• 한국전산유체공학회:학술대회논문집
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• 2005.04a
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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.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.227-239
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• 2015

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, doublefault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperature profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain coolable. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel-coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.14 no.6
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• pp.441-449
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• 2002

Single-phase convection and partial nucleate boiling in free-surface and submerged jet impingements of subcooled water ejected through a 2-mm-diameter circular pipe nozzle were investigated by local measurements. Effects of jet velocity and nozzle-to-imping-ing surface distance as well as heat flux on distributions of wall temperature and heat transfer coefficients were considered. Incipience of boiling began from far downstream in contrast with the cases of the planar water jets of high Reynolds numbers. Heat flux increase and velocity decrease reduced the temperature difference between stagnation and far downstream regions with the increasing influence of boiling in partial boiling regime. The chance in nozzle-to-impinging surface distance from H/d=1 to 12 had a significant effect on heat transfer around the stagnation point of the submerged jet, but not for the free-surface jet. The submerged jet provided the lower cooling performance than the free-surface jet due to the entrainment of the pool fluid of which temperature increased.

• Journal of the Korean Solar Energy Society
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• v.23 no.4
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• pp.37-43
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• 2003

The aim of this research is to study the feasibility of the solar(hot fluid) driven $NH_3/H_2O$ absorption chiller, made by re-manufacturing of Gas fired $NH_3/H_2O$ absorption chiller. This experimental study is performed with the temperature of the inlet hot fluid of generator. In order to determine the inlet temperature of the generator, which gives maximum COP, the experimental data are obtained with various hot fluid supply temperature in range of $130\sim170^{\circ}C$. Remodeled chiller is operated with periodical cooling effect, which due to mixture subcooled pool boiling, then the COP is evaluated in average. The maximum COP$(\sim0.36)$ is at $160^{\circ}C$. The temperature is stable operation temperature range of typical vacuum collector. It offers a feasibility of solar driven $NH_3/H_2O$ absorption chiller.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2005.11a
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• pp.394-399
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• 2005

A case history is presented pertaining to high frequency piping vibration and noise caused by globe valve in the spent fuel pool cooling system of nuclear power plant. Frequency analyses were performed on the system to diagnose the problem and develop a solution to reduce the piping vibration and noise. The source of the high frequency and noise energy was traced to the globe valve located immediately downstream of the centrifugal pump by performing valve throttling test. Measurements of vibration and noise are presented to show that the high frequency vibration and noise amplitude was dependent upon the valve disc position and flow rate. Strouhal vortex shedding frequencies were generated at the exit of the globe valve which exited structural resonance of valve disc and amplified the high frequency vibration and noise. The problem was identified as an interaction between the flow inside globe valve and the valve disc structure. Attempts to reduce the vibration and noise amplitudes of the piping system were successfully achieved by the modification of guide-disc diameter and disc-edge figure The valve disc was replaced by an alternative to eliminate the source of the harmful high frequency vibration and noise.

• Nuclear Engineering and Technology
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• v.40 no.1
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• pp.93-98
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• 2008

Several adhesive tapes have been studied in terms of their suitability for securing gold wires into positions for neutron flux measurements in the reactor core and irradiation facilities surrounding the core of the Open Pool Australian Light water (OPAL) reactor. Gamma ray spectrometry has been performed on each irradiated tape in order to identify and quantify activated components. Numerous metallic impurities have been identified in all tapes. Calculations relating to both the effective neutron shielding properties of the tapes and the error in measurement of the $^{198}Au$ activity caused by superfluous activity due to residual tape have been made. The most important identified effects were the prolonged cooling times required before safe enough levels of radioactivity to allow handling were reached, and extra activity caused by residual tape when measured with an ionisation chamber. Knowledge of the most suitable tape can allow a minimal contribution due to these effects, and the use of gamma spectrometry in preference to ionisation chamber measurements of the flux wires is shown to make all systematic errors due to the tape completely negligible.

