• Nuclear Engineering and Technology
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• 제51권3호
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• pp.769-775
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• 2019

The tensile strength of irradiated 3C-SiC, SiC with artificial point defects, SiC with symmetric tilt grain boundaries (GBs), irradiated SiC with GBs are investigated using molecular dynamics simulations at 300 K. For an irradiated SiC sample, the tensile strength decreases with the increase of irradiation dose. The Young's modulus decreases with the increase of irradiation dose which agrees well with experiment and simulation data. For artificial point defects, the designed point defects dramatically decrease the tensile strength of SiC at low concentration. Among the point defects studied in this work, the vacancies drop the strength the most seriously. SiC symmetric tilt GBs decrease the tensile strength of pure SiC. Under irradiated condition, the tensile strengths of all SiC samples with grain boundaries decrease and converge to certain value because the structures become amorphous and the grain boundaries disappear after high dose irradiation.

• Nuclear Engineering and Technology
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• 제42권5호
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• pp.576-581
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• 2010

KAERI (Korea Atomic Energy Research Institute) fabricated MOX (Mixed Oxide) fuel pellets as a cooperation project with PSI (Paul Scherrer Institut) for an irradiation test in the Halden reactor. The MOX pellets were fitted into fuel rods that included instrumentation for measurement in IFE (Institutt for Energiteknikk). The fuel rods were assembled into the test rig and irradiated in the Halden reactor up to 50 MWd/kgHM. The irradiated fuel rods were transported to the IFE, where ceramography was carried out. The fuel rods were cut transversely at the relatively higher burn-up locations and then the radial cross sections were observed. Micrographs were analyzed using an image analysis program and grain sizes along the radial direction were measured by the linear intercept method. Radial cracks in the irradiated MOX were observed that were generally circumferentially closed at the pellet periphery and open in the hot central region. A circumferential crack was formed along the boundary between the dark central and the outer regions. The inner surface of the cladding was covered with an oxide layer. Pu-rich spots were observed in the outer region of the fuel pellets. The spots were surrounded by many small pores and contained some big pores inside. Metallic fission product precipitates were observed mainly in the central region and in the inside of the Pu spots. The average areal fractions of the metallic precipitates at the radial cross section were 0.41% for rod 6 and 0.32% for rod 3. In the periphery, pore density smaller than 2 ${\mu}m$ was higher than that of the other regions. The grain growth occurred from 10 ${\mu}m$ to 12 ${\mu}m$ in the central region of rod 6 during irradiation.

• Nuclear Engineering and Technology
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• 제46권2호
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• pp.147-158
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• 2014

In order to investigate how the microstructure of fuel/matrix-interaction (FMI) layers change during irradiation, different U-7Mo dispersion fuel plates have been irradiated to high fission density and then characterized using scanning electron microscopy (SEM). Specifially, samples from irradiated U-7Mo dispersion fuel elements with pure Al, Al-2Si and AA4043 (~4.5 wt.%Si) matrices were SEM characterized using polished samples and samples that were prepared with a focused ion beam (FIB). Features not observable for the polished samples could be captured in SEM images taken of the FIB samples. For the Al matrix sample, a relatively large FMI layer develops, with enrichment of Xe at the FMI layer/Al matrix interface and evidence of debonding. Overall, a significant penetration of Si from the FMI layer into the U-7Mo fuel was observed for samples with Si in the Al matrix, which resulted in a change of the size (larger) and shape (round) of the fission gas bubbles. Additionally, solid fission product phases were observed to nucleate and grow within these bubbles. These changes in the localized regions of the microstructure of the U-7Mo may contribute to changes observed in the macroscopic swelling of fuel plates with Al-Si matrices.

• Nuclear Engineering and Technology
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• 제52권5호
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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.

• Nuclear Engineering and Technology
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• 제41권2호
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• pp.155-162
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• 2009

Since the 90's, EDF and AREVA-NP have irradiated, up to very high burnups, lead assemblies housing $M5^{(R)}$ cladded fuels. Post-irradiation examination of high burnup $UO_2$ pellets show an increase in the fission-gas release rate, an increase in fuel swelling, and formation of fission-gas bubbles throughout the pellets. Xenon abundances were quantified, and phenomena leading to this bubble formation were identified. All examinations provided valuable data on the complex state of the fuel during irradiation. They show the good behavior of these fuels, exhibiting various microstructures at very high burnups, none of which is likely to lead to problems during irradiation.

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.1002-1008
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• 2016

An understanding of the coarsening process of the large fission gas pores in the high burn-up structure (HBS) of irradiated $UO_2$ fuel is very necessary for analyzing the safety and reliability of fuel rods in a reactor. A numerical model for the description of pore coarsening in the HBS based on the Ostwald ripening mechanism, which has successfully explained the coarsening process of precipitates in solids is developed. In this model, the fission gas atoms are treated as the special precipitates in the irradiated $UO_2$ fuel matrix. The calculated results indicate that the significant pore coarsening and mean pore density decrease in the HBS occur upon surpassing a local burn-up of 100 GWd/tM. The capability of this model is successfully validated against irradiation experiments of $UO_2$ fuel, in which the average pore radius, pore density, and porosity are directly measured as functions of local burn-up. Comparisons with experimental data show that, when the local burn-up exceeds 100 GWd/tM, the calculated results agree well with the measured data.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.965-966
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• 2004

• Nuclear Engineering and Technology
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• 제45권7호
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• pp.847-858
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• 2013

Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea) in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4-5 $g-U/cm^3$ were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr), additional protective coatings (silicide or nitride), and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 춘계공동학술발표회요약집
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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.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.2792-2800
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• 2022

In the core of the WWR-K reactor, a long-term irradiation of tristructural isotopic (TRISO)-coated fuel particles (CFPs) with a UO₂ kernel was carried out under high-temperature gas-cooled reactor (HTGR)-like operating conditions. The temperature of this TRISO fuel during irradiation varied in the range of 950-1100 ℃. A fission per initial metal atom (FIMA) of uranium burnup of 9.9% was reached. The release of gaseous fission products was measured in-pile. The release-to-birth ratio (R/B) for the fission product isotopes was calculated. Aspects of fuel safety while achieving deep fuel burnup are important and relevant, including maintaining the integrity of the fuel coatings. The main mechanisms of fuel failure are kernel migration, silicon carbide corrosion by palladium, and gas pressure increase inside the CFP. The formation of gaseous fission products and carbon monoxide leads to an increase in the internal pressure in the CFP, which is a dominant failure mechanism of the coatings under this level of burnup. Irradiated fuel compacts were subjected to electric dissociation to isolate the CFPs from the fuel compacts. In addition, nondestructive methods, such as X-ray radiography and gamma spectrometry, were used. The predicted R/B ratio was evaluated using the fission gas release model developed in the high-temperature test reactor (HTTR) project. In the model, both the through-coatings of failed CFPs and as-fabricated uranium contamination were assumed to be sources of the fission gas. The obtained R/B ratio for gaseous fission products allows the finalization and validation of the model for the release of fission products from the CFPs and fuel compacts. The success of the integrity of TRISO fuel irradiated at approximately 9.9% FIMA was demonstrated. A low fuel failure fraction and R/B ratios indicated good performance and reliability of the studied TRISO fuel.