• Progress in Medical Physics
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• v.15 no.4
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• pp.215-219
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• 2004

Due to their excellence for the high-energy therapy range of photon beams, researchers show increasing interest in applying MOSFET dosimeters to low- and medium-energy applications. In this energy range, however, MOSFET dosimeter is complicated by the fact that the interaction probability of photons shows significant dependence on the atomic number, Z, due to photoelectric effect. The objective of this study is to develop a very detailed 3-dimensional Monte Carlo simulation model of a MOSFET dosimeter for radiological characterizations and calibrations. The sensitive volume of the High-Sensitivity MOSFET dosimeter is very thin (1 ${\mu}{\textrm}{m}$) and the standard MCNP tallies do not accurately determine absorbed dose to the sensitive volume. Therefore, we need to score the energy deposition directly from electrons. The developed model was then used to study various radiological characteristics of the MOSFET dosimeter. the energy dependence was quantified for the energy range 15 keV to 6 MeV; finding maximum dependence of 6.6 at about 40 keV. A commercial computer code, Sabrina, was used to read the particle track information from an MCNP simulation and count the tracks of simulated electrons. The MOSFET dosimeter estimated the calibration factor by 1.16 when the dosimeter was at 15 cm depth in tissue phantom for 662 keV incident photons. Our results showed that the MOSFET dosimeter estimated by 1.11 for 1.25 MeV photons for the same condition.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1109-1119
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• 2016

Although, the direct use of spent pressurized water reactor (PWR) fuel in CANda Deuterium Uranium (CANDU) reactors (DUPIC) cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by $UO_2$ enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1) the power distribution amongst the fuel elements of the bundle; (2) the coolant void reactivity; and (3) the reactor point-kinetics parameters.

• Radiation Oncology Journal
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• v.28 no.4
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• pp.224-230
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• 2010

Purpose: We designed the aft-multiple-slit (AMS) system to reduce scatter in cone-beam computed tomography (CBCT). As a preliminary study, we performed a Monte Carlo N-Particle Transport Code (MCNP) simulation to verify the effectiveness of this system. Materials and Methods: The MCNPX code was used to build the AMS geometry. An AMS is an equi-angled arc to consider beam divergence. The scatter-reduced projection images were compared with the primary images only and the primary plus scatter radiation images with and without AMS to evaluate the effectiveness of scatter reduction. To obtain the full 2 dimensional (2D) projection image, the whole AMS system was moved to obtain closed septa of the AMS after the first image acquisition. Results: The primary radiation with and without AMS is identical to all the slit widths, but the profiles of the primary plus scattered radiation varied according to the slit widths in the 2D projection image. The average scatter reduction factors were 29%, 15%, 9%, and 8% when the slit widths were 5 mm, 10 mm, 15 mm, and 20 mm, respectively. Conclusion: We have evaluated the scatter reduction effect of the AMS in CBCT imaging using the Monte Carlo (MC) simulations. A preliminary study based on the MCNP simulations showed a mount of scatter reduction with the proposed system.

• Journal of Radiation Protection and Research
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• v.26 no.2
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• pp.67-71
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• 2001

In order to shield the neutrons affecting the background of Low Level Gamma Ray Spectrometer, a neutron shielder was designed. The method used in this study for neutron shielding was the deceleration of fast neutrons by high density polyethylene(HDPE) and the absorption of those slowing-down neutrons by $B_4C$. The calculation results of neutron Interaction in HDPE using Monte Carlo simulation code MCNP4B showed that the thermal-neutron flux was maximum at 10 cm thickness of HDPE. The results also showed that 95% of the thermal neutrons were absorbed by 2 mm thickness of $B_4C$ absorber Consisted of 30 w% $B_4C$ and 70 w% polymer. The results of the Monte Carlo calculation were in good agreement with the experimental value obtained by a neutron shielding apparatus designed for this purpose.

