Abstract A heat pipe module was developed for the thermal hydraulics system code AC 2 /ATHLET to facilitate the analysis of heat pipe cooled-reactor concepts. The tool enables the transient simulation of liquid metal heat pipes. For validation, an experiment with a potassium heat pipe was successfully simulated. The simulation with different angles of inclination from horizontal to vertical is possible.
Abstract The presence of radioactive contamination in concrete structures poses a silent yet serious threat to both environmental safety and public health. Such contamination, if not properly localized, can lead to prolonged radiation exposure, difficulties in decommissioning, and unnecessary disposal of large amounts of low-level radioactive waste. This study presents a Monte Carlo simulation-based method to determine the type and exact position of a gamma-emitting contaminant located on or inside a concrete wall. Two collimated detectors are positioned on opposite sides of the wall, and the depth of the source is calculated using the ratio of photopeak counts measured simultaneously in a single simulation run. A proportional relationship between the count rates eliminates the need for prior knowledge of the wall materials attenuation coefficient. The simulated setup carefully models realistic experimental conditions, including detector geometry, wall composition, and source characteristics. It is demonstrated that the photopeak count is inversely proportional to the source distance from the wall surface. The simulation results show good agreement with experimental measurements, confirming the validity and reliability of the proposed non-destructive method for localized contamination assessment.
Abstract Miniature neutron source reactors (MNSRs) rely exclusively on buoyancy-driven natural circulation for core cooling, making steady-state behavior sensitive to heat rejection, inlet thermal conditions, and localized hydraulic resistance. In compact MNSR configurations, short flow paths and dominant minor losses create strong inlet-outlet thermal coupling, allowing outlet thermal energy to influence inlet conditions before complete heat removal and thereby reduce the effective buoyancy head. This work examines inlet thermal feedback and resistance-limited natural circulation in the Nigeria Research Reactor-1 (NIRR-1) using a mechanism-oriented, FSAR-consistent GOTHIC thermal-hydraulic framework. Geometry, dominant hydraulic resistance, nodalization, and numerical conditions are held fixed while external heat-rejection pathways are varied in a controlled manner. This design isolates the influence of loop-to-pool thermal coupling on circulation response without altering the hydraulic structure or introducing nonphysical coupling to the surrounding pool. The results show that strengthening external cooling progressively suppresses inlet thermal feedback. At low cooling strength, enhanced heat rejection reduces inlet temperature, increases the effective density difference, and amplifies circulation. Beyond a defined range of cooling conditions, further increases in heat rejection produce only marginal changes in mass flow rate and core temperature rise, indicating that the loop becomes constrained primarily by localized hydraulic resistance. Analysis of the chimney-to-pool temperature difference shows that this transition is governed by reduction of recirculated outlet thermal energy available for inlet reheating. These findings provide a physically interpretable basis for representing heat-rejection boundaries in system-level analyses of compact research reactors.
Abstract Metaheuristic optimization algorithms frequently struggle to maintain an effective balance between exploration and exploitation, particularly on high-dimensional problems where premature convergence and reduced population diversity degrade performance. Opposition-based Learning (OBL) is a widely used remedy, yet its established variants, including the Dynamic Opposition-based Learning (DOBL) method, construct opposite solutions that can still trap the search in local optima. This study proposes the Chaotic Sech-Tanh Dynamic Opposition-based Learning (CHSTDOBL) strategy, which integrates two complementary mechanisms into the dynamic opposition framework: pseudo-random sequences generated by the Ikeda chaotic map, which inject aperiodic variability to resist premature convergence, and hyperbolic secant and tangent functions, which compress these sequences into a bounded range that enables controlled local refinement. The strategy was integrated into the Whale Optimization Algorithm and compared against five established OBL variants on 500-dimensional benchmark functions, the CEC 2013 test suite, and 12 constrained engineering design problems. CHSTDOBL outperformed or matched its competitors on 20 of 24 multimodal and 21 of 24 unimodal benchmark functions and was ranked first by the Friedman test. In the CEC 2013 suite, it attained the best mean and objective values on 12 and 13 functions, respectively. For constrained design problems, it achieved the best feasible objective value in eight of 12 cases, failing to retain the best-known optimum in only four, while producing consistently tighter solution distributions than seven competing optimizers. Its generality was confirmed by embedding it into eight recently proposed metaheuristic algorithms, establishing CHSTDOBL as a competitive, algorithm-agnostic enhancement.
