HTGR Research Articles

Study of the behavior of dust particles in helium turbines considering the effects of particle deposition and resuspension

The deposition of graphite dust poses significant challenges to the helium turbines in high-temperature gas-cooled reactors. In this study, FLUENT, a Computational Fluid Dynamics (CFD) program was used with a discrete-phase model and a random-walk model to calculate the trajectories of particles (assumed spherical). Considering the interactions between particles and the wall as well as the resuspension effect of the fluid, a particle-deposition model was established and coupled to the flow-field calculations of blades with film cooling using user-defined functions. The influence of different deposition models, particle diameters, and blowing ratios on deposition were investigated. The results show rebounding and resuspending particles significantly affect the particle-deposition rate and its distribution. With increasing particle diameter, the deposition rate initially increases and then decreases. The influence of blowing ratio on deposition is complex; as the blowing ratio is increased, the deposition rate of small particles increases, while that of large particles decreases.

RELAP5-3D validation studies based on the High Temperature Test facility

In the spring and summer of 2019, experiments were conducted at the High Temperature Test Facility (HTTF) that form the basis of an upcoming high-temperature gas-cooled reactor (HTGR) thermal hydraulics (T/H) benchmark. HTTF is an integral effects test facility for HTGR T/H modeling validation. This paper presents RELAP5-3D models of two of those experiments: PG-27, a pressurized conduction cooldown (PCC); and PG-29, a depressurized conduction cooldown (DCC). These models used the RELAP5-3D model of HTTF originally developed by Paul Bayless as a starting point. The sensitivity analysis and uncertainty quantification code, RAVEN was used to perform calibration studies for the steady-state portion of PG-27. We developed four PG-27 calibrations based on steady-state conditions. These calibrations all used an effective thermal conductivity equal to 36 % of the measured thermal conductivity, but they differed with respect to the frictional pressure drops and radial conduction models. These models all captured the trends in steady-state temperature distributions and transient temperature behavior well. All four calibrations show room for improvement in predicting the transient temperature rise. The smallest error in temperature rise during the transient was a 21 % underprediction, and the largest was a 48 % underprediction. The errors in transient temperature rise are largely a result of a mismatch in power density between the RELAP5-3D model and the experiment due to the location of active heater rods along the boundary between heat structures in the model. The best of these calibrations was applied to PG-29 to model the DCC. Once again, temperatures during the transient were underpredicted but trends in temperature were captured. The RELAP5-3D model captured trends in the data but could not reproduce measured temperatures exactly. This result is not attributed to deficiencies in the experimental data or to RELAP5–3D itself. Rather, this result likely arises due to the some of the assumptions and decisions made when the RELAP5-3D model was first developed, prior to the execution of HTTF experiments. An agreement in prediction of temperature trends but challenges reproducing HTTF temperatures within measurement uncertainty is consistent with previous analyses of HTTF in the literature. Future RELAP5-3D validation activities centered around HTTF may be able to provide greater insight into the code’s capabilities for HTGR modeling with a more finely nodalized model.

An inversion model for burnup history of spent fuels in pebble bed HTGR based on optimization algorithm

The measurement of transuranic isotopes in spent nuclear fuel is a crucial technique for nuclear material accounting and nuclear non-proliferation. In a pebble bed high-temperature gas-cooled reactor, numerous fuel spheres circulate through the reactor core multiple times, making it unfeasible to document the power histories of individual fuel spheres. This complexity renders it impossible to employ traditional methods, such as solving the forward burnup equations used in conventional reactors, to calculate the transuranic isotopes.In this approach, optimization algorithms were coupled with the data obtained from the BUrnup Measurement System to establish a power history inversion model, which was used to solve for the power history of fuel spheres and subsequently calculate the transuranic isotopes.The paper also compared the pros and cons of coupling the inversion model with different optimization algorithms and evaluated models using various schemes of activity vector. This process helped determine the most suitable optimization algorithm and scheme of activity vector. Sensitivity analysis revealed that the established inversion model exhibited a degree of tolerance to noise in input data and the underlying physical model.

