The Leaching Behavior of Bituminized Radioactive Waste in the Geologic Disposal Conditions of the Boom Clay Formation

1997 ◽  
Vol 506 ◽  
Author(s):  
A. Sneyers ◽  
P. Van Iseghem
Keyword(s):  
Radioactive Waste ◽  
Crack Formation ◽  
Soluble Salts ◽  
Leaching Behavior ◽  
Boom Clay ◽  
Leach Rate ◽  
Diffusion Controlled ◽  
Geologic Disposal ◽  
Leaching Behaviour ◽  
Clay Formation

ABSTRACTAs part of the evaluation of the safety of geologic disposal, the leaching behaviour of two types of bituminized radioactive waste (Eurobitum and CEA bitumen) has been studied as a function of temperature, pressure, leachant composition and bitumen matrix type. Inactive and active bitumen samples were brought into contact with two test media, simulating the geologic disposal environment of the Boom clay formation. At contact with these media, the samples swelled and soluble salts and radionucides were leached. It was found that the leach rate is influenced by temperature, the leachant composition, and the physical characteristics of the bitumen matrix. The release of nitrate is interpreted as a diffusion controlled process, which can however be disturbed by crack formation. The leaching of 60Co, 90Sr, and total β is diffusion controlled. Low leach rates were measured for Pu and Am: the release of Pu and Am is limited by their solubility in the leachant. Pu and Am are preferentially sorbed to the Boom clay, the test container or the bitumen. The leached mobile Pu and Am concentrations are of the order of 10−10 to 1013 M at 23°C. The results of this study suggest that the integrity of bituminized waste packages is seriously affected due to the leaching of soluble salts: a full-size 220 litre Eurobitum drum is predicted to be depleted in NaNO3 in less than 20,000 years.

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).

Static Leaching of Fully Radioactive Waste Glass at 90°C in the Presence of Bentonite, Granite, and Stainless Steel Corrosion Products

1985 ◽  
Vol 50 ◽  
Author(s):  
Hans-Peter Hermansson ◽  
Inga-Kari Björner ◽  
Hilbert Christensen ◽  
Toshiaki Ohe ◽  
Lars Werme
Keyword(s):  
Stainless Steel ◽  
Radioactive Waste ◽  
Steel Corrosion ◽  
Corrosion Products ◽  
Leach Rate ◽  
Compacted Bentonite ◽  
Infrared Reflection ◽  
Leaching Experiments ◽  
Leaching Behaviour ◽  
Glass Leaching

AbstractThe results from the JSS phase II static glass leaching experiments are presented. The experiments have been performed in the presence of bentonite, granite and stainless steel corrosion products at 90°C for 28 days. The leaching behaviour of Cs has been analysed using gamma-spectrometry. The surface of the leached glass samples has been analysed by infrared reflection spectrometry.The presence of uncompacted bentonite seems to have a very small, increasing, effect on the leach rate of the glass compared to when no bentonite is present. In the presence of compacted bentonite, a lower leach rate has beep founy than in the prgsencelof uncompacted bentonite. (For Cs 0.3 g × m−2 × d−1 and 0.7 g × m−2 × d−1 respectively.) The addition of stainless steel corrosion products seems to have a slightly increasing effect on the leach rate in the presence of compacted bentonite. The beneficial effect of granite on the leach rate seems to be overshadowed by the presence of uncompacted bentonite.

scholarly journals Preparation and Study of Corrosion Behavior of Some Simulated Radioactive Waste Glasses

2019 ◽  
Vol 16 (3) ◽  
pp. 271-279
Author(s):  
Patit Paban Malik
Keyword(s):  
Radioactive Waste ◽  
Ground Water ◽  
Corrosion Behavior ◽  
Iron Phosphate ◽  
Leaching Behavior ◽  
Time Intervals ◽  
Safe Environment ◽  
Leach Rate ◽  
Weight Losses ◽  
Leachate Solution

To safe environment from radioactive waste it is important to fix them as radioactive waste glasses. The corrosion behavior of radioactive waste glasses in water is significantly important. Radionuclides return to the biosphere by means of leaching from waste form into ground water. Finally the ground water containing the radionuclide are transported to the surface. In this study, the preparation, characterization and leaching behavior of some borosilicate (BS) and lead iron phosphate (LIP) of different chemical composition doped with simulated nuclear waste oxide were investigated. We measured the pH found to be in the range from 6.78 up to 7.79 of the leachate solution at normal temperature and at varying time intervals. Leaching study of these glasses were conducted with the help of Soxhlet distillation apparatus with distilled water upto 24 hours and for BS9 - BS12 upto 100 hours duration. Weight losses were are measured with respect to time of leaching. Leach rate of some borosilicate glass samples loaded with uranium are calculated from surface area measurements. The results are reported in the range 1.34x10-4 g.m-2.hr-1 and 6.26 x 10-4 g.m-2.hr-1 respectively at 90°C.

