• Journal of the Korean Society of Radiology
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• v.15 no.5
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• pp.755-760
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• 2021

In this study, the shielding rate of major X-ray protective equipment used in the medical environment was calculated using X-ray spectrum data emitted from the diagnostic X-ray generator of The Institute of Physics and Engineering(IPEM) Report-78, and the applicability of radiation protection was investigated. Radiation shielding rates were calculated through reduction rates of air-kerma and total intensity for lead apron (0.3 mmPb), thyroid shield (0.5 mmPb), lead goggles (0.5 mmPb), and lead glass (1.8, 2.7, 3.3 mmPb) used for diagnostic X-ray protection. As a result, the shielding rate calculated as the air kerma reduction rate ranged from 96.31 to 100% at 80 kV, and 90.35 to 100% at 120 kV. In addition, the results of this calculation were well matched with the results of previous studies measuring the actual shielding rate, and it is expected that the X-ray spectrum data of IPEM Report-78 can be used for radiation protection.

• Journal of the Korean Society of Radiology
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• v.14 no.5
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• pp.545-552
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• 2020

The purpose of this study is to derive a lead thickness calculation formula for direct-shielded doors based on NCRP Report No.151 and IAEA Safety Report Series N0.47. After deriving the dose rate calculation formula for the direct shielded door, this formula was substituted for the lead shielding thickness calculation formula to derive the shielding thickness calculation formula at the door. The lead shielding thickness calculated from the derived direct shielded door shielding thickness calculation formula was about 6% lower than that calculated by the NCRP and IAEA secondary barrier shielding thickness calculation methods. This result is interpreted as meaning that the thickness calculation is more conservative from the NCRP and IAEA secondary barrier shielding thickness calculation methods and fits well for secondary beam shielding. In conclusion, it is thought that the formula for calculating lead shielding thickness of the direct shielded door derived in this study can be usefully used in the shield design of the door.

• Journal of the Korean Society of Radiology
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• v.15 no.6
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• pp.813-819
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• 2021

In this study, the shielding rate of L-block-type shielding equipment used for radiation protection when radioactive fluorine is injected into the human body and the dose distribution of the space in the injection room were calculated using the Monte Carlo method. The shielding rate of the body and window parts of the L-block-type shielding equipment was 99.99%. The dose distribution calculated at a distance of 1 m was relatively high at 135°, 45°, 225°, 315°, and 180° of the XZ plane, and was calculated to be very low at 0°, 90°, and 270°. In the YZ plane, it was relatively high at 135°, 180°, and 225°, and was calculated very low at the remaining angles. The AZ and BZ planes also showed similar results to the YZ plane. In addition, it was confirmed that the shielding rate was the best in the range of 225° to 315° through the dose distribution in the horizontal direction of the source and the 45° direction above the source. These results can be used as basic data necessary for radiation protection of radiation workers.

• Journal of radiological science and technology
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• v.40 no.2
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• pp.281-287
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• 2017

The purpose of this study is to calculate the thickness of shielding for concrete which is mainly used for radiation shielding and study of the walls constructed to shield medical linear accelerator. The optimal shielding thickness was calculated using MCNPX(Ver.2.5.0) for 10 MV of photon beam energy generated by linear accelerator. As a result, the TVL for photon shielding was formed at 50~100 cm for pure concrete and concrete with Boron+polyethylene at 80~100 cm. The neutron shielding was calculated 100~140 cm for pure concrete and concrete with Boron+polyethylene at 90~100 cm. Based on this study, the concrete is considered to be most efficient method of using steel plates and adding Boron+polyethylene th the concrete.

• 대한방사선방어학회:학술대회논문집
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• 2011.11a
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• pp.178-179
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• 2011

• Progress in Medical Physics
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• v.15 no.4
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• pp.210-214
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• 2004

Since production of radioactive isotope for using PET, a lot of neutrons were produced. The produced neutrons were mainly shielded by concrete facility. Secondary photons are generated and emitted from the concrete shielding wall of the PET cyclotron since the proton-generated neutrons are thermalized and absorbed in the concrete wall and emit secondary radiations, i.e., photons. This study calculated neutron dose and photon dose at outside of the accelerator facility using MCNPX code. As results of the calculation, total dose were calculated less than limited dose by law.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.275-281
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• 1996

