• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.921-928
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• 2013

A future nuclear energy system is being developed at Korea Atomic Energy Research Institute (KAERI), the system involves a Sodium Fast Reactor (SFR) linked with the pyro-process. The pyro-process produces a source material to fabricate a SFR fuel rod. Therefore, an isotopic fissile content assay is very important for fuel rod safety and SFR economics. A new technology for an analysis of isotopic fissile content has been proposed using a lead slowing down spectrometer (LSDS). The new technology has several features for a fissile analysis from spent fuel: direct isotopic fissile assay, no background interference, and no requirement from burnup history information. Several calculations were done on the designed spectrometer geometry: detection sensitivity, neutron energy spectrum analysis, neutron fission characteristics, self shielding analysis, and neutron production mechanism. The spectrum was well organized even at low neutron energy and the threshold fission chamber was a proper choice to get prompt fast fission neutrons. The characteristic fission signature was obtained in slowing down neutron energy from each fissile isotope. Another application of LSDS is for an optimum design of the spent fuel storage, maximization of the burnup credit and provision of the burnup code correction factor. Additionally, an isotopic fissile content assay will contribute to an increase in transparency and credibility for the utilization of spent fuel nuclear material, as internationally demanded.

• 한국전산유체공학회:학술대회논문집
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• 2008.03b
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• pp.264-267
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• 2008

Dual-cooled fuel with inner and outer flow channel was proposed for high burup, next generation nuclear fuel design. The annular cylinder of dual cooled fuel has higher structural strength compared to the conventional one, but also have concerns about flow induced vibration due to an additional flow of inner channel and the difference of flow velocity in between inner and outer channel. In this study, the dynamic stability of flexible, annular cylinder was evaluated according to the flow variation and compared to the that of the conventional PWR fuel rod. Centrifugal and Coriolis force by the additional flow in the inner channel were added in the dynamic equation of flexible beam in uniform, external, and axial flow. Complex eigenfrequency was calculated by the finite element method. Stability margin of annular cylinder compared to the solid cylinder and change of the dynamic characteristic are presented and discussed as a analysis results.

• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.4
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• pp.287-297
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• 2014

For several decades, many countries operating nuclear power plants have been studying the various disposal alternatives to dispose of the spent nuclear fuel or high-level radioactive waste safely. In this paper, as a direct disposal of spent nuclear fuels for deep geological disposal concept, the rod consolidation from spent fuel assembly for the disposal efficiency was considered and analyzed. To do this, a concept of spent fuel rod consolidation was described and the related concepts of disposal canister and disposal system were reviewed. With these concepts, several thermal analyses were carried out to determine whether the most important requirement of the temperature limit for a buffer material was satisfiedin designing an engineered barrier of a deep geological disposal system. Based on the results of thermal analyses, the deposition hole distance, disposal tunnel spacing and heat release area of a disposal canister were reviewed. And the unit disposal areas for each case were calculated and the disposal efficiencies were evaluated. This evaluation showed that the rod consolidation of spent nuclear fuel had no advantages in terms of disposal efficiency. In addition, the cooling time of spent nuclear fuels from nuclear power plant were reviewed. It showed that the disposal efficiency for the consolidated spent fuel rods could be improved in the case that cooling time was 70 years or more. But, the integrity of fuels and other conditions due to the longer term storage before disposal should be analyzed.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2001.04a
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• pp.651-654
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• 2001

In the dismantling process of nuclear spent fuels, the spent fuel rod cutting process, followed immediately by the decladding process, performs the cutting the spent fuel rods to a proper length for fast decladding operation. In this paper, we analyzed the chemical compositions, mechanical properties, and physical characteristics for SUS and zircaloy tubes in order to identify the feasibility of cutter-blade type in cutting SUS and zircaloy tubes. It is considered that material, shape and angle, and heat treatment for fabricating the highly durable cutter blade and also it is investigated that the round-shape sustenance of cross-section, amount of debris production, and fire occurrence for measuring the cutting performance on SUS and zircaloy tubes, spent fuel rod cutting device is designed to be operated automatically through the remote control system for use in Hot Cell(radioactive) area and the electro-driven mechanical parts are modularized for easy maintenance. Results from various experiments confirm the efficiency of this device.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.2
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• pp.163-170
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• 2023

Since spent nuclear fuel assemblies (SFA) are transported to interim storage or final disposal facility after cooling the decay heat, finite element analysis (FEA) with simplification is widely used to show their integrity against cladding failure to cause dispersal of radioactive material. However, there is a lack of research addressing the comprehensive impact of shape and element simplification on analysis results. In this study, for the optimization of a typical pressurized water reactor SFA, different types of finite element models were generated by changing number of fuel rods, fuel rod element type and assembly length. A series of FEA in use of these different models were conducted under a shock load data obtained from surrogate fuel assembly transportation test. Effects of number of fuel rods, element type and length of assembly were also analyzed, which shows that the element type of fuel rod mainly affected on cladding strain. Finally, an optimal finite element model was determined for other practical application in the future.

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1485-1494
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• 2023

The bow of the nuclear fuel assembly is a well-known phenomenon. One of the vital criteria during the history of nuclear fuel development has been fuel assembly's mechanical stability. Once present, the fuel assembly bow can lead to safety issues like excessive water gap and power redistribution or even incomplete rod insertion (IRI). The extensive bow can result in assembly handling and loading problems. This is why the fuel assembly's bow is one of the most often controlled geometrical factors during periodic fuel inspections for VVER when compared e.g. to on-site fuel rod gap measurements or other instrumental measurements performed on-site. Our proposed screening method uses existing video records for fuel inspection. We establish video frames normalization and aggregation for the purposes of bow measurement. The whole process is done by digital image processing algorithms which analyze rotations of video frames, extract angles whose source is the fuel set torsion, and reconstruct torsion schema. This approach provides results comparable to the commonly utilized method. We tested this new approach in real operation on 19 fuel assemblies with different campaign numbers and designs, where the average deviation from other methods was less than 2 % on average. Due to the fact, that the method has not yet been validated during full scale measurements of the fuel inspection, the preliminary results stand for that we recommend this method as a complementary part of standard bow measurement procedures to increase measurement robustness, lower time consumption and preserve or increase accuracy. After completed validation it is expected that the proposed method allows standalone fuel assembly bow measurements.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.304.2-304.2
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.227-227
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• 1998

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.212.1-212
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• 2002