Gotowe bibliografie tematyczne / Nuclear fuel pellet

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Gotowa bibliografia na temat „Nuclear fuel pellet”

Autor: Grafiati
Data publikacji: 1 lutego 2025

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Artykuły w czasopismach na temat "Nuclear fuel pellet"

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Dooley, Patricia, Dakota Contryman, Addie Hervey, Robert Ivers, Isabella Reddish i Yuze Song. "Design of an optimized nuclear fuel pellet". Nuclear Science and Technology Open Research 2 (9.01.2024): 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 i 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, nr 47-48 (grudzień 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, nr 3 (1.02.2018): 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 i Peter S. Zav’yalov. "High Productive Optoelectronic Pellets Surface Inspection for Nuclear Reactors". Key Engineering Materials 437 (maj 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 i Bright Kwame Afornu. "Effect of Dual Surface Cooling on the Temperature Distribution of a Nuclear Fuel Pellet". Key Engineering Materials 769 (kwiecień 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, i Jiří Martinec. "CALCULATION OF STRESS AND DEFORMATION IN FUEL ROD CLADDING DURING PELLET-CLADDING INTERACTION". Acta Polytechnica 55, nr 6 (31.12.2015): 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 i 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, nr 2 (1.09.2021): 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, i Sanghoon Lee. "Simplified Model of a High Burnup Spent Nuclear Fuel Rod under Lateral Impact Considering a Stress-Based Failure Criterion". Metals 11, nr 10 (14.10.2021): 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 i Volker Metz. "Actinides induced irradiation damage and swelling effect in irradiated Zircaloy-4 after 30 years of storage". Safety of Nuclear Waste Disposal 1 (10.11.2021): 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 i Mark Kelly. "Nuclear fuel pellet inspection using artificial neural networks". Journal of Nuclear Materials 264, nr 1-2 (styczeń 1999): 141–54. http://dx.doi.org/10.1016/s0022-3115(98)00464-4.

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Rozprawy doktorskie na temat "Nuclear fuel pellet"

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Kingdon, David Ross. "Safety characteristics of a suspended-pellet fission reactor system". Thesis, National Library of Canada = Bibliothèque nationale du Canada, 1998. http://www.collectionscanada.ca/obj/s4/f2/dsk1/tape11/PQDD_0001/NQ42856.pdf.

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Jernkvist, Lars Olof. "Modelling of pellet-cladding interaction induced failure of light water reactor nuclear fuel rods". Licentiate thesis, Luleå tekniska universitet, 1998. http://urn.kb.se/resolve?urn=urn:nbn:se:ltu:diva-26115.

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Konarski, Piotr. "Thermo-chemical-mechanical modeling of nuclear fuel behavior : Impact of oxygen transport in the fuel on Pellet Cladding Interaction". Thesis, Lyon, 2019. http://www.theses.fr/2019LYSEI080.

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L’objectif de cette thèse est d'étudier l'impact du transport de l’oxygène sur la thermochimie de l’interaction pastille-gaine. Pendant les rampes de puissance, le combustible nucléaire est exposé à des gradients de température élevés. Il subit des changements chimiques et structurels. Le gonflement du combustible entraîne un contact mécanique avec la gaine, provoquant des contraintes mécaniques élevées. Simultanément, des espèces chimiquement réactives sont libérées par le centre des pellets chauds et peuvent interagir avec la gaine. La combinaison de ces facteurs chimiques et mécaniques peut entraîner une défaillance de la gaine causée par la fissuration par corrosion sous contrainte. Il a été prouvé que le transport de l'oxygène sous des gradients de température élevés affecte la thermochimie, ce qui peut jouer un rôle important dans la fissuration par corrosion sous contrainte. Cette thèse présente des simulations 3D de rampes de puissance dans des réacteurs à eau sous pression avec le code de performance de combustible ALCYONE, qui fait partie de l'environnement informatique PLEIADES. Le code a été mis à jour pour associer la description de la thermochimie du carburant irradié déjà disponible au transport de l'oxygène en tenant compte de la thermo diffusion de l'oxygène. L'impact de la redistribution de l’oxygène pendant une période transitoire de puissance sur la thermochimie du combustible irradié et sur le relâchement de gaz chimiquement réactif provenant du combustible (I(g), I2(g), CsI(g), TeI2(g), Cs(g) et Cs2(g) principalement) est étudié. Les simulations montrent que la redistribution de l’oxygène, même modérée, conduit à la réduction des oxydes métalliques (dioxyde de molybdène, molybdates de césium, oxyde de chrome) au centre des pastilles de combustible et, par conséquent, au relâchement d’une quantité beaucoup plus importante de césium gazeux, en accord avec les examens post-irradiation. Les calculs tridimensionnels des quantités d'importance pour la fissuration par corrosion sous contrainte due à l'iode (contrainte circonférentielle, déformation circonférentielle, pression partielle d'iode sur gaine) sont ensuite utilisés dans des simulations de propagation de fissures de gaine
The goal of this thesis is to study the impact of oxygen transport on thermochemistry of nuclear fuel and pellet cladding interaction. During power ramps, nuclear fuel is exposed to high temperature gradients. It undergoes chemical and structural changes. The fuel swelling leads to a mechanical contact with the cladding causing high mechanical stresses in the cladding. Simultaneously, chemically reactive gas species are released from the hot pellet center and can interact with the cladding. The combination of these chemical and mechanical factors may lead to the cladding failure by iodine stress corrosion cracking. It has been proven that oxygen transport under high temperature gradients affects irradiated fuel thermochemistry, a phenomenon which may be of importance for stress corrosion cracking. This thesis presents 3D simulations of power ramps in pressurized water reactors with the fuel performance code ALCYONE, which is part of the computing environment PLEIADES. The code has been upgraded to couple the description of irradiated fuel thermochemistry already available with oxygen transport taking into account oxygen thermal diffusion. The impact of oxygen redistribution during a power transient on irradiated fuel thermochemistry in the fuel and on chemically reactive gas release from the fuel (consisting of I(g), I2(g), CsI(g), TeI2(g), Cs(g) and Cs2(g), mainly) is studied. The simulations show that oxygen redistribution, even if moderate in magnitude, leads to the reduction of metallic oxides (molybdenum dioxide, cesium molybdate, chromium oxide) at the fuel pellet center and consequently to the release of a much greater quantity of gaseous cesium, in agreement with post-irradiation examinations. The three-dimensional calculations of the quantities of importance for iodine stress corrosion cracking (hoop stress, hoop strain, iodine partial pressure at the clad inner wall) are then used in simulations of clad crack propagation
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Lemarignier, Paul. "Etude et mise au point d’un procédé de fabrication additive pour l’élaboration de combustibles nucléaires innovants". Electronic Thesis or Diss., Limoges, 2024. http://www.theses.fr/2024LIMO0108.

