• Nuclear Engineering and Technology
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• 제56권6호
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• pp.1975-1988
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• 2024

It is important to maintain cladding integrity in spent nuclear fuel management. This study proposes a numerical analysis method to evaluate the fracture resistance of irradiated zirconium alloy cladding under pinch load known to cause Mode-III failure. The mechanical behavior and fracture of the cladding under pinch loading can be evaluated by a Ring Compression Test (RCT). To simulate the fracture of hydride precipitates, zirconium matrix, and Zr/hydride interfaces under the stress field generated by RCT, a micro-structure crack propagation simulation method based on Continuum Damage Mechanics (CDM) has been proposed. Our RCT simulation model was constructed from microscopic images of irradiated cladding. In this study, we developed an automated process to generate a pixel-based finite element model by separating the hydride precipitates, zirconium matrix, and interfaces using an image segmentation method. The appropriate element size was selected to ensure the efficiency and accuracy of a crack propagation simulation. The load-displacement curves and strain energies from RCT were compared and analyzed with the simulation results of different element sizes. The finalized RCT simulation model can be used to establish the failure criterion of fuel rods under pinch loading. The advantages and limitations of the proposed method are fully discussed here.

• Nuclear Engineering and Technology
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• 제30권6호
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• pp.507-517
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• 1998

Unlike conventional fuel types, fuel constituent redistribution and sodium intrusion into the fuel slug are the unique phenomena of the irradiated metallic fuel. A thermal calculation model on metallic U-10 wt.%Zr fuel rod for LMRs is developed with considerations given to these phenomena. The amount of constituent redistribution is estimated based on the thermotransport process. The temperature profile of fuel slug is predicted by taking into account of Zr redistribution, porosity formation and sodium logging effects. A sample calculation is performed and compared to experimental data in literature. As a result, the predicted redistribution and temperature profile are well agreed with experimental data, assuming that 15 times increment of ex-reactor diffusivity, $Q_{r}$ $^{*}$ is -50 kJ/mole and sodium is infiltrated only outside of the fuel slug. Furthermore, the redistribution effects on fuel integrity and fuel temperature profile are discussed.d.

• 방사성폐기물학회지
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• 제2권2호
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• pp.125-133
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• 2004

최대 선출력 61 ㎾/m 및 평균 연소도 1,770 ㎿d/tU의 조건으로 하나로에서 조사한 DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) 핵 연료를 EPMA (Electron Probe Micro Analyzer)를 이용하여 핵분열 생성물을 분석하였다. EPMA의 정확한 분석 방법을 확립하고자, 핵분열생성물 대신 시약을 첨가하여 제조한 모의 DUPIC 핵연료로 EPMA 분석을 수행하였고, 그 결과를 습식 화학 분석의 결과와도 비교하여 평가하였다. 모의 DUPIC 핵연료 중심부의 금속 석출물은 약 1 $\mu\textrm{m}$ 정도의 크기로 관찰되었으며, 이들의 조성은 Mo-53.89 at.%, Ru-37.40 at.% 및 Pd+Rh-8.71 at%이었다. 모의 DUPIC 핵연료 시험에서 정립한 시험방법으로 조사한 DUPIC 핵연료 시편의 금속 석출물 특성을 분석하였다. 핵연료 중앙부에서 관찰된 금속 석출물들의 크기는 2∼2.5 $\mu\textrm{m}$ 정도이었으며, Mo-47.34 at.%, Ru-46 at.%, Pd+Rh-6.65 at.%의 조성임을 확인하였다. 이 실험을 위하여, 특별히 시료의 전도성을 향상시키기 위한 처리를 하였으며, 작은 금속 석출물에 EPMA의 전자빔을 정확히 조사할 수 있는 실험 조건을 제시하였다.

