Zeitschriftenartikel zum Thema „Captur of gaseous radionuclides“
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1
Stewart, A., A. Brudenell und C. D. Collins. „Deposition of gaseous radionuclides to fruit“. Journal of Environmental Radioactivity 52, Nr. 2-3 (Januar 2001): 175–89. http://dx.doi.org/10.1016/s0265-931x(00)00032-1.
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Collins, C. D., und G. Shaw. „Modelling the fate of gaseous radionuclides in crops“. Radioprotection 37, Nr. C1 (Februar 2002): C1–43—C1–48. http://dx.doi.org/10.1051/radiopro/2002083.
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Chung, C., und C. H. Tsai. „Rapid Monitoring of Gaseous Radionuclides Using a Portable Spectrometer“. Radiation Protection Dosimetry 61, Nr. 1-3 (01.08.1995): 137–40. http://dx.doi.org/10.1093/oxfordjournals.rpd.a082769.
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Chung, C., und C. H. Tsai. „Rapid Monitoring of Gaseous Radionuclides Using a Portable Spectrometer“. Radiation Protection Dosimetry 61, Nr. 1-3 (01.08.1995): 137–40. http://dx.doi.org/10.1093/rpd/61.1-3.137.
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Johnson, Chelsea, Nandini G, Santosh K. Balivada und Surya Prakash. „Radioactive Waste Management in a Medical Cyclotron Facility - A Review“. International Journal of Health Technology and Innovation 1, Nr. 03 (23.12.2022): 20–23. http://dx.doi.org/10.60142/ijhti.v1i03.53.
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The cyclotron is a device used to create radioactive atoms with a short half-life (radioactive isotopes) that can be utilised for research and medical imaging. When nuclear and radiation facilities are utilized, serviced, or decommissioned, radioactive waste is produced. The amount of radioactive waste produced is greatly decreased by good operating procedures. Iodine-123, Technetium-99m, Iodine-131, Gallium-67 Thallium-201 and fluorine-18 fluorodeoxyglucose are among the radionuclides utilised in medicine. The most widely used gaseous/aerosol radionuclides are (aerosolized) technetium-99m, xenon-133, and krypton-81m. The use of radionuclides (radioactive element) for industrial process control and instrumentation, medical diagnostic and therapeutic purposes, as well as numerous uses in research, education, agriculture, geological exploration, construction, and other human endeavors, results in radioactive waste. These applications generate a variety of radioactive waste, which can come from sealed sources and be in solid, liquid, or gaseous form. If the trash containing considerable amounts of radionuclides is not handled properly, there may be serious concerns to both the environment and human health. Due to the wide variety of waste kinds addressed, special consideration must be paid to safety concerns and regulatory management. This article will examine the fundamental procedures for managing radioactive waste in compliance with the regulatory agencies like AERB (Atomic Energy Regulatory Board) and IAEA (International Atomic Energy Agency).
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Andrews, Hunter B., Praveen K. Thallapally und Alexander J. Robinson. „Monitoring Xenon Capture in a Metal Organic Framework Using Laser-Induced Breakdown Spectroscopy“. Micromachines 14, Nr. 1 (29.12.2022): 82. http://dx.doi.org/10.3390/mi14010082.
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Molten salt reactor operation will necessitate circulation of a cover gas to remove certain evolved fission products and maintain an inert atmosphere. The cover gas leaving the reactor core is expected to contain both noble and non-noble gases, aerosols, volatile species, tritium, and radionuclides and their daughters. To remove these radioactive gases, it is necessary to develop a robust off-gas system, along with novel sensors to monitor the gas stream and the treatment system performance. In this study, a metal organic framework (MOF) was engineered for the capture of Xe, a major contributor to the off-gas source term. The engineered MOF column was tested with a laser-induced breakdown spectroscopy (LIBS) sensor for noble gas monitoring. The LIBS sensor was used to monitor breakthrough tests with various Xe, Kr, and Ar mixtures to determine the Xe selectivity of the MOF column. This study offers an initial demonstration of the feasibility of monitoring off-gas treatment systems using a LIBS sensor to aid in the development of new capture systems for molten salt reactors.
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Barbin, Nikolay M., Stanislav A. Titov, Dmitry I. Terentiev und Anton M. Kobelev. „Computer simulation of thermal processes involving Sr and Ca radionuclides in the process of heating radioactive graphite in an air atmosphere“. Nuclear Energy and Technology 9, Nr. 4 (19.12.2023): 273–79. http://dx.doi.org/10.3897/nucet.9.116661.
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The paper presents the results from a thermodynamic analysis of the behavior of Sr and Ca radionuclides in the process of heating radioactive graphite in an air atmosphere. The TERRA software package was used for the thermodynamic analysis in a temperature range of 300 to 3600 K to determine the possible composition of the ionized, gaseous and condensed phases. It has been found that strontium is in the form of condensed SrCl2(c) and gaseous SrCl2 in a temperature range of 300 to 1600 K, and in the form of gaseous SrCl2, SrO, SrCl and Sr and ionized SrCl+, Sr+ and SrO+ when the temperature is increased from 1600 to 3600 K. Calcium is in the form of condensed CaCl2(c), CaUO4(c), CaO(c) and gaseous CaCl2 in the temperature interval between 300 and 2100 K, and in the form of gaseous Ca, CaCl and CaO and ionized Ca+, CaO+ and CaCl+ when the temperature is increased from 2100 to 3600 K. The paper determines the key reactions within individual phases and among condensed, gaseous and ionized phases. The equilibrium constants of their reactions have been calculated. Based on the results obtained, dependence plots are presented for the Sr and Ca radionuclide distribution by phases.
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Quérel, A., P. Lemaitre, M. Monier, E. Porcheron, A. I. Flossmann und M. Hervo. „An experiment to measure raindrop collection efficiencies: influence of rear capture“. Atmospheric Measurement Techniques 7, Nr. 5 (19.05.2014): 1321–30. http://dx.doi.org/10.5194/amt-7-1321-2014.
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Abstract. In the case of severe accident with loss of containment in a nuclear plant, radionuclides are released into the atmosphere in the form of both gases and aerosol particles (Baklanov and Sørensen, 2001). The analysis of radioactive aerosol scavenged by rain after the Chernobyl accident highlights certain differences between the modelling studies and the environmental measurements. Part of these discrepancies can probably be attributed to uncertainties in the efficiencies used to calculate aerosol particle collection by raindrops, particularly drops with a diameter larger than one millimetre. In order to address the issue of these uncertainties, an experimental study was performed to close the gaps still existing for this key microphysical parameter. In this paper, attention is first focused on the efficiency with which aerosol particles in the accumulation mode are collected by raindrops with a diameter of 2 mm. The collection efficiencies measured for aerosol particle in the sub-micron range are quantitatively consistent with previous theoretical model developed by Beard (1974) and thus highlight the major role of rear capture in the submicron range.
9
张, 志远. „Calculation of the Radionuclides Concentration in Gaseous Effluents from Yangjiang Nuclear Power Plant“. Nuclear Science and Technology 04, Nr. 03 (2016): 78–87. http://dx.doi.org/10.12677/nst.2016.43010.
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Kuzin, R., S. N. Brykin und T. Tairov. „SOURCES OF RADIOACTIVE WASTE IN LEACH PLANTS PROCESSING URANIUM ORES“. Fine Chemical Technologies 11, Nr. 5 (28.10.2016): 21–25. http://dx.doi.org/10.32362/2410-6593-2016-11-5-21-25.
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A distinctive feature of enterprises for extracting and processing uranium ore is the inevitable pollution by solid, liquid and gaseous waste. The amount of radioactive waste (RW) is most significant in the nuclear fuel cycle. In spite of its relatively low activity it is the major contributor to the formation of radiation hazards to the people and environment. The radioactivity of uranium ores and of their processing waste is due to natural radionuclides of uranium (238U and 235U) and thorium (232Th) radioactive decay chains.
11
Danis, A., M. Oncescu und M. Ciubotariu. „System for calibration of track detectors used in gaseous and solid alpha radionuclides monitoring“. Radiation Measurements 34, Nr. 1-6 (Juni 2001): 155–59. http://dx.doi.org/10.1016/s1350-4487(01)00142-1.
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Yasukawa, Chisato, Shoko Aoki, Miki Nonaka, Masateru Itakura, Masaharu Tsubokura, Kei’ichi Baba, Hiroya Ohbayashi et al. „Intake of Radionuclides in the Trees of Fukushima Forests 1. Field Study“. Forests 10, Nr. 8 (02.08.2019): 652. http://dx.doi.org/10.3390/f10080652.