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Après l’accident nucléaire de Fukushima-Daiichi en mars 2011, une R&D pour l’amélioration du comportement des combustibles lors des situations accidentelles de refroidissement (dénommés « ATF » pour Accident Tolerant Fuels) a été relancée. Une des voies d’étude porte sur l’amélioration des propriétés thermiques. Du fait de la faible conductivité thermique de l’UO2, un important gradient de température radial s’établit au sein du combustible. Cette haute température à cœur réduit la marge à fusion et de fait le délai de grâce pour une intervention. L’ajout d’une phase plus conductrice sous la forme d’inserts possédant des géométries précises et optimisées permettrait, selon les modélisations, une nette augmentation de la conductivité thermique globale du combustible. Du fait de la géométrie complexe des inserts, la fabrication additive est la solution envisagée pour l’élaboration de ces pastilles composites CERMET. La technologie de fabrication additive retenue est la micro-extrusion, pour sa simplicité de mise en œuvre dans un contexte nucléaire et la possibilité d’imprimer simultanément plusieurs matériaux. Pour initier cette étude sur ces CERMET, l’alumine a été choisie comme matériau simulant technologique de l’UO2, et le molybdène comme phase conductrice. Les nombreux paramètres procédé concernant les formulations des pâtes, les paramètres d’impression et les traitements thermiques participant à la fabrication de pastilles CERMET ont été étudiés. Notamment pour rendre les pâtes extrudables par l’imprimante 3D, les formulations ont été optimisées du point de vue rhéologique permettant le respect de la géométrie fidèle au modèle CAO et un fonctionnement compatible à l’extrusion alternée des deux formulations. Les paramètres machine comme le diamètre des buses ou le débit d’extrusion ont été adaptés aux pièces à imprimer permettant d’obtenir des impressions de bonne qualité. Cependant, après déliantage puis frittage, le retrait différentiel des deux composants (alumine et molybdène) du fait d’un taux de charge et d’une cinétique de retrait différents entraine l’apparition de décohésion. Pour résoudre cette difficulté, la formulation de la phase métallique a été revue. Des formulations dites « hybrides », mélanges de diverses proportions d’alumine et de molybdène, ont apporté une nette amélioration de la cohésion du CERMET. Les propriétés thermiques de ces CERMET ont pu été évaluées selon deux méthodes de type laser flash. L’ensemble des travaux ont permis la démonstration de la faisabilité de l’impression des CERMET de structure interne complexe mais a aussi permis la mise en évidence des difficultés d’optimisation des très nombreux paramètres d’un procédé innovant du fait des nombreuses étapes de la formulation des pâtes aux différents traitements thermiques
In the wake of the Fukushima-Daiichi nuclear accident in March 2011, R&D to improve the behavior of fuels during accidental cooling situations (known as “ATF” for Accident Tolerant Fuels) was relaunched. One of the ways being explored is the improvement of thermal properties. Due to UO2's low thermal conductivity, a significant radial temperature gradient is established within the fuel. This high core temperature reduces the melting margin and hence the coping time for intervention. The addition of a more conductive phase in the form of inserts with precise, optimized geometries would, according to modelling, significantly increase the fuel's overall thermal conductivity. Given the complex geometry of the inserts, additive manufacturing is the solution envisaged for the production of these CERMET composite pellets. The additive manufacturing technology chosen is robocasting, for its simplicity of implementation in a nuclear context and the possibility of simultaneously printing several materials. To initiate this study on CERMETs, alumina was chosen as the technological simulant material for UO2, and molybdenum as the conductive phase. Numerous process parameters were studied, including paste formulations, printing parameters and heat treatments involved in the manufacture of CERMET pellets. In particular, to make the pastes extrudable by the 3D printer, the formulations were optimized from a rheological point of view, enabling them to respect the correct geometry of the CAD model, and to operate compatibly with the alternating extrusion of the two formulations. Machine parameters such as nozzle diameter and extrusion flow rate were adapted to the parts to be printed, resulting in good quality prints. However, after debinding and sintering, the differential shrinkage of the two components (alumina and molybdenum) due to different loading rates and shrinkage kinetics leads to decohesion. To solve this problem, the formulation of the metallic phase was reviewed. “Hybrid” formulations, blends of varying proportions of alumina and molybdenum, have brought a marked improvement in CERMET cohesion. The thermal properties of these CERMETs were assessed using two laser-flash methods. This work demonstrated the feasibility of printing CERMETs with a complex internal structure, but also highlighted the difficulties involved in optimizing the many parameters of an innovative process, due to the numerous stages from paste formulations to heat treatments
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SERAFIM, ANTONIO da C. "Estudo da densificação do combustível urânio - 7% gadolínio (Gd2O3) nanoestruturado". reponame:Repositório Institucional do IPEN, 2016. http://repositorio.ipen.br:8080/xmlui/handle/123456789/27502.

