• Journal of Korean Society of Industrial and Systems Engineering
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• v.47 no.3
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• pp.8-17
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• 2024

There is a demand for introducing a challenging and innovative R&D system to develop new technologies to generate weapon system requirements. Despite the increasing trend in annual core technology development tasks, the infrastructure expansion, including personnel in research management institutions, is relatively insufficient. This situation continuously exposes difficulties in task planning, selection, execution, and management. Therefore, there is a pressing need for strategies to initiate timely research and development and enhance budget execution efficiency through the streamlining of task agreement schedules. In this study, we propose a strategic model utilizing a flexible workforce model, considering constraints and optimizing workload distribution through resource allocation to minimize bottlenecks for efficient task agreement schedules. Comparative analysis with the existing operational environment confirms that the proposed model can handle an average of 67 more core technology development tasks within the agreement period compared to the baseline. In addition, the risk management analysis, which considered the probabilistic uncertainty of the fluctuating number of core technology research and development projects, confirmed that up to 115 core technology development can be contracted within the year under risk avoidance.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.937-946
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• 2020

The aim in the present work is to simulate accident scenarios of AP1000 during the small-break loss-of-coolant accident (SBLOCA) and investigate the performance and behavior of automatic depressurization system (ADS) during accidents by using MIDAC (The Module In-vessel Degradation severe accident Analysis Code). Four types of accidents with different hypothetical conditions were analyzed in this study. The impact on the thermal-hydraulic of the reactor coolant system (RCS), the passive core cooling system and core degradation was researched by comparing these types. The results show that the RCS depressurization becomes faster, the core makeup tanks (CMT) and accumulators (ACC) are activated earlier and the effect of gravity water injection is more obvious along with more ADS valves open. The open of the only ADS1-3 can't stop the core degradation on the basis of the first type of the accident. The open of ADS1-3 has a great impact on the injection time of ACC and CMT. The core can remain intact for a long time and the core degradation can be prevent by the open of ADS-4. The all results are significant and meaningful to understand the performance and behavior of the ADS during the typical SBLOCA.

• Proceedings of the SAREK Conference
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• 2004.06a
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• pp.202-202
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• 2004

• Proceedings of the Korean Society of Potoscience Conference
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• 2005.06a
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• pp.42-42
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• 2005

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.999-1008
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• 2023

Deformed fuel rods can cause a partial blockage of the flow area in a subchannel. Such flow blockage will influence the core coolant flow and further the core heat transfer during the reflooding phase and subsequent severe accidents. Nevertheless, most of the system analysis codes simulate the accident process based on the assumed flow blockage ratio, resulting in inconsistencies between simulated results and actual conditions. This paper aims to study the influence of flow blockage in severe accident scenario of the CAP1400 reactor. First, the flow blockage model of ISAA code is improved based on the FRTMB module. Then, the ISAA-FRTMB coupling system is adopted to model and calculate the QUENCH-LOCA-0 experiment. The correctness and validity of the flow blockage model are verified by comparing the peak cladding temperature. Finally, the DVI Line-SBLOCA accident is induced to analyze the influence of flow blockage on subsequent CAP1400 reactor core heat transfer and core degradation. From the results of the DVI Line-SBLOCA accident analysis, it can be concluded that the blockage ratio is in the range of 40%-60%, and the position of severe blockage is the same as that of cladding rupture. The blockage reduces the circulation area of the core coolant, which in turn impacts the heat exchange between the core and the coolant, leading to the early failure and collapse of some core assemblies and accelerating the core degradation process.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.561-574
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• 2009

First ex-vessel core catcher has been applied to the practical design of NPPs with VVER-1000 reactors built in China (Tyanvan) and India (Kudankulam) for severe accident management (SAM) and mitigation of SA consequences. The paper presents the concept and basic design of this crucible-type core catcher as well as an evaluation of its efficiency. The important role of oxidic sacrificial material is discussed. Insight into the behaviour of the molten pool, which forms in the catcher after core relocation from the reactor vessel, is provided. It is shown that heat loads on the water-cooled vessel walls are kept within acceptable limits and that the necessary margins for departure from nucleate boiling (DNB) and of vessel failure caused by thermo-mechanical stress are satisfactorily provided for.

• IE interfaces
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• v.21 no.1
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• pp.96-108
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• 2008

This paper proposes the systematic design approach for developing the evaluation indicators that evaluate the candidate proposals in the national defense core-technology R&D projects. To improve the validity and fairness of the evaluation indicators, the existing evaluation process in a military and a public sector are surveyed and also the existing evaluation system of the core-technology R&D programs for the national defense is analysed and discussed. A new system for the evaluation indicators is designed by using the axiomatic design, factor analysis and the analytic hierarchy process. It is expected that the proposed evaluation indicators contribute to enhance the fairness and the reliability of the evaluation process for the proposal of the national defense core-technology R&D projects.

• Design & Manufacturing
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• v.8 no.2
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• pp.41-45
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• 2014

Automated polishing apparatus based on the research have been developed. The research is improvement of polishing process for surface quality and uniformity improvement of preform injection mold core. Surface quality of preform core have influence on ejecting and product quality after injection molding. Thus, the current being made by hand to automate the polishing process, the surface of the preform to improve the quality and uniformity improvement. First made a division by analyzing manual process a step-by-step. And draw a mechanism for converting mechanical movement. Automated polishing apparatus for preform core was developed, through which shortens production time and were able to secure the safety of the worker.

• Journal of Magnetics
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• v.19 no.2
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• pp.146-150
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• 2014

The objective of this paper is to provide a comparison between two transverse flux rotary machines (TFRM) with different topologies of stator cores. Depending on how to make stator core with laminated steel sheets, the one topology is 'perpendicular stacking core' and the other is 'separated core'. Both of the two cores have been designed considering 3-dimensional (3-D) magnetic flux path with the same output power conditions, but the core losses are quite different and it causes different magnetic and thermal characteristics. For comparison of these two topologies of stator cores, therefore, core losses have been calculated and used as a heat source in no-load conditions, and the thermal stress has been also calculated. 3-D finite element method has been used for the magnetic field, thermal, and stress analysis to consider the 3-D flux path of the TFRM. After comparing the analysis results of the two topologies, experimental results are also presented and discussed.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.878-888
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• 2020

This paper presents the validation of UNIST in-house Monte Carlo code MCS used for the high-fidelity simulation of commercial pressurized water reactors (PWRs). Its focus is on the accurate, spatially detailed neutronic analyses of startup physics tests for the initial core of the Watts Bar Nuclear 1 reactor, which is a vital step in evaluating core phenomena in an operating nuclear power reactor. The MCS solutions for the Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) core physics benchmark progression problems 1 to 5 were verified with KENO-VI and Serpent 2 solutions for geometries ranging from a single-pin cell to a full core. MCS was also validated by comparing with results of reactor zero-power physics tests in a full-core simulation. MCS exhibits an excellent consistency against the measured data with a bias of ±3 pcm at the initial criticality whole-core problem. Furthermore, MCS solutions for rod worth are consistent with measured data, and reasonable agreement is obtained for the isothermal temperature coefficient and soluble boron worth. This favorable comparison with measured parameters exhibited by MCS continues to broaden its validation basis. These results provide confidence in MCS's capability in high-fidelity calculations for practical PWR cores.