• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.1-10
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• 2022

This study discusses the effect of using ²³²Th instead of ²³⁸U on the neutronic characteristics and the main operating parameters of the pressurized water reactor (PWR). MCNPX version 2.7 was used to compare the neutronic characteristics of UO₂ with (Th, ²³⁵U)O₂ and (Th, ²³³U) O₂. Firstly, the infinity multiplication factor (K_inf), thermal neutron flux, and power distribution have been studied for the investigated fuel types. Secondly, the effect of Gd₂O₃ and Er₂O₃ on the K_inf and on the radial thermal neutron flux and thermal power has been investigated to distinguish which of them is more suitable than the other in reactivity management. Thirdly, to illustrate the effectiveness of ²³²Th in decreasing the inventory of both the actinides and non-actinides, the concentration of plutonium (Pu) isotopes and minor actinides (MAs) has been simulated with the fuel burnup. Besides, due to their large thermal neutron absorption cross-section, the concentrations of ¹³⁵Xe, ¹⁴⁹Sm, and ¹⁵¹Sm with the fuel burnup have been investigated. Finally, the main safety parameters such as the reactivity worth of the control rods (ρ_CR), the effective delayed neutron fraction β_eff, and the Doppler reactivity coefficient (DRC) were calculated to determine to which extent these fuel types achieve the acceptable limits.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.1 no.1
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• pp.11-23
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• 2003

A study on the electrolytic dissolution of SUS-304 and Inconel-600 specimen was carried out in neutral salt electrolyte to evaluate the applicability of electrochemical decontamination process for recycle or self disposal with authorization of large amount of metallic wastes contaminated with uranium compounds generated by dismantling a retired uranium conversion plant in Korea. Although the best electrolytic dissolution performance for the specimens was observed in a Na2s04 electrolyte, a NaNO$_3$ neutral salt electrolyte, in which about 30% for SUS-304 and the same for Inconel-600 in the weight loss was shown in comparison with that in a Na$_2$SO$_4$ solution, was selected as an electrolyte for the electrochemical decontamination of metallic wastes with the consideration on the surface of system components contacted with nitric acid and the compatibility with lagoon wastes generated during the facility operation. The effects of current density, electrolytic dissolution time, and concentration of NaNO$_3$ on the electrolytic dissolution of the specimens were investigated. On the basis of the results obtained through the basic inactive experiments, electrochemical decontamination tests using the specimens contaminated with uranium compounds such as UO$_2$, AUC (ammonium uranyl carbonate) and ADU (ammonium diuranate) taken from an uranium conversion facility were performed in 1M NaNO$_3$ solution with the current density or In mA/$\textrm{cm}^2$. it was verified that the electrochemical decontamination of the metallic wastes contaminated uranium compounds was quite successful in a NaNO$_3$ neutral salt electrolyte by reducing $\alpha$ and $\beta$ radioactivities below the level of self disposal within 10 minutes regardless of the type of contaminants and the degree of contamination.

• The Journal of the Korean dental association
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• v.49 no.4
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• pp.203-210
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• 2011

The aims of this study were to evaluate the adequate thickness of opaque resins for situations such as an oral black cavity and discolored tooth structure, as well as the translucency of each opaque material at various thicknesses. Six opaque-shade composite resins (Z-350 OA3, Amelogen Universal A2O, Esthet-X A2O, Esthet-X A4O, Charmfil UO and Aelite Universal OA3) were prepared in metal molds with a hole of 8 mm in diameter and various thicknesses (0.5mm, 1.0mm, 1.5mm, 2.0mm, 2.5mm, 3.0 mm and 4.0mm). Four backgrounds (white tile, black tile, C4 shade porcelain and opaque resin itself) were used to determine the translucency parameter (between black and white backgrounds). and to mimic a black oral cavity (between black and opaque resin backgrounds) and a discolored tooth structure (between C4 and opaque resin backgrounds). Color measurements were made by a colorimeter to determine the CIELAB values of each specimen with each background and to calculate the translucency parameter and ${\Delta}E^*$ value difference among the specimens on the backgrounds. The translucency parameter and ${\Delta}E^*$ obtained between black and opaque resin backgrounds decreased in similar pattern as thickness increased. A C4 background was masked by resin thicknesses of 0.5-1.0mm, while a black background required thicknesses of 1.0-2.0mm. Adequate knowledge about differences in the optical character like translucency of the materials used is essential, together with the accumulated experience of the individual clinician.

