• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.23 no.1
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• pp.49-55
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• 2013

A reactor vessel internals comprehensive vibration assessment program(RVI CVAP) of an advanced power reactor 1400(APR1400) is being performed as a non-prototype category-2 type of reactor based on the US nuclear regulatory commission regulatory guide(NRC RG) 1.20. The aim of this paper is to present the results of structural response analysis and measuring locations of a upper guide structure(UGS) assembly of the APR1400 reactor. The analysis results of the UGS assembly show that the specified integrity levels meet the design acceptance criteria. Also, the measuring locations are determined by the analysis results of the UGS assembly and selection criteria of previous study. These analysis results and measuring locations will be used as a guide to design a measurement system for the APR1400 RVI CVAP.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.22 no.5
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• pp.474-479
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• 2012

A comprehensive vibration assessment program for the advanced power reactor 1400(APR1400) reactor vessel internals is established in accordance with the united states nuclear regulatory commission regulatory guide 1.20 revision 3. This paper is related to instruments and measurement locations based on the vibration and stress response analysis results of the inner barrel assembly top plate in APR1400. The analysis results of the inner barrel assembly top plate in the reactor show that the deterministic stress and deformation due to the reactor coolant pump induced pressure pulsations are larger than the random stress and deformation induced by the flow turbulence. The selection of the instruments and measurement locations at inner barrel assembly top plate in the reactor is essential requirements and very important study process for the vibration and stress measurement program in comprehensive vibration assessment program for APR1400 reactor vessel internals.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.4
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• pp.253-262
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• 2015

Radionuclide deposited on the surface of several internal and external systems in a nuclear power plant is created by the activation of corrosion products from nuclear reactor structural materials and fission products. Especially, the constant contact between water and the surface corrodes the inside where primary system makes coolants and corrosion products mixed. Also, these are circulated along the systems. For comparing models, CRUDTRAN, DISER, MIGA-RT and CPAIR codes are analyzed to predict the quantity of radionuclide and corrosion product of primary reactor that are used at the stage of designing. The corrosion products behavior of domestic PWR primary system was predicted by using CRUDTRAN. This study aims to increase the reliability of corrosion product evaluation model by comparing the actual values and calculated values with the data of a Westing House-type Nuclear Power Plant.

• Journal of the Mineralogical Society of Korea
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• v.25 no.4
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• pp.283-293
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• 2012

In order to have better insights into the chemical differentiation of Earth from its magma ocean phase to the current stratified structure, detailed information of crystallization kinetics of silicate melts consisting of the magma ocean is essential. The structural transitions in oxide glasses and melts upon crystallization provide improved prospects for a systematic and quantitative understanding of the crystallization processes. Here, we report the $^{27}Al$ 3QMAS NMR spectra for sol-gel synthesized $Al_2O_3$ glass with varying temperature and annealing time. The NMR spectra for the amorphous $Al_2O_3$ show well-resolved Al coordination environments, characterized with mostly $^{[4,5]}Al$ and a minor fraction of $^{[6]}Al$. The fraction of $^{[5]}Al$ in the alumina phase decreases with increasing annealing time at constant temperature. The NMR results of $Al_2O_3$ phases also imply that multiple processes (e.g., crystallization and/or changes in structural disorder within glasses) could involve upon its phase transition. The current results and method can be useful to understand crystallization kinetics of diverse natural and multi-component silicate glasses and melts. The potential result may yield atomic-level understanding of Earth's chemical evolution and differentiation from the magma ocean.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2012.10a
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• pp.300-305
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• 2012

