Gas-cooled Reactor Research Articles

Feasibility study of a prestressed-concrete containment vessel for a novel GFR nuclear reactor

The latest modern nuclear energy systems currently researched and developed are the Generation IV (Gen IV) nuclear reactors, and one of the more promising Gen IV designs is the Gas-cooled fast reactor (GFR). Among the number of challenging problems of the design of GFR, the development of passive safety systems ensuring passive decay heat removal (DHR) during loss-of-coolant-accidents (LOCA) is of the highest importance. The currently proposed DHR strategy relies on intentionally increasing pressure inside a gas-tight guard vessel containing the reactor to approximately 4 bar. This, however, presents a significant engineering challenge as the vessel must be capable of withstanding high pressure of up to 10 bar, due to pressure spikes and safety coefficients, while also withstanding increased temperatures during the LOCA. As no GFR guard vessel has ever been designed, this paper presents a feasibility study establishing whether such vessel can be designed. The feasibility study is aimed at the assessment of a preconceptual design of a prestressed concrete vessel previously created by the authors. In the study, stresses in the structure induced by normal-operation loads and LOCA loads are calculated for three types of extreme constraint conditions covering all possible real constraints of the structure. The calculations are based on heat transfer equations, Euler–Bernoulli theory, and Lamé equation implemented in a in-house Python code. In the results, stresses in the vessel are presented and compared to design criteria. The main conclusion of the paper is that the vessel can be designed; however, certain adverse phenomena which complicate the design process, such as non-linear creep of concrete and occurrence of tensile stress up to 1MPa, will have to be addressed and considered during future detailed design. Following on from this affirmative conclusion, the authors will create the detailed design using advanced three-dimensional modelling in their future work.

Thermal-hydraulic analysis of He–Xe gas mixture in 2×2 rod bundle wrapped with helical wires

Gas-cooled space reactor, which adopts He–Xe gas mixture as working fluid, is a better choice for megawatt power generation. In this paper, thermal-hydraulic characteristics of He–Xe gas mixture in 2 × 2 rod bundle wrapped with helical wires is numerically investigated. The velocity, pressure and temperature distribution of the coolant are obtained and analyzed. The results show that the existence of helical wires forms the vortexes and changes the velocity and temperature distribution. Hot spots are found at the contact corners between helical wires and fuel rods. The highest temperature of the hot spots reach 1600K, while the mainstream temperature is less than 400K. The helical wire structure increases the friction pressure drop by 20%–50%. The effect extent varies with the pitch and the number of helical wires. The helical wire structure leads to the reduction of Nusselt number. Comparing thermal-hydraulic performance ratios (THPR) of different structures, the THPR values are all less than 1. It means that gas-cooled space reactor adopting helical wires could not strengthen the core heat removal performance. This work provides the thermal-hydraulic design basis for He–Xe gas cooled space nuclear reactor.

Effect of impurity ratios on the high-temperature corrosion of Inconel 617 and Incoloy 800H in impure helium

Helium is utilized to be the coolant of the high-temperature gas-cooled reactor (HTGR), which contains trace impurities in the primary circuit. The impurities may cause various corrosion behaviors of superalloys at elevated temperatures. In this study, the oxidation and carburization of Inconel 617 and Incoloy 800H at 950 °C in impure helium were investigated. According to thermodynamics and kinetics, the different ratios PH2/PH2O and PCH4/PH2O were set in this research, which leads to different oxidation and carburization behaviors of alloys at 950 °C. When PH2O is at a constant concentration, high PH2 would reduce the oxidation capacity of the atmosphere, and high PCH4 will cause severe carburization. Due to the presence of the silicon oxide interlayer, the oxide layer of Incoloy 800H has poor adhesion and it is easier to be destroyed at high temperatures. It may be the reason that Incoloy 800H has weaker resistance to carburization.

Modeling of irradiated dimensional change strain in MSR graphite moderator

When irradiated in the core of a thermal spectrum nuclear reactor, graphite undergoes dimensional changes which depend on the neutron spectrum. The graphite lifespan is a limiting parameter in MSR designs; it correlates strongly with the graphite’s dimensional changes. Coupled modeling of the thermal–hydraulic, neutronic, and thermal-mechanics conditions is necessary to evaluate the graphite thermal and mechanical responses to the MSR core conditions. To assess graphite dimensional changes associated with irradiation, a computational model was developed using GeN-Foam. In particular, the thermo-mechanical region solver was expanded to account for the moderator’s response to the core’s neutron kinetics. These capabilities enable modeling of the thermal deformation of graphite moderator structures as well as calculation of the moderator temperature at the thermo-mechanical mesh cell level. The GeN-Foam graphite moderator model was benchmarked using the experimental and computational data of graphite moderator thermal response in high temperature gas-cooled reactors.

