• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2005년도 춘계 학술대회논문집
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• pp.70-73
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• 2005

The HANARO, multi-purpose research reactor, 30 MWth open-tank-in-pool type, is planning to produce a fission moly-99 of radio isotopes, a mother nuclide of Tc-99m, a medical isotope and is under developing a target handling tool for loading and unloading it in a circular flow tube (OR-5). A guide tube is extended from the reactor core to the top of the reactor chimney for easily loading the target under the reactor normal operation. But active coolant through the core can be quickly raised up to the top of the chimney through the guide tube. The jet flow was suppressed in the guide tube after reducing the inner diameter of a flow restriction orifice installed in the OR-5 flow tube for adding the pressure difference in the flow tube after unloading the target. This paper describes an analytical analysis to calculate the flow distribution in the core of the HANARO after suppressing the jet flow of the guide tube. As results, it was confirmed through the analysis results that the flow distribution in the core of the HANARO were not adversely affected.

• 한국전산유체공학회지
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• 제10권4호통권31호
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• pp.66-71
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• 2005

HANARO, a multi-purpose research reactor, 30 MWth open-tank-in-pool type, is planning to produce a fission moly-99 of radio isotopes, a mother nuclide of Tc-99m, a medical isotope and a target handling tool is under development for loading and unloading it in a circular flow tube (OR-5) of HANARO. A guide tube is extended from the reactor core to the top of the reactor chimney for easily loading the target under a normal operation of the reactor. But active coolant through the core can be quickly raised up to the top of the chimney through the guide tube. The jet flow was suppressed in the guide tube after reducing the inner diameter of a flow restriction orifice installed in the OR-5 flow tube for adding the pressure difference in the flow tube. This paper describes an analytical analysis to calculate the flow distribution in the core of HANARO after suppressing the jet flow of the guide tube. As results, it was confirmed through the analysis results that the flow distribution in the core of HANARO were not adversely affected.

• 한국기계가공학회지
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• 제16권3호
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• pp.16-23
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• 2017

For most machine tools, it is necessary to remove chips and coolant oil because it they will continue to be created during the manufacture of workpieces. Existing products that are in use are installed and used as they reflect depending on the characteristics of each device separately. This study proposes a method to remove the security chip as well as developing an integrated system capable of reducing coolant damage. The Leverage AutoCAD and CATIA program was used for 2D and 3D design, shapes were identified by utilizing the KeyShot program, and the load and displacement analysis of the development apparatus was performed utilizing the ANSYS program. After the prototype underwent sufficient design review, the mixed oil separation device had a complete sensor control program using the LabVIEW program. The chip design process for transferring experiments and experiments on the mixed oil cooling device were developed for performance tests of the product. The final product resulted in an increase in space utilization during commercialization, reduced installation costs, and caused social effects such as pulmonary flow reduction, which, through the economic costs, reduces pollution, resulting in various benefits to the industry, such as deceased errors in the workplace decreases.

• Journal of Advanced Marine Engineering and Technology
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• 제8권1호
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• pp.85-99
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• 1984

The stability of the two-phase flow in a heated channel is of great importance in the design and operation of the boilers and light water nuclear reactors, because it can cause flow oscillations and lead to a violation of thermal limits with resultant overheating of the channels and cladding. This paper presents a systematic evaluation to the variation effects of the basic four (4) dimensionless parameters in a homogeneous equilibrium model. The flow stability is examined on the ground of static characteristic curves. The complicated transfer function of flow dynamics which gives consideration to the transport lag of density wave is derived, and the transient flow stability is analysed by applying the Nyquist stability criterion in control engineering. The analysis results summed up as follows 1. The coolant flow becomes stable in large friction number and specific flow, while it is unstabale in small friction number and flow. 2. Large phase-change number and Froude number destabilize the two-phase flow, but small numbers stabilize it. The effect to variation of phase-change number is more dominant compared with Froude number. 3. The dynamic analysis is required to hold the sufficient safety of heated channels since only static results does not keep it. The special attention could be payed in the design and operation of heat engines, because the unstaable region exists within the stable boundary at small and middle phase-change number and Froude number.

• Nuclear Engineering and Technology
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• 제29권3호
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• pp.229-239
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• 1997

To investigate the realistic behavior of mass and energy release and resultant containment response during large break Loss of Coolant accident (LOCA), analyses are performed for Yonggwang (YGN) 3&4 nuclear power plants by using a merged version of RELAP5/CONTEMPT4 computer code. Comparative analyses by using conservative design computer codes are also peformed. The break types analyzed are the double-ended guillotine breaks at the cold leg and hot leg. The design analysis resulted in containment peak pressure during post-blowdown phase for the cold leg break. However, the RELAP5/CONTEMPT4 analyses show that the containment pressure has a peak during blowdown phase, thereafter it decreases monotonously without the second port-blowdown peak. For the hot leg break, revised design analysis shows much lower pressure than that reported in YGN 3&4 final safety analysis report. The RELAP5/CONTEMPT4 analysis shoos similar trend and confirmed that the bypass flow through the broken loop steam generator during post-blowdown is negligibly small compared to that of cold leg break. The low pressure and temperature predicted tv realistic analysis presented in this paper suggest that the design analysis methodology contains substantial margin and it can be improved to provide benefit in investment protection, such as, relaxing plant technical specifications and reducing containment design pressure.