• Journal of Radiation Protection and Research
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• 제13권2호
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• pp.88-94
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• 1988

조사된 경수로 핵연료의 도로 수송시, 방사선 영향을 INTERTRAN 코드로 평가하였다. 계산된 결과는 인구밀도로 구분한 각 지역의 소집단과 작업자에 따라 집단 선량당량으로 구하였다. 정상 수송과 수송 사고에 대한 평가 결과, 각 집단의 선량은 매우 낮은 수준을 나타내었다. 계산 프로그램은 다양한 입력 자료에 의하여 정량화 되므로 앞으로 입력 자료의 확보를 위한 연구가 필요하다.

• 한국조명전기설비학회:학술대회논문집
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• 한국조명전기설비학회 2009년도 춘계학술대회 논문집
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• pp.243-247
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• 2009

A Radioactive constraint of the nuclear fuel assembly irradiated by neutron during normal operation cycle of the nuclear power plant is typically on the order of about 3 kGy/h. In order to inspect nuclear fuel assembly using a VT (vision technology) system, the light such as halogen lamp is used. As the halogen lamp has lower color temperature than the sun light, the objects under halogen lamp illumination are seemed to be tinted with red. In this paper, high brightness LED is considered to be used as the light source of VT system. The high brightness LED, which is a key light source of the VT system, have been gamma irradiated at the dose rate of 4 kGy/h during two hours up to a total dose of 8 kGy. The radiation induced color-center in the LED housing cap made of plastics materials is observed.

• Nuclear Engineering and Technology
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• 제43권6호
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• pp.489-498
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• 2011

UK National Nuclear Laboratory's (NNL's) version of the ENIGMA fuel performance code is described, including details of the development history, the system modelled, the key assumptions, the thermo-mechanical solution scheme, and the various incorporated models. The recent development of ENIGMA in the areas of whole core analysis and dry storage applications is then discussed. With respect to the former, the NEXUS code has been developed by NNL to automate whole core fuel performance modelling for an LWR core, using ENIGMA as the underlying fuel performance engine. NEXUS runs on NNL's GEMSTONE high performance computing cluster and utilises 3-D core power distribution data obtained from the output of Studsvik Scandpower's SIMULATE code. With respect to the latter, ENIGMA has been developed such that it can model the thermo-mechanical behaviour of a given LWR fuel rod during irradiation, pond cooling, drying, and dry storage - this involved: (a) incorporating an out-of-pile clad creep model for irradiated Zircaloy-4; (b) including the ability to simulate annealing out of the clad irradiation damage; (c) writing of additional post-irradiation output; (d) several other minor modifications to allow modelling of post-irradiation conditions.

• Nuclear Engineering and Technology
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• 제50권7호
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• pp.1190-1197
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• 2018

Analysis of the GRSW-A model coupled into the FALCON code is extended by simulation of central void formation in fuel pellets due to high-temperature fuel restructuring. The extended calculation is verified against published, well-known experimental data. Good agreement with the data for a central void diameter in pellets of the rod irradiated in an Experimental Breeder Reactor is shown. The new calculation methodology is employed in comparative analysis of modern BWR fuel behavior under assumed high-power operation. The initial fuel porosity is shown to have a major effect on the predicted central void diameter during the operation in question. Discernible effects of a central void on peak fuel temperature and Pellet-Cladding Mechanical Interaction (PCMI) during a simulated power ramp are shown. A mitigating effect on PCMI is largely attributed to the additional free volume in the pellets into which the fuel can creep due to internal compressive stresses during a power ramp.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3741-3758
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• 2021

Outcomes of the FALCON code analysis-related part of the STARS-ENSI Service Project on Evaluation of the new U.S.NRC RIA Fuel Safety Criteria and Application to the Swiss Reactors are presented. Substantial conservatism of the updated safety limits for high-temperature and PCMI cladding failure, as proposed in the NRC Regulatory Guide RG 1.236, is confirmed. Applicability of the updated failure limits to fuel safety analysis in the Swiss PWRs, as applied to standard fuel designs using UO2 fuel pellets and SRA Zry-4 as cladding materials is discussed. Conducting of new integral RIA tests with irradiated samples using doped- and gadolinia fuel pellets to support appropriate fuel safety criteria for RIA events is recommended.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.273-273
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• 2001

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 춘계학술발표대회 요약집
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• pp.263.2-263.2
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• 2003