Relevant bibliographies by topics / Nuclear fuel cladding tube

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  1. Journal articles
  2. Dissertations / Theses
  3. Books
  4. Book chapters
  5. Conference papers
  6. Reports

Academic literature on the topic 'Nuclear fuel cladding tube'

Author: Grafiati
Published: 4 June 2021
Last updated: 12 February 2022

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Journal articles on the topic "Nuclear fuel cladding tube"

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Li, Jing, Sa Jian Wu, Yong Li Wang, Liang Yin Xiong, and Shi Liu. "Performance of 14Cr ODS-FeCrAl Cladding Tube for Accident Tolerant Fuel." Materials Science Forum 1016 (January 2021): 806–12. http://dx.doi.org/10.4028/www.scientific.net/msf.1016.806.

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In the framework of Accident tolerant fuel (ATF) program, several types of claddings and pellets with enhanced accident tolerance have been developed for light water reactors. Oxide dispersion strengthened (ODS) FeCrAl alloys have been considered as a promising candidate for cladding materials due to their good mechanical strength, excellent structural stability and chemical durability at high temperature. The out-of-pile performance of 14Cr ODS-FeCrAl cladding tube fabricated by cold-rolling, such as microstructure, thermophysical property, mechanical property, and corrosion resistance, has b
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Kim, Jin Seon, Yong Hwan Kim, Seung Jae Lee, and Young Ze Lee. "Fretting Wear Damage of the Corroded Fuel Cladding Tubes for Nuclear Fuel Rod against Supporting Grids." Key Engineering Materials 345-346 (August 2007): 709–12. http://dx.doi.org/10.4028/www.scientific.net/kem.345-346.709.

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Fuel cladding tubes in nuclear fuel assembly are held up by supporting grids because the tubes are long and slender. Fluid flows of high-pressure and high-temperature in the tubes cause oscillating motions between tubes and supports. This is called as FIV (flow induced vibration), which causes fretting wear in contact parts of tube-support. The fretting wear of tube-support can threaten the safety of nuclear power plant. Therefore, a research about the fretting wear characteristics of tube-support is required. The fretting wear tests were performed with supporting grids and cladding tubes, esp
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Kim, Young-Hwan, Yung-Zun Cho, and Jin-Mok Hur. "Experimental Approaches for Manufacturing of Simulated Cladding and Simulated Fuel Rod for Mechanical Decladder." Science and Technology of Nuclear Installations 2020 (January 24, 2020): 1–12. http://dx.doi.org/10.1155/2020/1905019.

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We are developing a practical-scale mechanical decladder that can slit nuclear spent fuel rod-cuts (hulls + pellets) on the order of several tens of kgf of heavy metal/batch to supply UO2 pellets to a voloxidation process. The mechanical decladder is used for separating and recovering nuclear fuel material from the cladding tube by horizontally slitting the cladding tube of a fuel rod. The Korea Atomic Energy Research Institute (KAERI) is improving the performance of the mechanical decladder to increase the recovery rate of pellets from spent fuel rods. However, because actual nuclear spent fu
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Park, Young Chang, Yong Hwan Kim, Seung Jae Lee, and Young Ze Lee. "The Predictions of the Fretting Wear between Supporting Grids and Cladding Tubes of Nuclear Fuel Rod." Key Engineering Materials 326-328 (December 2006): 1243–46. http://dx.doi.org/10.4028/www.scientific.net/kem.326-328.1243.

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Fretting can be defined as the oscillatory motion with very small amplitudes, which usually occur between two contacting surfaces. Fretting wear is the removal of material from contacting surfaces through fretting action. This fretting wear, which occurs between cladding tubes of nuclear fuel rod and grids, causes in damages the cladding tubes by flow induced vibration in a nuclear reactor. In this paper the fretting wear tests were performed with two types of cladding tubes and three types of supporting grids in water. Fretting wear tests were done using various applied loads. From the result
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FUJITA, Kazumi, and Tsutomu KAKUMA. "Fabrication system of zircaloy nuclear fuel cladding tube." Journal of the Atomic Energy Society of Japan / Atomic Energy Society of Japan 29, no. 6 (1987): 487–92. http://dx.doi.org/10.3327/jaesj.29.487.

