Artigos de revistas: "Nuclear fuel pellet"

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Dooley, Patricia, Dakota Contryman, Addie Hervey, Robert Ivers, Isabella Reddish e Yuze Song. "Design of an optimized nuclear fuel pellet". Nuclear Science and Technology Open Research 2 (9 de janeiro de 2024): 1. http://dx.doi.org/10.12688/nuclscitechnolopenres.17443.1.

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Background The design of an improved nuclear fuel pellet for use in the Westinghouse AP1000 reactor that is more powerful than existing pellets, is less expensive to manufacture, and meets Nuclear Regulatory Commission requirements for certification was undertaken to complete a senior design course in the ABET-certified nuclear engineering curriculum of Rensselaer Polytechnic Institute, Troy, NY. Methods The modeling team selected the Monte Carlo N-Particle (MCNP) program for assessing how well the pellet design achieves a k-effective value of 1, designed the base model consisting of a fuel pin inside a boron-water moderator with reflector, and ran MCNP tests on the base pellet. The design team modified the base pellet and tested it at different uranium-235 enrichments, with void spheres of varying volume and silicon carbide inclusions in the void volume. The simulation team selected K-code for testing the fuel pellets. The economics team analyzed the cost of manufacturing the improved pellet from cost of raw material through its tail assay in the form of Separative Work Unit (SWUs). The impacts team researched environmental, societal, governmental, political, and public affairs aspects of nuclear fuel production. Results Multiple configurations of uranium enrichment and silicon carbide volume inclusions in the nuclear fuel pellet achieved a k eff of 1, and the price per pellet, assuming fabrication costs comparable to existing manufacturing processes, was reduced by as much as about 50% when the volume of uranium oxide replaced by silicon carbide is 0.27 cm3. At smaller replacement volumes, the price per pellet is reduced by as little as 5%. Conclusions The goal of designing an optimized fuel pellet was met. Replacing a 0.27 cm3-volume sphere of uranium oxide with silicon carbide from the center of a pellet of 4%, 5%, or 6% uranium-235 enrichment reduced the cost of the pellet by approximately 50%.

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Heikinheimo, Janne, Teemu Kärkelä, Václav Tyrpekl, Matĕj̆ Niz̆n̆anský, Mélany Gouëllo e Unto Tapper. "Iodine release from high-burnup fuel structures: Separate-effect tests and simulated fuel pellets for better understanding of iodine behaviour in nuclear fuels". MRS Advances 6, n.º 47-48 (dezembro de 2021): 1026–31. http://dx.doi.org/10.1557/s43580-021-00175-1.

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Abstract Iodine release modelling of nuclear fuel pellets has major uncertainties that restrict applications in current fuel performance codes. The uncertainties origin from both the chemical behaviour of iodine in the fuel pellet and the release of different chemical species. The structure of nuclear fuel pellet evolves due to neutron and fission product irradiation, thermo-mechanical loads and fission product chemical interactions. This causes extra challenges for the fuel behaviour modelling. After sufficient amount of irradiation, a new type of structure starts forming at the cylindrical pellet outer edge. The porous structure is called high-burnup structure or rim structure. The effects of high-burnup structure on fuel behaviour become more pronounced with increasing burnup. As the phenomena in the nuclear fuel pellet are diverse, experiments with simulated fuel pellets can help in understanding and limiting the problem at hand. As fission gas or iodine release behaviour from high-burnup structure is not fully understood, the current preliminary study focuses on (i) sintering of porous fuel samples with Cs and I, (ii) measurements of released species during the annealing experiments and (iii) interpretation of the iodine release results with the scope of current fission gas release models. Graphical abstract

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Mirsalimov, Vagif. "Crack nucleation in rod-type nuclear fuel pellet". Mathematics and Mechanics of Solids 24, n.º 3 (1 de fevereiro de 2018): 668–85. http://dx.doi.org/10.1177/1081286517753977.

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A plane problem of fracture mechanics on crack nucleation in a rod-type nuclear fuel pellet is considered. Nuclear reactor fuel pellets in operation may be damaged in various ways; in particular, crack nucleation. We consider a problem for the case of a heat-releasing fuel pellet with cladding: as the heat release intensity increases, zones of heightened stress are formed in the nuclear fuel pellet. The heightened stress will promote the appearance of prefracture bands that are simulated as zones of weakened interparticle bonds of the material. Interaction of prefracture zone faces is simulated by placing bonds between faces that have a specified deformation pattern. The problem of equilibrium of a fuel pellet with prefracture zones is reduced to the solution of a system of singular integral equations. An analysis of the ultimate state of the zone of weakened interparticle bonds of the material is realized on the basis of the criterion of critical opening of prefracture zone faces.

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Beloborodov, Alexey V., Evgeny V. Vlasov, Leonid V. Finogenov e Peter S. Zav’yalov. "High Productive Optoelectronic Pellets Surface Inspection for Nuclear Reactors". Key Engineering Materials 437 (maio de 2010): 165–69. http://dx.doi.org/10.4028/www.scientific.net/kem.437.165.

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The results of development and investigation of computer-vision systems for inspection of the external view of fuel pellets for nuclear fuel elements are presented. The systems developed utilize CCD-cameras to record the images of a fuel pellet’s external view in reflected beams that ensures high contrast of the defects in the picture area. One has developed a database containing images of simulators, as well as real pellets. Tests of an experimental set for fuel pellet inspection have demonstrated its inspection productivity to be 1 pellet per second and its detection probability higher than 95%. The research team has also developed an experimental set with higher inspection productivity (at least 7 pellets per second).

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Joseph, Odii Christopher, Agyekum Ephraim Bonah e Bright Kwame Afornu. "Effect of Dual Surface Cooling on the Temperature Distribution of a Nuclear Fuel Pellet". Key Engineering Materials 769 (abril de 2018): 296–310. http://dx.doi.org/10.4028/www.scientific.net/kem.769.296.

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Heat removal from nuclear reactor core has been one of the major Engineering considerations in the construction of nuclear power plant. At the center of this consideration is the nuclear fuel pellet whose burning efficiency determines the rate of heat transfer to the coolant. This research, focuses on the study of temperature distribution of solid fuel, temperature distribution of annular fuel with external cooling and the temperature distribution of annular fuel with internal and external cooling. We analyzed the different distribution and made a conclusion on the possibility of improving temperature management of Nuclear fuel rod, by designing fuel pellets based on this geometrical and thermal Analysis. To date, a lot of studies has been done on the thermal and geometrical properties of Nuclear fuel pellet, it is observed that annular fuel pellet with simulteneous internal and external cooling can achieve better temperature distribution which leads to high linear heat generation rate, thus generating more power in the design [1]. It has also been observed that annular fuel pellets has low fission gas release [10]. In large LOCA, the peak cladding temperature of annular fuel is about 600 which is significantly less than that of solid fuel (920 ), this is due to the fact that annular fuel cladding has lower initial temperature and the thinner annular fuel can be cooled more efficiently than the solid fuel. One of drawbacks of annular fuel technology is “the fuel gap conductance assymmetry” which is caused by outward thermal expansion, it has a potential effect on the MDNBR (Minimum Departure from Nucleate Boiling Ratio), which is the minimum ratio of the critical to actual heat flux found in the core [10]. In this model, we used the ceramic fuel pellet of UO2 as our case study. All the parameters in this model are assumed parameters of UO2. The Heat Transfer tool (ANSYS APDL) was used to validate the Analytical Model of this research.

