interactions between water and concrete may well be dominated by dissolution kinetics of unstable amorphous and crystalline phases and precipitation kinetics of meta-stable or stable phases. The use of accelerated batch leaching tests, in order to qualify the different concretes to be used in deep repositories, enables data about the dissolution–precipitation processes to be acquired and to make comparisons between different cementitious materials. However, column leach tests are considered to be more representative of field leaching conditions because of continuous flux of the leaching solution through the monolithic material. There is enormous concern on determining the low-pH concrete behaviour under the long-term action of water in representative conditions of the real storage scenario. However, up to now, very little has been published on this subject. In general, results from the few leaching tests that have been made show good resistance of the low-pH cementitious materials against water aggression, although an altered front can be observed from the surface in all the tested samples. From leaching analyses using deionised water, Ca2+ released in the leaching solution has been described, as well as a decalcification process governed by diffusion, where the Ca2+ flux is not only balanced by the release of OH but also by that of sulphate ions (Yamamoto et al., 2007; Codina et al., 2008). However, decalcification of the low-pH cement pastes is much slower than that of OPC ones. Apart from a very low CaO/SiO2 ratio (0.3 to 0.4 for all pastes), a disappearance of ettringite and enrichment in a hydrotalcite-like phase have been reported near the leached surfaces of low-pH cement pastes (Codina et al., 2008). In leaching tests made on low-pH concretes using real groundwater (from the A¨spo¨ site) (Garcı´ a Calvo et al., 2009a) or simulated fresh groundwater (Yamamoto et al., 2007), leaching of Ca2+ is again clearly recognised. After the test periods, low-pH concretes show a small altered front (>100 kPa). None of the techniques can be used if the inflow of water from the rock is high since piping and erosion will make the backfill heterogeneous and fluid.

11.5

Long-term performance

11.5.1 Issues The most important issue for long-term isolation of HLW is the chemical stability of the smectite under hydrothermal conditions created in repositories. Another matter is the deformation of the buffer that it will undergo under stresses induced by the load of canisters, earthquakes and tectonics. We will consider mechanical strain first.

11.5.2 Time-dependent mechanical strain Theory The Terzaghi concept of consolidation and expansion has a quite obvious physical meaning for soils containing non-expanding clay mineral particles, like illite and kaolinite (Pusch and Yong, 2006). Thus, an increased total pressure Δp of water-saturated clay of these latter types generates a pore pressure of exactly the same magnitude, but if water is allowed to be drained the pressure Δp is successively transferred to the particle network, which is thereby compressed and densified. The same process takes place in smectiterich clay, leading to a closer distance between some of the stacks of lamellae and closer spacing of the lamellae in some stacks, by which the repulsion between the lamellae and stacks of lamellae increases. Unloading has the opposite effect, i.e. water is taken up in the interlamellar space and on the basal surfaces of the stacks of lamellae by the very strong hydration


Development and application of buffer and backfill materials

345

potential, yielding expansion. The load of a canister on the buffer hence causes compression of the buffer in conjunction with expulsion of porewater until equilibrium is reached between the imposed stresses and the mobilised stress resistance. Such equilibrium is reached relatively quickly but in a longterm perspective, considering a period of time of hundreds or thousands of years, another mechanism, i.e. shear-induced creep, is expected to dominate and cause settlement. It has the form of slip within and at the contacts of adjacent stacks of lamellae, taking place under constant volume conditions or in conjunction with compression. The accumulated time-dependent strain is termed ‘creep’. For smectite clay with insignificant friction between particles, creep is stronger than for non-expandable clays, but the microstructural strain mechanisms have the same character of stochastically distributed slip occurring where energy barriers are overcome. The empirically derived relationship in the following equation is being used in current buffer design work for describing time-dependent strain as a function of the principal stress state and stress conditions at failure (Pusch and Yong, 2006): n t g_ ¼ g_ 0 eaðs1 s3 Þ=ðs1 s3 Þfe ðaÞðs1 s3 Þ0 =ðs1 s3 Þf ½11:1 tr where t = time after stress change tr = reference time (105 s) (σ1σ3)0 = reference deviator stress (0.5 (σ1σ3)) (σ1σ3)f = deviator stress for actual condition (σ1σ3)fe = deviator stress at failure g_ = creep rate g_ 0 = creep rate at time tr n and α = parameters derived from laboratory tests The following equation expresses the shear strength q as a function of the mean effective stress p: ½11:2

q ¼ apb where a = q for p = 1 kPa b = inclination of curve in log p/log q diagrams

Equations [11.1] and [11.2] are completely empirical and not based on any conceptual microstructural model and are therefore not of general validity. For deriving reliable theoretical expressions of creep and creep rate as functions of stress and clay density the true involved strain mechanisms must be identified and taken as a basis of formulation of conceptual and


346


theoretical models (Pusch and Yong, 2006; Pusch 2008). Such work involves definition and concretisation of the barriers on the microstructural scale to slip, which are represented by van der Waals forces, hydrogen bonds and strong primary valence bonds. They form a spectrum with the weakest forces, van der Waals forces, at its lower end and strong primary valence bonds at its upper end (Pusch and Feltham, 1980). The response of the structure to a macroscopic shear stress is that the overall deformation of the entire network of particles changes by disintegration, translation and rotation of small, weaker aggregates while larger ones are less affected and stay strong. The breakdown of weak aggregates involves transformation to a laminated system of flaky particles. This microstructural organisation is believed to be the reason for the Newtonian rheological behaviour of smectite clay that has undergone a large strain in one or two directions. Using stochastic mechanics the bulk shear strain rate for low bulk shear stresses, i.e. lower than about 1/3 of conventionally determined shear strength, is found to be as defined below with t

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