WorldWideScience

Sample records for clad metals

  1. Metal-clad waveguide sensors

    DEFF Research Database (Denmark)

    Skivesen, Nina

    This work concerns planar optical waveguide sensors for biosensing applications, with the focus on deep-probe sensing for micron-scale biological objects like bacteria and whole cells. In the last two decades planar metal-clad waveguides have been brieflyintroduced in the literature applied...... for various biosensing applications, however a thorough study of the sensor configurations has not been presented, but is the main subject of this thesis. Optical sensors are generally well suited for bio-sensing asthey show high sensitivity and give an immediate response for minute changes in the refractive...... index of a sample, due to the high sensitivity of optical bio-sensors detection of non-labeled biological objects can be performed. The majority of opticalsensors presented in the literature and commercially available optical sensors are based on evanescent wave sensing, however most of these sensors...

  2. Optimization of metal-clad waveguide sensors

    DEFF Research Database (Denmark)

    Skivesen, N.; Horvath, R.; Pedersen, H.C.

    2005-01-01

    The present paper deals with the optimization of metal-clad waveguides for sensor applications to achieve high sensitivity for adlayer and refractive index measurements. By using the Fresnel reflection coefficients both the angular shift and the width of the resonances in the sensorgrams are taken...... into account. Our optimization shows that it is possible for metal-clad waveguides to achieve a sensitivity improvement of 600% compared to surface-plasmon-resonance sensors....

  3. Clad buffer rod sensors for liquid metals

    International Nuclear Information System (INIS)

    Clad buffer rods, consisting of a core and a cladding, have been developed for ultrasonic monitoring of liquid metal processing. The cores of these rods are made of low ultrasonic-loss materials and the claddings are fabricated by thermal spray techniques. The clad geometry ensures proper ultrasonic guidance. The lengths of these rods ranges from tens of centimeters to 1m. On-line ultrasonic level measurements in liquid metals such as magnesium at 700 deg C and aluminum at 960 deg C are presented to demonstrate their operation at high temperature and their high ultrasonic performance. A spherical concave lens is machined at the rod end for improving the spatial resolution. High quality ultrasonic images have been obtained in the liquid zinc at 600 deg C. High spatial resolution is needed for the detection of inclusions in liquid metals during processing. We also show that the elastic properties such as density, longitudinal and shear wave velocities of liquid metals can be measured using a transducer which generates and receives both longitudinal and shear waves and is mounted at the end of a clad buffer rod. (author)

  4. Deep-probe metal-clad waveguide biosensors

    DEFF Research Database (Denmark)

    Skivesen, Nina; Horvath, Robert; Thinggaard, S.;

    2007-01-01

    Two types of metal-clad waveguide biosensors, so-called dip-type and peak-type, are analyzed and tested. Their performances are benchmarked against the well-known surface-plasmon resonance biosensor, showing improved probe characteristics for adlayer thicknesses above 150-200 nm. The dip-type metal......-clad waveguide sensor is shown to be the best all-round alternative to the surface-plasmon resonance biosensor. Both metal-clad waveguides are tested experimentally for cell detection, showing a detection linut of 8-9 cells/mm(2). (c) 2006 Elsevier B.V. All rights reserved....

  5. Clad metals by roll bonding for SOFC interconnects

    Science.gov (United States)

    Chen, L.; Jha, B.; Yang, Zhenguo; Xia, Guang-Guang; Stevenson, Jeffry W.; Singh, Prabhakar

    2006-08-01

    High-temperature oxidation-resistant alloys are currently considered as a candidate material for construction of interconnects in intermediate-temperature solid oxide fuel cells. Among these alloys, however, different groups of alloys demonstrate different advantages and disadvantages, and few, if any, can completely satisfy the stringent requirements for the application. To integrate the advantages and avoid the disadvantages of different groups of alloys, cladding has been proposed as one approach in fabricating metallic layered interconnect structures. To examine the feasibility of this approach, the austenitic Ni-base alloy Haynes 230 and the ferritic stainless steel AL 453 were selected as examples and manufactured into a clad metal. Its suitability as an interconnect construction material was investigated. This paper provides a brief overview of the cladding approach and discusses the viability of this technology to fabricate the metallic layered-structure interconnects.

  6. Performance of HT9 clad metallic fuel at high temperature

    International Nuclear Information System (INIS)

    Steady-state testing of HT9 clad metallic fuel at high temperatures was initiated in EBR-II in November of 1987. At that time U-10 wt. % Zr fuel clad with the low-swelling ferritic/martensitic alloy HT9 was being considered as driver fuel options for both EBR-II and FFTF. The objective of the X447 test described here was to determine the lifetime of HT9 cladding when operated with metallic fuel at beginning of life inside wall temperatures approaching ∼660 degree C. Though stress-temperature design limits for HT9 preclude its use for high burnup applications under these conditions due to excessive thermal creep, the X447 test was carried out to obtain data on high temperature breach phenomena involving metallic fuel since little data existed in that area

  7. Properties of explosively bonded steel with monel as cladding metal

    International Nuclear Information System (INIS)

    In the heat exchangers using sea water, corrosion-resistant metal is clad on carbon steel structures for corrosion prevention. The explosion bonding established as corrosion-preventive lining method in chemical equipments was investigated to obtain high reliability by easy working in case of the heat exchangers for nuclear power plants by using it in combination with build-up welding. By this method, the area of build-up welding is remarkably reduced, and the finishing of welded surfaces is limited to easily workable parts. The quality control of explosively bonded Monel metal plates is easy. The explosion-clad steels of SF50 and NCuP and SPV 24 and NCuP were tested. The steels explosively bonded with Monel metal showed good workability in actual application, and had sufficient bending strength and corrosion resistance after the working and heat treatment. Supersonic wave flaw detection and other mechanical tests can be carried out in the as-explosively bonded state, and build-up welding can be used in combination. The explosion-clad steels satisfied the relevant standard values after the heat treatment below 650 deg C, and after cold bending, hot working, butt welding, build-up welding, welding of other members, and stress-relieving annealing. (Kako, I.)

  8. Metaliographic Analyses of Laser Cladded WC-Ni and WC-Co Hard-facing Metals

    Institute of Scientific and Technical Information of China (English)

    HKChikwanda; MChiremba; CVanRooyen

    2004-01-01

    Laser cladding is performed to improve the surface properties of metallic machine components. Extensive work is being conducted to investigate the relationships among the cladding parameters, clad powder characteristics and the quality of the clad layer. This work presents some of the metallographic analyses results of WC-Ni and WC-Co clad layers. The clad layers are chayacterised with non-uniform carbide par[icles, mostly WC imbedded in a more ductile matrix. The transition from the clad layer to the subslxate metal had a distinct dilution zone. The ratio of this zone to the clad height was in the range of 10-12% and this still needs robe refined.

  9. Prevention of microcracking by REM addition to alloy 690 filler metal in laser clad welds

    International Nuclear Information System (INIS)

    Effect of REM addition to alloy 690 filler metal on microcracking prevention was verified in laser clad welding. Laser clad welding on alloy 132 weld metal or type 316L stainless steel was conducted using the five different filler metals of alloy 690 varying the La content. Ductility-dip crack occurred in laser clad welding when La-free alloy 690 filler metal was applied. Solidification and liquation cracks occurred contrarily in the laser cladding weld metal when the 0.07mass%La containing filler metal was applied. In case of laser clad welding on alloy 132 weld metal and type 316L stainless steel, the ductility-dip cracking susceptibility decreased, and solidification/liquation cracking susceptibilities increased with increasing the La content in the weld metal. The relation among the microcracking susceptibility, the (P+S) and La contents in every weld pass of the laser clad welding was investigated. Ductility-dip cracks occurred in the compositional range (atomic ratio) of La/(P+S) 0.99(on alloy 132 weld metal), >0.90 (on type 316L stainless steel), while any cracks did not occur at La/(P+S) being between 0.21-0.99 (on alloy 132 weld metal) 0.10-0.90 (on type 316L stainless steel). Laser clad welding test on type 316L stainless steel using alloy 690 filler metal containing the optimum La content verified that any microcracks did not occurred in the laser clad welding metal. (author)

  10. Laser cladding of wear resistant metal matrix composite coatings

    International Nuclear Information System (INIS)

    A number of coatings with wear-resistant properties as well as with a low friction coefficient are produced by laser cladding. The structure of these coatings is determined by required performance and realized as metal matrix composite (MMC), where solid lubricant serves as a ductile matrix (e.g. CuSn), reinforced by appropriate ceramic phase (e.g. WC/Co). One of the engineered coating with functionally graded material (FGM) structure has a dry friction coefficient 0.12. Coatings were produced by coaxial injection of powder blend into the zone of laser beam action. Metallographic and tribological examinations were carried out confirming the advanced performance of engineered coatings

  11. Choice of methods and determination of fracture toughness for anticorrosion cladding metal

    International Nuclear Information System (INIS)

    Technique for fracture toughness determination within wide temperature range is chosen. Experiment results on austenitic anticorrosion cladding metal cracking resistance are given in comparison with temperature dependence low envelope of 15Ch2MFA steel fracture toughness. From the data obtained it follows, that crack propagation direction along cladding metal does not affect KIJ fracture toughness value. It is shown, that fracture toughness values of anticorrosion layer material are higher, than those of low-alloy steel for cladding

  12. On microstructure and flexural strength of metal-ceramic composite cladding developed through microwave heating

    Science.gov (United States)

    Sharma, Apurbba Kumar; Gupta, Dheeraj

    2012-05-01

    A domestic multimode microwave applicator was used to develop carbide reinforced (tungsten-based) metal-matrix composite cladding on austenitic stainless steel substrate. Cladding was developed through microwave irradiation of the preplaced clad materials at 2.45 GHz for 420 s. Clads show metallurgical bonding with substrate by partial dilution of materials. Back scattered images of clad section confirm uniformly distributed reinforced particles in the metallic matrix. Presence of WC, W2C, NiSi, NiW and Co3W3C phases was detected in the clad. Flexural characteristics show two distinct load transitions attributable to deformations of the matrix and the reinforced particles. Clads fail at the upper transition load; further load is taken by the SS-316 substrate. Clads exhibit good stiffness and good adhesion with the substrate. Multi directional cracks were observed at the clad surface; on further loading, cracks get propagated into the clad thickness without getting peeled-off. Mechanism of clad development has been introduced.

  13. Development of eutectic free cladding materials for metallic fuel

    International Nuclear Information System (INIS)

    Historically, it is well known that U base metallic fuel has a lower eutectic temperature with stainless steel cladding. In the phase diagram for the U-Fe binary system, the eutectic temperature is 998K. The eutectic reaction is a limiting factor for raising reactor operation temperature. For the purpose of development of eutectic-free cladding materials, three kinds of diffusion-couple tests with 10 mass%Zr alloy were conducted at a temperature of 1027K for 2250 hrs. We selected the following materials: (a) nitrogen charged zirconium foils, (b) vanadium foils of commercial grade, and (c) nitrogen charged ferritic stainless steel (HT-9). The results showed that typical Zr with layer was observed in all of these materials. Zr with layer appeared to act as a barrier against inter-diffusion of U, Fe. The barrier provided immunity to the eutectic reaction. Discussion was made on C-14 problems in relation to another desirable thermodynamic characteristics of Zr such as carbon-14 immobilization. EPMA analysis indicated relatively high nitrogen concentration at the barrier. The barrier is probably composed of ZrN. (author)

  14. Cladding failure margins for metallic fuel in the integral fast reactor

    International Nuclear Information System (INIS)

    The reference fuel for Integral Fast Reactor (IFR) is a ternary U-Pu-Zr alloy with a low swelling austenitic or ferritic stainless steel cladding. It is known that low melting point eutectics may form in such metallic fuel-cladding systems which could contribute to cladding failure under accident conditions. This paper will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel

  15. Aluminum clad ferritic stainless steel foil for metallic catalytic converter substrate applications

    Energy Technology Data Exchange (ETDEWEB)

    Chang, C.S.; Pandey, A.; Jha, B.

    1996-09-01

    A roll bonding process was developed to produce Al clad ferritic stainless steel foil for the metallic catalytic converter substrate application. Clad foils with different chemistry were produced and their properties were evaluated. Heat treatment conditions for the homogenization of clad foils were identified. This article includes results from oxidation tests and mechanical tests on as-rolled and heat treated clad foil. Results from commercial ingot metallurgy foil were also included for comparison. The oxidation weight gain study indicates that the Al content in the foils is directly related to the usable life of the foil. However, rare earth addition is necessary to improve the oxidation resistance of this material for the high temperature applications by slowing down the weight gain kinetics and thus extend the usable life of foils. The heat treated clad foil also exhibit excellent tensile ductility when compared to the ingot metallurgy foil.

  16. Review of fuel/cladding eutectic formation in metallic SFR fuel pins

    International Nuclear Information System (INIS)

    Sodium-cooled Fast Reactors (SFRs) remain a strong contender amongst the Generation IV reactor concepts. Metallic fuel has been a primary fuel option for SFR designers in the US and was used extensively in the first generation of SFRs. One of the benefits of metallic fuel is its chemical compatibility with the coolant; unfortunately this compatibility does not extend to steel cladding at elevated temperatures. It has been known that uranium, plutonium, and rare earths diffuse with cladding constituents to form a low melting point fuel/cladding eutectic which acts to thin the cladding once the interfacial temperature rises above the system liquidus temperature. Since the 1960's, many experiments have been performed and published to evaluate the rate of fuel/cladding eutectic formation and the temperature above which melting will begin as a function of fuel/cladding interfacial temperature, time at temperature, fuel constituents (uranium, fissium or uranium (plutonium) zirconium), cladding type (stainless steel 316, stainless steel 306, D9 or HT9), beginning of life linear power, plutonium enrichment and burnup. The results of these tests, however, remain scattered across conference and journal papers spanning 50 years. The tests used to collect this data also varied in experimental procedure throughout the years. This paper will consolidate the experimental data into four groups of similar test conditions and expand upon the testing performed for each group in detail. A companion paper in PSA 2011 will discuss predictive correlations formulated from this database. (authors)

  17. Micro structural evaluation of fuel clad chemical interaction for metallic fuels for fast reactor

    International Nuclear Information System (INIS)

    The neutronic performance of metal fuel based on binary U-Pu alloy or ternary U-Pu-Zr alloys are better than conventional uranium plutonium mixed oxide or high density carbide ceramic fuel. The growing energy demand in India needs faster growth of nuclear power and warrants introduction of fast reactors based on metallic fuels because of higher breeding ratio and lower doubling time. Two design concepts have been proposed: one based on sodium bonded ternary alloy fuel of U-Pu-Zr ( 2-10 wt%) in modified T91 cladding material and the other is U-Pu binary alloy mechanically bonded to modified T91 cladding material with 'Zircaloy', as a liner between the fuel alloy and the clad. The Zircaloy liner act as a barrier in reducing the fuel clad chemical interaction. It also helps in transfer of heat from the fuel to the clad. Fuel clad chemical interaction is a serious issue limiting the life of a fuel pin as a result of formation of low temperature eutectic between the fuel and components of the cladding material. The eutectic reaction temperature between T91 and Uranium were estimated by dilatometry, differential thermal analysis and high temperature microscopy. Diffusion couple experiments were also carried out between U/Zr/T91 and U/T91 by isothermal annealing of the couples between 550 deg C to 750 deg C for times up to 1500 hrs. to find out the extent of chemical interaction. These studies were supported by metallographic examination, micro hardness measurement, XRD, SEM/EDAX and EPMA. The eutectic temperature was found to be higher than the estimated fuel clad interface temperature under the reactor operating condition. The paper highlights the results of these studies and attempts to analyze them in the light of performance. The outcome of these studies has been useful to the fuel designer in optimizing the design features and predicting the in-reactor fuel behavior. (author)

  18. Inspection of Fuel Cladding and Metal Gasket in Metallic Dry Cask at Tokai No. 2 Power Station

    International Nuclear Information System (INIS)

    The metallic dry cask storage of spent fuel started in December 2001 at TOKAI No.2 power station. The cask that had served for 7 years was inspected in January 2009. The objective of this inspection is confirmation of fuel cladding and metal gasket integrity. This cask accommodates 8 × 8 zirconium liner type fuel. The gasket applied to this cask consists of aluminum outer lining and Inconel spring. This inspection confirmed that there had been no damage in fuel cladding and metal gasket during the storage for 7 years. (author)

  19. Metallic laser clad coatings : on the pirocessing-microstructure-property relationships

    NARCIS (Netherlands)

    Ocelik, V.; de Oliveira, U.; De Hosson, J. Th. M.; DeHosson, JTM; Brebbia, CA

    2009-01-01

    A thick metallic coating that is resistant against high loading impact, severe wear and corrosion at high temperatures can be produced through the laser clad method. This work introduces the Orientation Imagining Microscopy based on electron backscatter diffraction in a scanning electron microscope

  20. 6.9-kV, 63-KA metal-clad switchgear with vacuum circuit breakers for nuclear power plants

    International Nuclear Information System (INIS)

    The technological advancement for larger capacities of vacuum circuit breakers in recent years has truly been prominent. Based on data of a series of in-house tests made so far and on performance in the field, Toshiba Corporation has developed a metal-clad switchgear with vacuum circuit breakers for nuclear power plants. Metal-clad switchgears for nuclear power plants are strongly required to have high reliability and large capacities. This paper describes results of a series of reliability tests made on the vacuum circuit breaker and resistance to earthquakes of the metal-clad switchgear required for use in a nuclear power plant

  1. Three-Point Bending Test for Metal-Ceramic Hybrid Fuel Cladding Tubes

    International Nuclear Information System (INIS)

    A metal-ceramic hybrid cladding tube is one of these concepts. The tube forms composite ceramic layers on the zirconium alloy cladding tubes to enhance their stability during all accidents, as well as to prevent them from generating hydrogen gas under severe accidents. The new tubes can be made in the following two stages: first, producing zirconium alloy tubes, and second, forming ceramic composites on the tubes. The first stage of the fabrication is the same as a conventional manufacturing process for zirconium fuel cladding tubes. The inner tube was produced from a Zr-alloy ingot by repeating the pilgering and annealing. The second stage covers the fabrication of a SiC ceramic fiber composite. The SiC-fiber filament is wound on the Zr tube. The hetero interface between Zr metal and SiC ceramic can be adjusted with compliant media to minimize the material incoherence. The empty space within the fiber-wound perform is then filled with SiC-based preceramic polymer. Finally, the cladding tube is completed with surface coating

  2. A Temperature Sensor Based on a Symmetrical Metal-Cladding Optical Waveguide

    Institute of Scientific and Technical Information of China (English)

    ZHOU Guo-Rui; FENG Guo-Ying; ZHANG Yi; MA Zi; WANG Jian-Jun

    2012-01-01

    A compact temperature sensor based on a symmetrical metal-cladding optical waveguide using free-space coupling is proposed and demonstrated theoretically and experimentally. The symmetrical Au-cladding optical waveguide is based on a thin LiNbO3 slab sandwiched between two metal films, which serve as the coupling layer and reflecting panel, respectively. The sensitivity of this sensor of 9.08×10-2 deg/℃, 6.6 ×10-2 deg/℃ and 4.8 × 10-2 deg/℃ corresponding to 3238-order, 3237-order and 3236-order modes, respectively, are obtained. Higher resolution is predicted with a larger linear expansion coefficient material and a higher resolution θ/2θ goniometer.%A compact temperature sensor based on a symmetrical metal-cladding optical waveguide using free-space coupling is proposed and demonstrated theoretically and experimentally.The symmetrical Au-cladding optical waveguide is based on a thin LiNbO3 slab sandwiched between two metal films,which serve as the coupling layer and reflecting panel,respectively.The sensitivity of this sensor of 9.08 × 10-2 deg/℃,6.6 × 10-2 deg/℃ and 4.8 × 10-2 deg/℃ corresponding to 3238-order,3237-order and 3236-order modes,respectively,are obtained.Higher resolution is predicted with a larger linear expansion coefficient material and a higher resolution θ/2θ goniometer.

  3. Metal clad active fibres for power scaling and thermal management at kW power levels.

    Science.gov (United States)

    Daniel, Jae M O; Simakov, Nikita; Hemming, Alexander; Clarkson, W Andrew; Haub, John

    2016-08-01

    We present a new approach to high power fibre laser design, consisting of a polymer-free all-glass optical fibre waveguide directly overclad with a high thermal conductivity metal coating. This metal clad active fibre allows a significant reduction in thermal resistance between the active fibre and the laser heat-sink as well as a significant increase in the operating temperature range. In this paper we show the results of a detailed thermal analysis of both polymer and metal coated active fibres under thermal loads typical of kW fibre laser systems. Through several different experiments we present the first demonstration of a cladding pumped aluminium-coated fibre laser and the first demonstration of efficient operation of a cladding-pumped fibre laser at temperatures of greater than 400 °C. Finally, we highlight the versatility of this approach through operation of a passively (radiatively) cooled ytterbium fibre laser head at an output power of 405 W in a compact and ultralight package weighing less than 100 g. PMID:27505822

  4. Interfacial microstructure and properties of copper clad steel produced using friction stir welding versus gas metal arc welding

    Energy Technology Data Exchange (ETDEWEB)

    Shen, Z.; Chen, Y. [Mechanical and Mechatronics Engineering, University of Waterloo, Waterloo (Canada); Haghshenas, M., E-mail: mhaghshe@uwaterloo.ca [Mechanical and Mechatronics Engineering, University of Waterloo, Waterloo (Canada); Nguyen, T. [Mechanical Systems Engineering, Conestoga College, Kitchener (Canada); Galloway, J. [Welding Engineering Technology, Conestoga College, Kitchener (Canada); Gerlich, A.P. [Mechanical and Mechatronics Engineering, University of Waterloo, Waterloo (Canada)

    2015-06-15

    A preliminary study compares the feasibility and microstructures of pure copper claddings produced on a pressure vessel A516 Gr. 70 steel plate, using friction stir welding versus gas metal arc welding. A combination of optical and scanning electron microscopy is used to characterize the grain structures in both the copper cladding and heat affected zone in the steel near the fusion line. The friction stir welding technique produces copper cladding with a grain size of around 25 μm, and no evidence of liquid copper penetration into the steel. The gas metal arc welding of copper cladding exhibits grain sizes over 1 mm, and with surface microcracks as well as penetration of liquid copper up to 50 μm into the steel substrate. Transmission electron microscopy reveals that metallurgical bonding is produced in both processes. Increased diffusion of Mn and Si into the copper cladding occurs when using gas metal arc welding, although some nano-pores were detected in the FSW joint interface. - Highlights: • Cladding of steel with pure copper is possible using either FSW or GMAW. • The FSW yielded a finer grain structure in the copper, with no evidence of cracking. • The FSW joint contains some evidence of nano-pores at the interface of the steel/copper. • Copper cladding by GMAW contained surface cracks attributed to high thermal stresses. • The steel adjacent to the fusion line maintained a hardness value below 248 HV.

  5. Interfacial microstructure and properties of copper clad steel produced using friction stir welding versus gas metal arc welding

    International Nuclear Information System (INIS)

    A preliminary study compares the feasibility and microstructures of pure copper claddings produced on a pressure vessel A516 Gr. 70 steel plate, using friction stir welding versus gas metal arc welding. A combination of optical and scanning electron microscopy is used to characterize the grain structures in both the copper cladding and heat affected zone in the steel near the fusion line. The friction stir welding technique produces copper cladding with a grain size of around 25 μm, and no evidence of liquid copper penetration into the steel. The gas metal arc welding of copper cladding exhibits grain sizes over 1 mm, and with surface microcracks as well as penetration of liquid copper up to 50 μm into the steel substrate. Transmission electron microscopy reveals that metallurgical bonding is produced in both processes. Increased diffusion of Mn and Si into the copper cladding occurs when using gas metal arc welding, although some nano-pores were detected in the FSW joint interface. - Highlights: • Cladding of steel with pure copper is possible using either FSW or GMAW. • The FSW yielded a finer grain structure in the copper, with no evidence of cracking. • The FSW joint contains some evidence of nano-pores at the interface of the steel/copper. • Copper cladding by GMAW contained surface cracks attributed to high thermal stresses. • The steel adjacent to the fusion line maintained a hardness value below 248 HV

  6. Bonding Interface and Bending Deformation of Al/316LSS Clad Metal Prepared by Explosive Welding

    Science.gov (United States)

    Guo, Xunzhong; Fan, Minyu; Wang, Liuan; Ma, Fuye

    2016-06-01

    The morphology, elemental distribution, and phase analysis of the bonding interface were investigated by means of SEM, EDS, and XRD to evaluate the interface bonding properties of Al/316LSS clad metal prepared by explosive welding method. Furthermore, the micro-hardness and bending properties were also investigated. The results indicated that the linear and wavy bonding interfaces coexisted and intermetallic phases were present in the local interfacial zone. Moreover, the micro-hardness value at the bonding interface with intermetallic phases was higher than that at the interface without any intermetallic phases. In addition, bulk metal compounds could easily lead to the generation of micro-cracks during the bending forming process.

  7. Bonding Interface and Bending Deformation of Al/316LSS Clad Metal Prepared by Explosive Welding

    Science.gov (United States)

    Guo, Xunzhong; Fan, Minyu; Wang, Liuan; Ma, Fuye

    2016-04-01

    The morphology, elemental distribution, and phase analysis of the bonding interface were investigated by means of SEM, EDS, and XRD to evaluate the interface bonding properties of Al/316LSS clad metal prepared by explosive welding method. Furthermore, the micro-hardness and bending properties were also investigated. The results indicated that the linear and wavy bonding interfaces coexisted and intermetallic phases were present in the local interfacial zone. Moreover, the micro-hardness value at the bonding interface with intermetallic phases was higher than that at the interface without any intermetallic phases. In addition, bulk metal compounds could easily lead to the generation of micro-cracks during the bending forming process.

  8. Development of a metal-clad advanced composite shear web design concept

    Science.gov (United States)

    Laakso, J. H.

    1974-01-01

    An advanced composite web concept was developed for potential application to the Space Shuttle Orbiter main engine thrust structure. The program consisted of design synthesis, analysis, detail design, element testing, and large scale component testing. A concept was sought that offered significant weight saving by the use of Boron/Epoxy (B/E) reinforced titanium plate structure. The desired concept was one that was practical and that utilized metal to efficiently improve structural reliability. The resulting development of a unique titanium-clad B/E shear web design concept is described. Three large scale components were fabricated and tested to demonstrate the performance of the concept: a titanium-clad plus or minus 45 deg B/E web laminate stiffened with vertical B/E reinforced aluminum stiffeners.

  9. Ceramics reinforced metal base composite coatings produced by CO II laser cladding

    Science.gov (United States)

    Yang, Xichen; Wang, Yu; Yang, Nan

    2008-03-01

    Due to the excellent performance in high strength, anti-temperature and anti-wear, ceramics reinforced metal base composite material was used in some important fields of aircraft, aerospace, automobile and defense. The traditional bulk metal base composite materials are the expensive cost, which is limited in its industrial application. Development of laser coating of ceramics reinforced metal base composite is very interesting in economy. This paper is focused on three laser cladding ceramics coatings of SiC particle /Al matrix , Al IIO 3 powder/ Al matrix and WC + Co/mild steel matrix. Powder particle sizes are of 10-60μm. Chemical contents of aluminum matrix are of 3.8-4.0% Cu, 1.2-1.8% Mg, 0.3-0.99% Mn and balance Al. 5KW CO II laser, 5 axes CNC table, JKF-6 type powder feeder and co-axis feeder nozzle are used in laser cladding. Microstructure and performance of laser composite coatings have been respectively examined with OM,SEM and X-ray diffraction. Its results are as follows : Microstructures of 3C-,6H- and 5H- SiC particles + Al + Al 4SiC 4 + Si in SiC/Al composite, hexagonal α-Al IIO 3 + cubic γ-Al IIO 3 + f.c.c Al in Al IIO 3 powder/ Al composite and original WC particles + separated WC particles + eutectic WC + γ-Co solid solution + W IIC particles in WC + Co/steel coatings are respectively recognized. New microstructures of 5H-SiC in SiC/Al composite, cubic γ-Al IIO 3 in Al IIO 3 composite and W IIC in WC + Co/ steel composite by laser cladding have been respectively observed.

  10. Optical-assembly periodic structure of ferrofluids in a liquid core/metal cladding optical waveguide.

    Science.gov (United States)

    Wang, Xianping; Yin, Cheng; Sun, Jingjing; Han, Qingbang; Li, Honggen; Sang, Minghuang; Yuan, Wen; Cao, Zhuangqi

    2013-11-01

    We present a novel and simple mechanism for the fabrication of periodic microstructure based on a ferrofluids core/metal cladding optical waveguide chip. The ultrahigh-order modes excited in the millimeter scale guiding layer lead to the ordered particle aggregates in ferrofluids without applying a magnetic field. Since the absorption of photons by the extremely dilute ferrofluids is extremely small and the Soret effect is not noticeable, a tentative explanation in terms of the optical trapping effect is proposed. Furthermore, this scheme exhibits all-optically tunable reflectivity and lateral Goos-Hänchen shift, which potentially may be for practical use in novel optical devices. PMID:24216657

  11. Prospects and merits of metal-clad semiconductor lasers from nearly UV to far IR

    CERN Document Server

    Khurgin, Jacob B

    2015-01-01

    Using metal-clad (or plasmonic) waveguide structures in semiconductor lasers carries a promise of reduced size, threshold, and power consumption. This promise is put to a rigorous theoretical test, that takes into account increased waveguide loss, Auger recombination, and Purcell enhancement of spontaneous recombination. The conclusion is that purported benefits of metal waveguides are small to nonexistent for all the band-to-band and intersubband lasers operating from UV to Mid-IR range, with a prominent exception of far-IR and THz quantum cascade lasers. For these devices, however, metal waveguides already represent the state of the art, and the guiding mechanism in them has far more in common with a ubiquitous transmission line than with plasmonics.

