• Progress in Medical Physics
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• v.26 no.4
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• pp.280-285
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• 2015

In radiation therapy fields, a brachytherapy is a treatment that kills lesion of cells by inserting a radioisotope that keeps emitting radiation into the body. We currently verify the consistency of radiation treatment plan and dose distribution through film/screen system (F/S system), provide therapy after checking dose. When we check dose distribution, F/S systems have radiation signal distortion because there is low resolution by penumbra depending on the condition of film developed. In this study, We fabricated a $HgI_2$ Semiconductor radiation sensor for base study in order that we verify the real dose distribution weather it's same as plans or not in brachytherapy. Also, we attempt to evaluate the feasibility of QA system by utilizing and evaluating the sensor to brachytherapy source. As shown in the result of detected signal with various source-to-detector distance (SDD), we quantitatively verified the real range of treatment which is also equivalent to treatment plans because only the low signal estimated as scatters was measured beyond the range of treatment. And the result of experiment that we access reproducibility on the same condition of ${\gamma}$-ray, we have made sure that the CV (coefficient of variation) is within 1.5 percent so we consider that the $HgI_2$ sensor is available at QA of brachytherapy based on the result.

• Korean Journal of Digital Imaging in Medicine
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• v.14 no.2
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• pp.75-81
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• 2012

The purpose of this study was administered to the body for examination and treatment, high-energy radioactive isotope(F-18, I-131) in vitro discharge experiments. Increasing exposed dose of radiation to health professionals is caused by the increase of PET/CT use and a radioactive isotope. Therefore, the high-energy isotope F-18 and I-131 after administration about using Metabolite excretion was studied. As a results of this study, patients had plenty of fluids for testing and treatment alone administered radiopharmaceuticals can be more than twice as fast excretion induced emissions. Also was able to get a better image space to reduce the dose rate.

• Journal of Radiation Protection and Research
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• v.9 no.2
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• pp.76-89
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• 1984

Weighted committed dose equivalents($W_T\;H_{50}$) per intake of unit activity of four nuclides-I-131, I-133, Cs-134 and Cs-137-, which was based on the concepts of ICRP Pub. 30, are calculated for adult who is 70 kg and 25 years old and, for infant who is 10 kg and 1 year old. Metabolism of iodine taken through oral or inhalation pathway is described by using the three-compartment model which consists of inorganic, thyroid and organic compartment. After intake, the amount of iodine in every compartment is calculated by solving the transfer equations among the these compartments. As soon as caesium is taken into the body, it is distributed uniformly in the body through the transfer compartment. In this case, the amount of caesium in total body is calculated by using the total body compartment model which is divided into two tissue compartments because of their different biological half-lifes of caesium in body. As a result of calculations, whether oral or inhalation pathway, the values of ($W_T\;H_{50}$) per intake of unit activity of I-131 for infants are about ten times as much as those of adults. On the other hand, for Cs-134 and Cs-137, the values of $W_T\;H_{50}$ per intake of unit activity show that, whether adults of infants, they have almost the same values.

• Journal of the Korean Society of Radiology
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• v.11 no.7
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• pp.597-603
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• 2017

