scholarly journals Nature of C-(A)-S-H Phases Formed in the Reaction Bentonite/Portlandite

2014 ◽  
Vol 2014 ◽  
pp. 1-8 ◽  
Author(s):  
Raúl Fernández ◽  
Lorena González ◽  
Ana Isabel Ruiz ◽  
Jaime Cuevas
Keyword(s):  
Spent Fuel ◽  
Hydrothermal Conditions ◽  
Reaction Conditions ◽  
Cement Pastes ◽  
Molar Ratios ◽  
Deep Geological Disposal ◽  
Calcium Silicates ◽  
Engineered Barriers ◽  
High Level ◽  
Representative Material

Reactions between bentonite/montmorillonite and portlandite have been studied in the context of the engineered barriers of a purpose built repository for the deep geological disposal of spent fuel and high-level radioactive wastes. Portlandite was selected as the representative material of cement leaching in the early alkaline stage expected in a repository when conventional Ordinary Portland Cement (OPC) is used. Eight different batch experiments were performed for a reaction time near to two months, including bentonite or montmorillonite at montmorillonite/portlandite molar ratios of 2 : 1 and 3 : 1 under hydrothermal conditions. Temperatures of reactions were maintained constant at either 60 or 120°C. Calcium silicates hydrates with limited substitution of Al for Si (C-(A)-S-H phases with Al/Si <0.3) were formed with different structures and compositions as a function of the reaction conditions. Orthorhombic 11 Å-tobermorite-type phase was detected in experiments at 120°C while a more disordered monoclinic tobermorite formed at 60°C. These results are useful for the interpretation of experimental data in more complex experiments using concrete or cement pastes and bentonite, where C-(A)-S-H phases of variable compositions can precipitate, in addition to the characteristic cement hydrates and other secondary minerals carbonates.

scholarly journals Formation of C-A-S-H phases from the interaction between concrete or cement and bentonite

Clay Minerals ◽  
2016 ◽  
Vol 51 (2) ◽  
pp. 223-235 ◽  
Author(s):  
Raúl Fernández ◽  
Ana Isabel Ruiz ◽  
Jaime Cuevas
Keyword(s):  
Calcium Silicate Hydrate ◽  
Laboratory Experiments ◽  
Pore Scale ◽  
Lime Mortar ◽  
Cement Interface ◽  
Argillaceous Rocks ◽  
Deep Geological Disposal ◽  
Engineered Barriers ◽  
High Level ◽  
Increasing Temperature

AbstractConcrete and bentonite are being considered as engineered barriers for the deep geological disposal of high-level radioactive waste in argillaceous rocks. Three hydrothermal laboratory experiments of different scalable complexity were performed to improve our knowledge of the formation of calcium aluminate silicate hydrates (C-A-S-H) at the interface between the two materials: concretebentonite transport columns, lime mortar-bentonite transport columns and a portlandite- (bentonite and montmorillonite) batch experiment. Precipitation of C-A-S-H was observed in all experiments. Acicular and fibrous morphologies with certain laminar characteristics were observed which had smaller Ca/Si and larger Al/Si ratios with increasing temperature and lack of accessory minerals. The compositional fields of these C-A-S-H phases formed in the experiments are consistent with Al/(Si+Al) ratios of 0.2– 0.3 described in the literature. The most representative calcium silicate hydrate (C-S-H) phase from the montmorillonite–cement interface is Al-tobermorite. Structural analyses revealed a potential intercalation or association of montmorillonite and C-A-S-H phases at the pore scale.

Key Phenomena Governing HLW Glass Behavior in the French Deep Geological Disposal

2015 ◽  
Vol 1744 ◽  
pp. 127-138
Author(s):  
Stéphan Schumacher ◽  
Christelle Martin ◽  
Yannick Linard ◽  
Frédéric Angeli ◽  
Delphine Neff ◽  
...  
Keyword(s):  
Sustainable Management ◽  
Near Field ◽  
Steel Corrosion ◽  
Spent Fuel ◽  
Water Leaching ◽  
Residual Rate ◽  
Partially Saturated ◽  
Deep Geological Disposal ◽  
High Level ◽  
Glass Behavior

