scholarly journals Hydro-chemical modelling of in situ behaviour of bituminized radioactive waste in Boom Clay

2014 ◽  
Vol 400 (1) ◽  
pp. 117-134 ◽  
Author(s):  
N. Mokni ◽  
S. Olivella ◽  
E. Valcke ◽  
N. Bleyen ◽  
S. Smets ◽  
...  
Keyword(s):  
Radioactive Waste ◽  
Boom Clay ◽  
Chemical Modelling

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.

In Situ Migration Experiments in the Boom Clay at Mol ; Experimental Method and Preliminary Results

1988 ◽  
Vol 127 ◽  
Author(s):  
M. Put ◽  
M. Monsecour ◽  
A. Fonteyne ◽  
H. Yoshida ◽  
P. De Regge
Keyword(s):  
Radioactive Waste ◽  
First Generation ◽  
Far Field ◽  
Radioactive Tracers ◽  
Experimental Conditions ◽  
Boom Clay ◽  
Long Term Experiments ◽  
Clay Formation

ABSTRACTA first generation of underground migration experiments is described, consisting of labelled clay cores emplaced in boreholes drilled in the Boom clay formation. The boreholes are sealed by natural convergence of the clay and porewater percolates through the labelled clay cores which are consolidated in situ. After monitoring of the radioactive tracers in the percolating porewater, the experiment is retrieved from the borehole and the tracer profile is measured in the clay cores. With the exception of accelerated porewater flow, due to the existence of a high hydraulic head around the underground gallery, the experimental conditions are close to those expected in the far-field of a closed repository for radioactive waste. The main advantage of this approach is the availability of real porewater during relatively long-term experiments. Results are reported for the experiments performed with europium and strontium tracers.

In-Situ Interaction between Cement and Clay: Implications for Geological Disposal

2000 ◽  
Vol 663 ◽  
Author(s):  
A. Sneyers ◽  
M. Paul ◽  
M. Tyrer ◽  
F.P. Glasser ◽  
J. Fays ◽  
...  
Keyword(s):  
Radioactive Waste ◽  
Cementitious Materials ◽  
Buffering Capacity ◽  
Magnesium Aluminate ◽  
In Situ Tests ◽  
X Ray ◽  
Boom Clay ◽  
Materials Used ◽  
Clay Formation

ABSTRACTThe extent and the consequences of interactions between cementitious materials used in radioactive waste management and clay host rock are described. In-situ tests were performed on seven cement formulations representing materials applied in repository construction, for backfilling or for solidification of radioactive waste. Samples were exposed to realistic repository conditions of the Boom Clay Formation in the HADES underground laboratory. Chemical, physical and mineralogical changes across the cement-clay interface were identified by combined observations from Electron Probe Microanalysis, Infrared microscopy and X-Ray powder diffraction. Significant interactions in both the cement and the clay part were found in a zone extending up to several hundreds of microns. The most prominent features are (1) leaching of cement with loss of calcium and/or silicon; (2) development of a calcium-rich zone in Boom Clay close to or at contact; (3) the formation of a contact zone marked by the precipitation of a (hydrated) magnesium aluminate phase; (4) reduction in apparent porosity of initially porous/permeable materials and (5) precipitation of calcite within the cement. This elemental exchange tends to diminish pH and reduce the buffering capacity of the cement. Although hydroxide will diffuse into the clay, the development of an extensive alkaline halo in the surrounding clay is unlikely owing to the buffering capacity of the Boom Clay pore water.

New Technologies for Decontamination of Radioactive Substances Scattered by Nuclear Accident / Nowe Technologie Dekontaminacji Radioaktywnych Substancji Rozproszonych Przez Awarie Nuklearna

2013 ◽  
Vol 58 (1) ◽  
pp. 283-290 ◽  
Author(s):  
Y. Nishizaki ◽  
H. Miyamae ◽  
S. Ichikawa ◽  
K. Izumiya ◽  
T. Takano ◽  
...  
Keyword(s):  
Radioactive Waste ◽  
New Technologies ◽  
Ferric Hydroxide ◽  
Nuclear Accident ◽  
Radioactive Cesium ◽  
Cesium Ions ◽  
Naoh Solution ◽  
High Concentrations ◽  
Water Rinsing

Our effort for decontamination of radioactive cesium scattered widely by nuclear accident in March 2011 in Fukushima, Japan has been described. Radioactive cesium scattered widely in Japan has been accumulating in arc or plasma molten-solidified ash in waste incinerating facilities up to 90,000 Bq/kg of the radioactive waste. Water rinsing of the ash resulted in dissolution of cesium ions together with high concentrations of potassium and sodium ions. Although potassium inhibits the adsorption of cesium on zeolite, we succeeded to precipitate cesium by in-situ formation of ferric ferrocyanide and iron rust in the radioactive filtrate after rinsing of the radioactive ash with water. Because the regulation of no preservation of any kind of cyanide substances, cesium was separated from the precipitate consisting of cesium-captured ferric ferrocyanide and ferric hydroxide in diluted NaOH solution and subsequent filtration gave rise to the potassium-free radioactive filtrate. Cesium was captured by zeolite from the potassium-free radioactive filtrate. The amount of this final radioactive waste of zeolite was significantly lower than that of the arc-molten-solidified ash.

scholarly journals In situ grouting of a low-level radioactive waste trench

1987 ◽  
Author(s):  
R. D. Spence ◽  
T. T. Godsey ◽  
E. W. McDaniel
Keyword(s):  
Radioactive Waste ◽  
Low Level ◽  
Low Level Radioactive Waste

The Belgian Demonstration Programme for the Disposal of High-Level and Long-Lived Radioactive Waste

2006 ◽  
Vol 932 ◽  
Author(s):  
Bernier Frédéric ◽  
Demarche Marc ◽  
Bel Johan
Keyword(s):  
Radioactive Waste ◽  
Thermal Load ◽  
Large Scale ◽  
Mechanical Response ◽  
Boom Clay ◽  
Geological Formation ◽  
Important Input ◽  
Clay Layers ◽  
High Level ◽  
Large Scale Tests

ABSTRACTThe EIG EURIDICE is responsible for performing large-scale tests, technical demonstrations and experiments so as to assess the feasibility of a final disposal of vitrified radioactive waste in deep clay layers. This programme is part of the Belgian Research and Development programme managed by ONDRAF/NIRAS. The research infrastructure includes the Underground Research Facilities HADES (URF HADES) in the Boom Clay geological formation and surface facilities. The achievements of the demonstration programme are the demonstration of the construction of shafts and galleries at industrial scale, the characterisation of the hydro-mechanical response of the host rock, and the “OPHELIE mock-up” a large scale hydration test under thermal load of pre-fabricated bentonite blocks. The future works will consist mainly in the realisation of the “PRACLAY experiments” including a large scale heater test. The large scale heater test has to demonstrate that Boom Clay is suitable, in terms of performance of the disposal system, to undergo the thermal load induced by the vitrified waste. The combined effect of the excavation and the thermal load will be investigated. A long term (more than 10 years) large scale heater test would be representative of the most penalizing conditions that could be encountered in the real disposal. The results of this test will constitute an important input for the Safety and Feasibility Cases 1 (SFC-1, 2013) and 2 (SFC-2, 2020).

Sign in / Sign up

Export Citation Format

Share Document