• 한국의학물리학회지:의학물리
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• 제33권4호
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• pp.53-62
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• 2022

In this study, we have investigated the shielding evaluation methodology for facilities using kV energy generators. We have collected and analysis of safety evaluation criteria and methodology for overseas facilities using radiation generators. And we investigated the current status of shielding evaluation of domestic industrial radiation generators. According to the statistical data from the Radiation Safety Information System, as of 2022, a total of 7,679 organizations are using radiation generating devices. Among them, 6,299 facilities use these devices for industrial purposes, which accounts for a considerable portion of radiation. The organizations that use these devices evaluate whether the exposure dose for workers and frequent visitors is suitable as per the limit regulated by the Nuclear Safety Act. Moreover, during this process, the safety shields are evaluated at the facilities that use the radiation generating devices. However, the facilities that use radiating devices having energy less than or equal to 6 MV for industrial purposes are still mostly evaluated and analyzed according to the National Council on Radiation Protection and Measurements 49 (NCRP 49) report published in 1976. We have investigated the technical standards of safety management, including the maximum permissible dose and parameters assessment criteria for facilities using radiation generating devices, based on the NCRP 49 and the American National Standards Institute/Health Physics Society N.43.3 reports, which are the representative reports related to radiation shielding management cases overseas.

• 방사선산업학회지
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• 제18권1호
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• pp.35-42
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• 2024

Demand for RW transportation is expected to increase due to the continuous generation of RW from nuclear power plants and facilities, decommissioning of plants, and saturation of spent fuel temporary storage facilities. The locational aspect of plants and radiation protection optimization for the public have led to an increasing demand for maritime transportation, necessitating to apprehend the overseas and domestic current status. Given the potential long-term radiological impact on the public in the event of a sinking accident, a pre-transportation exposure assessment is necessary. The objective of this study is to investigate the overseas and domestic RW maritime transportation current status and overseas dose assessment cases for the public in sinking accident. Selected countries, including Japan, UK, Sweden, and Korea, were examined for transport cases, Japan and the U.S were chosen for dose assessment case in sinking accidents. As a result of the maritime transportation case analysis, it was performed between nuclear power plants and reprocessing facilities, from plants to disposal or intermediate storage facilities. HLW and MOX fuel were transported using INF 3 shipments, and all transports were performed low speed of 13 kn or less. As a result of the dose assessment for the public in sinking accident, japan conducted an assessment for the sinking of spent fuel and vitrified HLW, and the U.S conducted for the sinking of spent fuel. Both countries considered external exposure through swimming and working at seashore, and internal exposure through seafood ingestion as exposure pathway. Additionally, Japan considered external exposure through working on board and fishing, and the U.S considered internal exposure through spray inhalation and desalinized water and salt ingestion. Internal exposure through seafood ingestion had the largest dose contribution. The average public exposure dose was 20 years after the sinking, 0.04 mSv yr^-1 for spent fuel and 5 years after the sinking, 0.03 mSv yr^-1 for vitrified HLW in Japan. In the U.S, it was 1.81 mSv yr^-1 5 years after the sinking of spent fuel. The results of this study will be used as fundamental data for maritime transportation of domestic RW in the future.

• Journal of Radiation Protection and Research
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• 제43권2호
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• pp.50-58
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• 2018

Background: In terms of the Level 3 probabilistic safety assessment (Level 3 PSA), ingestion of food that had been exposed to radioactive materials is important to assess the intermediate- and long-term radiological dose. Because the ingestion dose is considerably dependent upon the agricultural and dietary characteristics of each country, the reliability of the assessment results may become diminished if the characteristics of a foreign country are considered. Thus, this study intends to evaluate and analyze the ingestion dose of Korean during a severe accident by completely considering the available agricultural and dietary characteristics in Korea. Materials and Methods: This study uses COMIDA2, which is a program based on dynamic food chain model. It sets the parameters that are appropriate to Korean characteristics so that we can evaluate the inherent ingestion dose of Korean. The results were analyzed by considering the accident date and food category with regard to the $^{137}Cs$. Results and Discussion: The dose and contribution of the food category depicted distinctive differences based on the accident date. Particularly, the ingestion dose during the first and second years depicted a considerable difference by the accident date. However, after the third year, the effect of foliar absorption was negligible and exhibited a similar tendency along with the order of root uptake rate based on the food category. Conclusion: In this study, the agricultural and dietary characteristics of Korea were analyzed and evaluated the ingestion dose of Korean during a severe accident using COMIDA2. By considering the inherent characteristics of Korean, it can be determined that the results of this study will significantly contribute to the reliability of the Level 3 PSA.

