• 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.

• 방사성폐기물학회지
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• 제20권2호
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• pp.161-170
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• 2022

The effect of oxidation time on the characteristics and mechanical properties of spent nuclear fuel cladding was investigated using Raman spectroscopy, tube rupture test, and tensile test. As oxidation time increased, the Raman peak associated with the tetragonal zirconium oxide phase diminished and merged with the Raman peak associated with the monoclinic zirconium oxide phase near 333 cm^-1. Additionally, the other tetragonal zirconium oxide phase peak at 380 cm^-1 decreased after 100 d of oxidation, whereas the zirconium monoclinic oxide peak became the dominant peak. The oxidation time had no effect on the tube rupture pressure of the oxidized zirconium alloy tube. However, the yield and tensile stresses of the oxidized nuclear fuel cladding tube decreased after 100 d of oxidation. The results of the scanning electron microscopy and transmission electron microscopy were represented with the in-situ Raman analysis result for the oxide characteristics generated on the cladding of spent nuclear fuel.

• 한국정밀공학회지
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• 제33권4호
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• pp.279-286
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• 2016

To enhance the safety of nuclear reactors after the Fukushima accident, researchers are developing various types of accident tolerant fuel (ATF) to increase the coping time and reduce the generation of hydrogen by oxidation. Coated cladding, an ATF concept, can be a promising technology in view of its commercialization. We applied 3D printing technology to the fabrication of coated cladding as well as of coated pellets. Direct metal tooling (DMT) in 3D printing technologies can create a coated layer on the tubular cladding surface, which maintains stability during corrosion, creep, and wear in the reactor. A 3D laser coating apparatus was built, and parameter studies were carried out. To coat pellets with erbium using this apparatus, we undertook preliminary experiments involving metal pellets. The adhesion test showed that the coated layer can be maintained at near fracture strength.

• 대한금속재료학회지
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• 제48권8호
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• pp.691-698
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• 2010

A study has been carried out to evaluate the interaction behavior between a lanthanide element and clad material in order to analyze the effect of the lanthanide element on the fuel cladding chemical interaction (FCCI). A diffusion couple test between Misch metal (70Ce-30La) and ferritic-martensitic steel (Gr.92) was performed at $660^{\circ}C$, followed by a microstructural analysis of the coupled sample. The results showed that Ce in the Misch metal, rather than La, reacted with the ferritic-martensitic steel (FMS) to form an interaction layer that penetrated the clad thickness. Fe diffused outside the clad interface to form an $Fe_2Ce$ compound, leaving a depletion of Fe caused by excess diffusion as well as by the formation of Cr-rich precipitation inside the interaction layer. The rate of growth followed the cubic rate law, which indicated that Fe depletion was caused by the diffusion of Fe and that the associated Cr-rich phase formation controlled the whole diffusion process.

• 대한금속재료학회지
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• 제47권10호
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• pp.621-628
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• 2009

Thermomechanical processes were carried out to evaluate their effects on the mechanical and the microstructural property of a ferritic-martensitic steel. Modified 9Cr-1Mo steels were hot-rolled at a temperature of either $780^{\circ}C$ or $850^{\circ}C$ after normalizing at $1050^{\circ}C$ and then were air-cooled. Continuous annealing at $850^{\circ}C$ for 2 hours immediately after the hot rolling was also performed and they were compared to the specimens without thermomechanical process. The result showed that there were little differences between the hot rolled specimens in terms of the precipitation density and size. However, V content inside the MX precipitates increased in the case of the specimen rolled at $850^{\circ}C$. The application of the continuous annealing induced coarsening of the Nb-rich MX precipitation as well as an increase in the amount of V-rich MX precipitation, which is expected to enhance high temperature mechanical properties of the ferritic-martensitic steel.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.1-13
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• 2020

At present, the Department of Energy (DOE) in Unite State are directing the efforts of developing accident tolerant fuel (ATF) technology. As the first barrier of nuclear fuel system, the material selection of fuel rod cladding for ATFs is a basic but very significant issue for the development of this concept. The advanced cladding is attractive for providing much stronger oxidation resistance and better in-pile behavior under sever accident conditions (such as SBO, LOCA) for giving more coping time and, of course, at least an equivalent performance under normal condition. In recent years, many researches on in-plie or out-pile physical properties of some suggested cladding materials have been conducted to solve this material selection problem. Base on published literatures, this paper introduced relevant research backgrounds, objectives, research institutions and their progresses on several main potential claddings include triplex SiC, FeCrAl and MAX phase material Ti3SiC2. The physical properties of these claddings for their application in ATF area are also reviewed in thermohydraulic and mechanical view for better understanding and simulating the behaviors of these new claddings. While most of important data are available from publications, there are still many relevant properties are lacking for the evaluations.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3367-3378
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• 2021