Electrochemical behavior of matrix graphite in nitric acid by cyclic voltammetry

Post-irradiation examination of irradiated spherical fuel elements in high-temperature gas-cooled reactors (HTGRs) requires the separation of integrated tri-structural isotropic (TRISO) particles from matrix graphite (MG). The electrochemical method is a promising approach for separating intact TRISO particles and exploring the fission product source terms in MG. In this work, the electrochemical behavior of MG and its three compositions, i.e., artificial graphite (AG), natural graphite (NG), and binder carbon (BC) were investigated experimentally by cyclic voltammetry using nitric acid as an electrolyte, and these three compositions showed very different electrochemical properties. Compared with BC, AG and NG exhibited relatively higher electrochemical reactivity controlled by diffusion, while BC performed almost identical adsorbing-controlled electrochemical behavior with MG composites, which probably resulted from encased AG and NG flakes by bulk BC. According to cyclic voltammetric and characterizing results of functional groups on the surface of the specimens, the deconsolidation reacting procedures of MG could be further interpreted into a two-step process. The first step was the electrochemical oxidation of BC by unsaturated , defective, and sp2 carbon atoms. In the second step, the NG and AG flakes were intercalated, oxidized and exfoliated by combined reactions of hydrolysis, intercalation, and oxygen evolution. The proposed electrochemical procedures obtained in this study provide the theoretical foundation to PIE investigations of HTGR spent fuels.

Life cycle assessment of nuclear hydrogen production processes based on high temperature gas-cooled reactor

Conventional hydrogen production technologies mostly fossil fuels as energy and material basis. The rapid development of nuclear energy in recent years offers a new opportunity. Clean electricity and process heat generated by nuclear reactors can provide energy for hydrogen production, effectively reducing the environmental burden. This study used life cycle assessment (LCA) method to sort out the inputs and outputs of the nuclear hydrogen production processes and analyze the environmental impacts based on local data in China. In this study, we constructed frameworks for two nuclear energy-based processes and created four different scenarios to compare the effect of energy efficiency. Six indicators were used to quantify the environmental impact. The results showed that: (1) electrolysis cell manufacturing and spent fuel disposal generate the largest emissions in hydrogen production. (2) S–I cycle is sensitive to heat transfer efficiency, while high-temperature electrolysis is more sensitive to power generation efficiency; (3) The environmental impact of high-temperature electrolysis (without carrier gas) is slightly lower than that of S–I cycle, but the advantage will disappear as energy efficiency increases. At present, high-temperature electrolysis offers a clean alternative to conventional technologies for hydrogen energy and hydrogen economy. The S–I cycle might have a better prospect in the future. Our study results will provide a scientific assessment of the possibilities of developing nuclear energy for hydrogen production in China and help to make some decisions and policies.

DESIGN OF HELICAL TYPE STEAM GENERATOR FOR EXPERIMENTAL POWER REACTOR

Reaktor Daya Eksperimental (RDE) is a high-temperature gas-cooled reactor (HTGR) for electricity generation, heat generation, and hydrogen production by Batan. Empirical and numerical calculations are needed to strengthen the existing design. The numerical method by computational fluid dynamic (CFD) analyzes temperature distribution and pressure drop along the pipe. The Batan RDE steam generator design has a seven-layer helical pipe model, while this research uses a one-layer helix pipe. In empirical calculations, the heat transfer region has three sections; single-phase liquid, two-phase, and single-phase vapor heat transfer. In numerical calculations, apply the assumption of constant heat flux and constant working fluid properties. The results of empiric calculations data showed that the helical pipe height was 3.98 m, shorter than the Batan design, which is 4.97 m. This considerable difference due to empirical calculations did not cover the safety factor. The results of numerical calculations show that in the single-phase, empiric calculation data were acceptable since the different values of numerical calculations for empiric calculations data were below 10%. Meanwhile, the case of the two-phase numerical calculations is not satisfactory and needs further research to obtain optimal results.

