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Journal articles on the topic 'Nuclear fuel cladding tube'

Author: Grafiati

Published: 4 June 2021

Last updated: 12 February 2022

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1

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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2

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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3

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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4

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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5

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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6

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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7

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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8

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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9

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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10

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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11

Kim, Do Sik, Sang Bok Ahn, Wan Ho Oh, et al. "Tensile Test Techniques for a Nuclear Fuel Cladding in a Hot Cell." Key Engineering Materials 345-346 (August 2007): 1561–64. http://dx.doi.org/10.4028/www.scientific.net/kem.345-346.1561.

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Modified transverse and longitudinal tensile test techniques are proposed to evaluate the mechanical properties of nuclear fuel cladding materials under a hoop and axial loading condition in a hot cell. The ring specimen with a gage length of 3 mm and a width of 2 mm is designed to limit a deformation within the gage section and to maximize the uniformity of a strain distribution at the gage section. The loading grip is designed such that a constant curvature of a specimen is maintained during a deformation. The contact surface is lubricated with a graphite lubricant (Model P-37, Molykote Co.)

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12

MINAMOTO, Hirofumi, Naoki OKADA, Masafumi NAKATSUKA, and Shozo KAWAMURA. "208 Simulation of Dynamic Buckling of a Nuclear Fuel Cladding Tube." Proceedings of Conference of Tokai Branch 2013.62 (2013): 87–88. http://dx.doi.org/10.1299/jsmetokai.2013.62.87.

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13

Shu, Linghong, and Yunqiao Dong. "Finite element analysis of temperature field of nuclear fuel cladding tube." Journal of Physics: Conference Series 1985, no. 1 (2021): 012061. http://dx.doi.org/10.1088/1742-6596/1985/1/012061.

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14

Cech, Miroslav, and Martin Sevecek. "MODELLING OF NUCLEAR FUEL CLADDING TUBES CORROSION." Acta Polytechnica CTU Proceedings 4 (December 16, 2016): 13. http://dx.doi.org/10.14311/ap.2016.4.0013.

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This paper describes materials made of zirconium-based alloys used for nuclear fuel cladding fabrication. It is focused on corrosion problems their theoretical description and modeling in nuclear engineering.

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15

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." Key Engineering Materials 345-346 (August 2007): 705–8. http://dx.doi.org/10.4028/www.scientific.net/kem.345-346.705.

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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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16

Herm, Michel, Ron Dagan, Ernesto González-Robles, Nikolaus Müller, and Volker Metz. "Comparison of calculated and measured radionuclide inventory of a Zircaloy-4 cladding tube plenum section." MRS Advances 3, no. 19 (2018): 1031–37. http://dx.doi.org/10.1557/adv.2018.274.

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ABSTRACTCladding tubes of water-cooled nuclear reactors are usually made of Zircaloy and are an important retaining element for radionuclides present in the fuel both during predisposal activities such as reloading of fuel assemblies from interim storage casks to final disposal casks and during final disposal in the case of canister breaching. However, cladding integrity is affected by various processes during reactor operation and beyond, e.g. fuel cladding chemical interaction and fission product precipitation onto the inner cladding surface. Using experimental and modelling methods, the rad

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17

Sagiroun, Mamoun I. A., Xin Rong Cao, Wasim M. K. Helal, and John N. Njoroge. "A Review of Development of Zirconium Alloys as a Fuel Cladding Material and its Oxidation Behavior at High-Temperature Steam." International Journal of Engineering Research in Africa 46 (January 2020): 7–14. http://dx.doi.org/10.4028/www.scientific.net/jera.46.7.

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Currently, Zr-alloys are widely used in nuclear power reactors for fuel cladding and structural components. Many types of zr-based alloys were developed to overcome the challenges encountered in the progress of nuclear reactors (high-burnup and high-duty). Oxygen diffused into the cladding, hydrogen absorbed in the cladding (breakaway oxidation and ruptured balloons) and rapid oxidation rate are results of chemical interaction of cladding material with steam at high temperature. Zirconium alloys seem to be the most suitable for use in fuel cladding, if they can overcome the rapid oxidation at

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18

Lee, Hyeon-Geun, Daejong Kim, Ji Yeon Park, and Weon-Ju Kim. "FEA Study on Hoop Stress of Multilayered SiC Composite Tube for Nuclear Fuel Cladding." Journal of the Korean Ceramic Society 51, no. 5 (2014): 435–41. http://dx.doi.org/10.4191/kcers.2014.51.5.435.

