• Journal of Information Processing Systems
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• v.1 no.1 s.1
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• pp.36-40
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• 2005

TRISO (Tri-Isotropic)-coated fuel particle is widely applied due to its higher stability at high temperature and its efficient retention capability for fission products in the HTGR (high temperature gas-cooled reactor), one of the highly efficient Generation IV reactors. The typical ball-type TRISO-coated fuel particle with a diameter of about 1 mm is composed of a nuclear fuel particle as a kernel and of outer coating layers. The coating layers consist of a buffer PyC, inner PyC, SiC, and outer PyC layer. In this study, a digital image processing algorithm is proposed to automatically measure the thickness of the coating layers. An FBP (filtered backprojection) algorithm was applied to reconstruct the CT image using virtual X-ray radiographic images for a simulated TRISO-coated fuel particle. The automatic measurement algorithm was developed to measure the coating thickness for the reconstructed image with noises. The boundary lines were automatically detected, then the coating thickness was circularly by the algorithm. The simulation result showed that the measurement error rate was less than 1.4%.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3874-3897
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• 2023

A high-fidelity computational fluid dynamics (CFD) analysis was performed using the Large Eddy Simulation (LES) model for the lower plenum of the High-Temperature Test Facility (HTTF), a ¼ scale test facility of the modular high temperature gas-cooled reactor (MHTGR) managed by Oregon State University. In most next-generation nuclear reactors, thermal stress due to thermal striping is one of the risks to be curiously considered. This is also true for HTGRs, especially since the exhaust helium gas temperature is high. In order to evaluate these risks and performance, organizations in the United States led by the OECD NEA are conducting a thermal hydraulic code benchmark for HTGR, and the test facility used for this benchmark is HTTF. HTTF can perform experiments in both normal and accident situations and provide high-quality experimental data. However, it is difficult to provide sufficient data for benchmarking through experiments, and there is a problem with the reliability of CFD analysis results based on Reynolds-averaged Navier-Stokes to analyze thermal hydraulic behavior without verification. To solve this problem, high-fidelity 3-D CFD analysis was performed using the LES model for HTTF. It was also verified that the LES model can properly simulate this jet mixing phenomenon via a unit cell test that provides experimental information. As a result of CFD analysis, the lower the dependency of the sub-grid scale model, the closer to the actual analysis result. In the case of unit cell test CFD analysis and HTTF CFD analysis, the volume-averaged sub-grid scale model dependency was calculated to be 13.0% and 9.16%, respectively. As a result of HTTF analysis, quantitative data of the fluid inside the HTTF lower plenum was provided in this paper. As a result of qualitative analysis, the temperature was highest at the center of the lower plenum, while the temperature fluctuation was highest near the edge of the lower plenum wall. The power spectral density of temperature was analyzed via fast Fourier transform (FFT) for specific points on the center and side of the lower plenum. FFT results did not reveal specific frequency-dominant temperature fluctuations in the center part. It was confirmed that the temperature power spectral density (PSD) at the top increased from the center to the wake. The vortex was visualized using the well-known scalar Q-criterion, and as a result, the closer to the outlet duct, the greater the influence of the mainstream, so that the inflow jet vortex was dissipated and mixed at the top of the lower plenum. Additionally, FFT analysis was performed on the support structure near the corner of the lower plenum with large temperature fluctuations, and as a result, it was confirmed that the temperature fluctuation of the flow did not have a significant effect near the corner wall. In addition, the vortices generated from the lower plenum to the outlet duct were identified in this paper. It is considered that the quantitative and qualitative results presented in this paper will serve as reference data for the benchmark.

