In the glass solidification process of high-level radioactive liquid waste (HLLW) in cold crucibles, the temperature of glass melt is a key parameter that determines the quality of glass matrix and the stable operation of the process. However, under the extreme coupled conditions of high temperature (> 1200°C), strong radiation, and severe corrosion, traditional contact-type thermocouple temperature measurement devices are confronted with problems such as high-temperature melting, severe corrosion, and mechanical fracture, resulting in short service life and high frequency of maintenance and replacement, which seriously restricts the stable operation of the process. To overcome the above limitations, this paper proposes a cold crucible virtual temperature measurement model based on a temporal neural network. The model takes the high-frequency power supply current and voltage, stirrer rotation speed and position, and historical temperature data collected during the 2000-h continuous operation of the cold crucible engineering prototype as training samples and constructs a temporal neural network with large-delay characteristics. This network can predict the glass melt temperature at the next moment based on the process data of the previous 5000 s, with a mean squared error (MSE) of 2.54 × 10−3 and a prediction accuracy of about 85% on the test set. To further improve the prediction performance, this study designs a two-stage hybrid neural network, which adds a short-term memory network on the basis of the original temporal neural network. This hybrid model can correct the temperature at the next moment based on the process data of the previous 500 s and the corresponding predicted temperature. It should be noted that during the start-up phase of this hybrid model, it is necessary to rely on the operation of a physical temperature measurement device for at least 500 s and use the temperature data output by the device as the model input to initiate temperature prediction. Experimental results show that the virtual temperature measurement model achieves an MSE of 8.97 × 10−4 and a prediction accuracy of about 95% on the test set, which is about 12% higher than that of the ordinary temporal neural network. This method effectively avoids the limitations of physical temperature measuring instruments in extreme environments, provides a new high-reliability and low-maintenance–frequency temperature measurement approach for the HLW glass solidification process, and has important theoretical and engineering significance for promoting the intelligence and operational safety of nuclear waste treatment equipment in China.
We present a brief overview of the current state of radioactive (or nuclear) waste (RW) management (RWM), focusing on its end point and international advice provided by the International Atomic Energy Agency (IAEA). RW, defined as waste that contains, or is contaminated with, radionuclides at activity concentrations greater than clearance levels set by the national regulatory organisations with guidance from the IAEA, is a by-product of nuclear energy generation, research, the production of life-saving radiopharmaceuticals and industrial applications (e.g., disused radioactive sources from geophysical logging instruments). RW is also produced from processing of materials containing naturally occurring radioactive materials (NORMs) such as those typically occurring in oil and gas production (e.g., mineral scale, produced water), ore beneficiation (e.g., tailings and sludges) and water purification (e.g., filter media, reverse osmosis membranes). RWM envisions all administrative and operational activities involved in the handling, pretreatment, treatment, conditioning, transport, storage and disposal of RW. Disposal involves emplacement of RW in a dedicated facility without the intention of retrieval, requiring waste matrices and waste packages, which conform with waste acceptance criteria (WAC); safety of disposal is demonstrated on the basis of operational and postclosure safety cases. By international consensus, the disposal is the end point of RWM activities. We discuss RW classification, give examples of RW inventories, summarise general considerations regarding RWM and provide an overview of global experience and current trends in geological disposal.
