• 한국진공학회:학술대회논문집
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• 한국진공학회 2009년도 제38회 동계학술대회 초록집
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• pp.314-314
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• 2010

The design of the blanket and shield play a key role in determining the size of a reactor since it has an impact on the various reactor components. The blanket should produce enough tritium for tritium self-sufficiency and the shield should provide sufficient protection for the superconducting TF coil. Neutronic optimization of the blanket and the shield is necessary, and we coupled the system analysis with a neutronic calculation to account for the interrelation of the blanket and shield with the plasma performance of a reactor system in a self-consistent manner. By using the coupled system analysis code, the operational space for a low aspect ratio (LAR) tokamak reactor with a superconducting toroidal field (TF) coil is investigated with an spect ratio in the range of 1.5 - 2.5. The minimum major radius which satisfies all the physics and engineering requirements increases with the magnetic field at the magnetic axis. A required inboard shield thickness is mainly determined by the requirement on the protection of the TF coil against radiation damage. It is shown that to have a fusion power bigger than 3,000 MW in the LAR tokamak with a superconducting TF coil, a major radius bigger than 4.0 m is required.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.989-995
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• 2018

Argonne National Laboratory of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have cooperated on the development, design, and construction of a neutron source facility. The facility was constructed at Kharkov, Ukraine, and its commissioning process is underway. The facility will be used for researches, producing medical isotopes, and training young nuclear specialists. The neutron source facility is designed with a provision to include a cryogenically cooled moderator system-a cold neutron source (CNS). This CNS provides low-energy neutrons, which will be used in the scattering experiment and material structures analysis. Cold neutron guides, coated with reflective material for the low-energy neutrons, will be used to transport the cold neutrons to the experimental site. The cold neutron guides would keep the cold neutrons within certain energy and angular space concentrated inside, while most of the gamma rays and high-energy neutrons are not affected by the cold neutron guides. For the KIPT design, the cold neutron guides need to extend several meters outside the main shield of the facility, and curved guides will also be used to remove the gamma and high-energy neutron. The neutron guides should be installed inside a shield structure to ensure an acceptable biological dose in the facility hall. Heavy concrete is the selected shielding material because of its acceptable performance and cost. Shield design analysis was carried out for the CNS guide hall. MCNPX was used as the major computation tool for the design analysis, with neutron and gamma dose calculated separately. Weight windows variance reduction technique was also used in the shield design. The goal of the shield design is to keep the total radiation dose below the $5.0{\mu}Sv/hr$ guideline outside the shield boundary. After a series of iterative MCNPX calculations, the shield configuration and parameters of CNS guide hall were determined and presented in this article.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 1994년도 가을 학술발표회 논문집
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• pp.399-404
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• 1994

Heavyweight(or High density) concrete, which is generally for shiedling structures, differs from normal weight concrete by having a higher density and special compositions to improve its attenuation properties. There are setting 7 Beam Ports around the reactor of the KMRR Project(Korea Multi-purpose Research Reactor) conducted by the KAERI(Korea Atomic Energy Research Institute). High density(p=5.0t/$\textrm{m}^3$) and Heavyweight(p=3.5t/$\textrm{m}^3$) concrete were placed around the Beam Ports in order to shield radiation. This paper was discussed about construction of High density concrete. High density concrete was placed with method of Preplace Aggregate. Coarse metallic aggregate(steel shot) was used. Boron, boron carbide(B4C), was used to capture effctively the neutrom. The mock-up test was carried out. And the consturction of High density concrete was performed successfully.

• 대한기계학회논문집
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• 제18권7호
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• pp.1833-1839
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• 1994

Experiments have been carried out to investigate the burning characteristics of waste tires in high temperature environments. The burning of waste tire chips consists of four stages ; evaporation of volatile matters, ignition, burning of volatile matters, and burning of solid carbon. Burning time of waste tire chips depends on the gas temperature and the initial weight of the chip. However, the environments. In the ceramic matrix burner with a ceramic radiation shield, the burning time of the waste tire chips becomes shorter than that without the shield. This is due to the increase in heat transfer to the tire chips by radiation.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 춘계학술발표회요약집
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• pp.399-399
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• 1999

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 추계학술발표대회 요약집
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• pp.446.2-446.2
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• 2003

• 대한방사선기술학회지:방사선기술과학
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• 제47권3호
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• pp.197-203
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• 2024

During paranasal sinus X-ray examinations in children, the radiological technologist's thyroid shield is often not implemented to shorten the examination time. This study measured the radiation exposure before and after the implementation of thyroid shielding by analyzing the difference in radiation exposure, the radiological technologist's could receive depending on the actual thyroid shielding. In the left TLD, when thyroid shielding was not performed(N), the radiation exposure dose(mSv) was 2.869 for the depth dose[Hp(10)] and 2.886 for the surface dose[H(3)], and when thyroid shielding was performed(Y), the Hp(10) was 0.033 and the H(3) was 0.034. In the right TLD, when thyroid shielding was not performed(N), the radiation exposure dose was 3.149 for Hp(10) and 3.137 for H(3), and when thyroid shielding was performed, the Hp(10) of (Y) was 0.013 and the H(3) was 0.015. The differences in the overall exposure dose measurement values are all statistically significant (p<0.05). The difference in radiation dose between when thyroid shielding was not performed and when thyroid shielding was performed was more than 99.2% in both cases, indicating a high radiation shielding rate.