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O processo de sinterização de pastilhas de UO2-Gd2O3 tem sido investigado devido à sua importância na indústria nuclear e ao comportamento complexo durante a sinterização. A sinterização é bloqueada a partir de 1300°C, quando a densificação é deslocada na direção de maiores temperaturas e a densidade final obtida é diminuída. Esta pesquisa contempla o desenvolvimento de combustíveis nucleares para reatores de potência visando aumentar a sua eficiência no núcleo do reator através da elevação da taxa de queima. Foi estudado o uso do Gd2O3 de tamanho nanométrico, na faixa de 10 a 30nm, o qual foi adicionado ao UO2, visando verificar a possibilidade de evitar-se o característico bloqueio da sinterização devido ao efeito Kirkendall observado em pesquisas anteriores. As amostras foram produzidas por meio da mistura mecânica a seco dos pós de UO2 e de 7% Gd2O3 (macroestruturado e nanométrico). Os pós foram compactados e as pastilhas foram sinterizadas a 1700°C sob atmosfera de H2. Os resultados indicam que o característico bloqueio da sinterização no sistema UO2-Gd2O3 macroestruturado, que ocorre na faixa de temperatura de 1300-1500°C, retardando a densificação, foi observado de forma menos intensa quando o Gd2O3 nanométrico foi utilizado, ocorrendo à temperatura de 900°C, e facilitando a densificação posterior. Os ensaios dilatométricos indicaram uma retração de 22, 18 e 20% respectivamente nas pastilhas de UO2, UO2-7%Gd2O3 macro e UO2-7% Gd2O3nanométrico. Foi verificada uma retração 2% maior quando o Gd2O3 nanométrico foi utilizado quando comparada com a obtida com o uso do Gd2O3 macro, usado comercialmente, resultando em pastilhas com densidade adequada para uso como combustível nuclear.
Dissertação (Mestrado em Tecnologia Nuclear)
IPEN/D
Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP
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Baurens, Bertrand. "Couplages thermo-chimie mécaniques dans le dioxyde d'uranium : application à l' intéraction pastille-gaine". Thesis, Aix-Marseille, 2014. http://www.theses.fr/2014AIXM4047/document.

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En rampe de puissance, le combustible nucléaire est soumis à d'importantes contraintes thermiques et mécaniques, et subit une modification profonde de son environnement chimique. Le combustible contraint fortement la gaine, notamment au niveau des zones inter-pastilles, ce qui, associé au relâchement de produits de fission corrosifs, peut conduire à sa rupture par corrosion sous contraintes. Les évolutions simultanées de la mécanique, de la thermique et de la chimie du combustible sont liées, et participent au bon ou mauvais comportement de l'UO2 en rampe de puissance. L'objectif de ce travail est de modéliser à l'échelle d'une pastille de combustible, l'évolution couplée de la chimie, de la thermique et de la mécanique, et de préciser l'impact de ces couplages sur le comportement de l'UO2 en rampe de puissance. La finalité est d'évaluer un terme source en relâchement d'iode pour alimenter les modèles de corrosion sous contraintes dédiés aux études d'Interaction Pastille-Gaine
Nuclear fuels under power transient undergo high thermal and mechanical stresses, as well as deep chemical modifications. Stresses on the cladding at the inter-pellet plane due to the pellet thermal expansion, associated to the corrosive fission product release, can lead to clad failures, resulting from a stress corrosion cracking mechanism. The thermal, mechanical and chemical properties of the UO2 irradiated fuel are closely dependent and play a major role on the behavior of the material during a power transient. The aim of this work is to model at the pellet scale the chemical, thermal and mechanical coupled changes of the UO2 fuel during a power transient scenario and to evaluate the consequences on the fuel behavior. The final objective is to obtain an evaluation of the iodine release source term to be used in I-SCC modelling codes dedicated to Pellet-Clad-Interaction studies
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REIS, REGIS. "Análise do comportamento sob irradiação do combustível nuclear a altas queimas com os programas computacionais FRAPCON e FRAPTRAN". reponame:Repositório Institucional do IPEN, 2014. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11797.

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Dissertação (Mestrado em Tecnologia Nuclear)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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NUNES, BEATRIZ G. "Determinação exerimental de razões espectrais e do espectro de energia dos nêutrons no combustível do reator nuclear IPEN/MB-01". reponame:Repositório Institucional do IPEN, 2012. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10069.

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Dissertação (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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Lage, Aldo Márcio Fonseca. "Modelagem geométrica computacional das etapas de prensagem e sinterização de pastilhas e de laminação de placas combustíveis em dispersão de microesferas de (Th,25%U) O2 em matriz de aço inoxidável". CNEN - Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte, 2005. http://www.bdtd.cdtn.br//tde_busca/arquivo.php?codArquivo=46.

