• Nuclear Engineering and Technology
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• v.24 no.2
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• pp.168-177
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• 1992

The importance of the study of two phase flow phenomena has increased for both fuel performance and safety analysis of nuclear power plants. In the analysis of two phase flow system, an accurate prediction of local void fractions is very important. In this study, a vertical rectangular subchannel having 4 electrically heated rods is constructed for the measurement of local void fraction under two phase flow. The measurement has been conducted by electrical conductivity probes and signal processing circuit which are known to be adequate to measuring local void fraction. Also experiments are performed with varying the inlet flow rate to search for radial void fraction profile accordingly to the different flow rate even with the same averaged void fraction. From the result of experiments, the validity of electrical conductivity probe and electrical circuit is confirmed.

• Proceedings of the KSME Conference
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• 2007.05a
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• pp.1803-1806
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• 2007

The General roles of a spacer grid(SG) are providing a lateral and vertical support for fuel rods, promoting a mixing of coolant and keeping guide tubes straight so as not to impede a control rod insertion under any normal or accidental conditions. To evaluate the impact characteristics of a SG such as impact velocity, critical buckling strength and duration time, a few types of impact tests for SGs have been conducted. In a previous study, a new welding method, a through-welding method, was proposed to increase critical buckling strength of a SG without any design change or material change and was verified by impact tests with $7{\times}7$ partial SG specimens.In this paper, the effect of through-welding method in case of a $16{\times}16$ full-size SG is investigated by pendulum impact tests with $16{\times}16$ SG specimens. And the increase of critical buckling strength for full-size SGs is measured by comparison with impact results of spot-welded and through-welded SGs.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.291-298
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• 2014

This paper concerns estimating uncertainties of the core neutronics design parameters of power reactors by direct sampling method (DSM) calculations based on the two-step McCARD/MASTER design system in which McCARD is used to generate the fuel assembly (FA) homogenized few group constants (FGCs) while MASTER is used to conduct the core neutronics design computation. It presents an extended application of the uncertainty propagation analysis method originally designed for uncertainty quantification of the FA FGCs as a way to produce the covariances between the FGCs of any pair of FAs comprising the core, or the covariance matrix of the FA FGCs required for random sampling of the FA FGCs input sets into direct sampling core calculations by MASTER. For illustrative purposes, the uncertainties of core design parameters such as the effective multiplication factor ($k_{eff}$), normalized FA power densities, power peaking factors, etc. for the beginning of life (BOL) core of Yonggwang nuclear unit 4 (YGN4) at the hot zero power and all rods out are estimated by the McCARD/MASTER-based DSM computations. The results are compared with those from the uncertainty propagation analysis method based on the McCARD-predicted sensitivity coefficients of nuclear design parameters and the cross section covariance data.

• The KSFM Journal of Fluid Machinery
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• v.17 no.6
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• pp.89-94
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• 2014

The 4th generation SFR is being designed with a milestone of construction by 2028. It is important to understand the subchannel flow characteristics in fuel assembly through the experimental investigations and to estimate the calculation uncertainties for insuring the confidence of the design code calculation results. The friction coefficient and the mixing coefficient are selected as primary parameters. The two parameters are related to the flow distribution and diffusion. To identify the flow distribution, an iso-kinetic method was developed based on the previous study. For the mixing parameters, a wire mesh system and a laser induced fluorescence methods were developed in parallel. The measuring systems were adopted on 37 rod bundle test geometry, which was developed based on the Euler number scaling. A scaling method for a design of experimental facility and the experimental identification techniques for the flow distribution and mixing parameters were developed based on the measurement requirement.

• Proceedings of the Korean Society of Marine Engineers Conference
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• 2005.06a
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• pp.1083-1090
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• 2005

Repeated ribs are used on heat exchange surfaces to promote turbulence and enhance convective heat transfer. Applications include fuel rods of gas-cooled nuclear reactors, inside cavities of turbine blades, and internal surfaces pipes used in heat exchangers. Despite the great number of literature papers, only few experimental data concern detailed distributions of friction factors and heat transfer coefficients in square channels varying the number of rough walls. This issue is tackled by investigating effects of different number of ribbed walls on heat transfer and friction characteristics in square channel. The rough wall have a 45$^{\circ}$ inclined square rib. Uniform heat flux is maintained on whole inner heat transfer channel area. The heat transfer coefficient and friction factor values increase with increasing the number of rough walls.

