Journal articles: 'Radioactive wastes – Thermal properties'

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El-Didamony, H., T. A. Bayoumi, and M. I. Sayed. "Evaluation of the Properties of Cemented Liquid Scintillator Wastes under Flooding Scenario in Various Aqueous Media." ISRN Chemical Engineering 2012 (December 26, 2012): 1–11. http://dx.doi.org/10.5402/2012/373795.

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Experiments simulating flooding scenario in various aqueous media for a long period were carried out to determine the adequacy of cement-clay composite for solidification/stabilization of spent organic radioactive liquid scintillator wastes. The final cement waste form blocks were immersed in three aqueous media, namely, seawater, groundwater, and tapwater. The immersion process lasted for increasing periods up to 540 days. Following each predetermined interval period, physical and mechanical evaluations of the immersed blocks were determined. In addition, the change in the hydration products was followed by X-ray diffraction and infrared spectroscopy as nondestructive analyses to recognize the deterioration in the microstructure that may occur due to the flooding event. Thermal analysis and scanning electron microscopy were performed to confirm the data obtained.

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Mallants, Dirk, Karl Travis, Neil Chapman, Patrick V. Brady, and Hefin Griffiths. "The State of the Science and Technology in Deep Borehole Disposal of Nuclear Waste." Energies 13, no. 4 (February 14, 2020): 833. http://dx.doi.org/10.3390/en13040833.

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This Special Issue of the Energies Journal on Deep Borehole Disposal of Nuclear Waste has delivered a timely update on the science and technology of borehole disposal and the types of radioactive wastes it could potentially accommodate. The Special Issue papers discuss (i) circumstances under which a national waste management programme might wish to consider deep borehole disposal; (ii) a status report of deep borehole disposal options in Germany; (iii) the analysis of corrosion performance of engineered barrier systems; (iv) a review of the potential cementing systems suitable for deep borehole disposal; (v) the thermal evolution around heat-generating waste for a wide range of material properties and disposal configurations; (vi) a geochemical analysis of deep brines focussed on fluid-rock interactions; (vii) post-closure performance assessment calculations for deep borehole disposal of Cs/Sr capsules and an example safety case for (viii) horizontal and (ix) vertical deep borehole disposal of nuclear wastes.

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Jiang, Zao, Zhong Hui Xu, Qin Shuai, Ping Li, and Ya Hong Xu. "Thermal Stability of Geopolymer - Sr Contaminated Zeolite A Blends." Key Engineering Materials 727 (January 2017): 1089–97. http://dx.doi.org/10.4028/www.scientific.net/kem.727.1089.

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Geopolymer have gradually become an alternative materials for hazardous and radioactive waste immobilization. However, the compatibility between radioactive waste and geopolymer in various extreme conditions are still not clear. In this study, the thermal stability of geopolymer solidified product of Sr contaminated zeolite A was investigated. For this purpose, the specimens after standard curing were subjected to temperatures of 600, 700 and 800 degrees C for 2h. Freeze-thaw cycles were also employed to test the thermal stability of geopolymer matrices. Meanwhile, changes in the leaching characteristics and physical properties were analyzed. Analysis showed that geopolymer - Sr contaminated zeolite A blends exhibited excellent thermal stability. Although elevated temperature and freeze-thaw cycles led to the deterioration of geopolymer matrices, the leaching rate and mechanical properties of solidified product were still acceptable. The resluts obtained indicated that geopolymer appeared to be very suitable for radioactive waste immobilization.

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Woignier, Thierry, Juan Primera, and Jerôme Reynes. "Nanoporous Glasses for Nuclear Waste Containment." Journal of Nanomaterials 2016 (2016): 1–10. http://dx.doi.org/10.1155/2016/4043632.

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Research is in progress to incorporate nuclear waste in new matrices with high structural stability, resistance to thermal shock, and high chemical durability. Interactions with water are important for materials used as a containment matrix for the radio nuclides. It is indispensable to improve their chemical durability to limit the possible release of radioactive chemical species, if the glass structure is attacked by corrosion. By associating high structural stability and high chemical durability, silica glass optimizes the properties of a suitable host matrix. According to an easy sintering stage, nanoporous glasses such as xerogels, aerogels, and composite gels are alternative ways to synthesize silica glass at relatively low temperatures (≈1,000–1,200°C). Nuclear wastes exist as aqueous salt solutions and we propose using the open pore structure of the nanoporous glass to enable migration of the solution throughout the solid volume. The loaded material is then sintered, thereby trapping the radioactive chemical species. The structure of the sintered materials (glass ceramics) is that of nanocomposites: actinide phases (~100 nm) embedded in a vitreous silica matrix. Our results showed a large improvement in the chemical durability of glass ceramic over conventional nuclear glass.

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Vejmelkova, E., M. Cachova, L. Scheinherrova, P. Konvalinka, M. Keppert, P. Bezdicka, and R. Cerny. "Mechanical and thermal properties of concrete suitable for radioactive waste disposal sites." IOP Conference Series: Materials Science and Engineering 385 (July 2018): 012061. http://dx.doi.org/10.1088/1757-899x/385/1/012061.

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Ghattas, N. K., N. E. Ikladious, and H. A. Shatta. "Radioactive Waste from Nuclear Power Plants Part II. Thermal Properties of Waste/Polymer Products." Isotopenpraxis Isotopes in Environmental and Health Studies 25, no. 3 (January 1989): 98–101. http://dx.doi.org/10.1080/10256018908624068.

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Kuznetsov, D. G., V. V. Ivanov, I. B. Popov, and B. G. Ershov. "Effect of thermal annealing on properties of borobasalt systems containing radioactive waste fractions." Radiochemistry 54, no. 2 (April 2012): 193–97. http://dx.doi.org/10.1134/s1066362212020178.

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Koťátková, Jaroslava, Monika Čáchová, Petr Bezdička, Eva Vejmelková, Petr Konvalinka, Lucie Zemanová, and Robert Černý. "Influence of Supplementary Cementitious Materials on the Properties of Concrete for Secondary Protection Barrier in Radioactive Waste Repositories." Key Engineering Materials 760 (January 2018): 96–101. http://dx.doi.org/10.4028/www.scientific.net/kem.760.96.

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The paper studies the basic physical, mechanical and thermal properties of concrete with three types of supplementary materials for use in waste repositories. The used supplementary materials were natural zeolite, brick dust and metashale. These have pozzolanic properties which modify the parameters of resulting matrix and the unreacted part of the material offers sites for sorption of hazardous species. Results showed rather negative influence of natural zeolite on compressive strength and porosity, as well as thermal properties when compared to brick dust and metashale.

