• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2008.04a
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• pp.641-646
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• 2008

Long and slender body with or without flexible supports under severe operating condition can be unstabilized even by the small cross flow. Turbulent flow mixer, which actually increases thermal-hydraulic performance of the nuclear fuel by boosting turbulence, disturbs the flow field around the fuel rod and affects dynamic behavior of the nuclear fuel rods. Few studies on this problem can be found in the literature because these effects depend on the specific natures of the support and the design of the system. This work shows how the dynamics of a multi-span fuel rod can be affected by the turbulent flow, which is discretely activated by a flow mixer. By solving a state-space form of the eigenvalue equation for a multi-span fuel rod system, the critical velocity at which a fuel rod becomes unstable was established. Based on the simulation results, we evaluated how stability of a multi-spanned nuclear fuel rod with mixing vanes can be affected by the coolant flow in an operating reactor core.

• Nuclear Engineering and Technology
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• v.29 no.4
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• pp.336-347
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• 1997

During postulated severe accidents in Light water Reactors, molten corium which was ejected from the reactor vessel bottom, may erode the concrete basemat of the containment and there by threaten the containment integrity. This study experimentally examines the molten core-concrete interaction (MCC) using 20kg of thermite melt (Fe + $Al_2$O$_3$) and the concrete, used in Yonggwang Nuclear Power Plant Units 3 and 4 (YGN 3 & 4) in Korea. The measured data are the downward heat fluxes, concrete erosion rate, gases and particle generation rates during MCCI. Transient results ore compared with those of TURCIT experiment conducted by SNL in USA. The peak downward heat flux to the concrete was measured to be about 2.1㎿/$m^2$. The initial concrete erosion rate was 175cm per hour, decreasing to 30cm per hour. It was shown from the post-test that the erosion was progressed downward up to 18mm in the concrete slug.

• Nuclear Engineering and Technology
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• v.47 no.4
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• pp.389-397
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• 2015

The effects of mixing vanes (MVs) attached to a grid spacer on the characteristics of air-water annular flows were experimentally investigated. To know the effects, a grid spacer with or without MV was inserted in a vertical circular pipe of 16-mm internal diameter. For three cases (i.e., no spacer, spacer without MV, and spacer with MV), the liquid film thickness, liquid entrainment fraction, and deposition rate were measured by the constant current method, single liquid film extraction method, and double liquid film extraction method, respectively. The MVs significantly promote the re-deposition of liquid droplets in the gas core flow into the liquid film on the channel walls. The deposition mass transfer coefficient is three times higher for the spacer with MV than for the spacer without MV, even for cases 0.3-m downstream from the spacer. The liquid film thickness becomes thicker upstream and downstream for the spacer with MV, compared with the thickness for the spacer without MV and for the case with no spacer.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.709-718
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• 2012

The pressurized water reactor APR1400 adopts DVI (Direct Vessel Injection) for the emergency cooling water in the upper downcomer annulus. The International Standard Problem number 50 (ISP-50) was launched with the aim to investigate thermal hydraulic phenomena during a 50% DVI line break scenario with best estimate codes making use of the experimental data available from the ATLAS facility located at KAERI. The present work describes the calculation results obtained for the ISP-50 using the RELAP5/MOD3.3 system code. The work aims at validation and assessment of the code to reproduce the observed phenomena and investigate about its limitations to predict complicated mixing phenomena between the subcooled emergency cooling water and the two-phase flow in the downcomer. The obtained results show that the overall trends of the main test variables are well reproduced by the calculations. In particular, the pressure in the primary system show excellent agreement with the experiment. The loop seal clearance phenomenon was observed in the calculation and it was found to have an important influence on the transient progression. Moreover, the collapsed water levels in the core are accurately reproduced in the simulations. However, the drop in the downcomer level before the activation of the DVI from safety injection tanks was underestimated due to multi-dimensional phenomena in the downcomer that are not properly captured by one-dimensional simulations.

• Proceedings of the KSME Conference
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• 2003.11a
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• pp.1133-1138
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• 2003

The pressure tubes, which contain high temperature heavy water and fuel, are within the core of a CANDU nuclear reactor, and are thus subjected to high stresses, temperature gradient, and neutron flux. Further, it is well known that pressure tubes of cold-worked Zr-2.5Nb materials result in hydrogen diffusion, which create fully-hydrided regions (frequently called Blister). Thus a proper investigation of hydrogen diffusion within zirconium-alloy nuclear components, such as CANDU pressure tube and fuel channels is essential to predict the structural integrity of these components. In this respect, this paper presents numerical investigation of hydrogen diffusion to quantify the hydrogen concentration for blister growth of CANDU pressure tube. For this purpose, coupled temperature-hydrogen diffusion analyses are performed by means of two-dimensional finite element analysis. Comparison of predicted temperature field and blister with published test data shows good agreement.

