Search Results (37)

In this paper, several optimized design results of the HTGR-based 10 MWth Reaktor Daya Eksperimental (RDE) (Experimental Power Reactor), so far conducted, are reviewed and compared from the neutronics, reactor types, refueling schemes, and fuel cycle points of view. The review covers the multipass and once-through-then-out (OTTO) pebble-bed cores, as well as block/prismatic type cores with several fuel shuffling options. As for the fuel cycle, uranium and thorium fuels are considered. The fuel burnup performance and power distribution are evaluated and compared among other important design parameters. Reactor physics codes, nuclear data libraries, and calculation models and procedures used for the design and analysis are reviewed, and challenges for future improvements are discussed. Full article

The High Temperature Gas-cooled Reactor (HTGR) is a Generation IV advanced nuclear reactor, which can realize inherent safety and prevent core melt. The Institute of Nuclear and New Energy Technology (INET) of Tsinghua University developed a commercial-scale 200 MWe High Temperature gas-cooled Reactor Pebble bed Module project (HTR-PM), which entered commercial operation on 6 December 2023. A passive Reactor Cavity Cooling System (RCCS) was designed for HTR-PM to export heat from the reactor cavity during normal operation and also in accident conditions, keeping the safety of the reactor pressure vessel (RPV) and reactor cavity. The RCCS of HTR-PM has been designed as three independent sets; the normal operation of two sets of RCCS can guarantee the safety of the PRV and reactor activity. The heat can be transferred from the RPV to the final heat sink atmosphere through thermal radiation and natural convection in the reactor cavity, and the natural circulation of water and air in the RCCS. The CAVCO code was developed by the INET to simulate the behavior of an RCCS. In this paper, assuming different RPV temperatures and different ambient temperatures, as well as assuming all or parts of the RCCS sets work, the performances of RCCS are studied by CAVCO to evaluate its operational reliability, so as to provide a reference for further optimization. The analysis results indicate that even under hypothetically extremely RPV temperatures, two sets of RCCS could effectively remove heat without causing water boiling or system failure. However, during the winter when ambient temperatures are low, particularly when the reactor operates at a lower RPV temperature, additional attention must be given to the operational safety of the system. It is crucial to prevent system failure caused by the freezing of circulating water and the potential cracking of water-cooling pipes due to freezing. Depending on the reactor status and ambient conditions, one or all three sets of RCCS may need to be taken offline. In addition, the maximum heat removal capacity of the RCCS with only two sets operational exceeds the design requirement of 1.2 MW. When the ambient temperature fluctuates significantly, it may be advisable to increase the number of available RCCS sets to mitigate the effect of abrupt changes in cooling water temperature on pipeline thermal stress. Full article

Nuclear power continues to be a great promise in the green revolution, as it is a cost-effective, low-emission, and safer alternative to fossil fuels that is capable of continuous operation. A preliminary design evaluation is presented for a submersible nuclear power station capable of operating under its own power during emergencies and routine maintenance. Because it is stationed at sea, it offers a resilient solution to natural disasters such as earthquakes and tsunamis, giving it the capability to disengage and sail to deeper waters in less than a half of an hour. In the present evaluation, the hull dimensions of a very large existing submarine and the turbomachinery layout of a Pebble Bed Modular Reactor cycle were used as baselines. The conceptual design of the submersible nuclear power station includes reactor and turbomachinery integration, preliminary sizing (4 pressure hull design; total length of 57.74 m), and propulsion system analysis, demonstrating the technical viability of the proposed submersible power station. Full article