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The earthquake and tsunami on 11 March 2011 led to a meltdown followed by a hydrogen explosion at the Fukushima–Daiichi nuclear power plant in Japan, causing the dispersal of abundant radionuclides into the atmosphere and ocean. The radionuclides were deposited onto trees and local residences in aerosol or gaseous forms that were partly absorbed by rain or melting snow. Here, we show that the radionuclides attached to the surfaces of trees, in which some radiocesium was incorporated into the xylem through ray cells and through symplastic pathways. The level of incorporated radiocesium varied based on tree species and age because of the ability of radiocesium to attach to the surface of the outer bark. After four years, the radiocesium level in the forest has been decreasing as it is washed out with rainwater into the sea and as it decays over time due to its half-life, but it can also be continuously recycled through leaf tissue, litter, mulch, and soil. As a result, the level of radiocesium was relatively increased in the heartwood and roots of trees at four years after the event. In private forest fields, most trees were left as afforested trees without being used for timber, although some trees were cut down. We discuss an interdisciplinary field study on the immediate effects of high radiation levels upon afforested trees in private forest fields.
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Mietelski, Jerzy W., Renata Kierepko, Kamil Brudecki, Paweł Janowski, Krzysztof Kleszcz und Ewa Tomankiewicz. „Long-range transport of gaseous 131I and other radionuclides from Fukushima accident to Southern Poland“. Atmospheric Environment 91 (Juli 2014): 137–45. http://dx.doi.org/10.1016/j.atmosenv.2014.03.065.
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Ludovici, Gian Marco, Andrea Chierici, Susana Oliveira de Souza, Francesco d’Errico, Alba Iannotti und Andrea Malizia. „Effects of Ionizing Radiation on Flora Ten Years after the Fukushima Dai-ichi Disaster“. Plants 11, Nr. 2 (15.01.2022): 222. http://dx.doi.org/10.3390/plants11020222.
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The aim of this work is to analyze the effects of ionizing radiation and radionuclides (like 137Cs) in several higher plants located around the Fukushima Dai-ichi Nuclear Power Plant (FNPP), evaluating both their adaptive processes and evolution. After the FNPP accident in March 2011 much attention was focused to the biological consequences of ionizing radiation and radionuclides released in the area surrounding the nuclear plant. This unexpected mishap led to the emission of radionuclides in aerosol and gaseous forms from the power plant, which contaminated a large area, including wild forest, cities, farmlands, mountains, and the sea, causing serious problems. Large quantities of 131I, 137Cs, and 134Cs were detected in the fallout. People were evacuated but the flora continued to be affected by the radiation exposure and by the radioactive dusts’ fallout. The response of biota to FNPP irradiation was a complex interaction among radiation dose, dose rate, temporal and spatial variation, varying radiation sensitivities of the different plants’ species, and indirect effects from other events. The repeated ionizing radiations, acute or chronic, guarantee an adaptation of the plant species, demonstrating a radio-resistance. Consequently, ionizing radiation affects the genetic structure, especially during chronic irradiation, reducing genetic variability. This reduction is associated with the different susceptibility of plant species to chronic stress. This would confirm the adaptive theory associated with this phenomenon. The effects that ionizing radiation has on different life forms are examined in this review using the FNPP disaster as a case study focusing the attention ten years after the accident.
15
Puchkov, A., D. Ponikarovskih und A. Olyukov. „Challenges in gaseous radwaste handling and in recording atmospheric leakages of radionuclides from high-pressure gas system“. TRANSACTIONS OF THE KRYLOV STATE RESEARCH CENTRE S-I, Nr. 2 (02.09.2019): 225–32. http://dx.doi.org/10.24937/2542-2324-2019-2-s-i-225-232.
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Li, Chuan, Wenqian Li, Lifeng Sun, Haoyu Xing und Chao Fang. „Chemical Forms of Important Fission Products in Primary Circuit of HTR-PM under Conditions of Normal Operation and Overpressure and Water Ingress Accidents: A Study with a Chemical Thermodynamics Approach“. Science and Technology of Nuclear Installations 2019 (25.07.2019): 1–12. http://dx.doi.org/10.1155/2019/4251280.
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The chemical forms of important fission products (FPs) in the primary circuit are essential to the source term analysis of high-temperature gas-cooled reactors because the volatility, transfer, and diffusion of these radionuclides are significantly influenced by their chemical forms. Through chemical reactions with gaseous impurities in the primary circuit, these FPs exist in diverse chemical forms, which vary under different operational conditions. In this paper, the chemical forms of cesium (Cs), strontium (Sr), silver (Ag), iodine (I), and tritium in the primary circuit of the Chinese pebble-bed modular high-temperature gas-cooled reactor (HTR-PM) under normal conditions and accident conditions (overpressure and water ingress accident) are studied with chemical thermodynamics. The results under normal conditions show that Cs exists mainly in the form of Cs2CO3 at 250°C and gaseous form at 750°C, and for I and Ag, Ag3I3 and Ag convert to gaseous CsI and AgO, respectively, with increasing temperature, while SrCO3 is the only main kind of compound for Sr. It is also observed that new compounds are generated under accidents: I exists in HI form when a water ingress accident occurs. Regarding tritium, the chemical forms of FPs change little, but compounds need higher temperature to convert. Furthermore, hazard of some FPs in different chemical forms is also discussed comprehensively in this paper. This study is significant for understanding the chemical reaction mechanisms of FPs in an HTR-PM, and furthermore it may provide a new point of view to analyze the interaction between FPs and structural materials in reactor as well as their hazards.
17
Williams, S. J. „An overview of gas research in support of the UK geological disposal programme“. Mineralogical Magazine 76, Nr. 8 (Dezember 2012): 3271–78. http://dx.doi.org/10.1180/minmag.2012.076.8.40.
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AbstractGases will be generated in waste packages during their transport to a geological disposal facility (GDF), this generation will continue during GDF operations and after GDF closure. The range of gases produced will include flammable, radioactive and chemotoxic species. These must be managed to ensure safety during transport and operations, and the post-closure consequences need to be understood. The two primary post-closure gas issues for a GDF are the need for the system pressure to remain below a value at which irreversible damage to the engineered barrier system and host geology could occur, and the need to ensure that any flux of gas (in particular gaseous radionuclides) to the biosphere does not result in unacceptable risk. This paper provides an overview of the research of the Nuclear Decommissioning Authority, Radioactive Waste Management Directorate into gas generation and its migration from a GDF.
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Bugai, D., und R. Avila. „Scenarios and Pathways of Radionuclide Releases from Near-Surface Waste Disposal Facilities: A Brief Overview of Historical Evidence“. Nuclear and Radiation Safety, Nr. 3(87) (15.09.2020): 21–27. http://dx.doi.org/10.32918/nrs.2020.3(87).03.
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The very low-level waste (VLLW) produced during decommissioning of nuclear facilities can be suitable for disposal in landfill type facilities. Considering the similarities in design, the experience gained in near-surface disposal of radioactive waste in trenches and vaults is relevant to the issue of VLLW disposal in landfills. This paper presents a brief review of internationally reported cases of radionuclide releases from near-surface disposal facilities. Based on this review, the conclusions are made that the following radionuclide release and exposure scenarios should be accounted for in safety assessment of VLLW disposal in landfills: i) leaching from waste to groundwater by atmospheric precipitations; ii) bath-tubing scenario; iii) scenarios caused by extreme meteorological and hydrological events (erosion, flooding, etc.); iv) human intrusion. The gaseous transport deserves attention for a number of relevant radionuclides, such as (C-14, Rn-222, etc.). In addition, the possibility of early degradation of engineered containment structures (soil covers, bottom seals) should be cautiously considered.
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Sarabu, Deepika Swetha. „A dentist’s perspective on scintigraphy about its applications as a diagnostic tool for a myriad of diseases affecting oral health“. Journal of Oral Medicine, Oral Surgery, Oral Pathology and Oral Radiology 9, Nr. 4 (15.12.2023): 202–6. http://dx.doi.org/10.18231/j.jooo.2023.043.
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A "branch or specialty of medicine & medical imaging that uses radionuclides and relies on the process of radioactive decay in the diagnosis and treatment of disease" is known as nuclear medicine. Assessing physiologic change, which follows biochemical changes, is made easier with the use of radionuclide imaging, which is sometimes referred to as a functional imaging approach. Scintigraphy is a diagnostic procedure used in nuclear medicine where radioisotopes are administered internally into the body in liquid or gaseous forms, and the distinctive radiation that emerges is captured by external detectors known as gamma cameras, producing two-dimensional pictures. Bone scintigraphy, lymphoscintigraphy, salivary gland scintigraphy, and radio-immuno-scintigraphy are a few examples of scintigraphy techniques.Nuclear imaging has the advantage of giving extremely high levels of diagnostic sensitivity, which is successful in identifying even the smallest pathophysiological changes that are highlighted in the process of diagnosing early-stage to progressive diseases, making it the preferred diagnostic modality.
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Smith, K., D. Jackson, G. Smith und S. Norris. „Comparison of modelled uptake to cereal crops of 14C from gaseous or groundwater mediated pathways“. Mineralogical Magazine 76, Nr. 8 (Dezember 2012): 3241–49. http://dx.doi.org/10.1180/minmag.2012.076.8.37.
