• Nuclear Engineering and Technology
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• v.28 no.1
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• pp.17-26
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• 1996

The in-reactor fuel rod support conditions against the fretting wear-induced damage can be evaluated by residual spacer grid spring deflection or rod-to-grid gap. In order to evaluate the impact of fuel design parameters on the fretting wear-induced damage, a simulation methodology of the in-reactor fuel rod supporting conditions as a function of burnup has been developed and implemented in the GRIDFORCE program. The simulation methodology takes into account cladding creep rate, initial spring deflection, initial spring force, and spring force relaxation rate as the key fuel design parameters affecting the in-reactor fuel rod supporting conditions. Based on the parametric studies on these key parameters, it is found that the initial spring deflection, the spring force relaxation rate and cladding creepdown rate are in the order of the impact on the in-reactor fuel rod supporting conditions. Application of this simulation methodology to the fretting wear-induced failure experienced in a commercial plant indicates that this methodology can be utilized as an effective tool in evaluating the capability of newly developed cladding materials and/or new spacer grid designs against the fretting wear-induced damage.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.734-743
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• 2017

In order to investigate the swelling behavior and fuel-cladding interaction mechanism of U-10Zr alloy metallic fuel at very low burnup, an irradiation experiment was concisely designed and conducted on the China Mianyang Research Reactor. Two types of irradiation samples were designed for studying free swelling without restraint and the fuel-cladding interaction mechanism. A new bonding material, namely, pure aluminum powder, was used to fill the gap between the fuel slug and sample shell for reducing thermal resistance and allowing the expansion of the fuel slug. In this paper, the concise irradiation rig design is introduced, and the neutronic and thermal-hydraulic analyses, which were carried out mainly using MCNP (Monte Carlo N-Particle) and FLUENT codes, are presented. Out-of-pile tests were conducted prior to irradiation to verify the manufacturing quality and hydraulic performance of the rig. Nondestructive postirradiation examinations using cold neutron radiography technology were conducted to check fuel cladding integrity and swelling behavior. The results of the preliminary examinations confirmed the safety and effectiveness of the design.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.824-831
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• 2024

In order to investigate the occurring processes during long-term dry storage of spent fuel assemblies, a joined project called SPIZWURZ, between the Karlsruhe Institute of Technology and the Gesellschaft für Anlagen-und Reaktorsicherheit (GRS), was started. Aim of the SPIZWURZ project is the determination and quantification of the influence of texture and elastic strain on diffusion and solubility of hydrogen in three different zirconium alloys used in western Europe during a long-term cooling transient (1 K/d) starting at 400 ℃. The strain in the cladding of an irradiated spent fuel rod shall be measured. Models predicting the formation of radial oriented hydrides will be validated, improved, and implemented in the GRS fuel rod performance code TESPA-ROD. This paper describes the SPIZWURZ project and already obtained first results.

• 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.

• Resources Recycling
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• v.26 no.5
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• pp.97-104
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• 2017

Nuclear fuel cladding tubes are manufactured through pilgering and the annealing process. In order to remove the oxidized layer and impurities on the surface of the tube, a pickling process is required. Zirconium (Zr) is dissolved in a HF and $HNO_3$ acid mixture during the process and the pickling waste acid, including the dissolved Zr, is completely discarded after neutralization. This study observes the effects of the residual impurities (Ba) in the pickling solution regenerated from the $BaF_2$ precipitation process on the waste pickling solution. In addition, the concentration of Ba and Zr for the actual nuclear fuel cladding tube process was optimized. The regenerated pickling solution was tested through a pilot plant pickling process device that simulates the commercial pickling process of nuclear fuel cladding tubes, and the pickling efficiency was analyzed through AFM analysis of the roughness of the cladding tube surface.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2016.10a
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• pp.193-194
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• 2016

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2565-2571
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• 2020

