• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2315-2324
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• 2023

Space nuclear reactors are becoming popular in deep space exploration owing to their advantages of high-power density and stability. Following the fourth-generation nuclear reactor technology, a conceptual design of the dual drum-controlled space molten salt reactor (D²-SMSR) is proposed. The reactor concept uses molten salt as fuel and heat pipes for cooling. A new reactivity control strategy that combines control drums and safety drums was adopted. Critical physical characteristics such as neutron energy spectrum, neutron flux distribution, power distribution and burnup depth were calculated. Flow and heat transfer characteristics such as natural convection, velocity and temperature distribution of the D²-SMSR under low gravity conditions were analyzed. The reactivity control effect of the dual-drums strategy was evaluated. Results showed that the D²-SMSR with a fast spectrum could operate for 10 years at the full power of 40 kWth. The D²-SMSR has a high heat transfer coefficient between molten salt and heat pipe, which means that the core has a good heat-exchange performance. The new reactivity control strategy can achieve shutdown with one safety drum or three control drums, ensuring high-security standards. The present study can provide a theoretical reference for the design of space nuclear reactors.

• Journal of the Korean Society of Safety
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• v.26 no.5
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• pp.46-53
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• 2011

The risk assessment of thermal behavior and runaway reaction cased by an exothermic batch process in manufacture of the vinyl acetate resin are described in the present paper. The aim of the study was to evaluate the risk of runaway reaction with operating parameters such as a reaction inhibitor, reaction temperature and a mount of methanol charged in the vinyl acetate polymerization process. The experiments were performed by a sort of calorimetry with the Multimax reactor system as a screening tool to investigate runaway reaction. From the experimental results, it was found that we could occur the auto acceleration for reaction of raw materials with operating parameters over $65^{\circ}C$ of reaction temperature in the vinyl acetate polymerization process.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2716-2727
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• 2021

The core shroud is one of the most important internal components of the reactor vessel internals because it meets the neutron fluence directly emitted by the nuclear fuel. In particular, dynamic effects for an earthquake should be evaluated with respect to the neutron irradiation flux. As a prerequisite to this study, simplified and detailed finite element models are developed for the core shroud using the ANSYS Design Parametric Language. Using the El Centro earthquake, seismic analyses are performed for the simplified and detailed core shroud models. Modal characteristics are obtained and their results are used for a time history analysis. Response spectrum analyses are also performed to access the degree of seismic excitation. The results of these analyses are compared to investigate the response characteristics between the simplified and detailed core shroud models from the time history and response spectrum analyses.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.759-766
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• 2013

As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop safety systems for 3rd-generation (GEN-III) nuclear power plants that are driven by passive systems. The Passive Auxiliary Feedwater System (PAFS) is one of several passive safety systems being designed for the Advanced Power Reactor Plus (APR+), and extensive studies are being conducted to complete its design and to verify its feasibility. Because the PAFS removes decay heat from the reactor core under transient and accident conditions, it is necessary to evaluate the heat removal capability of the PAFS under hypothetical accident conditions. The heat removal capability of the PAFS is strongly dependent on the heat transfer at the condensate tube in Passive Condensation Heat Exchanger (PCHX). To evaluate the model of heat transfer coefficient for condensation, the Multi-dimensional Analysis of Reactor Safety (MARS) code is used to simulate the experimental results from PAFS Condensing Heat Removal Assessment Loop (PASCAL). The Shah model, a default model for condensation heat transfer coefficient in the MARS code, under-predicts the experimental data from the PASCAL. To improve the calculation result, The Thome model and the new version of the Shah model are implemented and compared with the experimental data.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.685-686
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.560-561
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.381-382
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• 2004

• Structural Engineering and Mechanics
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• v.35 no.1
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• pp.99-117
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• 2010

Beam-column joints are the key structural elements, which dictate the behavior of structures subjected to earthquake loading. Though large experimental work has been conducted in the past, still various issues regarding the post-yield behavior, ductility and failure modes of the joints make it a highly important research topic. This paper presents experimental results obtained for eight beam-column joints of different sizes and configuration under cyclic loads along with the analytical evaluation of their response using a simple and effective analytical procedure based on nonlinear static pushover analysis. It is shown that even the simplified analysis can predict, to a good extent, the behavior of the joints by giving the important information on both strength and ductility of the joints and can even be used for prediction of failure modes. The results for four interior and four exterior joints are presented. One confined and one unconfined joint for each configuration were tested and analyzed. The experimental and analytical results are presented in the form of load-deflection. Analytical plots are compared with envelope of experimentally obtained hysteretic loops for the joints. The behavior of various joints under cyclic loads is carefully examined and presented. It is also shown that the procedure described can be effectively utilized to analytically gather the information on behavior of joints.

• Nuclear Engineering and Technology
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• v.40 no.4
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• pp.255-268
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• 2008

The greater demand for electricity and the available capacity within safety margins in some operating NPPs are prompting nuclear utilities to request license modification to enable operation at a higher power level, beyond their original license provisions. Such plant modifications require an in-depth safety analysis to evaluate the possible safety impact. The analysis must consider the thermo hydraulic, radiological and structural aspects, and the plant behavior, while taking into account the capability of the structures, systems and components, and the reactor protection and safeguard systems set points. The purpose of this paper is to introduce international experiences and approaches for implementation of power uprates related to the reactor thermal power of nuclear power plants. The paper is intended to give the reader a general overview of the major processes, work products, issues, challenges, events, and experiences in the power uprates program. The process of increasing the licensed power level of a nuclear power plants is called a power uprate. One way of increasing the thermal output from a reactor is to increase the amount of fissile material in use. It is also possible to increase the core power by increasing the performance of the high power bundles. Safety margins can be maintained by either using fuels with a higher performance, or through the use of improved methods of analysis to demonstrate that the required margins are retained even at the higher power levels. The paper will review all types of power uprates, from small to large, and across various reactor types, including light and heavy water, pressurized, and boiling water reactors. Generally, however, the content of the report focuses on power uprates of the stretch and extended type. The International Atomic Energy Agency (IAEA) is developing a technical guideline on power uprates and side effects of power uprates in nuclear power plants.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.38 no.3
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• pp.271-277
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• 2014

Internal structures, especially those located in the upstream of a reactor core, may have a significant influence on the core inlet flow rate distribution depending on both their shapes and the relative distance between the internal structures and the core inlet. In this study, to examine the effect of the reactor internal structure geometry treatment method on the prediction accuracy for the scale-down APR+ flow distribution, simulations with real geometry modeling were conducted using ANSYS CFX R.14, a commercial computational fluid dynamics software, and the predicted results were compared with those of the porous medium assumption. It was concluded that the core inlet flow distribution could be predicted more accurately by considering the real geometry of the internal structures located in the upstream of the core inlet. Therefore, if sufficient computational resources are available, an exact representation of these internal structures, for example, lower support structure bottom plate and ICI nozzle support plate, is needed for the accurate simulation of the reactor internal flow.