Earticle

Background: This study evaluated the atmospheric dispersion of radioactive material according to local weather conditions and emission conditions. Materials and Methods: Local weather conditions were defined as 8 patterns that frequently occur around the Kori Nuclear Power Plant and emission conditions were defined as 6 patterns from a combination of emission rates and the total number of particles of the 137Cs, using the WRF/HYSPLIT modeling system. Results and Discussion: The highest mean concentration of 137Cs occurred at 0900 LST under the ME4_1 (main wind direction: SSW, daily average wind speed: 2.8 ms-1), with a wide region of its high concentration due to the continuous wind changes between 0000 and 0900 LST; under the ME3 (NE, 4.1 ms-1), the highest mean concentration of 137Cs occurred at 1500 and 2100 LST with a narrow dispersion along a strong northeasterly wind. In the case of ME4_4 (S, 2.7 ms-1), the highest mean concentration of 137Cs occurred at 0300 LST because 137Cs stayed around the KNPP under low wind speed and low boundary layer height. As for the emission conditions, EM1_3 and EM2_3 that had the maximum total number of particles showed the widest dispersion of 137Cs, while its highest mean concentration was estimated under the EM1_1 considering the relatively narrow dispersion and high emission rate. Conclusion: This study showed that even though an area may be located within the same radius around the Kori Nuclear Power Plant, the distribution and levels of 137Cs concentration vary according to the change in time and space of weather conditions (the altitude of the atmospheric boundary layer, the horizontal and vertical distribution of the local winds, and the precipitation levels), the topography of the regions where 137Cs is dispersed, the emission rate of 137Cs, and the number of emitted particles.

Background: Boron Neutron Capture Therapy (BNCT) is a new radiation therapy. In BNCT, there exists some very critical problems that should be solved. One of the severest problems is that the treatment effect cannot be known during BNCT in real time. We are now developing a SPECT (single photon emission computed tomography) system (BNCT-SPECT), with a cadmium telluride (CdTe) semiconductor detector. BNCT-SPECT can obtain the BNCT treatment effect by measuring 478 keV gamma-rays emitted from the excited state of 7Li nucleus created by 10B(n,α) 7Li reaction. In the previous studies, we investigated the feasibility of the BNCTSPECT system. As a result, the S/N ratio did not meet the criterion of S/N > 1 because deterioration of the S/N ratio occurred caused by the influence of Compton scattering especially due to capture gamma-rays of hydrogen. Materials and Methods: We thus produced an arrayed detector with two CdTe crystals to test cross talk phenomenon and to examine an anti-coincidence detection possibility. For more precise analysis for the anti-coincidence detection, we designed and made a collimator having a similar performance to the real BNCT-SPECT. Results and discussion: We carried out experiments with the collimator to examine the effect of cross talk of scattering gamma-rays between CdTe elements more practically. As a result of measurement the coincidence events were successfully extracted. Conclusion: We are now planning to carry out evaluation of coincidence rate from the measurement and comparison of it with the numerical calculations.

Background: In case of radiation emergencies, internal exposure monitoring for the members of public will be required to confirm internal contamination of each individual. In-vivo monitoring technique using portable gamma spectrometer can be easily applied for internal exposure monitoring in the vicinity of the on-site area. Materials and Methods: In this study, minimum detectable doses (MDDs) for 134Cs, 137Cs, and 131I were calculated adjusting minimum detectable activities (MDAs) from 50 to 1,000 Bq to find out the optimal in-vivo counting condition. DCAL software was used to derive retention fraction of Cs and I isotopes in the whole body and thyroid, respectively. A minimum detectable level was determined to set committed effective dose of 0.1 mSv for emergency response. Results and Discussion: We found that MDDs at each MDA increased along with the elapsed time. 1,000 Bq for 134Cs and 137Cs, and 100 Bq for 131I were suggested as optimal MDAs to provide in-vivo monitoring service in case of radiation emergencies. Conclusion: In-vivo monitoring program for emergency response should be designed to achieve the optimal MDA suggested from the present work. We expect that a reduction of counting time compared with routine monitoring program can achieve the high throughput system in case of radiation emergencies.

