• Nuclear Engineering and Technology
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• 제36권6호
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• pp.528-539
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• 2004

It is essential in commercial reactors that the safety limits imposed on the fuel pellets and fuel clad barriers, such as the linear power density (LPD) and the departure from nucleate boiling ratio (DNBR), are not violated during reactor operations. In order to accurately monitor the safety limits of current reactor states, a detailed three-dimensional (3D) core power distribution should be estimated from the in-core detector signals. In this paper, we propose a calculation methodology for detailed 3D core power distribution, using in-core detector signals and core monitoring constants such as the 3D Coupling Coefficients (3DCC), node power fraction, and pin-to-node factors. Also, the calculation method for several core safety parameters is introduced. The core monitoring constants for the real core state are promptly provided by the core design code and on-line MASTER (Multi-purpose Analyzer for Static and Transient Effects of Reactors), coupled with the core monitoring program. through the plant computer, core state variables, which include reactor thermal power, control rod bank position, boron concentration, inlet moderator temperature, and flow rate, are supplied as input data for MASTER. MASTER performs the core calculation based on the neutron balance equation and generates several core monitoring constants corresponding to the real core state in addition to the expected core power distribution. The accuracy of the developed method is verified through a comparison with the current CECOR method. Because in all the verification calculation cases the proposed method shows a more conservative value than the best estimated value and a less conservative one than the current CECOR and COLSS methods, it is also confirmed that this method secures a greater operating margin through the simulation of the YGN-3 Cycle-1 core from the viewpoint of the power peaking factor for the LPD and the pseudo hot pin axial power distribution for the DNBR calculation.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1472-1482
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• 2017

Zirconium alloy cladding tube specimens were irradiated at $380^{\circ}C$ up to a fast neutron fluence of $7.5{\times}10^{24}n/m^2$ in a research reactor to investigate the effect of neutron irradiation on hydride reorientation and mechanical property degradation. Cool-down tests from $400^{\circ}C$ to $200^{\circ}C$ under 150 MPa tensile hoop stress were performed. These tests indicate that the irradiated specimens generated a smaller radial hydride fraction than did the unirradiated specimens and that higher hydrogen content generated a smaller radial hydride fraction. The irradiated specimens of 500 ppm-H showed smaller ultimate tensile strength and plastic strain than those characteristics of the 250 ppm-H specimens. This mechanical property degradation caused by neutron irradiation can be explained by tensile hoop stress-induced microcrack formation on the hydrides in the irradiation-damaged matrix and subsequent microcrack propagation along the hydrides and/or through the matrix.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2009년도 제40회 하계학술대회
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• pp.932_933
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• 2009

A technology based on thyristors will be used to manufacture the super-conducting coil AC/DC converters because of the low ratio of cost over installed power compared to a design based on GTO or similar technology. But phase-controlled converter suffers from fundamental disadvantage. They inject current harmonics into the input ac mains due to their nonlinear characteristics, thereby distort the supply voltage waveform, and demand reactive power from the associated ac power system at retarded angles. To overcome this disadvantage, in the case of two series converters at the DC side, connected to the same step-down transformer, apply for the sequence control. It is the most simple and efficient way to reduce the reactive power consumption at low cost. Analytical sequence control algorithm is suggested, the validity of the proposed scheme has been verified by experimental results with the small-scaled International Thermonuclear Experimental Reactor (ITER) Power Supply to minimize reactive power consumption.

• Nuclear Engineering and Technology
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• 제27권3호
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• pp.393-402
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• 1995

