• Journal of Radiation Protection and Research
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• v.41 no.2
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• pp.101-104
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• 2016

Background: To protect the public from natural radioactive materials, the 'Act on safety control of radioactive rays around living environment" was established in Korea. There is an annual effective dose limit of 1 mSv for products, but the activity concentration limit for products is not established yet. Materials and Methods: To suggest the activity concentration limits for consumer goods containing NORM, in this research, we assumed the "small room model" surrounding the ICRP reference phantom to simulate the consumer goods in contact with the human bodies. Using the Monte Carlo code MCNPX, we evaluate the effective dose rate for the ICRP reference phantom in a small room with dimension of phantom size and derived the activity concentration limit for consumer goods. Results and Discussion: The consumer goods have about 1600, 1200 and $19000Bq{\cdot}kg^{-1}$ for $^{226}Ra$, $^{232}Th$ and $^{40}K$, and the activity concentration limits are about six times comparing with the values of building materials. We applied the index to real samples, though we did not consider radioactivity of $^{40}K$, indexes of the some samples are more than 6. However, this index concept using small room model is very conservative, for the consumer goods over than index 6, it is necessary to reevaluate the absorbed dose considering real usage scenario and material characteristics. Conclusion: In this research, we derived activity concentration limits for consumer goods in contact with bodies and the results can be used as preliminary screening tool for consumer goods as index concept.

• Journal of Radiation Industry
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• v.5 no.4
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• pp.325-330
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• 2011

To predict the details of flow in industrial process unit, single photon emission computed tomography (SPECT) is a promising technique. Recently, industrial SPECT based on medical system has developed by researchers of the Korea Atomic Energy Research Institute (KAERI) and Hanyang University. In the present study, to confirm the effect of the number of detectors on image quality, and determine the optimal number of detectors in industrial SPECT, industrial SPECT system with various geometries were evaluated by the Monte Carlo simulation. CsI(Tl) detectors ($12mm{\times}12mm{\times}20mm$) with collimators (the geometric resolution of collimator $R_g$ was 4 cm at the center of the 30 cm diameter cylindrical vessel object) were modeled in a hexagonal array, and the point sources of $^{99m}Tc$, $^{68}Ga$, and $^{137}Cs$ were simulated at the center of the cylindrical vessel object using the MCNPX code. Then, the reconstruction images of each geometry were reconstructed using the expectation maximization (EM) algorithm. In this study, the reciprocity theorem was used to improve computation time required for system matrix of the EM algorithm. The result shows that the resolution of the reconstructed image was significantly improved by increasing the number of detectors in industrial SPECT system and more than 60 detectors will be required for the resolution of the reconstructed image.

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.875-883
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• 2015

The use of subcritical aqueous homogenous reactors driven by accelerators presents an attractive alternative for producing $^{99}Mo$. In this method, the medical isotope production system itself is used to extract $^{99}Mo$ or other radioisotopes so that there is no need to irradiate common targets. In addition, it can operate at much lower power compared to a traditional reactor to produce the same amount of $^{99}Mo$ by irradiating targets. In this study, the neutronic performance and $^{99}Mo$, $^{89}Sr$, and $^{131}I$ production capacity of a subcritical aqueous homogenous reactor fueled with low-enriched uranyl nitrate was evaluated using the MCNPX code. A proton accelerator with a maximum 30-MeV accelerating power was used to run the subcritical core. The computational results indicate a good potential for the modeled system to produce the radioisotopes under completely safe conditions because of the high negative reactivity coefficients of the modeled core. The results show that application of an optimized beam window material can increase the fission power of the aqueous nitrate fuel up to 80%. This accelerator-based procedure using low enriched uranium nitrate fuel to produce radioisotopes presents a potentially competitive alternative in comparison with the reactor-based or other accelerator-based methods. This system produces ~1,500 Ci/wk (~325 6-day Ci) of $^{99}Mo$ at the end of a cycle.

• Journal of Radiation Protection and Research
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• v.46 no.2
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• pp.58-65
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• 2021

Background: The evaluation of skyshine distribution, release of airborne radioactive nuclides, and soil activation and groundwater migration were required for radiological assessment of the impact on the environment surrounding In-Flight (IF)-system facility of the RAON (Rare isotope Accelerator complex for ON-line experiment) accelerator complex. Materials and Methods: Monte Carlo simulation by MCNPX code was used for evaluation of skyshine and activation analysis for air and soil. The concentration model was applied in the estimation of the groundwater migration of radionuclides in soil. Results and Discussion: The skyshine dose rates at 1 km from the facility were evaluated as 1.62 × 10^-3 μSv·hr^-1. The annual releases of ³H and ¹⁴C were calculated as 9.62 × 10^-5 mg and 1.19 × 10^-1 mg, respectively. The concentrations of ³H and ²²Na in drinking water were estimated as 1.22 × 10^-1 Bq·cm^-3 and 8.25 × 10^-3 Bq·cm^-3, respectively. Conclusion: Radiological assessment of environmental impact on the IF-facility of RAON was performed through evaluation of skyshine dose distribution, evaluation of annual emission of long-lived radionuclides in the air and estimation of soil activation and groundwater migration of radionuclides. As a result, much lower exposure than the limit value for the public, 1 mSv·yr^-1, is expected during operation of the IF-facility.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.167-179
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• 2024