The vibration and stress analysis program of comprehensive vibration assessment program (CVAP) is to verify theoretically the structural integrity of reactor vessel internals (RVI) and to provide the basis for selecting the locations monitored in measurement and inspection programs. This paper covers the verification of the vibration and stress analysis methodology of APR1400 RVI CVAP. The analysis methodology was developed to use 3-dimensional hydraulic and structural models with ANSYS and CFX. To validate the methodology, the hydraulic loads and structural reponses of OPR1000 were predicted and compared with the calculated and measured data in the OPR1000 RVI CVAP. Since the results predicted with this methodology were close to the measured values considerably, it was confirmed that the analysis methodology was developed properly.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.28 no.5
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• pp.535-542
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• 2015

A reactor containment acts as a final barrier to prevent leakage of radioactive material due to the possible reactor accidents into external environment. Because of the functional importance of the containment building, the SIT(Structural Integrity Test) for containments shall be performed to evaluate the structural acceptability and demonstrate the quality of construction. In this paper, numerical analyses are presented, which simulate the results obtained from the SIT for a prestressed concrete(PSC) structure. A sophisticate structural analysis model is developed to simulate the structural behavior during the SIT properly based on various preliminary analysis results considering contact condition among structural elements. From the comparison of the analysis and test results based on the acceptance criteria of ASME CC-6000, it can be concluded that the construction quality of the containment has been well maintained and the acceptable performance of new design features has been verified.

• Proceedings of the Korean Vacuum Society Conference
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• 2010.02a
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• pp.129-129
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• 2010

적외선 검출소자(Infrared Photodetector)는 근적외선에서 원적외선 영역에 이르는 광범위한 파장 범위의 적외선을 이용하는 기기로서 대상물이 방사하는 적외선 영역의 에너지를 흡수하여 이를 영상화할 수 있는 장비이다. 적외선 관련 기술은 2차 세계대전 기간에 태동하였으며, 현재에는 원거리 감지기술 등과 접목되면서 그 활용 분야가 다양해지고 있다. 특히 능동형 정밀 타격무기를 비롯한 감시 정찰 장비 및 지능형 전투 장비 시스템 등에 대한 요구를 바탕으로 보다 정밀하고 신속한 표적 감지 및 정보처리 기술에 관한 연구가 선진국을 통해서 활발히 진행되고 있다. 기존의 Bolometer 형식의 열 감지 소자는 반응 속도가 느리고 측정 감도가 낮은 단점이 있으며, MCT(HgCdTe)를 이용한 적외선 검출기의 경우 높은 기계적 결함과 77K 저온에서 동작해야하기 때문에 발생하는 추가 비용 등이 문제점으로 지적되고 있다[1]. 이에 반해 화합물 반도체 자기조립 양자점(self-assembled quantum dot)을 이용한 적외선 수광소자는 양자점이 가지는 불연속적인 내부 에너지 준위로 인하여, 높은 내부 양자 효율과 온도 안정성을 기대할 수 있으며, 고성능, 고속처리, 저소비전력 및 저소음의 실현이 가능하다. 본 연구에서는 적층 InAs/InGaAs dot-in-a-well 구조를 유기금속화학기상증착법을 이용하여 성장하고 이를 소자에 응용하였다. 균일한 적층 양자점의 성장을 위해서 원자현미경(atomic force microscopy)을 이용하여, 각 층의 양자점의 크기와 밀도를 관찰하였고, photoluminescence (PL)를 이용하여 발광특성을 연구하였다. 각 층간의 GaAs space layer의 두께와 온도 조절 과정을 조절함으로써 균일한 적층 양자점 구조를 얻을 수 있었다. 이를 이용하여 양자점의 전도대 내부의 에너지 준위간 천이(intersubband transition)를 이용하는 n-type GaAs/intrinsic InAs 양자점/n-type GaAs 구조의 양자점 적외선수광소자 구조를 성장하였다. 이 과정에서 상부 n-type GaAs의 성장 온도가 600도 이상이 되는 경우 발광효율이 급격히 감소하고, 암전류가 크게 증가하는 것을 관찰하였다. 이는 InAs 양자점과 주변 GaAs 간의 열에 의한 상호 확산에 의하여 양자점의 전자 구속 효과를 저해하는 것으로 설명된다.