Background: Evaluation of deoxyribonucleic acid (DNA)-strand break is important to elucidate the biological effect of ionizing radiations. The conventional methods for DNA-strand break evaluation have been achieved by Agarose gel electrophoresis and others using an electrical property of DNAs. Such kinds of DNA-strand break evaluation systems can estimate DNAstrand break, according to a molecular weight of DNAs. However, the conventional method needs pre-treatment of the sample and a relatively long period for analysis. They do not have enough sensitivity to detect the strand break products in the low-dose region. Materials and Methods: The sample is water, methanol and plasmid DNA solution. The plasmid DNA pUC118 was multiplied by using Escherichia coli JM109 competent cells. The resonance frequency and Q-value were measured by means of microwave dielectric absorption spectroscopy. When a sample is located at a center of the electric field, resonance curve of the frequency that existed as a standing wave is disturbed. As a result, the perturbation effect to perform a resonance with different frequency is adopted. Results and discussion: The resonance frequency shifted to higher frequency with an increase in a concentration of methanol as the model of the biological material, and the Q-value decreased. The absorption peak in microwave power spectrum of the double-strand break plasmid DNA shifted from the non-damaged plasmid DNA. Moreover, the sharpness of absorption peak changed resulting in change in Q-value. We confirmed that a resonance frequency shifted to higher frequency with an increase in concentration of the plasmid DNA. Conclusion: We developed a new technique for an evaluation of DNA damage. In this paper, we report the evaluation method of DNA damage using microwave dielectric absorption spectroscopy.

Background: Carbon ion therapy has achieved satisfactory results. However, patients have a risk to get a secondary cancer. In order to estimate the risk, it is essential to understand particle transportation and nuclear reactions in the patient’s body. The particle transport Monte Carlo simulation code is a useful tool to understand them. Since the code validation for heavy ion incident reactions is not enough, the experimental data of the elementary reaction processes are needed. Materials and Methods: We measured neutron production double-differential cross-sections (DDXs) on a carbon bombarded with 430 MeV/nucleon carbon beam at PH2 beam line of HIMAC facility in NIRS. Neutrons produced in the target were measured with NE213 liquid organic scintillators located at six angles of 15, 30, 45, 60, 75, and 90°. Results and Discussion: Neutron production double-differential cross-sections for carbon bombarded with 430 MeV/nucleon carbon ions were measured by the time-of-flight method with NE213 liquid organic scintillators at six angles of 15, 30, 45, 60, 75, and 90°. The cross sections were obtained from 1 MeV to several hundred MeV. The experimental data were compared with calculated results obtained by Monte Carlo simulation codes PHITS, Geant4, and FLUKA. Conclusion: PHITS was able to reproduce neutron production for elementary processes of carbon-carbon reaction precisely the best of three codes.

Background: Various characteristics such as size distribution, chemical component and radioactivity have been analyzed for radioactive aerosols released from Fukushima Daiichi Nuclear Power Plant. Measured results for radioactive aerosols suggest that the potential transport medium for radioactive cesium was non-sea-salt sulfate. This result indicates that cesium isotopes would preferentially attach with sulfate compounds. In the present work the attachment behavior of fission products to aqueous solution aerosols of sodium salts has been studied using a generation system of solution aerosols and spontaneous fission source of 248Cm. Materials and Methods: Attachment ratios of fission products to the solution aerosols were compared among the aerosols generated by different solutions of sodium salt. Results and Discussion: A significant difference according as a solute of solution aerosols was found in the attachment behavior. Conclusion: The present results suggest the existence of chemical effects in the attachment behavior of fission products to solution aerosols.

Background: To find out the leak characteristic of research reactor ‘HNANOR’ building in a typhoon condition Materials and Methods: MELCOR code which normally is used to simulate severe accident behavior in a nuclear power plant was used to simulate the leak rate of air and fission products from reactor hall after the shutdown of the ventilation system of HANARO reactor building. For the simulation, HANARO building was designed by MELCOR code and typhoon condition passed through Daejeon in 2012 was applied. Results and discussion: It was found that the leak rate is 0.1%·day-1 of air, 0.004%·day-1 of noble gas and 3.7 × 10-5%·day-1 of aerosol during typhoon passing. The air leak rate of 0.1%·day-1 can be converted into 1.36 m3·hr-1, but the design leak rate in HANARO safety analysis report was considered as 600 m3·hr-1 under the condition of 20 m·sec-1 wind speed outside of the building by typhoon. Conclusion: Most of fission products during the maximum hypothesis accident at HANARO reactor will be contained in the reactor hall, so the direct radiation by remained fission products in the reactor hall will be the most important factor in designing emergency preparedness for HANARO reactor.