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Bamonte, Patrick, and Pietro Gambarova. "Properties of Concrete Subjected to Extreme Thermal Conditions." Journal of Structural Fire Engineering 5, no. 1 (March 1, 2014): 47–62. http://dx.doi.org/10.1260/2040-2317.5.1.47.

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Durability, high-temperature resistance, impact and blast resilience, radiation-shielding properties, irradiation endurance and - of course - good mechanical properties are required of the cementitious composites to be used in a variety of high-performance structures. Among these, tall buildings, road and railway tunnels, off-shore platforms, gasification plants, wind and solar mills for the production of "clean" energy should be mentioned, as well as nuclear power plants, and radioactive- and hazardous-waste repositories. Hence, understanding, measuring and modelling concrete behavior under extreme environmental conditions is instrumental in making concrete structures safer and more efficient. To this end, the hot and residual properties associated with the exposure to high temperature, fire and thermal shock are treated in this paper. Reference is made to ordinary vibrated concrete (Normal-Strength Concrete - NSC), as well as to a number of innovative cementitious composites, such as Fiber-Reinforced Concrete - FRC, High-Performance/High-Strength Concrete - HPC/HSC, Ultra High-Performance/Very High-Strength Concrete - UHPC /VHSC, Self-Compacting/Consolidating Concrete - SCC, Light-Weight Concrete - LWC, shotcrete and high-strength mortars. It is shown that these materials can be "tailored" according to a variety of requirements and functions, even if several aspects of their behavior (like spalling in fire and long-term mechanical properties under sustained high temperature) are still open to investigation.

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Kuznetsov, D. G., V. V. Ivanov, I. B. Popov, and B. G. Ershov. "Effect of thermal annealing on the properties of boron-containing basalt systems containing radioactive waste components." Radiochemistry 51, no. 1 (February 2009): 73–76. http://dx.doi.org/10.1134/s1066362209010160.

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Jena, Hrudananda, R. Asuvathraman, K. V. Govindan Kutty, and P. R. Vasudeva Rao. "Comparison of electrical conductivity and thermal properties of borosilicate glass with and without simulated radioactive waste." Journal of Thermal Analysis and Calorimetry 115, no. 1 (May 21, 2013): 367–74. http://dx.doi.org/10.1007/s10973-013-3229-6.

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Sau, Núria, Enrique Romero, and Hervé Van Baelen. "Coupled thermo-hydro-mechanical behaviour of a deep clay." E3S Web of Conferences 92 (2019): 10001. http://dx.doi.org/10.1051/e3sconf/20199210001.

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An experimental study on Ypresian clays –one of the potential deep formations in Belgium for the geological disposal of heat-emitting radioactive waste– was undertaken to systematically study its thermal properties and coupled hydro-mechanical response during fast heating pulse tests. An accurate characterisation of the thermal properties is required for assessing the near-field perturbations around disposal galleries that the sedimentary host rock formation will undergo. A new experimental cell adapted to apply the high in situ stresses and with thermal flux sensors was used to directly measure the thermal conductivity at different sample orientations (heat flux orthogonal and parallel to bedding planes). A clear influence of the degree of saturation – despite being close to saturation – and anisotropic features on thermal conductivity have been detected. The study was complemented by performing fast heating pulse tests under constant volume on a new and fully-instrumented axisymmetric cell. The cell allowed recording the pore pressure build-up and dissipation along a heating pulse and under water-undrained conditions.

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Goj, Pawel, Aleksandra Wajda, Agata Stoch, Ireneusz Krakowiak, and Pawel Stoch. "An Insight into the Correlation between Chemical Composition Changes of Aluminum-Iron-Polyphosphate Glasses and Thermal Properties." Materials 14, no. 8 (April 20, 2021): 2065. http://dx.doi.org/10.3390/ma14082065.

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The present study aimed to investigate the influence of the gradual substitution of Fe2O3 by Al2O3 on the thermal properties of polyphosphate glasses. The conducted considerations based on differential scanning calorimetry (DSC) and heating microscopy thermal analysis provided much essential information about the correlation between glass chemical composition and its characteristic parameters, such as transformation temperature, specific heat, crystallization temperature, crystallization enthalpy, the activation energy of crystal growth, melting temperature, and Angell glass thermal stability. The obtained estimation of viscosity changes as a function of temperature could be very helpful for researchers to correctly plan the vitrification process and thus radioactive waste immobilization. A precise analysis of DSC curves and X-ray diffraction patterns revealed the possibility of crystallization process design in order to create materials with different levels of crystallinity and phase composition. The drawn conclusions allow choosing the glass with the optimal concentration of Al2O3 and Fe2O3, which ensures the relatively low melting temperature, viscosity, and glass crystallization ability, with application potential in nuclear waste immobilization.

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Zihms, S. G., and J. F. Harrington. "Thermal cycling: impact on bentonite permeability." Mineralogical Magazine 79, no. 6 (November 2015): 1543–50. http://dx.doi.org/10.1180/minmag.2015.079.6.29.

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AbstractDue to its favourable properties, in particular, low permeability and swelling capacity, bentonite has been favoured as an engineered-barrier and backfill material for the geological storage of radioactive waste. To ensure its safe long-term performance it is important to understand any changes in these properties when the material is subject to heat-emitting waste. As such, this study investigates the hydraulic response of bentonite under multi-step thermal loading subject to a constant-volume boundary condition, to represent a barrier system used in a crystalline or other hard-rock host rock. The experimental set up allows continuous measurement of the hydraulic and mechanical responses during each phase of the thermal cycle. After the initial hydration of the bentonite, the temperature was raised in 20°C increments from 20 to 80°C followed by a final step to reach 120°C. Each temperature was held constant for at least 7–10 days to allow the hydraulic transients to equilibrate. The data suggest that the permeability of bentonite appears to be sensitive to changes in temperature which may extend beyond those explained by simple changes in water viscosity. However, permeability may be boundary-condition dependent and this should be considered when designing experiments or applying these results to other repository host rocks. Either way, the magnitude of the change in permeability observed in this study is minor and its impact on the hydraulic performance of the barrier is negligible.

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Kuznetsov, D. G., V. V. Ivanov, I. B. Popov, and B. G. Ershov. "Effect of thermal annealing on the properties of borophosphate systems containing simulated Cs + Sr radioactive waste fractions." Radiochemistry 56, no. 1 (January 2014): 105–8. http://dx.doi.org/10.1134/s1066362214010202.

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Ivanov, V. V., D. G. Kuznetsov, and I. B. Popov. "Effect of thermal annealing on the properties of borobasalt systems containing simulated radioactive waste from pyroelectrochemical technology." Radiochemistry 59, no. 4 (July 2017): 402–6. http://dx.doi.org/10.1134/s1066362217040130.