  12. End Plug Welding of New FM Cladding Tube for SFR Metallic Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Won; Woo, Yoon Myeng; Kim, Hyung Tae; Kim, Ki Hwan; Kim, Sung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Ferritic/martensitic (FM) steel such as HT9 has been selected for a SFR fuel cladding tube material because of the excellent irradiation characteristics. But the HT9 material is not enough to satisfy the discharge burnup goal due to the high coolant outlet temperature and the low creep resistance. Recently new FM steel is under development to improve HT9 material properties. At KAERI, development of new FM steels, named FC92B and FC92N, has been in progress to improve thermal creep resistance. In this study, the qualification test of the end plug welding of new FM steels, FC92B and FC92N, for SFR metallic fuel rods was carried out based on the developed HT9 welding technology. Through the qualification test, the end plug welding of new FM steels was well verified. Through the qualification test, the weld quality of end plug welding of new FM cladding tubes, FC92B and FC92N, was evaluated according to the qualification test plan and satisfied the requirements on the weld. Consequently, the qualified welding process is ready to produce SFR metallic fuel rods.

  13. Resonant excitation of Rayleigh waves in a narrow fluid channel clad between two metal plates

    Science.gov (United States)

    Nagaraj, Nagaraj; Krokhin, Arkadii; Sánchez-Dehesa, José.; Garcia-Chocano, Victor M.

    2012-02-01

    We study extraordinary absorption of acoustic energy due to resonant excitation of Rayleigh waves in a narrow water channel clad between two unidentical metal plates with Brass plate on one side of the channel and Aluminium plate on the other. The extraordinary absorption is observed at discrete resonant frequencies. From the elastic properties of the metal plates we derive a dispersion equation for coupled Rayleigh waves. Two different types of resonances, corresponding to different polarizations of the coupled waves, are studied for different channel widths and are experimentally confirmed. We also present the experimental confirmation of coupling through measurements of change in transmission minima with channel aperture. Experimental, theoretical, and numerical results are in a good agreement.

  14. Laboratory studies of shear/leach processing of zircaloy clad metallic uranium reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Swanson, J.L.; Bray, L.A.; Kjarmo, H.E.; Ryan, J.L.; Matsuzaki, C.L.; Pitman, S.G.; Haberman, J.H.

    1985-12-01

    The safety aspects addressed centered on understanding and explaining the undesirable reactions, ''fires,'' observed in a few instances during earlier processing of such fuel at the Nuclear Fuels Services (NFS) plant at West Valley, New York. Consideration of the dissolver fires that occurred at NFS leads to the conclusion that they resulted from rapid reactions with uranium metal, rather than with zirconium metal or with sensitized weld beads. The fires observed at NFS during hulls handling operations may have involved sensitized weld beads as suggested by earlier investigators, but current results suggest that these fires also could have been caused by reactions involving uranium metal. Very little pyrophoric activity was observed in leeached cladding hulls, indicating a very low probability for safety problems resulting from the U-Zr intermetallic zone in N-Reactor fuel. Consideration of the potential role of hydrides in the fires observed at NFS indicates that they were also not important factors. Consideration was also given to protective atmospheres to be used during shearing to prevent excessive reaction during that operation. A water deluge during shearing will likely provide adequate safety while meshing well with other process considerations. Studies on the dissolution of metallic uranium in nitric acid show an initial slower reaction followed by a faster reaction that proceeds at a sustained rate for a prolonged period of time. At solution concentrations typical of those encountered in practical uranium dissolver conditions, this sustained rate is governed by an equation such as: Dissolution rate = K (surface area) ((HNO3)+2(U))/sup 2.6/. Little difference was found in dissolution rates of as-fabricated and of irradiated fuel. The transuranic element content of leached cladding hulls was found to be approx. 400 nCi/g. This is too high to allow disposal as low-level waste.

  15. Laboratory studies of shear/leach processing of zircaloy clad metallic uranium reactor fuel

    International Nuclear Information System (INIS)

    The safety aspects addressed centered on understanding and explaining the undesirable reactions, ''fires,'' observed in a few instances during earlier processing of such fuel at the Nuclear Fuels Services (NFS) plant at West Valley, New York. Consideration of the dissolver fires that occurred at NFS leads to the conclusion that they resulted from rapid reactions with uranium metal, rather than with zirconium metal or with sensitized weld beads. The fires observed at NFS during hulls handling operations may have involved sensitized weld beads as suggested by earlier investigators, but current results suggest that these fires also could have been caused by reactions involving uranium metal. Very little pyrophoric activity was observed in leeached cladding hulls, indicating a very low probability for safety problems resulting from the U-Zr intermetallic zone in N-Reactor fuel. Consideration of the potential role of hydrides in the fires observed at NFS indicates that they were also not important factors. Consideration was also given to protective atmospheres to be used during shearing to prevent excessive reaction during that operation. A water deluge during shearing will likely provide adequate safety while meshing well with other process considerations. Studies on the dissolution of metallic uranium in nitric acid show an initial slower reaction followed by a faster reaction that proceeds at a sustained rate for a prolonged period of time. At solution concentrations typical of those encountered in practical uranium dissolver conditions, this sustained rate is governed by an equation such as: Dissolution rate = K (surface area) ([HNO3]+2[U])/sup 2.6/. Little difference was found in dissolution rates of as-fabricated and of irradiated fuel. The transuranic element content of leached cladding hulls was found to be approx. 400 nCi/g. This is too high to allow disposal as low-level waste

  16. Performance of the Barrier between the Metallic Fuel and the Clad Material in Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Metallic fuel has been considered as one of the most probable candidates of the fuel system in the Sodium-cooled Fast Reactor (SFR) in that it has high thermal conductivity, proliferation resistance, and good compatibility between sodium. Addition of the alloying element such as chromium, molybdenum, zirconium and titanium was applied in order to increase the solidus temperature of the uranium-plutonium alloy. Among these, uranium-plutonium alloys with the addition of 10-20% zirconium have been considered in the design of the metallic fuel in SFR. However, actinide elements in metallic fuel like uranium and plutonium react with stainless steel at a temperature above 650 .deg. C to form eutectic compounds. Such eutectic reaction reduces cladding thickness so that mechanical integrity of the cladding gradually decreases as the fuel burnup proceeds. To mitigate such a circumstance, barrier layer, which prevents both fuel and clad elements from diffusing each other, has been developed. Metallic foil made of pure metal has been suggested as a barrier and its feasibility test has been carried out. The objectives in this study are to propose several kinds of the barrier material and to verify its performance under a fuel-clad interaction situation

  17. Terahertz multi-metal-wire hybrid-cladding hollow waveguide for refractive index sensing

    Science.gov (United States)

    Ying-Ying, Yu; Xu-You, Li; Kun-Peng, He; Bo, Sun

    2016-02-01

    We propose a design of terahertz refractive index sensing based on the multi-metal-wire (MMW) hybrid-cladding hollow waveguide. The proposed terahertz hybrid-cladding hollow waveguide comprises one air core in the center surrounding MMW surrounded dielectric. The central air core is used for filling lossless measurands and transmitting terahertz light. In particular, the refractive index sensing is realized by measuring the mode field area (MFA) variation of radially polarized mode. The modal effective refractive index, mode field intensity distribution, and mode field area properties responding to the measurand refractive indexes for different operating frequencies and structure dimensions are investigated, respectively. Simulations show that the proposed terahertz refractive index sensor can realize easily the measurement of the measurand refractive index. Meanwhile, the effects of operating frequency and structure parameters on sensitivity and measurement accuracy are also studied. In view of the trade-off between sensitivity and measurement accuracy, the reasonable choice of the operating frequency and structure parameters can optimize appropriately the sensitivity and measurement accuracy, and the sensitivity can reach approximately 0.585 mm2/RIU (RIU is short for refraction index units) with the proper frequency and structure parameter. Project supported by the National Natural Science Foundation of China (Grant No. 51309059).

  18. Chemical compatibility of uranium based metallic fuels with T91 cladding

    International Nuclear Information System (INIS)

    Highlights: ► Performance of Zr as FCCI barrier layer was evaluated by diffusion experiments. ► Rate constant for reaction at U/Zr interface was 2.07 × 10−8 m s−1/2 at 973 K. ► Rate constant for reaction at Zr/T91 interface was 1.95 × 10−8 m s−1/2 at 973 K. ► Activation energy for reaction at Zr/T91 interface was found to be 54.7 kJ mole−1. ► Interdiffusion between U–6Zr and T91 resulted in formation of three layers. - Abstract: Studies related to development of fast reactor fuels based on ternary U–Pu–Zr and binary U–Pu alloys has been initiated in India for building a data base on thermo-physical and thermodynamic properties, fuel-clad compatibility etc. which are very useful to the fuel-designer to optimize the design feature and to predict the in-reactor fuel behaviour. Fuel-clad chemical compatibility is considered as one of the major concerns for metallic fuels. In the present investigation, the performance of Zr as fuel-clad chemical interaction (FCCI) barrier layer between U and T91 was evaluated by diffusion couple experiments. The growth kinetics of reaction layers at U/Zr and Zr/T91 interfaces were established. The growth kinetics of the reaction zone at both the U/Zr and Zr/T91 interfaces were determined at 973 K from the plot of log (width) versus log (time). The value of reaction index n was found to be around 2 at both the U/Zr and Zr/T91 interfaces. The reaction constant (k) for the growth of reaction layer at the U/Zr interface was determined to be 2.07 × 10−8 m s−1/2 at 973 K. Similarly, the rate constant at the Zr/T91 interface was found to be 1.95 × 10−8 m s−1/2 at 973 K. The activation energy Q for the reaction at the Zr/T91 interface was determined and was found to be 54.7 kJ mole−1. The fuel-clad chemical compatibility between U–6Zr alloy and T91 steel was also investigated in the present study by diffusion couple experiments. The interdiffusion between U–6Zr and T91 at 973 K resulted in the

  19. Guided modes in asymmetric metal-cladding left-handed material waveguides

    Institute of Scientific and Technical Information of China (English)

    Ying He; Xia Zhang; Yanfang Yang; Chunfang Li

    2011-01-01

    @@ We investigate guided modes in the asymmetric waveguide structure with a left-handed material (LHM) layer surrounded by air and metal.A graphical method is proposed to determine the guided modes.New properties of the oscillating and surface guided modes, such as absence of the fundamental mode,coexistence of the oscillating and surface guided modes, fast attenuation of the surface guided modes, and mode degeneracy, are analyzed in detail.%We investigate guided modes in the asymmetric waveguide structure with a left-handed material (LHM) layer surrounded by air and metal. A graphical method is proposed to determine the guided modes. New properties of the oscillating and surface guided modes, such as absence of the fundamental mode, coexistence of the oscillating and surface guided modes, fast attenuation of the surface guided modes, and mode degeneracy, are analyzed in detail. We also investigate dispersive characteristics of the metal-LHMair optical waveguide. The propagation constant increases with decreasing slab thickness for the first-order oscillating mode, which is different from that in traditional metal-cladding waveguides.

  20. All-optical modulator based on a ferrofluid core metal cladding waveguide chip

    International Nuclear Information System (INIS)

    We propose a novel optical intensity modulator based on the combination of a symmetrical metal cladding optical waveguide (SMCW) and ferrofluid, where the ferrofluid is sealed in the waveguide to act as a guiding layer. The light matter interaction in the ferrofluid film leads to the formation of a regular nanoparticle pattern, which changes the phase match condition of the ultrahigh order modes in return. When two lasers are incident on the same spot of the waveguide chip, experiments illustrate all-optical modulation of one laser beam by adjusting the intensity of the other laser. A possible theoretical explanation may be due to the optical trapping and Soret effect since the phenomenon is considerable only when the control laser is effectively coupled into the waveguide. (electromagnetism, optics, acoustics, heat transfer, classical mechanics, and fluid dynamics)

  1. Electrophoretic Deposition for the Fabrication of High-Performance Metal-Ceramic Hybrid Cladding

    International Nuclear Information System (INIS)

    Metal-ceramic hybrid cladding consisting of a Zr liner and SiCf/SiC composite is one of the candidate systems. To achieve a high-performance metal-ceramic hybrid cladding, it is important to synthesize the SiCf/SiC composites with high flexural strength. The most common interphases, such as pyrolytic carbon (PyC) and boron nitride (BN) coating, have been applied on the surface of SiC fibers by chemical vapor deposition (CVD) or chemical vapor infiltration (CVI). In addition, the SiC matrix phase for SiCf/SiC composites has been commonly formed by CVI and polymer infiltration and pyrolysis (PIP), which are very costly and complicated processes. For this reason, the fabrication process of SiCf/SiC composites that is low-cost and simple has been strongly needed. In this study, weak phase coating using a commercial colloidal carbon black suspension was performed on SiC fibers through electrophoretic deposition (EPD), and carbon-coated SiCf/SiC composites were fabricated by EPD. The mechanical properties at room temperature were evaluated to investigate the effect of the carbon interfacial layer on the mechanical properties of carbon-coated SiCf/SiC composites. In this study, it was concluded that the EPD method is effective for homogeneous carbon black coating on SiC fibers, and that the carbon coating layer on SiC fibers plays an important role in optimizing the interface between fibers and the matrix, and enhances the toughness of carbon-coated SiCf/SiC composites during fracture

  2. On the cladding effect on metal resistance to hydrogen embrittlement of nuclear reactor vessel

    International Nuclear Information System (INIS)

    The pearlitic 15 Kh2NMFA steel with austenitic cladding is investigated. The distribution of hardness and local strength properties is investigated. It is shown that the austenitic cladding even if it contains stress concentrators of the crack type considerably increases the resistance to hydrogen embrittlement of vessels of WWER. Therefore, while there are no effective methods of protecting reactor vessels from hydrogenation, the application of cladding is necessary

  3. 6.9-kV, 63-kA metal-clad switchgear with vacuum circuit breakers for nuclear power plants

    International Nuclear Information System (INIS)

    Based on data of a series of in-house tests made so far and on performance in the field, Toshiba Corporation has developed a metal-clad switchgear with vacuum circuit breakers for nuclear power plants. Metal-clad switchgear for nuclear power plants are required to have high reliability and large capacities. This paper describes results of a series of reliability tests made on the vacuum circuit breaker and resistance to earthquakes of the metal-clad switchgear required for use in a nuclear power plant. 4 refs

  4. Chemical compatibility of uranium based metallic fuels with T91 cladding

    Energy Technology Data Exchange (ETDEWEB)

    Kaity, Santu [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Kutty, T.R.G., E-mail: tkutty@barc.gov.in [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Agarwal, Renu [Product Development Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Laik, Arijit [Materials Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Kumar, Arun [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Performance of Zr as FCCI barrier layer was evaluated by diffusion experiments. Black-Right-Pointing-Pointer Rate constant for reaction at U/Zr interface was 2.07 Multiplication-Sign 10{sup -8} m s{sup -1/2} at 973 K. Black-Right-Pointing-Pointer Rate constant for reaction at Zr/T91 interface was 1.95 Multiplication-Sign 10{sup -8} m s{sup -1/2} at 973 K. Black-Right-Pointing-Pointer Activation energy for reaction at Zr/T91 interface was found to be 54.7 kJ mole{sup -1}. Black-Right-Pointing-Pointer Interdiffusion between U-6Zr and T91 resulted in formation of three layers. - Abstract: Studies related to development of fast reactor fuels based on ternary U-Pu-Zr and binary U-Pu alloys has been initiated in India for building a data base on thermo-physical and thermodynamic properties, fuel-clad compatibility etc. which are very useful to the fuel-designer to optimize the design feature and to predict the in-reactor fuel behaviour. Fuel-clad chemical compatibility is considered as one of the major concerns for metallic fuels. In the present investigation, the performance of Zr as fuel-clad chemical interaction (FCCI) barrier layer between U and T91 was evaluated by diffusion couple experiments. The growth kinetics of reaction layers at U/Zr and Zr/T91 interfaces were established. The growth kinetics of the reaction zone at both the U/Zr and Zr/T91 interfaces were determined at 973 K from the plot of log (width) versus log (time). The value of reaction index n was found to be around 2 at both the U/Zr and Zr/T91 interfaces. The reaction constant (k) for the growth of reaction layer at the U/Zr interface was determined to be 2.07 Multiplication-Sign 10{sup -8} m s{sup -1/2} at 973 K. Similarly, the rate constant at the Zr/T91 interface was found to be 1.95 Multiplication-Sign 10{sup -8} m s{sup -1/2} at 973 K. The activation energy Q for the reaction at the Zr/T91 interface was determined and was found to be 54.7 kJ mole{sup -1

  5. Effect of yttrium additions on void swelling in Liquid Metal Fast Breeder Reactor candidate cladding alloys

    International Nuclear Information System (INIS)

    Candidate Liquid Metal Fast Breeder Reactor cladding alloys AL1 (Fe-26% Ni-9% Cr) and AL2 (Fe-35% Ni-12% Cr) without and with the addition of 0.1% yttrium were bombarded by 4 MeV56Fe2+ ions without and with simultaneous bombardment by 0.4 MeV 4He+ ions. These bombardments were conducted at various irradiation temperatures to determine the effect of yttrium on void swelling. The addition of yttrium decreased peak swelling for 4 MeV 56Fe2+ ion bombarded AL1 and AL2 by 28% and 20%, respectively. In all cases where similar sample comparisons were made (i.e., undoped with undoped and doped with doped) and where bombardment conditions were similar (i.e., single with single beam and dual with dual beam), AL1 showed less peak swelling than did AL2. Simultaneously implanting helium during heavy-ion bombardment increased peak swelling in undoped and doped AL1 by factors of 2.3 and 2.6, respectively

  6. Metal-clad waveguide characterization for contact-based light transmission into tissue

    Science.gov (United States)

    Chininis, Jeffrey; Whiteside, Paul; Hunt, Heather K.

    2016-02-01

    As contemporary laser dermatology procedures, like tattoo removal and skin resurfacing, become more popular, the complications of their operation are also becoming more prevalent. Frequent incidences of over-exposure, ocular injury, and excessive thermal damage represent mounting concerns for those seeking such procedures; moreover, each of these problems is a direct consequence of the standard, free-space method of laser transmission predominantly used in clinical settings. Therefore, an alternative method of light transmission is needed to minimize these problems. Here, we demonstrate and characterize an alternative method that uses planar waveguides to deliver light into sample tissue via direct contact. To do this, slab substrates made from glass were clad in layers of titanium and silver, constraining the light within the waveguide along the waveguide's length. By creating active areas on the waveguide surface, the propagating light could then optically tunnel into the tissue sample, when the waveguide was brought into contact with the tissue. SEM and EDS were used to characterize the metal film thickness and deposition rates onto the glass substrates. Laser light from a Q-switched Nd:YAG source operating at 532nm was coupled into the waveguide and transmitted into samples of pig skin. The amount of light transmitted was measured using photoacoustics techniques, in conjunction with a photodiode and integrating sphere. Transmitting light into tissue in this manner effectively resolves or circumvents the complications caused by free-space propagation methods as it reduces the operating distance to 0, which prevents hazardous back-reflections and allows for the ready incorporation of contact cooling technologies.

  7. Plasmonic gap-mode nanocavities with metallic mirrors in high-index cladding.

    Science.gov (United States)

    Cheng, Pi-Ju; Weng, Chen-Ya; Chang, Shu-Wei; Lin, Tzy-Rong; Tien, Chung-Hao

    2013-06-01

    We theoretically analyze plasmonic gap-mode nanocavities covered by a thick cladding layer at telecommunication wavelengths. In the presence of high-index cladding materials such as semiconductors, the first-order hybrid gap mode becomes more promising for lasing than the fundamental one. Still, the significant mirror loss remains the main challenge to lasing. Using silver coatings within a decent thickness range at two end facets, we show that the reflectivity is substantially enhanced above 95 %. At a coating thickness of 50 nm and cavity length of 1.51 μm, the quality factor is about 150, and the threshold gain is lower than 1500 cm(-1). PMID:23736601

  8. Structural cladding /clad structures

    DEFF Research Database (Denmark)

    Beim, Anne

    2012-01-01

    Structural Cladding /Clad Structures: Studies in Tectonic Building Practice A. Beim CINARK – Centre for Industrialized Architecture, Institute of Architectural Technology, The Royal Danish Academy of Fine Arts School of Architecture, Copenhagen, Denmark ABSTRACT: With point of departure...... of materials, the structural features and the construction details of building systems in selected architectural works. With a particular focus at heavy constructions made of solid wood and masonry, and light weight constructions made of wooden frame structures and steel profiles, it is the intention...... tightness in constructions. At the same time a need for longevity and effortless maintenance have lead to contemporary architectural structures, where the exterior walls and the building envelope most often are made of several layers of advanced materials and separate building elements. In most contemporary...

  9. Friction surface cladding: An exploratory study of a new solid state cladding process

    NARCIS (Netherlands)

    Liu, S.J.; Bor, T.C.; Stelt, van der A.A.; Geijselaers, H.J.M.; Kwakernaak, C.; Kooijman, A.M.; Mol, J.M.C.; Akkerman, R.; Boogaard, van den A.H.

    2015-01-01

    Friction surface cladding is a newly developed solid state cladding process to manufacture thin metallic layers on a substrate. In this study the influence of process conditions on the clad layer appearance and the mechanical properties of both the clad layer and the substrate were investigated. Thi

  10. Efficient phase-matched third harmonic generation in a metal-clad plasmonic double-slot waveguide

    International Nuclear Information System (INIS)

    We propose a metal-clad plasmonic double-slot waveguide with DDMEBT integrated into the slot region as the interactive material for third harmonic generation (THG) from the mid-IR (3600 nm) to the near-IR (1200 nm) region. Typically, an efficient THG process in a waveguide platform relies on three key aspects: high third-order nonlinear susceptibility of the interactive material, fulfillment of the phase-matching condition (PMC), and large pump-harmonic modal overlap. Although it has been theoretically predicted to be possible, designing waveguides to achieve the three key aspects simultaneously is still a major challenge. In the metal-clad plasmonic double-slot waveguide, the PMC between the zeroth mode at the fundamental wave (FW) and the second mode at the third harmonic (TH) is achieved. Taking advantage of the channel plasmon polariton (CPP), the electric fields at both FW and TH are tightly confined in the slot region. The specific slot waveguide structure is exploited to significantly enhance the pump-harmonic modal overlap by greatly reducing the counteractive electric field portion of the second mode at TH. According to our simulation, THG conversion efficiency reaches 1.4732 × 10−5 with 3 W pump power at a waveguide length of 12.3 μm. This THG efficiency is greatly enhanced because of the high third-order nonlinear optical susceptibility of the DDMEBT, the specific plasmonic slot waveguide structure, and the CPP nature of the guided modes—it is more than two times that obtained by simply considering a single slot under the same slot-width condition. (paper)

  11. Environmental Coating Layer for the Metal-Ceramic Hybrid Fuel Cladding Tubes

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Yangil; Kim, Sunhan; Park, Dongjun; Kim, Hyungil; Park, Jeongyong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The dissolution is much faster as the composition is off the stoichiometry. Since the SiC pre ceramic polymer impregnation results in excess oxygen or carbon in the composite matrix, the outer environmental barrier is indispensable to prevent the expected corrosion. Several candidates were considered as a surface coating material. Oxides such as ZrO{sub 2}, ZrO{sub 2}.Y{sub 2}O{sub 3}, SiO{sub 2}, Cr{sub 2}O{sub 3}, and Ta{sub 2}O{sub 5} are very stable in water. Also, the safety of fuel cladding against off-normal accidents could be increased, since the oxides can endure very high temperature. ZrO{sub 2} in a nuclear environment has been studied a lot since it is naturally formed on conventional fuel cladding tubes. In the case of SiO{sub 2}, the transformation of the crystal phase during irradiation was reported; however, their effect on the corrosion resistance was not investigated. Cr{sub 2}O{sub 3} and Ta{sub 2}O{sub 5} are quite good as a protective anti-corrosion layer, but inappropriate in terms of the neutron cross-section. A sol-gel based coating method and a synthesis of the precursor materials are being developed for the surface coating. Surface coating of of ZrO{sub 2} and ZrO{sub 2}.Y{sub 2}O{sub 3} on SiC was performed using a sol-gel dip coating method. Although the coated layer was impractical as a barrier of SiC dissolution, the preliminary test of coating materials can suggest the direction to further research.

  12. Slow light with symmetric gap plasmon guides with finite width metal claddings

    Indian Academy of Sciences (India)

    S Dutta Gupta

    2009-02-01

    We study the dispersion relation and the modes of a symmetric gap plasmon guide, where a dielectric planar slab is coated with finite metallic layers on both top and bottom. The finite conductivity of the metal is taken into account. The modes of the structure exhibit significant differences from those of dielectric waveguides with air or metal as the bounding media. Avoided level crossing phenomenon between the plasmon and the guided modes is shown to exist, leading to leaky modes. The structure sandwiched between two high index media is shown to lead to slow light in transmission. The group delay is shown to be larger for higher order modes.

  13. The analysis of irradiated nuclear fuel and cladding materials, determination of carbon, hydrogen and oxygen/metal ratio

    International Nuclear Information System (INIS)

    Equipment has been developed for the determination of carbon, hydrogen and oxygen/metal ratio on irradiated fuels, of carbon in stainless steel cladding materials and in graphite rich deposits, and of hydrogen in zircaloy. Carbon is determined by combustion to carbon dioxide which is collected and measured manometrically, hydrogen by vacuum extraction followed by diffusion through a palladium thimble, and oxygen/metal ratio by CO/CO2 equilibration. A single set of equipment was devised in order to minimise the time and work involved in changing to a different set of equipment in a separate box, for each type of analysis. For each kind of analysis, alterations to the apparatus are involved but these can be carried out with the basic set in position in a shielded cell, although to do so it is necessary to obtain access via the gloves on the fibre-glass inner glove box. This requires a removal of samples emitting radiation, by transfer to an adjoining cell. A single vacuum system is employed. This is connected through a plug in the lead wall of the shielded cell, and couplings in the glove box wall to the appropriate furnaces. Carbon may be determined, in stainless steel containing 400 to 800 ppm C, with a coefficient of variation of +- 2%. On deposits containing carbon, the coefficient of variation is better than +- 1% for 2 to 30 mg of carbon. Hydrogen, at levels between 30 and 200 ppm in titanium can be determined with a coefficient of variation of better than +- 5%. Titanium has been used in lieu of zircaloy since standardised zircaloy specimens are not available. The precision for oxygen/metal ratio is estimated to be +- 0.001 Atoms oxygen. Sample weights of 200 mg are adequate for most analyses. (author)

  14. Metal-clad switchgear with large capacity vacuum circuit breaker in two-tier arrangement for nuclear power plants

    International Nuclear Information System (INIS)

    Accompanying the increase of main machinery capacity in nuclear power stations, the short-circuit capacity for 6.9 kV in-house auxiliary machinery circuit has increased, and a 63 kA circuit breaker has become necessary. Although magnetic breakers have been used as large capacity breakers so far, vacuum breakers which are more suitable for the recent environmental conditions of power stations have become employed. Hitachi Ltd. has developed the metal-clad switchboard with vacuum breakers of 7.2 kV, 1,200 to 3,000 A, and breaking current of 63 kA in two-tier arrangement. The main features of this breaker are small size, light weight, long life, labour-saving in maintenance and inspection, simple construction, easy handling, high reliability and safety. In addition, in this paper, the construction of the breaker and switchboard, aseismic property, and test results are described. The tests include the withstand voltage test, elevated temperature test, short period current test, short-circuit test, low current breaking test, continuous on-off test, on-off surge combination test and short-circuit breaking test under the condition of vacuum failure in one phase. The aseismic property is guaranteed by analyzing the vibration characteristics and the strength using computer-aided finite element method so that the performance required is satisfied. (Wakatsuki, Y.)

  15. Fiber-to-Waveguide and 3D Chip-to-Chip Light Coupling Based on Bent Metal-Clad Waveguides

    CERN Document Server

    Lu, Zhaolin; Shi, Kaifeng

    2016-01-01

    Efficient fiber-to-waveguide light coupling has been a key issue in integrated photonics for many years. The main challenge lies in the huge mode mismatch between an optical fiber and a single mode waveguide. Herein, we present a novel fiber-to-waveguide coupler, named "L-coupler", through which the light fed from the top of a chip can bend 90{\\deg} with low reflection and is then efficiently coupled into an on-chip Si waveguide within a short propagation distance (<20{\\mu}m). The key element is a bent metal-clad waveguide with a big matched input port. According to our finite-difference time-domain (FDTD) simulation, the coupling efficiency is over 80% within a broad range of working wavelengths in the near-infrared regime for a transverse electric input Gaussian wave. The coupler is polarization-dependent, with very low coupling efficiency (6%-9%) for transverse magnetic waves. The coupler can also be used for three-dimensional (3D) chip-to-chip optical interconnection by efficiently coupling light into ...