CORALUS: An Integrated In Situ Corrosion Test on α-Active Glass

1997 ◽  
Vol 506 ◽  
Author(s):  
Pierre Van Iseghem ◽  
Elie Valcke ◽  
Nicole Godon ◽  
Norbert Jockwer
Keyword(s):  
Storage Conditions ◽  
Current Status ◽  
Research Facility ◽  
Geologic Repository ◽  
Boom Clay ◽  
Geologic Disposal ◽  
Different Types ◽  
Backfill Materials ◽  
Clay Formation

ABSTRACTAs part of the evaluation of the safety of the geologic disposal of HLW glass, a new in situ test on α-active glass will be started up in the underground research facility HADES (SCK•CEN, Mol, Belgium), situated in the Boom Clay formation, at a depth of 223 m. Besides a demonstration of the safe disposal of HLW glass containers in Boom Clay formation, the overall objective of this CORALUS test (CORrosion of Active gLass in Underground Storage conditions) is to study the performance of active HLW glass specimens in direct contact with different types of backfill materials, under conditions as representative as possible for those expected to prevail in a geologic repository in the Boom Clay formation. In order to meet this objective, the CORALUS test will integrate not only the experience of previous in situ tests performed at SCK•CEN and other research institutes, but also the expertise of the cooperating institutes. The paper presents the motivation for this new in situ test, its objectives, and the current status of its design

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.

Processes Related to the Water Uptake by EUROBITUM Bituminised Radioactive Waste: Theoretical Considerations and First Experimental Results

2008 ◽  
Vol 1107 ◽  
Author(s):  
An Mariën ◽  
Steven Smets ◽  
Xiangling Li ◽  
Elie Valcke
Keyword(s):  
Radioactive Waste ◽  
Water Uptake ◽  
Swelling Pressure ◽  
Management Program ◽  
Constant Volume ◽  
Constant Stress ◽  
Stress Tests ◽  
Permeable Membrane ◽  
Boom Clay ◽  
Clay Formation

AbstractAccording to the present Belgian radioactive waste management program, Eurobitum bituminised radioactive waste will be disposed of in a geologically stable underground clay formation. The Boom Clay is studied as a potential host formation because of its low diffusion and high retention properties towards radionuclides. The presence of the radioactive waste should not disturb these properties. Due to the presence of hygroscopic salts (25 to 30 weight% NaNO3), Eurobitum will take up pore water which will result in a swelling and possibly in a very high swelling pressure. First scoping calculations suggest that the swelling pressure exerted to Boom Clay should remain below 7 to 8 MPa to avoid the formation of fractures. If the bitumen in EUROBITUM behaved like a perfect semi-permeable membrane and if no swelling were allowed after the dissolution of NaNO3 into a saturated solution of 10.8 M, osmotic pressures of ∼50 MPa could be attained. To better understand the interaction between the swelling Eurobitum and the host formation, coupled hydro-chemical-mechanical constitutive laws for Eurobitum have to be developed. To this purpose, water uptake tests under constant volume (‘confined’) and constant stress (‘semi-confined’) conditions are being performed. After ∼2 years of hydration of small inactive Eurobitum samples in constant volume conditions, the swelling pressure has raised to ∼12 MPa. The volume of samples that can swell against counter pressures of 2.2, 3.3, or 4.4 MPa (constant stress tests) increased with ∼5 to 11 volume%, independently of the applied counter pressure. Approximately 10 weight% of the initial NaNO3 content has been leached.

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.

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.

Twenty-five years' geotechnical observation and testing in the Tertiary Boom Clay formation

Géotechnique ◽  
2007 ◽  
Vol 57 (2) ◽  
pp. 229-237 ◽  
Author(s):  
F. Bernier ◽  
X. L. Li ◽  
W. Bastiaens
Keyword(s):  
Boom Clay ◽  
Clay Formation
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