• Nuclear Engineering and Technology
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• v.22 no.2
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• pp.95-100
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• 1990

The exact expression for the 1151 force on a fuel assembly in a reactor core is derived in terms of calculable hydraulic parameters. The relation for the lift force. pressure drop, buoyancy force, viscous force. and fuel assembly weight is discussed. Based on the derived exact expression. error analysis is made for a simple expression applying COBRA IV-i to a typical PWR fuel assembly. The error analysis revealed that the error of the simple expression consists of four terms and the overall error depends on the flow rate change direction, and its magnitude is about 1%.

• Proceedings of the KSME Conference
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• 2007.05a
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• pp.1854-1858
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• 2007

The reliability assurance with respect to the test procedure and results of the out-pile mechanical performance test for the nuclear fuel assembly is an essential task to assure the test quality and to get a permission for fuel loading into the commercial reactor core. For the case of vibration test, which is carried out to obtain basic dynamic characteristics of the fuel assembly, proper management and appropriate calibration of instruments and devices used in the test, various efforts to minimize the possible error during the test and signal acquisition process are needed. Additionally, the deep understanding both of the theoretical assumption and simplification cation for the signal processing/modal analysis and of the functions of the devices used in the test were highly required. Finally, to verify the test result to represent the accurate natural characteristics of the structure, the proper correlation analysis between the theoretical and experimental method has to be carried out. In this study, the overall procedure and result of lateral vibration test for the fuel assembly's mechanical characterization were briefly introduced. A series of measures to assure and improve the reliability of the vibration test were discussed.

• Journal of Energy Engineering
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• v.24 no.3
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• pp.135-141
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• 2015

The core melts stratifies into lower mixture layer and upper metal layer by density in a severe accident condition. The decay heat generated from the mixture layer threatens the integrity of the reactor vessel. This study simulated the natural convection heat transfer of the mixture layer with volumetric heat source using the mass transfer system. $H_2SO_4-CuSO_4$ electroplating system was used as the mass transfer system. With the modified Rayleigh number of $3{\times}10^{14}$, the Nusselt number showed minimum at the bottom and increased along curvature to the top of the experimental apparatus.

• Microbiology and Biotechnology Letters
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• v.41 no.4
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• pp.469-471
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• 2013

In this study, we measured the anti-bacterial activity of radioisotope Lu-177 using a new laboratory instrument. The disk method used for the measurement of existing anti-bacterial antibiotics is drug diffusion into the medium. To measure the antimicrobial activity of a radioisotope, a new type of laboratory instrument is necessary to prevent the drug from spreading and the present invention was thus tested. In the medium, a space isolated separately for radioisotope injection was used to prevent the radioisotope from spreading and radioisotopes are actually injected by this system. We found that the antibacterial activity increased according to the radiation dose increases. It is expected that, through the present study, measuring the antibacterial activity of the other radioisotopes easily in the laboratory will be possible.

• Nuclear Engineering and Technology
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• v.11 no.1
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• pp.29-45
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• 1979

In this paper. nuclear characteristics of TRIGA Mark-III has been analyzed in detail for six different fuel options. Presently, 70 w/o enriched FLIP fuels are adopted for TRIGA core to improve fuel lifetime. However, such highly enriched fuels are not easily obtained due to nonproliferation treaty. This research examines the possible substitution for FLIP fuels with high density fuels without reducing the nuclear performance. This work will provide long-time plan for TRIGA operation.

• International Journal of Concrete Structures and Materials
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• v.10 no.1
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• pp.47-60
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• 2016

The objective of this study was to evaluate the capability of different strength-maturity models to account for the effect of the hydration heat on the in-place strength development of high-strength concrete specifically developed for nuclear facility structures under various ambient curing temperatures. To simulate the primary containment-vessel of a nuclear reactor, three 1200-mm-thick wall specimens were prepared and stored under isothermal conditions of approximately $5^{\circ}C$ (cold temperature), $20^{\circ}C$ (reference temperature), and $35^{\circ}C$ (hot temperature). The in situ compressive strengths of the mock-up walls were measured using cores drilled from the walls and compared with strengths estimated from various strength-maturity models considering the internal temperature rise owing to the hydration heat. The test results showed the initial apparent activation energies at the hardening phase were approximately 2 times higher than the apparent activation energies until the final setting. The differences between core strengths and field-cured cylinder strengths became more notable at early ages and with the decrease in the ambient curing temperature. The strength-maturity model proposed by Yang provides better reliability in estimating in situ strength of concrete than that of Kim et al. and Pinto and Schindler.

• Proceedings of the KIEE Conference
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• 2005.10b
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• pp.270-272
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• 2005

In this paper, we proposed a new method detecting actions of some components driven by the coil excitation. Nuclear power reactors are typically controlled by the movement of several neutron-absorbing control rods into or out of the reactor core. For moving control rods, we use an electromagnetic-jack-typed mechanism, which is called Control Rod Drive Mechanism. This mechanism moves control rods by the step composed of sequential actions of components. In case any mechanical problems happen in the mechanism, the orders for the control rod movement from the higher system cannot be performed properly. This abnormal state must be monitored and the sequential actions of the components can be the monitoring target. The actions of components generate some deviations in the profiles of the currents flowing into the coils in the mechanism. We focused on this phenomena and devised a new method of detecting the actions of the components in Control Rod Drive Mechanism by using the wavelet transform for observing the current profile.

• 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.