  16. 金属导爆索的爆炸水声特性%Underwater sound characteristics of metal-clad detonating cords

    Institute of Scientific and Technical Information of China (English)

    贾虎; 沈兆武

    2011-01-01

    为了研究金属导爆索的水声特性,进行了水下爆炸压力测试和气泡脉动实验,获得了金属导爆索水下爆炸冲击波传播和衰减特性及气泡脉动特性.研究了金属导爆索水下爆炸的声压级、声持续时间、混响效应和功率谱特性,结果表明:(1)金属导爆索水下爆炸声压级较高,具有很强的声功率;(2)金属导爆索水下爆炸后产生的气泡脉动和随后产生的大量小气泡持续时间长,能够显著提高水下爆炸的混响效应和持续时间;(3)金属导爆索水下爆炸声的频率范围非常广,在各频率范围内都有很强的能量,尤以低频段能量最高.可见,金属导爆索具有声压级高、频率范围广、混响效应强和声持续时间长的特点.%Underwater explosion pressure test and gas bubble pulsation experiment were conducted to obtain the transmission and attenuation properties as well as the gas bubble pulsation properties for the underwater explosion shockwaves of the metal-clad detonating cords. And some underwater explosion performances for the metal-clad detonating cords were explored such as the sound pressure level, sound duration, reverberation effect, and power spectrum. Results show that the metal-clad detonating cords display a high level of sound pressure and sound power; that the gas bubble pulsations following the underwater explosion of the metal-clad detonating cords and the ensuing huge amount of small bubbles will persist for a long time, which can observably increase the reverberation effect and sound duration of the underwater explosion; and that the underwater explosion sound of the metal-clad detonating cords covers a wide range of frequencies, the energy in each of which is powerful, and the energy in the low frequency range is the highest.

  17. Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

    Science.gov (United States)

    Carmack, W. J.; Chichester, H. M.; Porter, D. L.; Wootan, D. W.

    2016-05-01

    The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The MFF fuel operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in EBR-II experiments. Data from the MFF-3 and MFF-5 assemblies are most comparable to the data obtained from the EBR-II X447 experiment. The two X447 pin breaches were strongly influenced by fuel/cladding chemical interaction (FCCI) at the top of the fuel column. Post irradiation examination data from MFF-3 and MFF-5 are presented and compared to historical EBR-II data.

  18. Nuclear fuel element cladding

    International Nuclear Information System (INIS)

    Composite cladding for a nuclear fuel element containing fuel pellets is formed with a zirconium metal barrier layer bonded to the inside surface of a zirconium alloy tube. The composite tube is sized by a cold working tube reduction process and is heat treated after final reduction to provide complete recrystallization of the zirconium metal barrier layer and a fine-grained microstructure. The zirconium alloy tube is stress-relieved but is not fully recrystallized. The crystallographic structure of the zirconium metal barrier layer may be improved by compressive deformation such as shot-peening. (author)

  19. Cladding properties changes during operation

    International Nuclear Information System (INIS)

    Austenitic cladding was originally designed as a protection of ferritic/bainitic base materials of reactor pressure vessels against corrosion. Nevertheless, its existence must be taken into account into reactor pressure vessel integrity evaluation from several reasons: cladding has different thermal properties with respect to base metal which affect temperature fields in a vessel; cladding has different mechanical and thermal-mechanical properties comparing with base metal which affect stress field in a vessel; austenitic cladding has different fracture mechanics properties that base metal, but they are also changing during operation due to radiation damage. Austenitic cladding from WWER-440 reactor pressure vessels has been studied within an extended surveillance programme and some interesting results have been obtained. Austenitic cladding made from Nb-stabilized 18/10 type is characterized by some δ-ferrite content in its initial state which results in slight transition behaviour of fracture properties. These properties are changing after irradiation - fracture toughness is decreasing as well as tensile properties are increasing. This second trend was also supported by measurements realized during in-service inspections of inner vessel wall using instrumented indentation testing method. Knowledge of austenitic properties, mainly of its fracture mechanics parameters, is also necessary for a proper evaluation of reactor pressure vessel behaviour during PTS regimes. (author)

  20. Multilayer cladding with hyperbolic dispersion for plasmonic waveguides

    DEFF Research Database (Denmark)

    Babicheva, Viktoriia; Shalaginov, Mikhail Y.; Ishii, Satoshi; Boltasseva, Alexandra; Kildishev, Alexander V.

    2015-01-01

    We study the properties of plasmonic waveguides with a dielectric core and multilayer metal-dielectric claddings that possess hyperbolic dispersion. The waveguides hyperbolic multilayer claddings show better performance in comparison to conventional plasmonic waveguides. © OSA 2015....

  1. Dipole controlled metal gate with hybrid low resistivity cladding for gate-last CMOS with low Vt

    KAUST Repository

    Hinkle, Christopher L.

    2010-06-01

    In this contribution, NMOS and PMOS band edge effective work function (EWF) and correspondingly low Vt are demonstrated using standard fab materials and processes in a gate-last scheme. For NMOS, the use of an Al cladding layer results in Vt = 0.08 V consistent with NMOS EWF = 4.15 eV. Migration of the Al cladding into the TiN and a relatively low oxygen concentration near the TiN/HfO2 interface are responsible for the low EWF. For PMOS, employing a W cladding layer along with a post-TiN anneal in an oxidizing ambient results in elevated oxygen concentration near the TiN/HfO2 interface and Vt = -0.20 V consistent with a PMOS EWF = 5.05 eV. First-principles calculations indicate N atoms displaced from the TiN during the oxidizing anneal form dipoles at the TiN/HfO2 interface that play a critical role in determining the PMOS EWF. © 2010 IEEE.

  2. Nuclear-powered pacemaker fuel cladding study

    International Nuclear Information System (INIS)

    The fabrication of fuel capsules with refractory metal and alloy clads used in nuclear-powered cardiac pacemakers precludes the expedient dissolution of the clad in inorganic acid solutions. An experiment to measure penetration rates of acids on commonly used fuel pellet clads indicated that it is not impossible, but that it would be very difficult to dissolve the multiple cladding. This work was performed because of a suggestion that a 238PuO2-powered pacemaker could be transformed into a terrorism weapon

  3. Active Metal Brazing and Characterization of Brazed Joints in C-C and C-SiC Composites to Copper-Clad-Molybdenum System

    Science.gov (United States)

    Singh, M.; Asthana, R.

    2008-01-01

    Carbon/carbon composites with CVI and resin-derived matrices, and C/SiC composites reinforced with T-300 carbon fibers in a CVI SiC matrix were joined to Cu-clad Mo using two Ag-Cu braze alloys, Cusil-ABA (1.75% Ti) and Ticusil (4.5% Ti). The brazed joints revealed good interfacial bonding, preferential precipitation of Ti at the composite/braze interface, and a tendency toward delamination in resin-derived C/C composite. Extensive braze penetration of the inter-fiber channels in the CVI C/C composites was observed. The Knoop microhardness (HK) distribution across the C/C joints indicated sharp gradients at the interface, and a higher hardness in Ticusil than in Cusil-ABA. For the C/SiC composite to Cu-clad-Mo joints, the effect of composite surface preparation revealed that ground samples did not crack whereas unground samples cracked. Calculated strain energy in brazed joints in both systems is comparable to the strain energy in a number of other ceramic/metal systems. Theoretical predictions of the effective thermal resistance suggest that such joined systems may be promising for thermal management applications.

  4. Clad-coolant chemical interaction

    International Nuclear Information System (INIS)

    This paper provides an overview of the kinetics for zircaloy clad oxidation behaviour in steam and air during reactor accident conditions. The generation of chemical heat from metal/water reaction is considered. Low-temperature oxidation of zircaloy due to water-side corrosion is further described. (authors)

  5. Fuel-cladding chemical interaction

    International Nuclear Information System (INIS)

    The chemistry of the nuclear fuel is very complex. Its chemical composition changes with time due to the formation of fission products and depends on the temperature level history within the fuel pellet and the clad during operation. Firstly, in thermal reactors, zircaloy oxidation from reaction with UO2 fuel under high-temperature conditions will be addressed. Then other fuel-cladding interaction phenomena occurring in fast reactors will be described. Large thermal gradients existing between the centre and the periphery of the pellet induce the radial redistribution of the fuel constituents. The fuel pellet can react with the clad by different corrosion processes which can involve actinide and/or fission product transport via gas, liquid or/and solid phases. All these phenomena are briefly described in the case of different kinds of fuels (oxide, carbide, nitride, metallic) to be used in fast reactors. The way these phenomena are taken into account in fuel performance codes is presented. (authors)

  6. Stone cladding engineering

    CERN Document Server

    Sousa Camposinhos, Rui de

    2014-01-01

    This volume presents new methodologies for the design of dimension stone based on the concepts of structural design while preserving the excellence of stonemasonry practice in façade engineering. Straightforward formulae are provided for computing action on cladding, with special emphasis on the effect of seismic forces, including an extensive general methodology applied to non-structural elements. Based on the Load and Resistance Factor Design Format (LRDF), minimum slab thickness formulae are presented that take into consideration stress concentrations analysis based on the Finite Element Method (FEM) for the most commonly used modern anchorage systems. Calculation examples allow designers to solve several anchorage engineering problems in a detailed and objective manner, underlining the key parameters. The design of the anchorage metal parts, either in stainless steel or aluminum, is also presented.

  7. CAFE experiments on the flow and freezing of metal fuel and cladding melts (1) -Test conditions and overview of the results-

    International Nuclear Information System (INIS)

    In order to assess the progression and consequences of a core melt accident in a metal fueled fast reactor, knowledge of the flow and freezing behavior of the molten core material as it moves axially through and beyond the core is required. This information is important for predicting the fuel relocation process after cladding failure and post-accident heat removal requirements for achieving in-vessel retention. The Core Alloy Flow and Erosion (CAFE) experiments were performed in order to investigate the fundamental flow, metallurgical interaction, and freezing behavior of uranium-iron-type melts within iron-based trough-shaped flow channels and provide information that can support the initial development of mathematical models for such behavior. Four tests were conducted using molten uranium and two tests were performed using U-Fe eutectic mixture. In each test, 1 to 1.65 kg of melt was introduced into an inclined stainless steel trough. These test results provide understandings on fundamental flow and freezing behavior of melts including metallurgical interaction in the steel flow channels with a variety of melt and flow channel conditions and also offer useful information for developing analytical models to describe such behavior. (author)

  8. Development of high performance cladding

    International Nuclear Information System (INIS)

    The developments of superior next-generation light water reactor are requested on the basis of general view points, such as improvement of safety, economics, reduction of radiation waste and effective utilization of plutonium, until 2030 year in which conventional reactor plants should be renovate. Improvements of stainless steel cladding for conventional high burn-up reactor to more than 100 GWd/t, developments of manufacturing technology for reduced moderation-light water reactor (RMWR) of breeding ratio beyond 1.0 and researches of water-materials interaction on super critical pressure-water cooled reactor are carried out in Japan Atomic Energy Research Institute. Stable austenite stainless steel has been selected for fuel element cladding of advanced boiling water reactor (ABWR). The austenite stain less has the superiority for anti-irradiation properties, corrosion resistance and mechanical strength. A hard spectrum of neutron energy up above 0.1 MeV takes place in core of the reduced moderation-light water reactor, as liquid metal-fast breeding reactor (LMFBR). High performance cladding for the RMWR fuel elements is required to get anti-irradiation properties, corrosion resistance and mechanical strength also. Slow strain rate test (SSRT) of SUS 304 and SUS 316 are carried out for studying stress corrosion cracking (SCC). Irradiation tests in LMFBR are intended to obtain irradiation data for damaged quantity of the cladding materials. (M. Suetake)

  9. Metal-clad superconducting tapes with high critical current density in the Y-Ba-Cu-O and Bi(Pb)-Sr-Ca-Cu-O systems

    International Nuclear Information System (INIS)

    Superconducting powders representing the Y-Ba-Cy-O (YBCO) and Bi(Pb)-Sr-Ca-Cu-O (BPSCCO) systems were used either alone or combining them with different amounts of silver powder to fabricate metal-clad superconducting tapes by the powder-in-tube method. The metallic tubes used were nickel for the YBCO system and silver for the BPSCCO system. The initial tapes were produced by cold-rolling the tubes to various thicknesses and then processed following rigorous regimens of sintering, annealing and oxygenation schedules. The YBCO tapes were generally processed by a gradientless melt-texturizing process in an oxygen atmosphere, and the BPSCCO tapes were processed by a combination of rolling and sintering operations in air and in low oxygen partial pressure. This paper reports on the effects of various compositional and processing variables on the superconductive properties of the tapes measured by the DC magnetization, AC susceptibility and direct transport current measurements. Transport critical currents greater than 600 A at 4.2K were routinely carried by the YBCO tapes. Also, for the YBCO system, the effects of silver addition to control oxygenation and nickel poisoning of the superconducting phase were studied in detail with the help of magnetization measurements. Addition of silver up to 5 wt% was found to be beneficial for critical current density in YBCO tapes. However, direct contact of the superconductor with oxygen was essential for optimum current density. On the other hand, homogenization of the precursor powder by repeated calcination and the tapes of the lowest thickness offered the most improved critical current density values of the BPSCCO system

  10. Reactor physics assessment of alternate cladding materials

    International Nuclear Information System (INIS)

    A preliminary reactor physics assessment has been performed for candidate alternate cladding materials to replace zirconium alloys in enhanced accident tolerant fuel (ATF) concepts for light water reactors. Proposed ATF concepts seek to reduce severe accident risks by increasing the coping time available to operators for accident response and reducing the extent and rate of heat and hydrogen production from steam oxidation. Candidate materials in this neutronics-focused study included austenitic stainless steel 310SS, alumina-forming ferritic alloys (FeCrAl), and silicon carbide (SiC). Historic 304SS cladding and Zircaloy were considered as reference points. Initial results indicate that the metallic options require increased uranium enrichments and/or decreased cladding thicknesses to match the operating cycle lengths achieved with Zircaloy; FeCrAl offered the smallest reactivity penalty, whereas 310SS showed large negative impacts. Ceramic SiC cladding performed well if cladding thicknesses remained similar to those for Zircaloy, but large clad thickness increases led to negative impacts. Fuel pellet relative radial power distributions were similar for all clad materials analyzed. Finally, an economic assessment found that 310SS or FeCrAl could increase fuel pellet costs by 15–36%, while SiC fuel pellet costs were very similar to Zircaloy. (author)

  11. Efecto de los metales sobre microcrustáceos de agua dulce. Avances metodológicos y potencialidad de cladóceros y copépodos como organismos test

    OpenAIRE

    María Florencia Gutierrez; Ana María Gagneten

    2012-01-01

    El incremento de los metales en los cuerpos de agua dulce a causa de las actividades antropogénicas genera importantes alteraciones sobre la biota. Esta revisión analiza los efectos adversos de varios metales de relevancia ecotoxicológica sobre los microcrustáceos zooplanctónicos (cladóceros y copépodos), los avances experimentales en esta línea y las ventajas de cada grupo como organismos test. En general, la necesidad de obtener indicadores más sensibles y representativos que los tradiciona...

  12. Cladding Effects on Structural Integrity of Nuclear Components

    International Nuclear Information System (INIS)

    Based on this study, the following conclusions and recommendations can be made: Due to significant differences in the thermal and mechanical properties between the austenitic cladding and the ferritic base metal, residual stresses are induced in the cladding and the underlying base metal. These stresses are left in clad components even after Post-Weld Heat Treatment (PWHT). The different restraint conditions of the clad component have a minor influence on the magnitude of the cladding residual stresses in the cladding layer. The thickness of the clad object is the main impacting geometrical dimension in developing cladding residual stresses. A clad object having a base material thickness exceeding 10 times the cladding thickness would be practically sufficient to introduce cladding residual stresses of a thick reactor pressure vessel. For a clad component that received PWHT, the peak tensile stress is in the cladding layer, and the residual stresses in the underlying base material are negligible. However, for clad components not receiving PWHT, for instance the repair welding of the cladding, the cladding residual stresses of tensile type exist even in the base material. This implies a higher risk for underclad cracking for clad repairs that received no PWHT. For certain clad geometries, like nozzles, the profile of the cladding residual stresses depends on the clad thickness and position, and significant tensile stresses can also exist in the base material. Based on different measurements reported in the literature, a value of 150 GPa can be used as Young's Modulus of the austenitic cladding material at room temperature. The control measurements of small samples from the irradiated reactor pressure vessel head did not reveal a significant difference of Young's Modulus between the irradiated and the unirradiated cladding material condition. No significant differences between the axial and tangential cladding residual stresses are reported in the measurement of

  13. Cladding Effects on Structural Integrity of Nuclear Components

    Energy Technology Data Exchange (ETDEWEB)

    Sattari-Far, Iradi; Andersson, Magnus [lnspecta Technology AB, Stockholm (Sweden)

    2006-06-15

    Based on this study, the following conclusions and recommendations can be made: Due to significant differences in the thermal and mechanical properties between the austenitic cladding and the ferritic base metal, residual stresses are induced in the cladding and the underlying base metal. These stresses are left in clad components even after Post-Weld Heat Treatment (PWHT). The different restraint conditions of the clad component have a minor influence on the magnitude of the cladding residual stresses in the cladding layer. The thickness of the clad object is the main impacting geometrical dimension in developing cladding residual stresses. A clad object having a base material thickness exceeding 10 times the cladding thickness would be practically sufficient to introduce cladding residual stresses of a thick reactor pressure vessel. For a clad component that received PWHT, the peak tensile stress is in the cladding layer, and the residual stresses in the underlying base material are negligible. However, for clad components not receiving PWHT, for instance the repair welding of the cladding, the cladding residual stresses of tensile type exist even in the base material. This implies a higher risk for underclad cracking for clad repairs that received no PWHT. For certain clad geometries, like nozzles, the profile of the cladding residual stresses depends on the clad thickness and position, and significant tensile stresses can also exist in the base material. Based on different measurements reported in the literature, a value of 150 GPa can be used as Young's Modulus of the austenitic cladding material at room temperature. The control measurements of small samples from the irradiated reactor pressure vessel head did not reveal a significant difference of Young's Modulus between the irradiated and the unirradiated cladding material condition. No significant differences between the axial and tangential cladding residual stresses are reported in the

  14. Analysis of the behaviour of under-clad and surface cracks in cladded components

    International Nuclear Information System (INIS)

    The issue of the contribution is the characterization of under-clad and surface crack behaviour in ferritic steel components with an austenitic welded cladding. The experimental investigations were performed using large-scale samples. The residual stress field was determined in detail by a numerical simulation of the welding and heat treatment processes. These results were used for the numerical simulation of crack initiation and crack arrest. In all evaluated cases the crack was initiated in the ferritic material, while the cladding stayed intact even in case of a crack jump in the base metal. In the frame of case studies the results were transferred to application relevant geometries

  15. Spatial Mode Selective Waveguide with Hyperbolic Cladding

    CERN Document Server

    Tang, Y; Xu, M; Bäumer, S; Adam, A J L; Urbach, H P

    2016-01-01

    Hyperbolic Meta-Materials~(HMMs) are anisotropic materials with permittivity tensor that has both positive and negative eigenvalues. Here we report that by using a type II HMM as cladding material, a waveguide which only supports higher order modes can be achieved, while the lower order modes become leaky and are absorbed in the HMM cladding. This counter intuitive property can lead to novel application in optical communication and photonic integrated circuit. The loss in our HMM-Insulator-HMM~(HIH) waveguide is smaller than that of similar guided mode in a Metal-Insulator-Metal~(MIM) waveguide.

  16. Studies on fuel-clad chemical interaction of U-10Zr alloy with T91 cladding

    International Nuclear Information System (INIS)

    Fuel-clad chemical compatibility has been recognized as one of the major concerns about the performance of metallic fuel since it limits the life of the fuel pin due to formation of low melting eutectic. The fuel-clad compatibility between U-10Zr and T91 was studied by diffusion couple experiments at normal operating and transient conditions. The diffusion reaction between these two was strongly retarded due to formation of a Zr-rich layer at the interface. (author)

  17. Cobalt-based superalloy layers deposited on X38CrMoV5 steel base metal by explosion cladding process

    OpenAIRE

    LANGLOIS, Laurent; Bigot, Régis; ETTAQI, Saïd

    2008-01-01

    International audience A grade 25 cobalt-based superalloy in the form of a sheet 5 mm in thickness and a steel substrate of type X38CrMoV5 are joined by explosion cladding. The macrostructure and microstructure of the interface and of the co-based superalloy layers are studied. The interface presents the form of wavelets with a period of 1000 µm and an amplitude of 250 µm. The superalloy grains are deformed during the cladding process with several slip systems appearing. Near to the interf...

  18. Development and characterisations of WC–12Co microwave clad

    Energy Technology Data Exchange (ETDEWEB)

    Zafar, Sunny, E-mail: sunny.zafar9@gmail.com; Sharma, Apurbba Kumar, E-mail: akshafme@gmail.com

    2014-10-15

    In the present work, WC–12Co based cermet clad was developed on AISI 304 stainless steel using microwave hybrid heating technique. The experimental trials were carried out in a 1.4 kW industrial multimode microwave applicator. The paper explains the major events occurring during microwave irradiation and formation of clad. The developed clads were subsequently characterised through field emission scanning electron microscopy equipped with energy dispersive X-ray spectroscopy, X-ray diffraction, assessment of porosity and microhardness. The WC–12Co clads developed with an approximate thickness of 1 mm, illustrated excellent metallurgical bonding with substrate. The microstructure of the WC–12Co clad mainly consists of skeleton structured carbides embedded in tough metallic phase. The phase analysis of the developed clads indicate the presence of various stable and complex carbides like Co{sub 6}W{sub 6}C, Co{sub 3}W{sub 3}C and Fe{sub 6}W{sub 6}C. The uniform distribution of such carbides with skeleton-like morphology in the microstructure is indicative of high hardness of the clad. The developed clads were free from visible interfacial cracking and the clad porosity was found in the order of approximately 0.98%. The average microhardness of the WC–12Co microwave clads was observed to be 1135 ± 88 HV. - Highlights: • Microwave cladding of WC–12Co on AISI 304 stainless steel is carried out. • Skeleton-like structures of W–Co based carbides are embedded in metallic matrix. • Clad–substrate interface is free from un-melted and un-dissolved carbide particles. • Hardness of clad (1135 ± 88 HV) is 3.5 times that of the substrate (325 ± 49 HV)

  19. CAFE experiments on the flow and freezing of metal fuel and cladding melts (1) - Test conditions and overview of the results -

    International Nuclear Information System (INIS)

    For metal fueled fast reactors, assessment of the core disruptive accident (CDA) is necessary for both design and licensing. This assessment has generally addressed an initiating accident phase (fuel pin disruption and material relocation within subassemblies), followed by a transition phase (disruption of subassemblies and gross core-material relocation), and concluding in a post-accident heat removal phase (establishment of a stable, coolable configuration of core material). In order to assess the CDA, knowledge of the flow and freezing behavior of the molten core material as it moves axially through and beyond the core is required. This information is important for predicting fuel relocation during both the initiating and transition phases. For uranium-based metal fuel in stainless steel-clad fuel pins, the flow and freezing behavior of the core melt is complicated due to metallurgical interaction that will occur between the melt and the steel structures such as intact cladding, duct wall, or below-core structures. This interaction results in composition changes of the melt including eutectic formation and associated changes in the freezing temperature of the melt and other thermo-physical properties. The objectives of the Core Alloy Flow and Erosion (CAFE) experiments are to investigate the fundamental flow, metallurgical interaction, and freezing behavior of uranium-iron-type melts within iron-based trough-shaped flow channels and provide information that can support the development of mathematical models that describe the movements of molten fuel-bearing core materials during CDAs. The CAFE experiment apparatus consists principally of the induction heating system, melt flow system (comprising the crucible, flow controller cup, trough-shaped flow channel, and catch cup), confinement and ventilation system, instrumentation and control system. Melt produced in yttria-coated crucible by induction heating was received at the flow controller cup so as to prevent

  20. Mechanical interaction fuel/cladding

    International Nuclear Information System (INIS)

    There is a common agreement that rather large plastic cladding deformation may occur in fast breeder reactor conditions. In thermal irradiation experiments these deformations can be directly measured as cladding diameter increase. In case of fast flux, a distinction must be made between plastic strain and swelling due to pore formation. The separation of these two effects can be made by a combination of cladding diameter measurements and cladding density measurements. A simpler method to determine the mean plastic cladding expansion is to compare the increase of relative mean cladding diameter along the fuel element and the increase of relative cladding length. This comparison for the irradiation experiment in Rapsodie is shown

  1. Cladding material, tube including such cladding material and methods of forming the same

    Energy Technology Data Exchange (ETDEWEB)

    Garnier, John E.; Griffith, George W.

    2016-03-01

    A multi-layered cladding material including a ceramic matrix composite and a metallic material, and a tube formed from the cladding material. The metallic material forms an inner liner of the tube and enables hermetic sealing of thereof. The metallic material at ends of the tube may be exposed and have an increased thickness enabling end cap welding. The metallic material may, optionally, be formed to infiltrate voids in the ceramic matrix composite, the ceramic matrix composite encapsulated by the metallic material. The ceramic matrix composite includes a fiber reinforcement and provides increased mechanical strength, stiffness, thermal shock resistance and high temperature load capacity to the metallic material of the inner liner. The tube may be used as a containment vessel for nuclear fuel used in a nuclear power plant or other reactor. Methods for forming the tube comprising the ceramic matrix composite and the metallic material are also disclosed.

  2. Brittle fracture resistance of anti-corrosive cladding on pressure vessel

    International Nuclear Information System (INIS)

    This paper reports the estimation of brittle fracture resistance of austenitic-ferritic anticorrosive cladding metal, produced by submerged arc welding with the use of strip electrodes. The dependence of impact toughness and temperature both in as produced condition and after the exposure to a neutron fluence together with the temperature dependence of cladding metal static crack resistance were determined. The transition from ductile to brittle condition for cladding metal was found to be typical for a ferritic-perlitic steel

  3. Management of cladding hulls and fuel hardware

    International Nuclear Information System (INIS)

    The reprocessing of spent fuel from power reactors based on chop-leach technology produces a solid waste product of cladding hulls and other metallic residues. This report describes the current situation in the management of fuel cladding hulls and hardware. Information is presented on the material composition of such waste together with the heating effects due to neutron-induced activation products and fuel contamination. As no country has established a final disposal route and the corresponding repository, this report also discusses possible disposal routes and various disposal options under consideration at present

  4. Development of Cr Electroplated Cladding Tube for preventing Fuel-Cladding Chemical Interaction (FCCI)

    International Nuclear Information System (INIS)

    Metal fuel has been selected as a candidate fuel in the SFR because of its superior thermal conductivity as well as enhanced proliferation resistance in connection with the pyroprocessing. However, metal fuel suffers eutectic reaction (Fuel Cladding Chemical Interaction, FCCI) with the fuel cladding made of stainless steel at reactor operating temperature so that cladding thickness gradually reduces to endanger reactor safety. In order to mitigate FCCI, barrier concept has been proposed between the fuel and the cladding in designing fuel rod. Regarding this, KAERI has initiated barrier cladding development to prevent interdiffusion process as well as enhance the SFR fuel performance. Previous study revealed that Cr electroplating has been selected as one of the most promising options because of its technical and economic viability. This paper describes the development status of the Cr electroplating technology for the usage of fuel rod in SFR. This paper summarizes the status of Cr electroplating technology to prevent FCCI in metal fuel rod. It has been selected for the ease of practical application at the tube inner surface. Technical scoping, performance evaluation and optimization have been carried out. Application to the tube inner surface and in-pile test were conducted which revealed as effective

  5. Development of Cr Electroplated Cladding Tube for preventing Fuel-Cladding Chemical Interaction (FCCI)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jun Hwan; Woo, Je Woong; Kim, Sung Ho; Cheon, Jin Sik; Lee, Byung Oon; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Metal fuel has been selected as a candidate fuel in the SFR because of its superior thermal conductivity as well as enhanced proliferation resistance in connection with the pyroprocessing. However, metal fuel suffers eutectic reaction (Fuel Cladding Chemical Interaction, FCCI) with the fuel cladding made of stainless steel at reactor operating temperature so that cladding thickness gradually reduces to endanger reactor safety. In order to mitigate FCCI, barrier concept has been proposed between the fuel and the cladding in designing fuel rod. Regarding this, KAERI has initiated barrier cladding development to prevent interdiffusion process as well as enhance the SFR fuel performance. Previous study revealed that Cr electroplating has been selected as one of the most promising options because of its technical and economic viability. This paper describes the development status of the Cr electroplating technology for the usage of fuel rod in SFR. This paper summarizes the status of Cr electroplating technology to prevent FCCI in metal fuel rod. It has been selected for the ease of practical application at the tube inner surface. Technical scoping, performance evaluation and optimization have been carried out. Application to the tube inner surface and in-pile test were conducted which revealed as effective.

