• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.727-732
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• 2016

A simple and rapid spectrophotometric method has been developed to enable the determination of plutonium concentration in an irradiated fuel solution in the presence of all fission products. An excess of ceric ammonium nitrate solution was employed to oxidize all the valence states of plutonium to +6 oxidation state. Interference due to the presence of fission products such as ruthenium and zirconium, and corrosion products such as iron in the envisaged concentration range, as in the irradiated fuel solution, was studied in the determination of plutonium concentration by the direct spectrophotometric method. The stability of plutonium in +6 oxidation state was monitored under experimental conditions as a function of time. Results obtained are reproducible, and this method is applicable to radioactive samples resulting before the solvent extraction process during the reprocessing of fast reactor spent fuel. An analysis of the concentration of plutonium shows a relative standard deviation of <1.2% in standard as well as in simulated conditions. This reflects the fast reactor fuel composition with respect to uranium, plutonium, fission products such as ruthenium and zirconium, and corrosion products such as iron.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.1047-1048
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• 2004

Simulation Coated layers of a nuclear fuel particle were evaluated by field emission scanning electron microscopy and nano-indentation method to give basic data to estimate 'Amoeba effect' and give an optimum fabrication condition and high quality control. Coated layers on the fuel kernel are in the order of buffer pyrolytic carbon, inner pyrolytic carbon, silicon carbide and outer pyrolytic carbon layers, which average thicknesses are 95, 25, 30 and 28 ${\mu}m$, respectively. Their densities and hardnesses are 1.08, 1.15, 3.18, 1.82 $g/cm^3$ and 0.522, 0.874, 9.641, and 2.726 GPa, respectively. Comparing theoretical density of pyrolytic carbon of 2.22 $g/cm^3$, the relative amount of porosity in each layer is about 52% for buffer, 48% for inner PyC and 18% for outer PyC.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2792-2800
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• 2022

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 ℃. 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.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.167-179
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• 2024

This study explores the feasibility of employing (U, Th)-based accident tolerant fuels (ATFs), specifically (_0.8UO₂, _0.2ThO₂), (_0.8UN, _0.2ThN), and (_0.8UC, _0.2ThC). The investigation assesses the overall performance of these proposed fuel materials in comparison to the conventional UO₂, focusing on deep neutronic and thermal-hydraulic (Th) analyses. Neutronic analysis utilized the MCNPX code, while COMSOL Multiphysics was employed for thermal-hydraulic analysis. The primary objective of this research is to overcome the limitations associated with traditional UO₂ fuel by exploring alternative fuel materials that offer advantages in terms of abundance and potential improvements in performance and safety. Given the limited abundance of UO₂, long-term sustainable nuclear energy production faces challenges. From a neutronic standpoint, the U-Th based fuels demonstrated remarkable fuel cycle lengths, except (_0.8UN, _0.2ThN), which exhibited the minimum fuel cycle length and, consequently, the lowest fuel burn-up. Regarding thermal-hydraulic performance, (_0.8UN, _0.2ThN) exhibited outstanding performance with significant margins against fuel melting compared to the other materials. Overall, when considering the integrated performance, the most favourable results were obtained with the use of the (_0.8UC, _0.2ThC) fuel configurations. This study contributes valuable insights into the potential benefits of (U, Th)-based ATFs as a promising avenue for enhanced nuclear fuel performance.

• Journal of Powder Materials
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• v.7 no.2
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• pp.63-70
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• 2000

Thermal spray coating process has proven to be effective at producing hard, dense, wear resistance coatings on the relatively mild substrates. Among several spraying techniques, HVOF (High Velocity Oxygen Fuel) and plasma coating processes, which are preferentially used for the wear resistance application such as capstans, have been applied in this study. The effects of pre-treatment, it-process and post-treatment parameters on the wear and mechanical properties of WC+12%Co, Cr3C2 and Al2O3 powder coatings have been investigated and correlated with the microstructures. The results indicated that the carbide coating was more preferable to the oxide coatings and the post-treatments consisting of vacuum annealing and sealing on carbide coatings led to significant improvements in wear resistance, adhesive strength and coating phase stabilization over the other processing techniques in this application.

