IAEA's Coordinated Research Programmes on Performance of High Level Wastes in Deep Geological Repositories

2006 ◽  
Vol 932 ◽  
Author(s):  
J.L. Gonzalez ◽  
P. Van Iseghem
Keyword(s):  
Waste Form ◽  
Spent Fuel ◽  
Significant Progress ◽  
Energy Agency ◽  
Basic Understanding ◽  
Research Programmes ◽  
Deep Geological Repositories ◽  
And Performance ◽  
High Level ◽  
Form Development

ABSTRACTThe main conclusions of a Coordinated Research Programmes organized by the International Atomic Energy Agency on the performance of high-level wastes in deep geological repositories are summarized in this paper. The programme ran from 1997 till 2004. Glass, spent fuel and ceramics were the waste forms considered. The 13 participating countries reported their R&D including waste form development, basic understanding of the waste form properties and the performance in simulated disposal conditions. Significant progress has been achieved on the various issues. Recommendations are formulated as to integrate the R&D with the geological disposal conditions and performance assessment considerations, or to conceive generic studies in preparation of the above approach.

scholarly journals Spent Fuel as a Waste Form - Data Needs to Allow Long Term Performance Assessment Under Repository Disposal Conditions.

1986 ◽  
Vol 84 ◽  
Author(s):  
V. M. Oversby
Keyword(s):  
Performance Assessment ◽  
Waste Form ◽  
High Level Waste ◽  
Detailed Knowledge ◽  
Spent Fuel ◽  
Expected Performance ◽  
High Level ◽  
Term Performance ◽  
Physical And Chemical ◽  
Waste Repositories

AbstractPerformance assessment calculations are required for high level waste repositories for a period of 10,000 years under NRC and EPA regulations. In addition, the Siting Guidelines (IOCFR960) require a comparison of sites following site characterization and prior to final site selection to be made over a 100,000 year period. In order to perform the required calculations, a detailed knowledge of the physical and chemical processes that affect waste form performance will be needed for each site. While bounding calculations might be sufficient to show compliance with the requirements of IOCFR60 and 40CFRI91, the site comparison for 100,000 years will need to be based on expected performance under site specific conditions. The only case where detailed knowledge of waste form characteristics in the repository would not be needed would be where radionuclide travel times to the accessible environment can be shown to exceed 100,000 years. This paper will review the factors that affect the release of radionuclides from spemt fuel under repository conditions, summarize our present state of knowledge, and suggest areas where more work is needed in order to support the performance assessment calculations.

The Strategy of the Belgian Nuclear Research Centre in the Area of High-Level Waste Form Compatibility Research

2007 ◽  
Author(s):  
Karel Lemmens ◽  
Christelle Cachoir ◽  
Elie Valcke ◽  
Karine Ferrand ◽  
Marc Aertsens ◽  
...  
Keyword(s):  
Nuclear Research ◽  
Waste Form ◽  
Research Centre ◽  
High Level Waste ◽  
Spent Fuel ◽  
National Agency ◽  
Waste Forms ◽  
In Situ Experiments ◽  
High Level

The Belgian Nuclear Research Centre (SCK•CEN) has a long-standing expertise in research concerning the compatibility of waste forms with the final disposal environment. For high level waste, most attention goes to two waste forms that are relevant for Belgium, namely (1) vitrified waste from the reprocessing of spent fuel, and (2) spent fuel as such, referring to the direct disposal scenario. The expertise lies especially in the study of the chemical interactions between the waste forms and the disposal environment. This is done by laboratory experiments, supported by modeling. The experiments vary from traditional leach tests, to more specific tests for the determination of particular parameters, and highly realistic experiments. This results in a description of the phenomena that are expected upon disposal of the waste forms, and in quantitative data that allow a conservative long-term prediction of the in situ life time of the waste form. The predictions are validated by in situ experiments in the underground research laboratory HADES. The final objective of these studies, is to estimate the contribution of the waste form to the overall safety of the disposal system, as part of the Safety and Feasibility Case, planned by the national agency ONDRAF/NIRAS. The recent change of the Belgian disposal concept from an engineered barrier system based on the use of bentonite clay to a system based on a concrete buffer has caused a reorientation of the research programme. The expertise in the area of clay-waste interaction will however be maintained, to develop experimental methodologies in collaboration with other countries, and as a potential support to the decision making in those countries where a clay based near field is still the reference. The paper explains the current R&D approach, and highlights some recent experimental set-ups available at SCK•CEN for this purpose, with some illustrating results.

