The methodology followed in Belgium to investigate the compatibility with geological disposal of Eurobitum bituminized intermediate-level radioactive waste

2009 ◽  
Vol 1193 ◽  
Author(s):  
Elie Valcke ◽  
An Marien ◽  
Maarten Van Geet
Keyword(s):  
Radioactive Waste ◽  
Organic Molecules ◽  
Nuclear Research ◽  
Intermediate Level ◽  
Research Centre ◽  
Geological Disposal ◽  
Boom Clay ◽  
Multiple Processes ◽  
Intermediate Level Radioactive Waste ◽  
Clay Formation

AbstractIn Belgium, Eurobitum intermediate-level long-lived bituminized radioactive waste containing large amounts of NaNO3, which is a hygroscopic and soluble salt, is to be disposed of in an underground repository in a geologically stable clay formation. The Boom Clay is studied as a potential host formation because of its favourable properties to limit and delay the migration of the leached radionuclides and other contaminants (heavy metals, NaNO3, organic molecules) to the biosphere. The emplacement of the bituminized waste will induce multiple processes that could have a significant effect on the key properties of the clay. Because several of these processes are interdependent, the study of the compatibility of Eurobitum with geological disposal is complex. To structure the research and to identify possible knowledge gaps, the Belgian Radioactive Waste Management Agency ONDRAF/NIRAS developed a new methodology based on safety functions and safety statements. In this paper, this methodology is briefly explained, with reference to the disposal of Eurobitum. Experimental results obtained at the Belgian Nuclear Research Centre SCK•CEN are presented and discussed in the light of the safety functions and safety statements approach. The importance of the interdependence of the processes is highlighted. Special attention is given to the evolution of the disposal design as a result of the improved understanding of key processes.

The Belgian Approach Towards the Study of the Compatibility of Eurobitum With the Geological Disposal Environment

2007 ◽  
Author(s):  
Elie Valcke ◽  
Robert Gens
Keyword(s):  
Radioactive Waste ◽  
Waste Management ◽  
Water Uptake ◽  
Nuclear Research ◽  
Swelling Pressure ◽  
Research Centre ◽  
Soluble Salts ◽  
Geological Disposal ◽  
Management Agency ◽  
Clay Formation

In Belgium, EUROBITUM bituminized radioactive waste containing large amount of soluble salts (NaNO3) is to be disposed of in a final repository in a clay formation. Since the emplacement of the waste will induce many interdependent processes that could negatively affect the interesting radionuclide retarding properties of the clay, the study of the compatibility of EUROBITUM is very complex. To better structure the research and to identify possible knowledge gaps, NIRAS/ONDRAF, the Belgian Radioactive Waste Management Agency, developed the safety functions and safety statements approach. In this paper, we present the application of this approach for the case of EUROBITUM. The approach is illustrated with new and old results on water uptake, swelling, swelling pressure build-up, and ageing, obtained from tests performed in the laboratories of SCK•CEN, the Belgian Nuclear Research Centre.

Safety Assessment of Geological Disposal of High-Level Radioactive Waste in Boom Clay: Relation With the Radionuclide Inventory

2009 ◽  
Author(s):  
Pierre Van Iseghem ◽  
Jan Marivoet
Keyword(s):  
Radioactive Waste ◽  
Safety Assessment ◽  
High Level Waste ◽  
Spent Fuel ◽  
Geological Disposal ◽  
High Level Radioactive Waste ◽  
Boom Clay ◽  
High Level ◽  
Clay Formation ◽  
Parameter Values

This paper discusses the impact of the parameter values used for the transport of radionuclides from high-level radioactive waste to the far-field on the long-term safety of a proposed geological disposal in the Boom Clay formation in Belgium. The methodology of the Safety Assessment is explained, and the results of the Safety Assessment for vitrified high-level waste and spent fuel are presented. The radionuclides having the strongest impact on the dose-to-man for both HLW glass and spent fuel are 79Se, 129I, 126Sn, 36Cl, and 99Tc. Some of them are volatile during the vitrification process, other radionuclides are activation products, and for many of them there is no accurate information on their inventory in the waste form. The hypotheses in the selection of the main parameter values are further discussed, together with the status of the R&D on one of the main dose contributing radionuclides (79Se).

scholarly journals Geology, seismic activity and groundwater conditions at six potential disposal sites for radioactive waste from Risø, Denmark

1969 ◽  
Vol 28 ◽  
pp. 13-16
Author(s):  
Peter Gravesen ◽  
Bertel Nilsson ◽  
Merete Binderup ◽  
Tine B. Larsen ◽  
Stig A. Schack Pedersen
Keyword(s):  
Radioactive Waste ◽  
Data Base ◽  
Decision Process ◽  
Seismic Activity ◽  
Nuclear Research ◽  
Intermediate Level ◽  
Disposal Site ◽  
Research Facility ◽  
Intermediate Level Radioactive Waste ◽  
Selection Of

In 2003 it was decided by the Danish Parliament that lowand intermediate-level radioactive waste from the Danish nuclear research facility, Risø, is to be stored at a permanent disposal site (Ministeriet for Sundhed og Forebyggelse 2008; Gravesen et al. 2012a). Both the issue of storage and the selection of potential sites have caused considerable public debate. In this paper we report on the most recent geological investigations intended to further improve the data base for the selection and decision process, although no conclusions have been drawn at this stage.

