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Journal articles on the topic 'Pellet-Cladding mechanical Interaction'

Author: Grafiati

Published: 4 June 2021

Last updated: 7 February 2022

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1

Qi, Feipeng, Zhenhai Liu, Quan Li, et al. "Pellet-cladding mechanical interaction analysis of Cr-coated Zircaloy cladding." Nuclear Engineering and Design 367 (October 2020): 110792. http://dx.doi.org/10.1016/j.nucengdes.2020.110792.

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2

Seo, Sang Kyu, Sung Uk Lee, Eun Ho Lee, Dong Yol Yang, Hyo Chan Kim, and Yong Sik Yang. "3D Finite Element Simulation of Pellet-Cladding Mechanical Interaction." Transactions of the Korean Society of Mechanical Engineers A 40, no. 5 (2016): 437–47. http://dx.doi.org/10.3795/ksme-a.2016.40.5.437.

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3

Michel, B., J. Sercombe, G. Thouvenin, and R. Chatelet. "3D fuel cracking modelling in pellet cladding mechanical interaction." Engineering Fracture Mechanics 75, no. 11 (2008): 3581–98. http://dx.doi.org/10.1016/j.engfracmech.2006.12.014.

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4

Chao, Chlng-Kong, and Che-Chung Tseng. "A Power-Rate-Dependent Model for Pellet/Cladding Mechanical Interaction." Nuclear Technology 101, no. 2 (1993): 202–11. http://dx.doi.org/10.13182/nt93-a34781.

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5

Hong, Kisik, J. R. Barber, M. D. Thouless, and Wei Lu. "Effect of power history on pellet-cladding interaction." Nuclear Engineering and Design 358 (March 2020): 110439. http://dx.doi.org/10.1016/j.nucengdes.2019.110439.

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6

YANAGISAWA, Kazuaki, Shoji KATANISHI, and Toshio FUJISHIRO. "Pellet-Cladding Mechanical Interaction of PWR Fuel Rod Under Rapid Power Transient." Journal of Nuclear Science and Technology 31, no. 7 (1994): 671–76. http://dx.doi.org/10.1080/18811248.1994.9735208.

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7

Michel, B., J. Sercombe, and G. Thouvenin. "A new phenomenological criterion for pellet–cladding interaction rupture." Nuclear Engineering and Design 238, no. 7 (2008): 1612–28. http://dx.doi.org/10.1016/j.nucengdes.2008.01.012.

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8

Kim, Hyo Chan, Sang Kyu Seo, Sung Uk Lee, and Yong Sik Yang. "Development of NUFORM3D module with FRAPCON3.4 for simulation of pellet-cladding mechanical interaction." Nuclear Engineering and Design 318 (July 2017): 61–71. http://dx.doi.org/10.1016/j.nucengdes.2017.03.035.

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9

Saad, Djillali, Hocine Benkharfia, Mourad Kadouma, and Tahar Zidi. "Pellet-cladding mechanical interaction analysis of heavy water fuel rods under power ramps." Annals of Nuclear Energy 159 (September 2021): 108320. http://dx.doi.org/10.1016/j.anucene.2021.108320.

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10

Denis, Alicia, and Alejandro Soba. "Simulation of pellet-cladding thermomechanical interaction and fission gas release." Nuclear Engineering and Design 223, no. 2 (2003): 211–29. http://dx.doi.org/10.1016/s0029-5493(02)00390-4.

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11

Massih, A. R., T. Rajala, and L. O. Jernkvist. "Analyses of pellet-cladding mechanical interaction behaviour of different ABB atom fuel rod designs." Nuclear Engineering and Design 156, no. 3 (1995): 383–91. http://dx.doi.org/10.1016/0029-5493(94)00964-z.

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12

Cherezov, Alexey, Jinsu Park, Hanjoo Kim, Jiwon Choe, and Deokjung Lee. "A Multi-Physics Adaptive Time Step Coupling Algorithm for Light-Water Reactor Core Transient and Accident Simulation." Energies 13, no. 23 (2020): 6374. http://dx.doi.org/10.3390/en13236374.