ABSTRACTAccording to the Planning Act of 28th June 2006, Andra is in charge of ensuring the sustainable management of all radioactive waste generated in France, especially the high-level and long-lived vitrified waste produced from spent fuel recycling.Since 2006, all the studies and research related to the components of HLW cells have been incorporated into a broader R&D program which aims at characterizing and modeling (i) the glass matrix dissolution, (ii) the corrosion of the overpack and the lining, and (iii) the claystone evolution in the near field, considering all the interactions between these surrounding materials. This program, coordinated by Andra, has involved up to eighteen laboratories.After closure of disposal cells and overpack failure, glass alteration is expected to begin in partially saturated conditions due to hydrogen production resulting from carbon steel corrosion in anoxic conditions. Therefore, the glass should at least partially be hydrated by water vapor during thousands of years until complete saturation. A part of the studies aimed to determine the glass behavior in such conditions, the influence of the main parameters (temperature, relative humidity) and consequences of vapor hydration on subsequent radionuclides release by water leaching.In addition, the major part of the work focused on the influence of the environment on glass alteration. The effect of clay pore water on glass alteration rates (initial rate, rate drop and residual rate) was determined and particularly that of pH and magnesium. The nature of steel corrosion products and their interactions with glass alteration were also investigated. All these studies relied on experiments in surface laboratories, in Andra’s underground laboratory, together with natural or archeological analogs and modeling studies.

Modelling sorption data for the actinides Am(III), Np(V) and Pa(V) on montmorillonite

2006 ◽  
Vol 94 (9-11) ◽  
Author(s):  
Michael H. Bradbury ◽  
B. Baeyens
Keyword(s):  
Near Field ◽  
High Level Waste ◽  
Spent Fuel ◽  
Data Set ◽  
Sorption Data ◽  
Deep Geological Disposal ◽  
Retention Characteristics ◽  
High Level ◽  
Trace Concentrations ◽  
First Time

The retention characteristics of the bentonite near-field engineered barrier proposed in most of the concepts for the deep geological disposal of high-level waste and spent fuel are an important component in repository performance assessment studies. Montmorillonite generally constitutes 65 to 90 wt.% of the bentonite. Sorption edge measurements have been performed at trace concentrations for the actinides Am(III), Np(V) and Pa(V) on purified and conditioned SWy-1 montmorillonite under anoxic, carbonate free conditions. To the best of the author´s knowledge, this is the first time a sorption data set has been measured for

scholarly journals Preliminary modelling of radionuclide migration in the argillaceous sediments of the Sumer Formation (Northwestern Bulgaria)

2020 ◽  
Vol 49 (3) ◽  
pp. 13-18
Author(s):  
Dimitar Antonov ◽  
Madlena Tsvetkova ◽  
Doncho Karastanev
Keyword(s):  
Radioactive Waste Disposal ◽  
High Level Waste ◽  
Spent Fuel ◽  
Radionuclide Migration ◽  
Geological Disposal ◽  
Disposal Facility ◽  
Deep Geological Disposal ◽  
High Level ◽  
Host Rocks ◽  
Selection Of

In Bulgaria, from the preliminary analyses performed for site selection of deep geological disposal of high-level waste (HLW) and spent fuel (SF), it was concluded that the most promising host rocks are the argillaceous sediments of the Sumer Formation (Lower Cretaceous), situated in the Western Fore-Balkan Mts. The present paper aims to compare the transport of three major radionuclides from a hypothetical radioactive waste disposal facility, which incorporates an engineering barrier of bentonite into the argillaceous (marl) medium. The simulations were performed by using HYDRUS-1D computer programme. The results are used for a preliminary estimation of argillaceous sediments as a host rock for geological disposal of HLW.

scholarly journals Needs of countries with longer timescale for deep geological repository implementation

2020 ◽  
Vol 6 ◽  
pp. 22
Author(s):  
Bálint Nős
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
High Level Waste ◽  
Deep Geological Repository ◽  
Spent Fuel ◽  
International Consensus ◽  
Geological Disposal ◽  
Geological Repository ◽  
Deep Geological Disposal ◽  
High Level

Countries operating nuclear power plants have to deal with the tasks connected to spent fuel and high-level radioactive waste management. There is international consensus that, at this time, deep geological disposal represents the safest and most sustainable option as the end point of the management of high-level waste and spent fuel considered as waste. There are countries with longer timescale for deep geological repository (DGR) implementation, meaning that the planned date of commissioning of their respective DGRs is around 2060. For these countries cooperation, knowledge transfer, participation in RD&D programmes (like EURAD) and adaptation of good international practice could help in implementing their own programmes. In the paper the challenges and needs of a country with longer implementation timescale for DGR will be introduced through the example of Hungary.