• Journal of Power System Engineering
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• v.9 no.1
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• pp.30-35
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• 2005

Repeated ribs are used on heat exchanger surfaces to promote turbulence and to enhance convective heat transfer. Applications include fuel rods of gas-cooled nuclear reactors, inside cavities of turbine blades, and internal surfaces pipes used in heat exchangers. Despite the great number of literature papers, only few experimental data concerns detailed distributions of friction factors and heat transfer coefficients in square channels varying the number of rough walls. This issue was tackled by investigating effects of different number of ribbed walls on heat transfer and friction characteristics in square channel. The rough wall had a $45^{\circ}$ inclined square rib. Uniform heat flux was maintained on the whole inner heat transfer channel area. The heat transfer coefficient and friction factor values increased with increasing the number of rough walls.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.35-42
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• 2018

This article presents clean core criticality calculations and control rod worth calculations for TRIGA (Training, Research, Isotope production-General Atomics) Mark II research reactor benchmark cores using Winfrith Improved Multi-group Scheme-D/4 (WIMS-D/4) and Program for Reactor In-core Analysis using Diffusion Equation (PRIDE) codes. Cores 133 and 134 were analyzed in 2-D (r, ${\theta}$) and 3-D (r, ${\theta}$, z), using WIMS-D/4 and PRIDE codes. Moreover, the influence of cross-section data was also studied using various libraries based on Evaluated Nuclear Data File (ENDF/B-VI.8 and VII.0), Joint Evaluated Fission and Fusion File (JEFF-3.1), Japanese Evaluated Nuclear Data Library (JENDL-3.2), and Joint Evaluated File (JEF-2.2) nuclear data. The simulation results showed that the multiplication factor calculated for all these data libraries is within 1% of the experimental results. The reactivity worth of the control rods of core 134 was also calculated with different homogenization approaches. A comparison was made with experimental and reported Monte Carlo results, and it was found that, using proper homogenization of absorber regions and surrounding fuel regions, the results obtained with PRIDE code are significantly improved.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.29 no.3 s.234
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• pp.340-348
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• 2005

Repeated ribs are used on heat exchange surfaces to promote turbulence and enhance convective heat transfer. Applications include fuel rods of gas-cooled nuclear reactors, inside cavities of turbine blades, and internal surfaces pipes used in heat exchangers. Despite the great number of literature papers, only few experimental data concern detailed distributions of friction factors and heat transfer coefficients in square channels varying the number of rough walls. This issue is tackled by investigating effects of different number of ribbed walls on heat transfer and friction characteristics in square channel. The rough wall have a $45{\circ}C$ inclined square rib. Uniform heat flux is maintained on whole inner heat transfer channel area. The heat transfer coefficient and friction factor values increase with increasing the number of rough walls.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.776-783
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• 2019

Laser welding is usually a more effective method than electron-beam one since a vacuum chamber is not required. It is important for joining Zr-1%Nb (E110) alloy in a manufacturing process of nuclear fuel rods. In the present work, effect of energy parameters of pulsed laser welding on properties of butt joints of sheets with a thickness of 0.5 mm is investigated. The most efficient combination has been found (8-11 J pulse energy, 10-14 ms pulse duration, 780-810 W peak pulse power, 3 Hz pulse frequency, 1.12 mm/s welding speed). The results show that ultimate strength under static loading can not be used as a quality criterion for zirconium alloys welds. Increased shielding gas flow rate does not allow to protect weld metal totally and contributes to defect formation without using special nozzles. Several types of imperfections of the welds have been found, but the major problem is branching microcracks on the surface of the welds. It is difficult to identify the cause of their appearance without additional research on improving the welding zone protection (gas composition and flow rate as well as nozzle configuration) and studying the hydrogen content in the welds.

• Journal of Radiation Protection and Research
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• v.47 no.4
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• pp.226-236
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• 2022

Background: An effective communication strategy for reducing conflicts in South Korea has been designed through the analysis of public perception and communication variables on nuclear power under the conditions of rapidly changing nuclear power policies. Materials and Methods: This study conducted both qualitative research through group discussions based on social psychology and quantitative research through surveys. Results and Discussion: Nuclear power plant (NPP) area residents in favor of nuclear power indicated higher levels of communication, safety perception, and contribution than those against it. NPP area residents trusted the civilian expert groups (18.3%) and local government (17.3%) the most, while metropolitan city residents trusted the Nuclear Safety and Security Commission and the Korea Institute of Nuclear Safety (20.7%) the most. In determining nuclear power policy, both the NPP area residents (18.1%) and metropolitan city residents (17.1%) prioritized safety, health, and the environment. While metropolitan city residents thought that energy security and economic growth (16.4%) were important, NPP area residents thought the current issue of spent fuel rods (14.1%) to be important. Conclusion: It is necessary for the nuclear power industry to have and actively implement communication and conflict resolution strategies based on the patterns obtained in the study results.