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Dissertations / Theses on the topic 'Nuclear fuel cladding tube'

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Published: 4 June 2021

Last updated: 12 February 2022

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1

Paramonova, Ekaterina (Ekaterina D. ). "CRUD resistant fuel cladding materials." Thesis, Massachusetts Institute of Technology, 2013. http://hdl.handle.net/1721.1/82447.

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Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2013.<br>"June 2013." Cataloged from PDF version of thesis.<br>Includes bibliographical references (pages 27-29).<br>CRUD is a term commonly used to describe deposited corrosion products that form on the surface of fuel cladding rods during the operation of Pressurized Water Reactors (PWR). CRUD has deleterious effects on reactor operation and currently, there is no effective way to mitigate its formation. The Electric Power Research Institute (EPRI) CRUD Resistant Fuel Cladding project has the obje

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2

Drieux, Patxi. "Elaboration de tubes épais de SiC par CVD pour applications thermostructurales." Phd thesis, Université Sciences et Technologies - Bordeaux I, 2013. http://tel.archives-ouvertes.fr/tel-00958465.

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L'objectif de la thèse était de synthétiser des tubes de SiC monolithiques pour améliorer l'étanchéité de la structure composite SiC/SiC d'une gaine de combustible nucléaire. Des revêtements tubulaires de 8 mm de diamètre et quelques centaines de micromètres d'épaisseur ont été produits par dépôt chimique en phase vapeur à pression atmosphérique à partir d'un mélange CH3SiHCl2/H2. Le procédé a été développé de manière à réaliser en continu des tubes de SiC de plusieurs dizaines de centimètres de long. La composition chimique et la microstructure des tubes ont été déterminées par microsonde de

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3

Jarvis, Jennifer Anne. "Hydrogen entry in Zircaloy-4 fuel cladding : an electrochemical study." Thesis, Massachusetts Institute of Technology, 2015. http://hdl.handle.net/1721.1/103730.

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Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2015.<br>Cataloged from PDF version of thesis.<br>Includes bibliographical references (pages 291-297).<br>Corrosion and hydrogen pickup of zirconium alloy fuel cladding in water cooled nuclear reactors are life-limiting phenomena for fuel. This thesis studies the fate of hydrogen liberated by waterside corrosion of Zircaloy-4 fuel cladding in Pressurized Water Reactors (PWRs): are the adsorbed protons incorporated into the oxide and eventually the metal, or are they evolved into molecular hydr

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4

Lee, Youho. "Safety of light water reactor fuel with silicon carbide cladding." Thesis, Massachusetts Institute of Technology, 2013. http://hdl.handle.net/1721.1/86866.

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Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2013.<br>Cataloged from PDF version of thesis.<br>Includes bibliographical references (pages 303-314).<br>Structural aspects of the performance of light water reactor (LWR) fuel rod with triplex silicon carbide (SiC) cladding - an emerging option to replace the zirconium alloy cladding - are assessed. Its behavior under accident conditions is examined with an integrated approach of experiments, modeling, and simulation. High temperature (1100°C~1500°C) steam oxidation experiments demonstrated

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5

Reece, Warren Daniel. "Theory of cladding breach location and size determination using delayed neutron signals /." Diss., Georgia Institute of Technology, 1988. http://hdl.handle.net/1853/13317.

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6

Jena, Anupam S. M. Massachusetts Institute of Technology. "Wettability of candidate Accident Tolerant Fuel (ATF) cladding materials in LWR conditions." Thesis, Massachusetts Institute of Technology, 2020. https://hdl.handle.net/1721.1/127299.

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Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, May, 2020<br>Cataloged from the official PDF of thesis.<br>Includes bibliographical references (pages [69]-70).<br>Since the 2011 Fukushima accident, substantial research has been dedicated to developing accident tolerant fuel (ATF) cladding materials. These analyses have mostly concentrated on the capability of potential ATF materials to withstand runaway steam oxidation and preserve their mechanical strength and structural integrity under thermal shocks. However, knowledge is relatively defici

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7

Seshadri, Arunkumar. "Impact of reactor environment on quenching heat transfer of accident tolerant fuel cladding." Thesis, Massachusetts Institute of Technology, 2018. https://hdl.handle.net/1721.1/121711.

