Assessing the durability of nuclear glass with respect to silica controlling processes in a clayey underground disposal

2006 ◽  
Vol 932 ◽  
Author(s):  
Laurent De Windt ◽  
Stéphanie Leclercq ◽  
Jan van der Lee
Keyword(s):  
Host Rock ◽  
Reactive Transport ◽  
Transport Model ◽  
Glass Block ◽  
High Level Waste ◽  
Residual Rate ◽  
Glass Dissolution ◽  
Underground Disposal ◽  
High Level

ABSTRACTThe long-term behaviour of vitrified high-level waste in an underground clay repository was assessed by using the reactive transport model HYTEC with respect to silica diffusion, sorption and precipitation processes. Special attention was given to the chemical interactions between glass, corroded steel and the host-rock considering realistic time scale and repository design. A kinetic and congruent dissolution law of R7T7 nuclear glass was used assuming a first-order dissolution rate, which is chemistry dependent, as well as a long-term residual rate. Without silica sorption and precipitation, glass dissolution is diffusion-driven and the fraction of altered glass after 100,000 years ranges from 5% to 50% depending on the fracturation degree of the glass block. Corrosion products may limit glass dissolution by controlling silica diffusion, whereas silica sorption on such products has almost no effect on glass durability. Within the clayey host-rock, precipitation of silicate minerals such as chalcedony may affect glass durability much more significantly than sorption. In that case, however, a concomitant porosity drop is predicted that could progressively reduce silica diffusion and subsequent glass alteration.

Long-Term Radioactivity Release from Solidified High-Level Waste - Part II Parametric Study of Waste form Properties, Temperature and Time

1982 ◽  
Vol 15 ◽  
Author(s):  
Friedrich K. Altenhein ◽  
Werner Lutze ◽  
Rodney C. Ewing
Keyword(s):  
Parametric Study ◽  
Computer Code ◽  
Salt Dome ◽  
Waste Form ◽  
Glass Block ◽  
High Level Waste ◽  
Surface Layers ◽  
History Of ◽  
High Level

The computer code QTERM has been used to calculate the total released activity from a single glass block when in contact with brine in a salt dome repository as a function of: (1) waste form properties, (2) leaching mechanisms, (3) retention or precipitation of specific radionuclides in surface layers, (4) thermal history of the repository and (5) decreasing activity as a function of time.

Geological Repository for Nuclear High Level Waste in France From Feasibility to Design Within a Legal Framework

2007 ◽  
Author(s):  
Patrice Voizard ◽  
Stefan Mayer ◽  
Gerald Ouzounian
Keyword(s):  
Host Rock ◽  
Legal Framework ◽  
High Level Waste ◽  
National Review ◽  
Review Team ◽  
Geological Repository ◽  
Underground Disposal ◽  
High Level ◽  
Underground Research Laboratory ◽  
Future Work

Over the past 15 years, the French program on deep geologic disposal of high level and long-lived radioactive waste has benefited from a clear legal framework as the result of the December 30, 1991 French Waste Act. To fulfil its obligations stipulated in this law, Andra has submitted the “Dossier 2005 Argile” (clay) and “Dossier 2005 Granite” to the French Government. The first of those reports presents a concept for the underground disposal of nuclear waste at a specific clay site and focuses on a feasibility study. Knowledge of the host rock characteristics is based on the investigations carried out at the Meuse/Haute Marne Underground Research Laboratory. The repository concept addresses various issues, the most important of which relates to the large amount of waste, the clay host rock and the reversibility requirement. This phase has ended upon review and evaluation of the “Dossier 2005” made by different organisations including the National Review Board, the National Safety Authority and the NEA International Review Team. By passing the “new”, June 28, 2006 Planning Act on the sustainable management of radioactive materials and waste, the French parliament has further defined a clear legal framework for future work. This June 28 Planning Act thus sets a schedule and defines the objectives for the next phase of repository design in requesting the submission of a construction authorization application by 2015. The law calls for the repository program to be in a position to commission disposal installations by 2025.

