• Journal of computational fluids engineering
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• v.10 no.4 s.31
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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.

• Nuclear Engineering and Technology
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• v.53 no.7
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• pp.2207-2220
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• 2021

Lead alloy is used as coolant in Lead-based cooled Fast Reactor (LFR). The natural characteristics of lead alloy are combined with the simple structural design of LFR. This constitutes the inherent safety characteristics of LFR. The main work of this paper is to take the main coolant pump (MCP) in the lead-cooled fast reactor (LFR) as the research object, and to study the flow pattern distribution of the internal flow field under the reverse rotation pump condition, the reverse rotation positive-flow braking condition and the reverse rotation negative-flow braking condition. In this paper, the double-outlet volute type and the space guide vane are selected as the potential designs of the CLEAR-I MCP. In this paper, the CFD method is used to study the potential reverse accident of the MCP. It is found that the highest flow velocity in the impeller appears at the impeller outlet, and the Q-H curves of the two design programs basically coincide. The space guide vane type MCP has better hydraulic performance under the reverse rotation positive-flow condition, the Q-H curves of the two designs gradually separate with increasing flow rate, and the maximum flow velocity inside the space guide vane type MCP is obviously lower than that of the double-outlet volute type. For the reverse rotation test of MCP, only the condition of the forward rotating pump of the main coolant pump is tested and verified. For the simulation of the MCP in LBE medium, it proved that the turbulence model and basic settings selected in the simulation are reliable.

• Nuclear Engineering and Technology
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• v.44 no.5
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• pp.501-506
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• 2012

Many innovative design features are employed in the reactor vessel internals of SMART, a small integral-type pressurized water reactor, one of which is the flow skirt, which uniformly distributes flow and horizontally restrains the lower part of the core support barrel. This new design requires a comprehensive investigation of vibration characteristics. Therefore, in this study, modal characteristics of flow skirts are investigated with finite element analysis. Specifically, we investigate how the presence of holes, the presence of three rings attached to the flow skirt, and the thickness of the lowest shell effect vibration characteristics. In addition, the fluid effect is addressed, since the flow skirt is submerged in the fluid.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4477-4490
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• 2023

Fuel locking mechanisms (FLMs) are essential in upward-flow research reactors to prevent accidental fuel separation from the core during reactor operation. This study presents a novel design concept for a remotely controlled plate-type nuclear fuel locking mechanism. By employing electromagnetic field analysis, we optimized the design of the electromagnet for fuel unlocking, allowing the FLM to adapt to various research reactor core designs, minimizing installation space, and reducing maintenance efforts. Computational flow analysis quantified the drag acting on the fuel assembly caused by coolant upflow. Subsequently, we performed finite element analysis and evaluated the structural integrity of the FLM based on the ASME boiler and pressure vessel (B&PV) code, considering design loads such as dead weight and flow drag. Our findings confirm that the new FLM design provides sufficient margins to withstand the specified loads. We fabricated a prototype comprising the driving part, a simplified moving part, and a dummy fuel assembly. Through basic operational tests on the assembled components, we verified that the manufactured products meet the performance requirements. This remote-controlled micro locking mechanism holds promise in enhancing the safety and efficiency of plate-type nuclear fuel operation in upflow research reactors.

• 한국전산유체공학회:학술대회논문집
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• 2005.10a
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• pp.45-47
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• 2005

We have investigated the mixing behavior of a tubular heat exchanger reactor using CFD and compared its mixing performance with different type of reactors such as jet mixer and continuous stirred tank reactor (CSTR). The mixedness in each reactor was quantified introducing a statistical average value, the coefficient of variation (CoV), which is a normalized standard deviation of concentration of a component over the whole fluid domain. Through the analysis of the flow pattern and turbulent energy distribution, we suggested a simple but effective way to improve the mixing performance of the tubular heat-exchanger reactor, which include the addition of the internals and/or the increase of the recycle flow rate. It was found that the CoV value of the tubular reactor could be nearly equivalent to that of CSTR by applying those two alternatives suggested here.

