Search

In clearance measurements involving a single material type, a conversion factor was applied to convert measurement results to activity based on an assumed uniform density. However, this factor has been found to underestimate activity in material mixtures. In this study, we proposed a method to identify the location with the lowest detection sensitivity (minimum location) in a mixture and evaluated its applicability to the conversion factor. We calculated this minimum location by determining the weighted average of a two-dimensional distribution of line attenuation coefficients, obtained through X-ray computed tomography (X-ray CT). We were able to identify the angle in polar coordinates using the center of the image as the origin, but the distance remained indeterminate. However, with this method, the minimum location can be identified along any straight line passing through the center of the image. This method has the potential to enhance activity evaluation for mixtures in cases where conservative evaluation of activity was previously unfeasible., Competing Interests: Declaration of competing interest The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper., (Copyright © 2025. Published by Elsevier Ltd.)

Radioactive waste management is critical for maintaining the sustainability of nuclear fuel cycles. In this study, we propose a novel bismuth-based reduced graphene oxide (Bi 0 -rGO) composite for the immobilization of off-gas radioactive iodine. This material synthesized via a solvothermal route exhibited a low surface area (2.96 m 2 /g) combined with a maximum iodine sorption capacity of 1228 ± 25 mg/g at 200 °C. The iodine sorbent was mixed with Bi 2 O 3 powder and distilled water to fabricate waste matrices, which were cold-sintered at 300 °C under a uniaxial pressure of 500 MPa for 20 min to achieve a relative density of ∼98% and Vickers hardness of 1.3 ± 0.1 GPa. The utilized methodology reduced the iodine leaching rate by approximately 3 orders of magnitude through the formation of a chemically durable iodine-bearing waste form (BiOI). This study demonstrates the high potential of Bi 0 -rGO as an innovative solution for the immobilization of radioactive waste at relatively low temperatures.

This paper reports a method for the fabrication of mineral-like SrMoO 4 ceramics with a powellite structure, which is promising for the immobilization of the high-energy 90 Sr radioisotope. The reported method is based on the solid-phase "in situ" interaction between SrO and MoO 3 oxides initiated under spark plasma sintering (SPS) conditions. Dilatometry, XRD, SEM, and EDX methods were used to investigate the consolidation dynamics, phase formation, and structural changes in the reactive powder blend and sintered ceramics. The temperature conditions for SrMoO 4 formation under SPS were determined, yielding ceramics with a relative density of 84.0-96.3%, Vickers microhardness of 157-295 HV, and compressive strength of 54-331 MPa. Ceramic samples demonstrate a low Sr leaching rate of 10 -6 g/cm 2 ·day, indicating a rather high hydrolytic stability and meeting the requirements of GOST R 50926-96 imposed on solid radioactive wastes. The results presented here show a wide range of prospects for the application of ceramic matrixes with the mineral-like composition studied here to radioactive waste processing and radioisotope manufacturing.

The aim was to develop a web application for the estimation of the radioactivity contained in liquid and solid waste produced by in vitro nuclear medicine laboratories. Median percentages of solid and liquid waste were estimated from successive runs of 44 in vitro kits. This data was fed into a web-based app, which estimates the activity of the waste and provides detailed record keeping. Median liquid/solid waste percentages were 77%/23% for RIA and 95%/5% for IRMA assays. Total waste activities were below regulatory discharge limits., Competing Interests: Declaration of competing interest The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper., (Copyright © 2022 Elsevier Ltd. All rights reserved.)

The main goal of this study was to assess alternatives to the current challenges on environmental quality and circular economy. The former is here addressed by the treatment of radioactively contaminated solutions, and the latter by using abundant and low-cost biomass. In this paper, we examine the biosorption of hexavalent uranium (U(VI)) in a batch system using the macrophytes Limnobium laevigatum and Azolla sp. by three operational parameters: biomass dose, metal ion concentration, and contact time. Simulated solutions were firstly addressed with two biomasses, followed by studies with real liquid organic radioactive waste (LORW) with Azolla sp. The batch experiments were carried out by mixing 0.20 g biomass in 10 mL of the prepared solution or LORW. The total contact time employed for the determination of the equilibrium times was 240 min, and the initial U(VI) concentration was 0.63 mmol L -1 . The equilibrium times were 15 min for L. laevigatum and 30 min for Azolla sp. respectively. A wide range of initial U(VI) concentrations (0.25-36 mmol L -1 ) was then used to assess the adsorption capacity of each macrophyte. Isotherm models validated the adsorption performance of the biosorption process. Azolla sp. presented a much higher U(VI) uptake (0.474 mmol g -1 ) compared to L. laevigatum (0.026 mmol g -1 ). When in contact with LORW, Azolla sp. removed much less uranium, indicating an adsorption capacity of 0.010 mmol g -1 . In conclusion, both biomasses, especially Azolla sp., can be used in the treatment of uranium-contaminated solutions., (© 2022. The Author(s), under exclusive licence to Springer-Verlag GmbH Germany, part of Springer Nature.)

