• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회A
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• pp.285-290
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• 2007

KALIMER-600 is a sodium cooled fast reactor with a fast spectrum neutron reactor core. The NSSS design has three heat transport systems of a PHTS (Primary Heat Transport System), a IHTS (Intermediate Heat Transport System) and a SGS (Steam Generation System). PHTS is a pool type and has a large amount of sodium in the pool. The mechanical design targets are maintaining the enough structural integrity for a seismic load of SSE 0.3g and the thermal and mechanical loads by the high temperature environments and an economical competitiveness when compared with other reactor types.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2005년도 학술대회 논문집 정보 및 제어부문
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• pp.430-432
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• 2005

저에너지 중성자가 가톨리늄(Gd) 막에 입사되면 중성자 포획과정에서 전환전자가 생성된다. 이 전환전자에 의해 pMOSFET $SiO_2$ 산화층에서 발생된 전자-전공쌍이 발생되고, 이 가운데 정공은 산화층 내부에 쉽게 붙잡혀(Trap) 양전하 센터로 작용하게 된다. 이 축적된 전하는 pMOSFET의 문턱전압(Threshold voltage)을 변화시킨다. 본 연구에서는 이러한 간접측정 원리를 이용하여 열중성자를 실기간 탐지할 수 있는 반도체형 탐지소자를 개발하고 하나로(HANARO) 방사선장에서의 시험을 통해 성능을 검증하였다. 그리고 감도관련 변수의 최적화를 통하여 작업자가 사용 가능한 범위의 고감도 열중성자 선량계로 개선 제작하였다. 개발된 선량계는 소형으로 실시간 열중성자 측정이 가능하며 감마방사선으로부터 독립적으로 열중성자를 측정할 수 있는 장점도 지니고 있다.

• 한국지하수토양환경학회:학술대회논문집
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• 한국지하수토양환경학회 2003년도 추계학술발표회
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• pp.258-262
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• 2003

연안지역에서 해수침투대의 평가를 위하여 다양한 물리검층을 수행하였다. 특히, 해수침투대의 정량적인 평가에 활용될 수 있는 지층의 공극측정은 매우 어려운 문제중의 하나이다. 연안지역의 경우, 미고결지층에 대한 불교란 시료 채취가 어렵고 대부분의 관측정은 시추공 붕락방지를 위하여 내경 50mm 의 PVC 케이싱을 설치하는 경우가 많기 때문에 공극의 측정은 현실적으로 많은 어려움이 따른다. 본 연구에서는 전남 영광지역에서 각종 조사목적으로 굴착한 시추공에서 다양한 물리검층을 수행하여 공내수의 높은 전기전도도가 기원하는 지층을 확인하고자 하였다. 전자유도검층과 공극검층을 수행한 시추공(YK-4호공)에 대한 해석결과, 공내수의 높은 전기전도도는 물리검층법으로 추정한 사질층 공극수와 비슷한 범위를 보였다. 물리검층법으로 추정한 공극이나 높은 염수를 보이는 구간에 대한 해석결과는 보완해야할 많은 부분이 있지만 제한된 현장 상황에서 조사결과의 불화실성을 줄이는데 많은 기여를 할 것으로 기대된다.

• Corrosion Science and Technology
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• 제5권6호
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• pp.201-205
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• 2006

HYPER(Hybrid Power Extraction Reactor) is the accelerator driven subcritical transmutation system developed by KAERI(Korea Atomic Research Institute). HYPER is designed to transmute long-lived transuranic actinides and fission products such as Tc-99 and I-129. Liquid lead-bismuth eutectic (LBE). Has been a primary candidate for coolant and spallation neutron target due to its appropriate thermal-physical and chemical properties, However, it is very corrosive to the common steels used in nuclear installations at high temperature. This corrosion problem is one of the main factors considered to set the upper limits of temperature and velocity of HYPER system. In this study, a parametric study for a corrosion model was performed. And a preliminary corrosion model was also developed to predict the corrosion rate in isothermal Pb and LBE flow loops.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.605-610
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• 1997

A simple dynamic model is developed for the transient simulation of the nuclear power reactor. The dynamic model includes the normalized neutron kinetics model with reactivity feedback effects and the core thermal-hydraulics model. The main objective of this paper demonstrates the capability of the developed dynamic model to simulate various important variables of interest for a nuclear power reactor transient. Some representative results of transient simulations show the expected trends in all cases, even though no available data for comparison. In this work transient simulations are performed on a microcomputer using the DESIRE/N96T continuous system simulation language which is applicable to nuclear power reactor transient analysis.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1326-1338
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• 2017

