• Nuclear Engineering and Technology
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• 제46권4호
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• pp.547-556
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• 2014

Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding, $UO_2$ pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a $2{\times}10$ cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the $2{\times}10$ cask array, dose rates at the center point of the array and at the center of the casks' height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the $2{\times}10$ cask array showed exponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280m from the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than 25mrem/yr, due to the decay of radioactivity of spent fuel in storage.

• Journal of Radiation Protection and Research
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• 제47권4호
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• pp.214-225
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• 2022

Background: The conventional cerium-doped Gd₂Al₂Ga₃O₁₂ (GAGG(Ce)) scintillator-based gamma-ray imager has a bulky detector, which can lead to incorrect positioning of the gammaray source if the shielding against background radiation is not appropriately designed. In addition, portability is important in complex environments such as inside nuclear power plants, yet existing gamma-ray imager based on a tungsten mask tends to be weighty and therefore difficult to handle. Motivated by the need to develop a system that is not sensitive to background radiation and is portable, we changed the material of the scintillator and the coded aperture. Materials and Methods: The existing GAGG(Ce) was replaced with Bi₄Ge₃O₁₂ (BGO), a scintillator with high gamma-ray detection efficiency but low energy resolution, and replaced the tungsten (W) used in the existing coded aperture with lead (Pb). Each BGO scintillator is pixelated with 144 elements (12 × 12), and each pixel has an area of 4 mm × 4 mm and the scintillator thickness ranges from 5 to 20 mm (5, 10, and 20 mm). A coded aperture consisting of Pb with a thickness of 20 mm was applied to the BGO scintillators of all thicknesses. Results and Discussion: Spectroscopic characterization, imaging performance, and image quality evaluation revealed the 10 mm-thick BGO scintillators enabled the portable gamma-ray imager to deliver optimal performance. Although its performance is slightly inferior to that of existing GAGG(Ce)-based gamma-ray imager, the results confirmed that the manufacturing cost and the system's overall weight can be reduced. Conclusion: Despite the spectral characteristics, imaging system performance, and image quality is slightly lower than that of GAGG(Ce), the results show that BGO scintillators are preferable for gamma-ray imaging systems in terms of cost and ease of deployment, and the proposed design is well worth applying to systems intended for use in areas that do not require high precision.

• Journal of Radiation Protection and Research
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• 제6권1호
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• pp.8-24
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• 1981

This paper presents flux-to-dose-rate conversion factors for neutrons and gamma rays based on the American National Standard Institute(ANSI) N666. These data are used to calculated the dose rate distribution of neutron and gamma ray in radiation fields. Neutron flux-to-dose-rate conversion factors for energies from $2.5{\times}10^{-8}$ to 20 MeV are presented; the corresponding energy range for gamma rays is 0.01 to 15 MeV. Flux-to-dose-rate conversion factors were calculated, under the assumption that radiation energy distribution has nonlinearity in the phantom, have different meaning from those values obtained by monoetiergetic radiation. Especially, these values were determined with the cross section library. The flux-to-dose-rate conversion factors obtained in this work were in a good agreement to the values presented by ANSI. Those data will be a useful for the radiation shielding analysis and the radiation dosimetry in the case of continuous energy distributions.

• Nuclear Engineering and Technology
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• 제51권3호
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• pp.818-824
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• 2019

In the present work, we determined the gamma-ray attenuation characteristics of eight different polymers(Polyamide (Nylon 6) (PA-6), polyacrylonitrile (PAN), polyvinylidenechloride (PVDC), polyaniline (PANI), polyethyleneterephthalate (PET), polyphenylenesulfide (PPS), polypyrrole (PPy) and polytetrafluoroethylene (PTFE)) using transmission geometry utilizing the high resolution HPGe detector and different radioactive sources in the energy range 81-1333 keV. The experimental linear attenuation coefficient values are compared with theoretical data (WinXCOM data). The linear attenuation coefficient of all polymers reduced quickly with the increase in energy, at the beginning, while decrease more slowly in the region from 267 keV to 835 keV. The effective atomic number of PVDC and PTFE are comparatively higher than the $Z_{eff}$ of the remaining polymers, while PA-6 possesses the lowest effective atomic number. The half value layer results showed that PTFE ($C_2F_4$, highest density) is more effective to attenuate the gamma photons. Also, the theoretical results of macroscopic effective removal cross section for fast neutrons ($\sum_{R}$) were computed to investigate the neutron attenuation characteristics. It is found that the $\sum_{R}$ values of the eight investigated polymers are close and ranged from $0.07058cm^{-1}$ for PVDC to $0.11510cm^{-1}$ for PA-6.

