• Nuclear Engineering and Technology
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• 제54권2호
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• pp.601-607
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• 2022

In recent years, monitoring of spent nuclear fuel inside dry cask storage has become an important area of national security. Muon tomography is a useful method for monitoring spent nuclear fuel because it uses high energy muons that penetrate deep into the target material and provides a 3-D structure of the inner materials. We designed a muon tomography system consisting of four 2-D position sensitive detector and characterized and optimized the system parameters. Each detector, measuring 200 × 200 cm², consists of a plastic scintillator, wavelength shifting (WLS) fibers and, SiPMs. The reconstructed image is obtained by extracting the intersection of the incoming and outgoing muon tracks using a Point-of-Closest-Approach (PoCA) algorithm. The Geant4 simulation was used to evaluate the performance of the muon tomography system and to optimize the design parameters including the pixel size of the muon detector, the field of view (FOV), and the distance between detectors. Based on the optimized design parameters, the spent fuel assemblies were modeled and the line profile was analyzed to conduct a feasibility study. Line profile analysis confirmed that muon tomography system can monitor nuclear spent fuel in dry storage container.

• 방사성폐기물학회지
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• 제9권2호
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• pp.93-98
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• 2011

염분분위기에서의 부식은 사용후핵연료의 중간저장 기간동안 304 스테인레스 강재 건식저장용기의 주 열화기구들 중 하나다. 본 연구에서는 감소정도가 서로 다른 냉연 304 스테인레스 강 시편들에 0.5wt.%의 염화나트륨 연무를 분사시키면서 느린 변형속도시험(SSRT)과 중성염 분사시험(NSS)을 $85^{\circ}C$와 $200^{\circ}C$ 에서 수행하였다. $85^{\circ}C$에서 2000 시간 동안 시험한 NSS시편의 무게 변화는 $200^{\circ}C$에서 시험한 시편의 무게 변화와 크게 달랐다. NSS 시편의 $85^{\circ}C$에서 무게 감량은 미미하였지만, 냉연 감소율이 증가함에 따라서 무게 변화는 점진적으로 감소하였다. $85^{\circ}C$와 $200^{\circ}C$에서 그리고 염분분사 환경에서 가볍게 냉연 가공된 시편의 SSRT 시험으로부터 얻은 항복강도와 극한 인장응력의 값은 공기 중의 값보다 약간 낮았다. 그러나 염분 분위기에서 부식으로 인한 20% 감소 냉연시편의 강도는 더 이상 변화하지 않았다. 예비결과는 냉연 304 스테인레스 강의 질과 성능이 건식저장용기의 제작을 위한 조건에 맞는다는 것을 증명하였다. 그러나 냉연 스테인레스 강의 장기적인 성능을 더 잘 이해하기 위해서는 염분분위기에서 이 재질의 부식거동에 관한 더 많은 연구가 필요하다.

• Nuclear Engineering and Technology
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• 제45권7호
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• pp.969-978
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• 2013

A full-sized model for the horizontally oriented metal cask containing 21 spent fuel assemblies has been considered to evaluate the internal natural convection behavior within a dry shield canister (DSC) filled with helium as a working fluid. A variety of two-dimensional CFD numerical investigations using a turbulent model have been performed to evaluate the heat transfer characteristics and the velocity distribution of natural convection inside the canister. The present numerical solutions for a range of Rayleigh number values ($3{\times}10^6{\sim}3{\times}10^7$) and a working fluid of air are further validated by comparing with the experimental data from previous work, and they agreed well with the experimental results. The predicted temperature field has indicated that the peak temperature is located in the second basket from the top along the vertical center line by effects of the natural convection. As the Rayleigh number increases, the convective heat transfer is dominant and the heat transfer due to the local circulation becomes stronger. The heat transfer characteristics show that the Nusselt numbers corresponding to $1.5{\times}10^6$ < Ra < $1.0{\times}10^7$ are proportional to 0.5 power of the Rayleigh number, while the Nusselt numbers for $1.0{\times}10^7$ < Ra < $8.0{\times}10^7$ are proportional to 0.27 power of the Rayleigh number. These results agreed well with the trends of the experimental data for Ra > $1.0{\times}10^7$.

