• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2438-2446
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• 2023

To cool down a nuclear reactor core and prevent the fuel damage without a pump-driven active component during any anticipated accident, the passive emergency core cooling system (PECCS) was designed and adopted in an advanced light water reactor, i-POWER. In this study, for a validation of the cooling capability of PECCS, thermal-hydraulic integral effect tests were performed with the ATLAS facility by simulating intermediate and small break loss-of-coolant accidents (IBLOCA and SBLOCA). The test result showed that PECCS could effectively depressurize the reactor coolant system by supplying the safety injection water from the safety injection tanks (SITs). The result pointed out that the safety injection from IRWST should have been activated earlier to inhibit the excessive core heat-up. The sequence of the PECCS injection and the major thermal hydraulic transient during the SBLOCA transient was similar to the result of the IBLOCA test with the equivalent PECCS condition. The test data can be used to evaluate the capability of thermal hydraulic safety analysis codes in predicting IBLOCA and SBLOCA transients under an operation of passive safety system.

• 대한기계학회논문집A
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• 제41권4호
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• pp.333-336
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• 2017

스피커 진동판은 고유 특성에 의해 분할진동을 발생 시킨다. 이 분할진동은 진동판의 형상 변화를 가져올 정도로 뚜렷한 영향을 주는데, 본 논문에서는 IT 분야의 첨병인 스마트 폰을 포함한 초박형 멀티미디어 기기에서 많이 사용되는 마이크로 스피커를 그 대상으로 삼는다. 마이크로 스피커는 일반적인 스피커와 다른 평판형의 구조적인 형태와 공간적인 제약이 존재한다. 특히 구동 공간이 밀폐형으로 설계되어 무빙 코일에서 발생하는 열의 냉각이 열악하고 보조적인 서스펜션 구조를 갖추기 어렵다. 본 연구에서는 진동판의 열전달과 분할진동 모드의 연관성을 연구한다. 이를 위해 진동판의 레이저 스캔을 통한 분할진동 측정과 열화상 카메라 촬영을 통한 열변화 측정의 두 단계로 나누어 실험을 진행한다. 이를 통해 특정 주파수 범위에서 분할진동 모드와 열전달 형태를 비교함으로써, 열화상 카메라를 통한 촬영 결과로 진동판 분할진동 모드의 경향성을 빠르게 예상할 수 있어, 마이크로 스피커의 최적 설계에 도움이 되는 지표를 제공할 수 있을 것으로 기대한다.

• Elastomers and Composites
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• 제57권4호
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• pp.165-174
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• 2022

The effect of fumed silica loading on the thermal stability and mechanical properties of fluorosilicone (FVMQ) rubber was investigated. The distribution of fumed silica inside FVMQ was characterized using scanning electron microscopy, and the thermal stability of composites was evaluated using thermogravimetric analysis and by the changes in mechanical performance during thermo-oxidative aging. The function mechanism of fumed silica was studied by Fourier transform infrared spectroscopy. The results show that with increasing silica content, the crosslink density of composites, the modulus at 100%, and tensile strength also increased, whereas the elongation at break decreased. Furthermore, increasing the silica content of composites increased the initial decomposition temperature (T_d) and residual weight of the composite after exposure to nitrogen. In addition, the thermal oxidative aging experiment demonstrated improved aging resistance of the FVMQ composites, including lower change in tensile strength, elongation at break, and modulus at 100%.

• 한국전산유체공학회지
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• 제20권3호
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• pp.87-93
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• 2015

In this study, a numerical analysis was performed to simulate the thermal-hydraulic response of the secondary side of a steam generator(SG) model equipped with an orifice-type SG outlet flow restrictor to a main steam line break(MSLB) at a pressurized water reactor(PWR) plant. The SG analysis model includes the SG upper steam space and the part of the main steam pipe between the SG outlet and the broken pipe end. By comparing the numerical calculation results for the present SG model to those obtained for a simple SG model having no flow restrictor, the effects of the flow restrictor on the thermal-hydraulic response of SG to the MSLB were investigated.

