scholarly journals Cement as a component of the multi-barrier system: radionuclide retention in cementitious environments

2021 ◽  
Vol 1 ◽  
pp. 151-152
Author(s):  
Xavier Gaona ◽  
Marcus Altmaier ◽  
Iuliia Androniuk ◽  
Nese Çevirim-Papaioannou ◽  
Michel Herm ◽  
...  
Keyword(s):  
Nuclear Waste ◽  
Host Rock ◽  
Near Field ◽  
Quantitative Description ◽  
Cementitious Materials ◽  
Assessment Process ◽  
High Level Waste ◽  
Hardened Cement ◽  
Spectroscopic Techniques ◽  
Research Activities

Abstract. Safety concepts regarding nuclear waste disposal in underground repositories generally rely on a combination of engineered and geological barriers, which minimize the potential release of radionuclides from the containment-providing rock zone or even their transport into the biosphere. Cementitious materials are used for conditioning of certain nuclear waste types, as components of waste containers and overpacks, as well as being constituents of structural materials at the interface between backfilling and host rock in some repository concepts. For instance, the preferred option for the disposal of high-level waste (HLW) in Belgium is based on the supercontainer design, which consists of a carbon steel overpack surrounded by a thick concrete buffer (Bel et al., 2006). In the event of formation water interacting with cementitious materials, pore water solutions characterized by (highly) alkaline pH conditions will form. These boundary conditions define the chemical response of the radionuclides, but also influence the behaviour of neighbouring components of the multi-barrier system, e.g. bentonitic or argillaceous backfilling and host rock. Hardened cement paste or Sorel cement are considered to be main sorbing materials present in the near field of repositories for low- and intermediate-level waste (L/ILW). Hence, interactions of radionuclides with cementitious materials represent a very important mechanism retarding their mobility and potential migration from the near field (Wieland, 2014; Ochs et al., 2016). While the quantitative description of the sorption processes (usually in terms of sorption coefficients, i.e. Kd values) is a key input in the safety analysis of nuclear waste repositories, detailed mechanistic analysis and understanding of sorption phenomena provide additional scientific arguments and important process understanding, and thus enhance both the quality of safety arguments and the overall confidence in the safety assessment process. Research at KIT-INE dedicated to the interaction of cementitious materials with radionuclides is conducted in the context of different repository concepts, including clay (low- and high-ionic strength conditions), crystalline rock or rock salt. Experimental and theoretical studies are performed within the framework of national (GRAZ, BMWi) and international (CEBAMA and EURAD-CORI, EU Horizon 2020 Programme) projects, extending to third-party projects with several waste management organizations in Europe, e.g. SKB (Sweden), ONDRAF-NIRAS (Belgium) or BGE (Germany). The combination of classical experimental (wet chemistry) methods, advanced spectroscopic techniques and theoretical calculations provides both an accurate quantitative evaluation and a fundamental understanding of the sorption processes. Examples of recent studies at KIT-INE on radionuclide behaviour in cementitious systems in the context of both L/ILW and HLW will be presented in this contribution to explain methodologies, scientific approaches and results. The present state of knowledge as well as main remaining uncertainties affecting the retention processes of radionuclides in cementitious environments under different conditions will be critically discussed, also in view of current international research activities and repository projects.

Hydro-geochemical processes in the emplacement cavern of a low and intermediate-level waste repository in an indurated clay-rock

2021 ◽  
Author(s):  
Vanessa Montoya ◽  
Jaime Garibay-Rodriguez ◽  
Olaf Kolditz
Keyword(s):  
Chemical Evolution ◽  
Nuclear Waste ◽  
Host Rock ◽  
Numerical Models ◽  
Intermediate Level ◽  
High Level Waste ◽  
Nuclear Waste Repository ◽  
Case Scenario ◽  
Damaged Zone ◽  
Waste Repository

