• Journal of radiological science and technology
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• v.47 no.3
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• pp.197-203
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• 2024

During paranasal sinus X-ray examinations in children, the radiological technologist's thyroid shield is often not implemented to shorten the examination time. This study measured the radiation exposure before and after the implementation of thyroid shielding by analyzing the difference in radiation exposure, the radiological technologist's could receive depending on the actual thyroid shielding. In the left TLD, when thyroid shielding was not performed(N), the radiation exposure dose(mSv) was 2.869 for the depth dose[Hp(10)] and 2.886 for the surface dose[H(3)], and when thyroid shielding was performed(Y), the Hp(10) was 0.033 and the H(3) was 0.034. In the right TLD, when thyroid shielding was not performed(N), the radiation exposure dose was 3.149 for Hp(10) and 3.137 for H(3), and when thyroid shielding was performed, the Hp(10) of (Y) was 0.013 and the H(3) was 0.015. The differences in the overall exposure dose measurement values are all statistically significant (p<0.05). The difference in radiation dose between when thyroid shielding was not performed and when thyroid shielding was performed was more than 99.2% in both cases, indicating a high radiation shielding rate.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.3
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• pp.371-381
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• 2023

Evaluating the effectiveness of the radiation protection measures deployed at the Centralized Radioactive Waste Management Facility in Ghana is pivotal to guaranteeing the safety of personnel, public and the environment, thus the need for this study. Radiagem^TM 2000 was used in measuring the dose rate of the facility whilst the personal radiation exposure of the personnel from 2011 to 2022 was measured from the thermoluminescent dosimeter badges using Harshaw 6600 Plus Automated TLD Reader. The decay store containing scrap metals from dismantled disused sealed radioactive sources (DSRS), and low-level wastes measured the highest dose rate of 1.06 ± 0.92 µSv·h^-1. The range of the mean annual average personnel dose equivalent is 0.41-2.07 mSv. The annual effective doses are below the ICRP limit of 20 mSv. From the multivariate principal component analysis biplot, all the personal dose equivalent formed a cluster, and the cluster is mostly influenced by the radiological data from the outer wall surface of the facility where no DSRS are stored. The personal dose equivalents are not primarily due to the radiation exposures of staff during operations with DSRS at the facility but can be attributed to environmental radiation, thus the current radiation protection measures at the Facility can be deemed as effective.

• Journal of Radiation Protection and Research
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• v.48 no.3
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• pp.131-143
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• 2023

Background: This study examined the detection limit of thyroid screening monitoring conducted at the time of the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident in 2011 using a Monte Carlo simulation. Materials and Methods: We calculated the detection limit of a NaI(Tl) survey meter to measure ¹³¹I accumulation in the thyroid gland of children. Mathematical phantoms of 1- and 5-year-old children were developed in the simulation of the Particle and Heavy Ion Transport code System code. Contamination of the body surface with eight radionuclides found after the FDNPP accident was assumed to have been deposited on the neck and shoulder area. Results and Discussion: The detection limit was calculated as a function of ambient dose rate. In the case of 40 Bq/cm² contamination on the body surface of the neck, the present simulations showed that residual thyroid radioactivity corresponding to thyroid dose of 100 mSv can be detected within 21 days after intake at the ambient dose rate of 0.2 µSv/hr and within 11 days in the case of 2.0 µSv/hr. When a time constant of 10 seconds was used at the dose rate of 0.2 µSv/hr, the estimated survey meter output error was 5%. Evaluation of the effect of individual differences in the location of the thyroid gland confirmed that the measured value would decrease by approximately 6% for a height difference of ±1 cm and increase by approximately 65% for a depth of 1 cm. Conclusion: In the event of a nuclear disaster, simple measurements carried out using a NaI(Tl) scintillation survey meter remain effective for assessing ¹³¹I intake. However, it should be noted that the presence of short-half-life radioactive materials on the body surface affects the detection limit.

• Progress in Medical Physics
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• v.19 no.2
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• pp.113-119
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• 2008

In this study, we have estimated error of calculation results for 5-type RTP systems and investigated a toleration for error of the RTPs referenced from the evaluation items of AAPM Report-62. For this study, we have introduced the concept of 'normal dose rate(NDR)' and compared the results of experiment and calculation from RTPs at the same reference level. The results from all RTPs were satisfied at various field shapes and heterogeneous phantom materials except the surface irregularity.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.380-387
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• 2015

A lead slowing-down spectrometer (LSDS) system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as $^{235}U$, $^{239}Pu$, $^{241}Pu$, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea) is planned to utilize a high-flux ($>10^{12}n/cm^2{\cdot}s$) neutron source comprised of a high-energy (30 MeV)/high-current (~2 A) electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (< $0.06{\mu}Sv/h$), a few shielding materials [high-density polyethylene (HDPE)eBorax, $B_4C$, and $Li_2CO_3$] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in the near future.

