• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4731-4742
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• 2022

RBMK fuel assemblies differ from other LWR FA due to a specific arrangement of the fuel rods, the low enrichment, and the used burnable absorber - erbium. Therefore, there is a challenge to adapt modeling tools, developed for other LWR types, to solve RBMK problems. A set of 10 different depletion simulation schemes were tested to estimate the impact on reactivity and spent fuel composition of possible SCALE code options for the neutron transport modelling and the use of different nuclear data libraries. The simulations were performed using cross-section libraries based on both, VII.0 and VII.1, versions of ENDF/B nuclear data, and assuming continuous energy and multigroup simulation modes, standard and user-defined Dancoff factor values, and employing deterministic and Monte Carlo methods. The criticality analysis with burn-up credit was performed for the SFP loaded with RBMK-1500 FA. Spent fuel compositions were taken from each of 10 performed depletion simulations. The criticality of SFP is found to be overestimated by up to 0.08% in simulation cases using user-defined Dancoff factors comparing the results obtained using the continuous energy library (VII.1 version of ENDF/B nuclear data). It was shown that such discrepancy is determined by the higher U-235 and Pu-239 isotopes concentrations calculated.

• Nuclear Engineering and Technology
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• 제52권8호
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• pp.1756-1763
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• 2020

The paper studies spark plasma sintering (SPS) of industrially used UO₂-based fuel containing integral fuel burnable absorber (IFBA) of neutrons Gd₂O₃. Densification dynamics of pristine UO₂ powder and the one added with 2 and 8 wt% of Gd₂O₃ under ultrasonication in liquid has been studied under SPS conditions at 1050, 1250, and 1450 ℃. Effect of sintering temperature on phase composition as well as on O/U stoichiometry has been investigated for UO₂ SPS ceramics. Sintering of uranium dioxide added with Gd₂O₃ yields solid solution (U,Gd)O₂, which is isostructural to UO₂. SEM with EDX and metallography were implemented to analyze the microstructure of the obtained UO₂ ceramics and composite UO₂-Gd₂O₃ one, particularly, open porosity, defects, and Gd₂O₃ distribution were studied. Microhardness, compressive strength and density were shown to reduce after addition of Gd₂O₃. Obtained results prove the hypothesis on formation of stable pores in the system of UO₂-Gd₂O₃ due to Kirkendall effect that reduces sintering efficiency. The paper expands fundamental knowledge on pros and cons of fuel fabrication with IFBA using SPS technology.

• 한국유체기계학회 논문집
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• 제10권2호
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• pp.41-45
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• 2007

The shutoff unit was designed to provide rapid insertion of neutron absorbing material into the reactor core to shutdown the reactor quickly and also to withdraw the absorber slowly to avoid a log-rate trip. Four shutoff units were installed on the HANARO reactor but the half-core test facility was equipped with one shutoff unit. The reactor trip or shutdown is accomplished by four shutoff units by insertion of the shutoff rods. The shutoff rod(SOR) is actuated by a directly linked hydraulic cylinder on the reactor chimney, which is pressurized by a hydraulic pump. The rod is released to drop by gravity, when triplicate solenoid valves are de-energized to vent the cylinder. The hydraulic pump, pipe and air supply system are provided to be similar with the HANARO reactor. The shutoff rod drops for 647mm stroke within 1.13 seconds to shut down the reactor and it is slowly inserted to the full down position, 700mm, with a damping. We have conducted the drop test of the shutoff rod in order to show the performance and the structural integrity of operating system of the shutoff unit. The present paper deals with the 647mm drop time and the withdrawal time according to variation of the pool water temperature, the water level and the core flow.

• 한국전산유체공학회지
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• 제21권4호
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• pp.33-39
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• 2016

Generally, thermal analysis of spent fuel storage cask has been conducted using the porous media and effective thermal conductivity model to simplify the structural complexity of spent fuel assemblies. As the fuel assembly is composed of two regions; active fuel region corresponding to UO2 pellets and unactive fuel region corresponding to the top and bottom nozzle, the heat transfer performance can be influenced depending on porous media application at these regions. In this study, numerical analysis on concrete storage cask of spent fuel was performed to investigate heat transfer effects for two cases; one was porous media application only to active fuel region(case 1) and the other one was porous media to whole length of fuel assembly(case 2). Using computational fluid dynamics code, the three dimensional, 1/4 symmetry model was constructed. For two cases, maximum temperatures for each component were evaluated below the allowable limits. For the case 1, maximum temperatures for fuel cladding, neutron absorber and baskets inside the canister were slightly higher than those for the case 2. In particular, even though the helium flows with low velocity due to buoyant forces occurred at the top and bottom of unactive fuel region, treating only active fuel region as the porous media was ineffective in respect of the heat removal performance of concrete storage cask, implying a conservative result.