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Mohapatra, M., and B. S. Tomar. "Spectroscopic Investigations of Radiation Damage in Glasses Used for Immobilization of Radioactive Waste." Defect and Diffusion Forum 341 (July 2013): 107–28. http://dx.doi.org/10.4028/www.scientific.net/ddf.341.107.

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Borosilicate based glass formulations have been found suitable for vitrification of high level nuclear waste (HLW) generated during the reprocessing of spent nuclear fuel from nuclear reactors. These glasses possess desirable properties like high chemical, mechanical, thermal and radiation stability for HLW storage. Also, the amorphous nature of the glass helps accommodate the waste containing a variety of elements easily. Because of the presence of the radioactive components, such as, fission /activation products and minor actinides present in the waste, the glass containment experiences radiation damage that can significantly alter the glass structure which may influence their long term leaching behavior. Spectroscopic techniques provide direct and non-invasive method for investigating this radiation damage in the glasses. The present paper gives a glimpse of the current status and issues regarding the investigation of radiation damage in the glass matrices.

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Finsterle, Stefan, Richard Muller, Rod Baltzer, Joe Payer, and James Rector. "Thermal Evolution near Heat-Generating Nuclear Waste Canisters Disposed in Horizontal Drillholes." Energies 12, no. 4 (February 13, 2019): 596. http://dx.doi.org/10.3390/en12040596.

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We consider the disposal of spent nuclear fuel and high-level radioactive waste in horizontal holes drilled into deep, low-permeable geologic formations using directional drilling technology. Residual decay heat emanating from these waste forms leads to temperature increases within the drillhole and the surrounding host rock. The spacing of waste canisters and the configuration of the various barrier components within the horizontal drillhole can be designed such that the maximum temperatures remain below limits that are set for each element of the engineered and natural repository system. We present design calculations that examine the thermal evolution around heat-generating waste for a wide range of material properties and disposal configurations. Moreover, we evaluate alternative layouts of a monitoring system to be part of an in situ heater test that helps determine the thermal properties of the as-built repository system. A data-worth analysis is performed to ensure that sufficient information will be collected during the heater test so that subsequent model predictions of the thermal evolution around horizontal deposition holes will reliably estimate the maximum temperatures in the drillhole. The simulations demonstrate that the proposed drillhole disposal strategy can be flexibly designed to ensure dissipation of the heat generated by decaying nuclear waste. Moreover, an in situ heater test can provide the relevant data needed to develop a reliable prediction model of repository performance under as-built conditions.

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Nikitin, A. N., O. A. Pocheptsova, and S. Matthies. "Aspects of the thermal and transport properties of crystalline salt in designing radioactive waste storages in halogen formations." Crystallography Reports 55, no. 3 (May 2010): 450–57. http://dx.doi.org/10.1134/s1063774510030144.

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Romanenko, Iryna, Maryna Holiuk, Pavlo Kutsyn, Iryna Kutsyna, Hennadii Odynokin, Anatolii Nosovskyi, Vitalii Pastsuk, et al. "New composite material based on heavy concrete reinforced by basalt-boron fiber for radioactive waste management." EPJ Nuclear Sciences & Technologies 5 (2019): 22. http://dx.doi.org/10.1051/epjn/2019050.

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A new composite material with neutron radiation shielding properties is presented. This fiber reinforced concrete material incorporates basalt-boron fiber, with different concentrations of boron oxide in fiber, and is applicable to nuclear energy and nuclear waste management. The methodology for production of boron oxide (B2O3) infused basalt fiber has been developed. First experimental samples of basalt boron fiber containing 6% of B2O3 and 12% B2O3 have been produced in laboratory conditions. The concrete samples reinforced by two types of basalt-boron fiber with different dosages have been prepared for neutron experiment. The neutron experimental investigations on radiation shielding properties of concrete reinforced by basalt-boron fiber have been performed by means of Pu-Be neutron source. The prepared samples have been tested in the course of several series of tests. It is shown that basalt-boron fibers in concrete improve neutron radiation shielding properties for neutrons with different energies, but it appears to be most effective when it comes to thermal neutrons.

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Bukaemskiy, Andrey, Johannes Fachinger, and Dirk Bosbach. "Synthesis and Properties of Reaction-Bonded SiC Ceramic with Embedded UO₂ - TRISO Coated Particles." Advances in Science and Technology 73 (October 2010): 136–41. http://dx.doi.org/10.4028/www.scientific.net/ast.73.136.

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Ceramics based on SiC are characterized by extreme hardness, high thermal conductivity, relatively low thermal expansion and chemical durability. In principle, SiC ceramics can be considered as a long-term stable matrix for final disposal of radioactive waste, such as coated fuel particles (CP) separated from the graphite matrix from spent HTR (high-temperature gas-cooled reactor) fuel pebbles. In the present work, SiC-based ceramic with the embedded UO2 - TRISO (tristructural-isotropic) coated particles was synthesized by the reaction-bonding process. The synthesis was performed in standard SiC crucible. Several physico-mechanical properties of the synthesized samples were investigated. It was shown that the coated particles in the reaction-bonded silicon carbide (RBSiC) matrix are distributed homogeneously. The amount of pores is insignificant and the crippling of the coated particles is not observed. Besides, the junction between CP and RBSiC matrix and between RBSiC matrix and the SiC crucible is very good. For all ceramic components of the synthesized samples, namely, for the UO2-kernels, SiC-layers, SiC crucible wall and for the synthesized RBSiC ceramics, the values of microhardness and fracture toughness were measured and compared with the reference data. The strength properties, such as tensile strength of the synthesized samples, failure mechanism of the reaction-bonded SiC ceramic with embedded UO2 - TRISO coated particles, microstructure of the fracture surfaces, the peculiarities of the coated particles fracture were investigated in detail. Moreover, the diffusion of radioactive tracers (137Cs, 241Am, 36Cl, 3H) across synthesized ceramic was studied and the high safety characteristics of the synthesized ceramic were demonstrated.

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Golik, Vladimir I., Iurii I. Razorenov, Vladimir S. Vagin, and Vasilii I. Liashenko. "Study and development of hardening mixture composition based on unconventional industrial waste." Izvestiya vysshikh uchebnykh zavedenii. Gornyi zhurnal 1, no. 3 (May 14, 2021): 13–27. http://dx.doi.org/10.21440/0536-1028-2021-3-13-27.

