• Nuclear Engineering and Technology
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• 제38권5호
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• pp.405-410
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• 2006

Microstructural features and their related mechanical property changes in the 309L cladding and the heat affected zone (HAZ) of SA508 cl.3 steel were investigated through the use of TEM, tensile and small punch (SP) tests. The specimens were irradiated at 563 K up to the neutron fluences of $5.79{\times}10^{19}n/cm^2$ (>1MeV). The microstructure of the clad was mainly composed of a fcc ${\gamma}-phase$, a low percentage of bcc ${\delta}-ferrite$, and a brittle ${\sigma}-phase$. Along the weld fusion line there formed a heavy carbide precipitation with a width of $20{\sim}40{\mu}m$, showing preferential cracking during plastic deformation. The yield stress and ductile-to-brittle transition temperature (DBTT) of the irradiated clads increased. The origin of the hardening and the shift of the DBTT are discussed in terms of the irradiation-produced defect clusters of a fine size and brittle ${\sigma}-phase$.

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2565-2571
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• 2020

Background: Understanding the behaviour of nuclear fuel claddings by conducting burst test on single cladding tube under simulated loss-of-coolant accident conditions and developing theoretical cum empirical predictive computer codes have been the focus of several investigations. The developed burst criterion (a) assumes symmetrical deformation of cladding tube in contrast to experimental observation (b) interpolates the properties of Zircaloy-4 cladding in mixed α+β phase (c) does not account for azimuthal temperature variations. In order to overcome all these drawbacks of burst criterion, it is reasoned that artificial intelligence technique may be a better option to predict the burst parameters. Methods: Artificial neural network models based on feedforward backpropagation algorithm with logsig transfer function are developed. Results: Neural network architecture of 2-4-4-3, that is model with two hidden layers having four nodes in each layer is found to be the most suitable. The mean, maximum, and minimum prediction errors for this optimised model are 0.82%, 19.62%, and 0.004%, respectively. Conclusion: The burst stress, burst temperature, and burst strain obtained from burst criterion have average deviation of 19%, 12%, and 53% respectively whereas the developed neural network model predicted these parameters with average deviation of 6%, 2%, and 8%, respectively.

• 대한금속재료학회지
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• 제48권8호
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• pp.717-723
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• 2010

The degradation behavior of fuel cladding is a very import concern in nuclear power generation, because the operation of nuclear plants can be limited by fuel cladding degradation. In order to evaluate the hydride effect on failure of zirconium fuel claddings, a ring tensile test for the circumferential direction was carried out at room temperature for claddings having different hydride characteristics such as density and orientation; microstructural evaluation was also performed for those claddings. The circumferential failure of the claddings was promoted by increasing the hydride concentration in the matrix; however, the failure of the claddings was affected by the hydride orientation rather than by the hydride concentration in the matrix. From fracture surface observation, the cladding failure during the ring tensile test was matched with the hydride orientation.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3379-3397
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• 2021

In this study, a multi-physics modeling method was developed to analyze a nuclear fuel rod's thermo-mechanical behavior especially for high temperature anisotropic creep deformation during ballooning and burst occurring in Loss of Coolant Accident (LOCA). Based on transient heat transfer and nonlinear mechanical analysis, the present work newly incorporated the nuclear fuel rod's special characteristics which include gap heat transfer, temperature and burnup dependent material properties, and especially for high temperature creep with material anisotropy. The proposed method was tested through various benchmark analyses and showed good agreements with analytical solutions. From the validation study with a cladding burst experiment which postulates the LOCA scenario, it was shown that the present development could predict the ballooning and burst behaviors accurately and showed the capability to predict anisotropic creep behavior during the LOCA. Moreover, in order to verify the anisotropic creep methodology proposed in this study, the comparison between modeling and experiment was made with isotropic material assumption. It was found that the present methodology with anisotropic creep could predict ballooning and burst more accurately and showed more realistic behavior of the cladding.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.263-263
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• 2002

• Structural Engineering and Mechanics
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• 제20권1호
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• pp.45-67
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• 2005