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19

Mayuzumi, Masami, and Takeo Onchi. "Creep deformation of an unirradiated zircaloy nuclear fuel cladding tube under dry storage conditions." Journal of Nuclear Materials 171, no. 2-3 (1990): 381–88. http://dx.doi.org/10.1016/0022-3115(90)90384-y.

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20

Gávelová, Petra, Patricie Halodová, Ondřej Libera, Iveta Adéla Prokůpková, Věra Vrtílková, and Jakub Krejčí. "Experimental Verification of Phase Diagram Calculations of Zr-Based Alloys after High-Temperature Oxidation." Defect and Diffusion Forum 405 (November 2020): 351–56. http://dx.doi.org/10.4028/www.scientific.net/ddf.405.351.

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Zirconium-based alloys are commonly used as a material for nuclear fuel claddings in the light water reactors. The cladding material must function to fix a huge number of pellets, while conducting heat into the coolant that flows turbulently around the fuel rods. Cladding tubes can contain gaseous fission products that escape the fuel. Thus, by functioning as a sealed unit, it prevents a contamination of the coolant water with high-radioactive fission products. The integrity of claddings is always a critical issue during reactor operation and wet or dry storage and transport of the spent fuel

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21

Vingsbo, Olof, Ali R. Massih, and Stig Nilsson. "Evaluation of Fretting Damage of Zircaloy Cladding Tubes." Journal of Tribology 118, no. 4 (1996): 705–10. http://dx.doi.org/10.1115/1.2831598.

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The work is part of the search for a technique to study the fretting properties of the cladding of fuel rods in nuclear reactors. The strategy is to compare the fretting scars of specimens tested in a near-full-scale, nonradioactive simulator (so called fuel assembly endurance test) with scars on specimens tested under well controlled conditions in the laboratory. By systematic variations of the laboratory testing parameters it has proved possible to achieve scars with the same type of damage characteristics as those from the fuel assembly endurance test. This makes it possible to estimate, by

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22

Galvin, T., N. C. Hyatt, W. M. Rainforth, I. M. Reaney, and D. Shepherd. "Slipcasting of MAX phase tubes for nuclear fuel cladding applications." Nuclear Materials and Energy 22 (January 2020): 100725. http://dx.doi.org/10.1016/j.nme.2020.100725.

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23

Kim, Daejong, Hyun-Geun Lee, Ji Yeon Park, and Weon-Ju Kim. "Fabrication and measurement of hoop strength of SiC triplex tube for nuclear fuel cladding applications." Journal of Nuclear Materials 458 (March 2015): 29–36. http://dx.doi.org/10.1016/j.jnucmat.2014.11.117.

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24

Kim, Young-Hwan, Yung-Zun Cho, Young-Soon Lee, and Jin-Mok Hur. "Engineering Design of a Mechanical Decladder for Spent Nuclear Rod-Cuts." Science and Technology of Nuclear Installations 2019 (August 14, 2019): 1–16. http://dx.doi.org/10.1155/2019/9273503.

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A practical scale mechanical decladder that can slit spent nuclear fuel rod-cuts (hulls + pellets) of several tens of kg HM/batch is being developed to supply UO2 pellets to a voloxidation process. The mechanical decladder is an apparatus for separating and recovering fuel material and cladding tubes by horizontally slitting the cladding tube of a fuel rod and a defective irradiated fuel rod. In this study, we address the engineering design of the mechanical decladder for the pretesting of rod-cut slitting. To obtain the requirements of the mechanical decladder, we first manufactured a slitter

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25

INAGAKI, Masahisa, Kimihiko AKAHORI, Jirou KUNIYA, et al. "Effect of chemical composition on corrosion resistance of zircaloy fuel cladding tube for BWR." Journal of the Atomic Energy Society of Japan / Atomic Energy Society of Japan 32, no. 7 (1990): 728–40. http://dx.doi.org/10.3327/jaesj.32.728.