This study explores the feasibility of employing (U, Th)-based accident tolerant fuels (ATFs), specifically (_0.8UO₂, _0.2ThO₂), (_0.8UN, _0.2ThN), and (_0.8UC, _0.2ThC). The investigation assesses the overall performance of these proposed fuel materials in comparison to the conventional UO₂, focusing on deep neutronic and thermal-hydraulic (Th) analyses. Neutronic analysis utilized the MCNPX code, while COMSOL Multiphysics was employed for thermal-hydraulic analysis. The primary objective of this research is to overcome the limitations associated with traditional UO₂ fuel by exploring alternative fuel materials that offer advantages in terms of abundance and potential improvements in performance and safety. Given the limited abundance of UO₂, long-term sustainable nuclear energy production faces challenges. From a neutronic standpoint, the U-Th based fuels demonstrated remarkable fuel cycle lengths, except (_0.8UN, _0.2ThN), which exhibited the minimum fuel cycle length and, consequently, the lowest fuel burn-up. Regarding thermal-hydraulic performance, (_0.8UN, _0.2ThN) exhibited outstanding performance with significant margins against fuel melting compared to the other materials. Overall, when considering the integrated performance, the most favourable results were obtained with the use of the (_0.8UC, _0.2ThC) fuel configurations. This study contributes valuable insights into the potential benefits of (U, Th)-based ATFs as a promising avenue for enhanced nuclear fuel performance.

• Journal of Radiation Protection and Research
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• v.37 no.4
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• pp.181-190
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• 2012

Naturally occurring radioactive materials (NORM) in building materials are main sources of external radiation exposure to the general public. The objective of this study was to assess external radiation dose in Korean dwellings due to NORM in concrete walls. Reference room model for dose assessment was made by analyzing room structure and housing scale of Korean dwellings. In addition, dose assessments were made for varying room sizes. Absorbed doses to air and effective dose rates were calculated using radiation transport code MCNPX. Assuming a reference room of $3{\times}4{\times}2.8m^3$, absorbed dose rates in air were 0.80, 0.97, 0.08 nGy $h^{-1}$ per Bq $kg^{-1}$ for uranium series, thorium series, and $^{40}K$, respectively. Effective dose rates were 0.57, 0.69, 0.058 nSv $h^{-1}$ per Bq $kg^{-1}$, respectively. Radiation dose resulting from concrete of ceiling and floor increased with room area while radiation dose from concrete of walls decreased with room area. Therefore, total radiation doses were almost the same for the varying room area from 5 to $30m^2$. Effective dose in Korean dwellings was calculated based on measurement data of NORM concentration in concrete and occupancy fraction of Korean population by location. Annual effective dose was 0.59 mSv assuming that indoor occupancy fraction was 0.89 and concentrations of uranium series, thorium series and $^{40}K$ were 26, 39, 596 Bq $kg^{-1}$, respectively. Finally, annual effective dose in Korean dwellings can be calculated by the following equation: Effective dose=indoor occupancy fraction${\times}8760\;h\;y^{-1}{\times}(0.57C_U+0.69C_{Th}+0.058C_K)$.

• Journal of Radiation Protection and Research
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• v.41 no.1
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• pp.71-79
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• 2016

Background: It is necessary to analyze radioactivity of naturally occurring radioactive materials (NORM) in products to ensure radiological safety required by Natural Radiation Safety Management Act. The pretreatments for the existing analysis methods require high technology and time. Such destructive pretreatments including grinding and dissolution of samples make impossible to reuse products. We developed a rapid and simple procedure of radioactivity analysis for thorium series in the products containing NORM. Materials and Methods: The developed method requires non-destructive or minimized pretreatment. Radioactivity of the product without pretreatment is initially measured using gamma spectroscopy and then the measured radioactivity is adjusted by considering material composition, mass density, and geometrical shape of the product. The radioactivity adjustment can be made using scaling factors, which is derived by radiation transport Monte Carlo simulation. Necklace, bracelet, male health care product, and tile for health mat were selected as representative products for this study. The products are commonly used by the public and directly contacted with human body and thus resulting in high radiation exposure to the user. Results and Discussion: The scaling factors were derived using MCNPX code and the values ranged from 0.31 to 0.47. If radioactivity of the products is measured without pretreatment, the thorium series may be overestimated by up to 2.8 times. If scaling factors are applied, the difference in radioactivity estimates are reduced to 3-24%. Conclusion : The developed procedure in this study can be used for other products with various materials and shapes and thus ensuring radiological safety.

