• Proceedings of the KIPE Conference
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• 2006.06a
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• pp.470-474
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• 2006

This dissertation discuses about a ferrite core plasma source using low operating frequency without sputtering problem by the stored electric field. Compared with the conventional RF power system with 13.56MHz switching frequency, the proposed plasma power system is only separated at 400kHz, so that it makes possible to use of low cost switching elements, PWM control and soft switching. Moreover, it could improve the coupling efficiency for plasma and antenna by using the ferrite core in order to transfer the energy of the load This dissertation tried to analyze new plasma generation method for the plasma generation system by modeling the plasma load and grafting the concept of impedance matching in order to interpret it with the formula This dissertation verified the ferrite core inductive coupling plasma source authorized for 400kHz of low frequency power by applying to the semi-conductor ash process thru the measurement of ash capacity and uniformed plasma distribution on the actual wafer.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2458-2473
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• 2024

The turbulent flow in reactor pressure vessel (RPV) of pressurized water reactor (PWR) is important for the flow rate distribution at core inlet. Thus, it is vital to study the turbulent flow phenomena in RPV. However, the complicated fluid channel consisted of inner structures of RPV will block or refract the laser sheet of particle image velocimetry (PIV). In this work, the matched index of refraction (MIR) of sodium iodide (NaI) solution and acrylic was applied to support optical path for flow field measurements by PIV in the 1/10th scaled-down RPV model. The experimental results show detailed velocity field at different locations inside the scaled-down RPV model. Some interesting phenomena are obtained, including the non-negligible counterflow at the corner of nozzle edge, the high downward flowing stream in downcomer, large vortices above vortex suppression plate in lower plenum. And the intensity of counterflow and the strength of vortices increase as inlet flow rate increasing. Finally, the case of asymmetry flow was also studied. The turbulent flow has different pattern compared with the case of symmetrical inlet flow rate, which may affect the uniformity of flow distribution at the core inlet.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.677-690
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• 2009

A previous study demonstrated that the two-row fuel assembly has much more favorable neutron-physical and thermal-hydraulic behavior than the conventional one-row fuel assemblies. Based on the newly developed two-row fuel assembly, an SCWR core is proposed and analyzed. The performance of the proposed core is investigated with 3-D coupled neutron-physical and thermal-hydraulic calculations. During the coupling procedure, the thermal-hydraulic behavior is analyzed using a sub-channel analysis code and the neutron-physical performance is computed with a 3-D diffusion code. This paper presents the main results achieved thus far related to the distribution of some neutronic and thermal-hydraulic parameters. It shows that with adjustment of the coolant and moderator mass flow in different assemblies, promising neutron-physical and thermal-hydraulic behavior of the SCWR core is achieved. A sensitivity study of the heat transfer correlation is also performed. Since the pin power in fuel assemblies can be non-uniform, a sub-channel analysis is necessary in order to investigate the detailed distribution of thermal-hydraulic parameters in the hottest fuel assembly. The sub-channel analysis is performed based on the bundle averaged parameters obtained with the core analysis. With the sub-channel analysis approach, more precise evaluation of the hot channel factor and maximum cladding surface temperature can be achieved. The difference in the results obtained with both the sub-channel analysis and the fuel assembly homogenized method confirms the importance of the sub-channel analysis.

• Journal of Energy Engineering
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• v.24 no.2
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• pp.154-166
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• 2015

The in-core instrumentation measures core power distribution and coolant temperature in local regions of the core in pressurized water reactors. The installation types are distinguished by the designs of routing paths that exit either through reactor bottom mounted instrument nozzles or through reactor top mounted instrument nozzles. Although each type has unique advantages, it is generally known that top mounted design is more competitive with respect to emphasizing nuclear safety issues and ability to cope with severe accidents. The international nuclear vendors have provided various types of reactors with top mounted design. Nuclear power reactors in Korea, however, only have been designed to be applicable to the use of bottom mounted design, and it has been pointed out that the capabilities of Korean reactors against severe accidents should be further enhanced. The paper deals with technical issues on reactor internal and external design, in-core instrumentation, support assembly, sealing mechanism with nozzles, handling, and analytical issues in order to establish the ways of development.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.179-184
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• 1996

Power density distribution and criticality of a pressurized water reactor are calculated with a Monte Carlo calculation using the MCNP code. The MCNP model is based on one-eighth core symmetry. Individual fuel assemblies are modeled with fullscope three dimensional description except grid spacer. The fuel rod is divided into eight axial segments. Core internals above and below the active fuel region is represented as coolant. After 400 cycle calculations, the system converges to a k value of 1.09151$\pm$0.00066. Fission reaction rate in each rod is also calculated to use as the source term in pressure vessel fluence calculation.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3388-3400
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• 2023

