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Pore structure and hygric properties of composite materials for radionuclide protection barriers
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Many practical applications of concrete require adjustment of its properties. If concrete is used as casing for radioactive waste deposits, its main purpose is to mitigate radionuclide release and the mix design must conform to this requirement. In this article bentonite as a material with great sorption capacity is used as a concrete component. The main issue is to assess durability of designed mixtures. Therefore, the pore structure and hygric properties are investigated. The experimental results show that bentonite utilization in concrete mixtures leads to a significant growth of pore amount in the range of 0.01 to 1 μm and increase of water transport ability.
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Mass concrete has been commonly known for its thermal stresses which arise due to the entrapment of hydration temperature susceptible to thermal cracking. The utilization of mineral additives is a promising and widely adopted technique to mitigate such effects. This paper presents the thermal, physico-chemical, mechanical, and morphological behaviour of mass concrete with blends of bentonite (BT) and fly ash (FA). Apart from the rise in temperature due to hydration, the compressive strength, ultrasonic pulse velocity (UPV), differential thermal analysis (DTA), thermo-gravimetric analysis (TGA), X-ray diffraction (XRD) analysis, and microstructure were studied. The results of this study revealed that the substitution of BT and FA significantly improved the compressive strength and development rate of UPV in the mass concrete samples. The FA concrete (FC) specimen presented the lowest temperature during the peak hours compared to all other concrete mixes studied in this research. Bentonite concrete (BC) was also found to be more effective in controlling the escalation of temperature in mass concrete. Scan electron microscopy (SEM) micrographs presented partially reacted FA particles in a mix. XRD and DTA analysis indicated that the concentration of calcium hydroxide (CH) declined by substituting FA and BT, specifically in ternary blends, which was due to the dilution effect and consumption of CH through the pozzolanic reaction.
This review article presents the current state-of-knowledge of the use of cementitious materials for radioactive waste disposal. An overview of radwaste management processes with respect to the classification of the waste type is given. The application of cementitious materials for waste disposal is divided into two main lines: i) as a matrix for direct immobilization of treated waste form; and ii) as an engineered barrier of secondary protection in the form of concrete or grout. In the first part the immobilization mechanisms of the waste by cement hydration products is briefly described and an up-to date knowledge about the performance of different cementitious materials is given, including both traditional cements and alternative binder systems. The advantages, disadvantages as well as gaps in the base of information in relation to individual materials are stated. The following part of the article is aimed at description of multi-barrier systems for intermediate level waste repositories. It provides examples of proposed concepts by countries with advanced waste management programmes. In the paper summary, the good knowledge of the material durability due to its vast experience from civil engineering is highlighted however with the urge for specific approach during design and construction of a repository in terms of stringent safety requirements.
Radioactive wastes are generated from a wide range of sources, including the power industry, and medical and scientific research institutions, presenting a range of challenges in dealing with a diverse set of radionuclides of varying concentrations. Conditioning technologies are essential for the encapsulation and immobilisation of these radioactive wastes, forming the initial engineered barrier required for their transportation, storage and disposal. The need to ensure the long term performance of radioactive waste forms is a key driver of the development of advanced conditioning technologies. The Handbook of advanced radioactive waste conditioning technologies provides a comprehensive and systematic reference on the various options available and under development for the treatment and immobilisation of radioactive wastes. The book opens with an introductory chapter on radioactive waste characterisation and selection of conditioning technologies. Part one reviews the main radioactive waste treatment processes and conditioning technologies, including volume reduction techniques such as compaction, incineration and plasma treatment, as well as encapsulation methods such as cementation, calcination and vitrification. This coverage is extended in part two, with in-depth reviews of the development of advanced materials for radioactive waste conditioning, including geopolymers, glass and ceramic matrices for nuclear waste immobilisation, and waste packages and containers for disposal. Finally, part three reviews the long-term performance assessment and knowledge management techniques applicable to both spent nuclear fuels and solid radioactive waste forms. With its distinguished international team of contributors, the Handbook of advanced radioactive waste conditioning technologies is a standard reference for all radioactive waste management professionals, radiochemists, academics and researchers involved in the development of the nuclear fuel cycle. Provides a comprehensive and systematic reference on the various options available and under development for the treatment and immobilisation of radioactive wastes Explores radioactive waste characterisation and selection of conditioning technologies including the development of advanced materials for radioactive waste conditioning Assesses the main radioactive waste treatment processes and conditioning technologies, including volume reduction techniques such as compaction.