Numerical simulation of the transient graphite oxidation process based on a dynamic mesh method

The high-temperature gas-cooled reactor uses graphite-matrix spherical fuel elements to coat the tri-structural isotropic fuel particles (TRISO). In accident conditions, graphite oxidation may cause fuel particle exposure, which further leads to severe radioactive contamination. Therefore, it is very important to study the oxidation behavior of graphite during air-ingress accidents. This work aimed to develop a reaction kinetic model of graphite oxidation based on the Institute of Nuclear and New Energy Technology (INET) experiment correlations and to study the transient process of graphite oxidation using computational fluid dynamics and a dynamic mesh method. The results show that convective diffusion and mass diffusion dominate the distribution of reaction products at different flow rates. The reaction rate in the early stage of oxidation was high and then decreased rapidly until the oxygen consumption rate and supplementation rate reached equilibrium. After that, the oxidation rate decreased slowly due to the reduction of graphite surface area. The fluctuating flow around the graphite leads to anisotropic oxygen concentrations at different inflow angles on the graphite surface. The fluctuations caused a non-uniform ellipsoidal graphite structure, in which graphite has a smaller radius of curvature and pits on the windward side and a larger radius of curvature on the leeward side. The oxidation mass of graphite is in good agreement with the INET experimental data, which shows that this study can provide a method for the estimation of the mass and shape of graphite oxidation.

Comparison of unirradiated and irradiated AGR-2 TRISO fuel particle oxidation response

The silicon carbide (SiC) coating in a tristructural isotropic (TRISO) particle acts as a barrier to fission product release during reactor operation and accident scenarios. Oxidation and subsequent failure of the SiC layer during a rare air ingress event is a proposed mechanism for fission product release in a high-temperature gas-cooled reactor (HTGR). Although previous oxidation studies have analyzed unirradiated TRISO particle response, this study compared the oxidation behavior of irradiated and unirradiated TRISO particles from the second Advanced Gas Reactor Fuel Development and Qualification Program irradiation experiment (AGR-2). Particles with exposed SiC were subjected to six varying oxidizing tests in the Furnace for Irradiated TRISO Testing (FITT), examined for failure fraction with the Irradiated Microsphere Gamma Analyzer (IMGA) and characterized with focused ion beam and scanning/transmission electron microscopy techniques to analyze the oxide layer. Uncorrelated unirradiated particle failures throughout the series of exposures suggests that external factors inherent to the experiment increased particle failure sensitivity. However, irradiated particle observations indicated an increased failure response at 400 h 1400 °C in both 2% and 21% O2 atmospheres above failure associated with external factors. Oxide thickness measurements after 400 h at 1400 °C revealed a greater oxidation rate than predicted by parabolic growth, which was attributed to the increased complexity of the oxide structure at longer exposure times. Altering the atmosphere from 21% to 2% O2 reduced the average oxide thickness by approximately 12%–14% in both irradiated and unirradiated particles at 400 h 1400 °C. Overall, the minor variations observed between irradiated and unirradiated particles in this study led to the conclusion that unirradiated TRISO particles can be used to approximate irradiated TRISO oxidation kinetics.

Design analysis of a printed circuit heat exchanger for HTGR using a 3D finite elements model

The intermediate heat exchanger (IHX) is a critical component in most of the High Temperature Gas-cooled Reactor plant layouts. The printed circuit heat exchanger (PCHE) has been studied as an advantageous candidate for this application last years. The extreme operating conditions of the IHX and the higher structural integrity of the PCHE over other heat exchanger technologies have motivated these studies. In this work, the structural assessment of a zigzag channel PCHE under IHX conditions was carried out by means of computational fluid dynamics (CFD) and Finite Element Analysis (FEA). The ASME Boiling and Pressure Vessel Code, Sections III and VIII, were applied to evaluate the safety operation of the PCHE. The temperature field in steady state was obtained using a 3D CFD model. Then it was used as thermal load to calculate thermal stress using a FEA model. The influence of the zigzag angle and the bend radius on the thermal and mechanical stresses was analyzed. Thus, the values studied for the zigzag angle were 0° (straight), 15°, 30 and 45°, while for the bend radius were 0.5 mm, 2 mm, 4 mm and 6 mm. The importance of the 3D temperature field to calculate thermal stress in a zigzag channel was evidenced due to the appearance of local temperature differences in the channel cross-sections. The change in flow direction created areas of higher local temperatures where the flow lines change their directions impacting the channel wall. A significant thermal stress concentration was found in these areas, and this effect intensified with increasing zigzag angle. The bend radius was decisive to reduce the thermal stress concentration when the zigzag angle was higher than 15°.