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Halabuk, Dávid, e Jiří Martinec. "CALCULATION OF STRESS AND DEFORMATION IN FUEL ROD CLADDING DURING PELLET-CLADDING INTERACTION". Acta Polytechnica 55, n.º 6 (31 de dezembro de 2015): 384. http://dx.doi.org/10.14311/ap.2015.55.0384.

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The elementary parts of every fuel assembly, and thus of the reactor core, are fuel rods. The main function of cladding is hermetic separation of nuclear fuel from coolant. The fuel rod works in very specific and difficult conditions, so there are high requirements on its reliability and safety. During irradiation of fuel rods, a state may occur when fuel pellet and cladding interact. This state is followed by changes of stress and deformations in the fuel cladding. The article is focused on stress and deformation analysis of fuel cladding, where two fuels are compared: a fresh one and a spent one, which is in contact with cladding. The calculations are done for 4 different shapes of fuel pellets. It is possible to evaluate which shape of fuel pellet is the most appropriate in consideration of stress and deformation forming in fuel cladding, axial dilatation of fuel, and radial temperature distribution in the fuel rod, based on the obtained results.

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Nguyen, Van Tung, Trong Hung Nguyen, Thanh Thuy Nguyen e Duy Minh Cao. "Predicting behavior of AP-1000 nuclear reactor fuel rod under steady state operating condition by using FRAPCON-4.0 software". Nuclear Science and Technology 8, n.º 2 (1 de setembro de 2021): 43–50. http://dx.doi.org/10.53747/jnst.v8i2.90.

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This paper reports the results on the predictions of behavior of AP-1000 nuclear reactorfuel rod under steady state operating condition by using FRAPCON-4.0 software. The predictive items were the temperature distribution in the fuel rod, including fuel centerline temperature, fuel pellet surface temperature, gas temperature, cladding inside and outside temperature, oxide surface and bulk coolant temperature; and gap conductance and thickness.The predictive items also include deformation of fuel pellets, fission gas release and rod internal pressure, cladding oxidation and hydration. The predictive data were suggested the fuel rod behavior image in nuclear reactor.

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Kim, Seyeon, e Sanghoon Lee. "Simplified Model of a High Burnup Spent Nuclear Fuel Rod under Lateral Impact Considering a Stress-Based Failure Criterion". Metals 11, n.º 10 (14 de outubro de 2021): 1631. http://dx.doi.org/10.3390/met11101631.

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The inventory of spent nuclear fuel (SNF) generated in nuclear power plants is continuously increasing, and it is very important to maintain the structural integrity of SNF for economical and efficient management. The cladding surrounding nuclear fuel must be protected from physical and mechanical deterioration, which causes fuel rod breakage. In this study, the material properties of the simplified beam model of a SNF rod were calibrated for a drop accident evaluation by considering the pellet–clad interaction (PCI) of the high burnup fuel rod. In a horizontal drop, which is the most damaging during a drop accident of SNF, the stress in the cladding caused by the inertia action of the pellets has a great effect on the integrity of the fuel rod. The failure criterion for SNF was selected as the membrane plus bending stress through stress linearization in the cross-sections through the thickness of the cladding. Because the stress concentration in the cladding around the vicinity of the pellet–pellet interface cannot be simulated in a simplified beam model, a stress correction factor is derived through a comparison of the simplified model and detailed model. The applicability of the developed simplified model is checked through dynamic impact simulations. The developed model can be used in cask level analyses and is expected to be usefully utilized to evaluate the structural integrity of SNF under transport and in storage conditions.

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Marchetti, Mara, Michel Herm, Tobias König, Simone Manenti e Volker Metz. "Actinides induced irradiation damage and swelling effect in irradiated Zircaloy-4 after 30 years of storage". Safety of Nuclear Waste Disposal 1 (10 de novembro de 2021): 7–8. http://dx.doi.org/10.5194/sand-1-7-2021.

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Abstract. After several years in the reactor core, irradiated nuclear fuel is handled and subsequently stored for a few years under water next to the core, to achieve thermal cooling and decay of very short-lived radionuclides. Thereafter, it might be sent to dry-cask interim storage before final disposal in a deep geological repository. Here, the spent nuclear fuel (SNF) is subject to a series of physicochemical phenomena which are of concern for the integrity of the nuclear fuel cladding. After moving the SNF from wet to dry storage, the temperature increases, then slowly decreases, leading the hydrogen in solid solution in the cladding to precipitate radially with consequent hydride growth and cladding embrittlement (Kim, 2020). Another phenomenon affecting the physical properties of the cladding during interim dry storage is the irradiation damage produced in the inner surface of the cladding by the alpha decay of the actinides present at the periphery of the pellet, particularly when the burnup at discharge is high. SNF pellets with high average burnup present larger fuel volumes at the end of their useful life due to accumulation of insoluble solid fission products and noble gases, which leads to disappearance of the as-fabricated pellet–clad gap. Further swelling is expected as a consequence of actinide decay and the accumulation of helium. This leads to larger cladding hoop stress and larger alpha decay damage. The present work first investigates the variation in diameter caused by pellet swelling in an irradiated Zircaloy-4 cladding after chemical digestion of the uranium oxide (UOx) pellet. Second, the irradiation damage produced during the 30 years elapsed since the end of irradiation in terms of displacements per atom (dpa) is studied by means of the FLUKA Monte Carlo code. The irradiation damage produced by the decay of actinides in the inner surface of the cladding extends for less than 3 % in depth. The considered cladded UOx pellet was extracted from a pressurized water reactor (PWR) fuel rod consisting of five segments, with an average burnup at discharge of 50.4 GWd (tHM)−1.

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Keyvan, Shahla, Xiaolong Song e Mark Kelly. "Nuclear fuel pellet inspection using artificial neural networks". Journal of Nuclear Materials 264, n.º 1-2 (janeiro de 1999): 141–54. http://dx.doi.org/10.1016/s0022-3115(98)00464-4.

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KWON, Y. D., S. B. KWON, K. T. RHO, M. S. KIM e H. J. SONG. "THERMO-ELASTIC-PLASTIC-CREEP FINITE ELEMENT ANALYSES OF ANNULAR NUCLEAR FUELS". International Journal of Modern Physics: Conference Series 06 (janeiro de 2012): 379–84. http://dx.doi.org/10.1142/s2010194512003479.

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In this study, we tried to examine the pros and cons of the annular type of fuel concerning mainly with the temperatures and stresses of pellet and cladding. The inner and outer gaps between pellet and cladding may play an important role on the temperature distribution and stress distribution of fuel system. Thus, we tested several inner and outer gap cases, and we evaluated the effect of gaps on fuel systems. We conducted thermo-elastic-plastic-creep analyses using an in-house thermo-elastic-plastic-creep finite element program that adopted the 'effective-stress-function' algorithm. Most analyses were conducted until the gaps disappeared; however, certain analyses lasted for 1582 days, after which the fuels were replaced. Further study on the optimal gaps sizes for annular nuclear fuel systems is still required.