Abstract With the global resurgence of nuclear energy, the sustainable management of the back-end fuel cycle – specifically radioactive waste (RW) management and nuclear decommissioning (D&D) – has emerged as a critical technical and financial challenge. While technical capabilities are advancing, establishing robust, long-term financing mechanisms remains a primary hurdle for newcomer nuclear nations. This review study systematically evaluates Türkiye’s current approach by comparing its legal frameworks and funding mechanisms with those of established nuclear countries, alongside guidelines from the IAEA and OECD NEA. Moving beyond generic cost evaluations, the paper integrates the latest, highly specific national developments. This includes the recent operationalization of the Special Accounts Board and the official 2025 submission of the initial decommissioning cost plan for the Akkuyu NPP. Based on these findings and Türkiye’s official target of reaching 20 GWe nuclear capacity by 2050, the study identifies existing gaps regarding spent nuclear fuel (SNF) repatriation and open/closed fuel cycle contingencies. Consequently, we provide a novel financial and technical roadmap outlining specific strategies for near-surface disposal facilities (NSDF) and deep geological repositories (DGR) projected for the 2040–2050 period. Ultimately, this study offers actionable recommendations for newcomer nations to ensure fiscal readiness, intergenerational equity, and environmental safety.
Abstract Passive safety systems are now one of the fundamentals in designing advanced nuclear power plants, for enhancing safety and reliability by utilizing natural phenomena like condensation, natural circulation, gravity, and compressed air instead of active mechanical parts. This paper provides a thorough review of passive safety features in modern reactor designs, such as the AP1000, APR+, CPR1000, and different small modular reactors (SMRs). Particularly focuses on integrating heat pipe technology as a passive solution for decay and residual heat removal and reactor core cooling. The analysis focused on the design and operational performance of these systems. This study showed that different alternatives of passive safety features with advanced heat pipe applications can improve safety and thermal management in future nuclear power plants. This work presents a comprehensive review of passive safety systems and heat pipe applications in nuclear power plants, integrating foundational research from early passive reactor concepts with recent state-of-the-art studies published from 2022 to 2025.
Abstract This article aims to outline the legal framework for the current and future use of nuclear technology in spaceflight. Whereas binding international space law is enshrined in a system of treaties, foremost the Outer Space Treaty, this does not contain any particular provisions on the use of nuclear technologies. What we may call “nuclear law in space” is based on UN General Assembly Resolution 47/68 and the COPUOS/IAEA Safety Framework of 2009. Both documents require justification for the use of nuclear technology in space. For those uses which are justified, they contain requirements for safety and radiation protection which have thus far provided a good basis for the use of nuclear technology in space. With the advent of new nuclear technologies, however, such as modern (micro)reactors, they are considered to potentially require further development. There is no fundamental obstacle to such adjustments as long as the generic intentions of these regulations – prevention of damage to humans and to the Earth’s biosphere and prevention of contamination in outer space – are respected. Indeed, it may be assumed that novel nuclear technologies may warrant such protection even more effectively than existing ones.
Abstract Indonesia requires a long-lived nuclear reactor as part of its sustainable energy strategy, and this objective can be achieved through the modified CANDLE reactor. In this study, an assembly model with reflective boundary conditions was developed using the OpenMC code. The reactor core was modeled axially into six regions, each measuring 25 cm. The (U, Pu)N fuel was applied exclusively to the first and second regions, while the remaining regions consisted of natural uranium. This analysis investigates the influence of fuel rod configurations and power variations on reactor performance, focused on a single assembly under the aforementioned boundary conditions, in which neutrons reaching the boundary are not permitted to escape the system but are instead reflected back into the assembly volume being modeled, thereby eliminating neutron leakage. The simulation configuration employing 250 active batches, 50 inactive batches, and 20,000 particles was capable of achieving a standard deviation of 0.027 %. A 169-pin configuration was employed for the power variation study. Power levels of 1,632.7 kW and 1,959.2 kW yielded burnup levels of 240 and 308 GW d/T, with average power densities of 33 and 39 W/cm 3 , respectively, and peak power factor (PPF) values approaching unity. These findings provide a preliminary characterization of neutron behavior and the neutronic feasibility of the modified CANDLE design, analyzed within an assembly under reflective boundary conditions, serving as an initial step prior to its implementation and analysis in a full reactor core.