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Relevance. Phosphogypsum, the product of apatites chemical processing, is one of the most common mining wastes. Phosphogypsum utilization is not widespread yet, therefore its cost is low. Integrated research has been carried out to determine the technological capacity and economic feasibility of phosphogypsum, sludge, and dolomite utilization as binders in the conversion of ore production techniques. Research aim is to develop a hardening mixture composition based on unconventional industrial waste, determine the technological capacity and economic feasibility of utilizing phosphogypsum, sludge, dolomite, and other accessible low activity wastes as a substitute for expensive and relatively scarce binder material. Research methodology. Initial data are studied of the wastes possessing binding properties. The efficiency of admixing them is determined from the robustness of hardening backfill mixtures control samples that have been produced in laboratory conditions. Based on the research, a database is created to apply the results in practice for mining development. Results. The hardening materials compositions were obtained based on unconventional industrial waste including hydrometallurgical and dressing tailings, furnace clinker, low-grade sand, thermal power plants (TPP) and chemical industry ashes. The optimal composition of the mixture per cubic meter: tailings – 600–750 kg; TPP ash – 180–220 kg; cement dust – 250–315 kg; cement – 35–40 kg; tempering water – 450–515 l under the mixture’s fluidity of about 14 cm according to the readings of the mortar consistency measuring device (StroyTsNIL cone). To ensure radiation safety of the hardening mixture that is based on unconventional industrial waste, it is advisable to take into account not only their chemical and physical-mechanical indicators but the value of naturally radioactive nuclides’ effective activity as well. Conclusions. It has been stated that the robustness of mixtures containing gypsum is 1.5–2.0 times higher, and under the hardening time of 3, 6 and 12 months makes up 3.1; 5.7 and 7.6 MPa correspondingly. It has been shown that the compositions with the binder’s flow rate of 450 kg per cubic meter under the cement : sludge ratio of 1 : 2 show the robustness from 2.8 to 4.9 MPa in 28 days. The content of low-grade sand levigate particles reaches 20% and more. Classes with a specific area of 28.4 m2/kg refer to fine sand, and with a specific area of 27.7 m2/kg refer to medium sand. The robustness of the 28 days old composition reaches 0.5 MPa, 90 days – 0.9 MPa depending on the cement flow rate. It has been substantiated that binders based on fluorine gypsum, phosphogypsum, and belite sludge by mixed grinding of the granulated 26 "Izvestiya vysshikh uchebnykh zavedenii. Gornyi zhurnal". No. 3. 2021 ISSN 0536-1028 blast-furnace slug with the ferrochrome sludge and phosphogypsum up to 70%, 0.08 mm size, show the robustness of the binder up to 3.0 MPa with the flow rate of 450 kg per a cubic meter of the mixture. Keywords: industrial waste; hardening mixture; binder; PPT ash; ash and slag; cement; fluorine gypsum; phosphogypsum; non-reactive aggregate; belite sludge; naturally radioactive nuclides. Acknowledgements. Specialists from Platov South-Russian State Polytechnic University (Novocherkassk, Russia), Ukrainian Research and Design Institute of Industrial Technology (Zhovti Vody, Ukraine), Vostochnyy (Eastern) Mining and Enrichment Combine (Zhovti Vody, Ukraine), Dnipro University of Technology (Dnipro, Ukraine) and others took part in creating, improving, and introducing R&D.

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Delage, Pierre, Nabil Sultan, and Yu Jun Cui. "On the thermal consolidation of Boom clay." Canadian Geotechnical Journal 37, no. 2 (April 1, 2000): 343–54. http://dx.doi.org/10.1139/t99-105.

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When a mass of saturated clay is heated, as in the case of host soils surrounding nuclear waste disposal at great depth, the thermal expansion of the constituents generates excess pore pressures. The mass of clay is submitted to gradients of pore pressure and temperature, hydraulic and thermal flows, and changes in its mechanical properties. In this work, some of these aspects were experimentally studied in the case of Boom clay to help predict the response of the soil, in relation to investigations in the Belgian underground laboratory at Mol. Results of slow-heating tests with careful volume change measurements showed that a reasonable prediction of the thermal expansion of the clay-water system was obtained by using the thermal properties of free water. Despite the density of Boom clay, no significant effect of water adsorption was observed. The thermal consolidation of Boom clay was studied through fast-heating tests. A simple analysis shows that the hydraulic and thermal transfers are uncoupled. Experimental results from fast-heating tests showed that the consolidation coefficient does not change significantly with increased temperature, due to the opposite effect of increasing permeability and decreasing porosity. The changes of permeability with temperature were investigated by running constant head measurements at various temperatures. An indirect analysis, based on estimation of the coefficient of volume change mv, showed that the indirect method of estimating the permeability from consolidation tests should be considered carefully. Intrinsic permeability values were derived by considering the change of the viscosity of free water with temperature. A unique relationship between the intrinsic permeability and the porosity was observed, with no dependence on temperature, confirming that the flow involved in the permeability test only concerns free water.Key words: clays, thermal consolidation, adsorbed water, permeability, temperature effects, radioactive waste disposal.

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Skvortsov, I. V., V. V. Kalistratova, E. V. Belova, A. V. Rodin, I. P. Sokolov, and B. F. Myasoedov. "Thermal properties of 2,2′-bipyridine-6,6′-dicarboxylic acid bis(N-ethyl-4-hexylanilide), an extractant for radioactive waste components." Radiochemistry 59, no. 6 (November 2017): 612–17. http://dx.doi.org/10.1134/s1066362217060091.

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Jobmann, M., and A. Meleshyn. "Evaluation of temperature-induced effects on safety-relevant properties of clay host rocks with regard to HLW/SF disposal." Mineralogical Magazine 79, no. 6 (November 2015): 1389–95. http://dx.doi.org/10.1180/minmag.2015.079.6.14.

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AbstractDBE TECHNOLOGY, BGR and GRS are developing a methodology to demonstrate the safety of a repository for high-level waste and spent fuel (HLW/SF) in clays according to the requirements of the German regulating body. In particular, these requirements prescribe that the barrier effect of host rocks must not be compromised by a thermal impact resulting from HLW/SF emplacement. To substantiate and quantify this requirement, we carried out a literature survey of research on thermally-induced changes on clay properties. Effects thus compiled can be divided into thermo-hydro-mechanical and chemical-biological-mineralogical effects and were analysed with regard to their relevance to the integrity of clay host rocks. This analysis identified one effect of major influence within each group: thermal expansion and compaction as well as results of microbial activities. Importantly, it further revealed that a moderate temperature increase above 100°C cannot be expected to compromise the integrity of the geological barrier according to the current knowledge state. Evidence is presented in this paper that temperature increases up to 150°C can actually contribute to an improved performance of a radioactive waste repository by increasing the consolidation of the clay and sterilizing the repository's near-field to depress the deteriorative microbial effects. A quantitative temperature criterion for thermal impact of HLW/SF on clay host rocks is accordingly proposed.

