• Nuclear Engineering and Technology
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• v.44 no.8
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• pp.831-846
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• 2012

KAERI has developed a two-phase CFD code, CUPID, for a refined calculation of transient two-phase flows related to nuclear reactor thermal hydraulics, and its numerical models have been verified in previous studies. In this paper, the CUPID code is validated against experiments on the downcomer boiling and moderator flow in a Calandria vessel. Physical models relevant to the validation are discussed. Thereafter, multi-scale thermal hydraulic analyses using the CUPID code are introduced. At first, a component-scale calculation for the passive condensate cooling tank (PCCT) of the PASCAL experiment is linked to the CFD-scale calculation for local boiling heat transfer outside the heat exchanger tube. Next, the Rossendorf coolant mixing (ROCOM) test is analyzed by using the CUPID code, which is implicitly coupled with a system-scale code, MARS.

• Nuclear Engineering and Technology
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• v.29 no.3
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• pp.240-250
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• 1997

Liquid sodium is widely used as a coolant of LMR(Liquid Metal Reactor) because of its physical and nuclear properties. However, the liquid sodium is very chemically reactive with oxygen and water so that the study on the sodium fire plays an important role in the LMR safety analysis. In this study, a sodium fire model is suggested to analyze the sodium pool fire where both the flame and the reaction products are considered. And also, sodium pool fire analysis computer code, SOPA, is developed. The sensitivity study on the experimental parameters such as the thermal radiation from flame to atmospheric gas, the vessel cooling and the duration of sodium spill was performed. The results showed good agreements with experimental data in the literature.

• Nuclear Engineering and Technology
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• v.38 no.1
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• pp.33-44
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• 2006

This paper presents a review of small-to-medium-sized, pressurized-water-cooled nuclear power reactors whose major primary coolant systems are integrated into a reactor pressure vessel, the concepts categorized as Integrated Primary System Nuclear Reactors (IPSRs). Typical examples of these proposals of interest in this review are CAREM, SMART, IRIS and IMR, all of which are being aimed at the near term deployment. Emphasis is placed on thermal hydraulic aspects. A brief characterization of the IPSR concepts is made and comparisons of plant key parameters are shown. Discussions will follow for the core cooling under rated power conditions and natural circulation heat removal on the basis of the design data available in the public domain.

• Proceedings of the Korean Society of Marine Engineers Conference
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• 2012.06a
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• pp.152-153
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• 2012

While engine room arranging pipe which is used from the vessel, It measured the internal flow of 90 degree elbow which is used from the main condenser. Fluid flow in elbow of 90 degree is measured by PIV and Dewetron system. The Reynolds number adopts 50000 and experimental study of flow field in the elbow.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.240-252
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• 2015

In the original licensing application for the prototype fast-breeder reactor, MONJU, the event progression during an unprotected loss of flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, was evaluated. Through this evaluation, it was confirmed that radiological consequences could be suitably limited even if mechanical energy was released. Following the Fukushima-Daiichi accident, a new nuclear safety regulation has become effective in Japan. The conformity of MONJU to this new regulation should hence be investigated. The objectives of the present study are to conduct a preliminary evaluation of ULOF for MONJU, reflecting the knowledge obtained after the original licensing application through CABRI experiments and EAGLE projects, and to gain the prospect of in-vessel retention for the conformity of MONJU to the new regulation. The preliminary evaluation in the present study showed that no significant mechanical energy release would take place, and that thermal failure of the reactor vessel could be avoided by the stable cooling of disrupted-core materials. This result suggests that the prospect of in-vessel retention against ULOF, which lies within the bounds of the original licensing evaluation and conforms to the new nuclear safety regulation, will be gained.

• Proceedings of the KAIS Fall Conference
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• 2006.11a
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• pp.261-264
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• 2006

본 논문에서는 다양한 곡률반경의 연속에 의하여 살 두께 차가 큰 사출성형품에 불균일한 수축으로 인한 변형이 생성되어 이를 방지하기 위한 적정 CAE 냉각설계를 수행하였다. SAN 및 PMMA 재질의 Jar용기에 대한 균일냉각구조와 최적성형조건을 금형설계에 적용하고자 사출성형의 중요인자인 사출압력, 수지온도, 금형온도, 냉각조건 등을 moldflow 프로그램을 활용하여 연구를 수행하였다. 연구결과로서, 적정 변수인 사출압력 상승, 수지온도 낮춤, 급속냉각으로 후로우 등의 불량현상을 분석하였고 변형 및 불량을 극소화시킬 수 있는 냉각구조와 사이클 시간을 단축시킬 수 있는 사출성형조건을 제시하였고 적정 냉각모듈로부터 냉각시간을 단축하였다.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.2
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• pp.60-66
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• 2017

