• Journal of Radiation Protection and Research
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• 제45권2호
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• pp.81-87
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• 2020

Background: Kori unit #1 is permanently shut down after a 40-year lifetime. The Nuclear Safety and Security Commission recommends establishing initial decommissioning plans for all nuclear and radwaste treatment facilities. Therefore, the Korea Atomic Energy Research Institute (KAERI) must establish an initial and final decommissioning plan for radwaste-treatment facilities. Radiation safety assessment, which constitutes one chapter of the decommissioning plan, is important for establishing a decommissioning schedule, a strategy, and cost. It is also a critical issue for the government and public to understand. Materials and Methods: This study provides a method for assessing external radiation dose to workers during decommissioning. An external dose is calculated following each exposure scenario, decommissioning strategy, and working schedule. In this study, exposure dose is evaluated using the deterministic method. Physical characterization of the facility is obtained by both direct measurement and analysis of the drawings, and radiological characterization is analyzed using the annual report of KAERI, which measures the ambient dose every month. Results and Discussion: External doses are calculated at each stage of a decommissioning strategy and found to increase with each successive stage. The maximum external dose was evaluated to be 397.06 man-mSv when working in liquid-waste storage. To satisfy the regulations, working period and manpower must be managed. In this study, average and cumulative exposure doses were calculated for three cases, and the average exposure dose was found to be about 17 mSv/yr in all the cases. Conclusion: For the three cases presented, the average exposure dose is well below the annual maximum effective dose restriction imposed by the international and domestic regulations. Working period and manpower greatly affect the cost and entire decommissioning plan; hence, the chosen option must take account of these factors with due consideration of worker safety.

• Nuclear Medicine and Molecular Imaging
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• 제43권4호
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• pp.259-279
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• 2009

Radioiodine ablation therapy has been considered to be a standard treatment for patient with differentiated thyroid cancer after total thyroidectomy. Patients may need to be hospitalized to reduce radiation exposure of other people and relatives from radioactive patients receiving radioiodine therapy. Medical staffs, nursing staffs and technologists sometimes hesitate to contact patients in radioiodine therapy ward. The purpose of this paper is to introduce radiation dosimetry, estimate radiation dose from patients and emphasize the safety of radiation exposure from patients treated with high dose radioiodine in therapy ward. The major component of radiation dose from patient is external exposure. However external radiation dose from these patients treated with typical therapeutic dose of 4 to 8 GBq have a very low risk of cancer induction compared with other various risks occurring in daily life. The typical annual radiation dose without shielding received by patient is estimated to be 5 to 10 mSv, which is comparable with 100 to 200 times effective dose received by chest PA examination. Therefore, when we should keep in mind the general principle of radiation protection, the risks of radiation exposure from patients are low and the medical personnel are considered to be safe from radiation exposure.

• 한국콘텐츠학회논문지
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• 제17권7호
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• pp.498-504
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• 2017

치아를 칫솔질하거나 치약을 사용하는 것과 같은 구강 건강 행위는 구강 건강을 개선하며, 따라서 삶의 질을 향상시키는 중요한 부분이다. 그러나 화학 물질에 대한 연구도 필요한 실정이다. 따라서 본 연구는 구강 건강 행위로 인해 야기되어지는 불소 노출에 미치는 요인을 조사하여 정확한 구강 건강 지침을 제공하고자 한다. ConsExpo 5.0 모델에서 불소 화합물의 경구 노출을 적용한 결과, 일일 불소 인체노출량 추정은 성인남성의 모델 결과 oral external dose는 0.000196 mg/kg, oral acute(internal) dose는 0.000196 mg/kg, oral chronic(internal)dose는 0.000465 mg/kg/day로 추정되었다. 성인여성은 연구결과 oral external dose는 $4.1{\times}10^{-6}mg/kg$, oral acute(internal) dose $4.1{\times}10^{-6}mg/kg$, oral chronic(internal) dose $9.99{\times}10^{-6}mg/kg/day$로 추정되었다.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1781-1788
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• 2022

The containment building Kori Unit 1 may require sequential steps for full decommissioning. This study assumes that the containment building is to be used as an auxiliary building that handles nuclear power systems and materials during decommissioning before conversion into a greenfield. Through the derivation of guidelines and dose evaluation, it was confirmed whether the radiation workers were satisfied with the ALARA decision. The specific modeling of the external radiation exposure was performed based on the facility investigation procedures. The external radiation specific derived concentration guideline levels (DCGLs) for radiation workers in containment building were obtained using the RESRAD-BUILD code and were applied to the VISIPLAN 3D ALARA Planning Tool code to calculate the working dose and check worker safety. The derivation of site-specific and realistic DCGLs and dose evaluation via 3D modeling can contribute to the scenario development for the decommission and remediation of containment building.

