• International Journal of Fluid Machinery and Systems
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• v.2 no.4
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• pp.400-408
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• 2009

Optimization of seal geometries can reduce significantly the energetic losses in a hydraulic seal [1], especially for high head runner turbine. In the optimization process, a reliable prediction of the losses is needed and CFD is often used. This paper presents numerical experiments to determine an adequate CFD model for straight, labyrinth and stepped hydraulic seals used in Francis runners. The computation is performed with a finite volume commercial CFD code with a RANS low Reynolds turbulence model. As numerical computations in small radial clearances of hydraulic seals are not often encountered in the literature, the numerical results are validated with experimental data on straight seals and labyrinth seals. As the validation is satisfactory enough, geometrical optimization of hydraulic seals using CFD will be studied in future works.

• Animal Systematics, Evolution and Diversity
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• v.36 no.3
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• pp.274-275
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• 2020

Recently, the new genus Parkiana (Lecithoceridae) and its two new species, Parkiana matutinalis and Parkiana andasibensis, were described by the authors from Andasibe, Madagascar. Although the morphological descriptions and figures fully characterized the new genus with two new species in that paper, the journal issue in which the description appeared was published online-only, and the ZooBank LSID, required for validation of new names in electronic-only publications, was not included. Therefore, the present note serves to validate the names of the new genus and species, Parkiana, Parkiana matutinalis and Parkiana andasibensis, by fulfilling Code conditions for nomenclatural availability.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.1917-1925
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• 2020

New versions of ENDF/B and JEFF data libraries have been released during the past two years with significant updates in the neutron reaction sublibrary and the thermal neutron scattering sublibrary. In order to get a more comprehensive impression of the criticality quality of these two latest neutron data libraries, and to provide reference for the selection of the evaluated nuclear data libraries for the science and engineering applications of the Reactor Monte Carlo code RMC, the criticality benchmarking of the two latest neutron data libraries has been performed. RMC was employed as the computational tools, whose processing capability for the continuous representation ENDF/B-VIII.0 thermal neutron scattering laws was developed. The RMC criticality validation suite consisting of 116 benchmarks was established for the benchmarking work. The latest ACE format data libraries of the neutron reaction and the thermal neutron scattering laws for ENDF/B-VIII.0, ENDF/B-VII.1, and JEFF-3.3 were downloaded from the corresponding official sites. The ENDF/B-VII.0 data library was also employed to provide code-to-code validation for RMC. All the calculations for the four different data libraries were performed by using a parallel version of RMC, and all the calculated standard deviations are lower than 30pcm. Comprehensive analyses including the C/E values with uncertainties, the δk/σ values, and the metrics of χ² and < |Δ| >, were conducted and presented. The calculated keff eigenvalues based on the four data libraries generally agree well with the benchmark evaluations for most cases. Among the 116 criticality benchmarks, the numbers of the calculated keff eigenvalues which agree with the benchmark evaluations within 3σ interval (with a confidence level of 99.6%) are 107, 109, 112, and 113 for ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0 and JEFF-3.3, respectively. The present results indicate that the ENDF/B-VIII.0 neutron data library has a better performance on average.

• Journal of the Korea Institute of Information and Communication Engineering
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• v.20 no.3
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• pp.543-548
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• 2016

Public electronic certificates, as an important means for identification, have been used as the main economic transactions, including banking, e-government, e-commerce. identification. However, damage cases of certificates have been increased by Illegally issued and by hacking practices. Also the users have a difficult in ensuring that their certificates when and where to use. Therefore, the proposed system gives the organization code for the Institutions using OCSP services in advance, the organization code embedded in extensions of OCSP request message structure when institutions ask the validation of certificate to CA(Certificate Authority). Also, OCSP server can extract the organization code from OCSP request message, confirm the institution, and record it in certificate history management table of DB. In this paper, we presented a system that could determine the certificate history check using OCSP service, public electronic certificate validation service, and implemented to prevent and cope immediately with financial incidents.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1487-1503
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• 2019

