• Proceedings of the KIEE Conference
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• 2002.06a
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• pp.38-40
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• 2002

In the case of conventional high frequency inverter, with damage of switch by surge voltage when switch gets into compulsion extinction by load accident and so on because reactor is connected by series to switch, or there was problem of conduction loss by reactor's resistivity component, Also, it has controversial point of that can not ignore conduction loss of switch in complete work kind action of soft switching. In this paper, as high frequency induction heating power supply, we propose half bridge type multi resonance soft switching high frequency inverter topology that can realize high amplitude operation of load current with controlling switch current by multiplex resonance, mitigating surge voltage when switch gets into compulsion extinction and to be complete operation of zero current switching by opposit parallel connected reactor to inverter switch. and do circuit analysis for choice of most suitable circuit parameter of circuit

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.1013-1021
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• 2024

A series of experiments were performed to produce a fuel source for a molten salt reactor (MSR) through pyroprocessing technology. A simulated LiCl-KCl-UCl₃-NdCl₃ salt system was prepared, and the U element was fully recovered using a liquid cadmium cathode (LCC) by applying a constant current. As a result, the salt was purified with an UCl₃ concentration lower than 100 ppm. Subsequently, the U/RE ingot was prepared by melting U and RE metals in Y₂O₃ crucible at 1473 K as a surrogate for RE-rich ingot product from pyroprocessing. The produced ingot was sliced and used as a working electrode in LiCl-KCl-LaCl₃ salt. Only RE elements were then anodically dissolved by applying potential at - 1.7 V versus Ag/AgCl reference electrode. The RE-removed ingot product was used to produce UCl₃ via the reaction with NH4Cl in a sealed reactor.

• Progress in Superconductivity and Cryogenics
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• v.5 no.3
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• pp.57-61
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• 2003

In this investigation, the 6.6kV/200A Inductive Superconducting Fault Current Limiter (SFCL) was designed and fabricated. The type of DC reactor for Inductive SFCL was determined as solenoid type during the period of $1${st}$ year research. The 5 bobbins for DC reactor were fabricated and each bobbin was wound with 4 stacked High-Tc superconducting (HTS) tapes and the 5 bobbins were connected in series. The critical current and inductance of DC reactor were simulated by Finite Element method (FEM) and compared with the measured results. The characteristics of DC reactor were enhanced in sub-cooled nitrogen system rather than in liquid nitrogen system. The procedures to accomplish the sub-cooled nitrogen system and the experimental results were introduced in detail. Moreover, the design of sub-cooled nitrogen cryogenic system for next year research was introduced in brief.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.60 no.4
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• pp.880-884
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• 2011

The matrix-type superconducting fault current limiter (MSFCL) consists of the trigger and current-limiting parts. The trigger part with reactors connected in parallel improves the quenching characteristics by applying the external magnetic field into the superconducting units. The current-limiting part with superconducting units connected in parallel and shunt reactors connected in series limit the fault current when the fault occurs. We developed the integrated reactor with the trigger and the current-limiting parts to apply high external magnetic field into the superconducting units. This was composed of a superconducting unit for the trigger part and two superconducting units for the current-limiting parts. We confirmed that the external magnetic field generated in the MSFCL with an integrated reactor was larger than that of the MSFCL with the separated reactors. So the differences of voltages generated between superconducting units were decreased in the difference according to the increment of the applied voltage. The whole magnitude of the SFCL was reduced because the volume of an integrated reactor could be reduced by one-third than that of the separated reactors. We confirmed that the critical behavior between the superconducting units in the MSFCL with an integrated reactor was more improved than that of the MSFCL with the separated reactors.

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.257-272
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• 2007

Supercritical water-cooled reactors (SCWRs) are recognized as a Generation IV reactor concept. The Super LWR is a pressure-vessel type thermal spectrum SCWR with downward-flow water rods and is currently under study at the University of Tokyo. This paper reviews Super LWR safety. The fundamental requirement for the Super LWR, which has a once-through coolant cycle, is the core coolant flow rate rather than the coolant inventory. Key safety characteristics of the Super LWR inhere in the design features and have been identified through a series of safety analyses. Although loss-of-flow is the most important abnormality, fuel rod heat-up is mitigated by the "heat sink" and "water source" effects of the water rods. Response of the reactor power against pressurization events is mild due to a small change in the average coolant density and flow stagnation of the once-through coolant cycle. These mild responses against transients and also reactivity feedbacks provide good inherent safety against anticipated-transient-without-scram (ATWS) events without alternative actions. Initiation of an automatic depressurization system provides effective heat removal from the fuel rods. An "in-vessel accumulator" effect of the reactor vessel top dome enhances the fuel rod cooling. This effect enlarges the safety margin for large LOCA.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1715-1720
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• 2019

