• Journal of the Korean Society of Radiology
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• v.14 no.4
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• pp.455-465
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• 2020

The purpose of this study is to evaluate the photoneutron characteristics generated by the linear accelerator target and collimator. The computer simulation design firstly, consisted of a target, a single material target and a composite material target. Secondly, it consisted of a cone beam and a fan beam depending on the type of the collimator. Finally, the material of the fan beam collimator is composed of a single material composed of only lead (Pb) and a composite material collimator composed of tungsten (W) and lead (Pb). The research method calculated the photoneutron production rate and energy spectrum using F2 tally from the surface of a virtual sphere at a distance of 100 cm from the target. As a result, firstly the photoneutron production rate was 20% difference, depending on the target. Secondly, depending on the type of the collimator, there was a 10% difference. Finally, depending on the collimator material, there was a 40% difference. In the photoneutron energy spectrum, the average photoneutron flux tended to be similar to the photoneutron production rate. As a result, it was confirmed that the 9 MeV linear accelerator photoneutron are production increased more by the collimator than by the target, and by the material, not the type of the collimator. Selecting and operating targets and collimator with low photoneutron production will be the most active radiation protection. Therefore, it is considered that this research can be a useful data for introducing and operating and radiation protection of a linear accelerator for container security inspection.

• Radiation Oncology Journal
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• v.21 no.2
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• pp.158-166
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• 2003

Purpose: To propose a conceptual design of a novel source for intensity modulated brachytherapy. Materials and Methods: The source design incorporates both radioactive and shielding materials (stainless steel or tungsten), to provide an asymmetric dose intensity in the azimuthal direction. The intensity modulated intravascular brachytherapy was performed by combining a series of dwell positions and times, distributed along the azimuthal coordinates. Two simple designs for the beta-emitting sources, with similar physical dimensions to a $^{90}Sr/Y$ Novoste Beat-Cath source, were considered in the dosimetric feasibility study. In the first design, the radioactive and materials each occupy half of the cylinder and in the second, the radioactive material occupies only a quater of the cylinder. The radial and azimuthal dose distributions around each source were calculated using the MCNP Monte Carlo code. Results: The preliminary hypothetical simulation and optimization results demonstrated the 87$\%$ difference between the maximum and minimum doses to the lumen wall, due to off-centering of the radiation source, could be reduced to less than 7$\%$ by optimizing the azimuthal dwell positions and times of the partially shielded intravascular brachytherapy sources. Conclusion: The novel brachytherapy source design, and conceptual source delivery system, proposed in this study show promising dosimetric characteristics for the realization of intensity modulated brachytherapy in intravascular treatment. Further development of this concept will center on building a delivery system that can precisely control the angular motion of a radiation source in a small-diameter catheter.

• Journal of radiological science and technology
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• v.43 no.5
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• pp.389-395
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• 2020

The evaluation of radioactivated components of heavy-ion accelerator facilities affects the safety of radiation management and the exposure dose for workers. and this is an important issue when predicting the disposal cost of waste during maintenance and dismantling of accelerator facilities. In this study, the FLUKA code was used to simulate the proton treatment device nozzle and classify the radio-nuclides and total radioactivity generated by each component over a short period of time. The source term was evaluated using NIST reference beam data, and the neutron flux generated for each component was calculated using the evaluated beam data. Radioactive isotopes caused by generated neutrons were compared and evaluated using nuclide information from the International Radiation Protection Association and the Korea Radioisotope association. Most of the nuclides produced form of beta rays and electron capture, and short-lived nuclides dominated. However, In the case of ⁵⁴Mn, which is a radioactive product of iron, the effect of gamma rays should be considered. In the case of tritium generated from a material with a low atomic number, it is considered that handling care should be taken due to its long half-life.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.2
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• pp.58-63
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• 2019

