UO2 Kernels Research Articles

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.

Uncertainty Quantification of Engineering Parameters for a Nuclear Reactor Loaded with Dispersed Fuel Particles

Owing to their high intrinsic safety, dispersed fuel particles are an important fuel pattern for fourth-generation nuclear reactors. Due to the unique cladding layers and the random dispersion characteristics, dispersed fuel particles significantly differ from pressurized water reactors regarding operation-induced uncertainty. This study quantitatively analyzed overall uncertainty while considering a random distribution of dispersed fuel particles, material thickness, and fuel enrichment. The results demonstrated that, for all packing fractions, the uncertainty induced by the random dispersion of dispersed fuel particles was below 0.03%. For every packing fraction, the differences between the results obtained by the regular and the random distribution models increased, and then decreased, until reaching its maximum (1.297%) at 15%. Keff decreased as the radius of the UO2 kernel increased; Keff increased as the thickness of the cladding layer increased; the uncertainty of Keff was 1.003% when a random distribution of particles, material thickness, and fuel enrichment were taken into consideration; the uncertainty of the power distribution of reactor core assemblies was maximized (1.495%) at the edge of the reactor core. Quantitative analysis of uncertainty provides references for the optimization of design and safety margin analysis for reactors.

Application of vacuum membrane distillation-Fenton oxidation process for deep purification of low-level radioactive organic wastewater

During the production of UO2 kernel for the high temperature gas-cooled nuclear reactor, a large amount of low-level radioactive organic wastewater (LLROW) containing ammonia, uranium and organic compounds are generated. The current treatment process is suffered from the production of radioactive solid wastes, low adsorption efficiency of organic matters and the frequent membrane fouling during RO process. In this study, vacuum membrane distillation (VMD)-Fenton oxidation process was developed for the deep purification of the LLROW. The process parameters were optimized and the system performance was evaluated. As a result, all major contaminates such as uranium, NH3-N, TN and COD could meet the GB8978-1996 (secondary standard) and GB/T31962-2015 (B standard) for direct discharge. The membrane fouling was also studied by membrane autopsy as well as the quantum calculation. It was found that weak hydrogen bonds were formed between PTFE membrane and contaminants while the membrane fouling could be recovered by 0.1 M HCl cleaning.

Study on the effect of long-term high temperature irradiation on TRISO fuel

In the core of the WWR-K reactor, a long-term irradiation of tristructural isotopic (TRISO)-coated fuel particles (CFPs) with a UO2 kernel was carried out under high-temperature gas-cooled reactor (HTGR)-like operating conditions. The temperature of this TRISO fuel during irradiation varied in the range of 950–1100 °C. A fission per initial metal atom (FIMA) of uranium burnup of 9.9% was reached. The release of gaseous fission products was measured in-pile. The release-to-birth ratio (R/B) for the fission product isotopes was calculated. Aspects of fuel safety while achieving deep fuel burnup are important and relevant, including maintaining the integrity of the fuel coatings. The main mechanisms of fuel failure are kernel migration, silicon carbide corrosion by palladium, and gas pressure increase inside the CFP. The formation of gaseous fission products and carbon monoxide leads to an increase in the internal pressure in the CFP, which is a dominant failure mechanism of the coatings under this level of burnup. Irradiated fuel compacts were subjected to electric dissociation to isolate the CFPs from the fuel compacts. In addition, nondestructive methods, such as X-ray radiography and gamma spectrometry, were used. The predicted R/B ratio was evaluated using the fission gas release model developed in the high-temperature test reactor (HTTR) project. In the model, both the through-coatings of failed CFPs and as-fabricated uranium contamination were assumed to be sources of the fission gas. The obtained R/B ratio for gaseous fission products allows the finalization and validation of the model for the release of fission products from the CFPs and fuel compacts. The success of the integrity of TRISO fuel irradiated at approximately 9.9% FIMA was demonstrated. A low fuel failure fraction and R/B ratios indicated good performance and reliability of the studied TRISO fuel.