Printed Circuit Heat Exchangers (PCHEs): A Brief Review

Heat exchangers and other heat transfer devices/systems play vital roles of heat transfer in thermal fluid flow systems for industrial application. Sodium cooled fast reactors are normally designed to have two loops of sodium coolants and one loop of water coolant which generates steam for power production. The two loops of sodium coolants consist of primary cooling system of sodium which cools the fuel rods of the reactor core and secondary cooling system of sodium transferring heat from the sodium primary cooling system. The water-cooling system transfers heat from the secondary cooling system of sodium for steam generation. Lead cooled fast reactors on the other hand are designed to have primary cooling system of lead cooling the fuel rods in the reactor core and secondary cooling system of water transferring heat from the lead cooling primary system for steam generation. Water cooled Nuclear Power Plants used water to cool the reactor core in the primary system and the heat removed from the core is used for steam generation directly as in BWRs and SCWRs or in the secondary system of heat exchanger as in PWRs. Other reactor systems such as Gas-cooled fast reactor (GFR), Molten-salt reactor (MSR), High-temperature gas-cooled reactor (HTGR), and Small Modular Reactors (SMRs) also have various types of heat exchangers in their designs to support power/electricity generation. Appropriate heat exchangers are therefore needed for various stages of heat transfer in power generation systems. Thus, Heat exchangers and other heat transfer devices/systems play vital roles of heat transfer in thermal fluid flow systems for industrial applications. This study presents brief review of PCHEs which have comparable advantages over other types of heat exchangers. Recent studies on PCHEs and other heat exchanger types have been reviewed. Design and optimization of PCHEs, optimization of Brayton and Rankine circles, and fluid flow and heat transfer devices/systems have been discussed briefly. The review findings show that the design and optimization of PCHEs depends on the intended industrial application of the heat exchanger. The various channel types and channel cross-section types available for design and optimisation as well as the design and optimised system being able to withstand high pressure and temperature conditions in addition to its compact size for the intended industrial application make PCHEs unique among other types of heat exchangers.

Fusion-fission hybrid reactor facility: power profiling

The current state of research in the field of nuclear and thermonuclear power aimed at creating power generation plants makes it possible to predict the further development of modern power industry in the direction hybrid reactor power plants. Such hybrid systems include a tokamak with reactor technologies, worked out in detail in Russia, and systems with an additional source of neutrons. Power generation plants using tokamaks and accelerators with the required level of proton energy will be of exceptionally large size and power, which will postpone their construction on an industrial scale to the distant future. The ongoing research is aimed at the development of small generation and has the prospect of entering the field of energy use in a shorter period. The hybrid reactor facility under study consists of an axisymmetric assembly of fuel blocks of a high-temperature gas-cooled reactor and a linear plasma source of additional neutrons. The paper demonstrates the results of optimization plasma-physical, thermophysical and gas-dynamic studies, the purpose of which is to level the distortions of the power density field, which are formed in the volume of the multiplicating part of the facility due to the pulsed operation of the plasma source of D-T-neutrons. The studies on increasing the “brightness” of the source and modeling its operating modes were carried out using the DOL and PRIZMA programs. The thermophysical optimization and gas-dynamic calculations were performed using the verified SERPENT and FloEFD software codes. The calculations were made on a high-performance cluster of the Tomsk Polytechnic University.

A GPU-based DEM model for the pebble flow study in packed bed: Simulation scheme and validation

This study developed a GPU-based DEM (GPU-DEM) model for the pebble flow in the packed bed. The details of the GPU-DEM scheme have been illustrated. The uniform mesh neighbor-search method, unified memory addressing (UMA), and special boundary conditions were incorporated in the GPU-DEM model. Cases of the direct impact of two identical particles, the angle of repose of particles drained from a lifted hopper, and the circulating hopper to mimic a pebble flow in High-Temperature Gas-cooled Reactor (HTGR) reactors were carried out to systematically verify the applicability, computational efficiency, and accuracy of the GPU-DEM model. Related results, like the unloading speed, porosity, and velocity distribution, obtained by the GPU-DEM model have been validated too. In addition, the effects of single-precision floating-point number computation on GPU were discussed. The results demonstrated that the GPU-DEM can achieve 18– 20 times acceleration while maintaining accuracy, even when single-precision floating-point numbers were used for calculation. Therefore, it is a powerful tool to explore the particle flows in chemical engineering applications.