<p>By 2080, Germany will have to store around 600 000 m<sup>3</sup> of low and intermediate-level nuclear waste (L-ILW) with negligible heat generation. This kind of waste is largely made up of used parts of nuclear power stations such as pumps, pipelines, filters, etc. placed in various types of waste containers made from either steel, cast iron, or reinforced concrete in different designs and sizes (i.e. cylindrical or box shaped). It is already decided that a total of 303 000 of the 600 000 m<sup>3</sup> L-ILW will be disposed in a final storage facility in the former iron ore mine Schacht Konrad which is under construction. However, it is still not clear where the L-ILW emplaced in in the old salt mine Asse (200 000 m<sup>3</sup>) will be stored in the future. The situation is particularly critical, as the waste have to be retrieved from the instable mine shafts partially flooded with groundwater, causing strong socio-political concerns as radioactive waste could contaminate the water nearby. For this reason, the new search for a nuclear waste repository for high-level waste (HLW), started in 2017, should also consider the possibility to accommodate the waste from Asse. Obviously, this is still subject to critics as this will make finding a final repository more difficult as storing HLW and L-ILW together requires different concepts and designs for each other and, above all, much more space.</p><p>In this context, in this contribution we have defined conceptual and numerical models to assess the hydro-chemical evolution of a L-ILW disposal cell in indurated clay rocks, involving the interaction of different components/materials and the expected hydraulic and/or chemical gradients over 100 000 years. The L-ILW disposal cell leverages a multi-barrier concept buried 400 m below the surface. The multi-barrier system is comprised of the waste matrix (i.e. backfilling the waste drums), the disposal container, the mortar backfill in the emplacement tunnel (where the disposal containers are located) and the clay host rock. The dimensions and design of the emplacement tunnel (e.g. 11 × 13 m) and disposal cells represent and consider some aspects taken into account in the designs of some European countries. In addition, tunnel walls reinforced with a shotcrete liner and the Excavation Damaged Zone is considered in the concept. The model is implemented in OpenGeoSys-6, an open-source version-controlled scientific software based on Finite Element Method which is capable of handling fully coupled hydro-chemical models by coupling OpenGeoSys to iPHREEQC. First calculation results, demonstrate that the most important processes affecting the near-field chemical evolution are i) the degradation of the concrete and cementitious grouts with porewater migrating inwards from the host rock and ii) the significant quantities of reactive and non-reactive gases (i.e. hydrogen, carbon dioxide and methane) that are generated as a result of: i) the anaerobic corrosion of metals present in the waste and containers and ii) the degradation of organic compounds by microbial and chemical processes. As a first approximation, some assumptions and simplifications have been considered, probably resulting in a wort case scenario.</p>

The Salton Sea Geothermal Field as a Natural Analog for the Near-Field in a Salt High-Level Nuclear Waste Repository

1984 ◽  
Vol 44 ◽  
Author(s):  
Wilfred A. Elders ◽  
Judith B. Moody
Keyword(s):  
Nuclear Waste ◽  
Near Field ◽  
Geothermal Field ◽  
Salton Sea ◽  
High Level Waste ◽  
Nuclear Waste Repository ◽  
Natural Analog ◽  
Naturally Occurring ◽  
Wide Range ◽  
High Level

AbstractThe Salton Sea Geothermal Field (SSGF), on the delta of the Colorado River in southern California, is being studied as a natural analog for the near-field environment of proposed nuclear waste repositories in salt. A combination of mineralogical and geochemical methods is being employed to develop a three dimensional picture of temperature, salinity, lithology, mineralogy, and chemistry of reactions between the reservoir rocks and the hot brines. Our aim is to obtain quantitative data on mineral stabilities and on mobilities of the naturally occurring radionuclides of concern in Commercial High-Level Waste (CHLW). These data will be used to validate the EQ3/6 geochemical code under development to model the salt near-field repository behavior.Maximum temperatures encountered in wells in the SSGF equal or exceed peak temperatures expected in a salt repository. Brines produced from these wells have major element chemistry similar to brines from candidate salt sites. Relative to the rocks, these brines are enriched in Na, Mn, Zn, Sr, Ra and Po, depleted in Ba, Si, Mg, Ti, and Al, and strongly depleted in U and Th. However the unaltered rocks contain only about 2–3 ppm of U and 4–12 ppm of Th, largely in detrital epidotes and zircons. Samples of hydrothermally altered rocks from a wide range of temperature and salinity show rather similar uniform low concentrations of these elements, even when authigenic illite, chlorite, epidote and feldspar are present. These observations suggest that U and Th are relatively immobile in these hot brines. However Ra, Po, Cs and Sr are relatively mobile. Work is continuing to document naturally occurring radionuclide partitioning between SSGF minerals and brine over a range of temperature, salinity, and lithology.