Abstract Passive safety systems play a critical role in reactor accident mitigation by relying on inherent physical mechanisms rather than active components. Although passive systems are generally considered more reliable than conventional active safety systems, their failure mechanisms involve significant uncertainties, making accurate reliability quantification challenging for traditional probabilistic safety assessment (PSA) methods. In particular, under severe accident conditions such as large break loss-of-coolant accidents (LBLOCA), the failure probability of passive systems is extremely low, while strong phenomenological uncertainties further complicate reliable estimation. To address these challenges, this study applies an importance sampling (IS) – based method to calculate the failure probability of the Emergency Passive System (EPS) in the HPR1000 reactor. The approach leverages established IS techniques and adapts them to the EPS context, improving both computational efficiency and estimation accuracy, and enabling the quantification of extremely small failure probabilities with a limited number of simulation samples. Furthermore, three methods – direct importance sampling, importance sampling combined with artificial neural networks (IS-ANN), and subset simulation combined with neural networks (SS-ANN) – are systematically compared. The results demonstrate that the failure probabilities obtained using the proposed IS-based method are in good agreement with those derived from neural network–enhanced approaches. This confirms the effectiveness and superiority of the proposed method for evaluating the reliability of passive safety systems and provides a practical and efficient framework for reliability assessment under severe accident conditions.
The experiments of W. Bothe and P. Jensen from the year 1940 to determine of the thermal neutron absorption cross sections of graphite and carbon were recalculated with current neutron transport codes and latest nuclear data in order to find out the reasons for the negative results, which excluded the use of both graphite and pure carbon as moderators in a natural uranium reactor. One result of the recalculation is that the diffusion theory underlying the measurement of the graphite cross section was appropriate to the problem, that the boron content of the graphite investigated was at 6 appm, and that the thermal absorption cross section of 8 mb (or 7.58 mb in a re-evaluation of the experiment) as given by W. Bothe and P. Jensen was correct. On the other hand, the measured absorption cross section of the graphite impurities of 1.61 mb was too low by a factor of 2.64, resulting in a too high carbon absorption cross section of 6.4 mb. According to the current calculation, the absorption cross section of the carbon should only have been 3.38 mb. The recalculation of the ash disk experiment, on the basis of which W. Bothe and P. Jensen had determined the neutron absorption of the impurities, shows that there are four reasons for their underestimation, two of which can be quantified. The measurements of W. Bothe and P. Jensen take center stage of a more comprehensive investigation of the graphite work of the German Uranium Club (Uranverein) in the years 1939–1942. This shows, among others, that the negative result of the Bothe-Jensen measurement was not of decisive importance in the decision to reject graphite as a moderator in a nuclear reactor, but rather that it was erroneously assumed that it was practically impossible to remove the boron impurity from the graphite. The successful approach of L. Szilard and E. Fermi to solve the graphite problem during the early phase of the Manhattan Project is also escribed. Thereby, it reveals, that the measuring method attributed to W. Heisenberg, which was used in Germany by W. Bothe and P. Jensen to determine the thermal neutron absorption cross section of graphite, most probably originates from L. Szilard. Overall, it turns out that the members of the German Uranium Club, compared to the American scientists, made only minor or no efforts in order to develop procedures for the accurate characterization, the lossless ashing, and the sufficient purification of graphite, which are necessary for a successful application of this material in nuclear engineering.
The Modular Accident Analysis Program version 5 (MAAP5) is a code developed by Fauske & Associates, Inc. for simulating severe accident progression in light water reactors. Kuosheng Nuclear Power Plant (KSNPP) is currently in the early stage of decommissioning, during which the reactor pressure vessel (RPV) remains open with fuel still present in the core. This study presents the development, steady-state verification, and severe accident simulations of a MAAP5.06 model specifically configured for shutdown open-vessel conditions. The model was constructed using actual plant parameters and verified against both normal operation and shutdown steady-state conditions. Three severe accident scenarios were investigated: Station Blackout (SBO), Loss-of-Coolant Accident (LOCA), and LOCA with recovery injection (LPCI/LPCS). Results demonstrate that the upgraded MAAP5.06 model successfully reproduces thermal-hydraulic behavior, water level progression, cladding temperature evolution, and hydrogen generation. The developed model provides a technical basis for emergency planning, risk-informed decision-making, and safety evaluations during decommissioning.
High-radiation environments within fusion reactors, such as the complex internal structure of the divertor, necessitate remote maintenance with extreme precision. Even minor errors from the robotic manipulators can lead to catastrophic cost overruns and prolonged reactor down-time. While heavy duty 9DoF arm provides reach, its inherent scale introduces significant errors, necessitating a highly accurate multi-joint end effector. This paper addresses the critical challenge of error reduction in this final manipulator by proposing a novel 2P4R kinematic architecture. Through rigorous theoretical analysis and simulation, we demonstrate that the strategic placement of prismatic joints minimizes the number of active errors contributing joints during typical trajectories. A comparative study against a conventional 6R manipulator on the same path confirms that the proposed design achieves a substantially lower cumulative Cartesian error. The results conclusively show that this kinematic configuration provides an inherent advantage in achieving the sub-millimeter accuracy required for fusion reactor maintenance.