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Neste trabalho foi realizada a modelagem geométrica computacional das Cetapas de prensagem e sinterização da pastilha e da laminação da placa de combustível nuclear contendo microesferas de (Th,25%U)O2 dispersas em matriz de aço inoxidável com o objetivo de avaliar a distribuição destas microesferas nas diversas etapas do processamento. As regras de modelagem foram desenvolvidas baseadas nos parâmetros de cada etapa da fabricação da placa combustível. Para isto foram obtidas placas através do processamento por laminação de molduras de chapas de aço inoxidável, contendo pastilha fabricadas com microesferas de (Th,25%U)O2 com carregamentos de 10, 20 e 40% em peso de combustível disperso em matriz de aço inoxidável. Os dados das placas com carregamentos de 30 e 50% foram obtidos por interpolação da curva. As microesferas, obtidas pelo processo sol-gel, foram previamente secas, reduzidas e sinterizadas a 1700oC, durante 2 horas, sob atmosferas de hidrogênio. As microesferas sinterizadas alcançaram uma densidade de cerca de 98% da densidade teórica, e possuem um diâmetro médio de cerca de 300 mm e uma elevada resistência à fratura, de aproximadamente 40 N/microesfera. As regras implementadas neste modelo foram aplicadas nas coordenadas dos centros das esferas virtuais, que simulam as microesferas combustíveis de (Th,25%U)O2, obtendo-se novas coordenadas espaciais para cada uma delas nas etapas de prensagem e sinterização da pastilha e da laminação da placa combustível. Este modelo foi projetado com o uso de técnicas de análise de sistema estruturada, implementado utilizando a linguagem de programação Delphi e os resultados visualizados através do programa AutoCAD. Os resultados do modelo foram validados comparando-se as frações volumétricas experimentais em cada um dos carregamentos estudados com as frações simuladas. Este trabalho será de grande valia para o estudo do carregamento de microesferas na placa combustível, permitindo obter um combustível de elevado desempenho mecânico, térmico e neutrônico mesmo em mais alto carregamento.
The computational geometric modeling of the pressing, sintering and lamination stages for nuclear fuel plates composed by (Th,25%U)O2, microspheres dispersed into stainless steel matrix has been done in order to investigate the microspheres distribution in the various processing stages. The modeling standards were based on the parameters related to each fuel plate manufacturing stage. Accordingly, the plates were obtained through lamination processing of stainless steel plate frames comprising (Th,25%U)O2 microspheres pellets dispersed into stainless steel powder with loading of 10, 20 and 40% of microspheres dispersed into stainless steel matrix. The data for plates with loading of 30 and 50% have been obtained by linear interpolation. The microspheres produced by the sol-gel method were previously reduced and sintered at 1700 0C during 2 hours at hydrogen atmosphere. These sintered microspheres have reached about 98% of the theoretical density, with a mean diameter of 300 mm and a high resistance to fracture, near to 40 N/microsphere. The implemented standards in this model were applied at the virtual spheres center coordinates, which simulate the (Th,25%U)O2 fuel microspheres, and generate the new spatial coordinates to each of them in the pressing, sintering and lamination stages. This model was developed using structured system analysis techniques and it has been implemented using the Delphi programming language. The results were displayed through the AutoCAD program and validated comparing the experimental volumetric fractions in each of the studied loading, with the simulated fractions. The results indicate that this work could be a powerful tool in the investigation of microspheres loading in the fuel plate, allowing the attainment of a high mechanical and neutronic performance fuel, even for higher level loading.
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REZENDE, RENATO P. "Soldagem de juntas tubulares de aço inoxidável austenítico AISI 348 para varetas combustíveis em reatores nucleares". reponame:Repositório Institucional do IPEN, 2015. http://repositorio.ipen.br:8080/xmlui/handle/123456789/23883.

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Dissertação (Mestrado em Tecnologia Nuclear)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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Książki na temat "Nuclear fuel pellet"

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Commissariat à l'énergie atomique, Cadarache., Direction de l'énergie nucléaire, DEC, Electricité de France i OECD Nuclear Energy Agency, red. Pellet-clad interaction in water reactor fuels: Seminar proceedings, Aix-en-Provence, France, 9-11 March 2004. Paris: Nuclear Energy Agency, Organisation for Economic Co-operation and Development, 2005.

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Pellet Clad Interaction in Water Reactor Fuels (Nuclear Science). OECD, 2005.

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Części książek na temat "Nuclear fuel pellet"

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Zhou, Yunfei, Cheng Wang, Bin Cheng i Hongguang Yang. "Numerical Simulation of Fuel Pellet Cladding Interaction in Nuclear Reactor". W Advances in Energy Resources and Environmental Engineering, 181–88. Cham: Springer International Publishing, 2024. http://dx.doi.org/10.1007/978-3-031-42563-9_18.

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Hendricks, John S., Martyn T. Swinhoe i Andrea Favalli. "Examples for Nuclear Safeguards Applications". W Monte Carlo N-Particle Simulations for Nuclear Detection and Safeguards, 155–94. Cham: Springer International Publishing, 2022. http://dx.doi.org/10.1007/978-3-031-04129-7_3.