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Holdsworth, Eccles, Rowbotham, Brookfield, Collison, Bond, Kavi, and Edge. "The Effect of Gamma Irradiation on the Physiochemical Properties of Caesium-Selective Ammonium Phosphomolybdate–Polyacrylonitrile (AMP–PAN) Composites." Clean Technologies 1, no. 1 (September 19, 2019): 294–310. http://dx.doi.org/10.3390/cleantechnol1010020.

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Managing certain by-products of the nuclear fuel cycle, such as the radioactive isotopes of caesium: 134Cs, 135Cs and 137Cs is challenging due to their environmental mobility and radioactivity. While a great many materials can isolate Cs+ ions from neutral or basic aqueous solutions via ion exchange, few of these, with the exception of ammonium phosphomolybdate (AMP), function effectively in acidic media. The use of AMP, and its porous composite in polyacrylonitrile (PAN) for management of Cs radioisotopes in various nuclear wastes have been known for decades and are well studied, yet the effects of radiation on the physiochemical properties of such composites have only received limited attention to date. In a previous publication, we demonstrated that a 100 kGy gamma irradiation dose has negligible effect on the ion exchange performance of AMP and AMP–PAN with respect to capacity or kinetics under the Cs+ concentrations and acidity found in spent nuclear fuel (SNF) recycling. As a continuation of this prior study, in this publication we explore the effects of gamma irradiation on the physiochemical properties of AMP and AMP–PAN using a range of characterisation methods. The effects of the same gamma dose on the oxidation state of Mo in AMP and AMP–PAN, the thermal degradation of both AMP and AMP–PAN, combined with a first study into the high-temperature degradation AMP, are reported. The implications of irradiation, its possible mechanism, the conditions present in SNF recycling, and for the end-of-life disposal or recycling of these materials are also discussed.

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Motoshima, Takayuki, Sachie Iso, and Tomoyoshi Nishimura. "Behavior of compacted Ca-bentonite subjected to HMC loading: observations and interpretation." MATEC Web of Conferences 337 (2021): 04004. http://dx.doi.org/10.1051/matecconf/202133704004.

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This study presents result of HMC experimental test and discussions of chemical effort in Cabentonite. SWCC test, swelling pressure test and unconfined compression test as hydro-mechanical-chemical (HMC) test were applied to Cabentonite-sand mixture. Na-type bentonite has been studied to predict safety and confidence for artificial barrier layer consist of bentonite that many experimental research reports are represented, and can offer technology advanced comprehensive performance due to mathematical simulation models associated to physical parameters on thermal-hydration-mechanical-chemical properties. This study proposed a significant property for calcium bentonite to relate evaluating safety in radioactive waste disposal system, which emphasised much influence of suction and salinity chloride to hydration-mechanical behaviour for Calcium bentonite.

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Valter, M., and M. Plötze. "Characteristics of variably saturated granular bentonite after long-term storage at near-field relevant temperatures." Clay Minerals 48, no. 2 (May 2013): 343–61. http://dx.doi.org/10.1180/claymin.2013.048.2.14.

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AbstractBentonite is a potential material for use in the engineered barrier of radioactive waste repositories because of its low hydraulic permeability, self-sealing capability and retention capacity. It is expected that bentonite would react at the elevated temperatures accompanying the radioactive decay in the nuclear waste. The presented study was started in order to improve understanding of the coupled influence of temperature and (pore) water on the physicochemical and mineralogical properties of bentonite during thermal treatment under near-field relevant conditions. Granular Na-bentonite MX-80 was differently saturated (Sr = 1–0.05) and stored at different temperatures (50–150°C) in a closed system. Upon dismantling after different periods of time (3 to 18 months), mineralogical characteristics, cation exchange capacity and content of leachable cations, as well as physicochemical properties such as surface area and water adsorption were investigated.The results showed a high mineralogical stability. A slight conversion from the sodium to an earth alkali form of the bentonite was observed. However, considerable changes in the physicochemical properties of the bentonite were observed, particularly by treatment above the critical temperature of 120°C. The cation exchange capacity decreased during heating at 150°C by approximately. 10%. The specific surface area dropped by more than 50%. The water uptake capacity under free swelling conditions showed a slight tendency to lower values especially for samples heated for more than 12 months. The water vapour adsorption ability in contrast drops by 25% already within three months at T = 120°C. These changes are mostly related to the variations in the interlayer cation composition and to smectite aggregation processes. The observed alterations are rather subtle. However, temperatures ⩾ 120°C had a remarkable negative influence on different properties of MX-80.

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Amakawa, Tadashi, Kazuo Adachi, and Shinji Yasui. "Fundamental Research on Thermal Plasma Technology for Treatment of Low Level Radioactive Solid Waste Part II Influence of Waste Composition on Material Properties of Solidified Products." IEEJ Transactions on Fundamentals and Materials 119, no. 3 (1999): 357–64. http://dx.doi.org/10.1541/ieejfms1990.119.3_357.

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Liang, Haian, Tan Tang, Longpeng Zhang, Xiaodong Liu, Shuai Liu, Ting Yang, Xinjun Cheng, and Juan Zhang. "Stability Analysis of Host Claystone under T-M Coupling of HLW Disposal Repository in China: Experiments and Numerical Simulation." Geofluids 2021 (September 15, 2021): 1–11. http://dx.doi.org/10.1155/2021/8118021.

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This paper focuses on the stability of the high-level radioactive waste (HLW) in the proposed clay rock in Tamusu area of China. The in-situ stress as well as the variational characteristics of ambient temperature caused by nuclide decay during HLW storage should be noticeable. A series of thermal property tests and thermo-mechanical coupled strength (T-M) tests of rock samples in the target formation are carried out. Then the stability of surrounding rock of an HLW under the combination of heat release from HLW and in-situ stress is simulated and analyzed by numerical method. Thermal properties of Tamusu clay rock samples are obtained by testing their thermal conductivity. In order to obtain the characteristics and the failure modes of rock samples at different temperatures, the T-M coupling experiments in the temperature range of 100°C are conducted. Numerical model for simulating the state of operation of the nuclear waste tank buried in the tunnel within 100 years is constructed. A thermal boundary by the heat release equation of HLW and the real in-situ stress level in Tamusu area are considered in the model. While, the variation law of surrounding rock’s temperature, stress, and deformation corresponding to the embedding time is obtained from the numerical calculation. Finally, the stability of the deep geological repository is comprehensively evaluated. The results show that the temperature has a significant impact on the T-M coupling characteristics of Tamusu clay rock, and the proposed repository numerical model has no large deformation and failure problems in 100 years. However, the temperature of the surrounding rock of the repository may exceed the safety standard value during the operation period.