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Mahendra Prabhu, N., K. A. Gopal, S. Murugan, et al. "Determining the feasibility of identifying creep rupture of stainless steel cladding tubes on-line using acoustic emission technique." International Journal of Structural Integrity 6, no. 3 (2015): 410–18. http://dx.doi.org/10.1108/ijsi-08-2014-0038.

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Purpose – The purpose of this paper is to determine the feasibility of identifying the creep rupture of reactor cladding tubes using acoustic emission technique (AET). Design/methodology/approach – The creep rupture tests were carried out by pressuring stainless steel capsules upto 6 MPa at room temperature and then heating continuously in a furnace upto rupture. The acoustic emission (AE) signals generated during the creep rupture tests were recorded using a 150 kHz resonant sensor and analysed using AE Win software. Findings – When rupture occurs in the pressurized capsule tube representing
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Le Roux, S. D., and D. J. Van der Merwe. "Texture Analysis in Zircaloy Cladding Tube Material for Nuclear Fuel." Materials Science Forum 157-162 (May 1994): 1455–62. http://dx.doi.org/10.4028/www.scientific.net/msf.157-162.1455.

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Murugan, Aravind, R. Sai Santhosh, Ravikumar Raju, A. K. Lakshminarayanan, and Shaju K. Albert. "Dissimilar and Similar Laser Beam and GTA Welding of Nuclear Fuel Pin Cladding Tube to End Plug Joint." Advanced Engineering Forum 24 (October 2017): 40–47. http://dx.doi.org/10.4028/www.scientific.net/aef.24.40.

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The end plug to cladding tube of fast reactor fuel pin is normally welded using Gas Tungsten Arc Welding (GTAW) process. The GTAW process has large heat input and wide heat-affected-zone (HAZ) than high energy density process such as laser welding. In the present study Laser Beam Welding (LBW) is being considered as an alternative welding process to join end plug to clad tube. The characteristics of autogenous processes such as GTAW and pulsed Nd-YAG laser welding on fuel cladding tube to end plug joints have been investigated in this study. Dissimilar combinations of modified stainless steel
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Zelenskii, V. F., I. M. Neklyudov, B. P. Chernyi, et al. "Centrifugal vacuum gasting for fuel cladding tube blanks." Soviet Atomic Energy 67, no. 1 (1989): 531–33. http://dx.doi.org/10.1007/bf01126395.

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Dyk, Štěpán, and Vladimír Zeman. "Bifurcations in Mathematical Model of Nonlinear Vibration of the Nuclear Fuel Rod." Applied Mechanics and Materials 821 (January 2016): 207–12. http://dx.doi.org/10.4028/www.scientific.net/amm.821.207.

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The paper deals with nonlinear phenomena that occurs during vibration of nuclear fuel rod (FR). The FR is considered as a system consisting of two impact-interacting subsystems FR cladding (zircalloy tube) and fuel pellets stack placed inside FR cladding. Between both subsystems, there is a small radial clearance. The FR is bottom-end-fixed, and at eight equidistant levels, the FR cladding is supported by spacer grids (SG). Both subsystems are modelled by means of finite element method for one-dimensional Euler-Bernoulli continua. During fuel assembly (FA) motion caused by pressure pulsations
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Dissertations / Theses on the topic "Nuclear fuel cladding tube"

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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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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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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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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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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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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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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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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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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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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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Books on the topic "Nuclear fuel cladding tube"

1

Causey, A. R. Irradiation-enhanced creep of cold-worked Zr-2.5Nb tubes and helical-springs. Reactor Materials Research Branch, Chalk River Laboratories, 1993.

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Yegorova, L. A. Data base on the behavior of high burnup fuel rods with Zr-1% Nb cladding and UO2 fuel (VVER Type) under reactivity accident conditions. U.S. Nuclear Regulatory Commission, 1999.

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Sardjono, Ignatius Djoko. Study of burnout phenomena on a cylindrical heater tube simulating nuclear fuel rod. National Library of Canada, 1990.