• The KSFM Journal of Fluid Machinery
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• v.18 no.2
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• pp.54-59
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• 2015

The performance of UV reactor which is used in water treatment is strongly affected by UV fluence rate and water flow in the UV reactor. Therefore, CFD tools are widely used in designing process of UV reactors. This paper describes the development of a computational fluid dynamics (CFD) methodology that can be used to calculate the performance of open channel type UV reactor used in wastewater treatment plant. All computations were performed using commercial CFD code, CFX, by considering three dimensional, steady, incompressible flow. The Eulerian-Eulerian multi-phase method were used to capture the water-air interface. The MSSS model, provided by UVCalc3D, was used to calculate the UV intensity field. The numerical predictions and calculated UV Dose were compared with experimental dataset to validate the CFD methodology. The reactor performance based on MS2 log reduction was well matched with measurements within 6%.

• Journal of computational fluids engineering
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• v.20 no.3
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• pp.87-93
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• 2015

In this study, a numerical analysis was performed to simulate the thermal-hydraulic response of the secondary side of a steam generator(SG) model equipped with an orifice-type SG outlet flow restrictor to a main steam line break(MSLB) at a pressurized water reactor(PWR) plant. The SG analysis model includes the SG upper steam space and the part of the main steam pipe between the SG outlet and the broken pipe end. By comparing the numerical calculation results for the present SG model to those obtained for a simple SG model having no flow restrictor, the effects of the flow restrictor on the thermal-hydraulic response of SG to the MSLB were investigated.

• The KSFM Journal of Fluid Machinery
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• v.17 no.6
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• pp.77-83
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• 2014

The vessel of integral reactor contains its major primary components such as the fuel and core, pumps, steam generators, and a pressurizer, so its size is proportional to the required space for the installation of each component. The steam generators take up the largest volume of internal space of reactor vessel and their volumes is substantial for the overall size of reactor vessel. Reduction of installation space for steam generators can lead to much smaller reactor vessel with resultant decrease of overall cost for the components and related facilities. A printed circuit heat exchanger is one of the compact types of heat exchangers available as an alternative to conventional shell and tube heat exchangers. Its name is derived from the procedure used to manufacture the flat metal plates that form the core of the heat exchanger, which is done by chemical milling. These plates are then stacked and diffusion bonded, converting the plates into a solid metal block containing precisely engineered fluid flow passages. The overall heat transfer area and pressure drops are evaluated for the steam generator based on the concept of the printed circuit heat exchanger in this study. As the printed circuit heat exchanger is known to have much larger heat transfer area density per unit volume, we can expect significantly reduced steam generator compared to former shell and tube type of steam generator. For the introduction of new steam generator, two design requirements are considered: flow area ratio between primary and secondary flow paths, and secondary side parallel channel flow oscillation. The results show that the overall volume of the steam generator can be significantly reduced with printed circuit type of steam generator.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.134-140
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• 2016

The Control Rod Assembly (CRA) controls the reactor power by adjusting its position in the reactor core during normal operation and should be quickly inserted into the reactor core by free drop under scram condition to shut down chain reactions. Therefore, the drop time of the CRA is one of important factors for the safety of the nuclear reactor and must be experimentally verified. This study presents the drop performance test of the CRA which has been conceptually designed for the Proto-type Generation IV Sodium-cooled Fast Reactor. During the test, the CRA was free dropped from a height of 1 m under different flow rate conditions and its drop time was measured. The results showed that the drop time of the CRA increased as the flow rate increased; the average drop times of the CRA were approximately 1.527 seconds, 1.599 seconds and 1.676 seconds at 0%, 100% and 200% of design flow rates, respectively.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.5 no.3
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• pp.227-231
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• 2004

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 steam generator or the pump whose type is the axial flow. 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 the pump power of the axial flow pump has been introduced in this study. Up to now, we did not found out a precedent which the pump power is used for the flow-rate calculation at normal operation of the commercial nuclear power plants. The objective of the study is to embody the flow-rate calculation method by the measurement of the 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 flow-rate by the measurement of the pump motor inputs.