Atomistic and cluster dynamics modeling of fission gas (Xe) diffusivity in TRISO fuel kernels

TRISO fuel particles are candidates for use in next generation reactors including gas reactors, fluoride salt-cooled high temperature reactors, and micro-reactors. The UCO fuel kernel consists of a uranium dioxide (UO2) and uranium carbide mixture. The addition of UC2 helps suppress the formation of carbon monoxide gas, which led to failures during initial TRISO development. The addition of uranium carbide alters the chemistry of the UO2 kernel, which is known to influence performance parameters such as fission gas diffusivity, although the impact has not been quantified and no models exist that take the change in chemistry into account. Therefore, better understanding and more accurate models of the impact of chemistry on fuel performance are of high priority. In this paper, a first-principles density functional theory (DFT) and empirical potential based multi-scale study has been carried out to model the diffusivity of fission gas xenon (Xe) in UCO TRISO fuel kernels. The focus is on the UO2 component in the UCO fuel kernels, as that represents the largest volume fraction of the fuel kernels. The study relies on DFT and empirical potential calculations to determine Xe and point defect properties, which are then used in thermodynamic and kinetic models to predict diffusion for intrinsic conditions. In addition, the information is utilized in cluster dynamics simulations using the Centipede code to estimate the impact of irradiation on defect transport. The presence of UC2 or UC2−x in the UCO fuel kernels is shown to have a substantial impact on the UO2 non-stoichiometry by inducing oxygen vacancies and driving UO2 sub-stoichiometric, which causes much slower Xe diffusion in UCO compared to light water reactor UO2 fuel. The application of this model in fuel performance simulations using the Bison code is also demonstrated.

Numerical evaluation of AGR-2 fission product release

The AGR-2 experiment produced normal operation and accident condition fuel performance data for tri-structural isotropic (TRISO) particles with UCO and UO2 kernels. Data for compacts with no failed particles and for compacts with one or more failed particles are available. To model the fission product diffusion and release from these compacts, it is important to consider the computational mesh as well as the heat conduction and mass diffusion properties of the materials in a TRISO particle. Code and solution verification studies with the Bison fuel performance code were performed to show that the code is computing correct solutions. Bison comparisons to post-irradiation examination data and PARFUME predictions were made for several sets of compacts. These comparisons considered silver, cesium, strontium, and krypton with and without failed particles. Bison predictions closely match those of PARFUME, while both codes generally overpredict post-irradiation examination data.

Optimum coating layer thicknesses of a TRISO having an 800-μm UO2 kernel under normal operation conditions of a 10-MWth block-type HTGR

The study evaluates the optimal thicknesses of the coating layers of an 800-μm UO2 tri-structural isotropic coated fuel particle (TRISO) for a 10-MWth block-type micro modular high temperature gas-cooled reactor (HTGR) using a response surface methodology (RSM) and an HTGR fuel performance analysis code, COPA. The objective function to be minimized in the optimal design of a TRISO consists of the product of the packing fraction of TRISO particles in a compact, the fractional releases of major metallic fission products from a TRISO and the failure probability of the silicon carbide (SiC) layers. A constraint on the sum of the coating layer thicknesses was made using a maximum packing fraction. Two fuel performance analyses such as a failure fraction analysis and a fission product release analysis are applied to the optimal designs of a TRISO that the RSM produced. An 800-μm UO2 TRISO having buffer, inner high-density pyrocarbon (IPyC), SiC, outer high-density pyrocarbon (OPyC), each with a thickness of 110, 40, 57, 35 μm, or 120, 35, 60, 27 μm is suitable for the operation of a 10-MWth block-type micro modular HTGR within 7600 days when both the SiC and through-coatings failure fractions do not occur and the fractional release of silver, cesium, strontium and krypton are below 1 × 10−9. The packing fraction of the optimal TRISOs in a compact is 40%.

