• Journal of Radiation Protection and Research
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• v.47 no.2
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• pp.77-85
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• 2022

Background: After the Fukushima Daiichi Nuclear Power Station (FDNPS) accident, a model was developed to estimate the external exposure doses for residents who were expected to return to their homes after evacuation orders were lifted. However, the model's accuracy and uncertainties in parameters used to estimate external doses have not been evaluated. Materials and Methods: The model estimates effective doses based on the integrated ambient dose equivalent (H^*(10)) and life patterns, considering a dose reduction factor to estimate the indoor H^*(10) and a conversion factor from H^*(10) to the effective dose. Because personal dose equivalent (H_p(10)) has been reported to agree well with the effective dose after the FDNPS accident, this study validates the model's accuracy by comparing the estimated effective doses with H_p(10). The H_p(10) and life pattern data were collected for 36 adult participants who lived or worked near the FDNPS in 2019. Results and Discussion: The estimated effective doses correlated significantly with H_p(10); however, the estimated effective doses were lower than H_p(10) for indoor sites. A comparison with the measured indoor H^*(10) showed that the estimated indoor H^*(10) was not underestimated. However, the H_p(10) to H^*(10) ratio indoors, which corresponds to the practical conversion factor from H^*(10) to the effective dose, was significantly larger than the same ratio outdoors, meaning that the conversion factor of 0.6 is not appropriate for indoors due to the changes in irradiation geometry and gamma spectra. This could have led to a lower effective dose than H_p(10). Conclusion: The estimated effective doses correlated significantly with H_p(10), demonstrating the model's applicability for effective dose estimation. However, the lower value of the effective dose indoors could be because the conversion factor did not reflect the actual environment.

• Nuclear Engineering and Technology
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• v.45 no.1
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• pp.1-12
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• 2013

This paper describes the radiation doses to human and non-human biota in the Republic of Korea, as a result of the Fukushima nuclear accident. By using the measured airborne activity and ground deposition, the effective and thyroid doses of five human age groups (infant, 5 years, 10 years, 15 years and adult) were estimated by the ECOSYS code, and the whole body absorbed dose rate of the eight Korean reference animals and plants (RAPs) was estimated by the K-BIOTA (the Korean computer code to assess the risk of radioactivity to wildlife). The first-year effective and thyroid human doses ranged from 5.7E-5 mSv in the infant group to 2.0E-4 mSv in the 5 years group, and from 5.0E-4 mSv in the infant group to 3.4E-3 mSv in the 5 years group, respectively. The life-time (70 years) effective and thyroid human doses ranged from 1.5E-4 mSv in the infant group to 3.0E-4 mSv in the 5 years group, and from 6.0E-4 mSv in the infant group to 3.5E-3 mSv in the 5 years group, respectively. The estimated maximum whole body absorbed dose rate to the Korean RAPs was 6.7E-7 mGy/d for a snake living in soil (terrestrial biota), and 2.0E-5 mGy/d for freshwater fish (aquatic biota), both of which were far less than the generic dose criteria to protect biota from ionizing radiation. Also, the screening level assessment for ERICA's (Environmental Risks from Ionizing Contaminants: Assessments and management) limiting organisms showed that the risk quotient (RQ) for the estimated maximum soil and water activity was significantly less than unity for both the terrestrial and freshwater organisms. Conclusively, the radiological risk of the radioactivity released into the environment by the Fukushima nuclear accident to the public and the non-human biota in the republic of Korea is considered negligible.

• Journal of Radiation Protection and Research
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• v.41 no.3
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• pp.291-300
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• 2016