ROBBE is the first autonomous robotic system to combine real-time 3D perception, digital-twin trajectory planning, and ultra-high-pressure (UHD) waterjet technology for coating removal and decontamination of variant-rich steel components in nuclear dismantling. Conceived to replace ergonomically demanding and radiologically relevant manual work, the system scans previously unknown geometries, computes collision-free tool paths that honor process physics and robot kinematics, and executes coating removal and decontamination in a closed cabin without personnel exposure. The end-effector integrates a sealed laser-triangulation scanner behind a protective shutter with a high-performance rotating UHD nozzle; all media and signals are routed through a humidity-resistant 7-axis industrial robot. The planning stack fuses scan data, fixture separation, and a digital twin of robot, table, part, and cell environment to optimize nozzle distance, angle, and feed speed within experimentally derived efficacy windows. An IMU-based module is available for robust pose tracking during manual re-clamping, though most parts are efficiently processed using a "clean upper half + flip" strategy with selective re-cleaning. Development progressed from a full-scale prototype outside the controlled area (RWE01) to an integrated operational cell in the Biblis dismantling plant (RWE02). The UHD process runs at 2,200-2,400 bar (rated 2,500 bar) and 24l/min; removal performance peaks at approximately 50-60 mm stand-off, and feed is adapted online to reachability constraints. 12 months of operational testing on representative part classes (pipe sections and shells, vessel plates, profiles, machine and device components with free-form surfaces, undercuts, and bores) verified autonomy, repeatability, and robustness. ROBBE achieved >= 90 % decontamination on > 90 % of parts, with net Scan + UHD times < 12 min per part and total part times typically < 30 min including manual handling. No collisions occurred;arm, hoses, and rotary feedthroughs operated without failure over 390 h of arm-power-on. For aged, multi-layer coating systems with pronounced thermal hardening, decontamination and full coating removal were reliably achieved by lowering feed rates, applying multiple passes, and, where appropriate, using chemical pre-treatment. The integrated cabin enables continuous UHD operation without operator presence, improving safety (ALARA) and process consistency compared with manual methods that are limited by noise exposure and operator fatigue. Lessons learned emphasize the benefit of a dedicated robotic cell (optimized robot-to-table layout, servo-driven round table, automatic part feed), consolidated control and HMI (single GUI for all parameters), and targeted use of the IMU for complex re-clamping only. With a demonstrated technology readiness of TRL 7-8, ROBBE is ready for broader deployment across nuclear dismantling sites and is being extended in the ARRIVE project to laser ablation and integrated radiological pre-/free measurement for an end-to-end autonomous decontamination and documentation chain.
Nuclear energy is increasingly being considered as a means to diversify energy sources in the Gulf region. The United Arab Emirates already operates nuclear power plants, while Saudi Arabia is pursuing its own nuclear program to reduce dependence on fossil fuels. The introduction of nuclear energy requires strong safety measures and reliable methods to evaluate the consequences of potential accidents to ensure preparedness at both national and regional levels. This study examines two commonly used tools for nuclear accident assessment: Geant4, a particle transport simulation framework based on the Monte Carlo method, and MACCS, which evaluates large-scale consequences for regulatory purposes. The Barakah Nuclear Power Plant in the UAE, near the Saudi border, is used as a case study to explore potential cross-border effects. Four key radionuclides - Iodine-131, Cesium-137, Strontium-90, and Xenon-133 - were analyzed. Using NOAA HYSPLIT with GDAS meteorological data, a hypothetical 10 % reactor core release was simulated. Results indicate that 131I and 137Cs could exceed safety limits, with the plume potentially reaching Abu Dhabi in 9-15 h and Riyadh in 42-54 h. These findings highlight the need to integrate detailed simulations with policy-focused modeling to guide emergency planning.
