• Journal of Radiation Protection and Research
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• v.26 no.2
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• pp.93-99
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• 2001

Voxel head phantom for overcoming the limitation of mathematical phantom in depleting anatomical details was constructed and example dose calculation for BNCT was performed. The repeated structure algorithm of the general purpose Monte Carlo code, MCNP4B was applied for yokel Monte Carlo calculation. Simple binary yokel phantom and combinatorial geometry phantom composed of two materials were constructed for validating the voxel Monte Carlo calculation system. The tomographic images of VHP man provided by NLM(National Library of Medicine) were segmented and indexed to construct yokel head phantom. Comparison of doses for broad parallel gamma and neutron beams in AP and PA directions showed decrease of brain dose due to the attenuation of neutron in eye balls in case of yokel head phantom. The spherical tumor volume with diameter, 5cm was defined in the center of brain for BNCT dose calculation in which accurate 3 dimensional dose calculation is essential. As a result of BNCT dose calculation for downward neutron beam of 10keV and 40keV, the tumor dose is about doubled when boron concentration ratio between the tumor to the normal tissue is $30{\mu}g/g$ to $3{\mu}g/g$. This study established the voxel Monte Carlo calculation system and suggested the feasibility of precise dose calculation in therapeutic radiology.

• Journal of Radiation Protection and Research
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• v.18 no.1
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• pp.63-70
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• 1993

Double ionization of the K- shell accompanying K- shell electron capture of the 0.035 MeV transition of $^{125}I$ has been studied by counting coincidences between $K_{\alpha}$ hypersatellite X-rays and Ka satellite X-rays emitted when double vacancies are filled. The $^{125}I\;and\;^{125}Te^m$ source materials were used in the measurement. We obtained the coincidence spectrum using two NaI(T1) detectors and a Ge(Li) detector and TAC(Time-to-Amplitude Converter), and then analysed the measured coincidence number $N(K_{\alpha}^{II},\;K_{\alpha}^s)$, the total number $N(K_{\alpha})$ of K X-ray. The probability per K-shell electron capture that a double vacancy is formed, $P_{KK}$ is formed, $P_{KK}$ is found to be $2.15{\times}10^{-4}$.

• Journal of Radiation Protection and Research
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• v.45 no.3
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• pp.142-146
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• 2020

Background: Tritium (³H) analysis in groundwater was difficult because of its low activity. Therefore, the electrolytic enrichment method was used. To improve the detection limit and for performing simple analysis, a high-volume counting vial with the available liquid scintillation counter (LSC) was investigated. Further, it was compared with a conventional 20-mL counting vial. Materials and Methods: The LSC with the electrolytic enrichment method was used ³H analysis in groundwater. A high-volume 145-mL counting vial was compared with a conventional 20-mL counting vial to determine the counting characteristics of different LSCs. Results and Discussion: When a Quantulus LSC was used, the counting window between channels 35 and 250 was used. The background count was approximately 1.86 cpm, and the counting efficiency increased from 8% to 40% depending on the mixing ratio of the volume of sample and cocktail solution. For LSC-LB7, the optimum counting window was between 1 and 4.9 keV, which was selected by the factory (Hitachi Aloka Medical Ltd., Japan) by considering quenching using a standard external gamma source. The background count of LSC-LB7 was approximately 3.60 ± 0.29 cpm when the 145-mL vial was used and 2.22 ± 0.17 cpm when the 20-mL vial was used. The minimum detectable activity (MDA) of the 20-mL vial was greater for LSC-LB7 than for Quantulus. The MDA with the 145-mL vial was improved to 0.3 Bq/L when compared with the value of 1.6 Bq/L for the 20-mL vial. Conclusion: The counting efficiency when using the 145-mL vial was 27%, whereas it was 18% when using the 20-mL vial. This difference can be attributed to the vial volume. The figure of merit (FOM) of the 145-mL vial was four times greater than that of the 20-mL vial because the volume of the former vial is approximately seven times greater than that of the latter. Further, the MDA for ³H decreased from 1.6 to 0.3 Bq/L. The counting efficiency and FOM of LSC-LB7 was slightly less than those of Quantulus when the 20-mL vial was used. The background counting rate of the Quantulus was lower than that of the LSC-LB7.