UO2 Spheres Produce by External Gelation Process

UO2 kernels, a key component of fuel elements for high temperature gas cooled reactors, have usually been prepared by sol-gel methods. Sol-gel processes have a number of advantages, such as simple processes and facilities, and higher sphericity and density. In this study, to produce 900 μm-sized UO2 particles using an external gelation process, contact length extension of the NH3 gas of the broth droplets pass and the improvement of the gelation device capable of spraying 14 MNH4OH solution are used to form 3,000 μm-sized liquid droplets. To produce high-sphericity and high-density UO2 particles, HMTA, which promotes the gelation reaction in the uranium broth solution, is added to diffuse ammonium ions from the outside of the gelation solution during the aging process and generate ammonium ions from the inside of the ADU gel particles. Sufficient gelation inside of ADU gel particles is achieved, and the density of the UO2 spheres that undergo the subsequent treatment is 10.78 g/cm3; the sphericity is analyzed and found to be 0.948, indicating good experimental results.

THE CONCEPTUAL DESIGN FOR LIQUID EFFLUENT TREATMENT OF UO2 KERNEL FABRICATION

THE CONCEPTUAL DESIGN FOR LIQUID EFFLUENT TREATMENT OF UO2 KERNEL FABRICATION. The pebble fuel for HTGR is prepared through fabrication of UO2 kernel, coated particle and spherical element fuel. In the fabrication of UO2 kernel by external gelation method, a multicomponent of liquid effluent is generated. Therefore, the liquid effluent is required to be treated for safety reason before disposed to waste storage. In this paper, the conceptual design for the liquid effluent treatment of UO2 kernel fabrication is performed with the simulation process using CHEMCAD software. CHEMCAD is a software that can be utilized for chemical process design. The results showed that the proposed conceptual design is able to separate valuable components: isopropyl alcohol (IPA) and tetrahydrofurfuryl alcohol (THFA). The flowrate of IPA product is 5.28 kg/h with purity of 0.99 in mass fraction and the flowrate of THFA product is 1.01 kg/h with purity of 0.99 in mass fraction.Keywords: liquid effluent, UO2 kernel, CHEMCAD, HTGR.

3D Analysis of a multipurpose modular reactor for small energy production

The Fixed Bed Nuclear Reactor (FBNR) is a modular reactor for small energy needs (70-120 MWel). FBNR is pressurized water reactor, but uses GEN-IV reactors fuel. UO2 kernels of 500 μm diameter with 25 μm thick SiC cladding are embedded in graphite matrix form the spherical fuel elements with a diameter of 15 mm and covered by 300 μm thick Zircaloy-4 cladding. FBNR has high level safety features and operates without On-site Refueling. Approximately 1.62 million spherical fuel elements are suspended in a core height of 200 cm and inner and outer diameters of 32 cm and 224 cm by upwards flowing light water coolant with 280°C entry and 340°C exit temperatures. The unit lattice geometry is dodecahedron with fuel element in center and surrounded with light water coolant-moderator.In the present work, 9% enriched UO2 fuel is used. Neutronic analysis has been performed with the MCNPX-2.7.0 code. As, it is not practical to tackle millions of fuel elements separately, at first run unit fuel cell calculations are performed in spherical geometry for one single lattice consisting of separate fuel, cladding and volume-equivalent moderator regions. In the second run, the fuel cell is homogenized. Infinite cell calculations have been executed for variable moderator/fuel (H/235U) ratios in order to determine under- and over-moderated criticality values. The criticality values for the homogenized cell geometry are slightly lower and remains <5% range, that is, on the conservative side, which is acceptable for the purpose of this study. This allows us to homogenize the entire core for the full 3-D reactor with core and reflector regions. Reactor criticality increases with increasing H/235U ratios in the, under moderated region, where H/235U = 18 is selected for further investigations. Temporal variation of the reactor criticality is pursued for full reactor power of 210 MWth (~70 MWel net) with keff = 1.36 at the beginning of life down to keff = 1.02 over 1260 days with a plant factor of 1.0. The final fuel burn up was found as 57.3 GWd/ton.