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This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.<br>Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2018<br>Cataloged from student-submitted PDF version of thesis. Page 123 blank.<br>Includes bibliographical references (pages 106-116).<br>Development of accident tolerant fuels (ATF) for light water reactors (LWRs) came into focus for the nuclear engineering community after the accidents at Fukushima-Daiichi. The primary focus of the ATF program is

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8

Stempien, John D. (John Dennis). "Behavior of triplex silicon carbide fuel cladding designs tested under simulated PWR conditions." Thesis, Massachusetts Institute of Technology, 2011. http://hdl.handle.net/1721.1/76948.

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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011.<br>Cataloged from PDF version of thesis.<br>Includes bibliographical references (p. 101-107).<br>A silicon carbide (SiC) fuel cladding for LWRs may allow a number of advances, including: increased safety margins under transients and accident scenarios, such as loss of coolant accidents; improved resource utilization via a higher burnup beyond the present limit of 62 GWd/MTU; and improved waste management. The proposed design, referred to as Triplex, consists of three layers: an inner monolith,

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9

Al, Shater Abdulla Faisal. "Intergranular corrosion of sensitized 20Cr-25Ni-Nb stainless steel nuclear fuel cladding materials." Thesis, University of Manchester, 2010. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.706485.

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10

Auguste, Rasheed. "Quantifying the fouling resistance of Accident-Tolerant Fuel (ATF) cladding coatings with force spectroscopy." Thesis, Massachusetts Institute of Technology, 2017. http://hdl.handle.net/1721.1/112377.

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Thesis: S.B., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2017.<br>This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.<br>Cataloged from student-submitted PDF version of thesis.<br>Includes bibliographical references (pages 418-420).<br>CRUD (Chalk River Unidentified Deposits) is buildup of metal oxides on the interior of nuclear reactors. This is caused by corrosion in reactor internals, leading to problems such as coolant contamination in porous deposits le

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11

Li, Zhen. "SURFACE HARDENING OF AUSTENITIC FE–CR–NI ALLOYS FOR ACCIDENT-TOLERANT NUCLEAR FUEL CLADDING." Case Western Reserve University School of Graduate Studies / OhioLINK, 2018. http://rave.ohiolink.edu/etdc/view?acc_num=case150486174877088.

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12

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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13

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

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14

Mattingly, Brett T. (Brett Thomas). "Performance analysis of matrix fuel for a passive pressure tube light water reactor." Thesis, Massachusetts Institute of Technology, 1995. http://hdl.handle.net/1721.1/38099.

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15

Guenoun, Pierre S. M. Massachusetts Institute of Technology. "Design optimization of advanced PWR SiC/SiC fuel cladding for enhanced tolerance of loss of coolant conditions." Thesis, Massachusetts Institute of Technology, 2016. http://hdl.handle.net/1721.1/103649.

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Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2016.<br>This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.<br>Cataloged from student-submitted PDF version of thesis.<br>Includes bibliographical references (pages 64-68).<br>Limited data has been published (especially on experimental work) on integrated multilayer SiC/SiC prototypical fuel cladding. In this work the mechanical performance of three unique architectures of three-layer silicon carbide

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16

Fray, Elliott Shepard. "Materials testing and development of functionally graded composite fuel cladding and piping for the Lead-Bismuth cooled nuclear reactor." Thesis, Massachusetts Institute of Technology, 2013. http://hdl.handle.net/1721.1/82456.

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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2013.<br>Cataloged from PDF version of thesis.<br>Includes bibliographical references (pages 176-179).<br>This study has extended the development of an exciting technology which promises to enable the Pb-Bi eutectic cooled reactors to operate at temperatures up to 650-700°C. This new technology is a functionally graded composite steel which resists high temperature LBE corrosion. This composite steel consists of a Fel2Cr2Si protective layer weld overlaid on a T91 steel and then drawn to fuel claddin

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17

Gudipati, Mithun. "Computational fluid dynamics simulations of basket and fuel cladding temperatures within a rail cask during normal transport." abstract and full text PDF (free order & download UNR users only), 2007. http://0-gateway.proquest.com.innopac.library.unr.edu/openurl?url_ver=Z39.88-2004&rft_val_fmt=info:ofi/fmt:kev:mtx:dissertation&res_dat=xri:pqdiss&rft_dat=xri:pqdiss:1446432.