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26

Nagy, Richárd, Márton Király, Péter Petrik, and Zoltán Hózer. "Infrared observation of ballooning and burst of nuclear fuel cladding tubes." Nuclear Engineering and Design 371 (January 2021): 110942. http://dx.doi.org/10.1016/j.nucengdes.2020.110942.

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27

Yanwei, Zhang, Wang Rongshan, Bai Guanghai, Liu Erwei, and Mei Jinna. "Burst Test Research on Zirconium Alloy for Nuclear Fuel Cladding Tubes." Rare Metal Materials and Engineering 46, no. 6 (2017): 1491–96. http://dx.doi.org/10.1016/s1875-5372(17)30152-2.

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28

Kim, Ickchan, Fauzia Khatkhatay, Liang Jiao, et al. "TiN-based coatings on fuel cladding tubes for advanced nuclear reactors." Journal of Nuclear Materials 429, no. 1-3 (2012): 143–48. http://dx.doi.org/10.1016/j.jnucmat.2012.05.001.

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29

Choi, M. S., H. C. Kim, and M. S. Yang. "Propagation characteristics of elastic circumferential waves in nuclear fuel cladding tubes." Ultrasonics 30, no. 4 (1992): 213–19. http://dx.doi.org/10.1016/0041-624x(92)90079-2.

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30

Moon, Jong Han, Young Jun Lee, Jin Hang Lee, Jong Won Hong, and Jong Hyeon Lee. "Evaluation of SMUT Properties according to Nb Content in the Pickling Process of Nuclear Fuel Cladding Tube." Korean Journal of Materials Research 29, no. 8 (2019): 483–90. http://dx.doi.org/10.3740/mrsk.2019.29.8.483.

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31

Ford, I. J. "Axial crack propagation in fuel pin cladding tubes." Nuclear Engineering and Design 136, no. 3 (1992): 243–54. http://dx.doi.org/10.1016/0029-5493(92)90026-r.

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32

Kratochvílová, I., R. Škoda, J. Škarohlíd, et al. "Nanosized polycrystalline diamond cladding for surface protection of zirconium nuclear fuel tubes." Journal of Materials Processing Technology 214, no. 11 (2014): 2600–2605. http://dx.doi.org/10.1016/j.jmatprotec.2014.05.009.

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33

Andrieu, C., S. Ravel, G. Ducros, and C. Lemaignan. "Release of fission tritium through Zircaloy-4 fuel cladding tubes." Journal of Nuclear Materials 347, no. 1-2 (2005): 12–19. http://dx.doi.org/10.1016/j.jnucmat.2005.06.008.

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34

Rohmer, Eric, Eric Martin, and Christophe Lorrette. "Mechanical properties of SiC/SiC braided tubes for fuel cladding." Journal of Nuclear Materials 453, no. 1-3 (2014): 16–21. http://dx.doi.org/10.1016/j.jnucmat.2014.06.035.

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35

Lee, Jung Won, Jong Hwan Kim, Ki Hwan Kim, Jeong Yong Park, and Sung Ho Kim. "Development of End Plug Welding Technique for SFR Fuel Rod Fabrication." Science and Technology of Nuclear Installations 2016 (2016): 1–9. http://dx.doi.org/10.1155/2016/9549805.

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In Korea, R&D on a sodium-cooled fast reactor (SFR) was begun in 1997, as one of the national long-term nuclear R&D programs. As one fuel option for a prototype SFR, a metallic fuel, U-Zr alloy fuel, was selected and is currently being developed. For the fabrication of SFR metallic fuel rods, the end plug welding is a crucial process. The sealing of the end plug to the cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions, and parameters w

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36

Meng, Xin Ming, Fei Xue, Wei Wei Yu, and Hong Mei Guo. "Creep Tests Research on a New Zirconium Alloy for Nuclear Fuel Cladding Tubes." Applied Mechanics and Materials 151 (January 2012): 47–51. http://dx.doi.org/10.4028/www.scientific.net/amm.151.47.

