• Nuclear Engineering and Technology
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• v.41 no.6
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• pp.765-774
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• 2009

For over 30 years, investigations of the thermohydraulic behavior of pressurized-water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany. The PKL facility models the entire primary side and significant parts of the secondary side of a of pressurized water reactor at a height scale of 1:1. Volumes, power ratings and mass flows are scaled with a ratio of 1:145. The experimental facility consists of four primary loops with circulation pumps and steam generators (SGs) arranged symmetrically around the reactor pressure vessel (RPV). The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermohydraulic phenomena. The PKL tests began in the mid 1970s with the support of the German Research Ministry. Since the mid 1980s, the project has also been significantly supported by the German PWR operators. Since 2001, 25 partner organizations from 15 countries have taken part in the PKL investigations with the support and mediation of the OECD/ NEA (Nuclear Energy Agency). After an overview of PKL history and a short description of the facility, this paper focuses on the investigations carried out since the beginning of the international cooperation, and shows, by means of some examples, what insights can be derived from the tests.

• Proceedings of the Korea Information Processing Society Conference
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• 2012.04a
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• pp.79-82
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• 2012

기존의 서버 전원 모드 제어에는 서버를 Off할 때 정적 종료 방식을 사용하는 관계로 서버가 사용자 요청을 모두 처리하는 최적의 종료 시간을 찾는데 시간이 많이 걸리는 단점을 가진다. 이 시간이 짧게 되면 사용자 QoS를 보장할 수 없고, 반대로 이 시간이 길게 되면 전력 절감을 기대할 수 없다. 본 논문에서는 정적 종료 방식의 단점을 극복하는 동적 종료 방식을 제안한다. 제안된 방식은 최적의 종료 시간을 찾을 필요 없이 각 서버가 사용자의 요청을 모두 처리하였을 때 자동적으로 서버를 Off한다. 제안된 방법은 최적의 시간을 자동적으로 찾아내기 때문에 사용자 QoS를 보장하고, 전력을 절감한다. 실험은 30대의 PC 클러스터를 이용하여 수행되었고, 실험을 통하여 제안하는 동적 종료 방법이 기존의 정적 종료 방법에 비해 운영자의 수고 없이 자동적으로 전력 절감 및 사용자 QoS에 기여함을 확인하였다.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.908-914
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• 2019

Fusion power shutdown system (FPSS) is a safety system to stop plasma in case of accidents or incidents. The gas injection system for the FPSS presented in this work is designed to research the flow development in a closed system. As the efficiency of the system is a crucial property, plenty of experiments are executed to get optimum parameters. In this system, the flow is driven by the pressure difference between a gas storage tank and a vacuum vessel with a source pressure. The idea is based on a constant volume system without extra source gases to guarantee rapid response and high throughput. Among them, valves and gas species are studied because their properties could influence the velocity of the fluid field. Then source pressures and volumes are emphasized to investigate the volume flow rate of the injection. The source pressure has a considerable effect on the injected volume. From the data, proper parameters are extracted to achieve the best performance of the FPSS. Finally, experimental results are used as a quantitative benchmark for simulations which can add our understanding of the inner gas flow in the pipeline. In generally, there is a good consistency and the obtained correlations will be applied in further study and design for the FPSS.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.60-72
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• 2019

The air-cooled passive decay heat removal system (APDHR) was proposed to provide the ultimate heat sink for non-LOCA accidents. The APDHR is a modified one of Passive Auxiliary Feed-water system (PAFS) installed in APR+. The PAFS has a heat exchanger in the Passive Condensate Cooling Tank (PCCT) and can remove decay heat for 8 h. After that, the heat transfer rate through the PAFS drastically decreases because the heat transfer condition changes from water to air. The APDHR with a vertical heat exchanger in PCCT will be able to remove the decay heat by air if it has sufficient natural convection in PCCT. We conducted the thermal-hydraulic simulation by the MARS code to investigate the behavior of the APR + selected as a reference plant for the simulation. The simulation contains two phases based on water depletion: the early phase and the late phase. In the early phase, the volume of water in PCCT was determined to avoid the water depletion in three days after shutdown. In the late phase, when the number of the HXs is greater than 4089 per PCCT, the MARS simulation confirmed the long-term cooling by air is possible under extended Station Blackout (SBO).

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2977-2983
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• 2023

During normal operation, some parts of the fission product in the defective fuel rods can release into the primary loops in PWR and the escape rate coefficients are widely used to assess quantitatively the release behaviors of fission products in the industry. The escape rate coefficients have been standardized and have been validated by some drilling experiments before the 1970s. In the paper, the model to determine the escape rate coefficients of fission products has been established and the typical escape rate coefficients of noble gas and iodine have been deduced based on the measured radiochemical data in one operating PWR. The result shows that the apparent escape rate coefficients vary with the release-to-birth and decay constants for different fission products of the same element. In addition, it is found that the escape rate coefficients from the defective rod with large defects are much higher than the standard escape rate coefficients, i.e., averagely 4.4 times and 1.8 times for noble gas and iodine respectively. The enhanced release of fission products from the severe secondary hydriding of several defective fuel rods in one cycle may lead to the potential risk of the temporary shutdown of the operating reactors.