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AbstractCarbon-14 has been identified as one of the more significant radionuclides in solid radioactive wastes in a repository, due to the potential radiological impact arising if 14C were to be released and enter the biosphere. However, the assessment of radiation doses is complicated by the major role of carbon in biological processes, and this has tended to lead to the adoption of a cautious assessment approach.An international comparison of five models used to predict uptake of 14C to agricultural crops has been undertaken, within the BIOPROTA framework. Processes investigated include conversion of 14C-labelled CH4 into CO2 in soils, carbon accumulation in and release from soil carbon pools, gaseous emanation to, and dispersion from, the plant canopy atmosphere and, incorporation into plants by photosynthesis.For a unit rate of entry of 14C to soil, modelled activity concentrations in cereal crops differ by three to five orders of magnitude. This reflects, in part, differing assumptions for mixing and dispersion of air above the soil surface and within the crop canopy layer. For a unit activity concentration of 14C in air, the modelled uptake to cereal crops converges significantly. Following an assumed irrigation of crops with groundwater containing unit activity of 14C, the predicted uptake to crops varied by two to four orders of magnitude, again largely dominated by assumptions regarding the canopy atmosphere. In all cases, there is some convergence in model predictions as field size increases.A continuing programme of field research is being undertaken in parallel with the assessment work.
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Voutilainen, Mikko, Juuso Sammaljärvi, Eveliina Muuri, Jérôme Donnard, Samuel Duval und Marja Siitari-Kauppi. „Digital autoradiography on C-14-labelled PMMA impregnated rock samples using the BeaverTM“. MRS Advances 3, Nr. 21 (2018): 1161–66. http://dx.doi.org/10.1557/adv.2018.226.
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In Finland and Sweden the KBS-3 concept has been chosen for the disposal of spent nuclear fuel in crystalline rock. Recent experiments have shown that heterogeneity of rock may play a major role in the transport of radionuclides. Autoradiographic methods have been proven to be able to assist the characterization of heterogeneous structures. In this study we tested a novel filmless autoradiographic device called BeaverTM which applies a micro patterned gaseous detector in order to quantitatively map beta emissions from C-14 atoms. The studied samples were impregnated with C-14-labelled methylmethacrylate (C-14-MMA) and polymerized to C-14-PMMA with thermal initiator. The BeaverTM was then used to determine the spatial distribution of the C-14-PMMA by measuring the C-14 emissions. The porosity is determined from the amount of C-14-PMMA in the rock sample and results were compared to ones from phosphor imaging plate autoradiography. The resulting images show a heterogeneous distribution of porosity which arises from the different minerals. The samples were chosen from three sites that have been used recently for in situ diffusion experiments: Olkiluoto (Finland), Äspö (Sweden) and Grimsel (Switzerland).
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Svetlana, A. Garelina, B. Grigoryan Gagik, A. Zaharyan Robert und M. Sedrakyan Armen. „Dynamics of the Concentration of Gaseous Radionuclides 14CO2 and 14CH4 Released above the Burial Site of Operational Radioactive Waste at Nuclear Power Plants“. Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika 2024, Nr. 1 (März 2024): 107–18. http://dx.doi.org/10.26583/npe.2024.1.09.
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Herm, Michel, Elke Bohnert, Luis Iglesias Pérez, Tobias König, Volker Metz, Arndt Walschburger und Horst Geckeis. „Twenty years of research on spent nuclear fuel in the context of final disposal at KIT-INE in multinational collaborative projects“. Safety of Nuclear Waste Disposal 1 (10.11.2021): 237–38. http://dx.doi.org/10.5194/sand-1-237-2021.
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Abstract. Disposal of spent nuclear fuel (SNF) in deep geological repositories is considered a preferential option for the management of such wastes in many countries with nuclear power plants. With the aim to permanently and safely isolate the radionuclide inventory from the biosphere for a sufficient time, a multibarrier system consisting of technical, geotechnical and geological barriers is interposed between the emplaced waste and the environment. In safety assessments for deep underground repositories, access of water, followed by failure of canisters and finally loss of the cladding integrity is considered in the long-term. Hence, evaluating the performance of SNF in deep geological disposal systems requires process understanding of SNF dissolution and rates as well as quantification of radionuclides release from SNF under reducing conditions of a breached container. In order to derive a radionuclide source term, the SNF dissolution and alteration processes can be assigned to two steps: (i) instantaneous release of radionuclides upon cladding failure from gap and grain boundaries and (ii) a long-term release that results from dissolution of the fuel grains itself (Ewing, 2015). In this context, research at KIT-INE has focused for more than 20 years on the behavior of SNF (irradiated UO2 and MOX fuels) under geochemical conditions (pH, redox and ionic strength) representative of various repository concepts, including the interaction of SNF with backfill material, such as bentonite as well as the influence of iron corrosion products, e.g. magnetite and radiolytic reactions on SNF dissolution mechanisms. Since 2001, KIT-INE has contributed with experimental and theoretical studies on the behavior of SNF under repository relevant conditions to six Euratom projects viz SFS (2001–2004), NF-PRO (2004–2006), MICADO (2006–2009), RECOSY (2007–2011), FIRST-Nuclides (2012–2014) and DISCO (2016–2021). Moreover, since 2007, overall 4 consecutive projects for the Belgian waste management organization, ONDRAF-NIRAS, were performed on the behavior of SNF under conditions representative of the Belgian “Supercontainer” concept. In this contribution, we summarize major achievements of theses research projects to understand and quantify the radionuclide release from dissolving SNF under repository conditions. In particular, the dependence of radionuclide release on the chemical composition of the aqueous and gaseous phase in the proximity of repositories in different types of host rock is discussed.
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Yamashita, Yu, Hiromi Tanabe, Tomofumi Sakuragi, Ryota Takahashi und Michitaka Sasoh. „C-14 Release Behavior and Chemical Species from Irradiated Hull Waste under Geological Disposal Conditions“. MRS Proceedings 1665 (2014): 187–94. http://dx.doi.org/10.1557/opl.2014.645.
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ABSTRACTC-14 contained in Hull waste is one of the most important radionuclides in the safety assessment of transuranic (TRU) waste disposal. For more realistic safety assessment, it is important to clarify the release mechanism and chemical species of C-14 from Hull waste. In this research, leaching tests were conducted using an irradiated Zry cladding tube from a boiling-water reactor (BWR) to obtain leaching data and to investigate the relationship between Zry metal corrosion and C-14 release behavior. Both organic and inorganic C-14 compounds existed in the the liquid phase, and some C-14 moved to the gaseous phase. The release rate of C-14 obtained from the BWR cladding tube after two-year leaching tests was lower than the release rate from a pressurize water reactor (PWR) cladding tube. It is considered that the BWR cladding tube used in this test did not easily corrode since it used a comparatively new material. The release rate of C-14 was slightly lower as compared with the corrosion rate of unirradiated Zry. This is thought to be the result of improved corrosion resistance conferred by neutron irradiation, which encouraged the dissolution of grain boundary precipitation elements, such as Fe, Cr, and Ni, into the crystal grains. The leaching tests will be continued for 10 years.
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González-Robles, Ernesto, Markus Fuß, Elke Bohnert, Nikolaus Müller, Michel Herm, Volker Metz und Bernhard Kienzler. „Study of the release of the fission gases (Xe and Kr) and the fission products (Cs and I) under anoxic conditions in bicarbonate water“. MRS Proceedings 1744 (2015): 35–41. http://dx.doi.org/10.1557/opl.2015.345.
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ABSTRACTFor safety assessment analyses of the disposal of spent nuclear fuel (SNF) in deep geological repositories it is indispensable to evaluate the contribution of fission products to the instant release fraction (IRF). During the last three years the EURATOM FP7 Collaborative Project, “Fast / Instant Release of Safety Relevant Radionuclides from Spent Nuclear Fuel (CP FIRST-Nuclides)” was carried out to get a better understanding of the IRF.Within CP FIRST-Nuclides, a leaching experiment with a cladded SNF pellet was performed in bicarbonate water (19 mM NaCl + 1 mM NaHCO3) under Ar /H2 atmosphere over 333 days. The cladded SNF pellet was obtained from a fuel rod segment which was irradiated in the Gösgen pressurized water reactor; the average burn-up of the segment was 50.4 MWd/kgUO2. In the multi-sampling experiment, gaseous and liquid samples were taken periodically. The moles of the fission gases Kr and Xe released in the gas phase and those of 129I and 137Cs released in solution were measured. Cumulative release fractions of (1.6 ± 0.2)·10-1 fission gases, (1.6 ± 0.1)·10-1129I and (3.9 ± 0.2)·10-2 137Cs, respectively, were achieved after 333 days of leaching. Accordingly the release ratio of fission gases to 129I was 1:1 and the release ratio of fission gases to 137Cs was 4:1, respectively.