• Journal of Radiation Protection and Research
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• v.45 no.3
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• pp.101-107
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• 2020

Background: An important lesson learned from the Fukushima accident is that the transition to the mid- and long-term phases from the emergency-response phase requires less than a year, which is not very long. It is necessary to know how much radioactive material has been deposited in an urban area to establish mid- and long-term countermeasures after a radioactive accident. Therefore, an urban deposition model that can indicate the site-specific characteristics must be developed. Materials and Methods: In this study, the generalized urban deposition velocity and the subsequent variation in radionuclide contamination were estimated based on the characteristics of the Korean urban environment. Furthermore, the application of the obtained generalized deposition velocity in a hypothetical scenario was investigated. Results and Discussion: The generalized deposition velocities of ¹³⁷Cs, ¹⁰⁶Ru, and ¹³¹I for each residence type were obtained using three-dimensional (3D) modeling. For all residence types, the deposition velocities of ¹³¹I are greater than those of ¹⁰⁶Ru and ¹³⁷Cs. In addition, we calculated the generalized deposition velocities for each residential types. Iodine was the most deposited nuclide during initial deposition. However, the concentration of iodine in urban environment drastically decreases owing to its relatively shorter half-life than ¹⁰⁶Ru and ¹³⁷Cs. Furthermore, the amount of radioactive material deposited in nonresidential areas, especially in parks and schools, is more than that deposited in residential areas. Conclusion: In this study, the generalized urban deposition velocities and the subsequent deposition changes were estimated for the Korean urban environment. The 3D modeling was performed for each type of urban residential area, and the average deposition velocity was obtained and applied to a hypothetical accident. Based on the estimated deposition velocities, the decision-making systems can be improved for responding to radioactive contamination in urban areas. Furthermore, this study can be useful to predict the radiological dose in case of large-scale urban contamination and can support decision-making for long-term measurement after nuclear accident.

• Journal of Radiation Protection and Research
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• v.38 no.4
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• pp.172-178
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• 2013

In this study, neutron detection efficiency of the UNDA system, which has been developed for study on nuclear material accountancy in a future pyro-process facility, was evaluated by using the MCNPX code. The detection efficiency was evaluated as a function of (1) positions of $^{252}Cf$ neutron source in the axial and radial directions, and (2) thicknesses and locations of the container filled with the depleted uranium materials for two different designs of the UNDA. In the case of $^{252}Cf$ source positions, detection efficiency was distributed from 6.83% to 13.35%. As $^{252}Cf$ source was positioned at upper part in the axial direction, detection efficiency was decreased after a slight increase. On the other hands, as $^{252}Cf$ source was positioned at outer part in the radial direction, detection efficiency was increased. In the case of container thickness, there was a slight decline when the thickness was increased. As the container was located at upper part, detection efficiency was decreased and as the container was located at outer part, detection efficiency was increased. Detection efficiency was varied from 10.31% to 13.61%. These values were higher than that of $^{252}Cf$ source case. The UNDA with polyethylene cover has about 2% higher detection efficiency than the UNDA without the cover.

• Journal of Radiation Protection and Research
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• v.42 no.2
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• pp.114-129
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• 2017

