• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 1999년도 춘계 학술대회논문집
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• pp.205-212
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• 1999

Safety Depressurization System of the Korean Next Generation Reactor prevents the Reactor Coolant System from over-pressurization by discharging the coolant with high pressure and temperature into the In-containment Refueling Water Storage Tank(IRWST) during an accident. If temperature in the IRWST rises above the temperature limit of $200\;^{\circ}F$ due to the discharged coolant, an unstable steam condensation may occur and cause large load on the IRWST wall. To investigate whether this condition can be reached or not for the design basis accident, the flow and temperature distributions of water in the IRWST wire calculated by using CFX 4.2 computer code. The results show that the local water temperature does not exceeds the temperature limit within the transient time of 5 seconds.

• Corrosion Science and Technology
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• 제5권3호
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• pp.99-105
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• 2006

This paper is to investigate feasibility on quantitative aging state of epoxy coating on concrete wall in containment structure under operation of nuclear power plants. For evaluating the physical characteristics of the epoxy coating, adhesion strengths of two kinds of degraded epoxy coating systems on both steel surfaces and concrete surfaces were measured via accelerated aging. Comparatively impedance data taken by ultrasonic test were also taken to relate with adhesion data. After aging, in case of concrete, from half of specimens, aging of epoxy coating was developed. As for steel, on $4^{th}$ inspection day, adhesion force was failed. To improve reliability on quality degradation of epoxy, relationship between adhesion and impedance was analyzed. By tracing to co-respond to these data, it was possible to Fig. out physical state of as-built epoxy coating. The possibility to develop new methodology of time - dependent aging state on epoxy coating was found and discussed.

• Journal of Advanced Research in Ocean Engineering
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• 제3권1호
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• pp.41-52
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• 2017

This paper presents a numerical and experimental study of sloshing loads on liquefied natural gas (LNG) vessels. Conventional LNG carriers with membrane-type cargo systems have filling restrictions from 10% to 70% of tank height. The main reason for such restrictions is high sloshing loads around these filling depths. However, intermediate filling depths cannot be avoided for most LNG vessels except the LNG carrier. This study attempted to design a membrane-type LNG tank with a modified lower-chamfer shape that allows all filling operations. First, numerical sloshing analysis was carried out to find an efficient height of the lower-chamfer that can reduce sloshing pressure at partially filled conditions. The numerical sloshing analysis program SHI-SLOSH was used for numerical simulation; this program is based on SOLA-VOF. The effectiveness of the newly designed tanks was validated by 1:50-scale three-dimensional tank tests. A total of three different tanks were tested: a conventional tank and two modified tanks. As test conditions, various filling depths and wave periods were considered, and the same test conditions were applied to the three tanks. During the test, slosh-induced dynamic pressures were measured around the corners of the tank wall. The measured pressure data were post-processed and the pressures of the three different tanks were statistically compared in several ways. Experimental results show that the modified tanks were quite effective in reducing sloshing loads at low filling conditions. This study demonstrated the possibility of all filling operations for LNG cargo containment systems.

• 한국가스학회지
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• 제5권4호
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• pp.49-55
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• 2001

본 연구에서는 $9\%$ 니켈강재로 제작된 완전 방호식 내부탱크 구조물의 설계 안전성 문제를 유한요소법으로 해석하고자 한다. 내부탱크 구조물에 대한 수치적 해석결과에 의하면 내부탱크에 부착된 톱가더와 스티프너는 내부탱크의 변형율과 응력을 제어하는 중요한 역할을 하고 있다. 내부탱크에 작용하는 여러 하중에서 초저온 액체에 의한 유체정압은 $-162^{\circ}C$의 초저온 온도하중보다 더 큰 영향을 미치고 있음을 알 수 있다. 그러나, 내부탱크에 작용하는 자중량 좌굴하중에 의한 영향은 내부탱크 구조물이 충분한 강도로 이미 설계되어 있으므로 대단히 자게 나타났지만, 특히 톱가더나 스티프너의 역할은 미미한 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제34권3호
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• pp.187-201
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• 2002

