- •Preface
- •Acronyms
- •Introduction
- •Background and objectives
- •Content, format and presentation
- •Radioactive waste management in context
- •Waste sources and classification
- •Introduction
- •Radioactive waste
- •Waste classification
- •Origins of radioactive waste
- •Nuclear fuel cycle
- •Mining
- •Fuel production
- •Reactor operation
- •Reprocessing
- •Reactor decommissioning
- •Medicine, industry and research
- •Medicine
- •Industry
- •Research
- •Military wastes
- •Conditioning of radioactive wastes
- •Treatment
- •Compaction
- •Incineration
- •Conditioning
- •Cementation
- •Bituminisation
- •Resin
- •Vitrification
- •Spent fuel
- •Process qualification/product quality
- •Volumes of waste
- •Inventories
- •Inventory types
- •Types of data recorded
- •Radiological data
- •Chemical data
- •Physical data
- •Secondary data
- •Radionuclides occurring in the nuclear fuel cycle
- •Simplifying the number of waste types
- •Radionuclide inventory priorities
- •Material priorities
- •Inventory evolution
- •Assumptions
- •Errors
- •Uncertainties
- •Conclusions
- •Acknowledgements
- •References
- •Development of geological disposal concepts
- •Introduction
- •Historical evolution of geological disposal concepts
- •Geological disposal
- •Definitions and comparison with near-surface disposal
- •Development of geological disposal concepts
- •Roles of the geosphere in disposal options
- •Physical stability
- •Hydrogeology
- •Geochemistry
- •Overview
- •Alternatives to geological disposal
- •Introduction
- •Politically blocked options: sub-seabed and Antarctic icecap disposal
- •Sea dumping and sub-seabed disposal
- •Antarctic icesheet disposal
- •Technically impractical options; partitioning and transmutation, space disposal and icesheet disposal
- •Partitioning and Transmutation
- •Space disposal
- •Icesheets and permafrost
- •Non-options; long-term surface storage
- •Alternatives to conventional repositories
- •Introduction
- •Alternative geological disposal concepts
- •Utilising existing underground facilities
- •Extended storage options (CARE)
- •Injection into deep aquifers and caverns
- •Deep boreholes
- •Rock melting
- •The international option: technical aspects
- •Alternative concepts: fitting the management option to future boundary conditions
- •Conclusions
- •References
- •Site selection and characterisation
- •Introduction
- •Prescriptive/geologically led
- •Sophisticated/advocacy led
- •Pragmatic/technically led
- •Centralised/geologically led
- •Conclusions to be drawn
- •Lessons to be learned (see Table 4.2)
- •Site characterisation
- •Can we define the natural environment sufficiently thoroughly?
- •Sedimentary environments
- •Hydrogeology
- •The regional hydrogeological model
- •More local hydrogeological model(s)
- •Crystalline rock environments
- •Lithology and structure
- •Hydrogeology
- •Hydrogeochemistry
- •Any geological environment
- •References
- •Repository design
- •Introduction: general framework of the design process
- •Identification of design requirements/constraints
- •Concept development
- •Major components of the disposal system and safety functions
- •A structured approach for concept development
- •Detailed design/specifications of subsystems
- •Near-field processes and design issues
- •Design approach and methodologies
- •Design confirmation and demonstration
- •Interaction with PA/SA
- •Demonstration and QA
- •Repository management
- •Future perspectives
- •References
- •Assessment of the safety and performance of a radioactive waste repository
- •Introduction
- •The role of SA and the safety case in decision-making
- •SA tasks
- •System description
- •Identification of scenarios and cases for analysis
- •Consequence analysis
- •Timescales for evaluation
- •Constructing and presenting a safety case
- •References
- •Repository implementation
- •Legal and regulatory framework; organisational structures
- •Waste management strategies
- •The need for a clear policy and strategy
- •Timetables vary widely
- •Activities in development of a geological repository
- •Concept development
- •Siting
- •Repository design
- •Licensing
- •Construction
- •Operation