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Saleh, Mohd Nazmi, Rahimah Embong, Norida Ridzuan, Nor Hanimah Hamidi, Ricca Rahman Nasaruddin, Putri Nadzrul Faizura Megat Khamaruddin und Norasyikin Ismail. „Crude Oil Extraction and Technologically-Enhanced Naturally Occurring Radioactive Materials (TENORM) Pre-Treatment of Petroleum Sludge: A Review“. Materials Science Forum 1056 (14.03.2022): 105–10. http://dx.doi.org/10.4028/p-0of01c.
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Malaysia is known as one of the major petroleum producers in the Asia-Pacific region. Petroliam Nasional Berhad (PETRONAS) in 2018 stated that the average daily production was over 1.7 million barrels of oil equivalent while the remaining reserves were estimated at more than 5 billion barrels of oil equivalent within more than 400 oilfields. With the growing demand for petroleum-based products, significant contaminated scales and sludge are generated each year from the petroleum industry. During production, the extracted fluids from the oil reservoir tend to carry along the Technologically-Enhanced Naturally Occurring Radioactive Materials (TENORM) of the 238U and 232Th decay chains from the Earth’s crust. TENORM in the petroleum sludge will results in radionuclides’ precipitation with silicates and carbonates, thus lowering the amount of oil extracted. There is a need for further information regarding the elemental composition (metal and nonmetal) and the surface morphology. Such information will guide the choice for useful partitioning of heavy metals between solid and gaseous products and provide a basis for comparing product characteristics with the parent material. Thus, this paper presents a comprehensive analysis of petroleum sludge’s physical and chemical properties and its treatment and application. Profound evaluation of the extent of sludge treatment before disposal could be done and can significantly impact refinery and petrochemical industries.
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Lepore, L., L. Falconi, V. Fabrizio, A. Roberti, D. Formenton, M. G. Iorio und L. Sperandio. „Defected fuel rods identification in TRIGA Reactors: The experience at the ENEA Casaccia TRIGA RC-1 reactor“. EPJ Web of Conferences 288 (2023): 04005. http://dx.doi.org/10.1051/epjconf/202328804005.
Annotation:
Experience in running TRIGA reactors all around the world has shown that fuel cladding failures can occur. Fission products, especially in gaseous physical form, can exit the defected fuel rods being dispersed within the water primary coolant. Suspects of a cladding failure event can be confirmed by detection of short-lived fission products, i.e., Krypton, Xenon, and even Iodine isotopes in primary water, or within the ionic-exchange resins tank installed for the purification of the primary coolant loop. The magnitude of the release of those ‘key-indicators’ from the defected fuel element(s) is the driving method by which the defected rod(s) can be identified. ‘Significant’ releases can be detected with a direct online sampling of the water from the top of the suspected fuel rod with reactor at power, leading this water to an online high-resolution gamma spectrometry analysis system. By considering delays due to the water velocity in tubes and decay time of radionuclides identified within the gamma spectrum, it is possible to calculate the concentration of those radionuclides just emerging from the inspected fuel rod. Releases lower than the minimum detection capabilities of the previous online experimental configuration push to modify the detection method with an indirect identification of the release. This is the case when the ‘normal’ radioactive background of the activated water, when reactor is on power, is the dominant component in the gamma spectrum of the sampled water, and fission gases (even produced) are not identified promptly, i.e. a relationship to a specific fuel rod by the sampling circuit before could not be identified. The paper describes the experience carried out at the Italian ENEA TRIGA RC-1 reactor, deepening the technical aspects and solutions applied to solve the issue. In particular, a ‘significant’ release has been found in the instrumented fuel rod within ring B, i.e. the inner fuel ring, exposed to the maximum neutron flux of the reactor. The leaking element was found within a week, in two days of operation, being the sampling system designed on detection of minutes-shortlived fission gases (no need to ‘cool’ down the primary loop by waiting decay of hours-lived radionuclides). After the removal of the instrumented fuel rod in ring B, further days in searching other ‘significant’ release with the sampling circuit before have reported nothing detected. But in samples of water taken after reactor shutdown, some iodine emerged after decaying of ‘normal’ radioactive background of the activated water. This was sufficient evidence of another defected fuel rod in the pool. Identification of the latter defected fuel element is still ongoing, being based on: 1) selective removal of rod(s) from the reactor core, 2) run power of the reactor, 3) take samples of water after shutdown and measure iodine after decaying of ‘normal’ radioactive-background of the activated water, 4) identify the rod responsible of the remaining leakage when no further iodine is detected.
28
Kristiansen, N. I., A. Stohl und G. Wotawa. „Atmospheric removal times of the aerosol-bound radionuclides <sup>137</sup>Cs and <sup>131</sup>I measured after the Fukushima Dai-ichi nuclear accident – a constraint for air quality and climate models“. Atmospheric Chemistry and Physics 12, Nr. 22 (16.11.2012): 10759–69. http://dx.doi.org/10.5194/acp-12-10759-2012.
Annotation:
Abstract. Caesium-137 (137Cs) and iodine-131 (131I) are radionuclides of particular concern during nuclear accidents, because they are emitted in large amounts and are of significant health impact. 137Cs and 131I attach to the ambient accumulation-mode (AM) aerosols and share their fate as the aerosols are removed from the atmosphere by scavenging within clouds, precipitation and dry deposition. Here, we estimate their removal times from the atmosphere using a unique high-precision global measurement data set collected over several months after the accident at the Fukushima Dai-ichi nuclear power plant in March 2011. The noble gas xenon-133 (133Xe), also released during the accident, served as a passive tracer of air mass transport for determining the removal times of 137Cs and 131I via the decrease in the measured ratios 137Cs/133Xe and 131I/133Xe over time. After correction for radioactive decay, the 137Cs/133Xe ratios reflect the removal of aerosols by wet and dry deposition, whereas the 131I/133Xe ratios are also influenced by aerosol production from gaseous 131I. We find removal times for 137Cs of 10.0–13.9 days and for 131I of 17.1–24.2 days during April and May 2011. The removal time of 131I is longer due to the aerosol production from gaseous 131I, thus the removal time for 137Cs serves as a better estimate for aerosol lifetime. The removal time of 131I is of interest for semi-volatile species. We discuss possible caveats (e.g. late emissions, resuspension) that can affect the results, and compare the 137Cs removal times with observation-based and modeled aerosol lifetimes. Our 137Cs removal time of 10.0–13.9 days should be representative of a "background" AM aerosol well mixed in the extratropical Northern Hemisphere troposphere. It is expected that the lifetime of this vertically mixed background aerosol is longer than the lifetime of fresh AM aerosols directly emitted from surface sources. However, the substantial difference to the mean lifetimes of AM aerosols obtained from aerosol models, typically in the range of 3–7 days, warrants further research on the cause of this discrepancy. Too short modeled AM aerosol lifetimes would have serious implications for air quality and climate model predictions.
29
Sakuragi, Tomofumi, Hideaki Miyakawa, Tsutomu Nishimura und Tsuyoshi Tateishi. „Long-term Corrosion of Zircaloy-4 and Zircaloy-2 by Continuous Hydrogen Measurement under Repository Condition“. MRS Proceedings 1518 (2013): 173–78. http://dx.doi.org/10.1557/opl.2013.66.
Annotation:
ABSTRACTCorrosion behavior is a key issue for the waste disposal of irradiated metals, such as hulls and endpieces, and is considered to be a leaching source of radionuclides including C-14. However, little information about Zircaloy corrosion in anticorrosive conditions has been provided.In the present study, long-term corrosion tests of Zircaloy-4 and Zircaloy-2 were performed in assumed disposal conditions (dilute NaOH solution, pH 12.5, 303 K) by using the gas flow system for 1500 days. The corrosion rate, which was determined by measuring gaseous hydrogen and the hydrogen absorbed in Zircaloy, decreased with immersion time and was lower than the value of 2×10−2 μm/y used in performance assessment (1500-day values: 5.84×10−3 and 5.66×10−3 μm/y for Zircaloy-4, 1000-day values: 8.81×10−3 μm/y for Zircaloy-2). The difference in corrosion behavior between Zircaloy 4 and Zircaloy-2 was negligible. The average values of the hydrogen absorption ratios for Zircaloy-4 and Zircaloy-2 during corrosion were 91% and 94%, respectively.The hydrogen generation kinetics of both gas evolution and absorption into metal can be shown by a parabolic curve. This result indicates that the diffusion process controls the Zircaloy corrosion in the early corrosion stage of the present study, and that the thickness of the oxide film in this stage is limited to approximately 25 nm and may therefore be in the form of dense tetragonal zirconia.