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A new reactor core multi-physics system addresses the pellet-to-cladding heat transfer modeling to improve full-core operational transient and accident simulation used for assessment of reactor core nuclear safety. The rigorous modeling of the heat transfer phenomena involves strong interaction between neutron kinetics, thermal-hydraulics and nuclear fuel performance, as well as consideration of the pellet-to-cladding mechanical contact leading to dramatic increase in the gap thermal conductance coefficient. In contrast to core depletion where parameters smoothly depend on fuel burn-up, the co

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13

Yanagisawa, Kazuaki, and Hiroaki Saito. "A study on bamboo ridge deformation induced by pellet-cladding interaction." Nuclear Engineering and Design 97, no. 3 (1986): 339–46. http://dx.doi.org/10.1016/0029-5493(86)90169-x.

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14

D’Ambrosi, V., C. Destouches, G. Ricciardi, et al. "Fuel rod nonlinear vibrations to detect and characterize Pellet-Cladding Interaction." Nuclear Engineering and Design 379 (August 2021): 111214. http://dx.doi.org/10.1016/j.nucengdes.2021.111214.

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15

Vallejo, Isabel, Juan Blázquez, and Félix Barrio. "The use of the noise analysis for the detection of the pellet- cladding mechanical interaction." Nuclear Engineering and Design 180, no. 2 (1998): 169–74. http://dx.doi.org/10.1016/s0029-5493(97)00285-9.

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16

Soba, Alejandro, and Alicia Denis. "Simulation with DIONISIO 1.0 of thermal and mechanical pellet-cladding interaction in nuclear fuel rods." Journal of Nuclear Materials 374, no. 1-2 (2008): 32–43. http://dx.doi.org/10.1016/j.jnucmat.2007.06.020.

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17

Hellouin de Menibus, Arthur, Jerome Sercombe, Quentin Auzoux, and Christophe Poussard. "Thermomechanical loading applied on the cladding tube during the pellet cladding mechanical interaction phase of a rapid reactivity initiated accident." Journal of Nuclear Materials 453, no. 1-3 (2014): 210–13. http://dx.doi.org/10.1016/j.jnucmat.2014.06.046.

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18

Magnusson, Per, Anna-Maria Alvarez-Holston, Katja Ammon, et al. "Effects of Zr-hydride distribution of irradiated Zircaloy-2 cladding in RIA-simulating pellet-clad mechanical interaction testing." Nuclear Engineering and Technology 50, no. 2 (2018): 246–52. http://dx.doi.org/10.1016/j.net.2017.12.013.

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19

Udagawa, Yutaka, Tomoyuki Sugiyama, and Masaki Amaya. "Thresholds for failure of high-burnup LWR fuels by Pellet Cladding mechanical interaction under reactivity-initiated accident conditions." Journal of Nuclear Science and Technology 56, no. 12 (2019): 1063–72. http://dx.doi.org/10.1080/00223131.2019.1637795.

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20

Belov, Alexander I., Randy W. L. Fong, Brian W. Leitch, Thambiayah Nitheanandan, and Anthony Williams. "CHARACTERIZING HIGH-TEMPERATURE DEFORMATION OF INTERNALLY HEATED NUCLEAR FUEL ELEMENT SIMULATORS." CNL Nuclear Review 5, no. 1 (2016): 67–84. http://dx.doi.org/10.12943/cnr.2016.00005.

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The sag behaviour of a simulated nuclear fuel element during high-temperature transients has been investigated in an experiment utilizing an internal indirect heating method. The major motivation of the experiment was to improve understanding of the dominant mechanisms underlying the element thermo-mechanical response under loss-of-coolant accident conditions and to obtain accurate experimental data to support development of 3-D computational fuel element models. The experiment was conducted using an electrically heated CANDU fuel element simulator. Three consecutive thermal cycles with peak t

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21

Retel, Violaine, Frédérique Trivaudey, M. Lamine Boubakar, Dominique Perreux, and Philippe Thevenin. "Comparative effects of structural and material parameters variability on Pellet–Cladding Interaction in a PWR fuel rod." Nuclear Engineering and Design 228, no. 1-3 (2004): 35–46. http://dx.doi.org/10.1016/j.nucengdes.2003.06.004.

