• 대한방사선방어학회:학술대회논문집
• /
• 2009.11a
• /
• pp.230-231
• /
• 2009

• Journal of Radiation Protection and Research
• /
• v.10 no.2
• /
• pp.148-154
• /
• 1985

Radiation shield design for a shipping cask, KSC-1, was evaluated to verify that the cask can be used in the transportation of a spent fuel assembly discharged from KNU 5 & 6. Radiation source term of the spent fuel assembly was calculated with the computer program ORIGEN-79, QAD-CG, ANISN-KA and DOT 3.5 codes Were used in the shielding calculations and the nuclear cross section data needed was extracted from the DLC-23/CASK library. It is concluded that KSC-1 shipping cask satisfies the requirements specified in the relevant regulations under normal conditions of transport and under accident conditions in transport.

• Journal of the Korean Society of Radiology
• /
• v.11 no.6
• /
• pp.423-428
• /
• 2017

In this study, the calculation of the effective spatial dose distribution of the diagnostic imaging laboratory of K university was performed by the Monte Carlo simulation. The radiation generator has a maximum tube voltage of 150 kVp and a maximum current of 700 mA. Using the results, we compared the spatial effective dose distributions of diagnostic imaging laboratory when the shielding door was closed and opened. In conclusion, it was found that the effective dose in the operating room of the diagnostic imaging laboratory does not exceed the annual dose limit (6 mSv/y) of the student (occasional visitor) even when the door is opened. However, since the effective dose when the door is open is about 16 times higher in front of the lead glass window and about 3,000 times higher in front of the doorway than the case when the door is closed, closing the shielding door at the time of the practical exercising reduces unnecessary radiation exposure by great extent.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.05a
• /
• pp.263-268
• /
• 1996

CANDU 6 형 원자로의 반응도제어기구 설치대에 있는 수많은 반응도제어기구들은 원자로심에서 발생한 방사선의 흐름통로를 제공하므로 설치대에서의 방사선 피폭이 예상된다. 이런 반응도제어기구 설치대에서의 방사선량을 예측하기 위하여 1 차원 각분할 전산코드인 ANISN 과 2 차원 각분할 전산코드인 DOT를 사용하여 방사선 차폐해석을 수행하였다. 반응도제어기구 도관을 통과하는 방사선의 흐름에 기인한 월성 2호기 반응도제어기구 설치대 상단에서의 최대 선량율은 31$\mu$Sv/hr 로 설계 목표치 250$\mu$Sv/hr 보다 낮게 평가되었다.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.05a
• /
• pp.275-281
• /
• 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 Radiation Protection and Research
• /
• v.12 no.1
• /
• pp.34-39
• /
• 1987

Radiation shielding analysis for a 6MeV X-ray facility was carried out. The primary and leakage radiation for the facility can be evaluated based on the methodology in NCRP No. 49 and 51. The present study deals with radiation scattering analysis for the outside and inside door of the facility based on the albedo concept. The calculated dose rates were compared with the results of MORSE-CG code calculation and the measured data, resulting in a good agreement, even though there existed some deviation for the inside door. These results can be utilized to the radiation shielding design of the medical and industrial X and gamma ray facilities, and to the safety evaluation of these facilities.

• Journal of Radiation Protection and Research
• /
• v.18 no.2
• /
• pp.27-35
• /
• 1993

A shielding analysis of the shipping cask for transporting the CANDU spent fuel bundles has been studied. Radiation source term has been calculated on spent fuel with burn-up of 7,800 MWD/MTU and 5 years cooling time by ORIGEN2 code. The shielding calculation for the cask capable of transporting 378 bundles of CANDU spent fuel has been made by use of 1-D ANISN and 2-D DOT 4.2 codes. As a result of analysis, the optimum shield thickness of cask was obtained. And it is proved that the safety in radiation shielding under normal transport and hypothetical accident conditions is confirmed to satisfy the allowable values specified in IAEA Safety Series No. 6 and the Korean Atomic Law.

• Nuclear Engineering and Technology
• /
• v.25 no.4
• /
• pp.561-569
• /
• 1993

DOT4.2-QAD-CG coupling method was used to analyze the dose rates outside the side and the bottom shield system of CANDU 6 plant. The average dose rates at the main airlock and the new fuel loading area are approximately 6 $\mu$Sv/h as it is required. The calculated dose rates have a good agreement with the measurements at the operating CANDU 6 plant. The method used in this paper can be applied to the radiation shielding analysis of Wolsong 2, 3, and 4 CANDU 6 type plants which will be constructed in the near future.

• Radioisotope journal
• /
• v.19 no.4
• /
• pp.21-31
• /
• 2004

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.37 no.12
• /
• pp.1567-1572
• /
• 2013

To supply electric power in certain extreme environments such as a spacecraft or in military applications, a radioisotope thermoelectric generator has been highlighted as a useful energy source owing to its high energy density, long lifetime, and high reliability. A radioisotope thermoelectric generator generates electric power by using the heat energy converted from the radioactive energy of a radioisotope. In this study, FE analyses such as radiation shield analysis, heat transfer analysis, and power recovery rate analysis have been carried out to achieve an optimal design for a radioisotope thermoelectric generator using $SrTiO_2$.