A discrete ordinates (SN) method has been applied to the solution of the steady-state multigroup neutron transport equation in Cartesian geometry. Angular and spatial discretization were performed using the (SN) scheme and the finite difference method, respectively, with isotropic and anisotropic scattering treated up to arbitrary order. The algorithms were implemented in a FORTRAN code named n-DOTEC. The contribution of this work lies in the development and verification of a computational framework for one- and two-dimensional fixed-source transport problems, capable of handling multiple energy groups and anisotropic scattering expansions of arbitrary order, and using PETSc together with MUMPS for the solution of the resulting sparse linear systems. In the two-dimensional case, the code includes a product quadrature (PQ) developed by the authors, based on Gauss-Legendre quadrature for both polar and azimuthal angles, with point weights defined as the product of the corresponding one-dimensional weights. The code was validated against several one-dimensional benchmarks and a two-dimensional benchmark problem. Test cases include fixed-source problems with vacuum and reflective boundary conditions under demanding configurations that require high-order (SN) to achieve accuracy. The numerical results demonstrate good agreement with analytical solutions and reference codes such as DANTSYS and PARTISN. Reported mean relative errors were below 1% for most cases, and as low as 0.0005% for a two-group anisotropic scattering benchmark, confirming the robustness of the implementation. The method efficiently computes flux distributions for any number of energy groups, and its simple formulation makes it a reliable tool for transport calculations. In addition, n-DOTEC has potential applications in generating variance-reduction parameters, such as weight windows, for Monte Carlo simulations.
This paper introduces a simple and rapid separation method for the determination of 55 Fe and 63 Ni in lead–bismuth alloys. This method is based on chemical precipitation and solid‐phase extraction chromatography techniques, which can separate and purify individual radionuclides from sample matrix elements and other radionuclides. First, lead, bismuth, and iron are separated by chemical precipitation. On the one hand, nickel is separated on a NI‐resin column from the filtered solution. On the other hand, after the precipitation is digested, iron is further separated from lead and bismuth by chemical precipitation. This study explored the effects of solution pH, ammonium citrate concentration, and eluent concentration (acid concentration) on the retention behavior of the NI‐resin. The optimal separation efficiency for nickel was achieved under the following conditions: a solution pH of approximately 9, an ammonium citrate concentration of 0.2 mol/L, and an eluent (nitric acid) concentration of 3 mol/L. The recovery rates of Fe and Ni obtained by this method are 88.24% and 81.57%, respectively, and the decontamination factors of Pb and Bi are 6.5 × 10 2 and 1.4 × 10 4 , respectively.
The fluorination furnace is a key facility in uranium conversion for producing uranium hexafluoride. However, high‐speed uranium tetrafluoride powders often cause severe erosion in the feed pipe and reaction zone of the facility, leading to structural failure and potential leakage of radioactive and toxic substances. Despite its importance, erosion behavior under the facility remains poorly understood. In this study, a simplified 3D numerical model of the furnace was developed. Using multiphysics coupling methods, the erosion phenomena of varying inclination angles and inlet velocities in the feed pipe were analyzed. Results show that increasing the inclination angle intensifies erosion and localizes damage near the opposite wall of the feed pipe outlet in the reaction zone. An inlet velocity threshold (4 to 5 m/s) marks a transition from sliding to impact‐dominated wear. In contrast, erosion in the feed pipe remains uniform but increases near the outlet. This work provides qualitative insights into erosion‐prone hotspot and offers references for structural design and safety improvement of fluorination facility in nuclear fuel cycling.