Background: As new legislation has come into force implementing radiation safety management for the use of naturally occurring radioactive materials (NORM), it is necessary to establish a rapid and accurate measurement technique. Measurement of 238U and 232Th using conventional methods encounter the most significant difficulties for pretreatment (e.g. , purification, speciation, and dilution/enrichment) or require time-consuming processes. Therefore, in this study, the applicability of ED-XRF as a non-destructive and rapid screening method was validated for raw materials and by-product samples. Materials and Methods: A series of experiments was conducted to test the applicability for rapid screening of XRF measurement to determine activity of 238U and 232Th based on certified reference materials (e.g. , soil, rock, phosphorus rock, bauxite, zircon, and coal ash) and NORM samples commercially used in Korea. Statistical methods were used to compare the analytical results of ED-XRF to those of certified values of certified reference materials (CRM) and inductively coupled plasma mass spectrometry (ICP-MS). Results and Discussion: Results of the XRF measurement for 238U and 232Th showed under 20% relative error and standard deviation. The results of the U-test were statistically significant except for the case of U in coal fly ash samples. In addition, analytical results of 238U and 232Th in the raw material and by-product samples using XRF and the analytical results of those using ICP-MS (R2 ≥ 0.95) were consistent with each other. Thus, the analytical results rapidly derived using ED-XRF were fairly reliable. Conclusion: Based on the validation results, it can be concluded that the ED-XRF analysis may be applied to rapid screening of radioactivities (238U and 232Th) in NORM samples.

Background: Methodologies for a series of radiological consequence assessments show a distinctive difference according to the design principles of the original nuclear suppliers and their technical standards to be imposed. This is due to the uncertainties of the accidental source term, radionuclide behavior in the environment, and subsequent radiological dose. Both types of PWR and PHWR are operated in Korea. However, technical standards for evaluating atmospheric dispersion have been enacted based on the U.S. NRC’s positions regardless of the reactor types. For this reason, it might cause a controversy between the licensor and licensee of a nuclear power plant. Materials and Methods: It was modelled under the framework of the NRC Regulatory Guide 1.145 for light-water reactors, reflecting the features of heavy-water reactors as specified in the Canadian National Standard and the modelling features in MACCS2, such as atmospheric diffusion coefficient, ground deposition, surface roughness, radioactive plume depletion, and exposure from ground deposition. Results and Discussion: An integrated accident consequence assessment code, ACCESS (Accident Consequence Assessment Code for Evaluating Site Suitability), was developed by taking into account the unique regulatory positions for reactor types under the framework of the current Korean technical standards. Field tracer experiments and hand calculations have been carried out for validation and verification of the models. Conclusion: The modelling approaches of ACCESS and its features are introduced, and its applicative results for a hypothetical accidental scenario are comprehensively discussed. In an applicative study, the predicted results by the light-water reactor assessment model were higher than those by other models in terms of total doses.

Background: Experiments with small electrochemical apparatus were previously carried out for detecting low-energy neutrinos under irradiation of reactor neutrinos and under natural neutrino environment. The experimental result indicated that the output current of reactorneutrino irradiated detector was appreciably larger than that of natural environmental one. Usual interaction cross-sections of neutrinos are quite small, so that they do not explain the experimental result at all. Materials and Methods: To understand the experimental data, we propose that some biological products may generate AV-type scalar field B0, leading to a large interaction cross-section. The output current generation is ascribed to an electrochemical process that may be assisted by weak interaction phenomena. Dissolved oxygen concentrations in the detector solution were measured in this study, for the purpose of understanding the mechanism of the detector output current generation. Results and Discussion: It was found that the time evolution of experimental output current was mostly reproduced in simulation calculation on the basis of the measured dissolved oxygen concentration. Conclusion: We mostly explained the variation of experimental data by using the electrochemical half-cell analysis model based on the DO concentration that is consistent to the experiment.

