Academic literature on the topic 'Nuclear fuel cladding tube'
Create a spot-on reference in APA, MLA, Chicago, Harvard, and other styles
Consult the lists of relevant articles, books, theses, conference reports, and other scholarly sources on the topic 'Nuclear fuel cladding tube.'
Next to every source in the list of references, there is an 'Add to bibliography' button. Press on it, and we will generate automatically the bibliographic reference to the chosen work in the citation style you need: APA, MLA, Harvard, Chicago, Vancouver, etc.
You can also download the full text of the academic publication as pdf and read online its abstract whenever available in the metadata.
Journal articles on the topic "Nuclear fuel cladding tube"
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.
Full textKim, 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.
Full textKim, 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.
Full textPark, 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.
Full textFUJITA, 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.
Full textMahendra Prabhu, N., K. A. Gopal, S. Murugan, T. K. Haneef, C. K. Mukhopadhyay, S. Venugopal, and T. Jayakumar. "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 (June 8, 2015): 410–18. http://dx.doi.org/10.1108/ijsi-08-2014-0038.
Full textLe 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.
Full textMurugan, 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.
Full textZelenskii, V. F., I. M. Neklyudov, B. P. Chernyi, M. P. Zeidlits, A. F. Vanzha, V. G. Rubashko, L. N. Ryabchikov, et al. "Centrifugal vacuum gasting for fuel cladding tube blanks." Soviet Atomic Energy 67, no. 1 (July 1989): 531–33. http://dx.doi.org/10.1007/bf01126395.
Full textDyk, 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.
Full textDissertations / Theses on the topic "Nuclear fuel cladding tube"
Paramonova, Ekaterina (Ekaterina D. ). "CRUD resistant fuel cladding materials." Thesis, Massachusetts Institute of Technology, 2013. http://hdl.handle.net/1721.1/82447.
Full text"June 2013." Cataloged from PDF version of thesis.
Includes bibliographical references (pages 27-29).
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 objective to study the effect of different surface modifications of Zircaloy cladding on the formation of CRUD, and ultimately minimize its effects. This modification will alter the surface chemistry and therefore the CRUD formation rate. The objective of this study was to construct a pool boiling facility at atmospheric pressure and sub-cooled boiling conditions, and test a series of samples in simulated PWR water with a high concentration of nanoparticulate CRUD precursors. After testing, ZrC was the only material out of six that did not develop dark, circular spots, which are hypothesized to be the beginnings of CRUD boiling chimneys. Further testing will be needed to confirm that it is indeed more CRUD resistant, even under realistic PWR conditions in a parallel testing facility.
by Ekaterina Paramonova.
S.B.
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.
Full textJarvis, 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.
Full textCataloged from PDF version of thesis.
Includes bibliographical references (pages 291-297).
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 hydrogen and released into the coolant? Water chemistry modeling was used to understand effects of radiolysis and CRUD. Density functional theory (DFT) was used to investigate the role of oxidized Zr(Fe,Cr)2 second phase particles. Chemical potentials and the electron chemical potential were used to connect these two modeling efforts. A radiolysis model was developed for the primary loop of a PWR. Dose profiles accounting for fuel burnup, boron addition, axial power profiles, and a CRUD layer were produced. Dose rates to the bulk coolant increased by 21-22% with 12.5-75 pim thick CRUD layers. Radially-averaged core chemistry was compared to single-channel chemistry at individual fuel rods. Calculations showed that local chemistry was more oxidizing at high-power fuel and fuel with CRUD. Local hydrogen peroxide concentrations were up to 2.5 ppb higher than average levels of 5-8 ppb. Radiolysis results were used to compute chemical potentials and the corrosion potential. Marcus theory was applied to compare the band energies of oxides associated with Zircaloy-4 and the energy levels for proton reduction in PWR conditions. Hydrogen interactions with Cr203 and Fe203, both found in oxidized precipitates, were studied with DFT. Atomic adsorption of hydrogen was modeled on the Cr and Feterminated (0001) surfaces. Climbing Image-Nudged Elastic Band calculations were used to model the competing pathways of hydrogen migration into the subsurface and molecular hydrogen formation. A two-step mechanism for hydrogen recombination was identified consisting of: reduction of an adsorbed proton (H+) to a hydride ion (H-) and H2 formation from an adjacent adsorbed proton and hydride ion. Overall, results suggest that neither surface will be an easy entrance point for hydrogen ingress and that Cr203 is more likely to be involved in hydrogen evolution than the Fe203.
by Jennifer Anne Jarvis.