• 한국방사선학회논문지
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• 제12권7호
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• pp.807-812
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• 2018

본 연구에서는 국가에서 권고하고 있는 일발촬영 진단참고준위 설정에 사용된 조건을 조사하여 PCXMC v2.0 프로그램을 이용하여 유효선량을 측정하고 생물학적 평가를 해보고자 한다. 그 결과 ICRP 60에서 유효선량은 가장 높은 Pelvis AP는 0.794 mSv 가장 낮은 Chest PA는 0.050 mSv이었다. ICRP 103에서는 남성이 가장 높은 T-Spine AP는 0.733 mSv 가장 낮은 Chest PA는 0.057 mSv, 여성은 가장 높은 T-Spine AP는 0.906 mSv 가장 낮은 Chest PA는 0.052 mSv이었다. 남녀 성인 40세 기준으로 일반촬영별 유효선량을 평가 해 보았고, 선량한도의 제한을 받지 않는 의료피폭이라도 방사선위해의 확률적 영향을 최소화하기 위해서 선량을 권고량 이하로 유지하여 국민의 의료피폭을 줄이기 위해 노력이 필요할 것으로 사료된다.

• 대한방사선기술학회지:방사선기술과학
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• 제45권6호
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• pp.483-489
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• 2022

This study proposes a new skull lateral examination, and provides an improved examination environment for patients and radiologists. The study was divided into three groups. One group was divided into the SID (source to image receptor distance) 110 ㎝ and 180 ㎝ in the skull lateral posture, the other group The other group was divided into an position in contact with the detector and an position without contact with the detector, and the other group was divided into male and female groups, considering that the difference in shoulder width between adult males and females would affect the dose and image quality. For dose evaluation, the ESD (entrance surface dose) was measured at the EAM (external auditory meatus), and the conditions were applied equally at 70 ㎸p, 200 ㎃, and 10 ㎃s. For image quality evaluation, SNR (signal to noise ratio) and CNR (contrast to noise ratio) were measured in frontal sinus, EAM, and sella turcica. As a result of ESD comparison, when sid 110 ㎝ to sid 180 ㎝ was changed among the three groups, ESD values decreased the most to 729.18±4.62 μ㏉ and 224.18±0.74 μ㏉ at 180 ㎝ (p<0.01). The values of SNR and CNR were statistically significant (p<0.01), but there was no qualitative difference. This shows that when the SID is 180 ㎝, it is possible to reduce the dose without lowering the image quality. So, It is suggested that the SID 180 ㎝ is used without contacting the detector when examining the skull lateral.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.295-303
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• 2023

Whole-body counters are widely used to assess internal contamination after a nuclear accident. However, it is difficult to determine radioiodine activity due to limitations in conventional calibration phantoms. Inhaled or ingested radioiodine is heterogeneously distributed in the human body, necessitating time-dependent biodistribution for the assessment of the internal contamination caused by the radioiodine intake. This study aims at calculating counting efficiencies considering the biodistribution of ¹³¹I in whole-body counting measurement. Monte Carlo simulations with computational human phantoms were performed to calculate the whole-body counting efficiency for a realistic radioiodine distribution after its intake. The biodistributions of ¹³¹I for different age groups were computed based on biokinetic models and applied to age- and gender-specific computational phantoms to estimate counting efficiency. After calculating the whole-body counting efficiencies, the efficiency correction factors were derived as the ratio of the counting efficiencies obtained by considering a heterogeneous biodistribution of ¹³¹I over time to those obtained using the BOMAB phantom assuming a homogeneous distribution. Based on the correction factors, the internal contamination caused by ¹³¹I can be assessed using whole-body counters. These correction factors can minimize the influence of the biodistribution of ¹³¹I in whole-body counting measurement and improve the accuracy of internal dose assessment.

• Journal of Radiation Protection and Research
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• 제47권4호
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• pp.181-194
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• 2022