Background: Radiation dose rates in PRIDE facility is evaluated quantitatively for assessing radiation safety of workers because of large amounts of depleted uranium being handled in PRIDE facility. Even if direct radiation from depleted uranium is very low and will not expose a worker to significant amounts of external radiation. Materials and Methods: ORIGEN-ARP code was used for calculating the neutron and gamma source term being generated from depleted uranium (DU), and the MCNP5 code was used for calculating the neutron and gamma fluxes and dose rates. Results and Discussion: The neutron and gamma fluxes and dose rates due to DU on spherical surface of 30 cm radius were calculated with the variation of DU mass and density. In this calculation, an imaginary case in which DU density is zero was added to check the self-shielding effect of DU. In this case, the DU sphere was modeled as a point. In case of DU mixed with molten salt of 50-250 g, the neutron and gamma fluxes were calculated respectively. It was found that the molten salt contents in DU had little effect on the neutron and the gamma fluxes. The neutron and the gamma fluxes, under the respective conditions of 1 and 5 kg mass of DU, and 5 and 19.1 g.cm-3 density of DU, were calculated with the molten salt (LiCl+KCl) of 50 g fixed, and compared with the source term. As the results, similar tendency was found in neutron and gamma fluxes with the variation of DU mass and density when compared with source spectra, except their magnitudes. Conclusion: In the case of the DU mass over 5 kg, the dose rate was shown to be higher than the environmental dose rate. From these results, it is concluded that if a worker would do an experiment with DU having over 5 kg of mass, the worker should be careful in order not to be exposed to the radiation.

Background: For a verification of newly-developed neutron absorbers, one of guidelines on the qualification and acceptance of neutron absorbers is the neutron attenuation test. However, this approach can cause a problem for the qualifications that it cannot distinguish how the neutron attenuates from materials. Materials and Methods: In this study, an estimation method of neutron absorption performances for materials is proposed to detect both direct penetration and back-scattering neutrons. For the verification of the proposed method, MCNP simulations with the experimental system designed in this study were pursued using the polyethylene, iron, normal glass and the vitrified form. Results and Discussion: The results show that it can easily test neutron absorption ability using single absorber model. Also, from simulation results of single absorber and double absorbers model, it is verified that the proposed method can evaluate not only the direct thermal neutrons passing through materials, but also the scattered neutrons reflected to the materials. Therefore, the neutron absorption performances can be accurately estimated using the proposed method comparing with the conventional neutron attenuation test. Conclusion: It is expected that the proposed method can contribute to increase the reliability of the performance of neutron absorbers.

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 tetrahedralmesh 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 organaveraged 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 Tet- Gen 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.

Background: Liquid scintillation counters (LSCs) are commonly used as an analytical method for detecting 222Rn 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 222Rn in groundwater. Materials and Methods: The Quantulus 1220 and 300SL were used to detect the presence of 222Rn. 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 222Rn calibration source prepared with a 226Ra 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 0.08 Bq·L-1, while that of the 300SL was 0.20 Bq·L-1. The analytical results regarding 222Rn 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 222Rn in groundwater. Both LSCs can be applied to determine the levels of 222Rn in groundwater under the management of the Ministry of Environment.

Background: Ascertainment bias are common in epidemiologic studies to assess the association between thyroid cancer risk and living near nuclear power plants because many thyroid cancers are diagnosed by chance through health examination. We surveyed the ultra sonography (USG) examination history and conducted thyroid and breast USG in residents living near nuclear power plants. Materials and Methods: The study population comprised 2,421 residents living near nuclear power plants in Korea. Information on demographic characteristics, including diagnostic examination history, was collected by interview using questionnaires. USG examination was conducted to evaluate the presence of thyroid nodules and breast lesion. Study participants were divided into 3 groups according to the distance of their respective villages from a nuclear power plant. The proportions of USG examination history and prevalence of thyroid nodules and breast lesions were compared between groups. Results and Discussion: Examination histories of thyroid USG were 23.1%, 13.7%, and 10.5% in men and 31.3%, 26.7%, 18.3% in women in the short, intermediate, and long distance groups, respectively. There were significant inverse associations between thyroid USG history and the distance from nuclear power plants (P for trend = 0.001 for men and 0.017 for women). However, there was no association between the distance of villages from nuclear power plants and prevalence of thyroid nodules. Conclusion: Our results suggest that there may be an ascertainment bias in population-based studies examining the harmful effects of NPPs examination and researchers should pay attention to ascertainment bias resulted from differential health examination. Correction for ascertainment bias, active follow-up and examination for all study population to remove differential health examination is needed.

