Works matching DE "NUCLEAR fuel claddings"
Results: 284
In situ imaging of corrosion processes in nuclear fuel cladding.
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- Corrosion Engineering, Science & Technology, 2017, v. 52, n. 8, p. 596, doi. 10.1080/1478422X.2017.1344038
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Effect of temperature on corrosion and semiconducting properties of oxide films formed on M5 zirconium alloy.
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- Corrosion Engineering, Science & Technology, 2016, v. 51, n. 2, p. 104, doi. 10.1179/1743278215Y.0000000036
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基于局部干法的核电水池覆面缺陷焊接 修复工艺研究.
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- Metal Working (1674-165X), 2024, n. 10, p. 65
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Investigation on Neutronic Parameters of the KLT-40S Reactor Core with U<sub>3</sub>Si<sub>2</sub>-FeCrAl using SCALE Code.
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- Journal of Engineering & Technological Sciences, 2023, v. 55, n. 1, p. 22, doi. 10.5614/j.eng.technol.sci.2023.55.1.3
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Effect of surface oxides on tritium entrance and permeation in FeCrAl alloys for nuclear fuel cladding: a review.
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- Corrosion Reviews, 2023, v. 41, n. 2, p. 143, doi. 10.1515/corrrev-2022-0033
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Protective coatings on zirconium-based alloys as accident-tolerant fuel (ATF) claddings.
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- Corrosion Reviews, 2017, v. 35, n. 3, p. 141, doi. 10.1515/corrrev-2017-0010
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Performance of FeCrAl for accident-tolerant fuel cladding in high-temperature steam.
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- Corrosion Reviews, 2017, v. 35, n. 3, p. 167, doi. 10.1515/corrrev-2016-0067
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Performance degradation of candidate accidenttolerant cladding under corrosive environment.
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- Corrosion Reviews, 2017, v. 35, n. 3, p. 129, doi. 10.1515/corrrev-2017-0014
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Deconvoluting Cr states in Cr-doped UO<sub>2</sub> nuclear fuels via bulk and single crystal spectroscopic studies.
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- Nature Communications, 2023, v. 14, n. 1, p. 1, doi. 10.1038/s41467-023-38109-0
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EXPERIMENTAL RESEARCH CONCERNING WATERSIDE CORROSION OF FUEL CLADDING IN CANDU REACTOR.
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- Journal of Science & Arts, 2024, v. 24, n. 1, p. 211, doi. 10.46939/J.Sci.Arts-24.1-b03
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Radiation Effects in Nuclear Ceramics.
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- Advances in Materials Science & Engineering, 2012, p. 1, doi. 10.1155/2012/905474
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A COMPARISON OF ALGORITHMS EFFICIENCY FOR MANEUVERING POWER OF THE WWER-1000 REACTOR.
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- Naukovi visti NTUU - KPI, 2010, v. 2010, n. 5, p. 10
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- Article
Thermal‐hydraulic analysis of UO2 and MOX fuel considering different cladding materials at various burnup levels in pressurized water reactor.
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- Heat Transfer, 2021, v. 50, n. 7, p. 7215, doi. 10.1002/htj.22225
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Redistribution of Nb and other alloying elements in Nb-doped Zr alloy under high dose ion irradiation.
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- Journal of Nuclear Science & Technology, 2024, v. 61, n. 12, p. 1521, doi. 10.1080/00223131.2024.2364711
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A study on the fracture pattern change of high-burnup fuel cladding failed by pellet-cladding mechanical interaction failure under reactivity-initiated accident conditions.
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- Journal of Nuclear Science & Technology, 2024, v. 61, n. 9, p. 1265, doi. 10.1080/00223131.2024.2313553
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Approximation of photonic crystal fibres with large air holes by the step index fibre model.
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- Journal of Modern Optics, 2008, v. 55, n. 9, p. 1479, doi. 10.1080/09500340701668572
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Impact of Pressure Relief Holes on Core Coolability for a PWR During a Large-Break Loss-of-Coolant Accident with Core Blockage Using RELAP5-3D.
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- Nuclear Technology, 2016, v. 193, n. 1, p. 88, doi. 10.13182/NT14-147
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DEPENDENCE OF FIRE TIME OF CONCERN ON LOCATION OF A ONE-ASSEMBLY TRANSPORT PACKAGE.
