• Title, Summary, Keyword: cladding tube

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FEA Study on Hoop Stress of Multilayered SiC Composite Tube for Nuclear Fuel Cladding (핵연료 피복관용 다중층 SiC 복합체 튜브의 Hoop Stress 전산모사 연구)

  • Lee, Hyeon-Geun;Kim, Daejong;Park, Ji Yeon;Kim, Weon-Ju
    • Journal of the Korean Ceramic Society
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    • v.51 no.5
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    • pp.435-441
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    • 2014
  • Silicon carbide-based ceramics and their composites have been studied for application to fusion and advanced fission energy systems. For fission reactors, $SiC_f$/SiC composites can be applied to core structural materials. Multilayered SiC composite fuel cladding, owing to its superior high temperature strength and low hydrogen generation under severe accident conditions, is a candidate for the replacement of zirconium alloy cladding. The SiC composite cladding has to retain its mechanical properties and original structure under the inner pressure caused by fission products; as such it can be applied as a cladding in fission reactor. A hoop strength test using an expandable polyurethane plug was designed in order to evaluate the mechanical properties of the fuel cladding. In this paper, a hoop strength test of the multilayered SiC composite tube for nuclear fuel cladding was simulated using FEA. The stress caused by the plug was distributed nonuniformly because of the friction coefficient difference between the inner surface of the tube and the plug. Hoop stress and shear stress at the tube was evaluated and the relationship between the concentrated stress at the inner layer of the tube and the fracture behavior of the tube was investigated.

A novel approach for manufacturing oxide dispersion strengthened (ODS) steel cladding tubes using cold spray technology

  • Maier, Benjamin;Lenling, Mia;Yeom, Hwasung;Johnson, Greg;Maloy, Stuart;Sridharan, Kumar
    • Nuclear Engineering and Technology
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    • v.51 no.4
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    • pp.1069-1074
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    • 2019
  • A novel fabrication method of oxide dispersion strengthened (ODS) steel cladding tubes for advanced fast reactors has been investigated using the cold spray powder-based materials deposition process. Cold spraying has the potential advantage for rapidly fabricating ODS cladding tubes in comparison with the conventional multi-step extrusion process. A gas atomized spherical 14YWT (Fe-14%Cr, 3%W, 0.4%Ti, 0.2% Y, 0.01%O) powder was sprayed on a rotating cylindrical 6061-T6 aluminum mandrel using nitrogen as the propellant gas. The powder lacked the oxygen content needed to precipitate the nanoclusters in ODS steel, therefore this work was intended to serve as a proof-of-concept study to demonstrate that free-standing steel cladding tubes with prototypical ODS composition could be manufactured using the cold spray process. The spray process produced an approximately 1-mm thick, dense 14YWT deposit on the aluminum-alloy tube. After surface polishing of the 14YWT deposit to obtain desired cladding thickness and surface roughness, the aluminum-alloy mandrel was dissolved in an alkaline medium to leave behind a free-standing ODS tube. The as-fabricated cladding tube was annealed at $1000^{\circ}C$ for 1 h in an argon atmosphere to improve the overall mechanical properties of the cladding.

MODAL TESTING AND MODEL UPDATING OF A REAL SCALE NUCLEAR FUEL ROD

  • Park, Nam-Gyu;Rhee, Hui-Nam;Moon, Hoy-Ik;Jang, Young-Ki;Jeon, Sang-Youn;Kim, Jae-Ik
    • Nuclear Engineering and Technology
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    • v.41 no.6
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    • pp.821-830
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    • 2009
  • In this paper, modal testing and finite element modeling results to identify the modal parameters of a nuclear fuel rod as well as its cladding tube are discussed. A vertically standing full-size cladding tube and a fuel rod with lead pellets were used in the modal testing. As excessive flow-induced vibration causes a failure in fuel rods, such as fretting wear, the vibration level of fuel rods should be low enough to prevent failure of these components. Because vibration amplitude can be estimated based on the modal parameters, the dynamic characteristics must be determined during the design process. Therefore, finite element models are developed based on the test results. The effect of a lumped mass attached to a cladding tube model was identified during the finite element model optimization process. Unlike a cladding tube model, the density of a fuel rod with pellets cannot be determined in a straightforward manner because pellets do not move in the same phase with the cladding tube motion. The density of a fuel rod with lead pellets was determined by comparing natural frequency ratio between the cladding tube and the rod. Thus, an improved fuel rod finite element model was developed based on the updated cladding tube model and an estimated fuel rod density considering the lead pellets. It is shown that the entire pellet mass does not contribute to the fuel rod dynamics; rather, they are only partially responsible for the fuel rod dynamic behavior.