BackgroundTo address the safety threat of atmospheric dispersion of petrochemical hazardous substances to the nuclear steam supply plants (NSSPs) adjacent to petrochemical parks, this study systematically investigates the mountain barrier effect on hazardous substance dispersion and its safety impact on NSSP via 3D computational fluid dynamics (CFD) numerical simulation.MethodsThree-dimensional numerical models are established for 10 terrain scenarios, including flat terrain and 9 combined mountain conditions covering 3 heights (50, 80, and 110 m) coupled with 3 slope gradients (10 degrees, 20 degrees, and 30 degrees). Full transient simulations were conducted for 3 release velocities, 3 ambient wind regimes, and 3 hazardous substances with different densities. The highest-risk, most conservative working condition was identified, and the lower explosive limit (LEL) was adopted as the critical hazardous threshold for quantitative risk analysis.ResultsThe results show that (1) the combination of 20 m/s high-speed release, 3 m/s ambient wind speed at 10 m height (wind profile power exponent 0.4), and high-density diesel vapor is the most conservative high-risk condition, featured by the strongest pollutant gravity settlement, weakest turbulent dilution, and longest over-limit dispersion range. (2) The relative height between the mountain and reactor containment is the core dominant factor for NSSP pollution risk. The 110-m mountain (30 m higher than the containment) delivers the optimal protection: under 20 degrees slope, it guides over 92% of pollutants to diffuse above the 80-m core protection height, reduces the NSSP front pollutant mass fraction to 0.1% of the flat terrain baseline (only 2.6% of the hazardous threshold), and limits the over-limit range to within 500 m downwind of the source. The 80-m equal-height mountain can reduce the NSSP front concentration below the safety threshold at slopes >= 20 degrees. In contrast, the 50-m low mountain has no positive protection effect, with near-field (400-500 m) low-altitude pollutant accumulation amplified by up to 153.3% compared to flat terrain and NSSP front concentration amplified consistently 4.0-4.3 times the hazardous threshold. (3) Mountain slope exerts a significant nonlinear regulatory effect on protection performance, with 20 degrees being the optimal critical gradient for 110-m high mountains.ConclusionsThis study identifies that a mountain height not lower than the NSSP reactor containment is the prerequisite for an effective barrier effect and reveals the chain regulation mechanism of "flow field reconstruction, diffusion path alteration, and concentration redistribution" of mountain terrain on heavy pollutant dispersion. The results provide a quantitative basis and technical support for site selection and layout optimization of NSSP adjacent to petrochemical parks.
Scenario-based dose assessment tools such as VRdose are widely used for radiological planning of high-exposure maintenance tasks at nuclear power plants; however, these tools often yield conservative predictions that systematically exceed actual worker exposure. This study aims to quantify and interpret this discrepancy by deriving task-specific correction factors for the installation and removal of steam generator nozzle dams at Korean pressurized water reactors. Ten years (2012-2021) of occupational exposure data reported in Korea Hydro and Nuclear Power annual dose reports were analyzed and compared with VRdose simulation results. To ensure a consistent comparison independent of workforce size and worker exposure time, both measured collective doses and simulated results were converted into time-normalized dose rates. Annual correction factors were calculated as the ratio of actual to predicted dose rates, and their long-term average value and statistical uncertainty were evaluated using 95% and 99% confidence intervals. The results show that VRdose consistently overpredicted worker exposure across all analyzed years. Predicted dose rates exceeded measured values by approximately a factor of 4-10. This systematic overestimation reflects conservative planning assumptions related to exposure time, radiation field representation, and shielding conditions rather than modeling deficiencies. The derived correction factors, therefore, quantify the inherent conservatism of the VRdose scenario for this specific high-exposure task. Rather than serving as universal adjustment coefficients or deterministic predictors of future doses, the proposed correction factors provide interpretive planning references that allow radiation protection officers to contextualize scenario-based predictions within realistic operational bounds. By integrating these factors into outage planning, more balanced ALARA-driven decisions regarding manpower allocation, worker rotation, shielding, and task sequencing can be achieved. The methodology presented in this study offers a practical framework for transforming conservative scenario outputs into actionable radiological planning information and may be extended to other high-exposure maintenance tasks through task- and unit-specific implementation.
The tip clearance is an essential structure in multiphase pumps, and the vortex flow caused by the leakage at the tip of the blade disturbs the internal flow field of the pump, leading to energy losses, which in turn weaken the pump's pressure-boosting performance and reduce its work output efficiency. To clarify the energy loss behavior at various tip clearances, the study examines the effects of tip clearance and IGVF on energy dissipation using entropy production theory. The energy losses under the coupled operating conditions of gas-liquid two-phase flow and tip clearance are further revealed. Increasing tip clearance leads to greater impeller energy dissipation. The IGVF modifies the dissipation distribution on the impeller for different clearances. Entropy production on the impeller wall is primarily focused on the blade pressure side, at the inlet leading edge, within the tip clearance, and near the hub close to the inlet. In addition, a larger tip clearance tends to decrease the friction loss associated with the impeller blades. For different IGVFs, the turbulent dissipation inside the impeller shows relatively weak dependence on tip clearance. Tip clearance loss and turbulent dissipation loss exhibit different magnitudes of variation and dominance under different IGVFs. The maximum turbulent dissipation rate reached 4640 m2/s3, the maximum entropy production in the impeller reached 2.02 & times; 104 W/(m3 & centerdot;K) at IGVF = 15% and Rtc = 0.5 mm, and the tip clearance loss at IGVF = 5% and Rtc = 1.5 mm was nearly 50% higher than that under the other conditions.
