Search Results (30)

This article is devoted to an overview of the conducted work and the current status of decommissioning of the world’s first BN-350 industrial fast neutron reactor. The reactor was put into operation on 16 July 1973 in Aktau. In 1999, the government of Kazakhstan decided to shut down the reactor, and from that moment to the present, it has been in the decommissioning stage. All work on decommissioning the reactor facility was grouped into five stages. The first stage was completed in 2010 when the spent fuel of the BN-350 reactor was placed for long-term storage. The second stage is nearing completion. Research is currently underway to develop technologies for processing radioactive sodium. The goal of the third and fourth stages of the BN-350 reactor decommissioning is the comprehensive processing of liquid and solid radioactive waste. Now such waste is stored in special storage directly on the territory of the nuclear power plant. Full article

The decommissioning and dismantling of nuclear research reactors can lead to a large amount of low- and intermediate-level radioactive waste. For repositories, the materials must be kept confined and safety must be ensured for extended time spans. Waste is encapsulated in concrete, which leads to alkaline conditions with pH values of 12 and higher. This can be advantageous for some radionuclides due to their precipitation at high pH. For other materials, such as reactive metals, however, it can be disadvantageous because it might foster their corrosion. The Studsvik R2 research reactor contained an AlMg3.5 alloy with a composition close to that of commercial Al5154 for its core internals and the reactor tank. Aluminum corrosion is known to start rapidly due to the formation of an oxidation layer, which later functions as natural protection for the surface. The corrosion can lead to pressure build-up through the accompanied production of hydrogen gas. This can lead to cracks in the concrete, which can be pathways for radioactive nuclides to migrate and must therefore be prevented. In this study, unirradiated rod-shaped samples were cut from the same material as the original reactor tank manufacture. They were embedded in concrete with elevated water–cement ratios of 0.7 compared to regular commercial concrete (ca. 0.45) to ensure water availability throughout all of the experiments. The sample containers were stored in pressure vessels with attached high-definition pressure gauges to read the hydrogen-induced pressure build-up. A second set of samples were exposed in simplified artificial cement–water to study similarities in corrosion characteristics between concrete and cement–water. Additionally, the samples were exposed to concrete and cement–water in free-standing sample containers for deconstructive examinations. In concrete, the corrosion rates started extremely high, with values of more than 10,000 µm/y, and slowed down to less than 500 µm/y after 2000 h, which resulted in visible channels inside the concrete. In the cement–water, the samples showed similar behavior after early fluctuations, most likely caused by the surface coverage of hydrogen bubbles. These trends were further supported by mass loss evaluations. Full article

This study proposes a decision-making model based on the economic assessment of phased decommissioning of energy facilities, specifically focusing on a nuclear power plant (NPP). The objective of the research is to develop and validate an economic assessment methodology for comparing immediate and deferred dismantling strategies for a 1000 MW NPP unit. For economic justification, a comparison of the economic expenses is proposed based on the accumulation of radioactive waste, safety activities, and labour costs for the two options. The methods employed include a multifactorial analysis based on expert assessments, considering economic expenses related to radioactive waste accumulation, safety activities, and labour costs. Criteria with differences exceeding 10% for quantitative indicators and fundamental differences for qualitative indicators were deemed significant; each criterion’s acceptability was weighted accordingly. The key results show that deferred dismantling is economically preferable; the total score for deferred dismantling exceeds that of immediate dismantling by approximately 10% (14.16 points vs. 15.86 points). A comparison of block schedules for decommissioning, dynamics of labour costs, and annual volumes of reprocessed radioactive waste for the baseline and optimised deferred dismantling options shows that both options meet the continuity condition of the ‘active’ stages. At the same time, the optimised option demonstrates significant advantages in the uniformity of labour costs and workload of radioactive waste treatment plants during dismantling. The activities at the stage of power unit decommissioning are proposed to be carried out within the licence framework for its operation by the organisational and technical solutions to ensure safety during operation. The deterministic consequences and risks will align with the safety assessment, which will be determined based on the latest analysis results, taking into account sustainable operation. Full article