• Nuclear Engineering and Technology
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• v.45 no.4
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• pp.469-476
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• 2013

In a postulated severe core damage accident in a PHWR, multiple failures of core cooling systems may lead to the collapse of pressure tubes and calandria tubes, which may ultimately relocate inside the calandria vessel forming a terminal debris bed. The debris bed, which may reach high temperatures due to the decay heat, is cooled by the moderator in the calandria. With time, the moderator is evaporated and after some time, a hot dry debris bed is formed. The debris bed transfers heat to the calandria vault water which acts as the ultimate heat sink. However, the questions remain: how long would the vault water be an ultimate heat sink, and what would be the failure mode of the calandria vessel if the heat sink capability of the reactor vault water is lost? In the present study, a numerical analysis is performed to evaluate the thermal loads and the stresses in the calandria vessel following the above accident scenario. The heat transfer from the molten corium pool to the surrounding is assumed to be by a combination of radiation, conduction, and convection from the calandria vessel wall to the vault water. From the temperature distribution in the vessel wall, the transient thermal loads have been evaluated. The strain rate and the vessel failure have been evaluated for the above scenario.

• Journal of Powder Materials
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• v.28 no.5
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• pp.410-416
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• 2021

Ti-6Al-4V alloy has a wide range of applications, ranging from turbine blades that require smooth surfaces for aerodynamic purposes to biomedical implants, where a certain surface roughness promotes biomedical compatibility. Therefore, it would be advantageous if the high volumetric density is maintained while controlling the surface roughness during the LPBF of Ti-6Al-4V. In this study, the volumetric energy density is varied by independently changing the laser power and scan speed to document the changes in the relative sample density and surface roughness. The results where the energy density is similar but the process parameters are different are compared. For comparable energy density but higher laser power and scan speed, the relative density remained similar at approximately 99%. However, the surface roughness varies, and the maximum increase rate is approximately 172%. To investigate the cause of the increased surface roughness, a nonlinear finite element heat transfer analysis is performed to compare the maximum temperature, cooling rate, and lifetime of the melt pool with different process parameters.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1845-1850
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• 2022

In-vessel corium retention (IVR) during external reactor vessel cooling (ERVC) is a key severe accident management strategy adopted in advanced nuclear power plants. The injection of nanofluids has been regarded as a means of enhancing CHF when using the IVR-ERVC strategy to safeguard high-power nuclear reactors. However, a critical practical concern is that various types of debris flowing from the contaminant sump during operation of an ERVC system might degrade CHF enhancement by nanofluids. Our objective here was to experimentally assess the viability of nanofluid use to enhance CHF in practical ERVC contexts (e.g., when fluids contain various types of debris). The types and characteristics of debris expected during IVR-ERVC were examined. We performed pool boiling CHF experiments using nanofluids containing these types of debris. Notably, we found that debris did not cause any degradation of the CHF enhancement characteristics of nanofluids. The nanoparticles are approximately 1000-fold smaller than the debris particles; the number of nanoparticles in the same volume fraction is 1 billion-fold greater. Nanofluids increase CHF via porous deposition of nanosized particles on the boiling surface; this is not hindered by extremely large debris particles.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1911-1923
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• 2023

The nuclear safety assessment involving large transient simulations is forcing the community to develop methods for coupling thermal-hydraulics and neutronic codes and three-dimensional (3D) Computational Fluid Dynamics (CFD) codes. In this paper a set of dynamic boundary conditions are implemented in OpenFOAM^Ⓡ in order to apply zero-dimensional (0D) approaches coupling with 3D thermal-hydraulic simulation in a single framework. This boundary conditions are applied to model pipelines, tanks, pumps, and heat exchangers. On a first stage, four tests are perform in order to assess the implementations. The results are compared with experimental data, full 3D CFD, and system code simulations, finding a general good agreement. The semi-implicit implementation nature of these boundary conditions has shown robustness and accuracy for large time steps. Finally, an application case, consisting of a simplified open pool with a cooling external circuit is solved to remark the capability of the tool to simulate thermal hydraulic systems commonly found in nuclear installations.