Catalogue Search | MBRL
Are you sure you want to remove the book from the shelf?
{{itemTitle}}
510
result(s) for
"Gas cooled reactors."
Sort by:
Advanced Structural Materials for Gas-Cooled Fast Reactors—A Review
This review summarizes the development of the Gas-Cooled Fast Reactor (GFR) concept from the early 1970s until now, focusing specifically on structural materials and advanced fuel cladding materials. Materials for future nuclear energy systems must operate under more extreme conditions than those in the current Gen II or Gen III systems. These conditions include higher temperatures, a higher displacement per atom, and more corrosive environments. This paper reviews previous GFR concepts in light of several promising candidate materials for the GFR system. It also reviews the recent development of nuclear power and its use in the peaceful exploration of space. The final section focuses on the development and testing of new advanced materials such as SiCf/SiC composites and high entropy alloys (HEA) for the construction and development of GFRs.
Journal Article
Aerodynamic model based on similarity numbers and computational fluid dynamics for coolant mixing in HTGR bottom plenum
This work is devoted to the analysis of an experimental approach to modeling the processes of non-isothermal helium jet mixing at the outlet of the prismatic fuel assemblies and entering the bottom plenum of the high temperature gas cooled reactor core. A review of relevant literature and an analysis of similarity numbers characterizing the investigated process were performed. Based on an integrated approach, the characteristics of the experimental model and the operating parameters of the aerodynamic test facility assembled at the NSTU were selected. Computational studies of the three-dimensional coolant flow in a full-scale reactor fragment under normal operating conditions and in an aerodynamic model have validated the design of the experimental model and confirmed the operating parameters of the test facility.
Journal Article
Finite Element Simulation on Irradiation Effect of Nuclear Graphite with Real Three-Dimensional Pore Structure
The structural integrity of nuclear graphite is paramount for the lifespan of High-Temperature Gas-Cooled Reactors. The nuclear graphite components operate under extreme conditions involving high temperature, pressure, and intense neutron irradiation, leading to complex service behavior that is difficult to characterize only by experimental methods. This study employs the finite element method (FEM) to assess component stress and failure risk. The ManUMAT simulation method was first validated against irradiation data for Gilsocarbon graphite from an Advanced Gas-Cooled Reactor and was subsequently applied to stress–strain analysis of the nuclear graphite bricks in the HTR-PM side reflector layer. The 3D micropore structure of nuclear graphite was obtained via X-μCT and reconstructed in Avizo to establish an FEM model based on the actual pore geometry. Simulations of nuclear graphite over a 30 full-power-year service period predicted a significant contraction on the core-side and minimal thermal expansion on the out-side driven by the neutron doses. This research establishes a finite element framework that extends the ManUMAT approach by integrating a realistic pore structure model, thereby providing a foundation for quantifying the microstructural effects on macroscopic performance.
Journal Article
Selection of Planning Options of Electricity and Freshwater Cogeneration Method Based on High-Temperature Gas-Cooled Reactor
The lack of fresh water in the world has become a growing concern. As an open-source incremental technology for water resources, desalination has become an important method to solve the global water crisis. Based on the inherent safety, versatility, modularity, and advantages of high-temperature gas-cooled reactors, the Saudi Arabia desalination project is the relying background. This paper proposes a complete solution for the high-temperature gas-cooled reactor power and water coproduction project by selecting a combination of process-proven multi-effect distillation (MED) and reverse osmosis (RO). In the scheme, a tertiary circuit is designed for the isolation of radioactive entities. An innovative comparative analysis of the engineering investment and production costs of different desalination technologies, such as MED and RO, and a comparison of the investment estimates of the “thermal” and “membrane” methods for the production of 10,000 tonnes of fresh water per day are performed. The feasibility and energy efficiency of the multi-effect distillation–reverse osmosis (MED-RO) scheme are presented, demonstrating the feasibility and practicality of the above approach.
Journal Article
Computational Fluid Dynamics Modeling of Single Isothermal and Non-Isothermal Impinging Jets in a Scaled-Down High-Temperature Gas-Cooled Reactor Facility
In the current work, the flow characteristics of single isothermal and non-isothermal jets discharging into the upper plenum of a 1/16th scaled-down high-temperature gas-cooled reactor (HTGR) facility were studied. ANSYS Fluent simulations were carried out in the central plane of the jet water flow and the upper plenum for different Reynolds numbers (Re) ranging from 3413 to 12,819. Then, the statistical jet water flow characteristics, such as the mean velocity, root-mean-square fluctuating velocity, Reynolds stress, and the mean temperature in the upper plenum, were computed and presented. The current study’s results showed that the flow maximum velocity occurred far from the jet inlet. Finally, the temperature profiles were plotted, and it was found that the maximum temperature of the flow occurred close to the plume inlet and after that decreased downstream.