  6. Initial Cladding Condition

    International Nuclear Information System (INIS)

    The purpose of this analysis is to describe the condition of commercial Zircaloy clad fuel as it is received at the Yucca Mountain Project (YMP) site. Most commercial nuclear fuel is encased in Zircaloy cladding. This analysis is developed to describe cladding degradation from the expected failure modes. This includes reactor operation impacts including incipient failures, potential degradation after reactor operation during spent fuel storage in pool and dry storage and impacts due to transportation. Degradation modes include cladding creep, and delayed hydride cracking during dry storage and transportation. Mechanical stresses from fuel handling and transportation vibrations are also included. This Analysis and Model Report (AMR) does not address any potential damage to assemblies that might occur at the YMP surface facilities. Ranges and uncertainties have been defined. This analysis will be the initial boundary condition for the analysis of cladding degradation inside the repository. In accordance with AP-2.13Q, ''Technical Product Development Planning'', a work plan (CRWMS M andO 2000c) was developed, issued, and utilized in the preparation of this document. There are constraints, caveats and limitations to this analysis. This cladding degradation analysis is based on commercial Pressurized Water Reactor (PWR) fuel with Zircaloy cladding but is applicable to Boiling Water Reactor (BWR) fuel. Reactor operating experience for both PWRs and BWRs is used to establish fuel reliability from reactor operation. It is limited to fuel exposed to normal operation and anticipated operational occurrences (i.e. events which are anticipated to occur within a reactor lifetime), and not to fuel that has been exposed to severe accidents. Fuel burnup projections have been limited to the current commercial reactor licensing environment with restrictions on fuel enrichment, oxide coating thickness and rod plenum pressures. The information provided in this analysis will be used in

  7. Initial Cladding Condition

    Energy Technology Data Exchange (ETDEWEB)

    E. Siegmann

    2000-08-22

    The purpose of this analysis is to describe the condition of commercial Zircaloy clad fuel as it is received at the Yucca Mountain Project (YMP) site. Most commercial nuclear fuel is encased in Zircaloy cladding. This analysis is developed to describe cladding degradation from the expected failure modes. This includes reactor operation impacts including incipient failures, potential degradation after reactor operation during spent fuel storage in pool and dry storage and impacts due to transportation. Degradation modes include cladding creep, and delayed hydride cracking during dry storage and transportation. Mechanical stresses from fuel handling and transportation vibrations are also included. This Analysis and Model Report (AMR) does not address any potential damage to assemblies that might occur at the YMP surface facilities. Ranges and uncertainties have been defined. This analysis will be the initial boundary condition for the analysis of cladding degradation inside the repository. In accordance with AP-2.13Q, ''Technical Product Development Planning'', a work plan (CRWMS M&O 2000c) was developed, issued, and utilized in the preparation of this document. There are constraints, caveats and limitations to this analysis. This cladding degradation analysis is based on commercial Pressurized Water Reactor (PWR) fuel with Zircaloy cladding but is applicable to Boiling Water Reactor (BWR) fuel. Reactor operating experience for both PWRs and BWRs is used to establish fuel reliability from reactor operation. It is limited to fuel exposed to normal operation and anticipated operational occurrences (i.e. events which are anticipated to occur within a reactor lifetime), and not to fuel that has been exposed to severe accidents. Fuel burnup projections have been limited to the current commercial reactor licensing environment with restrictions on fuel enrichment, oxide coating thickness and rod plenum pressures. The information provided in this analysis

  8. Pellet cladding mechanical interactions of ceramic claddings fuels under light water reactor conditions

    Science.gov (United States)

    Li, Bo-Shiuan

    Ceramic materials such as silicon carbide (SiC) are promising candidate materials for nuclear fuel cladding and are of interest as part of a potential accident tolerant fuel design due to its high temperature strength, dimensional stability under irradiation, corrosion resistance, and lower neutron absorption cross-section. It also offers drastically lower hydrogen generation in loss of coolant accidents such as that experienced at Fukushima. With the implementation of SiC material properties to the fuel performance code, FRAPCON, performances of the SiC-clad fuel are compared with the conventional Zircaloy-clad fuel. Due to negligible creep and high stiffness, SiC-clad fuel allows gap closure at higher burnup and insignificant cladding dimensional change. However, severe degradation of SiC thermal conductivity with neutron irradiation will lead to higher fuel temperature with larger fission gas release. High stiffness of SiC has a drawback of accumulating large interfacial pressure upon pellet-cladding mechanical interactions (PCMI). This large stress will eventually reach the flexural strength of SiC, causing failure of SiC cladding instantly in a brittle manner instead of the graceful failure of ductile metallic cladding. The large interfacial pressure causes phenomena that were previously of only marginal significance and thus ignored (such as creep of the fuel) to now have an important role in PCMI. Consideration of the fuel pellet creep and elastic deformation in PCMI models in FRAPCON provide for an improved understanding of the magnitude of accumulated interfacial pressure. Outward swelling of the pellet is retarded by the inward irradiation-induced creep, which then reduces the rate of interfacial pressure buildup. Effect of PCMI can also be reduced and by increasing gap width and cladding thickness. However, increasing gap width and cladding thickness also increases the overall thermal resistance which leads to higher fuel temperature and larger fission

  9. Wear resistance and hot corrosion behaviour of laser cladding Co-based alloy

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    2Cr13 stainless steel was surface cladded with Co-based alloy using a high power carbon dioxide laser. The microstructure, wear resistance and corrosion properties of the clad layer were investigated. It is found that the high temperature corrosion behavior and wearing resistant property of the clad layer are 3 and 2.5 times higher than those of the parent metal. Under the high temperature molten lead sulphate salt corrosion condition, the clad layer fails by spalling which is caused by intergrannular corrosion within the clad layer. The fine dendritic structure and the oxide help to retard the penetration of the sulphur ion that induces the intergrannular corrosion.

  10. High Temperature Steam Corrosion of Cladding for Nuclear Applications: Experimental

    Energy Technology Data Exchange (ETDEWEB)

    McHugh, Kevin M; Garnier, John E; Sergey Rashkeev; Michael V. Glazoff; George W. Griffith; Shannong M. Bragg-Sitton

    2013-01-01

    Stability of cladding materials under off-normal conditions is an important issue for the safe operation of light water nuclear reactors. Metals, ceramics, and metal/ceramic composites are being investigated as substitutes for traditional zirconium-based cladding. To support down-selection of these advanced materials and designs, a test apparatus was constructed to study the onset and evolution of cladding oxidation, and deformation behavior of cladding materials, under loss-of-coolant accident scenarios. Preliminary oxidation tests were conducted in dry oxygen and in saturated steam/air environments at 1000OC. Tube samples of Zr-702, Zr-702 reinforced with 1 ply of a ß-SiC CMC overbraid, and sintered a-SiC were tested. Samples were induction heated by coupling to a molybdenum susceptor inside the tubes. The deformation behavior of He-pressurized tubes of Zr-702 and SiC CMC-reinforced Zr-702, heated to rupture, was also examined.

  11. Physics model describing the melting, the motion and the relocation of the clad after an undercooling accident in a liquid-metal fast breeder reactor

    International Nuclear Information System (INIS)

    In this note, we present a physics model describing after an accidental loss of flow in a annular channel, the melting, the motion and the relocalisation of the clad, with (or without) centering grids for the pin. The heating of the fuel and centering grids is also described. The equations of the model, their difference approximations and their resolution as well as the flow chart of the code are indicated. The possibilities of the code ALFA are illustrated by some examples and in particular with some calculations concerning the experiments CABRI B1 and CEFUS TR 3.1. The calculated and experimental results are in good agreement: the phenomena chronology is well described and the post morten geometry is found

  12. EPRI fuel cladding integrity program

    Energy Technology Data Exchange (ETDEWEB)

    Yang, R. [Electric Power Research Institute, Palo Alto, CA (United States)

    1997-01-01

    The objectives of the EPRI fuel program is to supplement the fuel vendor research to assure that utility economic and operational interests are met. To accomplish such objectives, EPRI has conducted research and development efforts to (1) reduce fuel failure rates and mitigate the impact of fuel failures on plant operation, (2) provide technology to extend burnup and reduce fuel cycle cost. The scope of R&D includes fuel and cladding. In this paper, only R&D related to cladding integrity will be covered. Specific areas aimed at improving fuel cladding integrity include: (1) Fuel Reliability Data Base; (2) Operational Guidance for Defective Fuel; (3) Impact of Water Chemistry on Cladding Integrity; (4) Cladding Corrosion Data and Model; (5) Cladding Mechanical Properties; and (6) Transient Fuel Cladding Response.

  13. Clad ballooning model in MELCOR

    International Nuclear Information System (INIS)

    Clad ballooning may substantially decrease the flow of fluids through the affected core region and may expose the inner cladding surface to oxidation in the vicinity of rupture sites. The cladding ballooning model was not included in MELCOR 1.8.4. and consideration of incorporating the cladding ballooning model is scheduled as a post-1.8.4 release activity. The purpose of this paper is to analyze the effect of the clad ballooning model by the modified MELCOR 1.8.4 with this model. The typical accident sequence of a large LOCA scenario is selected. The clad ballooning model accelerates the accident progression compared to that without the ballooning model. The amount of hydrogen does not change much and it may be caused by ignoring the effect of flow area change. Future study is planning to analyze the flow redistribution

  14. Fuel cladding tubes and fuel elements

    International Nuclear Information System (INIS)

    Purpose: To enable non-destructive measurement for the thickness of zirconium barriers. Constitution: Regions capable of non-destructive inspection are provided at the boundary between a fuel cladding tube made of zirconium alloy and the zirconium barrier lined to the inner circumference surface of the tube. As the regions being capable of distinguishing by ultrasonic wave reflection, solid materials, for example, non-metal materials different from that for the tube and the barrier are placed or gaps are provided at the boundary between the zirconium alloy cladding tube and the zirconium barrier. Since ultrasonic waves are reflected at each of the boundaries by the presence of these regions, thickness of the zirconium barrier can be measured in a non-destructive manner from either the inner or the outer surface of the tube. (Yoshino, Y.)

  15. GCFR core cladding temperature limits

    International Nuclear Information System (INIS)

    This paper reviews the phenomena that affect selection of the GCFR cladding faulted temperature limit. The limiting effects are determined to be clad melting, strength and oxidation rate. The selected temperature limit is 13000C (23700F). The limits for normal, upset and emergency events are also breifly reviewed, and some changes under consideration are discussed

  16. Evolution of Westinghouse fuel cladding

    International Nuclear Information System (INIS)

    As the nuclear power generating industry has matured, there is an increasing trend in core operating fuel duties. At the same time, refined requirements from regulators, e.g. in the areas of LOCA and RIA, must be fulfilled. This drives a continuing evolution of cladding materials, to provide more performance margin and support even higher fuel duty designs. Cladding performance, in particular with respect to in-reactor corrosion and hydrogen pickup, has improved dramatically since Zircaloy-2 and Zircaloy-4 were established in 1952 and 1960 respectively. For Westinghouse PWR cladding, the corrosion rate has decreased by more than one order of magnitude since; going from the original Zircaloy-4 to ZIRLO® and Optimized ZIRLO™ claddings. The next generation of Westinghouse PWR cladding, AXIOM™, shows further reduction of corrosion and hydrogen pickup, most notably at very high burnup, over 70 GWD/MTU. In Westinghouse BWR fuel, a carefully optimized variant of Zircaloy-2, LK3™ cladding, continues to demonstrate excellent performance under all operating conditions to date. In order to further reduce the hydrogen pickup, a new BWR cladding alloy, HiFi™, developed by NFI, is now being verified. Data indicate a reduction of the hydrogen absorption of around 50% with respect to Zircaloy-2. This paper describes the evolution of the different PWR and BWR cladding materials, providing details of their current experience base and post-irradiation examinations. (author)

  17. Aerogel-clad optical fiber

    Science.gov (United States)

    Sprehn, Gregory A.; Hrubesh, Lawrence W.; Poco, John F.; Sandler, Pamela H.

    1997-01-01

    An optical fiber is surrounded by an aerogel cladding. For a low density aerogel, the index of refraction of the aerogel is close to that of air, which provides a high numerical aperture to the optical fiber. Due to the high numerical aperture, the aerogel clad optical fiber has improved light collection efficiency.

  18. Analysis of cold crack of AP1000 steam generator tube sheet cladding

    International Nuclear Information System (INIS)

    This paper discusses the causes of AP1000 Steam Generator (SG) tube sheet underclad cracking. Base metal weldability, hydrogen influence, welding techniques and weld residue stresses are all discussed in details in contributing to the underclad cracking problems. Feasible and realistic improvement plans are proposed for the AP1000 SG tube sheet cladding, including the controlling forging procurement, welding process and cladding techniques. (authors)

  19. Optimization of process parameters in explosive cladding of titanium/stainless steel 304L plates

    International Nuclear Information System (INIS)

    Explosive cladding is a solid state welding process best suited for joining incompatible metals. The selection of process parameters viz., explosive mass ratio, stand off distance and initial angle of inclination dictate the nature of the cladding. Optimization of process parameters in explosive cladding of titanium-stainless steel 304L plates, based on two level three factorial design, is attempted to establish the influencing parameters. Analysis of variance was employed to find the linear, regression and interaction values. Mathematical models to estimate the responses-amplitude and wavelength were developed. The microstructure of the Ti-SS304L explosive clad interface reveals characteristic undulations concurrent with design expectations. (orig.)

  20. Implications and control of fuel-cladding chemical interaction for LMFBR fuel pin design

    International Nuclear Information System (INIS)

    Fuel-cladding-chemical-interaction (FCCI) is typically incorporated into the design of an LMFBR fuel pin as a wastage allowance. Several interrelated factors are considered during the evolution of an LMFBR fuel pin design. Those which are indirectly affected by FCCI include: allowable pin power, fuel restructuring, fission gas migration and release from the fuel, fuel cracking, fuel swelling, in-reactor cladding creep, cladding swelling, and the cladding mechanical strain. Chemical activity of oxygen is the most readily controlled factor in FCCI. Two methods are being investigated: control of total oxygen inventory by limiting fuel O/M, and control of oxygen activity with buffer metals

  1. Application of YAG laser cladding to the flange seating surface

    International Nuclear Information System (INIS)

    Stainless cladding on carbon steel is usually conducted by shielded metal arc welding (SMAW) or gas tungsten arc welding (GTAW). YAG ( Yttrium-Aluminum-Garnet) laser welding is superior to these methods of welding in the following respects : (1) The heat affected zone (HAZ) is narrower and there is less distortion. (2) YAG laser cladding has the required chemical compositions, even with possibly fewer welding layers under controlled dilution. (3) Greater welding speed. YAG laser cladding application to vessel flange seating surfaces was examined in this study and the results are discussed. The following objectives were carried out : (1) Determination of welding conditions for satisfactory cladding layers and (2) whether cladding would be adequately possible at a cornered section of a stair-like plate, assuming actual flange shape. (3) Measurement of welding distortion and heat affected zone in carbon steel. The welding conditions for producing no-crack deposit with low dilution in carbon steel were clarified and welding by which cladding at cornered section would be possible was achieved. welding distortion by YAG laser was found less than with GTAW and HAZ made by first layer welding could be tempered appropriately by second layer welding. (author)

  2. Mechanisms of fuel-cladding chemical interaction: US interpretation

    International Nuclear Information System (INIS)

    Proposed mechanisms of fuel-cladding chemical interaction (FCCI) in LMFBR fuel pins are reviewed and examined in terms of in-pile and out-of-pile data. From this examination several factors are identified which may govern the occurrence of localized deep intergranular penetrations of Type-316SS cladding. Using a plausible mechanistic hypothesis for FCCI, first steps have been taken towards developing a quantitative, physically-meaningful, mathematical method of predicting cladding wastage in operating fuel pins. Both kinetic and thermodynamic aspects of FCCI are considered in the development of this prediction method, together with a fuel chemistry model that describes the evolution of thermochemical conditions at the fuel-cladding gap. On the basis of results from recent fuel pin and laboratory tests a thermal transport mechanism has been proposed to explain the thermal gradient-induced migration of Fe, Cr, and Ni from cladding into the fuel. This mechanism involves chemical transport of the metallic cladding components (as tellurides) in liquid Cs-Te. (author)

  3. Residual stress measurements in laser clad aircraft aluminium alloys

    International Nuclear Information System (INIS)

    Fatigue and corrosion damage of structural components threatens the safety and availability of civil and military aircrafts. There is no sign of relief from these threats as civil and military aircrafts worldwide are continuously being pushed further into and past their initial design fatigue lives in tight financial circumstances. Given fatigue and corrosion damage often initiates at the surface and sub-surface of the components, there has been extensive research and development worldwide focused on advanced aircraft repair technologies and surface enhancement methods. The Deep Surface Rolling (DSR) is one of advanced surface enhancement technologies that can introduce deep compressive residual stresses into the surface of aircraft metallic structure to extend its fatigue life. For the development of cost-effective aircraft structural repair technologies such as laser cladding, in this study, aluminium alloy 7075-T651 specimens with simulated corrosion damage were repaired using laser cladding technology. The surface of the laser cladding region was then processed by DSR. The experimental results from subsequent fatigue testing of laser cladded baseline, DSR and post-heat treated laser cladded specimens discovered that the DSR process can significantly increase fatigue life in comparison with the ascladded baseline. The three dimensional residual stresses were measured by neutron diffraction and the results confirmed the beneficial compressive residual stresses at the cladding surface can be achieved in depth more than 1.0 mm.

  4. Multiresponse Optimization of Laser Cladding Steel + VC Using Grey Relational Analysis in the Taguchi Method

    Science.gov (United States)

    Zhang, Zhe; Kovacevic, Radovan

    2016-07-01

    Laser cladding of metal matrix composite coatings (MMCs) has become an effective and economic method to improve the wear resistance of mechanical components. The clad quality characteristics such as clad height, carbide fraction, carbide dissolution, and matrix hardness in MMCs determine the wear resistance of the coatings. These clad quality characteristics are influenced greatly by the laser cladding processing parameters. In this study, American Iron and Steel Institute (AISI) 420 + 20% vanadium carbide (VC) was deposited on mild steel with a high powder direct diode laser. The Taguchi-based Grey relational method was used to optimize the laser cladding processing parameters (laser power, scanning speed, and powder feed rate) with the consideration of multiple clad characteristics related to wear resistance (clad height, carbide volume fraction, and Fe-matrix hardness). A Taguchi L9 orthogonal array was designed to study the effects of processing parameters on each response. The contribution and significance of each processing parameter on each clad characteristic were investigated by the analysis of variance (ANOVA). The Grey relational grade acquired from Grey relational analysis was used as the performance characteristic to obtain the optimal combination of processing parameters. Based on the optimal processing parameters, the phases and microstructure of the laser-cladded coating were characterized by using x-ray diffraction (XRD) and scanning electron microscopy (SEM) with energy-dispersive spectroscopy (EDS).

  5. Complete Non-Radioactive Operability Tests for Cladding Hull Chlorination

    Energy Technology Data Exchange (ETDEWEB)

    Collins, Emory D [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Johnson, Jared A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hylton, Tom D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brunson, Ronald Ray [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunt, Rodney Dale [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); DelCul, Guillermo Daniel [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bradley, Eric Craig [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Spencer, Barry B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-04-01

    Non-radioactive operability tests were made to test the metal chlorination reactor and condenser and their accessories using batch chlorinations of non-radioactive cladding samples and to identify optimum operating practices and components that need further modifications prior to installation of the equipment into the hot cell for tests on actual used nuclear fuel (UNF) cladding. The operability tests included (1) modifications to provide the desired heating and reactor temperature profile; and (2) three batch chlorination tests using, respectively, 100, 250, and 500 g of cladding. During the batch chlorinations, metal corrosion of the equipment was assessed, pressurization of the gas inlet was examined and the best method for maintaining solid salt product transfer through the condenser was determined. Also, additional accessing equipment for collection of residual ash and positioning of the unit within the hot cell were identified, designed, and are being fabricated.

  6. Explosion Clad for Upstream Oil and Gas Equipment

    Science.gov (United States)

    Banker, John G.; Massarello, Jack; Pauly, Stephane

    2011-01-01

    Today's upstream oil and gas facilities frequently involve the combination of high pressures, high temperatures, and highly corrosive environments, requiring equipment that is thick wall, corrosion resistant, and cost effective. When significant concentrations of CO2 and/or H2S and/or chlorides are present, corrosion resistant alloys (CRA) can become the material of choice for separator equipment, piping, related components, and line pipe. They can provide reliable resistance to both corrosion and hydrogen embrittlement. For these applications, the more commonly used CRA's are 316L, 317L and duplex stainless steels, alloy 825 and alloy 625, dependent upon the application and the severity of the environment. Titanium is also an exceptional choice from the technical perspective, but is less commonly used except for heat exchangers. Explosion clad offers significant savings by providing a relatively thin corrosion resistant alloy on the surface metallurgically bonded to a thick, lower cost, steel substrate for the pressure containment. Developed and industrialized in the 1960's the explosion cladding technology can be used for cladding the more commonly used nickel based and stainless steel CRA's as well as titanium. It has many years of proven experience as a reliable and highly robust clad manufacturing process. The unique cold welding characteristics of explosion cladding reduce problems of alloy sensitization and dissimilar metal incompatibility. Explosion clad materials have been used extensively in both upstream and downstream oil, gas and petrochemical facilities for well over 40 years. The explosion clad equipment has demonstrated excellent resistance to corrosion, embrittlement and disbonding. Factors critical to insure reliable clad manufacture and equipment design and fabrication are addressed.

  7. Weld overlay cladding with iron aluminides

    Energy Technology Data Exchange (ETDEWEB)

    Goodwin, G.M. [Oak Ridge National Lab., TN (United States)

    1995-08-01

    The hot and cold cracking tendencies of some early iron aluminide alloy compositions have limited their use in applications where good weldability is required. Using hot crack testing techniques invented at ORNL, and experimental determinations of preheat and postweld heat treatment needed to avoid cold cracking, we have developed iron aluminide filler metal compositions which can be successfully used to weld overlay clad various substrate materials, including 9Cr-1Mo steel, 2-1/4Cr-1Mo steel, and 300-series austenitic stainless steels. Dilution must be carefully controlled to avoid crack-sensitive deposit compositions. The technique used to produce the current filler metal compositions is aspiration-casting, i.e. drawing the liquid from the melt into glass rods. Future development efforts will involve fabrication of composite wires of similar compositions to permit mechanized gas tungsten arc (GTA) and/or gas metal arc (GMA) welding.

  8. Elimination of Start/Stop defects in laser cladding

    NARCIS (Netherlands)

    Ocelik, V.; Eekma, M.; Hemmati, I.; De Hosson, J. Th. M.

    2012-01-01

    Laser cladding represents an advanced hard facing technology for the deposition of hard, corrosion and wear resistant layers of controlled thickness onto a selected area of metallic substrate. When a circular geometry is required, the beginning and the end of the laser track coincide in the same are

  9. Microstrain Determination in Individual Grains of Laser Deposited Cladding Layers

    NARCIS (Netherlands)

    de Oliveira, Uazir O. B.; Ocelik, Vaclav; De Hosson, Jeff T. M.; Chandra, T; Tsuzaki, K; Militzer, M; Ravindran, C

    2007-01-01

    The laser cladding technique makes the deposition of thick metallic, wear and corrosion resistant coatings feasible on weaker substrates. During the process, localized high thermal gradients generate internal stresses that may cause cracking when these overcome the fracture stress. To explain the fo

  10. Study on laser welding of fuel clad tubes and end plugs made of modified 9Cr-1Mo steel for metallic fuel of Fast Breeder Reactors

    Science.gov (United States)

    Harinath, Y. V.; Gopal, K. A.; Murugan, S.; Albert, S. K.

    2013-04-01

    A procedure for Pulsed Laser Beam Welding (PLBW) has been developed for fabrication of fuel pins made of modified 9Cr-1Mo steel for metallic fuel proposed to be used in future in India's Fast Breeder Reactor (FBR) programme. Initial welding trials of the samples were carried out with different average power using Nd-YAG based PLBW process. After analyzing the welds, average power for the weld was optimized for the required depth of penetration and weld quality. Subsequently, keeping the average power constant, the effect of various other welding parameters like laser peak power, pulse frequency, pulse duration and energy per pulse on weld joint integrity were studied and a procedure that would ensure welds of acceptable quality with required depth of penetration, minimum size of fusion zone and Heat Affected Zone (HAZ) were finalized. This procedure is also found to reduce the volume fraction delta-ferrite in the fusion zone.

  11. High power cladding light strippers

    Science.gov (United States)

    Wetter, Alexandre; Faucher, Mathieu; Sévigny, Benoit

    2008-02-01

    The ability to strip cladding light from double clad fiber (DCF) fibers is required for many different reasons, one example is to strip unwanted cladding light in fiber lasers and amplifiers. When removing residual pump light for example, this light is characterized by a large numerical aperture distribution and can reach power levels into the hundreds of watts. By locally changing the numerical aperture (N.A.) of the light to be stripped, it is possible to achieve significant attenuation even for the low N.A. rays such as escaped core modes in the same device. In order to test the power-handling capability of this device, one hundred watts of pump and signal light is launched from a tapered fusedbundle (TFB) 6+1x1 combiner into a high power-cladding stripper. In this case, the fiber used in the cladding stripper and the output fiber of the TFB was a 20/400 0.06/0.46 N.A. double clad fiber. Attenuation of over 20dB in the cladding was measured without signal loss. By spreading out the heat load generated by the unwanted light that is stripped, the package remained safely below the maximum operating temperature internally and externally. This is achieved by uniformly stripping the energy along the length of the fiber within the stripper. Different adhesive and heat sinking techniques are used to achieve this uniform removal of the light. This suggests that these cladding strippers can be used to strip hundreds of watts of light in high power fiber lasers and amplifiers.

  12. Technical committee meeting on fuel and cladding interaction. Summary report

    International Nuclear Information System (INIS)

    Experiments and experiences concerning fuel-cladding interaction in thermal and fast neutron flux burnup are dealt with. A number of results from in-pile and out-of pile experiments with different fuel pins with cladding made of different stainless steels showed the importance of corrosion process, dependent on the burnup, core temperature, metal-oxide ratio, and other steady state parameters in the core of fast reactors (most frequently LMFBRs). This is of importance for fuel pins design and fabrication. Mixed oxide fuel is treated in many cases

  13. Laser cladding of cobalt and boron free hardfacing materials for nuclear applications

    International Nuclear Information System (INIS)

    Full text: Most common hardfacing alloys are of stellite family, which are cobalt base alloys and borides are also being developed for the same purpose. Both cobalt and boron are not preferred in nuclear industry as cobalt becomes radioactive after irradiation and boron is a neutron poison. Therefore, there is a need to develop cobalt and boron free hardfacing alloys for nuclear application. Attempt has been made to develop cobalt - boron free hardfacing alloys for laser cladding. Laser cladding of three nickel based hardfacing materials, one metallic system (Ni-15Cr-32Mo) and two composite systems (Ni-20Cr)-40Cr2C3 and (Ni-20Cr)-40WC) has been attempted. These hardfacing materials were cladded onto 0.15C steel sheet by blown powder laser cladding. Laser cladding of stellite was also done for comparative purpose. The process parameters were optimised to obtain defect free cladding. The cladded samples were characterized by visual, optical microscopy and microhardness measurements. Wear testing of these claddings was done by the pin on disc method against 600-grit size abrasive paper. Comparative study of wear properties of these claddings was done. Results of these investigations are reported in this paper

  14. Deep surface rolling for fatigue life enhancement of laser clad aircraft aluminium alloy

    International Nuclear Information System (INIS)

    Highlights: • Deep surface rolling as a post-repair enhancement technology was applied to the laser cladded 7075-T651 aluminium alloy specimens that simulated corrosion damage blend-out repair. • The residual stresses induced by the deep surface rolling process were measured. • The deep surface rolling process can introduce deep and high magnitude compressive residual stresses beyond the laser clad and substrate interface. • Spectrum fatigue test showed the fatigue life was significantly increased by deep surface rolling. - Abstract: Deep surface rolling can introduce deep compressive residual stresses into the surface of aircraft metallic structure to extend its fatigue life. To develop cost-effective aircraft structural repair technologies such as laser cladding, deep surface rolling was considered as an advanced post-repair surface enhancement technology. In this study, aluminium alloy 7075-T651 specimens with a blend-out region were first repaired using laser cladding technology. The surface of the laser cladding region was then treated by deep surface rolling. Fatigue testing was subsequently conducted for the laser clad, deep surface rolled and post-heat treated laser clad specimens. It was found that deep surface rolling can significantly improve the fatigue life in comparison with the laser clad baseline repair. In addition, three dimensional residual stresses were measured using neutron diffraction techniques. The results demonstrate that beneficial compressive residual stresses induced by deep surface rolling can reach considerable depths (more than 1.0 mm) below the laser clad surface

  15. Development of Co-Pilgering Process for Manufacturing Double Clad Tubes for Accident Tolerant Fuel

    International Nuclear Information System (INIS)

    Accident Tolerant Fuels (ATF) are those that, in comparison with the standard UO2 - Zr system, can tolerate loss of active cooling in the core for a considerably longer time period (depending on the accident scenario), while maintaining or improving the fuel performance during normal operations. ATF cladding development efforts focus on materials with more benign steam reaction. For this, advanced steels (e.g. FeCrAl), refractory metals (e.g. Mo), ceramic cladding (SiC), Innovative alloys with dopants, zirconium alloy with coating or sleeve are being developed. Single material like zirconium alloy as clad may not be compatible with both fuel and coolant at elevated temperatures in accident scenario. Double clad tube is one of the prime concepts which has to be explored to develop ATF cladding. Two different clad materials- one oxidant resistant (like FeCrAl) and the other, fuel compatible (like Zr-4) constitute together as outer and inner tube to form ATF cladding. Bonding two different tubes in controlled thickness ratios and with almost no gap in between is utmost difficult. Different types of processes are available for production of double clad tubes such as coating, co-extrusion, co- drawing, internal expansion/external compaction, explosive bonding, co-pilgering etc,. Nuclear Fuel Complex (NFC), India has successfully demonstrated manufacturing of double clad tube by co-pilgering process where in outer cladding is of modified 9Cr-1Mo Steel and inner liner is of zircaloy-4. Considering different deformation behaviour of above materials during pilgering, fabrication of double clad tube is very critical. Optimization of tube dimensions like outer diameter and wall thickness at pre and final stages during pilgering is very important to achieve the required overall tube dimension and bonding between the tubes. This paper gives the methodology of manufacture of Double Clad Tubes by pilgering and the bonding between the two materials achieved in this process

  16. Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Siefken, Larry James

    1999-02-01

    Preliminary designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the clad-ding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert's law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the proposed models on the calculated behavior of fuel rods in severe accident conditions. This report is a revision and reissue of the report entitled; "Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents."