Gamma radiation shielding and liquid permeability coefficient of baked and unbaked kaolin powder were experimentally and theoretically investigated. A NaI(Tl) detector was used to measure attenuation properties of the samples. The experimental values for the linear coefficient were validated by theoretical values using WinXCOM software. The results established that unbaked kaolin with density 2.65 g/cm 3 yielded highest mass attenuation coefficients of 0.0736 cm 2 /g, 0.0561 cm 2 /g and 0.0527 cm 2 /g for 661.7, 1173.2 and 1332.5 keV respectively and liquid permeability coefficient of 2.33 x 10 -12  m/s. The study concludes that compacted unbaked kaolin be considered a valid alternative for radiation shielding due to its low liquid permeability coefficient for liquid radioactive waste management., (Copyright © 2021 Elsevier Ltd. All rights reserved.)

This article gives an overview of Russian legislation and international recommendations on the safe management regulation of the accumulated very low level radioactive waste (RW), provides the comparison of Russian and international criteria for RW attribution to this or that class and also the existing radiation safety requirements. This article also deals with the difficulties in management of production waste containing radionuclides. Criteria and standards were selected from international (ICRP and IAEA publications) and Russian (Federal Law on Management of Radioactive Waste, Government Decree on Radioactive Waste Classification, The Basic Sanitary Rules of Radiation Safety) documents for comparative analysis. International and Russian criteria for the radioactive materials attribution to RW and production waste containing radionuclides were compared during the analysis. A number of radionuclides for which it is necessary to establish such criteria have been identified. It was found that Russian classification of radioactive materials is generally consistent with IAEA classification. Practices of managing production waste containing radionuclides (protection from contact with the environment, the placement in the long-term storage facilities or disposal) depend on the period of potential hazard of radionuclides. The management of production waste containing radionuclides requires a separate system which should be developed in accordance with the RW management system., (© 2021 Society for Radiological Protection. Published on behalf of SRP by IOP Publishing Limited. All rights reserved.)

Iodine-131 is increasingly used for diagnostic and therapeutic applications. The excretion of radioactive iodine is primarily through the urine. The safe disposal of radioactive waste is an important component of overall hospital waste management. This study investigated the feasibility of using graphene oxide/chitosan (GO/CS) sponges as an adsorbent for the removal of iodine-131 from aqueous solutions. The adsorption efficiency was investigated using iodine-131 radioisotopes to confirm the results in conjunction with stable isotopes. The results revealed that the synthetic structure consists of randomly connected GO sheets without overlapping layers. The equilibrium adsorption data fitted well with the Langmuir model. The separation factor (R L ) value was in the range of 0-1, confirming the favorable uptake of the iodide on the GO/CS sponge. The maximum adsorption capacity of iodine-131 by GO/CS sponges was 0.263 MBq/mg. The highest removal efficiency was 92.6% at pH 7.2 ± 0.2. Due to its attractive characteristics, including its low cost, the ease of obtaining it, and its eco-friendly properties, the developed GO/CS sponge could be used as an alternative adsorbent for removing radioiodine from wastewater.

The Swiss disposal concept foresees that carbon-14 ( 14 C) is predominantly released from irradiated steel disposed of in a cement-based repository for low- and intermediate-level radioactive waste. To predict how 14 C migrates in the cementitious environment of the repository near field and subsequently in the host rock, knowledge about the carbon speciation during anoxic steel corrosion in alkaline conditions is therefore essential. To this end, batch-type corrosion experiments with carbon-containing zero-valent iron (ZVI) powders subject to oxidative pre-treatments were carried out in NaOH solution at pH 11 and 12.5. Alkanes and alkenes (C 1 -C 7 ) were identified in the gas phase and produced on the iron surface by a Fischer-Tropsch type mechanism. The kind of oxidative pre-treatment has an effect on the production rate of hydrocarbons (HCs). In the liquid phase, carboxylic acids were identified and produced during the oxidative pre-treatment of the ZVI powders. They are released instantaneously from the oxide layer upon contact with the alkaline solution. The kind of oxidative treatment and the exposure time to oxic conditions directly influence the amount of carboxylic acids accommodated in the oxide layer., (Copyright © 2020 Elsevier Ltd. All rights reserved.)