원자로냉각재계통의 설계를 위한 구조해석 분야에는 원자로의 정상운전 과정에서 발생하는 유체의 온도와 압력의 변화에 의해 냉각재계통에 발생하는 정적하중해석, 지진과 가상적인 분지관 파단사고에 의해 냉각재계통에 발생하는 동적하중해석분야로 구분할 수 있다. 원자로냉가재계통의 구조해석은 원자력발전소의 안전성 화보 측면을 중시하여 해석시 충분한 여유도를 고려한 보수적인 해석 방법을 원용한다. 지진이나 가상적인 분지관 파단사고에 의한 냉각재계통의 구조해석은 사고시 냉각재계통의 안전성을 유지하는 방어적인 개념으로서 기기의 건전성을 확보하기 위하여 충분한 보수성과 안전여유가 해석시 고려된다 정상운전에 의해 냉각재계통에 발생하는 하중은 원자력 발전소의 상존하는 하중의 개념으로서 냉각재계통의 기본 설계 하중으로 인식된다. 특히 고온 고압의 유체로 인하여 발생하는 냉각재 계통의 열팽창 현상은, 정상운전 하중으로 인하여 나타나는 전형적인 거동으로서, 냉각재계통 구조해석 결과읜 중요한 지표로서 인식된다. 따라서 냉각재계통의 열팽창 현상을 정확히 예측하는 것은 원자로 냉각재계통 구조해석의 가장 중요한 목표중의 하나이다. 본 연구에서는 정상운전 하중에 의한 원자로 냉가재계통의 열팽창 거동을 해석하기 위한 냉각재계통의 모델링 방법과 해석 방법을 제시하였다. 해석 결과의 타당성을 검토하기 위하여 최근 건설 완료 단계에 돌입한 표준형 1000 MWe 급 가압경수로(Pn)의 고온기능시험 (Hot Function Test)과정에서 실측한 자료를 근거로 하여 원자로냉각재계통의 열팽창 거동 해석의 타당성을 입증코자 하였다.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.249-256
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• 2019

The 9Cr-1Mo ferritic-martensitic ODS steel is a promising structural material for the next generation nuclear power plants including fast reactors for application in reactor vessels and nuclear fuel. The ODS steel was cooled down by furnace cooling, air cooling, oil quenching and water quenching, respectively, after normalizing it at $1150^{\circ}C$ for 1 h and then tempering at $780^{\circ}C$ for 1 h. It is found that grain size, a relative portion of ferrite and martensite, martensitic lath configuration, behaviors of carbide precipitates, and hardness of the ODS steel are strongly dependent on a cooling rate. The grain size and martensitic lath width become smaller with the increase in a cooling rate. The carbides were precipitated at the grain boundaries formed between the ferrite and martensite phases and at the martensitic lath interfaces. In addition, the carbide precipitates become smaller and more widely dispersed with the increase in a cooling rate, resulting in that the faster cooling rate generated the higher hardness of the ODS steel.

• 센서학회지
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• 제18권5호
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• pp.402-407
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• 2009

In this study, plastic optical fiber sensors which can measure thermal neutron dose in a mixed neutron-gamma field are developed and characterized. Using $^{252}Cf$ and $^{60}Co$ sources, the scintillators suitable for thermal neutron detection, are tested and the scintillating lights generated from a plastic optical fiber sensor in the Kyoto University Critical Assembly (kuca) core are measured. Also, the distributions of thermal neutron and gamma-ray are measured in a mixed field as a function of the distance from the center of the reactor core at KUCA and the distribution of thermal neutron is obtained using a subtraction method. Sensitivity of the fiber-optic radiation sensor system is about 0.49 V/mW according to power of the KUCA core and its relative error is about 1.2 %.

• Nuclear Engineering and Technology
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• 제41권4호
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• pp.493-520
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• 2009

During the last four decades, 16 Pressurized Water Reactors (PWR) and 4 Pressurized Heavy Water Reactors (PHWR) have been constructed and operated in Korea, and nuclear fuel technology has been developed to a self-reliant state. At first, the PWR fuel design and manufacturing technology was acquired through international cooperation with a foreign partner. Then, the PWR fuel R&D by Korea Atomic Energy Research Institute (KAERI) has improved fuel technology to a self-reliant state in terms of fuel elements, which includes a new cladding material, a large-grained $UO_2$ pellet, a high performance spacer grid, a fuel rod performance code, and fuel assembly test facility. The MOX fuel performance analysis code was developed and validated using the in-reactor test data. MOX fuel test rods were fabricated and their irradiation test was completed by an international program. At the same time, the PWR fuel development by Korea Nuclear Fuel (KNF) has produced new fuel assemblies such as PLUS7 and ACE7. During this process, the design and test technology of fuel assemblies was developed to a self-reliant state. The PHWR fuel manufacturing technology was developed and manufacturing facility was set up by KAERI, independently from the foreign technology. Then, the advanced PHWR fuel, CANFLEX(CANDU Flexible Fuelling), was developed, and an irradiation test was completed in a PHWR. The development of the CANFLEX fuel included a new design of fuel rods and bundles.. The nuclear fuel technology in Korea has been steadily developed in many national R&D programs, and this advanced fuel technology is expected to contribute to a worldwide nuclear renaissance that can create solutions to global warming.