An Integrated Safety Assessment System for Near-Surface Disposal of Low- and Intermediate-Level Radioactive Waste in Korea

2003 ◽  
Author(s):  
JooWan Park ◽  
Chang-Lak Kim ◽  
Jin Beak Park ◽  
Eun Yong Lee ◽  
Youn Myoung Lee ◽  
...  
Keyword(s):  
Radioactive Waste ◽  
Safety Assessment ◽  
Assessment System ◽  
Intermediate Level ◽  
Radioactive Waste Disposal ◽  
System Level ◽  
Assessment Methodology ◽  
Regulatory Requirements ◽  
Near Surface ◽  
Intermediate Level Radioactive Waste

An integrated safety assessment system to be used for evaluation of near-surface disposal concept has been developed within the framework of safety assessment methodology taken for low- and intermediate-level radioactive waste disposal in Korea. It is to provide an evaluation of the safety of the disposal system in a clear, comprehensive and well-documented manner, and to integrate the results into a defensible package showing reasonable assurance of compliance with regulatory requirements for licensing application. This system is made up of two key components, a system-level safety assessment code and an input database/quality assurance module for safety assessment, which both are interfaced with each other.

Methodology in Corrosion Testing of Container Materials for Radioactive Waste Disposal in a Geological Clay Repository

2001 ◽  
Author(s):  
Bruno Kursten ◽  
Frank Druyts ◽  
Pierre Van Iseghem
Keyword(s):  
Radioactive Waste ◽  
Corrosion Behaviour ◽  
Open Circuit ◽  
Geological Disposal ◽  
Boom Clay ◽  
Engineered Barriers ◽  
High Level ◽  
Aerobic And Anaerobic

Abstract The current worldwide trend for the final disposal of conditioned high-level, medium-level and long-lived alpha-bearing radioactive waste focuses on deep geological disposal. During the geological disposal, the isolation between the radioactive waste and the environment (biosphere) is realised by the multibarrier principle, which is based on the complementary nature of the various natural and engineered barriers. One of the main engineered barriers is the metallic container (overpack) that encloses the conditioned waste. In Belgium, the Boom Clay sediment is being studied as a potential host rock formation for the final disposal of conditioned high-level radioactive waste (HLW) and spent fuel. Since the mid 1980’s, SCK•CEN has developed an extensive research programme aimed at evaluating the suitability of a wide variety of metallic materials as candidate overpack material for the disposal of HLW. A multiple experimental approach is applied consisting of i) in situ corrosion experiments, ii) electrochemical experiments (cyclic potentiodynamic polarisation measurements and monitoring the evolution of ECORR as a function of time), and iii) immersion experiments. The in situ corrosion experiments were performed in the underground research facility, the High Activity Disposal Experimental Site, or HADES, located in the Boom clay layer at a depth of 225 metres below ground level. These experiments aimed at predicting the long-term corrosion behaviour of various candidate container materials. It was believed that this could be realised by investigating the medium-term interactions between the container materials and the host formation. These experiments resulted in a change of reasoning at the national authorities concerning the choice of over-pack material from the corrosion-allowance material carbon steel towards corrosion-resistant materials such as stainless steels. The main arguments being the severe pitting corrosion during the aerobic period and the large amount of hydrogen gas generated during the subsequent anaerobic period. The in situ corrosion experiments however, did not allow to unequivocally quantify the corrosion of the various investigated candidate overpack materials. The main shortcoming was that they did not allow to experimentally separate the aerobic and anaerobic phase. This resulted in the elaboration of a new laboratory programme. Electrochemical corrosion experiments were designed to investigate the effect of a wide variety of parameters on the localised corrosion behaviour of candidate overpack materials: temperature, SO42−, Cl−, S2O32−, oxygen content (aerobic - anaerobic),… Three characteristic potentials can be derived from the cyclic potentiodynamic polarisation (CPP) curves: i) the open circuit potential, OCP, ii) the critical potential for pit nucleation, ENP, and iii) the protection potential, EPP. Monitoring the open circuit potential as a function of time in clay slurries, representative for the underground environment, provides us with a more reliable value for the corrosion potential, ECORR, under disposal conditions. The long-term corrosion behaviour of the candidate overpack materials can be established by comparing the value of ECORR relative to ENP and EPP (determined from the CPP-curves). The immersion tests were developed to complement the in situ experiments. These experiments aimed at determining the corrosion rate and to identify the corrosion processes that can occur during the aerobic and anaerobic period of the geological disposal. Also, some experiments were elaborated to study the effect of graphite on the corrosion behaviour of the candidate overpack materials.

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