• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.1109-1112
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• 2017

The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prerequisite for the long-term operation of nuclear power plants beyond their original design life. The expiration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels. The applicability of the TTS model to predict the embrittlement of Korean RPVs made of SA533B-1 plates and welds was investigated in this study. It was concluded that the TTS model of 10 CFR 50.61a matched the trends of the radiation embrittlement in the SA533B-1 plates and welds better than did that of Regulatory Guide (RG) 1.99 Rev. 2. This is attributed to the fact that the prediction performance of 10 CFR 50.61a was enhanced by considering the difference in radiation embrittlement sensitivity among the different types of RPV materials.

• Journal of the Korean Society of Safety
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• v.25 no.5
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• pp.7-14
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• 2010

Ultra high pressure system, which can be generally increased over 1,000bar, needs to have sealing mechanism to protect leakage and selection of the materials used in the intensifier. Components such as pressure vessel, hydraulic hose assembly, accumulator, hydraulic cylinder, hydraulic valve, pipe, etc., are tested under the impulse-pressure conditions. Components need to be tested under 1.5 to 3 times of rated pressure to check the tolerance even though rated pressure range of these components are not ultra high pressure. So, the ultra high pressure system needs to be equiped to test components. In this study, safety assessments of ultra high pressure system which are using failure analysis of components, changing the types of the control system, and finite element analysis with static condition, are investigated.

• 유체기계공업학회:학술대회논문집
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• 2004.12a
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• pp.492-500
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• 2004

The purpose of this study is localization of safety relief valves fur Nuclear Service through technical development with overall design, fabrication, inspection, capacity certification test and functional qualification test of safety relief valves in accordance with KEPIC MN Code(or ASME Sec.III ). The safely relief valve is the important equipment used to protect the pressure vessel, the steam generator and the other pressure facility from overpressure by discharging the operating medium when the pressure of system is reaching the design pressure of the system. But we're depending on technology of the other country up to the present time. Because we don't have our own technologies, we have been spent the great time and money on installing and repairing safety relief valve at nuclear power plant. Therefore we have to achieve the development of safety relief valves for Nuclear Service with our own technologies.

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.257-272
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• 2007

Supercritical water-cooled reactors (SCWRs) are recognized as a Generation IV reactor concept. The Super LWR is a pressure-vessel type thermal spectrum SCWR with downward-flow water rods and is currently under study at the University of Tokyo. This paper reviews Super LWR safety. The fundamental requirement for the Super LWR, which has a once-through coolant cycle, is the core coolant flow rate rather than the coolant inventory. Key safety characteristics of the Super LWR inhere in the design features and have been identified through a series of safety analyses. Although loss-of-flow is the most important abnormality, fuel rod heat-up is mitigated by the "heat sink" and "water source" effects of the water rods. Response of the reactor power against pressurization events is mild due to a small change in the average coolant density and flow stagnation of the once-through coolant cycle. These mild responses against transients and also reactivity feedbacks provide good inherent safety against anticipated-transient-without-scram (ATWS) events without alternative actions. Initiation of an automatic depressurization system provides effective heat removal from the fuel rods. An "in-vessel accumulator" effect of the reactor vessel top dome enhances the fuel rod cooling. This effect enlarges the safety margin for large LOCA.

• 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

• Journal of the Korean Society of Marine Environment & Safety
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• v.24 no.3
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• pp.352-359
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• 2018

As a fuel for ship propulsion, liquefied natural gas (LNG) is currently considered a proven and reasonable solution for meeting the IMO emission regulations, with gas engines for the LNG-fueled ship covering a broad range of power outputs. For an LNG-fueled ship, the LNG bunkering process is different from the HFO bunkering process, in the sense that the cryogenic liquid transfer generates a considerable amount of boil-off gas (BOG). This study investigated the effect of the temperature difference on boil-off gas (BOG) production during ship-to-ship (STS) LNG bunkering to the receiving tank of the LNG-fueled ship. A concept design was resumed for the cargo/fuel tanks in the LNG bunkering vessel and the receiving vessel, as well as for LNG handling systems. Subsequently, the storage tank capacities of the LNG were $4,500m^3$ for the bunkering vessel and $700m^3$ for the receiving vessel. Process dynamic simulations by Aspen HYSYS were performed under several bunkering scenarios, which demonstrated that the boil-off gas and resulting pressure buildup in the receiving vessel were mainly determined by the temperature difference between bunkering and the receiving tank, pressure of the receiving tank, and amount of remaining LNG.

• 한국가스학회:학술대회논문집
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• no.5
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• pp.119-128
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• 2002

• Journal of Mechanical Science and Technology
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• v.19 no.9
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• pp.1761-1772
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• 2005

In blood flow passing through the mechanical heart valve (MHV) and elastic blood vessel, hemolysis and platelet activation causing thrombus formation can be seen owing to the shear stress in the blood. Also, fracture and deformation of leaflets can be observed depending on the shape and material properties of the leaflets which is opened and closed in a cycle. Hence, comprehensive study is needed on the hemodynamics which is associated with the motion of leaflet and elastic blood vessel in terms of fluid-structure interaction. In this paper, a numerical analysis has been performed for a three-dimensional pulsatile blood flow associated with the elastic blood vessel and curved bileaflet for multiple cycles in light of fluid-structure interaction. From this analysis fluttering phenomenon and rebound of the leaflet have been observed and recirculation and regurgitation have been found in the flow fields of the blood. Also, the pressure distribution and the radial displacement of the elastic blood vessel have been obtained. The motion of the leaflet and flow fields of the blood have shown similar tendency compared with the previous experiments carried out in other studies. The present study can contribute to the design methodology for the curved bileaflet mechanical heart valve. Furthermore, the proposed fluid-structure interaction method will be effectively used in various fields where the interaction between fluid flow and structure are involved.

• Journal of the Korean Society for Precision Engineering
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• v.29 no.9
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• pp.1012-1019
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• 2012

In this study, CNG composite vessel is analyzed by using coupled model with liner and composite layer. For the coupled model, a method using theoretical analysis and FEA is suggested: elastic solution for laminated tube is used for theoretical analysis of the composite vessel, FEA is performed to the model of CNG composite vessel in actual conditions. On the basis of these results, optimal thickness and winding angle of the composite layer considering the material properties and thickness of the liner are determined. The results of theoretical analysis and FEA are compared with those carried out in previous studies for verifying the suggested analysis method.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.395-400
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• 1996

Scoping analyses for the Safety Injection System (SIS) configuration for Korean Next Generation Reactor (KNGR) are peformed in this study. The KNGR SIS consists of four mechanically separated hydraulic trains. Each hydraulic train consisting of a High Pressure Safety Injection (HPSI) pump and a Safety Injection Tank (SIT) is connected to the Direct Vessel Injection (DVI) nozzle located above the elevation of cold leg and thus injects water into the upper portion of reactor vessel annulus. Also, the KNGR is going to adopt the advanced design feature of passive fluidic device which will be installed in the discharge line of SIT to allow more effective use of borated water during the transient of large break LOCA. To determine the feasible configuration and capacity of SIT and HPSI pump with the elimination of the Low Pressure Safety Injection (LPSI) pump for KNGR, licensing design basis evaluations are performed for the limiting large break LOCA. The study shows that the DVI injection with the fluidic device SIT enhances the SIS performance by allowing more effective use of borated water for an extended period of time during the large break LOCA.