Role of microstructure on CO corrosion of SiC layer in UO2-TRISO fuel

The Advanced Gas Reactor Fuel Qualification and Development (AGR) program has focused on qualification of UCO kernel tristructural-isotropic (TRISO) particle fuel relative to UO2 kernel TRISO particle fuel. However, a UO2 kernel variant was included in the second AGR irradiation experiment (AGR-2) for comparison and to connect to historic fuel irradiation data. The development of a multiscale, post-irradiation examination (PIE) analysis approach through the AGR Program has allowed for a comprehensive understanding of individual particle failure. This approach has been applied to gain an understanding of SiC layer failure in UO2 kernel TRISO fuel from AGR-2 after safety testing at 1600–1700 °C. Particle failure by intergranular CO corrosion, facilitated by a compromised inner pyrolytic carbon layer, has been confirmed through the combined application of modern x-ray computed tomography and electron microscopy techniques. Additionally, a relationship between grain boundary character and CO corrosion has been identified. This finding provides an opportunity to develop mitigating strategies to improve the resilience of the SiC layer to CO corrosion in UO2 TRISO fuel.

Characteristics of Uranium Dioxide (UO2) Kernel Produced by Sintering Process using Modified Sintering Reactor

Sintering process is the final stage of fuel kernel manufacturing prior to the coating process. This stage is very important part of the whole process, because it will determine the feasibility of UO2 kernel to comply with the specifications of HTR reactor fuel. The objective of this research was to obtain UO2 kernel with the density of ≥ 95% TD. The results showed that the highest density reached 92.56% TD or about 10.1441 g/cm3. This condition of sintering was gained at the temperature of 1400 °C with sintering time of 2 hours. The sintering product diameter gained was around 919 μm the specific surface area 4.4213 m2/g, and a total pore volume 4,751 x 10−3 cm3/g. The density of UO2 kernel produced from this research is the best compared to previous finding because of its density already approaches the HTR fuel specification requirements.

The Selection of Geometry and Flow Rate on The Fluidized Bed Reactor for Coating Particle

The “Reaktor Daya Eksperimental” has been designated to be built in Indonesia and the selected type of this reaktor is a high-temperature gas-cooled reactor. The fuel system used in this reactor is pebble bed fuel. The process of coating the sintered kernel into coated particles, is one of the processes in the manufacture of gas cooled high temperature reactor fuels. The geometry of the fluidization reactor determines the level of contact between the kernel and the filling gas. The flow rate of the fluidizing gas also affects fluidization at the rate at which stable fluidization occurs. Both of the above thing are an important role in the success of the coating process. Based on data and information from previous researcheres, designed a tool for coating process. The design of the sintered UO2 kernel coating reactor into coated particles is carried out based on the coating process data, the purpose of the reactor making and the selection of its capacity. The reactor prototype is used to obtain useful process data for the manufacture of a production reactor. The type of fluidization reactor made by the prototype is a conical spout bed reactor or conical reactor. The fluidization reactor prototype was chosen to have a 2 cm diameter cylinder, 0.3 cm inlet diameter, and 60° conical angle. The flow rate for the reactor operation is between the maximum fluidization flow rate at the inlet and the minimum fluidization velocity on the cylinder. The gas stream acording to the calculation above 2.5279 m/sec at the inlet pipe or 1.0716 L/min. up to the flow 1.38 m/s or 26.1255 L/men. Comparison of other research results with modeling using CFD, its flow rate is about 10 m/s or 4.239 L/min.