Microstructural analysis of tristructural isotropic particles in high-temperature steam mixed gas atmospheres

High-temperature gas-cooled reactors (HTGRs) use tristructural isotropic (TRISO) particles embedded in a graphitic matrix material to form the integral fuel element. Potential off-normal reactor conditions for HTGRs include steam ingress with temperatures above 1,000 °C. Fuel element exposure to steam can cause the graphitic matrix material to evolve, forming an atmosphere composed of oxidants and oxidation products and potentially exposing the TRISO particles to these conditions. Investigating the oxidation response of TRISO particles exposed to a mixed gas atmosphere will provide insight into the stability under off-normal conditions. In this study, surrogate TRISO particles were exposed to high temperatures (T = 1,200 °C) in flowing steam (3% < pH2O < 21%) and CO (pCO < 1%) to determine the oxidation behavior of the SiC layer when exposed to various mixed gas atmospheres. Scanning electron microscopy, x-ray diffraction, and focused ion beam milling was used to determine the impact of CO and steam on the oxidation behavior of the SiC layer. The data presented demonstrates how the SiC layer showed strong oxidation resistance due to limited SiO2 growth and maintained its structural integrity under these off-normal conditions.

Improved Methods for Calculating the Neutron Diffusion Coefficient of General Pebble Bed

There exists a neutron streaming effect in the high-temperature gas-cooled pebble-bed reactor (HTGR) pebble-bed core caused by the spatial heterogeneity of the neutron’s free path, which has a remarkable impact on neutron leakage. It is necessary to take into consideration the streaming effect in evaluating the homogenized diffusion coefficient of the pebble bed, prior to the whole-core diffusion calculation. In this paper, two methods are proposed for calculating the homogenized multigroup diffusion coefficient of the pebble bed based on migration area conservation theory and Benoist’s theory, respectively. Compared with existing methods, the newly proposed methods are adaptable to a general pebble bed consisting of multitype pebbles and nonvacuum atmosphere. Numerical results demonstrate the proposed methods’ effectiveness and consistency in evaluation of the pebble-bed homogenized diffusion coefficient.

Design of a laser welding plugging system for the steam generators of high-temperature gas-cooled reactor

As fourth-generation nuclear power equipment, high-temperature gas-cooled reactor (HTGR) is attracting more and more attention both in industry and academia. The core outlet temperature of the steam generator (SG) in HTGR is much higher than that of pressurized water reactor steam generators, which is more likely to cause damage. Till now, welding plugging is still one of the important means to prevent the leakage of damaged heat transfer tubes. However, how to improve the automation of welding plugging to reduce the short-range operation time in a radioactive environment is still difficult. This work proposes the design of a laser welding plugging system for the HTGR. Firstly, the design requirements are analyzed and the overall design scheme is given. The whole system consists of a laser generator, a multi-axis robot, and supporting devices. Secondly, the key supporting devices including a laser welding head, a pneumatic clamping jaw, and a single-axis motion platform are explained in detail. The structure simulation based on SolidWorks is adopted, which provides the basis for the design of the single-axis motion platform. Thirdly, the control system which takes the PLC control system as the main control unit and the industrial computer as the monitoring data processing unit is described. Finally, an experimental system is built and the laser welding experiment is carried out to test the feasibility of the designed laser welding plugging system.

Inventories of Short-Lived Fission Gas Nuclides in Nuclear Reactors

Taking inventories in reactor cores is critical for understanding their radioactive source terms and establishing the relationship between the activity concentration in the primary loop and the status of the reactor core’s fuel. However, there is a niche in which a simple but accurate relationship between reactor conditions and nuclide inventories can reliably predict the fission gas nuclide activities of the reactor core in the primary loop. In this study, a simple and efficient model called “Inventories of a Point Reactor for Fission Gas Nuclides” (IPRFGN) was proposed to calculate and interpret such inventories, in which a 10 MW high-temperature gas-cooled experimental reactor (HTR-10) was used as the test case. The present study findings were consistent with those of a general point–depletion burnup code such as the KORIGEN code. Here, the relative error was &lt;1%. Based on the application of the IPRFGN model in HTR-10, the results indicate that the proposed IPRFGN model has provided the relationship between the inventories of fission gas nuclides in the core and the reactor conditions in all types of nuclear fission reactors. In the future, the IPRFGN model will be used for calculating fission gas nuclide inventories in various reactors.