The KN-12 cask is designed to transport 12 PWR spent nuclear fuels and to comply with the requirements of Korea Atomic Energy Act, IAEA Safety Standards Series No.57-1 and US 10 CFR Part 71 for a Type B(U)F package. It provides containment, radiation shielding, structural integrity, criticality control and heat removal for normal transport and hypothetical accident conditions. W.H 14$\times$14, 16$\times$16 and 17$\times$17 fuel assemblies with maximum allowable initial enrichment of 5.0 wt.%, maximum average burn-up of 50,000 MWD/MTU and minimum cooling time of 7 years being used in Korea will be loaded and subsequently transported under dry and wet conditions. A forged cylindrical cask body which constitutes the containment vessel is closed by a cask lid. Polyethylene rods for neutron shielding are arranged in two rows of longitudinal bore holes in the cask body wall. A fuel basket to accommodate up to 12 PWR fuel assemblies provides support of the fuels, control of criticality and a path to dissipate heat. Impact limiters to absorb the impact energy under the hypothetical accident conditions are attacked at the top and at the bottom side of the cask during transport. Handling weight loaded with water is 74.8 tons and transport weight loaded with water with the impact limiters is 84.3 tons. The cask will be licensed in accordance with Korea Atomic Energy Act 3nd fabricated in Korea in accordance with ASME B&PV Code Section 111, Division 3.

• 한국지반공학회논문집
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• 제23권7호
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• pp.47-55
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• 2007

본 연구에서는 흙-벤토나이트월과 같은 차단벽체에 대한 전기전도도 모니터링 기법의 적용성을 평가하였다. 벤토나이트 함량, 투수계수 및 중금속 농도에 따른 전기전도도 변화를 파악하기 위하여 투수실험과 주상실험을 실시하였다. 흙-벤토나이트 혼합 다짐시편의 전기전도도는 벤토나이트 함량이 증가함에 따라 선형적으로 증가하는 경향을 나타내었다. 또한 흙-벤토나이트 차단벽체의 투수계수는 벤토나이트 함량과 반비례하는 경향을 나타내었으며, 전기전도도와도 반비례하는 경향을 나타내었다. 주상실험을 통하여 전기전도도 파과곡선과 농도 파과곡선을 동시에 구할 수 있었다. 이 결과는 전기전도도 측정기법이 벽체 내부의 임의 지점에서 중금속의 이동을 감지하고 오염물질의 누출 가능성을 확인하는 유용한 방법으로 활용될 수 있음을 나타낸다. 본 연구결과에 의하면 전기전도도 측정기법이 차단벽체 모니터링에 효과적으로 적용될 수 있을 것으로 기대된다.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.132-140
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• 2024

In the event of a severe accident in Sodium Cooled Fast Reactors (SFR), the sodium combustion aerosols along with fission product aerosols would migrate to the environment through leak paths of the Reactor Containment Building (RCB) concrete wall under positive pressure. Understanding the characteristics of sodium aerosol transport through concrete leak paths is important as it governs the environmental source term. In this context, experiments are conducted to study the influence of various parameters like pressure, initial mass concentration, leak path diameter, humidity etc., on the transport and deposition of sodium aerosols in straight leak paths of concrete. The leak paths in concrete specimens are prepared by casting and the diameter of the leak path is measured using thermography technique. Aerosol transport experiments are conducted to measure the transported and plugged aerosol mass in the leak paths and corresponding plugging times. The values of differential pressure, aerosol concentration and relative humidity taken for the study are in the ranges 10-15 kPa, 0.65-3.04 g/m³ and 30-90% respectively. These observations are numerically simulated using 1-Dimensional transport equation. The simulated values are compared with the experimental results and reasonable agreement among them is observed. From the safety assessment view of reactor, the approach presented here is conservative as it is with straight leak paths.

• 한국지하수토양환경학회:학술대회논문집
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• 한국지하수토양환경학회 2005년도 총회 및 춘계학술발표회
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• pp.202-210
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• 2005