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AbstractFour simplified examples of actual safeguard calculations are provided. First is a combined (α,n) and spontaneous fission source in a nuclear fuel assembly in a water tank with fission detectors. Second is an HLNC2 high level neutron coincidence counter for spontaneous fission and (α,n) neutron multiplicity counting. Third is an NaI scintillation detector for a photon pulse-height tally measurement of a photon source from a cylindrical UO2 pellet. Fourth is a Cf shuffler, utilizing a 252Cf spontaneous fission source, used to measure a nuclear material item, with 3He tubes measuring the time-dependent delayed neutron buildup signature. The MCNP output from these calculations provides the data for safeguard analysis.
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Yang, Xiaoliang, Xuequan Wang, Zhe Pan, Jie Liu i Jiandong Luo. "Preliminary Application of CT Technology in Non-destructive Testing of Nuclear Fuel Elements". W Springer Proceedings in Physics, 98–106. Singapore: Springer Nature Singapore, 2023. http://dx.doi.org/10.1007/978-981-99-1023-6_10.

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AbstractWith the emergence of various novel fuel elements, traditional X-ray test technologies refer to national standards that have gradually been unable to meet the non-destructive testing (NDT) requirements for these novel fuel elements. As a new NDT technology, industrial computed tomography (CT) has great potential for NDT of nuclear fuel elements. In this paper, through a personalized transformation of self-developed X-ray equipment, we carried out CT scanning imaging experiments up to more than 400 kV on pellet-shell gap in rod-shaped fuel elements, a high-density annular component, and a tungsten-based workpiece. Not only that, after three-dimensional reconstruction and image analysis, it was found that sub-millimeter internal void defects could be detected. Furthermore, size measurements were carried out through image analysis which achieved a relative error of 5%. A conservative conclusion can be drawn from this research: industrial CT, including but not limited to micro-CT, high-energy X-ray CT, etc., has an optimistic future in testing internal defects and measuring internal dimensions of novel fuel elements.
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Kim, Ki Hwan, Jong Man Park, Don Bae Lee, Chul Goo Chi i Chang Kyu Kim. "Fabrication of Monolithic UAl2 Pellet for High-Density Nuclear Fuel". W Advanced Materials Research, 925–28. Stafa: Trans Tech Publications Ltd., 2007. http://dx.doi.org/10.4028/0-87849-463-4.925.

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Panakkal, J. P., J. K. Ghosh i P. R. Roy. "Nondestructive Characterization of Mixed Oxide Pellets in Welded Nuclear Fuel Pins by Neutron Radiography and Gamma-autoradiography". W Nondestructive Characterization of Materials, 832–38. Berlin, Heidelberg: Springer Berlin Heidelberg, 1989. http://dx.doi.org/10.1007/978-3-642-84003-6_96.

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Onder, E. Nihan. "Fuel Pellet, Element and Assembly". W Fundamentals of Nuclear Fuel, 85–98. ASME, 2023. http://dx.doi.org/10.1115/1.887158_ch6.

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Fuels that are fabricated for water-cooled power reactors are based on ceramic uranium dioxide (UO2), sintered from the powder to form fuel pellets. Fuel pellets, usually about 10 mm to 12 mm in diameter and 10 to 15 mm in length, are stacked in zirconium alloy (i.e., zircaloy) tubes, forming fuel elements. Numerous fuel elements form a fuel assembly. Note that assembly is the term used for LWR fuel and bundle is the term used for pressure-tube PHWR; however, assembly is used here in general terms. The specifications of pellets, fuel elements and assemblies are discussed next.
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Kato, Masato. "Fuel Design and Fabrication: Pellet-Type Fuel". W Encyclopedia of Nuclear Energy, 298–307. Elsevier, 2021. http://dx.doi.org/10.1016/b978-0-12-819725-7.00107-0.

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Piro, Markus H. A., Dion Sunderland, Steve Livingstone, Jerome Sercombe, R. Winston Revie, Aaron Quastel, Kurt A. Terrani i Colin Judge. "Pellet-Clad Interaction Behavior in Zirconium Alloy Fuel Cladding". W Comprehensive Nuclear Materials, 248–306. Elsevier, 2020. http://dx.doi.org/10.1016/b978-0-12-803581-8.09799-x.

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Onder, E. Nihan. "Advanced Fuel Concept". W Fundamentals of Nuclear Fuel, 203–56. ASME, 2023. http://dx.doi.org/10.1115/1.887158_ch10.

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Before the nuclear accident at the Fukushima Daiichi plant, the world-wide development efforts of advanced nuclear fuels focused on improving nuclear fuel performance in terms of reduced fuel failures, increased power density, and increased burnup. However, during the nuclear accident at the Fukushima Daiichi plant in March 2011, some undesirable performance characteristics of the conventional fuel system (i.e., UO2 fuel with -zircaloy-4 or zircaloy-2), such as rapid hydrogen production (i.e., liberation of hydrogen due to steam oxidation of zircaloy cladding, which in turn later resulted in hydrogen explosions in the Fukushima Daiichi plant), were identified. Since then, new concepts of pellet and cladding have been proposed and tested as Accident Tolerant Fuel (ATF) that can tolerate the loss of active cooling for a longer period of time, and produce less or no hydrogen, while maintaining or improving the fuel performance during normal operations, as compared to the standard UO2 – zirconium alloy fuel system [84]. ATF was originally proposed for existing LWRs, because the existing LWR fleet don’t have the improved safety features (e.g., passive safety system) as the advanced reactors (i.e., new generation reactors, Gen-IV and small modular reactors, SMR). It is difficult to implement these improved safety features into the existing reactors because of the requirements of significant design changes. Therefore, safety can be further improved for the existing reactors by utilizing a combination of fuel innovations (i.e., ATF) and operational and/or reactor design changes. Nevertheless, ATF can also be used to fuel the advanced reactors, designed to utilize similar fuels to the conventional fuel (i.e., UO2) or fuel system (i.e., UO2 - zirconium alloy).
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Onder, E. Nihan. "Nuclear Power Reactors and Their Fuels". W Fundamentals of Nuclear Fuel, 3–6. ASME, 2023. http://dx.doi.org/10.1115/1.887158_ch2.