Ph. D.
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.
Full textCataloged from PDF version of thesis.
Includes bibliographical references (pages 303-314).
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 that the oxidation of monolithic SiC is about three orders of magnitude slower than that of zirconium alloys, and with a weaker impact on mechanical strength. This, along with the presence of the environmental barrier coating around the load carrying intermediate layer of SiC fiber composite, diminishes the importance of oxidation for cladding failure mechanisms. Thermal shock experiments showed strength retention for both [alpha]-SiC and [beta]-SiC, as well as A1₂O₃ samples quenched from temperatures up to 1260°C in saturated water. The initial heat transfer upon the solid - fluid contact in the quenching transient is found to be a controlling factor in the potential for brittle fracture. This implies that SiC would not fail by thermal shock induced fracture during the reflood phase of a loss of coolant accident, which includes fuel-cladding quenching by emergency coolant at saturation conditions. A thermo-mechanical model for stress distribution and Weibull statistical fracture of laminated SiC cladding during normal and accident conditions is developed. It is coupled to fuel rod performance code FRAPCON-3.4 (modified here for SiC) and RELAP-5 (to determine coolant conditions). It is concluded that a PWR fuel rod with SiC cladding can extend the fuel residence time in the core, while keeping the internal pressure level within the safety assurance limit during steady-state and loss of coolant accidents. Peak burnup of 93 MWD/kgU (10% central void in fuel pellets) at 74 months of in-core residence time is found achievable with conventional PWR fuel rod design, but with an extended plenum length (70 cm). An easier to manufacture, 30% larger SiC cladding thickness requires an improved thermal conductivity of the composite layer to reduce thermal stress levels under steady-state operation to avoid failure at the same burnup. A larger Weibull modulus of the SiC cladding improves chances of avoiding brittle failure.
by Youho Lee.
Ph. D.
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.
Full textJena, 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.
Full textCataloged from the official PDF of thesis.
Includes bibliographical references (pages [69]-70).
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 deficient for the thermal-hydraulic properties of these materials, particularly under light water reactor (LWR) operating conditions. The surface wettability is particularly important, as it affects the dynamics of the boiling heat transfer process, and consequently, the critical heat flux (CHF) and rewetting temperatures, which are important thermal limits for LWRs. Surface wettability determines nucleation site density, bubble departure diameter, and bubble departure frequency.
Therefore, it is essential to quantify the surface wettability of candidate ATF cladding materials to determine their thermal-hydraulic behavior compared to conventional Zircaloy claddings. The surface wettability is usually quantified through the sessile droplet contact angle, which is the angle formed between the liquid-vapor and the liquid-solid interface. The contact angle depends on the fluid, solid, surface finish, and operating conditions, i.e., temperature and pressure. However, most of the measurements available in the literature are performed at low pressure and in an inert atmosphere, which is quite different from the operating conditions of LWRs (i.e., in a steam-saturated atmosphere at a pressure as high as 15.5 MPa or 155 bars).
To close this gap, in this study, we designed and built an autoclave-type facility capable of measuring static, advancing, and receding contact angle in steam-saturated atmospheres, from sub-atmospheric conditions up to the critical point of water, i.e., 22.1 MPa (221 bar or 3200 psi) and 374°C. We measured the static contact angle of conventional Zircaloy-4 and candidate ATF cladding materials (e.g., Cr-coated Zr-4, FeCrAl, and SiC). The contact angle decreases with an increase in temperature for all the materials. Rough surfaces showed higher wettability, i.e., lower contact angle, compared to the smooth surfaces. These trends are expected from theory. All the materials showed different wettability under the same temperature and pressure conditions. Individual correlations for temperature dependence for each of them are proposed.
by Anupam Jena.
