• Nuclear Engineering and Technology
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• v.47 no.1
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• pp.85-93
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• 2015

Background: $^{241}Am$, $^{243}Am$, and $^{237}Np$ isotopes are among the most radiotoxic components of spent nuclear fuel. Recently, researchers have planned different incineration scenarios for the highly radiotoxic elements of nuclear waste in critical reactors. Computational methods are widely used to predict burnup rates of such nuclear wastes that are used under fuel matrixes in critical reactors. Methods: In this work, the Monte Carlo N-particle transport code was used to calculate the neutronic behavior of a transuranic (TRU)-bearing CANada Deuterium Uranium 6 reactor. Results: The computational data showed that the 1.0% TRU-containing thorium-based fuel matrix presents higher proliferation resistance and TRU depletion rate than the other investigated fuel Matrixes. The fuel matrix includes higher negative temperature reactivity coefficients as well. Conclusion: The investigated thorium-based fuel matrix can be successfully used to decrease the production of highly radiotoxic isotopes.

• Journal of the Korean Ceramic Society
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• v.55 no.4
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• pp.400-405
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• 2018

FCM nuclear fuel, a concept proposed as an accident tolerant fuel in light water reactors, consists of TRISO fuel particles embedded in a SiC matrix. The uniform dispersion of internal TRISO fuel particles in the FCM fuel is very important for improving the fuel efficiency. In this study, FCM sintered pellets with various volume ratios of TRISO-coated particles were prepared by hot press sintering. The distribution of TRISO-coated particles was quantitatively analyzed using X-ray ${\mu}CT$ and expressed as a dispersion uniformity index. TRISO-coated particles were most uniformly dispersed in the FCM pellets prepared using only overcoated TRISO particles without mixing of additional SiC matrix powder. FCM pellets with uniformly dispersed TRISO particle volume fraction of up to 50% were prepared using overcoated TRISO particles with varying thickness.

• Nuclear Engineering and Technology
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• v.29 no.2
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• pp.99-109
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• 1997

A dynamic model for hydrogen generation by Fuel-Coolant Interactions(FCI) is developed with separate models for each FCI stage, coarse mixing and stratification. The model includes the physical concept of FCI, semi-empirical heat and mass transfer correlation and the concentration diffusion equation with the general non-zero boundary condition. The calculated amount of hydrogen, which is mainly generated in stratification, is compared with the FITS experiments. The model developed in this study shows a good agreement within a range of 10 % fuel oxidation rate and predicts the controlled mechanism of the chemical reaction very well. And this model predicts more accurately than the previous works. It is shown from the sensitivity study that the higher initial temperature of fuel particle is, the larger the reaction rate is. Up to 2700 K of temperature of the particle, the reaction rate increases rapid, which can lead to metal ignition.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2616-2628
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• 2021

The results of effective irradiation swelling in a wide range of burnup levels are numerically obtained for an inert matrix fuel, which are verified with DART model. The fission gas swelling of fuel particles is calculated with a mechanistic model, which depends on the external hydrostatic pressure. Additionally, irradiation and thermal creep effects are included in the inert matrix. The effects of matrix creep strains, external hydrostatic pressure and temperature on the effective irradiation swelling are investigated. The research results indicate that (1) the above effects are coupled with each other; (2) the matrix creep effects at high temperatures should be involved; and (3) ranged from 0 to 300 MPa, a remarkable dependence of external hydrostatic pressure can be found. Furthermore, an explicit multi-variable mathematic model is established for the effective irradiation swelling, as a function of particle volume fraction, temperature, external hydrostatic pressure and fuel particle fission density, which can well reproduce the finite element results. The mathematic model for the current volume fraction of fuel particles can help establish other effective performance models.

• Korean Journal of Materials Research
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• v.29 no.8
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• pp.483-490
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• 2019

Currently, the Korean nuclear industry uses ZIRLO as material for nuclear fuel cladding(zirconium alloy). KEPCO Nuclear Fuel is in the process of developing a HANA alloy to enable domestic production of cladding. Cladding manufacture involves multistage heat treatments and pickling processes, the latter of which is vital for the removal of defects and impurities on the cladding surface. SMUT that forms on the cladding surface during such pickling process is a source of surface defects during heat treatment and post-treatment processes if not removed. This study analyzes ZIRLO, HANA-4, and HANA-6 alloy claddings to extensively study the SEM/EDS, XRD, and particle size characteristics of SMUT, which are second phase particles that are formed on the cladding surface during pickling processes. Using the analysis results, this study observes SMUT formation characteristics according to Nb concentration in Zr alloys during the washing process following the pickling process. In addition, this study observes SMUT removal characteristics on cladding surfaces according to concentrations of nitric acid and hydrofluoric acid in the acid solution.

• Nuclear Engineering and Technology
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• v.28 no.4
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• pp.366-372
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• 1996

The simulated spent PWR fuel pellet which is corresponding to the turnup of 33,000 MWD/MTU is prepared by adding 11 fission-product elements to UO$_2$. The simulated spent fuel pellet is treated at 40$0^{\circ}C$ in air (oxidation), at 110$0^{\circ}C$ in air (high-temperature treatment), and at $600^{\circ}C$ in hydrogen (reduction). The product is treated through additional addition and reduction up to 3 cycles. Pellets are completely pulverized by the first oxidation, and the high-temperature treatment causes particle and crystallite to grow and surface to be smooth, and thus particle size significantly increases and surface area decreases. The reduction following the high-temperature treatment decreases much the particle size by means of the formation of intercrystalline cracks. The particle size decreases a little during the second oxidation and reduction cycle and then remains nearly constant during the third and fourth cycles. Surface area of pounder increases progressively with the repetition of oxidation and reduction cycles, mainly due to the formation of Surface cracks. The degradation of surface area resulting from high-temperature treatment is restored by too subsequent resulting oxidation and reduction cycles.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.947-948
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.441-442
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.306-307
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.249-251
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• 2005