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Ceramic oxide fuels, in particular, uranium dioxide (UO2) in the form cylindrical pellets, with zircaloy as cladding are used in the majority of operating nuclear power reactors. However, some reactors employ mixed uranium-plutonium oxide (mixed oxide, MOX2; (U,Pu)O2±x). The two major isotopes of natural uranium are 235U (0.711 wt%) and 238U 99.284 wt% with traces of 234U isotope. In the light water-cooled reactors (LWR), including pressurized water (PWR), boiling water (BWR) and water-cooled and water-moderated (VVER) power reactors, the fuel is enriched to increase the abundance of fissile 235U in the naturally occurring uranium up to slightly less than 5 wt% because the main contributor to fissioning is 235U. In the pressurized heavy3 water reactors (PHWR), excluding the fuel used in the German pressurized vessel design in Atucha in Argentina, natural uranium is used in fabricating the UO2 pellets. Due to its lower neutron absorption, heavy water (D2O) is one of the most efficient moderators to be used with natural uranium. For LWR applications, fuel needs to be enriched (up to 4.9 wt% 235U in U) because light water (H2O) not only scatters but absorbs many neutrons to be an effective moderator to be used with natural uranium.
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Streszczenia konferencji na temat "Nuclear fuel pellet"

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Ambrosek, Richard G., Robert C. Pedersen i Amanda Maple. "Modeling of MOX Fuel Pellet-Clad Interaction Using ABAQUS". W 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22142.

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Post-irradiation examination (PIE) has indicated an increase in the outer diameter of fuel pins being irradiated in the Advanced Test Reactor (ATR) for the MOX irradiation program. The diameter increase is the largest in the region between fuel pellets. The fuel pellet was modeled using PATRAN and the model was evaluated using ABAQUS, version 6.2. The results from the analysis indicate the non-uniform clad diameter is caused by interaction between the fuel pellet and the clad. The results also demonstrate that the interaction is not uniform over the pellet axial length, with the largest interaction occurring in the region of the pellet-pellet interface. Results were obtained for an axi-symmetric model and for a 1/8 pie shaped segment, using the coupled temperature-displacement solution technique.
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Klouzal, Jan, i Martin Dostál. "Modelling of the Impact of Local Effects on Fuel-Cladding Interaction During Power Ramp". W 2014 22nd International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/icone22-30807.

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The power increase rate of the reactor is often derived using the fuel performance code. Too restrictive rates are not desirable since they lead to the loss of production. On the other hand fast increase or not well controlled axial power oscillation may result in the rod failure due to pellet-cladding interaction. Most of the currently used fuel performance codes treat the stack of the fuel pellets using a simplified “1.5D” approach where individual pellets are not distinguished and the fuel stack is taken to be symmetrical. In reality, several effects must be taken into account when more accurate description is required. These local effects include contact at pellet-pellet interface, fuel pellet cracking under thermal stress, fabrication defects pellets or azimuthal asymmetry in the heat generation or heat transfer conditions due to rod bowing or presence of control rods. Detailed models of the local phenomena are therefore being developed at ÚJV Řež using the ABAQUS 6.12 code and used to improve the predictions of the codes routinely used for the core design assessment. For example the impact of the use of the advanced pellet materials on the peak loading that the cladding will experience during the power ramp has been quantified.
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Jiang, Hao, Jy-An John Wang i Hong Wang. "Potential Impact of Interfacial Bonding Efficiency on Used Nuclear Fuel Vibration Integrity During Normal Transportation". W ASME 2014 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/pvp2014-29067.

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Finite element analysis (FEA) was used to investigate the impacts of interfacial bonding efficiency at pellet–pellet and pellet–clad interfaces on surrogate of used nuclear fuel (UNF) vibration integrity. The FEA simulation results were also validated and benchmarked with reversible bending fatigue test results on surrogate rods consisting of stainless steel (SS) tubes with alumina-pellet inserts. Bending moments (M) are applied to the FEA models to evaluate the system responses of the surrogate rods. From the induced curvature, κ, the flexural rigidity EI can be estimated as EI=M/κ. The impacts of interfacial bonding efficiency include the moment carrying capacity distribution between pellets and clad and cohesion influence on the flexural rigidity of the surrogate rod system. The result also indicates that the immediate consequences of interfacial de-bonding are a load carrying capacity shift from the fuel pellets to the clad and a reduction of the composite rod flexural rigidity. Therefore, the flexural rigidity of the surrogate rod and the bending moment bearing capacity between the clad and fuel pellets are strongly dependent on the efficiency of interfacial bonding at the pellet–pellet and pellet–clad interfaces. FEA models will be further used to study UNF vibration integrity.
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Gitzhofer, F., K. Mailhot, M. I. Boulos, I. H. Jung, J. S. Lee i H. S. Park. "Fabrication of Simulated Nuclear Fuel Pellets by Induction Plasma Deposition". W ITSC 1998, redaktor Christian Coddet. ASM International, 1998. http://dx.doi.org/10.31399/asm.cp.itsc1998p1283.

