• Korean Journal of Materials Research
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• v.28 no.7
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• pp.398-404
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• 2018

Fatigue crack growth retardation of 304 L stainless steel is studied using a neutron diffraction method. Three orthogonal strain components(crack growth, crack opening, and through-thickness direction) are measured in the vicinity of the crack tip along the crack propagation direction. The residual strain profiles (1) at the mid-thickness and (2) at the 1.5 mm away from the mid-thickness of the compact tension(CT) specimen are compared. Residual lattice strains at the 1.5 mm location are slightly higher than at the mid-thickness. The CT specimen is deformed in situ under applied loads, thereby providing evolution of the internal stress fields around the crack tip. A tensile overload results in an increased magnitude of the compressive residual stress field. In the crack growth retardation, it is found that the stresses are dispersed in the crack-wake region, where the highest compressive residual stresses are measured. Our neutron diffraction mapping results reveal that the dominant mechanism is by interrupting the transfer of stress concentration at the crack tip.

• Korean Journal of Materials Research
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• v.25 no.12
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• pp.690-693
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• 2015

Fatigue crack growth experiments were carried out on a 304 L stainless steel compact-tension(CT) specimen under load control mode. Neutron diffraction was employed to quantitatively measure the residual strains/stresses and the evolution of stress fields in the vicinity of a propagating fatigue-crack tip. Three principal stress components (i.e. crack growth, crack opening, and through-thickness direction stresses) were examined in-situ under loading as a function of distance from the crack tip along the crack-propagation path. The stress/strain fields, measured both at the mid-thickness and near the surface of the CT specimen, were compared. The results show that much higher compressive residual stress fields developed in front of the crack tip near the surface than developed at the mid-thickness area. The change of the stresses ahead of the crack tip under loading is more significant at the mid-thickness area than it is near the surface.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1769-1780
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• 2022

A new method of dose calculation algorithm, called GPU-accelerated Monte Carlo and collapsed cone Convolution (GMCC) was developed to improve the calculation speed of BNCT treatment planning system. The GPU-accelerated Monte Carlo routine in GMCC is used to simulate the neutron transport over whole energy range and the Collapsed Cone Convolution method is to calculate the gamma dose. Other dose components due to alpha particles and protons, are calculated using the calculated neutron flux and reaction data. The mathematical principle and the algorithm architecture are introduced. The accuracy and performance of the GMCC were verified by comparing with the FLUKA results. A water phantom and a head CT voxel model were simulated. The neutron flux and the absorbed dose obtained by the GMCC were consistent well with the FLUKA results. In the case of head CT voxel model, the mean absolute percentage error for the neutron flux and the absorbed dose were 3.98% and 3.91%, respectively. The calculation speed of the absorbed dose by the GMCC was 56 times faster than the FLUKA code. It was verified that the GMCC could be a good candidate tool instead of the Monte Carlo method in the BNCT dose calculations.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1304-1310
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• 2021

Two-phase flow, especially gas-liquid two-phase flow, has a wide application in industrial field. The diagnosis of two-phase flow parameters, which directly determine the flow and heat transfer characteristics, plays an important role in providing the design reference and ensuring the security of online operation of two-phase flow system. Computer tomography (CT) is a good way to diagnose such parameters with imaging method. This paper has proposed a novel image reconstruction method for thermal neutron CT of two-phase flow with improved simulated annealing (ISA) algorithm, which makes full use of the prior information of two-phase flow and the advantage of stochastic searching algorithm. The reconstruction results demonstrate that its reconstruction accuracy is much higher than that of the reconstruction algorithm based on weighted total difference minimization with soft-threshold filtering (WTDM-STF). The proposed method can also be applied to other types of two-phase flow CT modalities (such as X(𝛄)-ray, capacitance, resistance and ultrasound).

• Korean Journal of Soil Science and Fertilizer
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• v.35 no.4
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• pp.197-206
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• 2002

Volumetric soil water contents through a soil profile were monitored to identify the effects of tillage systems and soil physico-chemical characteristic on soil water movement from the soil profile. Water content profiles under no tillage (NT) and conventional tillage (CT) practices were compared at two commercial farms in central Illinois from 1992 through 1994, using neutron-scattering techniques in weekly intervals during each growing season. The volumetric water content of surface soil layers was affected more by tillage systems and rainfall amounts, whereas that of the subsoil layers was more strongly affected by soil types. Soil water percolated faster through Saybrook and Catlin soils than through Drummer, Flanagan, and Ipava soils because Saybrook and Catlin soils have lower clay content and water-retention capacity and higher permeability than Drummer, Flanagan, and Ipava soils. Increased soil organic matter (SOM) in Drummer, Flanagan, and Ipava soils would be attributable to the higher soil water retention than other soil types. Soil water contents in the corn root zone were consistently higher under CT plots than under NT plots.