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Coordinated Research Programme Meeting on Investigation of Fuel Element Cladding Interaction with Water Coolant in Power Reactors (1986 Bhabha Atomic Research Centre). Proceedings of Coordinated Research Programme Meeting on Investigation of Fuel Element Cladding Interaction with Water Coolant in Power Reactors and Topical Meeting on Water Chemistry in Nuclear Energy Systems, Bhabha Atomic Research Centre, Bombay 400085, November 24-28, 1986. The Centre, 1986.

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R, Theaker J., and International Symposium on Zirconium in the Nuclear Industry (10th : 1994), eds. Fabrication of Zr-2.5Nb pressure tubes to minimise the harmful effects of trace elements. Fuel Channel Components Branch, Chalk River Laboratories, 1994.

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1935-, Causey A. R., and Nuclear Industry Conference (1993 : Baltimore, Md.), eds. On the anisotropy of in-reactor creep of Zr-2.5Nb tubes. Reactor Materials Research Branch, Chalk River Laboratories, 1993.

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Almarshad, Abdullah I. A. A model for waterside oxidation of zircaloy fuel cladding in pressurized water reactors. 1990.

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Ren, Yongli. Fracture toughness behavior of Zircaloy-4 in the form of fuel cladding tubing in nuclear reactors. 2004.

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P, Moeller M., U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation. Division of Safety Review and Oversight., Pacific Northwest Laboratory, and U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Reactor Accident Analysis., eds. The Impact of fuel cladding failure events on occupational radiation exposures at nuclear power plants: Case study. Division of Safety Review and Oversight, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 1986.

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The Impact of fuel cladding failure events on occupational radiation exposures at nuclear power plants: Case study. Division of Safety Review and Oversight, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 1986.

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Book chapters on the topic "Nuclear fuel cladding tube"

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Park, Young Chang, Sung Hoon Jeong, Yong Hwan Kim, Seung Jae Lee, and Young Ze Lee. "Influence of Temperature on the Fretting Wear of Advanced Nuclear Fuel Cladding Tube against Supporting Grid." In The Mechanical Behavior of Materials X. Trans Tech Publications Ltd., 2007. http://dx.doi.org/10.4028/0-87849-440-5.705.

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Kim, Jin Seon, Yong Hwan Kim, Seung Jae Lee, and Young Ze Lee. "Fretting Wear Damage of the Corroded Fuel Cladding Tubes for Nuclear Fuel Rod against Supporting Grids." In The Mechanical Behavior of Materials X. Trans Tech Publications Ltd., 2007. http://dx.doi.org/10.4028/0-87849-440-5.709.

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Sakamoto, Kan, Masafumi Nakatsuka, and Toru Higuchi. "Simulation of Outside-in Cracking in Boiling Water Reactor Fuel Cladding Tubes under Power Ramp." In Zirconium in the Nuclear Industry: 16th International Symposium. ASTM International, 2010. http://dx.doi.org/10.1520/stp49293t.

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Sakamoto, Kan, Masafumi Nakatsuka, and Toru Higuchi. "Simulation of Outside-in Cracking in Boiling Water Reactor Fuel Cladding Tubes under Power Ramp." In Zirconium in the Nuclear Industry: 16th International Symposium. ASTM International, 2010. http://dx.doi.org/10.1520/stp49391s.

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Sakamoto, Kan, Masafumi Nakatsuka, and Toru Higuchi. "Simulation of Outside-in Cracking in Boiling Water Reactor Fuel Cladding Tubes under Power Ramp." In Zirconium in the Nuclear Industry: 16th International Symposium. ASTM International, 2010. http://dx.doi.org/10.1520/stp152920120041.

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Yagnik, Suresh K., Jen-Hung Chen, and Roang-Ching Kuo. "Effect of Hydride Distribution on the Mechanical Properties of Zirconium-Alloy Fuel Cladding and Guide Tubes." In Zirconium in the Nuclear Industry: 17th Volume. ASTM International, 2014. http://dx.doi.org/10.1520/stp154320120192.