• 방사성폐기물학회지
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• 제2권1호
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• pp.1-11
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• 2004

한국원자력연구소에서는 고온의 용융염 매질 하에서 사용 후 핵연료를 환원시키는 차세대관리종합공정 연구를 수행 중에 있다. 추후 본 기술개발을 실증시험 하기 위해서는 방사선 차폐능이 확보된 핫셀이 필수적이며, 핫셀은 최대 1,385TBq의 방사능량에 대한 차폐 안전성을 가져야 한다. 최대 방사선원에 대한 핫셀의 차폐능을 확보하기 위하여, 본 연구에서는 실증시험 시 사용후핵연료부터 발생하는 중성자 및 감마선에 의한 선량률이 법적 허용선량치보다 낮게 유지되도록 핫셀의 차폐 설계에 대한 안전성을 평가하였다. QAD-CGGP 및 MCNP-4C 코드를 이용하여 핫셀 차폐체의 설계치에 대한 차폐 계산을 수행하였다. 작업구역에 대한 감마선 차폐계산 결과 QAD-CGGP 코드는 2.10${\times}$$10^{-3}$, 2.97${\times}$$10^{-3}$ mSv/h, MCNP-4C 코드는 1.60${\times}$$10^{-3}$, 2.99${\times}$$10^{-3}$ mSv/h 이었으며, 서비스 구역은 1.01${\times}$$10^{-2}$, 7.88${\times}$$10^{-2}$ mSv/h 로 평가되었다. 그리고 MCNP-4C코드를 이용하여 중성자에 의한 선량률을 계산한 결과, 중성자에 의한 선량률은 감마에 의한 선량률의 약 20% 이하치를 나타내었다. 따라서 선량률 대부분은 감마선에 의한 영향임을 알 수 있었다. 본 연구를 통하여 핫셀의 차폐 설계치가 작업구역의 선량 제한치 0.01 mSv/h 와 서비스 구역에서의 선량 제한치 0.15 mSv/h를 만족시키는 것을 확인할 수 있었다.

• Nuclear Engineering and Technology
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• 제3권2호
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• pp.65-75
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• 1971

국내에서 산출되는 각종 광물골재를 사용하여 방사선 차폐용 중차폐 콩크리트를 제조하고 감마선에 대한 차폐 효과를 실험한 결과 최적하다고 판단된 자철광 중차폐 콩크리트를 대상으로 60Co 감마선의 Broad beam을 사용하여 방사선 차폐 효과를 측정하였다. 본 실험을 통하여 실험적으로 차폐체내의 방사선의 감쇄곡선으로부터 차폐 체 두께의 변화에 따르는 방사선 투과율과의 상호관계에 관한 수식을 다음과 같이 유도해냈다. I (x) = I (ο) exp(-$\mu$X) exp(1.03$\times$$10^{-1}$X-3.38$\times$$10^{-3}$X$^2$＋5.29$\times$$10^{-5}$X$^3$) X< 20 cm 때, I (x) =I (ο) exp(-$\mu$X) exp(4.66$\times$$10^{-2}$ X＋2.12$\times$$10^{-1}$) X＞20 cm 때. 이와같이 얻은 결과식에서 오른쪽 첫번째항은 최초 감마선의 감쇄를 표시하고 그 다음항은 차폐체 내에서의 감마선 재생계수를 나타낸다. 이 실험에 첨가하여 차폐체의 실제 설계에 입각한 입방형 자철광 구조체 (두께 8 cm, 내부공간 40$\times$40$\times$40cm)에 대한 차폐효과를 측정한 결과 평판 차폐체를 사용할 때 보다 투과 방사선이 증가됨을 알았다.

• 전기학회논문지
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• 제63권2호
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• pp.284-290
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• 2014

A fiber-optic beta/gamma dual detector system with two types of sensing probes was fabricated to detect the beta- and gamma-rays simultaneously. As scintillators of the sensing probe type 1, two different inorganic scintillators, $CaF_2(Eu)$ and LYSO(Ce) crystals, were used to obtain the each scintillating efficiency with respect to beta-and gamma-rays and the inherent energy spectra of radioactive isotopes. In the case of the sensing probe type 2, which is composed of two identical inorganic scintillators and a beta shielding material based on the lead, it could discriminate beta- and gamma-rays using a subtraction method. In conclusion, we demonstrated that the proposed fiber-optic beta/gamma dual detector could measure and discriminate beta- and gamma-rays using both energy spectroscopy and subtraction method.

