Search Results (555)

Nuclear power is a low-emission and economically competitive energy source, yet the effective disposal and management of its associated radioactive waste can be challenging. Radioactive waste can be categorised as high-level waste (HLW), intermediate-level waste (ILW), and low-level waste (LLW). LLW primarily comprises materials contaminated during routine clean-up, such as mop heads, paper towels, and floor sweepings. While LLW is less radioactive compared to HLW and ILW, the management of LLW poses significant challenges due to the large volume that requires processing and disposal. The volume of LLW can be significantly reduced through sorting, which is typically performed manually in a labour-intensive way. Smart management techniques, such as computer vision (CV) and machine learning (ML), have great potential to help reduce the workload and human errors during LLW sorting. This paper provides a comprehensive review of previous research related to LLW sorting and a summative review of existing applications of CV in solid waste management. It also discusses state-of-the-art CV and ML algorithms and their potential for automating LLW sorting. This review lays a foundation for and helps facilitate the applications of CV and ML techniques in LLW sorting, paving the way for automated LLW sorting and sustainable LLW management. Full article

Some of nuclear power’s primary detractors are the unique environmental challenges and impacts of radioactive wastes generated during fuel cycle operations. Key benefits of spent fuel reprocessing (SFR) are reductions in primary high active waste (HAW) masses, volumes, and lengths of radiotoxicity at the expense of secondary waste generation and high capital and operational costs. By employing advanced waste management and resource recovery concepts in SFR beyond the existing standard PUREX process, such as minor actinide and fission product partitioning, these challenges could be mitigated, alongside further reductions in HAW volumes, masses, and duration of radiotoxicity. This work assesses various current and proposed SFR and fuel cycle options as base cases, with further options for fission product partitioning of the high heat radionuclides (HHRs), rare earths, and platinum group metals investigated. A focus on primary waste outputs and the additional energy that could be generated by the reprocessing of high-burnup PWR fuel from Gen III(+) reactors using a simple fuel cycle model is used; the effects of 5- and 10-year spent fuel cooling times before reprocessing are explored. We demonstrate that longer cooling times are preferable in all cases except where short-lived isotope recovery may be desired, and that the partitioning of high-heat fission products (Cs and Sr) could allow for the reclassification of traditional raffinates to intermediate level waste. Highly active waste volume reductions approaching 50% vs. PUREX raffinate could be achieved in single-target partitioning of the inactive and low-activity rare earth elements, and the need for geological disposal could potentially be mitigated completely if HHRs are separated and utilised. Full article

Nuclear energy can help stop climate change by generating large amounts of emission-free electricity. Nuclear reactor designs are continually being developed to be more fuel efficient, safer, easier to construct, and to produce less nuclear waste. The term advanced nuclear reactors refers either to Generation III+ and Generation IV or small modular reactors. Every reactor is associated with the nuclear fuel cycle that must be economically viable and competitive. An important matter is optimization of fissile materials used in reactor and/or reprocessing of spent fuel and reuse. Currently operating reactors use the open cycle or partially closed cycle. Generation IV reactors are intended to play a significant role in reaching a fully closed cycle. At the same time, we can observe the growing interest in development of small modular reactors worldwide. SMRs can adopt either fuel cycle; they can be flexible depending on their design and fuel type. Spent nuclear fuel management should be an integral part of the development of new reactors. The proper management methods of the radioactive waste and spent fuel should be considered at an early stage of construction. The aim of this paper is to highlight the challenges related to reprocessing of new forms of nuclear fuel. Full article

This article provides a comprehensive review of gas flow behavior in low-permeability geological media, focusing on its implications for the long-term performance of engineered barriers in underground radioactive waste repositories. Key mechanisms include two-phase flow and gas-driven fracturing, both critical for assessing repository safety. Understanding the generation and migration of gas is crucial for the quantitative assessment of repository performance over extended timescales. The article synthesizes the current research on various types of claystone considered as potential host rocks for repositories, providing a comprehensive analysis of gas transport mechanisms and constitutive models. In addressing the challenges related to multi-field coupling, the article provides practical insights and outlines potential solutions and areas for further research, underscoring the importance of interdisciplinary collaboration to tackle these challenges and push the field forward. In addition, the article evaluates key research projects, such as GMT, FORGE, and DECOVALEX, shedding light on their methodologies, findings, and significant contributions to understanding gas migration in low-permeability geological media. In this context, mathematical modeling becomes indispensable for predicting long-term repository performance under hypothetical future conditions, enhancing prediction accuracy and supporting long-term safety assessments. Finally, the growing interest in gas-driven fracturing is explored, critically assessing the strengths and limitations of current numerical simulation tools, such as TOUGH, the phase-field method, and CODE_BRIGHT. Noteworthy advancements by the CODE_BRIGHT team in gas injection simulation are highlighted, although knowledge gaps remain. The article concludes with a call for innovative approaches to simulate gas fracturing processes more effectively, advocating for advanced modeling techniques and rigorous experimental validation to address existing challenges. Full article

