• Proceedings of the Korean Nuclear Society Conference
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• 1977.06a
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• pp.112-112
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• 1977

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.1-15
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• 2016

For a long time, a top priority in the nuclear industry was the safe, reliable, and economic operation of light water reactors. However, the development of accident-tolerant fuel (ATF) became a hot topic in the nuclear research field after the March 2011 events at Fukushima, Japan. In Korea, innovative concepts of ATF have been developing to increase fuel safety and reliability during normal operations, operational transients, and also accident events. The microcell $UO_2$ and high-density composite pellet concepts are being developed as ATF pellets. A microcell $UO_2$ pellet is envisaged to have the enhanced retention capabilities of highly radioactive and corrosive fission products. High-density pellets are expected to be used in combination with the particular ATF cladding concepts. Two concepts-surface-modified Zr-based alloy and SiC composite material-are being developed as ATF cladding, as these innovative concepts can effectively suppress hydrogen explosions and the release of radionuclides into the environment.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1140-1151
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• 2023

Fully Ceramic Microencapsulated (FCM) fuel is emerging advanced fuel material for the future nuclear reactors. The fuel pellet in the FCM fuel is composed of matrix and a large number of TRistructural-ISOtopic (TRISO) fuel particles which are randomly dispersed in the SiC matrix. The minimum layer thickness in a TRISO fuel particle is on the order of 10^-5 m, and the length of the FCM pellet is on the order of 10^-2 m. Hence, the heat transfer in the FCM pellet is a multi-scale phenomenon. In this study, three multi-scale heat conduction models including the Multi-region Layered (ML) model, Multi-region Non-layered (MN) model and Homogeneous model for FCM pellet were constructed. In the ML model, the random distributed TRISO fuel particles and coating layers are completely built. While the TRISO fuel particles with coating layers are homogenized in the MN model and the whole fuel pellet is taken as the homogenous material in the Homogeneous model. Taking the results by the ML model as the benchmark, the abilities of the MN model and Homogenous model to predict the maximum and average temperature were discussed. It was found that the MN model and the Homogenous model greatly underestimate the temperature of TRISO fuel particles. The reason is mainly that the conventional equivalent thermal conductivity (ETC) models do not take the internal heat source into account and are not suitable for the TRISO fuel particle. Then the improved ETCs considering internal heat source were derived. With the improved ETCs, the MN model is able to capture the peak temperature as well as the average temperature at a wide range of the linear powers (165 W/cm~ 415 W/cm) and the packing fractions (20%-50%). With the improved ETCs, the Homogenous model is better to predict the average temperature at different linear powers and packing fractions, and able to predict the peak temperature at high packing fractions (45%-50%).

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2990-3002
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• 2021

The mechanical interaction between the fuel pellet and the cladding tube of a nuclear fuel rod is a very important for safety studies as this phenomenon could lead to fuel failure and release of radioactivity. To investigate the ductility of cladding tubes used in WWER type nuclear power plants, several mandrel tests were performed in the Centre for Energy Research (EK). This modified mandrel test was used to model the mechanical interaction between the fuel pellet and the cladding using a segmented tool. The tests were conducted at room temperature and at 300 ℃ with inactive as-received and hydrogenated cladding ring samples. The results show a gradual decrease in ductility as the hydrogen content increases, the ductile-brittle transition was seen above 1500 ppm hydrogen absorbed.

• Nuclear Engineering and Technology
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• v.20 no.3
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• pp.189-196
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• 1988

This study investigates the effect of fuel pellet eccentricity on fuel rod thermal performance under the steady state condition. The governing equations in the fuel pellet and the cladding region are set up in 2-dimensional cylindrical coordinate (r, $\theta$) and are solved by finite element method. The angular-dependent heat transfer coefficient in the gap region is used in order to account for the asymmetry of gap width. Material propeties are used as a function of temperature and volumetric heat generation as a function of radial position. The results show the increase of maximum local heat flux at the cladding outer surface and the decrease of maximum and average fuel temperatures due to eccentricity. The former is expected to affect the uncertainties in the minimum DNBR calculation. The latter two are expected to reduce the possibility of fuel melting and the fuel stored energy. Also, the fuel pellet eccentricity introduces asymmetry in fuel pellet temperature and movement of the location of maximum fuel pellet temperature.

