• Journal of Radiation Protection and Research
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• v.28 no.3
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• pp.215-223
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• 2003

Against major release of radioactive material in nuclear power plant, Emergency Planning Zone(EPZ)s are typically established around nuclear power plants to effectively perform the public protective measures. The domestic methodology to determine the size of the EPZ is similar to that of Japan established in 1980, where calculations were based on the conservative accident source term. The objective of this study is to re-evaluate the validity of established EPZ, the area within the radius of $8{\sim}10km$ around domestic nuclear power plants, using the source terms covering full spectrum of accidents obtained from PSA study of ULJIN 3&4. To evaluate the risks of health effects, the computer code MACCS2(MELCOR Accident Consequence Code System2) was used. The result shows that the existing EPZ can reduce the probability of early fatality adequately for most of the source term categories(STCs) except for STC-14 and STC-19. In case of STC-14 and 19, the evacuation distance of 16km and 13km, respectively, are required. These distances can be reduced by improving emergency preparedness since the sensitivity studies for the public protective actions show that the magnitude of early fatality is largely affected by the time delays in notification and evacuation.

• Journal of Radiation Protection and Research
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• v.43 no.1
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• pp.29-38
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• 2018

Background: Radon contributing about 42% of annual average dose, mainly comes from soil. In this paper, standard measurement procedures for soil radon exhalation rate are suggested and their measurement uncertainties are analyzed. Materials and Methods: We used accumulation method for estimating surface exhalation rate. The closed-loop measurement system was made up with a RAD7 detector and a surface chamber. Radon activity concentrations in the system were observed as a function of time, with data collection of 5 and 15-minute and the measurement time of 4 hours. Linear and exponential fittings were used to obtain radon exhalation rates from observed data. Standard deviations of measurement uncertainties for two approaches were estimated using usual propagation rules. Results and Discussion: The exhalation rates (E) from linear approach, with 30 minutes measurement time were $44.8-48.6mBq{\cdot}m^{-2} {\cdot}s^{-1}$ or $2.14-2.32atom{\cdot}cm^{-2}{\cdot}s^{-1}$ with relative measurement uncertainty of about 10%. The contributions of fitting parameter A, volume (V) and surface (S) to the estimated measurement uncertainty of E were 59.8%, 30.1% and 10.1%, in average respectively. In exponential fitting, at 3-hour measurement we had E ranged of $51.6-69.2mBq{\cdot}m^{-2} {\cdot}s^{-1}$ or $2.46-3.30atom{\cdot}cm^{-2}{\cdot}s^{-1}$ with about 15% relative uncertainty. Fitting with 4-hour measurement resulted E about $51.3-68.2mBq{\cdot}m^{-2} {\cdot}s^{-1}$ or $2.45-3.25atom{\cdot}cm^{-2}{\cdot}s^{-1}$ with 10% relative uncertainty. The uncertainty contributions in exponential approach were 75.1%, 13.4%, 8.7%, and 2.9% for total decay constant k, fitting parameter B, V, and S, respectively. Conclusion: In obtaining exhalation rates, the linear approach is easy to apply, but by saturation feature of radon concentrations, the slope tends to decrease away from the expected slope for extended measurement time. For linear approach, measurement time of 1-hour or less was suggested. For exponential approach, the obtained exhalation rates showed similar values for any measurement time, but measurement time of 3-hour or more was suggested for about 10% relative uncertainty.

• 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.22 no.2
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• pp.97-101
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• 1997

During the period of March 1995 to December 1995, tritium concentrations of tritiated water vapor (HTO) and tritiated hydrogen (HT) in the atmosphere in Taejon were measured to evaluate present background levels of tritium in the atmosphere. Air samples were collected continuously for three weeks with a sampling system for tritium in the atmosphere and were analyzed by a liquid scintillation counting system. The range of the atmospheric HTO concentrations was 3.2-36 mBq $m^{-3}$ with a mean value of 16.2 mBq $m^{-3}$. The atmospheric HTO concentrations were the highest in summer and the lowest in winter. This trend was similar to the variation of atmospheric absolute humidity. The specific activities of tritium in atmospheric water vapor in Taejon ranged from 0.62 Bq $L^{-1}$ to 3.82 Bq $L^{-1}$ with a mean value of 2.04 Bq $L^{-1}$. The atmospheric HT concentrations were in the range of 35.7 mBq $m^{-3}$ to 48.9 mBq $m^{-3}$ with a mean value of 41.1 mBq $m^{-3}$.

