• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3549-3558
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• 2023

The jet pump can be used in a test device of a nuclear reactor for high flow amplification as it reduces inlet flow requirement and thereby size of the process components. In the present work, a miniature jet pump was designed to meet high flow amplification greater than 3. Subsequently, experiments were carried out using a test setup for design validation and performance evaluation of the jet pump for different parameters. It was observed that a minimum pressure of 0.6 bar (g) was required for the secondary fluid inside the jet pump to ensure cavitation free performance at high amplification. Spacing between the nozzle tip and the mixing chamber entry point had significant effect on the performance of the jet pump. Variation in primary flow, temperature and area ratio also affected the performance. It was observed that at high flow amplification, the analytical solution differed significantly from experimental results due to very large velocities encountered in the miniature size jet pump.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.735-744
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• 2012

In order to quantify the flow distribution characteristics of APR+ reactor, a test was performed on a test facility, ACOP ($\underline{A}$PR+ $\underline{C}$ore Flow & $\underline{P}$ressure Test Facility), having a length scale of 1/5 referring to the prototype plant. The major parameters are core inlet flow and outlet pressure distribution and sectional pressure drops along the major flow path inside reactor vessel. To preserve the flow characteristics of prototype plant, the test facility was designed based on a preservation of major flow path geometry. An Euler number is considered as primary dimensionless parameter, which is conserved with a 1/40.9 of Reynolds number scaling ratio. ACOP simplifies each fuel assembly into a hydraulic simulator having the same axial flow resistance and lateral cross flow characteristics. In order to supply boundary condition to estimate thermal margins of the reactor, the distribution of inlet core flow and core exit pressure were measured in each of 257 fuel assembly simulators. In total, 584 points of static pressure and differential pressures were measured with a limited number of differential pressure transmitters by developing a sequential operation system of valves. In the current study, reactor flow characteristics under the balanced four-cold leg flow conditions at each of the cold legs were quantified, which is a part of the test matrix composing the APR+ flow distribution test program. The final identification of the reactor flow distribution was obtained by ensemble averaging 15 independent test data. The details of the design of the test facility, experiment, and data analysis are included in the current paper.

• Transactions of the KSME C: Technology and Education
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• v.3 no.4
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• pp.299-305
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• 2015

Nuclear hydrogen production technology is being developed for the future energy supply system. The sulfur-iodine thermo-chemical hydrogen production process directly splits water by using of the heat generated from very high temperature gas-cooled reactor, a typical Generation IV nuclear system. Nuclear hydrogen key technologies are composed of VHTR simulation technology at elevated temperature, computational tools, TRISO fuel, and sulfur iodine hydrogen production technology. Key technology for nuclear hydrogen production system were developed and demonstrated in a laboratory scale test facility. Technical challenges for the commercial hydrogen production system were discussed.

• Journal of Environmental Science International
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• v.19 no.5
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• pp.647-653
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• 2010

The objectives of this research are to evaluate and compare the oxygen transfer coefficients($K_{La}$) in both a general bubbles reactor and a micro-nano bubbles reactor for effective operation in sewage treatment plants, and to understand the effect on microbial kinetic parameters of biomass growth for optimal biological treatment in sewage treatment plants when the micro-nano bubbles reactor is applied. Oxygen transfer coefficients($K_{La}$) of tap water and effluent of primary clarifier were determined. The oxygen transfer coefficients of the tap water for the general bubbles reactor and micro-nano bubbles reactor were found to be 0.28 $hr^{-1}$ and 2.50 $hr^{-1}$, respectively. The oxygen transfer coefficients of the effluent of the primary clarifier for the general bubbles reactor and micro-nano bubbles reactor were found be to 0.15 $hr^{-1}$ and 0.91 $hr^{-1}$, respectively. In order to figure out kinetic parameters of biomass growth for the general bubbles reactor and micro-nano bubbles reactor, oxygen uptake rates(OURs) in the saturated effluent of the primary clarifier were measured with the general bubbles reactor and micro-nano bubbles reactor. The OURs of in the saturated effluent of the primary clarifier with the general bubbles reactor and micro-nano bubbles reactor were 0.0294 mg $O_2/L{\cdot}hr$ and 0.0465 mg $O_2/L{\cdot}hr$, respectively. The higher micro-nano bubbles reactor's oxygen transfer coefficient increases the OURs. In addition, the maximum readily biodegradable substrate utilization rates($K_{ms}$) for the general bubbles reactor and micro-nano bubbles reactor were 3.41 mg COD utilized/mg active VSS day and 7.07 mg COD utilized/mg active VSS day, respectively. The maximum specific biomass growth rates for heterotrophic biomass(${\mu}_{max}$) were calculated by both values of yield for heterotrophic biomass($Y_H$) and the maximum readily biodegradable substrate utilization rates($K_{ms}$). The values of ${\mu}_{max}$ for the general bubbles reactor and micro-nano bubbles reactor were 1.62 $day^{-1}$ and 3.36 $day^{-1}$, respectively. The reported results show that the micro-nano bubbles reactor increased air-liquid contact area. This method could remove dissolved organic matters and nutrients efficiently and effectively.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.02a
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• pp.114-129
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• 2004