- •Monitoring
- •Research and development
- •The staging process
- •Attributes of adaptive staging
- •The decision-making process
- •Status of geological disposal programmes
- •Overview
- •Status of geological disposal projects in selected countries
- •International repositories
- •Costs and financing
- •Cost estimates
- •Financing
- •Conclusions
- •Acknowledgements
- •References
- •Research and development infrastructure
- •Introduction: Management of research and development
- •Drivers for research and development
- •Organisation of R&D
- •R&D in specialised (nuclear) facilities
- •Introduction
- •Inventory
- •Release of radionuclides from waste forms
- •Solubility and sorption
- •Waste form dissolution
- •Colloids
- •Organic degradation products
- •Gas generation
- •Conventional R&D
- •Engineered barriers
- •Corrosion
- •Buffer and backfill materials
- •Container fabrication
- •Natural barriers
- •Geochemistry and groundwater flow
- •Gas transport and two-phase flow
- •Biosphere
- •Radionuclide concentration and dispersion in the biosphere
- •Climate change
- •Landscape change
- •Underground rock laboratories
- •URLs in sediments
- •Nature’s laboratories: studies of the natural environment
- •General
- •Corrosion
- •Cement
- •Clay materials
- •Degradation of organic materials
- •Glass corrosion
- •Radionuclide migration
- •Model and database development
- •Conclusions
- •References
- •Building confidence in the safe disposal of radioactive waste
- •Growing nuclear concerns
- •Communication systems in waste management programmes
- •The Swiss programme
- •The Japanese programme
- •Examples of communication styles in other countries
- •Finland
- •Sweden
- •France
- •United Kingdom
- •Comparisons between communication styles in Finland, France, Sweden and the United Kingdom
- •Lessons for the future
- •What is the way forward?
- •Acknowledgements
- •References
- •A look to the future
- •Introduction
- •Current trends in repository programmes
- •Priorities for future efforts
- •Waste characterisation
- •Operational safety
- •Emplacement technologies
- •Knowledge management
- •Alternative designs and optimisation processes
- •Materials technology
- •Novel construction/immobilisation materials: the example of low pH cement
- •Future SA code development
- •Implications for environmental protection: disposal of other wastes
- •Conclusions
- •References
- •Index
Waste sources and classification |
29 |
2.7.2.1. Radiological data
Individual radionuclide and total activities A summary of some of the methods used to determine radionuclide inventories is given in Table 2.5.
This information is required not only as input for a repository safety assessment, but is also used for dose rate, heat output and radiotoxicity calculations.
The number of radionuclides presented in a radionuclide inventory is programmespecific, e.g., in the Swiss programme, for model inventories all radionuclides of halflife greater than 60 days plus important short-lived daughters are initially considered (i.e., a total of 147 radionuclides). Following a simplified safety analysis (screening exercise), this list is reduced to those radionuclides that are safety-relevant and only for this inventory is the full safety assessment study performed.
As an example, the nuclides selected after the simplified safety assessment, those selected after the analysis of the near-field and, finally, those that dominated releases from the far-field in the case of the studies performed for a potential Swiss L/ILW repository (Nagra, 1993) are given in Table 2.6.
Heat output Heat outputs are determined using WBq 1 conversion factors based on the determined radionuclide inventory. For L/ILW, heat output is not normally of great interest (apart from ensuring that it is not too high for ILW-LL). However, for HLW wastes, this parameter determines the size of the repository, in that disposal canister heat output limits are required to ensure that the engineered barrier system (EBS) functions as required. These limits are a compromise between the spacing between waste packages and the dimensions and spacing of the tunnels (see also chapters 3 and 5).