A context-ensembled refinement network for image segmentation of coated fuel particles

The thickness of the coating layer of coated fuel particles plays a vital role in the reliable operation of a high-temperature gas-cooled reactor. Coating thickness remains extremely challenging to measure it efficiently, accurately, and robotically. The existing methods struggle with poorly refined segmentation results and easy adhesion of boundaries, complicating the accurate image segmentation of coating layers, which is fundamental to achieving the measurement goal. To overcome these issues, we designed the Context-Ensembled Refinement Network. This network refines the segmentation results to accurately extract different coating regions, leveraging its context fusion module and boundary refinement module to effectively utilize multiscale contextual information and provide strong boundary refinement capability. Experimental results demonstrate that our proposed method outperforms recent state-of-the-art methods, achieving an mIoU value of 95.8 %, and holds promising potential for automatic measurement of coating thickness.

Study on Conventional Island Retrofit Strategies for Converting Coal-Fired Power Plants to Nuclear Power Stations in China

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.

Temperature Effects of Nuclear and Electronic Stopping Power on Si and C Radiation Damage in 3C-SiC.

Silicon carbide has been considered a material for use in the construction of advanced high-temperature nuclear reactors. However, one of the most important design issues for future reactors is the development of structural defects in SiC under a strong irradiation field at high temperatures. To understand how high temperatures affect radiation damage, SiC single crystals were irradiated at room temperature and after being heated to 800 °C with carbon and silicon ions of energies ranging between 0.5 and 21 MeV. The number of displaced atoms and the disorder parameters have been estimated by using the channeling Rutherford backscattering spectrometry. The experimentally determined depth profiles of induced defects at room temperature agree very well with theoretical calculations assuming its proportionality to the electronic and nuclear-stopping power values. On the other hand, a significant reduction in the number of crystal defects was observed for irradiations performed at high temperatures or for samples annealed after irradiation. Additionally, indications of saturation of the crystal defect concentration were observed for higher fluences and the irradiation of previously defected samples.

Roadmap of Graphite Moderator and Graphite-Matrix TRISO Fuel Management Options

Most high-temperature reactors use graphite as a moderator and structural material. This includes high-temperature gas-cooled reactors with helium cooling and TRi-structural ISOtropic (TRISO) fuel particles embedded in graphite, as well as fluoride salt–cooled high-temperature reactors with clean salt coolant and TRISO fuel particles embedded in graphite and thermal spectrum molten salt reactors with a graphite moderator and fuel dissolved in the salt. The largest volume radioactive waste stream from these reactors is the irradiated graphite. We describe herein a roadmap for management of these graphite wastes that contain radioactive 14C, tritium, and other radionuclides. There may be some graphite wastes with sufficiently low radioactivity levels that can be treated as nonradioactive waste and managed like other graphite waste. Management options for the graphite include (1) direct disposal, (2) recycled back to the reactor or other nuclear applications, and (3) oxidizing the graphite with release as an effluent or underground sequestration of the carbon dioxide. Cosequestration of this carbon dioxide with carbon dioxide from industrial, biological, and cement production processes can isotopically dilute the 14C before sequestration to eliminate the possibility of exceeding individual radiation exposure limits. We also describe options for processing graphite-matrix TRISO fuel, including separating the bulk graphite to reduce the volumes of used fuel for disposal or processing to recover fissile materials. The inventories of radioactive isotopes in different carbon wastes vary by many orders of magnitude; thus, there is no single economic option for the management of all graphite waste.

Static and dynamic fracture toughness of graphite materials with varying grain sizes

Graphite materials play critical roles as moderators, reflectors, and core structural components in high-temperature gas-cooled nuclear reactors. During the reactor operation, graphite materials may experience a variety of loads, including thermal, radiation, fatigue, and dynamic loads, potentially leading to crack initiation and propagation. Therefore, it is imperative to investigate their fracture properties. Despite this, there remains a paucity of comprehensive studies on the fracture toughness of graphite materials with varying grain sizes, particularly concerning dynamic fracture toughness. This study addresses this gap by employing a digital-image-correlation-based virtual extensometer to analyze crack propagation in graphite materials of different grain sizes, enabling precise measurement of crack propagation length and fracture toughness. Findings reveal that static fracture toughness increases with larger grain sizes, while under dynamic loading, smaller grain sizes exhibit greater fracture toughness. Additionally, dynamic fracture toughness shows a near-linear increase with impact speed. Scanning electron microscopy analysis of fracture surface morphology highlights the impact of grain size and impact speed on fracture toughness. Nuclear graphite specimens with larger grain sizes have more irregular grain and pore distributions, enhancing crack deflection and propagation resistance, thereby increasing fracture toughness. The observed loading rate dependence of dynamic fracture toughness is attributed to a gradual transition from intergranular to transgranular fracture modes with increasing impact speed.