Jozef Stefan Institute (JSI), TRIGA Mark II reactor employs the homogeneous mixture of uranium and zirconium hydride fuel type. Since its upgrade, a series of fresh fuel steady state experimental benchmarks have been conducted. The benchmark results have provided data for testing computational neutronics codes which are important for reactor design and safety analysis. In this work, we investigated the JSI TRIGA Mark II reactor neutronics characteristics: the effective multiplication factor and two safety parameters, namely the control rod worth and the fuel temperature reactivity coefficient using SuperMC. The modeling and real-time cross section generation methods of SuperMC were evaluated in the investigation. The calculation analysis indicated the following: the effective multiplication factor was influenced by the different cross section data libraries; the control rod worth evaluation was better with Monte Carlo codes; the experimental fuel temperature reactivity coefficient was smaller than calculated results due to change in water temperature. All the results were in good agreement with the experimental values. Hence, SuperMC could be used for the designing and benchmarking of other TRIGA Mark II reactors.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4179-4188
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• 2021

This study identifies efficient earthquake intensity measures (IMs) for seismic performances and fragility evaluations of the reactor containment building (RCB) in the advanced power reactor 1400 (APR1400) nuclear power plant (NPP). The computational model of RCB is constructed using the beam-truss model (BTM) for nonlinear analyses. A total of 90 ground motion records and 20 different IMs are employed for numerical analyses. A series of nonlinear time-history analyses are performed to monitor maximum floor displacements and accelerations of RCB. Then, probabilistic seismic demand models of RCB are developed for each IM. Statistical parameters including coefficient of determination (R²), dispersion (i.e. standard deviation), practicality, and proficiency are calculated to recognize strongly correlated IMs with the seismic performance of the NPP structure. The numerical results show that the optimal IMs are spectral acceleration, spectral velocity, spectral displacement at the fundamental period, acceleration spectrum intensity, effective peak acceleration, peak ground acceleration, A95, and sustained maximum acceleration. Moreover, weakly related IMs to the seismic performance of RCB are peak ground displacement, root-mean-square of displacement, specific energy density, root-mean-square of velocity, peak ground velocity, Housner intensity, velocity spectrum intensity, and sustained maximum velocity. Finally, a set of fragility curves of RCB are developed for optimal IMs.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2011.06a
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• pp.889-890
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• 2011

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.287-298
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• 2007

As part of the $AP1000^{TM}$ pressurized water reactor design certification program, a series of integral systems tests of the nuclear steam supply system was performed at the APEX-1000 test facility at Oregon State University. These tests provided data necessary to validate Westinghouse safety analysis computer codes for AP1000 applications. In addition, the tests provided the opportunity to investigate the thermal-hydraulic phenomena expected to be important in AP1000 small-break loss of coolant accidents (SBLOCAs). The APEX-1000 facility is a 1/4-scale pressure and 1/4-scale height simulation of the AP1000 nuclear steam supply system and passive safety features. A series of eleven tests was performed in the APEX-1000 facility as part of a U.S. Department of Energy contract. In all, four SBLOCA tests representing a spectrum of break sizes and locations were simulated along with tests to study specific phenomena of interest. The focus of this paper is the SBLOCA tests. The key thermal-hydraulic phenomena simulated in the APEX-1000 tests, and the performance and interactions of the passive safety-related systems that can be investigated through the APEX-1000 facility, are emphasized. The APEX-1000 tests demonstrate that the AP1000 passive safety-related systems successfully combine to provide a continuous removal of core decay heat and the reactor core remains covered with considerable margin for all small-break LOCA events.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.37 no.9
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• pp.855-862
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• 2013

A series of 1/5 scale-down reactor flow distribution tests had been conducted to determine the hydraulic characteristics of an APR+ (Advanced Power Reactor Plus), which were used as the input data for an open core thermal margin analysis code. In this study, to examine the applicability of computational fluid dynamics with the porous model to the analysis of APR+ internal flow, simulations were conducted using the commercial multi-purpose computational fluid dynamics software ANSYS CFX V.14. It was concluded that the porous domain approach for some reactor internal structures could adequately predict the flow characteristics inside a reactor in a qualitative manner. If sufficient computational resources are available, the predicted core inlet flow distribution is expected to be more accurate by considering the real geometry of the internal structures, especially upstream of the core inlet.