Surface and in-depth structural changes in nuclear graphite irradiated with noble gases described with Raman imaging

4th Generation high-temperature gas-cooled nuclear reactors (HTGR) are regarded as possible sources of industrial heat in Poland and Europe, allowing for a substantial reduction of the dependency on gas and coal import. It is mainly due to their safety of use, reliability and economy in a current energetic crisis. In this work, graphite, as a primary construction material and neutron moderator in HTGR, was evaluated before and after ion irradiation since its properties depend on the material’s structure and purity. Commercial graphite materials (IG-110, NBG-17) and the laboratory’s in-home material were chosen for the exemplary samples. The structural damage in HTGR was simulated with energetic Ar+ and He+ ions with fluencies from 1E12 to 2E17 ion/cm2. Raman imaging was chosen to assess radiation damage build-up: the crystallites’ evolution, occurrence and types of defects. The recorded evolution showed stronger disordering of the material with heavier Ar+ ions than with He+.

Research on non-LWRs operational intervention level based on HTR

Nuclear emergency is the final line of defense for humanity and the environment, and operational intervention level (OIL) becomes a key technical element pre-specified in emergency preparedness and guides emergency response actions. The International Atomic Energy Agency (IAEA) issued a new version of OIL system based on light water reactors (LWRs) in 2017. To explore its applicability for non-LWRs, this paper presents a set of specified OIL determination procedures for high temperature gas-cooled reactors (HTRs). Considering a representative OIL based on ground monitoring (OIL1) as the case study, it concludes that the default OIL1 is 500 μSv/h for HTR-PM, half of the IAEA recommended value. In addition, through sensitivity analysis, we find that the slow release of radionuclides does not affect the determination of the default OIL1 value. From the perspective of OIL, this research further confirms that HTR reactor types can technically simplify off-site emergency requirements.

Development of a hybrid model for fuel performance analysis of spherical fuel elements under normal conditions

The spherical fuel element (SFE) dispersed with many tri-isotropic fuel (TRISO) particles has been primarily employed in the high-temperature gas-cooled reactor. Owing to the complex structure of SFE, behaviors will be evolved for both TRISO particles and the fuel matrix with the increasing burnup. To evaluate the fuel performance of SFE, the homogenization method has been conventionally adopted for SFE scale modeling, and dispersed TRISO particles are analyzed subsequently with fixed temperature boundary conditions. However, owing to the varying properties of TRISO particles under reactor operation, the homogenization models dedicated to certain circumstances can hardly reflect the properties evolution of TRISO with high fidelity, which may lead to uncertainties in analysis results. To this end, a hybrid model, namely the conventional homogenization model informed by the lower scale model online, has been developed in this paper, which may increase modeling fidelity for both SFE and particles. Specifically, property models for the TRISO particle and graphite matrix have been implemented in COMSOL. Subsequently, based on the linking of COMSOL and MATLAB, the hybrid model for SFE analysis was established. Through the hybrid model, the temperature field and stress distribution of TRISO, the evolution of effective properties, and the thermal and mechanical state of the SFE were obtained and analyzed. The results show that within the scope of this paper, the central temperature of TRISO in the innermost region rises with the increasing burnup for the decreasing thermal conductivity of materials, and the peak value is about 1324 K, which is about 60 K higher than that in the outermost region. The difference in the maximum plenum pressure is more than 5 MPa among TRISO particles, which is mainly caused by the variations in CO production. Consequently, the hoop stress of SiC in the outermost region is slightly elevated. The distribution of effective thermal conductivity (ETC) for TRISO is nearly 4.3 W/(m·K) at the beginning of life (BOL), while it was reduced to 3.6 W/(m·K) at the end of life (EOL). There are marginal variations along the radial direction for the ETC of TRISO. Meanwhile, the ETC of SFE decreases from about 21 W/(m·K) for fresh fuel to 13 W/(m·K) for spent fuel. The predicted effective elastic modulus and coefficient of thermal expansion of SFE increase gradually versus time. The former is caused by the densification of pyrolytic carbon, and the latter is mainly determined by the high volume fraction of matrix and the changes of CTE for pyrolytic carbon. In addition, the comparison in the temperature field between the proposed model and the Differential Effective Material Theory (D-EMT) model may further demonstrate the feasibility and effectiveness of the hybrid model.