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18

Bell, Benjamin. "The influence of alloying elements on the corrosion of Zr-based nuclear fuel cladding using density functional theory." Thesis, Imperial College London, 2016. http://hdl.handle.net/10044/1/51546.

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Zr-based alloys are used primarily as fuel cladding in water-cooled nuclear fission reactors. This is due to their good thermal and mechanical properties and low capture cross section for thermal neutrons. In this work, density functional theory (DFT) simulations were performed to investigate the behaviour of dopant elements in the cladding alloy oxide, with the aim of furthering the understanding of corrosion and hydrogen pick-up in Zr-based alloys. Simulations were performed in both monoclinic and tetragonal ZrO2 on single isolated defects and defect clusters, with the effect of compressive

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19

Macdonald, Vincent. "Détermination d’un critère de rupture des gaines de Zircaloy-4 détendu hydruré contenant un blister d’hydrures, en conditions d’accident d’injection de réactivité." Thesis, Paris Sciences et Lettres (ComUE), 2016. http://www.theses.fr/2016PSLEM038/document.

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Cette étude porte sur la détermination d’un critère de rupture des gaines de combustible de Zircaloy-4 détendu hydruré contenant un blister d’hydrures, en conditions accidentelles représentatives d’un accident d’injection de réactivité. Deux plages de comportement différentes en fonction de la température ont clairement été mises en évidence grâce à l’étude bibliographique, aux différentes campagnes d’essais mécaniques et aux analyses des faciès de rupture des éprouvettes rompues : une rupture de type fragile pour la gaine à 25°C et une rupture ductile à 350°C.A 25°C, la rupture fragile a été

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20

Matsson, Ingvar. "Studies of Nuclear Fuel Performance Using On-site Gamma-ray Spectroscopy and In-pile Measurements." Doctoral thesis, Uppsala : Acta Universitatis Upsaliensis : Univ.-bibl. [distributör], 2006. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-6912.

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21

Haurais, Florian. "Evaluate the contribution of the fuel cladding oxidation process on the hydrogen production from the reflooding during a potential severe accident in a nuclear reactor." Thesis, Université Paris-Saclay (ComUE), 2016. http://www.theses.fr/2016SACLS375/document.

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En centrales nucléaires, un accident grave est une séquence très peu probable d’événements durant laquelle des composants du réacteur sont significativement endommagés, par interactions chimiques et/ou fusion, à cause de très hautes températures. Cela peut mener à des rejets radiotoxiques dans l’enceinte et à une entrée d’air dans le réacteur. Dans ce contexte, ce travail de thèse mené chez EDF R&D visait à modéliser la détérioration du gainage combustible, en alliages de zirconium, en conditions accidentelles : haute température et soit vapeur soit mélange air-vapeur. L’objectif final éta

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22

Morgan, Andrew. "JOINING AND HERMETIC SEALING OF SILICON CARBIDE USING IRON, CHROMIUM, AND ALUMINUM ALLOYS." VCU Scholars Compass, 2014. http://scholarscompass.vcu.edu/etd/3529.

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Silicon Carbide (SiC) is increasingly gaining attention as a potential fuel cladding material, on account of its favorable thermo-mechanical and neutronic properties. The major limitations of such a cladding is currently associated with joining and hermetic sealing. The work presented here investigated the use of Al, Cr and Fe metals and a specialized alloy (FeCrAl) to achieve hermetic sealing of SiC tubes as well as a joining technology of SiC. Major part of solving this issue requires addressing joining of ceramic and metallic components, which are largely dissimilar in both thermal and mech

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23

Pereira, Luiz Alberto Tavares. "Desenvolvimento de processos de reciclagem de cavacos de Zircaloy via refusão em forno elétrico a arco e metalurgia do pó." Universidade de São Paulo, 2014. http://www.teses.usp.br/teses/disponiveis/85/85134/tde-27052014-090225/.