The methods and performance of a pin-level nuclear reactor core thermal-hydraulics (T/H) code ESCOT employing the drift-flux model are presented. This code aims at providing an accurate yet fast core thermal-hydraulics solution capability to high-fidelity multiphysics core analysis systems targeting massively parallel computing platforms. The four equation drift-flux model is adopted for two-phase calculations, and numerical solutions are obtained by applying the Finite Volume Method (FVM) and the Semi-Implicit Method for Pressure-Linked Equation (SIMPLE)-like algorithm in a staggered grid system. Constitutive models involving turbulent mixing, pressure drop, and vapor generation are employed to simulate key phenomena in subchannel-scale analyses. ESCOT is parallelized by a domain decomposition scheme that involves both radial and axial decomposition to enable highly parallelized execution. The ESCOT solutions are validated through the applications to various experiments which include CNEN $4{\times}4$, Weiss et al. two assemblies, PNNL $2{\times}6$, RPI $2{\times}2$ air-water, and PSBT covering single/two-phase and unheated/heated conditions. The parameters of interest for validation include various flow characteristics such as turbulent mixing, spacer grid pressure drop, cross-flow, reverse flow, buoyancy effect, void drift, and bubble generation. For all the validation tests, ESCOT shows good agreements with measured data in the extent comparable to those of other subchannel-scale codes: COBRA-TF, MATRA and/or CUPID. The execution performance is examined with a mini-sized whole core consisting of 89 fuel assemblies and for an OPR1000 core. It turns out that it is about 1.5 times faster than a subchannel code based on the two-fluid three field model and the axial domain decomposition scheme works as well as the radial one yielding a steady-state solution for the OPR1000 core within 30 s with 104 processors.

• International Journal of High-Rise Buildings
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• v.2 no.2
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• pp.123-130
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• 2013

In the past two decades, researchers from different countries have conducted series of experimental and theoretical studies to investigate the behaviour of structures in fire. Many new insights, data and calculation methods have been reported, which form the basis for modern interdisciplinary structural fire engineering. Some of those methods are now adopted in quantitative performance-based codes and have been migrated into practice. Mainly based on the achievements in structural fire research at China, the Chinese national code for fire safety of steel structures in buildings has been drafted and approved, and will be released in this year. The code is developed to prevent steel structures subjected to fire from collapsing, ensure safe evacuation of building occupants, and reduce the cost for repairing the damages of the structure caused by fire. This paper presents the main contents of the code, which includes the fire duration requirements of structural components, fundamental requirements on fire safety design of steel components, temperature increasing of atmosphere and structural components in fire, loading effect and capacity of various components in fire, and procedure for fire-resistant check and design of steel components. The analytical approaches employed in the code and their validation works are also presented.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1218-1230
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• 2019

A sub-channel solver, named ${\underline{S}}teady$ and ${\underline{T}}ransient$ ${\underline{A}}nalyzer$ for ${\underline{R}}eactor$ ${\underline{T}}hermal$ hydraulics (START), has been developed using the homogenous model for two-phase conditions of light water reactors. The code is developed as a fast and accurate TH-solver for coupled and multi-physics calculations. START has been validated against the NUPEC PWR Sub-channel and Bundle Test (PSBT) database. Tests like single-channel quality and void-fraction for steady state, outlet fluid temperature for steady state, rod-bundle quality and void-fraction for both steady state and transient conditions have been analyzed and compared with experimental values. Results reveal a good accuracy of solution for both steady state and transient scenarios. Axially different values for turbulent mixing coefficient are used based on different grid-spacer types. This provides better results as compared to using a single value of turbulent mixing coefficient. Code-to-code evaluation of PSBT results by the START code compares well with other industrial codes. The START code has been parallelized with the OpenMP algorithm and its numerical performance is evaluated with a large whole PWR core. Scaling study of START shows a good parallel performance.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.161-161
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.67-68
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• 2005

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2852-2859
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• 2020

This paper deals with the simulation-based design optimization and experimental validation of the characteristics of an in-core silver Self-Powered Neutron Detector (SPND). Optimized dimensions of the SPND are determined by combining Monte Carlo simulations and analytical methods. As a first step, the Monte Carlo transport code MCNPX is used to follow the trajectory and fate of the neutrons emitted from an external source. This simulation is able to seamlessly integrate various phenomena, including neutron slowing-down and shielding effects. Then, the expected number of beta particles and their energy spectrum following a neutron capture reaction in the silver emitter are fetched from the TENDEL database using the JANIS software interface and integrated with the data from the first step to yield the origin and spectrum of the source electrons. Eventually, the MCNPX transport code is used for the Monte Carlo calculation of the ballistic current of beta particles in the various regions of the SPND. Then, the output current and the maximum insulator thickness to avoid breakdown are determined. The optimum design of the SPND is then manufactured and experimental tests are conducted. The calculated design parameters of this detector have been found in good agreement with the obtained experimental results.