Dose rates If no production data exist, these can be calculated with varying degrees of sophistication. A simple, user-friendly code is Microshield (Grove, 1998) but, like all tools, the limitations must be understood if it is to be used for more than simple modelling. If more accurate values are required, and especially if complicated
Table 2.5
Overview of methods used to determine radionuclide inventories
Waste category |
Radionuclide inventory determination |
|
|
Reactor operational |
Correlation factors (measurement of samples of raw waste to obtain a radionuclide |
wastes |
spectrum. As the relationship between the nuclides is relatively stable, this |
|
spectrum can then be correlated to a few key measured nuclides or dose rates). |
|
Reactor modelling. In Belgium, e.g., reactor parameters (cooling water chemistry, |
|
etc.) have been exhaustively correlated to the wastes produced and hence |
|
inventories can be determine in this manner. |
Reactor core |
Modified fuel depletion code validated by a few measurements. |
components |
|
Reactor bioshield |
Neutron transport code supplemented by a few measurements. |
Reprocessing waste |
Fuel depletion codes, information from the reprocesser (BNFL, Cogema). |
Spent fuel |
Fuel depletion codes. |
Accelerator wastes |
Spallation, medium energy reaction codes supplemented by a few measurements. |
Medicine, industry and |
Production declarations, government statistics concerning import and/or production |
research |
of sources. |
|
|
30 |
D.F. McGinnes |
Table 2.6
Summary of nuclides in the initial (reference) inventories and at subsequent stages of the safety assessment for a L/ILW repository in Switzerland.
Nuclides |
Nuclides |
Nuclides |
Nuclides |
Nuclides |
Nuclides |
remaining |
selected after |
dominating |
remaining |
selected after |
dominating |
after initial |
the analysis |
releases from |
after initial |
the analysis |
releases from |
pre-screening |
of near-field |
the far-field |
pre-screening |
of near-field |
the far-field |
exercise |
releases |
|
exercise |
releases |
|
|
|
|
|
|
|
H-3 |
|
|
Be-10 |
|
|
C-14 |
C-14 |
C-14 |
Na-22 |
|
|
Cl-36 |
Cl-36 |
Cl-36 |
Ar-42 |
|
|
K-40 |
K-40 |
K-40 |
Ca-41 |
|
|
Fe-55 |
|
|
Fe-60 |
|
|
Co-60 |
|
|
Ni-59 |
Ni-59 |
|
Ni-63 |
Ni-63 |
|
Se-79 |
Se-79 |
Se-79 |
Sr-90 |
Sr-90 |
|
Zr-93 |
|
|
Nb-94 |
|
|
Mo-93 |
Mo-93 |
Mo-93 |
Tc-99 |
|
|
Ru-106 |
|
|
Pd-107 |
|
|
Ag-108M |
Ag-108M |
|
Sn-126 |
|
|
I-129 |
I-129 |
|
Cs-134 |
|
|
Cs-135 |
|
|
Cs-137 |
Cs-137 |
|
Sm-151 |
|
|
Eu-152 |
|
|
Ho-166M |
|
|
Pb-210 |
|
|
Po-210 |
|
|
Ra-226 |
Ra-226 |
Ra-226 |
Ra-228 |
Ra-228 |
|
Ac-227 |
Ac-227 |
|
Th-228 |
Th-228 |
|
Th-229 |
Th-229 |
Th-229 |
Th-230 |
Th-230 |
|
Th-232 |
Th-232 |
|
Pa-231 |
Pa-231 |
Pa-231 |
U-232 |
U-232 |
|
U-233 |
U-233 |
U-233 |
U-234 |
U-234 |
|
U-235 |
U-235 |
|
U-236 |
U-236 |
|
U-238 |
U-238 |
|
Np-237 |
Np-237 |
Np-237 |
Pu-238 |
Pu-238 |
|
Pu-239 |
Pu-239 |
Pu-239 |
Pu-240 |
Pu-240 |
Pu-240 |
Pu-241 |
Pu-241 |
|
Pu-242 |
Pu-242 |
|
Am-241 |
Am-241 |
Am-241 |
Am-242M |
Am-242M |
|
Am-243 |
Am-243 |
|
Cm-243 |
Cm-243 |
|
Cm-244 |
Cm-244 |
|
Cm-245 |
Cm-245 |
|
Cm-246 |
Cm-246 |
|
geometries are involved, more sophisticated codes such as RANKERN (AEA, 2003) or MCNP (LANL, 2005) should be considered.
Surface contamination If no production data exist, these are normally set at the handling limits that are approved by the competent regulator.
Radiolytic gas production This can be measured but, more often, is based on conservative calculations of the theoretical gas production rate using so-called G-values. These G-values are experimentally determined for different types of materials, e.g., concrete, PVC, paper, rubber, etc., and describe the amount of gas produced from the deposition of a discrete amount of radiation in this material. Generally, these values are determined for - and -radiation, with having proportionally higher values due to the