• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.292-303
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• 2004

The previous exponential experiment system has been improved for the automatical and accurate axial movement of the neutron source and detector with attaching the automatical control system which consists of a Programmable Logical Controller(PLC) and a stepping motor set. The automatic control program which controls MCA and PLC consistently has been also developed on the basis of GENIE 2000 Library. The exponential experiments have been carried out for Kori 1 unit spent fuel assemblies, Cl4, Jl4 and G23, and Kori 2 unit spent fuel assembly, J44, using the improved systematical measurement system. As the results, the average exponential decay constants for 4 assemblies are determined to be 0.1302, 0.1267, 0.1247, and 0.1210, respectively, with the application of Poisson regression.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.43-51
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• 2004

In order to decommission the shielding concrete of KRR(Korea Research Reactor) -1＆2, it must be exactly determined activated level and range by neutron irradiation during operation. To determine the activated level and range, it must be sampled and analyzed the core sample. But, there are difficulties in sample preparation and determination of the measurement efficiency by self-absorption. In the study, the full energy efficiency of the HPGe detector was compared with the measured value using standard source and the calculated one using Monte Carlo simulation. Also. self-absorption effects due to the density and component change of the concrete were calculated using the Monte Carlo method. Its results will be used radioactivity analysis of the real concrete core sample in the future.

• Nuclear Engineering and Technology
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• v.34 no.5
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• pp.494-501
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• 2002

Radioactive Corrosion Products (CRUD) which are generated by the neutron activation of general corrosion products at the nuclear power plant are the major source of occupational radiation exposure. Most of the CRUD has a characteristic of showing strong ferrimagnetisms. Along with the new development and production of permanent magnet (rare earth magnet) which generates much stronger magnetic field than the conventional magnet, new type of magnetic filter that can separate CRUD efficiently and eventually reduce radiation exposure of personnel at nuclear power plant is suggested. This separator consists of inner and outer magnet assemblies, coolant channel and container surrounding the outer magnet assembly. The rotational motion of the inner and outer permanent magnet assemblies surrounding the coolant channel by driving motor system produces moving alternating magnetic fields in the coolant channel. The CRUD can be separated from the coolant by the moving alternating magnetic field. This study describes the results of preliminary experiment performed with the different flow rates of coolant and rotation velocities of magnet assemblies. This new magnetic filter shows better performance results of filtering the magnetite at coolant (water). How rates, rotating velocities of magnet assemblies and particle sizes turn out to be very important design parameters.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1535-1540
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• 2022

A solid tungsten target was used at the China Spallation Neutron Source (CSNS) with 100 kW proton beam power. To improve the lifetime, hot isostatic pressing (HIP) process was selected to bond tantalum cladding with tungsten plates. Radioactive isotope ¹⁸²Ta, an activation product of tantalum, was found in the cooling water after a period of operation, however, no radioactive isotopes of ¹⁸⁷W was found, which shows the tantalum layer remained mostly intact. The CSNS Target-1 had been operating safely for three years and was replaced by Target-2 in August 2020.

• Journal of Radiation Protection and Research
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• v.36 no.1
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• pp.44-51
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• 2011

Activation foil-based Bonner sphere spheres are used to obtain neutron energy spectra of nuclear power plants or accelerator-produced neutrons. The position and the foil mass dependence of response functions should be studied carefully before measurement of Bonner spheres. This study showed that the normal incidence to the foil surface made a large shift of responses while parallel and isotropic incidence made no position dependence. The correlation between foil mass and response was not linear. Therefore, the response functions of activation-foil based Bonner spheres should be calculated for every different foil mass and the direction of Bonner spheres for parallel incidence will be preferred for radioactive neutron source or accelerator target produced neutrons.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.915-920
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• 2001

A spent fuel problem has prevented the nuclear power from claiming to be a completely clean energy source. The nuclear transmutation technology to incinerate the long lived radioactive nuclides and produce energy during the incineration process is believed to be one or the best solutions. HYPER(Hybrid Power Extraction Reactor) is the accelerator driven transmutation system which is being developed by KAERI(Korea Atomic Energy Research Institute). Some major feature of HYPER have been developed and employed. On-power fueling concepts are employed to keep system power constant with minimum variation of accelerator power. A hollow cylinder-type metal fuel is designed for the on-line refueling concept. Lead-bismuth(Pb-Bi) is adopted as a coolant and Spallation target material. HYPER is a subcritical reactor which needs an external neutron source. 1GeV proton beam is irradiated to Lead-bismuth(Pb-Bi) target inside HYPER, and spallation neutrons are produced. When proton beams are irradiated, much heat is also deposited in the Pb-Bi target and beam window which separates Pb-Bi and accelerator vacuum. Therfore, an effective cooling is needed for HYPER target. In this paper, we performed the thermal-hydraulic analysis of HYPER target using FLUENT code, and also calculated thermal and mechanical stress of the beam window using ANSYS code.

