• Nuclear Engineering and Technology
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• v.30 no.3
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• pp.273-286
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• 1998

As there has arisen a concern that failure of the high burnup fuel under the reactivity-insertion accident(RIA) may occur at the energy lower than the expected, fuel behavior under the rod ejection accident in a typical Westinghouse-designed 950 MWe PWR was analyzed by using the three dimensional nodal transient neutronics code, PANBOX2 and the transient fuel rod performance analysis code, FRAP-T6. Fuel failure criteria versus the burnup was conservatively derived taking into account available test data and the possible fuel failure mechanisms. The high burnup and longer cycle length fuel loading scheme of a peak rod turnup of 68 MWD/kgU was selected for the analysis. Except three dimensional core neutronics calculation, the analysis used the same core conditions and assumptions as the conventional zero dimensional analysis. Results of three dimensional analysis showed that the peak fuel enthalpy during the rod ejection accident is less than one third of that calculated by the conventional zero dimensional analysis methodology and the fraction of fuel failure in the core is less than 4 %. Therefore, it can be said that the current design limit of less than 10 percent fuel failure and maintaining the core coolable geometry would be adequately satisfied under the rod ejection accident, even though the conservative fuel failure criteria derived from the test data are applied.

• Proceedings of the KIEE Conference
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• 2006.07b
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• pp.875-876
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• 2006

Superconducting fault currents(SFCLs) are expected to improve not only reliability but also stability of power systems. The analysis on current limiting operations of the flux-lock type SFCL, which consists of a flux-lock reactor wound an iron core and a YBCO thin film, was compared the open-loop with the closed-loop iron core of the subtractive polarity winding. In the SFCL, operation characteristics could be controlled by adjusting the inductances and the winding directions of the coils, then magnetic field induced in the iron core. The current limiting characteristics under the same experimental conditions were generated regardless of the iron core conditions. We confirmed that capacity of the SFCL was increased effectively by the closed-loop iron core. However, the power burden of the system could be lowered by the open-loop iron core.

• Journal of KSNVE
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• v.9 no.4
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• pp.747-753
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• 1999

This paper presents results of vibration analysis of a generator-stator core for 500 MW fossil power plant. A finite element analysis using a commercial S/W is performed to estimate alternating electromagnetic forces, mainly of 120 Hz in 60 Hz machines, acting on the core, and then to calculate forced response of the core. Results are compared with design requirements.

• Journal of the Korean Magnetics Society
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• v.5 no.5
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• pp.395-398
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• 1995

For the core loss measurement under arbitrary waveform of magnetic induction, we have constructed a single sheet core loss measuring system which consists of yoke apparatus for single sheet of $10\;cm{\times}10\;cm$, arvitrary waveform synthesizer, B-feedback system, and two channel transient recorder. Using the constructed measuring system, we can measure core loss including higher harmonics up to 2 kHz. Core loss of non-oreinted electrical steel was increased exponentially when higher harmonic frequency was increased or amplitude of harmonic induction was increased.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.54-67
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• 2018

There have been recent efforts to establish methods for high-fidelity and multi-physics simulation with coupled thermal-hydraulic (T/H) and neutronics codes for the entire core of a light water reactor under accident conditions. Considering the computing power necessary for a pin-by-pin analysis of the entire core, subchannel-scale T/H analysis is considered appropriate to achieve acceptable accuracy in an optimal computational time. In the present study, the applicability of in-house code CUPID of the Korea Atomic Energy Research Institute was extended to the subchannel-scale T/H analysis. CUPID is a component-scale T/H analysis code, which uses three-dimensional two-fluid models with various closure models and incorporates a highly parallelized numerical solver. In this study, key models required for a subchannel-scale T/H analysis were implemented in CUPID. Afterward, the code was validated against four subchannel experiments under unheated and heated single-phase incompressible flow conditions. Thereafter, a subchannel-scale T/H analysis of the entire core for an Advanced Power Reactor 1400 reactor core was carried out. For the high-fidelity simulation, detailed geometrical features and individual rod power distributions were considered in this demonstration. In this study, CUPID shows its capability of reproducing key phenomena in a subchannel and dealing with the subchannel-scale whole core T/H analysis.