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Hoop creep tests and axial creep tests were carried out on a new zirconium alloy for nuclear fuel cladding tubes at 375 °C. The result indicated that the proportion of the second creep stage decreased as the stress increasing, meanwhile, the steady-state creep rate increased in both creep test modes. However, the ability of creep rupture resistance in the hoop direction is more strengthen than that in axial direction. The SEM analysis showed that the creep fracture was mainly dimple toughness fracture. The dimple for low stress was intensive and small. The fitting curve by the θ function was w

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37

Bérerd, N., H. Catalette, A. Chevarier, N. Chevarier, H. Faust, and N. Moncoffre. "Zirconium surface modification under fission product irradiation. Application to nuclear fuel cladding tubes." Surface and Coatings Technology 158-159 (September 2002): 473–76. http://dx.doi.org/10.1016/s0257-8972(02)00290-6.

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38

Jia, Qi, Li Xun Cai, and Chen Bao. "Strain Fatigue Behavior of Thin-Walled Tubes of Zr-1Nb and Zr-4 and Thin Plates of N18 at Elevated Temperatures." Applied Mechanics and Materials 69 (July 2011): 39–44. http://dx.doi.org/10.4028/www.scientific.net/amm.69.39.

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Thin plates and small thin-walled tubes of zircaloys are used as fuel cladding materials of nuclear reactors. In order to prevent buckling of plate or tube specimens of zircaloys under cyclic strain loading, a set of self-invented clamps were used for strain fatigue tests of small thin-walled tubes of Zr-4 and Zr-1Nb at elevated temperature. Depending on axial control strain, a new strain fatigue test method named as (equivalent strain)-SF(strain fatigue) was developed for thin funnel-like plate specimens of N18 zircaloy with double side notches. A series of monotonic tension tests for the thr

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39

Oh, Dong Seok, Young Ho Lee, Chang Hwan Shin, Tae Hyun Chun, Hyung Kyu Kim, and Kye Bock Lee. "Development of a Bi-Axial Acceleration-Detecting Device for a Tube." Key Engineering Materials 326-328 (December 2006): 1483–86. http://dx.doi.org/10.4028/www.scientific.net/kem.326-328.1483.

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A new device for measuring an acceleration in a fuel rod has been developed. The primary purpose is to apply it to the experiments for a nuclear fuel fretting, which is caused by a fuel rod and grid interaction due to a flow-induced vibration of the rods. A bi-axial accelerationdetecting device of a cylindrical shape for an insertion into a cladding tube is designed. Two unimorph piezoelectric accelerometers of small size and for special use in a high temperature condition were attached to the housing’s inner wall of the mounting device, which were oriented perpendicularly with each other to a

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40

Xie, Miaoxia, Xiangtao Shang, Yanxin Li, Zehui Zhang, Minghui Zhu, and Jiangtao Xiong. "Rotary Friction Welding of Molybdenum without Upset Forging." Materials 13, no. 8 (2020): 1957. http://dx.doi.org/10.3390/ma13081957.

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A large instantaneous axial forging load is required to be applied for the final stage of rotary friction welding (RFW), which is usually conducive to obtaining clean, compact, and high-quality joints. However, for slender fuel claddings made of molybdenum (Mo) with low stiffness, the instantaneous axial forging load cannot be applied at the final stage of welding. This study carried out RFW tests without upset forging on Mo in the atmospheric environment and investigated the effects of welding time on joint morphology, axial shortening, microstructures, microhardness, tensile strength, and te

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41

Busser, Vincent, Jean Desquines, Stéphanie Fouquet, Marie Christine Baietto, and Jean Paul Mardon. "Modelling of Corrosion Induced Stresses during Zircaloy-4 Oxidation in Air." Materials Science Forum 595-598 (September 2008): 419–27. http://dx.doi.org/10.4028/www.scientific.net/msf.595-598.419.