• Nuclear Engineering and Technology
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• 제37권4호
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• pp.363-374
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• 2005

A heterogeneous thorium-based Kyung Hee Thorium Fuel (KTF) assembly design was assessed for application in the APR-1400 to study the feasibility of using thorium fuel in a conventional pressurized water reactor (PWR). Thermal hydraulic safety was examined for the thorium-based APR-1400 core, focusing on the Departure from Nucleate Boiling Ratio (DNBR) and Large Break Loss of Coolant Accident (LBLOCA) analysis. To satisfy the minimum DNBR (MDNBR) safety limit condition, MDNBR>1.3, a new grid design was adopted, that enabled grids in the seed and blanket assemblies to have different loss coefficients to the coolant flow. The fuel radius of the blanket was enlarged to increase the mass flow rate in the seed channel. Under transient conditions, the MDNBR values for the Beginning of Cycle (BOC), Middle of Cycle (MOC), and End of Cycle (EOC) were 1.367, 1.465, and 1.554, respectively, despite the high power tilt across the seed and blanket. Anticipated transient for the DNBR analysis were simulated at conditions of $112\%$ over-power, $95\%$ flow rate, and $2^{\circ}C$ higher inlet temperature. The maximum peak cladding temperature (PCT) was 1,173K for the severe accident condition of the LBLOCA, while the limit condition was 1,477K. The proliferation resistance potential of the thorium-based core was found to be much higher than that of the conventional $UO_2$ fuel core, $25\%$ larger in Bare Critical Mass (BCM), $60\%$ larger in Spontaneous Neutron Source (SNS), and $155\%$ larger in Thermal Generation (TG) rate; however, the radio-toxicity of the spent fuel was higher than that of $UO_2$ fuel, making it more environmentally unfriendly due to its high burnup rate.

• 한국전산유체공학회지
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• 제14권2호
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• pp.59-67
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• 2009

An accurate prediction of the bypass flow is of great importance in the VHTR core design concerning the fuel thermal margin. Nevertheless, there has not been much effort in evaluating the amount and the distribution of the core bypass flow. In order to evaluate the behavior and the distribution of the coolant flow, a unit-cell experiment was carried out. Unit-cell is the regular triangular section which is formed by connecting the centers of three hexagonal blocks. Various conditions such as the inlet mass flow rate, block combinations and the size of bypass gap were examined in the experiment. CFD analysis was carried out to analyze detailed characteristics of the flow distribution. Commercial CFD code FLUENT 6.3 was validated by comparing with the experimental results. In addition, SST model and standard k-$\varepsilon$ model were validated. The results of CFD simulation show good agreements with the experimental results. SST model shows better agreement than standard k-$\varepsilon$ model. Results showed that block combinations and the size of the bypass gap have an influence on the bypass flow ratio but the inlet mass flow rate does not.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1658-1668
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• 2019

Nuclear power generates a large portion of the energy used today and plays an important role in energy development. To ensure safe nuclear power generation, it is essential to conduct an accurate analysis of reactor structural integrity. Accordingly, in this study, a methodology for obtaining accurate structural responses to the combined seismic and reactor coolant loads existing prior to the shutdown of a nuclear reactor is proposed. By applying the proposed analysis method to the reactor vessel internals, it is possible to derive the seismic responses considering the influence of the hydraulic loads present during operation for the first time. The validity of the proposed methodology is confirmed in this research by using the finite element method to conduct seismic and hydraulic load analyses of the advanced APR1400 1400 MWe power reactor, one of the commercial reactors. The structural responses to the combined applied loads are obtained using displacement-based and stress-based superposition methods. The safety of the subject nuclear reactor is then confirmed by analyzing the design margin according to the American Society for Mechanical Engineers (ASME) evaluation criteria, demonstrating the promise of the proposed analysis method.

• 소음진동
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• 제10권2호
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• pp.261-268
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• 2000

The instrumented capsule is subjected to flow-induced vibration(FIV) due to the flow of the primary coolant and then the structural integrity of the capsule during irradiation in the HANARO reactor is an issue of major concern. For this purpose the acceleration was measured by four accelerometers attached to the protection tube of the capsule mainbody and the displacement of test holes was calcultated using commercial finite element program ANSYS to evaluate the structural interference with the neighboring flow tubes under the reactor operating condition. The calculated displacements of test holes in the reactor in-core were found to be lower than the values of allowable design criteria.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2477-2487
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• 2021

To improve safety analysis technology for a nuclear reactor containment considering an interaction between a reactor coolant system (RCS) and containment, this study aims at an experimental investigation on the integrated simulation of the RCS and containment, with an integral effect test facility, ATLAS-CUBE. For a realistic simulation of a pressure and temperature (P/T) transient, the containment simulation vessel was designed to preserve a volumetric scale equivalently to the RCS volume scale of ATLAS. Three test cases for a steam line break (SLB) transient were conducted with variation of the initial condition of the passive heat sink or the steam flow direction. The test results indicated a stratified behavior of the steam-gas mixture in the containment following a high-temperature steam injection in prior to the spray injection. The test case with a reduced heat transfer on the passive heat sink showed a faster increase of the P/T inside the containment. The effect of the steam flow direction was also investigated with respect to a multi-dimensional distribution of the local heat transfer on the passive heat sink. The integral effect test data obtained in this study will contribute to validating the evaluation methodology for mass and energy (M/E) and P/T transient of the containment.