Background: It is necessary to assess radiation dose to workers due to inhalation of airborne particulates containing naturally occurring radioactive materials (NORM) to ensure radiological safety required by the Natural Radiation Safety Management Act. The objective of this study is to develop an internal dose assessment procedure for workers at industries using raw materials containing natural radionuclides. Materials and Methods: The dose assessment procedure was developed based on harmonization, accuracy, and proportionality. The procedure includes determination of dose assessment necessity, preliminary dose estimation, airborne particulate sampling and characterization, and detailed assessment of radiation dose. Results and Discussion: The developed dose assessment procedure is as follows. Radioactivity concentration criteria to determine dose assessment necessity are $10Bq{\cdot}g^{-1}$ for $^{40}K$ and $1Bq{\cdot}g^{-1}$ for the other natural radionuclides. The preliminary dose estimation is performed using annual limit on intake (ALI). The estimated doses are classified into 3 groups ( < 0.1 mSv, 0.1-0.3 mSv, and > 0.3 mSv). Air sampling methods are determined based on the dose estimates. Detailed dose assessment is performed using air sampling and particulate characterization. The final dose results are classified into 4 different levels ( < 0.1 mSv, 0.1-0.3 mSv, 0.3-1 mSv, and > 1 mSv). Proper radiation protection measures are suggested according to the dose level. The developed dose assessment procedure was applied for NORM industries in Korea, including coal combustion, phosphate processing, and monazite handing facilities. Conclusion: The developed procedure provides consistent dose assessment results and contributes to the establishment of optimization of radiological protection in NORM industries.

• The Journal of Korean Society for Radiation Therapy
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• v.10 no.1
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• pp.30-44
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• 1998

Accurate delivery of doses using a high dose rate(HDR) brachytherapy, remote afterloading system(RALS) depends on knowing the strength of the radioactive source at the time of treatment, the precision and consistency of the timer, and the ability of the unit to position the source at the proper dwell location along the applicator. Periodic Quality Assurance(QA) on HDR machines is a part of the standard protocol of any user. The safety of the patient & staff, positional accuracy, temporal accuracy, and dose delivery accuracy are periodically(weekly, quarterly, monthly) estimated using HDR source(Ir-192), treatment planning devices, measurement devices, and overall treatment devices with regard to treatment delivery. The overall measurement results are estimated successfully and assessed its clinical significance. As a result, our HDR brachytherapy units has been very accurate until now. The QA program protocol permits routine clinical use and provides a high confidence level in the accurate operation of HDR units. Therefore, regular QA of HDR brachytherapy is essential for successful treatment.

• Journal of the Korean Society of Safety
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• v.8 no.4
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• pp.165-174
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• 1993

For the safety assessment of the demonstration-scale incineration plant for treating the combustible radioactive wastes, radiation doses of a worker and a residential individual were estimated. The demonstration plant showed a good performance of trial-burn tests using non-radioactive tracers with resulting In high mass reduction of around 40 times and very low emmission of dusts through a stack, which promised a high decontamination factor in an order of 10$^{7}$ . Based on the result s obtained from the trial-burns in the process, the estimation of radiation dose for workers and general publics near the plant was made using dose pathway calculation theories. The parametric values for calculation were selected from design and operational results of the process and from more conservative conditions In reference data. The estimated annual doses for workers and residential indivisuals were 3.07 $\times$ 10$^{-4}$ and 4.35 X 10$^{-8}$ $\mu$Sv/y, respectively, which were high enough to operate the process when comparing with the allowable dose limit in the regulation. The dose calculation models were quite applicable with showing an excellent safety for the process.

• Progress in Medical Physics
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• v.28 no.2
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• pp.45-48
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• 2017

This paper evaluated the amount of radiation generated by wedge filters during radiation therapy using a high-energy linear accelerator, and the dose to the worker during wedge replacement. After 10-MV photon beam was irradiated with wedge filter, the wedge was removed from the linear accelerator, and the dose rate and energy spectrum were measured. The initial measurement was approximately 1 uSv/h, and the radiation level was reduced to 0.3 uSv/h after 6 min. The effective half-life derived from the dose rate measurement was approximately 3.5 min, and the influence of AI-28 was about 53%. From the energy spectrum measurements, a peak of 1,799 keV was measured for AI-28, while the peak for Co-58 was not measured in the control room. The peaks for Au-106 and Cd-105 were found only measurement was done without wedge removement from the linear accelerator. The additional doses received by the radiation worker during wedge replacement were estimated to be 0.08-0.4 mSv per year.