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Sampaio Ribeiro, Luciana, Francisco Javier Rios e Armindo Santos. "Porous Stainless Steel Microsphere Synthesis by a Nonconventional Powder Metallurgy Process Useful in the Cermet-Type Advanced Nuclear Fuel Fabrication". Journal of Nanomaterials 2023 (29 de abril de 2023): 1–22. http://dx.doi.org/10.1155/2023/3555763.

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The fabrication of SS (stainless steel)-UO2 cermet-type advanced nuclear fuel pellets suitable for use in power reactors depends on the development of metallic (SS), ceramic (UO2), and cermet (SS-UO2) microspheres with special characteristics. In this work, a nonconventional powder metallurgy process was developed to produce porous SS microspheres aiming to contribute to solve the bottlenecks found in the SS-UO2 cermet pellet manufacturing. SS, UO2, and SS-UO2 microspheres and SS-UO2 cermet pellets were fabricated and characterized (XRD, EDX, EDS, and SEM). Hard (153 ± 5 µm; 132.2 ± 24.7 MPa; 72% TD) and soft (216 ± 10 µm; 1.3 ± 0.4 MPa; 17% TD) SS, hard (176 ± 6 µm; 147.4 ± 25.0 MPa; 99% TD) UO2, and cermet (SS-UO2) microspheres were obtained. The soft porous SS microspheres did not micronize properly in situ, but their high compressibility favors the compaction of the green SS-UO2 cermet pellet; in this pellet, the UO2 microspheres behaved as rigid inclusions. This favored the obtainment of sintered SS-UO2 cermet pellets with high geometric densities (93% TD), excellent metal–ceramic interaction, and the preservation of the physical integrity of the UO2 microspheres. The usage of high fractions of the SS-UO2 cermet microspheres obtained mixed with low fractions of the said soft porous SS microspheres is already under development. This will enable the fabrication of SS-UO2 cermet pellets with a volume fraction greater than 42 vol% UO2, a homogeneous distribution of UO2 microspheres in the metallic matrix, and null connection between them. The oxide–metal reduction mechanisms were discussed. The applicability of the process is already being explored in the manufacture of porous NdFeB microspheres.

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Kim, Young-Hwan, Yung-Zun Cho e Jin-Mok Hur. "Experimental Approaches for Manufacturing of Simulated Cladding and Simulated Fuel Rod for Mechanical Decladder". Science and Technology of Nuclear Installations 2020 (24 de janeiro de 2020): 1–12. http://dx.doi.org/10.1155/2020/1905019.

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We are developing a practical-scale mechanical decladder that can slit nuclear spent fuel rod-cuts (hulls + pellets) on the order of several tens of kgf of heavy metal/batch to supply UO2 pellets to a voloxidation process. The mechanical decladder is used for separating and recovering nuclear fuel material from the cladding tube by horizontally slitting the cladding tube of a fuel rod. The Korea Atomic Energy Research Institute (KAERI) is improving the performance of the mechanical decladder to increase the recovery rate of pellets from spent fuel rods. However, because actual nuclear spent fuel is dangerously toxic, we need to develop simulated spent fuel rods for continuous experiments with mechanical decladders. We describe procedures to develop both simulated cladding tubes and simulated fuel rod (with physical properties similar to those of spent nuclear fuel). Performance tests were carried out to evaluate the decladding ability of the mechanical decladder using two types of simulated fuel (simulated tube + brass pellets and zircaloy-4 tube + simulated ceramic fuel rod). The simulated tube was developed for analyzing the slitting characteristics of the cross section of the spent fuel cladding tube. Simulated ceramic fuel rod (with mechanical properties similar to the pellets of actual PWR spent fuel) was produced to ensure that the mechanical decladder could slit real PWR spent fuel. We used castable powder pellets that simulate the compressive stress of the real spent UO2 pellet. The production criteria for simulated pellets with compressive stresses similar to those of actual spent fuel were determined, and the castables were inserted into zircaloy-4 tubes and sintered to produce the simulated fuel rod. To investigate the slitting characteristics of the simulated ceramic fuel rod, a verification experiment was performed using a mechanical decladder.

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Vlasov, E. V., A. V. Beloborodov, P. S. Zav'yalov e D. G. Syretskiy. "Control of the appearance of fuel pellets ends surfaces in a conveyor production". Дефектоскопия, n.º 7 (15 de julho de 2023): 33–43. http://dx.doi.org/10.31857/s0130308223070047.

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The article deals with the problem of quality control of fuel pellets for nuclear reactors. During the development of the control system, various methods for obtaining and processing images of pellet surfaces were investigated. The main difficulty of this task is the imperfect quality of the resulting image of the inspected object, as well as the limited time for its processing. Software and hardware tools and algorithms have been developed for high-performance inspection of fuel pellet geometry, which significantly increase the reliability of inspection results. As a result of the work, stable images with a high degree of repeatability and sufficient resolution have been obtained, suitable for subsequent high-performance, reliable mathematical processing. A high degree of independence of the image and processing results from the individual characteristics of individual products and their batches has been achieved.

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Cantini, Federico, Martina Adorni e Francesco D’Auria. "Nuclear Fuel Modelling During Power Ramp". Journal of Energy - Energija 62, n.º 1-4 (18 de julho de 2022): 68–80. http://dx.doi.org/10.37798/2013621-4219.

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Fuel rods operating for several years in a LWR can experience fuel-cladding gap closure as a result of the phenomena due to temperature and irradiation. Local power increase induces circumferential stresses in the cladding as a result of the different expansion in the cladding and the pellet. In presence of corrosive fission products (i.e. Iodine) and beyond specific stress threshold and level of burnup, cracks may grow-up from the internal to the external cladding surface, causing fuel rod failure. The phenomenon, known as pellet cladding interaction-stress corrosion cracking PCI/SCC, or PCI, has been identified as a problem since the 70's. The PWR Super-Ramp experiment (part of OECD/NEA “International Fuel Performance Experiments (IFPE) database”) twenty eight fuel rods behaviour has been simulated using TRANSURANUS code version “v1m1j11”. Two sets (“Reference” and “Improved”) of suitable input decks modelling the fuel rods, based on the available literature are used to run the simulations. Focus is given to the main phenomena which are involved or may influence the cladding failure. Systematic comparison of the code results with the experimental data are performed for the parameters relevant for the PCI phenomenon. Sensitivity calculations on fission gas release models implemented in TRANSURANUS code are also performed in order to address the impact on the results. The results show the ability of TRANSURANUS version “v1m1j11” in conservatively predicting the rods failure due to PCI in PWR fuel and Zircaloy-4 cladding. Increased availability of experimental data would help to perform a deeper analysis.

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Reigel, M., C. Donohoue, Douglas Burkes, John J. Moore e J. R. Kennedy. "Application of Combustion Synthesis to the Production of Actinide Bearing Nitride Ceramic Nuclear Fuels". Materials Science Forum 561-565 (outubro de 2007): 1749–52. http://dx.doi.org/10.4028/www.scientific.net/msf.561-565.1749.

