• Journal of the Korean institute of surface engineering
• /
• v.52 no.6
• /
• pp.321-325
• /
• 2019

Triso-type coating layers of silicon carbide and graphite on UO₂ paticulate nuclear fuel were prepared by using fluidized bed type chemical vapor deposition and self-propagating high temperature synthesis methods to make a coated nuclear fuel of a power plant for hydrogen mass-production. The source and carrier gases were the mixture of methyltrichlorosilane and propane, and inert argon. Chemical analysis and microstructure observation showed that the coated layers were inner graphite, middle silicon carbide and outer graphite. The elastic modulus and nano-hardness of the silicon carbide layer were 503 [GPa] and 36 [GPa], respectively.

• Nuclear Engineering and Technology
• /
• v.43 no.6
• /
• pp.477-488
• /
• 2011

A formulation is proposed for modelling the process of intra-granular diffusion of fission gas during irradiation of $UO_2$ under both normal operating conditions and power transients. The concept represents a simple extension of the formulation of Speight, including an estimation of the contribution of bubble motion to fission gas diffusion. The resulting equation is formally identical to the diffusion equation adopted in most models that are based on the formulation of Speight, therefore retaining the advantages in terms of simplicity of the mathematical-numerical treatment and allowing application in integral fuel performance codes. The development of the new model proposed here relies on results obtained by means of molecular dynamics simulations as well as finite element computations. The formulation is proposed for incorporation in the TRANSURANUS fuel performance code.

• Nuclear Engineering and Technology
• /
• v.37 no.4
• /
• pp.309-316
• /
• 2005

The contents of a new book currently in preparation are described. The dearth of books in the field of nuclear materials has left both students in nuclear materials classes and professionals in the same field without a resource for the broad fundamentals of this important sub-discipline of nuclear engineering. The new book is devoted entirely to materials problems in the core of light-water reactors, from the pressure vessel into the fuel. Key topics deal with the $UO_2$ fuel, Zircaloy cladding, stainless steel, and of course, water. The restriction to LWR materials does not mean a short monograph; the enormous quantity of experimental and theoretical work over the past 50 years on these materials presents a challenge of culling the most important features and explaining them in the simplest quantitative fashion. Moreover, LWRs will probably be the sole instrument of the return of nuclear energy in electric power production for the next decade or so. By that time, a new book will be needed.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.05a
• /
• pp.40-45
• /
• 1996

A new concept of equivalent homogenization is proposed. The concept employs new set of homogenized parameters: homogenized cross sections (XS) and interface matrix (IM), which relates partial currents at the celt interfaces. The idea of interface matrix generalizes the idea of discontinuity factors (DFs), proposed and developed by K.Koebke and K.Smith. The method of K.Smith can be simulated within framework of new method, while the new method approximates heterogeneous cell better in case of the steep flux gradients at the cell interfaces. The attractive shapes of new concept are: improved accuracy, simplicity of incorporation in the existing codes, equal numerical expenses in comparison to the K.Smith's approach. The new concept is useful for: (a) explicit reflector/baffle simulation; (b) control blades simulation; (c) mixed UO2/MOX core simulation, The offered model has been incorporated in the finite difference code and in the nodal code PANBOX. The numerical results show good accuracy of core calculations and insensitivity of homogenized parameters with respect to in- core conditions.

• Nuclear Engineering and Technology
• /
• v.53 no.11
• /
• pp.3741-3758
• /
• 2021

Outcomes of the FALCON code analysis-related part of the STARS-ENSI Service Project on Evaluation of the new U.S.NRC RIA Fuel Safety Criteria and Application to the Swiss Reactors are presented. Substantial conservatism of the updated safety limits for high-temperature and PCMI cladding failure, as proposed in the NRC Regulatory Guide RG 1.236, is confirmed. Applicability of the updated failure limits to fuel safety analysis in the Swiss PWRs, as applied to standard fuel designs using UO2 fuel pellets and SRA Zry-4 as cladding materials is discussed. Conducting of new integral RIA tests with irradiated samples using doped- and gadolinia fuel pellets to support appropriate fuel safety criteria for RIA events is recommended.

• Nuclear Engineering and Technology
• /
• v.52 no.1
• /
• pp.109-114
• /
• 2020

This work presents the results of laboratory scale tests of the CARBOFLUOREX (CARBOnate FLUORide EXtraction) process - a novel technology for the recovery of U and Pu from the solid fluorides residue (fluorination ash) of Fluoride Volatility Method (FVM) reprocessing of spent nuclear fuel (SNF). To study the oxidative leaching of U from the fluorination ash (FA) by Na₂CO₃ or Na₂CO₃-H₂O₂ solutions followed by solvent extraction by methyltrioctylammonium carbonate in toluene and purification of U from the fission products (FPs) impurities we used a surrogate of FA consisting of UF₄ or UO₂F₂, and FPs fluorides with stable isotopes of Ce, Zr, Sr, Ba, Cs, Fe, Cr, Ni, La, Nd, Pr, Sm. Purification factors of U from impurities at the solvent extraction refining stage reached the values of 10⁴-10⁵, and up to 10⁶ upon the completion of the processing cycle. Obtained results showed a high efficiency of the CARBOFLUOREX process for recovery and separating of U from FPs contained in FA, which allows completing of the FVM cycle with recovery of U and Pu from hardly processed FA.