• The Journal of Korean Society for Radiation Therapy
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• v.25 no.2
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• pp.145-151
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• 2013

Purpose: In this study, we analyzed how the dose change by field size effects on atomic number of shielding materials while using 6 MeV election beam. Materials and Methods: The parallel plate chamber is mounted in $25{\times}25cm^2$ the phantom such that the entrance window of the detector is flush with the phantom surface. phantom was covered laterally with aluminum, copper and lead which thickness have 5% of allowable transmission and then the doses were measured in field size $6{\times}6$, $10{\times}10$ and $20{\times}20cm^2$ respectively. 100 cGy was irradiated using 6 MeV electron beam and SSD (Source Surface Distance) was 100 cm with $10{\times}10cm^2$ field size. To calculate the photon flux, electron flux and Energy deposition produced after pass materals respectively, MCNPX code was used. Results: The results according to the various shielding materials which have 5% of allowable transmission are as in the following. Thickness change rate with field size of $6{\times}6cm^2$ and $20{\times}20cm^2$ that compared to the field size of $10{\times}10cm^2$ found to be +0.06% and -0.06% with aluminum, +0.13% and -0.1% with copper, -1.53% and +1.92% with lead respectively. Compare to the field size $10{\times}10cm^2$, energy deposition for $6{\times}6cm^2$ and $20{\times}20cm^2$ had -4.3% and +4.85% respectively without shielding material. With aluminum it had -0.87% and +6.93% respectively and with lead it had -4.16% and +5.57% respectively. When it comes to photon flux with $6{\times}6cm^2$ and $20{\times}20cm^2$ of field sizes the chance -8.95% and +15.92% without shielding material respectively, with aluminum the number -15.56% and +16.06% respectively and with copper the chance -12.27% and +15.53% respectively, with lead the number +12.36% and -19.81% respectively. In case of electron flux in the same condition, the number -3.92% and +4.55% respectively without shielding material respectively, with aluminum the number +0.59% and +6.87% respectively, with copper the number -1.59% and +3.86% respectively, with lead the chance -5.15% and +4.00% respectively. Conclusion: In this study, we found that the required thickness of the shielding materials got thinner with low atomic number substance as the irradiation field is increasing. On the other hand, with high atomic number substance the required thickness had increased. In addition, bremsstrahlung radiation have an influence on low atomic number materials and high atomic number materials are effected by scattered electrons.

• Journal of Radiation Protection and Research
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• v.45 no.2
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• pp.69-75
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• 2020

Background: Dose conversion coefficients (DCCs) have been commonly used to estimate radiation-dose absorption by human organs based on physical measurements of fluence or kerma. The International Commission on Radiological Protection (ICRP) has reported a library of DCCs, but few studies have been conducted on their applicability to non-Caucasian populations. In the present study, we collected a total of 8 Korean pediatric and adult voxel phantoms to calculate the organ DCCs for idealized external photon-irradiation geometries. Materials and Methods: We adopted one pediatric female phantom (ETRI Child), two adult female phantoms (KORWOMAN and HDRK Female), and five adult male phantoms (KORMAN, ETRI Man, KTMAN1, KTMAN2, and HDRK Man). A general-purpose Monte Carlo radiation transport code, MCNPX2.7 (Monte Carlo N-Particle Transport extended version 2.7), was employed to calculate the DCCs for 13 major radiosensitive organs in six irradiation geometries (anteroposterior, posteroanterior, right lateral, left lateral, rotational, and isotropic) and 33 photon energy bins (0.01-20 MeV). Results and Discussion: The DCCs for major radiosensitive organs (e.g., lungs and colon) in anteroposterior geometry agreed reasonably well across the 8 Korean phantoms, whereas those for deep-seated organs (e.g., gonads) varied significantly. The DCCs of the child phantom were greater than those of the adult phantoms. A comparison with the ICRP Publication 116 data showed reasonable agreements with the Korean phantom-based data. The variations in organ DCCs were well explained using the distribution of organ depths from the phantom surface. Conclusion: A library of dose conversion coefficients for major radiosensitive organs in a series of pediatric and adult Korean voxel phantoms was established and compared with the reference data from the ICRP. This comparison showed that our Korean phantom-based data agrees reasonably with the ICRP reference data.

• Journal of radiological science and technology
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• v.40 no.1
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• pp.93-99
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• 2017

Dose enhancement effects at megavoltage (MV) X and ${\gamma}-ray$ energies, and the effects of different energy levels on incident energy, dose enhancement agents, and concentrations were analyzed using Monte Carlo simulations. Gold, gadolinium, Iodine, and iron oxide ($Fe_2O_3$) were compared as dose enhancement agents. For incident energy, 4, 6, 10 and 15 MV X-ray spectra produced by a linear accelerator and a Co-60 ${\gamma}-ray$ were used. The dose enhancement factor (DEF) was calculated using an ICRU Slab phantom for concentrations of 7, 18, and 30 mg/g. The DEF was higher at higher concentrations of dose enhancement agents and at lower incident energies. The calculated DEF ranged from 1.035 to 1.079, and dose enhancement effects were highest for iron oxide, followed by iodine, gadolinium, and gold. Thus, this study contributes to improving the therapeutic ratio by delivering larger doses of radiation to tumor volume, and provides data to support further in vivo and in vitro studies.