• Proceedings of the Korean Magnestics Society Conference
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• 2000.05a
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• pp.86-87
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• 2000

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
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• 1997.11a
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• pp.447-450
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• 1997

Temperature distribution measurements in the mockup apparatus of reactor vessel were performed to determine the effective thermal conductivity of Al powder porous media where stainless steel tubes were installed with different geometry. The temperature distributions at four separated sections with different arrangements of porous media have different slopes according to the geometrical configuration. From the measured temperature distribution, effective thermal conductivity have been derived using the least square fitting method.

• Journal of the Korean Society of Manufacturing Process Engineers
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• v.11 no.4
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• pp.160-167
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• 2012

This paper is concerned with a particular fracture mechanics parameter ${\Delta}K_{th}$, known as the 'threshold stress intensity range', or 'fatigue threshold'. This threshold ${\Delta}K_{th}$ constitutes, as it were, a hinge between the notion of crack initiation and the notion of crack growth. It has often been thought that, like the endurance limit, it could be an intrinsic criterion of the material. The study was conducted on a SA-508 pressure vessel steel used in the nuclear power industry. This material exhibits a typical threshold effect in the range of the crack growth rates which were determined; that is, below approximately $da/dN=10^{-6}mm/cycle$, the slope of the da./dN versus ${\Delta}K$ curve is almost vertical. The value of ${\Delta}K_{th}$ was determined at a growth rate of $10^{-7}$ mm/cycle according to the ASTM Standard for threshold testing. The fatigue threshold values are in the range 21 $kg/mm^{3/2}$ to 12 $kg/mm^{3/2}$ depending on the stress ratio effect.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.91-100
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• 2016

The neutron activation inventories in reactor vessel and its internals, and bio-shield of a PWR nuclear power plant were calculated to evaluate the effect of impurity elements contained in the structural materials on the activation inventory. Carbon steel is, in this work, used as the reactor vessel material, stainless steel as the reactor vessel internals, and ordinary concrete as the bio-shield. For stainless steel and carbon steel, one kind of impurity concentration was employed, and for ordinary concrete five kinds were employed in this study using MCNP5 and FISPACT for the calculation of neutron flux and activation inventory, respectively. As the results, specific activities for the cases with impurity elements were calculated to be more than twice than those for the cases without impurity elements in stainless and carbon steel. Especially, the specific activity for the concrete material with impurity elements was calculated to be 30 times higher than that without impurity. Neutron induced reactions and activation inventories in each material were also investigated, and it is noted that major radioactive nuclide in steel material is Co-60 from cobalt impurity element, and, in concrete material, Co-60 and Eu-152 from cobalt and europium impurity elements, respectively. The results of this study can be used for nuclear decommissioning plan during activation inventory assessment and regulation, and it is expected to be used as a reference in the design phase of nuclear power plant, considering the decommissioning of nuclear power plants or nuclear facilities.

• Proceedings of the KWS Conference
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• 2002.10a
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• pp.418-424
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• 2002

Korean Fusion Reactor(KSTAR) system consists of a vacuum vessel, in-vessel components, cryostat, thermal shield, super-conducting magnets and magnet supporting structures. These systems are in the final stage of engineering design with the involvement of industrial manufacturers. The overall configuration and the detailed dimensions of the KSTAR structure have been determined and the first stage of manufacturing is progressing now. In this study, the fabrication and assembly sequence were evaluated in viewpoint of high strengthening joints and very high accuracy. Especially for this purpose, the special cleaning process and welding process were proposed for high strengthening austenitic stainless steel which shall be used at cryogenic temperature. The draft procedure qualification data for welding process are presented with precise welding data including special narrow groove design. For the cooling line attachment on the surface of inside wall of magnet structure case, Induction brazing technology is introduced with some special jigging system and some consumables.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.138-143
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• 2001

To perform the rigorous integrity evaluation of RPV, it is necessary to consider metallurgical factors such as microstructure evolution during multi-pass welding process and PWHT. The microstructures of the heat affected zone(HAZ) of SA508 steel were predicted by a combination of simulated thermal analysis and a simple kinetic models for austenite grain growth and austenite-ferrite transformation. Phase equilibrium of SA508 steel were calculated using a Thermo-Calc package. Carbide growth in th HAZ were predicted by a empirical model, taking into account the predicted microstructure evolution.

• Journal of the Korean Society for Heat Treatment
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• v.25 no.3
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• pp.121-125
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• 2012

The large forgings used in chemical plants or nuclear power plants are produced by complex heat treatment. because of thickness up to 200~300 mm and weight up to 200~300 ton, setting proper heat treatment cycle is so difficult. In addition, defects of products make companies wasting large money and valuable time. In this study, to reduce try & err, when setting heat treatment of reactor pressure vessel steel SA508Gr.3, carrying out the basic mechanical property test of SA508 Gr.3 and testing hardness of SA508Gr.3 in various tempering temperature. and calculating temper curve with Hollomon-Jaffe parameter.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.301.1-301
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• 2002

• Proceedings of the KWS Conference
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• 1998.05a
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• pp.140-143
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• 1998

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.301.2-301
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• 2002