• Nuclear Engineering and Technology
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• 제27권3호
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• pp.327-335
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• 1995

원자로의 가동중지 중이나. 재장전시 중기 발생기의 세관검사 및 보수작업을 병행하면 원전의 운전정지보수기 간을 현저하게 단축할 수 있다. 이때 원자로가 설치되어 있는 수조의 냉가수가 중기발생기내로 유입되는 것을 막는 장비로써 노즐댐이 있다. 노즐댐의 설치는 고방사선환경과 제한된 공간에서 작업을 해야 하는 특수성 때문에 작업자들이 기피하는 현상을 보인다. 현재 쓰이고 있는 무거운 노즐댐은 노즐댐설치 및 제거작업에 장애가 되는 가장 큰 요인이다. 본 논문에서는 노즐댐의 재질선정과 구조설계를 병행하여 현재 쓰이고 있는 노즐댐보다 가벼우면서도 굽힘강성 대 무게비와 비 강도가 증가된 노즐댐을 설계하였으며, 탄소섬유강화 복합재료로 경량노즐댐을 제작 완료하였다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2003년도 하계학술대회 논문집 D
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• pp.2215-2217
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• 2003

In this paper we describe a Control Rod Control System(CRCS) with the various functions for the test and operation of Control Rod Drive Mechanism(CRDM). The CRCS controls the motion of the full length rod drive mechanisms in response to signals from the Reactor Operator and the Reactor Regulating System. The mechanisms are grouped and identified as being for either Shutdown Banks or Control Banks. The CRCS also provides information regarding rod motion, rod position, and status of the Rod Control System. Also we have implemented the diverse functions in the developed CRCS. Due to the developed CRCS, we are assured that the commercial operation by this system be made before long.

• 전기학회논문지
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• 제66권2호
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• pp.305-314
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• 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
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• 제41권6호
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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.

• 한국정보처리학회:학술대회논문집
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• 한국정보처리학회 2012년도 춘계학술발표대회
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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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• 제51권3호
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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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• 제51권1호
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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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• 제55권8호
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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.

• 한국안전학회지
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• 제38권6호
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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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• 제55권12호
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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.