Glass Degradation in Performance Assessment Models1

2015 ◽  
Vol 1744 ◽  
pp. 163-172 ◽  
Author(s):  
William L. Ebert
Keyword(s):  
Performance Assessment ◽  
Reactive Transport ◽  
Intrinsic Property ◽  
Secondary Phase ◽  
Simple Expression ◽  
Glass Waste ◽  
Residual Rate ◽  
Glass Dissolution ◽  
Kinetics Of

ABSTRACTThe interface with reactive transport models used in performance assessment calculations is described to identify aspects of the glass waste form degradation model important to long-term predictions. These are primarily the conditions that trigger the change from the residual rate to the Stage 3 rate and the values of those rates. Although the processes triggering the change and controlling the Stage 3 rate are not yet understood mechanistically, neither appears related to an intrinsic property of the glass. The sudden and usually significant increase in the glass dissolution rate suggests the processes that trigger the increase are different than the processes controlling glass dissolution prior to that change. Application of a simple expression that was derived for mineral transformation to represent the kinetics of coupled glass dissolution and secondary phase precipitation reactions is shown to be consistent with experimental observations of Stage 3 and useful for modeling long-term glass dissolution in a complex disposal environment.

Chemical durability of high-level waste glass in repository environment: main conclusions and remaining uncertainties from the GLASTAB and GLAMOR projects

2006 ◽  
Vol 932 ◽  
Author(s):  
P. Van Iseghem ◽  
K. Lemmens ◽  
M. Aertsens ◽  
S. Gin ◽  
I. Ribet ◽  
...  
Keyword(s):  
Near Field ◽  
Chemical Durability ◽  
High Level Waste ◽  
Model Parameters ◽  
Residual Rate ◽  
Geological Disposal ◽  
Glass Dissolution ◽  
Geologic Disposal ◽  
High Level

ABSTRACTThis paper reviews the main conclusions of two European Commission funded projects, GLASTAB and GLAMOR, on the durability of HLW glass in geological disposal, and the remaining uncertainties. The progress in GLASTAB relates to the characterization of the alteration layer, the modeling of glass dissolution, the interaction of glass with near field materials, the behaviour of radionuclides, and the calculation of HLW glass dissolution as part of geological disposal. The GLAMOR programme was focused on the decrease of the glass dissolution rate upon achieving silica saturation in solution. Two conclusions from this programme are the importance of the residual rate measured beyond silica saturation and the need to reduce the uncertainties in model parameters. The overall conclusion from the two projects is that strong progress has been achieved on basic dissolution mechanisms and interaction with disposal environments - HLW glass will act as a good barrier to the release of radionuclides in geologic disposal.

scholarly journals Mineralogical, numerical and analytical studies of the coupled oxidation of pyrite and coal

2003 ◽  
Vol 67 (2) ◽  
pp. 381-398 ◽  
Author(s):  
K. A. Evans ◽  
C. J. Gandy ◽  
S. A. Banwart
Keyword(s):  
Reactive Transport ◽  
Transport Model ◽  
Pyrite Oxidation ◽  
Spoil Heap ◽  
Reaction Fronts ◽  
Order Of Magnitude ◽  
Magnitude Variation ◽  
Discharge Chemistry ◽  
Analytical Expressions

Mineralogical, bulk and field leachate compositions are used to identify important processes governing the evolution of discharges from a coal spoil heap in County Durham. These processes are incorporated into a numerical one-dimensional advective-kinetic reactive transport model which reproduces field results, including gas compositions, to within an order of magnitude. Variation of input parameters allows the effects of incorrect initial assumptions on elemental profiles and discharge chemistry to be assessed. Analytical expressions for widths and speeds of kinetic reaction fronts are developed and used to predict long-term development of mineralogical distribution within the heap. Results are consistent with observations from the field site. Pyrite oxidation is expected to dominate O2 consumption in spoil heaps on the decadal timescale, although C oxidation may stabilize contaminants in effluents on the centennial scale.

scholarly journals Reactive transport calculations to evaluate sulphide fluxes in the near-field of a SF/HLW repository

2019 ◽  
Vol 98 ◽  
pp. 10005
Author(s):  
Marek Pękala ◽  
Paul Wersin ◽  
Veerle Cloet ◽  
Nikitas Diomidis
Keyword(s):  
Reactive Transport ◽  
Spent Nuclear Fuel ◽  
Near Field ◽  
Opalinus Clay ◽  
High Level Waste ◽  
Waste Containment ◽  
Geological Repository ◽  
High Level ◽  
Reducing Bacteria ◽  
Hlw Repository