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Laufek, František, Irena Hanusová, Jiří Svoboda, Radek Vašíček, Jan Najser, Magdaléna Koubová, Michal Čurda, et al. "Mineralogical, Geochemical and Geotechnical Study of BCV 2017 Bentonite—The Initial State and the State following Thermal Treatment at 200 °C." Minerals 11, no. 8 (August 12, 2021): 871. http://dx.doi.org/10.3390/min11080871.

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Bentonites are considered to be the most suitable materials for the multibarrier system of high-level radioactive waste repositories. Since BCV bentonite has been proved to be an ideal representative of Czech Ca-Mg bentonites in this respect, it has been included in the Czech Radioactive Waste Repository Authority (SÚRAO) buffer and backfill R&D programme. Detailed knowledge of processes in the material induced by thermal loading provides invaluable assistance regarding the evolution of the material under repository conditions. Samples of both original BCV 2017 bentonite and the same material thermally treated at 200 °C were characterised by means of chemical analysis, powder X-ray diffraction, infrared spectroscopy, thermal analysis, cation exchange capacity, specific surface area (BET) measurements, the determination of the swell index, the liquid limit, the swelling pressure and water retention curves. The smectite in BCV 2017 bentonite comprises Ca-Mg montmorillonite with a significant degree of Fe3+ substitution in the octahedral sheet. Two main transformation processes were observed following heating at 200 °C over 27 months, the first of which comprised the dehydration of the montmorillonite and the subsequent reduction of the 001 basal distance from 14.5 Å (the original BCV 2017) to 9.8 Å, thus indicating the absence of water molecules in the interlayer space. The second concerned the dehydration and partial dehydroxylation of goethite. With the exception of the dehydration of the interlayer space, the PXRD and FTIR study revealed the crystallochemical stability of the montmorillonite in BCV 2017 bentonite under the selected experimental conditions. The geotechnical tests indicated no major changes in the mechanical properties of the thermally treated BCV 2017 bentonite, as demonstrated by the similar swelling pressure values. However, the variation in the swell index and the gradual increase in the liquid limit with the wetting time indicated a lower hydration rate. The retention curves consistently showed the lower retention capacity of the thermally treated samples, thus indicating the incomplete re-hydration of the thermally treated BCV 2017 exposed to air humidity and the difference in its behaviour compared to the material exposed to liquid water.

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Romanenko, I., M. Holiuk, A. Nosovsky, T. Vlasenko, and V. Gulik. "New Composite Material Based on Heavy Concrete and Basalt-Boron Fiber for Neutron Radiation Shielding Properties." Nuclear and Radiation Safety, no. 2(82) (June 6, 2019): 19–25. http://dx.doi.org/10.32918/nrs.2019.2(82).04.

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It is necessary to have reliable radiation protection for safe operation of different radiation sources. Radiation shielding properties have been studied for a long time both in our country and abroad. However, there is a strong necessity to develop new composite materials, which will provide protection against radiation and have improved mechanical and economic characteristics. The paper describes a new composite material for neutron radiation shielding properties based on heavy concrete with serpentinite aggregate and with basalt-boron fiber with different concentrations of fiber boron oxide for using in biological shielding in nuclear industry. Protective properties of the new composite material were investigated with different neutron sources: 1) neutrons with 14 MeV energy; 2) fast fission neutrons for U-235; 3) fast fission neutrons for U-235 after passing a water layer. The simulation of the neutron radiation in presented composite material with adding crushed stone aggregate and serpentinite aggregate is performed by Monte Carlo Serpent code. It is shown that basalt-boron fibers in concrete improve the protective properties of concrete against neutron irradiation for neutrons with different energies, but the most effective is the addition of a basalt-boron fiber in the case of thermal neutrons. This research was supported by Horizon 2020 ERA-NET Support Programme, Research Grant Agreement No 7.9-3/18/7 (“Development of Boron-Infused Basalt-Fiber Reinforced Concrete for Nuclear and Radioactive Waste Management Applications”). Implementation of activities described in the Roadmap to Fusion during Horizon 2020 through a joint programme of the members of the EUROfusion consortium (2014-2020), Work Package PMI. Also, this research was carried out with the financial support of the IAEA, within the terms and conditions of the Research Contract20638 in the framework of the Coordinated Research Project (CRP) “Accelerator Driven Systems (ADS) Applications and Use of Low-Enriched Uranium in ADS (T33002)’’ within the Project “The Two-Zone Subcritical Systems with Fast and Thermal Neutron Spectra for Transmutation of Minor Actinides and Long-Lived Fission Products”.

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Deptuła, A., J. Chwastowska, Wiesława Łada, T. Olczak, D. Wawszczak, E. Sterlinska, B. Sartowska, and K. C. Goretta. "Sol-Gel-Derived Hydroxyapatite and its Application to Sorption of Heavy Metals." Advances in Science and Technology 45 (October 2006): 2198–203. http://dx.doi.org/10.4028/www.scientific.net/ast.45.2198.

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Hydroxyapatite (HA) microspheres of diameter <70 μm have been synthesized by solgel processing. The starting sols were prepared by ultrasonic mixing of concentrated solutions of calcium acetate (1.7M) with 85% H3PO4, followed by emulsification in dehydrated 2-ethyl-1- hexanol. Drops of emulsion were solidified by extraction of water with this solvent. The final thermal treatment was a 2 h soaking in air at 900°C. Properties such as hydraulic resistance and sedimentation rate, which are important for application in ion-exchangers, were superior for our prepared microspheres in comparison with irregularly shaped commercial HA. Adsorption of the following metals was investigated: U, Zn, Fe, Cu, Ni, Co, Cd, Pb, Mn, Al, Cr, As, Sb, Bi and Mo. Retention was ≈100% for of all the metals studied when pH > 3–4. On average, ≈1/20 moles of metal reacted with 1 mole of HA. Adsorbed metals could be desorbed with efficiencies of 60–90%. In addition, nuclear-waste-saturated beds of HA could be transformed to insoluble ceramics by thermal treatment. The cost of producing HA microspheres was estimated to be comparable to current prices of irregularly shaped commercial hydroxyapatite powders. For radioactive contaminants such as U, for which resorption need not be considered, porous monoliths were produced by use of industrial reagents. The retention capacity was determined to be 30 mg of U per 1 mg of monolith.