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Self-propagating high temperature (combustion) synthesis (SHS) is being used to develop several synthesis and processing routes for the next generation of ceramic nuclear fuels. These fuels are based on an actinide nitride within an inert matrix. The application of SHS is particularly important in the synthesis of americium (Am) based ceramics; since the rapid heating and cooling cycles used in this process will help to minimize vaporization loss of Am, which is a major problem in synthesizing Am-based ceramics. Manganese, praseodymium, and dysprosium are being used as physical and chemical surrogates for various actinides. Actinide nitride powders produced using auto-ignition combustion synthesis (AICS) are subsequently reacted with zirconium powder using SHS to produce a final fuel pellet. This paper will discuss the research to date on the synthesis of Am-N powders as well as the production of dense Zr-Am-N pellets as a model ceramic fuel system.

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Ferry, C., J. Radwan e H. Palancher. "Review about the Effect of He on the Microstructure of Spent Nuclear Fuel in a Repository". MRS Advances 1, n.º 62 (2016): 4147–56. http://dx.doi.org/10.1557/adv.2017.202.

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ABSTRACTHelium is produced in spent nuclear fuel by α-decays of actinides. After 10,000 years, the concentration of He accumulated in UO2 spent fuel is about 0.23 at.%. For direct disposal of spent nuclear fuel, consequences of helium build-up on the fuel matrix microstructure must be evaluated since it can modify the radionuclide release when water comes into contact with the spent fuel surface, after breaching of the disposal canister. An operational model has been proposed in order to evaluate the effect of helium on the microstructure of spent fuel in a repository. Based on conservative assumptions and different scenarios of bubble population, the calculated helium critical concentration, that could lead to a partial loss of integrity of the spent fuel pellet, is 0.37 at.%. However, observations on He-implanted UO2, α-doped UO2 pellets and natural analogues evidence a macroscopic damage only for He concentrations, which are more than one order of magnitude higher.

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Chernov, Igor, Аnton Kushtym, Volodymyr Tatarinov e Dmytro Kutniy. "Manufacturing Features and Characteristics of Uranium Dioxide Pellets for Subcritical Assembly Fuel Rods". 3, n.º 3 (2 de setembro de 2022): 59–66. http://dx.doi.org/10.26565/2312-4334-2022-3-08.

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The influence of technological processes and manufacturing of uranium dioxide fuel pellets for fuel elements for experimental fuel assembly (FA-X) which was designed as an alternative fuel for the nuclear research installation (NRI) "Neutron Source Controlled by Electron Accelerator" were investigated. Unlike standard production processes of UO2 pellets, the special feature fabrication process of this nuclear fuel type is production of uranium dioxide powder with enrichment of 4.4 %wt. of 235U achieved by mixing of two batches of powders with different uranium contents: 0.4 %wt. 235U and 19.7%wt. 235U, as well as ensuring the required tolerance of fuel pellets without the use of machining operations. A set of design and process documentation were developed in the R&D Center at NSC KIPT. Experimental stack of fuel pellets, fuel elements and a pilot fuel assembly FA-X were fabricated and designed to be compatible and interchangeable with VVR-M2 fuel assembly adopted as a standard assembly for the first fuel loading at the "Neutron Source Driven by an Electron Accelerator" FA. As opposition to the variant of VVR-M2 fuel assembly which consisted of three fuel rods of tubular shape with dispersion composition UO2‑Al, FA-X accommodates six fuel rods of pin-type with UO2 pellet which located in the zirconium cladding (E110) as the closest analogue of fuel rods of VVER-1000 power reactor. Inside cladding locate a 500 mm high fuel stack which is secured against displacement by a spacer. In the basic variant of FA-X the fuel pellets are made of UO2 with 235U enrichment near 4.4 %wt.

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Demarco, Gustavo L., e Armando C. Marino. "3D Finite Elements Modelling for Design and Performance Analysis of Pellets". Science and Technology of Nuclear Installations 2011 (2011): 1–10. http://dx.doi.org/10.1155/2011/843491.

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The geometry of a fuel pellet is a compromise among the intention to maximize UO2content and minimize the temperature gradient taking into account the thermomechanical behaviour, the economy, and the safety of the fuel management during and after irradiation. “Dishings”, “shoulders”, “chamfers”, and/or “a central hole” on a cylinder with an improvedl/drelation (length of the pellet/diameter) are introduced in order to optimize the shape of the pellet. The MeCom tools coupled with the BaCo code constitutes a complete system for the 3D analysis of the stress strain state of the pellet under irradiation. CANDU and PHWR MOX fuel will be used to illustrate the excellent qualitative agreement between experimental data and calculations by using these computational tools.

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Forsberg, K., L. O. Jernkvist e A. R. Massih. "Modeling oxygen redistribution in UO2+ fuel pellet". Journal of Nuclear Materials 528 (janeiro de 2020): 151829. http://dx.doi.org/10.1016/j.jnucmat.2019.151829.

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Kusumoputro, Benyamin, Rozandi Prarizky, Wahidin Wahab, Dede Sutarya e Li Na. "Assesment of Quality Classification of Green Pellets for Nuclear Power Plants Using Improved Levenberg-Marquardt Algorithm". Advanced Materials Research 608-609 (dezembro de 2012): 825–34. http://dx.doi.org/10.4028/www.scientific.net/amr.608-609.825.

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Cylindrical uranium dioxide pellets, which are the main components for nuclear fuel elements in Light Water Reactor, should have a high density profile, uniform shape and quality for the safety used as a reactor fuel component. The quality of green pellets is conventionally monitored through a laboratory measurement of the physical pellets characteristics followed by a graphical chart classification technique. However, this conventional classification method shows some drawbacks, such as the difficulties on its usage, low accuracy and time consuming, and does not have the ability to adress the non-linearity and the complexity of the relationship between the pellet’s quality variables and the pellett’s quality. In this paper, an Improved Levenberg-Marquard based neural networks is used to classify the quality process of the green pellets. Robustness of this learning algorithm is evaluated by comparing its recognition rate to that of the conventional Back Propagation neural learning algorithm. Results show that the Improved Levenberg-Marquard algorithm outperformed the Back Propagation learning algorthm for various percentage of training/testing paradigm, showing that this system could be applied effectively for classification of pellet quality.

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Kusumoputro, Benyamin, Dede Sutarya e Li Na. "Nuclear Power Plant Fuel’s Quality Classification Using Ensemble Back Propagation Neural Networks". Advanced Materials Research 685 (abril de 2013): 367–71. http://dx.doi.org/10.4028/www.scientific.net/amr.685.367.

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Nuclear power plants fuel production is very crucial and highly complex processes, involving numerous variables. For the safety used in the Light Water Nuclear Reactor, the cylindrical uranium dioxide pellets as the main fuel element should shows uniform shape, uniform quality and a high density profile. Therefore, the assesment of the quality classification of these pellets is important for improving the efficiency of the production process. The quality of green pellets is conventionally monitored through a laboratory measurement of the physical pellets characteristics followed by a graphical chart classification technique. This method, however, is difficult to use and shows low accuracy and time consuming, since its lack of the ability to adress the non-linearity and the complexity of the relationship between the pellet’s quality variables and the pellett’s quality. In this paper, an intelligent technique is develop to classify the pellets quality by using a computational intelligence methods. Instead of a Single Back Propagation neural networks that ussualy used, an Ensemble Back Propagation neural networks is proposed. It is proved in the experimental results that the Ensemble Back Propagation neural networks show higher classification rate compare with that of Single Back Propagation neural networks, showing that this system could be applied effectively for classification of pellet quality in its fabrication process.