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- Nuclear Technology, 2015, v. 192, n. 2, p. 142, doi. 10.13182/NT14-156
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- Article
FUEL PERFORMANCE ANALYSIS OF A (ThU)O<sub>2</sub>-FUELED, REDUCED MODERATION BOILING WATER REACTOR.
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- Nuclear Technology, 2015, v. 191, n. 3, p. 268, doi. 10.13182/NT14-104
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A CORE DESIGN STUDY ON THE FUEL DISPLACEMENT OPTIONS FOR AN EFFECTIVE TRANSITION BETWEEN BREAKEVEN AND TRU BURNING SODIUM-COOLED FAST REACTORS.
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- Nuclear Technology, 2014, v. 186, n. 3, p. 390, doi. 10.13182/NT13-90
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URANIUM STARTUP FAST REACTORS WITH METAL FUEL USING ONCE-THROUGH FUEL CYCLE.
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- Nuclear Technology, 2014, v. 186, n. 3, p. 378, doi. 10.13182/NT13-21
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FUEL MANAGEMENT OF PWR CORES WITH SILICON CARBIDE CLADDING.
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- Nuclear Technology, 2014, v. 186, n. 3, p. 353, doi. 10.13182/NT12-131
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DEVELOPMENT AND APPLICATION OF AN INTEGRATED FUEL PERFORMANCE AND SUBCHANNEL MODEL FOR ANALYSIS OF SODIUM FAST REACTORS.
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- Nuclear Technology, 2013, v. 184, n. 1, p. 63, doi. 10.13182/NT13-A19869
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CHALLENGES ASSOCIATED WITH THE MITIGATION OF UNPROTECTED LOSS OF FLOW IN SODIUM-COOLED FAST REACTOR CORES.
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- Nuclear Technology, 2013, v. 181, n. 1, p. 56, doi. 10.13182/NT13-A15756
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ANALYSIS OF THE BR2 LOSS-OF-FLOW TEST A.
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- Nuclear Technology, 2011, v. 176, n. 1, p. 93, doi. 10.13182/NT11-A12545
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INSTRUMENT CAPABILITIES AND DEVELOPMENTS AT THE HALDEN REACTOR PROJECT.
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- Nuclear Technology, 2011, v. 173, n. 1, p. 78, doi. 10.13182/NT11-A11486
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ANALYSIS OF THE WATERSIDE CORROSION KINETICS OF ZIRCALOY-4 FUEL CLADDING IN FRENCH PWRs.
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- Nuclear Technology, 2010, v. 170, n. 3, p. 444, doi. 10.13182/NT10-A10330
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COMPARISON OF THE HIGH-TEMPERATURE STEAM OXIDATION KINETICS OF ADVANCED CLADDING MATERIALS.
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- Nuclear Technology, 2010, v. 170, n. 1, p. 272, doi. 10.13182/NT10-A9464
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LARGE BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS OF VVER-1000 REACTOR USING CATHARE CODE.
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- Nuclear Technology, 2010, v. 170, n. 1, p. 123, doi. 10.13182/NT10-A9451
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DISSOLUTION OF ZIRCALOY-2-CLAD UO<sub>2</sub> COMMERCIAL REACTOR FUEL.
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- Nuclear Technology, 2010, v. 169, n. 3, p. 263, doi. 10.13182/NT10-A9378
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USE OF GEOMETRICALLY ACCURATE FUEL MODELS TO PREDICT CLADDING AND BASKET TEMPERATURES WITHIN A TRUCK CASK DURING NORMAL TRANSPORT.
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- Nuclear Technology, 2009, v. 167, n. 2, p. 313, doi. 10.13182/NT09-A8966
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A radial heat flow apparatus for thermal conductivity characterisation of cylindrical samples.
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- Journal of Thermal Analysis & Calorimetry, 2017, v. 130, n. 3, p. 1855, doi. 10.1007/s10973-017-6578-8
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Analysis of the technological process of welding a membrane wall with Inconel 625 nickel alloy.
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- International Journal of Advanced Manufacturing Technology, 2023, v. 127, n. 5/6, p. 3031, doi. 10.1007/s00170-023-11499-7
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A precision grinding technology for zirconium alloy tubes based on ultrasonic wall thickness automatic measurement system.
