Static Leaching of Radioactive Glass Under Conditions Simulating a Granitic Repository for High-Level Waste: Phase 1.

1983 ◽  
Vol 26 ◽  
Author(s):  
Hans-Peter Hermansson ◽  
Hilbert Christensen ◽  
David E Clark ◽  
Inga-Kari Björner ◽  
Hayaichi Yokoyama ◽  
...  
Keyword(s):  
Phase 1 ◽  
High Level Waste ◽  
Borosilicate Glasses ◽  
Joint Research ◽  
Weight Losses ◽  
Infrared Reflection ◽  
Joint Research Project ◽  
Actinide Oxides ◽  
Leaching Experiments ◽  
High Level

ABSTRACTA joint research project with participation from Japan, Switzerland and Sweden is underway at Studsvik (The JSS-project). The project concerns investigations on the leaching of fully radioactive glass (containing 12 wt% fission product oxides and actinide oxides) manufactured by CEA/Marcoule.So far the glass has been leached in doubly distilled water and in silicate water at 90°C. Some leaching experiments involved the presence of crushed Stripa granite in the same containers as the glass.Due to strong radiation and the presence of plutonium the leaching was carried out in a specially designed lead cave using gilded stainless steel containers.Weight losses, pH and elemental mass losses were determined together with infrared reflection spectral changes. These data are compared to those obtained from a simulated nonradioactive glass of nearly the same composition and to similar alkali borosilicate glasses previously investigated at Studsvik.

The Long-Term Corrosion and Modelling of Two Simulated Belgian Reference High-Level Waste Glasses

1987 ◽  
Vol 112 ◽  
Author(s):  
P. Van Iseghem ◽  
B. Grambow
Keyword(s):  
Corrosion Behaviour ◽  
Stable Condition ◽  
High Level Waste ◽  
Forward Rate ◽  
Borosilicate Glasses ◽  
Al Content ◽  
Rate Of Dissolution ◽  
High Level ◽  
High Al Content

AbstractThe corrosion behaviour in distilled water of two simulated candidate high level waste borosilicate glasses (SAN602519L3C2 and SM58LW11) his been investigated at 90°C for different SA/V condition's 10, 100, 7800 m−1). The experimental data were modelled using the PHREEQE and GLASSOL computer codes. The model is quite successful for describing the corrosion behaviour, using experimentally derived values for the forward rate, silica saturation and the final rate. GWss SAN60 is more stable than glass SM58 at SA/V values of 10 and 100 m−1, but in the long term the relative performance is inverse. Indeed, the high Al content of SAN60 induces the creation of analcime crystals after SiO2 has reached its saturation concentration in solution, which cause an enhancement of the final rate of dissolution of the glass; for SM58 on the contrary the SiO2 solution is a stable condition.

Predicting Pu Concentrations in Solutions Contacting Geologic Materials

1981 ◽  
Vol 6 ◽  
Author(s):  
Richard G. Strickert ◽  
Dhanpat Rai
Keyword(s):  
Nuclear Waste ◽  
Safety Assessment ◽  
High Level Waste ◽  
Stable Phase ◽  
Borosilicate Glasses ◽  
Geologic Materials ◽  
Solid Phases ◽  
High Level ◽  
Waste Repositories

ABSTRACTKnowledge of Pu solid phases present in nuclear wastes is important for predicting the geochemical behavior of Pu. Thermodynamic data and experimental measurements using discrete Pu compounds, Pu-doped borosilicate glasses (simulating a high-level waste form), and Pu contaminated sediments suggest that PuO2(c) is very stable and is expected to be present in the repository. The solubility of the stable phase, such as PuO2(c), can be used to predict the maximum Pu concentration in solutions for long-term safety assessment of nuclear waste repositories.

Effect of Alpha Decay on Nuclear Borosilicate Glass Properties

2004 ◽  
Vol 824 ◽  
Author(s):  
Sylvain Peuget ◽  
Christophe Jégou ◽  
Véronique Broudic ◽  
Danièle Roudil ◽  
Xavier Deschanels ◽  
...  
Keyword(s):  
Alpha Decay ◽  
High Level Waste ◽  
Minor Actinide ◽  
Borosilicate Glasses ◽  
Leaching Behavior ◽  
Doped Glasses ◽  
High Burnup ◽  
High Level ◽  
Doped Glass ◽  
Glass Properties

AbstractThe properties of actinide glasses are studied in the context of high-level waste management programs. Reprocessing high burnup fuels in particular will increase the minor actinide content in the glass package, resulting in higher cumulative alpha decay doses in the glass, and raising the question of the glass matrix behavior and especially its containment properties. The effect of alpha self-irradiation on the glass behavior is evaluated by doping the glass with a short-lived actinide (244Cm) to reach in several years the alpha dose received by the future glass packages over several thousand years. “R7T7” borosilicate glasses were doped with 3 different curium contents (0.04 wt%, 0.4 wt% and 1.2 wt% 244CmO2). The glass homogeneity was characterized by SEM and XRD. Chemical analysis after dissolution of the glass confirmed the chemical composition. The curium content and its distribution in the glass specimens were checked by calorimetry measurements and gamma spectrometry on 243Cm and 245Cm isotopes. The density and mechanical properties of the curium-doped glasses were characterized up to 2 × 1018 α/g, revealing only a slight evolution of the macroscopic behavior of R7T7 glass in this range. The leaching behavior of curium-doped glass was also studied by Soxhlet tests. The results do not show any significant evolution of the initial alteration rate with the alpha dose.

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