In the design of nuclear power plant systems, the thermal-hydraulic response in the containment during a loss-of-coolant accident (LOCA) plays a crucial role. In this study, the advanced small modular reactor NHR200-II is selected as the research object. The transient responses of the containment during a typical small break LOCA (SBLOCA) with different containment models are modeled and compared, including both the lumped parameter approach with a single volume and the distributed parameter approach with subdivided volumes. The analysis results revealed that the containment pressure responses-a key thermal-hydraulic parameter during SBLOCA transients in the NHR200-II design-show only minor differences when modeled using the two approaches. These findings indicate that the lumped parameter model can be sufficient to satisfy the requirements for predicting containment pressure responses in the NHR200-II design. However, further details of thermal-hydraulic phenomena and processes can be achieved through the distributed containment model, such as temperature distribution and fluid velocity within the containment. Notably, it was observed that heat transfer between the vapor and containment shell is a significant heat removal mechanism during the SBLOCA transients in the NHR200-II. This finding emphasizes the importance of considering heat transfer through the containment shell in future analyses of containment response.
Online monitoring of nuclear power plant (NPP) operational status is critical for ensuring safe and reliable operation. Given the complexity of NPP systems and the large number of variables to be monitored, conventional monitoring approaches which rely heavily on operators’ experience are facing significant challenges in fault detection and identification. This research focuses on developing and applying an improved reconstruction-based contribution (RBC) analysis for NPP status monitoring and fault variables identification. Although traditional PCA-based RBC methods are widely employed for identifying fault variables in multivariate transient deviations, there are still several limitations that compromise diagnostic accuracy: optimizing one statistic may lead to an increase in another; “dragging-tail effect” causing nonfault variables to be identified; and statistics converging to local optima after partial fault direction reconstruction. To address these issues, this paper proposes an improved approach through the introduction of a joint target function with the weight coefficient empirically determined to achieve optimal balance, as well as a variable contribution sorting and truncation method to enhance the accuracy and reliability of fault variables identification.
Radioactive sources have many beneficial uses across various socioeconomic sectors, but these sources pose safety and security risks. The detection and measurement of ionizing radiation constitute an important pillar not only to ensure radioprotection but above all to prevent illicit trafficking of radioactive materials. However, to guarantee the reliability of data from measuring equipment, these devices should be subjected to periodic calibration. The aim of our study is to undertake alternative calibration of radiation survey meters for calibration factor determination by means of gamma radiation emitted by the 137Cs radionuclide. To do that, calibration equipment has been designed and fully locally set up, and then calibration protocols and procedures have been established. Thirteen (13) survey meters were locally calibrated, and for evaluation of the performance of our designed calibration system, four (04) of them were sent to Ghana Secondary Standard Dosimetry Laboratory (SSDL) for calibration factor comparison purposes. For each survey meter, four (04) series of measurements have been carried out. The results showed calibration factors varying from 1.02 +/- 0.10 to 2.24 +/- 0.12. The designed calibration tool demonstrates good stability, with a coefficient of variation less than 05%. In addition, according to the statistical indicators criteria, all calibration factors below 1.50 were consistent with SSDL results, whereas the only calibration factor above this value was not. Therefore, the designed calibration tool, together with the calibration protocol and procedures, complies with ISO 4037-3:2019 requirements.
