• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1584-1602
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• 2024

It is vital to output the required electrical power following various task requirements when the space reactor power supply is operating in orbit. The dynamic performance of the closed Brayton cycle thermoelectric conversion system is initially studied and analyzed. Based on this, a load tracking power regulation method is developed for the liquid metal cooled space reactor power system, which takes into account the inlet temperature of the lithium on the hot side of the intermediate heat exchanger, the filling quantity of helium and xenon, and the input amount of the heat pipe radiator module. After comparing several methods, a power regulation method with fast response speed and strong system stability is obtained. Under various changes in power output, the dynamic response characteristics of the ultra-small liquid metal lithium-cooled space reactor concept scheme are analyzed. The transient operation process of 70 % load power shows that core power variation is within 30 % and core coolant temperature can operate at the set safety temperature. The second loop's helium-xenon working fluid has a 65K temperature change range and a 25 % filling quantity. The lithium at the radiator loop outlet changes by less than ±7 K, and the system's main key parameters change as expected, indicating safety. The core system uses less power during 30 % load power transient operation. According to the response characteristics of various system parameters, under low power operation conditions, the lithium working fluid temperature of the radiator circuit and the high-temperature heat pipe operation temperature are limiting conditions for low-power operation, and multiple system parameters must be coordinated to ensure that the radiator system does not condense the lithium working fluid and the heat pipe.

• Journal of the Institute of Electronics and Information Engineers
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• v.51 no.9
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• pp.215-220
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• 2014

Study and application result of power controller to research reactor is presented. Considering safety-oriented design concept and other control environment, we developed a simple closed-loop controller that provides limiting function of power-change-rate as well as low-overshoot and fine tracking performance. The design result has been well-proven via simulation and actual application to a research reactor.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1231-1240
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• 2019

Unlike land-based nuclear power plants, a marine or floating reactor is affected by external forces due to ocean conditions. These external forces can cause additional accelerations and affect each system and equipment of the marine reactor. Therefore, in designing a marine reactor and evaluating its performance and stability, a thermal hydraulic safety analysis code is necessary to consider the thermal hydrodynamic effects of ship motion. MARS, which is a reactor system analysis code, includes a dynamic motion model that can simulate the thermal-hydraulic phenomena under three-dimensional motion by calculating the body force term included in the momentum equation. In this study, it was verified that the dynamic motion model can simulate fluid motion with reasonable accuracy using conceptual problems. In addition, two modifications were made to the dynamic motion model; first, a user-supplied table to simulate a realistic ship motion was implemented, and second, the flow regime map determination algorithm was improved by calculating the volume inclination information at every time step if the dynamic motion model was activated. With these modifications, MARS could simulate the thermal-hydraulic phenomena under ocean motion more realistically.

• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.476-483
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• 2017

The decay heat that is produced by nuclear reactor spent fuel must be cooled in a spent fuel storage pool. A wickless heat pipe or a vertical two-phase closed thermosyphon (TPCT) is used to remove this decay heat. The objective of this research is to investigate the thermal performance of a prototype model for a large-scale vertical TPCT as a passive cooling system for a nuclear research reactor spent fuel storage pool. An experimental investigation and numerical simulation using RELAP5/MOD 3.2 were used to investigate the TPCT thermal performance. The effects of the initial pressure, filling ratio, and heat load were analyzed. Demineralized water was used as the TPCT working fluid. The cooled water was circulated in the water jacket as a cooling system. The experimental results show that the best thermal performance was obtained at a thermal resistance of $0.22^{\circ}C/W$, the lowest initial pressure, a filling ratio of 60%, and a high evaporator heat load. The simulation model that was experimentally validated showed a pattern and trend line similar to those of the experiment and can be used to predict the heat transfer phenomena of TPCT with varying inputs.