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22

Cinbiz, M. Nedim, Takaaki Koyanagi, Gyanender Singh, Yutai Katoh, Kurt A. Terrani, and Nicholas R. Brown. "Failure behavior of SiC/SiC composite tubes under strain rates similar to the pellet-cladding mechanical interaction phase of reactivity-initiated accidents." Journal of Nuclear Materials 514 (February 2019): 66–73. http://dx.doi.org/10.1016/j.jnucmat.2018.11.023.

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23

Scha¨ffler, I., P. Geyer, P. Bouffioux, and P. Delobelle. "Thermomechanical Behavior and Modeling Between 350°C and 400°C of Zircaloy-4 Cladding Tubes From an Unirradiated State to High Fluence (0 to 85s˙1024 nm−2,E>1 MeV)." Journal of Engineering Materials and Technology 122, no. 2 (1999): 168–76. http://dx.doi.org/10.1115/1.482783.

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This paper first describes the effect of neutron irradiation on the thermomechanical behavior of stress-relieved Zircaloy-4 fuel tubes that have been analyzed after exposure to five different fluences ranging from nonirradiated material to high burnup. In the second part, a viscoplastic model is proposed to simulate, for different isotherms, 350°C<T<400°C, out-of-flux anisotropic mechanical behavior of the cladding tubes over the fluence range 0<ϕ<100s˙1024 nm−2E>1 MeV. The model, identified for tests conducted at 350°C, has been validated from tests made at 380°C and 400°C. The

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24

D’Ambrosi, Veronica, Stephane Breaud, Christophe Destouches, et al. "Experimental characterization of PCI impact on vibrating fuel rod under axial turbulent flow representative of JHR irradiation device ADELINE: Set-up conception and measurement method." EPJ Web of Conferences 225 (2020): 04007. http://dx.doi.org/10.1051/epjconf/202022504007.

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Mechanical loading originating due to Pellet- Cladding Interaction (PCI) in PWR nuclear fuel rod constitutes a first order importance phenomenon when attempting at increasing the flexibility o f n uclear r eactors o peration t o f ollow grid demand. The need to improve the understanding of such complex multi-physics phenomena motivates the realization of particular irradiation sequences carried out in highly instrumented devices located in Material Testing Reactors. Among this kind, ADELINE irradiation device is being developed and will be installed in the future Jules Horowitz Reactor. In the

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25

Sercombe, Jérôme, Renaud Masson та Thomas Helfer. "Stress concentration during pellet cladding interaction: Comparison of closed-form solutions with 2D(r,θ) finite element simulations". Nuclear Engineering and Design 260 (липень 2013): 175–87. http://dx.doi.org/10.1016/j.nucengdes.2013.03.019.

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26

Cordara, Theo, Hannah Smith, Ritesh Mohun, et al. "Hot Isostatic Pressing (HIP): A novel method to prepare Cr-doped UO2 nuclear fuel." MRS Advances 5, no. 1-2 (2020): 45–53. http://dx.doi.org/10.1557/adv.2020.62.

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ABSTRACTThe addition of Cr2O3 to modern UO2 fuel modifies the microstructure so that, through the generation of larger grains during fission, a higher proportion of fission gases can be accommodated. This reduces the pellet-cladding mechanical interaction of the fuel rods, allowing the fuels to be “burned” for longer than traditional UO2 fuel, thus maximising the energy obtained. We here describe the preparation of UO2 and Cr-doped UO2 using Hot Isostatic Pressing (HIP), as a potential method for fuel fabrication, and for development of analogue materials for spent nuclear fuel research. Chara

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27

Nawaz, Amjad, Yoshikawa Hidekazu, Ming Yang, and Anwar Hussain. "Thermal behavior analysis of PWR fuel during RIA at various fuel burnups using modified theatre code." Nuclear Technology and Radiation Protection 31, no. 4 (2016): 307–17. http://dx.doi.org/10.2298/ntrp1604307n.