Pebble Bed Reactors (PBRs) represent a new generation of nuclear reactors. However, modeling TRi-structural ISOtropic (TRISO) fuel particles employed in PBRs presents a unique challenge in comparison to most conventional reactor designs. Rapid generation of different possible fuel particle configurations for Monte-Carlo simulations provides improved insights into the effects of particle distribution irregularities on the neutron economy. Defective pebbles could cause changes in the neutron flux in a nuclear reactor due to increased or decreased moderating effects. Different configurations of particle fuel also impact isotope production within the nuclear reactor. This study simulates several TRISO configurations representing limited capabilities of randomization algorithms, manufacturing defects configurations and/or special pebble design. All predictions are compared to an equivalent homogenized model used as baseline. The results show that the TRISO configuration has a non-negligible impact on the parameters under consideration. To explain these results, the ratio of the thermal flux of each model to the thermal flux of the homogeneous model is calculated. A clear pattern is observed in the data: as irregularities in the moderator medium emerge due to the distribution of TRISO particles, the neutron spectrum softens, leading to higher values of k_∞ and better fuel utilization. This dependence of the spectrum on the TRISO configuration is used to explain the pattern observed in the depletion calculation. The results open the possibility of optimizing the TRISO configuration in manufactured pebbles for fuel utilization and safeguards. Future work should focus on full core simulations to determine the extent of these findings. Full article

The global push for clean hydrogen production has identified nuclear energy, particularly high-temperature gas-cooled reactors (HTGRs), as a promising solution due to their ability to provide high-temperature heat. This study conducted a techno-economic analysis of hydrogen production in Saudi Arabia using the pebble bed modular reactor (HTR-PM), focusing on two methods: high-temperature steam electrolysis (HTSE) and the sulfur–iodine (SI) thermochemical cycle. The Hydrogen Economic Evaluation Program (HEEP) was used to assess the economic viability of both methods, considering key production factors such as the discount rate, nuclear power plant (NPP) capital cost, and hydrogen plant efficiency. The results show that the SI cycle achieves a lower levelized cost of hydrogen (LCOH) at USD 1.22/kg H₂ compared to HTSE at USD 1.47/kg H₂, primarily due to higher thermal efficiency. Nonetheless, HTSE offers simpler system integration. Sensitivity analysis reveals that variations in the discount rate and NPP capital costs significantly impact both production methods, while hydrogen plant efficiency is crucial in determining overall economics. The findings contribute to the broader discourse on sustainable hydrogen production technologies by highlighting the potential of nuclear-driven methods to meet global decarbonization goals. The paper concludes that the HTR-PM offers a viable pathway for large-scale hydrogen production in Saudi Arabia, aligning with the Vision 2030 objectives. Full article

The High-Temperature Gas-Cooled Pebble Bed Reactor (HTG-PBR) is notable in the advanced reactor realm for its online refueling capabilities and inherent safety features. However, the multiphysics coupling nature of HTG-PBR, involving neutronic analysis, pebble flow movement, and thermo-fluid dynamics, creates significant challenges for its development, optimization, and safety analysis. This study focuses on the high-fidelity neutronic modelling and analysis of HTG-PBR with an emphasis on achieving an equilibrium state of the reactor for long-term operations. Computational approaches are developed to perform high-fidelity neutronics analysis by coupling the superior modelling capacities of the Monte Carlo Method (MCM) and Discrete Element Method (DEM). The MCM-based code OpenMC and the DEM-based code LIGGGHTS are employed to simulate the neutron transport and pebble movement phenomena in the reactor, respectively. To improve the computational efficiency to expedite the equilibrium core search process, the reactor core is discretized by grouping pebbles in axial and radial directions with the incorporation of the pebble position information from DEM simulations. The OpenMC model is modified to integrate fuel circulation and fresh fuel loading. All of these measures ultimately contribute to a successful generation of an equilibrium core for HTG-PBR. For demonstration, X-energy’s Xe-100 reactor—a 165 MW thermal power HTG-PBR—is used as the model reactor in this study. Starting with a reactor core loaded with all fresh pebbles, the equilibrium core search process indicates the continuous loading of fresh fuel is required to sustain the reactor operation after 1000 days of fuel depletion with depleted fuel circulation. Additionally, the model predicts 213 fresh pebbles are needed to add to the top layer of the reactor to ensure the k_eff does not reduce below the assumed reactivity limit of 1.01. Full article

