• 한국지하수토양환경학회:학술대회논문집
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• 한국지하수토양환경학회 2003년도 추계학술발표회
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• pp.267-271
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• 2003

대수층 저장 및 회수(Aquifer Storage and Recovery, 이하 ASR) 기법은 지표수가 여유가 있는 기간동안 지하 대수층에 정수된 물을 주입하여 두었다가, 첨두 수요(peak demand)가 필요한 한시적인 기간동안 대수층 내에 저장되어 있던 물을 양수하여 공급하는 새로운 개념의 수자원 공급 체계의 일환이다. 본 발표에서는 미국 Oregon 주의 대표적인 3개 도시에서 수행된 수자원 공급 사업과 이 수자원 공급사업에서 ASR 기법이 어떻게 적용되었는지를 살펴보았다. 이들 도시에서 수행된 ASR 기법을 활용한 수자원 공급은, 기존 지표 저류조 저장 방식에 비해 월등히 경제적이며, 다양한 이점을 가지고 있는 것으로 드러났다. 또한 이들 도시에서의 ASR 기법 적용에 따른 효율성이 2001년도 및 2002년도에 분명하게 나타나 현재 이들 도시는 ASR 시설에 대한 확충 계획을 수립, 진행하고 있다. 한편, 이들 지역의 ASR 기법 활용 용수 공급 사례는, 국내의 경우에도 투수성이 우수한 화산암 지역, 제3기 지층 및 석회암 지역 등에서 활용될 경우, 충분한 효과를 발휘 할 수 있을 것으로 기대된다.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.170-175
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• 2019

Displacement cross-sections for an advanced assessment of radiation damage rates were obtained for a number of materials using the arc-dpa model at neutron incident energies from $10^{-5}eV$ to 10 GeV. Evaluated data files, CEM03 and ECIS codes, and an approximate approach were applied for the calculation of recoil energy distributions in neutron induced reactions. Three sets of displacement cross-sections based on the use of low-energy data from JEFF-3.3, ENDF/B-VIII.0, and JENDL-4.0u were prepared. Files contain also cross-sections calculated using the standard NRT model. Special efforts were made to estimate the uncertainty of obtained displacement cross-sections.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3260-3271
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• 2022

Experience shows that the solution to waste management in any national programme is lengthy and burdened with uncertainties. There are several uncertainties that contribute to the costs associated with spent fuel management. In this work, we have analysed the impact of the current nuclear data on the isotopic composition of the spent fuel and consequently their influence on the main spent fuel observables such as decay heat, activity, neutron multiplication factor, and neutron and photon source terms. Nuclear libraries based on the most general nuclear data ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0 and JEFF-3.3 are considered. A typical NPP Krško fuel assembly is analysed using the Monte Carlo code Serpent 2. The analysis considers burnup of up to 60 GWd/tU and cooling times of up to 100 years. The comparison of results showed significant differences, which should be taken into account when selecting the library and evaluating the uncertainty in determining the characteristics of the spent fuel.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1460-1471
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• 2024

2D and 3D ultrasonic imaging has so far not been demonstrated in pure molten lead in the open literature. In this study the development of such an ultrasonic device for imaging is outlined and results from testing at 380 ℃ in lead are presented. The main difficulties were found to be achieving then maintaining suitable wetting while ensuring suitable durability of the device, both due to the harsh nature of molten lead and the elevated temperatures. The successful detection and imaging (2D and 3D), of differently shaped targets, where the features were above the size of the transmitted ultrasound beam was demonstrated.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.304-309
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• 2023

In the last decade, nuclear medicine appears to be a good choice of medicine. ⁵⁸Co, ⁹⁹Mo, ⁹⁹Tc, ⁹⁹Re, ¹³³Xe and ¹⁸⁶Re are very important radionuclides for nuclear medicine. In this study, the excitation functions of ⁵⁸Ni (n, p) ⁵⁸Co, ⁹⁹Tc (n, p) ⁹⁹Mo, ⁹⁹Ru (n, p) ⁹⁹Tc, ¹³¹Xe (n, p) ¹³¹I, ¹³³Cs (n, p) ¹³³Xe and ¹⁸⁶Os (n, p) ¹⁸⁶Re nuclear reactions were calculated at neutron energies between 1 and 20 MeV using TALYS 1.95 and EMPIRE 3.2 nuclear codes. Furthermore, the cross sections were calculated with the empirical formula derived in our past study at 14-15 MeV. The obtained results were compared with the measured values in EXFOR library, and with the evaluated data of (JENDL-4.0/HE, JEFF-3.3, TENDL-2019, ENDF/B-VIII.0, IRDFF-II, JENDL/ImPACT-18). The results are in good agreement with those of the evaluated data libraries and experimental results and indicates that these radioisotopes can be produced by smaller cyclotrons.