Background: Understanding the behaviour of nuclear fuel claddings by conducting burst test on single cladding tube under simulated loss-of-coolant accident conditions and developing theoretical cum empirical predictive computer codes have been the focus of several investigations. The developed burst criterion (a) assumes symmetrical deformation of cladding tube in contrast to experimental observation (b) interpolates the properties of Zircaloy-4 cladding in mixed α+β phase (c) does not account for azimuthal temperature variations. In order to overcome all these drawbacks of burst criterion, it is reasoned that artificial intelligence technique may be a better option to predict the burst parameters. Methods: Artificial neural network models based on feedforward backpropagation algorithm with logsig transfer function are developed. Results: Neural network architecture of 2-4-4-3, that is model with two hidden layers having four nodes in each layer is found to be the most suitable. The mean, maximum, and minimum prediction errors for this optimised model are 0.82%, 19.62%, and 0.004%, respectively. Conclusion: The burst stress, burst temperature, and burst strain obtained from burst criterion have average deviation of 19%, 12%, and 53% respectively whereas the developed neural network model predicted these parameters with average deviation of 6%, 2%, and 8%, respectively.

• Nuclear Engineering and Technology
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• v.41 no.4
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• pp.493-520
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• 2009

During the last four decades, 16 Pressurized Water Reactors (PWR) and 4 Pressurized Heavy Water Reactors (PHWR) have been constructed and operated in Korea, and nuclear fuel technology has been developed to a self-reliant state. At first, the PWR fuel design and manufacturing technology was acquired through international cooperation with a foreign partner. Then, the PWR fuel R&D by Korea Atomic Energy Research Institute (KAERI) has improved fuel technology to a self-reliant state in terms of fuel elements, which includes a new cladding material, a large-grained $UO_2$ pellet, a high performance spacer grid, a fuel rod performance code, and fuel assembly test facility. The MOX fuel performance analysis code was developed and validated using the in-reactor test data. MOX fuel test rods were fabricated and their irradiation test was completed by an international program. At the same time, the PWR fuel development by Korea Nuclear Fuel (KNF) has produced new fuel assemblies such as PLUS7 and ACE7. During this process, the design and test technology of fuel assemblies was developed to a self-reliant state. The PHWR fuel manufacturing technology was developed and manufacturing facility was set up by KAERI, independently from the foreign technology. Then, the advanced PHWR fuel, CANFLEX(CANDU Flexible Fuelling), was developed, and an irradiation test was completed in a PHWR. The development of the CANFLEX fuel included a new design of fuel rods and bundles.. The nuclear fuel technology in Korea has been steadily developed in many national R&D programs, and this advanced fuel technology is expected to contribute to a worldwide nuclear renaissance that can create solutions to global warming.

• Journal of the Korean Ceramic Society
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• v.51 no.5
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• pp.453-458
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• 2014

SiC and its composites have been considered for use as nuclear fuel cladding materials of pressurized light water reactors. In this study, a $SiC_f$/SiC composite as a constituent layer of SiC triplex fuel cladding was fabricated using a chemical vapor infiltration (CVI) process in which tubular SiC fiber preforms were prepared using a filament winding method. To enhance the matrix density of the composite layer, winding patterns, deposition temperature, and gas input ratio were controlled. Fiber arrangement and porosity were the main parameters influencing densification behaviors. Final density of the composites decreased as the SiC fiber volume fraction increased. The CVI process was optimized to densify the tubular preforms with high fiber volume fraction at a high $H_2$/MTS ratio of 20 at $1000^{\circ}C$; in this process, surface canning of the composites was effectively retarded.

• Korean Journal of Materials Research
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• v.9 no.10
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• pp.1000-1005
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• 1999

To investigate the effects of V and Sb on the recrystallization of Zr-0.8Sn alloy, the microstructure of heat-treated specimens was observed by optical microscope, SEM, and TEM. Microhardness tests were also carried out for the annealed specimens. From microstructural studies, the V or Sb additions were found to delay recrystallization process as well as grain growth. Especially, Sb was more effective in delaying the recrystallization. This delay of recrystallization and grain growth by V or Sb additions may be due to the interference in the movement of dislocation and crystal interface by V or Sb precipitates.