This study numerically investigated the effects of various contact strategies on the thermal hydraulic behavior within a structured bed of 100 explicitly modeled pebbles. Four contact strategies and two thermal hydraulic conditions were considered. The strategies to avoid contact singularities include decreasing the pebble diameter, increasing the pebble diameter, bridging the pebble surfaces near the contact region, and capping the pebble surfaces near the contact region. One strategy, Strategy 3a, which involves bridging with a cylinder equal to 10% of the pebble diameter, was selected as the baseline strategy because it addressed the contact singularity while minimizing the geometric changes that affect the bed porosity. The two thermal hydraulic conditions were full-power operation (Case 1) and pressurized loss of forced cooling or PLOFC (Case 2). Simulations of the conjugate heat transfer within the structured bed were performed using the Reynolds-averaged Navier–Stokes approach with the realizable k-$ϵ$ turbulence model and two-layer all $y^{+}$ wall treatment. The thermal-fluid quantities of interest were compared between the contact strategies for each case. In Case 1, the hydraulic behavior was sensitive to the contact strategy, with large differences in the pressure drop (30%) and volume-average velocity (4%). The thermal behavior was not sensitive, with less than a 0.5% difference across the strategies. To better understand the separate effects of each heat transfer mode, Case 2 was divided into the following subcases: conduction (Case 2a); conduction/radiation (Case 2b); and conduction/radiation/convection (Case 2c). Case 2a represents an early phase of the PLOFC transient. Case 2b represents an intermediate phase of the PLOFC transient, with the pebble temperatures sufficiently high for the radiative heat transfer to be non-negligible. Case 2c represents a late phase of the PLOFC transient after the establishment of the natural circulation of the heat transfer fluid. For Case 2, large differences in the contact strategy were observed only in Case 2a with only conduction. The difference in the maximum pebble temperature was 23% in Case 2a, 2% in Case 2b, and 0.3% in Case 2c. Full article

The porous media approach is regarded as an appropriate methodology for hydraulic simulations of complex pebble beds in fusion reactors. In order to determine the parameters (permeability α and inertial loss coefficient C) of the porous media accurately, two methods are proposed: the correction method and the fitting method. In this paper, a single-channel model with sequentially packed pebbles is constructed in order to obtain the pressure drop gradient against superficial velocities. Two methods, the correction method and fitting method, are employed to determine the permeability and inertial loss coefficient, and the results are evaluated with comparisons. Based on the results, both the correction method and fitting method are deemed feasible for the parameter determinations. In consideration of the consumption of resources and time for simulation, the fitting method is recommended during the preliminary design phase to shorten the duration of design, while the correction method is suggested to obtain precise results when the design is accomplished. Both of the methods would be evaluated with the data obtained from experiments in the future. Full article

The last decade has seen the development and application of data-driven methods taking off in nuclear engineering research, aiming to improve the safety and reliability of nuclear power. This work focuses on developing a reinforcement learning-based control sequence optimization framework for advanced nuclear systems, which not only aims to enhance flexible operations, promoting the economics of advanced nuclear technology, but also prioritizing safety during normal operation. At its core, the framework allows the sequence of operational actions to be learned and optimized by an agent to facilitate smooth transitions between the modes of operations (i.e., load-following), while ensuring that all safety significant system parameters remain within their respective limits. To generate dynamic system responses, facilitate control strategy development, and demonstrate the effectiveness of the framework, a simulation environment of a pebble-bed high-temperature gas-cooled reactor was utilized. The soft actor-critic algorithm was adopted to train a reinforcement learning agent, which can generate control sequences to maneuver plant power output in the range between 100% and 50% of the nameplate power through sufficient training. It was shown in the performance validation that the agent successfully generated control actions that maintained electrical output within a tight tolerance of 0.5% from the demand while satisfying all safety constraints. During the mode transition, the agent can maintain the reactor outlet temperature within $\pm 1.5$ °C and steam pressure within 0.1 MPa of their setpoints, respectively, by dynamically adjusting control rod positions, control valve openings, and pump speeds. The results demonstrate the effectiveness of the optimization framework and the feasibility of reinforcement learning in designing control strategies for advanced reactor systems. Full article

