• Journal of Magnetics
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• 제18권1호
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• pp.43-49
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• 2013

A pulsed ferrite magnetic field generator (FMFG) was designed for the use in the 1000 m long through-the-earth (TTE) communication system for mining disaster situations. To miniaturize the TTE system, a ferrite core having 10,000 of permeability was used for the FMFG. Attenuation of the magnetic field intensity from the FMFG (200-turn and 0.18 m diameter) was calculated to be 89.95 dB at 1000 m depth soil having 0.1 S/m of conductivity. This attenuation was lower than 151.13 dB attenuation of 1 kHz electromagnetic wave at the same conditions. Therefore, the magnetic-field was found to be desirable as a signal carrier source for TTE communications as compared to the electromagnetic wave. The designed FMFG generates the magnetic field intensity of $1{\times}10^{-10}$ Tesla at 1000 m depth. This magnetic field is detectable by compact magnetic sensors such as flux gate or magnetic tunneling junction sensor. Therefore, the miniature FMFG TTE communication system can replace the conventional electromagnetic wave carrier type TTE system and allow reliable signal transmission between rescuer and trapped miners.

• 한국통신학회논문지
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• 제19권6호
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• pp.1046-1055
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• 1994

본 논문에서는 영상 신호처리에 적합한 고속 1 line VRAM을 ASIC화 설계하기 위하여 엑세스 시간특성 및 집적도가 우수한 3-TR dual-port 다이나믹 셀을 채용하여 메모리 코어를 설계하였다. 고속 파이프라인 동작을 위하여 서브어레이 1로부터 첫 행을 분리하였고, TM기 비트 라인에 데이터 래치 구조를 채용하여 한 번지의 동시 입.출력이 가능하도록 설계하였다. 주변 회로로 번지 선택기, 1/2V 전압 발생기를 각각 설계하여 개선된 동작특성을 확인한 후 1.5[ m] CMOS 설계규칙을 이용하여 ASIC화 설계하였다.

• Nuclear Engineering and Technology
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• 제31권1호
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• pp.68-79
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• 1999

The YGN Units 3&4 plant conditions during shutdown operation were reviewed to identify the possible event scenarios following the loss of shutdown cooling (SDC) event. For the five cases of typical reactor coolant system (RCS) configurations under the worst event sequence, such as unavailable secondary cooling and no RCS inventory makeup, the thermal hydraulic analyses were performed using the RELAP5/MOD3.2 code to investigate the plant behavior following the event. The thermal hydraulic analyses include the estimation of time to boil, time to core uncovery, and time to core heat up to determine the containment closure time to prevent the uncontrolled release of fission products to atmosphere. The result indicates that the containment closure is recommended to be achieved within 42 minutes after the loss of SDC for the steam generator (SG) inlet plenum manway open case or the large cold leg open case under the worst event sequence. The containment closure time is significantly dependent on the elevation and size of the opening and the SG secondary water level condition. It is also found that the containment closure needs to be initiated before the boiling time to ensure the survivability of the workers in the containment. These results will provide useful information to operators to cope with the loss of SDC event.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.589-596
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• 2013

In a pressurized heavy water reactor, following loss of the primary coolant, severe core damage would begin with the depletion of the liquid moderator, exposing the top row of internally-voided fuel channels to steam cooling conditions on the inside and outside. The uncovered fuel channels would heat up, deform and disassemble into core debris. Large inventories of water passively reduce the rate of progression of the accident, prolonging the time for complete loss of engineered heat sinks. The efficacy of available backup and ultimate heat sinks, available in a CANDU 6 reactor, in mitigating the consequences of a prolonged station blackout scenario was analysed using the MAAP4-CANDU code. The analysis indicated that the steam generator secondary side water inventory is the most effective heat sink during the accident. Additional heat sinks such as the primary coolant, moderator, calandria vault water and end shield water are also able to remove decay heat; however, a gradually increasing mismatch between heat generation and heat removal occurs over the course of the postulated event. This mismatch is equivalent to an additional water inventory estimated to be 350,000 kg at the time of calandria vessel failure. In the Enhanced CANDU 6 reactor ~2,040,000 kg of water in the reserve water tank is available for prolonged emergencies requiring heat sinks.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.402-409
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• 2019

When a loss of coolant accident (LOCA) occurs in a nuclear power plant, accident scenarios which can prevent core damage are defined based on break size. Current probabilistic safety assessment evaluates that core damage can be prevented under small-break LOCA (SBLOCA) and steam generator tube rupture (SGTR) with rapid cool down (RCD) strategy when all safety injection systems are unavailable. However, previous research has pointed out a limitation of RCD in terms of initiation time. Therefore, RCD success criteria estimation based on allowable coping time under a SBLOCA or SGTR when all safety injection systems are unavailable was performed based on time-line and thermal-hydraulic analyses. The time line analysis assumed a single emergency operating procedure flow, and the thermal hydraulic analysis utilized MARS-KS code with variables of break size, cooling rate, and operator allowable time. Results show while RCD is possible under SGTR, it is impossible under SBLOCA at the APR1400's current cooling rate limitation of 55 K/hr. A success criteria map for RCD under SBLOCA is suggested without cooling rate limitation.

• Nuclear Engineering and Technology
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• 제54권2호
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• pp.554-564
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• 2022

The natural circulation phenomena occurring in fully integrated nuclear reactors are associated with a unique formation mechanism. The phenomenon results from a structural feature of these reactors involving upward flow from the core, located in the central-bottom region of a single vessel, and downward flow to the steam generator in the annulus region. In this study, to understand the natural circulation in a single vessel involving a multi-layered flow path, single-phase and two-phase natural circulation tests were performed using the SMART-ITL facility, and validation analysis of the TASS/SMR-S code was performed by comparing the corresponding test results. Three single-phase natural circulation tests were sequentially conducted at 15%, 10%, and 5% of full-scaled core-power without RCP operation, following which a two-phase natural circulation test was successively conducted with an artificial discharge of coolant inventory. The simulation capability of the TASS/SMR-S code with respect to the natural circulation phenomena was validated against the test results, and somewhat conservative but reasonably comparative results in terms of overall thermalhydraulic behavior were shown.

