• Journal of the Korea Institute of Information and Communication Engineering
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• v.17 no.8
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• pp.1857-1864
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• 2013

In this paper, we improve the quantum cryptography system using a dual photon transmission plaintext user password algorithmwas designed to implementthe exchange. Existing quantum cryptographic key transport protocols, algorithms, mainly as a quantum cryptography system using the paper, but it improved the way the dual photon transmission through the quantum algorithm re not getting transmitted plaintext.

• Progress in Superconductivity and Cryogenics
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• v.18 no.4
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• pp.5-8
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• 2016

We report on the fabrication and measurement results of the electrical transport properties of superconductor-normal metal-superconductor (SNS) weak links, made of PbIn superconductor and Au metal. The maximum supercurrent reaches up to ${\sim}6{\mu}A$ at T = 2.3 K and the supercurrent persists even at T = 4.7 K. Magnetic field dependence of the critical current is consistent with a theoretical fit using the narrow junction model. The superconducting quantum interference device (SQUID) was also fabricated using two PbIn-Au-PbIn junctions connected in parallel. Under perpendicular magnetic field, we clearly observed periodic oscillations of dV/dI with a period of magnetic flux quantum threading into the supercurrent loop of the SQUID. Our fabrication methods would provide an easy and simple way to explore the superconducting proximity effects without ultra-low-temperature cryostats.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3073-3084
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• 2022

The propagation of radiation source uncertainties in spent nuclear fuel (SNF) cask shielding calculations is presented in this paper. The uncertainty propagation employs the depletion and source term outputs of the deterministic code STREAM as input to the transport simulation of the Monte Carlo (MC) codes MCS and MCNP6. The uncertainties of dose rate coming from two sources: nuclear data and modeling parameters, are quantified. The nuclear data uncertainties are obtained from the stochastic sampling of the cross-section covariance and perturbed fission product yields. Uncertainties induced by perturbed modeling parameters consider the design parameters and operating conditions. Uncertainties coming from the two sources result in perturbed depleted nuclide inventories and radiation source terms which are then propagated to the dose rate on the cask surface. The uncertainty analysis results show that the neutron and secondary photon dose have uncertainties which are dominated by the cross section and modeling parameters, while the fission yields have relatively insignificant effect. Besides, the primary photon dose is mostly influenced by the fission yield and modeling parameters, while the cross-section data have a relatively negligible effect. Moreover, the neutron, secondary photon, and primary photon dose can have uncertainties up to about 13%, 14%, and 6%, respectively.

• Journal of Radiation Protection and Research
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• v.45 no.2
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• pp.69-75
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• 2020

Background: Dose conversion coefficients (DCCs) have been commonly used to estimate radiation-dose absorption by human organs based on physical measurements of fluence or kerma. The International Commission on Radiological Protection (ICRP) has reported a library of DCCs, but few studies have been conducted on their applicability to non-Caucasian populations. In the present study, we collected a total of 8 Korean pediatric and adult voxel phantoms to calculate the organ DCCs for idealized external photon-irradiation geometries. Materials and Methods: We adopted one pediatric female phantom (ETRI Child), two adult female phantoms (KORWOMAN and HDRK Female), and five adult male phantoms (KORMAN, ETRI Man, KTMAN1, KTMAN2, and HDRK Man). A general-purpose Monte Carlo radiation transport code, MCNPX2.7 (Monte Carlo N-Particle Transport extended version 2.7), was employed to calculate the DCCs for 13 major radiosensitive organs in six irradiation geometries (anteroposterior, posteroanterior, right lateral, left lateral, rotational, and isotropic) and 33 photon energy bins (0.01-20 MeV). Results and Discussion: The DCCs for major radiosensitive organs (e.g., lungs and colon) in anteroposterior geometry agreed reasonably well across the 8 Korean phantoms, whereas those for deep-seated organs (e.g., gonads) varied significantly. The DCCs of the child phantom were greater than those of the adult phantoms. A comparison with the ICRP Publication 116 data showed reasonable agreements with the Korean phantom-based data. The variations in organ DCCs were well explained using the distribution of organ depths from the phantom surface. Conclusion: A library of dose conversion coefficients for major radiosensitive organs in a series of pediatric and adult Korean voxel phantoms was established and compared with the reference data from the ICRP. This comparison showed that our Korean phantom-based data agrees reasonably with the ICRP reference data.