Complex oxides with the general formula Gd₂(Ti_1−xZr_x)₂O₇ are promising candidates for radioactive waste immobilization due to their capacity to withstand radiation by dissipating part of the free energy driving defect creation and phase transitions. In this study, samples with varying zirconium content (xZr = 0.00, 0.15, 0.25, 0.375, 0.56, 0.75, 0.85, 1.00) were synthesized via the sol–gel method and thermally treated at 500 °C to obtain nanosized powders mimicking the defective structure of irradiated materials. Synchrotron-based techniques were employed to investigate their structural properties: High-Resolution X-ray Powder Diffraction (HR-XRPD) was used to assess long-range structure, while Pair Distribution Function (PDF) analysis and Extended X-ray Absorption Fine Structure (EXAFS) spectroscopy provided insights into the local structure. HR-XRPD data revealed that samples with low Zr content (xZr ≤ 0.25) are amorphous. Increasing Zr concentration led to the emergence of a crystalline phase identified as defective fluorite (xZr = 0.375, 0.56). Samples with the highest Zr content (xZr ≥ 0.75) were fully crystalline and exhibited only the fluorite phase. The experimental G(r) functions of the fully crystalline samples in the low r range are suitably fitted by the Weberite structure, mapping the relaxations induced by structural disorder in defective fluorite. These structural insights informed the subsequent EXAFS analysis at the Zr-K and Gd-L₃ edges, confirming the splitting of the cation–cation distances associated with different metal species. Moreover, EXAFS provided a local structural description of the amorphous phases, identifying a consistent Gd-O distance across all compositions. Full article

This study systematically explored the performance of a vacuum membrane distillation (VMD) system equipped with polytetrafluoroethylene (PTFE) hollow fiber membranes for treating simulated, acidic, low-level radioactive liquid waste. By focusing on key operational parameters, including feed temperature, vacuum pressure, and flow velocity, an orthogonal experiment was designed to obtain the optimal parameters. Considering the potential application scenarios, the following two factors were also studied: the initial nuclide concentrations (0.5, 5, and 50 mg·L⁻¹) and tributyl phosphate (TBP) concentrations (0, 20, and 100 mg·L⁻¹) in the feed solution. The results indicated that the optimal operational parameters for VMD were as follows: a feed temperature of 70 °C, a vacuum pressure of 90 kPa, and a flow rate of 500 L·h⁻¹. Under these parameters, the VMD system demonstrated a maximum permeate flux of 0.9 L·m⁻²·h⁻¹, achieving a nuclide rejection rate exceeding 99.9%, as well as a nitric acid rejection rate of 99.4%. A significant negative correlation was observed between permeate flux and nuclide concentrations at levels above 50 mg·L⁻¹. The presence of TBP in the feed solution produced membrane fouling, leading to flux decline and a reduced separation efficiency, with severity increasing with TBP concentration. The VMD process simultaneously achieved nuclide rejection and nitric acid concentration in acidic radioactive wastewater, demonstrating strong potential for nuclear wastewater treatment. Full article

This study examines methods to improve the energy efficiency of radiofrequency inductively coupled plasma (RF-ICP) torches for radioactive waste treatment, with a focus on surpassing the typical energy efficiency limit of approximately 70%. To improve energy efficiency and plasma performance, this research investigates the transition from axial gas flow to vortex gas flow patterns using COMSOL Multiphysics software v6.2. Key plasma parameters, including energy efficiency, number of gas vortices, heat transfer, and temperature distribution, were analyzed to evaluate the improvements. The results indicate that adopting a vortex flow pattern increases energy conversion efficiency, increases heat flux, and reduces charge losses. Furthermore, optimizing the torch body design, particularly the nozzle, chamber volume, and gas entry angle, significantly improves plasma properties and energy efficiency by up to 90%. Improvements to RF-ICP torches positively impact waste decomposition by creating better thermal conditions that support resource recovery and potential material recycling. In addition, these improvements contribute to reducing secondary waste, mitigating environmental risks, and fostering long-term public support for nuclear technology, thereby promoting a more sustainable approach to waste management. Simulation results demonstrate the potential of RF-ICP flares as a cost-effective and sustainable solution for the thermal treatment of low- to intermediate-level radioactive waste. Full article