• Journal of Radiation Protection and Research
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• v.44 no.4
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• pp.156-160
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• 2019

Background: To compare the dose of radiation received by the fetus in a pregnant patient irradiated for head and neck cancer using helical tomotherapy and three-dimensional conformal radiation therapy (3DCRT). Materials and Methods: The patient was modeled with a humanoid phantom to mimic a gestation of 26 weeks. Radiotherapy with a total dose of 2 Gy was delivered with both tomotherapy (2.5 and 5.0 cm jaw size) and 3DCRT. The position of the fetus was predicted to be 45 cm from the field edge at the time of treatment. The delivered dose was measured according to the distance from the field edge and the fetus. Results and Discussion: The accumulated dose to the fetus was 1.6 cGy by 3DCRT and 2 and 2.3 cGy by the 2.5 and 5 cm jaw tomotherapy plans. For tomotherapy, the fetal dose with the 2.5 cm jaw was lower than that with the 5 cm jaw, although the radiation leakage was greater for 2.5 cm jaw plan due to the 1.5 fold longer beam-on time. At the uterine fundus, tomotherapy with a 5 cm jaw delivered the highest dose of 2.4 cGy. When the fetus moves up to 35 cm at the 29th week of gestation, the resultant fetal doses for 3DCRT and tomotherapy with 2.5 and 5 cm jaws were estimated as 2.1, 2.7, and 3.9 cGy, respectively. Conclusion: For tomotherapy, scattering radiation was more important due to the high monitor unit values. Therefore, selecting a smaller jaw size for tomotherapy may reduce the fetal dose. however, evaluation of risk should be individually performed for each patient.

• Nuclear Engineering and Technology
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• v.42 no.3
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• pp.313-318
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• 2010

Alnine pellets were installed in a nuclear power plant for one or two operation cycles and measured by electron spin resonance (ESR) spectrometers for dosimetry. Dose and "x/y ratio", i.e., satellite peak over main center peak ratio, were measured for the returned alanine dosimeters from the nuclear power plant and compared to the values of reference alanine dosimeters exposed only to gamma rays. The variation of the x/y ratio change depended on the population of radicals from each radiation component with different LET. The gamma dose in a mixed radiation field was estimated by an additive gamma ray irradiation experiment and the measured dose rate at specified locations in the containment building.

• Journal of Radiation Protection and Research
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• v.32 no.3
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• pp.111-115
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• 2007

A deuterium oxide leakage accident occurred on October 4, 1999, at nuclear power plant in Korea. The concentration of tritium in air increased and 22 workers were exposed by tritium at that time. It is well known that tritium causes internal exposure. Therefore, we examined complete blood cell count, physical and biological dosimetry fur 13 workers among whole 22 workers to check the health effect and to evaluate the dose estimation of tritium exposure. The leukocyte count test, one of general blood test, was normal. The estimated doses were 0 - 4.44 mSv by physical dosimetry and 0-37 mGy by biological dosimetry. This dose does not exceed radiation dose limit, and the clinical symptoms of the exposed workers were not shown. The consistency between clinical sign and estimated dose means that physical and biological dosimetry were very useful especially in accident evaluation.

• Journal of Radiation Industry
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• v.17 no.1
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• pp.111-118
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• 2023

The International Commission on Radiological Protection (ICRP) 103 recommends a cost-benefit analysis method as an auxiliary tool for scientific and rational decision-making for the principle of optimization of radiological protection. In order to conduct a cost-benefit analysis, the safety improvement of nuclear power by regulation must be measured and converted into monetary terms. The improvement of nuclear safety can be measured by reducing the radiation exposure dose of the people, and it is necessary to determine the coefficient to convert the radiation exposure dose into money. The monetary coefficient is calculated as the product of the statistical life value (VSL) and the nominal risk coefficient. In order to derive the monetary coefficient, the willingness to pay (WTP) can be estimated using the contingent valuation method (CVM), which quantifies the value of non-market goods by converting them into monetary units. WTP can be estimated based on the random utility model, which is the basic model for bivariate selection type conditional value measurement data. Statistical life value can be calculated using the estimated WTP and reduction in early mortality, and a monetary coefficient can be derived.