Electrochemical Study of the Pitting Corrosion of Stainless Steel Candidate Overpack Materials for the Disposal of High-Level Radioactive Waste in Boom Clay

1998 ◽  
Vol 289-292 ◽  
pp. 1083-1090
Author(s):  
D. Druyts ◽  
B. Kursten ◽  
P. Van Iseghem
1992 ◽  
Vol 294 ◽  
Author(s):  
K. Osada ◽  
S. Muraoka

ABSTRACTThe corrosion behavior of type 304 stainless steel was studied under gamma irradiation as part of the evaluation for the long-term durability of high-level radioactive waste (HLW) disposal containers. Gamma rays, generated from fission products in high-level radioactive waste, are considered to change the environment around the canisters and overpacks. The redox potentials for NaCl solutions and corrosion potentials of stainless steel were measured to consider the effects of gamma irradiation, by using an electrochemical method. The pitting potentials of stainless steel for NaCl solutions were also measured to examine the pitting corrosion under gamma irradiation. As a result of this experiment, it is concluded that the oxidizing properties as a result of the formation of H2O2 and H2 produced by gamma irradiation depended on the concentration of Cl−, and that the strength of oxidizing properties of 1M (mol·dm−3) NaCl solution was particularly high. and the pitting corrosion was found for 1M NaCI solution under gamma irradiation at the dose rate of 2.6×102 C/kg·h (1.0×106 R/h) at 60°C, by using an electrochemical method.


2006 ◽  
Vol 932 ◽  
Author(s):  
Bruno Kursten ◽  
Frank Druyts

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.


Author(s):  
Pierre Van Iseghem ◽  
Jan Marivoet

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).


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