Background: The purpose of this paper is to confirm the event timings and the magnitude of fission product aerosol release from the Fukushima accident. Over a few hundreds of technical papers have been published on the environmental impact of Fukushima Daiichi accident since the accident occurred on March 11, 2011. However, most of the research used reverse or inverse method based on the monitoring of activities in the remote places and only few papers attempted to estimate the release of fission products from individual reactor core or from individual spent fuel pool. Severe accident analysis code can be used to estimate the radioactive release from which reactor core and from which radionuclide the peaks in monitoring points can be generated. Materials and Methods: The basic material used for this study are the initial core inventory obtained from the report JAEA-Data/Code 2012-018 and the given accident scenarios provided by Japanese Government or Tokyo Electric Power Company (TEPCO) in official reports. In this research a forward method using severe accident progression code is used as it might be useful for justifying the results of reverse or inverse method or vice versa. Results and Discussion: The release timing and amounts to the environment are estimated for volatile radioactive fission products such as noble gases, cesium, iodine, and tellurium up to 184 hours (about 7.7 days) after earthquake occurs. The in-plant fission product behaviors and release characteristics to environment are estimated using the severe accident progression analysis code, MELCOR, for Fukushima Daiichi accident. These results are compared with other research results which are summarized in UNSCEAR 2013 Report and other technical papers. Also it may provide the physically based arguments for justifying or suspecting the rationale for the scenarios provided in open literature. Conclusion: The estimated results by MELCOR code simulation of this study indicate that the release amount of volatile fission products to environment from Units 1, 2, and 3 cores is well within the range estimated by the reverse or inverse method, which are summarized in UNSCEAR 2013 report. But this does not necessarily mean that these two approaches are consistent.

• Journal of Radiation Protection and Research
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• v.41 no.3
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• pp.245-252
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• 2016

Background: Ingestion of $^{210}Po$ laden seafood accounts for a substantial amount of the effective dose of $^{210}Po$. Among seafood items, mollusks, especially domestically produced oysters and mussels, are highly enriched in $^{210}Po$ and are consumed in large quantities in Korea. Materials and Methods: Oysters and mussels around the Korean coasts were collected from major farm areas in November 2013. Samples were spiked with an aliquot of $^{209}Po$ as a yield tracer, and they were digested with $6mol{\cdot}L^{-1}$ $HNO_3$ and $H_2O_2$. The $^{210}Po$ and $^{209}Po$ were spontaneously deposited onto a silver disc in an acidic solution of $0.5mol{\cdot}L^{-1}$ HCl and measured using an alpha spectrometer. The activity concentrations of $^{210}Pb$ and $^{210}Po$ were decay corrected to the sampling date, accounting for the possible in-growth and decay of $^{210}Po$. Results and Discussion: $^{210}Po$ activity concentrations in oysters were in a range from 41.3 to $206Bq{\cdot}(kg-ww)^{-1}$ and mussels in a range from 42.9 to $46.7Bq{\cdot}(kg-ww)^{-1}$. The $^{210}Po$ activity concentration of oysters in the turbid Western coast was higher than the Southern coast. The $^{210}Po$ activity concentration of the oysters was positively correlated ($R^2=0.89$) with those of the suspended particulate matter in the surface water. The calculated annual effective dose of $^{210}Po$ from oysters and mussels consumed by the Korean population was 21-104 and $5.01-5.46{\mu}Sv{\cdot}y^{-1}$. The combined effective dose due to the consumption of oysters and mussels appears to account for about $35{\pm}19%$ of that arising from seafood consumption in the Korean population. Conclusion: The annual effective dose of $^{210}Po$ for oysters in the Korean population was found to be higher than other countries. The total annual effective dose of $^{210}Po$ due to consumption of oysters and mussels consumed in Korea was found to be $76{\pm}42{\mu}Sv{\cdot}y^{-1}$, accounting for $28{\pm}16%$ of the total effective dose of $^{210}Po$ from food in Korea.

• Journal of Radiation Protection and Research
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• v.41 no.4
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• pp.389-394
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• 2016