This study presents a comprehensive analysis of the microstructural characteristics and chemical composition of base and weld materials from reactor pressure vessels in the first (units 1 and 2) and second (unit 8) generations of Russian VVER 440 reactors at the Greifswald nuclear power plant. We measured the specific activities of ⁶⁰Co and ¹⁴C in activated samples from units 1 and 2. ⁶⁰Co, with its shorter half-life (t_1/2 = 5.27 a), is a key dose-contributing radionuclide during decommissioning, while ¹⁴C (t_1/2 = 5700 a) plays an important role in a geological repository for low- and intermediate-level radioactive waste. Our findings reveal differences in the proportions of trace elements between the base and weld materials as well as between the two reactor generations. Microstructural analysis identified Mo-rich precipitates and (Mn, S)-rich inclusions containing secondary micro-inclusions in the unit 1 and 2 samples. Raman spectroscopy confirmed iron oxides (γ-Fe₂O₃, Fe₃O₄), silicates (Mn-SiO₃), and Cr₂O₃/NiCr₂O₄ in the base metal as well as MnFe₂O₃ in the weld metal. X-ray photoelectron spectroscopy identified Mn inclusions as MnS, MnS₂, or mixed Mn, Fe sulfides, and the Mo precipitates as MoSi₂. These findings offer valuable insights into the speciation of elements and the potential release of radionuclides through corrosion processes under repository conditions. Full article

The Chalk River Laboratories (CRL) site in Ontario, Canada, has long been a hub for nuclear research, which has resulted in the accumulation of legacy nuclear waste, including radioactive materials such as uranium, plutonium, and other radionuclides. Effective management of this legacy requires precise contamination and risk assessments, with a particular focus on the concentration levels of fissile materials such as ${}^{235}\mathrm{U}$. These assessments are essential for maintaining nuclear criticality safety. This study estimates the upper bounds of ${}^{235}\mathrm{U}$ concentrations. We investigated the use of a hybrid parametric bootstrapping method and robust statistical techniques to analyze datasets with outliers, then compared these outcomes with those derived from nonparametric bootstrapping. This study underscores the significance of measuring ${}^{235}\mathrm{U}$ for ensuring safety, conducting environmental monitoring, and adhering to regulatory compliance requirements at nuclear legacy sites. We used publicly accessible ${}^{235}\mathrm{U}$ data from the Eastern Desert of Egypt to demonstrate the application of these statistical methods to small datasets, providing reliable upper limit estimates that are vital for remediation and decommissioning efforts. This method seeks to enhance the interpretation of sensor data, ultimately supporting safer nuclear waste management practices at legacy sites such as CRL. Full article

Nuclear power plant decommissioning requires the rapid and accurate classification of radioactive waste in narrow spaces and under time constraints. Photon-counting detector technology offers an effective solution for the quick classification and detection of radioactive hotspots in a decommissioning environment. This paper characterizes a 5 mm CdTe Timepix3 detector and evaluates its feasibility as a single-layer Compton camera. The sensor’s electron mobility–lifetime product and resistivity are studied across bias voltages ranging from −100 V to −3000 V, obtaining values of ${\mathrm{\mu }}_e{\tau }_e$ = (1.2 ± 0.1) × 10⁻³ cm²V⁻¹, and two linear regions with resistivities of ${\rho }_I=(5.8\pm 0.2) \mathrm{G}\Omega \mathrm{cm}$ and ${\rho }_{II}=(4.1\pm 0.1) \mathrm{G}\Omega \mathrm{cm}$. Additionally, two calibration methodologies are assessed to determine the most suitable for Compton applications, achieving an energy resolution of 16.3 $\mathrm{k}$$\mathrm{e}\mathrm{V}$ for the ¹³⁷Cs photopeak. The electron’s drift time in the sensor is estimated to be (122.3 ± 7.4) ns using cosmic muons. Finally, a Compton reconstruction of two simultaneous point-like sources is performed, demonstrating the detector’s capability to accurately locate radiation hotspots with a ∼51 cm resolution. Full article

