• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.355.1-355
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.899-905
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• 1998

The analysis priority makings the recommendation to reduce the total core damage frequency (CDF) of Wolsong nuclear Power Plant nits 2/3/4 was Performed in this paper. In order to derive the recommendation, the sensitivity analysis of CDF on which major contributors effect m performed based on the accident quantification results during Level 1 Probabilistic safety assessment (PSA). Priorities were ranked in tile way that compares the CDF reduction rate with efforts required to implement those recommendations using risk matrix

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2670-2677
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• 2023

In the nuclear safety studies, a new trend refers to the evaluation of uncertainties as a mandatory component of best-estimate safety analysis which is a modern and technically consistent approach being known as BEPU (Best Estimate Plus Uncertainty). The major objectives of this study consist in performing a study of uncertainties/sensitivities of the major analysis results for a generic CANDU 6 Nuclear Power Plant during Station Blackout (SBO) progression to understand and characterize the sources of uncertainties and their effects on the key figure-of-merits (FOMs) predictions in severe accidents (SA). The FOMs of interest are hydrogen mass generation and event timings such as the first fuel channel failure time, beginning of the core disassembly time, core collapse time and calandria vessel failure time. The outcomes of the study, will allow an improvement of capabilities and expertise to perform uncertainty and sensitivity analysis with severe accident codes for CANDU 6 Nuclear Power Plant.

• JSTS:Journal of Semiconductor Technology and Science
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• v.12 no.3
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• pp.297-312
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• 2012

For future multi-core processors, 3D integration is regarded as one of the most promising techniques since it improves performance and reduces power consumption by decreasing global wire length. However, 3D integration causes serious thermal problems since the closer proximity of heat generating dies makes existing thermal hotspots more severe. Conventional air cooling schemes are not enough for 3D multi-core processors due to the limit of the heat dissipation capability. Without more efficient cooling methods such as liquid cooling, the performance of 3D multi-core processors should be degraded by dynamic thermal management. In this paper, we examine the architectural impact of cooling methods on the 3D multi-core processor to find potential benefits of liquid cooling. We first investigate the thermal behavior and compare the performance of two different cooling schemes. We also evaluate the leakage power consumption and lifetime reliability depending on the temperature in the 3D multi-core processor.

• Nuclear Engineering and Technology
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• v.47 no.2
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• pp.165-175
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• 2015

The application of neutron noise analysis (NNA) to the ex-core neutron detector signal for monitoring the vibration characteristics of a reactor core support barrel (CSB) was investigated. Ex-core flux data were generated by using a nonanalog Monte Carlo neutron transport method in a simulated CSB model where the implicit capture and Russian roulette technique were utilized. First and third order beam and shell modes of CSB vibration were modeled based on parallel processing simulation. A NNA module was developed to analyze the ex-core flux data based on its time variation, normalized power spectral density, normalized cross-power spectral density, coherence, and phase differences. The data were then analyzed with a fuzzy logic module to determine the vibration characteristics. The ex-core neutron signal fluctuation was directly proportional to the CSB's vibration observed at 8Hz and15Hzin the beam mode vibration, and at 8Hz in the shell mode vibration. The coherence result between flux pairs was unity at the vibration peak frequencies. A distinct pattern of phase differences was observed for each of the vibration models. The developed fuzzy logic module demonstrated successful recognition of the vibration frequencies, modes, orders, directions, and phase differences within 0.4 ms for the beam and shell mode vibrations.

• Journal of Power System Engineering
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• v.3 no.3
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• pp.22-28
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• 1999

The injected liquid does not break-up instantly after injection for diesel engine. There is some unbroken portion, which is the liquid core(The length of liquid core is called the break-up length) in the spray. If the liquid core is longer than the depth of the bowl in the small DI diesel engine, the liquid core impinges on the surface of the piston. Once the liquid core impinges on the surface, it cannot ignite or burn rapidly and thus prolongs burning time with a degradation in thermal efficiency. The break-up length of a diesel spray in a compressure vessel was measured by an electric resistance method, A voltage was applied between the nozzle and screen, bar, needle electrode inserted at various axial and radial positions into atomizing sprays. As a result, a current flows not only in the region of liquid core but also through the droplets of the spray. It is found that the break-up length measured with screen electrode is overestimated. The break-up length of the spray is found to be proportional to the square root of the density ratio of fuel and surrounding gas. The break-up length of the spray decreases as the injection pressure and the back pressure increase.

• 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.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.448-456
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• 2016

Severe accident scenarios in nuclear reactors, such as nuclear meltdown, reveal that an extremely hot molten core may fall into the nuclear reactor cavity and seriously affect the safety of the nuclear containment vessel due to the chain reaction caused by the reaction between the molten core and concrete. This paper reports on research focused on the type and amount of vapor produced during the reaction between a high-temperature molten core and concrete, as well as on the erosion rate of concrete and the heat transfer characteristics at its vicinity. This study identifies themass fraction and melting temperature as the most influential properties of concrete necessary for a safety analysis conducted in relation to the thermal interaction between the molten core and the basemat concrete. The types of concrete that are actually used in nuclear reactor cavities were investigated. The $H_2O$ content in concrete required for the computation of the relative amount of gases generated by the chemical reaction of the vapor, the quantity of $CO_2$ necessary for computing the cooling speed of the molten core, and the melting temperature of concrete are evaluated experimentally for the molten core-concrete interaction analysis.

• 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.

• KEPCO Journal on Electric Power and Energy
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• v.7 no.1
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• pp.21-24
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• 2021