• 콘크리트학회논문집
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• 제26권1호
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• pp.71-78
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• 2014

사용후핵연료의 중간저장시설인 콘크리트 캐스크(cask)는 해안부근에 입지할 가능성이 크기 때문에 염해에 대한 문제가 크게 우려된다. 그리고 염해에 의한 철근의 부식 및 균열발생은 철근콘크리트구조물의 방사선 차폐기능뿐 아니라 구조성능 저하의 주요 원인이기 때문에 염해에 대한 평가는 매우 중요한 사항이다. 특히 염해환경과 함께 콘크리트 캐스크 내부에서는 사용후핵연료의 발열에 의해 $60^{\circ}C$정도의 고온 환경이 예상되기 때문에 고온에서의 염해에 대한 검토가 요구되지만, 기존 콘크리트 구조물의 염해평가에서는 온도에 대한 영향이 전혀 고려되어 있지 않아 고온에서 염해에 노출된 철근콘크리트구조물들의 내구설계 및 수명예측을 위해 참고할 만한 자료가 거의 없다. 이 연구에서는 다양한 온도환경에서의 염수(NaCl)침지시험을 통해 콘크리트의 염화물이온 확산계수를 측정하고 염화물이온 확산계수와 온도의 관계를 규명하고자 하였다. 실험 결과, 콘크리트의 염화물이온 확산계수는 온도의 증가에 따라 현저히 증대하여 고온환경에서의 염해 발생가능성이 매우 큰 것으로 조사되었다. 또한 크리트의 염화물이온 확산계수는 물시멘트(W/C)비가 낮아질수록 감소하였고, 이 경향은 온도가 증가(고온환경)하여도 동일하게 나타났다. 염화물이온 확산계수의 온도의존성은 아레니우스식(Arrhenius equation)으로 나타내어졌고 회귀분석 결과, 확산계수의 대수 값은 절대온도의 역수와 선형관계를 나타내었다. 또한 온도의존성을 나타내는 활성화에너지(activation energy)는 물시멘트(W/C)비가 낮을수록 높게 나타났다.

• Nuclear Engineering and Technology
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• 제41권2호
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• pp.179-186
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• 2009

This paper presents a thermal creep test under internal pressure and post-test characterization performed on high burnup (68 MWd/kgU) ZIRLO. This research has been done by the CSN, ENRESA, and ENUSA in order to investigate the behavior of advanced cladding materials in contemporary PWRs at higher burnup under dry cask storage conditions. Also, to investigate the hydride reorientation, the cool-down of the samples after the test has been done in a coordinated manner with the internal pressure. The creep results obtained are consistent with the expected behavior from reference CWSR material, Zr-4. During the test, the material retained significant ductility: one specimen leaked during the test at an engineering strain of the tube section of 17%; remarkably, the crack closed due to de-pressurization. Although significant hydride reorientation occurred during the cool-down under pressure, no specimen failed during the cool-down.

• 방사성폐기물학회지
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• 제19권4호
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• pp.435-446
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• 2021

This article examines the consequences of a significant spent fuel management decision or event in the United States, South Korea and Taiwan. For the United States, it is the financial impact of the Department of Energy's inability to take possession of spent fuel from commercial nuclear power companies beginning in 1998 as directed by Congress. For South Korea, it is the potential financial and socioeconomic impact of the successful construction, licensing and operation of a low and intermediate level waste disposal facility on the siting of a spent fuel/high level waste repository. For Taiwan, it is the operational impact of the Kuosheng 1 reactor running out of space in its spent fuel pool. From these, it draws six broad lessons other countries new to, or preparing for, nuclear energy production might take from these experiences. These include conservative planning, treating the back-end of the fuel cycle holistically and building trust through a step-by-step approach to waste disposal.

• Journal of Radiation Protection and Research
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• 제40권2호
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• pp.92-100
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• 2015