• Nuclear Engineering and Technology
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• v.52 no.10
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• pp.2173-2182
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• 2020

Control valves are widely used to regulate fluid flux in nuclear power plants, and there are more than 1500 control valves in the primary circuit of one nuclear power plant. With their help, the flux can be regulated to a specific level of water or steam to guarantee the energy efficiency and safety of the nuclear power plant. The flux characteristics of the control valve mainly depend on the valve core shape. In order to analyze the effects of valve core shapes on flux characteristics of control valves, this paper focuses on the valve core shapes. To begin with, numerical models of different valve core shapes are established, and results are compared with the ideal flux characteristics curve for the purpose of validation. Meanwhile, the flow fields corresponding to different valve core shapes are investigated. Moreover, relationships between the valve core opening and the outlet flux under different valve core shapes are carried out. The flux characteristics curve and equation are proposed to predict the outlet flux under different valve core openings. This work can benefit the further research of the flux control and the optimization of the valve core for control valves in nuclear power plants.

• The Journal of Korean Institute of Communications and Information Sciences
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• v.17 no.8
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• pp.816-822
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• 1992

In this paper, propagation characteristics of trapezoidal-shaped segmented core single mode fibers is investigated theoretically as a function of relative Index difference ratio( =p) under the condition of zero dispersion at i=1.,isrm, and bending loss of trape zoidalshaped segmented core single mode fiber is greatly decreased less than that of conventional single mode fibers ( triangular Index, dual shape core). And mode field distribution In core Is confined 2H% stronger than that of a tapezoidal Index fiber In addition, the advantages of trapezoldal-shaped segmented core fibers are compared with t hose of conventional triangular -shaped segmented core fibers.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.459-464
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• 1996

The 3-D power distribution synthesis scheme was implemented in Totally Integrated Core Operation Monitoring System (TICOMS), which is under development as the next generation core monitoring system. The on-line 3-D core power distribution obtained from the measured fixed incore detector readings is used to construct the hot pin power as well as the core average axial power distribution. The core average axial power distribution and the hot pin power of TICOMS were compared with those of the current digital on-line core monitoring system, COLSS, which construct the core average axial power distribution and the pseudo hot pin power. The comparison shows that TICOMS results in the slightly more accurate core average axial power distribution and the less conservative hot pin power. Therefore, these results increased the core operating margins. In addition, the on-line 3-D power distribution is expected to be very useful for the core operation in the future.

• Nuclear Engineering and Technology
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• v.32 no.5
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• pp.425-432
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• 2000

It has been an essential part for the safety assessment of nuclear power plants to understand various phenomena associated with the molten core-concrete interaction(MCCI) under severe accidents. In this study, the severe accident analysis code MELCOR was used to simulate the MCCI experiments such as SWISS and SURC test series which had been performed in Sandia National Laboratories(SNL). The calculation results were compared with corresponding experimental data such as melt temperature, concrete ablation distance, gas generation rate, and aerosol release rate. Good agreements were observed between MELCOR calculation and experimental data. The melt pool was sustained within the range of high temperature and the concrete ablation occurred continuously. The gas generation and aerosol release were under the influence of melt temperature and overlying water pool, respectively.

• Transactions of the Korean Society of Automotive Engineers
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• v.21 no.3
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• pp.165-173
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• 2013

Core loss has a major effect on heat generation in synchronous motors with surface-mounted permanent magnets (SPMs). It is essential to perform heat transfer analysis considering core loss in SPM because core loss is seriously affected by torque and speed of motors. In the present study, mechanical loss, core loss and coil loss are evaluated by measuring input and output energies under various driving conditions. For a better understanding heat transfer paths in synchronous motors, we developed a lumped thermal system analysis model. Subsequently, heat transfer analysis has been performed based on acquired energy loss, temperature data and thermal resistance with three types of SPM. It is shown that the torque constants decrease by Max. 10% as speed increase. At the rated torque, the core loss is Max. 10.9 times greater than the coil loss and the hysteresis loss of magnets is dominant in total loss.