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Rowe, R. Kerry, F. B. Abdelaal, M. Zafari, M. S. Morsy, and D. G. Priyanto. "An approach to high-density polyethylene (HDPE) geomembrane selection for challenging design requirements." Canadian Geotechnical Journal 57, no. 10 (October 2020): 1550–65. http://dx.doi.org/10.1139/cgj-2019-0572.

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An approach for selecting a high-density polyethylene (HDPE) geomembrane (GMB) for a long design life is described and illustrated for five 2 mm thick textured GMBs when immersed in a simulated municipal solid waste leachate (L3) and two simulated leachates representative of low-level radioactive waste leachates (L7 and L9) for 9–16 months at a range of temperatures. Although made from the same nominal resin, substantial differences are reported in both the initial properties and the rate of antioxidant depletion for the five GMBs. At an expected operating liner temperature of 10 °C and immersed in L3, the projected time to antioxidant depletion for the five GMBs ranges from 125 to over 2000 years. The antioxidant depletion in leachates L7 and L9 were similar or slower than in leachate L3. There was no evidence of traditional thermal-oxidative degradation reported over the 9–16 months of monitoring; however, there was a significant reduction in stress crack resistance due to physical ageing ranging between 30% and 70% of the initial value. Two GMBs are considered highly likely to have service life well in excess of the required design life of 550 years. It is suggested that the proposed approach could be adopted for selecting GMBs for other projects that require a long design life.

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Nishimura, Tomoyoshi, and Masaaki Fukaya. "Evaluation of direct shear strength of compacted bentonite having pore-water pressure." E3S Web of Conferences 195 (2020): 02028. http://dx.doi.org/10.1051/e3sconf/202019502028.

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Safety of great deep repository design has been investigated for high-level radioactive waste disposal system in several countries such as Belgium, Canada, China, France, Germany, Japan, Sweden and Switzerland. The repository of the disposal is in most cases based on the concept of a multi-barrier system using the host rock barrier formation and a man-made barrier formation. The man-made barrier consists of high expansive bentonite. Thermal-hydro-mechanical behaviour simulation models were developed, including some parameters described by experimental works. The complex phenomena due to the transition into saturation and chemical reactions at the bentonite barrier system have been explained. This study focused on direct shear strength for compacted bentonite related to some factors induced by uncertainty problems such as hydration effect and pore-water pressure. Measured shear strength properties of compacted bentonite had been determined at high suction values. Also, it is clear that there is some influence of direct shear speed on direct shear strength both under unsaturated-saturated conditions. A modified direct shear apparatus was used in this study to observe changes in shear strength with increments of pore-water pressures.

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Kashcheev, V. A., P. P. Polu�ktov, and A. S. Polyakov. "Thermal-diffusion aging in solidified radioactive wastes." Soviet Atomic Energy 60, no. 2 (February 1986): 173–76. http://dx.doi.org/10.1007/bf01371185.

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Moiseenko, E. V., N. I. Drobyishevsky, R. A. Butov, and Yu N. Tokarev. "Concept of Large-scale Thermomechanical URL Experiments in the Nizhnekanskiy Rock Massif." Radioactive Waste 12, no. 3 (2020): 101–11. http://dx.doi.org/10.25283/2587-9707-2020-3-101-111.

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Numerical simulation of thermomechanical processes in a deep underground radioactive waste repository requires information on the host rock and the engineered barriers properties at a scale of dozens of centimeters, meters and more. However, the extrapolation of the values obtained on small-scale samples in surface laboratories yields excessive uncertainties. The materials behavior is also influenced by conditions that cannot be reliably reproduced in a surface laboratory, such as water content or initial stress-strain state. Following experiments are planned to study the host rock and the engineered barriers behavior during heating under conditions similar to those expected in the repository, as well as to assess their large-scale thermomechanical properties. In the experiment focused on the excavation damaged zone thermal mechanics, the behavior of reinforced drift walls and vaults under heating will be studied. The experimental facility will involve two drifts with the same orientation as the planned repository ones. As a result, the spatial distribution of excavation damaged zone thermomechanical parameters and their evolution due to heating will be identified. The second experiment focuses on the host rock mass behavior under spatially nonuniform unsteady heating. The facility will feature two vertical boreholes with heaters. The experiment will be divided into several stages: study of the host rock initial state, estimation of the rock main thermomechanical properties, study of the temporal evolution of the stress field due to 3D temperature gradients and of the processes in the host rock occurring during its cooling and re-saturation with water. Following the completion of the separate-effect test program, an integrated experiment should be carried out to study the coupled processes with respect to their mutual influence. The obtained results will be used to refine the values of input parameters for numerical simulations and their uncertainty ranges, as well as to validate the computer codes.

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Doudou, S., E. K. Swain-Phipps, A. J. Fuller, S. M. Wickham, N. Daniels, E. Fourcy, M. Fournier, et al. "Strategic Study of Thermal Treatment of European Radioactive Wastes." IOP Conference Series: Materials Science and Engineering 818 (April 23, 2020): 012002. http://dx.doi.org/10.1088/1757-899x/818/1/012002.

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TANG, Y. S. "EXPERIENCE IN THERMAL TREATMENT OF LOW-LEVEL RADIOACTIVE WASTES." Chemical Engineering Communications 103, no. 1 (April 1991): 83–98. http://dx.doi.org/10.1080/00986449108910864.

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Mabrouk, M., F. Lemont, and J. M. Baronnet. "Incineration of radioactive organic liquid wastes by underwater thermal plasma." Journal of Physics: Conference Series 406 (December 17, 2012): 012002. http://dx.doi.org/10.1088/1742-6596/406/1/012002.

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Yang, Junqiang, Keliang Shi, Xuejie Sun, Xiaoqing Gao, Peng Zhang, Zhiwei Niu, and Wangsuo Wu. "An approach for the efficient immobilization of 79Se using Fe-OOH modified GMZ bentonite." Radiochimica Acta 108, no. 2 (January 28, 2020): 113–26. http://dx.doi.org/10.1515/ract-2019-3151.