• Journal of Radiation Protection and Research
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• v.22 no.4
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• pp.309-317
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• 1997

The measured radioactivity of gamma-emitting radionuclides in each radioactive waste drum using the non-destructive waste assay method is underestimated than real radioactivity in radioactive waste drum because the gamma-rays are attenuated within the medium. Therefore, the measured radioactivity should be corrected for the attenuation of gamma-rays. For the correction of the attenuation of gamma-rays, the attenuation correction method should be applied differently by considering the distribution and density of medium in radioactive wastes drum generated from nuclear power plants. In this study, the model drums were fabricated for simulating five types of radioactive waste drums generated from nuclear power plant and the optimum methods of the attenuation correction were experimentally determined to analyze the activity of radionuclides in the waste drum accurately using the segmented gamma scanning system. With the determination of the attenuation correction methods from the experimental results the transmission method and the average density method for the miscellaneous waste drum, the transmission method and the differential peak absorption method for the shielded miscellaneous waste drum were used to measure the density of medium in waste drums. Also, the average density method and the differential peak absorption method for the spent resin drum, the paraffin solidified drum, and the spent filter drum were used.

• Journal of Radiation Protection and Research
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• v.21 no.4
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• pp.313-327
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• 1996

As an at-reactor(AR) storage method o( spent fuel, there are horizontal concrete module type, metal storage cask type, concrete storage cask type, dual purpose (transportation and storage) cask type and multi-purpose (transportation, storage and disposal) cask type. All other types except multi-purpose one have been already used for AR dry storage of spent fuels after obtaining operation license in various foreign countries. Also the development of multi-purpose type has been continued for operation license. In America, Japan, Germany, Canada, Spain, Switzerland, and Czech Republic, etc., AR dry storage facilities are under operation or on propulsion, and spent fuels are transported to interim storage facility or reprocessing plant after dry storage at reactor temporarily. At Wolsung site, in case of Korea, concrete silo type has already been introduced, and it is believed to be inevitable to store spent fuels at reactor temporarily, considering the reality that storage capacity of spent fuel is approaching to the limit in some nuclear power plants. In this report, the system characteristics, design requirements, technical standards and status of AR storage system, which is suitable for domestic site such as Kori, have been studied. In most cases, the licensed period of storage cask is limited up to 20 years and the integrity of material and maintenance of leaktightness are required during the whole service life.

• Journal of Radiation Protection and Research
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• v.13 no.1
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• pp.8-20
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• 1988

The annual individual and collective doses to the thyroids of four age-dependent groups due to the in-take of I-131 released from the Younggwang nuclear power plant NU-1 & 2, Korea, are estimated using the model presented in ICRP 29. Sensitivity and robustness of the model are analyzed. In case of 0.12% fuel defect during normal operation, the collective dose is founded to be 3.05${\times}10^{-3}$man-thyroid-Sv, which is higher than the value calculated by the GASPAR code, 2.3${\times}10^{-3}$man-thyroid-Sv. The maximal individual annual doses resulting from an acute release are higher than those calculated under the assumption of continuous release by $1.4{\sim}1.7$ times. The most important pathway to the infant is milk and, in contrast, that to child, teen and adult is ingestion of crops. The model used is the calculation appears to be influenced by the variables such as roubstness-index. The weighted committed dose equivalent obtained by the ICRP 29 model is slightly higher than that calculated by the three-compartment model.