30
Dragunova, Anastasiya V., Mikhail S. Morkin und Vladimir V. Perevezentsev. „Features of methods for monitoring the fuel cladding tightness in lead-cooled fast breeder reactors“. Nuclear Energy and Technology 7, Nr. 4 (17.12.2021): 319–25. http://dx.doi.org/10.3897/nucet.7.78372.
Annotation:
To timely detect failed fuel elements, a reactor plant should be equipped with a fuel cladding tightness monitoring system (FCTMS). In reactors using a heavy liquid-metal coolant (HLMC), the most efficient way to monitor the fuel cladding tightness is by detecting gaseous fission products (GFP). The article describes the basic principles of constructing a FCTMS in liquid-metal-cooled reactors based on the detection of fission products and delayed neutrons. It is noted that in a reactor plant using a HLMC the fuel cladding tightness is the most efficiently monitored by detecting GFPs. The authors analyze various aspects of the behavior of fission products in a liquid-metal-cooled reactor, such as the movement of GFPs in dissolved and bubble form along the circuit, the sorption of volatile FPs in the lead coolant (LC) and on the surfaces of structural elements, degassing of the GFPs dissolved in the LC, and filtration of cover gas from aerosol particles of different nature. In addition, a general description is given of the conditions for the transfer of GFPs in a LC environment of the reactor being developed. Finally, a mathematical model is presented that makes it possible to determine the calculated activity of reference radionuclides in each reactor unit at any time after the fuel element tightness failure. Based on this model, methods for monitoring the fuel cladding tightness by the gas activity in the gas volumes of the reactor plant will be proposed.
31
Weyand, Torben, Holger Seher und Guido Bracke. „Geochemical benchmark tests to validate the conversion of thermodynamic data for TOUGHREACT“. Safety of Nuclear Waste Disposal 1 (10.11.2021): 161–62. http://dx.doi.org/10.5194/sand-1-161-2021.
Annotation:
Abstract. According to the ongoing site selection process for a repository for high-level radioactive waste in Germany, rock salt, clay and crystalline rock are possible host rocks. The pore water of these rocks contains saline solutions with high ionic strengths. To model the speciation and/or migration of radionuclides in long-term safety analyses for nuclear waste disposal, a geochemical code that includes thermodynamic data suitable for saline solutions is needed. Thermodynamic equilibrium in saline solutions with high ionic strengths is usually modelled using the Pitzer approach (Pitzer, 1991). Within the context of nuclear waste disposal, the THEREDA project (Moog et al., 2015) provides thermodynamic data for some widely used geochemical codes (PHREEQC, Geochemist's Workbench, ChemApp, and EQ 3/6) using the Pitzer approach; however, for modelling in long-term safety analyses for nuclear waste disposal, another geochemical code, TOUGHREACT, is used. Therefore, scripts were developed to convert thermodynamic data of the THEREDA project to be applicable in TOUGHREACT. The scripts were validated by benchmark tests and by comparing calculations using PHREEQC and TOUGHREACT (Weyand et al., 2021). In total, 50 different benchmark tests were performed considering 3 specific geochemical systems, which are relevant to long-term safety analyses: (1) oceanic salt system, polythermal: K, Mg, Ca, Cl, SO4, H2O(l), (2) actinide system, isothermal: Am(III), Cm(III), Nd(III), Na, Mg, Ca, Cl, OH, H2O(l) and (3) carbonate system, isothermal: Na, K, Mg, Ca, Cl, SO4, HCO3/CO2(g), H2O(l). Each benchmark test considered specific ion concentrations in solution and in gaseous phases in the presence of specific minerals. The benchmark tests derived the geochemical equilibria and the results of both codes were compared to each other and to experimental data. The results of the calculations using both codes showed a good correlation. Remaining deviations can be explained by technical differences of the codes.
32
Dibb, J. E., L. D. Ziemba, J. Luxford und P. Beckman. „Bromide and other ions in the snow, firn air, and atmospheric boundary layer at Summit during GSHOX“. Atmospheric Chemistry and Physics Discussions 10, Nr. 5 (31.05.2010): 13609–42. http://dx.doi.org/10.5194/acpd-10-13609-2010.
Annotation:
Abstract. Measurements of gas phase soluble bromide in the boundary layer and in firn air, and Br− in aerosol and snow, were made at Summit, Greenland (72.5° N, 38.4° W, 3200 m a.s.l.) as part of a larger investigation into the influence of Br chemistry on HOx cycling. The soluble bromide measurements confirm that photochemical activation of Br− in the snow causes release of active Br to the overlying air despite trace concentrations of Br− in the snow (means 15 and 8 nmol Br− kg−1 of snow in 2007 and 2008, respectively). Mixing ratios of soluble bromide above the snow were also found to be very small (mean <1 ppt both years, with maxima of 3 and 4 ppt in 2007 and 2008, respectively), but these levels clearly oxidize and deposit long-lived gaseous elemental mercury and may perturb HOx partitioning. Concentrations of Br− in surface snow tended to increase/decrease in parallel with the specific activities of the aerosol-associated radionuclides 7Be and 210Pb. Earlier work has shown that ventilation of the boundary layer causes simultaneous increases in 7Be and 210Pb at Summit, suggesting there is a pool of Br in the free troposphere above Summit in summer time. Speciation and the source of this free tropospheric Br are not well constrained, but we suggest it may be linked to extensive regions of active Br chemistry in the Arctic basin which are known to cause ozone and mercury depletion events shortly after polar sunrise. If this hypothesis is correct, it implies persistence of the free troposphere Br− for several months after peak Br activation in March/April. Alternatively, there may be a~ubiquitous pool of Br− in the free troposphere, sustained by currently unknown sources and processes.
33
Dibb, J. E., L. D. Ziemba, J. Luxford und P. Beckman. „Bromide and other ions in the snow, firn air, and atmospheric boundary layer at Summit during GSHOX“. Atmospheric Chemistry and Physics 10, Nr. 20 (20.10.2010): 9931–42. http://dx.doi.org/10.5194/acp-10-9931-2010.
Annotation:
Abstract. Measurements of gas phase soluble bromide in the boundary layer and in firn air, and Br− in aerosol and snow, were made at Summit, Greenland (72.5° N, 38.4° W, 3200 m a.s.l.) as part of a larger investigation into the influence of Br chemistry on HOx cycling. The soluble bromide measurements confirm that photochemical activation of Br− in the snow causes release of active Br to the overlying air despite trace concentrations of Br− in the snow (means 15 and 8 nmol Br− kg−1 of snow in 2007 and 2008, respectively). Mixing ratios of soluble bromide above the snow were also found to be very small (mean <1 ppt both years, with maxima of 3 and 4 ppt in 2007 and 2008, respectively), but these levels clearly oxidize and deposit long-lived gaseous elemental mercury and may perturb HOx partitioning. Concentrations of Br− in surface snow tended to increase/decrease in parallel with the specific activities of the aerosol-associated radionuclides 7Be and 210Pb. Earlier work has shown that ventilation of the boundary layer causes simultaneous increases in 7Be and 210Pb at Summit, suggesting there is a pool of Br in the free troposphere above Summit in summer time. Speciation and the source of this free tropospheric Br− are not well constrained, but we suggest it may be linked to extensive regions of active Br chemistry in the Arctic basin which are known to cause ozone and mercury depletion events shortly after polar sunrise. If this hypothesis is correct, it implies persistence of the free troposphere Br− for several months after peak Br activation in March/April. Alternatively, there may be a ubiquitous pool of Br− in the free troposphere, sustained by currently unknown sources and processes.
34
Kristiansen, N. I., A. Stohl und G. Wotawa. „Atmospheric removal times of the aerosol-bound radionuclides <sup>137</sup>Cs and <sup>131</sup>I during the months after the Fukushima Dai-ichi nuclear power plant accident – a constraint for air quality and climate models“. Atmospheric Chemistry and Physics Discussions 12, Nr. 5 (14.05.2012): 12331–56. http://dx.doi.org/10.5194/acpd-12-12331-2012.
Annotation:
Abstract. Caesium-137 (137Cs) and iodine-131 (131I) are radionuclides of particular concern during nuclear accidents, because they are emitted in large amounts and are of significant health impact. 137Cs and 131I attach to the ambient accumulation-mode (AM) aerosols and share their fate as the aerosols are removed from the atmosphere by scavenging within clouds, precipitation and dry deposition. Here, we estimate their removal times from the atmosphere using a unique high-precision global measurement data set collected over several months after the accident at the Fukushima Dai-ichi nuclear power plant in March 2011. The noble gas xenon-133 (133Xe), also released during the accident, served as a passive tracer of air mass transport for determining the removal times of 137Cs and 131I via the decrease in the measured ratios 137Cs/133Xe and 131I/133Xe over time. After correction for radioactive decay, the 137Cs/133Xe ratios reflect the removal of aerosols by wet and dry deposition, whereas the 131I/133Xe ratios are also influenced by aerosol production from gaseous 131I. We find removal times for 137Cs of 10.0–13.9 days and for 131I of 17.1–24.2 days during April and May 2011. We discuss possible caveats (e.g. late emissions, resuspension) that can affect the results, and compare the 137Cs removal times with observation-based and modeled aerosol lifetimes. Our 137Cs removal time of 10.0–13.9 days should be representative of a "background" AM aerosol well mixed in the extratropical Northern Hemisphere troposphere. It is expected that the lifetime of this vertically mixed background aerosol is longer than the lifetime of AM aerosols originating from surface sources. However, the substantial difference to the mean lifetimes of AM aerosols obtained from aerosol models, typically in the range of 3–7 days, warrants further research on the cause of this discrepancy. Too short modeled AM aerosol lifetimes would have serious implications for air quality and climate model predictions.