In this paper, we study the influence of densified microsilica and colloidal nanosilica admixtures on the mechanical strength and the microstructural characteristics of special mortars used for immobilizing radioactive concrete waste. The experimental program focused on the replacement of cement with micro- and/or nanosilica, in different proportions, in the basic composition of a mortar made with recycled aggregates. The technical criteria imposed for such cementitious systems, used for the encapsulation of low-level radioactive waste, imply high fluidity, increased mechanical strength and lack of segregation and of bleeding. We aimed to increase the structural compactness of the mortars by adding micro- and nanosilica, all the while maintaining the technical criteria imposed, to obtain a cement matrix with high durability and increased capacity for immobilizing radionuclides. The samples from all the compositions obtained were analyzed from the point of view of mechanical strength. Also, micro- and nanosilica as well as samples of the optimal mortar compositions were analyzed physically and microstructurally. Experimental data showed that the mortar samples present maximum compressive strength for a content between 6 and 7.5% wt. of microsilica, respectively, for a content of 2.25% wt. nanosilica. The obtained results suggest a synergistic effect of micro- and nanosilica when they are used simultaneously in cementitious compositions. Thus, among the analyzed compositional variants, the mortar composition with 3% wt. microsilica and 2.25% wt. nanosilica showed the best performance, with an increase in compressive strength of 23.5% compared to the control sample (without micro- and nanosilica). Brunauer–Emmett–Teller (BET) analysis and scanning electron microscopy (SEM) images highlighted the decrease in pore diameter and the increase in structural compactness, especially for mortar samples with nanosilica content or a mixture of micro- and nanosilica. This study is useful in the field of recycling radioactive concrete resulting from the decommissioning of nuclear research or nuclear power reactors. Full article

The Dismantling and Decommissioning (D&D) of nuclear facilities poses several challenges for radioactivity measurement laboratories involved in environmental radiation monitoring plans. One of them is the definition of the detection limits to be achieved for the radionuclides analysis in different samples. The detection limits should be set in such a way that the obtained concentration values for each radionuclide are easily discriminated from certain maximum activity concentration levels. These maximum activity concentration levels are usually set in view of the respective dose contributions from each radionuclide. There are some national legislations that settle detection limits for drinking water. However, there is no regulation containing detection limits for groundwater or surface water. In this way, different institutions or companies require very different detection limits for radioactivity concentration assessment in those types of water associated with D&D activities. In this work, we focus on the detection limits required for the D&D activities in rainwater, surface water and groundwater. We propose detection limits obtained by applying the WHO methodology for maximum activity concentration levels and compare with those requested by radioactive waste management agencies and regulatory bodies. Some real cases where our proposal allows identification of events are analysed and conclusions are extracted. Full article

Due to various historical events, in the Russian Federation, in addition to the radioactive waste storage facilities used in world practice, there are various nuclear and radiation hazardous facilities that require special procedures for monitoring and decommissioning. One of these facilities is the disposal site for LRW on the territory of the JSC Siberian Chemical Plant, where specially prepared waste is injected into sand reservoirs lying at depths of 300–350 m between clayey strata. This study examines in detail the features of the lithological and mineral composition of reservoir sands and aquitards. The processes of environmental transformation in reservoir sands, which lead to changes in the composition and structure of rocks, were characterized. These processes manifest themselves in the form of the development of leaching zones and their “healing” with newly formed smectite, the destruction of terrigenous grains, including the development of cracks, and the growth of newly formed smectite in the pore space of reservoirs. The forms of occurrence and localization of authigenic smectite formed as a result of technogenic impact are described. It has been shown that, despite the obvious impact of highly reactive solutions accompanying liquid radioactive waste, the insulating properties of the geological environment are maintained and even improved to some extent. Full article