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Park, Young Chang, Yong Hwan Kim, Seung Jae Lee, and Young Ze Lee. "The Predictions of the Fretting Wear between Supporting Grids and Cladding Tubes of Nuclear Fuel Rod." In Experimental Mechanics in Nano and Biotechnology. Trans Tech Publications Ltd., 2006. http://dx.doi.org/10.4028/0-87849-415-4.1243.

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Kubo, Toshio, Hiroaki Muta, Shinsuke Yamanaka, Masayoshi Uno, and Keizo Ogata. "In Situ Scanning Electron Microscope Observation and Finite Element Method Analysis of Delayed Hydride Cracking Propagation in Zircaloy-2 Fuel Cladding Tubes." In Zirconium in the Nuclear Industry: 16th International Symposium. ASTM International, 2011. http://dx.doi.org/10.1520/stp49270t.

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Kubo, Toshio, Hiroaki Muta, Shinsuke Yamanaka, Masayoshi Uno, and Keizo Ogata. "In Situ Scanning Electron Microscope Observation and Finite Element Method Analysis of Delayed Hydride Cracking Propagation in Zircaloy-2 Fuel Cladding Tubes." In Zirconium in the Nuclear Industry: 16th International Symposium. ASTM International, 2011. http://dx.doi.org/10.1520/stp49368s.

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Kubo, Toshio, Hiroaki Muta, Shinsuke Yamanaka, Masayoshi Uno, and Keizo Ogata. "In Situ Scanning Electron Microscope Observation and Finite Element Method Analysis of Delayed Hydride Cracking Propagation in Zircaloy-2 Fuel Cladding Tubes." In Zirconium in the Nuclear Industry: 16th International Symposium. ASTM International, 2011. http://dx.doi.org/10.1520/stp152920120018.

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Conference papers on the topic "Nuclear fuel cladding tube"

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Dabney, Tyler, Hwasung Yeom, Kyle Quillin, Nick Pocquette, and Kumar Sridharan. "Cold Spray Technology for Oxidation-Resistant Nuclear Fuel Cladding." In ITSC2021, edited by F. Azarmi, X. Chen, J. Cizek, et al. ASM International, 2021. http://dx.doi.org/10.31399/asm.cp.itsc2021p0167.

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Abstract Light water reactors (LWR) use zirconium-alloy fuel claddings; the tubes that hold the uranium-dioxide fuel pellets. Zr-alloys have very good neutron transparency; but during a loss of coolant accident or beyond design basis accident (BDBA) they can undergo excessive oxidation in reaction with the surrounding steam environment. Relatively thin oxidationresistant coatings on Zr-alloy fuel cladding tubes can potentially buy coping time in these off-normal scenarios. In this study; cold spraying; solid-state powder-based materials deposition technology has been developed for deposition o
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Qin, Zhou, Li Jiwei, Dang Yu, and Ding Yang. "Research on Nickel-Plated Hydrogen-Absorption Device in Fuel Rod and Performance Testing." In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-67112.

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The hydrogen may be introduced into the fuel rod during the process of production and manufacture. During the operation in reactor, the irradiated fuel pellets also produce radioactive isotopes of hydrogen and tritium. Under the operating condition in pile, the hydrogen in fuel rod will enter the zirconium alloy cladding tube forming hydride, lead to the hydrogen brittleness of cladding tube, and severe cases can lead to the cladding tube broken. The radioactive tritium inside fuel rod has high activity, and it possibly goes through the cladding tube by diffusion penetration into the reactor c
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Abe, Hiroshi, Seung Mo Hong, and Yutaka Watanabe. "High-Temperature Steam Oxidation Kinetics and Mechanism of SCWR Fuel Cladding Candidate Materials." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-16432.