S.M.
S.M. Massachusetts Institute of Technology, Department of Nuclear Science and Engineering
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.
Full textThesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2018
Cataloged from student-submitted PDF version of thesis. Page 123 blank.
Includes bibliographical references (pages 106-116).
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 to identify alternative fuel and cladding technologies that may provide enhanced safety, competitiveness, and economics. The new fuel design must also be compatible with present-day LWR design. For near-term applications, coatings on the nominal Zirconium-based cladding material and other metallic materials are being considered to improve the corrosion resistance and reduce the generation of hydrogen at high temperatures. Major ATF coating choices under consideration include chromium as a coating, iron-chromium-aluminum alloys (FeCrAl) as cladding and molybdenum as a coating, which have demonstrated better mechanical and oxidation behavior during the experimental testing.
Thermal-fluids characteristics are pivotal for a robust testing of ATF concepts as the proposed candidates may have an entirely different thermal-hydraulic behavior when compared to Zircaloy-4. ATF coatings may display very different boiling characteristics as a result of different microstructures and surface characteristics. In the present work, transient boiling heat transfer during quenching of the candidate ATF claddings on vertical rodlets is studied experimentally. The candidate ATF material (chromium, FeCrAl, and molybdenum) are applied on Zircaloy-4 rodlets. The vertical solid rodlets are heated to temperatures up to 1000 °C and are quenched in a saturated pool of water at atmospheric pressure. The temperature variation during the quenching of rodlets was recorded insitu with synchronized visualization of boiling regimes over the test specimen using a high-speed video camera.
The quench performance of the ATF coatings was analyzed based on the examination of various surface parameters such as wettability, roughness, emissivity and capillary wicking. In order to obtain a more realistic picture of the candidate performance during the emergency cooling reflood phase in a nuclear reactor, the coated rodlets are also oxidized in an autoclave before quenching. The performance of the candidate claddings is evaluated after oxidation and the surface characterized. It was observed from the post-test analysis that the surface characteristics and oxidation had a significant impact on the quench performance of ATF coatings, which varied between different coating materials. In order to better understand the thermal margins in a reactor specific environment, an analysis was performed on samples after exposing them to gamma rays. The gamma rays tend to change the surface wettability through a phenomenon called Radiation Induced Surface Activation.
A Gammacell 220E irradiator that uses 12 cobalt-60 pencil sources, arranged axially in a sample chamber at MIT, was used to irradiated the samples. The results of water quenching and contact angle studies showed a higher Leidenfrost temperature and wettability in both samples exposed to gamma irradiation. The detailed microscopic analysis attributed the enhanced wettability to oxidation of the surface under gamma irradiation.
by Arunkumar Seshadri.
S.M.
S.M. Massachusetts Institute of Technology, Department of Nuclear Science and Engineering
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.
Full textCataloged from PDF version of thesis.
Includes bibliographical references (p. 101-107).