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The fuel irradiation and burnup causes geometrical and dimensional changes in the fuel rod which affects its thermal resistance and ultimately affects the fuel rod behavior during steady-state and transient conditions. The consistent analysis of fuel rod thermal performance is essential for precise evaluation of reactor safety in operational transients and accidents. In this work, analysis of PWR fuel rod thermal performance is carried out under steady-state and transient conditions at different fuel burnups. The analysis is performed by using thermal hydraulic code, THEATRe. The code is modif

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28

Lee, Sanghoon, and Seyeon Kim. "Development of Equivalent Beam Model of High Burnup Spent Nuclear Fuel Rods under Lateral Impact Loading." Metals 10, no. 4 (2020): 470. http://dx.doi.org/10.3390/met10040470.

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Spent nuclear fuel (SNF) is nuclear fuel that has been irradiated and discharged from nuclear reactors. During the whole management stages of SNF before it is, in the end, disposed in a deep geological repository, the structural integrity of fuel rods and the assemblies should be maintained for safety and economic reasons. In licensing applications for the SNF storage and transportation, the integrity of SNF needs to be evaluated considering various loading conditions. However, this is a challenging task due to the complexity of the geometry and properties of SNF. In this paper, a simple and e

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29

Valentine, Timothy, Maria Avramova, Michael Fleming, et al. "OVERVIEW OF THE OECD-NEA EXPERT GROUP ON MULTI-PHYSICS EXPERIMENTAL DATA, BENCHMARKS AND VALIDATION." EPJ Web of Conferences 247 (2021): 06048. http://dx.doi.org/10.1051/epjconf/202124706048.

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The OECD Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC) established the Expert Group on Multi-physics Experimental Data, Benchmarks and Validation (EGMPEBV) in 2014 to bridge the gap between advanced, multi-physics simulation capabilities and the relatively low availability of dedicated, high-fidelity experimental data and benchmarks specifically for multi-physics modelling and simulation tools. The EGMPEBV was mandated to establish mechanisms for the certification of experimental data and benchmark models and to establish the processes and procedures for the validation of multi-p

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30

Mirsalimov, Vagif. "Crack nucleation in rod-type nuclear fuel pellet." Mathematics and Mechanics of Solids 24, no. 3 (2018): 668–85. http://dx.doi.org/10.1177/1081286517753977.

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A plane problem of fracture mechanics on crack nucleation in a rod-type nuclear fuel pellet is considered. Nuclear reactor fuel pellets in operation may be damaged in various ways; in particular, crack nucleation. We consider a problem for the case of a heat-releasing fuel pellet with cladding: as the heat release intensity increases, zones of heightened stress are formed in the nuclear fuel pellet. The heightened stress will promote the appearance of prefracture bands that are simulated as zones of weakened interparticle bonds of the material. Interaction of prefracture zone faces is simulate

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31

Zeng, Zitao, Yongyu Pan, Xi Chen, et al. "Three-Dimensional Modeling of Thermal-Mechanical Behavior of Accident Tolerant Fuels." Frontiers in Energy Research 9 (March 17, 2021). http://dx.doi.org/10.3389/fenrg.2021.636502.

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Considering the safety issues of the traditional UO2-Zr fuel, a variety of accident-tolerant fuel (ATF) candidates have been proposed in recent years. Among the several ATFs, U3Si2, and UN are the two promising candidates for fuel materials owing to their high thermal conductivity and high uranium density. The FeCrAl alloy and the SiC/SiC composite material are the two promising candidates for cladding owing to their high oxidation resistance and high strength. In order to quantitatively evaluate the performance of ATFs, this study summarizes the physical models of typical ATF cladding materia

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32

Deng, Yangbin, Yingwei Wu, Dalin Zhang, Wenxi Tian, G. H. Su, and Suizheng Qiu. "Simulation on Pellet–Cladding Mechanical Interaction of Accident Tolerant Fuel With Coated Cladding." Journal of Nuclear Engineering and Radiation Science 5, no. 1 (2019). http://dx.doi.org/10.1115/1.4041194.