• Journal of Radiation Protection and Research
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• v.46 no.3
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• pp.112-119
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• 2021

Background: This work aims to develop a new imaging system based on a pulse shape discrimination-capable Cs₂LiYCl₆:Ce (CLYC) scintillation detector combined with the rotational modulation collimator (RMC) technique for dual-particle imaging. Materials and Methods: In this study, a CLYC-based RMC system was designed based on Monte Carlo simulations, and a prototype was fabricated. Therein, a rotation control system was developed to rotate the RMC unit precisely, and a graphical user interface-based software was also developed to operate the data acquisition with RMC rotation. The RMC system was developed to allow combining various types of collimator masks and detectors interchangeably, making the imaging system more versatile for various applications and conditions. Results and Discussion: Operational performance of the fabricated system was studied by checking the accuracy and precision of the collimator rotation and obtaining modulation patterns from a gamma-ray source repeatedly. Conclusion: The prototype RMC system showed reliability in its mechanical properties and reproducibility in the acquisition of modulation patterns, and it will be further investigated for its dual-particle imaging capability with various complex radioactive source conditions.

• Nuclear Engineering and Technology
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• v.46 no.4
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• pp.547-556
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• 2014

Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding, $UO_2$ pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a $2{\times}10$ cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the $2{\times}10$ cask array, dose rates at the center point of the array and at the center of the casks' height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the $2{\times}10$ cask array showed exponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280m from the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than 25mrem/yr, due to the decay of radioactivity of spent fuel in storage.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.584-589
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• 2003

The hot cell facility which is designed to permit safe handling of source materials with radioactivity levels up to 1,385 TBq, is planned to be built. To meet this goal, the facility is designed to keep gamma and neutron radiation lower than the recommended dose-rate in normally occupied areas. The calculations performed with QAD-CGGP and MCNP-4C are used to evaluate the proposed engineering design concepts that would provide acceptable dose-rates during a normal operation in hot cell facility. The maximum effective gamma dose-rates on the surfaces of the facility at operation area and at service area calculated by QAD-CGGP are estimated to be $2.10{\times}10^{-3}$, $2.97{\times}10^{-2}$ and $1.01{\times}10^{-1}$ mSv/h, respectively. And those calculated by MCNP-4C are $1.60{\times}10^{-3}$, $2.99{\times}10^{-3}$ and $7.88{\times}10^{-2}$ mSv/h, respectively The dose-rates contributed by neutrons are one order of magnitude less than that of gamma sources, and penetration and toboggan will be partly reinforced by lead shield.

• Journal of Energy Engineering
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• v.25 no.2
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• pp.52-64
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• 2016

The production of nuclear energy from thorium which is non-fissile material was a main issue until the middle of 1970's, because of the thorium's abundance as energy resources, its capability of breeding fissile material U233, and the reduction of long-lived actinides. However, to use thorium as nuclear fuel, some obstacles such as the necessities of external neutron source and long-term neutron irradiation for effective breeding, and the production of high radioactive isotopes in the course of thorium breeding cycle should be overcome. The difficulties to resolve these cons of thorium cycle became the reason of interruption of the related researches in the middle of 1970's. But in the 21st century, the change of societal perspective regarding nuclear energy and the appearance of accelerator-driven nuclear reactor shift those cons into pros and rehabilitate the study of thorium. The high activity of thorium cycle turned out to be a good option as higher resistance and easier detectibility of nuclear proliferation and the employment of subcritical accelerator-driven reactor as external neutron sources is considered to enhance the nuclear safety. In this study we compare the thorium cycle with the currently-used uranium cycle and analyze the technical status and perspective of thorium researches which use accelerator-driven reactors.