• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1148-1153
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• 2018

The Leaking Fuel Experiment test facility was designed to simulate the activity release from spent leaking fuel rods under steady state and transient conditions in the spent fuel pool. The experimental rig included an electrically heated fuel rod with different defects and a cooling system. The fission product transport was simulated by potassium-chloride. The conductivity changes of the water in the cooling system were measured to provide information about the amount of released solution. Defects of different sizes and positions were applied, together with a wide range of rod powers to simulate decay heat. The produced data can be used for predicting the activity release from leaking fuel under storage conditions and for the interpretation of fuel examination procedures.

• Nuclear Engineering and Technology
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• v.41 no.2
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• pp.187-198
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• 2009

In this paper, three power ramp tests performed on high burn-up Re-crystallized Zircaloy2 - UO2 BWR fuel rods (56 to 63 MWd/kgU) within the SCIP project are simulated with METEOR and ALCYONE 3D. Two of the ramp tests are of staircase type up to Linear Heat Rates of 420 and 520 W/cm and with long holding periods. Failure of the 420 W/cm fuel rod was observed after 40 minutes. The third ramp test consisted of a more standard ramp test with a constant power rate of 80 W/cm/min up to 410 W/cm with a short holding time. The tests were first simulated with the METEOR 1D fuel rod code, which gave accurate results in terms of profilometry and fission gas releases. The behaviour of a fuel pellet fragment and of the cladding piece on top of it was then investigated with ALCYONE 3D. The size and the main characteristics of the ridges after base irradiation and power ramp testing were recovered. Finally, the failure criteria validated for PWR conditions and fuel rods with low-to-medium burn-ups were used to analyze the failure probability of the KKL rodlets during ramp testing.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.375-388
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• 2024

In this study, to evaluate the shock and vibration load characteristics of used fuel, a sea transportation test was conducted using simulated fuel assemblies under normal transport conditions. An overall test data analysis was performed based on the measured strain and acceleration data obtained from cruise, rotation, acceleration, braking, depth of water, and rolling tests. In addition, shock response spectrum and power spectral densities were obtained for each test case. Amplification and attenuation characteristics were investigated based on the load path. The load was amplified as it passed from the overpack to the simulated used fuel-assembly. As a result of the RMS trend analysis, the fuel-loading position of the transportation package affected the measured strain in the fuel rod, and the maximum strains were obtained at the spans with large spacing. However, even these maximum strains were very small compared to the fatigue strength and the cladding yield strength. Moreover, the fuel rods located on the side exhibited a larger strain value than those at the center.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.33-38
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• 1997

Criticality experiments with fixed neutron poison plates for water moderated and reflected low enriched(2.35 and 4.31 wt%) UO$_2$fuel rod clusters were evaluated to validate calculation techniques employed in analyzing fuel shipping and storage systems having steel, boral, or cadmium shield. Measurements were obtained for both the 2.35 wt% and the 4.31 wt% enriched rods in square pitched, water flooded lattices. The critical experiments with the 2.35 wt% enriched rods consists of three 20$\chi$ 16 or 20$\chi$ 17 fuel cluster. Critical separation were used in the experiments with the 4.31 wt% enriched fuel rods. In the experiments, the poison plates were placed on both sides of the centrally located fuel cluster. Critical separation between the three sub-critical fuel clusters were then measured for varying plate thicknesses and distances of the plates to the center fuel cluster. Calculations were performed for thirty eight critical configuration using KENO-V. a and MCNP. All of the results were within 1.23% in $\Delta$k when individually compared with the critical value of 1.0. Discrepancies of the code results are probably due to uncertainties in experiments and/or analytical modeling experiments. In general, MCNP predictions were observed to be in best agreement with the experiments.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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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.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2007.06a
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• pp.537-538
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• 2007

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2002.05a
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• pp.684-687
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• 2002

