• 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.

• Journal of radiological science and technology
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• v.26 no.3
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• pp.51-57
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• 2003

Mitigation of fast neutron irradiation damage on reactor vessel and improvement of mechanical integrity are desired for the successful plant life-time extension. In this study, the performance of metallic hydride for this application is reviewed and compared. First, selected prospective metallic hydrides are evaluated by MCNP code and put into the attenuation test using Cf-252 neutron source. Since for the reactor application high moderation and reflection with no absorption are favored, Z factor is introduced for the evaluation. According to the Z value estimation $ZrD_2$ and $TiD_2$ are turned out to be the most favorable fast neutron shielding materials. More thorough evaluation by computer simulation and experimentally, will be followed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.1
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• pp.11-15
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• 2008

In order to transport spent resin in a high integrated container made of high density polyethylene, a method for determining transportation grade by radioactivity analysis was developed. Ratios of radioisotopes in spent resin were derived from radioactivity analysis on spent resin. Associated curie-to-dose factors were determined to estimate radioisotope inventory from surface dose rates of spent resin. From the results, Activity limit of type A package was derived to be 1.19 TBq for HIC, and the corresponding surface dose rate was found to be 124.2 mSv/h.

• 전자공학회논문지 IE
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• v.45 no.2
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• pp.6-12
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• 2008

In this study, because of the importance of asphalt content, the target is to develop a asphalt content detector that the influence of the jurisdiction and usage of radioisotopes are limited of 100[${\mu}Ci$] or less. For development of a asphalt content detector, it is divided into three parts that is affected the change of asphalt content due to the position of neutron source. First part is to decrease the distance between asphalt mixture and neutron source, second is to install the reflector, and last is to change the history number. It performed to analysis by computer simulation and satisfactory results of error range can use to design data for development of a asphalt content detector.

• Journal of Radiation Protection and Research
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• v.21 no.4
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• pp.263-271
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• 1996

The ANSI N13.32 recommends that a study of the angular response of a dosimeter be carried out once, although no pass/fail criterion is given for angular response. Gamma dose equivalent conversion and angular dependence factors were calculated by using MCNP code for the case of ANSI N13.32 extremity phantoms(finger and arm) at the depth of $7mg/cm^2$. Those extremity dosimeters were assumed to be irradiated from both monoenergitic photons and ISO X-ray narrow beams. These calculated gamma dose equivalent conversion and angular dependence factors were compared to B. Grosswendt's result calculated by using X-ray beams. The result showed that the dose equivalent conversion factors of this study agreed well with that of B. Grosswendt for all energies within 2% except 7% in the case of the low energies. In the case of angular dependence factors comparison, they agreed within 3%. It was shown that angular dependence factors of the finger phantom decreased as the horizontal angle of the phantom increased for the ISO X-ray beams less than 60keV. For the higher energy X-ray beams range they decreased slightly around 40 degree, but then increased from this energy to 90 degree.

• Journal of Radiation Protection and Research
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• v.18 no.2
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• pp.71-79
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• 1993

Form the pure Maxwellian distribution(kT= 1.42MeV), the effects upon calibration factors of encapsulating a $^{252}Cf$ spontaneous fission neutron source were investigated to establish a standard neutron field in the Secondary Standard Dosimetry Laboratory at Korea Atomic Energy Research Institute(KAERI). A Monte Carlo code MCNP was used in simulating the encapsulation SR-Cf-100 and SR-Cf-1273 to be real conditions. The anisotropy(FI) and fluence-to-dose equivalents conversion factors$(H/{\Phi})$ were evaluated and compared with other results. As the results, the FI was determined to be 1.061 at ${\theta}=90^{\circ}$ with ${\pm}0.2%$ statistical error and the $(H/{\Phi})$ was evaluated to be $333.9 [pSv\;cm^2]\;with\;{\pm}0.5%$ statistical error, which is lower by 1.8% than that recommended by the ISO 8529. This means physically that the neutron spectrum of the unmoderated $^{252}Cf$ source in KAERI is a little more softened than that by the ISO.