This paper presents a new approach to predict the racking load-displacement response of plasterboard clad walls found in Australian light-framed residential structures under monotonic racking load. The method is based on a closed-form mathematical model, described herein as the 'Modularised' Closed-Form Mathematical model or MCFM model. The model considers the non-linear behaviour of the connections between the plasterboard cladding and frame. Furthermore, the model is flexible as it enables incorporation of different nailing patterns for the cladding. Another feature of this model is that the shape of stud deformation is not assumed to be a specific function, but it is computed based on the strain energy approach to take account of the actual load deformation characteristics of particular walls. Verification of the model against the results obtained from a detailed Finite Element (FE) model is also reported. Very good agreement between the closed form solution and that of the FE model was achieved.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3154-3165
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• 2022

During a loss-of-coolant accident (LOCA) in the pressurized water reactor (PWR), there is a possibility that high temperature and internal pressure of the fuel rods lead to ballooning of the cladding, which causes a partial blockage of flow area in a subchannel. Such flow blockage would influence the core coolant flow, thus affecting the core heat transfer during a reflooding phase and subsequent severe accident. However, most of the system analysis codes simulate the accident process based on the assumed channel blockage ratio, resulting in the fact that the simulation results are not consistent with the actual situation. This paper integrates the developed core Fuel Rod Thermal-Mechanical Behavior analysis (FRTMB) module into the self-developed severe accident analysis code ISAA. At the same time, the existing flow blockage model is improved to make it possible to simulate the change of flow distribution due to fuel rod deformation. Finally, the ISAA-FRTMB is used to simulate the QUENCH-LOCA-0 experiment to verify the correctness and effectiveness of the improved flow blockage model, and then the effect of clad ballooning on core heat transfer and subsequent parts of core degradation is analyzed.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(2)
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• pp.175-180
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• 1998

The U-Zr metallic alloy with low swelling HT9 cladding is the candidate for the KALIMER fuel rod. The fuel rod should be able to maintain the structural integrity during its lifetime in the reactor. In a typical metallic fuel rod, load is mainly applied by internal gas pressure, and the deformation is primarily caused by creep of the cladding. The three-dimensional FEM modelling of a fuel rod is important to predict the structural behavior in concept design stage. Using the ANSYS code, the 3-D structure analyses were performed for various configuration, element and loads. It has been shown that the present analysis model properly evaluate the structural integrity of fuel rod. The present analysis results show that the fuel rod is expected to maintain its structural integrity during normal operation.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.1031-1043
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• 2017

A structural model for stress distributions of coated Zircaloy subjected to realistic incore pressure difference, thermal expansion, irradiation-induced axial growth, and creep has been developed in this study. In normal operation, the structural integrity of coating layers is anticipated to be significantly challenged with increasing burnup. Strain mismatch between the zircaloy and the coated layer, due to their different irradiation-induced axial growth, and creep deformation are found to be the most dominant causes of stress. This study suggests that the compatibility of the high temperature irradiation-induced strains (axial growth and creep) between zircaloy and the coating layer and the capability to undergo plastic strain should be taken as key metrics, along with the traditional focus on chemical protectiveness.

• 소성∙가공
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• 제17권5호
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• pp.364-372
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• 2008

In this study, the circumferential creep behaviors ofpilgered advanced Zirconium alloy tubes such as Zr-Nb-O and Zr-Nb-Sn-Fe were investigated in the temperature range of $400\sim500^{\circ}C$ and in the stress range of 80$\sim$150MPa. The test results indicate that the stress exponent for the steady-state creep rate of the Zr-Nb-Sn-Fe alloy decreases with the increase of stress(from 6$\sim$7 to 4), while that of the Zr-Nb-O alloy is nearly independent of stress(5$\sim$6). The activation energy of creep deformation is found to be nearly the same as the activation energy for Zr self diffusion. This indicates that the creep deformation may be controlled by dislocation climb mechanism in Zr-Nb-O. On the other hand, the transition of stress exponent(from 6-7 to 4) in Zr-Nb Sn-Fe strongly suggests the transition of the rate controlling mechanism at high stresses. The lower stress exponent at high stresses in Zr-Nb-Sn-Fe can be explained by the dynamic deformation aging effect caused by interaction of dislocations with Sn substitutional atoms.