Technical Innovation in the Project of 5 MW Low Temperature Nuclear Heating Reactor and 10 MW High Temperature Gas-Cooled Experimental Reactor of Nuclear Research Institute of Tsinghua University

Technical Innovation in the Project of 5 MW Low Temperature Nuclear Heating Reactor and 10 MW High Temperature Gas-Cooled Experimental Reactor of Nuclear Research Institute of Tsinghua University

Evolution of radiation-induced damage in nuclear graphite – A comparative structural and microstructural study

Graphite, as a material for high-temperature gas-cooled reactors (HTGR), will be exposed to harsh environment. The stability of graphite structure under irradiation is of a key importance for efficiency, reliability and security of the Generation IV nuclear reactors. Three types of nuclear grade graphite were subjected to irradiation in this research – two commercially manufactured (IG-110 and NBG-17) and the laboratory's in-home material (NCBJ). The samples were exposed to 150 keV Ar+ and He+ ions bombardment at 400 °C with fluences ranging from 1E12 to 2E17 ion/cm2 in order to simulate in-reactor conditions. For analysis of the level of structure damage, type of created defects and crystallite size changes under ion irradiation ex-situ Raman spectroscopy was used. The methodology of spectra fitting was developed. Furthermore, SEM observation of irradiated materials was performed. Results showed structural degradation of materials by the means of amorphisation: slight at a low fluence level, rising rapidly at higher irradiation values. Furthermore, stronger structural disorder was found in the materials irradiated with heavier Ar+ ions than with lighter He+. Microstructural evolution of the nuclear graphites aligned with the structural deterioration in its stepwise character.

Evaluation of nanocrystalline ZrC under gamma irradiation and high pressure

Zirconium carbide (ZrC) is known for its exceptional properties, positioning it as a leading ultrahigh-temperature ceramic for diverse applications such as coating, oxygen-gettering, and inert matrix utilization in advanced high-temperature reactor fuels. Its impressive resistance to fission product corrosion, superior fission product retention capabilities, and resilience against hydrogen corrosion further highlight its significance in these fuel systems. In this work, the structural properties of zirconium carbide crystals (particle size of 20 nm) irradiated with 60Co gamma source were investigated at absorption doses from 300 kGy up to 3000 kGy using X-ray diffraction (XRD). Furthermore, the crystallinity properties of the crystals under high-pressure conditions were examined using the Small-angle X-ray Scattering (SAXS)/Wide-angle X ray scattering (WAXS) Xeuss 3.0 system. A comparative analysis of theoretical calculations and experimental results was conducted. The crystalline size experienced a decrease, while the microhardness showed variation. ZrC was identified to possess an Face Centered Cubic (FCC) – B1 structure of transition metal carbides (TMCs), with specific atomic arrangements. Under high pressure conditions, ZrC exhibits a shift to higher two-theta angles with compression of all diffraction peaks, without the formation of new peaks, up to a pressure of 9.0 GPa.

Implementation of thorium-based fuel for Indonesia Micro Reactor (IMR)

Indonesia has committed to reducing greenhouse gas emissions in the framework of net zero emissions, and nuclear energy will be a part of the energy mix. Indonesia has been developing a 10 MWt micro reactor of high temperature gas-cooled reactor (HTGR) with pebble fuel containing 17 % enriched UO2 kernel with OTTO cycle scheme. For the government’s interest to use thorium, this paper assesses the possibility of replacing uranium with thorium. The assessment is performed by calculating the effective neutron multiplication factor (keff). Neutrons captured by 232Th lead to the formation of two intermediate isotopes, namely 233Th and 233Pa, the latter being a significant neutron absorption cross-section. Neutronics analysis is performed using parameters of thorium mass fraction, core temperature, and active core height. The thorium mass fractions varies from 0 % to 90 % Th. The core temperatures are in the range of the normal operating temperature at 900 K, assumed superheat temperature at 1200 K, and shutdown temperature at 300 K. The active core height varies at 197, 180, 150, 130, and 125 cm. This paper aims to determine the keff of the IMR at maximum thorium/uranium mass fraction composition, maximum active core height, core temperature, and fuel burn-up. The MCNP code is used for analysis. The results show that keff decreases with increasing thorium mass fraction and core temperature, and decreasing active core height. Calculations with thorium mass fraction of 15 %, core active height of 197 cm in 20 steps show that the reactor may be continuously operated for more than 920 days without refueling. The analysis results show that IMR design has negative temperature reactivity, inherent safety characteristics, and the ability to use thorium-based fuels.