This computational study quantifies organ-specific absorbed doses for diagnostic PET with 89Zr-labeled hNd2 using ICRP adult male/female reference voxel phantoms and the MIRD/ICRP framework implemented in IDAC-Dose 2.1, cross-validated in WinAct 1.0; organ time-activity curves were parameterized for antibody-type kinetics, integrated to residence times, and converted to absorbed doses using IDAC S-values. In the female phantom, the heart received the highest absorbed dose (similar to 5.0 mGy/MBq), followed by the spleen (similar to 3.8 mGy/MBq) and lung (similar to 3.4 mGy/MBq); the kidneys and right colon were intermediate (similar to 2.1 and similar to 2.4 mGy/MBq, respectively), while the liver and stomach were similar to 1.3-1.6 mGy/MBq, and the left colon was similar to 0.5 mGy/MBq. In the male phantom, the heart (similar to 4.1 mGy/MBq) again dominated, followed by the spleen (similar to 3.2 mGy/MBq) and lung (similar to 2.6 mGy/MBq); the kidneys (similar to 1.7 mGy/MBq) and right colon (similar to 2.0 mGy/MBq) were intermediate; the liver and stomach ranged similar to 1.1-1.2 mGy/MBq, the left colon similar to 0.3 mGy/MBq, and the testes similar to 1.0 mGy/MBq. These results contextualize organ dose patterns for Zr-89-hNd2 PET and, together with the uncertainty analysis and cross-validation, support protocol optimization (e.g., activity selection and hydration strategies) that balances image quality with radiation safety.
Water scarcity poses a mounting constraint on Ghana's socioeconomic development, intensified by climate-driven hydrological variability, rapid urbanization, and industrial expansion. This study presents a Ghana-specific techno-economic and policy-integrated assessment of nuclear-powered seawater desalination, moving beyond generic applications of existing tools. A 300 MWe small modular reactor (SMR) operating in cogeneration mode with reverse osmosis (RO) and multieffect distillation (MED) is analyzed using the IAEA Desalination Economic Evaluation Program (DEEP), calibrated to Ghanaian regulatory compliance costs, local labor indices, financing structures, coastal siting constraints, and water conveyance distances. The proposed configuration delivers approximately 120,000 m3/day of potable water, sufficient to serve over 1.2 million people, while diverting about 23% of reactor thermal output without compromising electricity generation. The levelized cost of water is estimated at USD 0.84-1.05/m3, with sensitivity analysis identifying discount rate and plant availability as dominant uncertainty drivers. Environmental assessment extends beyond CO2 mitigation to include brine management, thermal plume dispersion, coastal ecosystem vulnerability, and potential radiological pathways under defense-in-depth design. Comparative analysis indicates that nuclear desalination provides superior cost stability and scalability for baseload urban water supply in Ghana when contrasted with solar-driven desalination and advanced wastewater recycling.
The nuclear power industry preferably aims to a closed nuclear fuel cycle and a transition to a two-component structure based on the joint operation of thermal and fast neutron reactors. This approach makes it possible to significantly reduce the volume of radioactive waste and increase the efficiency of using uranium fuel. Sodium fast neutron reactors currently demonstrate the highest technical readiness for commercial operation. This status has been achieved through the accumulation of operational experience from research reactors to pilot demonstration projects. Many of these reactors are currently in the process of decommissioning. Decommissioning of sodium fast neutron reactor facilities is associated with serious intellectual, material, and technological difficulties. There is still no overall experience in the world of completing the decommissioning of such facilities. Nevertheless, it is believed that the experience and competencies accumulated in these activities are key to forming a strategy for decommissioning of the future nuclear power plants with fast neutron reactors as an important final stage of the life cycle of such nuclear facilities. This is very important because decommissioning is a complex process that requires careful planning and coordination to ensure the safety of personnel, the public, and the environment in the future. In this regard, this review is devoted to an overview of the current progress in decommissioning various experimental sodium fast neutron reactor facilities. The accumulated experience, existing problems, and prospects for further work have been analyzed.