The safe disposal of high-level radioactive waste (HLW) is imperative for sustaining China’s rapidly expanding nuclear power sector, with deep geological repositories requiring rigorous site evaluation via underground research laboratories (URLs). This study presents a controlled-source audio-frequency magnetotellurics (CSAMT) survey at the Xinchang site in China’s Beishan area, a region dominated by high-resistivity metamorphic rocks. To overcome electrical data acquisition challenges in such resistive terrains, salt-saturated water was applied to transmitting and receiving electrodes to enhance grounding efficiency. Using excitation frequencies of 9600 Hz to 1 Hz, the survey achieved a 1000 m investigation depth. Data processing incorporated static effect removal via low-pass filtering and smoothness-constrained 2D inversion. The results showed strong consistency between observed and modeled data, validating inversion reliability. Borehole correlations identified a 600-m-thick intact rock mass, confirming favorable geological conditions for URL construction. The study demonstrates CSAMT’s efficacy in characterizing HLW repository sites in high-resistivity environments, providing critical geophysical insights for China’s HLW disposal program. These findings advance site evaluation methodologies for deep geological repositories, though integrated multidisciplinary assessments remain essential for comprehensive site validation. This work underscores the feasibility of the Xinchang site while establishing a technical framework that is applicable to analogous challenging terrains globally. Full article

This study investigates the mechanical and microstructural performance of three alkali-activated geopolymer formulations, constituted of metakaolin (MK), blast furnace slag (BFS), and a ternary blend of MK, BFS, and fly ash (MIX), for the immobilization of simulated radioactive liquid organic waste (RLOW). Thermal ageing tests were performed to evaluate geopolymer durability, including fire exposure (800 °C) and climatic chamber cycles (from −20 to 40 °C). Characterization through thermogravimetric analysis (TGA), compression tests, and scanning electron microscopy with energy-dispersive spectroscopy (SEM-EDS) was carried out to assess material degradation after thermal ageing. Preliminary results showed substantial strength and microstructural degradation in oil-loaded specimens after cyclic climatic ageing, while fire-exposed blank matrices retained partial mechanical integrity. BFS matrices exhibited the best thermal resistance, attributable to the formation of Ca-Al-Si-hydrate (C-A-S-H) gels. These findings support the use of optimized geopolymer formulations for safe RLOW immobilization, while contributing to the advancement of knowledge on sustainable and regulatory-compliant direct conditioning technology. Full article

In this work, hybrid membranes were fabricated by depositing polyvinyl chloride (PVC) fibers onto PET track-etched membranes (TeMs) using the electrospinning technique. The resulting structures exhibited enhanced hydrophobicity, with contact angles reaching 155°, making them suitable for applications in both water–oil mixture separation and membrane distillation processes involving low-level liquid radioactive waste (LLLRW), saline solutions, and natural water sources. The use of hybrids of TeMs and nanofiber membranes has significantly increased productivity compared to TeMs only, while maintaining a high degree of purification. Permeate obtained after MD of LLLRW and river water was analyzed by conductometry and the atomic emission spectroscopy (for Sr, Cs, Al, Mo, Co, Sb, Ca, Fe, Mg, K, and Na). The activity of radioisotopes (for ¹²⁴Sb, ⁶⁵Zn, ⁶⁰Co, ⁵⁷Co, ¹³⁷Cs, and ¹³⁴Cs) was evaluated by gamma-ray spectroscopy. In most cases, the degree of rejection was between 95 and 100% with a water flux of up to 17.3 kg/m²·h. These membranes were also tested in the separation of cetane–water emulsion with productivity up to 47.3 L/m²·min at vacuum pressure of 700 mbar and 15.2 L/m²·min at vacuum pressure of 900 mbar. Full article