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Abstract A study on induction plasma shape forming with ceramic materials, yttria stabilized zirconia ZrO2-Y2O3 (YSZ), was conducted in order to develop a new method for nuclear fuel fabrication. YSZ was selected because of its similar melting point then U02. Nuclear fuel consists of pellets Ø 10mm, thickness 12mm with a density over 96% theoretical density (TD). Process optimization was done with two different approaches: A large induction plasma flame (70mm), high number of pellets simultaneously sprayed (up to 108 pellets) and a medium power (40-50 kW) small supersonic induction plasma flame (10mm) for 1 pellet fabrication at a time. Process optimization was intensively done for the large induction plasma flame using chamber pressure, plasma plate power, powder spraying distance, sheath gas composition, probe position and particle size. The best results were 97.11% TD for 5mm thick pellets and 94% TD with the multi-pellets mold wheel type. For the single pellet approach, density as high as 99% TD has been obtained as measured on 12mm thick free standing pellets.
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Kubáň, Jan, i Radek Škoda. "Utilization of Thorium in LWR Fuels Aiming at Thermal Conductivity Improvements". W 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60300.

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One of the main drawbacks of uranium dioxide, which is used in almost all nuclear power reactors, is its low thermal conductivity. As a consequence, temperature at the center of fuel pellet is relatively high, because heat is poorly conducted away. To reach a higher level of safety, maximal temperature in any fuel pellet is one of the main limiting parameters, which restrict the fuel thermal output. This paper deals with the use of thorium in LWR fuels with the objective of fuel pellet maximal temperature reduction. Research investigating homogenous distribution of thorium dioxide (thoria) in uranium dioxide fuel has already been done and did not lead to considerable thermal conductivity improvements. The aim of this study is to investigate heterogeneous distribution of thorium in commonly used uranium dioxide fuel in the form of uranium and thorium pellets placed together.
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Tang, Changbing, Yongjun Jiao, Wenjie Li, Tao Qing, Yifei Miao i Ping Chen. "Numerical Simulation of Different Sizes Missing Pellet Surface Effects on Thermal-Mechanical Behaviors in Nuclear Fuel Rods". W 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60116.

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Nuclear fuel rods is mainly composed of uranium dioxide pellets and zirconium alloy cladding, there is a gap between pellets and cladding, which is filled with helium. Under the reactor operation conditions, pellets produce a lot of heat by nuclear fission reactions and at the same time also produce lots of radioactive fission products. Cladding serve as the first barrier to accommodate radioactive fission product, needs to maintain its structural integrity under the reactor operation conditions. Cladding stresses can be effectively limited by controlling power increase rates. However, pellet manufacturing defects such as missing pellet surface (MPS), may lead to cladding local stress significantly high to cause cladding failure. Simulating the impact of these defects correctly can help prevent these types of failure. MPS defects are 3D phenomenon, needs 3D modeling method to study the influence of these defects on the cladding .In this paper, stress update algorithm is derived, with the help of ABAQUS (a commercial finite element software), simulated the thermal-mechanical behaviors of the MPS defects fuel rod with a 3D FEM and completed the sensitivity analysis of MPS defects size for the fuel performance. The models included in this simulation, including pellet irradiation swelling (fission gas products induced swelling and fission solid products induced swelling), pellet densification, pellet relocation, pellet thermal expansion, pellet irradiation creep, pellet irradiation hardening, cladding irradiation growth, cladding thermal expansion, cladding thermal creep, cladding irradiation creep, cladding irradiation hardening and gap heat transfer (gas heat conduction, radiation heat transfer and contact heat conduction) etc. Furthermore, considering the effects of irradiation and temperature on the material parameters such as thermal conductivity, specific heat and young’s modulus etc. According to the simulation result, showing that MPS defects have a large impact on the performance of fuel rods, this impact will be more obvious with the size of MPS defects increase. The MPS defects cause larger gap distance between pellet and cladding, higher gap distance causes smaller gap conductance, and then causes elevated temperature at the center of the pellet and in the region of the pellet adjacent to the defect. The cladding temperature is reduced in the area immediately across from the defect, and is elevated in neighboring areas. Meanwhile, MPS defects clearly have a significant effect on stress distribution and maximum stress of the cladding, cause high tensile stresses in the inner surface of the cladding and high compressive stresses on the outer surface of the cladding at the center of the defect. Around the boundaries of the defect, the stresses are reversed, with high compressive stresses on the cladding interior and high tensile stresses on the cladding exterior.
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Li, Songyang, Dingqu Wang, Wenli Guo i Yueyuan Jiang. "Analysis and Prospect of the Duplex Fuel Pellets of LOWI Type for Water-Cooled Reactors". W 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60505.

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The duplex pellets under a “Low-Interact” (LOWI) nuclear fuel design, which consist of an outer enriched annulus and a depleted or natural core, can provide lower center temperature and reduced probability of pellet-clad mechanical interact (PCMI). Analysis and experiments were done in 1970s to examine the benefits and cost of LOWI design for water-cooled reactors. Results showed that the additional economic cost of this design should not be neglected in spite of the benefits. However, due to the improvement of nuclear fuel fabrication technology in the past 30 years, the benefits of LOWI design become more significant, especially when the potential of other methods to elevate the power density and overcome the constraints on ramp rates in power reactors is running out. In order to evaluate the feasibility of deploying the LOWI fuel in commercial and research reactors, neutronics and thermal calculations are made to figure out the performance of duplex UO2 pellets in particular reactors. It is shown that the center temperature of pellet has been greatly reduced without any change on assembly and core geometry, which means the opportunity of less fission gas production, higher power density and more adequate safety margin. A mechanical analysis of a typical LOWI design is also done. The challenges on duplex pellet manufacture are also discussed. Several fabrication techniques are presented to show the potential of cutting the cost of pellet production.
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Gamble, Kyle A. L., Anthony F. Williams i Paul K. Chan. "A Three-Dimensional Analysis of the Local Stresses and Strains at the Pellet Ridges in a Horizontal Nuclear Fuel Element". W 2014 22nd International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/icone22-30023.