scholarly journals Radionuclide geochemistry: solubility and thermodynamics in an HLW repository

2021 ◽  
Vol 1 ◽  
pp. 149-150
Author(s):  
David Fellhauer ◽  
Xavier Gaona ◽  
Marcus Altmaier ◽  
Horst Geckeis
Keyword(s):  
Nuclear Waste ◽  
Host Rock ◽  
Thermodynamic Data ◽  
Near Field ◽  
Reference Database ◽  
Chemical Information ◽  
Energy Agency ◽  
Complex Formation Constants ◽  
The Impact ◽  
Hlw Repository

Abstract. Deep geological disposal is the internationally favoured option to isolate high-level nuclear waste (HLW) from the biosphere and to minimise the potential radiological risk for future generations. Potentially contacting aqueous solutions such as groundwater may, however, lead to the corrosion of the solid casks containing the nuclear waste, and the formation of aqueous radionuclide systems in the near-field of the emplacement rooms. As dissolved species, radionuclides can in principle further migrate into the far-field and finally reach the biosphere on medium and long timescales. Like all chemical species, the radionuclides are subject to fundamental (geo)chemical laws. Relevant reactions that control retention and release, and hence, the migration behaviour and fate of radionuclides in a repository, are solubility equilibria, formation of soluble complexes, redox reactions, sorption on and incorporation into mineral surfaces, transport phenomena etc. These processes depend directly on the (geo)chemical boundary conditions, and, consequently, can differ greatly for various host rock systems such as clay rock, rock salt, and crystalline rock. Many of the radionuclides in HLW are heavy metals that are sparingly soluble under various repository-relevant conditions, e.g. actinides, lanthanides, transition metals, so that only partial dissolution (mobilisation) from the solid waste matrices is expected. This underlines the importance of evaluating the radionuclide solubility within a geochemically based safety assessment for repositories as it provides reliable upper-limit concentrations of the mobile, potentially migrating radionuclide fraction in the near-field. In this contribution, we discuss relevant aspects related to the topic radionuclide solubility and thermodynamics in a HLW repository. This includes a summary of recent laboratory studies on the solubility behaviour and speciation of key radionuclides in repository-relevant solutions, which are an important basis for obtaining (geo)chemical information and models, and the corresponding fundamental thermodynamic constants on aqueous radionuclide systems. National and international thermodynamic database projects, where quality-assured thermodynamic data (solubility products, complex formation constants, and ion-interaction parameters) are evaluated and compiled, e.g. the Nuclear Energy Agency Thermochemical Database (http://www.oecd-nea.org, last access: 1 November 2021) or the Thermodynamic Reference Database (http://www.thereda.de, last access: 1 November 2021), are highlighted and the main remaining uncertainties discussed. The experimental information and the quantitative thermodynamic data are applied within a generic case study to demonstrate the impact of different geochemical solution conditions representing different host rock systems considered as HLW repositories in Germany on the solubility and speciation of selected radionuclides.

scholarly journals Alteration of vitrified intermediate level nuclear waste in alkaline media: effects of cementitious materials, pH and temperature

RSC Advances ◽  
2018 ◽  
Vol 8 (66) ◽  
pp. 37665-37680 ◽  
Author(s):  
Tomo Suzuki-Muresan ◽  
A. Abdelouas ◽  
C. Landesman ◽  
A. Ait-Chaou ◽  
Y. El Mendili ◽  
...  
Keyword(s):  
Cement Paste ◽  
Nuclear Waste ◽  
Media Effects ◽  
Cementitious Materials ◽  
Intermediate Level ◽  
Alkaline Media ◽  
Hardened Cement ◽  
Hardened Cement Paste

Alteration experiments involving intermediate level nuclear waste (ILW) glass in contact with hardened cement paste (HCP) were performed to assess its behavior under simulated repository conditions.