35
Semerak, M., S. Lys und T. Kovalenko. „Analysis of the Plasma Recycling Process of Radioactive Waste“. Nuclear and Radiation Safety, Nr. 1(81) (12.03.2019): 23–29. http://dx.doi.org/10.32918/nrs.2019.1(81).04.
Annotation:
The possibility of the plasma processing of low-level or intermediatelevel radioactive wastes in the reactor equipped with arc plasmatrons is shown. The reactor design for the plasma processing of the radioactive wastes that allows promoting the efficiency of the plasma processing of the radioactive wastes (RAW) by the increasing of the speed and the intensity of the plasma pyrolysis is proposed. The various methods for RAW preparation, dosage and supply into the plasmochemical reactor have been investigated. The waste which is supplied to the reactor can be in various aggregate states (solid, liquid or gaseous) depending on which different kinds of preparation, dosage, and supply of RAW materials to the plasmochemical reactor are used. The solid waste must be ground for increasing of the phase separation surface. The degree of grinding of the wastes depends on their further reprocessing. The reactor allows processing of the mixed-type radioactive waste, which includes both combustible and non-combustible components. The wastes can be packed or ground up. The selected technological regimes should provide temperature from 1500 °C in the melting chamber to 250 °C in the upper part in the pyrogas exit zone to prevent the flow-out of volatile compounds of a series of radionuclides and heavy metals from the furnace and to process the waste and merge slag melt without adding of fluxes. The fused slag is a basaltiform monolith, where the content of aluminum oxide reaches 28%; silicon oxide up to 56%; sodium oxide from 2.5 to 11 %. The resulting radioactive slag is extremely resistant to the chemical influence. The pyrogas produced in the shaft furnace will have a heating value of about 5 MJ/nm3. This allows, after initial heating by plasmatron, maintaining the required temperature in the combustion chamber due to the heat released during combustion of the pyrogas, when the plasma heating source is switched off, and burning the resin and soot effectively. It is proved that the plasma technology for RAW reprocessing allows a significant reduction in waste volumes and waste placement for long-term storage with the most efficient use of storage facilities.
36
Paul, Matthew J., Steven R. Biegalski, Derek A. Haas und Justin D. Lowrey. „Adsorptive transport of noble gas tracers in porous media“. International Journal of Modern Physics: Conference Series 48 (Januar 2018): 1860124. http://dx.doi.org/10.1142/s2010194518601242.
Annotation:
The transport of noble gas radionuclides in porous media is relevant to the detection of underground nuclear detonations as well as the sequestration of reprocessing off-gases. However, in field tests releasing radioxenon underground, the quantity of radioxenon observed at the surface has fallen well below expectations.[Formula: see text] This study examined the diffusivity of noble gases (Kr and Xe) and the inert molecular gas sulfur hexafluoride (SF[Formula: see text] in porous media to observe any unexpected behavior. To replicate the transport of radiogenic noble gases in underground media, a two-bulb gaseous diffusion apparatus was constructed. The two bulbs were connected with a column of 10–30 Ottawa sand and ordinary atmosphere filled both the bulbs and pore spaces. The tracer gases were diluted in an isolated bulb to approximately 1000 ppm. Once released, the gases were allowed to diffuse through the column. Aliquots were withdrawn at regular time intervals from both bulbs and concentrations were quantified using a Shimadzu QP2010 SE gas chromatograph-mass spectrometer. The effective diffusivity was then calculated using a maximum likelihood estimate on the quasi-steady state model. The effective diffusivity of Xe in the silica sand was observed to be 135.2% that of SF[Formula: see text] whereas the effective diffusivity of Kr was observed to be 161.4% that of sulfur hexafluoride. These findings are consistent with the binary diffusivities in N[Formula: see text]: 132.6% and 161.7%, respectively. However, the apparent volume of the system was inconsistent amongst the species, with Xe converging at slightly lower gas-phase concentrations than Kr or SF[Formula: see text]. This apparent reduction in gas-phase concentration occurred within the first few measurements and is consistent with transient accumulation of an adsorbed phase. As the effective diffusivities in the silica sand were shown to be consistent with the binary diffusivities in N[Formula: see text], a porosity-tortuosity model appears to be sufficient when considering similar geological materials. However, with the observation of significant gas adsorption, consideration of adsorbed-phase accumulation is necessary when scaling to larger geological systems.
37
Narkuniene, Asta, Gintautas Poskas und Gytis Bartkus. „Modelling of the Corrosion-Induced Gas Impact on Hydraulic and Radionuclide Transport Properties of Geological Repository Barriers“. Minerals 14, Nr. 1 (19.12.2023): 4. http://dx.doi.org/10.3390/min14010004.
Annotation:
The geological disposal of high-level radioactive waste is the final step in the nuclear fuel cycle. It is realized via isolating the high-level radioactive waste in the geological environment with an appropriate system of engineered barriers. Radionuclides-containing materials must be isolated from the biosphere until the radioactivity contained in them has diminished to a safe level. In the case of high-level radioactive waste, it could take hundreds of thousands of years. Within such a long timescale, a number of physical and chemical processes will take part in the geological repository. For the assessment of radionuclide migration from a geological repository, it is necessary to predict the repository’s behavior once placed in the host rock as well as the host-rock response to disturbances due to construction. In this study, the analysis of repository barriers (backfill, concrete, inner excavation disturbed zone (EDZ), outer EDZ, host rock) thermo–hydraulic–mechanical (THM) evolution was performed, and the scope of gas-induced desaturation was analyzed with COMSOL Multiphysics. The analysis was based on modelling of a two-phase flow of miscible fluid (water and H2) considering important phenomena such as gas dissolution and diffusion, advective–diffusive transport in the gaseous phase, and mechanical deformations due to thermal expansion of water and porous media. The importance of proper consideration of temperature-dependent thermodynamic properties of water and THM couplings in the analysis of near-field processes was also discussed. The modelling demonstrated that such activities as 50 years’ ventilation of the waste disposal tunnel in initially saturated porous media, and such processes as gas generation due to corrosion of waste package or heat load from the waste, also led to desaturation of barriers. H2 gas generation led to the desaturation in engineered barriers and in a part of the EDZ close to the gas generation place vanishing soon after finish of gas generation, while the host rock remained saturated during the gas generation phase (50–100,000 years). Radionuclide transport properties in porous media such as effective diffusivity are highly dependent on the water content in the barriers determined by their porosity and saturation.
38
Pavliuk, Alexander, Sergey Kotlyarevskiy, Evgeny Bespala und Yuliya Bespala. „Potential of application of IRT-T research reactor as the solution of the problem of graphite radwaste disposal“. Nuclear Energy and Technology 4, Nr. 2 (26.11.2018): 127–33. http://dx.doi.org/10.3897/nucet.4.30771.
Annotation:
Aspects of handling irradiated graphite during decommissioning uranium-graphite reactors (UGR) of different types were investigated. It was demonstrated that handling reactor graphite is complicated by the presence in the composition of graphite of long-lived radionuclides, especially 14C, which may get entrained in biological cycles since carbon constitutes one of the main components of biological chains. Practical implementation of the process of selective separation of 14С can significantly reduce potential danger represented by graphite radioactive wastes due to the reduction of graphite activity as related to the isotope in question, as well as due to the reduction of the leaching rate by separating 14С isotope which is the most weakly bound within the graphite structure. Conclusion was formulated that analytical measurement methodologies and calculation methods allow reliably estimating only the total quantity of 14C accumulated in graphite, the contribution of 14C accumulation channel from 13C(n, γ)14C reaction, as well as the total contribution of 14N(n, p)14C reaction on nitrogen impurities and on nitrogen contained in purge gas. Method was suggested for estimating the values of contributions of different channels of accumulation on nitrogen impurities and nitrogen contained in purge gas using IRT-T research reactor (Tomsk, Tomsk Region). Parallel irradiation of batches of samples of non-irradiated (fresh) reactor-grade graphite contained in different gaseous media constitutes the basis of the study. Algorithm was suggested for calculating contributions of all channels of 14C accumulation according to the results of measurements to be obtained in the proposed studies. Recommendations were formulated on the use of all brands of graphite applied for manufacturing elements of graphite stacks of uranium-graphite reactors designed in Russia for determining selectively separated fraction of 14C for all types of graphite radioactive wastes by the companies in the RF which operated (are operating) the uranium-graphite reactors. Time of exposure of samples of irradiated graphite in the GEK-4 horizontal experimental channel of the IRT-T reactor was calculated and was found to be equal to ~ 10 days. Methodology was suggested for conducting a series of experiments for determining the values of contributions of 14C accumulation channels in the irradiated reactor graphite. The methodology suggested can be applied for determining fraction of selectively separated 14C in irradiated graphite elements of practically all uranium-graphite nuclear reactors, including reactors operated abroad Russia, under the condition of maintaining carbon dioxide gas atmosphere in one of the irradiated containers.