For the first time, the deposition area of heavy metals and other trace elements (Al, Ba, Cd, Co, Cr, Cu, Fe, Mn, Ni, P, Pb, S, Sr, Sb, V, Zn, and Hg) on the territory surrounding a landfill of domestic (municipal) waste at the 9th km of the Vilyuisky tract of Yakutsk within a radius of 51 km was assessed using the method of moss biomonitors and ICP-OES as an analytical technique. Mosses were analyzed for radionuclide content (⁴⁰K, ¹³⁷Cs, ²¹² Pb, ²¹⁴Pb, ²¹²Bi, ²¹⁴Bi, ²⁰⁸Tl, ⁷Be, and ²²⁸Ac) in a number of selected samples by semiconductor gamma spectrometry. The results of the examination of moss samples by ICP-OES indicate the presence of large amounts of toxic Ba and metal debris (Al, Co, Cr, Fe, S, and Pb) at the landfill. In addition, it is shown that the investigated samples contain elements such as Cd, Co, Cr, Cu, Cu, Mn, Ni, Pb, Sr, V, Zn, and Hg. The method of gamma spectrometry revealed that the studied samples contain such radioactive elements as ¹³⁷Cs, daughter products of ²³⁸U and ²³²Th. Detection of the same heavy metals and radionuclides in the atmospheric air of the city and in the vegetation near the landfill may indicate that one of the sources of environmental pollution may be products of incineration of the landfill contents at the 9th km of the Vilyuisky tract. Full article

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

The geological containment of high-level radioactive waste has become widely accepted among international organizations, and it has been adopted by many countries as part of their national nuclear waste disposal plan. The multi-barrier system, including the compacted bentonite blocks or pellets serving as human-made containment or buffer media, is the key component of high-level radioactive waste disposal, which contains a waste canister that isolates the nuclear waste from a human being geosphere for one million years. The bentonite clay surrounding the nuclear waste capsule is subjected to prolonged exposure to elevated temperatures because of the continuous decay of radioactivity. Long-term heating at high temperatures could change the buffers’ microstructural characteristics and physicochemical and hydromechanical properties, which can influence their self-sealing ability. This paper offers a comprehensive overview of the current understanding of thermal effects on bentonite-based buffer systems. The thermal impact on the microstructure, Atterberg limits, and swelling pressure of bentonite clay are intensely reviewed, and the findings are summarized. This review paper highlights new insights into the design of multi-layered containment approaches for high-level radioactive waste isolation. Full article

Different from general underground engineering, the micro-damage prior to failure of the surrounding rock has a significant influence on the geological disposal of high-level radioactive waste. However, the quantitative research on pre-failure dilatancy damage characteristics and stress path influence of hard brittle rocks under high stress levels is insufficient currently, and especially, the stress path under simultaneous unloading of axial and confining pressures is rarely discussed. Therefore, three representative mechanical experimental studies were conducted on the Beishan granite in the pre-selected area for high-level radioactive waste (HLW) geological disposal in China, including increasing axial pressure with constant confining pressure (path I), increasing axial pressure with unloading confining pressure (path II), and simultaneous unloading of axial and confining pressures (path III). Using the deviatoric stress ratio as a reference, the evolution laws and characteristics of stress–strain relationships, deformation modulus, generalized Poisson’s ratio, dilatancy index, and dilation angle during the path bifurcation stage were quantitatively analyzed and compared. The results indicate that macro-deformation and the plastic dilatancy process exhibit strong path dependency. The critical value and growth gradient of the dilatancy parameter for path I are both the smallest, and the suppressive effect of the initial confining pressure is the most significant. The dilation gradient of path II is the largest, but the degree of dilatancy before the critical point is the smallest due to its susceptibility to fracture. The critical values of the dilatancy parameters for path III are the highest and are minimally affected by the initial confining pressure, indicating the most significant dilatancy properties. Establish the relationship between the deformation parameters and the crack-induced volumetric strain and define the damage variable accordingly. The critical damage state and the damage accumulation process under various stress paths were examined in detail. The results show that the damage evolution is obviously differentiated with the bifurcation of the stress paths, and three different types of damage curve clusters are formed, indicating that the damage accumulation path is highly dependent on the stress path. The research findings quantitatively reveal the differences in deformation response and damage characteristics of Beishan granite under varying stress paths, providing a foundation for studying the nonlinear mechanical behavior and damage failure mechanisms of hard brittle rock under complex loading conditions. Full article