Internal dosimetry is a discipline which brings together a set of knowledge, tools and procedures for calculating the dose received after incorporation of radionuclides into the body. Several steps are necessary to calculate the committed effective dose (CED) for workers or members of the public. Each step uses the best available knowledge in the field of radionuclide biokinetics, energy deposition in organs and tissues, the efficiency of radiation to cause a stochastic effect, or in the contributions of individual organs and tissues to overall detriment from radiation. In all these fields, knowledge is abundant and supported by many works initiated several decades ago. That makes the CED a very robust quantity, representing exposure for reference persons in reference situation of exposure and to be used for optimization and assessment of compliance with dose limits. However, the CED suffers from certain limitations, accepted by the International Commission on Radiological Protection (ICRP) for reasons of simplification. Some of its limitations deserve to be overcome and the ICRP is continuously working on this. Beyond the efforts to make the CED an even more reliable and precise tool, there is an increasing demand for personalized dosimetry, particularly in the medical field. To respond to this demand, currently available tools in dosimetry can be adjusted. However, this would require coupling these efforts with a better assessment of the individual risk, which would then have to consider the physiology of the persons concerned but also their lifestyle and medical history. Dosimetry and risk assessment are closely linked and can only be developed in parallel. This paper presents the state of the art of internal dosimetry knowledge and the limitations to be overcome both to make the CED more precise and to develop other dosimetric quantities, which would make it possible to better approximate the individual dose.

• 한국방사선학회논문지
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• 제9권3호
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• pp.175-181
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• 2015

본 연구는 디지털 흉부엑스선 검사에서 화질의 저하 없이 환자선량을 감소시키기 위한 부가필터와 Ion chamber 센서 조합을 알아보고자 하였다. 실험은 관전압 125 kVp, 관전류 320 mA, AEC모드로 하여 부가필터와 Ion chamber의 센서를 네 가지 조합으로 나누어 선량을 측정하고, PCXMC를 이용하여 장기선량을 산출하였다. 또한 MTF로 물리적 화질을 평가하였다. 그 결과 동일 부가필터의 조건하에서 Ion chamber의 좌우 양쪽 센서 모두를 선택했을 때 입사표면 선량과 장기선량이 가장 낮게 나타났으며, 화질평가에서는 좌우 Ion chamber의 선택과 0.1 mmCu 필터를 선택했을 때 공간주파수 값이 2.494 lp/mm로 가장 높게 나타났다. 결론적으로 디지털 흉부촬영 시 Ion chamber의 좌우 양쪽 센서와 0.1 mmCu 필터를 선택하는 것이 우수한 화질의 영상을 획득하고 환자선량 저감에 도움이 될 것이다.

• 방사성폐기물학회지
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• 제21권2호
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• pp.255-269
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• 2023

Compared to operational wastes, nuclear power plant (NPP) decommissioning wastes are generated in larger quantities within a short time and include diverse types with a wider range of radiation characteristics. Currently used 200 L drums and IP-2 type transport containers are inefficient and restrictive in packaging and transporting decommissioning wastes. Therefore, new packaging and transport containers with greater size, loading weight, and shielding performance have been developed. When transporting radioactive materials, radiological safety should be assessed by reflecting parameters such as the type and quantity of the package, transport route, and transport environment. Thus far, safety evaluations of radioactive waste transport have mainly targeted operational wastes, that have less radioactivity and a smaller amount per transport than decommissioning wastes. Therefore, in this study, the possible radiation effects during the transport from NPP to disposal facilities were evaluated to reflect the characteristics of the newly developed containers and decommissioning wastes. According to the evaluation results, the exposure dose to transport workers, handling workers, and the public was lower than the domestic regulatory limit. In addition, all exposure dose results were confirmed, through sensitivity analysis, to satisfy the evaluation criteria even under circumstances when radioactive materials were released 100% from the container.

• Journal of Radiation Protection and Research
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• 제39권4호
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• pp.176-181
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• 2014

The object of this paper is to evaluate the fission product inventories and radiological doses in a non-LOCA event, based on the U.S. NRC's regulatory methodologies recommended by the TID-14844 and the RG 1.195. For choosing a non-LOCA event, one fuel assembly was assumed to be melted by a channel blockage accident. The Hanul nuclear power reactor unit 6 and the CE $16{\times}16$ fuel assembly were selected as the computational models. The burnup cross section library for depletion calculations was produced using the TRITON module in the SCALE6.1 computer code system. Based on the recently licensed values for fuel enrichment and burnup, the source term calculation was performed using the ORIGEN-ARP module. The fission product inventories released into the environment were obtained with the assumptions of the TID-14844 and the RG 1.195. With two kinds of source terms, the radiological doses of public in normal environment reflecting realistic circumstances were evaluated by applying the average condition of meteorology, inhalation rate, and shielding factor. The statistical analysis was first carried out using consecutive three year-meteorological data measured at the Hanul site. The annual-averaged atmospheric dispersion factors were evaluated at the shortest representative distance of 1,000 m, where the residents are actually able to live from the reactor core, according to the methodology recommended by the RG 1.111. The Korean characteristic-inhalation rate and shielding factor of a building were considered for a series of dose calculations.