• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1204-1212
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• 2024

To accurately calculate the heating distribution of the fast reactor, a neutron-photon library in MATXS format named Knight-B7.1-1968n × 94γ was processed based on the ENDF/B-VII.1 library for ultrafine groups. The neutron cross-section processing code MGGC2.0 was used to generate few-group neutron cross sections in ISOTXS format. Additionally, the self-developed photon cross-section processing code NGAMMA was utilized to generate photon libraries for neutron-photon coupled heating calculations, including photo-atom cross sections for the ISOTXS format, prompt photon production cross sections, and kinetic energy release in materials (KERMA) factors for neutrons and photons, and the self-shielding effect from the capture and fission cross sections of neutron to photon have been taken into account when the photon source generated by neutron is calculated. The interface code GSORCAL was developed to generate the photon source distribution and interface with the DIF3D code to calculate the neutron-photon coupling heating distribution of the fast reactor core. The neutron-photon coupled heating calculation route was verified using the ZPPR-9 benchmark and the RBEC-M benchmark, and the results of the coupled heating calculations were analyzed in comparison with those obtained from the Monte Carlo code MCNP. The calculations show that the library was accurately processed, and the results of the fast reactor neutron-photon coupled heating calculations agree well with those obtained from MCNP.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.11-18
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• 2022

In this research, curium oxide (CmO₂) is studied as fuel for VVER-1200 reactor to get an attention to its energy value and possibilities. For this purpose, CmO₂ is used in fuel rods or integrated burnable absorber (IBA) rods with and without UO₂ and then compared with the conventional fuel assembly of VVER-1200 reactor. It is burned to 60 GWd/t by using SRAC-2006 code and JENDL-4.0 data library. From these studies, it is found that CmO₂ is competent like UO₂ as a fuel due to higher fission cross-section of ²⁴³Cm and ²⁴⁵Cm isotopes and neutron capture cross-section of ²⁴⁴Cm and ²⁴⁶Cm isotopes. As a result, when some or all of the UO₂ of fuel rods or IBA rods are replaced by CmO₂, we get a similar k-inf like the reference even with lower enrichment UO₂ fuels. These studies show that the use of CmO₂ as IBA rods is more effective than the fuel rods considering the initially loaded amount, power peaking factor (PPF), fuel temperature and void coefficient, and the quality of spent fuel. From a detailed study, 3% CmO₂ with inert material ZrO₂ in IBA rods are recommended for the VVER-1200 reactor assembly from the once through concept.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.423-429
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• 2021

The investigation of the utilization of enriched ²⁰⁸Pb as a coolant to enhance the performance of a long-life fast reactor with a Modified CANDLE (Constant Axial shape of Neutron flux, nuclide densities, and power shape During Life of Energy production) burnup scheme has performed. The analyzes were performed on a reactor with thermal power of 800 MegaWatt Thermal (MWTh) with a refueling process every 15 years. Uranium Nitride (enriched ¹⁵N), ²⁰⁸Pb, and High-Cr martensitic steel HT-9 were employed as fuel, coolant, and cladding materials, respectively. One of the Pb-nat isotopes, ²⁰⁸Pb, has the smallest neutron capture cross-section (0.23 mb) among other liquid metal coolants. Furthermore, the neutron-producing cross-section (n, 2n) of ²⁰⁸Pb is larger than sodium (Na). On the other hand, the inelastic scattering energy threshold of ²⁰⁸Pb is the highest among Na, ^natPb, and Bi. The small inelastic scattering cross-section of ²⁰⁸Pb can harden the neutron energy spectrum. Therefore, ²⁰⁸Pb is a better neutron multiplier than any other liquid metal coolant. The excess neutrons cause more production than consumption of ²³⁹Pu. Hence, it can reduce the initial fuel loading of the reactor. The selective photoreaction process was developing to obtain enriched ²⁰⁸Pb. The neutronic was calculated using SRAC and JENDL 4.0 as a nuclear data library. We obtained that the modified CANDLE reactor with enriched ²⁰⁸Pb as coolant and reflector has the highest k-eff among all reactors. Meanwhile, the ^natPb cooled reactor has the lowest k-eff. Thus, the utilization of the enriched ²⁰⁸Pb as the coolant can reduce reactor initial fuel loading. Moreover, the enriched ²⁰⁸Pb-cooled reactor has the smallest power peaking factor among all reactors. Therefore, the enriched ²⁰⁸Pb can enhance the performance of a long-life Modified CANDLE fast reactor.

• Structural Engineering and Mechanics
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• v.52 no.4
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• pp.787-813
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• 2014

Starting with Hamilton's variational principle, the governing field equations for the steady state response of thin-walled beams under harmonic forces are derived. The formulation captures shear deformation effects due to bending and warping, translational and rotary inertia effects and as well as torsional flexural coupling effects due to the cross section mono-symmetry. The equations of motion consist of four coupled differential equations in the unknown displacement field variables. A general closed form solution is then developed for the coupled system of equations. The solution is subsequently used to develop a family of shape functions which exactly satisfy the homogeneous form of the governing field equations. A super-convergent finite element is then formulated based on the exact shape functions. Key features of the element developed include its ability to (a) isolate the steady state response component of the response to make the solution amenable to fatigue design, (b) capture coupling effects arising as a result of section mono-symmetry, (c) eliminate spatial discretization arising in commonly used finite elements, (d) avoiding shear locking phenomena, and (e) eliminate the need for time discretization. The results based on the present solution are found to be in excellent agreement with those based on finite element solutions at a small fraction of the computational and modelling cost involved.

