• 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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.475-479
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• 1995

During the hypothetical loss of coolant accident in the nuclear power plant the emergency core cooling water could not penetrate to the reactor core when the steam flow rate from the reactor core exceeds CCFL (Countercurrent flow limitation). The CCFL generated by earlier investigators are developed under the steady gas flow. However the flow instability in the reactor loop could generate oscillatory steam flow, hence their applicability under oscillating flow should be investigated. In this work, an experimental investigation of countercurrent flow in the vertical flow channel has been conducted under oscillatory gas flow. Pulsation of gas under oscillatory flow disturbs the flow pattern significantly and prevents flooding (CCFL) when its minimum value is less than the threshold gas flow rate value.

• Korean Chemical Engineering Research
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• 제59권1호
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• pp.85-93
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• 2021

For accurate and reliable process design for phenol oxidation in a plug flow reactor with supercritical water, modeling can be very insightful. Here, the velocity and density distribution along the reactor have been predicted by a numerical model and variations of temperature and phenol mass fraction are calculated under various flow conditions. The numerical model shows that as we proceed along the length of the reactor the temperature falls from above 430 ℃ to approximately 380 ℃. This is because the generated heat from the exothermic reaction is less that the amount lost through the walls of the reactor. Also, along the length, the linear velocity falls to less than one-third of the initial value while the density more than doubles. This is due to the fall in temperature which results in higher density which in turn demands a lower velocity to satisfy the continuity equation. Having a higher oxygen concentration at the reactor inlet leads to much faster phenol destruction; this leads to lower capital costs (shorter reactor will be required); however, the operational expenditures will increase for supplying the needed oxygen. The phenol destruction depends heavily on the kinetic parameters and can be as high as 99.9%. Using different kinetic parameters is shown to significantly influence the predicted distributions inside the reactor and final phenol conversion. These results demonstrate the importance of selecting kinetic parameters carefully particularly when these predictions are used for reactor design.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회B
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• pp.2088-2093
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• 2008

In this study, flow field measurement of the Pebble Bed Reactor(PBR) for the High Temperature Gas-cooled Reactor(HTGR) was performed. Large number of pebbles in the core of PBR provides complicated flow channel. Due to the complicated geometries, numerical analysis has been intensively made rather than experimental observation. However, the justification of computational simulation by the experimental study is crucial to develop solid analysis of design method. In the present study, a wind tunnel installed with pebbles stacked was constructed and equipped with the Particle Image Velocimetry(PIV). We designed the system scaled up to realize the room temperature condition according to the similarity. The PIV observation gave us stagnation points, low speed region so that the suspected high temperature region can be identified. With the further supplementary experimental works, the present system may produce valuable data to justify the Computational Fluid Dynamics(CFD) simulation method.

• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3320-3325
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• 2023

The scaled water-cooled Reactor Cavity Cooling System (RCCS) experimental facility reproduces a passive safety feature to be implemented in Generation IV nuclear reactors. It keeps the reactor cavity and other internal structures in operational conditions by removing heat leakage from the reactor pressure vessel. The present work uses Flownex one-dimensional thermal-fluid code to model the facility and predict the experimental thermal-hydraulic behavior. Two representative steady-state cases defined by the bulk volumetric flow rate are simulated (Re = 2,409 and Re = 11,524). Results of the cavity outlet temperature, risers' temperature profile, and volumetric flow split in the cooling panel are also compared with the experimental data and RELAP system code simulations. The comparisons are in reasonable agreement with the previous studies, demonstrating the ability of Flownex to simulate the RCCS behavior. It is found that the low Re case of 2,409, temperature and flow split are evenly distributed across the risers. On the contrary, there's an asymmetry trend in both temperature and flow split distributions for the high Re case of 11,524.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 1995년도 추계학술대회논문집; 한국종합전시장, 24 Nov. 1995
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• pp.201-207
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• 1995

A series of vibration assessment programs has been performed for Yonggwang Nuclear Power Plant Unit 4 (YGN 4) in order to verify the structural integrity of the reactor internals for flow induced vibration prior to its commercial operation. The structural analysis was done to provide the basis for measurement and the theoretical evidence for the structural integrity of the reactor internals. The actual flow induced hydraulic loads and reactor internals vibration response data were measured and recorded during pre-core hot functional testing of the plant. Then, the measured data have been reduced and analyzed, and compared with the analysis results such as the frequency contents, stresses, strains and displacements. It is concluded that the structural analysis methodology performed for vibration response of the reactor internals due to the flow induced vibration is appropriately conservative, and also that the structural integrity of YGN 4 reactor internals to flow induced vibration is acceptable for long term operation.