• Journal of Radiation Protection and Research
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• v.47 no.1
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• pp.50-57
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• 2022

Background: There are several proton therapy facilities in operation or planned in Taiwan, and these facilities are anticipated to not only treat cancer but also provide beam services to the industry or academia. The simplified approach based on the Monte Carlo-based data sets (source terms and attenuation lengths) with the point-source line-of-sight approximation is friendly in the design stage of the proton therapy facilities because it is intuitive and easy to use. The purpose of this study is to expand the Monte Carlo-based data sets to allow the simplified approach to cover the application of proton beams more widely. Materials and Methods: In this work, the MCNP6 Monte Carlo code was used in three simulations to achieve the purpose, including the neutron yield calculation, Monte Carlo-based data sets generation, and dose assessment in simple cases to demonstrate the effectiveness of the generated data sets. Results and Discussion: The consistent comparison of the simplified approach and Monte Carlo simulation results show the effectiveness and advantage of applying the data set to a quick shielding design and conservative dose assessment for proton therapy facilities. Conclusion: This study has expanded the existing Monte Carlo-based data set to allow the simplified approach method to be used for dose assessment or shielding design for beam services in proton therapy facilities. It should be noted that the default model of the MCNP6 is no longer the Bertini model but the CEM (cascade-exciton model), therefore, the results of the simplified approach will be more conservative when it was used to do the double confirmation of the final shielding design.

• Journal of the Korean Society of Radiology
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• v.14 no.5
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• pp.563-575
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• 2020

The purpose of this study is to evaluate the activation characteristics that occur in a linear accelerator for container security inspection. In the computer simulation design, first, the targets consisted of a tungsten (Z=74) single material target and a tungsten (Z=74) and copper (Z=29) composite target. Second, the fan beam collimator was composed of a single material of lead (Z=82) and a composite material of tungsten (Z-74) and lead (Z=82) depending on the material. Final, the concrete in the room where the linear accelerator was located contained magnetite type and impurities. In the research method, first, the optical neutron flux was calculated using the MCNP6 code as a F4 Tally for the linear accelerator and structure. Second, the photoneutron flux calculated from the MCNP6 code was applied to FISPACT-II to evaluate the activation product. Final, the decommissioning evaluation was conducted through the specific activity of the activation product. As a result, first, it was the most common in photoneutron targets, followed by a collimator and a concrete 10 cm deep. Second, activation products were produced as by-products of W-181 in tungsten targets and collimator, and Co-60, Ni-63, Cs-134, Eu-152, Eu-154 nuclides in impurity-containing concrete. Final, it was found that the tungsten target satisfies the permissible concentration for self-disposal after 90 days upon decommissioning. These results could be confirmed that the photoneutron yield and degree of activation at 9 MeV energy were insignificant. However, it is thought that W-181 generated from the tungsten target and collimator of the linear accelerator may affect the exposure when disassembled for repair. Therefore, this study presents basic data on the management of activated parts of a linear accelerator for container security inspection. In addition, When decommissioning the linear accelerator for container security inspection, it is expected that it can be used to prove the standard that permissible concentration of self-disposal.

• Journal of Radiation Industry
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• v.11 no.3
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• pp.157-165
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• 2017

In this paper, we calculate the light photons collection efficiency of large-volume plastic scintillation detector mainly used for radiation portal monitor (RPM). A Monte Carlo light photon transport code, DETECT2000, were used to quantitatively evaluate light collection efficiency of plastic scintillation detector. DETECT2000 calculated the placement of light collection efficiency based on the energy spectrum. We calculated the light collection efficiency relative to the position of the energy spectrum that proportional to the placement of the source. The $850{\times}285{\times}65mm^3$ size of polyvinyl toluene (PVT) scintillator was used for measurements. Through DETECT2000 simulation, the light collection efficiency of $5{\times}5$ arrays were calculated and verification was performed by comparing with experimentally measured. And then, the corrected MCNP simulation by applying the light collection efficiency in $21{\times}13$ arrays was compared and analyzed. Comparing the Monte Carlo simulation with measured results, it shows an average difference of 10.1% in $5{\times}5$ arrays. Particularly, about twice of the difference was found in the edge of first column, which coupled with PMT. In whole $5{\times}5$ array, the overall ratio was the same except for the first column. And then comparing the energy spectra of the $21{\times}13$ array with and without the light collection efficiency, it shows a difference of 6.69% in Compton edge area. The DETECT2000 based light collection efficiency simulation showed well agreement with the point source experiment. And comparing with measured energy spectra, we could compare the differences according to whether or not the light collection efficiency was applied. As a results, it is possible to increase the accuracy and reliability of Monte Carlo simulation results by pre-calculating the light collection efficiency according to the PVT geometry by using the DETECT2000.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.3051-3057
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• 2021