39
Rong, Qiuyu, Jie Jin, Suhua Wang und Xiangke Wang. „Recent progress of covalent organic frameworks in high selective separation of radionuclides“. Carbon Research 3, Nr. 1 (30.05.2024). http://dx.doi.org/10.1007/s44246-024-00137-w.
Annotation:
AbstractThe utilization of nuclear energy power and nuclear weapon tests not only releases large amounts of radionuclides into environment, but also needs 235U as nuclear fuel for nuclear energy generation. Covalent organic frameworks (COFs) have the advantages of tunable porous structures, adjustable active sites and enough special functional groups, which assure the high selective preconcentration of target radionuclides from complex solutions. In this perspective, the selective extraction of radionuclides (U(VI) as representative cationic ion, Tc(VII) as representative anionic ion, I2 as gaseous nuclide and other nuclides) by COFs through sorption, and photocatalytic strategies are described, and the results show the high efficiency of COFs in target radionuclides removal. The perspective and challenges for the real applications of COFs in future are discussed in the end. Graphical Abstract
40
Zhou, W., P. L. Chambré, T. H. Pigford und W. W. L. Lee. „Heat-Pipe Effect on the Transport of Gaseous Radionuclides Released from a Nuclear Waste Container“. MRS Proceedings 212 (1990). http://dx.doi.org/10.1557/proc-212-855.
Annotation:
ABSTRACTWe present analytic studies of the transport of gaseous species released from a spent-fuel waste package, as affected by a heat-pipe of counterflowing liquid and vaporized water in the surrounding rock. A heat-pipe is caused by heating which vaporizes pore water near the waste, releasing vapor into the fractures. Driven by its pressure gradient, the vapor flows away from the waste and condenses where the rock is cooler. Because of capillary pressure gradient due to non-uniform liquid saturation, condensate flows towards the waste through the porous rock. We first develop analytic solutions for the time-dependent transport of energy and fluid from the waste container to the surrounding fractured porous rock. From the mass fluxes of liquid and vapor, we solve the advective-diffusive transport of a gaseous species released from the waste. The major assumptions are quasi-steady-state, local thermodynamic equilibrium, no noncondensable gases, and no sorption. Our results include the extent of the heat-pipe zone as function of time, the vapor velocity distribution in the heat-pipe zone, radionuclide concentration in water vapor, and the flux of radionuclide at the waste surface normalized to the surface concentration. We find that the vapor velocity in the heat-pipe zone is 1000-fold greater than the local air velocity if there were no heat pipe. If the gaseous species release mechanism maintains a near-constant concentration of gaseous species in the gas outside and near the waste container surface, the mass rate of transport of that species would be increased 1.3 to 7 times greater than if there were no heat pipe. However, if the release rate of the gaseous species is affected little by the concentration of that species outside the container, the heat-pipe can have little affect on the transport rate of that species.
41
Barbin, Nikolai M., Dmitri I. Terentiev, Sergei G. Alekseyev, Marat A. Tuktarov und A. A. Romenkov. „Modeling of Radioactive Graphite Oxidation in Molten Salts: Computer Experiment“. MRS Proceedings 1193 (2009). http://dx.doi.org/10.1557/proc-1193-359.
Annotation:
AbstractGraphite is used as the neutron moderator and reflector in many nuclear reactors. Obsolete graphite nuclear reactors are put out of operation, leading to formation of a large quantity of radioactive graphite waste.It is proposed that irradiated reactor graphite is processed by high-temperature chemical oxidation in salt melts with an oxidant, which is part of the salt melt, leading to formation of exhaust gases: gaseous compounds of carbon and oxygen (CO2 and CO).This study deals with carbon oxidation and physical-chemical transformations of radioactive elements during the interaction between graphite waste of the atomic power industry and salt melts. The method of thermodynamic simulation is used. The carbon melt decreases the transfer of radionuclides to the gaseous phase as compared to incineration of graphite in the atmosphere.
42
Maugeri, Emilio Andrea, Jörg Neuhausen, Robert Eichler, Rugard Dressler, Kim Rijpstra, Stefaan Cottenier, David Piguet, Alexander Vögele und Dorothea Schumann. „Adsorption of volatile polonium and bismuth species on metals in various gas atmospheres: Part I – Adsorption of volatile polonium and bismuth on gold“. Radiochimica Acta 104, Nr. 11 (01.01.2016). http://dx.doi.org/10.1515/ract-2016-2573.
Annotation:
AbstractPolonium isotopes are considered the most hazardous radionuclides produced during the operation of accelerator driven systems (ADS) when lead–bismuth eutectic (LBE) is used as the reactor coolant and as the spallation target material. In this work the use of gold surfaces for capturing polonium from the cover gas of the ADS reactor was studied by thermochromatography. The results show that gaseous monoatomic polonium, formed in dry hydrogen, is adsorbed on gold at 1058 K. Its adsorption enthalpy was calculated as –250±7 kJ mol
43
Barbin, Nikolaj Mihajlovich, Anton Mihajlovich Kobelev, Dmitrij Ivanovich Terent'ev und Sergej Gennad'evich Alekseev. „Thermodynamic Analysis of Radioactive Graphite Oxidation in NiO-NaCl-KCl-Na2CO3-K2CO3 Melt in the Atmosphere of Argon“. KnE Materials Science, 31.12.2020. http://dx.doi.org/10.18502/kms.v6i1.8129.
Annotation:
Behavior of U, Pu radionuclides was investigated when heating radioactive graphite in NaCl – KCl – Na2CO3 – K2CO3 melt with NiO additives using the thermodynamic modeling method. Calculations were made by the TERRA software that is used for the determination of phase composition, thermodynamic and transport properties, taking into account chemical and phase changes in temperature range 373 – 3273 K. Calculation of equilibrium phase composition and parameters of equilibrium was carried out using reference information about properties of the individual substances (INVATERMO, HSC, etc.). This study demonstrates that at a temperature of 1273 K the condensed carbon burns down with the formation of CO and CO2. Increasing temperature to 1673 K causes the condensed compounds of uranium to evaporate. This study determined that uranium exists in the form of ionized UO−3 in temperature range from 1673 to 3273 K. Plutonium exists in the form of gaseous PuO2, PuO in temperature range 2373 – 3273 K.
Keywords: thermodynamic modeling, radionuclides, radioactive graphite
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Robshaw, Thomas J., Joshua Turner, Sarah Kearney, Brant Walkley, Clint A. Sharrad und Mark D. Ogden. „Capture of aqueous radioiodine species by metallated adsorbents from wastestreams of the nuclear power industry: a review“. SN Applied Sciences 3, Nr. 11 (13.10.2021). http://dx.doi.org/10.1007/s42452-021-04818-8.
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Abstract Abstract Iodine-129 poses a significant challenge in the drive towards lowering radionuclide emissions from used nuclear fuel recycling operations. Various techniques are employed for capture of gaseous iodine species, but it is also present, mainly as iodide anions, in problematic residual aqueous wastestreams, which have stimulated research interest in technologies for adsorption and retention of the radioiodine. This removal effort requires specialised adsorbents, which use soft metals to create selectivity in the challenging chemical conditions. A review of the literature, at laboratory scale, reveals a number of organic, inorganic and hybrid adsorbent matrices have been investigated for this purpose. They are functionalised principally by Ag metal, but also Bi, Cu and Pb, using numerous synthetic strategies. The iodide capacity of the adsorbents varies from 13 to 430 mg g−1, with ion-exchange resins and titanates displaying the highest maximum uptakes. Kinetics of adsorption are often slow, requiring several days to reach equilibrium, although some ligated metal ion and metal nanoparticle systems can equilibrate in < 1 h. Ag-loaded materials generally exhibit superior selectivity for iodide verses other common anions, but more consideration is required of how these materials would function successfully in industrial operation; specifically their performance in dynamic column experiments and stability of the bound radioiodine in the conversion to final wasteform and subsequent geological storage. Article highlights Metallated adsorbents for the capture and retention of radioiodine in the nuclear industry are assessed. The strengths and weaknesses of organic, inorganic and hybrid support matrices and loading mechanisms are discussed. Pathways for progression of this technology are proposed. Graphic abstract
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Bury, Dominika, Michał Jakubczak, Rajiv Kumar, Dominika Ścieżyńska, Jan Bogacki, Piotr Marcinowski und Agnieszka Maria Jastrzębska. „Cleaning the environment with MXenes“. MRS Bulletin, 28.03.2023. http://dx.doi.org/10.1557/s43577-023-00507-6.
