First performance assessment of the disposal of spent fuel in a clay layer

Directorate-General For Research And Innovation European Commission

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Nuclear energy and safety

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• Publié par: DIRECTORATE-GENERAL-FOR-RESEARCH-AND-INNOVATION-EUROPEAN-COMMISSION
• Nombre de lectures: 15
• EAN13: 928276124
• Langue: English
• Poids de l'ouvrage: 4 Mo

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ISSN 1018-5593
* *
European Commission
III
First performance assessment of the
disposal of spent fuel in a clay layer European Commission
nuclear science
and technology
First performance assessment of the
disposal of spent fuel in a clay layer
J. Marivoet, G. Volckaert, A. Snyers, J. Wibin
CEN/SCK
B-2400 Mol
Contract No FI2W-CT90-0016 with the EC
and
Contract No CCHO-89-094 with NIRAS/ONDRAF
Final report
Work performed as part of the European Atomic Energy Community's
shared-cost programme (1990-94) on 'Management and
storage of radioactive waste'
Task 5: Methods of evaluating the safety of disposal systems
Directorate-General
Science, Research and Development
1996 EUR 16752 EN LEGAL NOTICE
Neither the European Commission nor any person acting on
behalf of the Commission is responsible for the use which might be made of the
following information
Cataloguing data can be found at the end of this publication
Luxembourg: Office for Official Publications of the European Communities, 1996
ISBN 92-827-6124-X
© ECSC-EC-EAEC, Brussels · Luxembourg, 1996
Printed in Luxembourg ABSTRACT
Hitherto the performance assessments of geological disposal in clay layers considered essentially
waste forms resulting from the reprocessing of the spent fuel. The low prices of fresh uranium
and the use of mixed oxides (MOX) fuel make that the direct disposal of spent fuel is becoming
a realistic option for the nuclear fuel cycle.
The inventories of the various spent fuel types, uranium oxide fuel as well as MOX fuel, that can
be expected to arise from the Belgian nuclear programme are estimated. Available repository
concepts for the disposal of reprocessing waste are adapted to allow the placement of the large
spent fuel containers. Little information is available about the behaviour of spent fuel in the
conditions prevailing in a repository in clay and about the long-term behaviour of spent MOX
fuels in general.
Many models that have been developed for earlier assessments are reapplied. However the
modelling of the essential release processes requires the development of a specific near field
model which takes into account the different corrosion rates of the various components of the
disposed spent fuel assemblies and the radiolytic oxidation of the uranium oxide matrix. An
examination of the methodological aspects indicates that the calculations of this first assessment
of spent fuel disposal in clay should focus on the analysis of the normal evolution scenario.
Detenninistic as well as stochastic calculations are performed.
An essential conclusion of the study is that the calculated maximum dose rate is about one order
of magnitude higher than the one estimated for the case of high-level waste disposal. On the other
hand the maximum dose is still one order of magnitude lower than the one estimated for the case
of the disposal of the iodine captured at the reprocessing plant. Finally recommendations are
formulated for R&D programmes as wells as for more advanced performance assessments. FOREWORD
The authors like to thank the scientific officers of the European Commission Dr. N Cadelli and
Dr. H. von Maravic and of NIRAS/ONDRAF ir. J. Van Miegroet and Dr. P. De Prêter for the
fruitful discussions of the draft versions of the various chapters of this report.
We also thank our colleagues at SCK'CEN Charles De Raedt and Martin Put for making
contributions to this report by calculating respectively the radionuclide inventories and the
temperature distributions. We also have to mention our colleagues from the performance
assessment team of S CK· CEN Isabelle Wemaere and Theo Zeevaert for their continuous
contributions to respectively the aquifer and biosphere modelling in our assessments. Finally we
thank Frans Sleegers for the preparation of many figures and Monique Van Geel for carefully
typing the text of this report.
V Table of contents
ABSTRACT
III
FOREWORD V
CHAPTER 1 INTRODUCTION 1
1.1 Background
1.2 The objectives of the study
1.3 Lay-out of the report 2
References 3
CHAPTER 2 WASTE INVENTORIES 5
2.1 Introduction
2.2 Assumptions about the nuclear power production and fuel
2.3 Characteristics of the spent fuel assemblies 6
2.4 Reference time
2.5 Radionuclide inventories of the spent fuels
2.6 Total radionuclide inventories 22
2.7 Conclusions3
References
CHAPTER 3 DISPOSAL SCENARIOS AND REPOSITORY CONCEPT 31
3.1 Introduction 3
3.2 Disposal times
3.3 Scenarios for the conditioning of the spent fuel 31
3.4 Description of the waste package3
3.5 Preliminary repository concept
3.6 Thermal calculations
3.7 Repository concept5
3.7 Conclusions
References6
CHAPTER 4 METHODOLOGICAL ASPECTS 47
4.1 Introduction 4
4.2 Summary of the scenario selection applied for the EVEREST project 4
4.3 Reexamination of the FEP catalogue9
4.3.1 FEP's due to natural phenomena 50
4.3.2 FEP's due to human activities
4.3.3 FEP's due to waste and repository effects
4.4 Adaptations of the scenarios3
4.5. Scenarios to be analyzed in the present study4
4.6 Description of the normal evolution scenario
VII 4.7 Organization of consequence analyses 55
4.8 Conclusions 5
References7
CHAPTER BEHAVIOUR OF SPENT FUEL IN REPOSITORY CONDITIONS 61
5.1. Introduction 6
5.2. Characteristics of spent fuel
5.2.1. Composition and radionuclide inventory of spent fuel 61
5.2.2. Chemical form of radionuclides in spent fuel2
5.3. Processes that influence the release of radionuclides from spent fuel
5.3.1. The release of radionuclides from the spent fuel cladding
5.3.2. Thee ofs from the gap between the fuel matrix and the
cladding3
5.3.3. The release of radionuclides from irradiated fuel in contact with fluids 64
5.3.3.1. The release of radionuclides from the fuel matrix 6
5.3.3.2. The dissolution ofs from grain boundaries in spent
67 fuel
5.3.3.3. The chemical form of uranium in repository fluids8
5.4. The assessment of performance and the safety of spent fuel in geological disposal
conditions 6
5.4.1. The release of radionuclides from the cladding 6
5.4.2. Thee ofs from the gap between the fuel matrix and the
cladding9
5.4.3. The release of radionuclides from the grain boundary
5.4.4. Thee ofs from the spent fuel matrix
5.4.4.1. The mass transport limited dissolution of the spent fuel matrix 70
5.4.4.2. The reaction rate limitedn of the spent fuel matrix 71
5.5. Conclusions 73
References
CHAPTER 6 MODELS, CODES AND DATA 81
6.1 Introduction
6.2 The near field and clay modelling
6.2.1 The near field model
6.2.1.1 Release model for the gap
6.2.1.2el for the cladding2
6.2.1.3 Release model for the structural parts 8
6.2.1.4el for the grain boundaries
6.2.1.5 Release model for the uranium oxide grains
6.2.2 The clay model 83
6.2.2.1 Phenomena
6.2.2.2 Model4
6.2.3 The NUCDSPF code
6.2.3.1 Introduction
6.2.3.2 The NUCDIS code5
6.2.3.3 Adaptations to NUCDIS 8
VIM