• Nuclear Engineering and Technology
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• v.38 no.3
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• pp.225-230
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• 2006

With the advent of inexpensive computing power over the past two decades, applications of Monte Carlo radiation transport techniques have proliferated dramatically. At Los Alamos, the Monte Carlo codes MCNP5 and MCNPX are used routinely on personal computer platforms for radiation shielding analysis and dosimetry calculations. These codes feature a rich palette of variance reduction (VR) techniques. The motivation of VR is to exchange user efficiency for computational efficiency. It has been said that a few hours of user time often reduces computational time by several orders of magnitude. Unfortunately, user time can stretch into the many hours as most VR techniques require significant user experience and intervention for proper optimization. It is the purpose of this paper to outline VR strategies, tested in practice, optimized for several common radiation shielding tasks, with the hope of reducing user setup time for similar problems. A strategy is defined in this context to mean a collection of MCNP radiation transport physics options and VR techniques that work synergistically to optimize a particular shielding task. Examples are offered in the areas of source definition, skyshine, streaming, and transmission.

• Nuclear Engineering and Technology
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• v.22 no.4
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• pp.380-388
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• 1990

The MM-22 medical microtron at Korea Cancer Center Hospital is now operational for high energy electron and photon therapy, This microtron is designed to produce 5.3-22.5 MeV electron beams and deliver these to the treatment head through beam transport system with an intensity and stability suitable for cancer treatment. The availability of high quality radiation modalities from the MM-22 shows new possibilities in the treatment of deep seated tumours. Principle of operation, system structures and operating characteristics of the MM-22 are described in this paper.

• Journal of the Korean Institute of Electrical and Electronic Material Engineers
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• v.13 no.3
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• pp.265-272
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• 2000

In this paper the corona discharge in SF$_{6}$ gas used as insulating material in lots o high voltage equipment, is simulated by finite element method with Flux-Corrected Transport(FCT) method. By application of proposed method the negative corona discharge characteristics in needle to plane electrode is analyzed with time step. For the accuracy of analysis the secondary electron emission by photon and ion are also considered as well as the accuracy of analysis the secondary electron emission by photon and ion are also considered as well as townsend first ionization and electron attachment. The calculated results show that the electric field intensity between anode and ion group is decreased as times go-by according to field distortion by those space charge. Accordingly the electron density is decreased strongly by the attatchment effect of SF6 gas so that the corona discharge becomes extinguished abruptly.y.

• Nuclear Engineering and Technology
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• v.40 no.4
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• pp.319-326
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• 2008

To transport process wastes efficiently from the Advanced Spent Fuel Conditioning Pyro-process Facility (ACPF) at the Korea Atomic Energy Research Institute (KAERI), a new hot cell cask has been designed based on an existing hot cell padirac transport cask, with not only a neutron absorber for improved shielding capability, but also a docking facility for an easy docking system. In the new hot cell cask, two kinds of materials have been considered as shielding materials, polyethylene and resin. To verify the transport compatibility of the waste and spent fuel for the ACPF, neutron and photon shielding calculations were performed using the MCNPX code. The source term was evaluated by the ORIGEN-ARP code system based on spent PWR fuel. From the calculation, it was found that the maximum surface dose rates of the hot cell cask with the two candidates were estimated within the limit (2 mSv/hr).

• Progress in Medical Physics
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• v.19 no.1
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• pp.21-34
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• 2008

The parallel Monte Carlo electron and photon transport (PMCEPT) code [Kum and Lee, J. Korean Phys. Soc. 47, 716 (2006)] for calculating electron and photon beam doses has been developed based on the three dimensional geometry defined by computed tomography (CT) images and implemented on the Beowulf PC cluster. Understanding the limitations of Monte Carlo codes is useful in order to avoid systematic errors in simulations and to suggest further improvement of the codes. We evaluated the PMCEPT code by comparing its normalized depth doses for electron and photon beams with those of MCNP5, EGS4, DPM, and GEANT4 codes, and with measurements. The PMCEPT results agreed well with others in homogeneous and heterogeneous media within an error of $1{\sim}3%$ of the dose maximum. The computing time benchmark has also been performed for two cases, showing that the PMCEPT code was approximately twenty times faster than the MCNP5 for 20-MeV electron beams irradiated on the water phantom. For the 18-MV photon beams irradiated on the water phantom, the PMCEPT was three times faster than the GEANT4. Thus, the results suggest that the PMCEPT code is indeed appropriate for both fast and accurate simulations.

• Nuclear Engineering and Technology
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• v.44 no.2
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• pp.161-176
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• 2012

McCARD is a Monte Carlo (MC) neutron-photon transport simulation code. It has been developed exclusively for the neutronics design of nuclear reactors and fuel systems. It is capable of performing the whole-core neutronics calculations, the reactor fuel burnup analysis, the few group diffusion theory constant generation, sensitivity and uncertainty (S/U) analysis, and uncertainty propagation analysis. It has some special features such as the anterior convergence diagnostics, real variance estimation, neutronics analysis with temperature feedback, $B_1$ theory-augmented few group constants generation, kinetics parameter generation and MC S/U analysis based on the use of adjoint flux. This paper describes the theoretical basis of these features and validation calculations for both neutronics benchmark problems and commercial PWR reactors in operation.