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AbstractBecause of high mobility, the immobilization of long-lived fission product 79Se (often existed as 79Se(IV) and 79Se(VI) anions) is a critical consideration in the repository of high-level radioactive waste. In this work, a Fe-OOH modified bentonite (Fe-OOH-bent) was synthesized as a potential filling material in the repository site for effective adsorption and present the migration of different species of 79Se. The adsorbent was characterized using FT-IR, XRD, XFS, zeta potential and BET to clarify its physical properties, compositions and structures. A good thermal and radiation stabilities of Fe-OOH-bent was confirmed by its stable uptake ratio for Se(IV) and Se(VI) compared to original samples. The batch experimental results show that Se(IV) and Se(VI) can be efficiently removed from aqueous by Fe-OOH-bent within 60 min with maximum adsorption capacities of 68.45 mg/g for Se(IV) and 40.47 mg/g for Se(VI) in the optimal conditions, indicating its high potential application in consideration of its simple synthesis process, low cost and high adsorption capacity in view of immobilization of 79Se. The surface species and variation of oxide state of Fe as well as Se(IV) and Se(VI) onto Fe-OOH-bent were investigated by XPS analysis. The values of relative area of Se(IV)–O and Se(VI)–O in XPS spectra followed the same tendency as their adsorption ratio with the variation of system pH, suggesting that the formation of complexes between selenium species and Fe-OOH-bent surface.

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Nieminen, Matti, Markus Olin, Jaana Laatikainen-Luntama, Stephen M. Wickham, Slimane Doudou, Adam J. Fuller, Jenny Kent, et al. "Thermal treatment for radioactive waste minimisation." EPJ Nuclear Sciences & Technologies 6 (2020): 25. http://dx.doi.org/10.1051/epjn/2019040.

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Safe management of radioactive waste is challenging to waste producers and waste management organisations. Deployment of thermal treatment technologies can provide significant improvements: volume reduction, waste passivation, organics destruction, safety demonstration facilitation, etc. The EC-funded THERAMIN project enables an EU-wide strategic review and assessment of the value of thermal treatment technologies applicable to Low and Intermediate Level waste streams (ion exchange media, soft operational waste, sludges, organic waste, and liquids). THERAMIN compiles an EU-wide database of wastes, which could be treated by thermal technologies and documents available thermal technologies. Applicability and benefits of technologies to the identified waste streams will be evaluated through full-scale demonstration tests by project partners. Safety case implications will also be assessed through the study of the disposability of thermally treated waste products. This paper will communicate the strategic aims of the ongoing project and highlight some key findings and results achieved to date.

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Yudintsev, S. V., and A. A. Shiryaev. "Thermal Stability of Glass with Simulators of Chloride Highly Radioactive Wastes." Doklady Physics 63, no. 12 (December 2018): 513–16. http://dx.doi.org/10.1134/s1028335818120042.

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Cagnon, Hugo, Thierry Vidal, Alain Sellier, and Jean Michel Torrenti. "Transient Thermal Creep at Moderate Temperature." Key Engineering Materials 711 (September 2016): 885–91. http://dx.doi.org/10.4028/www.scientific.net/kem.711.885.

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The Cigéo project initiated by Andra (the French national agency for nuclear wastes management) concerns a deep geological disposal situated in an impermeable layer of clay approximately 500 m under ground. The High Performance Concrete (HPC) galleries will allow Intermediate Level Long-Life Nuclear Wastes to be stored. These exothermic radioactive wastes will induce a temperature rise which may reach 70°C. The very few studies that have investigated the effect of temperatures below 100°C on concrete behaviour have shown that temperature amplifies long-term deformations more than instantaneous ones. The present paper deals with the dimensional variations of HPC when subjected to temperature changes in sealed conditions. The aim is to provide accurate data to understand the causes of transient thermal creep. These experimental data may allow Thermo-Hydro-Mechanical models to be improved and thus provide accurate prediction of concrete long term behavior in such in situ conditions.

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Yudintsev, S., and A. Shiryayev. "Thermal Stability of Glass with Imitators of Chloride High Level Radioactive Wastes." Доклады академии наук 483, no. 4 (December 2018): 385–88. http://dx.doi.org/10.31857/s086956520003272-2.

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YAMAZAKI, SEIIRO. "Industrial applications of thermal plasma. 5. Treatment of low level radioactive wastes using thermal plasma." Journal of the Institute of Electrical Engineers of Japan 123, no. 2 (2003): 93–95. http://dx.doi.org/10.1541/ieejjournal.123.93.

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Enes, Teresa, José Aranha, Teresa Fonseca, Domingos Lopes, Ana Alves, and José Lousada. "Thermal Properties of Residual Agroforestry Biomass of Northern Portugal." Energies 12, no. 8 (April 12, 2019): 1418. http://dx.doi.org/10.3390/en12081418.

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Biomass from forestry and agricultural sector provides an important contribution to encounter the government’s targets for increasing bioenergy production and utilization. Characterization of agricultural and forest wastes are critical for exploiting and utilizing them for energy purpose. In the present work agricultural and forest wastes and shrubs were sampled in two sites in north Portugal (Ave and Sabor basin) and subjected to Higher Heating Value (HHV) and chemical composition quantification. The HHV was evaluated according to the methodology described in Standard DD CEN/TS14918:2005. For the lignin content, the procedure was made by the Klason method and the extractives content was determined with the Soxhlet method. For agricultural and forest wastes the HHV values are identical with a range of 17 to 21 MJ·kg−1. However, shrubs biomass presentx slightly higher and statistically different values from agricultural and forest wastes, varying between 19 and 21 MJ·kg−1. Forest wastes contain higher levels of holocellulose compared to agricultural wastes and, with respect to extractive contents, this trend is the reverse. There is a general tendency for the woody components present thermo-chemical properties more suited for energy purposes, than the residues formed by the branches and leaves.

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Ghiloufi, Imed, and Jacques Amouroux. "Electrolysis Effects on the cesium Volatility DURING thermal Plasma Vitrification of radioactive Wastes." High Temperature Material Processes (An International Quarterly of High-Technology Plasma Processes) 14, no. 1-2 (2010): 77–88. http://dx.doi.org/10.1615/hightempmatproc.v14.i1-2.60.

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YAMAKAWA, Hidetsugu, and Yutaka KOUNO. "Thermal Integrity of Packages Containing Vitrified High-Level Radioactive Wastes under Sea Surface Fire." Proceedings of the National Symposium on Power and Energy Systems 2000.7 (2000): 463–68. http://dx.doi.org/10.1299/jsmepes.2000.7.463.

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Iwasaki, Toshiyuki, Shigeru Kato, Yoichi Kodera, Akimichi Hatta, Masahiro Kikuzato, Akihiro Yamasaki, Takuya Ito, Seiichi Suzuki, and Toshinori Kojima. "Thermal Decomposition of Woody Wastes Contaminated with Radioactive Materials using Externally-Heated Horizontal Kiln." KAGAKU KOGAKU RONBUNSHU 41, no. 1 (2015): 62–66. http://dx.doi.org/10.1252/kakoronbunshu.41.62.

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Journal articles on the topic 'Radioactive wastes – Thermal properties'

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