• 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.

• 유체기계공업학회:학술대회논문집
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• 유체기계공업학회 2003년도 유체기계 연구개발 발표회 논문집
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• pp.161-166
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• 2003

It is the common features of the integral reactors that the main components of the RCS are installed within the reactor vessel, and so there are no any flow pipes connecting the coolant pumps or steam generators. Due to no any flow pipes, it is impossible to measure the differential pressure at the RCS of the integral reactors, and it also makes impossible measure the flow-rate of the reactor coolant. As a alternative method, the method by the measurement of coolant pump power has been introduced in this study. Up to now, we did not found out a precedent which the coolant pump power is used for the real-time flow-rate calculation at normal operation of the commercial nuclear power plants. The objective of the study is to embody the real-time flow-rate calculation method by the measurement of coolant pump power in an integral reactor. As a result of the study, we could theoretically reason that the capacity-head curve and capacity-shaft power curve around the rated capacity with the high specific-speeded axial flow pumps have each diagonally steep incline but show the similar shape. Also, we could confirm the above theoretical reasoning from the measured result of the pump motor inputs, So, it has been concluded that it is possible to calculate the real-time flow-rate by the measurement of pump motor inputs. In addition, the compensation for a above new method can be made by HBM being now used in the commercial nuclear power plants.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1604-1615
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• 2023

In an open-pool type research reactor with a downward forced flow in the core, pipes can be under sub-atmospheric pressure because of the large pressure drop at the reactor core in the atmospheric pool. Sub-atmospheric pressure can result in air inflow into the pipe from the pressure difference between the atmosphere and the inside of the pipe, which in a postulated pipe break scenario can lead to the breakdown of the cooling pump. In this study, a plant-scale experiment was conducted to study air inflow in large piping systems by considering the actual operational conditions of an advanced research reactor. The air inflow rate was measured, and the entrained air was visualized to investigate the behavior of air inflow and flow regime depending on the pipe break size. In addition, the developed drift-flux model for a large vertical pipe with a diameter of 600 mm was compared with other correlations. The flow regime transition in a large vertical pipe under downward flow was also studied using the newly developed drift-flux model. Consequently, the characteristics of two-phase flow in a large vertical pipe were found to differ from those in small vertical pipes where liquid recirculation was not dominant.

• Nuclear Engineering and Technology
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• 제52권7호
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• pp.1417-1428
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• 2020

The design of a nuclear reactor core requires basic thermal-hydraulic information concerning the heat transfer regime at which onset of nucleate boiling (ONB) will occur, the pressure drop and flow rate through the reactor core, the temperature and power distributions in the reactor core, the departure from nucleate boiling (DNB), the condition for onset of flow instability (OFI), in addition to, the critical velocity beyond which the fuel elements will collapse. These values depend on coolant velocity, fuel element geometry, inlet temperature, flow direction and water column above the top of the reactor core. Enough safety margins to ONB, DNB and OFI must-emphasized. A heat transfer package is used for calculating convection heat transfer coefficient in single phase turbulent, transition and laminar regimes. The main objective of this paper is to study the possibility of power upgrading of WWR-S research reactor from 2 to 10 MW_th. This study presents a one-dimensional mathematical model (axial direction) for steady-state thermal-hydraulic design and analysis of the upgraded WWR-S reactor in which two types of plate fuel elements are employed. FOR-CONV computer program is developed for the needs of the power upgrading of WWR-S reactor up to 10 MW_th.