EBS Behaviour Immediately After Repository Closure in a Clay Host Rock: The HE-E Experiment (Mont Terri URL)

2011 ◽  
Author(s):  
Irina Gaus ◽  
Klaus Wieczorek ◽  
Juan Carlos Mayor ◽  
Thomas Trick ◽  
Jose´-Luis Garcia` Sin˜eriz ◽  
...  
Keyword(s):  
Thermal Conductivity ◽  
Host Rock ◽  
Large Scale ◽  
Process Models ◽  
Research Activities ◽  
Hydraulic Behaviour ◽  
Validation Experiment ◽  
Engineered Barrier ◽  
Mont Terri

The evolution of the engineered barrier system (EBS) of geological repositories for radioactive waste has been the subject of many research programmes during the last decade. The emphasis of the research activities was on the elaboration of a detailed understanding of the complex thermo-hydro-mechanical-chemical processes, which are expected to evolve in the early post closure period in the near field. It is important to understand the coupled THM-C processes and their evolution occurring in the EBS during the early post-closure phase so it can be confirmed that the safety functions will be fulfilled. Especially, it needs to be ensured that interactions during the resaturation phase (heat pulse, gas generation, non-uniform water uptake from the host rock) do not affect the performance of the EBS in terms of its safety-relevant parameters (e.g. swelling pressure, hydraulic conductivity, diffusivity). The 7th Framework PEBS project (Long Term Performance of Engineered Barrier Systems) aims at providing in depth process understanding for constraining the conceptual and parametric uncertainties in the context of long-term safety assessment. As part of the PEBS project a series of laboratory and URL experiments are envisaged to describe the EBS behaviour after repository closure when resaturation is taking place. In this paper the very early post-closure period is targeted when the EBS is subjected to high temperatures and unsaturated conditions with a low but increasing moisture content. So far the detailed thermo-hydraulic behaviour of a bentonite EBS in a clay host rock has not been evaluated at a large scale in response to temperatures of up to 140°C at the canister surface, produced by HLW (and spent fuel), as anticipated in some of the designs considered. Furthermore, earlier THM experiments have shown that upscaling of thermal conductivity and its dependency on water content and/or humidity from the laboratory scale to a field scale needs further attention. This early post-closure thermal behaviour will be elucidated by the HE-E experiment, a 1:2 scale heating experiment setup at the Mont Terri rock laboratory, that started in June 2011. It will characterise in detail the thermal conductivity at a large scale in both pure bentonite as well as a bentonite-sand mixture, and in the Opalinus Clay host rock. The HE-E experiment is especially designed as a model validation experiment at the large scale and a modelling programme was launched in parallel to the different experimental steps. Scoping calculations were run to help the experimental design and prediction exercises taking the final design into account are foreseen. Calibration and prediction/validation will follow making use of the obtained THM dataset. This benchmarking of THM process models and codes should enhance confidence in the predictive capability of the recently developed numerical tools. It is the ultimate aim to be able to extrapolate the key parameters that might influence the fulfilment of the safety functions defined for the long term steady state.

scholarly journals Effect of a Nitrite/Nitrate-Based Accelerator on the Strength Development and Hydrate Formation in Cold-Weather Cementitious Materials

Materials ◽  
2021 ◽  
Vol 14 (4) ◽  
pp. 1006
Author(s):  
Akira Yoneyama ◽  
Heesup Choi ◽  
Masumi Inoue ◽  
Jihoon Kim ◽  
Myungkwan Lim ◽  
...  
Keyword(s):  
Calcium Nitrate ◽  
Hydrate Formation ◽  
Frost Damage ◽  
Cementitious Materials ◽  
Heat Of Hydration ◽  
Cold Weather ◽  
Strength Development ◽  
Hardened Cement ◽  
Tricalcium Aluminate ◽  
Nitrite Nitrate