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Reatores PWR empregam, como combustível nuclear, pastilhas de UO2 acondicionadas em tubos de ligas de zircônio, chamados de encamisamento. Na sua fabricação são gerados cavacos de usinagem que não podem ser descartados, pois a reciclagem deste material é estratégica quanto aos aspectos de tecnologia nuclear, econômicos e ambientais. As ligas nucleares têm altíssimo custo e não são produzidas no Brasil, sendo importadas para a fabricação do combustível nuclear. Neste trabalho são abordados dois métodos para reciclar os cavacos de Zircaloy. No primeiro, os cavacos foram fundidos utilizando um fo

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24

Tioka, Jakub. "Výpočetní a experimentální analýzy jaderných paliv nové generace." Master's thesis, Vysoké učení technické v Brně. Fakulta elektrotechniky a komunikačních technologií, 2021. http://www.nusl.cz/ntk/nusl-442550.

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The search for Accident tolerant fuels (ATF) which is the first part of this thesis is currently one of the most actual topics in the field of nuclear fuels. These fuels must be first successfully tested in operational and also accident conditions for their possible inclusion in commercial use. Following part of the thesis specifically focuses on the boiling crisis in nuclear reactors which can damage the nuclear fuel cladding. Therefore, it is necessary to know the critical heat flux value and the departure from nuclear boiling ratio. Calculations which determine critical heal flux value are

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25

Halabuk, Dávid. "Zhodnocení termomechanického chování perspektivních jaderných paliv při havárii s vnosem reaktivity." Master's thesis, Vysoké učení technické v Brně. Fakulta strojního inženýrství, 2016. http://www.nusl.cz/ntk/nusl-242902.

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The objective of this master’s thesis is to simulate thermo-mechanical behaviour of nuclear fuel in a pressurized water reactor during a reactivity initiated accident. An important part of this work is focused on examination of processes which occur during such accident and on creation of a detailed overview of material properties of nuclear fuel and fuel cladding which are necessary for simulations that closely reflect reality. Simulations in this thesis examine cases of fresh or irradiated nuclear fuel for two types of fuel cladding, Zircaloy-4, a material that is currently used in nuclear r

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26

Quaranta, Delphine. "Étude de voies potentielles pour le recyclage du zirconium des gaines en Zircaloy des combustibles nucléaires usés." Thesis, Toulouse 3, 2019. http://www.theses.fr/2019TOU30038.

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Le Zircaloy-4 est un alliage à base de zirconium (~ 98 % massique) constituant le gainage des assemblages nucléaires. Actuellement, les gaines de Zircaloy irradiées sont destinées au stockage géologique profond en raison de leur contamination en radioéléments (contamination issue du séjour en réacteur ainsi que du procédé de traitement). Elles sont classées en déchet de moyenne activité à vie longue suivant les recommandations de l'ANDRA (radioactivité : 10 6 - 10 9 Bq/g, périodes > 31 ans). Les gaines de Zircaloy irradiées représentent une part importante de l'inventaire des assemblages, ~

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27

Čásar, Ondřej. "Výpočet chování paliva reaktorů VVER programem FEMAXI-6." Master's thesis, Vysoké učení technické v Brně. Fakulta elektrotechniky a komunikačních technologií, 2019. http://www.nusl.cz/ntk/nusl-400560.

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The submitted master thesis deals with finding the right combination of models describing the cladding behavior implemented in the FEMAXI-6 computational code and then comparing it with the benchmark Zaporoshye, Novovoronezh and the modified FERMAXI 6 program with implemented models describing the E110 behavior used for VVER-type reactor fuel rods. Initially, there is a description of the FEMAXI-6 nuclear fuel analysis program including its structure, calculation mechanics and input file description. Furthermore, the work presents the benchmarks used to evaluate individual combinations of fuel

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28

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

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29

Yang, Kuo-ching, and 楊國慶. "Study of nuclear fuel cladding failure mechanisms." Thesis, 2010. http://ndltd.ncl.edu.tw/handle/64347835755762097066.

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博士<br>國立臺灣科技大學<br>機械工程系<br>98<br>The delayed hydride cracking and creep rupture are the failure mechanisms of spent fuel cladding for concerns during dry storage. The regulation requires that the spent fuel be readily retrievable from the storage system, therefore, the cladding integrity during dry storage should be well demonstrated. Concerning the delayed hydride cracking, experimental data and finite element computer code are applied and hydride reorientation is considered in our analysis. For creep rupture, C* is applied in the crack stability evaluation for a 20-year period of storage. Hy

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30

Hsu, Hsiao-Hung, and 許曉弘. "Hydrogen Embrittlement on Fracture Behavior of Zircaloy Nuclear Fuel Cladding." Thesis, 2012. http://ndltd.ncl.edu.tw/handle/87566640817187408566.

