• Journal of ILASS-Korea
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• v.19 no.2
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• pp.64-68
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• 2014

A driving vehicle performance which is driven by FTP-75 mode was simulated by computer. Throttle valve position, engine speed, air mass flow rate, fuel consumption et al. were computer simulated. A set of engine part load performance data, automatic transmission shift map and vehicle specifications were used for the computer simulation. Throttle valve position, engine speed, air mass flow rate et al. measured for evaluating the computer simulation results by driving the vehicle with FTP-75 mode on a chassis dynamometer. GT-Power$^{(R)}$ software was used for the computer simulation of the driving vehicle performance. Experimental fuel consumption rate was measured by using an ECU HILS fuel injection system. The experimental data and simulation results were compared. The computer simulation of the driving vehicle performance predicts the measured data well comparatively.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1339-1345
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• 2017

In the framework of the EU funded project NURESAFE, the subchannel code CTF and the neutronics code DYN3D were integrated and coupled on the NURESIM platform. The developments achieved during this 3-year project include assembly-level and pin-by-pin multiphysics thermal hydraulics/neutron kinetics coupling. In order to test this coupling, a PWR rod ejection transient was simulated on a MOX/UOX minicore. The transient is simulated using two different models of the minicore. In the first simulation, both codes model the core with an assembly-wise resolution. In the second simulation, a pin-by-pin fuel-centered model is used in CTF for the central assembly, and a pin power reconstruction method is applied in DYN3D. The analysis shows the influence of the different models on global parameters, such as the power and the average fuel temperature, but also on local parameters such as the maximum fuel temperature.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2704-2710
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• 2024

Nuclear archaeology research provides scientific methods to reconstruct the operating histories of fissile material production facilities to account for past fissile material production. While it has typically focused on analyzing material in permanent reactor structures, spent fuel or high-level waste also hold information about the reactor operation. In this computational study, we explore a Bayesian inference framework for reconstructing the operational history from measurements of isotope ratios from a sample of nuclear waste. We investigate two different inference models. The first model discriminates between three potential reactors of origin (Magnox, PWR, and PHWR) while simultaneously reconstructing the fuel burnup, time since irradiation, initial enrichment, and average power density. The second model reconstructs the fuel burnup and time since irradiation of two batches of waste in a mixed sample. Each of the models is applied to a set of simulated test data, and the performance is evaluated by comparing the highest posterior density regions to the corresponding parameter values of the test dataset. Both models perform well on the simulated test cases, which highlights the potential of the Bayesian inference framework and opens up avenues for further investigation.

• Journal of the Korean Society of Manufacturing Process Engineers
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• v.12 no.3
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• pp.41-47
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• 2013

This study was carried out to improve the design of fuel injection nozzle for marine medium speed diesel engine. For this purpose, fuel injection nozzle was modeled and simulated using CATIA V5R19 and FLUENT & MSC Nastran. Analyses of flow and heat transfer, respectively, were performed to find the optimal design of fuel injection nozzle. As the results, big pressure drop, which may lead to cavitation damage, was occurred at inlet of fuel injection hole with diameter 0.3mm. Furthermore, it was confirmed that the increase of mean temperature of fuel injection nozzle was almost a half in comparison with that of fuel injection nozzle tip.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1148-1153
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• 2018

The Leaking Fuel Experiment test facility was designed to simulate the activity release from spent leaking fuel rods under steady state and transient conditions in the spent fuel pool. The experimental rig included an electrically heated fuel rod with different defects and a cooling system. The fission product transport was simulated by potassium-chloride. The conductivity changes of the water in the cooling system were measured to provide information about the amount of released solution. Defects of different sizes and positions were applied, together with a wide range of rod powers to simulate decay heat. The produced data can be used for predicting the activity release from leaking fuel under storage conditions and for the interpretation of fuel examination procedures.

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 2008.05a
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• pp.488-490
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• 2008

This study aims to design the high-capacity vol-oxidizer using simulated fuels instead of spent nuclear fuels. Simulated fuels are fabricated by blending tungsten powder with silicon carbide powder, and thereafter, paraffin coating covers simulated fuels to increase their strength. An oxidation experiment using simulated fuels have been carried out in order to analyze oxidation characteristics similar to spent fuels. After oxidation, simulated fuels were almost oxidized to be powders. Increased volume of simulated fuels approached to spent fuels. These results can be utilized as important informations for designing a high-capacity vol-oxidizer.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2007.11a
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• pp.421-422
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• 2007

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.973-979
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• 2024

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered. Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up. The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.1045-1046
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• 2004

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2009.06a
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• pp.757-758
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• 2009