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, a central composite, and an outer environmental barrier coating (EBC). The inner monolith consists of dense SiC which provides strength and hermeticity to contain fission products. The composite layer is made of SiC fibers, woven around the monolith, and then infiltrated with a SiC matrix. The composite layer adds strength to the monolith and provides a pseudo-ductile failure mode. The EBC is a thin coating of SiC applied to the outside of the composite to protect it against corrosion. The ends of the tubes may be sealed via the bonding of SiC end caps to the SiC tube. Triplex tube samples, monolith-only samples, and SiC/SiC bonding samples (consisting of two blocks bonded together) were tested in three phases as part of an evaluation of the SiC cladding system. A number of samples were exposed to PWR coolant and neutronic conditions using an incore loop in the MIT research reactor (MITR-II). Other samples remained in their as-fabricated states for comparison. First, mechanical testing revealed significant strength reduction in the Triplex samples due to irradiation-induced point defects, corrosive pitting of the monolith, and possible differences in the behavior of the Triplex components. Some manufacturing abnormalities were also discovered which could have compromised strength. The Triplex samples tested here were not as strong as reported in a previous study. SEM analysis was able to follow the propagation of cracks from initiation, at the monolith inner surface, to termination, upon breaching the EBC. The composite layer was found to be key in dissipating the energy driving the crack formation. Second, three SiC/SiC bonding methods (six samples total) were tested in the MITR-II to 0.2 dpa, and five of the six samples failed. SEM analysis indicates radiation induced degradation of the bond material. Dimensional and volume measurements established the anisotropic swelling of the two SiC blocks in each bond sample, which would have caused shear stresses on the bonds, contributing to their failure. Finally, thermal diffusivity measurements of the Triplex samples show substantial decreases with irradiation (saturating at about 1 dpa) due to the accumulation of phonon-scattering defects and corrosion of SiC. By 1 dpa, the thermal diffusivity/conductivity of this SiC cladding design is diminished to a value lower than that of Zircaloy. In the as-fabricated state, a large difference exists between the monolith-only and Triplex samples due to the phonon scattering centers at the interfaces of the layers. With irradiation this difference decreases, suggesting that similar corrosion and radiation damage effects exist in both the monolith and Triplex samples.
by John D. Stempien.
S.M.
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.
Full textAuguste, 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.
Full textThis electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.
Cataloged from student-submitted PDF version of thesis.
Includes bibliographical references (pages 418-420).
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 left by CRUD. CRUD has forced many nuclear reactors into temporary shutdown or production downgrades, costing millions of dollars US per reactor. If the CRUD growth factors could be fully understood, they could be controlled, and the CRUD problem could be eliminated altogether. Atomic force microscopy can be used to measure the force, or the strength of the CRUD-clad bond with different materials. This research focuses on answering this question: How does the force change between CRUD particles and different materials that could be used for reactor cladding? This study will analyze lab-grown CRUD samples on different substrate materials and characterize CRUD growth on each. It was found the CRUD-bond forces (from least to greatest) on silicon carbide (SiC), Titanium aluminum carbide (Ti2AlC), and max-phase zirconium alloy 211(Zr4M211) behaved similarly in air and in water. The forces on each surface increased with increasing dwell time for the Fe3O4 particle AFM tip; in contrast, most adhesion forces stayed constant with the NiO AFM tip. Furthermore, these CRUD forces were compared to other non-accident tolerant fuels, and there are cases in which non-ATF materials show more CRUD resistance (less adhesive force) than ATF-materials. This study's analysis could be applied to other materials to be used for reactor cladding. Once the material with the lowest-strength CRUD bond is identified and installed, the nuclear industry could save millions of dollars US per reactor fuel cycle.
by Rasheed Auguste.
S.B.
Books on the topic "Nuclear fuel cladding tube"
Causey, A. R. Irradiation-enhanced creep of cold-worked Zr-2.5Nb tubes and helical-springs. Chalk River, Ont: Reactor Materials Research Branch, Chalk River Laboratories, 1993.
Find full textYegorova, 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. Washington, D.C: U.S. Nuclear Regulatory Commission, 1999.
Find full textSardjono, Ignatius Djoko. Study of burnout phenomena on a cylindrical heater tube simulating nuclear fuel rod. Ottawa: National Library of Canada, 1990.
Find full textCoordinated 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. Bombay: The Centre, 1986.
Find full textR, 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. Chalk River, Ont: Fuel Channel Components Branch, Chalk River Laboratories, 1994.
Find full text1935-, Causey A. R., and Nuclear Industry Conference (1993 : Baltimore, Md.), eds. On the anisotropy of in-reactor creep of Zr-2.5Nb tubes. Chalk River, Ont: Reactor Materials Research Branch, Chalk River Laboratories, 1993.
Find full textAlmarshad, Abdullah I. A. A model for waterside oxidation of zircaloy fuel cladding in pressurized water reactors. 1990.