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Belov, Alexander I., Randy W. L. Fong, Brian W. Leitch, Thambiayah Nitheanandan e Anthony Williams. "CHARACTERIZING HIGH-TEMPERATURE DEFORMATION OF INTERNALLY HEATED NUCLEAR FUEL ELEMENT SIMULATORS". CNL Nuclear Review 5, n.º 1 (junho de 2016): 67–84. http://dx.doi.org/10.12943/cnr.2016.00005.

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The sag behaviour of a simulated nuclear fuel element during high-temperature transients has been investigated in an experiment utilizing an internal indirect heating method. The major motivation of the experiment was to improve understanding of the dominant mechanisms underlying the element thermo-mechanical response under loss-of-coolant accident conditions and to obtain accurate experimental data to support development of 3-D computational fuel element models. The experiment was conducted using an electrically heated CANDU fuel element simulator. Three consecutive thermal cycles with peak temperatures up to ≈1000 °C were applied to the element. The element sag deflections and sheath temperatures were measured. On heating up to 600 °C, only minor lateral deflections of the element were observed. Further heating to above 700 °C resulted in an element multi-rate creep and significant permanent bow. Post-test visual and X-ray examinations revealed a pronounced necking of the sheath at the pellet-to-pellet interface locations. A wall thickness reduction was detected in the necked region that is interpreted as a sheath longitudinal strain localization effect. The sheath cross-sectioning showed signs of a “hard” pellet–cladding interaction due to the applied cycles. A 3-D model of the experiment was generated using the ANSYS finite element code. As a fully coupled thermal mechanical simulation is computationally expensive, it was deemed sufficient to use the measured sheath temperatures as a boundary condition, and thus an uncoupled mechanical simulation only was conducted. The ANSYS simulation results match the experiment sag observations well up to the point at which the fuel element started cooling down.

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Eidelpes, Elmar, Luis Francisco Ibarra e Ricardo Antonio Medina. "Ring compression tests on un-irradiated nuclear fuel rod cladding considering fuel pellet support". Journal of Nuclear Materials 510 (novembro de 2018): 446–59. http://dx.doi.org/10.1016/j.jnucmat.2018.08.009.

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Cherezov, Alexey, Jinsu Park, Hanjoo Kim, Jiwon Choe e Deokjung Lee. "A Multi-Physics Adaptive Time Step Coupling Algorithm for Light-Water Reactor Core Transient and Accident Simulation". Energies 13, n.º 23 (2 de dezembro de 2020): 6374. http://dx.doi.org/10.3390/en13236374.

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A new reactor core multi-physics system addresses the pellet-to-cladding heat transfer modeling to improve full-core operational transient and accident simulation used for assessment of reactor core nuclear safety. The rigorous modeling of the heat transfer phenomena involves strong interaction between neutron kinetics, thermal-hydraulics and nuclear fuel performance, as well as consideration of the pellet-to-cladding mechanical contact leading to dramatic increase in the gap thermal conductance coefficient. In contrast to core depletion where parameters smoothly depend on fuel burn-up, the core transient is driven by stiff equation associated with rapid variation in the solution and vulnerable to numerical instability for large time step sizes. Therefore, the coupling algorithm dedicated for multi-physics transient must implement adaptive time step and restart capability to achieve prescribed tolerance and to maintain stability of numerical simulation. This requirement is met in the MPCORE (Multi-Physics Core) multi-physics system employing external loose coupling approach to facilitate the coupling procedure due to little modification of constituent modules and due to high transparency of coupling interfaces. The paper investigates the coupling algorithm performance and evaluates the pellet-to-cladding heat transfer effect for the rod ejection accident of a light water reactor core benchmark.

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Mori, Y., K. Ishii, R. Hanayama, S. Okihara, Y. Kitagawa, Y. Nishimura, O. Komeda et al. "Ten hertz bead pellet injection and laser engagement". Nuclear Fusion 62, n.º 3 (3 de fevereiro de 2022): 036028. http://dx.doi.org/10.1088/1741-4326/ac3d69.

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Abstract A laser inertial fusion energy (IFE) reactor requires repetitive injection of fuel pellets and laser engagement to fuse fusion fuel beyond a few Hz. We demonstrate 10 Hz free-fall bead pellet injection and laser engagement with γ-ray generation. Deuterated polystyrene beads with a diameter of 1 mm were engaged by counter illuminating ultra-intense laser pulses with an intensity of 5 × 1017 W cm−2 at 10 Hz. The spatial distribution of free-fall beads was 0.86 mm in the horizontal direction and 0.18 mm in the vertical direction. The system operated for more than 5 min and 3500 beads were supplied with achieved frequencies of 2.1 Hz for illumination on the beads and 0.7 Hz for γ-ray generation; these frequencies were three times greater than with the previous 1 Hz injection system. The duration of operation was limited by the pellet supply. This injection and engagement system could be used for laser IFE research platforms.

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Centeno-Pérez, J., C. G. Aguilar-Madera, G. Espinosa-Paredes, E. C. Herrera-Hernández e A. D. Pérez-Valseca. "Upscaled elasticity modulus for nuclear fuel pellet (UO2) with porosity effects". Journal of Nuclear Materials 568 (setembro de 2022): 153875. http://dx.doi.org/10.1016/j.jnucmat.2022.153875.

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Tsibulskiy, S. "COMPARISON OF HOMOGENEOUS AND HETEROGENEOUS USE OF ENERGY PLUTONIUM IN VVER". PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS 2019, n.º 2 (26 de junho de 2019): 64–67. http://dx.doi.org/10.55176/2414-1038-2019-2-64-67.

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There are separate of fissile isotopes in a fuel rod in the fuel load admit the transition to a closed fuel circle in the nuclear power industry. For this solution to consider the formation of heterogeneous fuel pellet in which the fissile and raw isotope are separate and not mixed together. The raw material isotope is located in the peripheral zone. Fissile isotope, placed in the center of fuel pin holds 10% of the volume. There is row isotope in the periphery. All calculations for this research were made with UNK code. This code was complemented with a module, which allows calculating temperatures in pin cell. At present UNK allows solving of neutron-physical task of neutron transport, burn-up task (isotope composition evolution) and temperature calculation. Initial loading is plutonium, separated from spent fuel of the same cell of previous company. Irradiated plutonium of the central part is sent to either long term storage or final disposal. Energy plutonium location extracted from spent fuel of VVER in the central part of fuel pellet (heterogeneous pellet) decreases plutonium loading approximately 20 % in condition of the same company duration in comparison with UOX fuel. In case of heterogeneous location, there is depleted uranium in the periphery on the pellet. And after irradiation received plutonium isotope composition can be used repeatedly to form new loadings. This solution is qualitatively different from MOX case, which does not allow further using.

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MAHDAVI, M., e B. JALALY. "EFFECTS OF DEUTERIUM–LITHIUM FUSION REACTION ON INTERNAL TRITIUM BREEDING". International Journal of Modern Physics E 19, n.º 11 (novembro de 2010): 2123–32. http://dx.doi.org/10.1142/s0218301310016545.

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The optimal usage of designed fuel pellets is one of the very important parameters in inertial confinement fusion (ICF) systems. In this research, time-dependent dynamical equations for D/D fuel are written by considering impurity of 6 Li . Then dependency of gain on temperature, density and pellet radius is studied using Runge–Kutta method. The obtained results show that the energy gain will be maximized at the initial temperature 35 keV, density, 5000 g/cm3 and ratio impurity of 6 Li , 0.05.

