• Nuclear Engineering and Technology
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• v.41 no.1
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• pp.11-20
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• 2009

Radiation effects on materials are inherently multiscale phenomena in view of the fact that various processes spanning a broad range of time and length scales are involved. A multiscale modeling approach to embrittlement of pressure vessel steels is presented here. The approach includes an investigation of the mechanisms of defect accumulation, microstructure evolution and the corresponding effects on mechanical properties. An understanding of these phenomena is required to predict the behavior of structural materials under irradiation. We used molecular dynamics (MD) simulations at an atomic scale to study the evolution of high-energy displacement cascade reactions. The MD simulations yield quantitative information on primary damage. Using a database of displacement cascades generated by the MD simulations, we can estimate the accumulation of defects over diffusional length and time scales by applying kinetic Monte Carlo simulations. The evolution of the local microstructure under irradiation is responsible for changes in the physical and mechanical properties of materials. Mechanical property changes in irradiated materials are modeled by dislocation dynamics simulations, which simulate a collective motion of dislocations that interact with the defects. In this paper, we present a multi scale modeling methodology that describes reactor pressure vessel embrittlement in a light water reactor environment.

• Nuclear Engineering and Technology
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• v.27 no.5
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• pp.710-720
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• 1995

The propagation of pump-induced pressure pulsation in a reactor is important because of the potential for vibration and resultant damage of reactor internals. A hydrodynamic model has been developed to obtain the pressure fluctuation due to the operation of pumps in the annulus(between the core support barrel and reactor vessel of a pressurized water reactor) including the coolant inlet pipe. The mathematical analysis is formulated in accordance with the linearized Navier-Stokes equation by assuming a compressible, inviscid flow. Two regions are considered separately and by coupling the solutions of the inlet pipe and the annulus, the inlet nozzle pressure(pressure at pipe and annulus interface) is to be calculated without assumptions. The geometric parameter effect on the pump-induced pressure pulsation is evaluated. Comparison of predicted and measured inlet nozzle pressure values for each forcing frequency shows good order of magnitude agreement.

• Journal of KSNVE
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• v.10 no.1
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• pp.129-137
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• 2000

This paper presents the vibration characteristics of reactor internals of Ulchin-1 nuclear power plant, which are identified by using the conventional and the phase separated spectral analysis of the pressure vessel acceleration and ex-core neutron signals. These identified vibration characteristics show excellent agreement with those of Tricastin-1 nuclear power plant that is the prototype of Ulchin-1. And the trend of ex-core neutron signals has been observed during one reactor cycle. These results can be used as basic data for fault diagnosis of reactor internals.

• Proceedings of the Korean Vacuum Society Conference
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• 2008.02a
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• pp.53-53
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• 2008

• Nuclear Engineering and Technology
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• v.34 no.4
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• pp.344-357
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• 2002

The present paper reports spontaneous steam explosions observed in fuel coolant interaction experiments using prototypic reactor materials. Pure ZrO$_2$ and a mixture of UO$_2$ and ZrO$_2$ are used. A high temperature molten material in the form of a jet is poured into a subcooled water pool located in a pressure vessel. An induction skull melting technique is used for the melting of the reactor material. In both tests using pure ZrO$_2$ and a mixture of UO$_2$ and ZrO$_2$, either a quenching or a spontaneous steam explosion was observed. The morphology of debris and pressure profile clearly indicate the differences between the qunching cases and explosion cases. The dynamic pressure. dynamic impulse, water temperature, melt temperature, and static pressure Inside the containment chamber were measured . As the spontaneous steam explosion for the reactor material is firstly observed in the present experiments, the results of present experiments could be a siginificant step forward the understanding the explosion of the reactor material.

• Nuclear Engineering and Technology
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• v.47 no.5
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• pp.534-545
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• 2015

In the very high temperature reactor (VHTR), which is a next generation nuclear reactor system, ceramics are used as a fuel coating material and graphite is used as a core structural material. Even if a depressurization accident occurs and the reactor power goes up instantly, the temperature of the core will change only slowly. This is because the thermal capacity of the core is so high. Therefore, the VHTR system can passively remove the decay heat of the core by natural convection and radiation from the surface of the reactor pressure vessel. The objectives of this study are to investigate the heat transfer characteristics of natural convection of a one-side heated vertical channel with inserted porous materials of high porosity and also to develop the passive cooling system for the VHTR. An experiment was carried out using a one-side heated vertical rectangular channel. To obtain the heat transfer and fluid flow characteristics of the vertical channel with inserted porous material, we have also carried out a numerical analysis using a commercial Computational Fluid Dynamics (CFD) code. This paper describes the thermal performances of the one-side heated vertical rectangular channel with an inserted copper wire of high porosity.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.717-724
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• 2008

Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source for nuclear hydrogen generation, which can produce hydrogen from water or natural gas. A primary hot gas duct (HGD) as a coaxial double-tube type cross vessel is a key component connecting the reactor pressure vessel and the intermediate heat exchanger for the VHTR. In this study, structural sizing methodology for the primary HGD with a coaxial double-tube of the VHTR that produces heat at temperatures in the order of $950^{\circ}C$ was suggested and a structural pre-sizing of it was carried out as an example.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.327-327
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.239-239
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.225-225
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• 1999