Journal Article
Assessment of design approximation impact on neutronic characteristics of a high-temperature gas-cooled reactor fuel assembly
A comparative study has been conducted to find out how design approximations and simulation methods of a prismatic-type fuel block of a high-temperature gas-cooled reactor (HTGR) may affect the calculation accuracy of neutronic characteristics of fuel assemblies. To study the impact, a detailed three-dimensional computational model of a typical fuel block including fuel compacts, burnable absorber compacts, and coolant passages was developed. Changes in neutronic characteristics in the process of fuel assembly irradiation were calculated. The burnup was analyzed based on the SCALE 6.2.4 software package using a calculation module implementing the Monte Carlo method with a multigroup library of cross-sections on the basis of ENDF/B-VII.1 files of assessed nuclear data and the ORIGEN burnup analysis module included in this package. Different ways of modeling fuel compacts and burnable absorber compacts have been considered: using a built-in tool (DOUBLEHET cell type), by specifying fuel particles in the graphite matrix, and their combination. The calculations were made using the 252-group library of constants except for the option in which fuel compacts and burnable absorber compacts were simulated explicitly by particles in the graphite matrix. In the latter case, a library with a pointwise (CE) representation of cross-sections was used. A series of calculations were also made to assess the way computational statistic parameters affect the results. The results confirm correct operation of the SCALE complex built-in tool, i.e. cells with the DOUBLEHET-type double heterogeneity, and its prospective use to calculate neutronic characteristics of HTGR fuel. The calculations have also shown that it is acceptable to model burnable absorber compacts both by setting a DOUBLEHET-type cell and explicitly by particles in the graphite matrix. In general, the calculation results for these options agree quite well, within 1-2%, with the direct calculation using the pointwise library of cross-sections. Based on the computational statistic parameters, it may be recommended to set at least 200,000 histories and the number of particles in a generation or the number of generations should be at least 250.
Journal Article
Fatigue Life Analysis of Key Equipment in High Temperature Reactor Considering Nuclear Irradiation
The key equipment of high-temperature gas-cooled reactors is always exposed to nuclear irradiation, while the influence of nuclear irradiation on equipment fatigue life is one of the targets to be studied. Based on the expansion mechanism of fatigue cracks of fracture mechanics, combined with actual material parameters, the critical parts of the equipment, the position and time of occurrence of the working tolerance limit are pre-critically analyzed, and the fatigue life model under the influence of nuclear irradiation are corrected to calculate the influence of nuclear radiation on fatigue life which has certain guiding significance for the estimation of fatigue life of key equipment in high temperature gas cooled reactor.
Journal Article
Development of Two Novel Processes for Hydrogenation of CO2 to Methanol over Cu/ZnO/Al2O3 Catalyst to Improve the Performance of Conventional Dual Type Methanol Synthesis Reactor
Conventional methanol synthesis process (CR configuration) consists of water-cooled and gas-cooled reactors in which methanol and water are condensed inside the gas-cooled reactor which deactivates the catalyst. In this study, two novel configurations (AW and ACW configurations) are represented to address this problem in which the gas-cooled reactor is replaced with adiabatic reactor. Moreover, a condenser is applied between adiabatic and water-cooled reactors in ACW configuration. Results show that temperature increases somewhat along the adiabatic reactor that prevents gas condensate formation. Besides, the adiabatic reactor maximum temperature is less than that of first reactor in CR configuration which prevents copper based catalyst thermal sintering. Moreover, a high cross section-to-length ratio of the adiabatic reactor leads to negligible pressure drop along the reactor and improvement in CO2 conversion to methanol that has positive environmental effects. Also, water mole fraction decreases along the reactors of AW and ACW configurations to prevent the deactivation of catalyst active sites. Eventually, methanol production rates by AW and ACW configurations are improved around 25.5% and 43.1% in comparison with CR configuration. So, novel AW and ACW configurations provide many benefits including improvement in catalyst activity and durability, CO2 conversion, and the methanol production rate.
Journal Article