• Nuclear Engineering and Technology
• /
• v.53 no.6
• /
• pp.1931-1938
• /
• 2021

Uranium ingot is produced by metallothermic reduction of uranium tetrafluoride using magnesium or calcium as reductant. Presence of oxygen containing compounds viz. uranyl fluoride and uranium oxide in the starting uranium fluoride has a significant effect on the firing time, final temperature of the charge, slag-metal separation and hence the metal recovery. As reported in the literature, the maximum tolerable limit for uranyl fluoride in the UF₄ is 2.5 wt% and limit for uranium oxide content is in the range 2-3 wt%. No theoretical or experimental basis is available till date for these limits. Analyses have been carried out in this study to understand the effect of UO₂F₂ concentration in the starting fluoride on the final temperature of the products and thus the reduction characteristics. UF₄ having uranyl fluoride concentration, less than as well as more than 2.5 wt%, have been investigated. Thermodynamic calculations have been carried out to arrive at a general expression for the final temperature attained by the products during calciothermic reduction of UF₄. Finally, an upper limit for the oxygen containing impurities has been estimated using the CaO-CaF₂ phase diagram.

• Nuclear Engineering and Technology
• /
• v.56 no.3
• /
• pp.1022-1028
• /
• 2024

A neutronics study of a supercritical CO₂-cooled fast reactor core for nuclear propulsion has been performed in this work. The thermal power of the reactor core is 30 MWth and a ceramic UO₂ fuel can be used to achieve a 20-year lifetime without refueling. In order to make a compact core with inherent safety features, the drum-type reactivity control system and folding-type shutdown system are adopted. In addition, we suggest a cold shutdown system using gadolinium as a spectral shift absorber (SSA) against flooding. Although there is a penalty of U-235 enrichment for the core embedded with the cold shutdown system, it effectively mitigates the increment of reactivity at the flooding of seawater. In this study, the neutronics analyses have been performed by using the continuous energy Monte Carlo Serpent 2 code with the evaluated nuclear data file ENDF/B-VII.1 Library. The supercritical CO₂-cooled fast reactor core is characterized in view of important safety parameters such as the reactivity worth of reactivity control systems, fuel temperature coefficient (FTC), coolant temperature coefficient (CTC), and coolant temperature-density coefficient (CTDC). We can say that the suggested core has inherent safety features and enough flexibility for load-following operation.

• Nuclear Engineering and Technology
• /
• v.41 no.2
• /
• pp.187-198
• /
• 2009

In this paper, three power ramp tests performed on high burn-up Re-crystallized Zircaloy2 - UO2 BWR fuel rods (56 to 63 MWd/kgU) within the SCIP project are simulated with METEOR and ALCYONE 3D. Two of the ramp tests are of staircase type up to Linear Heat Rates of 420 and 520 W/cm and with long holding periods. Failure of the 420 W/cm fuel rod was observed after 40 minutes. The third ramp test consisted of a more standard ramp test with a constant power rate of 80 W/cm/min up to 410 W/cm with a short holding time. The tests were first simulated with the METEOR 1D fuel rod code, which gave accurate results in terms of profilometry and fission gas releases. The behaviour of a fuel pellet fragment and of the cladding piece on top of it was then investigated with ALCYONE 3D. The size and the main characteristics of the ridges after base irradiation and power ramp testing were recovered. Finally, the failure criteria validated for PWR conditions and fuel rods with low-to-medium burn-ups were used to analyze the failure probability of the KKL rodlets during ramp testing.

• Journal of the Korean Chemical Society
• /
• v.58 no.2
• /
• pp.169-178
• /
• 2014

A new series of metal complexes of V(IV), Pd(II), Pt(IV), Ce(IV) and U(VI) with 3-[(3-chlorophenyl)-hydrazono]-pentane-2,4-dione (Cphpd) were synthesized and characterized by elemental analysis, molar conductivity, magnetic moment measurements, UV-vis, FT-IR and $^1H$ NMR as well as TG-DTG techniques. The data indicated that the Cphpd acts as a bidentate ligand through the hydrazono nitrogen and one keto oxygen. The kinetic parameters have been evaluated by using Coats Redfern (CR) and Horowitz-Metzeger (HM) methods. The thermodynamic data reflected the thermal stability for all complexes. The calculated bond length and the bond stretching force constant, F(U=O), values for $UO_2$ bond are $0.775{\AA}$ and $286.95Nm^{-1}$. The bond lengths, bond angles, dipole moment and the lowest energy model structure of the complexes have been determined with DFT calculations. The antimicrobial activity of the synthesized ligand and its complexes were screened.