  17. Cladding of pressure vessel steel with corrosion resistant filler material

    International Nuclear Information System (INIS)

    Pressure vessels are often on the inside clad with corrosion resistant material. Of the various cladding processes surfacing by welding has proved to be most useful, especially for large thick-walled pressure vessels. Submerged arc welding with strip electrode is the most common method. Rather promising results have also been obtained by plasma hot wire welding. In general, Nb-alloyed austenitic stainless steel, over-alloyed with Cr and Ni, is used as filler material. Henceforth, also nickel alloys, e.g. Inconel 600, are used. The surfacing is made in one or several layers, following the requirements on the clad surface and the welding process used. The most dangerous welding defects in the surface are various types of cracks. The corrosion resistance of the cladding can show rather high local variations, depending on the composition of the filler material and various welding process factors. It is proved that the surface layer comparises areas with low chromium martensite. To ensure the corrosion resistance of the cladding, the generation of low-chromium martensite must be prevented by using suitable welding parameters, welding equipment and filler metal. It is also possible to eliminate the negative influence on the corrosion resistance from the low-chromium martensite, e.g. by welding in two layers. In the case of the high demands on quality a welding procedure test should always be made prior to production welding.(author)

  18. Local strain in cladding tube due to radial pellet cracking

    International Nuclear Information System (INIS)

    A study was made to develop a method for evaluation of the local strain in a cladding tube of the Advanced Thermal Reactor due to radial cracking of a UO2 fuel pellet. Effects of the number of cracks, initial crack width and the friction coefficient of a pellet-clad interface on behaviors of the local strain in a cladding tube were evaluated with a modelized experiment. A Zircaloy-2 ring specimen with inner diameter of 95 mm, height of 25 mm and wall thickness of 5 mm was expanded at room temperature with equally divided peripheral dice of a tool steel set in a specimen. The dice were divided into 8, 12 or 16 pieces. For each dividing number, two dice edge geometries were prepared, that is, not chamfered and chamfered by 2 mm. Strains of an external surface of the specimen were measured with 28 wire strain gages with gage length of 0.3 mm. The friction coefficient on the pellet-clad contact surface was not measured, but two friction conditions were prepared. One was metal-metal contact and the other was a contact surface coated with teflon film. The estimated friction coefficient was 0.1 for the former and 0.05 for the latter. An elastic-plastic analysis was carried out in order to evaluate the membrane hoop strain in the cladding tube. The analysis was made under two conditions. One was a plane stress condition of a radial and hoop stress which resembled the state of stress-strain developed in the ring specimen. The other was a plane strain condition of a radial and hoop strain which approximated the stress-strain state in a cladding tube

  19. Properties and features of structure formation CuCr-contact alloys in electron beam cladding

    Energy Technology Data Exchange (ETDEWEB)

    Durakov, Vasiliy G., E-mail: electron@ispms.tsc.ru [Institute of Strength Physics and Materials Science SB RAS, Tomsk, 634055 (Russian Federation); Dampilon, Bair V., E-mail: dampilon@ispms.tsc.ru, E-mail: gnusov@rambler.ru; Gnyusov, Sergey F., E-mail: dampilon@ispms.tsc.ru, E-mail: gnusov@rambler.ru [Institute of Strength Physics and Materials Science SB RAS, Tomsk, 634055, Russia and National Research Tomsk Polytechnic University, Tomsk, 634050 (Russian Federation)

    2014-11-14

    The microstructure and properties of the contact CuCr alloy produced by electron-beam cladding have been investigated. The effect of the electron beam cladding parameters and preheating temperature of the base metal on the structure and the properties of the coatings has been determined. The bimodal structure of the cladding coating has been established. The short circuit currents tests have been carried out according to the Weil-Dobke synthetic circuit simulating procedure developed for vacuum circuit breakers (VCB) test in real electric circuits. Test results have shown that the electron beam cladding (EBC) contact material has better breaking capacity than that of commercially fabricated sintered contact material. The application of the technology of electron beam cladding for production of contact material would significantly improve specific characteristics and reliability of vacuum switching equipment.

  20. Development of specimen preparation techniques for pitting potential measurement of irradiated fuel cladding tubes

    International Nuclear Information System (INIS)

    By the effect of the Great East Japan Earthquake, seawater was injected into spent fuel pools in unit 2, 3 and 4 at Fukushima Daiichi nuclear plant in order to cool spent fuels. It is known that chloride ion contained in seawater could cause pitting corrosion for metallic materials. It was concerned that radioactive products inside of fuel cladding tubes might be escaped through the pits. Therefore we have investigated the pit initiation condition of fuel cladding tubes by measuring pitting potential in order to evaluate stability of the enclosure function of fuel cladding tubes in spent fuel pools containing sea salt. In this report, we describe the development of specimen preparation techniques for pitting measurement of spent fuel cladding tubes having high radioactivity. By accomplishing of the development of the specimen preparation techniques, we could evaluate pit initiation condition of spent fuel cladding tubes in water containing sea salt. (author)

  1. Clad Degradation - FEPs Screening Arguments

    International Nuclear Information System (INIS)

    The purpose of this report is to document the screening of the cladding degradation features, events, and processes (FEPs) for commercial spent nuclear fuel (CSNF). This report also addresses the effect of some FEPs on both the cladding and the CSNF, DSNF, and HLW waste forms where it was considered appropriate to address the effects on both materials together. This report summarizes the work of others to screen clad degradation FEPs in a manner consistent with, and used in, the Total System Performance Assessment-License Application (TSPA-LA). This document was prepared according to ''Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of LA'' (BSC 2004a [DIRS 167796])

  2. Clad Degradation - FEPs Screening Arguments

    Energy Technology Data Exchange (ETDEWEB)

    E. Siegmann

    2004-03-17

    The purpose of this report is to document the screening of the cladding degradation features, events, and processes (FEPs) for commercial spent nuclear fuel (CSNF). This report also addresses the effect of some FEPs on both the cladding and the CSNF, DSNF, and HLW waste forms where it was considered appropriate to address the effects on both materials together. This report summarizes the work of others to screen clad degradation FEPs in a manner consistent with, and used in, the Total System Performance Assessment-License Application (TSPA-LA). This document was prepared according to ''Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of LA'' (BSC 2004a [DIRS 167796]).

  3. Pin clad strains in Phenix

    International Nuclear Information System (INIS)

    The Phenix reactor has operated for 4 years in a satisfactory manner. The first 2 sub-assembly loadings contained pins clad in solution treated 316. The principal pin strains are: diametral strain (swelling and irradiation creep), ovality and spiral bending of the pin (interaction of wire and pin cluster and wrapper). A pin cluster irradiated to a dose of 80 dpa F reached a pin diameter strain of 5%. This strain is principally due to swelling (low fission gas pressure). The principal parameters governing the swelling are instantaneous dose, time and temperature for a given type of pin cladding. Other types of steel are or will be irradiated in Phenix. In particular, cold-worked titanium stabilised 316 steel should contribute towards a reduction in the pin clad strains and increase the target burn-up in this reactor. (author)

  4. Preliminary design report for modeling of hydrogen uptake in fuel rod cladding during severe accidents

    International Nuclear Information System (INIS)

    Preliminary designs are described for models of the interaction of Zircaloy and hydrogen and the consequences of this interaction on the behavior of fuel rod cladding during severe accidents. The modeling of this interaction and its consequences involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer at the cladding external surface, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental and theoretical results are presented that show the uptake of hydrogen in the event of dissolution of the oxide layer occurs rapidly and that show the release of hydrogen in the event of cracking of the cladding occurs rapidly. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert's law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for Zr-H interaction into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the Zr-H interaction models on the calculated behavior of fuel rods in severe accident conditions

  5. Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents

    International Nuclear Information System (INIS)

    Preliminary designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert's law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the proposed models on the calculated behavior of fuel rods in severe accident conditions. This report is a revision and reissue of the report entitled; ''Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents''

  6. Preliminary design report for modeling of hydrogen uptake in fuel rod cladding during severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Siefken, L.J.

    1998-08-01

    Preliminary designs are described for models of the interaction of Zircaloy and hydrogen and the consequences of this interaction on the behavior of fuel rod cladding during severe accidents. The modeling of this interaction and its consequences involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer at the cladding external surface, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental and theoretical results are presented that show the uptake of hydrogen in the event of dissolution of the oxide layer occurs rapidly and that show the release of hydrogen in the event of cracking of the cladding occurs rapidly. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert`s law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for Zr-H interaction into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the Zr-H interaction models on the calculated behavior of fuel rods in severe accident conditions.

  7. Models for the Configuration and Integrity of Partially Oxidized Fuel Rod Cladding at High Temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Siefken, L.J.

    1999-01-01

    Models were designed to resolve deficiencies in the SCDAP/RELAP5/MOD3.2 calculations of the configuration and integrity of hot, partially oxidized cladding. These models are expected to improve the calculations of several important aspects of fuel rod behavior. First, an improved mapping was established from a compilation of PIE results from severe fuel damage tests of the configuration of melted metallic cladding that is retained by an oxide layer. The improved mapping accounts for the relocation of melted cladding in the circumferential direction. Then, rules based on PIE results were established for calculating the effect of cladding that has relocated from above on the oxidation and integrity of the lower intact cladding upon which it solidifies. Next, three different methods were identified for calculating the extent of dissolution of the oxidic part of the cladding due to its contact with the metallic part. The extent of dissolution effects the stress and thus the integrity of the oxidic part of the cladding. Then, an empirical equation was presented for calculating the stress in the oxidic part of the cladding and evaluating its integrity based on this calculated stress. This empirical equation replaces the current criterion for loss of integrity which is based on temperature and extent of oxidation. Finally, a new rule based on theoretical and experimental results was established for identifying the regions of a fuel rod with oxidation of both the inside and outside surfaces of the cladding. The implementation of these models is expected to eliminate the tendency of the SCDAP/RELAP5 code to overpredict the extent of oxidation of the upper part of fuel rods and to underpredict the extent of oxidation of the lower part of fuel rods and the part with a high concentration of relocated material. This report is a revision and reissue of the report entitled, Improvements in Modeling of Cladding Oxidation and Meltdown.

  8. Finite element elastic--plastic analysis of residual stresses due to clad welding in reactor vessels

    International Nuclear Information System (INIS)

    Residual stresses due to the weld deposited cladding on the inside of a typical Westinghouse pressurized water reactor vessel are investigated using an axisymmetric finite element elastic-plastic analysis. At the beginning of the analysis, one bead of the weld cladding is assumed to lie on the reactor vessel wall at melting temperature (26000F), but in the solid phase, while the vessel remains at 3000F (preheat temperature). All material properties used in the calculations are taken as temperature-dependent. Temperature profiles are obtained in the cladding and base metal at several discrete time intervals. These temperature profiles are used to obtain the stress distribution for the same time intervals. Residual hoop tensile stresses of approximately 25 ksi were found to exist in the cladding. Peak tensile stresses in the hoop direction occur in the base metal near the cladding interface and reach a value of 60 ksi at the end of the transient. The tensile stress decreases very rapidly through the thickness of the base metal and becomes insignificant at about two inches from the inside surface. In order to lower residual stresses, a post-weld heat treatment is performed by uniformly heating the vessel to 11000F, holding at that temperature for a specified period of time and then cooling slowly. The analysis shows that after this treatment, the peak stresses in the base metal decrease from 60 ksi to 32 ksi, while the stress in the cladding does not change significantly

  9. Weld overlay cladding with iron aluminides

    Energy Technology Data Exchange (ETDEWEB)

    Goodwin, G.M. [Oak Ridge National Lab., TN (United States)

    1997-12-01

    The author has established a range of compositions for these alloys within which hot cracking resistance is very good, and within which cold cracking can be avoided in many instances by careful control of welding conditions, particularly preheat and postweld heat treatment. For example, crack-free butt welds have been produced for the first time in 12-mm thick wrought Fe{sub 3}Al plate. Cold cracking, however, still remains an issue in many cases. The author has developed a commercial source for composite weld filler metals spanning a wide range of achievable aluminum levels, and are pursuing the application of these filler metals in a variety of industrial environments. Welding techniques have been developed for both the gas tungsten arc and gas metal arc processes, and preliminary work has been done to utilize the wire arc process for coating of boiler tubes. Clad specimens have been prepared for environmental testing in-house, and a number of components have been modified and placed in service in operating kraft recovery boilers. In collaboration with a commercial producer of spiral weld overlay tubing, the author is attempting to utilize the new filler metals for this novel application.

  10. Factors controlling hydrogen cracking during cladding of nuclear vessel steels

    International Nuclear Information System (INIS)

    During cladding of low alloy steels in nuclear pressure vessels for corrosion resistance, a potential problem exists of underclad hydrogen cracking. Research was undertaken to gain a better insight into the factors controlling underclad hydrogen cracking during cladding A508 Cl 3 nuclear vessel steels and to ensure the continued development of safe welding procedures in this critical application. The project was divided into three experimental phases. Phase I studied the potential and deposit hydrogen levels in Type 309 austenitic stainless steel and Ni alloy consumables and weld metals. Phase II incorporated implant testing of the A508 Cl 3 base material. A large test panel was fabricated in Phase III to approach the conditions of restraint and heat sink that are present in the pressure vessel cladding operation, but not necessarily those of the most critical components, such as nozzles where the cylindrical geometry may increase the overall restraint. The A508 Cl 3 test material was electron beam welded into the center of the test block which was then submerged arc-strip clad using very severe welding conditions in an attempt to generate underclad hydrogen cracks. It was found that for the shielded metal-arc welding (SMAW) and submerged arc welding (SAW) processes, deposit hydrogen levels were primarily controlled by flux moisture content. With single layer deposition, the implant test did not show evidence of the influence of segregation on cold cracking. All SMAW implant tests, without preheat and regardless of consumable, gave lower critical stress thresholds below about 51 ksi. A preheat of 150 deg.C increased this threshold to 80 ksi with Type 306 consumables. Even under welding conditions favorable for cracking, underclad hydrogen cracks could not be developed in a large-scale simulation of a cladding operation, indicating that very high total system restraint is needed to induce cracking

  11. Clad plates for construction of apparatus

    International Nuclear Information System (INIS)

    Importance of clad plates on the field of the construction of apparatus for the chemistry and petrol chemistry. Description of a cladding process to bond permanently and integrally ferritic steels and corrosion resistant and heat resistant materials by rolling. Information on available combinations of materials and gauge as well as on indispensable requirements to be met by the quality of the material. Results of tests carried out on the bond. Distribution of the elements between the clad and the base material. Bond properties, corrosion behaviour, toughness values and tensile properties of clad plates, heat treatment, cutting and welding of clad plates. Demonstration of applications. (orig.)

  12. Fuel cladding tubes and manufacture thereof

    International Nuclear Information System (INIS)

    Purpose: To enable smooth contaction between fuel pellets and cladding tubes, as well as prevent chemical reaction for the fission products released from the pellets. Method: The inner surface of a cladding tube is coated with a copper film and further provided thereover with a graphite film. The graphite film is formed through electrophoretic coating as follows: A cladding tube is supported rotatably in an electrophoretic coating tank containing coating solution incorporated with graphite powder and connected to an anode. A cathode is attached to the inside of the cladding tube. Coating current is supplied while rotating the cladding tube and the graphite film is formed through electrophoresis. (Ikeda, J.)

  13. Study of laser cladding nuclear valve parts

    International Nuclear Information System (INIS)

    The mechanism of laser cladding is discussed by using heat transfer model of laser cladding, heat conduction model of laser cladding and convective transfer mass model of laser melt-pool. Subsequently the laser cladding speed limit and the influence of laser cladding parameters on cladding layer structure is analyzed. A 5 kW with CO2 transverse flow is used in the research for cladding treatment of sealing surface of stop valve parts of nuclear power stations. The laser cladding layer is found to be 3.0 mm thick. The cladding surface is smooth and has no such defects as crack, gas pore, etc. A series of comparisons with plasma spurt welding and arc bead welding has been performed. The results show that there are higher grain grade and hardness, lower dilution and better performances of resistance to abrasion, wear and of anti-erosion in the laser cladding layer. The new technology of laser cladding can obviously improve the quality of nuclear valve parts. Consequently it is possible to lengthen the service life of nuclear valve and to raise the safety and reliability of the production system

  14. Technical basis for storage of Zircaloy-clad spent fuel in inert gases

    International Nuclear Information System (INIS)

    The technical bases to establish safe conditions for dry storage of Zircaloy-clad fuel are summarized. Dry storage of fuel with zirconium alloy cladding has been licensed in Canada, the Federal Republic of Germany, and Switzerland. Dry storage demonstrations, hot cell tests, and modeling have been conducted using Zircaloy-clad fuel. The demonstrations have included irradiated boiling water reactor, pressurized heavy-water reactor, and pressurized water reactor fuel assemblies. Irradiated fuel has been emplaced in and retrieved from metal casks, dry wells, silos, and a vault. Dry storage tests and demonstrations have involved about 15,000 fuel rods, and about 5600 rods have been monitored during dry storage in inert gases with maximum cladding temperatures ranging from 50 to 5700C. Although some tests and demonstrations are still in progress, there is currently no evidence that any rods exposed to inert gases have failed (one PWR rod exposed to an air cover gas failed at about 2700C). Based on this favorable experience, it is concluded that there is sufficient information on fuel rod behavior, storage conditions, and potential cladding failure mechanisms to support licensing of dry storage in the US. This licensing position includes a requirement for inert cover gases and a maximum cladding temperature guideline of 3800C for Zircaloy-clad fuel. Using an inert cover gas assures that even if fuel with cladding defects were placed in dry storage, or if defects develop during storage, the defects would not propagate. Tests and demonstrations involving Zircaloy-clad rods and assemblies with maximum cladding temperatures above 4000C are in progress. When the results from these tests have been evaluated, the viability of higher temperature limits should be examined. Acceptable conditions for storage in air and dry storage of consolidated fuel are issues yet to be resolved

  15. Pressurized water reactor fuel performance problems connected with fuel cladding corrosion processes

    International Nuclear Information System (INIS)

    environment. Our interpretation of the sub-cooled nucleate boiling impact on the PWR fuel cladding's chemical environment explains the most important change of the fuel cladding's environmental characteristics - the creation of oxidizing conditions. This conclusion requires new criteria for corrosion resistance properties of fuel cladding materials. These materials must have high corrosion resistance by oxidizing conditions. The presence of oxidizing conditions in PWR fuel cladding's environment explains also in practice observed fact that the nickel oxide formation dominates over the metallic nickel in corrosion product depositions on fuel elements' cladding. Our conclusions offer a logically explanation of prime root cause for Axial Offset Anomaly in US Westinghouse NPPs - the lower corrosion resistance of Zircaloy-4 cladding materials by oxidizing conditions occurring by sub-cooled nucleate boiling. (authors)

  16. Production and inspection of zircaloy fuel cladding tubes

    International Nuclear Information System (INIS)

    Zircaloy fuel cladding tubes are used for light and heavy water reactors. The tubes are basically produced in accordance with the ASTM B353 ''Standard specification for wrought zirconium and zirconium alloy seamless and welded tubes for nuclear service''. The production procedure for the zircaloy tubes is composed of consumable electrode are melting, forging, heat treatment, extruding, cold rolling, annealing, final roll reform and surface grinding. Concerning these producing procedure, the key points of each process relating to the material characteristics and the producing machines are presented. Next, the inspection of zircaloy fuel cladding tubes is outlined. The inspection standard of ASTM B 353-77 is tabulated as an example. Ultrasonic inspection and surface visual inspection as the flaw inspection methods, the dimensional inspection by ultrasonic pulse method for measuring the diameter and the wall thickness, electric and air micrometers for measuring the inner and outer diameters, and the ultrasonic resonance method for measuring the wall thickness, and the straightness inspection of tubes using a surface plate are explained. The mechanical tests for the zircaloy cladding tubes, such as the tensile test and the burst test, are described. The metal structure test, the corrosion test and the chemical analysis are outlined, and the characteristics of zircaloy cladding tubes for BWRs and PWRs are tabulated. (Nakai, Y.)

  17. Direct Laser Cladding , Current Status and Future Scope of Application

    Science.gov (United States)

    Weisheit, A.; Gasser, A.; Backes, G.; Jambor, T.; Pirch, N.; Wissenbach, K.

    During the last decades Direct Laser Cladding has become an established technique in many industrial fields for applying wear and corrosion protection layers on metallic surfaces as well as for the repair of high value-added components. The most important application fields are die and tool making, turbine components for aero engines and power generation, machine components such as axes and gears, and oil drilling components. Continuous wave (CW) lasers with a power up to 18 kW are used on automated machines with three or more axes, enabling 3D cladding . The outstanding feature of DLC is the high precision which leads to a minimum heat input into the work piece and a very low distortion. Due to the high cooling rates a fine grained microstructure is achieved during solidification. A new development in laser cladding is micro cladding in a size range below 50 \\upmum especially for electronic and medical applications. Furthermore, additive manufacturing is coming again into focus as a clean and resource-efficient method to manufacture and modify functional prototypes as well as unique and small lot parts.

  18. Production and quality control of fuel cladding tubes for LWRs

    International Nuclear Information System (INIS)

    This paper reviews the recent fabrication technology and corrosion resistance study of fuel cladding tubes for LWRs conducted by Sumitomo Metal Industries Ltd. started the research on zircaloy in 1957. In 1980, the factory exclusively for the production of cladding tubes was founded, and the mass production system on full scale was established. Thereafter, the various improvement of the production technology, the development of new products, and the heightening of the performance mainly on the corrosion resistance have been tested and studied. Recently, the works in the production processes were almost automated, and the installation of the production lines advanced, and the stabilization of product quality and the rationalization of costs are promoted. Moreover, the development of the zircaloy cladding tubes having high corrosion resistance has been advanced to cope with the long term cycle operation of LWRs hereafter. The features of zircaloy cladding tubes, the manufacturing processes, the improvement of the manufacturing technology, the improvement of the corrosion resistance and so on are reported. (K.I.)

  19. Development of Diffusion barrier coatings and Deposition Technologies for Mitigating Fuel Cladding Chemical Interactions (FCCI)

    Energy Technology Data Exchange (ETDEWEB)

    Sridharan, Kumar; Allen, Todd; Cole, James

    2013-02-27

    The goal of this project is to develop diffusion barrier coatings on the inner cladding surface to mitigate fuel-cladding chemical interaction (FCCI). FCCI occurs due to thermal and radiation enhanced inter-diffusion between the cladding and fuel materials, and can have the detrimental effects of reducing the effective cladding wall thickness and lowering the melting points of the fuel and cladding. The research is aimed at the Advanced Burner Reactor (ABR), a sodium-cooled fast reactor, in which higher burn-ups will exacerbate the FCCI problem. This project will study both diffusion barrier coating materials and deposition technologies. Researchers will investigate pure vanadium, zirconium, and titanium metals, along with their respective oxides, on substrates of HT-9, T91, and oxide dispersion-strengthened (ODS) steels; these materials are leading candidates for ABR fuel cladding. To test the efficacy of the coating materials, the research team will perform high-temperature diffusion couple studies using both a prototypic metallic uranium fuel and a surrogate the rare-earth element lanthanum. Ion irradiation experiments will test the stability of the coating and the coating-cladding interface. A critical technological challenge is the ability to deposit uniform coatings on the inner surface of cladding. The team will develop a promising non-line-of-sight approach that uses nanofluids . Recent research has shown the feasibility of this simple yet novel approach to deposit coatings on test flats and inside small sections of claddings. Two approaches will be investigated: 1) modified electrophoretic deposition (MEPD) and 2) boiling nanofluids. The coatings will be evaluated in the as-deposited condition and after sintering.

  20. CLAD DEGRADATION - FEPS SCREENING ARGUMENTS

    Energy Technology Data Exchange (ETDEWEB)

    R. Schreiner

    2004-10-21

    The purpose of this report is to evaluate and document the screening of the clad degradation features, events, and processes (FEPs) with respect to modeling used to support the Total System Performance Assessment-License Application (TSPA-LA). This report also addresses the effect of certain FEPs on both the cladding and the commercial spent nuclear fuel (CSNF), DOE-owned spent nuclear fuel (DSNF), and defense high-level waste (DHLW) waste forms, as appropriate to address the effects on multiple materials and both components (FEPs 2.1.09.09.0A, 2.1.09.11.0A, 2.1.11.05.0A, 2.1.12.02.0A, and 2.1.12.03.0A). These FEPs are expected to affect the repository performance during the postclosure regulatory period of 10,000 years after permanent closure. Table 1-1 provides the list of cladding FEPs, including their screening decisions (include or exclude). The primary purpose of this report is to identify and document the analysis, screening decision, and TSPA-LA disposition (for included FEPs) or screening argument (for excluded FEPs) for these FEPs related to clad degradation. In some cases, where a FEP covers multiple technical areas and is shared with other FEP reports, this report may provide only a partial technical basis for the screening of the FEP. The full technical basis for shared FEPs is addressed collectively by the sharing FEP reports. The screening decisions and associated TSPA-LA dispositions or screening arguments from all of the FEP reports are cataloged in a project-specific FEPs database.