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Wang, Qibiao, Yushi Luo, Yong Sun, Yang Wu, Bin Tang, Shuming Peng e Xianguo Tuo. "Weak-Edge Extraction of Nuclear Plate Fuel Neutron Images at Low Lining Degree". Applied Sciences 13, n.º 8 (19 de abril de 2023): 5090. http://dx.doi.org/10.3390/app13085090.

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Neutron imaging is an effective nondestructive testing (NDT) technique widely applied to detect structural defects and the enrichment of nuclear fuel elements due to its high penetration and nuclide-sensitive properties. Since the fuel element pellet is sealed in the cladding, the transmission imaging result is a superposition of the two parts. Therefore, the attenuation of neutrons by the cladding is interference that must be considered in the enrichment analysis. It is necessary to extract and separate cladding and pellets using an edge extraction method. However, the low neutron cross-section of the cladding material (e.g., aluminum and zirconium) leads to poor grayscale contrast at the cladding edge in the imaging result, and the intensity of the cladding edge is significantly lower than that of the pellet edge. In addition, affected by the noise from the imaging environment, the boundaries of targets are further blurred, making edge detection more challenging. Traditional detection algorithms extract the weak edges of cladding incompletely, and the results are discontinuous, with obvious edge breaks and missing areas. This paper proposes a method to extract edges in neutron images based on phase congruency (PC). This study utilized the classical perceptual field model to improve contrast at weak edges. The enriched edge map was generated using our PC model from six directions, allowing more weak edges to be detected accurately. The non-maximum suppression ensured precise localization and avoided edge breaks. Furthermore, the edge results were optimized by eliminating noise through morphological operations. The experimental results demonstrate that the proposed method effectively detects the weak edges of the cladding, is superior in accuracy and integrity to traditional detection, and is able to obtain stable and reliable results with different materials of neutron images. The edge integrity improved by 64.1%, and the edge localization accuracy reached 94.3%. The extracted edge information is useful in the next stage of the high-precision enrichment analysis.

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Kuzmin, Ilya V., Anton Yu Leshchenko, Sergey V. Pavlov, Rinat N. Shamsutdinov e Yuriy S. Mochalov. "Test bench for gas-dynamic studies in the furnace channel for nuclear fuel pellet sintering *". Nuclear Energy and Technology 5, n.º 2 (21 de junho de 2019): 171–75. http://dx.doi.org/10.3897/nucet.5.36479.

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Nuclear fuel pellets are sintered in high-temperature furnaces in an atmosphere with strictly defined requirements for the composition of the gas environments in the furnace’s different temperature zones. The preset process conditions in the mixed nitride uranium-plutonium (MNUP) fuel pellet sintering furnace is achieved through the respective gas supply arrangement and by the design of the barriers between the temperature zones and that of the gas supply and discharge units. A CFD model was created in the Ansys Fluent package and validated for testing the functionality of the design concepts used to develop the MNUP fuel sintering furnace channel. A mockup of the sintering furnace channel, which makes a part of the gas-dynamic test bench, was developed and fabricated for the analytical model validation. The paper presents a description of the test bench design and performance for measuring the concentration of gases in the channel simulating the nitride nuclear fuel sintering furnace channel. The results of the test bench gas-dynamic studies were used for the computational and experimental justification of the approaches used to develop the sintering furnace channel. The functionality of the barriers for the sintering furnace channel division into zones with the preset composition of the gas environments and the gas supply and discharge units has been tested experimentally. The obtained experimental data on the distribution of the process gas concentration makes it possible to validate computational thermophysical and gas-dynamic CFD models of the MNUP fuel sintering furnace channel.

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Lee, Sanghoon, e Seyeon Kim. "Development of Equivalent Beam Model of High Burnup Spent Nuclear Fuel Rods under Lateral Impact Loading". Metals 10, n.º 4 (3 de abril de 2020): 470. http://dx.doi.org/10.3390/met10040470.

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Spent nuclear fuel (SNF) is nuclear fuel that has been irradiated and discharged from nuclear reactors. During the whole management stages of SNF before it is, in the end, disposed in a deep geological repository, the structural integrity of fuel rods and the assemblies should be maintained for safety and economic reasons. In licensing applications for the SNF storage and transportation, the integrity of SNF needs to be evaluated considering various loading conditions. However, this is a challenging task due to the complexity of the geometry and properties of SNF. In this paper, a simple and equivalent analysis model for SNF rods is developed using model calibration based on optimization and process integration. The spent fuel rod is simplified into a hollow beam with a homogenous isotropic material, and the model parameters thus found are not dependent on the length of the reference fuel rod segment that is considered. Two distinct models with different interfacial conditions between the fuel pellets and cladding are used in the calibration to account for the effect of PCMI (Pellet-Clad Mechanical Interaction). The feasibility of the models in dynamic impact simulations is examined, and it is expected that the developed models can be utilized in the analysis of assembly-level analyses for the SNF integrity assessment during transportation and storage.

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Kim, Ki Hwan, Jong Man Park, Don Bae Lee, Chul Goo Chi e Chang Kyu Kim. "Fabrication of Monolithic UAl₂ Pellet for High-Density Nuclear Fuel". Advanced Materials Research 26-28 (outubro de 2007): 925–28. http://dx.doi.org/10.4028/www.scientific.net/amr.26-28.925.

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Uranium powder, which can be obtained by the atomization process, was used to fabricate UAl2 by using powder metallurgy technology. Uranium powder and Al powder were blended, extruded, and annealed into a UAl2 rod in a mold. Sound UAl2 rods were fabricated by the powder metallurgical process. The relative density of the UAl2 pellet formed by an annealing was at about 94%. The density increased with higher constraints on the mold and a smaller particle size of the uranium powder. A coarse uranium powder of about 80 μm in average diameter represented the remaining un-reacted uranium phase. On the other hand, a fine uranium powder of about 50 μm in average diameter could achieve a pure UAl2 phase without a uranium phase. The analysis by an X-ray diffraction pattern confirmed that the annealed specimens had interacted to form a UAl2 phase. Conclusively, the sintered UAl2 pellet is expected to be useful in developing advanced fuels.

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Yusibani, Elin, Fitria Helmiza, Fashbir Fashbir e Sidik Permana. "Simulation on the Effect of Coolant Inlet Temperature and Mass-Flowrate Variations to the Temperature Distribution in Single Pellet Thermal Reactor Core". Jurnal Penelitian Fisika dan Aplikasinya (JPFA) 11, n.º 1 (23 de julho de 2021): 63–71. http://dx.doi.org/10.26740/jpfa.v11n1.p63-71.

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An important factor in the development of nuclear energy is reactor safety. The performance of heat transfer from nuclear fuel to coolant is the main key to the reactor safety. This paper presents simulation on temperature distribution in two-dimensional laminar flow for single pellet thermal reactor with variation on temperature inlet and mass-flowrate. The OpenFoam platform (SimFlow 3.1) has been used for the computational and numerical analysis. The simulation is carried out on a single pellet with an aspect ratio of 1.2. The variations in the mass velocity of the coolant flow are 10, 100, and 14300 kg×s-1 with a constant coolant temperature of 552 K, and the variations of the input coolant temperature are 300, 552, and 1000 K with a constant mass-flowrate of 10 kg×s-1. The results obtained from the simulation show that for variations in the input coolant temperature of 300, 552, and 1000 K, the fuel temperature can be reduced respectively by 34, 26, and 14 K. At the fastest variation in the coolant mass-flowrate of 14300 kg×s-1, the coolant temperature around the pellet rises by 396 K. The decrease in fuel temperature is significant if the mass-flowrate of the input coolant flow is relatively low.