Alloy 600, a nickel-based material, is widely used in steam generator tubing and penetration nozzles of pressurized water reactors (PWRs) due to its exceptional corrosion resistance and excellent mechanical properties under high-temperature and high-pressure water conditions. However, conventional uniaxial testing, which is commonly employed to obtain plastic parameters, exhibits significant limitations in practical structural assessments of Alloy 600 due to its destructive nature. To investigate the local mechanical properties of Alloy 600 while accounting for the uncertainty in material parameters during indentation testing, this study proposes a probabilistic inverse method combining Bayesian inference with spherical indentation testing. To enhance the computational efficiency of the likelihood function in Bayesian theory, a preprocessing and postprocessing program for finite element (FE) software was developed to enable batch computations. Furthermore, an adaptive PC-Kriging surrogate model was employed to replace FE simulations, achieving efficient approximation of high computational cost models. Finally, Markov Chain Monte Carlo (MCMC) sampling was systematically applied to analyze the probabilistic distribution characteristics of Alloy 600’s plastic properties. Comparisons with traditional uniaxial tensile tests and other Bayesian methods confirm the validity and reliability of the proposed approach. The results demonstrate that this method provides probabilistic estimations of plastic properties, including probability distributions and confidence intervals, with stress-strain curves showing strong consistency with uniaxial tensile test data. The novel approach exhibits remarkable effectiveness in the probabilistic characterization of Alloy 600’s plastic properties.
The breeder blanket, as a critical component in fusion reactors, undertakes three essential functions: tritium breeding, neutron shielding, and thermal energy conversion. Serving as the primary plasma-facing component, it operates under extreme conditions—including sustained high temperatures, intense neutron irradiation fluxes, significant thermomechanical stresses induced by thermal cycling and plasma loading, and complex interactions with magnetic fields. These coupled operational challenges impose stringent reliability requirements, as blanket failure may lead to unplanned reactor shutdowns, structural integrity degradation, and heightened safety risks. This study establishes a systematic framework based on probabilistic risk assessment (PRA) to quantify breeder blanket failure risks and evaluate associated environmental radioactive contamination. The methodology comprises three key steps: (1) multidimensional failure mode analysis covering material degradation, manufacturing defects, assembly errors, environmental stressors, and accident scenarios; (2) quantitative risk annual frequency estimation via fault tree analyses to identify dominant failure pathways; and (3) proposing targeted mitigation strategies (e.g., advanced material selection, enhanced quality control, and real-time monitoring) to improve blanket reliability and reduce accident probabilities. The proposed framework provides a comprehensive approach for failure risk quantification and preventive design, offering scientific support to enhance the safety and operational sustainability of fusion reactors.
The Ghana Research Reactor-1 facility would certainly be decommissioned, either at the end of its life or by a decision to discontinue its operation. In planning for GHARR-1’s eventual decommissioning, estimating the overall decommissioning cost is a major activity that needs to be undertaken to support budgeting and decision-making for the project. To minimize the risks associated with inaccurate decommissioning cost estimates to the success of such a critical project, it is imperative to create an accurate and thorough cost estimate for GHARR-1’s decommissioning, utilizing a dependable approach and industry-standard tools like the CERREX-D2 software. This strategy is essential to guaranteeing a decommissioning process that is safe, effective, economical, and compliant with regulations and has the least negative environmental impact. In the current study, the cost for decommissioning GHARR-1 has been estimated based on the IAEA’s CERREX-D2 software. This involves identifying and analyzing decommissioning activities, evaluating workforce requirements, estimating direct and indirect costs associated with each decommissioning activity, and compiling data from similar projects for benchmarking. This approach results in a detailed data-driven cost estimate incorporating activity analysis and workforce assessments. A detailed estimate ensures proper resource allocation, safe waste disposal, and adherence to legal standards, while a well-defined budget helps allocate funds efficiently. The study covers key aspects of decommissioning activities, including dismantling, decontamination, spent fuel management, waste management, and site restoration. The scope includes data collection (reactor SSC details and industry benchmarks), cost estimation (using CERREX-D2 software to estimate inventories, unit factors, cost elements, and sensitivity analysis), and validation and verification (comparing estimates to similar projects and conducting reviews for accuracy). Despite the various benefits this project offers, it is limited to the quality of data provided, software capabilities, and the applicability of benchmark data. The findings show that the total decommissioning cost estimate for GHARR-1 is approximately $8.1 million, with management of radioactive waste and regulatory and safety compliance identified as the key cost drivers. These findings are supported by comparisons with similar projects and sensitivity analysis. This study provides insightful information about the financial planning required for the safe and effective decommissioning of GHARR-1 and may be used as a reference for the actual decommissioning cost estimates to be included in the detailed decommissioning plan for GHARR-1 in the future.