• Journal of the Korean Society of Safety
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• v.38 no.6
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• pp.72-78
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• 2023

The purpose of this paper is to derive a quantified approach for Operator Manual Actions (OMAs) based on the existing fire Human Reliability Analysis (HRA) methodology developed by the Korea Atomic Energy Research Institute (KAERI). The existing fire HRA method was reviewed, and supplementary considerations for OMA quantification were established through a comparative analysis with NUREG-1852 criteria and the review of the existing literature. The OMA quantification approach involves a timeline that considers the occurrence of Multiple Spurious Operations (MSOs) during a Main Control Room Abandonment (MCRA) determination and movement towards the Remote Shutdown Panel (RSP) in the event of a Main Control Room (MCR) fire. The derived failure probability of an OMA from the approach proposed in this paper is expected to enhance the understanding of its reliability. Therefore, it allows moving beyond the deterministic classification of "reliable" or "unreliable" in NUREG-1852. Also, in the event of a nuclear power plant fire where multiple OMAs are required within a critical time range, it is anticipated that the OMA failure probability could serve as a criterion for prioritizing OMAs and determining their order of importance.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4335-4349
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• 2023

Nuclear marine propulsion has been emerging as a next generation carbon-free power source, for which proper passive residual heat removal systems (PRHRSs) are needed for long-term safety. In particular, the characteristics of unlimited operation time and compact design are crucial in maritime applications due to the difficulties of safety aids and limited space. Accordingly, a compact and long-term-operable PRHRS has been proposed with the key design concept of using both air cooling and seawater cooling in tandem. To confirm its feasibility, this study conducted system design and a transient analysis in an accident scenario. Design results indicate that seawater cooling can considerably reduce the overall system size, and thus the compact and long-term-operable PRHRS can be realized. Regarding the transient analysis, the Multi-dimensional Analysis of Reactor Safety (MARS-KS) code was used to analyze the system behavior under a station blackout condition. Results show that the proposed design can satisfy the design requirements with a sufficient margin: the coolant temperature reached the safe shutdown condition within 36 h, and the maximum cooling rate did not exceed 40 ℃/h. Lastly, it was assessed that both air cooling and seawater cooling are necessary for achieving long-term operation and compact design.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.922-932
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• 2024

This paper presents the design features and preliminary design analysis results of the Wielenga Innovation Static Salt Reactor (WISSR). The WISSR incorporates features that make it both flexible and inherently safe. It is based on innovative technology that controls a nuclear reactor by moving molten salt fuel into or out of the core. The reactor is a low-pressure, fast spectrum transuranic (TRU) burner reactor. Inherent shutdown is achieved by a large negative reactivity feedback of the liquid fuel and by the expansion of fuel out of the core. The core is made of concentric, thin annular fuel chambers containing molten fuel salt. A molten salt coolant passes between the concentric fuel chambers to cool the core. The core has both fixed and variable volume fuel chambers. Pressure, applied by helium gas to fuel reservoirs below the core, pushes fuel out of a reservoir and up into a set of variable volume chambers. A control system monitors the density and temperature of the fuel throughout the core. Using NaCl-(TRU,U)Cl₃ fuel and NaCl-KCl-MgCl₂ coolant, a road-transportable compact WISSR core design was developed at a power level of 1250 MWt. Preliminary neutronics and thermal-hydraulics analyses demonstrate the technical feasibility of WISSR.

• The Journal of the Institute of Internet, Broadcasting and Communication
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• v.23 no.2
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• pp.157-162
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• 2023

Among the failures of photovoltaic power generation facilities, failures caused by surges account for 20% of the total failure rate, and energy emissions of tens to hundreds [A] during power generation and electrical damage to inverters and connection boards lead to electrical safety accidents. In particular, in the case of lightning, an abnormal voltage is induced in an electric circuit to destroy insulation, and the current flowing at this time causes a fire and acts as a factor that accelerates the deterioration of parts. Due to this action, the problem of electrical safety of solar power generation devices spreading from outside the city center to the inside of the city center such as houses, apartments, and government offices is emerging. Since lightning strikes cause both field-based and conducted electrical interference, this effect increases with increasing cable length or conductor loops. In addition, surge damages not only solar modules, inverters and monitoring devices, but also building facilities, which can eventually cause operational shutdown due to fire of the photovoltaic power generation system and consequent financial loss. Therefore, in this paper, a lightning protection system for solar power generation devices is studied for the purpose of reducing property damage and human casualties due to the increase in fire and electrical safety accidents caused by lightning strikes in photovoltaic power generation systems.

• The Journal of the Korea institute of electronic communication sciences
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• v.15 no.6
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• pp.1069-1074
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• 2020

Since DC has no zero point, an arc occurs when the DC circuit breaker performs a shutdown operation. In this case, a fatal accident may occur in the circuit breaker or in the grid, depending on the magnitude of the arc. Therefore, the shutdown performance and the reliability of the circuit breaker are important in the commercialization of HVDC. In this study, a superconducting LC circuit breaker was proposed to improve the performance and the reliability of the DC circuit breaker. The superconducting LC circuit breaker applied a superconducting coil to the inductor of the existing LC circuit breaker. Other than limiting the initial fault current, it also creates a stable zero point in the event of a fault current. To verify this, simulation was performed through EMTDC/PSCAD. Furthermore, the superconducting LC circuit breaker was compared with the LC circuit breaker with a normal coil. As a result, it was found that the LC circuit breaker with the superconducting coil limited the initial fault current further by approximately 12 kA compared to the LC circuit breaker with a normal coil. This reduced the arc extinguish time by approximately 0.16 sec, thereby decreasing the elctrical power burden on the circuit breaker.