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博士<br>國立臺灣海洋大學<br>材料工程研究所<br>100<br>Hydrogen embrittlement is one of the major degradation mechanisms for high burnup Zircaloy fuel cladding found during both reactor operation and spent fuel dry storage. The effects of heat treatment, zirconium hydride, and temperature on the fracture toughness of Zircaloy-4 cladding were evaluated by an X-specimen test in this study. The X-specimen test was designed to measure the fracture toughness of thin-walled tube like fuel cladding. The results obtained from Zircaloy-4 claddings by X-specimen tests show good reproducibility and are in agreement with li

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31

Li, Wan-Yun, and 李宛芸. "The Development and Application of the Nuclear Fuel Cladding Behavior Evaluation Methodology." Thesis, 2018. http://ndltd.ncl.edu.tw/handle/xwr5p4.

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Ren, Yongli. "Fracture toughness behavior of Zircaloy-4 in the form of fuel cladding tubing in nuclear reactors." 2004. http://link.library.utoronto.ca/eir/EIRdetail.cfm?Resources__ID=81168&T=F.

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Kelley, Randy Dean. "Design of an Integrated System to Recycle Zircaloy Cladding Using a Hydride-Milling-Dehydride Process." Thesis, 2010. http://hdl.handle.net/1969.1/ETD-TAMU-2010-08-8487.

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A process for recycling spent nuclear fuel cladding, a zirconium alloy (Zircaloy), into a metal powder that may be used for advanced nuclear fuel applications, was investigated to determine if it is a viable strategy. The process begins with hydriding the Zircaloy cladding hulls after the spent nuclear fuel has been dissolved from the cladding. The addition of hydrogen atoms to the zirconium matrix stresses the lattice and forms brittle zirconium hydride, which is easily pulverized into a powder. The dehydriding process removes hydrogen by heating the powder in a vacuum, resulting in a zirconi

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(9187205), Jonova Thomas. "MICROSTRUCTURAL CHARACTERIZATION AND MECHANICAL PROPERTY ASSESSMENT OF A NEUTRON IRRADIATED URANIUM-ZIRCONIUM NUCLEAR FUEL AND HT9 CLADDING." Thesis, 2020.

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<div>Metallic uranium-10 weight percent zirconium (U-10wt.%Zr) nuclear fuels are classified as potential fuels for fast breeder reactors as they possess a high fissile density and have increased compatibility with sodium, a frequently used reactor coolant. Despite their advantages when exposed to neutron irradiation in reactors, the fuels are subject to damage cascades and microstructural alterations. Fuel constituent re-distribution, phase transformation, fuel swelling, and fuel cladding chemical interactions (FCCI) are a few of the major interdependent microstructural alterations that occur

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Holik, III Eddie Frank (Trey). "Simulation results of an inductively-coupled rf plasma torch in two and three dimensions for producing a metal matrix composite for nuclear fuel cladding." 2008. http://hdl.handle.net/1969.1/ETD-TAMU-2363.

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I propose to develop a new method for the synthesis of metal matrix composites (MMC) using aerosol reactants in a radio frequency (RF) plasma torch. An inductivelycoupled RF plasma torch (ICPT) may potentially be designed to maintain laminar flow and a radial temperature distribution. These two properties provide a method by which a succession of metal layers can be applied to the surface of SiC fibers. In particular, the envisaged method provides a means to fully bond any desired metal to the surface of the SiC fibers, opening the possibility for MMCs in which the matrix metal is a highstreng

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"Zircaloy-4 and Incoloy 800H/HT Alloys for the Current and Future Nuclear Fuel Claddings." Thesis, 2015. http://hdl.handle.net/10388/ETD-2015-01-1885.

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Fuel cladding is one of the most critical components of nuclear reactors; so it is important to improve our understanding of various properties and behaviors of the cladding under different conditions approximating the nuclear reactor environment. Moreover, the efficiency of energy production, in addition to safety concerns, has resulted in progressive improvement of nuclear reactors design from Generation I to Generation IV. To complement this progressive trend, materials used for fuel cladding need to be improved or new materials should be developed. In this thesis, I address problems in the

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