• Proceedings of the Materials Research Society of Korea Conference
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• 2003.11a
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• pp.119-119
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• 2003

탄화규소나 열분해 탄소는 고온 특성 및 화학적인 안정성 이 우수하여 단미 혹은 코팅재로로 소재의 성능을 향상시키기 위하여 에너지 관련 분야, 반도체 치구 분야, 방위산업 및 항공우주 분야와 원자력 분야에서 다양하게 사용된다. 특히 원자력 분야에서는 고온형 원자로의 노심 요소 부품으로 적용 및 개발을 고려하고 있으며, 대표적인 예로 수소생산용 초고온 가스냉각로의 코팅 핵연료 입자를 들 수 있다. 일반적으로 TRISO라 불리는 가스냉각로 핵연료는 구형 $UO_2$ kemel의 주변을 PyC-SiC -PyC의 삼중 코팅층으로 둘러싸는 구조를 하고 있으며, 이 코팅층들은 kernel물질이 분열하는 동안 발생되는 내부 기체 압력을 견디는 압력용기 역할과 기체나 금속 핵분열 생성물들을 가두는 확산 장벽 역할을 하게 된다. 본 연구에서는 구형의 $UO_2$대신 선행연구를 위하여 구형 ZrO$_2$를 이용하여 증착온도나 시간 및 입력기체비 등의 화학증착 변수로 조절하여 SiC 및 PyC을 코팅하고, 각 변수들에 의한 증착층의 거동을 고찰하고자 하였다.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.28 no.5
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• pp.523-533
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• 2015

A reactor containment acts as a final barrier to prevent leakage of radioactive material due to the possible reactor accidents into external environment. Because of the functional importance of the containment building, the SIT(Structural Integrity Test) for containments shall be performed to evaluate the structural acceptability and demonstrate the quality of construction. An initial numerical analysis was performed to simulate the results obtained from the SIT for a prestressed concrete(PSC) structure. But the analysis results by the initial model expected smaller displacements than the measured ones by 30% at some locations. Accordingly, the research and development to improve the initial model to corelate the measured results of the SIT more properly have been performed. In this paper, the effects of the loss of concrete due to duct for tendons and the contact of duct and tendons in un-bonded tendon system are mainly evaluated based on the preliminary analysis results. In addition, the importances of the proper definition of mesh connectivity among structural elements of concrete, liner plates, rebars and tendons are discussed.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.25 no.1
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• pp.48-57
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• 2015

In a sodium-cooled fast reactor, which is a Generation-IV reactor, refueling is conducted by rotating, but not opening, the reactor head to prevent a reaction between the sodium, water and air. Therefore, an inspection technique that checks for the presence of any obstacles between the reactor core and the upper internal structure, which could disturb the rotation of the reactor head, is essential prior to the refueling of a sodium-cooled fast reactor. To this end, an ultrasound-based inspection technique should be employed because the opacity of the sodium prevents conventional optical inspection techniques from being applied to the monitoring of obstacles. In this study, a ranging inspection technique using a plate-type ultrasonic waveguide sensor was developed to monitor the presence of any obstacles between the reactor core and the upper internal structure in the opaque sodium. Because the waveguide sensor installs an ultrasonic transducer in a relatively cold region and transmits the ultrasonic waves into the hot radioactive liquid sodium through a long waveguide, it offers better reliability and is less susceptible to thermal or radiation damage. A 10 m horizontal beam waveguide sensor capable of radiating an ultrasonic wave horizontally was developed, and beam profile measurements and basic experiments were carried out to investigate the characteristics of the developed sensor. The beam width and propagation distance of the ultrasonic wave radiated from the sensor were assessed based on the experimental results. Finally, a feasibility test using cylindrical targets (corresponding to the shape of possible obstacles) was also conducted to evaluate the applicability of the developed ranging inspection technique to actual applications.