경수로 사용후핵연료 건식 중간저장시설의 방사선영향평가 효율성 개선을 목적으로 '선원항 지정방법에 따른 민감도 평가', '2-Step 계산'기법 개발 및 '냉각기간 이득효과' 적용에 따른 방사선 영향평가를 수행하였다. 본 연구에서는 저장건물의 용기배열 순서에 따라 순차적으로 선원항을 지정하여 직접선량에 미치는 민감도를 평가하였으며, 차폐건물 외벽에서의 방사선량은 내벽과 인접한 최근접 2개 열에 의한 영향이 지배적임을 확인하였다. 또한, 저장시설에 차폐 건물이 도입될 경우, 막대한 전산해석 시간을 감소시키기 위해 '2-Step 계산'기법을 수립하여 평가한 결과는 절반가량의 해석시간으로 직접(1-Step) 계산결과와 유사한 결과를 도출하였다. 마지막으로, 저장시설에 순차적으로 저장되는 저장용기의 보관기간을 사용후핵연료의 실제 냉각기간을 적용하면 건물 외벽에서의 방사선량이 냉각기간을 모두 동일하게 설정한 계산값에 비해 40% 정도 낮게 평가됨을 확인하였다. 본 연구는 중간저장시설의 방사선 영향평가를 위한 몬테칼로 차폐해석 방법의 효율성을 향상시키고자 수행되었으며, 좀 더 다양한 사례에 대한 평가를 통하여 신뢰성을 향상시킨다면 저장시설의 설계 및 부지경계 기준설정에 활용할 수 있을 것이다.

• Nuclear Engineering and Technology
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• 제55권2호
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• pp.690-695
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• 2023

Recommendations are given to improve the efficiency of radiation protection of transport casks for SNF transportation. The attenuation of ${\gamma}$-quanta of long-lived isotopes ¹³⁴Cs, ^137mBa(¹³⁷Cs), ¹⁵⁴Eu and ⁶⁰Co by optimizing the thicknesses and arrangement of layers of Fe and Pb radiation shields of transport casks is studied. The fixed radiation shielding mass (fixed mass thickness) is chosen as the main optimization criterion. The effect of the placement order of Fe and Pb layers in a combined two-layer radiation shield with an equivalent thickness of 30 cm is studied in detail. It is shown that with the same mass thicknesses of the Fe and Pb layers, the placement of Fe in the first layer, and Pb - in the second one provides more than twofold attenuation of ${\gamma}$-quanta compared to the reverse placement: Pb - in the first layer, Fe - in the second. The increase in the efficiency of attenuation of ${\gamma}$-quanta for TC with combined shielding of Fe and Pb is shown to be achieved by designing the first layer of radiation shielding around the canister with SNF from Fe of the maximum possible thickness.

• Nuclear Engineering and Technology
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• 제55권3호
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• pp.1009-1014
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• 2023

Muon tomography is a useful method for monitoring special nuclear materials (SNMs) such as spent nuclear fuel inside dry cask storage. Multiple Coulomb scattering of muons can be used to provide information about the 3-dimensional structure and atomic number(Z) of the inner materials. Tomography using muons is less affected by the shielding material and less harmful to health than other measurement methods. We developed a muon detector for muon tomography, which consists of a plastic scintillator, 64 long wavelength-shifting (WLS) fibers attached to the top of the plastic scintillator, and silicon photomultipliers (SiPMs) connected to both ends of each WLS fiber. The muon detector can acquire X and Y positions simultaneously using a position determination algorithm. The design parameters of the muon detector were optimized using DETECT2000 and Geant4 simulations, and a muon detector prototype was built based on the results. Spatial resolution measurement was performed using simulations and experiments to evaluate the feasibility of the muon detector. The experimental results were in good agreement with the simulation results. The muon detector has been confirmed for use in a muon tomography system.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1439-1447
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• 2023

The fuel cycle characteristics of the IVG.1M reactor were studied within the framework of the research reactor conversion program to modernize the IVG.1M reactor. Optimum use of the nuclear fuel and reactor was achieved through routine methods which included partial fuel reloading combined with scheduled maintenance operations. Since, the additional problem in planning the fuel cycle of the IVG.1M reactor was the poisoning of the beryllium parts of the core, reflector, and control system. An assessment of the residual power and composition of spent fuel is necessary for the selection and justification of the technology for its subsequent management. Computational studies were performed using the MCNP6.1 program and the neutronics model of the IVG.1M reactor. The proposed scheme of annual partial fuel reloading allows for maintaining a high reactor reactivity margin, stabilizing it within 2-4 β_eff for 20 years, and achieving a burnup of 9.9-10.8 MW × day/kg U in the steady state mode of fuel reloading. Spent fuel immediately after unloading from the reactor can be placed in a transport packaging cask for shipping or safely stored in dry storage at the research reactor site.