The authors have developed two devices which cuts the spend fuel rod to an optimal size and extracts fuel pellet from the pieces of cut fuel rods. These devices are so important to reduce radioactive wastes that some advanced countries developed their own methods and devices. The authors have benchmarked from these methods and devices. For spent fuel rod cutting, the tube cutting method has been chosen. some mechanical properties of the fuel tube and pellet has been carefully considered for an optimal cutting size. For fuel pellet extraction, a mechanically extracting method has been adopted. The existing chemical method have turned out to be inappropriate because it produced large amount of radioactive wastes, in spite of its high fuel recovery characteristics. The developed method has an advantage that it can be applied to other fuel rods that have different shapes and sizes. The two devices are set up and operated in the hot cell where people can not go in, so that the devices have been designed to be controlled remotely and modulated for easy maintenance. And the performance of the devices has been tested by using simulated fuel rod. From the experimental results, the devices are supposed to be useful for reducing radioactive wastes.

• Journal of Mechanical Science and Technology
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• v.14 no.8
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• pp.816-822
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• 2000

In the core of the nuclear power plant of PWR, several cases of fuel failure by unknown causes have been experienced for various fuel types. From the common features of the failure pattern, failure lead time, flow conditions, and flow induced vibration characteristics in nuclear fuel bundles, it is deduced that the fretting wear failure of the fuel rod at the spacer grid position is due to the fluidelastic vibration. In the past, fluidelastic vibration was simulated by quasi -static semi-analytical model, so called the static model, which could not account for the interaction between the rods within a bundle. To overcome this defect and to provide for more flexibilities applicable to the fuel bundle, Tanaka's unsteady model was modified to accomodate the geometrical differences and governing parameter changes during the operations such as the number of rods, pitch to diameter ratio (P/D), spring force, damping coefficient, etc. The critical velocity was calculated by solving the governing equations with the MATLAB code. A comparison between the estimated critical velocity and the test result shows a good agreement. Finally, the level of decrease of the critical velocity due to the reduction in the spring force and reduced damping coefficient due to the radiation exposure is also estimated.

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 2006.05a
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• pp.141-144
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• 2006

In this study, we developed a decladding device which separates 250 mm length of simulated nuclear spent fuel rod into the pallets and the pieces of the hulls after inserting the rod cut into the module with several pairs of blades. To improve the performance of the equipment, we considered some mechanisms to prevent the rod cut from being exposed or bounced into the hot-cell, to reduce the operation time, and to insert the rods automatically. It is expected that the newly developed system will contribute to prevent radioactive pollution in the hot-cell, reduce the operation time, and to increase the safety of the operators. As a result of the performance test for some mockup fuel rod cuts in the ACP(Advanced Spent Fuel Control Process) facility, it was verified that the decladding device could be applied to the actual fuel rod cut. And it will be able to use for a scale-up facility in the future.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2003.11a
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• pp.41-49
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• 2003

Tube oscillation behaviour is experimentally investigated for the study on the fuel rod fretting that is caused by the flow-induced vibration in nuclear reactor. The experiment was conducted in all at room temperature. The specimen of tube assembly was supported by plate springs which simulated the spacer grids and fuel rods of a fuel assembly. To investigate the influence of contact condition between the grids and rods, normal load of 10 and 5 N, gaps of 0.1 and 0.3 mm were applied. The range of the oscillation at the center of the fuel rod specimen was varied as 0.2, 0.3 and 0.4 mm to simulate the fuel rod vibration due to flow. Displacements near the contact were measured with four displacement sensors during the tube oscillation. As results, the shape of oscillation (phase) varied depending on the contact condition. The oscillation displacement increased considerably from the contact to gap condition. The displacement increased further as the gap size increased. It is regarded that the spring shape influences the tube oscillation behaviour. Simple calculation showed that the slip displacement was very small. Therefore, cumulative damage concept is necessary for the fuel rod wear. The mechanism of plowing is thought required to explain the severe wear in the case of gap existence.