• Journal of radiological science and technology
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• v.16 no.2
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• pp.27-38
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• 1993

To investigate relationships between image guality and exposure dose, Chest X-ray films were evaluated for the following points:how much scattered radiation can affect reduction in image quality and can be permissible diagnostically? For this purpose using a test charts and Burger's phantoms. The visual evaluation of their X-ray films and the measurements of scattered radiation were carried out. The dose of scattered radiation ranging from 20 to 25% was found to be for nothing in any diagnostic obstacle. In this range, surface doses were low of 17, 21, and $25{\mu}Gy$ for The thickness of the chest of 15, 20 and 25 cm respectively. Comparison of these high voltage X-ray films with low voltage ones showed a surface dose rate of 1:11.7. Therefore, X-ray quality, photosensitive materials(film and screen) and grid should be selected very carefully for the purpose of reduction in exposure dose.

• Journal of Radiation Protection and Research
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• v.16 no.2
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• pp.17-26
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• 1991

Beta ray $(^{90}Sr+^{90}Y)$ absorbed dose at tissue surface was measured from the distance of 30cm by use of extrapolation chamber. In the measurement, following factors were considered: effective area of collecting electrode, polarity effect, ion recombination and window attenuation. The measured absorbed dose rate at tissue surface was $1.493{\mu}Gy/sec$ with ${\pm}2.9%$.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.1
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• pp.123-132
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• 2021

Radiological impact analyses were carried out for a near-surface radioactive waste repository at Gyeongju in South Korea. The RESRAD-ONSITE code was applied for the estimation of maximum exposure doses by considering various exposure pathways based on a land area of 2,500 ㎡ with a 0.15 m thick contamination zone. Typical influencing input parameters such as shield depth, shield materials' density, and shield erosion rate were examined for a sensitivity analysis. Then both residential farmer and industrial worker scenarios were used for the estimation of maximum exposure doses depending on exposure duration. The radiation dose evaluation results showed that 60Co, 137Cs, and 63Ni were major contributors to the total exposure dose compared with other radionuclides. Furthermore, the total exposure dose from ingestion (plant, meat, and milk) of the contaminated plants was more significant than those assessed for inhalation, with maximum values of 5.5×10-4 mSv·yr-1 for the plant ingestion. Thus the results of this study can be applied for determining near-surface radioactive waste repository conditions and providing quantitative analysis methods using RESRAD-ONSITE code for the safety assessment of disposing radioactive materials including decommissioning wastes to protect human health and the environment.

• The Journal of Engineering Geology
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• v.30 no.4
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• pp.577-588
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• 2020

We evaluated the distributions of primordial radionuclides and effective dose rate of the Ogcheon Group, which includes rocks with high uranium content. Terrestrial gamma radiation was measured at 421 points using a portable gamma ray spectrometer. Dividing the study area into five geological units (og1, og2, og3, og4, and igneous rocks) revealed no significant difference in the concentration of surface radioactivity among the types. The concentrations of 40K, eU, and eTh for all samples ranged from 0.7% to 10.3% (average 5.2%), 0.6 to 287.0 ppm (average 8.5 ppm), and 4.0 to 102.4 ppm (average 31.3 ppm), respectively. The absorbed dose rate in the study area (calculated from the activity concentrations of 40K, eU, and eTh) was in the range of 28.84 to 1,714.5 nGy/h (average 195.4 nGy/h). Among the five geological units, the lowest average was 166.3 nGy/h (for og1) and the highest average was 233.3 nGy/h (for og2; median 198.1 nGy/h). The outdoor effective dose rate for the area obtained from the absorbed dose rate was in the range of 0.04 to 2.10 mSv/y (average 0.24 mSv/y). Except for the four sites located in the uranium-bearing coal bed of og2, none of the studied sites exceeded 1 mSv/y.

• Radiation Oncology Journal
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• v.10 no.1
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• pp.85-93
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• 1992

We have studied the dosimetric properties of electron beam using Lyon intraoperative device for intraoperative radiation therapy. The dosimetry data had compiled in such a way that a quick and correct decision regarding the cone shape, energy, and accurate calculations could be made. Using 3 dimensional water phantom, we have got the following data: cone output ratios, surface dose, $d_{max}$, $d_{90}$, flatness, symmetry, beam profiles, isodose curve, and SSD correction factors. The cone output ratios were measured with straight and bevelled cone, respectively. As the cone size and the energy were reduced, the cone output ratios decreased rapidly. With the flattening filter, the surface dose increased by electron beam to $85.3\%$, $89.2\%$, and $93.4\%$, for 6 MeV, 9 MeV, and 12 MeV, respectively. It is important to increase the surface dose to $90\%$ or more. Inspite of diminishing dose rate and beam penetration, this flattening filter increases the treatment volume significantly. With the combination of the three levels collimation and the flattening filter, we achieved good homogeneity of the beam and better flatness and the diameter of the 90$\%$ isodose curve was increased. It is important to increase the area that is included in the $90\%$ isodose level. The value of measured and calculated SSD correction factors did not agree over the clinically important range from 100 cm to 110 cm.