Development of Crystallizer for Advanced Aqueous Reprocessing Process

2006 ◽  
Author(s):  
Tadahiro Washiya ◽  
Toshiaki Kikuchi ◽  
Atsuhiro Shibata ◽  
Takahiro Chikazawa ◽  
Shunji Homma
Keyword(s):  
Experimental Studies ◽  
Small Scale ◽  
Test Results ◽  
Spent Fuel ◽  
Design Study ◽  
Energy Agency ◽  
Scale Test ◽  
Crystallization Conditions ◽  
And Performance ◽  
Fuel Reprocessing

Crystallization is one of the remarkable technologies for future fuel reprocessing process that has safety and economical advantages. Japan Atomic Energy Agency (JAEA) (former Japan Nuclear Cycle Development Institute), Mitsubishi Material Corporation and Saitama University have been developing the crystallization process. In previous study, we carried out experimental studies with uranium, MOX and spent fuel conditions, and flowsheet analysis was considered. [1, 2, 3] In association with these studies, an innovative continuous crystallizer and its system was developed to ensure high process performance. From the design study, an annular type continuous crystallizer was selected as the most promising design, and performance was confirmed by small-scale test and engineering scale demonstration at uranium crystallization conditions. In this paper, the design study and the demonstration test results are described.

Chemical durability and performance of assessment of spent fuel and high level waste forms under simulated repository conditions – Spanish contribution to IAEA project

2006 ◽  
Vol 932 ◽  
Author(s):  
Aurora Martínez-Esparza ◽  
José Antonio Gago ◽  
Javier Quiñones ◽  
Eduardo Iglesias ◽  
Esther Cera ◽  
...  
Keyword(s):  
Experimental Studies ◽  
Great Influence ◽  
Storage Conditions ◽  
Chemical Durability ◽  
High Level Waste ◽  
Spent Fuel ◽  
Geologic Storage ◽  
International Projects ◽  
And Performance ◽  
High Level

ABSTRACTChemical durability of spent fuel under repository conditions is one of the main topics of interest in national and international projects from the last two decades. During the last decade there have been growing the activities in Enresa related to the deep disposal concept with the aim of developing a spent fuel alteration model for the understanding of the behaviour of this nuclear waste under repository conditions. In this context, the development and utilisation of models and sub-models based on experimental work have been of great importance.Experimental studies with spent fuel(in Collaboration Agreement with ITU) and spent fuel analogues in several environmental conditions have been carried out into Enresa R+D Programmes in order to reach a better knowledge of the relevant processes and to quantify the spent fuel chemical durability under repository conditions.In this work, it is showed the utility of data provided from experiments with spent fuel analogues to test the mechanisms and the influence of relevant parameters in the spent fuel alteration under repository conditions.The evolution of irradiated fuel under interim storage conditions and in deep geologic storage and its oxygen to metal ratio (O/M) before the water access to the fuel is another factor of great influence on enhanced spent fuel leaching. This effect has been also studied by means of spent fuel analogues and by using simulated (artificially) aged fuel.

scholarly journals Solubility Limits. on Radionuclide Dissolution

1984 ◽  
Vol 44 ◽  
Author(s):  
Jerry F. Kerrisk
Keyword(s):  
Yucca Mountain ◽  
Waste Form ◽  
High Level Waste ◽  
Spent Fuel ◽  
Dissolution Rates ◽  
Diffusion Limited ◽  
Dissolution Model ◽  
High Level

AbstractThis paper examines the effects of solubility in limiting dissolution rates of a number of important radionuclides from spent fuel and high-level waste. Two simple dissolution models were used for calculations that would be characteristic of a Yucca Mountain repository. A saturation-limited dissolution model, in which the water flowing through the repository is assumed to be saturated with each waste element, is very conservative in that it overestimates dissolution rates. A diffusion-limited dissolution model, in which element-dissolution rates are limited by diffusion of waste elements into water flowing past the waste, is more realistic, but it is subject to some uncertainty at this time. Dissolution rates of some elements (Pu, Am, Sn, Th, Zr, Sm) are always limited by solubility. Dissolution rates of other elements (Cs, Tc, Np, Sr, C, I) are never solubility limited; their release would be limited by dissolution of the bulk waste form. Still other elements (U, Cm, Ni, Ra) show solubility-limited dissolution under some conditions.