• 한국자동차공학회논문집
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• 제20권3호
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• pp.20-25
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• 2012

As a kind of distributed energy system, the co-generation system based Diesel engine using after-treatment device was devised for its environmental friendly and economic qualities. It is utilized in that the electric power is produced by the generator connected to the Diesel engine, and waste heat is recovered from both the exhaust gases and the engine itself by the finned tube and shell & tube heat exchangers. An after-treatment device composed ceramic heater and DOC(Diesel Oxidation Catalyst) is installed at the engine outlet in order to completely reignite the unburned fuel from the Diesel engine. In this study, mutual relation of each experimental condition was derived through minimum number of experiment using Taguchi Design and ANOVA recently used in the various fields. It is found that the total efficiency (thermal efficiency plus electric power generation efficiency) of this system reaches maximum 94.4% which is approximately higher than that of the typical diesel engine exhaust heat recovery system.

• 한국압력기기공학회 논문집
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• 제9권1호
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• pp.15-19
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• 2013

The auxiliary feedwater is an important to remove the heat from the reactor core when the main feedwater system is unavailable. In most initiating events in Probabilistic Safety Assessment(PSA), the operaton of this system is required to mitigate the accidents. For one of domestic nuclear power plants, a design change of a turbine-driven auxiliary feedwater pump(TD-AFWP), pipe, and valves in the auxiliary system is implemented due to the aging related deterioration by long time operation. This change includes the replacement of the TD-AFWP, the relocation of some valves for improving the system availability, a new cross-tie line, and the installation of manual valves for maintenance. The design modification affects the PSA because the system is critical to mitigate the accidents. In this paper, the safety effect of the change of the auxiliary feedwater system is assessed with regard to the PSA view point. The results demonstrate that this change can supply the auxiliary feedwater from the TD-AFWP in the accident with the motor-driven auxiliary feedwater pump(MD-AFWP) unavailable due to test or maintenance. In addition, the change of MOV's normal position from "close" to "open" can deliver the water to steam generator in the loss of offsite power(LOOP) event. Therefore, it is confirmed that the design change of the auxiliary feedwater system reduces the total core damage frequency(CDF).

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3275-3285
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• 2021

A Nuclear Power Plant (NPP) is a complex dynamic system-of-systems with highly nonlinear behaviors. In order to control the plant operation under both normal and abnormal conditions, the different systems in NPPs (e.g., the reactor core components, primary and secondary coolant systems) are usually monitored continuously, resulting in very large amounts of data. This situation makes it possible to integrate relevant qualitative and quantitative knowledge with artificial intelligence techniques to provide faster and more accurate behavior predictions, leading to more rapid decisions, based on actual NPP operation data. Data-driven models (DDM) rely on artificial intelligence to learn autonomously based on patterns in data, and they represent alternatives to physics-based models that typically require significant computational resources and might not fully represent the actual operation conditions of an NPP. In this study, a feed-forward backpropagation artificial neural network (ANN) model was trained to simulate the interaction between the reactor core and the primary and secondary coolant systems in a pressurized water reactor. The transients used for model training included perturbations in reactivity, steam valve coefficient, reactor core inlet temperature, and steam generator inlet temperature. Uncertainties of the plant physical parameters and operating conditions were also incorporated in these transients. Eight training functions were adopted during the training stage to develop the most efficient network. The developed ANN model predictions were subsequently tested successfully considering different new transients. Overall, through prompt prediction of NPP behavior under different transients, the study aims at demonstrating the potential of artificial intelligence to empower rapid emergency response planning and risk mitigation strategies.

• 한국산학기술학회논문지
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• 제17권11호
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• pp.755-760
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• 2016

근래에는 신재생에너지를 이용한 독립적인 발전기의 수요가 증가하고 있는 추세이며 그 중에서 소형 풍력발전기의 개발 또한 활발하게 이루어지고 있다. 이러한 소형 풍력발전기는 목적에 따라 단순화 및 소형화가 가능하도록 영구자석이 주로 쓰인다. 하지만 영구자석 동기기는 구조적인 원인으로 인하여 코깅토크를 수반하고 이는 소음과 진동의 원인이 된다. 코깅토크는 영구자석이나 코어의 형상에 의해 변하며 적절한 설계기법으로 코깅토크를 저감시킬 수 있다. 본 논문에서는 영구자석의 형상변화를 통해 소형 풍력발전기에 많이 사용되는 표면부착형 영구자석 동기전동기의 코깅토크를 저감시키는 설계기법을 제시하였다. 코깅토크를 줄일 수 있는 영구자석의 형상을 구하는 데에는 확률론적 최적화기법의 일종인 진화론적 최적화기법을 사용했다. 최적화 기법을 적용할 때에 설계변수로는 영구자석의 폭을 조절하는 각도와, 영구자석의 외경을 조절하는 반지름을 설정하였다. 제시된 설계기법을 사용해서 극/슬롯의 조합이 8극/18슬롯이고 출력이 300W급인 풍력발전기를 설계하고 코깅토크와 출력전압 등의 특성을 계산했다. 계산결과에 의하면 초기모델에 비해 최적화모델에서 코깅토크와 토크리플 모두가 감소해서, 본 연구에서 제시한 설계기법이 코깅토크를 줄이는 데에 효과가 있음을 확인하였다.