The design phase and safety assessment of Generation IV liquid metal-cooled fast reactors calls for the improvement of fuel pin performance codes, in particular the enhancement of their predictive capabilities towards uranium-plutonium mixed oxide fuels and stainless-steel cladding under irradiation in fast reactor environments. To this end, the current capabilities of fuel performance codes must be critically assessed against experimental data from available irradiation experiments. This work is devoted to the assessment of three European fuel performance codes, namely GERMINAL, MACROS and TRANSURANUS, against the irradiation of two fuel pins selected from the SUPERFACT-1 experimental campaign. The pins are characterized by a low enrichment (~ 2 wt.%) of minor actinides (neptunium and americium) in the fuel, and by plutonium content and cladding material in line with design choices envisaged for liquid metal-cooled Generation IV reactor fuels. The predictions of the codes are compared to several experimental measurements, allowing the identification of the current code capabilities in predicting fuel restructuring, cladding deformation, redistribution of actinides and volatile fission products. The integral assessment against experimental data is complemented by a code-to-code benchmark focused on the evolution of quantities of engineering interest over time. The benchmark analysis points out the differences in the code predictions of fuel central temperature, fuel-cladding gap width, cladding outer radius, pin internal pressure and fission gas release and suggests potential modelling development paths towards an improved description of the fuel pin behaviour in fast reactor irradiation conditions.

• Nuclear Engineering and Technology
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• 제35권5호
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• pp.471-481
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• 2003

In order to evaluate the microstructural characteristics of irradiated fuel claddings (Zircaloy-4), two irradiated specimens having different burnups (18 GWD/MTU and 42 GWD/MTU) were prepared from the G23-M4 fuel rods, which were loaded in Kori Unit 1 for 4 fuel cycles. The oxide thickness, hydride morphology and hardness change were characterized by an optical microscope and a micro-hardness tester after preparing the irradiated specimens in the PIE (Post Irradiation Examination) facilities. The dislocation loops and the amorphous transformation of precipitates induced by the neutron irradiation in the nuclear plant were also examined by a TEM. As the burnup increased from 18 GWD/MTU to 42 GWD/MTU, the oxide thickness increased from $6.0{\mu}m\;to\;25.3{\mu}m$ and the contents of hydrogen pick-up in the Zr matrix also greatly increased. In the comparison of the hardness of the unirradiated fuel cladding, the amount of hardness increase was nearly $9.3\%\;and\;24.2\%$ for the irradiated samples of 18 and 42 GWD/MTU, respectively. Both -type dislocation loops and -type dislocation components were observed simultaneously in the irradiated specimens and the densities of the dislocation was increased by increasing the burnup. The precipitates in both the irradiated specimens were amorphously transformed by the neutron irradiation and the trend of the amorphous transformation of the precipitates was enhanced at a higher burnup.

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 2001년도 제34회 추계학술대회 개최
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• pp.245-251
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• 2001

A fretting wear tester is developed for experimental study on the fuel fretting problem of light water reactor. The feature of the developed tester is it can simulate the existence of gap between spring and fuel rod as well as different contacting force including the just-contact condition (0 N on the contact). Used are a servo-motor, an eccentric cylinder and lever mechanism for driving system. A spacer grid cell is constituted with four strap segments (each segment has a spring). This fretting wear tester can also be used as a fatigue tester of a spacer grid spring with the frequency of more than 10 Hz. It is required to simulate the frequency of the vibrating fuel rod due to flow-induced vibration in a reactor. In fretting wear test, up to two span-length of a fuel cladding tube can be accommodated. A specimen of cladding tube of one span-length is specially designed, which can be extended for two-span test. For .fatigue test, a device for clamping the spring fixture is installed additionally, Presently, the tester is designed for the condition of air environment and room temperature. The variation of the reciprocal distance is measured to check the stability of input force, which will be exerted to the cladding (for fretting wear. test) and the spring (for fatigue test) specimen.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 춘계공동학술발표회요약집
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• pp.242-242
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• 2002

Two MIMAS MaX fuel rods base-irradiated in a commercial PWR have been reinstrumented and irradiated at a test reactor. The fabrication data for two MOX roda are characterized together with base irradiation information. Both Rods were reinstrumented to be fitted with thermocouple to measure centerline temperature of fuel. One rod was equipped with pressure transducer for rod internal pressure whereas the other with cladding elongation detector. The post irradiation examinations for various items were performed to determine fuel and cladding in-pile behavior after base irradiation. By using well characterized fabrication and re-instrumentation data and power history, the fuel performance code, COSMOS, is verified with measured in-pile and PIE information. The COMaS code shows good agreement for the cladding oxidation and creep, and fission gas release when compared with PIE dad a after base irradiaton. Based on the re-instrumention information and power history measured in-pile, the COSMOS predicts re-instrumented in-pile thermal behaviour during power up-ramp and steady operation with acceptable accuracy. The rod internal pressure is also well simulated by COSMOS code. Therfore, with all the other verification by COSMOS code up to now, it can be concluded that COSMOS fuel performance code is applicable for the design and license for MaX fuel rods up to high burnup.