• KSBB Journal
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• 제20권5호
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• pp.341-344
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• 2005

섬유상 담체를 이용한 2단 소형 바이오필터를 이용한 악취가스 처리공정의 운전조건을 확립하기 위하여 BTX를 처리대상 물질로 바이오필터를 연속적으로 운전하여 섬유상 바이오필터 담체에 악취가스 처리용 미생물을 부착시킨 결과 우수한 부착된 생물막을 형성하여 높은 처리효율 $93\%$를 얻을 수 있었다. 또한, 악취 유입 농도 400 ppm 미만에서 체류시간 10 sec일 때 1단 반응기에서 $76\%$ 정도 제거율을 보이다가 2단 반응기에서 약 $93\%$의 제거율을 나타내었다. 그러나 체류시간이 15 sec 이상인 경우에는 제거효율이 $97\%$ 이상 유지되었다.

• 한국환경과학회지
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• 제20권3호
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• pp.291-299
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• 2011

The main objectives of this research are to investigate characteristics of ozone solubility due to low solubility of conventional bubbles-ozone generators, evaluate the treatment characteristics of reclaiming textile wastewater for industrial water by means of micro/nano bubbles-dissolved ozone flotation(MNB-DOF) process. The textile wastewater used in this research was obtained from final effluent of the textile wastewater in B city. There is a 400L reactor which consists of a micro-nano bubble system and a ozone generator for experiments. As a result of generating micro-nano bubbles (below $0.5{\mu}m$) by using of MNB-DOF process, it improved ozone solubility due to higher ozone transfer rates. Consequently, the shorter ozonation time clearly indicates the lower power costs. The reported results clearly indicated that MNB-DOF process can be effectively and inexpensively. Results of the experiments through MNB-DOF process in this study satisfy all reclaiming standards as industrial water: pH 6.5~8.5, SS 10 mg/L or below, $BOD_5$) 6 mg/L or below, turbidity 10 NTU or below, Coliforms 1,000/100 mL or below. Therefore there is a possibility of the reclaiming of the textile wastewater as industrial water.

• Ecology and Resilient Infrastructure
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• 제4권3호
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• pp.142-148
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• 2017

본 연구에서는 부영양화의 원인이 되는 질소와 인의 제거 효율을 향상시킬 목적으로 광물질을 활용한 여재를 이용하여 호기, 혐기의 흐름조건을 만들어주기 위한 수직 수평 흐름 인공습지를 고안하여 아크릴 반응조로 제작한 후 성능평가를 진행하였다. 수직 수평 흐름형 인공습지의 경우 호기 및 혐기조건을 평가하기 위해서 반응조 내 용존산소(DO) 농도를 측정해본 결과 호기상태에서는 2.7 mg/L, 혐기상태에서는 N.D로 확인되어 목적에 부합된 결과가 확인되었다. 실험결과에서는 SS 저감효과가 140 min, 80 min, 60 min의 운전시간에서 각각 94%, 91%, 61%의 효율을 보였고, T-P의 경우 각각의 운전시간에 따라 84%, 71%, 63%의 저감효율을 보였다. 또한 T-N의 경우 각각의 운전시간에 따라 63%, 49%, 42%의 저감효율을 보여 기존의 습지가 12~24 hr 체류시간을 가지는 것에 비하여 짧은 운전시간에도 높은 처리효율을 보이는 것으로 확인되었다. 본 연구에서는 수직 수평 흐름 방식을 복합 적용하여 기존 인공습지의 단점을 보완하기 위해 기술개발을 진행한 것으로 어떠한 기능적 효과를 갖는지 확인하였고, 향후 이에 대한 현장적용을 위한 운영 및 관리적 차원의 메커니즘 연구가 추가로 진행 될 필요가 있다.