• Journal of Radiation Protection and Research
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• 제41권4호
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• pp.378-383
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• 2016

Background: Radiation dose rates in PRIDE facility is evaluated quantitatively for assessing radiation safety of workers because of large amounts of depleted uranium being handled in PRIDE facility. Even if direct radiation from depleted uranium is very low and will not expose a worker to significant amounts of external radiation. Materials and Methods: ORIGEN-ARP code was used for calculating the neutron and gamma source term being generated from depleted uranium (DU), and the MCNP5 code was used for calculating the neutron and gamma fluxes and dose rates. Results and Discussion: The neutron and gamma fluxes and dose rates due to DU on spherical surface of 30 cm radius were calculated with the variation of DU mass and density. In this calculation, an imaginary case in which DU density is zero was added to check the self-shielding effect of DU. In this case, the DU sphere was modeled as a point. In case of DU mixed with molten salt of 50-250 g, the neutron and gamma fluxes were calculated respectively. It was found that the molten salt contents in DU had little effect on the neutron and the gamma fluxes. The neutron and the gamma fluxes, under the respective conditions of 1 and 5 kg mass of DU, and 5 and $19.1g{\cdot}cm^{-3}$ density of DU, were calculated with the molten salt (LiCl+KCl) of 50 g fixed, and compared with the source term. As the results, similar tendency was found in neutron and gamma fluxes with the variation of DU mass and density when compared with source spectra, except their magnitudes. Conclusion: In the case of the DU mass over 5 kg, the dose rate was shown to be higher than the environmental dose rate. From these results, it is concluded that if a worker would do an experiment with DU having over 5 kg of mass, the worker should be careful in order not to be exposed to the radiation.

• 대한방사선기술학회지:방사선기술과학
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• 제33권2호
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• pp.121-126
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• 2010

영상의학과 검사실을 비롯하여 병원에서 의료방사선 차폐제로 사용되는 대표적인 물질이 납이다. 납은 재질이 연하고 오래 동안 변질되지 않으며, 특히 X(${\gamma}$)선에 대한 선흡수계수가 커서 방사선 차폐제로 매우 유용하다. 그러나 납은 생물학적 구조와 기능에 필요하지 않는 부분이 많아 인체에 과다하게 노출되면 위험하므로 카드뮴, 수은, 비소 등과 같이 중금속으로 분류되어 있다. 이러한 위험성에서 벗어나기 위해서 납과 같은 방사선 차폐능력을 가지고 어떠한 형태로도 가공이 가능한 방사선 차폐물질을 개발하려고 노력하고 있다. 본 연구에서는 인체에 무해한 황산바륨을 이용하여 섬유, 고무, 실리콘에 함유하여 의료방사선 차폐시트를 개발하였고 이를 대상으로 의료방사선 차폐능력을 비교 평가하였다. 평가 결과에 있어서 실리콘에 바륨을 함유하여 제조한 시트가 가장 우수한 차폐능을 보였다.

• 한국방사선학회논문지
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• 제15권7호
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• pp.957-964
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• 2021

PET/CT 검사 시 사용되는 F-18은 양전자를 방출하는 방사성동위원소이며, 높은 에너지의 소멸감마선과 베타선은 방사선작업종사자의 피폭을 초래하는 원인이 된다. 본 연구에서는 핵의학과에서 근무하는 방사선작업종사자의 피폭선량 저감방안의 일환으로, F-18에 대한 Apron의 낮은 차폐 효율의 원인을 규명하고, 아크릴로 이중 차폐한 Apron의 실효성을 평가하였다. L-Block, Apron+아크릴, Apron, 아크릴+Apron, 아크릴 다섯 개의 차폐체를 이용하여 선량을 측정하고, 몬테카를로 시뮬레이션을 수행하여 경향성을 비교하였다. 그 결과, 아크릴로 이중 차폐한 Apron의 차폐율이 Apron 단독 차폐한 경우보다 약 4~8% 높은 차폐효과가 있는 것으로 나타났다. 사용자의 활동성에 크게 영향이 가지 않는 범위 안에서 적절한 두께의 아크릴로 이중 차폐한 개인방호복은 방사선 작업종사자의 피폭저감화에 도움을 줄 수 있을 것이다.