Chloride-induced corrosion in reinforcing steel is one of the major durability problems in reinforced concrete (RC) structures. Fick’s 2nd law can be simply applied to submerged RC structures however it has a very limited application to partially saturated condition. Furthermore, if RC structures have cracks on their surface in partially saturated condition, additional diffusion and permeation due to crack width should be considered for the quantitative evaluation of chloride penetration.In this paper, an analytical model is proposed for an evaluation of chloride behavior in cracked concrete. Both chloride diffusion and water permeation in a Representative Element Volume (REV) with crack are considered, which assume averaged transport of ion mass. Through rapid chloride penetration test (RCPT) for specimens with different crack width, crack effect on diffusion is analyzed considering crack width (0.1–0.4mm). Utilizing the crack effect on diffusion and permeation, an analysis technique for chloride behavior in cracked concrete is proposed based on the framework considering thermodynamic coupling of hydration, moisture transport, and micro-structure formation. The proposed technique shows a possibility of evaluation for chloride penetration in partially saturated-cracked concrete through the comparison with the results from salt spraying test (SST), which are in good agreement.
The leach rate of 60Co and 137Cs from two different ion exchange resins: (a) spent cation exchange resins and (b) spent mix bead ion exchange resins in cement–bentonite matrix has been studied. The solidification matrix was a standard Portland cement mixed with 290–350 kg/m3 spent cation exchange resins, with or without 2–5% of bentonite clay. The leach rates from the cement–bentonite matrix as 60Co: (4.2–7.3)×10−5 cm/d, and for 137Cs: (3.2–6.6)×10−5 cm/d, after 245 days were measured. From the leaching data the apparent diffusivity of cobalt and cesium in cement–bentonite clay matrix with a waste load of 290–350 kg/m3 spent cation exchange resins was measured as 60Co: (1.0–4.0)×10−6 cm2/d and for 137Cs: (0.5–2.6)×10−4 cm2/d after 245 days. These results are part of a 20-year mortar and concrete testing project which will influence the design of radioactive waste management for a future Serbian radioactive waste disposal center.
Free water intake experiment is analyzed from the point of view of the applicability of its output for material characterization. Several experimental setups are presented and discussed. Characteristic examples of sorptivity plots for different materials are given and their main features have been described. The application of the experiment for characterization of material damage and for explaining the water transport mechanisms in two-layer systems is presented. It is concluded that the free water intake experiment basically can be used as a means of material characterization but its strong and weak points should always be considered if conclusions on the behavior of a particular material are to be drawn.
To assess the safety for disposal of radioactive mortar-waste composition, the leaching of 137Cs, 60Co and 85Sr from immobilized radioactive evaporator concentrate, into a surrounding fluid has been studied. This paper also presents results from studies made to determine the influence of natural sorbents (bentonite and zeolite) on immobilization of evaporator concentrate in cement matrix, by measuring leachability of 137Cs, 60Co and 85Sr. Leaching tests were carried out in accordance with a method recommended by IAEA, under 300 days. From the leaching data the apparent diffusivity of cobalt, cesium and strontium in cement–bentonite matrix with a waste load of 701 kg/m3 evaporator concentrate was measured. The uniaxial compressive strength of 11 mortar samples is determined following the ASTM standards. These results are part of a 30-year mortar and concrete testing project which will influence the design of radioactive waste management for a future Serbian radioactive waste disposal center.
Moisture diffusivity is a transport property that is frequently used in the hygrothermal analysis of building envelope components. This property is dependent on the local moisture content. The experiments that lead to the detailed information on the dependence of diffusivity on moisture content are often very sophisticated. However, two recent exercises, as a part of the activities of an International Energy Agency Annex, have shown that a correct estimate of the magnitude of the moisture diffusivity can provide useful information with regard to its application in hygrothermal analysis. This technical note presents results from a simple moisture absorption measurement that led to a good estimate of the moisture diffusivity of building materials. Results from measurements on a sample of spruce are presented.
Zeolites are preferred ion exchange materials for the removal of radioisotopes from aqueous nuclear wastes because of their selectivity, radiation and temperature stability, and good compatibility with the cement matrix. Loaded materials, which are not regenerated, are preferably embedded in a solid matrix prior to ultimate storage in a repository. The aim of the present study is to investigate the possibility of solidifying exhausted synthetic zeolite A, loaded with (137)Cs and/or (90)Sr radionuclides, in Ordinary Portland Cement (OPC). Several factors affecting the characteristics of the final solidified waste product towards safe disposal such as mechanical strength and leaching behavior of the radionuclides have been studied. A simplified mathematical model based on diffusion mechanism for cylindrical geometry waste matrix has been simulated to predict the release rates of the investigated radionuclides from cement matrix. The predicted values are discussed in relation to experimentally observed leach rates to confirm the proposed mechanism in the model. The obtained results showed that the presence of zeolite A in the final cemented wastes improve the mechanical characteristics of the solidified cement matrix (mechanical strength and setting times) towards the safety requirements and reduce considerably the radionuclides leach rates.
The behaviours of cementitious materials in long term storage and disposal of radioactive waste: results of a coordinated research project
The behaviours of cementitious materials in long term storage and disposal of
radioactive waste: results of a coordinated research project, International Atomic
Energy Agency, Vienna (2013)