• Nuclear Engineering and Technology
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• 제50권7호
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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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• 제41권2호
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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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• 제56권8호
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• pp.3405-3424
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• 2024

Reactivity worth of fuel rods at the JSI TRIGA research reactor was measured. Differently burned fuel rods were chosen to validate fuel burnup calculations. Two methods of measuring reactivity worth of fuel rods are used, traditional method is compared to newly introduced method using fuel rods swapping. Connection between both methods is described theoretically and the theory is validated experimentally. Fuel rod worth calculated using the newly introduced fuel rod swap method was within 1σ of worth measured using the traditional method. In addition to the recently performed experiments, weekly measurements of reactor core reactivity throughout the operational history are used for validation. The measured data were used to validate the fuel burnup and core criticality calculations. Fuel burnup calculations are performed using three different computer codes: the deterministic TRIGLAV, the Monte Carlo Serpent-2, and the hybrid RAPID. Great agreement was observed for Serpent-2 and RAPID by simulating fuel rod worth and its burnup, indicating that the fuel burnup and criticality calculations are accurate and that reactivity changes due to small burnup differences on the order of 10 pcm can be accurately simulated. In addition it was shown using ex-core detectors and large fission chamber that detector response changes due to fuel swapping are evident for fuel rod burnup differences of 20 MWd/kg. Fuel burnup calculations were further validated on excess reactivity measurements for three mixed TRIGA cores. The calculated burnup reactivity coefficient Δρ_BU using Serpent-2 and RAPID was within 1σ of the measurements, showing both codes are capable of calculating burnup for different TRIGA fuel types.

• Nuclear Engineering and Technology
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• 제56권2호
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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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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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.

• 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.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2007년도 춘계학술대회 논문집
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• pp.537-538
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• 2007

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2002년도 춘계학술대회 논문집
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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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• 제14권8호
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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.

• 한국소성가공학회:학술대회논문집
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• 한국소성가공학회 2006년도 춘계학술대회 논문집
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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.