GAMMA DOSE RATE ANALYSIS IN BIOLOGICAL SHIELDING OF HTGR-10 MWth PEBBLE-BED REACTOR

GAMMA DOSE RATE ANALYSIS IN BIOLOGICAL SHIELDING OF HTGR-10 MWth PEBBLE BED REACTOR. HTGR-10 MWth is a high-temperature gas-cooled reactor. The fuel and moderator are pebble shaped with a radius of 3 cm. One fuel pebble consists of thousands of UO2 kernels with a density of 10.4 gram/cc and the enrichment rate of 17%. The core of HTGR-10 MWth is the center of origin of neutrons and gamma radiation resulting from the interaction of neutrons with pebble fuel, moderator and biological shield. The various types of radiations generated from such nuclear reactions should be monitored to ensure the safety of radiation workers. This research was conducted using MCNP-6 Program package with the aim to calculate and analyze gamma radiation dose in biological shield of HTGR-10 MWth. In this study, the biological shield is divided into 10 equal segments. The first step of the research is to benchmark the created program against the critical height of HTR-10. The results of the benchmarking show an error rate of ± 1.1327%, while the critical core height of HTGR 10 MWth for the ratio of pebble fuel and pebble moderator (F:M) of 52: 48 occurs at a height of 134 cm. The rate of gamma dose at the core is 3.0052E + 05 mSv/hr. On the biological shield made of regular concrete with a density of 2.3 grams/cc, the rate of gamma dose decreases according to an equation y = 0.0042 e-0.03x. Referring to Perka Bapeten no 4 of 2013, the safe limits for workers and radiation protection officers will be achieved if the minimum thickness of biological shield is 115 cm with gamma dose rate of 0 mSv/hour.Keywords: Gamma dose rate, HTGR 10 MWth, biological shield, pebble

Neutron dose rate analysis on HTGR-10 reactor using Monte Carlo code

The HTGR-10 reactor is cylinder-shaped core fuelled with kernel TRISO coated fuel particles in the spherical pebble with helium cooling system. The outlet helium gas coolant temperature outputted from the reactor core is designed to 700 °C. One advantage HTGR type reactor is capable of co-generation, as an addition to generating electricity, the reactor was designed to produce heat at high temperature can be used for other processes. The spherical fuel pebble contains 8335 TRISO UO2 kernel coated particles with enrichment of 10% and 17% are dispersed in a graphite matrix. The main purpose of this study was to analysis the distribution of neutron dose rates generated from HTGR-10 reactors. The calculation and analysis result of neutron dose rate in the HTGR-10 reactor core was performed using Monte Carlo MCNP5v1.6 code. The problems of double heterogeneity in kernel fuel coated particles TRISO and spherical fuel pebble in the HTGR-10 core are modelled well with MCNP5v1.6 code. The neutron flux to dose conversion factors taken from the International Commission on Radiological Protection (ICRP-74) was used to determine the dose rate that passes through the active core, reflectors, core barrel, reactor pressure vessel (RPV) and a biological shield. The calculated results of neutron dose rate with MCNP5v1.6 code using a conversion factor of ICRP-74 (2009) for radiation workers in the radial direction on the outside of the RPV (radial position = 220 cm from the center of the patio HTGR-10) provides the respective value of 9.22E-4 μSv/h and 9.58E-4 μSv/h for enrichment 10% and 17%, respectively. The calculated values of neutron dose rates are compliant with BAPETEN Chairman’s Regulation Number 4 Year 2013 on Radiation Protection and Safety in Nuclear Energy Utilization which sets the limit value for the average effective dose for radiation workers 20 mSv/year or 10μSv/h. Thus the protection and safety for radiation workers to be safe from the radiation source has been fulfilled. From the result analysis, it can be concluded that the model of calculation result of neutron dose rate for HTGR-10 core has met the required radiation safety standards.

The uranium recovery from UO2 kernel production effluent

For the fabrication of coated particle fuel elements of high temperature gas cooled reactors, the ceramic UO2 kernels are prepared through chemical gelation of uranyl nitrate solution droplets, which produces radioactive effluent with components of ammonia, uranium, organic compounds and ammonium nitrate. In this study, a flow sheet including evaporation, flocculation, filtration, adsorption, and reverse osmosis was established for the effluent treating. The uranium recovery could reach 99.9% after the treatment, with almost no secondary pollution produced.

Preparation of UCO Microspheres by Internal Gelation Process

UCO kernels were used for High Temperature Reactor (HTR) fuel, which can restrain kernels migration and avoid corrosion of cladding layer under the high burnup compared with UO2 or UC kernels. In this paper, UCO kernels were prepared by internal gelation process containing dissolving, broth preparation, dispersing, AWD (aging, washing and drying) and sintering procedures. The broth composition was studied based on the fabrication of UO2 kernels before. Effects of the carbon black content and sintering atmosphere on the microstructure and performance of microspheres were mainly studied. Surface morphology of dried microsphere was observed by integrated microscope. Grain size and the degree of densification of UCO microspheres were observed by SEM.

