• Journal of Radiation Protection and Research
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• 제47권2호
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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.

• Journal of Radiation Protection and Research
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• 제11권1호
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• pp.37-43
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• 1986

ICRU 구(球)를 피사체(被射體)로 하여 300 keV 중성자(中性子)의 방사선량(放射線量) 관계량(關係量)을 평가(評價)하였다. 피사체내(被射體內)의 선량당량(線量當量) 분포(分布)를 직접(直接) 산정(算定)하기 위해 중성자(中性子)-광자(光子)-하전입자(荷電粒子) 결합수송(結合輸送)을 다룰 수 있는 몬테칼로 코드 NEDEP을 사용하였다. 계산결과(計算結果) 얻은 방사선량(放射線量) 관계량(關係量)은 다음과 같다. 심부선량당량지수(深部線量當量指數) $H_{I,d}:1.78{\times}10^{11}\;Sv-cm^2$ 표층선량당량지수(表層線量當量指數) $H_{I,s}:2.08{\times}10^{-1}\;Sv-cm^2$ 주위선량당량(周圍線量當量) $H^*(0.07):1.70{\times}10^{-11}\;Sv-cm^2$ 주위선량당량(周圍線量當量) $H^*(10):1.78{\times}10^{-11}\;Sv-cm^2$ 실효선질계수(實效線質係數) $\bar{Q}^*(10):12.4$

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 추계학술논문요약집
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• pp.549-550
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• 2018

• Nuclear Engineering and Technology
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• 제51권8호
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• pp.1991-1997
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• 2019

In this manuscript, we present a method for the direct calculation of an ambient dose equivalent (H^* (10)) for the external gamma-ray exposure with an energy range of 40 keV to 2 MeV in an electronic personal dosimeter (EPD). The designed EPD consists of a 3 × 3 ㎟ PIN diode coupled to a 3 × 3 × 3 ㎣ CsI (Tl) scintillator block. The spectrum-to-dose conversion function (G(E)) for estimating H^* (10) was calculated by applying the gradient-descent method based on the Monte-Carlo simulation. The optimal parameters for the G(E) were found and this conversion of the H^* (10) from the gamma spectra was verified by using ²⁴¹Am, ¹³⁷Cs, ²²Na, ⁵⁴Mn, and ⁶⁰Co radioisotopes. Furthermore, gamma spectra and H^* (10) were obtained for an arbitrarily mixed multiple isotope case through Monte-Carlo simulation in order to expand the verification to more general cases. The H^* (10) based on the G(E) function for the gamma spectra was then compared with H^* (10) calculated by simulation. The relative difference of H^* (10) from various single-source spectra was in the range of ±2.89%, and the relative difference of H^* (10) for a multiple isotope case was in the range of ±5.56%.

• Journal of Radiation Protection and Research
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• 제42권1호
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• pp.42-47
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• 2017

Background: One of the most urgent issues following the accident at the Fukushima Daiichi nuclear power plant (FDNPP) was the remediation of the land, in particular, for residential area contaminated by the radioactive materials discharged. In this study, the effect of decontamination on reduction of ambient dose equivalent outdoors and indoors was evaluated. The latter is essential for residents as most individuals spend a large portion of their time indoors. Materials and Methods: From December 2012 to November 2014, thirty-seven Japanese single-family detached wooden houses were investigated before and after decontamination in evacuation zones. Outdoor and indoor dose measurements (n = 84 and 114, respectively) were collected based on in situ measurements using the NaI (Tl) scintillation surveymeter. Results and Discussion: The outdoor ambient dose equivalents [$H^*(10)_{out}$] ranged from 0.61 to $3.71{\mu}Sv\;h^{-1}$ and from 0.23 to $1.32{\mu}Sv\;h^{-1}$ before and after decontamination, respectively. The indoor ambient dose equivalents [$H^*(10)_{in}$] ranged from 0.29 to $2.53{\mu}Sv\;h^{-1}$ and from 0.16 to $1.22{\mu}Sv\;h^{-1}$ before and after decontamination, respectively. The values of reduction efficiency (RE), defined as the ratio by which the radiation dose has been reduced via decontamination, were evaluated as $0.47{\pm}0.13$, $0.51{\pm}0.13$, and $0.58{\pm}0.08$ ($average{\pm}{\sigma}$) when $H^*(10)_{out}$ < $1.0{\mu}Sv\;h^{-1}$, $1.0{\mu}Sv\;h^{-1}$ < $H^*(10)_{out}$ < $2.0{\mu}Sv\;h^{-1}$, and $2.0{\mu}Sv\;h^{-1}$ < $H^*(10)_{out}$, respectively, indicating the values of RE increased as $H^*(10)_{out}$ increased. It was found that the values of RE were $0.53{\pm}0.12$ outdoors and $0.41{\pm}0.09$ indoors, respectively, indicating RE was larger outdoors than indoors. Conclusion: Indoor dose is essential as most individuals spend a large portion of their time indoors. The difference between outdoors and indoors should be considered carefully in order to estimate residents' exposure dose before their returning home.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.277-284
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• 2017

This paper presents the establishment and characterization of a neutron calibration field using a bare $^{252}Cf$ source of low neutron source strength in Vietnam. The characterization of the field in terms of neutron flux spectra and neutron ambient dose equivalent rates were performed by Monte Carlo simulations using the MCNP5 code. The anisotropy effect of the source was also investigated. The neutron ambient dose equivalent rates at three reference distances of 75, 125, and 150 cm from the source were calculated and compared with the measurements using the Aloka TPS-451C neutron survey meters. The discrepancy between the calculated and measured values is found to be about 10%. To separate the scattered and the direct components from the total neutron flux spectra, an in-house shadow cone of 10% borated polyethylene was used. The shielding efficiency of the shadow cone was estimated using the MCNP5 code. The results confirmed that the shielding efficiency of the shadow cone is acceptable.