Hundreds of thousands of tons of waste are generated from decommissioned nuclear- power facilities, and it has become a critical global issue to secure technology for reducing and recycling this waste. Concrete waste (CW) is estimated to comprise 60–80% of the total waste, and concrete-waste powder (CWP) includes enough inorganic substances used as effective materials for waste treatment. Accordingly, it can be used to produce recycled cement (RC). This study aimed to evaluate the performance of a solidification agent manufactured using recycled cement (SRC) for the safe packing of radioactive wastes, such as coarse aggregates of CW, waste soil, and metal wastes originating from decommissioned nuclear facilities. The experimental results indicated that the most relevant incineration temperature of CWP for RC was 700 °C. The optimum water-to-binder ratio was determined to be 0.4, and the most relevant substitution ratio of ground granulated blast furnace slag for CWP was determined to be 15%. In addition, calcium silicate hydrate is the most effective hydration product for improving the compressive strength of SRC. The maximum packing capacities of the SRC for coarse aggregates, waste soil, and metal waste, which were simulated as radioactive wastes, were determined to be 30, 5, and 7 wt%, respectively. The results of leaching tests using SRC containing radioactive wastes contaminated with Co, Cs, and Sr indicated that their leachability indices met the acceptance level for disposal. Consequently, the RC composed of CWP can be used as a solidifying agent to safely dispose of radioactive wastes, such as coarse aggregates, waste soil, and metal waste. Full article

The decommissioning process of nuclear power facilities renders hundreds of thousands of tons of various types of waste. Of these different waste types, the amount of concrete waste (CW) varies greatly depending on the type of facility, operating history, and regulation standards. From the previous decommissioning projects, CW was estimated to comprise 60–80 wt.% of the total weight of radioactive wastes. This represents a significant technical challenge to any decommissioning project. Furthermore, the disposal costs for the generated concrete wastes are a substantial part of the total budget for any decommissioning project. Thus, the development of technologies effective for the reduction and recycling of CW has become an urgent agenda globally. Blast furnace slag (BFS) is an industrial byproduct containing a sufficient amount (higher than 30%) of CaO and it can be used as a substitute for ordinary Portland cement (OPC). However, there have been few studies on the application of BFS for the treatment of radioactive waste from decommissioning processes. This study was conducted to evaluate the performance of the solidification agent using ground granulated BFS (SABFS) to pack radioactive wastes, such as the coarse aggregates of CW (CACW), waste soil (WS), and metal waste (MW). The analytical results indicated that the CaO content of the ground granulated BFS was 36.8% and it was confirmed that calcium silicate hydrate (CSH) could be activated as the precursor of the hydration reactions. In addition, the optimum water-to-binder ratio was determined to be 0.25 and Ca(OH)₂ and CaSO₄ were found to be the most effective alkaline and sulfate activators for improving the compressive strength of the SABFS. The maximum packing capacities of the SABFS were determined to be 9 and 13 wt.% for WC and WM, respectively, when the content of CW was fixed at 50 wt.%. The results of the leaching tests using SABFS containing radioactive wastes contaminated with Co, Cs, and Sr indicated that their leachability indices met the acceptance level for disposal. Consequently, the SABFS can be used as a solidifying agent for the safe disposal of radioactive waste. Full article

Kori Unit 1, which was permanently shut down on 18 June 2017, is planned to undergo full-system decontamination prior to major decommissioning activities. One of the advantages of performing FSD is the downgrading of the classification level of radioactive waste. From this perspective, the impact on the steam generator (SG) tubes, which account for a considerable portion of the total surface area during FSD operation, was examined. Initially, the CRUDTRAN code was used to predict the radioactivity inventory of the Kori Unit 1 SG tubes, which turned out to be approximately 21% more conservative than the measured value. To estimate the radioactivity in the tubes after FSD, decontamination factor values from overseas cases in the SG tubes section were selected and applied. Then, the regulations for radioactive waste in Korea were reviewed, and the specific activity was calculated by predicting the mass of the SG tubes. As a conclusion, it was confirmed that the SG tubes will be classified as low-level radioactive waste, whether FSD is performed or not. Furthermore, it was observed that even if a high efficiency of FSD is achieved, if stored, it would take more than 50 years for clearance. Full article