The conversion of coal-fired power plants to nuclear power stations is a potential method for decarbonizing coal power and offers a pathway for low-carbon development in China’s power industry. This paper focuses on retrofitting China’s coastal coal-fired power stations and compares the potential nuclear reactor technologies for the retrofit: China’s mainstream pressurized water reactor and the commercially operated fourth-generation high-temperature gas-cooled reactor (HTGR). The analysis compares the degree of matching between the two technologies and coal-fired power stations in terms of unit capacity, thermal system parameters, unit speed, structural dimensions, and weight, which significantly impact the retrofit scheme. The results indicate that HTGR is more compatible with coal-fired power plants and is recommended as the type of nuclear reactor technology to be retrofitted. The study selected the 210 MWe High-Temperature Gas-Cooled Reactor Pebble-Bed Module (HTR-PM) as the reactor technology for retrofitting a typical 300 MW class subcritical coal-fired unit. Based on the concept of subcritical parameters upgrading, the potential analysis and strategy study of retrofit is carried out in terms of the turbine, the main heat exchange equipment, the main pumps, and the main thermal system pipelines in the conventional island. The results indicate that the conventional island of the HTR-PM nuclear power plant has significant potential for retrofitting, which can be a crucial research direction for nuclear retrofitting of coal-fired power plants. Full article

In a tritium-breeding blanket of a fusion reaction, helium, used as a tritium-purging gas, will purge the tritium breeder pebble beds to extract the tritium in blanket. The purge gas flow characteristics will affect the tritium extraction efficiency. The effect of the fixed wall on the pebble packing structures and purge gas flow characteristics was investigated by combining the discrete element method (DEM) and computational fluid dynamics (CFD) method. The results indicate that the fixed wall leads to a regular packing of the pebbles adjacent to the fixed wall in association with drastic fluctuations in the porosity of the pebble bed, which can affect the purge gas flow behaviors. Further analyses of helium flow behaviors show that the helium pressure in the pebble bed decreases in a linear manner along the flow direction, whereas the pressure drop gradient of helium increases gradually with an increase in the packing factor. The reduction in porosity in the pebble bed leads to a notable escalation in helium flow velocity. Concerning the direction perpendicular to the helium gas flow, the evolution of the cut-plane averaged velocity of helium is similar to that of the porosity, except in the region immediately adjacent to the wall. The pressure drop and flow characteristics obtained in this study can serve as input for the thermohydraulic analysis of the tritium blowing systems in the tritium-breeding blanket of a fusion reactor. Full article

Two-phase lithium ceramic Li₂TiO₃-Li₄SiO₄ is considered as a tritium multiplier for use in the solid blanket of fusion reactors. To date, the most accurate understanding of the processes of tritium and helium production and release occurring in the breeder blanket materials under neutron irradiation can only be obtained from experiments in fission research reactors. At that, irradiations in vacuum give the possibility to register even very fast gas release processes (bursts) from the ceramics’ voids and pores, although it reduces the thermal conductivity of the pebble bed. The purpose of this work was to simulate the heating of mono-sized pebble bed (1 mm in diameter) of two-phase lithium ceramic 25 mol%Li₂TiO₃+75 mol%Li₄SiO₄ in an ampoule device during neutron irradiation at the WWR-K research reactor under vacuum conditions, and to determine experimental parameters in order to prevent heating of the lithium ceramics up to the Li₄SiO₄-Li₂SiO₃ phase transition temperatures (>900 °C). For the first time, it was obtained that the effective thermal conductivity of a 1 mm mono-sized pebble bed of 25 mol%Li₂TiO₃+75 mol%Li₄SiO₄ significantly decreases (four times) when it is irradiated with neutrons in a vacuum (at a helium pressure of approximately 10 Pa), compared to a similar calculation at 100 kPa of helium (when the He sweep is used). It was concluded that it is difficult to evaluate the maximal temperature of the ceramics in the capsule by measuring the temperature of its outer metal wall (according to thermocouple readings) without using the results of thermophysical calculations for each type of ceramic, taking into account its quantity, specific heat release and pebble size(s). To control the temperature of the ceramics during an irradiation experiment in a vacuum, an in-capsule thermocouple should be used, placed in the center of the pebble bed. Measuring the temperature of the pebble bed based on the capsule wall temperature can lead to overheating of the ceramics and phase changes. Full article