Passive Containment Cooling System (PCCS) is proposed in advanced nuclear reactor designs for removing heat from the containment in case of an accident scenario. One of the PCCS designs features a system with condensers immersed in the water pool, serving as an ultimate heat sink. Steam mixes with air, enters vertical-tube condensers, and condenses via annular film forced flow of steam hindered by noncondensable gases. In the first part of the article, experimental studies have been covered. The experimental studies performed by researchers cover the range of inlet air mass fraction from 0 to 50 %, steam flow rate from 0.0016 to 0.0183 kg/s and total pressure from 0.03 to 7.5 MPa. In the second part, various correlations developed are discussed, which account for thermal-hydraulic parameters and nondimensional numbers. An assessment of these correlations is performed. In the third part, parametric effects and results obtained from the theoretical model, along with a comparison with experimental data and correlations, are discussed. The analysis performed covers the range of inlet air mass fraction from 11 to 52 %, steam flow rate from 0.004 to 0.03 kg/s and total pressure from 0.21 to 0.49 MPa. It has been found that heat transfer coefficient depends strongly on the mixture Reynolds numbers. It decreases sharply in the initial length of the tube and then slowly as the mass fraction of noncondensable gas increases along the length. The effect of film roughness and developing length is found to be negligible on the heat transfer coefficient.
Hellma Materials GmbH has built a Single Plane Compton Camera and developed a data analysis method suitable to locate radiation hot spots in nuclear facilities. The system consists of a small group of scintillation detectors and provides directional and spectrometric information on gamma-ray emission sources. A measurement campaign conducted at the former Rheinsberg NPP demonstrated the applicability of the technique for nuclear facilities.
The ongoing decommissioning of nuclear facilities in Germany requires a highly qualified workforce, yet the sector faces a growing shortage of trained personnel due to retirements, restricted facility access, and declining educational capacity. Immersive virtual reality (VR) provides an effective means of supporting competence development where safety or operational constraints limit real-world training. However, existing VR-based training applications for the nuclear power sector remain limited in scope and often lack participatory development involving domain experts. The present study reports on the participatory, design-based development of an immersive VR training system and presents findings from an evaluation with experienced nuclear professionals. Using a convergent mixed-methods design, quantitative data from standardized questionnaires were combined with qualitative insights from semi-structured interviews. Thirteen radiation-protection workers and project engineers tested two radiation-protection and two safeguards scenarios. Participants reported consistently high enjoyment, clear task structures, and substantial perceived didactic value. Spatial presence was rated at a moderate level. The VR training was perceived as particularly beneficial for novices, supporting conceptual understanding, procedural awareness, and safety culture. Qualitative findings also identified several areas for refinement, including enhanced preparatory guidance, improved interaction fidelity, richer contextual information, and the potential introduction of multi-user modes. Overall, the results indicate that immersive VR can serve as an accessible, engaging, and instructionally meaningful complement to existing training practices in the nuclear sector. The study provides empirically grounded design exemplars for site-independent VR training and highlights future directions for expanding scenario fidelity, validating learning outcomes, and embedding VR within structured training procedures.
The CANDLE reactor, as a kind of travelling wave reactor, in which the shapes of neutron flux and power distribution remain constant at equilibrium state, which moves along the axial direction, has innovative burnup characteristics. In this paper, a CANDLE reactor with UMo alloy as fuel and lead as coolant is proposed. Neutronics calculations and performance analysis of the reactor are carried out using RMC with ENDF/B-VII database, including Mo content in the fuel, 208Pb content in the lead coolant, ignition zone length and ignition zone fuel enrichment. The results show that the CANDLE reactor with 10.5 % enrichment U8Mo alloy as the fuel, 100 % 208Pb as the coolant and an ignition zone length of 75 cm exhibits optimal neutronics properties. After about 20 years, the power distribution tends to be stable and moves along the axis in a fixed distribution with a rate of about 3.0 cm/year, with a sufficient critical safety margin and stable breeding performance. This provides a feasible technical path for efficient utilization of uranium resources.
Against the backdrop of extended storage times in German interim storage facilities and the increase in low- and medium-level radioactive waste, monitoring container integrity is of critical importance. The ZIKA (Automated non-destructive internal corrosion detection on radioactive drums/15S9446A) research project, carried out as part of the FORKA funding initiative, presents an automated system for non-destructive testing (NDT) of radioactive waste drums. The primary goal is the reliable early detection of internal corrosion in order to identify safety risks before the integrity of the containers is compromised by externally visible degradation. As a further development of the predecessor project EMOS, the system architecture has been optimized for mobile use in a compact 10-foot container. The novel design integrates complex lifting mechanisms and robotics to ensure complete inspection of the entire drum surface, including the bottom. The multi-sensory approach combines laser scanners for topographic mapping, smart cameras for detecting external defects, and active thermography for identifying internal corrosion. Experimental validations by the Federal Institute for Materials Research and Testing (BAM) showed that laser thermography is more robust and reliable in defect detection than flash thermography. Machine learning algorithms were implemented for data analysis, with random forest models achieving the highest accuracy in defect classification and artifact suppression. The ZIKA system thus represents a significant contribution to increasing long-term safety standards in nuclear interim storage facilities.