• Nuclear Engineering and Technology
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• v.56 no.9
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• pp.3835-3850
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• 2024

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.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.729-738
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• 2009

Given the evolution of High-Temperature Gas-cooled Reactor(HTGR) designs, the source terms for licensing must be developed. There are three potential source terms: fission products, actinides in the fuel and tritium in the coolant. It is necessary to provide first an inventory of the source terms under normal operations. An analysis of source terms has yet to be performed for HTGRs. The previous code, which can estimate the inventory of the source terms for LWRs, cannot be used for HTGRs because the general data of a typical neutron cross-section and flux has not been developed. Thus, this paper uses a combination of the MCNP, ORIGEN, and MONTETEBURNS codes for an estimation of the source terms. A method in which the HTR-10 core is constructed using the unit lattice of a body-centered cubic is developed for core modeling. Based on this modeling method by MCNP, the generation of fission products, actinides and tritium with an increase in the burnup ratio is simulated. The model developed by MCNP appears feasible through a comparison with models developed in previous studies. Continuous fuel management is divided into five periods for the feeding and discharging of fuel pebbles. This discrete fuel management scheme is employed using the MONTEBURNS code. Finally, the work is investigated for 22 isotope fission products of nuclides, 22 actinides in the core, and tritium in the coolant. The activities are mainly distributed within the range of $10^{15}{\sim}10^{17}$ Bq in the equilibrium core of HTR-10. The results appear to be highly probable, and they would be informative when the spent fuel of HTGRs is taken into account. The tritium inventory in the primary coolant is also taken into account without a helium purification system. This article can lay a foundation for future work on analyses of source terms as a platform for safety assessment in HTGRs.

• Journal of the Korean Society of Safety
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• v.24 no.3
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• pp.54-58
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• 2009

A substitute energy development has been required due to the exhaust of the fossil fuel and an environmental problem. Consequently, possible technologies producing hydrogen from water that does not release carbon is a very promising technology. Also, Iodine-Sulfur(IS) thermochemical water decomposition is one of the promising processes that are used to produce hydrogen efficiently using the high temperature gas-cooled reactor(HTGR) as an energy source that is possible to supply heat over 900$^{\circ}C$. In this study, to make a initiating events identification for the IS process, Master Logic Diagram(MLD) is used and 9 initiating events that cause a leakage of the chemical material are identified. Also, 6 events are identified among 9 initiating events above and are quantified using event tree.

• Journal of Energy Engineering
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• v.15 no.2 s.46
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• pp.107-117
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• 2006

The status of water splitting thermochemical cycle for hydrogen production was reviewed in this article. Mass production of hydrogen could be possible using the thermochemical process which is similar to the concept of conventional chemical reaction system if the high temperature heat source is available. The mediators (chemicals and reagents) should be used to split chemically stable water, and should be recycled in a closed cycle in order to be environmentally acceptable. Though there is no process to reach commercial stage, IS cycle, two-step cycles based on metallic oxide such as ZnO/Zn, $Fe_3O_4/FeO$ and the associated cycles are attracted due to their possibilities of application. Development of materials for high temperature and/or corrosive conditions during thermochemical process is still important topic in some thermochemical processes.

• Journal of the Korean Ceramic Society
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• v.50 no.1
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• pp.37-42
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• 2013

Tristructural-isotropic (TRISO)-coated particles were fabricated by a fluidized-bed chemical vapor deposition (FBCVD) method for use in a very high temperature gas-cooled reactor (VHTR). ZrC as a constituent layer of TRISO coating layers was deposited by a chloride process using $ZrCl_4$ and $CH_4$ source gases in a temperature range of $1400^{\circ}C$ and $1550^{\circ}C$. The change in the microstructure of ZrC depending on the deposition temperature and its effect on the hardness were evaluated. As the deposition temperature increased to $1500^{\circ}C$, the grain size of the ZrC increased and the hardness of the ZrC decreased according to the Hall-Petch relationship. However, at $1550^{\circ}C$, the ZrC layer was highly non-stoichiometric and carbon-rich and did not obey the Hall-Petch relationship in spite of the decrease of the grain size. A considerable amount of pyrolytic carbon at the grain boundaries of the ZrC as well as coarse granular pyrolytic carbon were locally distributed in the ZrC layer deposited at $1550^{\circ}C$. Therefore, the hardness decreased largely due to the formation of a large amount of pyrolytic carbon in the ZrC layer.