• Structural Engineering and Mechanics
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• v.56 no.2
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• pp.169-188
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• 2015

There are two types of nonlinear analysis methods for building frameworks depending on the method of modeling the plastification of members including lumped plasticity and distributed plasticity. The lumped plasticity method assumes that plasticity is concentrated at a zero-length plastic hinge section at the ends of the elements. The distributed plasticity method discretizes the structural members into many line segments, and further subdivides the cross-section of each segment into a number of finite elements. When a reinforced concrete member experiences inelastic deformations, cracks tend to spread form the joint interface resulting in a curvature distribution. The program IDARC includes a spread plasticity formulation to capture the variation of the section flexibility, and combine them to determine the element stiffness matrix. In this formulation, the flexibility distribution in the structural elements is assumed to be the linear. The main objective of this study is to evaluate the accuracy of linear flexibility distribution assumed in the spread inelasticity model. For this purpose, nonlinear analysis of two reinforced concrete frames is carried out and the linear flexibility models used in the elements are compared with the real ones. It is shown that the linear flexibility distribution is incorrect assumption in cases of significant gravity load effects and can be lead to incorrect nonlinear responses in some situations.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2019-2028
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• 2024

Advanced nuclear reactors, especially the newly developed small and micro-reactors have complex neutron spectrum, which makes the deterministic reactor core calculations sensitive to the energy group structure of few-group cross-sections. To avoid significantly increasing the cost of energy discretization in the core calculation, two energy group structures with 31 groups and 33 groups were adopted for typical thermal and fast reactor cores, respectively. Then, an adaptive scheme of group division for reactor cores with a medium neutron spectrum was proposed. The works were based on the full spectrum nuclear reactor analysis code SARAX/TULIP. An equivalent one-dimensional model of the core was proposed to capture the key neutron spectrum features of the reactor core. Such features were used to adaptively determine a few-group structure for the following reactor core calculations. Then, the neutron spectrum in different zones with more details was calculated. With this spectrum, the cross-sections were condensed into the determined energy groups. Three tests based on different neutron spectrum were calculated to verify the schemes. The results show that using the adaptive energy group division scheme, the following core calculation can meet the accuracy requirement of different reactors with different neutron spectra.

• Nuclear Engineering and Technology
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• v.21 no.1
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• pp.12-17
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• 1989

The correction factor of neutron fraction absorbed by $^{55}$ Mn in the MnSO$_4$ bath was determined for the absolute measurement of neutron emission rate by using the solution circulation-type manganese sulfate bath system. For the determination of this correction factor, I/f, the atomic number desnsity and the effective neutron capture cross section data of Mn, S and impurity elements in the MnSO$_4$ solution were determined. For the atomic number density determination, the MnSO$_4$ solution concentration was determined by using the volumetric EDTA titration and gravimetric method. The impurity contents were analyzed by using the ICP method. For the calculation of effective neutron capture cross sections, a FORTRAN computer program EASCAL was developed in this study. in which Westcott's parameters and Axton's empirical relations are used.

• 한국가시화정보학회:학술대회논문집
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• 2002.11a
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• pp.19-22
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• 2002

A process of 3-D particle image velocimetry, called here, as '3-D stereo PIV' was developed for the measurement of a section field of 3-D complex flows. The present method includes modeling of camera by a calibrator based on the homogeneous coordinate system, transfromation of oblique-angled image to transformed image, identification of 2-D velocity vectors by 2-D cross-correlation equation, stereo matching of 2-D velocity vectors of two cameras, accurate calculation of 3-D velocity vectors by homogeneous coordinate system and finally 3-D animation as the post processing. In principle, as two frame images only are necessary for the single instantaneous analysis of a section field of 3-D flow, more effective vectors are obtainable contrary to the previous multi-frame vector algorithm. An experimental system was also used for the application of the proposed method. Three analog CCD cameras and an Argon-Ion Laser(300mW) for illumination were adopted to capture the wake flow behind a bluff obstacle.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2000.06a
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• pp.71-77
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• 2000

An axisymmetric supersonic jet is simulated at a Mach number of 1.5 and a Reynolds number of $10^5$ to identify the mechanism of sound radiation from the jet. The present simulation is performed based on the high-order accuracy and high-resolution ENO(Essentially Non-Oscillatory) schemes to capture the time-dependent flow structure representing the sound source. In this simulation, optimum expansion jet is selected as a target, where the pressure at nozzle exit is equal to that of the ambient pressure, to see pure shear layer growth without effect of change in jet cross section due to expansion or shock wave generated at nozzle exit. Shock waves are generated near vortex rings, and discernible pressure waves called Mach wave are radiated in the downstream direction with an angle from the jet axis, which is characteristic of high speed jet noise. Furthermore, vortex roll-up phenomena are observed through the visualization of vorticity contours.

• Journal of computational fluids engineering
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• v.20 no.4
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• pp.51-57
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• 2015

In the present study, two-dimensional numerical investigation of flow past a square cylinder beneath the free surface has been performed to identify the effects of presence of the free surface. An immersed boundary method was adopted for implementation of the cylinder cross-section in a Cartesian grid system. Also, a level-set method was used to capture the interface of two fluids. To prevent transition to three-dimensional flow, Reynolds number chosen for this simulation was 150. The cases for Froude number 0.2 and gap ratio(h/D) between 0.25 and 5.00 were examined. At the specific Reynolds number, we study the effects of gap ratio on flow characteristics around a square cylinder by computing flow fields, force coefficients and Strouhal number.