Residual Stress Analysis of Laser Cladding Repair for Nuclear Steam Generator Damaged Tubes (원전 증기발생기 레이저 클래딩 보수부위 잔류응력 해석)

  • Han, Won-Jin;Lee, Sang-Cheol;Lee, Seon-Ho
    • Proceedings of the KSME Conference
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    • pp.56-60
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    • 2008
  • Laser cladding technology was studied as a method for upgrading the present repair procedures of damaged tubes in a nuclear steam generator and Doosan subsequently developed and designed a new Laser Cladding Repair System. One of the important features of this newly developed Laser Cladding Repair System is that molten metal can be deposited on damaged tube surfaces using a laser beam and filler wire without the need to install sleeves inside the tube. Laser cladding qualification tests on the steam generator tube material, Alloy 600, were performed according to ASME Section IX. Residual stress analyses were performed for weld metal and heat affected zone of as-welded and PWHT with SYSWELD software.

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The wire laser cladding system for repairing inner side of Alloy 600 tubes (Alloy 600 전열관 내면 보수용 와이어 송급 레이저 클래딩 장치 개발)

  • Han, Won-Jin;Kim, U-Seong;Lee, Sang-Cheol;Lee, Seon-Ho;Lee, Chang-Yeol
    • Proceedings of the KWS Conference
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    • pp.196-198
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    • 2007
  • Laser cladding technology was studied as a method for upgrading the present repair procedures of damaged tubes in a nuclear steam generator and Doosan subsequently developed and designed a new Laser Cladding Repair System. One of the important features of this newly developed Laser Cladding Repair System is that molten metal can be deposited on damaged tube surfaces using a laser beam and filler wire without the need to install sleeves inside the tube. Laser cladding qualification tests on the steam generator tube material, Alloy 600, were performed according to ASME Section IX.

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A Study on the Laser Beam Weldability Using Zircaloy-4 Cladding Tube (지르칼로이-4 피복관을 이용한 레이저용접성 연구)

  • 박진석;김동균;김상태;양명승;김수성;이정원
    • Journal of Welding and Joining
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    • v.20 no.6
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    • pp.796-801
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    • 2002
  • Corrosion and tensile properties of zircaloy-4 cladding tube having a laser welding part in elevated temperature are studied to present the criterion of quality evaluation in nuclear reactor and find the scientific basis of SCC, with laser welding method using by coupling up cladding tube to end cap. In the result of tensile test($400^{\circ}C$), the fracture is not happened in the welding part but base metal and the result of corrosion test($400^{\circ}C$ 1500psi steam), corrosion rate of the molten zone and PMZ is a little higher than the other zone.

A Study on Mechanical Properties of Fuel Cladding Materials (원자로용 핵연료 피복재의 인장특성에 관한 연구)

  • Bae, Bong-Kook;Song, Chun-Ho;Seok, Chang-Sung
    • Proceedings of the KSME Conference
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    • pp.489-494
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    • 2001
  • The fuel of light water reactor used far several years at high temperature and pressure, so it needs to clad with high corrosion resistance material. The cladding materials need low absorption of a neutron and high corrosion resistance. Cladding materials used Zircaloy-2 in Boiling Water Reactor, Zircaloy-4 in Pressurized Water Reactor and Zirlo has good for long term corrosion. If fracture of cladding tube occured during operation, it caused disaster. So it is needed to estimate of integrity fur cladding materials. In this paper, tension characteristics of cladding materials are investigate which is basic research far fracture characteristic. Also analysis of residual stress effect between tube type(original type) specimen and flattened type specimen.

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Impact of Anisotropy in Creep and Irradiation Growth on the KOFA Zircaloy-4 Cladding tube Deformation Behavior (크립 및 조사성장 이방성이 KOFA Zircaloy-4 피복관의 변형거동에 미치는 영향)

  • Kim, Gi-Hang;Lee, Chan-Bok;Kim, Gyu-Tae
    • Korean Journal of Materials Research
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    • v.4 no.4
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    • pp.445-452
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    • 1994
  • Three-axial deformation behavior of the Zircaloy cladding tube under the irradiation condition of the fuel in pressurized water reactor can be analyzed by the anisotropy in the creep and the irra- diation growth, which depends on the texture parameter. A methodology to evaluate the impact of the anisotropic creep and irradiation growth on the strain in each axial direction of the cladding tube has been proposed. Based on the measured strains after irradiation and predicted ones with the help of a fuel performance analysis code, it is found that a tangential strain of the cladding tube is caused mainly by the creep, whereas a axial strain of the cladding is caused mainly by the irradiation growth but with a considerable contribution of the creep at low irradiation.

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Fretting Wear Characteristics of the Corroded Fuel Cladding Tubes for Nuclear Fuel Rod against Supporting Girds (부식된 핵연료 피복관과 지지격자 사이의 프레팅 마멸 특성)

  • Kim, Jin-Seon;Park, Se-Min;Kim, Yong-Hwan;Lee, Seung-Jae;Lee, Young-Ze
    • Tribology and Lubricants
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    • v.23 no.3
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    • pp.130-133
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    • 2007
  • 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 and support. The fretting wear of tube and 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, especially after corrosion treatment on tubes, in water. The tests were done using various applied loads with fixed amplitude. From the results of fretting tests, the wear amounts of tube materials can be predictable by obtaining the wear coefficient using the work rate model. Due to stick phenomena the wear depth was changed as increasing load and temperature. The maximum wear depth was decreased as increasing the water temperatures. At high temperatures there are the regions of some severe adhesion due to stick phenomena.