For the last 30 years, the NEA Thermochemical Database (TDB) Project (www.oecd-nea.org/dbtdb/) has been developing a chemical thermodynamic database for elements relevant to the safety of radioactive waste repositories, providing data that are vital to support the geochemical modeling of such systems. The recommended data are selected on the basis of strict review procedures and are characterized by their consistency. The results of these efforts are freely available, and have become an international point of reference in the field. As a result, a number of important national initiatives with regard to waste management programs have used the NEA TDB as their basis, both in terms of recommended data and guidelines. In this article we describe the fundamentals and achievements of the project together with the characteristics of some databases developed in national nuclear waste disposal programs that have been influenced by the NEA TDB. We also give some insights on how this work could be seen as an approach to be used in broader areas of environmental interest., (Copyright © 2017 Elsevier Ltd. All rights reserved.)

The Brazilian Nuclear Energy Agency (CNEN) is responsible for any radioactive waste storage and disposal in the country. The storage of radioactive waste is carried out in the facilities under CNEN regulation and its disposal is operated, managed and controlled by the CNEN. Oil NORM (Naturally Occurring Radioactive Materials) in this article refers to waste coming from oil exploitation. Oil NORM has called much attention during the last decades, mostly because it is not possible to determine its primary source due to the actual absence of a regulatory control mechanism. There is no efficient regulatory tool which allows determining the origin of such NORM wastes even among those facilities under regulatory control. This fact may encourage non-authorized radioactive material transportation, smuggling and terrorism. The aim of this project is to provide a geochemical signature for oil NORM waste using its naturally occurring isotopic composition to identify its origin. The here proposed method is the modeling of radioisotopes normally present in oil pipe contamination such as 228 Ac, 214 Bi and 214 Pb analyzed by gamma spectrometry. The specific activities of elements from different decay series are plotted in a scatter diagram. This method was successfully tested with gamma spectrometry analyses of oil sludge NORM samples from four different sources obtained from Petrobras reports for the Campos Basin/Brazil., (Copyright © 2018 Elsevier Ltd. All rights reserved.)

This review article presents the current state-of-knowledge of the use of cementitious materials for radioactive waste disposal. An overview of radwaste management processes with respect to the classification of the waste type is given. The application of cementitious materials for waste disposal is divided into two main lines: i) as a matrix for direct immobilization of treated waste form; and ii) as an engineered barrier of secondary protection in the form of concrete or grout. In the first part the immobilization mechanisms of the waste by cement hydration products is briefly described and an up-to date knowledge about the performance of different cementitious materials is given, including both traditional cements and alternative binder systems. The advantages, disadvantages as well as gaps in the base of information in relation to individual materials are stated. The following part of the article is aimed at description of multi-barrier systems for intermediate level waste repositories. It provides examples of proposed concepts by countries with advanced waste management programmes. In the paper summary, the good knowledge of the material durability due to its vast experience from civil engineering is highlighted however with the urge for specific approach during design and construction of a repository in terms of stringent safety requirements., (Copyright © 2017 Elsevier Ltd. All rights reserved.)

In this article we investigate whether long-term radioactive waste management by means of geological disposal can be understood as a social experiment. Geological disposal is a rather particular technology in the way it deals with the analytical and ethical complexities implied by the idea of technological innovation as social experimentation, because it is presented as a technology that ultimately functions without human involvement. We argue that, even when the long term function of the 'social' is foreseen to be restricted to safeguarding the functioning of the 'technical', geological disposal is still a social experiment. In order to better understand this argument and explore how it could be addressed, we elaborate the idea of social experimentation with the notion of co-production and the analytical tools of delegation, prescription and network as developed by actor-network theory. In doing so we emphasize that geological disposal inherently involves relations between surface and subsurface, between humans and nonhumans, between the social, material and natural realm, and that these relations require recognition and further elaboration. In other words, we argue that geological disposal concurrently is a social and a technical experiment, or better, a long-term socio-technical experiment. We end with proposing the idea of 'actor-networking' as a sensitizing concept for future research into what geological disposal as a socio-technical experiment could look like.