• Journal of the Korean institute of surface engineering
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• v.56 no.1
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• pp.84-93
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• 2023

The spent nuclear fuel is burned during the planned cycle in the plant and then generates elements such as actinide series, fission products, and plutonium with a long half-life. An 'interim storage' step is needed to manage the high radioactivity and heat emitted by nuclides until permanent-disposal. In the case of Korea, there is no space to dispose of high-level radioactive waste after use, so there is a need for a period of time using interim storage. Therefore, the intensity of neutrons and gamma-ray must be determined to ensure the integrity of spent nuclear fuel during interim storage. In particular, the most important thing in spent nuclear fuel is burnup evaluation, estimation of the source term of neutrons and gamma-ray is regarded as a reference measurement of the burnup evaluation. In this study, an analysis of spent nuclear fuel was conducted by setting up a virtual fuel burnup case based on CE16×16 fuel to check the total amount and spectrum of neutron, gamma radiation produced. The correlation between BU (burnup), IE (enrichment), and CT (cooling time) will be identified through spent nuclear fuel burnup calculation. In addition, the composition of nuclide inventory, actinide and fission products can be identified.

• Journal of the Korea Concrete Institute
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• v.28 no.1
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• pp.59-65
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• 2016

Commercial synthetic resins with great amount of hydrogen atoms were investigated for neutron shielding aggregates. Total three types of resins were considered in this study: high density polyethylene (HDPE), polypropylene (PP), and ultra molecular weight polyethylene (UPE). When these resins replaced 20, 40, 60 vol% of fine aggregates, mechanical properties were first evaluated including compressive and tensile strengths, and then image/microstructure analyses such as cross-section analysis, SEM, and X-ray CT were performed. The results showed that the compressive and tensile strengths decreased with the increase of replacement ratio of HDPE and PP, which was found through image analysis that it was closely related to the distribution of resins at the failure surface of test specimens. The strength reduction of UPE was quite small compared to HDPE and PP but it abruptly increased when the replacement level exceeded 60 vol%. The results of microstructure analyses indicated that the replacement level significantly affected the amount of air void so that it is critical to determine the reasonable amount of UPE to make cementitous materials for neutron shielding.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.292-300
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• 2024

Kinetic simulations have been performed on an Inertial Electrostatic Confinement Fusion (IECF) device. These simulations were performed using the particle-in-cell (PIC) method to analyze the behavior of ions in an IEC device and the effects of some parameters on the Confinement Time (CT). CT is an essential factor that significantly contributes to the IEC's performance as a nuclear fusion device. Using the PIC method, the geometry of a two-grided device with variable grid radius, the number of cathode grid rings, variable pressure and different dielectric thickness for the feed stalk was simulated. In this research, with the development of previous works, the interaction of particles was simulated and compared with previous results. The simulation results are in good agreement with the previous results. In these simulations, it was found that with the increase of the dielectric thickness of the feed stalk, the electric field was weakened and as a result, the confinement time was reduced. On the other hand, with the increase of the cathode radius, the confinement time increased. Using the results, an IEC device can be designed with higher efficiency and more optimal CT for ions.

• Progress in Medical Physics
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• v.27 no.4
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• pp.232-235
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• 2016

The purpose of this study is to develop the treatment planning system (TPS) based on Monte-Carlo simulation for BNCT. In this paper, we will propose a method for dose estimation by Monte-Carlo simulation using the CT image, and will evaluate the accuracy of dose estimation of this TPS. The complicated geometry like a human body allows defining using the lattice function in MCNPX. The results of simulation such as flux or energy deposition averaged over a cell, can be obtained using the features of the tally provided by MCNPX. To assess the dose distribution and therapeutic effect, dose distribution was displayed on the CT image, and dose volume histogram (DVH) was employed in our developed system. The therapeutic effect can be efficiently evaluated by these evaluation tool. Our developed TPS could be effectively performed creating the voxel model from CT image, the estimation of each dose component, and evaluation of the BNCT plan.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.523-531
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• 2022

In order to study the sensitivity and the uncertainty of the Moroccan research reactor TRIGA Mark II, a model of this reactor has been developed in our ERSN laboratory for use with the N-Particle MCNP Monte Carlo transport codes (version 6). In this article, the sensitivities of the effective multiplication factor of this reactor are evaluated using the ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0 libraries and in 44 energy groups, for the cross sections of the fuel (U-235 and U-238) and the moderator (H-1 and O-16). However, the quantification of the uncertainty of the nuclear data is performed using the nuclear code NJOY99 for the generation and processing of covariance matrices. On the one hand, the highest uncertainty deviations, calculated using the ENDFB-VII.1 and JENDL4.0 evaluations, are 2275, 386 and 330 pcm respectively for the reactions U₂₃₅(n, f), $ U_{235}(n\bar{\nu})$ and H₁(n, γ). On the other hand, these differences are very small for the neutron reactions of O-16 and U-238. Regarding the neutron spectra, in CT-mid plane, they are very close for the three evaluations (ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0). These spectra present two peaks (thermal and fission) around the energies 0.05 eV and 1 MeV.