Thermohydraulic Analysis of Gas Generation in a Disposal Facility for Vitrified High-Level Radioactive Waste in Boom Clay

2011 ◽  
Vol 174 (3) ◽  
pp. 401-410 ◽  
Author(s):  
Janez Perko ◽  
Eef Weetjens
Keyword(s):  
Radioactive Waste ◽  
Gas Generation ◽  
High Level Radioactive Waste ◽  
Boom Clay ◽  
Disposal Facility ◽  
High Level

A Review of Corrosion Considerations of Container Materials Relevant to Underground Disposal of High-Level Radioactive Waste in Belgium

2006 ◽  
Vol 932 ◽  
Author(s):  
Bruno Kursten ◽  
Frank Druyts
Keyword(s):  
Radioactive Waste ◽  
Large Scale ◽  
General Corrosion ◽  
Spent Fuel ◽  
High Level Radioactive Waste ◽  
Boom Clay ◽  
Nuclear Site ◽  
In Situ Experiments ◽  
Underground Disposal ◽  
High Level

ABSTRACTThe underground formation that is currently being considered in Belgium for the permanent disposal of high-level radioactive waste and spent fuel is a 30-million-year-old argillaceous sediment (Boom Clay layer). This layer is located in the northeast of Belgium and extending under the Mol-Dessel nuclear site at a depth between 180 and 280 meter.Within the concept for geological disposal (multibarrier system), the metallic container is the primary engineered barrier. Its main goal is to contain the radioactive waste and to prevent the groundwater from coming into contact with the wasteform by acting as a tight barrier. The corrosion resistance of container materials is an important aspect in ensuring the tightness of the metallic container and therefore plays an important role in the safe disposal of HLW. The metallic container has to provide a high integrity, i.e. no through-the-wall corrosion should occur, at least for the duration of the thermal phase (500 years for vitrified HLW and 2000 years for spent fuel).An extensive corrosion evaluation programme, sponsored by the national authorities and the European Commission, was started in Belgium in the mid 1980's. The main objective was to evaluate the long-term corrosion performance of a broad range of candidate container materials. In addition, the influence of several parameters, such as temperature, oxygen content, groundwater composition (chloride, sulphate and thiosulphate), γ-radiation, … were investigated. The experimental approach consisted of in situ experiments (performed in the underground research facility, HADES), electrochemical experiments, immersion experiments and large scale demonstration tests (OPHELIE, PRACLAY). Degradation modes considered included general corrosion, localised corrosion (pitting) and stress corrosion cracking.This paper gives an overview of the more relevant experimental results, gathered over the past 25 years, of the Belgian programme in the field of container corrosion.

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).

Swelling characteristics of bentonites in artificial seawater

2009 ◽  
Vol 46 (2) ◽  
pp. 177-189 ◽  
Author(s):  
Hideo Komine ◽  
Kazuya Yasuhara ◽  
Satoshi Murakami
Keyword(s):  
Radioactive Waste ◽  
Swelling Pressure ◽  
Radioactive Waste Disposal ◽  
Exchangeable Cation ◽  
Deformation Characteristics ◽  
High Level Radioactive Waste ◽  
Disposal Facility ◽  
Buffer Material ◽  
Swelling Characteristics ◽  
High Level

Bentonite is currently designated for use as a buffer material for the repository of high-level radioactive waste because such a material requires swelling characteristics to seal the waste. A high-level radioactive waste disposal facility may be built in a coastal area of Japan because transportation of this waste by ships is feasible. Therefore, it is important to investigate the effects of seawater on a bentonite-based buffer. This study is intended to investigate the influence of seawater on the swelling pressure and swelling deformation characteristics of five common types of bentonite. The experiments described herein clarify the relations between the influence of seawater grade and compaction density, type of exchangeable cation, montmorillonite content of the bentonite, and vertical pressure condition. Based on experimental results, suitable specifications have been defined for a bentonite-based buffer that can withstand the effects of seawater.

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