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A three-dimensional finite element model is being developed for a quarter fuel element, which is equivalent to a full fuel element using symmetry. The model uses the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework developed at Idaho National Laboratory. The model facilitates an in-depth investigation into a variety of deformation phenomena for a horizontal nuclear fuel element including bowing, sagging, and stresses and strains. This paper presents a preliminary analysis of the local stresses and strains of the sheath (clad) at the pellet-to-pellet interfaces for low, normal and high linear powers. During irradiation the fuel pellets thermally expand and take on an hourglass shape. The hourglassing behaviour leads to higher local stresses and strains in the sheath at the locations of the pellet-to-pellet interfaces. The purpose of this work is to quantify these stresses and strains for varying linear powers, and to illustrate the effect that the material model chosen for the cladding has on the results. Preliminary results are presented for two sheath types: elastic, and elastic including diffusional creep. These models are benchmarked against a validated industry code called ELESTRES. The results indicate that the predicted sheath hoop strain is about half of what is determined by ELESTRES in both the elastic and elastic-creep cases. This highlights the requirement of a pellet cracking model in three-dimensional simulations. The elastic-creep model predicts less stress within the sheath than the elastic model as expected.
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Li, Jiwei, Yang Ding, Wentao Liu, Guangwen Bi, Ruirui Zhao i Qin Zhou. "Out-of-Pile Properties Investigation of UO2-BeO Fuel Pellet". W 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-66585.

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In order to increasing the thermal conductivity of fuel pellet used in nuclear power plants, a UO2-BeO composite fuel was developed. The fuel pellets with different beryllia addition were manufactured by sintering and the physical properties were tested and compared with uranium dioxide fuel. The metallograph show that the continuous phase of beryllia was formed. The results show that the thermal conductivity was obviously larger and the elasticity modulus and the coefficient of thermal expansion were a little smaller than uranium dioxide. The thermal conductivity and the elasticity modulus were increased and the coefficient of thermal expansion was decreased by increasing the ratio of beryllia contents.
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Zhu, Wang, Zhang Chunyu, Li Aolin i Yuan Cenxi. "Three Dimensional Modeling of the Thermo-Mechanical Performance of the Fuel Rods of a PWR". W 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-66010.

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The fuel rods of pressurized water reactors operate under complex radioactive, thermal and mechanical conditions. Multiphysics has to be taken into account in order to evaluate their performance. Many existing fuel rod codes make great simplifications on analyzing the behavior of fuel rods. The present study develops a three dimensional module within the framework of a general-purpose finite element solver, i.e. ABAQUS, for modeling the thermo-mechanical performance of the fuel rods. A typical fuel rod is modeled and the temperature as well as the stress within the pellets are computed. The results show that the burnup levels have an important influence on the fuel temperature. The swelling of fission products cause dramatically increasing of pellet strain. The change of the cladding stress and radial displacement with the axial length can be reasonably predicted. It is shown that a quick power ramp or a reactivity insertion accident can induce high tensile stress to the outer regime of the pellet and may cause further fragmentation to the pellets.
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Raporty organizacyjne na temat "Nuclear fuel pellet"

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S. Keyvan. Intelligent Automated Nuclear Fuel Pellet Inspection System. Office of Scientific and Technical Information (OSTI), listopad 1999. http://dx.doi.org/10.2172/754854.

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Wang, Jy-An, Bruce Bevard, John Scaglione i Rose Montgomery. Fracture toughness evaluations for spent nuclear fuel dry storage canister welds and spent nuclear fuel clad-pellet structures. Office of Scientific and Technical Information (OSTI), kwiecień 2021. http://dx.doi.org/10.2172/1782033.

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Kips, R. Argentina-LLNL-LANL Comparative Sample Analysis on UO2 fuel pellet CRM-125A for Nuclear Forensics. Office of Scientific and Technical Information (OSTI), grudzień 2017. http://dx.doi.org/10.2172/1413178.

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Battaglia, Francine. Detailed Reaction Kinetics for CFD Modeling of Nuclear Fuel Pellet Coating for High Temperature Gas-Cooled Reactors. Office of Scientific and Technical Information (OSTI), listopad 2008. http://dx.doi.org/10.2172/942124.

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Asgari, Mehdi, Jake Hirschhorn, Eva Davidson, Dave Kropaczek, Andrew Godfrey i Ryan Sweet. Final Summary Report on the Feasibility and the Benefits of the Advanced Nuclear Fuel Pellet Designs with Radially Varying Fuel Zoning and Burnable Poison Concentration. Office of Scientific and Technical Information (OSTI), lipiec 2022. http://dx.doi.org/10.2172/1958390.

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D.E. Clark i D.C. Folz. Microwave Processing of Simulated Advanced Nuclear Fuel Pellets. Office of Scientific and Technical Information (OSTI), sierpień 2010. http://dx.doi.org/10.2172/992637.

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