• Journal of the Korean Ceramic Society
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• v.51 no.5
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• pp.435-441
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• 2014

Silicon carbide-based ceramics and their composites have been studied for application to fusion and advanced fission energy systems. For fission reactors, $SiC_f$/SiC composites can be applied to core structural materials. Multilayered SiC composite fuel cladding, owing to its superior high temperature strength and low hydrogen generation under severe accident conditions, is a candidate for the replacement of zirconium alloy cladding. The SiC composite cladding has to retain its mechanical properties and original structure under the inner pressure caused by fission products; as such it can be applied as a cladding in fission reactor. A hoop strength test using an expandable polyurethane plug was designed in order to evaluate the mechanical properties of the fuel cladding. In this paper, a hoop strength test of the multilayered SiC composite tube for nuclear fuel cladding was simulated using FEA. The stress caused by the plug was distributed nonuniformly because of the friction coefficient difference between the inner surface of the tube and the plug. Hoop stress and shear stress at the tube was evaluated and the relationship between the concentrated stress at the inner layer of the tube and the fracture behavior of the tube was investigated.

• 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.

• Nuclear Engineering and Technology
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• v.37 no.4
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• pp.327-362
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• 2005

The design, construction, and operation of a light water reactor (LWR) are subject to compliance with safety criteria specified for accident situations, such as loss-of-coolant accident (LOCA) and reactivity-initiated accident (RIA). Because reactor fuel is the primary source of radioactivity and heat generation, such a criterion is established on the basis of the characteristics and performance of fuel under the specific accident condition. As such, fuel behavior under accident situations impact many aspects of fuel design and power generation, and in an indirect manner, even spent fuel storage and management. This paper provides a comprehensive review of: the history of the current LOCA criteria, results of LOCA-related investigations on conventional and new classes of fuel, and status of on-going studies on high-burnup fuel under LOCA situations. The objective of the paper is to provide a better understanding of important issues and an insight helpful to establish new LOCA criteria for modem LWR fuels.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.1079-1092
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• 2020

A parametric study of several parameters relevant to design safety on the spent nuclear fuel (SNF) rod response under a drop accident is presented. In the view of the complexity of interactions between the independent safety-related parameters, a factorial design of experiment is employed as an efficient method to investigate the main effects and the interactions between them. A detailed single full-length fuel rod is used with consideration of post-irradiated fuel conditions under horizontal and vertical free-drops onto an unyielding surface using finite-element analysis. Critical drop heights and critical g-loads that yield the threshold plastic strain in the cladding are numerically estimated to evaluate the fuel rod structural resistance to impact load. The combinatory effects of four uncertain parameters (pellet-cladding interfacial bonding, material properties, spacer grid stiffness, rod internal pressure) and the interactions between them on the fuel rod response are investigated. The principal finding of this research showed that the effects of above-mentioned parameters on the load-carrying capacity of fuel rod are significantly different. This study could help to prioritize the importance of data in managing and studying the structural integrity of the SNF.

• Corrosion Science and Technology
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• v.2 no.6
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• pp.253-259
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• 2003

Zirconium based alloys have been extensively used as a cladding material for fuel rods in nuclear reactors, due to their low thermal neutron absorption cross-section, excellent corrosion resistance and good mechanical properties at high temperatures. However, a cladding material for fuel rods in nuclear reactors was contact water during long time at high-temperature, so it is necessary to improve the wear and corrosion resistance of the fuel cladding, At ambient environment, there are few data or paper on the characteristic of corrosion in chloride solution and acidic solution. The specimens used in this work are pure Zr and Zircaloy-4. Zircaloy-4 is a specific zirconium-based alloy containing, on a weight percent basis, 1.4% Sn, 0.2% Fe, 0.1% Cr. Pitting corrosion resistance of two alloys by ASTM G48 is higher than that of electrochemical method. Passive film formed on Zircaloy-4 is mainly composed of $ZrO_2$, metallic Sn, and iron species regardless of formation environments. Also, passive film formed on Zr alloys shows n-type semiconductic property on the base of Mott-Schottky plot.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.217.2-217
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.97-102
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• 1996

A methodology for estimating the number of failed fuel rods bused on the primary coolant activity in operating PWRs has been developed. This method deals with both the diffusion and the kinetic models. In case of small or medium cladding failures, the diffusion model which can consider different sizes of failure is used, whereas for large cladding failures the kinetic model is used. From the kinetic model, the release-to-birth rate ratio (R/B) is represented as a linear function of the number of failed fuel rods. This has been done by expressing the escape rate coefficient in terms of the slope of log(R/B) versus $log\;{\lambda}$. The present method has been applied to the cases of 26 cycles of several nuclear power plants for which ultrasonic testings were performed. The results show that the present method gives better predictions than the existing computer codes such as IODYNE and CADE.

• 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.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1096-1108
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• 2016

Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U-transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

• Nuclear Engineering and Technology
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• v.32 no.4
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• pp.372-378
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• 2000

The effect of alloying elements on the creep resistance of Zr alloys was investigated using thermal creep tests that were performed as a part of advanced fuel cladding development. The creep tests were conducted at 40$0^{\circ}C$ and 150 MPa for 240 hr. A statistical model was derived from the relationship between the steady-state creep rate and the content of individual alloying elements. The creep strengthening effect decreased in the following sequence : Nb, Sn, Mn, Cr, Mo, Fe and Cu. The high creep resistance of Sn and the opposite effect of Fe on zirconium alloys seem to be associated with their lowering and enhancing, respectively, the self-diffusivity of the zirconium matrix.