Find full textRen, Yongli. Fracture toughness behavior of Zircaloy-4 in the form of fuel cladding tubing in nuclear reactors. 2004.
Find full textP, 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. Washington, D.C: Division of Safety Review and Oversight, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 1986.
Find full textThe Impact of fuel cladding failure events on occupational radiation exposures at nuclear power plants: Case study. Washington, D.C: Division of Safety Review and Oversight, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 1986.
Find full textBook chapters on the topic "Nuclear fuel cladding tube"
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, 705–8. Stafa: Trans Tech Publications Ltd., 2007. http://dx.doi.org/10.4028/0-87849-440-5.705.
Full textKim, 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, 709–12. Stafa: Trans Tech Publications Ltd., 2007. http://dx.doi.org/10.4028/0-87849-440-5.709.
Full textSakamoto, 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, 1054–72. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2010. http://dx.doi.org/10.1520/stp49293t.
Full textSakamoto, 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, 1054–72. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2010. http://dx.doi.org/10.1520/stp49391s.
Full textSakamoto, 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, 1054–72. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2010. http://dx.doi.org/10.1520/stp152920120041.
Full textYagnik, 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, 1077–106. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2014. http://dx.doi.org/10.1520/stp154320120192.
Full textPark, 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, 1243–46. Stafa: Trans Tech Publications Ltd., 2006. http://dx.doi.org/10.4028/0-87849-415-4.1243.
Full textKubo, 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, 433–65. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2011. http://dx.doi.org/10.1520/stp49270t.
Full textKubo, 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, 433–65. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2011. http://dx.doi.org/10.1520/stp49368s.
Full textKubo, 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, 433–65. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2011. http://dx.doi.org/10.1520/stp152920120018.
Full textConference papers on the topic "Nuclear fuel cladding tube"
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, C. Cojocaru, B. Jodoin, H. Koivuluoto, Y. C. Lau, et al. ASM International, 2021. http://dx.doi.org/10.31399/asm.cp.itsc2021p0167.
Full textQin, 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.
Full textAbe, 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.
Full textPark, 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.
Full textChen, 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.
Full textPengbo, 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.
Full textWang, 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.
Full textSoloviev, Sergei L., Boris A. Gabaraev, Leonid M. Parafilo, Dmitry V. Kruchkov, Oleg Yu Novoselsky, Vladimir N. Filinov, and Oleg I. Melikhov. "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.
Full textLi, 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.
Full textAly, 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.
Full textReports on the topic "Nuclear fuel cladding tube"
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), December 2006. http://dx.doi.org/10.2172/896213.
Full textJaramillo, 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), March 2007. http://dx.doi.org/10.2172/931509.
Full textKristine Barrett and Shannon Bragg-Sitton. Advanced LWR Nuclear Fuel Cladding System Development Trade-Off Study. Office of Scientific and Technical Information (OSTI), September 2012. http://dx.doi.org/10.2172/1057698.
Full textLister, 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), December 2018. http://dx.doi.org/10.2172/1492831.
Full textYang, 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), November 2017. http://dx.doi.org/10.2172/1413204.
Full textHeuser, Brent, James Stubbins, Tomasz Kozlowski, Rizwan Uddin, Dallas Trinkle, Thoms Downar, Gary Was, Yong ang, Simon Phillpot, and piyush Sabharwall. Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel. Office of Scientific and Technical Information (OSTI), July 2017. http://dx.doi.org/10.2172/1391853.
Full textLu, 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), January 2014. http://dx.doi.org/10.2172/1116513.
Full textSimon, 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), September 2017. http://dx.doi.org/10.2172/1473586.
Full textBahney, Robert. Effective thermal conductivity method for predicting spent nuclear fuel cladding temperatures in a dry fill gas. Office of Scientific and Technical Information (OSTI), December 1997. http://dx.doi.org/10.2172/757327.
Full textGreiner, 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), March 2017. http://dx.doi.org/10.2172/1358184.
Full text