This paper presents preliminary analysis and results on IASCC sensitivity of a core shroud in the reactor pressure vessel. First, neutron irradiation flux distribution of the reactor internals was calculated by using the Monte Carlo simulation code, MCNP6.1 and the nuclear data library, ENDF/B-VII.1. Second, based on the neutron irradiation flux distribution, temperature and stress distributions of the core shroud during normal operation were determined by performing finite element analysis using the commercial finite element analysis program, ABAQUS, considering irradiation aging-related degradation mechanisms. Last, IASCC sensitivity of the core shroud was assessed by using the IASCC sensitivity definition of EPRI MRP-211 and the finite element analysis results. As a result of the preliminary analysis, it was found that the point at which the maximum IASCC sensitivity is derived varies over operating time, initially moving from the shroud plate located in the center of the core to the top shroud plate-ring connection brace over operating time. In addition, it was concluded that IASCC will not occur on the core shroud even after 60 years of operation (40EFPYs) because the maximum IASCC sensitivity is less than 0.5.

• 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.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4226-4230
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• 2022

The Korea Atomic Energy Research Institute (KAERI) has designed a Hybrid-Low Power Research Reactor (H-LPRR) which can be used for critical assembly and conventional research reactor as well. It is an open tank-in-pool type research reactor (Thermal Power: 50 kW_th) of which the most important applications are Neutron Activation Analysis (NAA), Radioisotope (RI) production, education and training. There are eight irradiation holes on the edge of the reactor core: IR (6 holes for RI production) and NA (2 holes for NAA) holes. In order to quantify the elemental concentration in target samples through the Instrumental Neutron Activation Analysis (INAA), it is necessary to measure neutron spectrum parameters such as thermal neutron flux, the deviation from the ideal 1/E epithermal neutron flux distribution (α), and the thermal-to-epithermal neutron flux ratio (f) for the irradiation holes. In this study, the MCNP6.1 code and FORTRAN 90 language are applied to determine the parameters for the two irradiation holes (NA-SW and NA-NW) in H-LPRR, and in particular its α and f parameters are compared to values of other research reactors. The results confirmed that the neutron irradiation holes in H-LPRR are designed to be sufficiently applied to neutron activation analysis, and its performance is comparable to that of foreign research reactors including the TRIGA MARK II.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3525-3534
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• 2023

Non-uniformity of Pu and Cm composition in used nuclear fuel was analyzed to determine its effect on Pu accountancy in pyroprocessing, while employing the Pu-to-²⁴⁴Cm ratio method. Burnup simulation of a typical pressurized water reactor fuel assembly, required for the analysis, was carried out using MCNP code. Used fuel nuclide composition, as a function of nine axial and two radial meshes, were evaluated. The axial variation of neutron flux and self-shielding effects were found to affect the uniformity of Pu and Cm compositions and in turn the Pu-to-²⁴⁴Cm ratio. However, the results of the study showed that these non-uniformities do not affect the use of Pu-to-²⁴⁴Cm ratio method for Pu accountancy, if the measurement samples are drawn from the voloxidized powder at the feed step of pyroprocessing. 'Material Unaccounted For' and its uncertainty estimates are also presented for a pyrprocessing facility to verify safeguards monitoring requirements of the IAEA.

• Radiation Oncology Journal
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• v.19 no.1
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• pp.66-73
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• 2001

Purpose : For the research of Boron Neutron Capture Therapy (BNCT), fast neutrons generated from the MC-50 cyclotron with maximum energy of 34.4 MeV in Korea Cancer Center Hospital were moderated by 70 cm paraffin and then the dose characteristics were investigated. Using these results, we hope to establish the protocol about dose measurement of epi-thermal neutron, to make a basis of dose characteristic of epi-thermal neutron emitted from nuclear reactor, and to find feasibility about accelerator-based BNCT. Method and Materials : For measuring the absorbed dose and dose distribution of fast neutron beams, we used Unidos 10005 (PTW, Germany) electrometer and IC-17 (Far West, USA), IC-18, ElC-1 ion chambers manufactured by A-150 plastic and used IC-l7M ion chamber manufactured by magnesium for gamma dose. There chambers were flushed with tissue equivalent gas and argon gas and then the flow rate was S co per minute. Using Monte Carlo N-Particle (MCNP) code, transport program in mixed field with neutron, photon, electron, two dimensional dose and energy fluence distribution was calculated and there results were compared with measured results. Results : The absorbed dose of fast neutron beams was $6.47\times10^{-3}$ cGy per 1 MU at the 4 cm depth of the water phantom, which is assumed to be effective depth for BNCT. The magnitude of gamma contamination intermingled with fast neutron beams was $65.2{\pm}0.9\%$ at the same depth. In the dose distribution according to the depth of water, the neutron dose decreased linearly and the gamma dose decreased exponentially as the depth was deepened. The factor expressed energy level, $D_{20}/D_{10}$, of the total dose was 0.718. Conclusion : Through the direct measurement using the two ion chambers, which is made different wall materials, and computer calculation of isodose distribution using MCNP simulation method, we have found the dose characteristics of low fluence fast neutron beams. If the power supply and the target material, which generate high voltage and current, will be developed and gamma contamination was reduced by lead or bismuth, we think, it may be possible to accelerator-based BNCT.