Background: Tetrahedral-mesh geometries can be used in the MCNP code, but the MCNP code accepts only the geometry in the Abaqus input file format; hence, the existing tetrahedral-mesh models first need to be converted to the Abacus input file format to be used in the MCNP code. In the present study, we developed a simple but useful computer program, TET2MCNP, for converting TetGen-generated tetrahedral-mesh models to the Abacus input file format. Materials and Methods: TET2MCNP is written in C++ and contains two components: one for converting a TetGen output file to the Abacus input file and the other for the reverse conversion process. The TET2MCP program also produces an MCNP input file. Further, the program provides some MCNP-specific functions: the maximum number of elements (i.e., tetrahedrons) per part can be limited, and the material density of each element can be transferred to the MCNP input file. Results and Discussion: To test the developed program, two tetrahedral-mesh models were generated using TetGen and converted to the Abaqus input file format using TET2MCNP. Subsequently, the converted files were used in the MCNP code to calculate the object- and organ-averaged absorbed dose in the sphere and phantom, respectively. The results show that the converted models provide, within statistical uncertainties, identical dose values to those obtained using the PHITS code, which uses the original tetrahedral-mesh models produced by the TetGen program. The results show that the developed program can successfully convert TetGen tetrahedral-mesh models to Abacus input files. Conclusion: In the present study, we have developed a computer program, TET2MCNP, which can be used to convert TetGen-generated tetrahedral-mesh models to the Abaqus input file format for use in the MCNP code. We believe this program will be used by many MCNP users for implementing complex tetrahedral-mesh models, including computational human phantoms, in the MCNP code.

• Journal of Radiation Protection and Research
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• v.41 no.4
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• pp.395-401
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• 2016

Background: Liquid scintillation counters (LSCs) are commonly used as an analytical method for detecting $^{222}Rn$ in groundwater because they involve a simple sample pretreatment and allow high throughout with an autosampler. The Quantulus 1220 is the best-selling LSC in Korea, but its production was stopped. Recently, a new type of LSC, the 300SL, was introduced. In this study, the 300SL was compared with the Quantulus 1220 in order to evaluate the ability of each apparatus to detect $^{222}Rn$ in groundwater. Materials and Methods: The Quantulus 1220 and 300SL were used to detect the presence of $^{222}Rn$. Radon gas was extracted from a groundwater sample using a water-immiscible cocktail in a LSC vial. The optimal analytical conditions for each LSC were determined using a $^{222}Rn$ calibration source prepared with a $^{226}Ra$ source. Results and Discussion: The optimal pulse shape analysis level for alpha and beta separation was 80 for the Quantulus 1220, and the corresponding pulse length index was 12 in the 300SL. The counting efficiency of the Quantulus 1220 for alpha emissions was similar to that of the 300SL, but the background count rate of the Quantulus 1220 was 10 times lower than that of the 300SL. The minimum detectable activity of the Quantulus 1220 was $0.08Bq{\cdot}L^{-1}$, while that of the 300SL was $0.20Bq{\cdot}L^{-1}$. The analytical results regarding $^{222}Rn$ in groundwater were less than 10% different between these LSCs. Conclusion: The 300SL is an LSC that is comparable to the Quantulus 1220 for detecting $^{222}Rn$ in groundwater. Both LSCs can be applied to determine the levels of $^{222}Rn$ in groundwater under the management of the Ministry of Environment.

• Journal of Radiation Protection and Research
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• v.29 no.3
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• pp.187-196
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• 2004

The health effects resulting from severe accidents of typical 1,000MWe KSNP(Korea Standard Nuclear Plant) PWR and typical 600MWe CANDU(CANada Deuterium Uranium) plants were estimated and compared. The population distribution of the site extending to 80km for both site were considered. The releaese fraction for various source term categories(STC) and core inventories were used in the estimation of the health effects risks by using the MACCS2(MELCOR Accident Consequence Code System2) code. Individuals are assumed to evacuate beyond 16km from the site. The health effects considered in this comparative study are early and cancer fatality risk, and the results are presented as CCDF(Complementary Cumulative Distribution Function) curves considering the occurrence probability of each STC's. According to the results, the early and cancer fatality risks of PHWR plants we lower than those of PWR plants. This is attributed the fact that the amount of radioactive mateials that released to the atmosphere resulting from the postulated severe accidents of PHWR plants are smaller than that of PWR plants. And, the dominating initiating event of STC that shows maximum early and cancer fatality risk is SGTR(Steam Generator Tube Rupture) for both plants. Therefore, the appropriated actions must be taken to reduce the occurrence probability and the amounts of radioactive materials released to the environment in order to protect the public for both PWR and PHWR plants.