In this study, two kinds of reference drums were developed. One type was constructed with nine layers of large-area sources filled with different materials having five different densities. The other type of reference drums was constructed with nine rod sources filled with the same materials of different densities. The efficiency calibration of a plastic scintillation counter in 4π counting geometry using these two kinds of drums showed that rod-source drums resulted in higher counting efficiency than layered source drums. The counting rates obtained from rod-source drums were closer to those obtained from a standard drum with water solution than counting rates from drums with layered sources. The results of this study recommend to use reference drums with rod-sources to compensate the drawbacks of standard drums with water solution of not being able to adjust the density of material. The proposed reference drums improve the accuracy of radioactivity analysis for waste drums of different densities., (Copyright © 2015 Elsevier Ltd. All rights reserved.)

A Europallet-sized calibration standard composed of 12 grey cast iron tubes contaminated with (60)Co and (110m)Ag with a mass of 246kg was developed. As the tubes were produced through centrifugal casting it was of particular concern to study the distribution of radionuclides in the radial direction of the tubes. This was done by removing 72 small samples (swarf) of ~0.3g each on both the inside and outside of the tubes. All of the samples were measured in the underground laboratory HADES., (Copyright © 2015 The Authors. Published by Elsevier Ltd.. All rights reserved.)

This study investigated the evolution of uranium distribution and speciation in mill tailings from the COMINAK mine (Niger), in production since 1978. A multi-scale approach was used, which combined high resolution remote sensing imagery, ICP-MS bulk rock analyses, powder X-ray diffraction, Scanning Electron Microscopy, Focused Ion Beam--Transmission Electron Microscopy and X-ray Absorption Near Edge Spectroscopy. Mineralogical analyses showed that some ore minerals, including residual uraninite and coffinite, undergo alteration and dissolution during tailings storage. The migration of uranium and other contaminants depends on (i) the chemical stability of secondary phases and sorbed species (dissolution and desorption processes), and (ii) the mechanical transport of fine particles bearing these elements. Uranium is stabilized after formation of secondary uranyl sulfates and phosphates, and adsorbed complexes on mineral surfaces (e.g. clay minerals). In particular, the stock of insoluble uranyl phosphates increases with time, thus contributing to the long-term stabilization of uranium. At the surface, a sulfate-cemented duricrust is formed after evaporation of pore water. This duricrust limits water infiltration and dust aerial dispersion, though it is enriched in uranium and many other elements, because of pore water rising from underlying levels by capillary action. Satellite images provided a detailed description of the tailings pile over time and allow monitoring of the chronology of successive tailings deposits. Satellite images suggest that uranium anomalies that occur at deep levels in the pile are most likely former surface duricrusts that have been buried under more recent tailings., (Copyright © 2015 Elsevier B.V. All rights reserved.)

In 2011 the joint research project Metrology for Radioactive Waste Management (MetroRWM)(1) of the European Metrology Research Programme (EMRP) started with a total duration of three years. Within this project, new metrological resources for the assessment of radioactive waste, including their calibration with new reference materials traceable to national standards will be developed. This paper gives a review on national, European and international strategies as basis for science-based metrological requirements in clearance and acceptance of radioactive waste., (© 2013 Elsevier Ltd. All rights reserved.)

Safety in nuclear power plants (NPPs) in India is a very important topic and it is necessary to dissipate correct information to all the readers and the public at large. In this article, I have briefly described how the safety in our NPPs is maintained. Safety is accorded overriding priority in all the activities. NPPs in India are not only safe but are also well regulated, have proper radiological protection of workers and the public, regular surveillance, dosimetry, approved standard operating and maintenance procedures, a well-defined waste management methodology, proper well documented and periodically rehearsed emergency preparedness and disaster management plans. The NPPs have occupational health policies covering periodic medical examinations, dosimetry and bioassay and are backed-up by fully equipped Personnel Decontamination Centers manned by doctors qualified in Occupational and Industrial Health. All the operating plants are ISO 14001 and IS 18001 certified plants. The Nuclear Power Corporation of India Limited today has 17 operating plants and five plants under construction, and our scientists and engineers are fully geared to take up many more in order to meet the national requirements.