• Korean Journal of Pharmacognosy
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• v.29 no.3
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• pp.243-247
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• 1998

A diverse panel of human tumor cell lines and a mouse melanoma cell line (B16-F1) were used for the cytotoxicity test of the nine sesquiterpene lactones with ${\beta}-methylene-{\gamma}-lactone$ group isolated from Hemisteptia lyrata, Chrysanthemum zawadskii and Chrysanthemum boreale. In the cell adhesion inhibitory activity test against B16-F1 mouse melanoma cell, hemistepcin B, cumambrin B, costunolide and tulipinolide were shown significant activities with $IC_{50}$ range of 2.2, 4.1, 0.9 and $0.3\;{\mu}g/ml$, respectively. In the cytotoxicity test against human tumor cells, the most active compound was costunolide having $IC_{50}$ values of below $0.3\;{\mu}g/ml$ against all the tested cell lines except for UACC62. Cumambrin A, hendelin and costunolide exhibited more strong activity against HCT15 and UO-31 cell lines than a positive control, adriamycin. All tested compounds showed an $IC_{50}$ values of below $5.0\;{\mu}g/ml$ against all the tested cell lines.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.632-635
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• 2003

For the disposal of low-level radwaste from nuclear power plant need the determination of levels of radio nuclides in radwaste. These nuclides include the difficult-to-measure nuclides, so indirect methodology for the determination of the difficult-to-measure nuclides have to be developed. In this work, for the determination of $^129I(t_{1/2}=1.57{\times}10^7 years)$ in low-level radwaste from nuclear power plant is investigated. Recovery of Iodide in simulated waste($UO_2$ pellet) as a soluble and radwaste(resin, woolen fabric)as a insoluble samples are measured. After pretreatment of sample, $I_2$ are extracted from aqueous solution with $CCl_4$. Then I are extracted from $CCl_4$ with 0.1M $NaHSO_3$ aqueous solution. iodide in aqueous solution are determined by ion chromatography. The overall recovery yield is 76.7 (RSD 1.7%) for mixed-acid digestion method. Incase of woolen fabrics, overall recovery yield is 74.3 (RSD 2.2%) and recovery of iodide in resin 56.5(RSD 5.6%) for alkaline fusion method.

• Journal of the Korean Ceramic Society
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• v.51 no.5
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• pp.418-423
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• 2014

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 $UO_2$ 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 $Al_2O_3-Y_2O_3$ 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.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.499-507
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• 2020

Research reactors for radioisotope production, fuel and material testing and research activities are designed, constructed and operated based on the society's needs. In this study, neutronic and thermal hydraulic design of a high neutron flux research reactor core for radioisotope production is presented. Main parameters including core excess reactivity, reactivity variations, power and flux distribution during the cycle, axial and radial power peaking factors (PPF), Pu₂₃₉ production and minimum DNBR are calculated by nuclear deterministic codes. Core calculations performed by deterministic codes are validated with Monte Carlo code. Comparison of the neutronic parameters obtained from deterministic and Monte Carlo codes indicates good agreement. Finally, subchannel analysis performed for the hot channel to evaluate the maximum fuel and clad temperatures. The results show that the average thermal neutron flux at the beginning of cycle (BOC) is 1.0811 × 10¹⁴ n/㎠-s and at the end of cycle (EOC) is 1.229 × 10¹⁴ n/㎠-s. Total Plutonium (Pu₂₃₉) production at the EOC evaluated to be 0.9487 Kg with 83.64% grade when LEU (UO₂ with 3.7% enrichment) used as fuel. This designed reactor which uses LEU fuel and has high neutron flux and low plutonium production could be used for peaceful nuclear activities based on nuclear non-proliferation treaty concepts.

