Currently, the Korea Atomic Energy Research Institute is conducting research on the development of technology to reduce the disposal area for SF (Spent nuclear Fuel). If the main radionuclides contained in SF can be separated and recovered according to their characteristics (long half-life, high mobility and high heat load) and uranium oxide which is expected to be the final residue, can be made into solids, the burden of the permanent disposal area of the SF will be greatly reduced. The waste form that end up in the repository must be verified for ease of manufacture and stability of the block. And, in order to increase the loading efficiency, a large block manufacturing technology is needed. This study describes the background of introducing PSA (Particle Size Analyzer) which is one of the necessary equipment for manufacturing UO2 blocks using slip casting, the method of using the equipment and performance verification of the equipment using standard samples. The particle size affects the sintering quality by the way the particles rearrange themselves during sintering. Powders of small particles are generally less free flowing and more difficult to compress, they form thin pores between the particles and sinter to higher density. In contrast, larger particle has a lower sintered density. Therefore, accurate particle size measurement and the selection of a suitable particle size are important. For this purpose, a PSA was installed in nuclear cycle experiment research center. To verify the performance of the equipment, a standard sample of 1.025 μm was analyzed. We got an average particle size of 1.0293 μm and standard deviation of 0.0668 μm. This value was within the uncertainty(±0.018 μm) of the sample’s certificate. In the future, this equipment will measure the size of UO2 (depleted uranium) powder and to produce large scale uranium oxide blocks.

NFDC (Nuclear Fuel and materials Data Center) is designated as a one of the data center of National Standard Reference Center from Ministry of Trade Industry and Energy at Dec. 30 2008. The fields of designation were nuclear fuel and energy materials. NFDC produces standard reference data of nuclear fuel and materials. To ensure reliability of experimental data uncertainty should be estimated. There are two kinds of uncertainty: A-type uncertainty from tester and B-type uncertainty from experimental equipments. To reduce the former, the measurement should be repeated for sufficient amount of times, and to reduce the latter type uncertainty all equipment have to be calibrated. In this study self calibration process of thermo-mechanical analyzer (TMA) was established to ensure the B-type uncertainty. The self calibration was performed using the standard reference material and correction factor was obtained. The correction factor was defined as the ratio of the thermal expansion value of the standard reference material reported in the certificate and the thermal expansion value measured using TMA. It is believed that the uncertainty evaluation process of TGA data developed in this study will be helpful for increasing reliability and stability evaluation of nuclear fuel and spent fuel.

When damaged nuclear fuel is stripped and re-fabricated into stabilized pellets, it is necessary to analyze the characteristics of the stabilized pellets, such as density, leaching behavior, and compressive strength, for final disposal. In this study, simulated nuclear fuel with UO2 and burn-up of 35 GWd/tU and 55 GWd/tU was used to measure the compressive strength of the stabilization pellet. In order to change the density of the sintered pellet, a sintered pellet was prepared by heat treatment at 1,550°C and 1,700°C for 6 hours in a reducing atmosphere of 4% H2/Ar. In the case of UO2, the density was 10.4 g/cm3 (94.5% of T.D.) and 10.6 g/cm3 (96.6% of T.D.) depending on the sintering temperature (1,550°C, 1,700°C). In the case of simulated fuel with a burn-up of 35 GWd/tU, the density was 8.8 g/cm3 (80.9% of T.D.) and 10.2 g/cm3 (93.6% of T.D.) depending on the sintering temperature (1,550°C, 1,700°C). In the case of simulated fuel with a burn-up of 55 GWd/tU, the density was 8.3 g/cm3 (77.0% of T.D.) and 10.0 g/cm3 (92.3% of T.D.) depending on the sintering temperature (1,550°C, 1,700°C). It was found that the compressive strength of simulated nuclear fuel decreased with increasing burn-up and increased with increasing density. In the case of UO2, the compressive strengths were 717.8 MPa and 897.4 MPa when the densities were 10.4 g/cm3 and 10.6 g.cm3, respectively. In the case of simulated nuclear fuel with a burn-up of 35 GWd/tU, the compressive strengths were 472.1 MPa and 732.3 MPa when the densities were 8.8 g/cm3 and 10.2 g/cm3. In the case of simulated nuclear fuel with a burn-up of 55 GWd/tU, the compressive strengths were 301.4 MPa and 515.5 MPa when the densities were 8.3 g/cm3 and 10.0 g/cm3, respectively.

National Standard Reference Center from Ministry of Trade Industry and Energy at Dec.30 2008. The fields of designation were nuclear fuel and energy materials. NFDC produces standard reference data of nuclear fuel and materials. To ensure reliability of experimental data uncertainty should be estimated. There are two kinds of uncertainty: A-type uncertainty from tester and B-type uncertainty from experimental equipments. To reduce the former, the measurement should be repeated for sufficient amount of times, and to reduce the latter type uncertainty all equipment have to be calibrated. In this study the uncertainty evaluation process of thermo-gravimetric analyzer (TGA) data was developed. The self calibration was performed using the standard mass and correction factor was obtained. The measurement model of oxidation was established, factors affected to uncertainty was analyzed, uncertainty of each factor using sensitivity coefficient was evaluated, combined uncertainty was calculated, and expanded uncertainty using coverage factor was calculated. It is believed that the uncertainty evaluation process of TGA data developed in this study will be helpful for increasing reliability and stability evaluation of nuclear fuel and spent fuel.

NFDC (Nuclear Fuel and materials Data Center) developed standard reference data for oxidation of HANA-6 cladding material. Thermo-gravimetric analyzer (TGA) was used to measure oxidation, and the measuring device was self-calibrated using standard materials. The oxidation amount of the HANA6 cladding was measured in an oxidizing atmosphere in the temperature range of 400 to 700°C. Through this, oxidation data, oxidation rate model equation, and graph were developed. The uncertainty factors were analyzed from the oxidation model. The expanded uncertainty of oxidation data was calculated by evaluating the uncertainty for each uncertainty factor. The oxidation data produced in this study was self-rated through deliberation by a specialized committee of NFDC and third experts. It was finally registered as a reference standard through the technical committee of the National Reference Standards Center. It is believed that the standard reference data developed in this study will be helpful for increasing reliability and stability evaluation of nuclear fuel and spent fuel.

In general, if a nuclear fuel cladding tube is damaged during reactor operation, it is called fuel failure. If the cladding tube is damaged, the function of sealing the nuclear fuel material is lost, and the fission products accumulated inside the nuclear fuel rod may leak into the coolant. The causes are the most damage caused by foreign substances in a coolant such as small iron wires, and GTRF (Gridto- Rod Wear) due to a grid, end-plug welding defect, PCMI (pellet cladding mechanical interaction), and oxidation corrosion damage. In this study, a device of simulating friction damage and debris induced damage between grid-fuel rods, which are the main causes of cladding tube damage, was developed. An air vibrator was installed as a function to induce vibration of the nuclear fuel rod. Sandpaper was installed between the grid and the fuel rod to induce friction between the grid-fuel rods. Saw teeth were installed on the grid to induce damage to foreign substances. It is believed that the simulated damaged nuclear fuel rod can be manufactured through on-study to provide the simulated damaged nuclear fuel rod necessary for the stabilization study of the damaged nuclear fuel rod.