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Fidalgo, Alexandre Barreiro, Olivia Roth, Anders Puranen, Lena Z. Evins e Kastriot Spahiu. "Aqueous leaching of ADOPT and standard UO2 spent nuclear fuel under H2 atmosphere". MRS Advances 5, n.º 3-4 (2020): 167–75. http://dx.doi.org/10.1557/adv.2020.69.

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ABSTRACTLeaching results to compare the dissolution behavior of a new type of fuel with additives (Advanced Doped Pellet Technology, ADOPT) with standard UO2 fuel are presented. Both fuels were irradiated in the same assembly of a commercial boiling water reactor to a local burnup of ∼58 MWd/kgU. Fuel fragments are leached in simplified groundwater in two autoclaves under hydrogen atmosphere, representing conditions in a canister failure scenario resulting in water intrusion for a spent nuclear fuel repository. Preliminary results indicate the uranium concentration decreased to 3-4x10-8 M after 421 days, slightly above the solubility of amorphous UO2. Xe has been detected in the gas phase of both autoclaves. The concentration of Cs and I seems to gradually approach constant values, yet the redox sensitive elements continue to slowly increase with time. The preliminary data obtained supports the hypothesis that there is no major difference in leaching behavior between the two fuels.

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Si, Shengyi. "Multiphysics Model Development and the Core Analysis for In Situ Breeding and Burning Reactor". Science and Technology of Nuclear Installations 2013 (2013): 1–14. http://dx.doi.org/10.1155/2013/154706.

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The in situ breeding and burning reactor (ISBBR), which makes use of the outstanding breeding capability of metallic pellet and the excellent irradiation-resistant performance of SiCf/SiC ceramic composites cladding, can approach the design purpose of ultralong cycle and ultrahigh burnup and maintain stable radial power distribution during the cycle life without refueling and shuffling. Since the characteristics of the fuel pellet and cladding are different from the traditional fuel rod of ceramic pellet and metallic cladding, the multiphysics behaviors in ISBBR are also quite different. A computer code, named TANG, to model the specific multiphysics behaviors in ISBBR has been developed. The primary calculation results provided by TANG demonstrate that ISBBR has an excellent comprehensive performance of GEN-IV and a great development potential.

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STANKUNAS, GEDIMINAS. "FRACTAL MODEL OF FISSION PRODUCT RELEASE IN NUCLEAR FUEL". International Journal of Modern Physics C 23, n.º 09 (setembro de 2012): 1250057. http://dx.doi.org/10.1142/s012918311250057x.

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A model of fission gas migration in nuclear fuel pellet is proposed. Diffusion process of fission gas in granular structure of nuclear fuel with presence of inter-granular bubbles in the fuel matrix is simulated by fractional diffusion model. The Grunwald–Letnikov derivative parameter characterizes the influence of porous fuel matrix on the diffusion process of fission gas. A finite-difference method for solving fractional diffusion equations is considered. Numerical solution of diffusion equation shows correlation of fission gas release and Grunwald–Letnikov derivative parameter. Calculated profile of fission gas concentration distribution is similar to that obtained in the experimental studies. Diffusion of fission gas is modeled for real RBMK-1500 fuel operation conditions. A functional dependence of Grunwald–Letnikov derivative parameter with fuel burn-up is established.

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YANAGISAWA, Kazuaki, e Harald DEVOLD. "Pellet-cladding interaction on light water reactor fuel. (II) BWR type fuel rod." Journal of the Atomic Energy Society of Japan / Atomic Energy Society of Japan 28, n.º 8 (1986): 771–82. http://dx.doi.org/10.3327/jaesj.28.771.

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YANAGISAWA, Kazuaki, Yoshiaki KONDO e Erik KOLSTAD. "Pellet-cladding interaction on light water reactor fuel, (I)". Journal of the Atomic Energy Society of Japan / Atomic Energy Society of Japan 28, n.º 7 (1986): 641–57. http://dx.doi.org/10.3327/jaesj.28.641.

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Marchal, N., C. Campos e C. Garnier. "Finite element simulation of Pellet-Cladding Interaction (PCI) in nuclear fuel rods". Computational Materials Science 45, n.º 3 (maio de 2009): 821–26. http://dx.doi.org/10.1016/j.commatsci.2008.10.015.

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Lin, Wei Keng, Jong Rong Wang, Yung Shin Tseng e Jui En Chang. "Using CFD Couple with Visual Basic to Investigate the Thermal Behavior for Fuel Rod Bowing Problem". Advanced Materials Research 651 (janeiro de 2013): 688–93. http://dx.doi.org/10.4028/www.scientific.net/amr.651.688.

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Nuclear fuel elements assemblies are generally consists of fuel rod bundles; each bundle is a concentric cylinder with three layers. Taking GE-8X8 for example, there is pellet, gap, and cladding from inside to outside. The diameter for each concentric cylinder is 9.26cm, 9.47cm, and 10.71cm respectively. In reality, a structure deformation may happen to those components due to the reason of radiation result in the high temperature of the bundles system. For the space of gap decreases by the expansion of pellet, the thermal conductivity might be under predicted and there is not enough study about this topic yet. To improve the accuracy of PRAs, more studies of the shrink phenomena on the gap between pellet and cladding are necessary. In this study, we had developed a program on the purpose of processes improvement for CFD simulation about spent fuel dry storage system. The program can adjust the dimension for each part of formation very friendly. We think it can also do some help on the needs if we want to compare the performance on heat transfer for different fraction on each part of bundle. In addition, the axial power distributions of the rod were also defined file by the user very easy, the results shown no obviously temperature difference between the full gap and 90% reduction of the gap.

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Cordara, Theo, Hannah Smith, Ritesh Mohun, Laura J. Gardner, Martin C. Stennett, Neil C. Hyatt e Claire L. Corkhill. "Hot Isostatic Pressing (HIP): A novel method to prepare Cr-doped UO2 nuclear fuel". MRS Advances 5, n.º 1-2 (2020): 45–53. http://dx.doi.org/10.1557/adv.2020.62.

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ABSTRACTThe addition of Cr2O3 to modern UO2 fuel modifies the microstructure so that, through the generation of larger grains during fission, a higher proportion of fission gases can be accommodated. This reduces the pellet-cladding mechanical interaction of the fuel rods, allowing the fuels to be “burned” for longer than traditional UO2 fuel, thus maximising the energy obtained. We here describe the preparation of UO2 and Cr-doped UO2 using Hot Isostatic Pressing (HIP), as a potential method for fuel fabrication, and for development of analogue materials for spent nuclear fuel research. Characterization of the synthesised materials confirmed that high density UO2 was successfully formed, and that Cr was present as particles at grain boundaries and also within the UO2 matrix, possibly in a reduced form due to the processing conditions. In contrast to studies of Cr-doped UO2 synthesised by other methods, no significant changes to the grain size were observed in the presence of Cr.