• Journal of the Korean Physical Society
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• 제73권10호
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• pp.1535-1540
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• 2018

In this study, we prepared and characterized Nb-doped $Li_7La_3Zr_{2-x}O_{12}$ (LLZNO) powder and pellets with a cubic garnet structure by using a modified sol-gel synthesis and hot pressing. LLZNO powder with a very small grain size and cubic structure without secondary phases could be obtained by using a synthesis method in which Li and La sources in a propanol solvent were mixed together with Zr and Nb sources in 2-methoxy ethanol. A pure cubic phase LLZNO pellet could be fabricated from the prepared LLZNO and an additional 6-wt% of $Li_2CO_3$ powder by hot pressing at $1050^{\circ}C$ and 15.8 MPa. The hot-pressed LLZNO pellet with a relative density of 99% exhibited a very dense surface morphology. The total Li ionic conductivity of the hot-pressed LLZNO was $7.4{\times}10^{-4}S/cm$ at room temperature, which is very high level compared to other reported values. The activation energy for ionic conduction was estimated to be 0.40 eV.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1593-1603
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• 2023

In deterministic and Monte Carlo transport codes, b-delayed emission is included using a group structure where all of the precursors are grouped together in 6 groups or families, but given the increase in computational power, nowadays there is no reason to keep this structure. Furthermore, there have been recent efforts to compile and evaluate all the available b-delayed neutron emission data and to measure new and improved data on individual precursors. In order to be able to perform a transient Monte Carlo simulation, data from individual precursors needs to be implemented in a transport code. This work is the first step towards the development of a tool to explore the effect of individual precursors in a fissile system. In concrete, individual precursor data is included by expanding the capabilities of the open source Monte Carlo code OpenMC. In the modified code - named Time Dependent OpenMC or OpenMC(TD)- time dependency related to β-delayed neutron emission was handled by using forced decay of precursors and combing of the particle population. The data for continuous energy neutron cross-sections was taken from JEFF-3.1.1 library. Regarding the data needed to include the individual precursors, cumulative yields were taken from JEFF-3.1.1 and delayed neutron emission probabilities and delayed neutron spectra were taken from ENDF-B/VIII.0. OpenMC(TD) was tested in a monoenergetic system, an energy dependent unmoderated system where the precursors were taken individually or in a group structure, and in a light-water moderated energy dependent system, using 6-groups, 50 and 40 individual precursors. Neutron flux as a function of time was obtained for each of the systems studied. These results show the potential of OpenMC(TD) as a tool to study the impact of individual precursor data on fissile systems, thus motivating further research to simulate more complex fissile systems.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.642-649
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• 2016

Important safety parameters such as the fuel temperature coefficient (FTC) and the power coefficient of reactivity (PCR) of the CANada Deuterium Uranium (CANDU-6) reactor have been evaluated using the Monte Carlo method. For accurate analysis of the parameters, the Doppler broadening rejection correction scheme was implemented in the MCNPX code to account for the thermal motion of the heavy uranium-238 nucleus in the neutron-U scattering reactions. In this work, a standard fuel lattice has been modeled and the fuel is depleted using MCNPX. The FTC value is evaluated for several burnup points including the mid-burnup representing a near-equilibrium core. The Doppler effect has been evaluated using several cross-section libraries such as ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1.1, and JENDL-4.0. The PCR value is also evaluated at mid-burnup conditions to characterize the safety features of an equilibrium CANDU-6 reactor. To improve the reliability of the Monte Carlo calculations, we considered a huge number of neutron histories in this work and the standard deviation of the k-infinity values is only 0.5-1 pcm.

• KSII Transactions on Internet and Information Systems (TIIS)
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• 제3권5호
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• pp.423-443
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• 2009

We present the design and implementation of a unique sensing and actuation application -- the Illuminator: a sensor network-based intelligent light control system for entertainment and media production. Unlike most sensor network applications, which focus on sensing alone, a distinctive aspect of the Illuminator is that it closes the loop from light sensing to lighting control. We describe the Illuminator's design requirements, system architecture, algorithms, implementation and experimental results. The system uses the Illumimote, a multi-modal and high fidelity light sensor module well-suited for wireless sensor networks, to satisfy the high-performance light sensing requirements of entertainment and media production applications. The Illuminator system is a toolset to characterize the illumination profile of a deployed set of fixed position lights, generate desired lighting effects for moving targets (actors, scenic elements, etc.) based on user constraints expressed in a formal language, and to assist in the set up of lights to achieve the same illumination profile in multiple venues. After characterizing deployed lights, the Illuminator computes optimal light settings at run-time to achieve a user-specified actuation profile, using an optimization framework based on a genetic algorithm. Uniquely, it can use deployed sensors to incorporate changing ambient lighting conditions and moving targets into actuation. Experimental results demonstrate that the Illuminator handles various high-level user requirements and generates an optimal light actuation profile. These results suggest that the Illuminator system supports entertainment and media production applications.