Volume 90, Issue 3

Following the German government’s decision to phase out the use of nuclear energy after the reactor accident in Fukushima in March 2011, 27 prototype and power reactors are currently undergoing a dismantling process in Germany. No nuclear power plants have been in operation in Germany since April 2023 (BMUV – Bundesministerium für Umwelt 2023). The dismantling of pipelines, which are difficult to access from the outside due to limited space or because they are installed in concrete structures, is associated with many challenges and additional work for the personnel on site. Among other things, there is no universal pipe cutting system that can be used to separate pipes made of different materials, diameters and wall thicknesses from the inside of the pipe. Additional functions such as remote operation or the ability to decontaminate the machines, are of great importance for use in a nuclear environment. Most of the devices available on the market focus on pipe decontamination and are limited to certain diameters and materials. In addition, the developments are based on conventional applications and must be further adapted to the particular conditions in a nuclear facility (Bachmann 2024). Based on this, the “RoTre” project and the resulting dissertation examined the internal pipe cutting process with disk cutters, circular saw blades and cut-off wheels. The investigations presented are part of the research project “Development of a novel universally internal pipe separator for the dismantling of (contaminated) pipelines (RoTre)”, funded by the Federal Ministry of Education and Research (BMBF).

The regulator is one of the most important equipment in the uranium enrichment centrifugal cascade, of which the main function is to automatically control the pressure and flowrate within the certain range. The regulator must operate under a low-pressure vacuum condition. Compression and expansion as well as the transonic and supersonic flow of gaseous Uranium hexafluoride (UF 6 ) in the low-pressure vacuum are drastic and complicated, which makes it difficult and significant to analyze the flow field in the regulator. The computational fluid dynamics (CFD) technique has been introduced in the analysis of complex flow in recent years. The flow field characteristics in the regulator of six opening ratio under high and low inlet pressure conditions were studied based on the flow field numerical simulation. The transonic and supersonic flow field characteristics as well as the flow separation (boundary layer separation) were studied utilizing standard k-ε turbulence model. The results show that the velocity and Mach number oscillate when the opening ratio is Greater than or equal to 40 % for high inlet pressure condition, which indicates the opening ratio range of stable flow. The flow separation near the boundary layer emerges downstream of the nozzle at 20 % opening ratio for both high and low inlet pressure conditions. The regulator has piecewise linear characteristic. The research can provide guidance for design and application of the regulator in the uranium enrichment centrifugal cascade.

Nuclear waste management presents significant challenges due to its long-term environmental impact, high radioactivity, and extreme toxicity. To address these concerns, the development of innovative materials can offer promising solutions for improving environmental sustainability, safety, and durability in waste treatment and immobilization. This paper explores advanced materials – such as glass-ceramics, advanced ceramics, geopolymers, metal-organic frameworks (MOFs), and nanomaterials – that exhibit exceptional chemical stability, durability, and resistance to radiation-induced degradation. These materials have the potential to efficiently encapsulate high-level radioactive waste while overcoming the limitations of traditional methods like vitrification and cementation, which, though effective, struggle with long-term stability and containment capacity. The review highlights the superior thermal and chemical resilience, increased waste-loading capacities, and enhanced resistance to radiation damage provided by these cutting-edge materials. Furthermore, it examines the advantages, challenges, and future potential of these novel materials, emphasizing their role in enhancing the safety, efficiency, and sustainability of nuclear waste management.

Breed-and-burn reactors are a type of fast reactor that can directly burn depleted uranium or nature uranium with once through fuel cycle. The breed-and-burn reactors are composed of ignition zones and breeding zones. One of the significant challenges in practice is that high peak neutron flux levels lead to high radiation damage for solid fuels and cladding, and also lead to high peak power density. In this study, double solitary waves reactor (DSWR), likely breed-and-burn reactors but features double separated ignition zones, were proposed to reduce peak neutron flux and peak power density by half in order to provide the feasible of practical applications, and still could not reduce total output power. The breed-and-burn reactors with double separated ignition zones would generate double solitary waves propagating in nonlinear system. The double solitary waves were observed when DSWR was operating normally, and a nonlinear superposition effect was discovered. The DSWR was a fast spectrum reactor cooled by liquid mental Pb–Bi. The wave amplitude or neutron flux, the evolution of nuclides and the nonlinear superposition effect were analyzed to evaluated the feasibility of DSWR. Finally, the optimized scheme was adopted then discharged fuel burnup decreased from 380 GWd/T to ∼ 200 GWd/T in breeding zone to yield to engineering technology.

Driven by the demand for compact, portable, clean, and environmentally friendly energy sources, heat pipe-cooled microreactors have gained increasing popularity and research interest. Over the past five decades, several design alternatives have been proposed. One of the most popular designs is the monolith stainless steel core structure. Others have sought to improve safety and efficiency by introducing liquid metal cores (Na/Pb) with promising superior safety and high energy output advantages. In this work, we advance the development of heat pipe-cooled reactors through a comparative case study of 300 kW monolith stainless steel (solid core) and a lead liquid bath (liquid core). A detailed comparative analysis was conducted to evaluate their neutronic and thermal performance. The Neutronic analysis was conducted using the SuperMC-CAS Monte Carlo code, while ANSYS (CFD) was used for thermal analysis. The results confirmed the feasibility of the stainless steel design and revealed the inherent advantages of the lead liquid core. Both cores exhibit sufficient excess reactivity to sustain reactor operation for 10 years without refueling. The lead liquid core enhances inherent safety, exhibits high negative reactivity feedback, and ensures a more uniform power distribution. Thermal analysis reveals that the lead liquid core maintains lower peak temperatures and gradients during normal operation and heat pipe failure, minimizing thermal stress and enhancing safety margins. This study affirmed the robustness of a proposed lead liquid core as a viable alternative to conventional designs providing enhanced safety and performance for heat pipe microreactors.