Detailed analysis of the neutron pathway through matter inside the nuclear reactor core is exceedingly needed for safety and economic considerations. Due to the constant development of high-performance computing technologies, neutronics analysis using computer codes became more effective and efficient to perform sophisticated neutronics calculations. In this work, a commercial pressurized water reactor (PWR) presented by Virtual Environment for Reactor Applications (VERA) Core Physics Benchmark are modeled and simulated using a high-fidelity simulation of OpenMC code in terms of criticality and fuel pin power distribution. Various problems have been selected from VERA benchmark ranging from a simple two-dimension (2D) pin cell problem to a complex three dimension (3D) full core problem. The development of the code capabilities for reactor physics methods has been implemented to investigate the accuracy and performance of the OpenMC code against VERA SCALE codes. The results of OpenMC code exhibit excellent agreement with VERA results with maximum Root Mean Square Error (RMSE) values of less than 0.04% and 1.3% for the criticality eigenvalues and pin power distributions, respectively. This demonstrates the successful utilization of the OpenMC code as a simulation tool for a whole core analysis. Further works are undergoing on the accuracy of OpenMC simulations for the impact of different fuel types and burnup levels and the analysis of the transient behavior and coupled thermal hydraulic feedback.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.59.1-59
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• 1999

• Journal of the Korean Institute of Electrical and Electronic Material Engineers
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• v.28 no.10
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• pp.670-675
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• 2015

Degradation diagnosis of cable is one of major issues for operation and maintenance in overhead distribution power lines. The diagnostic system for overhead power lines is composed of three parts in functional aspect - a travelling unit, a sensing unit and a communication unit. Among them, sensor detects the defects such as corrosion and disconnecting of power lines. Performance of sensor is very important, and besides, the size and structure of sensor is restricted for installation to small and lightweight diagnostic system. This paper suggests an optimal eddy current sensor best suit for small and lightweight diagnostic system in consideration of detecting performance, size and ease of installation and so on. Proposed sensor has been designed by Drum core structure and can be applied to the all domestic overhead power lines regardless of the cross-sectional areas. Also, it is showed that results of mock environmental test are satisfied.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.369-376
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• 2019

A complete solution for a soluble-boron-free (SBF) small modular reactor (SMR) is pursued with a new burnable absorber concept, namely centrally-shielded burnable absorber (CSBA). Neutronic flexibility of the CSBA design has been discussed with fuel assembly (FA) analyses. Major design parameters and goals of the SBF SMR are discussed in view of the reactor core design and three CSBA designs are introduced to achieve both a very low burnup reactivity swing (BRS) and minimal residual reactivity of the CSBA. It is demonstrated that the core achieves a long cycle length (~37 months) and high burnup (~30 GWd/tU), while the BRS is only about 1100 pcm and the radial power distribution is rather flat. This research also introduces a supplementary reactivity control mechanism using stainless steel as mechanical shim (MS) rod to obtain the criticality during normal operation. A further analysis is performed to investigate the local power peaking of the CSBA-loaded FA at MS-rodded condition. Moreover, a simple $B_4C$-based control rod arrangement is proposed to assure a sufficient shutdown margin even at the cold-zero-power condition. All calculations in this neutronic-thermal hydraulic coupled investigation of the 3D SBF SMR core are completed by a two-step Monte Carlo-diffusion hybrid methodology.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1126-1139
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• 2016

In this research, for the first time, a new optimization method, i.e., strength Pareto evolutionary algorithm II (SPEA-II), is developed for the burnable poison placement (BPP) optimization of a nuclear reactor core. In the BPP problem, an optimized placement map of fuel assemblies with burnable poison is searched for a given core loading pattern according to defined objectives. In this work, SPEA-II coupled with a nodal expansion code is used for solving the BPP problem of Kraftwerk Union AG (KWU) pressurized water reactor. Our optimization goal for the BPP is to achieve a greater multiplication factor ($K_{eff}$) for gaining possible longer operation cycles along with more flattening of fuel assembly relative power distribution, considering a safety constraint on the radial power peaking factor. For appraising the proposed methodology, the basic approach, i.e., SPEA, is also developed in order to compare obtained results. In general, results reveal the acceptance performance and high strength of SPEA, particularly its new version, i.e., SPEA-II, in achieving a semioptimized loading pattern for the BPP optimization of KWU pressurized water reactor.