Fourth-generation nuclear power systems are based on high-temperature gas-cooled reactors, in which the pebble fuel is the primary energy carrier. In this regard, applying protective pyrocarbon coatings on granulated fuel is an essential problem in ensuring the reliability of nuclear power plants. The article’s main idea is to research rational technological parameters of forming a pyrocarbon protective coating on the granules of a nuclear fuel model. For this purpose, granulated Al₂O₃ with the protective pyrocarbone coating was applied as a fuel model. The article’s aim is to study the effect of thermophysical parameters on applying a protective pyrocarbon coating on granulated Al₂O₃. During the experimental studies, thermal imaging of the pyrolysis process was used. The scientific novelty of the work is the equilibrium curves for the systems Al₂O₃:CH₄, Al₂O₃:CH₄:N₂, and Al₂O₃:CH₄:Ar. Their analysis allowed for evaluating rational thermochemical parameters of the pyrolysis process. As a result, rational thermophysical parameters of coating granulated Al₂O₃ with a pyrocarbon layer were evaluated, and the practical possibility of applying the pyrocarbon coating to granulated Al₂O₃ in the electrothermal fluidized bed was experimentally proven. It was shown that nitrogen does not significantly affect the target reaction product under a temperature of less than 1500 K. Also, the rational conditions for the pyrocarbon coating at a pressure of 0.1 MPa were realized at a temperature of 900–1500 K and using argon. Moreover, pyrocarbon was precipitated from hydrocarbon at 1073–1273 K. Overall, the need to add an inert gas for reducing the carbon black formation was proven to prevent a reduction of natural gas efficiency. Full article

The EU conducted the pre-conceptual design (PCD) phase of the demonstration reactor (DEMO) during 2014–2020 under the framework of the EUROfusion consortium. The current strategy of DEMO design is to bridge the breeding blanket (BB) technology gaps between ITER and a commercial fusion power plant (FPP) by playing the role of a “Component Test Facility” for the BB. Within this strategy, a so-called driver blanket, with nearly full in-vessel surface coverage, will aim at achieving high-level stakeholder requirements of tritium self-sufficiency and power extraction for net electricity production with rather conventional technology and/or operational parameters, while an advanced blanket (or several of them) will aim at demonstrating, with limited coverage, features that are deemed necessary for a commercial FPP. Currently, two driver blanket candidates are being investigated for the EU DEMO, namely the water-cooled lithium lead and the helium-cooled pebble bed breeding blanket concepts. The PCD phase has been characterized not only by the detailed design of the BB systems themselves, but also by their holistic integration in DEMO, prioritizing near-term solutions, in accordance with the idea of a driver blanket. This paper summarizes the status for both BB driver blanket candidates at the end of the PCD phase, including their corresponding tritium extraction and removal (TER) systems, underlining the main achievements and lessons learned, exposing outstanding key system design and R&D challenges and presenting identified opportunities to address those risks during the conceptual design (CD) phase that started in 2021. Full article

The HCPB TBS (Helium-Cooled Pebble Bed Test Blanket System) is one of the two European TBSs that will be installed and tested in the ITER reactor. The use of flanged connections in the Helium Coolant System and the Tritium Extraction System of the HCPB TBS would make the remote maintenance operations easier and faster. Therefore, investigating the helium leakage from flanges becomes a fundamental step toward the control of the tritium activity in the Port Cell, as the helium flow will contain a variable but not negligible amount of tritium. The first set of experiments on helium leakages from flanged connections is described in this paper. The experiments were performed in a HeFUS3 facility, an eight-shaped helium loop designed to work at HCPB-TBS-relevant conditions. The facility can provide a helium mass flow rate in the range of 0.27–1.4 kg/s and can reach a pressure as high as 80 bar and a temperature up to 530 °C. Two types of gaskets were tested in this campaign: a spiral-wound gasket and an oval ring joint. The gasket/flange assemblies are described in detail in this paper, together with the test section that hosts them and the performed commissioning tests. The tests were carried out at 500 °C and 80 bar. In these conditions, the leak rate from the flange with the oval ring joint resulted in being, on average, 1.42·10⁻⁶ mbar∙L/s, while the leak rate from the flange with the spiral-wound gasket resulted in being, on average, 3.73·10⁻³ mbar∙L/s. Full article