Across the world, any activity associated with the nuclear fuel cycle such as nuclear facility operation and decommissioning that produces radioactive materials generates ultramodern civilian radioactive waste, which is quite hazardous to human health and the ecosystem. Therefore, the development of effectual and commanding management is the need of the hour to make certain the sustainability of the nuclear industries. During the management process of waste, its immobilization is one of the key activities conducted with a view to producing a durable waste form which can perform with sustainability for longer time frames. The cementation of radioactive waste is a widespread move towards its encapsulation, solidification, and finally disposal. Conventionally, Portland cement (PC) is expansively employed as an encapsulant material for storage, transportation and, more significantly, as a radiation safeguard to vigorous several radioactive waste streams. Cement solidification/stabilization (S/S) is the most widely employed treatment technique for radioactive wastes due to its superb structural strength and shielding effects. On the other hand, the eye-catching pros of cement such as the higher mechanical strength of the resulting solidified waste form, trouble-free operation and cost-effectiveness have attracted researchers to employ it most commonly for the immobilization of radionuclides. In the interest to boost the solidified waste performances, such as their mechanical properties, durability, and reduction in the leaching of radionuclides, vast attempts have been made in the past to enhance the cementation technology. Additionally, special types of cement were developed based on Portland cement to solidify these perilous radioactive wastes. The present paper reviews not only the solidification/stabilization technology of radioactive wastes using cement but also addresses the challenges that stand in the path of the design of durable cementitious waste forms for these problematical functioning wastes. In addition, the manuscript presents a review of modern cement technologies for the S/S of radioactive waste, taking into consideration the engineering attributes and chemistry of pure cement, cement incorporated with SCM, calcium sulpho–aluminate-based cement, magnesium-based cement, along with their applications in the S/S of hazardous radioactive wastes. Full article

For the medical linear accelerators (linac) that utilize more than 10 MV of photon energy, components inside the linac head become radioactivate during the 10–15-year operating cycle. Prior to disposal, radioactive waste must be evaluated for activity, and the same procedure should be followed for medical linacs. In the Republic of Korea, regulation and methodology for the radioactivity evaluation for the medical linac is not established yet. In this study, we employed gamma spectroscopy and a survey meter for evaluating the radioactivity of medical linac components. The components of the Siemens linac considered in this study were classified after decommissioning, and dose rates were measured to up to a 5 cm distance from the component surfaces by using a survey meter. Radionuclides from components were detected using an in situ HPGe detector. Based on the type of radionuclides and dose rate, we estimated the radioactivity of the components. We studied the feasibility of the methodology for disposing of radioactive components by using the in situ HPGe detector. Full article

Separation of hydrated cement paste from aggregate is a key technology to reduce the amount of radioactive concrete waste during the decommissioning process. If separated cement-paste portions can be recycled as a solidifying agent for other radioactive waste, the amount of radioactive concrete waste could be close to “zero”. A study was conducted to achieve circular economy in the area of concrete decommissioning and found it to be successfully used as a solidifying agent for immobilization of liquid radioactive waste. However, previous work used a process that requires large amounts of energy (heat treatment was applied to most of the concrete fraction) because the objective was to completely remove hydrated cement powder from the aggregate. In this work, the separation system was modified to increase energy efficiency (heat treatment was applied to separated powder only), but such a change decreased the surface area of the recycled cement powder due to a higher inclusion of aggregate powder. A relatively lower solution to binder ratio could have been achieved for the preparation of wasteform specimens, and as a result, a 28 day compressive strength of wasteform could have become higher, but the final leachability indices were lower than the results observed from previous work. The results from 28 day compressive strength, thermal cycling and 90 day leaching experiments met the acceptance criteria for wasteform, indicating that this modified system can also be used for immobilization of liquid radioactive waste to meet the “zero” production of concrete waste during the decommissioning of a nuclear power plant. It should be noted that accurate monitoring of aggregate content in recycled cement powder during production is important to maintain proper reactivity of recycled cement powder. Full article