The measurement of silver diffusivity in zirconium carbide to study the release behavior of 110mAg in the ZrC TRISO-coated nuclear fuel particle

The tri-structural isotropic (TRISO) coated particle fuel has been developed and used for high temperature gas-cooled reactors (HTGRs). It provides a unique robustness of the first barrier for the fission products. The TRISO fuel particle has typically consisted of a UO2 or UCO kernel, surrounded by successive layers of porous carbon, dense inner pyrocarbon, silicon carbide, and dense outer pyrocarbon. During operation, however, the SiC layer has been known to release radioactive silver 110mAg which makes maintenance more difficult and thus costly. Zirconium carbide has been considered as a promising alternative to the SiC fission product barrier. ZrC exhibits high temperature stability and possibly possesses superior Pd resistance, while the retention properties especially for silver have not been adequately studied.To help elucidate the diffusive behavior of silver in the ZrC coating of the TRISO-coated particle, a new diffusion experimental technique, called the encapsulating source method, has been developed by constructing a constant source diffusion couple between ZrC and Ag gas originated from Zr–Ag solid solution. Scanning electron microscopy (SEM), wavelength-dispersive X-ray spectroscopy (WDS), electron backscatter diffraction (EBSD) and optical methods were used to analyze the diffusion couple annealed at 1500 °C. The resultant diffusion coefficient of Ag in single-crystalline ZrC0.84 at 1500 °C was experimentally determined to be about 2.8 (±1.2) × 10−17 m2/s.

Separation of ammonia from radioactive wastewater by hydrophobic membrane contactor

The radioactive wastewater produced in the manufacturing process of UO2 kernel for high temperature gas-cooled reactor (HTR) contains high concentration of ammonia. The ammonia has to be removed effectively for further treatment of the wastewater. In this study, the hydrophobic membrane contactor (HMC) was adopted to remove and recover the ammonia from radioactive wastewater at room temperature. The operating parameters such as feed velocity and initial ammonia concentration were determined. In addition, the effect of wastewater composition on ammonia separation was studied. The experiment results showed that ammonia removal efficiency could reach above 90% after 120 min operation when pH was not adjusted. While the initial pH of wastewater was adjusted to 12.0, ammonia removal efficiency could reach above 95%. The ammonia mass transfer coefficient increased with increase of feed velocity and tended to an asymptotic value when the feed velocity reached 0.049 m/s. When the initial ammonia concentration was 2211.6 mg/L, 5864.6 mg/L and 23,898.7 mg/L, the ammonia removal efficiency was 95.0%, 94.2% and 94.1%, respectively after 120 min operation, i.e. the initial concentration of ammonia in wastewater had almost no effect on ammonia separation. In addition, the coexisting substances, such as urea and tetrahydrofurfuryl alcohol (THFA), also had no effect on ammonia separation. The HMC is a promising way to separate ammonia from radioactive wastewater.

TRISO 입자를 포함하는 SiC 복합소결체의 소결 및 특성 평가

Fully ceramic micro encapsulated (FCM) nuclear fuel is a concept recently proposed for enhancing the stability of nuclear fuel. FCM nuclear fuel consists of tristructural-isotropic (TRISO) fuel particles within a SiC matrix. Each TRISO fuel particle is composed of a UO₂ kernel and a PyC/SiC/PyC tri-layer which protects the kernel. The SiC ceramic matrix is created by sintering. In this FCM fuel concept, fission products are protected twice, by the TRISO coating layer and by the SiC ceramic. The SiC ceramic has proven attractive for fuel applications owing to its low neutron-absorption cross-section, excellent irradiation resistivity, and high thermal conductivity. In this study, a SiC-matrix composite containing TRISO particles was sintered by hot pressing with Al₂O₃-Y₂O₂ additive system. Various sintering conditions were investigated to obtain a relative density greater than 95%. The internal distribution of TRISO particles within the SiC-matrix composite was observed using an x-ray radiograph. The fracture of the TRISO particles was investigated by means of analysis of the cross-section of the SiC-matrix composite.