The decontamination of high-level radioactive nuclear storage tanks with complex internal devices poses critical challenges in nuclear decommissioning engineering. This study develops a computational fluid dynamics (CFD) framework integrating the volume of fluid-discrete phase model (VOF-DPM) with the Eulerian wall film model (EWFM) to systematically evaluate spray decontamination processes in a prototypical stainless steel tank. The analysis covers critical performance indicators including transient pressure distribution across serpentine pipe surfaces, spraying liquid coverage rate and the effectiveness of liquid film formation. The data offer valuable insights and references for nuclear decontamination engineering applications, which can be applied to estimate the minimum spraying cycle for the tank, provide cases for spraying point layout, and optimize process parameters.
The electric melter system is a critical component in the high-level radioactive liquid waste (HLLW) vitrification process, with a safety classification of radiochemical safety (RS) level. At an operational site, the electric melter system is designed to be powered by an emergency power supply system, and manual switching of the power supply is employed to address power failure incidents. During long-term operation, the operating company identified several critical issues associated with the manual power switching process: excessive workload for operators, low operational efficiency, and poor timeliness in power restoration. In this study, rigorous theoretical analyses were conducted to diagnose the root causes of the aforementioned problems. Subsequently, two distinct and innovative electrical technical improvement schemes were proposed, each accompanied by detailed operational procedures for implementation and a reliable execution framework to ensure on-site applicability. A comprehensive evaluation was then performed on both schemes, considering factors such as economy, reliability, and on-site implementability. The optimally selected scheme was successfully implemented at the operational site, and its performance was further verified through subsequent HLLW vitrification operation cycles. The results demonstrate that the proposed technical improvement scheme fully resolves the practical challenges and demands of the operating company, achieving the expected objectives with excellent effectiveness. Furthermore, the implementation of the proposed technical improvement scheme has mitigated the potential risks induced by operational delays or intricate procedures, enhanced the certainty of power supply restoration under accident conditions, and thus exerted a crucial role in safeguarding the stable operation of the HLLW vitrification process. Another prominent contribution of this study lies in the provision of electrical single-line diagrams derived from the proposed technical improvement schemes. These diagrams can serve as a standardized design template and provide crucial references for the power supply system design of new HLLW vitrification projects.
Innovations in nuclear technology have the potential to play a transformative role in Ghana’s pursuit of a secure, reliable and low-carbon energy future. This review examines recent global advances in nuclear reactor technologies, fuel cycle systems, waste management approaches, regulatory practices and digital applications and assesses their relevance to Ghana’s emerging nuclear power programme. Particular attention is given to advanced reactor concepts, including small modular reactors, which offer scalable and flexible options aligned with Ghana’s projected energy demand. Developments in advanced fuels, reprocessing technologies and waste minimisation strategies are discussed as pathways for improving resource efficiency and long-term environmental safety. The paper also highlights the role of digitalisation and artificial intelligence in enhancing operational safety and efficiency, alongside associated cybersecurity and workforce challenges. By situating these innovations within Ghana’s institutional and regulatory context, the paper identifies key strategic and policy considerations for the responsible deployment of nuclear energy in support of national development goals.