• Journal of Radiation Protection and Research
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• v.16 no.2
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• pp.41-54
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• 1991

This paper aimed to analyse dose sensitivity to the controllable parameters of indoor radon $(^{222}Rn)$ and its decay products (Rn-D) by applying the input~output linear system theory. Physical behaviors of $^{222}Rn\;&\;Rn-D$ were analyzed in terms of $(^{222}Rn)$ gas -generation, -migation and -infiltration to indoor environments, and the performance output-function, i. e. mean dose equivalent to Tracho-Bronchial (TB) lung region, was assessed to the following extented ranges of the controllable paramenters; a) the ventilation rate $constant({\lambda}_v)\;:\;0{\sim}50[h^{-l}].\;b)$ the attachment rate $constant({\lambda}_a)\;:\;0{\sim}500[h^{-l}].\;c)$ the unattached-deposition rate constant (${\lambda}^u_d)\;:\;0-50[h-l]$. A linear input-output model was reconstructed from the original models in literatures, as follows, which was modified into the matrices consisting of 111 nodal equations; a) indoor $^{222}Rn\;&\;Rn-D$ Behaviour; Jacobi-Porstendoerfer-Bruno model.

• Journal of the Korean Society of Radiology
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• v.13 no.4
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• pp.653-659
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• 2019

The purpose of this study is to measure the (air dose rate of radiation dose) the discharged patient who was administrated high dose $^{131}I$ treatment, and to predict exposure radiation dose in public person. The dosimetric evaluation was performed according to the distance and angle using three copper rings in 30 patients who were treated with over 200mCi high dose Iodine therapy. The two observer were measured using a GM surverymeter with 8 point azimuth angle and three difference distance 50, 100, 150cm for precise radion dose measurement. We set up three predictive simulations to calculate the exposure dose based on this data. The most highest radiation dose rate was showed measuring angle $0^{\circ}$ at the height of 1m. The each distance average dose rate was used the azimuth angle average value of radiation dose rate. The maximum values of the external radiation dose rate depending on the distance were $214{\pm}16.5$, $59{\pm}9.1$ and $38{\pm}5.8{\mu}Sv/h$ at 50, 100, 150cm, respectively. If high dose Iodine treatment patient moves 5 hours using public transportation, an unspecified person in a side seat at 50cm is exposed 1.14 mSv radiation dose. A person who cares for 4days at a distance of 1 meter from a patient wearing a urine bag receives a maximum radiation dose of 6.5mSv. The maximum dose of radiation that a guardian can receive is 1.08mSv at a distance of 1.5m for 7days. The annual radiation dose limit is exceeded in a short time when applied the our developed radiation dose predictive modeling on the general public person who was around the patients with Iodine therapy. This study can be helpful in suggesting a reasonable guideline of the general public person protection system after discharge of high dose Iodine administered patients.

• Journal of Radiation Protection and Research
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• v.10 no.2
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• pp.96-108
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• 1985

A study for the assessment of natural environmental radiation exposure at a flat and open field of about $10,000m^2$ in area in CNU Daeduk campus has been carried out by means of gamma-ray scintillation spectrometry and thermoluminescence dosimetry for one year period of time from October 1984. The detectors used were 3'${\phi}{\times}$3' NaI(T1) and two different types of LiF TLD, namely, chip sealed in plastic sheet which tightly pressed on two open holes of a metal plate and Teflon disk. Three 24-hour cycles of in-situ spectrometry, and two 3-month and one 1-month cycles of field TL dosimetry were performed. All the spectra measured were converted into exposure rate by means of G(E) opertaion, and therefrom exposure rate due to terrestrial component of environmental radiation was figured out. Exposure rate determined by the spectrometry was, on average, $(10.54{\pm}2.96){\mu}R/hr$, and the rates of $(12.0{\pm}3.4){\mu}R/hr$ and $(11.0{\pm}3.6){\mu}R/hr$ were obtained from chip and disk TLD, respectively. Fluctuations in diurnal variation of the exposure rate measured by the spectrometry were noticeable sometime even in a single cycle of 24 hours. It is concluded that appropriately combined use of TLD with iu-sitn gamma-ray spectrometry system can give more accurate and precise measure of environmental radiation exposure, and further study for more adequate and sensitive TLD for environmental dosimetry, including improvement and elevation of accuracy in data assessment through inter-laboratory or international intercomparison is necessary.