• Korean Journal of Metals and Materials
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• v.48 no.5
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• pp.410-416
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• 2010

Studies were carried out to establish the technology for sodium-clad compatibility and to analyze the compatibility behavior of the Sodium-cooled Fast Reactor (SFR) cladding material under a flowing sodium environment. The natural circulation facility caused by the thermal convection of the liquid sodium was constructed and the 316-series stainless steels were exposed at $650{^{\circ}C}$ liquid sodium for 1458 hours. The weight change and related microstructural change were analyzed. The results showed that the quasi-dynamic facility represented by the natural convection exhibited similar results compared to the conventional dynamic facility. Selective leaching and local depletion of the chromium, re-distribution of the carbide, and the decarburization process took place in the 316-series stainless steel under a flowing sodium environment. This process decreased as the sodium flowed along the channel, which was caused by the change in the dissolved oxygen and carbon activity in the liquid sodium.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3095-3105
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• 2022

Solid Microencapsulated Fuel (SMF) is a type of solid fuel rod design that disperses TRISO coated fuel particles directly into a kind of matrix. SMF is expected to provide improved performance because of the elimination of cladding tube and associated failure mechanisms. This study focused on the neutronics and some of the fuel cycle characteristics of SMF by using OpenMC. Two kinds of SMFs have been designed and evaluated - fuel particles dispersed into a silicon carbide matrix and fuel particles dispersed into a zirconium matrix. A 7×7 fuel assembly with increased rod diameter transformed from the standard NHR200-II 9×9 array was also introduced to increase the heavy metal inventory. A preliminary study of two kinds of burnable poisons (Erbia & Gadolinia) in two forms (BISO and QUADRISO particles) was also included. This study found that SMF requires about 12% enriched UN TRISO particles to match the cycle length of standard fuel when loaded in NHR200-II, which is about 7% for SMF with increased rod diameter. Feedback coefficients are less negative through the life of SMF than the reference. And it is estimated that the average center temperature of fuel kernel at fuel rod centerline is about 60 K below that of reference in this paper.

• Journal of Power Electronics
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• v.11 no.4
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• pp.606-611
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• 2011

Power electronics is a key technology for electric, hybrid, plug-in hybrid, and fuel cell vehicles. Typical power electronics converters used in electric drive vehicles include dc/dc converters, inverters, and battery chargers. New semiconductor materials such as silicon carbide (SiC) and novel topologies such as the Z-source inverter (ZSI) have a great deal of potential to improve the overall performance of these vehicles. In this paper, a Z-source inverter for fuel cell vehicle application is examined under three different scenarios. 1. a ZSI with Si IGBT modules, 2. a ZSI with hybrid modules, Si IGBTs/SiC Schottky diodes, and 3. a ZSI with SiC MOSFETs/SiC Schottky diodes. Then, a comparison of the three scenarios is conducted. Conduction loss, switching loss, reverse recovery loss, and efficiency are considered for comparison. A conclusion is drawn that the SiC devices can improve the inverter and inverter-motor efficiency, and reduce the system size and cost due to the low loss properties of SiC devices. A comparison between a ZSI and traditional PWM inverters with SiC devices is also presented in this paper. Based on this comparison, the Z-source inverter produces the highest efficiency.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.1009-1014
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• 2016

SiC-based ceramic composites are currently being considered for use in fuel cladding tubes in light-water reactors. The joining of SiC ceramics in a hermetic seal is required for the development of ceramic-based fuel cladding tubes. In this study, SiC monoliths were diffusion bonded using a Ti foil interlayer and additional Si powder. In the joining process, a very low uniaxial pressure of ~0.1 MPa was applied, so the process is applicable for joining thin-walled long tubes. The joining strength depended strongly on the type of SiC material. Reaction-bonded SiC (RB-SiC) showed a higher joining strength than sintered SiC because the diffusion reaction of Si was promoted in the former. The joining strength of sintered SiC was increased by the addition of Si at the Ti interlayer to play the role of the free Si in RB-SiC. The maximum joint strength obtained under torsional stress was ~100 MPa. The joint interface consisted of $TiSi_2$, $Ti_3SiC_2$, and SiC phases formed by a diffusion reaction of Ti and Si.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.769-775
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• 2019

The tensile strength of irradiated 3C-SiC, SiC with artificial point defects, SiC with symmetric tilt grain boundaries (GBs), irradiated SiC with GBs are investigated using molecular dynamics simulations at 300 K. For an irradiated SiC sample, the tensile strength decreases with the increase of irradiation dose. The Young's modulus decreases with the increase of irradiation dose which agrees well with experiment and simulation data. For artificial point defects, the designed point defects dramatically decrease the tensile strength of SiC at low concentration. Among the point defects studied in this work, the vacancies drop the strength the most seriously. SiC symmetric tilt GBs decrease the tensile strength of pure SiC. Under irradiated condition, the tensile strengths of all SiC samples with grain boundaries decrease and converge to certain value because the structures become amorphous and the grain boundaries disappear after high dose irradiation.