• Journal of international Conference on Electrical Machines and Systems
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• v.3 no.3
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• pp.282-288
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• 2014

A large capacity power conversion system constructed by using two or more existing power converters has a lot of flexibility in how the power converters are used. However, at the same time, it has a problem of cross current flows between power converters. The cross current must be suppressed by controlling the system while miniaturizing the combination reactor. This paper focuses on two current control methods of a power conversion system constructed by using two power converters connected in parallel supplying the same power. In order to elucidate the control performance of cross current, each control method which are aimed at controlling cross current and not directly controlling it are examined in simulations.

• The Transactions of the Korean Institute of Electrical Engineers
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• v.33 no.3
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• pp.91-99
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• 1984

Time-optimal control for dmping a one-dimensional xenon-induced spatial power oscillations in pressurized water reactor is studied. Linearized system equations describing the spatial xenon oscillations have been derived based on lambda mode analysis. Optimal control strategies, eventually bang-bang controls, have been drawn applying Pontryagins Minimum Principle, subject to a band constraint on available contros strength. Validity of the linearized system equations and optimal control strategies derived has been demonstrated through conputer simulations which incorporate the finite difference method for one dimensional axial geometry, for the soulution of the two-group neutron diffusion equations. The results obtained through computer simulations show that xenon-induced transients can be suppressed successfully with bang-bang control.

• Journal of Environmental Health Sciences
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• v.24 no.2
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• pp.9-19
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• 1998

Laboratory experiments were carried out to investigate the performance of anaerobic sequencing batch reactor(ASBR) for digestion of a municipal sludge. Each cycle of the ASBR comprised feeding, two-or three-day reaction, one-day thickening, and withdrawal. The reactors were operated at an HRT of 10days and 5days with an equivalent organic loading rate of 0.8-1.54 gVS/l/d, 1.81-3.56 gVS/l/d at 35$\circ$C, respectively. Solids accumulation was remarkable in the ASBR during start-up period, and directly affected by settleable solids in the feed sludge. Floatation thickening occured in the ASBRs, and Solids profiles at the end of thickening step dramatically changed at solid-liquid interface. Slight difference in solids concentrations was observed within thickened sludge bed. Efficiencies through floatation thickening were comparable to that of additional thickening of the completely mixed control reactor. Average solids concentrations in the ASBRs were 2.2-2.6 times higher than that in the control throughout the total operation period. The dehydrogenase activity had a strong correlation with the solids concentration. Organics removals based on clarified effluent of the ASBRs were consistently above 86%. Remarkable increase in equivalent gas production of 27-52% was observed at the ASBRs compared with the control though the control and ASBRs showed similiar effluent quality. Thus, digestion of a municipal sludge was possible using the ASBR in spite of high concentration of solids in the sludge.

• Journal of Radiation Protection and Research
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• v.14 no.1
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• pp.34-45
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• 1989

The habitability of a reactor control room in a French 1300 MWe P'4 type PWR has been evaluated through the exposure dose assessment for the reactor operator following a Loss of Coolant Accident. The main hypotheses adopted in this evaluation are based on the French Standard Safety Analysis Report. A simple computer program, named COREX(Control Room EXposure), was developed to calculate : the time-integrated radioactivities released from the reactor building, the volume factors for radionuclides concerned and the resulting time-integrated external whole body and internal thyroid doses to the reactor operators staying in the control room up to 30 days following the LOCA. The results obtained in this study, on the whole, well agreed with those proposed by the EDF(Electricite de France) except for the case of the whole body exposure, which was attributed to the differences in the volume factors for the radionuclides concerned.

• 한국신재생에너지학회:학술대회논문집
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• 2010.11a
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• pp.133.2-133.2
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• 2010

Fischer-Tropsch synthesis (FTS) was carried out in heat-exchanger typed reactor with cobalt metallic foam catalyst. Considering the heat and mass transfer limitations in the cobalt catalyst, a Co-foam catalyst with an inner metallic foam frame and an outer cobalt catalyst was developed. The Co-foam catalyst was highly selective toward liquid hydrocarbon production and the liquid hydrocarbon productivity at $203^{\circ}C$ reached to $52.5ml/(kg_{cat}{\cdot}h)$, which was higher than that obtained by the Co-pellet. Furthermore, the heat-exchanger typed reactor was developed to efficiently control the highly exothermic reaction heat. The reaction heat generated in the FTS reaction on the cobalt active site was easily transferred to reactor wall by the metallic foam in the catalyst and the transferred reaction heat was directly removed by the hot oil which circulated the wall side of the heat-exchanger typed reactor.