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Effects of temperature, dissolved oxygen (DO), and degree of cold work (CW) on the oxidation kinetics of supercritical-water-cooled reactor (SCWR) fuel cladding candidate materials, including three types of 15Cr-20Ni austenitic stainless steels (1520 SSs), in superheated steam have been investigated assuming power-law kinetics. Characteristics of oxide layers and its relation to oxidation behaviors are also discussed. The effect of DO on the weight gain behavior in superheated steam at 700 °C was minor for all specimens at least up to 200 ppb DO. The tube-shaped specimens of 1520 SSs showed ve
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Park, Young-Chang, Yong-Hwan Kim, Seung-Jae Lee, and Young-Ze Lee. "Fretting Wear Between Supporting Grids and Cladding Tubes of Nuclear Fuel Rod." In ASME 2006 International Mechanical Engineering Congress and Exposition. ASMEDC, 2006. http://dx.doi.org/10.1115/imece2006-13625.

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The experimental investigation was performed to find the associated changes in characteristics of fretting wear with various water temperatures. Fretting can be defined as the oscillatory motion with very small amplitudes, which usually occur between two contacting surfaces. The fretting wear, which occurs between cladding tubes of nuclear fuel rod and grids, causes in damages the cladding tubes by flow induced vibration in a nuclear reactor. In this paper, the fretting wear tests were carried out using the zirconium alloy tubes and the grids with increasing the water temperature. The tube mat
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Chen, Liang, Lili Liu, Xiaoming Song, and Hua Pang. "A Theoretical Model of the Stress Intensity Factor Threshold of DHC for Fuel Cladding Tube." In 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-81665.

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Delayed hydride cracking (DHC) is the result of a mechanism of crack initiation and slow propagation. In DHC, hydrogen diffusion in the metal is required. Gradients of concentration, temperature, and stress are all important factors in controlling diffusivity. The classic theory of DHC still has potential to be modified. In this study, a calculation formula for DHC SIF threshold is established with consideration of the temperature history, the temperature field and the stress field induced by the temperature gradient and the external mechanical loading. Moreover, the temperature gradient on cr
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Pengbo, Ji, Lu Hanghang, Wang Zhaosong, Zhang Zhoufeng, and Liu Bin. "Microstructure and Properties Analysis of Inner Cladding Tube and End Plug Girth Weld of Annular Fuel Element." In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-67243.

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Annular fuel element is a kind of new double-sided fuel element, and the welding between the inner cladding tube and end plug of it belongs to a new kind of welding. Actually the conception was first proposed by Massachusetts Institute of Technology and the China Institute of Atomic Energy of Reactor Engineering makes the physical design of annular fuel element product. Through this research, we now basically master the method of welding annular fuel element. In this research, we designed a girth-welding fixture for welding the inner cladding tube and end plug of annular fuel element. The infl
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Wang, Hong, Jy-an John Wang, and Hao Jiang. "Fatigue Behavior of Spent Nuclear Fuel Rods in Simulated Transportation Environment." In ASME 2017 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/pvp2017-65842.

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Nuclear fuel rod is composed of cladding tube and a specified number of fuel pellets contained. In the United States, spent nuclear fuel (SNF) is expected to be transported to at least one storage facility before permanent disposal. The fatigue behavior of spent nuclear fuel (SNF) rods under reversed cyclic bending must be understood in order to evaluate their vibration integrity in a transportation environment. This is especially important for high-burnup SNFs (&gt;45 GWd/MTU). This report presents the experimental results related to Zircaloy (Zry)-4-based surrogate rods and high-burnup SNFs,
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Soloviev, Sergei L., Boris A. Gabaraev, Leonid M. Parafilo, et al. "Integrated Analysis of Mechanical and Thermal Hydraulic Behavior of Graphite Stack in Channel-Type Reactors in Case of a Fuel Channel Rupture Accident." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22338.

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The paper discusses the methodology and a computational exercise analyzing the processes taking place in the graphite stack of an RBMK reactor in case of a pressure tube rupture caused by overheating. The methodology of the computational analysis is implemented in integrated code U_STACK which models thermal-hydraulic and mechanical processes in the stack with a varying geometry, coupled with the processes going on in the circulation loop and accident localization (confinement) system. Coolant parameters, cladding and pressure tube temperatures, pressure tube ballooning and rupture, coolant ou
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Li, Feng, Takeshi Mihara, Yutaka Udagawa, and Masaki Amaya. "Biaxial-EDC Test Attempts With Pre-Cracked Zircaloy-4 Cladding Tubes." In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-67602.