• Nuclear Engineering and Technology
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• 제23권3호
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• pp.352-362
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• 1991

가압 경수로의 노심 설계에 있어서 제한된 우라늄 자원의 효율적인 이용을 위한 다양한 방안으로 장주기 운전, 고연소도 및 저누출 장전 모형 통을 강구하고 있는 추세이다. 이러한 노심들은 원자로 운전 주기 전반에 걸친 공간적 출력 분포 제어와 잉여 반응도 제어를 위해 가연성 독물질을 사용하고 있다. 이와 관련하여 가연성 독물질 관리의 최적화 연구가 다각도로 진행되고 있다. 본 연구에서는 1990년도부터 국내 가압 경수로에 국산 핵연료가 장전되기 시작하면서 가도리니아 독봉을 사용하고 있으며 장차 주된 가연성 독물질로 쓰일 예정이므로 이에 대해서 분석을 수행하였다. 분석 결과 가도리니아 독봉은 열중성자 흡수 단면적이 매우 큰데서 기인한 특이한 연소 특성을 보이고 있다. 특히 집합체 내에서의 가도리니아 독봉의 위치에 따라 매우 다양한 출력 분포를 보이고 있다. 이러한 다양한 출력 분포 중에서 노심의 반경 방향 첨두 출력을 가능한 낮게하는 집합체 내에서의 가도리니아봉 위치 최적 선정을 위한 방법론을 제시하였다.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1708-1714
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• 2019

The purpose of the critical evaluation of the spent fuel pool (SFP) is to verify that the maximum effective multiplication factor ($K_{eff}$) is less than the critical safety limit at 100% stored condition of the spent fuel with the maximum reactivity. At nuclear power plants, the storage standard of spent fuel, ie, the loading curve, is established to prevent criticality from being generated in SFP. Here, the loading curve refers to a graph showing the minimum discharged burnup versus the initial enrichment of spent fuel. Recently, US NRC proposed the new critical safety assessment guideline (DSS-ISG-2010-01, Revision 0) of PWR SFPs and most of utilities in US is following it. Of course, the licensed criterion of the maximum effective multiplication factor of SFP remains unchanged and it should be less than 0.95 from the 95% probability and the 95% confidence level. However, the new guideline is including the new evaluation methodologies like the application of the axial burnup profile, the validation of depletion and criticality code, and trend analysis. Among the new evaluation methodologies, the most important factor that affects $K_{eff}$ is the axial burnup profile of spent fuel. US NRC recommends to consider the axial burnup profiles presented in NUREG-6801 in criticality analysis. In this paper, criticality effect was evaluated considering three profiles, respectively: i) Axial burnup profiles presented in NUREG-6801. ii) Representative PWR axial burnup profile. iii) Uniform axial burnup profile. As the result, the case applying the axial burnup profiles presented in NUREG-6801 showed the highest $K_{eff}$ among three cases. Therefore, we need to introduce a new methodology because it can be issued if the axial burnup profiles presented in NUREG/CR-6801 are applied to the domestic nuclear power plants without any other consideration.

• 공업화학
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• 제24권3호
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• pp.227-230
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• 2013

$(U_{1-x}Nd_x)O_2$ 소결체를 $500^{\circ}C$에서 산화하여 얻은 $(U_{1-x}Ndx)_3O_8$ 분말의 상변화를 $900{\sim}1500^{\circ}C$의 공기 중에서 고온 산화 열처리를 하여 조사하였다. $1100^{\circ}C$ 이상의 온도로 산화 열처리할 경우에 Nd 농도가 높은 $(U_{1-y}Nd_y)O_{2+z}$ 상과 $U_3O_8$ 상이 생성됨을 확인하였으며, 산화 열처리 온도가 높아질수록 $(U_{1-y}Nd_y)O_{2+z}$ 상에서의 Nd 농도는 감소하였다. 산화 열처리 온도의 증가에 따라서 $U_3O_{8-w}$ 입자로부터 $(U_{1-y}Nd_y)O_{2+z}$ 입자로의 U 양이온 및 Nd 양이온이 두 입자의 계면을 통해 농도 구배에 따른 확산에 의해서 $(U_{1-y}Nd_y)O_{2+z}$ 상 내에 U의 농도는 증가하고 Nd의 농도는 감소하게 된다. 이러한 현상은 산화 열처리 온도증가에 따라서 $U_3O_8$ 상에 대한 $(U_{1-y}Nd_y)O_{2+z}$ 상의 X-선 회절피크의 적분강도비 증가와 $(U_{1-y}Nd_y)O_{2+z}$ 상의 입자가 커지는 것과 연관하여 해석할 수 있었다.

• 접착 및 계면
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• 제19권3호
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• pp.113-117
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• 2018

보론카바이드는 하드니스가 다이아몬드나 보론 나이트라이드 보단 낮지만 30 GPa이상의 높은 경도를 갖고 있으며, 높은 경도로 인해 탱크 장갑, 탄피 제조에 사용되고 있다. 또한 중성자를 흡수하는 능력이 있어 중성자 흡수제로 많이 사용되고 있어, 핵 발전 관련 사업에 활용도가 증가하고 있다. 중성자는 전자와의 상호작용이 없으며, 물질을 통과하는 과정에서도 상호작용 없이 통과하는 것으로 알려져 있다. 보론 카바이드와 함께 중성자와 상호작용이 높은 원자는 수소이며, 보론을 포함하는 수소 농도가 높은 폴리에스터, 에폭시 고분자 등이 원자력 발전 폐기물 보관을 위한 제품 제조를 위한 소재로 사용되고 있다. 본 논문에서는 보론 카바이드의 표면을 철산화물과 텅스텐으로 처리하여, 개질된 보론 카바이드와 에폭시 소재와의 상호작용을 향상시켰다. XRD, XPS를 이용하여 표면개질 되었음을 확인하였고, 처리된 보론카바이드의 함량에 따른 기계적 강도는 만능시험기(UTM)로 측정하였으며, 동역학 분석기(DMA)를 사용하여 경화물의 동적 특성을 관찰하였다.