• Nuclear Engineering and Technology
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• v.14 no.3
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• pp.103-109
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• 1982

An empirical formula for determining water content as functions of uranium concentration and nitric acid normalities in uranyl nitrate solutions has been derived from a least-squares analysis of experimental data, i.e., uranium concentration, nitric acid normalities and solution densities for a large number of UO$_2$(NO$_3$)$_2$ solutions. The formula derived is Q=1-0.3628C-0.0327H$^{+}$ where Q, C, and H$^{+}$ stand for water content (g/cc), uranium concentration (g/cc), ana nitric acid normality, respectively. Atom number densities and nuclear criticality for hypothetical uranyl nitrate solutions have been calculated by using the empirical formula, ana compared with the results obtained on the basis of uranium concentration, nitric acid normality, and solution density. The empirical formula derived in this study seems to be useful in uranium concentrations ranging from 0.295g/cc down to 0.004g/cc and nitric acid normality from 5.06 to 1.00..00.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3059-3072
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• 2022

A Jacobian-Free Newton Krylov Two-Nodal Coarse Mesh Finite Difference algorithm based on Nodal Expansion Method (NEM_TNCMFD_JFNK) is successfully developed and proposed to solve the three-dimensional (3D) and multi-group reactor physics models. In the NEM_TNCMFD_JFNK method, the efficient JFNK method with the Modified Incomplete LU (MILU) preconditioner is integrated and applied into the discrete systems of the NEM-based two-node CMFD method by constructing the residual functions of only the nodal average fluxes and the eigenvalue. All the nonlinear corrective nodal coupling coefficients are updated on the basis of two-nodal NEM formulation including the discontinuity factor in every few newton steps. All the expansion coefficients and interface currents of the two-node NEM need not be chosen as the solution variables to evaluate the residual functions of the NEM_TNCMFD_JFNK method, therefore, the NEM_TNCMFD_JFNK method can greatly reduce the number of solution variables and the computational cost compared with the JFNK based on the conventional NEM. Finally the NEM_TNCMFD_JFNK code is developed and then analyzed by simulating the representative PWR MOX/UO₂ core benchmark, the popular NEACRP 3D core benchmark and the complicated full-core pin-by-pin homogenous core model. Numerical solutions show that the proposed NEM_TNCMFD_JFNK method with the MILU preconditioner has the good numerical accuracy and can obtain higher computational efficiency than the NEM-based two-node CMFD algorithm with the power method in the outer iteration and the Krylov method using the MILU preconditioner in the inner iteration, which indicates the NEM_TNCMFD_JFNK method can serve as a potential and efficient numerical tool for reactor neutron diffusion analysis module in the JFNK-based multiphysics coupling application.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.603-610
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• 2021

The major drawback of zirconia-based materials, in view of their applications as targets for minor actinide transmutation, is their poor thermal conductivity. The addition of MgO, which has high thermal conductivity, to zirconia-based materials is expected to improve their thermal conductivity. On these grounds, the present study aims at phase characterization and thermophysical property evaluation of neodymium-substituted zirconia (Zr_0.8Nd_0.2O_1.9; using Nd₂O₃ as a surrogate for Am₂O₃) and its composites with MgO. The composite was prepared by a solid-state reaction of Zr_0.8Nd_0.2O_1.9 (synthesized by gel combustion) and commercial MgO powders at 1773 K. Phase characterization was carried out by X-ray diffraction and the microstructural investigation was performed using a scanning electron microscope equipped with energy dispersive spectroscopy. The linear thermal expansion coefficient of Zr_0.8Nd_0.2O_1.9 increases upon composite formation with MgO, which is attributed to a higher thermal expansivity of MgO. Similarly, specific heat also increases with the addition of MgO to Zr_0.8Nd_0.2O_1.9. Thermal conductivity was calculated from measured thermal diffusivity, temperature-dependent density and specific heat values. Thermal conductivity of Zr_0.8Nd_0.2O_1.9-MgO (50 wt%) composite is more than that of typical UO₂ fuel, supporting the potential of Zr_0.8Nd_0.2O_1.9-MgO composites as target materials for minor actinides transmutation.