• Journal of Radiation Protection and Research
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• 제26권3호
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• pp.183-190
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• 2001

The activity concentration of primordial radionuclides i.e., $^{238}U$ series, $^{232}Th$ series and $^{40}K$, in soil samples collected from Udagamandalam environment, have been measured by employing NaI (Tl) Gamma ray Spectrometer. The absorbed gamma dose rate has also been simultaneously measured by using both Environmental Radiation Dosimeter at each soil sampling location (ambient gamma dose) as well as from the gamma dose derived from the activity concentration of the primordial radionuclides. The results of activity concentration of each radio nuclides in soil, absorbed dose rate in air due to soil activity and possible cosmic radiation at each location along with human effective dose equivalent for Udagamandalam environment are presented and discussed.

• Journal of Radiation Protection and Research
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• 제38권1호
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• pp.37-43
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• 2013

TLD를 이용하여 중성자 선량을 측정할 경우, TLD는 중성자 에너지에 대한 반응도 차이가 크기 때문에 현장 중성자장의 스펙트럼 특성에 맞는 에너지 반응도 보정이 반드시 필요하다. 본 실험에는 소형으로 가공된 TLD 소자를 사용하여 $^{252}Cf$ 중성자장에 설치된 내부구조가 복잡하고 좁은 Long-Counter (중성자 검출기) 내외부에서의 중성자 주위선량당량(ambient dose equivalent)을 측정하였다. 측정결과는 입자수송해석코드(MCNPX)를 이용한 계산결과와 비교하였다. 기존의 TLD 교정 선원인 $D_2O$ 감속 $^{252}Cf$만으로 교정하여 판독한 결과값은 전산모사 계산값과 많은 차이를 보였다. 그러나 bare 및 $D_2O$ 감속 $^{252}Cf$ 선원을 사용하여 생산한 두 교정인자를 혼용한 판독값은 계산값과 비슷하였다. 결과적으로, TLD 소자는 사용 현장과 비슷한 특성을 가지는 중성자장에서 교정되어야지만 올바른 선량평가가 가능함을 확인하였다.

• Journal of Radiation Protection and Research
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• 제41권3호
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• pp.211-215
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• 2016

Background: The moderated neutron calibration fields using $^{241}Am$-Be sources and a graphite moderator have been constructed at the Facility of Radiation Standard (FRS) in the Japan Atomic Energy Agency (JAEA). Materials and Methods: The neutron spectra of the fields were evaluated by the Monte-Carlo calculations and measurements using the Bonner Multi-sphere Spectrometer. Results and Discussion: The fields have continuous neutron spectra from several MeV to thermal neutron energy, with fluence-averaged energies of 0.84 MeV and 0.60 MeV. Reference values of fluence rates and ambient/personal dose equivalent rates were determined from neutron spectra by measurements. Conclusion: Currently, the fields are available for calibration or performance test of neutron measuring instruments.

• Nuclear Medicine and Molecular Imaging
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• 제41권4호
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• pp.309-316
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• 2007

감시 림프절 생검은 유방암 수술에서 림프절 전이 상태를 알기 위한 표준 시술이다. 환자는 방사성 콜로이드를 주사 받은 후 수술을 받게 된다. 이 과정에서 검사를 하는 핵의학과, 수술장, 유방암 검체를 다루는 병리과의 관계자는 미량이나마 환자와 검체에 의해서 방사선 피폭을 받을 수 있다. 이 연구의 목적은 감시 림프절 생검 과정, 특히 병리처리 과정에서 받는 방사선피폭을 정량하여 그 안전성을 확인하고 병리 시설과 폐기물에 대해서도 방사선 관련 안전성을 확인하는 것이다. 대상 및 방법 : 감시림프절 생검은 방사성 콜로이드를 이용하여 일반적인 임상적 방법으로 시행되었다. 병리기사, 핵의학 기사 및 핵의학 의사의 피폭량을 열형광선량계를 이용하여 1달간 측정하였다. 또한 작업과정중의 잔존 방사능량, 흡수선량, 작업시간, 작업거리, 조직폐기물 및 병리검사실의 공간선량을 측정하였다. 결과 전신 및 손의 피폭량은 병리기사에서 각각 0.21 및 0.85 uSv/study이었고 핵의학과 의사 및 핵의학과 기사의 전신피폭량은 각각 0.2 및 2.3 uSv/study 이었다. 일반인 기준(1000 uSv/year)으로 병리기사는 년간 약 1100건 감시림프절 관련 검체 처리를 할 수 있었다. 각 과정의 잔존방사성 및 피폭거리, 시간으로 측정한 피폭량은 수술의사는 전신/손의 피폭량이 건당 2.47/22.4 uSv 이었고 수술장간호사는 건당 0.22/0 uSv 이었다. 병리실의 공간선량률은 0.02-0.03 mR/hr로 방사성 관리구역의 설정 기준에 도달하지 않았다. 폐기되는 검체 조직의 방사능은 거의 측정되지 않아 100 Bq/g에 훨씬 미치지 않았다. 결론: 방사성동위원소를 이용한 감시림프절 검사에 관계된 병리처리과정은 방사선안전측면에서 일반적으로 안전하며 별도의 안전관리나 시설 없이 이루어 질 수 있다.