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In this study, based on the code Fuel ROd Behavior Analysis (FROBA), a thermal–mechanical analysis code initially developed for traditional UO2-Zr fuel elements by our research group, a modified version was developed to perform the fuel performance simulation of accident tolerant fuels (ATFs), named FROBA-ATF. Compared with initial version, the cladding could be divided into arbitrary number control volumes with different materials in the new code, so it can be used to perform the calculation for multilayer coatings. In addition, a new nonrigid pellet–cladding mechanical interaction (PCMI) cal

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33

Solonin, Vladimir, Anatoly Sotnikov, and Ivan Sergienko. "Development of Mechanical Loading Device for Testing the Zirconium Cladding Under the Pellet-Cladding Interaction Conditions." Science and Education of the Bauman MSTU 14, no. 06 (2014). http://dx.doi.org/10.7463/0614.0715393.

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34

Dostál, M., J. Zymák, and M. Valach. "Physical and Numerical Difficulties in Computer Modelling of Pellet-Cladding Contact Problems for Burned-Up Fuel." Acta Polytechnica 45, no. 5 (2005). http://dx.doi.org/10.14311/756.

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The importance of fuel reliability is growing due to the deregulated electricity market and the demands on operability and availability to the electricity grid of nuclear units. Under these conditions of fuel exploitation, the problems of PCMI (Pellet-Cladding Mechanical Interaction) are very important from the point of view of fuel rod integrity and reliability. Severe loading is thermophysically and mechanically expressed as a greater probability of cladding failure especially during power maneuvering. We have to be able to make a realistic prediction of safety margins, which is very difficu

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35

Luley, Jakub, Branislav Vrban, Stefan Cerba, Filip Osuský, and Vladimir Necas. "Fuel Performance Modelling At High Burn-Up by FEMAXI-6 Code." Journal of Nuclear Engineering and Radiation Science, September 24, 2021. http://dx.doi.org/10.1115/1.4052519.

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Abstract The scope of current research in the field of fuel performance is primary aimed to an improvement of the operating reliability, safety and cost effectiveness of the reactors in operation. The current requirement of nuclear industry is to have fuel suitable for load follow operation. Fission gas release, Pellet-Cladding Mechanical Interaction and stress corrosion cracking are the main phenomena that limit the variability of reactor operation from a safety perspective. To reasonable predict the fuel performance limits it is necessary to benchmark the computational tools against high qua

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36

Mutiara, Etty, Winter Dewayatna, and Tri Yulianto. "PEMODELAN DAN SIMULASI KINERJA PIN UJI BAHAN BAKAR PWR DENGAN UO2 DIPERKAYA." Urania Jurnal Ilmiah Daur Bahan Bakar Nuklir 25, no. 3 (2019). http://dx.doi.org/10.17146/urania.2019.25.3.5692.

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PEMODELAN DAN SIMULASI KINERJA PIN UJI BAHAN BAKAR PWR DENGAN UO2 DIPERKAYA. Pemodelan dan simulasi pin uji bahan bakar PWR telah dilakukan menggunakan kode komputer termomekanik FEMAXI-6. Simulasi dilakukan untuk memprediksi kinerja pin uji dengan variasi pengayaan pelet UO2. Dari simulasi ini diharapkan akan diperoleh data mengenai pengaruh tingkat pengayaan bahan bakar, fluks netron yang diterima pin uji, durasi iradiasi dan tingkat pencapaian burn-up pada awal terjadinya kontak antara pelet UO2 dengan kelongsong zircaloy (pellet cladding mechanical interaction, PCMI) pada pin uji. Data ter

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