• 한국방사선학회논문지
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• 제18권3호
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• pp.283-292
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• 2024

안와림프종의 전자선 치료 시 안구 부속기관의 방사선 부작용 및 합병증으로 인해 후낭하백내장 발생률이 높아질 수 있다. 이에 본 연구는 몬테카를로 전산모사를 활용하여 의료용 선형가속기와 안구의 수학적 모델을 모사하고, 안구 부속기관에 대한 선량평가와 수정체 차폐체 두께에 따른 안구 부속기관별 차폐율을 비교 분석하였다. 안구 부속기관의 선량평가 결과, 차폐체 미사용 시 선량을 기준으로 수정체의 민감 영역에서 가장 높은 흡수선량 분포를 보였으며, 그 외 수정체의 비민감 영역, 전방, 유리체, 각막, 눈꺼풀 순으로 점차 낮아지는 경향을 나타내었다. 차폐체 사용 시 선량 분석 결과, 2 mm 두께 사용 시 수정체의 민감 영역에서 90% 이상의 선량감소효과를 나타냈으며, 비민감 영역과 전방은 83% 이상, 유리체, 각막, 눈꺼풀에서는 30 ~ 62%의 선량감소효과를 보였다. 안와림프종의 전자선 치료 시 수정체 민감 영역의 선량 저감을 위해서는 최소 2 mm 이상의 차폐체 사용이 필요할 것으로 판단되며, 수정체 이외 안구 부속기관에 대한 차폐체 두께 및 면적을 고려한 차폐 방안이 필요할 것으로 사료된다.

• Journal of Radiation Protection and Research
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• 제36권1호
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• pp.16-23
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• 2011

보다 정확한 우라늄 농축도 분석은 핵물질 관리를 위하여 중요하다. 본 연구에서는 감마선을 이용한 우라늄 농축도 분석에서 시료와 검출기 사이에 차폐체가 있는 경우와 측정시간 변화에 따른 분석결과에 대한 정확성 평가 및 오차분석을 수행하였다. 우라늄 농축도를 분석하기 위하여 FRAM (Fixed energy Response function Analysis with Multiple efficiencies)을 이용하였다. FRAM에 의한 분석결과의 정확성은 화학 분석 결과와의 비교를 통해 평가되었다. 연구결과 선원과 검출기사이에 차폐체가 존재할 경우 차폐체의 두께변화에 따른 감마선의 세기는 지수함수 형태로 감소하며 감마선에너지 피이크의 반치폭 (FWHM : Full Width at Half Maximum)은 차폐체의 두께에 영향을 거의 받지 않는 것을 보였다. 따라서 시료와 검출기 사이에 차폐체가 있는 경우의 우라늄 농축도 분석에서 FRAM을 이용한 감마선분광 분석은 유용하게 활용될 수 있다. 본 연구 결과는 차폐체 안에 우라늄이 있는 경우 핵물질 분석에 기여할 수 있을 것이 기대된다.

• 에너지공학
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• 제1권1호
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• pp.66-75
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• 1992

이중창의 열전달 성능을 수치해석과 실험적 방법을 통하여 조사하였다. 이중창 사이의 간격이 0.5cm에서 10cm사이의 실제적인 범위에서 이중창 사이의 간격이 작을 때에는 전도에 의한 열전달이 두드러졌으며 이중창 사이의 간격이 넓어지고 Rayleigh 수가 높은 영역에서는 대류에 의한 열전달이 크게 나타나 이중창 간격이 넓어진 이점을 크게 상쇄시켰다. 이와 같은 이중창 단열문제에 대한 보완책으로 이중창 중앙에 하나의 유리창을 더 설치한 3중창 경우에 대해 대류열전달 감소효과와 복사열차폐(radiation shield) 효과를 검토하였으며 또한 이중창 중간 높이에 스페이서(spacer)를 설치하여 대류유동 강도를 줄이는 방안 등에 대한 분석이 이루어졌다. 3중창을 사용한 경우에는 이중창에 비해 30-50%, 이중창에 스페이서를 설치한 경우에는 l0% 정도의 에너지 절약 효과를 기대할 수 있다.