• Journal of the Korean Ceramic Society
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• v.42 no.9 s.280
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• pp.618-623
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• 2005

HTGR (High Temperature Gas-Cooled Reactor) is highlighted to next generation power plant for producing the clean hydrogen gas. In this study, the spherical $UO_2$ kernel via $UO_3$ gel particles was prepared by the sol-gel process. Raw material of slightly Acid Deficient Uranyl Nitrate (ADUN) solution, which has pH = 1.10 and $[NO_3]/[U]$ mole ratio = 1.93, was obtained from dissolution of $U_3O_8$ powder with conc.-$HNO_3$. The surface of these spherical $UO_3$ gel particles, which was prepared from the broth solution, consisted of 1 M-uranium, 1 M-HMTA, and urea, were covered with the fine crystallite aggregates, and these particles were so hard that crushed well. But the other $UO_3$ gel particles prepared with the broth solution, consisted of 2 M-uranium, 2 M-HMTA, and urea, have soft surface characteristics and an amorphous phase. This type of $UO_3$ gel particles is some chance of doing possibility of high density from the compaction. The amorphous $UO_3$ gel particles was converted to $U_3O_8$ and then $UO_2$ by calcination at $600^{\circ}C\;in\;4\%\;-\;H_2\;+\;N2$ atmosphere.

• Journal of the Korean Ceramic Society
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• v.44 no.7
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• pp.368-374
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• 2007

CVD-SiC coating has been introduced as a protective layer in TRISO nuclear fuel particle of high temperature gas cooled reactor (HTGR) due to its excellent mechanical stability at high temperature. In order to prevent the failure of the TRISO particles, it is important to evaluate the fracture strength of the SiC coating layer. It is needed to develop a new simple characterization technique to evaluate the mechanical properties of the coating layer as a pre-irradiation step. In present work, direct strength measurement method with the specimen of hem i-spherical shell configuration was suggested. The indentation experiment on a hemisphere shell with a plate indenter was conducted. The fracture strength of the coating layer is related with the critical load for radial cracking of the shell. The finite element analysis was used to drive the semi-empirical equation for the strength measurement. The SiC hemispherical shells were successfully recovered from the section-grinding of TRISO coated particle and successive heat treatment in air. The strength of CVD-SiC coating layer was evaluated from the experimentally measured critical load during the indentation on SiC hemisphere shell. Weibull diagram of fracture strength was also constructed. This study suggested a new strength equation and experimental method to measure the fracture strength of CVD-SiC coating of TRISO coated fuel particles.

• Journal of the Korean Society for Nondestructive Testing
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• v.26 no.2
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• pp.69-76
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• 2006

TRISO(tri-isotropic)-coated fuel particle technology is utilized owing to its higher stability at a high temperature and Its efficient retention capability for fission products In the HTGR(high temperature gas-reeled reactor). The typical spherical TRISO fuel panicle with a diameter of about 1mm is composed of a nuclear fuel kernel and outer coating layers. The outer coating layers consist of a buffer PyC(pyrolytic carbon) layer, Inner PyC(1-PyC) layer, SiC layer, and outer PyC(O-PyC) layer Most of the Inspection Items for the TRTSO-coated fuel particle depend on destructive methods. The coating thickness of the TRISO fuel particle can be nondestructively measured by the X-ray radiography without generating radioactive wastel. In this study, the coaling thickness for the simulated TRISO-coated fuel particle with $ZrO_2$ kernel Instead of $%UO_2$ kernel was measured by using micro-focus X-ray radiography with micro-focus X-ray generator and flat panel detector The radiographic image was also enhanced by image processing technique to acquire clear boundary lines between coating layers. The coaling thickness wat effectively measured by applying the micro-focus X-ray radiography The inspection process for the TRISO-coated fuel particles will be improved by the developed micro-focus X-ray radiography and digital image processing technology.