  1. CLAD DEGRADATION - FEPS SCREENING ARGUMENTS

    International Nuclear Information System (INIS)

    The purpose of this report is to evaluate and document the screening of the clad degradation features, events, and processes (FEPs) with respect to modeling used to support the Total System Performance Assessment-License Application (TSPA-LA). This report also addresses the effect of certain FEPs on both the cladding and the commercial spent nuclear fuel (CSNF), DOE-owned spent nuclear fuel (DSNF), and defense high-level waste (DHLW) waste forms, as appropriate to address the effects on multiple materials and both components (FEPs 2.1.09.09.0A, 2.1.09.11.0A, 2.1.11.05.0A, 2.1.12.02.0A, and 2.1.12.03.0A). These FEPs are expected to affect the repository performance during the postclosure regulatory period of 10,000 years after permanent closure. Table 1-1 provides the list of cladding FEPs, including their screening decisions (include or exclude). The primary purpose of this report is to identify and document the analysis, screening decision, and TSPA-LA disposition (for included FEPs) or screening argument (for excluded FEPs) for these FEPs related to clad degradation. In some cases, where a FEP covers multiple technical areas and is shared with other FEP reports, this report may provide only a partial technical basis for the screening of the FEP. The full technical basis for shared FEPs is addressed collectively by the sharing FEP reports. The screening decisions and associated TSPA-LA dispositions or screening arguments from all of the FEP reports are cataloged in a project-specific FEPs database

  2. Fracture Toughness Of Zircaloy Claddings

    International Nuclear Information System (INIS)

    Zirconium-based alloys (Zircaloy) have been used as cladding material in Light Water Reactors for many years. During fabrication, or in in-reactor service, crack-type defects can be formed, posing questions regarding mechanical integrity. As claddings change their mechanical properties (mainly toughness) during service as a result of irradiation-induced degradation, oxidation and hydride formation, it is essential for integrity considerations to provide parameters for the assessment of the influence of flaws on rupture behaviour. Usually, fracture-mechanics parameters are employed such as the fracture toughness, KIC, or, for high plastic strains, the J-integral, JIC. The applicability of these parameters is, however, limited by the dimensions of the samples (e.g. thickness). In claddings with a wall thickness of below 1 mm, determination of toughness necessitates an extension of the J-integral concept. A method based on the traditional J-approach, but applicable to thin-walled structures, is presented in this paper. (author)

  3. Fuel behavior in severe accidents and Mo-alloy based cladding designs to improve accident tolerance

    International Nuclear Information System (INIS)

    The severe accidents at TMI-2 and Fukushima-Daiichi led to core meltdown and hydrogen explosions. The main source of energy causing core melting is the decay heat from β-, β+, and γ decays of short-lived isotopes following a power scram. The exothermic reaction of Zr-alloy cladding can further increase the cladding temperature leading to rapid cladding corrosion and hydrogen production. The most effective mitigation to minimize core damage in a severe accident is to extend the duration of heat removal capacity via battery-supported passive cooling for as long as practically possible. Replacing the Zr-alloy cladding with a higher heat resistant cladding with lower enthalpy release rate may also provide additional coping time for accident management. Such a heat resistant cladding may also overcome the current licensing concerns about Zr-alloy hydriding and post quench ductility issues in a design base loss of coolant accident (LOCA). Zr-alloy cladding, while has been optimized for normal operation in high pressure water and steam of light water reactors, will rapidly lose its corrosion resistance and tensile and creep strength in high pressure steam. Evaluation of alternate cladding materials and designs have been performed to search for a new fuel cladding design which will substantially improve the safety margins at elevated temperatures during a severe accident, while maintaining the excellent fuel performance attributes of the current Zr-alloy cladding. The screening criteria for the evaluation include neutronic properties, material availability, adaptability and operability in current LWRs, resistance to melting. The new designs also need to be fabricable, maintain sufficient strength and resist to attack by high pressure steam. Engineering metals, alloys and ceramics which can meet some or most of these requirements are limited. Following review of the properties of potential candidates, it is concluded that molybdenum alloys may potentially achieve the

  4. Theory of the frictional interaction between nuclear fuel cladding and a cracked ceramic pellet

    International Nuclear Information System (INIS)

    A summary is presented of the outcome of theoretical work detailed in five publications, reproduced as appendices, which is concerned with the tendency for the cladding tube of nuclear fuel elements to fracture as the result of power cycling or after a sudden upward power excursion. The relationship is shown between the properties of the clad, those of UO2 pellets, and the tendency of the clad to fail during upward power excursions. The role of interfacial friction is explored and the benefit to be obtained by reducing it is calculated for cases where a soft metal interlayer is present. It is shown that the experimentally-confirmed magnitude of the strain-concentration in the arc of cladding over a radial pellet crack could not arise if there were interfaceons present. Accordingly, these defects, although they do occur in some sliding situations, are thought to be absent from the pellet clas interface in fuel pins. (author)

  5. The state-of-the-art laser bio-cladding technology

    Science.gov (United States)

    Liu, Jichang; Fuh, J. Y. H.; Lü, L.

    2010-11-01

    The current state and future trend of laser bio-cladding technology are discussed. Laser bio-cladding is used in implants including fabrication of metal scaffolds and bio-coating on the scaffolds. Scaffolds have been fabricated from stainless steel, Co-based alloy or Ti alloy using laser cladding, and new laser-deposited Ti alloys have been developed. Calcium phosphate bioceramic coatings have been deposited on scaffolds with laser to improve the wear resistence and corrosion resistence of implants and to induce bone regeneration. The types of biomaterial devices currently available in the market include replacement heart valve prosthesis, dental implants, hip/knee implants, catheters, pacemakers, oxygenators and vascular grafts. Laser bio-cladding process is attracting more and more attentions of people.

  6. Advanced Fuels Campaign Cladding & Coatings Meeting Summary

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2013-03-01

    The Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) organized a Cladding and Coatings operational meeting February 12-13, 2013, at Oak Ridge National Laboratory (ORNL). Representatives from the U.S. Department of Energy (DOE), national laboratories, industry, and universities attended the two-day meeting. The purpose of the meeting was to discuss advanced cladding and cladding coating research and development (R&D); review experimental testing capabilities for assessing accident tolerant fuels; and review industry/university plans and experience in light water reactor (LWR) cladding and coating R&D.

  7. Ultrasonic monitoring of material processing using clad buffer rod sensors

    Science.gov (United States)

    Ramos Franca, Demartonne

    Ultrasonic sensors and techniques are developed for in-line monitoring of polymer extrusion, cleanliness of molten metals and liquid flow speed at elevated temperature. Pulse-echo mode is used for the first two processes, while the through-transmission mode is applied in the third one. The ultrasonic probe consists of high performance clad buffer rods with different dimensions to thermally isolate the commercial ultrasonic transducer from materials at high temperature. The clad buffer rods are made of steel, polymer and ceramic. Steel clad buffer rods are introduced for in-line monitoring of polymer extrusion processes. Owing to its superior performance in pulse-echo mode, for the first time such a probe is installed and performs ultrasonic monitoring in the die of a co-extrusion machine and in the barrel section of a twin-screw extruder. It can reveal a variety of information relevant to process parameters, such as polymer layer thickness, interface location and adhesion quality, stability, or polymer composition change. For the ultrasonic monitoring of polymer processes, probes with acoustic impedance that matches that of the processed polymer may offer certain advantages such as quantitative viscoelastic evaluation; thus high temperature polymer clad buffer rods, in particular PEEK, are developed. It is demonstrated that this new probe exhibits unique advantages for in-line monitoring of the cure of epoxies and polymer extrusion process. Long steel clad buffer rods with a spherical focus lens machined at the probing end are proposed for cleanliness evaluation of molten metals. The potential of this focusing probe is demonstrated by means of high-resolution imaging and particles detection in molten zinc at temperatures higher than 600°C, using a single probe operated at pulse-echo mode. A contrapropagating ultrasonic flowmeter employing steel clad buffer rods is devised to operate at high temperature. It is demonstrated that these rods guide ultrasonic signals

  8. Pulsed Magnetic Welding for Advanced Core and Cladding Steel

    Energy Technology Data Exchange (ETDEWEB)

    Cao, Guoping [Univ. of Wisconsin, Madison, WI (United States); Yang, Yong [Univ. of Florida, Gainesville, FL (United States)

    2013-12-19

    To investigate a solid-state joining method, pulsed magnetic welding (PMW), for welding the advanced core and cladding steels to be used in Generation IV systems, with a specific application for fuel pin end-plug welding. As another alternative solid state welding technique, pulsed magnetic welding (PMW) has not been extensively explored on the advanced steels. The resultant weld can be free from microstructure defects (pores, non-metallic inclusions, segregation of alloying elements). More specifically, the following objectives are to be achieved: 1. To design a suitable welding apparatus fixture, and optimize welding parameters for repeatable and acceptable joining of the fuel pin end-plug. The welding will be evaluated using tensile tests for lap joint weldments and helium leak tests for the fuel pin end-plug; 2 Investigate the microstructural and mechanical properties changes in PMW weldments of proposed advanced core and cladding alloys; 3. Simulate the irradiation effects on the PWM weldments using ion irradiation.

  9. Cladding Alloys for Fluoride Salt Compatibility

    Energy Technology Data Exchange (ETDEWEB)

    Muralidharan, Govindarajan [ORNL; Wilson, Dane F [ORNL; Walker, Larry R [ORNL; Santella, Michael L [ORNL; Holcomb, David Eugene [ORNL

    2011-06-01

    This report provides an overview of several candidate technologies for cladding nickel-based corrosion protection layers onto high-temperature structural alloys. The report also provides a brief overview of the welding and weld performance issues associated with joining nickel-clad nickel-based alloys. From the available techniques, two cladding technologies were selected for initial evaluation. The first technique is a line-of-sight method that would be useful for cladding large structures such as vessel interiors or large piping. The line-of-sight method is a laser-based surface cladding technique in which a high-purity nickel powder mixed into a polymer binder is first sprayed onto the surface, baked, and then rapidly melted using a high-power laser. The second technique is a vapor phase technique based on the nickel-carbonyl process that is suitable for cladding inaccessible surfaces such as the interior surfaces of heat exchangers. An initial evaluation for performed on the quality of nickel claddings processed using the two selected cladding techniques.

  10. Development of advanced zirconium fuel cladding

    International Nuclear Information System (INIS)

    This report includes the manufacturing technology developed for HANATM claddings, a series of their characterization results as well as the results of their in-pile and out-of pile performances tests which were carried out to develop some fuel claddings for a high burn-up (70,000MWd/mtU) which are competitive in the world market. Some of the HANATM claddings, which had been manufactured based on the results from the 1st and 2nd phases of the project, have been tested in a research reactor in Halden of Norway for an in-pile performance qualification. The results of the in-pile test showed that the performance of the HANATM claddings for corrosion and creep was better than 50% compared to that of Zircaloy-4 or A cladding. It was also found that the out-of pile performance of the HANATM claddings for such as LOCA and RIA in some accident conditions corrosion creep, tensile, burst and fatigue was superior or equivalent to that of the Zircaloy-4 or A cladding. The project also produced the other many data which were required to get a license for an in-pile test of HANATM claddings in a commercial reactor. The data for the qualification or characterization were provided for KNFC to assist their activities to get the license for the in-pile test of HANATM Lead Test Rods(LTR) in a commercial reactor

  11. Constraint effects of clad on underclad crack

    International Nuclear Information System (INIS)

    The finite element method is applied to two-dimensional elastic-plastic analyses for underclad crack problems. The analyses are performed for rectangular specimens with an underclad crack, which are composed of A533B class 1 steel and a clad material, to obtain the fracture mechanics parameter J-integral and the stress distribution ahead of a crack tip. The Q-factor proposed by O'Dowd and Shih is calculated from the stress distribution ahead of a crack tip, and the constraint effect of a crack tip due to a clad material or the effect of a clad material on the fracture toughness of a base material is discussed in terms of the Q-factor. Clad thickness, crack length and the material property of a clad material are varied to examine their effects

  12. FY 2014 Status Report: of Vibration Testing of Clad Fuel (M4FT-14OR0805033)

    Energy Technology Data Exchange (ETDEWEB)

    Bevard, Bruce Balkcom [ORNL

    2014-03-28

    The DOE Used Fuel Disposition Campaign (UFDC) tasked Oak Ridge National Laboratory (ORNL) to investigate the behavior of light-water-reactor (LWR) fuel cladding material performance related to extended storage and transportation of UNF. ORNL has been tasked to perform a systematic study on UNF integrity under simulated normal conditions of transportation (NCT) by using the recently developed hot-cell testing equipment, Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT). To support the testing on actual high-burnup UNF, fast-neutron irradiation of pre-hydrided zirconium-alloy cladding in the High Flux Isotope Reactor (HFIR) at elevated temperatures will be used to simulate the effects of high-burnup on fuel cladding for help in understanding the cladding materials properties relevant to extended storage and subsequent transportation. The irradiated pre-hydrided metallic materials testing will generate baseline data to benchmark hot-cell testing of the actual high-burnup UNF cladding. More importantly, the HFIR-irradiated samples will be free of alpha contamination and can be provided to researchers who do not have hot cell facilities to handle highly contaminated high-burnup UNF cladding to support their research projects for the UFDC.

  13. Calculation of hydrogen and oxygen uptake in fuel rod cladding during severe accidents using the integral diffusion method -- Final Design Report

    International Nuclear Information System (INIS)

    Final designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. A description is given of the implementation of the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5/MOD3.3 code

  14. Calculation of hydrogen and oxygen uptake in fuel rod cladding during severe accidents using the integral diffusion method -- Final Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Siefken, L.J.

    1999-05-01

    Final designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. A description is given of the implementation of the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5/MOD3.3 code.

  15. Calculation of Hydrogen and Oxygen Uptake in Fuel Rod Cladding During Severe Accidents Using the Integral Diffusion Method - Final Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Siefken, Larry James

    1999-06-01

    Final designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. A description is given of the implementation of the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5/MOD3.3 code.

  16. Quality of bimetal Al-Cu joint after explosive cladding

    Directory of Open Access Journals (Sweden)

    S. Berski

    2007-05-01

    Full Text Available Purpose: An analysis of quality of bimetallic joint between aluminium and copper layers of billet for extrusion process is the subject of the work.Design/methodology/approach: For preparing the quality analysis of particular layer of bimetal, the shearing test were done. During the tests the maximal stress for particular sets of the bimetal was established. For chosen cases the metallographic research of Al-Cu joint were done.Findings: The geometry of the cylindrical set and explosive cladding process parameters which allow to obtain the cylindrical bimetallic billets without cracks and delaminations and also with uniform cladded layer along and across the billet.Research limitations/implications: The analysis is concerning the explosive joint of pure aluminium Al995 and electrolytic copper M1E 99,97 in cylindrical sets. In the future research the analysis of this kind of joints after direct extrusion process is planning.Practical implications: The analysis could be helpful for more effective designing of the bimetal billets through the explosive cladding process and next for the plastic working processes.Originality/value: During the extrusion process with high value of extrusion ratio the delamination of the billet layers especially on the boundary of the layers is observed, this fact causes that joint after the metal working processes has lower strength even than components of the bimetal billet. So very important task is establishing the geometry set and explosive parameters to obtain the best quality of the joint.

  17. Metallography of pitted aluminum-clad, depleted uranium fuel

    International Nuclear Information System (INIS)

    The storage of aluminum-clad fuel and target materials in the L-Disassembly Basin at the Savannah River Site for more than 5 years has resulted in extensive pitting corrosion of these materials. In many cases the pitting corrosion of the aluminum clad has penetrated in the uranium metal core, resulting in the release of plutonium, uranium, cesium-137, and other fission product activity to the basin water. In an effort to characterize the extent of corrosion of the Mark 31A target slugs, two unirradiated slug assemblies were removed from basin storage and sent to the Savannah River Technology Center for evaluation. This paper presents the results of the metallography and photographic documentation of this evaluation. The metallography confirmed that pitting depths varied, with the deepest pit found to be about 0.12 inches (3.05 nun). Less than 2% of the aluminum cladding was found to be breached resulting in less than 5% of the uranium surface area being affected by corrosion. The overall integrity of the target slug remained intact

  18. Overlaid and rolled clad steel using high efficiency welding technique

    International Nuclear Information System (INIS)

    Stainless steel clad steel plates have been used widely as the economical material with excellent corrosion resistance, but since the environment of their use has become severe, the higher reliability of joining has been demanded. In these clad steel plates, generally the thickness of coating more than 2 mm is used, but in the structures subjected to mild corrosion, the plates with coating thinner than 1 mm are demanded. The overlaying and rolling method reported in this paper was developed to meet such demand, and by the use of super-wide electrodes, the process is efficient and economical. The thickness of coating can be controlled freely to 10 μm or more. The coating and parent materials are fused perfectly by overlaying welding, accordingly, the reliability of joining is very high. In this method, the coating material is overlaid by Maglay welding method on the surface of a slab, then it is hot-rolled. The features and principle of the Maglay welding method, and the properties of welded metal are reported. SUS 309 L was overlaid on a SM 41A slab of 90 mm thickness by two layers of 5 mm, and the slab was rolled to 18 mm. As the result, the mechanical properties and microstructure were satisfactory. This method can be applied to more complex forms other than plates. The mechanical properties of the welded joints of this clad plates were also examined. (Kako, I.)

  19. A finite element elastic-plastic analysis of residual stresses due to clad welding in reactor vessels

    International Nuclear Information System (INIS)

    Residual stresses due to the weld deposited cladding on the inside of a typical Westinghouse pressurized water reactor vessel are investigated using an axisymmetric finite element elastic-plastic analysis. At the beginning of the analysis, one bead of the weld cladding is assumed to lay on the reactor vessel wall at melting temperature (26000F), but in the solid phase, while the vessel remains at 3000F (preheat temperature). All material properties used in the calculations are taken as temperature-dependent. Temperature profiles are obtained in the cladding and base metal at several discrete time intervals. These temperature profiles are used to obtain the stress distribution for the same time intervals. Residual hoop tensile stresses of approximately 25 ksi were found to exist in the cladding. Peak tensile stresses in the hoop direction occur in the base metal near the cladding interface and reach a value of 60 ksi at the end of the transient. The tensile stress decreases very rapidly through the thickness of the base metal and becomes insignificant at about two inches from the inside surface. In order to lower residual stresses, a post-weld heat treatment is performed by uniformly heating the vessel to 11000F, holding at that temperature for a specified period of time and then cooling slowly. The analysis shows that after this treatment, the peak stresses in the base metal decrease from 60 ksi to 32 ksi, while the stress in the cladding does not change significantly. (author)

  20. A finite element elastic-plastic analysis of residual stresses due to clad welding in reactor vessels

    International Nuclear Information System (INIS)

    Residual stresses due to weld deposited cladding on the inside of a typical Westinghouse pressurized water reactor vessel are investigated using an axisymmetric finite element elastic-plastic analysis. At the beginning of the analysis, one head of the weld cladding is assumed to lie on the reactor vessel wall at melting temperature (2600degF), but in the solid phase, while the vessel remains at 300degF (preheat temperature). All material properties used in the calculations are taken as temperature-dependent. Temperature profiles are obtained in the cladding and base metal at several discrete time intervals. These temperatures profiles are used to obtain the stress distribution for the same time intervals. Residual hoop tensile stresses of approximately 25 ksi were found to exist in the cladding. Peak tensile stresses in the hoop direction occur in the base metal near the cladding interface and reach a value of 60 ksi at the end of the transient. The tensile stress decreases very rapidly through the thickness of the base metal and becomes insignificant at about two inches from the inside surface. In order to lower residual stresses, a post-weld heat treatment is performed by uniformly heating the vessel to 1100degF, holding at that temperature for a specified period of time and then cooling slowly. The analysis shows that after this treatment, the peak stresses in the base metal decrease from 60 ksi to 32 ksi, while the stress in the cladding does not change significantly. (author)

  1. Rapid manufacturing by laser sintering and laser cladding; Rapid Manufacturing durch Lasersintern und 3D-Laserstrahl-Auftragschweissen

    Energy Technology Data Exchange (ETDEWEB)

    Haferkamp, H. [Laser Zentrum Hannover e.V. (Germany); Alvensleben, F. von [Laser Zentrum Hannover e.V. (Germany); Gerken, J. [Laser Zentrum Hannover e.V. (Germany)

    1995-06-01

    Among the technologies which are under development for the direct production of metal components, the laser-supported techniques laser sintering and laser cladding offer positive expectations for industrial use. Founded on extensive work in the field of laser cladding of functional layers [1,2], results have been gathered at the Laser Zentrum Hannover (LZH) concerning the direct manufacturing of metal parts by laser supported techniques [3,4]. The different processes and first results concerning the build-up of metal parts mainly by laser sintering are described in this paper. During the investigation, the suitability of metals such as copper, nickel, aluminium and aluminium-bronze alloy for laser sintering without binders was tested. In addition, metal parts produced by laser cladding and a possibility of process monitoring are shown. For more details see 5 Extended Abstract. (orig.)

  2. TEC – Thin Environmental Cladding

    Directory of Open Access Journals (Sweden)

    Alan Tomasi

    2015-05-01

    Full Text Available Permasteelisa Group developed with Fiberline Composites a new curtain wall system (Thin Environmental Cladding or TEC, making use of pultruded GFRP (Glass Fiber Reinforced Polymer material instead of traditional aluminum. Main advantages using GFRP instead of aluminum are the increased thermal performance and the limited environmental impact. Selling point of the selected GFRP resin is the light transmission, which results in pultruded profiles that allow the visible light to pass through them, creating great aesthetical effects. However, GFRP components present also weaknesses, such as high acoustic transmittance (due to the reduced weight and anisotropy of the material, low stiffness if compared with aluminum (resulting in higher facade deflection and sensible fire behavior (as combustible material. This paper will describe the design of the TEC-facade, highlighting the functional role of glass within the facade concept with regards to its acoustic, structural, aesthetics and fire behavior.

  3. Technical development of double-clad process for thin strip casting of carbon steel

    Energy Technology Data Exchange (ETDEWEB)

    Brown, H.L.; Forkel, C.E.; Knudson, D.L.

    1984-08-01

    This report documents the technical development for a patent disclosure of a double-clad process for the continuous casting of thin-strip carbon steel. The fundamental idea of the disclosure is to form a product strip by depositing molten steel between two, cooled, clad strips of the same material. The claimed benefits include: (a) the conservation of energy in steel making through the elimination of soaking pits and reheat cycles, and (b) an improved surface on both sides of the as-cast product such that it will be suitable for direct feed to a cold-reduction mill. However, the process as conceived is not necessarily limited to the casting of carbon steel, but may be also applied to other metals and alloys. The work is described under three headings as follows. Preliminary Considerations and Scoping Analysis presents the basic idea of the double-clad, thin-strip casting process; the energy conservation potential; scoping heat transfer calculations for the casting process; and independent review of this work. Thermal Analysis for Roller Configuration of Double-Clad Process, presents the development, results, and independent review of a finite-element thermal analysis for the casting process as originally conceived (using only chilled rollers in direct contact with the clad material of the product strip). Further Considerations for Belt Configuration of Double-Clad Process deals with a modified equipment design which interposes two product support belts, one on each side of the product, between the clad strip and the rollers. In addition to the process description, this section presents the preliminary mechanical calculations for the endless metal belts and the work scope and results for the computer model revision and thermal analysis for the modified concept.

  4. Investigation of in-pile formed corrosion films on zircaloy fuel-rod claddings by impedance spectroscopy and galvanostatic anodization

    International Nuclear Information System (INIS)

    Hot-cell investigations have been executed to study the corrosion behaviour of irradiated Zircaloy fuel-rod claddings by impedance spectroscopy and galvanostatic anodization. The thickness of the compact oxide at the metal/oxide interface and the thickness of the minimum barrier oxide have been determined at different positions along the claddings. As shown by analysis, both quantities first increase and then decrease with increasing thickness of the total oxide. (author) 6 figs., 33 refs

  5. Method to improve the compatibility of oxide nuclear fuels with the fuel element cladding materials around them

    International Nuclear Information System (INIS)

    The compatibility of oxide nuclear fuels with their cladding material can be improved by adding getters which have a higher oxygen affinity than the cladding material. According to the invention, the getter is to be added to the nuclear fuel in very fine distribution. Powders of the metals titanium, zirconium, thorium, vanadium and niobium are named as suitable, as well as their oxides of the lower oxidation states or mixtures of the above substances. (UWI) 891 HP/UWI 892 CKA

  6. Ductile-to-brittle transition temperature for high-burnup cladding alloys exposed to simulated drying-storage conditions

    Science.gov (United States)

    Billone, M. C.; Burtseva, T. A.; Einziger, R. E.

    2013-02-01

    Structural analyses of dry casks containing high-burnup fuel require cladding mechanical properties and failure limits to assess fuel behavior. Pre-storage drying-transfer operations and early stage storage subject cladding to higher temperatures and much higher pressure-induced tensile hoop stresses relative to in-reactor operation and pool storage. Under these conditions, radial hydrides may precipitate during slow cooling and provide an additional embrittlement mechanism as the cladding temperature decreases below the ductile-to-brittle transition temperature (DBTT). A test procedure was developed to simulate the effects of drying-storage temperature histories. Following drying-storage simulation, samples were subjected to ring-compression test (RCT) loading, which was used as a ductility screening test and to simulate pinch-type loading that may occur during cask transport. RCT samples with 50% wall cracking were assessed as brittle. Prior to testing high-burnup cladding, many tests were conducted with pre-hydrided Zircaloy-4 (Zry-4) and ZIRLO™ to determine target 400 °C hoop stresses for high-burnup rodlets. Zry-4 cladding segments, from a 67-GWd/MTU fuel rod, with 520-620 wppm hydrogen and ZIRLO™ cladding segments from a 70-GWd/MTU fuel rod, with 350-650 wppm hydrogen were defueled and tested. Following drying-storage simulation, the extent of radial-hydride precipitation was characterized by the radial-hydride continuity factor. It was found that the DBTT was dependent on: cladding material, irradiation conditions, and drying-storage histories (stress at maximum temperature). High-burnup ZIRLO™ exhibited higher susceptible to radial-hydride formation and embrittlement than high-burnup Zry-4. It was also observed that uniformly pre-hydrided, non-irradiated cladding was not a good surrogate for high-burnup cladding because of the high density of circumferential hydrides across the wall and the high metal-matrix ductility for pre-hydrided cladding.

  7. Elastic-plastic deformation of a nuclear fuel cladding specimen under the internal pressure of a polymer pellet

    International Nuclear Information System (INIS)

    Full text: During the operation of light water reactors, corrosion results into the development of an oxide layer, on the external surface of zirconium alloy fuel cladding, and the introduction of hydrogen into the metal (Zr+2H2O→ZrO2+2H2). Initially, hydrogen is in solid solution and diffuses towards regions of low hydrogen concentration, high hydrostatic stress and low temperature. However, with increasing time of reactor operation, the hydrogen concentration may exceed its terminal solid solubility and brittle zirconium hydrides may precipitate. Indeed hydrides are present in high burn-up fuel cladding, which is therefore more susceptible to failure, depending on hydride volume fraction and existing defects in the oxide layer. The expansion due to compression (EDC) test has been developed for the study of irradiated and hydrided cladding failure, under high hoop strain rates, which are expected during a reactivity initiated accident (RIA). During this test, a piece of cladding tube is circumferentially loaded in tension due to the expansion of a polymer pellet, axially compressed inside the tube. A finite element simulation of the EDC-test is discussed. The objective of the study is: (i) to understand the deformation of the cladding, during the experiment, including the effect of cladding material properties, and (ii) to provide information, necessary for the development of failure criteria. The distributions of important field quantities with respect to the damage of the cladding are derived together with the evolution of their maximum values, during loading. It is shown that, before cladding yielding as well as after substantial plastic deformation, the radial displacement, on the external surface, and the total energy per unit volume, when appropriately normalized, vary along the cladding axis according to specific distributions, which do not depend on the level of loading. This characteristic of cladding deformation is useful for the interpretation of

  8. CALCULATION OF STRESS AND DEFORMATION IN FUEL ROD CLADDING DURING PELLET-CLADDING INTERACTION

    Directory of Open Access Journals (Sweden)

    Dávid Halabuk

    2015-12-01

    Full Text Available The elementary parts of every fuel assembly, and thus of the reactor core, are fuel rods. The main function of cladding is hermetic separation of nuclear fuel from coolant. The fuel rod works in very specific and difficult conditions, so there are high requirements on its reliability and safety. During irradiation of fuel rods, a state may occur when fuel pellet and cladding interact. This state is followed by changes of stress and deformations in the fuel cladding. The article is focused on stress and deformation analysis of fuel cladding, where two fuels are compared: a fresh one and a spent one, which is in contact with cladding. The calculations are done for 4 different shapes of fuel pellets. It is possible to evaluate which shape of fuel pellet is the most appropriate in consideration of stress and deformation forming in fuel cladding, axial dilatation of fuel, and radial temperature distribution in the fuel rod, based on the obtained results.

  9. Instrumentation. Nondestructive Examination for Verification of Canister and Cladding Integrity - FY2013 Status Update

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jones, Anthony M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pardini, Allan F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Denslow, Kayte M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Crawford, Susan L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Larche, Michael R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-09-30

    This report documents FY13 efforts for two instrumentation subtasks under storage and transportation. These instrumentation tasks relate to developing effective nondestructive evaluation (NDE) methods and techniques to (1) verify the integrity of metal canisters for the storage of used nuclear fuel (UNF) and to (2) characterize hydrogen effects in UNF cladding to facilitate safe storage and retrieval.

  10. Microstructure and abrasive wear studies of laser clad Al-Si/SiC composite coatings

    NARCIS (Netherlands)

    Anandkumar, R.; Colaco, R.; Ocelik, V.; De Hosson, J. Th. M.; Vilar, R.; Gyulai, J; Szabo, PJ

    2007-01-01

    Surface coatings of Al-Si/SiC metal-matrix composites were deposited on Al-7 wt. % Si alloy substrates by laser cladding. The microstructure of the coatings was characterized by optical microscopy, scanning electron microscopy (SEM) and X-ray diffraction (XRD). The microstructure of the coating mate

  11. CREEP STRAIN CORRELATION FOR IRRADIATED CLADDING

    International Nuclear Information System (INIS)

    In an attempt to predict the creep deformation of spent nuclear fuel cladding under the repository conditions, different correlations have been developed. One of them, which will be referred to as Murty's correlation in the following, and whose expression is given in Henningson (1998), was developed on the basis of experimental points related to unirradiated Zircaloy cladding (Henningson 1998, p. 56). The objective of this calculation is to adapt Murty's correlation to experimental points pertaining to irradiated Zircaloy cladding. The scope of the calculation is provided by the range of experimental parameters characterized by Zircaloy cladding temperature between 292 C and 420 C, hoop stress between 50 and 630 MPa, and test time extending to 8000 h. As for the burnup of the experimental samples, it ranges between 0.478 and 64 MWd/kgU (i.e., megawatt day per kilogram of uranium), but this is not a parameter of the adapted correlation

  12. GSGG edge cladding development: Final technical report

    International Nuclear Information System (INIS)

    The objectives of this project have been: (1) Investigate the possibility of chemical etching of GSGG crystal slabs to obtain increased strength. (2) Design and construct a simplified mold assembly for casting cladding glass to the edges of crystal slabs of different dimensions. (3) Conduct casting experiments to evaluate the redesigned mold assembly and to determine stresses as function of thermal expansion coefficient of cladding glass. (4) Clad larger sizes of GGG slabs as they become available. These tasks have been achieved. Chemical etching of GSGG slabs does not appear possible with any other acid than H3PO4 at temperatures above 3000C. A mold assembly has been constructed which allowed casting cladding glass around the edges of the largest GGG slabs available (10 x 20 x 160 mm) without causing breakage through the annealing step

  13. Fracture predictions in Zircaloy fuel cladding

    International Nuclear Information System (INIS)

    Predictions of the maximum initial allowable temperature required to achieve a 40-year life in dry storage are made for Zircaloy clad spent fuel. Maximum initial temperatures of 360 to 4050C for irradiated spent fuel cladding (wet pool storage) are predicted. The technique utilized in this work is based on the deformation and fracture map methodology. Maps are presented for temperatures between 50 and 8500C and stresses between 5 and 500 MPa. These maps are then combined with both the known temperature history (an exponentially decaying one) of Zircaloy fuel cladding in dry storage and a life fracture rule to predict the rupture life of the cladding in dry storage. Predictions of the deformation and fracture map methodology are shown to be in good agreement with constant stress-constant temperature data

  14. Explosive Cladding of Titanium and Aluminium Alloys on the Example of Ti6Al4V-AA2519 Joints / Wybuchowe Platerowanie Stopów Tytanu I Aluminium Na Przykładzie Połączenia Ti6Al4V-AA2519

    Directory of Open Access Journals (Sweden)

    Gałka A.