The economic performance of a nuclear reactor is fundamentally linked to optimizing fuel efficiency, extending burn-up, and ensuring safe operational parameters. In this regard, this research explores the influence of lattice computational models on macroscopic cross-section generation for VVER-1000 Fuel Assemblies (FAs) containing Burnable Absorber Rods (BARs), focusing on their impact on neutronic parameters throughout the first operational cycle. The study addresses the spectral hardening effect of BARs, which causes distinct neutronic behaviors in fuel rods based on their proximity to the absorber. Three FA simulation arrangements are compared to assess the neutronic consequences of these proximity effects. Results highlight the increased plutonium isotope production and reduced uranium-235 consumption in fuel rods adjacent to BARs. A comparative study of key core parameters during the cycle, such as boron concentration and power peaks, is presented to demonstrate the effects of BAR modeling. The WIMD-D4 transport lattice code, with an updated library by the IAEA, and PARCS, which leverages advanced nodal kernel and CITATION finite difference codes for core calculation, are utilized in this research.

Data prediction and classification are the cornerstones of modern data analysis, especially when it comes to enhancing operational safety in critical systems. The SIMCA software offers robust, scalable tools for these tasks, leveraging the partial least squares (PLS) method for both classification and regression. This study applied regression and classification analyses to a thermal-hydraulic dataset generated using the COBRA-IV code, focusing on the calculation of the Departure from Nucleate Boiling Ratio (DNBR). The dataset was categorized into two subsets: a steady-state random dataset and a transient dataset, incorporating scenarios such as Rod Cluster Control Assembly (RCCA) operations and Loss of Flow Rate Accidents (LOFA). The steady-state dataset was developed through the variation of five key features – power, flow rate, enthalpy rise peaking factor (F ΔH ), temperature, and pressure – and computing the corresponding DNBR for each case. Hyper parameter optimization was conducted to identify the optimal number of PLS components, ensuring comprehensive information retention across datasets. For each dataset, the PLS regression metrics, including variable importance in projection (VIP), R 2 X, R 2 Y, Q 2 Y, and PLS coefficients, were analyzed to evaluate the predictive accuracy and interpretability of the model. The results demonstrated a strong similarity with COBRA-IV outputs and confirmed the findings of prior work which employed artificial neural networks (ANNs). These outcomes highlight the reliability of PLS as a predictive tool and its potential for advancing the analysis of thermal-hydraulic datasets.

It is proposed to simulate the injection process of glassy nuclear waste in ANSYS Workbench simulation platform combined with life and death element method. After verifying the feasibility of the simulation method in COMSOL Multiphysics software, the influence and law of different cooling environments on the thermal shock behavior of the container are discussed, and the thermal-structural characteristics of the container during the injection process of high temperature glassy nuclear waste are understood. The results show that the temperature change trend of the model obtained by the heat transfer coefficient function simulation method and the simulated external flow method is consistent. In order to simplify the calculation, the heat transfer coefficient method is used to simulate the injection process of the container. The thermal calculation results show that the temperature on the side wall of the container is not uniform during the injection process, and the material above the liquid surface will produce large stress due to mutual restraint, and the container will undergo plastic deformation when heated. Compared with the results of injection in the natural environment, the external wall of the container is cooled more obviously after forced convection cooling, which effectively alleviates the stress concentration in the upper head and reduces the deformation of the upper head. The water cooling method makes the upper head almost no downward concave deformation, but the stress concentration at the lower head position is more obvious.

In response to the heightened global focus on hydrogen management following the Fukushima nuclear disaster, the Maanshan Nuclear Power Plant (NPP) in Taiwan has initiated safety enhancements by installing Passive Autocatalytic Recombiners (PARs) to reduce the risk of hydrogen explosions during severe accidents. This study utilizes the MAAP5 software to refine the containment model of Maanshan NPP, introducing updates in compartment divisions, flow paths, and heat sinks to simulate gas behavior, specifically hydrogen and carbon monoxide distribution, under accident conditions. A small break loss-of-coolant accident (LOCA) scenario was modeled to evaluate the containment’s performance before and after these updates, as well as after PAR installation. The results demonstrate that the updated model improves gas transfer and diffusion simulations, while the PARs effectively reduce hydrogen and carbon monoxide concentrations below the flammability limit (<10 %), lowering the risk of detonation. Additionally, the Adiabatic Isochoric Complete Combustion (AICC) pressure remains within safety standards, which is below 8.48 bar. These findings confirm that the containment model enhancements and PAR installations significantly improve nuclear safety at Maanshan NPP.

Blankets are an important component of fusion reactors and the key object of thermal design and safety analysis. The helium-cooled solid breeder (HCSB) blanket is one of the three candidate tritium breeding blanket concepts for the Chinese Fusion Engineering Test Reactor (CFETR). Fusion reactors’ thermal design and safety analysis are mostly based on fission reactor procedures worldwide. In this paper, based on the system safety analysis module cosSyst in the COSINE code package of China’s independently developed system safety analysis code for pressurized water reactors, functional development was carried out for the helium-cooled blanket of fusion reactor, and helium physical property envelope and corresponding heat transfer model were added. The modified code was used to analyze three typical accidents with the blanket and the combined helium cooling system (HCS): In-Vessel LOCA, In-Box LOCA, and Ex-Vessel LOCA. The research findings indicate that under the three accident scenarios, the first wall and the cooling loop system maintained their structural integrity, with no melting or exceeding pressure limits.