Thermodynamic and economic analysis of hierarchical gas–gas nuclear power plants and CHPs with high-temperature reactors and helium as the circulating medium

This paper presents the results of thermodynamic and economic analyses of innovative, gas–gas nuclear power plants and CHP plants with high-temperature gas-cooled reactors (HTGRs). In these power plants, two Joule cycles are performed in a hierarchical system. One is performed in the high-temperature range, the other in the low-temperature range. The reactor coolant is helium. It is also the circulating medium in the whole gas–gas system. The results of the thermodynamic and economic analyses presented in this paper cover a wide range of thermal capacities of HTGRs, from 100 to 3000 MW.

A tree-based automated machine learning approach of the obstructed view factor of thermal radiation in nuclear pebble beds

For the nuclear pebble bed of the high temperature gas-cooled reactor (HTGR), a tree-based automated machine learning approach is developed to discuss the complicated thermal radiation behaviors. The AutoML model for calculating the obstructed view factor between any two particles in the pebble bed includes the gradient boosting regression tree model with the fine-tuned hyperparameters and the analytical base model. Using the discrete packing of the HTR-PM nuclear reactor and the leaped Halton sequence, a large dataset of the view factor under different conditions is generated to train the AutoML model. On the same platform, numerical results show that AutoML model is approximately 5 × 105 times faster than the traditional approach. The Pareto front of the AutoML model indicates that the mean squared error decreases with the model complexity until it reaches the optimal solution. Then, the trained AutoML model is applied to the engineering pebble bed. It is shown that the average radiation exchange factor near the wall is lower than that in the bulk region. In addition, the packing height is also an important parameter for evaluating the radiative behaviors. In particular, when the height of the nuclear pebble bed is less than 20 times particle diameter, it is necessary to consider the contribution of the packing height to the radiation effective thermal conductivity.

Impact of temperature variations on BISON predictions of Ag release in AGR-1 and AGR-2 experiments

Understanding and quantifying the release of fission products like silver (Ag) from TRistructural ISOtropic (TRISO) fuel particles is important to assess the safe operation of advanced high temperature reactors. Although the silicon carbide (SiC) layer of TRISO particles is effective as the main fission product barrier, Ag can be released from intact TRISO particles. A mechanistic model for the effective Ag diffusivity, Deff, was previously developed as a function of temperature and microstructure variables informed by atomistic modeling of Ag diffusivity. In this study, we use this model to explore how experimental temperature uncertainties impact the overall predicted Ag release. This analysis indicates that calculated temperature uncertainties have a significant impact on the overall Ag release predictions. Furthermore, we show that the time average volume average (TAVA) temperature is not an appropriate proxy for the detailed temperature histories to predict fission product release due to the Arrhenius dependence of Ag diffusivity with respect to temperature. The detailed temperature histories, therefore, provide the most accurate results and are of most importance for modeling efforts. Overall, this work shows the importance of considering the experimental uncertainty of the temperature on computational predictions of fission product transport and release and the need for more accurate temperature histories from future experiments.

Modeling passive/active transition in high-temperature dry oxidation of reaction-sintered SiC for innovative fuel matrix in high-temperature gas-cooled reactors

Modeling passive/active transition in high-temperature dry oxidation of reaction-sintered SiC for innovative fuel matrix in high-temperature gas-cooled reactors

Multi‐scale study of fluidized bed‐chemical vapour deposition process in nuclear fuel coated particle fabrication for high‐temperature gas‐cooled reactor: A review

AbstractFluidized bed‐chemical vapour deposition (FB‐CVD) is a kind of key technology used widely in many application fields, such as semiconductors, nuclear energy, energy storage, and catalysts. In recent years, it has drawn much attention in the preparation of nuclear fuel coated particles (CP). It also has long played a crucial role in the preparation of high‐temperature gas‐cooled reactor (HTGR) fuel pebbles. The multi‐scale study of FB‐CVD technology has paid attention to the industrial fabrication of nuclear fuel particles at a large scale. In this paper, the recent FB‐CVD studies of different application fields are summarized first. Then, the recent works of our group in the field of FB‐CVD process in nuclear fuel particle fabrication are summarized. The FB‐CVD process in nuclear fuel particle fabrication and the multi‐scale study of the FB‐CVD process are overviewed in detail. Molecular dynamics (MD) simulation is used to study the CVD process of preparing the coating layer at the micro‐scale. Computational fluid dynamics–discrete element model (CFD‐DEM) simulation is used to study the high‐density particle fluidization, mixing particle fluidization, and particle coating process at the particle scale. Process simulation is used to study the entire FB‐CVD production line at the macro scale. Finally, the great application potential of the multi‐scale coupling study of the FB‐CVD process in the industrial fabrication of nuclear fuel particles is revealed. This paper is helpful to develop the academic research field of fluidized beds. It also has inspiration and reference significance for the expansion of other industrial applications of FB‐CVD.