The standard master curve (MC) approach has the major limitation because it is only applicable to homogeneous datasets. In nature, materials are macroscopically inhomogeneous and involve scatter of fracture toughness data due to various deterministic material inhomogeneity and random inhomogeneity. RPV(reactor pressure vessel) steel has different fracture toughness with varying distance from the inner surface of the wall due to cooling rate in manufacturing process; deterministic inhomogeneity. On the other hand, reference temperature, $T_0$, used in the evaluation of fracture toughness is acting as a random parameter in the evaluation of welding region; random inhomogeneity. In the present paper, four regions, the surface, 1/8T, 1/4T and 1/2T, were considered for fracture toughness specimens of KSNP (Korean Standard Nuclear Plant) SA508 Gr. 3 steel to investigate deterministic material inhomogeneity and random inhomogeneity. Fracture toughness tests were carried out for four regions and three test temperatures in the transition region. Fracture toughness evaluation was performed using the bimodal master curve (BMC) approach which is applicable to the inhomogeneous material. The results of the bimodal master curve analyses were compared with that of conventional master curve analyses. As a result, the bimodal master approach considering inhomogeneous materials provides better description of scatter in fracture toughness data than conventional master curve analysis. However, the difference in the $T_0$ determined by two master curve approaches was insignificant.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.35 no.4
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• pp.425-429
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• 2011

In the unlikely of nuclear reactor meltdown, the leaked core melt or corium must be contained in a device called core-catcher so that the corium can be cooled and stabilized. The ex-vessel behavior of corium involves complex physical and chemical mechanisms of flow propagation, heat transfer, and reactions with sacrificial substrates. In this study, the detailed characteristics of corium flow and heat transfer were investigated by using a commercial CFD code for VULCANO VE-U7 test reported in the literature. The volume-of-fluid (VOF) model was used to predict the interfacial surface formation of corium and the surrounding air, and the discrete ordinate model was adopted to calculate radiation between corium and the surroundings. It was found that cooling via radiation through the top surface of corium had a dominant effect on the temperature and viscosity profiles at the front of the corium flow.

• Journal of the Society of Naval Architects of Korea
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• v.53 no.5
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• pp.420-427
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• 2016

The marine equipment companies expanding facility investment in accordance with the booming economy are suffering from the reduced demand and the growth of chinese businesses. In this regard, the risk of overinvestment and the importance of prudent equipment investment must be reconsidered. Thus, in this study we performed a productivity and economical efficiency analysis in order to evaluate the investment value on production facilities in a company under the present conditions. The freezer of a fishing vessel manufactured by N company is selected as the subject of our study, while the assembly and welding cooling plates are configured as the scope of automation. Analysis on productivity and economical efficiency was conducted through CYCLONE (Cyclic Operation Network) simulation and economic analysis methods after analyzing the production process of freezer. The proposed analytical technique can be used to support the investment decision in production automation equipment of fishing vessels freezer.

• The KSFM Journal of Fluid Machinery
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• v.17 no.6
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• pp.109-114
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• 2014

After the Fukushima accident in 2011, it was revealed that nuclear power plant has the vulnerability to SBO accident and its extension situation without sufficient cooling of reactor core resulting core meltdown and radioactive material release even after reactor shutdown. Many safety systems had been developed like PAFS, hybrid SIT, and relocation of RPV and IRWST as a part of steps for the Fukushima accident, however, their applications have limitation in the situation that supply of feedwater into reactor is impossible due to high pressure inside reactor pressure vessel. The concept of hybrid heat pipe with control rod is introduced for breaking through the limitation. Hybrid heat pipe with control rod is the passive decay heat removal system in core, which has the abilities of reactor shutdown as control rod as well as decay heat removal as heat pipe. For evaluating the cooling performance hybrid heat pipe, a commercial CFD code, ANSYS-CFX was used. First, for validating CFD results, numerical results and experimental results with same geometry and fluid conditions were compared to a tube type heat pipe resulting in a resonable agreement between them. After that, wall temperature and thermal resistances of 2 design concepts of hybrid heat pipe were analyzed about various heat inputs. For unit length, hybrid heat pipe with a tube type of $B_4C$ pellet has a decreasing tendency of thermal resistance, on the other hand, hybrid heat pipe with an annular type $B_4C$ pellet has an increasing tendency as heat input increases.