    2015-12-01

    Full Text Available Explosive cladding is currently one of the basic technologies of joining metals and their alloys. It enables manufacturing of the widest range of joints and in many cases there is no alternative solution. An example of such materials are clads that include light metals such as titanium and aluminum. ach new material combination requires an appropriate adaptation of the technology by choosing adequate explosives and tuning other cladding parameters. Technology enabling explosive cladding of Ti6Al4V titanium alloy and aluminum AA2519 was developed. The clads were tested by means of destructive and nondestructive testing, analyzing integrity, strength and quality of the obtained joint.

  15. Determination of fuel-clad thermal gap conductance by use of Kalman filter methods

    International Nuclear Information System (INIS)

    Kalman filter methodology has been applied to an in-pile liquid-metal fast breeder reactor simulation experiment to obtain estimates of the fuel-clad thermal gap conductance. A transient lumped parameter model of the experiment is developed. An optimal estimate of the state vector chosen to characterize the experiment is obtained through the use of the Kalman filter. From this estimate, the fuel-clad thermal gap conductance is calculated as a function of time into the test and axial position along the length of the fuel pin

  16. TEC – Thin Environmental Cladding

    Directory of Open Access Journals (Sweden)

    Alan Tomasi

    2014-06-01

    Full Text Available Corresponding author: Alan Tomasi, Group R&D Project Manager, Permasteelisa S.p.A., viale E. Mattei 21/23 | 31029 Vittorio Veneto, Treviso, Italy. Tel.: +39 0438 505207; E-mail: a.tomasi@permasteelisagroup.com; www.permasteelisagroup.com Permasteelisa Group developed with Fiberline Composites a new curtain wall system (Thin Environmental Cladding or TEC, making use of pultruded GFRP (Glass Fiber Reinforced Polymer material instead of traditional aluminum. Main advantages using GFRP instead of aluminum are the increased thermal performance and the limited environmental impact. Selling point of the selected GFRP resin is the light transmission, which results in pultruded profiles that allow the visible light to pass through them, creating great aesthetical effects. However, GFRP components present also weaknesses, such as high acoustic transmittance (due to the reduced weight and anisotropy of the material, low stiffness if compared with aluminum (resulting in higher facade deflection and sensible fire behavior (as combustible material. This paper will describe the design of the TEC-facade, highlighting the functional role of glass within the facade concept with regards to its acoustic, structural, aesthetics and fire behavior.

  17. Gate-last TiN/HfO2 band edge effective work functions using low-temperature anneals and selective cladding to control interface composition

    KAUST Repository

    Hinkle, C. L.

    2012-04-09

    Silicon N-metal-oxide-semiconductor (NMOS) and P-metal-oxide-semiconductor (PMOS) band edge effective work functions and the correspondingly low threshold voltages (Vt) are demonstrated using standard fab materials and processes in a gate-last scheme employing low-temperature anneals and selective cladding layers. Al diffusion from the cladding to the TiN/HfO2interface during forming gas anneal together with low O concentration in the TiN enables low NMOS Vt. The use of non-migrating W cladding along with experimentally detected N-induced dipoles, produced by increased oxygen in the TiN, facilitates low PMOS Vt.

  18. Management of zirconium rod claddings with the process of electrochemical breakdown

    International Nuclear Information System (INIS)

    One of the wastes varieties resulted from reprocessing of irradiated fuel issued from nuclear-power plant are fuel rods claddings remaining in the apparatus after fuel dissolution. The most common technique of used cladding isolation is its cementing and disposal in metallic containers. In view of the high cost of the cladding material (zirconium), there have been a number of proposals for this material to be recovered and re-used. However, the residual contamination of the claddings following the dissolution of the fuel and the chemical stability of the zirconium militate against any proposal for a recycling process that might be economically justifiable. There is, however, information to be found in the literature on the synthesis of mineral-like materials based on zirconium. The idea of using zirconium, which is contained in fuel rod claddings and irradiated fuel itself to synthesize compounds that would be suited to long-term storage or final disposal, is conceptually attractive. The object of this work was to carry out a study on process of fuel rod claddings electrochemical dissolution in nitric acid solutions. In the authors' opinion an electrochemical dissolution of fuel rods claddings is the most worth-while technique for preparation of zirconium solutions intended for ceramic matrices synthesis. The dependences of dissolution rate on nitric acid concentration, temperature and electrolyte composition are presented. It is shown that the highest sample dissolution rate was observed in solutions of dilute nitric acid in presence of calcium nitrate at higher temperatures. The results of sample surface oxide layers analysis and the composition of precipitates, forming in the course of the dissolution, are given. (author)

  19. Iodine-oxygen and cadmium-induced stress corrosion cracking of Zr-4 cladding tube

    International Nuclear Information System (INIS)

    On the basis of iodine-induced stress corrosion cracking (SCC) experiments the authors did before, iodine-oxygen and cadmium-induced SCC was studied on Zr-4 cladding tube. Specimens used in experiments are cladding tubes of a reactor fuel element made by Institute of Nonferrous Metal of China. The tube which has a length of 145 mm and an outside diameter of 15.3 mm and an inside diameter of 14.9 mm was annealed at 620 K for two hours, and then it had a fine, stress-relieved microstructure. Two end-caps were welded on the cladding tube. There was a hole of 0.8 mm diameter in a protruding melting-welding platform on one end-cap of the specimen. Before welding the end-caps, a glass ampoule filled with a certain amount of oxygen and a piece of Zr-4 material which can dash the glass ampoule were put into the cladding tube. After plug-hole welding in high pressure argon, the cladding tube was shaken in order to make the piece of Zr-4 material dash the ampoule and the oxygen fill up the space inside the cladding tube. A certain amount of iodine was charged into the cladding tube from the hole before the plug-hole welding. The plug-hole welding in high pressure argon was performed on a specially prepared equipment within 0.1-0.5 second. At a certain temperature, the pressure of argon determines the mechanical load (stress). The SCC experiments were controlled within +-3 degree C by a thermocouple welded on the specimen. The cracking of the specimen or the leak of gas was sensitively supervised and timed by vacuum alarm system. Under various conditions of stress, the experiments for 28 specimens of iodine-oxygen agent and 5 specimens of cadmium agent were undertaken

  20. Clad Degradation- Summary and Abstraction for LA

    International Nuclear Information System (INIS)

    The purpose of this model report is to develop the summary cladding degradation abstraction that will be used in the Total System Performance Assessment for the License Application (TSPA-LA). Most civilian commercial nuclear fuel is encased in Zircaloy cladding. The model addressed in this report is intended to describe the postulated condition of commercial Zircaloy-clad fuel as a function of postclosure time after it is placed in the repository. Earlier total system performance assessments analyzed the waste form as exposed UO2, which was available for degradation at the intrinsic dissolution rate. Water in the waste package quickly became saturated with many of the radionuclides, limiting their release rate. In the total system performance assessments for the Viability Assessment and the Site Recommendation, cladding was analyzed as part of the waste form, limiting the amount of fuel available at any time for degradation. The current model is divided into two stages. The first considers predisposal rod failures (most of which occur during reactor operation and associated activities) and postdisposal mechanical failure (from static loading of rocks) as mechanisms for perforating the cladding. Other fuel failure mechanisms including those caused by handling or transportation have been screened out (excluded) or are treated elsewhere. All stainless-steel-clad fuel, which makes up a small percentage of the overall amount of fuel to be stored, is modeled as failed upon placement in the waste packages. The second stage of the degradation model is the splitting of the cladding from the reaction of water or moist air and UO2. The splitting has been observed to be rapid in comparison to the total system performance assessment time steps and is modeled to be instantaneous. After the cladding splits, the rind buildup inside the cladding widens the split, increasing the diffusion area from the fuel rind to the waste package interior. This model report summarizes the

  1. Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings - phase II

    Energy Technology Data Exchange (ETDEWEB)

    Blough, J.L.; Stanko, G.J. [Foster Wheeler Development Corp., Livingston, NJ (United States)

    1996-08-01

    In Phase I a variety of developmental and commercial tubing alloys and claddings were exposed to laboratory fireside corrosion testing simulating a superheater or reheater in a coal-fired boiler. Phase II (in situ testing) has exposed samples of 347, RA-8511, HR3C, 253MA, Fe{sub 3}Al + 5Cr, 310 modified, 800HT, NF 709, 690 clad, and 671 clad for over 10,000 hours to the actual operating conditions of a 250-MW coal-fired boiler. The samples were installed on an air-cooled, retractable corrosion probe, installed in the reheater cavity, and controlled to the operating metal temperatures of an existing and advanced-cycle coal-fired boiler. Samples of each alloy will be exposed for 4000, 12,000, and 16,000 hours of operation. The results will be presented for the metallurgical examination of the corrosion probe samples after 4000 hours of exposure.

  2. Optimal forming zone length in continuous extrusion of lead-clad glass fiber wire

    Institute of Scientific and Technical Information of China (English)

    李霞; 唐景林; 王丽薇; 高明

    2008-01-01

    Forming zone length (FZL) is a key parameter of the lead-clad glass fiber extrusion dies, and an unsuitable FZL will lead to breakage of the glass fiber and/or unacceptable geometric and metallographic qualities of the product. The optimal FZL was determined theoretically based on a mathematical model established by upper bound method, and accepted Pb-GF wire was actually obtained experimentally by symmetric side-feed extrusion at a much lower temperature than that published before. The wire has features of fine grains, uniform diameter, good coaxiality and satisfied mechanical property. The results and conclusions obtained in the research can be used to design the forming tools for lead-clad glass fiber extrusion and have significance to further research on the extrusion of other complex wires of metal-clad brittle core.

  3. Study on deposition rate and laser energy efficiency of Laser-Induction Hybrid Cladding

    Science.gov (United States)

    Wang, DengZhi; Hu, QianWu; Zheng, YinLan; Xie, Yong; Zeng, XiaoYan

    2016-03-01

    Laser-Induction Hybrid Cladding (LIHC) was introduced to prepare metal silicide based composite coatings, and influence of different factors such as laser type, laser power, laser scan speed and induction preheating temperature on the coating deposition rate and laser energy efficiency was studied systematically. Compared with conventional CO2 laser cladding, fiber laser-induction hybrid cladding improves the coating deposition rate and laser energy efficiency by 3.7 times. When a fiber laser with laser power of 4 kW was combined with an induction preheating temperature of 850 °C, the maximum coating deposition rate and maximum laser energy efficiency reaches 71 g/min and 64% respectively.

  4. Fabrication and Lasing Property of Yb~(3+)-doped Double-Clad Fibers with Novel Inner Cladding

    Institute of Scientific and Technical Information of China (English)

    2003-01-01

    The Yb3+-doped double-clad fibers with novel inner cladding have been made by using MCVD process, solution-doping method and optical machining together. The laser power and slope efficiency of the fiber lasers are higher than 1.8W and 50% respectively.

  5. A cladding pumped Ytterbium-doped fiber laser with holey inner and outer cladding

    OpenAIRE

    Furusawa, Kentaro; Malinowski, A.N.; Price, Jonathan H.V.; Monro, Tanya M.; Jayanta K. Sahu; Nilsson, Johan; Richardson, David J

    2001-01-01

    We have fabricated an ytterbium doped all-glass double-clad large mode area holey fiber. A highly efficient cladding pumped single transverse mode holey fiber laser has been demonstrated, allowing continuous-wave output powers in excess of 1W with efficiencies of more than 80%. Furthermore both Q-switched and mode-locked operation of the laser have been demonstrated.

  6. Examination of stainless steel-clad Connecticut Yankee fuel assembly S004 after storage in borated water

    International Nuclear Information System (INIS)

    A Connecticut Yankee fuel assembly (S004) was tested nondestructively and destructively. It was concluded that no obvious degradation of the 304L stainless steel-clad spent fuel from assembly S004 occurred during 5 y of storage in borated water. Furthermore, no obvious degradation due to the pool environment occurred on 304 stainless steel-clad rods in assemblies H07 and G11, which were stored for shorter periods but contained operationally induced cladding defects. The seam welds in the cladding on fuel rods from assembly S004, H07, and G11 were similar in that they showed a wrought microstructure with grains noticeably smaller than those in the cladding base metal. The end cap welds showed a dendritically cored structure, typical of rapidly quenched austenitic weld metal. Some intergranular melting may have occurred in the heat-affected zone (HAZ) in the cladding adjacent to the end cap welds in rods from assemblies S004 and H07. However, the weld areas did not show evidence of corrosion-induced degradation

  7. Effect of annealing on two different niobium-clad stainless steel PEMFC bipolar plate materials

    Institute of Scientific and Technical Information of China (English)

    Sung-Tae HONG; Dae-Wook KIM; Yong-Joo YOU; K.Scott WEIL

    2009-01-01

    Niobium (Nb)-clad stainless steels(SS) produced via roll bonding are being considered for use in the bipolar plates of polymer electrolyte membrane fuel celI(PEMFC) stacks. Because the roll bonding process induces substantial work hardening in the constituent materials, thermal annealing is used to restore ductility to the clad sheet so that it can be subsequently blanked, stamped and dimpled in forming the final plate component. Two roll bonded materials, niobium clad 340L stainless steel (Nb/340L SS) and niobium clad 434 stainless steel (Nb/434 SS) were annealed under optimized conditions prescribed by the cladding manufacturer. Comparative mechanical testing conducted on each material before and after annealing shows significant improvement in ductility in both cases. However, corresponding microstructural analyses indicate an obvious difference between the two heat treated materials. During annealing, an interlayer with thick less than 1 μm forms between the constituent layers in the Nb/340L SS, whereas no interlayer is found in the annealed Nb/434 SS material. Prior work suggests that internal defects potentially can be generated in such an interlayer during metal forming operations. Thus, Nb/434 SS may be the preferred candidate material for this application.

  8. Stability of LMR oxide pins and blanket rods during run-beyond-cladding-break (RBCB) operation

    International Nuclear Information System (INIS)

    Since 1981, the U.S. Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan have collaborated on an operational reliability testing program in the Experimental Breeder Reactor II. The tests were designed to determine the irradiation behavior of liquid-metal reactor (LMR) oxide pins and blanket rods during steady-state, transient, and run-beyond-claddin-breach (RBCB) operation. Phase I tests completed in 1987 involved current LMR oxide designs and claddings; the phase II tests begun in 1988 concentrate on advanced LMR designs, large-diameter pins (7.5 mm), and advance cladding alloys. The cladding breaches in these tests have been readily detected by fission-gas and delayed-neutron (DN) precursor release. The condition of the fuel pin has been monitored by these releases during RBCB operation. A variety of failures have been intentionally studied in the RBCB portion of the program for operating times of up to 142 full-power days; also, several failure types have been incidentally experienced during the transient tests. Types of failure have included those induced by gas-pressure loading either naturally or by prethinning of the cladding defects, and fuel-cladding mechanical interaction (FCMI)-induced failures or secondary failures caused by the formation of low-density fuel-sodium reaction product (FSRP). This paper summarizes this experience with regard to LMR oxide fuel stability during RBCB operation

  9. Fuel cladding behavior under rapid loading conditions

    Science.gov (United States)

    Yueh, K.; Karlsson, J.; Stjärnsäter, J.; Schrire, D.; Ledergerber, G.; Munoz-Reja, C.; Hallstadius, L.

    2016-02-01

    A modified burst test (MBT) was used in an extensive test program to characterize fuel cladding failure behavior under rapid loading conditions. The MBT differs from a normal burst test with the use of a driver tube to simulate the expansion of a fuel pellet, thereby producing a partial strain driven deformation condition similar to that of a fuel pellet expansion in a reactivity insertion accident (RIA). A piston/cylinder assembly was used to pressurize the driver tube. By controlling the speed and distance the piston travels the loading rate and degree of sample deformation could be controlled. The use of a driver tube with a machined gauge section localizes deformation and allows for continuous monitoring of the test sample diameter change at the location of maximum hoop strain, during each test. Cladding samples from five irradiated fuel rods were tested between 296 and 553 K and loading rates from 1.5 to 3.5/s. The test rods included variations of Zircaloy-2 with different liners and ZIRLO, ranging in burn-up from 41 to 74 GWd/MTU. The test results show cladding ductility is strongly temperature and loading rate dependent. Zircaloy-2 cladding ductility degradation due to operational hydrogen pickup started to recover at approximately 358 K for test condition used in the study. This recovery temperature is strongly loading rate dependent. At 373 K, ductility recovery was small for loading rates less than 8 ms equivalent RIA pulse width, but longer than 8 ms the ductility recovery increased exponentially with increasing pulse width, consistent with literature observations of loading rate dependent brittle-to-ductile (BTD) transition temperature. The cladding ductility was also observed to be strongly loading rate/pulse width dependent for BWR cladding below the BTD temperature and Pressurized Water Reactor (PWR) cladding at both 296 and 553 K.

  10. Current status of the Run-Beyond-Cladding Breach (RBCB) tests for the Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    This paper describes the results from the Integral Fast Reactor (IFR) metallic fuel Run-Beyond-Cladding-Breach (RBCB) experiments conducted in the Experimental Breeder Reactor II (EBR-II). Included in the report are scoping test results and the data collected from the prototypical tests as well as the exam results and discussion from a naturally occurring breach of one of the lead IFR fuel tests. All results showed a characteristic delayed neutron and fission gas release pattern that readily allows for identification and evaluation of cladding breach events. Also, cladding breaches are very small and do not propagate during extensive post breach operation. Loss of fuel from breached cladding was found to be insignificant. The paper will conclude with a brief description of future RBCB experiments planned for irradiation in EBR-II

  11. Development of advanced claddings for suppressing the hydrogen emission in accident conditions. Development of advanced claddings for suppressing the hydrogen emission in the accident condition

    International Nuclear Information System (INIS)

    The development of accident-tolerant fuels can be a breakthrough to help solve the challenge facing nuclear fuels. One of the goals to be reached with accident-tolerant fuels is to reduce the hydrogen emission in the accident condition by improving the high-temperature oxidation resistance of claddings. KAERI launched a new project to develop the accident-tolerant fuel claddings with the primary objective to suppress the hydrogen emission even in severe accident conditions. Two concepts are now being considered as hydrogen-suppressed cladding. In concept 1, the surface modification technique was used to improve the oxidation resistance of Zr claddings. Like in concept 2, the metal-ceramic hybrid cladding which has a ceramic composite layer between the Zr inner layer and the outer surface coating is being developed. The high-temperature steam oxidation behaviour was investigated for several candidate materials for the surface modification of Zr claddings. From the oxidation tests carried out in 1 200 deg. C steam, it was found that the high-temperature steam oxidation resistance of Cr and Si was much higher than that of zircaloy-4. Al3Ti-based alloys also showed extremely low-oxidation rate compared to zircaloy-4. One important part in the surface modification is to develop the surface coating technology where the optimum process needs to be established depending on the surface layer materials. Several candidate materials were coated on the Zr alloy specimens by a laser beam scanning (LBS), a plasma spray (PS) and a PS followed by LBS and subject to the high-temperature steam oxidation test. It was found that Cr and Si coating layers were effective in protecting Zr-alloys from the oxidation. The corrosion behaviour of the candidate materials in normal reactor operation condition such as 360 deg. C water will be investigated after the screening test in the high-temperature steam. The metal-ceramic hybrid cladding consisted of three major parts; a Zr liner, a ceramic

  12. Stress corrosion testing of irradiated cladding tubes

    International Nuclear Information System (INIS)

    Samples from two fuel rods with different cladding have been stress corrosion tested by closed-end argon-iodine pressurization at 3200C. The fuel rods with stress relieved and recrystallized Zircaloy-2 had received burnups of 10.000 and 20.000 MWd/ton UO2, respectively. It was found that the SCC failure stress was unchanged or slightly higher for the irradiated than for the unirradiated control tubes. The tubes failed consistently in the end with the lowest irradiation dose. The diameter increase of the irradiated cladding during the test was 1.1% for the stress-relieved samples and 0.24% for the recrystallized samples. SEM examination revealed no major differences between irradiated and unirradiated cladding. A ''semi-ductile'' fracture zone in recrystallized material is described in some detail. (author)

  13. Residual stress and crack initiation in laser clad composite layer with Co-based alloy and WC + NiCr

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Changmin; Park, Hyungkwon; Yoo, Jaehong [Division of Materials Science and Engineering, Hanyang University, Seoul 133-791 (Korea, Republic of); Lee, Changhee, E-mail: chlee@hanyang.ac.kr [Division of Materials Science and Engineering, Hanyang University, Seoul 133-791 (Korea, Republic of); Woo, WanChuck [Neutron Science Division, Korea Atomic Energy Research Institute, Daejeon 305-353 (Korea, Republic of); Park, Sunhong [Research Institute of Industrial Science & Technology, Hyo-ja-dong, Po-Hang, Kyoung-buk, San 32 (Korea, Republic of)

    2015-08-01

    Highlights: • Major problem, clad cracking in laser cladding process, was researched. • Residual stress measurements were performed quantitatively by neutron diffraction method along the surface of specimens. • Relationship between the residual stress and crack initiation was showed clearly. • Ceramic particle effect in the metal matrix was showed from the results of residual stress measurements. • Initiation sites of generating clad cracks were specifically studied in MMC coatings. - Abstract: Although laser cladding process has been widely used to improve the wear and corrosion resistance, there are unwanted cracking issues during and/or after laser cladding. This study investigates the tendency of Co-based WC + NiCr composite layers to cracking during the laser cladding process. Residual stress distributions of the specimen are measured using neutron diffraction and elucidate the correlation between the residual stress and the cracking in three types of cylindrical specimens; (i) no cladding substrate only, (ii) cladding with 100% stellite#6, and (iii) cladding with 55% stellite#6 and 45% technolase40s. The microstructure of the clad layer was composed of Co-based dendrite and brittle eutectic phases at the dendritic boundaries. And WC particles were distributed on the matrix forming intermediate composition region by partial melting of the surface of particles. The overlaid specimen exhibited tensile residual stress, which was accumulated through the beads due to contraction of the coating layer generated by rapid solidification, while the non-clad specimen showed compressive. Also, the specimen overlaid with 55 wt% stellite#6 and 45 wt% technolase40s showed a tensile stress higher than the specimen overlaid with 100% stellite#6 possibly, due to the difference between thermal expansion coefficients of the matrix and WC particles. Such tensile stresses can be potential driving force to provide an easy crack path ways for large brittle fractures

  14. Residual stress and crack initiation in laser clad composite layer with Co-based alloy and WC + NiCr

    International Nuclear Information System (INIS)

    Highlights: • Major problem, clad cracking in laser cladding process, was researched. • Residual stress measurements were performed quantitatively by neutron diffraction method along the surface of specimens. • Relationship between the residual stress and crack initiation was showed clearly. • Ceramic particle effect in the metal matrix was showed from the results of residual stress measurements. • Initiation sites of generating clad cracks were specifically studied in MMC coatings. - Abstract: Although laser cladding process has been widely used to improve the wear and corrosion resistance, there are unwanted cracking issues during and/or after laser cladding. This study investigates the tendency of Co-based WC + NiCr composite layers to cracking during the laser cladding process. Residual stress distributions of the specimen are measured using neutron diffraction and elucidate the correlation between the residual stress and the cracking in three types of cylindrical specimens; (i) no cladding substrate only, (ii) cladding with 100% stellite#6, and (iii) cladding with 55% stellite#6 and 45% technolase40s. The microstructure of the clad layer was composed of Co-based dendrite and brittle eutectic phases at the dendritic boundaries. And WC particles were distributed on the matrix forming intermediate composition region by partial melting of the surface of particles. The overlaid specimen exhibited tensile residual stress, which was accumulated through the beads due to contraction of the coating layer generated by rapid solidification, while the non-clad specimen showed compressive. Also, the specimen overlaid with 55 wt% stellite#6 and 45 wt% technolase40s showed a tensile stress higher than the specimen overlaid with 100% stellite#6 possibly, due to the difference between thermal expansion coefficients of the matrix and WC particles. Such tensile stresses can be potential driving force to provide an easy crack path ways for large brittle fractures

  15. Calculation of hydrogen and oxygen uptake in fuel rod cladding during severe accidents using the integral diffusion method -- Preliminary design report

    International Nuclear Information System (INIS)

    Preliminary designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert's law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the proposed models on the calculated behavior of fuel rods in severe accident conditions. This report is a revision and reissue of the report entitled; ''Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents.''

  16. Calculation of hydrogen and oxygen uptake in fuel rod cladding during severe accidents using the integral diffusion method -- Preliminary design report

    Energy Technology Data Exchange (ETDEWEB)

    Siefken, L.J.

    1999-02-01

    Preliminary designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert's law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the proposed models on the calculated behavior of fuel rods in severe accident conditions. This report is a revision and reissue of the report entitled; ``Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents.''

  17. Metallurgical and mechanical behaviours of PWR fuel cladding tube oxidised at high temperature

    International Nuclear Information System (INIS)

    Zirconium alloys are used as cladding materials in Pressurized Water Reactors (PWR). As they are submitted to very extreme conditions, it is necessary to check their behaviour and especially to make sure they meet the safety criteria. They are therefore studied under typical in service-loadings but also under accidental loadings. In one of these accidental scenarios, called Loss of Coolant Accident (LOCA) the cladding temperature may increase above 800 C, in a steam environment, and decrease before a final quench of the cladding. During this temperature transient, the cladding is heavily oxidised, and the metallurgical changes lead to a decrease of the post quench mechanical properties. It is then necessary to correlate this drop in residual ductility to the metallurgical evolutions. This is the problem we want to address in this study: the oxidation of PWR cladding materials at high temperature in a steam environment and its consequences on post quench mechanical properties. As oxygen goes massively into the metallic part - a zirconia layer grows at the same time - during the high temperature oxidation, the claddings tubes microstructure shows three different phases that are the outer oxide layer (zirconia) and the inner metallic phases (α(O) and 'ex β') - with various mechanical properties. In order to reproduce the behaviour of this multilayered material, the first part of this study consisted in creating samples with different - but homogeneous in thickness - oxygen contents, similar to those observed in the different phases of the real cladding. The study was especially focused on the β-->α phase transformation upon cooling and on the resulting microstructures. A mechanism was proposed to describe this phase transformation. For instance, we conclude that for our oxygen enriched samples, the phase transformation kinetics upon cooling are ruled by the oxygen partitioning between the two allotropic phases. Then, these materials were mechanically tested at

  18. Inpile (in PWR) testing of cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, B.J.; Kim, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-04-01

    As an introduction, the reasons to perform inpile tests are depicted. An overview over general inpile test procedure is given, and test details which are necessary for the development of new alloys for high burnup claddings, like sample geometries and measuring techniques for inpile corrosion testing, are described in detail. Tests for the creep and length change behavior of cladding tubes are described briefly. Finally, conclusions are drawn and literature citations for further test details are given. (author). 9 refs., 2 tabs., 17 figs.

  19. Comparison of corrosion behavior between fusion cladded and explosive cladded Inconel 625/plain carbon steel bimetal plates

    International Nuclear Information System (INIS)

    Highlights: ► Both explosive and fusion cladding aggravate the corrosion resistance of Inconel 625. ► Fusion cladding is more detrimental to nonuniform corrosion resistance. ► Single-layered fusion coat does not show any repassivation ability. ► Adding more layers enhance the corrosion resistance of fusion cladding Inconel 625. ► High impact energy spoils the corrosion resistance of explosive cladding Inconel 625. -- Abstract: One of the main concerns in cladding Inconel 625 superalloy on desired substrates is deterioration of corrosion resistance due to cladding process. The present study aims to compare the effect of fusion cladding and explosive cladding procedures on corrosion behavior of Inconel 625 cladding on plain carbon steel as substrate. Also, an attempt has been made to investigate the role of load ratio and numbers of fusion layers in corrosion behavior of explosive and fusion cladding Inconel 625 respectively. In all cases, the cyclic polarization as an electrochemical method has been applied to assess the corrosion behavior. According to the obtained results, both cladding methods aggravate the corrosion resistance of Inconel 625. However, the fusion cladding process is more detrimental to nonuniform corrosion resistance, where the chemical nonuniformity of fusion cladding superalloy issuing from microsegregation, development of secondary phases and contamination of clad through dilution hinders formation of a stable passive layer. Moreover, it is observed that adding more fusion layers can enhance the nonuniform corrosion resistance of fusion cladding Inconel 625, though this resistance still remains weaker than explosive cladding superalloy. Also, the results indicate that raising the impact energy in explosive cladding procedure drops the corrosion resistance of Inconel 625.