Recently, there has been increased use of calcium-nitrite and calcium-nitrate as the main components of chloride- and alkali-free anti-freezing agents to promote concrete hydration in cold weather concreting. As the amount of nitrite/nitrate-based accelerators increases, the hydration of tricalcium aluminate (C3A phase) and tricalcium silicate (C3S phase) in cement is accelerated, thereby improving the early strength of cement and effectively preventing initial frost damage. Nitrite/nitrate-based accelerators are used in larger amounts than usual in low temperature areas below −10 °C. However, the correlation between the hydration process and strength development in concrete containing considerable nitrite/nitrate-based accelerators remains to be clearly identified. In this study, the hydrate composition (via X-ray diffraction and nuclear magnetic resonance), pore structures (via mercury intrusion porosimetry), and crystal form (via scanning electron microscopy) were determined, and investigations were performed to elucidate the effect of nitrite/nitrate-based accelerators on the initial strength development and hydrate formation of cement. Nitrite/nitrate-AFm (aluminate-ferret-monosulfate; AFm) was produced in addition to ettringite at the initial stage of hydration of cement by adding a nitrite/nitrate-based accelerator. The amount of the hydrates was attributed to an increase in the absolute amounts of NO2− and NO3− ions reacting with Al2O3 in the tricalcium aluminate (C3A phase). Further, by effectively filling the pores, it greatly contributed to the enhancement of the strength of the hardened cement product, and the degree of the contribution tended to increase with the amount of addition. On the other hand, in addition to the occurrence of cracks due to the release of a large amount of heat of hydration, the amount of expansion and contraction may increase, and it is considered necessary to adjust the amount used for each concrete work.

scholarly journals New Approach for the Determination of Radiological Parameters on Hardened Cement Pastes with Coal Fly Ash

Materials ◽  
2021 ◽  
Vol 14 (3) ◽  
pp. 475
Author(s):  
Ana María Moreno de los Reyes ◽  
José Antonio Suárez-Navarro ◽  
Maria del Mar Alonso ◽  
Catalina Gascó ◽  
Isabel Sobrados ◽  
...  
Keyword(s):  
Fly Ash ◽  
Activity Concentration ◽  
Gamma Ray ◽  
Coal Fly Ash ◽  
Cementitious Materials ◽  
New Method ◽  
Supplementary Cementitious Materials ◽  
Hardened Cement ◽  
Cement Pastes ◽  
Activity Concentrations

Supplementary cementitious materials (SCMs) in industrial waste and by-products are routinely used to mitigate the adverse environmental effects of, and lower the energy consumption associated with, ordinary Portland cement (OPC) manufacture. Many such SCMs, such as type F coal fly ash (FA), are naturally occurring radioactive materials (NORMs). 226Ra, 232Th and 40K radionuclide activity concentration, information needed to determine what is known as the gamma-ray activity concentration index (ACI), is normally collected from ground cement samples. The present study aims to validate a new method for calculating the ACI from measurements made on unground 5 cm cubic specimens. Mechanical, mineralogical and radiological characterisation of 28-day OPC + FA pastes (bearing up to 30 wt % FA) were characterised to determine their mechanical, mineralogical and radiological properties. The activity concentrations found for 226Ra, 212Pb, 232Th and 40K in hardened, intact 5 cm cubic specimens were also statistically equal to the theoretically calculated values and to the same materials when ground to a powder. These findings consequently validated the new method. The possibility of determining the activity concentrations needed to establish the ACI for cement-based materials on unground samples introduces a new field of radiological research on actual cement, mortar and concrete materials.

Sorption Mechanism of Carbon-14 by Hardened Cement Paste

1995 ◽  
Vol 412 ◽  
Author(s):  
K. Noshita ◽  
T. Nishi ◽  
M. Matsuda ◽  
T. Izumida
Keyword(s):  
Electrostatic Force ◽  
Cementitious Materials ◽  
Hardened Cement ◽  
Sorption Mechanism ◽  
Carbon 14 ◽  
Adsorption Sites ◽  
Radioactive Waste Repositories ◽  
Waste Repositories ◽  
Positively Charged

AbstractCarbon-14 sorption by cementitious materials should be enhanced to ensure the long term safety of radioactive waste repositories. The sorption mechanism of inorganic C- 14 (CO32- was investigated using batch sorption experiments and zeta potential measurements. The results suggested that C-14 was adsorbed onto the cement surface by an electrostatic force, due to the reaction between SiO2 and CaO contained in the cementitious composition. That is, SiO2 was originally negatively charged (SiO-) in cement, but became positively charged through the interaction of Ca2+. These positive sites on the SiO2 surface adsorbed inorganic C-14. Ordinary Portland cement (OPC) did not contain enough SiO2 compared with its CaO content to produce sufficient numbers of C-14 adsorption sites. The C-14 distribution coefficient (Kd) was increased from 2,000 to 7,000 mL/g by adding SiO2 to OPC.

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