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In the frame of its research work on nuclear fuel safety, the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN) has highlighted the importance of cladding tube oxidation on its thermomechanical behavior. The occurrence of radial cracking and spallation has been observed as the main mechanisms for the zirconia layer degradation during transient experiments. A study of these two mechanisms has been jointly launched by IRSN and Areva-NP. Thus laboratory air oxidations of fully recrystallized or stress-relieved low-tin Zircaloy-4 cladding tubes have been performed. Representative

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42

Yakushkin, Aleksey A., and Philip I. Vysikaylo. "MODIFICATION OF THE SURFACE AND COATING APPLICATION ON FUEL CLADDING TUBES FOR nuclear reactors." Bulletin of the Moscow State Regional University (Physics and Mathematics), no. 4 (2018): 92–111. http://dx.doi.org/10.18384/2310-7251-2018-4-92-111.

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43

Duan, Zhengang, Huilong Yang, Yuhki Satoh, et al. "Current status of materials development of nuclear fuel cladding tubes for light water reactors." Nuclear Engineering and Design 316 (May 2017): 131–50. http://dx.doi.org/10.1016/j.nucengdes.2017.02.031.

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44

Shinohara, Yasunari, Hiroaki Abe, Toshiya Kido, Takeo Iwai, and Naoto Sekimura. "In Situ TEM Observation of Precipitation of Zirconium Hydrides in Zircaloy-4 under Hydrogen Ion Implantation." Materials Science Forum 561-565 (October 2007): 1765–68. http://dx.doi.org/10.4028/www.scientific.net/msf.561-565.1765.

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The formation of hydrides in zirconium alloy has been one of the essential matters of discussion to maintain mechanical strength of nuclear fuel cladding tubes. In this work, we examined the precipitation process of zirconium hydride by transmission electron microscopy under hydrogen ion irradiation. Zircaloy-4, which has been used extensively as nuclear fuel cladding, was irradiated with hydrogen ion at room temperature to achieve enough hydrogen concentration for precipitation. The growth of hydrides accompanied with dislocations around hydrides was observed under hydrogen implantation. The

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45

Sidelev, Dmitrii, Sergey Ruchkin, and Egor Kashkarov. "High-Temperature Oxidation of Cr-Coated Resistance Upset Welds Made from E110 Alloy." Coatings 11, no. 5 (2021): 577. http://dx.doi.org/10.3390/coatings11050577.

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The resistance upset welds (RUW) made from E110 alloy without and with Cr coatings were oxidized in air atmosphere at 1100 °C for 2, 10 and 30 min. The cross-section microstructure, elemental composition and hardness were studied before and after oxidation using optical and scanning electron microscopy, and indentations in welding region. The RUW welding does not noticeably change oxidation kinetics of E110 alloy. The most crucial effect has surface non-regularities formed after welding, which prevent uniform coating deposition on full surface of welded cladding tube and end plug. Cr coating d

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46

TSUCHIE, Yasuo, and Toshio KODAMA. "Post-irradiation examination of Tsuruga fuel using cladding tubes manufactured in Japan." Journal of the Atomic Energy Society of Japan / Atomic Energy Society of Japan 29, no. 3 (1987): 219–43. http://dx.doi.org/10.3327/jaesj.29.219.

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Király, Márton, Márta Horváth, Richárd Nagy, Nóra Vér, and Zoltán Hózer. "Segmented mandrel tests of as-received and hydrogenated WWER fuel cladding tubes." Nuclear Engineering and Technology 53, no. 9 (2021): 2990–3002. http://dx.doi.org/10.1016/j.net.2021.03.019.

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Mayuzumi, Masami, and Takeo Onchi. "Creep deformation and rupture properties of unirradiated Zircaloy-4 nuclear fuel cladding tube at temperatures of 727 to 857 K." Journal of Nuclear Materials 175, no. 1-2 (1990): 135–42. http://dx.doi.org/10.1016/0022-3115(90)90280-z.

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Yumura, Takanori, and Masaki Amaya. "Effects of ballooning and rupture on the fracture resistance of Zircaloy-4 fuel cladding tube after LOCA-simulated experiments." Annals of Nuclear Energy 120 (October 2018): 798–804. http://dx.doi.org/10.1016/j.anucene.2018.06.046.

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Senevat, J., and P. Mainy. "Eddy current examination technique during manufacturing of Zircaloy-4 fuel cladding tubes." Journal of Nuclear Materials 178, no. 2-3 (1991): 315–20. http://dx.doi.org/10.1016/0022-3115(91)90403-t.

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