• 전기학회논문지
• /
• 제66권2호
• /
• pp.305-314
• /
• 2017

Most of the existing researches on systemwide optimization of generator maintenance scheduling do not consider the equivalent operating hours(EOHs) mainly due to the difficulties of calculating the EOHs of the CCGTs in the large scale system. In order to estimate the EOHs not only the operating hours but also the number of start-up/shutdown during the planning period should be estimated, which requires the mathematical model to incorporate the economic dispatch model and unit commitment model. The model is inherently modelled as a large scale mixed-integer nonlinear programming problem and the computation time increases exponentially and intractable as the system size grows. To make the problem tractable, this paper proposes an EOH calculation based on demand grouping by K-means clustering algorithm. Network congestion is also considered in order to improve the accuracy of EOH calculation. This proposed method is applied to the actual Korean electricity market and compared to other existing methods.

• Nuclear Engineering and Technology
• /
• 제41권6호
• /
• pp.765-774
• /
• 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.

• 한국정보처리학회:학술대회논문집
• /
• 한국정보처리학회 2012년도 춘계학술발표대회
• /
• pp.79-82
• /
• 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
• /
• 제51권3호
• /
• pp.908-914
• /
• 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
• /
• 제51권1호
• /
• pp.60-72
• /
• 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
• /
• 제55권8호
• /
• pp.2977-2983
• /
• 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.

• 한국안전학회지
• /
• 제38권6호
• /
• pp.72-78
• /
• 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
• /
• 제55권12호
• /
• pp.4335-4349
• /
• 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
• /
• 제56권3호
• /
• pp.922-932
• /
• 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.

• 한국인터넷방송통신학회논문지
• /
• 제23권2호
• /
• pp.157-162
• /
• 2023

태양광발전 설비의 고장 중 서지에 의한 고장이 전체 고장률의 20% 차지하고 있으며 발전 중 수십에서 수백[A]의 에너지 방출과 인버터, 접속반 등의 전기적 손상은 전기안전사고로 이어지고 있다. 특히 낙뢰의 경우 전기회로에 이상 전압이 유기되어 절연을 파괴할 뿐만 아니라 이때 흐르는 전류는 화재의 원인이 되고 부품의 열화를 촉진하는 요인으로 작용한다. 이러한 작용으로 도심 밖에서 주택, 아파트, 관공서 등의 도심 내부로 확산하고 있는 태양광 발전장치의 전기 안전 문제가 대두되고 있다. 낙뢰는 필드 기반 및 전도성 전기 간섭을 유발하기에 이 효과는 케이블 길이 또는 도체 루프 증가와 관련하여 증가한다. 또한 서지는 태양광 모듈, 인버터 및 모니터링장치뿐만 아니라 건물 설비의 장치도 손상하기에 최종적으로는 태양광발전시스템의 화재로 인한 운영 중단과 이에 따른 재정손실을 유발하게 시킬 수 있다. 따라서 본 논문에서는 태양광발전시스템의 낙뢰발생으로 인한 화재 및 전기안전사고 증가로 인하여 재산피해 및 인명피해를 줄일 수 있는 목적으로 태양광발전장치의 낙뢰보호 시스템을 연구하고자 한다.