• Solar Energy
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• v.18 no.2
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• pp.51-67
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• 1998

In this paper, various solar systems are introduced according to usage pattern considering national weather condition and hot water comsumption pattern. First, basic principle and components of solar hot heater are presented, and then operational machanism of various solar hot water systems are presented according to the case of domestic and industrial use with network of piping.

• Computational Structural Engineering
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• v.8 no.1
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• pp.13-18
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• 1995

원전의 주요 구조물은 극심한 재해를 초래할 수 있는 자연현상(지진, 쓰나미 등)이 수반하는 극한하중에 견딜 수 있어야 하고 내부의 안전관련 기기의 기능과 만일의 사고(주증기나 고에너지 배관파단사고)시에도 구조적 안전성이 보장되도록 설계.시공되고 있다. 또한 운전원의 조작실수나 오판을 극복할 수 있도록 원자로 고유의 안전성과 자연법칙에 의한 정상유지를 할 수 있는 신기술이 도입되고 있다. 품질보증 계획에 의한 최선의 품질보장, 엄격한 방사선 감시와 온배수 영향의 최소화, 최신방사성 폐기물 처분기술 적용 등을 통하여 기술의존형 에너지인 원자력의 활용증대가 불가피하다고 판단된다.

• Journal of Energy Engineering
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• v.8 no.1
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• pp.59-66
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• 1999

원자력발전의 고유한 문제점을 해결할 수 있는 새로운 집적폐회로형 AMBIDEXTER 원자력시스템 개념을 제안하였다. 이 복합시스템은 일체형 원자로를 중심으로 열/에너지 변환회로와 방사선/물질 수송회로를 서로 독립적으로 구성하므로 최소 방사선 위험부담 아래서 원자력에너지의 잇점을 극대화하는 설계이다. 특히 방사선/물질 수송회로로부터 분리된 고준위 방사성 폐기물에서 고부가가치 동위원소나 방사선원을 선별적으로 용이하게 추출, 활용할 수 있다. 원자로 계통은 얇고 큰 Hastelloy 합금 원자로용기 내부를 노심, 침니, 열교환기, 다운캄어 및 입구플레넘 콤파트먼트로 분할하여 배관이나 벨브관이 없으므로 기기파손으로 인한 방사성물질의 대량 외부 누출은 불가능하다. Th/233U 용융염핵연료의 핵물리 및 열수력학적 특성을 살려 AMBIDEXTER 노심의 핵적 자활성 설계에 융통성을 부여하는 변성핵연료주기를 도입하면 핵연료자원의 공급 안정화나 핵확산방지의 투명성 제고에 큰 효과가 있다. AMBIDEXTER 설계연구에 관련된 핵심기술들은 일찍이 미국 ORNL에서 시작한 MSR 프로그램을 통해 개발되어 이미 대부분 상용화하고 있기 때문에 현재 추진 중인 250 MWth급 원형로 모듈의 개념개발에서는 주로 시스템 통합에 관한 문제들이 중점적으로 다루어진다.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.204-214
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• 1993

Work on fracture mechanics has provided a technical basis for elimination of main coolant loop double ended guillotine breaks from the structural design basis of reactor coolant system. Without main coolant loop pipe breaks, the tributary pipe breaks must be considered as design bases until further fracture mechanics work could eliminate some of these breaks from design consideration. This paper determines the core support barrel shell responses for the 3 inch pressurizer spray line nozzle break which is expected to be the only inlet break remaining in the primary side after leak-before-break evaluation is extended to smaller size pipes in the near future. The responses are compared with those due to 14 inch safety injection nozzle break and main coolant loop pipe break. The results show that, when the leak-before-break concept is applied to the primary side piping systems with a diameter of 10 inches or over, the core support barrel shell responses due to pipe breaks in the primary side are negligible for the faulted condition design.

• Journal of the Korean Institute of Gas
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• v.7 no.2 s.19
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• pp.48-53
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• 2003

This study is to look at the effect for deformation of carbon steel for high-presure pipe, on the AE signals produced by tensile test. Acoustic emission(AE) has been widely used in various fields because of its extreme sensitivity, dynamic detection ability and location of growing defects. We investigated a relationship failure mode and AE signals by tensile test, From the tensile test, we could divide into four ranges of the failure modes of elastic range, yield range, plastic range before $\sigma$u, plastic range after $\sigma$u. And failure behaviors of elastic range, yield range, plastic range before $\sigma$u, plastic range after $\sigma$u could be evaluated in tensile test by AE counts, accumulation counts and time frequency analysis. It is expected to be basic data that can protect a risk according to tensile test and bending of pipe material for pressure vessel, as a real time test of AE.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.2
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• pp.51-59
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• 2017

Steam jets ejected from a rupture zone of high energy pipes may cause damage to adjacent structures. This event could lead to more serious accidents in nuclear power plants. Therefore, to prevent serious accidents, high energy pipes of nuclear power plants are designed according to the ANSI / ANS 58.2 technical standard. However, the US Nuclear Regulatory Commission (USNRC) has recently pointed out non-conservatism in existing high energy pipe fracture evaluation methods, and required the assessment of the unsteady load of the jet caused by a potential feedback mechanism as well as the impact range of steam jet, the jet impact loads and the blast wave effects at the initial breakage stage. The potential feedback mechanism refers to a phenomenon in which a vortex formed by impingement jets amplifies vortex itself and induces jet vibration in a shear layer. In this study, CFD methodology using the LES turbulence model is established and numerical analysis is carried out to evaluate the dynamic behavior of impingement jets and the potential feedback mechanism during jet impingement. Obtained results have been compared with an empirical correlation and experiment.