High-dose $^{131}I$ therapy has been generally carried out to remove remaining thyroid tissue or to cure metastasize lesion of patients who received full thyroidectomy due to differentiated thyroid cancers. In case high-dose $^{131}I$ therapy is carried out for a patient, the patient should be hospitalized being isolated for a certain period in order to restrict the amount of exposure to radiation of people at large from the patient within the limit of a level of radiation. Effective half-life is an important value to calculate how family members are exposed to radiation from a patient or to decide the period of isolation of the patient from the family members. Therefore, in this study we calculated the effective decay constant, effective half-life and period of isolation of high-dose $^{131}I$ therapy patient using NM670 SPECT/CT. As a result of carrying out this study, the effective half-life of high-dose $^{131}I$ therapy patients was derived and the time to reach the discharge level of 1.2 GBq was confirmed. When they were compared with each other in each of curative doses, the effective half-life did not have significant difference, but the time when the level of radiation remaining in the interior of the body to reach the criteria of isolation and discharge showed significant difference and it could be confirmed that the higher the curative dose the longer the period of isolation becomes. When the effective half-lives in each type of preparation were compared with each other, they did not show significant difference. However, When the times to reach the level of radiation that is the criteria of isolation and discharge in each type of preparations, they showed significant difference. The cause of the shortening of the isolation period for rhTSH patients group is decided to be low curative dose. Accordingly, if the current national health insurance (the insurance is applied to using of rhTSH in 3.7 GBq or lower) is maintained, while discerning them in each of types of preparation, we would be able to discharge patients at the time earlier than the current period of isolation (2 nights and 3 days).

• Journal of Radiation Protection and Research
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• v.29 no.4
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• pp.213-219
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• 2004

Soybean plants were exposed to HTO vapor in an exposure box for 1 hour at different growth stages. Relative concentrations of TFWT at the end of exposure (percent ratios of TFWT concentrations to mean HTO concentrations in air moisture in the box during exposure) decreased on the whole in the order of leaf > shell > seed > stem with the highest values of 40.2% and 6.4% for leaf and stem, respectively. TFWT concentrations reduced by factors of several thousands to several hundred-thousands from the end of exposure till the harvest. The reduction factor decreased in the order of leaf > shell > seed > stem. Relative OBT concentrations at harvest (ratios of the OBT concentration in the dry plant part at harvest to the initial leaf TFWT concentration, ml $g^{-1}$) were in the range of $2.2{\times}10^{-5}{\sim}9.5{\times}10^{-3}$ for seeds being the highest when the exposure was performed at the actively seed-developing stage. The exposure time-dependent variation in the OBT concentration was much greater in seeds and shells than in leaves and stems. It was indicated that OBT would contribute to almost all the radiation dose due to the consumption of soybean seeds in most cases after an acute exposure of growing plants to HTO vapor. Present results are applicable to establishing and validating soybean $^3H$ models for an acute accidental release of HTO.

• Journal of Radiation Protection and Research
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• v.25 no.2
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• pp.75-80
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• 2000

The predicted results of INDAC code were compared with measured $^3H$ concentrations in air and pine-needle around the Wolsung site. The optimal sets of input data to INDAC were in addition selected by comparing the measured values with the predicted values of INDAC based on various conditions such as the release modes of effluents into the environment, the classification of wind classes, and the consideration of terrain. The predicted $^3H$ concentrations in air and pine-needle were shown to have good agreement with measured values, although there are some limitations such as uncertainties in measured values, complex topology around the site, and the land-sea breeze effects. The assumption on the $^3H$ behavior in vegetables or plants that the ratio of $^3H$ concentration in plant water to $^3H$ concentration in atmospheric water is 1/2 was shown to be conservative in terms of the audit calculation performed by the regulator. It was also found that data sets based on mixed mode and no terrain data were not appropriate for the audit calculation ensuring the compliance with regulations. Thus, if the mixed mode is considered as the release mode of effluents into the environment, meteorological data measured at 58 m height and terrain data should be used to evaluate the atmospheric dispersion factor.

• Journal of Radiation Protection and Research
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• v.34 no.2
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• pp.87-94
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• 2009

Tritium is the one of the dominant contributors to the internal radiation exposure of workers at pressurized heavy water reactors (PHWRs). This nuclide is likely to release to work places as tritiated water vapor (HTO) from a nuclear reactor and gets relatively easily into the body of workers by inhalation. Inhaled tritium usually reaches the equilibrium of concentration after approximately 2 hours inside the body and then is excreted from the body with a half-life of 10 days. Because tritium inside the body transports with body fluids, a whole body receives radiation exposure. Internal radiation exposure at PHWRs accounts for approximately 20-40% of total radiation exposure; most internal radiation exposure is attributed to tritium. Thus, tritium is an important nuclide to be necessarily monitored for the radiation management safety. In this paper, metabolism for tritium is established using its excretion rate results in urine samples of workers at PHWRs and an effective half-life, a key parameter to estimate the radiation exposure, was derived from these results. As a result, it was found that the effective half-life for workers at Korean nuclear power plants is shorter than that of International Commission on Radiological Protection guides, a half-life of 10 days.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.473-478
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• 2018