• Journal of Radiation Protection and Research
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• v.28 no.2
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• pp.87-95
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• 2003

The fabrication process and the structure of PIN semiconductor detectors have been designed optimally by simulation for doping concentration and width of p+ layer, impurities re-contribution due to annealing and the current distribution due to guard ring at the sliced edges. The characteristics to radiation response has been also simulated in terms of Monte Carlo Method. The device has been fabricated on n type, $400\;{\Omega}cm$, orientation <100>, Floating-Zone silicon wafer using the simulation results. The leakage current density of $0.7nA/cm^2/100{\mu}m$ is achieved by this process. The good linearity of radiation response to Cs-137 was kept within the exposure ranges between 5 mR/h and 25 R/h. This proposed process could be applied for fabricating a PIN semiconductor detector for measuring individual dose.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.4
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• pp.375-387
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• 2019

In this study, the radiation dose rates for the design basis fuel of 360 assemblies CANDU spent nuclear fuel transportation cask were evaluated, by measuring radiation source terms for the design basis fuel of a pressurized heavy water reactor. Additionally, radiological safety evaluation was carried out and the validity of the results was determined by radiological technical standards. To select the design basis fuel, which was the radiation source term for the spent fuel transportation cask, the design basis fuels from two spent fuel storage facilities were stored in a spent fuel transportation cask operating in Wolsung NPP. The design basis fuel for each transportation and storage system was based on the burnup of spent fuel, minimum cooling period, and time of transportation to the intermediate storage facility. A burnup of 7,800 MWD/MTU and a minimum cooling period of 6 years were set as the design basis fuel. The radiation source terms of the design basis fuel were evaluated using the ORIGEN-ARP computer module of SCALE computer code. The radiation shielding of the cask was evaluated using the MCNP6 computer code. In addition, the evaluation of the radiation dose rate outside the transport cask required by the technical standard was classified into normal and accident conditions. Thus, the maximum radiation dose rates calculated at the surface of the cask and at a point 2 m from the surface of the cask under normal transportation conditions were respectively 0.330 mSv·h^-1 and 0.065 mSv·h^-1. The maximum radiation dose rate 1 m from the surface of the cask under accident conditions was calculated as 0.321 mSv·h^-1. Thus, it was confirmed that the spent fuel cask of the large capacity heavy water reactor had secured the radiation safety.

• Journal of the Korean Society of Radiology
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• v.15 no.6
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• pp.803-811
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• 2021

In this study, the exposure amount of IASCC test worker was evaluated by applying the process simulation technology. Using DELMIA Version 5, a commercial process simulation code, IASCC test facility, hot cells, and workers were prepared, and IASCC test activities were implemented, and the cumulative exposure of workers passing through the dose-distributed space could be evaluated through user coding. In order to simulate behavior of workers, human manikins with a degree of freedom of 200 or more imitating the human musculoskeletal system were applied. In order to calculate the worker's exposure, the coordinates, start time, and retention period for each posture were extracted by accessing the sub-information of the human manikin task, and the cumulative exposure was calculated by multiplying the spatial dose value by the posture retention time. The spatial dose for the exposure evaluation was calculated using MCNP6 Version 1.0, and the calculated spatial dose was embedded into the process simulation domain. As a result of comparing and analyzing the results of exposure evaluation by process simulation and typical exposure evaluation, the annual exposure to daily test work in the regular entrance was predicted at similar levels, 0.388 mSv/year and 1.334 mSv/year, respectively. Exposure assessment was also performed on special tasks performed in areas with high spatial doses, and tasks with high exposure could be easily identified, and work improvement plans could be derived intuitively through human manikin posture and spatial dose visualization of the tasks.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2_spc
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• pp.317-325
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• 2020

The Korea Radioactive Waste Agency plans to expand the storage capacity of radioactive waste by constructing a radioactive waste inspecting building to solve the problem of the lack of inspection space and drum-handling space in the radioactive waste receipt and storage building for the first-stage disposal facility. In this study, the exposure doses of radiation workers that handle new disposal containers for decommissioning waste in the storage areas of the radioactive waste inspecting building were calculated using the Monte Carlo N-particle transport code. The annual collective dose was calculated as a total of 84.8 man-mSv for 304 new disposal containers and an estimated annual 306 working hours for the radiation work. When the 304 new disposal containers (small/medium type) were stored in the storage areas, it was found that 25 radiation workers should be involved in acceptance/disposal inspection, and the estimated exposure dose per worker was calculated as an average annual value of 3.39 mSv. When the radiation workers handle the small containers in high-radiation dose areas, the small containers should be shielded further by increasing the concrete liner thickness to improve the work efficiency and radiation safety of the radiation workers. The results of this study will be useful in establishing the optimal radiation working conditions for radiation workers using the source term and characteristics of decommissioning waste based on actual measurements.