• Nuclear Engineering and Technology
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• v.40 no.1
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• pp.1-8
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• 2008

One of the major current challenges to nuclear energy lies in its competitiveness. To stay competitive the industry needs to reduce maintenance and fuel cycle costs, while enhancing safety features. Extended burnup is one of the methods applied to meet these objectives However, there are a number of potential fuel failure causes related to increased burnup, as follows: l) Corrosion of zirconium alloy cladding and the water chemistry parameters that enhance corrosion; 2) Dimensional changes of zirconium alloy components, 3) Stresses that challenge zirconium alloy ductility and the effect of hydrogen (H) pickup and redistribution as it affects ductility, 4) Fuel rod internal pressure, 5) Pellet-cladding interactions (PCI) and 6) pellet-cladding mechanical interactions (PCMI). This paper discusses current and potential failure mechanisms of these failure mechanisms.

• Nuclear Engineering and Technology
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• v.42 no.1
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• pp.79-88
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• 2010

Spent $UO_2$ nuclear fuel discharged from a nuclear power plant (NPP) contains fission products, U, Pu, and other actinides. Due to neutron capture by $^{238}U$ in the rim region and a temperature gradient between the center and the rim of a fuel pellet, a considerable increase in the concentration of fission products, Pu, and other actinides are expected in the pellet periphery of high burnup fuel. The characterization of the radial profiles of the various isotopic concentrations is our main concern. For an analysis, spent nuclear fuels originating from the Yeonggwang-2 pressurized water reactor (PWR) were chosen as the test specimens. In this work, the distributions of some actinide isotopes were measured from center to rim of the spent fuel specimens by a radiation shielded laser ablation inductively coupled plasma mass spectrometer (LA-ICP-MS) system. Sampling was performed along the diameter of the specimen by reducing the sampling intervals from 500 ${\mu}m$ in the center to 100 ${\mu}m$ in the pellet periphery region. It was observed that the isotopic concentration ratios for minor actinides in the center of the specimen remain almost constant and increase near the pellet periphery due to the rim effect apart from the $^{236}U$ to $^{235}U$ ratio, which remains approximately constant. In addition, the distributions of local burnup were derived from the measured isotope ratios by applying the relationship between burnup and isotopic ratio for plutonium and minor actinides calculated by the ORIGEN2 code.

• Journal of the Korean Society for Nondestructive Testing
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• v.30 no.4
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• pp.324-329
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• 2010

Fuel rods using in nuclear power plants consist of uranium dioxide pellets enclosed in zirconium alloy(zircaloy) tubes. It is vitally important for the pellet surface to remain free from pits, cracks and chipping defects after it is loaded into the tubes to prevent local hot spots during reactor operation. This paper investigates the feasibility study for detecting surface flaws of pellets contained within nuclear fuel rod through X-ray tomography simulation. Reconstructed images used by parallel and fan-beam filtered back projection method were presented and confirmed the accessibility between simulation data and MPS(missing pellet surface) image data.

• 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.

• Corrosion Science and Technology
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• v.18 no.5
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• pp.175-181
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• 2019

A failed spent fuel rod with 53,000 MWd/tU from a nuclear power plant was characterized, and the fission products and oxygen layer in the pellet-clad mechanical interaction region were observed using an EPMA (Electron Probe Micro-Analyzer). A sound fuel rod burned under similar conditions was used to compare and analyze, the results of the failed fuel rod. In the failed fuel rod, the oxide layer represented $10{\mu}m$ of the boundary of the cladding, and $35{\mu}m$ of the region outside the cladding. By comparison, in the sound fuel rod, the oxide layer was $8{\mu}m$, observed in the cladding boundary region. The cladding inner surface corrosion and the resulting fuel-cladding bonding were investigated using an EPMA. Zirconium existed in the bonding layer of the (U, Zr)O compound beyond the pellet cladding interaction gap of $20{\mu}m$, and composition of UZr2O3 was observed in the failed fuel rod. This paper presents the results of the EPMA examination of a spent fuel specimen, and a technique to analyze fission products in the pellet-clad mechanical interaction region.