• 한국산업보건학회지
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• 제29권2호
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• pp.167-175
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• 2019

Objectives: The purpose of this study was to evaluate airborne radon and thoron levels and estimate the effective doses of workers who made household goods and mattresses using monazite. Methods: Airborne radon and thoron concentrations were measured using continuous monitors (Rad7, Durridge Company Inc., USA). Radon and thoron concentrations in the air were converted to radon doses using the dose conversion factor recommended by the Nuclear Safety and Security Commission in Korea. External exposure to gamma rays was measured at the chest height of a worker from the source using real-time radiation instruments, a survey meter (RadiagemTM 2000, Canberra Industries, Inc., USA), and an ion chamber (OD-01 Hx, STEP Co., Germany). Results: When using monazite, the average concentration range of radon was $13.1-97.8Bq/m^3$ and thoron was $210.1-841.4Bq/m^3$. When monazite was not used, the average concentration range of radon was $2.6-10.8Bq/m^3$ and the maximum was $1.7-66.2Bq/m^3$. Since monazite has a higher content of thorium than uranium, the effects of thoron should be considered. The effective doses of radon and thoron as calculated by the dose conversion factor based on ICRP 115 were 0.26 mSv/yr and 0.76 mSv/yr, respectively, at their maximum values. The external radiation dose rate was $6.7{\mu}Sv/hr$ at chest height and the effective dose was 4.3 mSv/yr at the maximum. Conclusions: Regardless of the use of monazite, the total annual effective doses due to internal and external exposure were 0.03-4.42 mSv/yr. Exposures to levels higher than this value are indicated if dose conversion factors based on the recently published ICRP 137 are applied.

• Journal of Radiation Protection and Research
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• 제45권2호
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• pp.53-68
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• 2020

Background: International organizations such as the World Health Organization (WHO) and the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) reported public exposure doses due to radionuclides released in the Fukushima nuclear accident a few years after the event. However, the reported doses were generally overestimated due to conservative assumptions such as a longer stay in deliberate areas designated for evacuation than the actual stay. After these reports had been published, more realistic dose values were reported by Japanese scientists. Materials and Methods: The present paper reviews those reports, including the most recently published articles; and summarizes estimated effective doses (external and internal) and issues related to their estimation. Results and Discussion: External dose estimation can be categorized as taking two approaches-estimation from ambient dose rate and peoples' behavior patterns-and measurements using personal dosimeters. The former approach was useful for estimating external doses in an early stage after the accident. The first 4-month doses were less than 2 mSv for most (94%) study subjects. Later on, individual doses came to be monitored by personal dosimeter measurements. On the basis of these measurements, the estimated median annual external dose was reported to be < 1 mSv in 2011 for 22 municipalities of Fukushima Prefecture. Internal dose estimation also can be categorized as taking two approaches: estimation from whole-body counting and estimation from monitoring of environmental samples such as radioactivity concentrations in food and drinking water. According to results by the former approach, committed effective dose due to ¹³⁴Cs and ¹³⁷Cs could be less than 0.1 mSv for most residents including those from evacuated areas. Conclusion: Realistic doses estimated by Japanese scientists indicated that the doses reported by WHO and UNSCEAR were generally overestimated. Average values for the first-year effective doses for residents in two affected areas (Namie Town and Iitate Village) were not likely to reach 10 mSv, the lower end of the doses estimated by WHO.

• Nuclear Engineering and Technology
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• 제53권1호
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• pp.273-281
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• 2021

The radiological safety of the spent resin treatment facility with a¹⁴C treatment capacity of 1 ton/day was evaluated in terms of the external and internal exposure of worker according to operation scenario. In terms of external dose, the annual dose for close work for 1 h/day at a distance of more than 1 m (19.8 mSv) satisfied the annual dose limit. For 8 h of close work per day, the annual dose exceeded the dose limit. For remote work of 2000 h/year, the annual dose was 14.4 mSv. Lead shielding was considered to reduce exposure dose, and the highest annual dose during close work for 1 h/day corresponded to 6.75 mSv. For close work of 2000 h/year and lead thickness exceeding 1.5 cm, the highest value of annual dose was derived as 13.2 mSv. In terms of internal exposure, the initial year dose was estimated to be 1.14E+03 mSv when conservatively 100% of the nuclides were assumed to leak. The allowable outflow rate was derived as 7.77E-02% and 2.00E-01% for the average limit of 20 mSv and the maximum limit of 50 mSv, respectively, where the annual replacement of the worker was required for 50 mSv.