Radioactive waste is planned to be disposed in a deep geological repository in the Opalinus Clay (OPA) rock formation in Switzerland. Cu coating of the steel disposal canister is considered as potential a measure to ensure complete waste containment of spent nuclear fuel (SF) and vitrified high-level waste (HLW) or a period of 100,000 years. Sulphide is a potential corroding agent to Cu under reducing redox conditions. Background dissolved sulphide concentrations in pristine OPA are low, likely controlled by equilibrium with pyrite. At such concentrations, sulphide-assisted corrosion of Cu would be negligible. However, the possibility exists that sulphate reducing bacteria (SRB) might thrive at discrete locations of the repository’s near-field. The activity of SRB might then lead to significantly higher dissolved sulphide concentrations. The objective of this work is to employ reactive transport calculations to evaluate sulphide fluxes in the near-field of the SF/HLW repository in the OPA. Cu canister corrosion due to sulphide fluxes is also simplistically evaluated.

scholarly journals The initial dissolution rates of simulated UK Magnox–ThORP blend nuclear waste glass as a function of pH, temperature and waste loading

2015 ◽  
Vol 79 (6) ◽  
pp. 1529-1542 ◽  
Author(s):  
N. Cassingham ◽  
C.L. Corkhill ◽  
D.J. Backhouse ◽  
R.J. Hand ◽  
J.V. Ryan ◽  
...  
Keyword(s):  
Dissolution Kinetics ◽  
Intrinsic Rate ◽  
Waste Glass ◽  
High Level Waste ◽  
Dissolution Mechanism ◽  
Dissolution Rates ◽  
Forward Rates ◽  
Glass Dissolution ◽  
High Level ◽  
Kinetics Of

AbstractThe first comprehensive assessment of the dissolution kinetics of simulant Magnox–ThORP blended UK high-level waste glass, obtained by performing a range of single-pass flow-through experiments, is reported here. Inherent forward rates of glass dissolution were determined over a temperature range of 23 to 70°C and an alkaline pH range of 8.0 to 12.0. Linear regression techniques were applied to the TST kinetic rate law to obtain fundamental parameters necessary to model the dissolution kinetics of UK high-level waste glass (the activation energy (Ea), pH power law coefficient (η) and the intrinsic rate constant (k0)), which is of importance to the post-closure safety case for the geological disposal of vitreous products. The activation energies based on B release ranged from 55 ± 3 to 83 ± 9 kJ mol–1, indicating that Magnox–THORP blend glass dissolution has a surface-controlled mechanism, similar to that of other high-level waste simulant glass compositions such as the French SON68 and LAW in the US. Forward dissolution rates, based on Si, B and Na release, suggested that the dissolution mechanism under dilute conditions, and pH and temperature ranges of this study, was not sensitive to composition as defined by HLW-incorporation rate.

scholarly journals Effect of Fe2O3 on the Immobilization of High-Level Waste with Magnesium Potassium Phosphate Ceramic

2019 ◽  
Vol 2019 ◽  
pp. 1-10
Author(s):  
Hailin Yang ◽  
Mingjiao Fu ◽  
Bobo Wu ◽  
Ying Zhang ◽  
Ruhua Ma ◽  
...  
Keyword(s):  
Sintering Temperature ◽  
Potassium Phosphate ◽  
High Level Waste ◽  
Radioactive Nuclide ◽  
Hydration Reaction ◽  
Sintering Process ◽  
Positive Role ◽  
Ceramic Structure ◽  
High Level

For the proposed novel procedure of immobilizing HLW with magnesium potassium phosphate cement (MKPC), Fe2O3 was added as a modifying agent to verify its effect on the solidification form and the immobilization of the radioactive nuclide. The results show that Fe2O3 is inert during the hydration reaction. It slows down the hydration reaction and lowers the heat release rate of the MKPC system, leading to a 3°C-5°C drop in the mixture temperature during hydration. Early comprehensive strength of Fe2O3 containing samples decreased slightly while the long-term strength remained unchanged. For the sintering process, Fe2O3 played a positive role, lowering the melting point and aiding the formation of ceramic structure. CsFe(PO4)2, or CsFePO4, was generated by sintering at 900°C. These products together with the ceramic structure and absorption benefit the immobilization of Cs+. The optimal sintering temperature for heat treatment is 900°C; it makes the solidification form a fired ceramic-like structure.

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