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AbstractRapid global industrialization constantly impacts the environment by discharging pollutants. Therefore, various materials are currently being investigated for environmental applications, including two-dimensional (2D) MXenes. Herein, we focus on MXene-enabled technologies for removing inorganic and organic contaminants present in gaseous and liquid forms, especially organic dyes, pharmaceuticals, and solid pollutants. We foresee a considerable potential for MXene-enabled technologies to remove heavy ions and radionuclides and recover precious elements. We show that MXenes could efficiently inactivate microorganisms without harming the environment. Finally, we discuss the associated opportunities and challenges in MXenes’ surface chemistry, semiconducting activity, interfacial effects, adsorption, and photocatalysis. Altogether, this article showcases outstanding opportunities for MXenes in the rapidly growing field of environmental applications. Graphical abstract
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Ashworth, Edward T., Ryotaro Ogawa, Juliana Nguyen, Chloe Afif, Rui C. Sá, Kim Butts Pauly, David R. Vera und Peter Lindholm. „A novel method for tracking hyperbaric nitrogen kinetics in vivo using radioactive nitrogen-13 gas and positron emission tomography“. Journal of Applied Physiology, 29.02.2024. http://dx.doi.org/10.1152/japplphysiol.00859.2023.
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Decompression sickness (DCS) is caused by gaseous nitrogen dissolved in tissues forming bubbles during decompression. To date no method exists to identify nitrogen within tissues, but with advances in PET technology it may be possible to track gaseous radionuclides into tissues. We aimed to develop a method to track nitrogen movement in vivo that could then be used to further our understanding of DCS using nitrogen-13 (13N2). A single anesthetized female Sprague Dawley rat, was exposed to 625 kPa, composed of air, isoflurane and 13N2 for 10 min. The PET scanner recorded 13N2 with energy windows of 250-750 keV. The PET showed an increase in 13N2 concentration in the lung, heart and abdominal regions, which all reached a plateau after ~4 min. This showed that it is possible to gain non-invasive in vivo measurements of nitrogen kinetics through the body while at hyperbaric pressures. Tissue samples showed radioactivity above background levels in the blood, brain, liver, femur and thigh muscle when assessed using a gamma counter.The method can be used to evaluate an array of challenges to our understanding of decompression physiology by providing a quantitative assessment method. Further development of the method will improve the specificity of the measured outcomes, and enable it to be used with larger mammals, including humans.
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Vulinović, Jelena, Srđan Vuković, Svetlana Pelemiš und Danijela Rajić. „RADON IN THE WATER“. Contemporary Materials 11, Nr. 1 (10.01.2020). http://dx.doi.org/10.7251/comen2001062v.
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Man and his environment are constantly exposed to the effects of ionizing radiation. Most of this radiation comes from natural and artificial radionuclides and the biggest radioecological problem is the 222Rn radioactive gas. Natural radioactivity comes from unstable radioisotopes that were present during the formation of the Earth, and are present today. According to the research by UNSCEAR(United Nations Scientific Committee on the Effects of Atomic Radiation) it is estimated that the radiation dose, which comes from natural radionuclides and to which man is exposed, is 2.4 mSv per year. Natural sources of radioactivity are cosmic radiation and Earth’s crust that contains primordial radioactive elements including those that are sources of radon (uranium). Radon is a natural inert radioactive gas without smell and taste. It is soluble in water and can easily diffuse with the gaseous and aqueous phase and in this way forms significant concentrations. The techniques and methods most commonly used to detect and determine the activities of radon in water are alpha spectrometry, gamma spectrometry and measurement techniques on a liquid scintillation detector. Throughout epidemiological studies, the World Health Organization has provided convincing evidence of the correlation of exposure to indoor radon and the development of lung cancer. Radon and its decomposition products are considered to be the second cause of lung cancer after consuming tobacco.
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Vučković, Biljana. „Negative effects of population exposure to radon from water“. ОДРЖИВИ РАЗВОЈ И УПРАВЉАЊЕ ПРИРОДНИМ РЕСУРСИМА РЕПУБЛИКЕ СРПСКЕ 9, Nr. 9 (16.12.2023). http://dx.doi.org/10.7251/eoru2309527v.
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The presence of radionuclides in drinking water, as well as in water used for other purposes, plays a significant role in the exposure of the population. Water monitoring is carried out in order to measure the content of natural radionuclides, primarily radon – 222Rn, and its short-lived descendants. Radon is the only gaseous radioactive product of the uranium series and a direct descendant of radium. The half-life of radon is 3.82 days, and its decay is followed by four descendants: 218Po, 214Pb, 214Bi and 214Po. The solubility of radium in water and its long-distance transport allow radon to accumulate in it, especially if it dissolves uranium-rich rocks. The resulting radon can be introduced into the water by the effect of entrainment and can be transported by underground flowing water to greater distances and entered into the body by ingestion. Radon is moderately soluble in water; it easily leaves it and thus increases its concentration in the air in closed rooms, and is taken into the body by inhalation. Epidemiological studies have shown that radon in drinking water is a secondary source of indoor radiation. The risk of manifesting carcinogenic effects due to exposure to radon, which is introduced into the body in different ways, is represented by chronic daily intake (CDI) of radon. The risk is quantitatively assessed for each entry route, and the simple sum of all entries determines the total exposure risk.
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Pastushkov, V. G., A. V. Molchanov, V. P. Serebryakov, T. V. Smelova und I. N. Shestoperov. „Technology and Equipment Based on Induction Melters with.Cold. Crucible for Reprocessing Active Metal Waste“. MRS Proceedings 663 (2000). http://dx.doi.org/10.1557/proc-663-59.
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ABSTRACTThe operation and, particularly, the decommissioning of NPPs and radiochemical plants result in substantial arisings of radioactive metal waste (RAMW) having different activity levels (from 5×10−4 to ≍ 40 Ci/kg).This paper reviews the specific features of the technology and equipment used to melt RAMW in electric arc and induction furnaces with ceramic or cold crucibles. The experimentally determined and calculated data are given on the level to which RAMW is decontaminated from the main radionuclides as well as on the distribution of the latter in the products of melting (ingot, slag, gaseous phase).Special attention is focused on the process and the facility for the induction-slag melting of RAMW in furnaces equipped with cold crucibles. The work described is under way at SSC RF VNIINM to master the technology of melting simulated high activity level Zr-alloy and stainless steel waste.
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OSMANLIOGLU, Ahmet. „Removal of Radioactive Gas by Zeolite Filter From Nuclear Power Plants“. Natural and Applied Sciences Journal, 13.06.2022. http://dx.doi.org/10.38061/idunas.844243.
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During the normal operation of nuclear power plants, some radioactive wastes are produced in the form of particles or aerosol gas in the reactor building. Particulate radioactive aerosols can be produced in a wide variety of particle sizes, possibly in combination with non-radioactive aerosols. Emission of corrosion products and fission products that are activated by the effect of nuclear rays generate aerosols from two sources. These; are created by the adsorption of gases generated by radioactive decay and volatile radionuclides formed during the fission process on the present suspended material. The most important volatile radionuclides that form the gaseous radioactive waste produced during the normal operation of nuclear power plants are halogens, noble gases, tritium, and carbon-14. The composition and amount of radioactivity present in the various airborne waste streams depend largely on the reactor type and release path. All gaseous waste from nuclear power plants must be treated before discharging into the atmosphere. In this paper, radioactive radon gas was used to represent the radioactive gases generated from nuclear power plants and natural zeolite was used as adsorbent material for radon removal. A series of experiments were conducted to measure the performance of the filter made in the zeolite. First of all, an approximate particle distribution size in the range of 1 to 3 mm was obtained by grinding natural zeolite. The ground material was then compressed in cylindrical adsorbent moulds of 35 mm diameter and 10 mm height. After the moulds were filled with the material, they were dried by heating to 110 ° C for 24 hours. At the end of the heat treatment, the adsorbent beds were cooled and connected to the test apparatus. RAD7 radon test device was used in the experiments. The RAD7 is a portable instrument that uses a solid-state alpha detector to measure radon gas concentrations in the range of 4.0-750,000 Bq / m3. The sampler of the RAD7 device works by drawing an air sample from an inlet filter into a 0.7 L sample cell covered with an electrical conductor. At the centre of the hemisphere, the cell is a planar silicon detector implanted with an ion to measure radioactivity. As result of the experiments, it shows that the zeolite filter absorbs 85% radioactive radon gas and can be used as an air filter in nuclear power plants.