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When the pellet-cladding mechanical interaction (PCMI) occurs in a reactivity-initiated accident (RIA), the states of stress and strain in the fuel cladding varies in a range depending on the friction and degree of bonding between cladding and pellet. Japan Atomic Energy Agency has developed the improved Expansion-due-to-compression (EDC) test apparatus to investigate the PCMI failure criterion of high-burnup fuel under such conditions. In this study, the failure behavior of cladding tube was investigated by using the improved EDC test apparatus. Cold-worked, stress-relieved and recrystallized
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Aly, Ahmed, Victor Petrov, Maria Avramova, Annalisa Manera, and Kostadin Ivanov. "Evaluation of the Mixing Vanes Effect on the Hydrogen Diffusion and Hydride Formation in the Fuel Cladding." In 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-82431.

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The fuel cladding is an important barrier to the release of fission products to the environment. Its integrity must be conserved during the in-reactor lifetime and during the spent fuel pool and dry cask storage. The corrosive interaction between the cladding and the water coolant in light water reactors leads to the oxidation of the zirconium-based cladding. A fraction of the hydrogen released due to those corrosive interactions or the radiolysis of the water coolant is picked-up by the fuel cladding. It diffuses inside the cladding driven by the concentration and temperature gradients. Event
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Reports on the topic "Nuclear fuel cladding tube"

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Macdonald, Digby, Mirna Urquidi-Macdonald, Yingzi Chen, Jiahe Ai, Pilyeon Park, and Han-Sang Kim. Oxidation of Zircaloy Fuel Cladding in Water-Cooled Nuclear Reactors. Office of Scientific and Technical Information (OSTI), 2006. http://dx.doi.org/10.2172/896213.

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Jaramillo, Roger A., WILLIAM R. Hendrich, and Nicolas H. Packan. Tensile Hoop Behavior of Irradiated Zircaloy-4 Nuclear Fuel Cladding. Office of Scientific and Technical Information (OSTI), 2007. http://dx.doi.org/10.2172/931509.

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Kristine Barrett and Shannon Bragg-Sitton. Advanced LWR Nuclear Fuel Cladding System Development Trade-Off Study. Office of Scientific and Technical Information (OSTI), 2012. http://dx.doi.org/10.2172/1057698.

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Lister, Tedd E., and Michael V. Glazoff. Transition of Spent Nuclear Fuel to Dry Storage: Modeling activities concerning aluminum spent nuclear fuel cladding integrity. Office of Scientific and Technical Information (OSTI), 2018. http://dx.doi.org/10.2172/1492831.

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Yang, Yong, and Simon Phillpot. Innovative coating of nanostructured vanadium carbide on the F/M cladding tube inner surface for mitigating the fuel cladding chemical interactions. Office of Scientific and Technical Information (OSTI), 2017. http://dx.doi.org/10.2172/1413204.

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Heuser, Brent, James Stubbins, Tomasz Kozlowski, et al. Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel. Office of Scientific and Technical Information (OSTI), 2017. http://dx.doi.org/10.2172/1391853.

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Lu, Hongbing, Satish Bukkapatnam, Sandip Harimkar, Raman Singh, and Scott Bardenhagen. Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels. Office of Scientific and Technical Information (OSTI), 2014. http://dx.doi.org/10.2172/1116513.

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Simon, Pierre Clement, Michael Tonks, Arthur Motta, and Long Qing Chen. Development of a fully validated quantitative model of hydride morphology in zirconium alloy nuclear fuel cladding. Office of Scientific and Technical Information (OSTI), 2017. http://dx.doi.org/10.2172/1473586.

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Bahney, Robert. Effective thermal conductivity method for predicting spent nuclear fuel cladding temperatures in a dry fill gas. Office of Scientific and Technical Information (OSTI), 1997. http://dx.doi.org/10.2172/757327.

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Greiner, Miles. Development and Experimental Benchmark of Simulations to Predict Used Nuclear Fuel Cladding Temperatures during Drying and Transfer Operations. Office of Scientific and Technical Information (OSTI), 2017. http://dx.doi.org/10.2172/1358184.

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