• 제어로봇시스템학회:학술대회논문집
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• 2004.08a
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• pp.1116-1120
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• 2004

In the nuclear power plant, there are several cylindrical vessels such as reactor vessel, pressuriser and so on. The vessels are usually constructed by welding large rolled plates, forged sections or nozzle pipes together. In order to assure the integrity of the vessel, these welds should be periodically inspected using sensors such as ultrasonic transducer or visual cameras. This inspection is usually conducted under water to minimize exposure to the radioactively contaminated vessel walls. The inspections have been performed by using a conventional inspection machine with a big structural sturdy column, however, it is so huge and heavy that maintenance and handling of the machine are extremely difficult. It requires much effort to transport the system to the site and also requires continuous use of the utility's polar crane to move the manipulator into the building and then onto the vessel. Setup beside the vessel requires a large volume of work preparation area and several shifts to complete. In order to resolve these problems, we have developed an underwater mobile robot guided by the laser pointer, and performed a series of experiments both in the mockup and in the real reactor vessel. This paper introduces our robotic inspection system and the laser guidance of the mobile robot as well as the results of the functional test.

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.3175-3180
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• 2008

A metal-fueled pool-type liquid metal fast reactor (LMFR) provides large margins to sodium boiling and fuel damage under accident conditions. The favorable passive safety results are obtained by both a reactivity feedback mechanism in the core and a passive decay heat removal system. Among the various reactivity feedbacks, the ones by a thermal expansion of a radial dimension of the core and by the control rod drivelines are strongly dependent on the flow conditions in the core and the hot pool, respectively. The effects of multidimensional thermal hydraulic characteristics on these reactivity feedbacks are investigated by the system-wide safety analysis code SSC-K with advanced thermal hydraulics models. Particularly a detailed three dimensional thermal hydraulics reactor core model is integrated into SSC-K for use in a whole system analysis of the passive safety aspects of LMR designs. The model provides fuel and cladding temperatures for every fuel pin in a reactor and coolant temperatures for every coolant sub-channel in the reactor.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4560-4570
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• 2022

Dynamic fault tree (DFT) and its related research methods have received extensive attention in safety analysis and reliability engineering. DFT can perform reliability modelling for systems with sequential correlation, resource sharing, and cold and hot spare parts. A technical modelling method of DFT is proposed for modelling ship collision accidents and loss-of-coolant accidents (LOCAs). Qualitative and quantitative analyses of DFT were carried out using the cutting sequence (CS)/extended cutting sequence (ECS) method. The results show nine types of dynamic fault failure modes in ship collision accidents, describing the fault propagation process of a dynamic system and reflect the dynamic changes of the entire accident system. The probability of a ship collision accident is 2.378 × 10^-9 by using CS. This failure mode cannot be expressed by a combination of basic events within the same event frame after an LOCA occurs in a marine nuclear reactor because the system contains warm spare parts. Therefore, the probability of losing reactor control was calculated as 8.125 × 10^-6 using the ECS. Compared with CS, ECS is more efficient considering expression and processing capabilities, and has a significant advantage considering cost.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1644-1651
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• 2022

This paper focuses on the power-level control of nuclear power plants (NPPs) in the presence of lumped disturbances. An adaptive second-order nonsingular terminal sliding mode control (ASONTSMC) scheme is proposed by resorting to the second-order nonsingular terminal sliding mode. The pre-existing mathematical model of the nuclear reactor system is firstly described based on point-reactor kinetics equations with six delayed neutron groups. Then, a second-order sliding mode control approach is proposed by integrating a proportional-derivative sliding mode (PDSM) manifold with a nonsingular terminal sliding mode (NTSM) manifold. An adaptive mechanism is designed to estimate the unknown upper bound of a lumped uncertain term that is composed of lumped disturbances and system states real-timely. The estimated values are then added to the controller, resulting in the control system capable of compensating the adverse effects of the lumped disturbances efficiently. Since the sign function is contained in the first time derivative of the real control law, the continuous input signal is obtained after integration so that the chattering effects of the conventional sliding mode control are suppressed. The robust stability of the overall control system is demonstrated through Lyapunov stability theory. Finally, the proposed control scheme is validated through simulations and comparisons with a proportional-integral-derivative (PID) controller, a super twisting sliding mode controller (STSMC), and a disturbance observer-based adaptive sliding mode controller (DO-ASMC).

• Journal of Korean Society of Industrial and Systems Engineering
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• v.39 no.2
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• pp.113-118
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• 2016

Project management is a tool for smooth operation during a full cycle from the design to normal operation including the schedule, document, and budget management, and document management is an important work for big projects such as the JRTR (Jordan Research and Training Reactor). To manage the various large documents for a research reactor, a project management system was resolved, a project procedure manual was prepared, and a document control system was established. The ANSIM (Advanced Nuclear Safety Information Management) system consists of a document management folder, document container folder, project management folder, organization management folder, and EPC (Engineering, Procurement and Construction) document folder. First, the system composition is a computerized version of the Inter-office Correspondence (IOC), the Document Distribution for Agreement (DDA), Design Documents, and Project Manager Memorandum (PM Memo) works prepared for the research reactor design. Second, it reviews, distributes, and approves design documents in the system and approves those documents to register and supply them to the research reactor user. Third, it integrates the information of the document system-using organization and its members, as well as users' rights regarding the ANSIM document system. Throughout these functions, the ANSIM system has been contributing to the vitalization of united research. Not only did the ANSIM system realize a design document input, data load, and search system and manage KAERI's long-period experience and knowledge information properties using a management strategy, but in doing so, it also contributed to research activation and will actively help in the construction of other nuclear facilities and exports abroad.