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
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• 2005.11a
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• pp.167-168
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• 2005

The operational characteristics of the integrated three-phase flux-lock type superconducting fault current limiter (SFCL) were analyzed. The suggested three-phase SFCL consisted of a three-phase flux-lock reactor and three high-Tc superconducting (HTSC) elements. The former has three windings wound on an iron core, each of which has the same turn's ratio between coil 1 and coil 2. The latter are connected in series with coil 2 of each phase. The integrated three-phase flux-lock type SFCL showed the operational characteristics that the fault phase could affect the sound phase, which resulted in quenching the HTSC element in the sound phase. Through the computer simulation applying numerical analysis for its three-phase equivalent circuit, the fault current limiting characteristics of the integrated three-phase flux-lock type SFCL according to the ground fault types were compared.

• Proceedings of the KIEE Conference
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• 1999.11d
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• pp.1054-1056
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• 1999

It is necessary to install the arc generation facility to obtain the important technology for the design of breakers and switches, and the improvement of their performance and reliability. With this facility it is possible to study the characteristics of Arc in air/gas/vacuum insulation environment. The arc generation facility briefly consists of capacitor bank which can charge enormous energy, an air-core reactor, and several measurement equipments. This facility can simulates the arc phenomena in breakers and switches by means of generating high currents. In order to the protect electrode damage during the arcing time in arc extinguishing chamber, we installed auxiliary current source in addition to main capacitor bank, This auxiliary current source produces relatively small arc between electrodes before high current generation by main capacitor bank. Therefore it is possible to observe and measure the arcing phenomenon without damage of electrodes.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.17 no.9
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• pp.819-825
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• 2007

Vibration characteristics of an duel-cooling cylindrical fuel rod, which was proposed as a candidate design of fuel's cross section for the ultra-high burnup nuclear fuel, according to the cross-sectional dimensions and the number of supports or the span length were analytically studied. Finite element(FE) modeling for the annular cross sectional fuel was based on the methodology, that have been proven by the test verification, for the conventional PWR nuclear fuel rod. A commercial FEA code, ABAQUS, was used for the FE modeling and analysis. A planar beam element (B21) that uses a linear interpolation was used for the fuel rod and a linear spring element for the spring and dimple of the SG. Natural frequencies and mode shape were calculated according to the preliminary design candidates for the fuel's cross sectional dimension and the number of span. From the analysis results, the design scheme of the annular fuel compatible to the present PWR nuclear reactor core was discussed in terms of the number of supports and fuel's cross section.

• Proceedings of the KWS Conference
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• 2005.11a
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• pp.70-72
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• 2005

Nuclear fuel assemblies for pressurized water reactors(PWR) are loaded in the reactor core throughout the residence time of three to five years. A spacer grid assembly, which is an interconnected array of slotted grid straps and is welded at the intersections to form an egg crate structure, is one of the main structural components of the nuclear fuel assembly. The spacer grid assembly is structurally required to have enough buckling strength under various kinds of lateral loads acting on the nuclear fuel assembly so as to keep the nuclear fuel assembly straight. To meet this requirement, it is necessary to weld the welding parts carefully and precisely. In this study, laser welding qualities of the Zircaloy spacer grid assembly welded by two welding companies, such as weld strength, weld penetration depth, and weld bead size, are examined and compared.

• Journal of Radiation Protection and Research
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• v.37 no.2
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• pp.70-74
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• 2012

Silicon carbide (SiC) is a promising material for neutron detection at harsh environments because of its capability to withstand strong radiation fields and high temperatures. Two PIN-type SiC semiconductor neutron detectors, which can be used for nuclear power plant (NPP) applications, such as in-core reactor neutron flux monitoring and measurement, were designed and fabricated. As a preliminary test, MCNPX simulations were performed to estimate reaction probabilities with respect to neutron energies. In the experiment, I-V curves were measured to confirm the diode characteristic of the detectors, and pulse height spectra were measured for neutron responses by using a $^{252}Cf$ neutron source at KRISS (Korea Research Institute of Standards and Science), and a Tandem accelerator at KIGAM (Korea Institute of Geoscience and Mineral Resources). The neutron counts of the detector were linearly increased as the incident neutron flux got larger.