Probabilistic safety and reliability studies toward licensing and deploying HTGR technology in the Polish chemical industry

The pivotal aspects of the Probabilistic Safety Assessment (PSA) and reliability studies for the foreseen HTGR-based nuclear-chemical (Nuc-Chem) facilities are discussed. The traditional PSA methodology has been extended to include the mutual dependence of nuclear and chemical systems of the hybrid Nuc-Chem installations, where HTGR serves as a heat source for various chemical processes. The Dynamic Bayesian Network (DBN) approach has been utilized for modeling time sequences of failures, including interactions between nuclear and non-nuclear systems, as well as external hazards. The Global Variance Method (GVM), based on the Sobol indices, has been proposed for uncertainty and sensitivity studies concerning numerous sources of variability in a consistent manner. The paper presents the outcomes and achievements from research oriented toward licensing and deploying High Temperature Gas-cooled Reactors (HTGR) within the Polish chemical industry. The general framework developed in this work for PSA of the hybrid Nuc-Chem facilities can be applied to support the licensing process and safe operation of the Polish chemical plants to be equipped with HTGRs.

Pebble flow in the HTR-PM reactor core by GPU-DEM simulation: Effect of friction

The high-temperature gas-cooled reactor (HTGR) with spherical fuel elements contains complex pebble flow. The flow behavior of pebbles is influenced by various factors, such as pebble density, friction coefficient, wall structure, and discharge port size. Using a GPU-DEM numerical model, the effects of the friction coefficient on the cyclic loading and unloading of pebbles in the full-scale HTR-PM are studied. Numerical simulations with up to 420,000 spherical pebbles are conducted. Four sets of friction coefficient values are determined for comparative analysis based on experimental measurements. Discharging speed, residence time, stress, porosity, and velocity distribution are quantitatively analyzed. In addition, a comparison with the CT-PFD experiment is carried out to validate the numerical model. The results show that near-wall retention phenomena are observed in the reactor core only when using large friction coefficients. However, using friction coefficient values closer to the measured experimental values, the pebble bed in HTR-PM exhibited good flow characteristics. Furthermore, the friction coefficient also influences the porosity and velocity distribution of the pebble bed, with lower friction coefficients resulting in lower overall stress in the bed. The discharge outlet's influence varies with different friction coefficient values. In summary, this study demonstrates that the value of the friction coefficient has a complex influence on the pebble flow in HTR-PM, which provides important insights for future numerical and experimental studies in this field.

Passivity-based control of fluid flow networks with capacitance

Fluid flow networks (FFNs) widely exist in industrial energy systems tightly related with the industrial production and residents living, such as district heating networks (DHN), mine-ventilation systems and gas transportation systems. The FFN operation is to drive the flowrates and pressures to their desired values, and the flowrate-pressure coordinated control of FFN is crucial in providing safe, stable and efficient operation. Although fluid capacitance has the same importance as fluid resistance and fluid inductance in giving dynamical characteristics of FFN, fluid capacitance is still not fully considered in the flowrate and pressure control design of FFN. In this paper, a nonlinear dynamic model of general FFN with fluid resistance, inductance and capacitance is first given. By using this nonlinear model, it is shown that the general FFN dynamics is strictly passive, and the corresponding generalized Hamiltonian form (GHF) is also given. Based on the passivity of FFN, a decentralized adaptive flowrate-pressure coordinated control is newly proposed, taking a simple proportional-integral (PI) form while providing globally asymptotic regulation for network flowrates and pressures. This FFN control method is applied to design the flowrate-pressure controller of an auxiliary startup circuit (ASC) equipped to a modular high temperature gas-cooled reactor (mHTGR), and simulation results in the cases of heating-up and cooling-down show the feasibility of control design.