• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.28 no.5
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• pp.186-194
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• 2016

As a turbulence-enhancing device, a mixing vane installed at a spacer grid of the fuel assembly plays a role in improving the convective heat transfer by generating either swirl flow in the subchannels or cross flow between fuel rod gaps. Therefore, both configuration and arrangement pattern of a mixing vane are important factors that determine the performance of a mixing vane. In this study, in order to examine the flow distribution features inside $5{\times}5$ fuel assembly with swirl-type mixing vanes used in benchmark calculation of OECD/NEA, simulations were conducted with commercial CFD software ANSYS CFX R.14. Predicted results were compared to data measured from MATiS-H (Measurement and Analysis of Turbulent Mixing in Subchannels-Horizontal) test facility. In addition, the effect of swirl-type mixing vanes on flow pattern inside the fuel assembly was described.

• Journal of the Korean Society for Nondestructive Testing
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• v.30 no.4
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• pp.324-329
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• 2010

Fuel rods using in nuclear power plants consist of uranium dioxide pellets enclosed in zirconium alloy(zircaloy) tubes. It is vitally important for the pellet surface to remain free from pits, cracks and chipping defects after it is loaded into the tubes to prevent local hot spots during reactor operation. This paper investigates the feasibility study for detecting surface flaws of pellets contained within nuclear fuel rod through X-ray tomography simulation. Reconstructed images used by parallel and fan-beam filtered back projection method were presented and confirmed the accessibility between simulation data and MPS(missing pellet surface) image data.

• Journal of Energy Engineering
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• v.13 no.3
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• pp.181-190
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• 2004

Computational models for analyzing the in-reactor behavior of metallic fuel pins under transient conditions in liquid-metal reactors are developed and implemented in the TRAMAC (TRAnsient thermo-Mechanical Analysis Code) for a metal fuel rod under transient operation conditions. Not only the basic models for a fuel rod performance but also some sub-models used for transient condition are installed in TRAMAC. Among the models, a fission gas release model, which takes the multi-bubble size distribution into account to characterize the lenticular bubble shape and the saturation condition on the grain boundary and the cladding deformation model have been developed based mainly on the existing models in the MAC-SIS code. Finally, cladding strains are calculated from the amount of thermal creep, irradiation creep, and irradiation swelling. The cladding strain model in TRAMAC predicts well the absolute magnitudes and gen-eral trends of their predictions compared with those of experimental data. TRAMAC results for the FH-1,2,6 pins are more conservative than experimental data and relatively reasonable than those of FPIN2 code. From the calculation results of TRAMAC, it is apparent that the code is capable of predicting fission gas release, and cladding deformation for LMR metal fuel finder transient operation conditions. The results show that in general, the predictions of TRAMAC agree well with the available irradiation data.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2771-2782
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• 2022

When assessing the radiological consequences of postulated accident scenarios, it is of primary interest to determine the amount of radioactive fission gas accumulated in the fuel rod free volume. The state-of-the-art semi-empirical approach (ANS 5.4-2010) is reviewed and compared with a mechanistic approach to evaluate the release of radioactive fission gases. At the intra-granular level, the diffusion-decay equation is handled by a spectral diffusion algorithm. At the inter-granular level, a mechanistic description of the grain boundary is considered: bubble growth and coalescence are treated as interrelated phenomena, resulting in the grain-boundary venting as the onset for the release from the fuel pellets. The outcome is a kinetic description of the release of radioactive fission gases, of interest when assessing normal and off-normal conditions. We implement the model in SCIANTIX and reproduce the release of short-lived fission gases, during the CONTACT 1 experiments. The results show a satisfactory agreement with the measurement and with the state-of-the-art methodology, demonstrating the model soundness. A second work will follow, providing integral fuel rod analysis by coupling the code SCIANTIX with the thermo-mechanical code TRANSURANUS.

• Nuclear Engineering and Technology
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• v.20 no.3
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• pp.189-196
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• 1988

This study investigates the effect of fuel pellet eccentricity on fuel rod thermal performance under the steady state condition. The governing equations in the fuel pellet and the cladding region are set up in 2-dimensional cylindrical coordinate (r, $\theta$) and are solved by finite element method. The angular-dependent heat transfer coefficient in the gap region is used in order to account for the asymmetry of gap width. Material propeties are used as a function of temperature and volumetric heat generation as a function of radial position. The results show the increase of maximum local heat flux at the cladding outer surface and the decrease of maximum and average fuel temperatures due to eccentricity. The former is expected to affect the uncertainties in the minimum DNBR calculation. The latter two are expected to reduce the possibility of fuel melting and the fuel stored energy. Also, the fuel pellet eccentricity introduces asymmetry in fuel pellet temperature and movement of the location of maximum fuel pellet temperature.

• Proceedings of the KSME Conference
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• 2003.04a
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• pp.1372-1377
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• 2003

The stabilization of propane/air lean premixed flames by a heated cylindrical rod is investigated experimentally. The flame stability limits, heat flux, surface temperatures, equivalence ratios, and mixture velocities are measured in order to understand the role of heat flux or surface temperature on the flame stabilization of lean premixed flames. The flame stability limits are lowered by a heated cylindrical rod and extended even below the flammability limit of propane/air mixture when sufficient heat flux is provided. The flame stability limit decreases with the increase of heat flux or surface temperature and decreases with the higher mixture velocity. The diameter of cylindrical rod, however, dose not significantly affect the flame stability limit. The laminar flame speed has been measured for ultra lean propane/air premixed flames. The flame stabilization by a heated cylindrical rod provides the useful tool for the measurement of flame speed under very fuel-lean conditions.

• Journal of the Korean Society for Precision Engineering
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• v.30 no.2
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• pp.223-230
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• 2013

When a new nuclear fuel is developed, an irradiation test needs to be carried out in the research reactor to analyze the performance of the new nuclear fuel. In order to check the performance of a nuclear fuel during the irradiation test in the test loop of a research reactor, sensors need to be attached in and out of the fuel rod and connect them with instrumentation cables to the measuring device located outside of the reactor pool. In particular, to check the temporary temperature change at the center of a nuclear fuel during the irradiation test, a thermocouple should be instrumented at the center of the fuel rod. Therefore, a hole needs to be made at the center of fuel pellet to put in the thermocouple. However, because the hardness and the density of a sintered $UO_2$ pellet are very high, it is difficult to make a small fine hole on a sintered $UO_2$ pellet using a simple drilling machine even though we use a diamond drill bit made by electro deposition. In this study, an automated drilling machine using a CVD diamond drill has been developed to make a fine hole in a fuel pellet without changing tools or breakage of workpiece. A sintered alumina ($Al_2O_3$) block which has a higher hardness than a sintered $UO_2$ pellet is used as a test specimen. Then, it is verified that a precise hole can be drilled off without breakage of the drill bit in a short time.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.746-751
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• 2001

Design requirements for the nuclear fuel assembly grid of the pressurized water reactor are reviewed from the mechanical/structural point of view. And mechanical/structural tests and numerical analyses on the various spacer grid candidates that has been uniquely designed by KAERI are carried out to find out their mechanical/structural performance. As a result, the results from the numerical analyses are good agreements with test results.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2009.06a
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• pp.757-758
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• 2009

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2009.10a
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• pp.413-414
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• 2009