• Journal of Biosystems Engineering
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• v.39 no.3
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• pp.205-214
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• 2014

Purpose: Phytosanitary irradiation treatment can effectively control regulated pests while maintaining produce quality. The objective of this study was to establish the best irradiation treatment for mangosteen, a popular tropical fruit, using a Monte Carlo simulation. Methods: Magnetic resonance image (MRI) data were used to generate a 3-D geometry to simulate dose distributions in a mangosteen using a radiation transport code (MCNP5). Microsoft Excel with visual basic application (VBA) was used to divide the image data into seed, flesh, and rind. Radiation energies used for the simulation were 10 MeV (high-energy) and 1.35 MeV (low-energy) for the electron beam, 5 MeV for X-rays, and 1.25 MeV for gamma rays from Co-60. Results: At 5 MeV X-rays and 1.25 MeV gamma rays, all areas (seeds, flesh, and rind) were irradiated ranging from 0.3 ~ 0.7 kGy. The average doses decreased as the number of fruit increased. For a 10 MeV electron beam, the dose distribution was biased: the dose for the rind where the electrons entered was $0.45{\pm}0.03$ kGy and the other side was $0.24 {\pm}0.10$ kGy. Use of an electron kinetic energy absorber improved the dose distribution in mangosteens. For the 1.35 MeV electron beam, the dose was shown only in the rind on the irradiated side; no significant dose was found in the flesh or seeds. One rotation of the fruit while in front of the beam improved the dose distribution around the entire rind. Conclusion: These results are invaluable for determining the ideal irradiation conditions for phytosanitary irradiation treatment of tropical fruit.

• Progress in Medical Physics
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• v.33 no.4
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• pp.129-135
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• 2022

Purpose: A full-energy-peak (FEP) efficiency correction is required through a Monte Carlo simulation for accurate radioactivity measurement, considering the geometrical characteristics of the detector and the sample. However, a relative deviation (RD) occurs between the measurement and calculation efficiencies when modeling using the data provided by the manufacturers due to the randomly generated dead layer. This study aims to optimize the structure of the detector by determining the dead layer thickness based on Monte Carlo simulation. Methods: The high-purity germanium (HPGe) detector used in this study was a coaxial p-type GC2518 model, and a certified reference material (CRM) was used to measure the FEP efficiency. Using the MC N-Particle Transport Code (MCNP) code, the FEP efficiency was calculated by increasing the thickness of the outer and inner dead layer in proportion to the thickness of the electrode. Results: As the thickness of the outer and inner dead layer increased by 0.1 mm and 0.1 ㎛, the efficiency difference decreased by 2.43% on average up to 1.0 mm and 1.0 ㎛ and increased by 1.86% thereafter. Therefore, the structure of the detector was optimized by determining 1.0 mm and 1.0 ㎛ as thickness of the dead layer. Conclusions: The effect of the dead layer on the FEP efficiency was evaluated, and an excellent agreement between the measured and calculated efficiencies was confirmed with RDs of less than 4%. It suggests that the optimized HPGe detector can be used to measure the accurate radioactivity using in dismantling and disposing medical linear accelerators.