Testing Safety and Performance Indicators in an Assessment of the Long-Term Performance of the Geological Disposal of Spent Fuel in Boom Clay

2003 ◽  
Author(s):  
Xavier Sillen ◽  
Jan Marivoet ◽  
Wim Cool ◽  
Peter de Preter
Keyword(s):  
Dose Rate ◽  
Performance Indicators ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Boom Clay ◽  
And Performance ◽  
The Individual ◽  
High Level ◽  
Clay Formation

The classical numerical output, or indicator, from assessments of the long-term safety of geological disposal systems for high-level radioactive waste is the individual effective dose rate. This indicator is an estimate of the possible individual health detriment and it is commonly compared to regulatory limits for assessing the safety of other nuclear activities as well, such as medical and industrial activities. As a safety indicator, the individual dose rate provides an estimate of the overall safety of the disposal system. However, because of the time frames involved in safety assessments of geological disposal systems, the need arises of complementary safety indicators that could be less affected by uncertainties like those associated with future human behaviour or the effects of climate change on the biosphere and the aquifers. Such alternative safety indicators can be, for example, radionuclide concentrations in the groundwater or fluxes to the biosphere due to a repository. Safety indicators only tell how globally safe a disposal system is. For confidence building, performance indicators can be used in addition to tell how the system works. In particular, performance indicators such as fluxes, activities or activity concentrations of selected radionuclides can show how the different components of the system fulfil their safety functions and contribute to the overall safety. The SPIN project of the European Commission assessed the usefulness of seven safety indicators and fourteen performance indicators by testing them in four actual assessments of disposal systems in granite formations. In this paper, indicators calculated from an assessment of the disposal of spent fuel in the poorly indurated Boom Clay formation are presented. Conclusions from the SPIN project that hold for repositories in clays are highlighted, as well as results that illustrate differences between the granite and clay disposal options. Finally, various performance and safety indicators are combined into a logical sequence to comprehensively present, and explain, the results of a safety assessment.

Development of the Supercontainer Design for Deep Geological Disposal of High-Level Heat Emitting Radioactive Waste in Belgium

2006 ◽  
Vol 932 ◽  
Author(s):  
Johan J.P. Bel ◽  
Stephen M. Wickham ◽  
Robert M.F. Gens
Keyword(s):  
Selection Process ◽  
Metal Corrosion ◽  
Preliminary Evaluation ◽  
Spent Fuel ◽  
Deep Geological Disposal ◽  
Reference Design ◽  
And Performance ◽  
High Level ◽  
Design Options ◽  
Future Work

ABSTRACTONDRAF-NIRAS has recently selected a Supercontainer with a Portland Cement (PC) buffer as the preferred new reference design for disposal of HLW and spent fuel. The selection process involved a multi-criteria analysis of alternative design options, which were evaluated against a range of long-term safety and feasibility criteria. A PC concrete has been chosen for the buffer because this will provide a highly alkaline chemical environment, which will last for thousands of years. In this environment the external surface of the overpack will be passivated and overpack corrosion will be inhibited. The concrete buffer also has low-hydraulic conductivity to slow the infiltration of external fluids to the overpack surface, and provides radiological shielding.ONDRAF-NIRAS has made a preliminary evaluation of the viability of the reference Supercontainer design. The following areas were reviewed and investigated: radiolysis, thermo-hydraulic (TH) behaviour of the concrete buffer, metal corrosion, the chemical and mineralogical evolution of the concrete buffer, and relevant industrial experience. This paper describes the main findings, and identifies remaining design and performance uncertainties. Prioritisation and recommendations for future work are also given.

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