ENDF/B-VI 핵단면적자료를 기초로 생성된 BUGLE93$^{[1]}$ 라이브러리를 이용하여 울진 3.4호기 원자로 주변의 콘크리트 일차차폐벽에 대한 방사선차폐해석을 수행하였다. 중성자 및 감마선 수송계산은 일차원 각분할 해석코드인 ANISN-ORNL$^{[2]}$ 을 이용하였다. 또한, 기존의 영광 3.4호기 설계에 이용하였던 CASK$^{[3]}$ 라이브러리를 대체할 경우 예상되는 차폐효과의 변화를 평가하기 위하여 노심으로부터 일차차폐벽 사이의 모든 매질에 대한 중성자 및 감마선속을 계산하고. 계산결과를 비교.분석하여 제시하였다. 중성자선속에 대한 분석결과, BUGLE93을 이용한 계산결과는 원자로용기 내부에서는 CASK를 이용한 결과보다 적은, 보다 현실적인 결과를 제공하지만 일차차폐벽내에서는 CASK를 이용한 결과보다 오히려 큰 선속을 보였다. 그러나 이차감마선에 의한 분석결과는 원자로용기 내부에서의 큰 차이에도 불구하고 일차차폐벽을 통과하면서 두결과가 거의 일치하였다. 이것은 BUGLE93 라이브러리가 노심 및 철성분에 대해서는 증가된 핵단면적을 제공하지만 콘크리트 성분에 대한 핵단면적은 오히려 감소하였기 때문이다. 결론적으로. 최소 7피트 두께의 일차차폐벽 외부에서 중성자선속은 감마선속에 비하여 무시할 수 있을 정도이므로. 원자로 내부영역에서 CASK 라이브러리와는 다른 결과를 보이는 BUGLE93 라이브러리를 원자로 일차차폐벽의 방사선차폐해석에 사용할 경우 기존의 CASK 라이브러리를 이용한 해석결과와 동일한 결과를 보이는 것으로 평가되었다.

• Journal of Astronomy and Space Sciences
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• v.19 no.2
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• pp.151-162
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• 2002

Two approximate calculation models for a cosmic radiation shielding in satellite are compared with detailed 3-dimensional calculation results. One is a sectoring method and the other is a chord-length distribution method. Shielding caltulation is performed for KITSAT-1 under the assumed environment at SAA (South Atlantic Anomaly) location with AP-8 radiation spectrum model. When both approximate models are applied, calculation error is expected compared with 3-D detailed geometry calculation because of straight knock-on assumption neglecting the deflection of incident proton. However, both approximate models showed good agreements with 3-dimensional detailed Monte Carlo calculation in two dose detector locations.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.269-274
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• 1996

월성원자력발전소 2호기와 같은 CANDU 6형 원자로의 반응도제어기구 설치대에는 여러 반응도제어기구가 삽입되기때문에 원자로심으로부터의 방사선흐름현상으로 인한 방사선피폭이 예상될 수 있는 위치이다. 좁고 긴 반응도제어기구 도관에서의 방사선 흐름으로 인한 반응도제어기구 설치대에서의 방사선량을 예측하기 위해 몬테 칼로 MCNP 코드를 1차원 각분할법 코드인 ANISN과 연계하여 사용하였다. 월성원자력2호기의 상단차폐해석을 위한 ANISN 계산, 도관의 방사선흐름을 평가하기 위한 MCNP 계산, 그리고 반응도제어기구 설치대에서의 방사선량율 평가를 위한 MCNP 계산등 3단계 계산 기법의 적응이 시도되었다.

• Journal of the Korean Society of Radiology
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• v.12 no.4
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• pp.467-474
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• 2018

CBCT is useful for improving the accuracy of the treatment site, but Repeated use increases the exposure dose. In this study, we aimed to provide basic data for dose reduction in CBCT implementation by dataization the simulating and dose reduction effect using shielding substance. Material in this study, Analyzation the photon beam by simulate the CBCT Through MCNPX and then calculate the absorption dose of body organ at shooting moment of thoracic abdominal position as target UF-Revise simulated body. At this time. Dose reduction effects at this time were evaluated according to the texture of materials and presence of shielding materials( lead, antimony, barium, sulfate, tungsten, bismuth). When CBCT was taken without shielding, the dose was calculated to be high in the breast and spine, and the dose in the esophagus and lung was calculated to be low. The doses according to the shield material were calculated as barium sulfate, antimony, bismuth, lead, and tungsten. The shielding rate was the highest in the thymus (73.6%) and the breast (59.9%) compared with the dose reduction according to presence or absence of the shield. However, it showed the lowest shielding rate in lung (2.1%) and spine (12.6%).