• Journal of computational fluids engineering
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• v.17 no.4
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• pp.1-8
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• 2012

During a reactor normal operation, two parallel 50% capacity cooling pumps circulate primary coolant to remove the fission reaction heat of the reactor through heat exchangers cold by a cooling tower. When one pump is failure, the other pump shall continuously circulate the coolant to remove the residual heat generated by the fuels loaded in the reactor after reactor shutdown. It is necessary to estimate how much flow rate will be supplied to remove the residual heat. We carried out a flow network analysis for the parallel primary pumps based on the piping network of the primary cooling system in HANARO. As result, it is estimated that the flow rate of one pump increased about 1.33 times the rated flow of one pump and was maintained within the limit of the cavitation critical flow.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.921-926
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• 2001

The reliability and accuracy of the information on control rod position are very important to the reactor safety and the design of the core protection system. In this study, a thorough investigation on the RSPT(Reed Switch Position Transmitter) type control rod position indication system and its actual implementation in the exiting nuclear power plants in Korea was performed first. A design of the control rod position indication system using reed switch for the CEDM on the system-integrated reactor SMART was developed based on the position indicator technology identified through the investigation. The feasibility of the design was evaluated by test of manufactured control rod position indicator using reed switch for SMART.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.657-658
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• 2004

The reactor protection system (RPS) of HANARO is a safety class system. The reactor is tripped by dropping four shut off rods (SOR). The SOR system consists of a SOR, hydraulic pump, hydraulic cylinder, solenoid valves and a power supply unit which has the AC coil contactor as a switching component. The hydraulic pump lifts up the SOR. The SOR drops by loss of the hydraulic pressure in the hydraulic circuit at the occurrence of voltage sags or interruptions. From this experiment, we knew that the magnitude of the voltage sag which impacts on this system is 70V, 500msec. The reactor regulation system (RRS) of HANARO has four CARs which are connected to the driver through a magnetic clutch. The CAR drops by loss of electromagnetic force of the magnetic clutch when the deeper voltage sags to lower than 10V, 500msec.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.37 no.12
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• pp.1175-1183
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• 2013

Even if some CFD software developers and its users think that a state-of-the-art CFD software can be used to reasonably solve at least single-phase nuclear reactor safety problems, there remain limitations and uncertainties in the calculation result. From a regulatory perspective, the Korea Institute of Nuclear Safety (KINS) is presently conducting the performance assessment of commercial CFD software for nuclear reactor safety problems. In this study, to examine the prediction performance of commercial CFD software with the porous model in the analysis of the scale-down APR (Advanced Power Reactor Plus) internal flow, a simulation was conducted with the on-board numerical models in ANSYS CFX R.14 and FLUENT R.14. It was concluded that depending on the CFD software, the internal flow distribution of the scale-down APR was locally somewhat different. Although there was a limitation in estimating the prediction performance of the commercial CFD software owing to the limited amount of measured data, CFX R.14 showed more reasonable prediction results in comparison with FLUENT R.14. Meanwhile, owing to the difference in discretization methodology, FLUENT R.14 required more computational memory than CFX R.14 for the same grid system. Therefore, the CFD software suitable to the available computational resource should be selected for massively parallel computations.

• Journal of the Korean Society of Safety
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• v.25 no.2
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• pp.24-28
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• 2010

The nuclear power Plant onsite AC electrical power sources are required to supply power to the engineering safety facility buses if the offsite power source is lost. Typically, Diesel Generators are used as the onsite power source. The 125 VAC buses are part of the onsite Class 1E AC and DC electrical power distribution system. The DC power distribution system ensure the availability of DC electrical power for system required to shutdown the reactor and maintain it in a safety condition after an anticipated operational occurrence or a postulated Design Base Accident. Recently, onsite DC power supply system trip occurs the loss of system function. To obtain the performance such as reliability and availability, we analyzed the cause of battery charger trip and described the improvement of DC power supply system reliability. Finally, we provide reliability performance criteria of charger in order to ensure the probabilistic goals for the safety of the nuclear power plants.

• 유체기계공업학회:학술대회논문집
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• 2005.12a
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• pp.509-514
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• 2005

Main coolant pump (MCP) is a canned-motor-type axial pump to circulate the primary coolant between nuclear fuel rods and steam generators in an integral type reactor. The reactor is designed to operate under condition of 310 oC and 14.7 MPa. Thus MCP has to be tested under same operating condition as reactor design condition in order to verify its performance and safety. In present work, a test loop to simulate real operating situation of the reactor has been designed and constructed to test MCP. And then, as a part of qualification test, canned motor functional test and pump hydraulic performance test have been carried out upon a prototype MCP. Canned motor efficiency and pump hydraulic characteristics including homologous curves and NPSH curves were obtained from the qualification test.