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Francon, Virginie, Marion Fregonese, Hiroshi Abe e Yutaka Watanabe. "Iodine-Induced Stress Corrosion Cracking of Zircaloy-4: Identification of Critical Parameters Involved in Intergranular to Transgranular Crack Propagation". Solid State Phenomena 183 (dezembro de 2011): 49–56. http://dx.doi.org/10.4028/www.scientific.net/ssp.183.49.

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During power transient conditions in nuclear reactors, uranium oxide pellets expand and crack due to the increase in temperature and their poor thermal conductivity. Moreover, the cladding undergoes creep because of the external pressure, and its diameter shortens. These antagonistic phenomena lead to the establishment of a contact between the pellet and the cladding, called the pellet-cladding interaction. The synergistic effect of the hoop tensile stress and strain imposed on the cladding by fuel thermal expansion and corrosion by iodine released from the UO2 fuel as a fission product at the same time can lead to Iodine-induced Stress Corrosion Cracking (I-SCC) of the Zircaloy-4 cladding. I-SCC failures of zirconium alloys are usually described in three steps: initiation of cracks, intergranular subcritical propagation, and critical propagation with a brittle transgranular propagation mode [1]. Transgranular propagation occurs as soon as the stress intensity factor overshoots a threshold value KI,SCC. It is the critical step and leads to the final ductile failure of the cladding. Transgranular cracks propagate by cleavage-like fracture on basal planes of the hexagonal lattice and fluting; it is the result of a competition between a plastic accommodation of the applied strain and the brittle fracture of basal planes by iodine assisted cleavage.

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Johnston, Craig M. T., e G. Cornelis van Kooten. "Economic consequences of increased bioenergy demand". Forestry Chronicle 90, n.º 05 (outubro de 2014): 636–42. http://dx.doi.org/10.5558/tfc2014-128.

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Although wind, hydro and solar are the most discussed sources of renewable energy, countries will need to rely much more on biomass if they are to meet renewable energy targets. In this study, a global forest trade model is used to examine the global effects of expanded demand for wood pellets fired with coal in power plants. Positive mathematical programming is used to calibrate the model to 2011 bilateral trade flows. To assess the impact of increased demand for wood pellets on global forest products, we consider a scenario where demand for wood pellets doubles. Findings indicate that production of lumber and plywood is likely to increase in most of the 20 model regions, but outputs of fibreboard, particleboard and pulp will decline as these products must compete with wood pellets for residual fibre. Ultimately, policies promoting aggressive renewable energy targets cause wood pellet prices to more than double in our scenarios, which could increase the cost of generating electricity to such an extent that, in some regions, electricity producers will continue to use fossil fuels as their primary fuel, while some others might find it worthwhile to rely more on nuclear energy for base load power.

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Nakamura, H., T. Kubo, T. Karino, H. Kato e S. Kawata. "Fuel pellet injection into heavy-ion inertial fusion reactor". High Energy Density Physics 35 (junho de 2020): 100741. http://dx.doi.org/10.1016/j.hedp.2019.100741.

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Pauzi, Anas Muhamad, Hector Iacovides e Andrea Cioncolini. "Pragmatic modelling of axial flow-induced vibration (FIV) for nuclear fuel rods". IOP Conference Series: Materials Science and Engineering 1285, n.º 1 (1 de julho de 2023): 012001. http://dx.doi.org/10.1088/1757-899x/1285/1/012001.

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Abstract Flow-induced vibration (FIV) at the spacer grid in the fuel assembly of a Light Water Reactor (LWR) is the leading cause of fuel failure. This project aims to produce a simulation benchmark on the experimental campaign at MACE on axial FIV of a cantilever beam in an annular tube, that mimics the configuration and environment of a typical LWR. The nuclear fuel rod, which consists of fuel pellets filled in Zirconium alloy cladding is modelled in the experiment as a steel rod filled with lead shots that closely approximates the filling density of the fuel pellet. To reduce the complexity and increase the efficiency of the simulation, further simplification is applied by on the geometry and assuming the solid domain as a single material instead of multiple materials. The first mode of frequency of the rod vibrating in quiescent water, which had been validated via the Euler-Bernoulli beam theory against experimental measurements, was used to design the single-material solid domain. Two models were proposed, firstly a solid rod with lower density and stiffness (SLE), and secondly an empty cladding with high density and low stiffness (EHD). Both solid and fluid domains were discretised using the cell-centred finite volume (FV) method and coupled with strong two-way fluid-structure interaction (FSI). Results on the frequency of vibration in quiescent water and in axial flow showed good agreement with experimental measurement, and the computational efficiency is analyzed for different rod models and changes in parameters of the solid domains.

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Kim, Dong-Joo, Keon Sik Kim, Dong Seok Kim, Jang Soo Oh, Jong Hun Kim, Jae Ho Yang e Yang-Hyun Koo. "Development status of microcell UO2 pellet for accident-tolerant fuel". Nuclear Engineering and Technology 50, n.º 2 (março de 2018): 253–58. http://dx.doi.org/10.1016/j.net.2017.12.008.

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Konashi, Kenji, e Michio Yamawaki. "Utilization of Hydride Materials in Nuclear Reactors". Advances in Science and Technology 73 (outubro de 2010): 51–58. http://dx.doi.org/10.4028/www.scientific.net/ast.73.51.

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Metal hydrides have high hydrogen atom density, which is equivalent to that of liquid water. Fast neutrons are efficiently moderated by hydrogen in metal hydrides. Metal hydrides have been studied for their potential application as nuclear materials in fast reactors (FRs). Two types of the utilizations of metal hydride in FRs are discussed in this paper. One is the utilization for transmutation target of long-lived nuclear wastes. Hydride fuel containing 237Np, 241Am and 243Am has been studied as a candidate transmutation target to reduce the radioactivity of long-lived nuclides included in reprocessed nuclear wastes. An application of the hafnium hydride has been investigated as neutron absorber in FRs. The core design has been performed to examine its characteristics and to evaluate the cost reduction effect. Demonstration of fabrication of hydride pin has been done with hydride pellets and stainless steel cladding. Coating technique of inner cladding surface has been also developed to reduce the permeation of hydrogen through stainless steel cladding. Physical and chemical properties of the pellet have been measured for designing the hydride pin. The integrity of the pellets at high temperature has been tested and their compatibility with sodium has also been tested. Irradiation test of hydrides has been performed in the fast experimental reactor, JOYO, at Japan Atomic Energy Association (JAEA).

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Keyvan, Shahla, Mark L. Kelly e Xiaolong Song. "Feature Extraction for Artificial Neural Network Application to Fabricated Nuclear Fuel Pellet INSPECTION". Nuclear Technology 119, n.º 3 (setembro de 1997): 269–75. http://dx.doi.org/10.13182/nt97-a35402.

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Zhang, Bin, Mengmeng Liu, Yongzhi Tian, Ge Wu, Xiaohui Yang, Songyang Shi e Jianning Li. "Defect inspection system of nuclear fuel pellet end faces based on machine vision". Journal of Nuclear Science and Technology 57, n.º 6 (2 de janeiro de 2020): 617–23. http://dx.doi.org/10.1080/00223131.2019.1708827.

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