In this study, the fly ash was employed as a possible alternative to the bentonite for its high sorption capacity against cationic heavy metal. To consider the constituents of barrier possibly used, the specimens were mixed with different material contents (fly ash : weathered soil : bentonite), then sorption test was performed. Also the specimens were molded on the wet side of optimum moisture contents like mixing ratio of sorption test and their hydraulic conductivities were measured in flexible-wall permeameters. And to confirm the effect of dissolved cations, the hydraulic conductivity tests were repeated by converting the permeant liquids from water to $Cd^{2+}$ solution. Finally, the Cd-concentration at the effluent was analyzed for 500hrs to compare the effectiveness of each specimen in contaminant retardation. Test results showed that the more the ratio of fly ash increase, the more Kd value increase, and the hydraulic conductivity of weathered soil/bentonite (95:5) mixture was the lowest $(2.9*10^{-8}cm/sec)$, and specimens made of fly ash and fly ash/weathered soil mixtures showed similar hydraulic conductivity. Although the permeant liquid was changed from water to $Cd^{2+}$ solution, the hydraulic conductivity of all specimens except for weathered soil maintained similarly like before. Consequently, the initial breakthrough point of Cd in weathered soil specimen was observed at about 5hrs after the test started while that of fly ash specimens was not observed during the whole test period of 500hrs. The results implied that fly ash had a sufficient retardation capacity against contaminant transport possibly by its high sorption capacity although it showed little effect on the reduction of hydraulic conductivity. Based on the test results, it could be concluded that the fly ash can be possibly used as a suitable barrier material in containment system to attenuate the contaminant transport for its high retardation capacity and for the low cost.

• 에너지공학
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• 제27권4호
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• pp.27-35
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• 2018

피동원자로건물냉각계통(PCCS)은 사고 발생 시 원자로건물로 방출된 열을 제거하여 원전의 건전성을 보장하기 위해 설계되었다. PCCS의 열제거 성능은 증기-공기 혼합물의 응축열전달에 의해 결정된다. 본 연구에서는 응축열전달계수의 예측 정확도를 향상시키기 위해 새로운 상관식을 이식한 MARS-KS 코드를 사용하여 PCCS의 열제거 성능을 평가하였다. MARS-KS 코드에 사용된 새로운 상관식은 압력, 벽면과냉도, 비응축성 기체 질량분율 및 응축튜브의 종횡비와 같은 열전달계수에 영향을 미치는 변수들을 이용하여 개발하였고, 이는 MARS-KS코드의 기본 응축 모델인 Colburn-Hougen 모델을 대체하여 적용되었다. 대형파단 냉각재상실사고 발생 시 PCCS의 운전에 따른 다양한 열수력학적 변수들을 분석하였고, 열제거 성능 평가를 위해 새로운 상관식이 적용된 MARS-KS 코드의 원자로건물 압력거동 계산결과와 기존의 응축모델을 이용한 해석결과를 비교하였다.

• 한국가스학회지
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• 제14권5호
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• pp.13-18
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• 2010

본 연구는 9% 니켈강재식 LNG 저장탱크용 통합제어안전관리시스템을 개발하고자 한다. 새로운 통합제어안전관리 시스템은 기존의 측정 및 제어시스템에 비해 압력, 변위, 하중을 측정할 수 있는 장치를 추가하였다. 또한, 측정된 데이터는 새로운 제어장치와 안전관리 시스템에 의해 통합되고 분석하는 프로세스를 동시적으로 진행하는 것이다. 초대형 완전밀폐식 LNG 저장탱크의 안전성과 효율성을 증가시키기 위해 통합제어안전관리시스템은 압력계이지를 추가하고, 내부탱크의 외측벽과 스티프너, 톱거더의 용접지역, 코너 프로텍션 탱크의 외측벽에 새로운 변위센서/압력센서를 설치하였다. 변위센서와 하중센서는 내부탱크와 코너 프로텍션 탱크의 9% 니켈강재 구조물에 대한 파손징후를 제공할 수 있고, 내부탱크로부터 누설되는 LNG를 감시할 수 있다. 기존의 누설센서는 내부탱크와 코너 프로텍션 사이의 단열지역에 설치한 누설감지기에 의해 경고신호가 접수될 때까지도 9% 니켈강재 탱크의 파괴에 대한 적절한 정보를 제공하지 못한다는데 문제가 있다. 따라서, 새로운 통합제어안전관리시스템은 온도, 압력, 변위, 하중, LNG의 밀도 데이터를 수집하고 분석하기 위한 것으로, 탱크시스템의 안전성과 내부탱크의 누설을 제어하기 위한 시스템이다. 또한, 디지털 데이터는 9% 니켈강재로 제작한 탱크의 안전성에 관련된 변위와 하중, LNG의 액위와 밀도, 쿨다운 공정, 누설, 압력 등을 제어하기 위해 측정한다.