Radioactive aerosol generated in cutting and melting work during the NPP decommissioning process can cause internal exposure to body through workers' breath. Thus, it is necessary to assess worker internal exposure due to the radioactive aerosol during decommissioning. The actually measured value of the working environment is needed for accurate assessment of internal exposure, but if it is difficult to actually measure that value, the internal exposure dose can be estimated through recommended values such as the fraction of amount of intake and the size of particles suggested by the International Committee on Radiological Protection (ICRP). As for the selection of particle size, this study applied a value of $5{\mu}m$, which is the size of particles considering the worker recommended by the ICRP. As for the amount of generation, the amount of intake was estimated using data on the mass of aerosol generated in a melting facility at a site in Kozloduy, Bulgaria. In addition, using these data, this study calculated the level of radioactivity in the worker's body and stool and conducted an assessment of internal exposure using the BiDAS computer code. The internal exposure dose of Type M was 0.0341 mSv, that of Type S was 0.0909 mSv. The two types of absorption showed levels that were 0.17% and 0.45% of the domestic annual dose limit, respectively.

• Journal of Radiation Protection and Research
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• v.1 no.1
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• pp.22-30
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• 1976

The perturbation of dose distribution adjacent to cavities in high energy electron has shown that the percentage of dose increase varies markedly as a function of the build-up layer, the length and thickness of the cavities, and the electron energy. The dose distribution showed that cavities similar in size to those encountered in the head and neck measured by industrial film dosimetry and corrected by ionization chambers. The most increased doses by measuring are resulted in a localized dose of up to 130% of that measured at the depth of maximum dose within a homogeneous tissue equivalent phantom. The measured values and correction factors of dose perturbation due to air cavities showed in diagrams and would be summarized as follows. 1. In $8{\sim}12MeV$ electron beams, the most marked dose is observed when the build-up layer thickness is 0.5cm and cavity volume is $2{\times}2{\times}2cm^3$. 2. The highest dose point is located under cavity when the energy is increased and cavity length is longer. 3. The cavity length at which the maximum percentage dose occurs decreases with increasing energy. 4. The highest percentage cavity doses are obtained when the energy is high, the build-up layer is thin, the thickness of the cavity is large, and the length of the cavity is approximately 1 to 3cm. 5. The doses of upper portion of cavity are less than the standard dose distribution as 5 to 10%. 6. The maximum range of electron beam are extended as much as thickness of cavity. 7. A cavity having a length of 5cm closely approximates a cavity of infinite length.

• Journal of Radiation Protection and Research
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• v.39 no.1
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• pp.1-6
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• 2014

A whole body counter (WBC) is used in nuclear power plants (NPP) to identify and measure internal radioactivity of workers who is likely to ingest or inhale radionuclides. WBC has several counting geometry, i.e. the thyroid, lung, whole body and gastrointestinal tract, considered with the location where radionuclides are deposited in the body. But only whole body geometry is used to detect internal radioactivity during whole body counting at NPPs. It is overestimated internal exposure dose because this measured values are indicated as the most conservative radioactivity values among the them of others geometry. In this study, experiments to measure radioactivity depending on the counting geometry of WBC were carried out using a WBC, a phantom, and standard radiation sources in order to improve overestimated internal exposure dose. Quantitative criteria, could be selected counting geometry according to ratio of count rates of the upper and lower detectors of the WBC, are provided through statistical analysis method.