The 3D simulations for the Rotary Specimen Rack (RSR), the shielding concret, and the reactor core dismantling processes in the Korea Research Reactor-1&2(KRR-1&2) were carried out in the present work. The four main dismantling items, which are the RSR, reactor core, beam tube, and the thermal column and the shield concrete, were selected among the many components in the KRR-2 by consideration of the activation, worker training, difficulty of the work and so on. On the basis of these, we built 3D CAD models, selected the proper dismantling technologies, and reviewed their dismantling processes. In this study, the 3D simulation results of the shielding concrete, and the reactor core dismantling processes are also presented and discussed.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.290-295
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• 1996

The purpose of this study is to establish the pressure/temperature curves of Reactor Coolant System for brittle fracture prevention. The pressure/temperature curve is the basis to select RC Pump and limits to operate the plant. Based on the plant operation experience, this curve should be re-generated periodically in order to ensure the structural integrity using data from the test of reactor vessel surveilance materials to compensate for the irradiation effects. This study provides the procedure of pressure/temperature curve generation in term of brittle fracture prevention of reactor vessel. Using the UCN 3&4 data, the sample pressure/temperature curve was generated, and it was compared with those of YGN 3&4 based on the stress and $RT_{NDT}$value.

• Journal of Applied Reliability
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• v.10 no.1
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• pp.57-64
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• 2010

The HANARO nuclear research reactor had been operated successfully and smoothly up to the year 2008 since its first year of instability in year 1995 just after the completion of construction. But the reliability of the reactor has been degraded from the year 2009 due to new experimental facilities such as Feul Test Loop(FTL) and Cold Neutron Source(CNS) which were installed in the HANARO plant. It turned out that these new facilities contributed unexpected stoppage of the plant. This paper describes causes of stoppage and suggestions to improve the reliability of the plant.

• Nuclear Engineering and Technology
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• v.32 no.2
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• pp.99-107
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• 2000

A probabilistic integrity analysis method is presented for a reactor vessel under pressurized thermal shock(PTS) based on Monte Carlo simulation. This method can be applied to the structural integrity assessment of a reactor vessel subjected to pressurized thermal shock where the coolant temperature transient cannot be expressed explicitly as a time function. An axially or circumferentially oriented infinite length surface crack is assumed to be in the beltline weld region of the rector vessel's inside surface. The random variables are the initial crack depth, neutron fluence on the vessel's inside surface, the copper and nickel content of the vessel materials, R $T_{NDT}$ , $K_{IC}$ , and K/aub la/. The reliability of a sample reactor vessel under PTS is assessed quantitatively and the influence of the amount of neutron fluence is also examined by applying the present method.sent method.

• Journal of the Korea Safety Management & Science
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• v.8 no.3
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• pp.67-75
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• 2006

Scheduling of dismantling old research reactor need to consider time, cost and safety for the worker. The biggest issue when dismantling facility for research reactor is safety for the worker and cost. Large portion of a budget is spending for the labor cost. To save labor cost for the worker, reducing a lead time is inevitable. Several algorithms applied to reduce read time, and safety considered as the most important factor for this project. This research presents three different dismantling scheduling scenarios. Best scenario shows the specific scheduling for worker and machine, so that it could save time and cost.

• Proceedings of the KIEE Conference
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• 2007.07a
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• pp.1626-1627
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• 2007

This paper deals with the design and manufacturing of power cabinet for reactor power control system(PCS). The PCS provides the control signals and motive power to operate the CEDMs(Control Element Drive Mechanism). The CEDM is raise and lower the CEAs(Control Element Assemblies) in the reactor core. The CEAs are constructed with the Boron-10 isotope which has a high microscopic cross section of absorption for thermal neutrons. This characteristic causes the addition of negative reactivity when a CEA is inserted and positive reactivity when it is withdrawn from the reactor core.