• Journal of Energy Engineering
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• v.13 no.2
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• pp.93-100
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• 2004

The smear media possible to sampling and radiation detection was prepared and evaluated for the surface contamination using indirect method. The films were made by impregnating Cerium Activated Yttrium Silicate (CAYS) in a polysulfone membrane. The membranes used solution as a dimethylformamide (DMF) and methylene chloride (MC), polysulfone as a polymer matrix and CAYS as a inorganic scintillator. The proximity membranes were prepared with single- and double-layered structure. The solidified methods were immersion to the nonsolvent bath such at water and ethanol and solvent evaporation. The measurement of the photon produced by interaction with radiation and inorganic scintillator used a photomultiflier tube (PMT), amplifier, and counter. In the comparison with the low background alpha/beta counter, the counter rate using inorganic scintillator proximity membrane for the $\^$14/C surface contamination was about 50%. Also. the $^3$H counting results revealed that the prepared membranes were efficient to monitor the surface contaminated with the low energy be-ray emitter nuclides.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.2
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• pp.77-87
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• 2018

In the case of high-energy line breaks in nuclear power plants, supersonic steam jet is formed due to the rapid depressurization. The steam jet can cause impingement load on the adjacent structures, piping systems and components. In order to secure the design integrity of the nuclear power plant, it is necessary to evaluate the load characteristics of the steam jet generated by high-energy pipe rupture. In the design process of nuclear power plant, jet impingement load evaluation was usually performed based on ANSI/ANS 58.2. However, U.S. NRC recently pointed out that ANSI/ANS 58.2 oversimplifies the jet behavior and that some assumptions are non-conservative. In addition, it is recommended that dynamic analysis techniques should be applied to consider transient load characteristics. Therefore, it is necessary to establish an evaluation methodology that can analyze the dynamic load characteristics of steam jet ejected when high energy pipe breaks. This research group has developed and validated the CFD analysis methodology to evaluate the transient behavior of supersonic impinging jet in the previous study. In this study, numerical study on the transient load characteristics of supersonic steam jet impingement was carried out and amplitude and frequency analysis of transient jet load was performed.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.2
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• pp.1-10
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• 2018

Structures, systems and components of nuclear power plants should be able to maintain safety even in the event of design-basis accidents such as high-energy line breaks. The high-pressure steam jet ejected from the broken pipe may cause damage to the adjacent structures. The ANSI/ANS 58.2 code has been adopted as a technical standard for evaluating the jet impingement load. Recently, the U.S. NRC pointed out the non-conservativeness of the ANSI/ANS 58.2, because it does not take into account the blast wave effect, dynamic behavior of the jet, and oversimplifies the shape and load characteristics of the supersonic steam jet. Therefore, it is necessary to improve the evaluation method for the high-energy line break accident. In order to evaluate the behavior of supersonic steam jet, an appropriate numerical analysis technique considering compressible flow effect is needed. In this study, numerical analysis methodology for evaluating supersonic jet impingement load was developed and verified. In addition, the conservativeness of the ANSI/ANS 58.2 model was investigated using the numerical analysis methodology. It is estimated that the ANSI jet model does not sufficiently reflect the physical behavior of under-expanded supersonic steam jet and evaluates the jet impingement load lower than CFD analysis result at certain positions.

• 한국신재생에너지학회:학술대회논문집
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• 2011.05a
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• pp.179.2-179.2
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• 2011

바이오매스 가스화 프로세스 개발에 있어서 가장 기본적인 해결과제는 고발열량의 합성가스 제조, 냉가스 효율의 향상, 타르 발생량 저감 및 제거이다. 가스화 효율 향상에 대한 연구는 국내외 적으로 많이 이루어지고 있으나, 타르 발생량 저감에 대한 연구는 많이 이루어져 있지 않다. 타르는 분자량이 큰 방향적 탄화수소로 응축되면 점성이 높아 배관폐쇄, 정제설비의 압력손실 증가로 인해 운전정지 및 가스화율 저하의 원인이 된다. 가스화로에서 타르 발생량을 저감시키는 방법 중에는 Ni계 촉매를 이용하는 방법이 있으나, 카본 누적에 의한 활성저하, 알칼리금속에 의한 응집 등의 문제가 발생할 수 있다. 한편 철산화물은 합성가스 중의 C2-C3계의 타르를 분해하는데 효과가 알려져 있다. 따라서 본 연구에서는 적벽돌, 염색슬러지 회재 등에는 철산화물이 다량 함유되어 있는 것에 착안하여 폐기물중의 폐금속을 이용한 바이오매스 가스화에 대한 연구를 수행하였다. 점토광물계 폐기물인 적벽돌 파쇄물($SiO_2$ 67.2%, $Al_2O_3$ 19.7%, $Fe_2O_3$ 8.7%, $K_2O$ 2.0%, $TiO_2$ 1.2%, MgO 0.7%)을 전처리 한 후 유동매체로하여 우드펠렛을 가스화한 결과, 가스 생성량이 증가하고, 타르 및 탄화수소류가 감소하는 경향을 나타내었다. 특히 타르는 후단의 타르 트랩에서 타르가 거의 검출이 되지 않았다. 전처리를 하지 않은 적벽돌 파쇄물은 반응시간이 경과한 후에 가스화율이 증가함에 따라 철화합이 가스화로내에서 환원되어 타르를 분해하는데에는 어느 정도의 반응시간이 필요한 것을 확인하였다.