• 한국방사선학회논문지
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• 제11권5호
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• pp.321-328
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• 2017

D-Shuttle (Chiyoda Technol Corporation, Tokyo, Japan) 선량계를 이용하여 개인피폭관리 및 자연방사선량의 모니터링을 위한 기초자료를 제공하는데 연구의 목적이 있다. D-Shuttle을 이용하여 선량을 산출하였다. 선량보고서에서 400 일 노출되었을 때에 1.346 mSv 이었고, 연간선량 (annual dose per year)은 1.228 mSv/year, 평균시간선량 (average dose per hour)은 $0.014{\mu}Sv/hr$ 이었다. 국내의 개인 외부피폭선량 (1.295 mSv/year =Korea average natural individual external dose), 국내의 연간부가선량 (additional dose per year)은 -0.0663 mSv/year 이다. D-Shuttle은 방사선모니터링을 위한 개인선량계로 방사선의 검출성능 우수한 기능, 실시간 방사선 피폭관리, 방사선 작업의 경보 기능, 효율적이고 사용이 편리한 개인 방사선선량의 피폭관리로 ALARA에 매우 유용한 선량계로 사용할 수 있다. 방사선작업종사자와 지역주민의 방사선모니터링 측정기기로 병원, 산업, 의료현장, 원전사고 지역과 비파괴 분야의 위험한 지역에서 방사선모니터링으로 활용될 수 있다.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1225-1233
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• 2024

Level 3 probabilistic safety assessment (PSA) is performed to calculate radionuclide concentrations and exposure dose resulting from nuclear power plant accidents. To calculate the external exposure dose from the released radioactive materials, the radionuclide concentrations are multiplied by two factors of dose coefficient and a finite cloud dose correction factor (FCDCF), and the obtained values are summed. This indicates that a standard set of FCDCFs is required for external exposure dose calculations. To calculate a standard set of FCDCFs, the effective distance from the release point to the receptor along the wind direction should be predetermined. The TID-24190 document published in 1968 provides equations to calculate FCDCFs and the resultant standard set of FCDCFs. However, it does not provide any explanation on the effective distance required to calculate the standard set of FCDCFs. In 2021, Sandia National Laboratories (SNLs) proposed a method to predetermine finite effective distances depending on the atmospheric stability classes A to F, which results in six standard sets of FCDCFs. Meanwhile, independently of the SNLs, the authors of this paper discovered that an infinite effective distance assumption is a very reasonable approach to calculate one standard set of FCDCFs, and they implemented it into the multi-unit radiological consequence calculator (MURCC) code, which is a post-processor of the level 3 PSA codes. This paper calculates and compares short- and long-range FCDCFs calculated using the TID-24190, SNLs method, and MURCC method, and explains the strength of the MURCC method over the SNLs method. Although six standard sets of FCDCFs are required by the SNLs method, one standard sets of FCDCFs are sufficient by the MURCC method. Additionally, the use of the MURCC method and its resultant FCDCFs for level 3 PSA was strongly recommended.

• Journal of Radiation Protection and Research
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• 제19권3호
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• pp.209-221
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• 1994

원자력발전소의 중대 사고시 대기로 방출된 방사성물질에 의해 피폭자가 사고후 일생동안 받게 될 전신 피폭선량과 핵종의 상대적 중요도를 방출점으로부터 거리에 따라 각 피폭경로에 대해 평가하였다. 방사능운과 지표에 침적된 방사성물질에 의한 외부피폭, 호흡과 오염된 음식물섭취에 의한 내부피폭이 피폭경로로 고려되었다. 오염된 음식물섭취에 의한 영향은 우리나라 환경을 고려하여 개발된 동적 삽식경로모델 KORFOOD을 사용하여 침적시점과 침적후 시간에 따른 음식물내 방사성물질의 농도 변화를 고려하였다. 방출점으로부터 80km까지 피폭선량을 평가한 결과, 오염된 음식물섭취에 의한 영향이 가장 높았다. 핵종별 기여도는 방사능운에 의한 외부피폭과 호흡에 의한 내부피폭의 경우 I, 침적된 방사성물질에 의한 외부피폭의 경우 Cs에 의한 영향이 가장 높았다. 오염된 음식물섭취에 의한 내부피폭의 경우 Cs은 여름철 침적, Sr은 겨울철 침적에 보다 중요한 영향을 미쳤다.