• Nuclear Engineering and Technology
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• 제31권1호
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• pp.17-28
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• 1999

A thermal-hydraulic analysis is conducted on the loss-of-coolant-accident (LOCA) during shutdown operation of YGN Units 3/4. Based on the review of plant-specific characteristics of YGN Units 3/4 in design and operation, a set of analysis cases is determined, and predicted by the RELAP5/MOD3.2 code during LOCA in the hot-standby mode. The evaluated thermal-hydraulic phenomena are blowdown, break flow, inventory distribution, natural circulation, and core thermal response. The difference in thermal-hydraulic behavior of LOCA at shutolown condition from that of LOCA at full power is identified as depressurization rate, the delay in peak natural circulation timing and the loop seal clearing (LSC) timing. In addition, the effect of high pressure safety injection (HPSI) on plant response is also evaluated. The break spectrum analysis shows that the critical break size can be between 1% to 2% of cold leg area, and that the available operator action time for the Sl actuation and the margin in the peak clad temperature (PCT) could be reduced when considering uncertainties of the present RELAP5 calculation.

• Nuclear Engineering and Technology
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• 제44권6호
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• pp.561-586
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• 2012

This paper describes a summary of final prediction results by system-scale safety analysis codes during the OECD/NEA/CSNI ISP-50 exercise, targeting a 50% Direct Vessel Injection (DVI) line break integral effect test performed with the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS). This ISP-50 exercise has been performed in two consecutive phases: "blind" and "open" phases. Quantitative comparisons were performed using the Fast Fourier Transform Based Method (FFTBM) to compare the overall accuracy of the collected calculations. Great user effects resulting from the combination of the possible reasons were found in the blind phase, confirming that user effect is still one of the major issues in connection with the system thermal-hydraulic code application. Open calculations showed better prediction accuracy than the blind calculations in terms of average amplitude (AA) value. A total of nineteen organizations from eleven countries participated in this ISP-50 program and eight leading thermal-hydraulic system analysis codes were used: APROS, ATHLET, CATHARE, KORSAR, MARS-KS, RELAP5/MOD3.3, TECH-M-97, and TRACE.

• 한국유체기계학회 논문집
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• 제17권5호
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• pp.37-42
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• 2014

In Korea while the dam or reservoir is an important water resource, the value of this water resource is deteriorating by thermal-induced stratification. To ameliorate the water quality of reservoir by breaking stratification the use of air diffuser system is now widespread in Korea. According to the previous research, dynamics of bubble plume and destratification efficiency depended upon two dimensionless groupings; Mh and Pn suggested by Asaeda et al (1993). However, these two variables only include Q, N, H, g, u. and installed Boryeong reservior in appropriate width of water aeration, air dose and number of installations after calculating by applying these figures. This paper is performed to find out effect analysis about water aeration improvement to break thermal stratification.

• 한국안전학회지
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• 제18권4호
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• pp.57-63
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• 2003

Thermal degradation of Acrylonitrile butadiene rubber(NBR), which is used for O-ring material as elastomeric sealed diaphragm value in the nuclear power plants, is examined. The thermal degradation is accelerated at 130$^{\circ}C$ by Arrhenius exploit method using the activation energy calculated by thermogravimetric analysis. The weight loss temperature and glass transition temperature are verified for thermally aged NBR. The relationship between dynamic mechanical properties and elongation at break are also investigated. The threshold alue of thermally aged NBR is a ten year in the change of elongation at break.

• Nuclear Engineering and Technology
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• 제33권1호
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• pp.12-24
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• 2001

The post-LOCA long term cooling (LTC) performance of the Korean Standard Nuclear Power Plant (KSNPP) is analyzed for both small break loss-of-coolant accidents (LOCA) and large break LOCA at cold leg. The RELAP5/MOD3.2.2 beta code is used to calculate the LTC sequences based on the LTC plan of the Korean Standard Nuclear Power Plants (KSNPP). A standard input model is developed such that LOCA and the followed LTC sequence can be calculated in a single run for both small break LOCA and large break LOCA. A spectrum of small break LOCA ranging from \ulcorner.02 to 0.5 k2 of break area and a double-ended guillotine break are analyzed. Through the code calculations, the thermal-hydraulic behavior and the boron behavior are evaluated and the effect of the important action including the safety injection tank (SIT isolation and the simultaneous injection in LTC procedure is investigated. As a result, it is found that the sufficient margin is available in avoiding the boron precipitation in the core. It is also found that a further specific condition for the SIT isolation action need to be setup and it is recommended that the early initiation of the simultaneous injection be taken for larger break LTC sequences.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 춘계공동학술발표회요약집
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• pp.148.2-148
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• 2002