The concrete is considered as an important radiation shielding material employed widely in nuclear reactors, particle accelerators, laboratory hot cells and other different radiation sources. The present research is dedicated to the shielding properties study of the ordinary concrete reinforced with different weight fractions of lead oxide micro/nano particles. Lead oxide particles were fabricated by chemical synthesis method and their properties including the average size, morphological structure, functional groups and thermal properties were characterized by XRD, FESEM-EDS, FTIR and TGA analysis. The gamma ray mass attenuation coefficient of concrete composites has been calculated and measured by means of the Monte Carlo simulation and experimental methods. The simulation process was based on the use of MCNP Monte Carlo code where the mass attenuation coefficient (μ/ρ) has been calculated as a function of different particle sizes and filler weight fractions. The simulation results showed that the employment of the lead oxide filler particles enhances the mass attenuation coefficient of the ordinary concrete, drastically. On the other hand, there are approximately no differences between micro and nano sized particles. The mass attenuation coefficient was increased by increasing the weight fraction of nanoparticles. However, a semi-saturation effect was observed at concentrations more than 10 wt%. The experimental process was based on the fabrication of concrete slabs filled by different weight fractions of nano lead oxide particles. The mass attenuation coefficients of these slabs were determined at different gamma ray energies using ²²Na, ¹³⁷Cs and ⁶⁰Co sources and NaI (Tl) scintillation detector. The experimental results showed that the HVL parameter of the ordinary concrete reinforced with 5 wt% of nano PbO particles was reduced by 64% at 511 keV and 48% at 1332 keV. Reasonable agreement was obtained between simulation and experimental results and showed that the employment of nano PbO particles is more efficient at low gamma energies up to 1Mev. The proposed concrete is less toxic and could be prepared in block form instead of toxic lead blocks.

• Journal of the Korean Society of Radiology
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• v.14 no.3
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• pp.193-201
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• 2020

The purpose of this study is to evaluate the photon characteristics according to the material and thickness of the electrons incidented through a linear accelerator. The computer simulation design is a linear accelerator target consisting of a 2 mm thick tungsten single material and a 1.8 mm and 2.3 mm thick tungsten and copper composite material. In the research method, First, the behavior of primary particles in the target was evaluated by electron fluence and electron energy deposition. Second, photons occurring within the target were evaluated by photon fluence. Finally, the photon angle-energy distribution at a distance of 1 m from the target was evaluated by photon fluence. As a result, first, electrons, which are primary particles, were not released out of the target for electron fluence and energy deposition in the target of a single material and a composite material. Then, electrons were linearly attenuated negatively according to the target thickness. Second, it was found that the composite material target had a higher photon generation than the single material target. This confirmed that the material composition and thickness influences photon production. Finally, photon fluence according to the angular distribution required for shielding analysis was calculated. These results confirmed that the photon generation rate differed depending on the material and thickness of the linear accelerator target. Therefore, this study is necessary for designing and operating a linear accelerator use facility for container security screening that is being introduced in the country. In addition, it is thought that it can be used as basic data for radiation protection.

• Journal of Radiation Protection and Research
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• v.24 no.4
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• pp.215-223
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• 1999

Characteristics of radiation field in the steam generator(S/G) water chamber of a PWR were investigated and the anticipated effective dose rates to the worker in the S/G chamber were evaluated by Monte Carlo simulation. The results of crud analysis in the S/G of the Kori nuclear power plant unit 1 were adopted for the source term. The MCNP4A code was used with the MIRD type anthropomorphic sex-specific mathematical phantoms for the calculation of effective doses. The radiation field intensity is dominated by downward rays, from the U-tube region, having approximate cosine distribution with respect to the polar angle. The effective dose rates to adults of nominal body size and of small body size(The phantom for a 15 year-old person was applied for this purpose) appeared to be 36.22 and 37.06 $mSvh^{-1}$) respectively, which implies that the body size effect is negligible. Meanwhile, the equivalent dose rates at three representative positions corresponding to head, chest and lower abdomen of the phantom, calculated using the estimated exposure rates, the energy spectrum and the conversion coefficients given in ICRU47, were 118, 71 and 57 $mSvh^{-1}$, respectively. This implies that the deep dose equivalent or the effective dose obtained from the personal dosimeter reading would be the over-estimate the effective dose by about two times. This justifies, with possible under- or over- response of the dosimeters to radiation of slant incidence, necessity of very careful planning and interpretation for the dosimetry of workers exposed to a non-regular radiation field of high intensity.