• Journal of Radiation Protection and Research
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• v.19 no.1
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• pp.69-80
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• 1994

The TCMI(Three-Compartment Model for iodine) computer code has been developed, which is based on the three-compartment model and the respiratory model recommended in ICRP publication 54. This code is able to evaluate the thyroid burden, dose equivalent, committed dose equivalent and urinary excretion rate as time-dependent functions from the input data: working time and the radioiodine concentration in air. Using the TCMI code, the time-dependent thyroid burdens, the thyroid doses and the urinary excretion rates were calculated for three specific exposure patterns : acute, chronic and periodic. Applicability as an internal dose evaluation method has been assessed by comparing the results with some operational experiences. Simple equations and tables are provided to be used in the evaluation of the thyroid burden and the resulting doses for given I-131 concentration in air and the working time.

• Journal of Radiation Protection and Research
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• v.20 no.1
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• pp.37-44
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• 1995

Since provisions on the technical criteria for personnel dosimetry was amended three years ago, several improvements in the technique of monitoring personnel doses by TLD have taken place, but for the photograpfic film as a personnel monitor, additional investigations should be carried out for its accuracy of dose estimates because of its wide use in the radiation involved industries. So, this paper describes the methods to develope dose evaluation algorithm for photographic film using ISO reference radiations by i) empirical formula, ii) degree-of-fit method, and iii) matrix approximation. These methods show a good agreement between irradiated and calculated dose within tolerance level represented in ANSI N13. 11, and can be used for the dose evaluation of X, ${\gamma}$ and/or radiation fields.

• 대한방사선방어학회:학술대회논문집
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• 2009.11a
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• pp.220-221
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• 2009

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2
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• pp.157-167
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• 2020

Removing radioactive concrete is crucial in the decommissioning of nuclear power plants. However, this process generates radioactive aerosols, exposing workers to radiation. Although large amounts of radioactive concrete are generated during decommissioning, studies on the internal exposure of workers to radioactive aerosols generated from the cutting of radioactive concrete are very limited. In this study, therefore, we calculate the internal radiation doses of workers exposed to radioactive aerosols during activities such as drilling and cutting of radioactive concrete, using previous research data. The electrical-mobility-equivalent diameter measured in a previous study was converted to aerodynamic diameter using the Newton-Raphson method. Furthermore, the specific activity of each nuclide in radioactive concrete 10 years after nuclear power plants are shut down was calculated using the ORIGEN code. Eventually, we calculated the committed effective dose for each nuclide using the IMBA software. The maximum effective dose of ¹⁵²Eu constituted 83.09% of the total dose; moreover, the five highest-ranked elements (¹⁵²Eu, ¹⁵⁴Eu, ⁶⁰Co, ²³⁹Pu, ⁵⁵Fe) constituted 99.63%. Therefore, we postulate that these major elements could be measured first for rapid radiation exposure management of workers involved in decommissioning of nuclear power plants, even if all radioactive elements in concrete are not considered.

• Journal of Radiation Protection and Research
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• v.24 no.4
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• pp.215-223
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• 1999

Characteristics of radiation field in the steam generator(S/G) water chamber of a PWR were investigated and the anticipated effective dose rates to the worker in the S/G chamber were evaluated by Monte Carlo simulation. The results of crud analysis in the S/G of the Kori nuclear power plant unit 1 were adopted for the source term. The MCNP4A code was used with the MIRD type anthropomorphic sex-specific mathematical phantoms for the calculation of effective doses. The radiation field intensity is dominated by downward rays, from the U-tube region, having approximate cosine distribution with respect to the polar angle. The effective dose rates to adults of nominal body size and of small body size(The phantom for a 15 year-old person was applied for this purpose) appeared to be 36.22 and 37.06 $mSvh^{-1}$) respectively, which implies that the body size effect is negligible. Meanwhile, the equivalent dose rates at three representative positions corresponding to head, chest and lower abdomen of the phantom, calculated using the estimated exposure rates, the energy spectrum and the conversion coefficients given in ICRU47, were 118, 71 and 57 $mSvh^{-1}$, respectively. This implies that the deep dose equivalent or the effective dose obtained from the personal dosimeter reading would be the over-estimate the effective dose by about two times. This justifies, with possible under- or over- response of the dosimeters to radiation of slant incidence, necessity of very careful planning and interpretation for the dosimetry of workers exposed to a non-regular radiation field of high intensity.

• 대한방사선방어학회:학술대회논문집
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• 2011.11a
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• pp.234-235
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• 2011

• 대한방사선방어학회:학술대회논문집
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• 2009.11a
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• pp.88-89
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• 2009

• Journal of Radiation Protection and Research
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• v.39 no.2
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• pp.89-95
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• 2014

In potential accident consequence assessments for the licensing approval of LWRs, the ground deposition of radionuclides released into the environment is not allowed into the models, as recommended in the U. S. Nuclear Regulatory Commission's regulatory guide. Meanwhile, it is allowed into the assessment models for the licensing approval of PHWRs with consideration of more detailed physical processes of radionuclides in the atmosphere. Under these backgrounds, importance of exposure dose by ground deposition was quantitatively evaluated and comprehensively discussed. For potential accidental releases of $^{137}Cs$ and $^{131}I$, total exposure doses were more conservative in case of without consideration of ground deposition than in case of with its consideration. It was because of that the depletion of air concentration resulting from ground deposition is more influential in the contribution to total exposure doses than additional doses from contaminated ground. The exposure doses by the inhalation of contaminated air showed the contribution of more than 90% in total exposure doses, depending on atmospheric stability, release period of radionuclides and distance from a release point. The exposure doses from contaminated ground showed less than 10% at most in contribution of total exposure doses. The ratios of total exposure doses in case of with consideration of deposition to without its consideration for $^{131}I$ were distinct than those for $^{137}Cs$. As the atmosphere is more stable, release duration of radionuclides is longer, distance from a release point is longer, it was more distinct.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.497-502
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• 2003

Landfill is an widely applied alternative for clearance of non-contaminated waste or slightly contaminated waste generated from nuclear facilities. In this study, exposure dose is estimated for a worker and a resident at the landfill area in Young-gwang nuclear power plant. Based on evaluated dose, clearance concentrations of each radionuclide are determinated for dose criteria of 10 $\muSv/y$. The results of age-dependent dose are 1.02 $\muSv$ per year for resident and 0.471 $\muSv$ per year for worker. Clearance concentrations for each radionuclide are evaluated from $1.33{\times}10_{-1}$ Bq per gram to $2.85{\times}10^2$ Bq per gram.

• Journal of radiological science and technology
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• v.33 no.2
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• pp.149-154
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• 2010

Exposed doses to the patient's caregiver and their house due to the 131I from patients discharged from the hospital were measured using OSL dosimeters. Usually, 3.37-5.55 GBq (100-150 mCi) of $^{131}I$ administrated patients are discharged from the hospital after 3 or 4 days of hospitalization in Korea. In addition, after 5 to 8 days, the accumulated doses of the patient's caregiver and house after hospitalization of the patient were measured using OSL dosimeters. The results of the measured average accumulated doses were 0.1 mSv, which is 10% of 1 mSv, the public dose limit in the Korean Atomic Energy Law. And it's standard deviation was 0.087 mSv. Based on the results of this study, we anticipate that we could assure the compliance of the regulation requirement 5 mSv of MEST (Ministry of Education, Science and Technology) Notice No. 2008-45 for the patient's caregiver or family, even if we reduce the 3-4 days of hospitalization to 1-2 days or less.