Background: Airborne radioactive particulate in many important nuclear facilities (particularly nuclear power plants) will have a strong impact on the relative public dose if they are released into the corresponding environment traversing the stack or vents. The radiation protection researchers have regarded the relative environment assessing and estimation of public doses. And the model of assessing impact of discharges radioactive substance to the environment have been recommended by many international organizations (e.g. IAEA) with the nuclear energy safety and radiation protection. Materials and methods: This paper introduced the generic models that were suggested by International Atomic Energy Agency (IAEA), for use in assessing the impact of discharges of radioactive substances to the environment (e.g. IAEA Safety Report Series No.19). Results and Discussion: The writers of this paper, based on the recommend methods, assessed the discharge limits in some airborne radioactive substances discharging standards. The reasons that IAEA method are introduced are mainly the following considerations: IAEA is one of international organizations with some authorities in the nuclear energy safety and radiation protection; and, more important, the recommend modes are operational methods rather than the methods having little operations such as that have used by some researchers. Conclusion: It is wish that the introduced methods in this paper can be referenced in draft or revise of the standards related to discharges of radioactive substances to the environment.

Background: This work was to get information about 90Sr contamination of the environment by using soil and moss from selected areas in Jeju Island, Korea. Materials and Methods: The activities of 90Sr in soil and moss samples were investigated at nine locations of Jeju island, Korea. The soil samples have been collected at 4 sites of Jeju island during June to August of 2013, analyzed for vertical distribution of 90Sr activities. The moss samples have been collected at 5 sites of Jeju island during November of 2011 to June of 2012, and analyzed for radioactive 90Sr. Results and Discussion: The 90Sr vertical concentrations in the investigated soil samples were 2.77 to 18.24 Bq·kg-1 in eastern part, 1.69 to 18.27 Bq·kg-1 in northern part, 3.76 to 13.46 Bq·kg-1 in the western part and 1.09 to 8.70 Bq·kg-1 in the southern part of the Mt. Halla in Jeju island, respectively. Activities of 90Sr show the highest value at the surface soil and decrease with depth. The activity concentration measured was in the range of 79.6 to 363 Bq·kg-1 -dry moss. Conclusion: This material is expected to be basis reference for survey of environmental radioactivity in Jeju Island.

Background: The International Commission on Radiological Protection (ICRP) recommendations and the Federal Guidance Report (FGR) published by the U.S. Environmental Protection Agency (EPA) have been widely applied worldwide in the fields of radiation protection and dose assessment. The dose conversion coefficients of the ICRP and FGR are widely used for assessing exposure doses. However, before the coefficients are used, the user must thoroughly understand the derivation process of the coefficients to ensure that they are used appropriately in the evaluation. Materials and Methods: The ICRP provides recommendations to regulatory and advisory agencies, mainly in the form of guidance on the fundamental principles on which appropriate radiological protection can be based. The FGR provides federal and state agencies with technical information to assist their implementation of radiation protection programs for the U.S. population. The system of radiation dose assessment and dose conversion coefficients in the ICRP and FGR is reviewed in this study. Results and Discussion: A thorough understanding of their background is essential for the proper use of dose conversion coefficients. The FGR dose assessment system was strongly influenced by the ICRP and the U.S. National Council on Radiation Protection and Measurements (NCRP), and is hence consistent with those recommendations. Moreover, the ICRP and FGR both used the scientific data reported by Biological Effects of Ionizing Radiation (BEIR) and United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) as their primary source of information. The difference between the ICRP and FGR lies in the fact that the ICRP utilized information regarding a population of diverse races, whereas the FGR utilized data on the American population, as its goal was to provide guidelines for radiological protection in the US. Conclusion: The contents of this study are expected to be utilized as basic research material in the areas of radiation protection and dose assessment.

Background: Although nuclear disaster is considered rare, its effects are serious, and we must prepare a system to enable an effective response. Materials and Methods: Since 2010, we have been offering a two-day seminar to provide current nurses and radiological technologists with basic knowledge and train them in radiation emergency medicine (REM) techniques. This training offers lectures to deepen each specialty from the perspective of REM, as well as exercises on ways to handle irradiated and/or contaminated patients. Participants were expected to treat patients according to the concept of REM. Results and Discussion: All participants learn to assess and decontaminate contaminated wounds through drills. The questionnaire survey for participants indicated that participants were satisfied with this training and wanted to attend again. Conclusion: We believe that this training course will provide a valuable opportunity for medical professionals to gain knowledge and expertise in REM.