• Title/Summary/Keyword: cladding tube

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Crystallization Behavior of Amorphous Ti-Be Alloys as Filler Metals for Joining Zircaloy-4 Tubes and Microstructures of the Brazed Zones (지르칼로이-4 브레이징용 비정질 Ti-Be 용가재의 결정화 거동 및 접합부 미세조직)

  • Kim, Sang-Ho;Go, Jin-Hyeon;Park, Chun-Ho
    • Korean Journal of Materials Research
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    • v.12 no.4
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    • pp.259-263
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    • 2002

  • Three different ribbons of amorphous $Til_{1-x}Be_x$ alloys such as $Ti_{0.59}Be_{0.41},\;Ti_{0.61}Be_{0.39}\;and\;Ti_{0.63}Be_{0.37}$ were made by melt-spinning method to be used as brazing filler metals for joining Zircaloy-4 nuclear fuel cladding tubes, and their crystallization behavior as well as microstructure of the brazed zone were examined. The crystallization behavior was investigated in teams of thermal stability, crystallization temperature and activation energy. The crystallization of the $Ti_{1-x}Be_x$ alloys proceeded in two steps by the formation of ${\alpha}$-Ti at a lower temperature and of TiBe at a higher temperature. The crystallization temperature and activation energy of $Ti_{1-x}Be_x$ alloys were higher and larger than those of $Zr_{1-x}Be_x$ alloys and PVD Be. Those resulted thinner joining layer with $Ti_{1-x}Be_x$ alloys, which kept sound thickness of Zircaloy-4 nuclear fuel cladding tubes after brazing. But in the brazed zones made by $Ti_{1-x}Be_x$ filler metals, a little solid-solution layers composed of Zr and Ti were formed toward the Zr cladding tube and Zr was detected in the brazed zones. Microstructure of brazed zone was changed from globular to dentrite with decreasing Be content in the $Ti_{1-x}Be_x$ filler metal.

  • https://doi.org/10.3740/MRSK.2002.12.4.259 인용 PDF KSCI

Evaluation of the Tensile Properties of Fuel Cladding at High Temperatures Using a Ring Specimen (링 시험편을 이용한 피복관의 고온 인장특성 평가)

  • Bae Bong-Kook;Koo Jae-Mean;Seok Chang-Sung
    • Transactions of the Korean Society of Mechanical Engineers A
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    • v.29 no.4 s.235
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    • pp.600-605
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    • 2005

  • In this study, the ring tensile test at high temperature was suggested to evaluate the hoop tensile properties of small tube such as the cladding in the nuclear reactor Using the Arsene's ring model, the ring tensile test was performed and the test data were calibrated. From the result of the ring test with strain gauge and the numerical analysis with 1/8 model, LCRR(load-displacement conversion relationship of ring specimen) was determined. We could obtain the hoop tensile properties by means of applying the LCRR to the calibrated data of the ring tensile test. A few difference was observed in view of the shape of fractured surface and the fracture mechanism between at the high temperature and at the room temperature.

  • https://doi.org/10.3795/KSME-A.2005.29.4.600 인용 PDF KSCI

Improving Accident Tolerance of Nuclear Fuel with Coated Mo-alloy Cladding

  • Cheng, Bo;Kim, Young-Jin;Chou, Peter
    • Nuclear Engineering and Technology
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    • v.48 no.1
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    • pp.16-25
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    • 2016

  • In severe loss of coolant accidents (LOCA), similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconiumalloy fuel claddingmaterials are rapidlyheateddue to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident management, an accident tolerant fuel (ATF) design may extend coping and recovery time for operators to restore emergency power, and cooling, and achieve safe shutdown. An ATF is required to possess high resistance to steam oxidation to reduce hydrogen generation and sufficient mechanical strength to maintain fuel rod integrity and core coolability. The initiative undertaken by Electric Power Research Institute (EPRI) is to demonstrate the feasibility of developing an ATF cladding with capability to maintain its integrity in $1,200-1,500^{\circ}C$ steam for at least 24 hours. This ATF cladding utilizes thin-walled Mo-alloys coated with oxidation-resistant surface layers. The basic design consists of a thin-walled Mo alloy structural tube with a metallurgically bonded, oxidation-resistant outer layer. Two options are being investigated: a commercially available iron, chromium, and aluminum alloy with excellent high temperature oxidation resistance, and a Zr alloy with demonstratedcorrosionresistance.Asthese composite claddings will incorporate either no Zr, or thin Zr outer layers, hydrogen generation under severe LOCA conditions will be greatly reduced. Key technical challenges and uncertainties specific to Moalloy fuel cladding include: economic core design, industrial scale fabricability, radiation embrittlement, and corrosion and oxidation resistance during normal operation, transients, and severe accidents. Progress in each aspect has been made and key results are discussed in this document. In addition to assisting plants in meeting Light Water Reactor (LWR) challenges, accident-tolerant Mo-based cladding technologies are expected to be applicable for use in high-temperature helium and molten salt reactor designs, as well as nonnuclear high temperature applications.

  • https://doi.org/10.1016/j.net.2015.12.003 인용 PDF KSCI