The continuous energy Monte Carlo neutron transport code, MC21, was coupled to the CTF subchannel thermal-hydraulics code using a combination of Consortium for Advanced Simulation of Light Water Reactors (CASL) tools and in-house Python scripts. An MC21/CTF solution for VERA Core Physics Benchmark Progression Problem 6 demonstrated good agreement with MC21/COBRA-IE and VERA solutions. The MC21/CTF solution for VERA Core Physics Benchmark Progression Problem 7, Watts Bar Unit 1 at beginning of cycle hot full power equilibrium xenon conditions, is the first published coupled Monte Carlo neutronics/subchannel T-H solution for this problem. MC21/CTF predicted a critical boron concentration of 854.5 ppm, yielding a critical eigenvalue of $0.99994{\pm}6.8E-6$ (95% confidence interval). Excellent agreement with a VERA solution of Problem 7 was also demonstrated for integral and local power and temperature parameters.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1339-1345
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• 2017

In the framework of the EU funded project NURESAFE, the subchannel code CTF and the neutronics code DYN3D were integrated and coupled on the NURESIM platform. The developments achieved during this 3-year project include assembly-level and pin-by-pin multiphysics thermal hydraulics/neutron kinetics coupling. In order to test this coupling, a PWR rod ejection transient was simulated on a MOX/UOX minicore. The transient is simulated using two different models of the minicore. In the first simulation, both codes model the core with an assembly-wise resolution. In the second simulation, a pin-by-pin fuel-centered model is used in CTF for the central assembly, and a pin power reconstruction method is applied in DYN3D. The analysis shows the influence of the different models on global parameters, such as the power and the average fuel temperature, but also on local parameters such as the maximum fuel temperature.

• Journal of Radiation Protection and Research
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• 제12권2호
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• pp.19-27
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• 1987

황산망간 용액조장치의 $^{56}Mn\;{\gamma}$선 검출효율을 결정하는데 $^{56}Mn$용액의 방사능을 절대측정하는 것은 필수적이다 $^{56}Mn$시료를 제작하기 위하여 99.99%의 순도를 갖는 Mn금속조각 13.718mg되는 시료를 한국에너지연구소 TRIGA MARK-II 원자로의 중성자선속이 약 $10^{13}n/cm^2{\cdot}s$되는 열중성자장에서 12분간 조사시켰다. 중성자 방사화된 $^{56}Mn$금속시료를 0.1N-HCI 용액 50ml 용해시켜서 $^{56}Mn$시료를 제작하여 $4{\pi}{\beta}-{\gamma}$ 동시계수기술로 방사능을 측정한 결과 불확도 0.366%를 갖는 값으로서 1987년 10월 15일 0 시를 기준하여 408.070kBq/mg을 얻었다.

• Nuclear Engineering and Technology
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• 제51권8호
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• pp.2018-2025
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• 2019

This study systematically compared two hybrid deterministic/Monte Carlo transport codes, ADVANTG/MCNP and MAVRIC, in solving a difficult shielding problem for a real-world spent fuel storage cask. Both hybrid codes were developed based on the consistent adjoint driven importance sampling (CADIS) methodology but with different implementations. The dose rate distributions on the cask surface were of primary interest and their predicted results were compared with each other and with a straightforward MCNP calculation as a baseline case. Forward-Weighted CADIS was applied for optimization toward uniform statistical uncertainties for all tallies on the cask surface. Both ADVANTG/MCNP and MAVRIC achieved substantial improvements in overall computational efficiencies, especially for gamma-ray transport. Compared with the continuous-energy ADVANTG/MCNP calculations, the coarse-group MAVRIC calculations underestimated the neutron dose rates on the cask's side surface by an approximate factor of two and slightly overestimated the dose rates on the cask's top and side surfaces for fuel gamma and hardware gamma sources because of the impact of multigroup approximation. The fine-group MAVRIC calculations improved to a certain extent and the addition of continuous-energy treatment to the Monte Carlo code in the latest MAVRIC sequence greatly reduced these discrepancies. For the two continuous-energy calculations of ADVANTG/MCNP and MAVRIC, a remaining difference of approximately 30% between the neutron dose rates on the cask's side surface resulted from inconsistent use of thermal scattering treatment of hydrogen in concrete.

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.847-858
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• 2016

An overview is given on the work of the Laboratory of Nuclear Energy Systems at ETH, Zurich (ETHZ) and of the Laboratory of Thermal Hydraulics at Paul Scherrer Institute (PSI), Switzerland on tight-lattice bundles. Two-phase flow in subchannels of a tight triangular lattice was studied experimentally and by computational fluid dynamics simulations. Two adiabatic facilities were used: (1) a vertical channel modeling a pair of neighboring sub-channels; and (2) an arrangement of four subchannels with one subchannel in the center. The first geometry was equipped with two electrical film sensors placed on opposing rod surfaces forming the subchannel gap. They recorded 2D liquid film thickness distributions on a domain of $16{\times}64$ measuring points each, with a time resolution of 10 kHz. In the bubbly and slug flow regime, information on the bubble size, shape, and velocity and the residual liquid film thickness underneath the bubbles were obtained. The second channel was investigated using cold neutron tomography, which allowed the measurement of average liquid film profiles showing the effect of spacer grids with vanes. The results were reproduced by large eddy simulation + volume of fluid. In the outlook, a novel nonadiabatic subchannel experiment is introduced that can be driven to steady-state dryout. A refrigerant is heated by a heavy water circuit, which allows the application of cold neutron tomography.