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- International Journal of Advanced Manufacturing Technology, 2022, v. 121, n. 1/2, p. 323, doi. 10.1007/s00170-022-09314-w
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Energy consumption modeling of additive-subtractive hybrid manufacturing based on cladding head moving state and deposition efficiency.
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- International Journal of Advanced Manufacturing Technology, 2022, v. 120, n. 11/12, p. 7755, doi. 10.1007/s00170-022-09265-2
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Effect of pre-cladding dust layer on filtration and pulse-jet cleaning performance of filter media.
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- Process Safety & Environmental Protection: Transactions of the Institution of Chemical Engineers Part B, 2024, v. 183, p. 1, doi. 10.1016/j.psep.2023.12.064
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Hoop Strain Measurement During a SiC/SiC Ceramic Composite Tube Burst Test by Digital Volume Correlation of X-Ray Computed Tomographs.
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- Experimental Mechanics, 2023, v. 63, n. 2, p. 275, doi. 10.1007/s11340-022-00916-9
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Nondestructive Elemental Diagnostics of the Fuel-Rod Cladding Surface by the Ion-Beam and X-Ray Analytical Methods.
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- Instruments & Experimental Techniques, 2021, v. 64, n. 1, p. 63, doi. 10.1134/S0020441221010085
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Impact of Environmental Exposure on the Service Life of Façade Claddings—A Statistical Analysis.
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- Buildings (2075-5309), 2021, v. 11, n. 12, p. 615, doi. 10.3390/buildings11120615
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Stochastic Simulation of Mould Growth Performance of Wood-Frame Building Envelopes under Climate Change: Risk Assessment and Error Estimation.
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- Buildings (2075-5309), 2021, v. 11, n. 8, p. 333, doi. 10.3390/buildings11080333
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High-Temperature Tensile Properties and Serrated Flow Behavior of FeCrAl Alloy for Accident-Tolerant Fuel Cladding.
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- Applied Sciences (2076-3417), 2024, v. 14, n. 24, p. 11748, doi. 10.3390/app142411748
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Weak-Edge Extraction of Nuclear Plate Fuel Neutron Images at Low Lining Degree.
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- Applied Sciences (2076-3417), 2023, v. 13, n. 8, p. 5090, doi. 10.3390/app13085090
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Numerical Study of Coupled Fluid and Solid Wave Propagation Related to the Cladding Failure of a Nuclear Fuel Rod.
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- Applied Sciences (2076-3417), 2022, v. 12, n. 4, p. 1784, doi. 10.3390/app12041784
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Interface Characterization within a Nuclear Fuel Plate.
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- Applied Sciences (2076-3417), 2019, v. 9, n. 2, p. 249, doi. 10.3390/app9020249
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Protective Cr Coatings with ZrO 2 /Cr Multilayers for Zirconium Fuel Claddings.
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- Coatings (2079-6412), 2022, v. 12, n. 10, p. 1409, doi. 10.3390/coatings12101409
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Oxidation of Silicon Carbide Composites for Nuclear Applications at Very High Temperatures in Steam.
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- Coatings (2079-6412), 2022, v. 12, n. 7, p. N.PAG, doi. 10.3390/coatings12070875
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Friction and Wear Properties of Cr-N x Coatings for Nuclear Fuel Cladding.
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- Coatings (2079-6412), 2022, v. 12, n. 2, p. 163, doi. 10.3390/coatings12020163
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On the Use of Chromium Coating for Inner-Side Fuel Cladding Protection: Thickness Identification Based on Fission Fragments Implantation and Damage Profile.
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- Coatings (2079-6412), 2021, v. 11, n. 6, p. 710, doi. 10.3390/coatings11060710
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High-Temperature Oxidation of Cr-Coated Resistance Upset Welds Made from E110 Alloy.
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- Coatings (2079-6412), 2021, v. 11, n. 5, p. 577, doi. 10.3390/coatings11050577
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Recent Advances in Protective Coatings for Accident Tolerant Zr-Based Fuel Claddings.
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- Coatings (2079-6412), 2021, v. 11, n. 5, p. 557, doi. 10.3390/coatings11050557
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