Common Aspects of the PNC and Nagra Assessments of Deep Repositories for Vitrified HLW

1994 ◽  
Vol 353 ◽  
Author(s):  
P.A. Smith ◽  
H. Umeki ◽  
F. Neall ◽  
I.G. McKinley
Keyword(s):  
Mathematical Models ◽  
State Of The Art ◽  
Performance Assessments ◽  
High Level Waste ◽  
Geological Disposal ◽  
Common Features ◽  
Comprehensive Performance ◽  
Deep Geological Disposal ◽  
Engineered Barriers ◽  
High Level

AbstractThe repository concepts developed by PNC and Nagra for the disposal of vitrified high-level waste show many common features; both concepts involve deep geological disposal, with massive engineered barriers of similar design. PNC and Nagra have recently published comprehensive performance assessments based on their repository concepts and these are compared in order to gain assurance that the conceptual and mathematical models employed are state-of-the-art and to build confidence in the datasets in the assessment results. The predicted performance of the engineered barriers is more favourable to safety in the case of the Nagra Assessment, whereas the opposite is true of the geological barriers. These differences in predicted performance can be traced to a small number of differences in model assumptions and data. It is concluded that the differences are consistent with the somewhat different requirements of the two assessments and that the comparison enhances confidence in both assessments.

Repository Engineering and Demonstration: Special Challenges for TRU

2010 ◽  
Author(s):  
Ian G. McKinley ◽  
Hiroyasu Takase
Keyword(s):  
Rock Properties ◽  
High Level Waste ◽  
Heat Output ◽  
Spent Fuel ◽  
Diverse Range ◽  
Deep Geological Disposal ◽  
Safety Case ◽  
High Level ◽  
A Site ◽  
Term Performance

The diverse range of long-lived radioactive wastes without significant heat output specified for deep geological disposal (here termed TRU) pose challenges that are potentially more serious than those from vitrified high-level waste and spent fuel. Despite this, the latter tend to be the focus of R&D in national programmes. Such challenges are particularly severe for the case for countries that are not considering evaporite host rocks or have a volunteering approach to siting and for those with inventories of TRU resulting from reprocessing of spent fuel. While there is little doubt that safe disposal of TRU is feasible, it is tricky to develop a convincing safety case for a site during early stages of characterisation as, compared to HLW/SF, less credit can be taken for robust, long-term performance of current designs of the engineered barrier systems. In order to improve this situation and increase flexibility with respect to host rock properties, two different options are available — improving the conditioning of particular waste streams or improving the overall repository safety concept. Although the former has been a focus for work in some countries (particularly Japan), much less effort has been invested in the latter and hence this will be illustrated by some examples. These options are compared in terms of their pros and cons with respect to practicality of implementation, environmental impact and cost. Additionally, the ease with which the resulting safety case can be supported by demonstrations of key arguments will be discussed, which may indicate the likely degree of acceptance by stakeholders.

Development of the Supercontainer Design for Deep Geological Disposal of High-Level Heat Emitting Radioactive Waste in Belgium

2006 ◽  
Vol 932 ◽  
Author(s):  
Johan J.P. Bel ◽  
Stephen M. Wickham ◽  
Robert M.F. Gens
Keyword(s):  
Selection Process ◽  
Metal Corrosion ◽  
Preliminary Evaluation ◽  
Spent Fuel ◽  
Deep Geological Disposal ◽  
Reference Design ◽  
And Performance ◽  
High Level ◽  
Design Options ◽  
Future Work

ABSTRACTONDRAF-NIRAS has recently selected a Supercontainer with a Portland Cement (PC) buffer as the preferred new reference design for disposal of HLW and spent fuel. The selection process involved a multi-criteria analysis of alternative design options, which were evaluated against a range of long-term safety and feasibility criteria. A PC concrete has been chosen for the buffer because this will provide a highly alkaline chemical environment, which will last for thousands of years. In this environment the external surface of the overpack will be passivated and overpack corrosion will be inhibited. The concrete buffer also has low-hydraulic conductivity to slow the infiltration of external fluids to the overpack surface, and provides radiological shielding.ONDRAF-NIRAS has made a preliminary evaluation of the viability of the reference Supercontainer design. The following areas were reviewed and investigated: radiolysis, thermo-hydraulic (TH) behaviour of the concrete buffer, metal corrosion, the chemical and mineralogical evolution of the concrete buffer, and relevant industrial experience. This paper describes the main findings, and identifies remaining design and performance uncertainties. Prioritisation and recommendations for future work are also given.

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