• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2006년도 학술논문요약집
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• pp.44-45
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• 2006

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2006년도 학술논문요약집
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• pp.71-72
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• 2006

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.785-794
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• 2016

The Argonne National Laboratory of the United States and the Kharkov Institute of Physics and Technology of the Ukraine have been collaborating on the design, development and construction of a neutron source facility at Kharkov Institute of Physics and Technology utilizing an electron-accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100-MeV electrons. The facility was designed to perform basic and applied nuclear research, produce medical isotopes, and train nuclear specialists. The biological shield of the accelerator building was designed to reduce the biological dose to less than 5.0e-03 mSv/h during operation. The main source of the biological dose for the accelerator building is the photons and neutrons generated from different interactions of leaked electrons from the electron gun and the accelerator sections with the surrounding components and materials. The Monte Carlo N-particle extended code (MCNPX) was used for the shielding calculations because of its capability to perform electron-, photon-, and neutron-coupled transport simulations. The photon dose was tallied using the MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is very small, ~0.01 neutron for 100-MeV electron and even smaller for lower-energy electrons. This causes difficulties for the Monte Carlo analyses and consumes tremendous computation resources for tallying the neutron dose outside the shield boundary with an acceptable accuracy. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were utilized for this study. The generated neutrons were banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron dose. The weight windows variance reduction technique was also utilized for both neutron and photon dose calculations. Two shielding materials, heavy concrete and ordinary concrete, were considered for the shield design. The main goal is to maintain the total dose outside the shield boundary less than 5.0e-03 mSv/h during operation. The shield configuration and parameters of the accelerator building were determined and are presented in this paper.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 2004년도 춘계 학술발표회 제16권1호
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• pp.608-611
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• 2004

As the use of various electronic equipments has been increased recently according to industrialization and information network establishment, concern about electromagnetic wave exposed environment has also been increased. Therefore, this study aims to verify electromagnetic wave absorbing effects of inorganic paint that is made of carbon, electro-conductive materials with regard to its physical characteristics, its electromagnetic wave absorbing rate through a mock-up test for proving its effects in the indoor condition. The results are as follows: The results of running tests on electromagnetic wave absorbing inorganic paints for checking their requirements as painting material such as adherence degree, resistance to fine crack, resistance to washing, alkali-resistance, discoloration-resistance, etc. show that inorganic paints have the physical characteristics meeting the requirements for painting materials. In addition, it shows that the electromagnetic wave absorbing effect, in line with the number of paintings and the thickness of paintings, secures $75\~89\%$ of efficiency. And the mock-up test shows that the electromagnetic wave absorbing effect inside building is directly proportional to the distance from the source of electromagnetic wave such as electronic equipments.

• 한국전자파학회논문지
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• 제24권7호
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• pp.748-751
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• 2013

본 논문은 유한 차분 시간 영역(FDTD: Finite-Difference Time-Domain)법을 이용하여 간단한 건물 구조의 광대역 복사성 펄스 결합에 관한 연구를 수행하였다. 이를 위해 건물을 콘크리트와 유리로 구성하였으며, 각 물질의 전기적 특성을 수치적으로 모델링하였다. 본 논문에서는 본 연구팀에서 개발한 분산 FDTD 알고리즘을 이용하여 건물의 전자파 특성 해석을 수행하고, 건물 구조에 따른 차폐 효과를 50 MHz~1 GHz 대역에서 분석하였다.

• 한국산학기술학회논문지
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• 제11권3호
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• pp.1035-1039
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• 2010

고에너지 X-ray를 투시 방사선원으로 사용한 컨테이너 화물검색시스템에서 생성되는 광중성자에 의해 주변 콘크리트 차폐벽에서 발생되는 방사화생성물을 평가하였다. 몬테카를로 전산해석 코드인 MCNPX2.5.0을 사용하였으며, 참조시스템은 국내 주요 항만에 설치된 9MeV X-ray 고정식 양방향 컨테이너 화물검색시스템이다. 9MeV X-ray 조사에 따라 생성되는 광중성자의 (n,$\gamma$) 반응에 의한 방사화생성물 재고량을 계산하고 이에 따라 야기되는 방사선 피폭선량을 계산하였다.

• Nuclear Engineering and Technology
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• 제46권1호
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• pp.73-80
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• 2014

A concrete cask is an option for spent nuclear fuel interim storage. A concrete cask usually consists of a metallic canister which confines the spent nuclear fuel assemblies and a concrete overpack. When the overpack undergoes a missile impact, which might be caused by a tornado or an aircraft crash, it should sustain an acceptable level of structural integrity so that its radiation shielding capability and the retrievability of the canister are maintained. A missile impact against a concrete overpack produces two damage modes, local damage and global damage. In conventional approaches [1], those two damage modes are decoupled and evaluated separately. The local damage of concrete is usually evaluated by empirical formulas, while the global damage is evaluated by finite element analysis. However, this decoupled approach may lead to a very conservative estimation of both damages. In this research, finite element analysis with material failure models and element erosion is applied to the evaluation of local and global damage of concrete overpacks under high speed missile impacts. Two types of concrete overpacks with different configurations are considered. The numerical simulation results are compared with test results, and it is shown that the finite element analysis predicts both local and global damage qualitatively well, but the quantitative accuracy of the results are highly dependent on the fine-tuning of material and failure parameters.

• 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.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 2010년도 춘계 학술대회 제22권1호
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• pp.191-192
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• 2010

본 연구는 한국원자력연구원에서 사용한 방사선 차폐용 고밀도 중량콘크리트를 현장 적용한 사례로써 제시된 목표 품질 규격을 만족하기 위한 재료 검토, 생산 준비, 현장 적용 제품을 통한 품질 특성에 대하여 각각의 진행 과정에 대하여 정리해 보았다.

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2607-2612
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• 2020

The present work was aimed to show the possibility of using hydrogel (acrylic/boric acid) for evaluation of the neutron radiation shielding. The influence of acrylic acid concentration, different gamma doses and relative contents of boric acid were studied. The physical properties and the thermomechanical stability of the studied samples were investigated. The shielding property of the composite for neutron was tested by Pu-Be neutron source (5 Ci) under room temperature. The neutron fluence rates and gamma fluxes were measured using a stilbene organic scintillator. The macroscopic effective removal cross-section Σ_R (cm^-1) of fast neutrons and total attenuation coefficient μ (cm^-1) of gamma rays has been studied experimentally. The transmission parameters, the relaxation length (??) and the half-value layer (HVL) were obtained. The obtained results indicated that the addition of boric acid to acrylic acid tends to increase the macroscopic effective removal cross-section Σ_R (cm^-1) to 0.141 compared to 0.094 of ordinary concrete.