  20. The Absorption Characteristics of Inhomogeneous Double-Clad Fibers

    Institute of Scientific and Technical Information of China (English)

    Hui Zhang; Zihua Wang; Zhongyin Xiao

    2003-01-01

    The absorption characteristics of radially inhomogeneous double-clad fiber (DCF) are investigated firstly with the method of caustic radius, combined with the method of WKBJ. The results are significant for double-clad optical fiber lasers and amplifiers.

  1. Hydrogen motion in Zircaloy-4 cladding during a LOCA transient

    Science.gov (United States)

    Elodie, T.; Jean, D.; Séverine, G.; M-Christine, B.; Michel, C.; Berger, P.; Martine, B.; Antoine, A.

    2016-04-01

    Hydrogen and oxygen are key elements influencing the embrittlement of zirconium-based nuclear fuel cladding during the quench phase following a Loss Of Coolant Accident (LOCA). The understanding of the mechanisms influencing the motion of these two chemical elements in the metal is required to fully describe the material embrittlement. High temperature steam oxidation tests were performed on pre-hydrided Zircaloy-4 samples with hydrogen contents ranging between 11 and 400 wppm prior to LOCA transient. Thanks to the use of both Electron Probe Micro-Analysis (EPMA) and Elastic Recoil Detection Analysis (μ-ERDA), the chemical elements partitioning has been systematically quantified inside the prior-β phase. Image analysis and metallographic examinations were combined to provide an average oxygen profile as well as hydrogen profile within the cladding thickness after LOCA transient. The measured hydrogen profile is far from homogeneous. Experimental distributions are compared to those predicted numerically using calculations derived from a finite difference thermo-diffusion code (DIFFOX) developed at IRSN.

  2. Novel Accident-Tolerant Fuel Meat and Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Robert D. Mariani; Pavel G Medvedev; Douglas L Porter; Steven L Hayes; James I. Cole; Xian-Ming Bai

    2013-09-01

    A novel accident-tolerant fuel meat and cladding are here proposed. The fuel meat design incorporates annular fuel with inserts and discs that are fabricated from a material having high thermal conductivity, for example niobium. The inserts are rods or tubes. Discs separate the fuel pellets. Using the BISON fuel performance code it was found that the peak fuel temperature can be lowered by more than 600 degrees C for one set of conditions with niobium metal as the thermal conductor. In addition to improved safety margin, several advantages are expected from the lower temperature such as decreased fission gas release and fuel cracking. Advantages and disadvantages are discussed. An enrichment of only 7.5% fully compensates the lost reactivity of the displaced UO2. Slightly higher enrichments, such as 9%, allow uprates and increased burnups to offset the initial costs for retooling. The design has applications for fast reactors and transuranic burning, which may accelerate its development. A zirconium silicide coating is also described for accident tolerant applications. A self-limiting degradation behavior for this coating is expected to produce a glassy, self-healing layer that becomes more protective at elevated temperature, with some similarities to MoSi2 and other silicides. Both the fuel and coating may benefit from the existing technology infrastructure and the associated wide expertise for a more rapid development in comparison to other, more novel fuels and cladding.

  3. Cumulative damage estimation of LMR fuel cladding under transient conditions

    International Nuclear Information System (INIS)

    Related to in-pile behaviors of LMR (Liquid Metal Reactor) fuel cladding, the objectives of this study are to establish the design criteria in probabilistic approach for the transient condition based on the data analysis and estimation. Cumulative damage is estimated in this study, which probabilistic estimation method is proposed and the probability of the effective thickness reduction of eutectic penetration depth is determined. As a result of this study, it is found that assumption of 100% cladding reduction of eutectic penetration depth is quite conservative. Basically, CDF (Cumulative Damage Fraction) calculation requires transient condition experiment data, and thus WPF (Whole Pin Furnace) data that is performed at Argon National Lab. are used. By probabilistic estimation of calculated CDF, Weibull plot of LMR transient condition is obtained for the unreliability to CDF. CDF of WPF FM5, simulating experiment of EBR-II's LOF (Loss Of Flow) accident, is 2.7657x10-2 and this value does not exceed the design limit (= 0.2) of CDF. CDF of FM5 does not exceed also the 10% probabilistic design limit, which is estimated in Weibull plot. Also, CDF under 10% probabilistic design limit is changed rapidly by Weibull shape parameter, β. Since probabilistic determination methods in CDF include the uncertainty of correlation and over-estimation tendency, it is concluded that this proposed method may have higher reliability than correlative estimation

  4. Corrosion Resistant Cladding by YAG Laser Welding in Underwater Environment

    International Nuclear Information System (INIS)

    It is known that stress-corrosion cracking (SCC) will occur in nickel-base alloys used in Reactor Pressure Vessel (RPV) and Internals of nuclear power plants. A SCC sensitivity has been evaluated by IHI in each part of RPV and Internals. There are several water level instrumentation nozzles installed in domestic BWR RPV. In water level instrumentation nozzles, 182 type nickel-base alloys were used for the welding joint to RPV. It is estimated the SCC potential is high in this joint because of a higher residual stress than the yield strength (about 400 MPa). This report will describe a preventive maintenance method to these nozzles Heat Affected Zone (HAZ) and welds by a corrosion resistant cladding (CRC) by YAG Laser in underwater environment (without draining a reactor water). There are many kinds of countermeasures for SCC, for example, Induction Heating Stress Improvement (IHSI), Mechanical Stress Improvement Process (MSIP) and so on. A YAG laser CRC is one of them. In this technology a laser beam is used for heat source and irradiated through an optical fiber to a base metal and SCC resistant material is used for welding wires. After cladding the HAZ and welds are coated by the corrosion resistant materials so their surfaces are improved. A CRC by gas tungsten arc welding (GTAW) in an air environment had been developed and already applied to a couple of operating plants (16 Nozzles). This method was of course good but it spent much time to perform because of an installation of some water-proof working boxes to make a TIG-weldability environment. CRC by YAG laser welding in underwater environment has superior features comparing to this conventional TIG method as follows. At the viewpoint of underwater environment, (1) an outage term reduction (no drainage water). (2) a radioactive exposure dose reduction for personnel. At that of YAG laser welding, (1) A narrower HAZ. (2) A smaller distortion. (3) A few cladding layers. A YAG laser CRC test in underwater

  5. Effects of temperature history during cooling process on cladding ductility reduction under lost of coolant accident conditions

    International Nuclear Information System (INIS)

    In a Loss-of-Coolant Accident (LOCA) of LWRs, heated fuel rods are quenched by water injected from the Emergency Core Cooling System. Existing analyses indicate that fuel rods are cooled slowly before the quench and the cooling rate at that stage depends on scenario of the accident. Microstructure can be different between fuel claddings quenched directly from high temperatures and those quenched after slow-cooling process. It has been suggested that ductility reduction is enhanced in the latter case. Although reduction in cladding ductility is important for reactor safety in a LOCA, the effects of cooling conditions on the cladding ductility reduction has not been sufficiently investigated and mechanism for the influence of cooling rate has not been clarified. In the present study, samples cut from non-irradiated 17x17-type Zircaloy-4 cladding tubes for PWRs were isothermally oxidized at 1373 and 1473 K, slowly cooled and then quenched, changing temperature at which quenching process is started and rate of the slow cooling. The oxidized and quenched samples were subjected to ring compression test in order to evaluate effects of the cooling conditions on the cladding ductility reduction. In addition, metallographic examination, Vickers hardness test and computer code analysis were conducted to examine the mechanisms. Due to oxidation, ZrO2 layer is formed at the cladding surface, oxygen-stabilized alpha phase layer is formed beneath the oxide layer, and oxygen concentration increases in the central metallic layer. As the metallic layer is cooled slowly from the isothermal oxidation temperatures, oxygen-rich α phase precipitates in the β phase layer. It was found out that reduction in cladding ductility strongly depends on area fraction of α phase region precipitated in the metallic layer. Both size and hardness of α-phase region are increased as the rate of the slow cooing decreases, while the area fraction is nearly constant. Accordingly, the reduction in

  6. Evaluation of neutron economical effect of new cladding materials in light water reactors

    International Nuclear Information System (INIS)

    The present paper aims to contribute from a neutronic aspect to activities for new cladding material development for light water reactors (LWRs) that can reduce the risk of hydrogen gas explosion. Iron (Fe), nickel (Ni), titanium (Ti), niobium (Nb) and vanadium (V) are selected as possible component elements to cover a variety of new cladding materials for LWRs. The effect of larger thermal absorption cross sections of these elements than those of zirconium (Zr), together with those of silicon carbide (SiC), on the neutron economy in LWRs is evaluated by performing pin cell burnup calculations for a conventional pressurized water reactor (PWR), a low-moderation high-burnup LWR (LM-LWR) and a high-moderation high-burnup LWR (HM-LWR). As can be anticipated from the thermal cross sections, SiC has excellent neutron economy. The materials other than SiC largely decrease discharge burnup for all three types of LWRs in comparison with Zircaloy-4. Among such elements of larger thermal absorption cross section, Nb has neutron economical advantage over the other materials except SiC in softer neutron spectrum reactors such as HM-LWR in which the atomic number ratio of hydrogen to heavy metal is 6. In conventional LWRs, stainless steel of low Ni contents has the neutron economic advantage as well as Nb for cladding material. The results of the calculations are summarized for the purpose to provide reference data for new cladding material development studies, in terms of the relation between fuel enrichment and cladding thickness from the viewpoint to achieve the same discharge burnup as the Zircaloy cladding. (author)

  7. Pellet-clad mechanical interactions: Pellet-clad bond failure and strain relief

    International Nuclear Information System (INIS)

    The effects of pellet-clad mechanical interaction would be expected to be particularly severe in the presence of bonding between the fuel and the cladding. However, such bonding is observed far more frequently than is corresponding cladding damage. It has recently been shown that the radial stress in the bond during power changes is very large and tensile, and thus likely to cause failure of the bond. In this paper the likely azimuthal extent of this de-bonding is considered, and the relief of hoop stress which this offers is assessed. It is shown that the magnitude of this relief is such as to provide an explanation of the low cladding failure rate observed. (orig.)

  8. Eutectic reaction analysis between TRU-50%Zr alloy fuel and HT-9 cladding, and temperature prediction of eutectic reaction under steady-state

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woan; Lee, Byoung Oon; Lee, Bong Sang; Park, Won Seok

    2001-02-01

    Blanket fuel assembly for HYPER contains a bundle of pins arrayed in triangular pitch, which has hexagonal bundle structure. The reference blanket fuel pin consists of the fuel slug of TRU-50wt%Zr alloy and the cladding material of ferritic martensite steel, HT-9. Chemical interaction between fuel slug and cladding is one of the major concerns in metallic fuel rod design. The contact of metallic fuel slug and stainless steel cladding in a fuel rod forms a complex multi-component diffusion couple at elevated temperatures. The potential problem of inter-diffusion of fuel and cladding components is essentially two-fold weakening of cladding mechanical strength due to the formation of diffusion zones in the cladding, and the formation of comparatively low melting point phases in the fuel/cladding interface to develop eutectic reaction. The main components of fuel slug are composed of zirconium alloying element in plutonium matrix, including neptunium, americium and uranium additionally. Therefore basic eutectic reaction change of Pu-Fe binary system can be assessed, while it is estimated how much other elements zirconium, uranium, americium and neptunium influence on plutonium phase stability. Afterwards it is needed that eutectic reaction is verified through experimental necessarily.

  9. Advanced ceramic cladding for water reactor fuel

    International Nuclear Information System (INIS)

    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of approximately 60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies ge50% would be examined

  10. The measurement of residual stresses in claddings

    International Nuclear Information System (INIS)

    The ring core method, a variation of the hole drilling method for the measurement of biaxial residual stresses, has been extended to measure stresses from depths of about 5 to 25mm. It is now possible to measure the stress profiles of clad material. Examples of measured stress profiles are shown and compared with those obtained with a sectioning technique. (author)

  11. Experimental research of irradiated nuclear fuel cladding failure processes: OECD Studsvik Cladding Integrity Project II

    International Nuclear Information System (INIS)

    The following 4 partial tasks were addressed: V001: Experimental results and knowledge of the effect of the material properties of the cladding and pellet on the phenomena of mechanical fuel-cladding interaction under the effect of radiation, at different temperatures and RAMP power load; V002: Knowledge based on the analysis of experimental data concerning the effect of iodine on the development of cracks on the fuel pin cladding tubes; V003: Processing the results of experiments to determine the primary cause of delayed hydride cracking (DHC) initiation in modern cladding alloys with low hydrogen concentrations; and V004: Analysis of the result of research into the effect of hydrides and hydrogen in the solid solution on the extension of nuclear fuel pin cladding. The results corroborated the prediction capabilities of the FEMAXI-6 code. The calculations were performed both for the reactor ramp tests and for the relaxation tests of the cladding materials, where MKP SW was the dominant tool. MKP was used for calculations within the bilateral relations with Studsvik Nuclear in the preparation of a new mechanical test for investigation of DHC, and basic MKP analyses were performed for the off-reactor test with an expansion mandrel. The theoretical generalization of the unique experimental data is documented through analysis and description of the final validation phase within the Quantum Technologies MKP model. (P.A.)

  12. Microprocessor: controlled extensometer for fuel cladding strain measurements

    International Nuclear Information System (INIS)

    An instrument for measuring strain of cladding for nuclear fuels has been developed for use during in-core characterization tests of candidate materials. Employing a microwave sensor of coaxial geometry, a strain of several microinches can be detected. At the opposite extreme, the large dynamic range of the instrument permits measuring strain equal to ten percent of the original specimen dimension. In all configurations tested, the sensor comprises a hollow metal cylinder enclosing a test specimen positioned coaxially within the cylinder. Microwave energy is absorbed by the specimen/sensor unit at particular frequencies related to size of the specimen. Consequently, determination of specimen size results from measurement of the resonant absorption frequencies of the sensor

  13. Fuel- and clad-motion diagnostics: licensing needs

    International Nuclear Information System (INIS)

    The paper addresses the current state of uncertainty with respect to fuel and clad motion during a hypothetical core-disruptive accident in a liquid metal fast breeder reactor as it relates to licensing needs. It should be noted that the paper does not represent an official position of the U.S. Nuclear Regulatory Commission, but rather, represents, in part, opinions and conclusions of its contractors. Particular attention is given to the needs for an assessment of the course of events during a hypothetical core-disruptive accident in the Clinch River Breeder Reactor. However, some of the issues discussed are likely to be relevant to larger breeder reactors as well. The issues addressed are related to the needs associated with analyses of the loss-of-flow (LOF) accident without scram and the transient overpower (TOP) accident, without scram

  14. Influence of texture on fracture toughness of zircaloy cladding

    International Nuclear Information System (INIS)

    The correlation between texture and fracture toughness of Zircaloy 2 cladding has been investigated in connection with axial cracks in fuel rods. The texture of the cladding determines the anisotropy of plasticity of the cladding which, in turn, should influence the strain conditions at the crack-tip. Plastic strains in the cladding under uniaxial tension were characterised by means of the anisotropy constants F, G and H calculated according to Hill's theory. Test temperatures between 20 and 300 deg C do not influence the F, G and H values. Any significant effect of hydrogen (about 500 wtppm) on the anisotropy constants F, G and H has not been revealed at a test temperature of 300 deg C. The results, obtained for stress-relieved and recrystallized cladding with different texture, show an obvious influence of texture on the fracture toughness of Zircaloy cladding. A higher fracture toughness has been found for cladding with more radial texture

  15. Correlation of waterside corrosion and cladding microstructure in high-burnup fuel and gadolinia rods

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M. (Argonne National Lab., IL (USA))

    1989-09-01

    Waterside corrosion of the Zircaloy cladding has been examined in high-burnup fuel rods from several BWRs and PWRs, as well as in 3 wt % gadolinia burnable poison rods obtained from a BWR. The corrosion behavior of the high-burnup rods was then correlated with results from a microstructural characterization of the cladding by optical, scanning-electron, and transmission-electron microscopy (OM, SEM, and TEM). OM and SEM examination of the BWR fuel cladding showed both uniform and nodular oxide layers 2 to 45 {mu}m in thickness after burnups of 11 to 30 MWd/kgU. For one of the BWRs, which was operated at 307{degree}C rather than the normal 288{degree}C, a relatively thick (50 to 70 {mu}m) uniform oxide, rather than nodular oxides, was observed after a burnup of 27 to 30 MWd/kgU. TEM characterization revealed a number of microstructural features that occurred in association with the intermetallic precipitates in the cladding metal, apparently as a result of irradiation-induced or -enhanced processes. The BWR rods that exhibited white nodular oxides contained large precipitates (300 to 700 nm in size) that were partially amorphized during service, indicating that a distribution of the large intermetallic precipitates is conductive to nodular oxidation. 23 refs., 9 figs.

  16. Automatic measuring system of zirconium thickness for zirconium liner cladding tubes

    International Nuclear Information System (INIS)

    An automatic system of pure zirconium liner thickness for zirconium-zircaloy cladding tubes has been successfully developed. The system consists of three parts. (1) An ultrasonic thickness measuring method for mother tubes before cold rolling. (2) An electromagnetic thickness measuring method for the manufactured tubes. (3) An image processing method for the cross sectional view of the manufactured cut tube samples. In Japanese nuclear industry, zirconium-zircaloy cladding tubes have been tested in order to realize load following operation in the atomic power plant. In order to provide for the practical use in the near future, Sumitomo Metal Industries, Ltd. has been studied and established the practical manufacturing process of the zirconium liner cladding tubes. The zirconium-liner cladding tube is a duplex tube comprising an inner layer of pure zirconium bonded to zircaloy metallurgically. The thickness of the pure zirconium is about 10 % of the total wall thickness. Several types of the automatic thickness measuring methods have been investigated instead of the usual microscopic viewing method in which the liner thickness is measured by the microscopic cross sectional view of the cut tube samples

  17. Effect of additives on corrosion resistance of Zirconium alloy for extended burn-up fuel cladding

    International Nuclear Information System (INIS)

    Sumitomo Metal Industries, Ltd. (SMI) supplies Zircaloy cladding tubes and has been developing high corrosion resistance Zr alloys for extended burn-up fuel claddings for BWR and PWR, respectively. For BWR cladding tube, small addition of IVb and Vb elements to Low Sn Zircaloy-2 improved nodular corrosion resistance. It was observed by Transmission Electron Microscopy that these additives complied with a Zr(Cr, Fe)2 type intermetallic compound and those size were finer than that precipitated in a conventional Zircaloy-2. That was assumed to result in suppressing nodular corrosion occurrence. For PWR cladding tube, small addition of Ni and Nb to extremely low Sn zirconium alloy improved uniform corrosion resistance and suppressed hydrogen pick-up. As this results Zr-1.0Sn-0.27Fe-0.16Cr-0.1Nb-0.01Ni were selected as a candidate alloy. In spite of extremely low Sn content, its mechanical properties were almost same as conventional Zircaloy-4. (author)

  18. 77 FR 36980 - Migratory Bird Hunting; Application for Approval of Copper-Clad Iron Shot as Nontoxic for...

    Science.gov (United States)

    2012-06-20

    ... of Copper-Clad Iron Shot as Nontoxic for Waterfowl Hunting AGENCY: Fish and Wildlife Service, Interior. ACTION: Notice of application for nontoxic shot approval. SUMMARY: We, the U.S. Fish and Wildlife Service, announce that Environ- Metal, Inc., of Sweet Home, Oregon, has applied for our approval of...

  19. 基于双面金属包覆光波导的传感器温度特性研究及实验验证∗%Analysis and exp erimental investigation of the temp erature prop erty of sensors based on symmetrical metal-cladding optical waveguide

    Institute of Scientific and Technical Information of China (English)

    罗雪雪; 陈家璧; 胡金兵; 梁斌明; 蒋强

    2015-01-01

    Symmetrical metal-cladding waveguide (SMCW) is a kind of new waveguide construction, and it consists of a planar glass slab sandwiched in two metal films with different thicknesses. The metal in this structure is usually a noble metal, such as Au, Ag and Cu etc. One of the characteristics of the glass is the sub-millimeter thickness, which is useful for exciting the ultrahigh order mode. Since the SMCW structure was proposed, it has received much attention from the researchers for its excellent characteristics of free-space coupling technique and ultrahigh order mode excitation. This free-space coupling technology has a higher sensitivity compared with the end-face coupling, prism coupling and grating coupling techniques. The ultrahigh order mode is very sensitive to the incident light wavelength, the thickness of guiding layer and the refractive index, but not sensitive to polarization. Based on the thermal-optical effect and thermal expansion effect of metal film and guiding layer materials, we research the temperature property of the SMCW structure. Researching methods include simulation analysis and experimental demonstration. First, we calculate the relation of the thickness and dielectric property of metal films, and the thickness and refractive index of the guiding layer with the temperature. Results show that these four factors are nearly proportional to the temperature difference. Then, we simulate the relationship of the reflectivity of the SMCW structure with those four factors by means of single-factor investigation under spectral and angular interrogation mode of operation, and find that the temperature-dependence of thickness of the guiding layer makes the chief contribution to the waveguide function of SMCW. Meanwhile, we analyze the sensitivity of the sensors based on SMCW structure, and the result shows that the sensitivity of this kind of sensor can be up to 21.89 pm/K (spectral mode) and 1.449 × 10−3 rad/K (angular mode). Finally, we

  20. Conditioning of nuclear cladding wastes by melting

    International Nuclear Information System (INIS)

    This paper discusses a cold-crucible induction melting process to condition cladding waste from irradiated fast breeder reactor fuel. The process has been developed by the CEA at Marcoule (France) as part of a major R and D program. It has been qualified at industrial scale on nonradioactive waste, and at laboratory scale on radioactive waste: several radioactive ingots have been produced from actual stainless steel or zircaloy hulls. The results confirm the numerous advantages of this containment method

  1. LASER CLADDING ON ALUMINIUM BASE ALLOYS

    OpenAIRE

    Pilloz, M.; Pelletier, J; Vannes, A.; Bignonnet, A.

    1991-01-01

    laser cladding is often performed on iron or titanium base alloys. In the present work, this method is employed on aluminum alloys ; nickel or silicon are added by powder injection. Addition of silicon leads to sound surface layers, but with moderated properties, while the presence of nickel induces the formation of hard intermetallic compounds and then to an attractive hardening phenomena ; however a recovery treatment has to be carried out, in order to eliminate porosity in the near surface...

  2. Pellet-Clad Mechanical Interaction Analysis with ANSYS Mechanical Module

    International Nuclear Information System (INIS)

    Pellet-Clad Mechanical Interaction (PCMI) has been known as a potential threat in fuel cladding integrity during power ramp conditions and high burn-up scenario. As the fuel outer surface contact with clad inner surface, the local stress become increased. Moreover, fuel pellet have much higher temperature in operation and have much greater expansion effects than clad, which occur additional contact pressure on clad inner surface, the cladding pellet deforms into a shape reflecting that of the pellet. This mechanical interaction between fuel pellet and clad depends on gap size, burn-up, friction coefficient between clad and pellet. Moreover, recent field result shows that nearly PCI-induced failures are thought to have developed at a missing pellet surface (MPS), where the tangential stress has its maximum and the cladding temperature has its minimum. For the additional study on PCMI, it is very important and valuable to find geometric parameters of MPS which make critical safety issue on cladding material safety. Followings are result and conclusion of the parametric studies

  3. Cladding failure by local plastic instability

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, J.M.; Deitrich, L.W.

    1977-12-01

    Cladding failure is one of the major considerations in analysis of fuel-pin behavior during hypothetical accident transients since time, location, and nature of failure govern the early postfailure material motion and reactivity feedback. Out-of-pile thermal transient tests of both irradiated and unirradiated fast-reactor cladding show that local plastic instability, or bulging, often precedes rupture and that the extent of local instability limits the initial rip length. To investigate the details of bulge formation and growth, a perturbation analysis of the equations governing large deformation of a cylindrical shell has been developed, resulting in a set of linear differential equations for the bulge geometry. These equations have been solved along with appropriate constitutive equations and various constraints on the ends of the cladding. Sources for bulge formation that have been considered include initial geometric imperfections and thermal perturbations due to either eccentric fuel pellets or nonsymmetric cooling. Of these, only the first is relevant to out-of-pile burst tests. Here it has been found that the most likely imperfection that will grow unstably to failure leads to a bulge around half the circumference with an axial length 1.1 times the deformed diameter. This is in general agreement with burst-test results. For the case of in-reactor fuel pins, it has been found that thermal perturbations can significantly affect local instability, particularly if the deformation process is thermally activated with a high activation energy.

  4. Development of advanced zirconium fuel cladding

    International Nuclear Information System (INIS)

    This report includes a series of the characterization results of candidate alloys, the manufacturing description of the advanced sample cladding tubes, and both the summary of out-of pile tests and the overview plan of in-pile test for them. Ten(10) kinds of the second candidate alloys, which had been selected at the first stage of the project, were comprehensively tested for their out-of pile performance. Six(6) kinds of the alloys were selected of the second ones as the final candidates through the screening tests. The out-of pile performance of the final candidates were superior to that of zircaloy-4. The advanced sample cladding tubes were made of the final candidates and tested for their out-of pile performances. The corrosion behaviors of the tubes were evaluated though the corrosion tests in water at 360 .deg. C, steam at 400 .deg. C and LiOH solution at 360 .deg. C. The mechanical properties such as creep, tensile and burst were also evaluated for each tube. The textures, microstructures, precipitates and hydrides of each tube were analyzed as well as the phase transformation was studied for each tube. In general, the test results showed that the performance of the advanced sample cladding tubes was improved over 30% in corrosion and 20% in mechanical property than that of zircaloy-4. The in-pile test of the tubes for the first phase was arranged from January 2003 to March 2007

  5. PWR cladding optimization for enhanced performance margins

    International Nuclear Information System (INIS)

    As the nuclear power generating industry has matured there is an increasing trend in core operating fuel duties. This drives a continuing evolution of cladding materials, to provide performance margin and support even higher fuel duty designs. Westinghouse has developed an optimized version of ZIRLOTM, with a thin level reduced from the nominal standard ZIRLO level of 1% to a range of 0.6% to 0.8%. The lower tin level has been shown to reduce the clad corrosion of fuel rods during reactor core operation by 30% or more while still providing the mechanical and off-normal corrosion protection benefits associated with tin alloy additions. Peak oxide levels of only 20-30 μm are observed at burnups up to 63 MWd/kgU. Using relatively small changes in the final annealing temperature, the clad creep can be adjusted to meet target ranges. In-reactor measurements of creep and growth of Optimized ZIRLOTM verify mechanical characteristics equivalent to standard ZIRLO. (author)

  6. Laser Clad Nickel Based Superalloys: Microstructure Evolution And High Temperature Oxidation Studies

    Science.gov (United States)

    Sircar, S.; Ribaudo, C.; Mazumder, J.

    1988-10-01

    Application of alloy coatings with superior oxidation resistance at elevated temperatures (1200°C) on superalloy components is of interest at present. There is a general consensus that the addition of rare earths such as hafnium (Hf) to these alloys has a pronounced effect on their performance. An in situ laser cladding technique was used to produce Ni-Al-Cr-Hf alloys on a nickel alloy substrate. Scanning Electron Microscope (SEM), Transmission Electron Microscope (TEM), and Scanning Transmission Electron Microscope (STEM) attached with Energy Dispersive X-ray (EDX) analyzers were employed for microstructural evolution studies of alloys produced during the laser cladding process. The microstructure of these alloys mainly consists of dendrites of Y' of the Ni3Al type with about 11-14 wt% Hf and an interdendritic eutectic phase. Electron microscopy in the dendritic zones reveals ordered domains whose morphology depends on laser cladding process parameters. Variation in these parameters produced only subtle changes in the composition and cell spacing of the dendritic phase. The eutectic constituent consists of a Hf-rich phase and a Hf-lean phase in an alternating lamellar structure. Convergent beam diffraction and x-ray spectroscopy techniques were used to characterize the constituents. A possible phase transformation sequence has been suggested. Differential Thermal Analysis (DTA) work indicates that the Y' dissolution temperature for the claddings is at least as high as the substrate material (Rene 80). Single cycle oxidation tests of eight hours at 1200°C in slowly flowing air reveal that the claddings have a lower weight gain rate than the substrate itself. Microchemistry and microstructure of the oxidized samples are examined using SEM attached with EDX and Auger Electron Spectroscopic (AES) techniques. The improvement in the oxidation resistance is believed to be at least partially due to the mechanical pegging between alumina coated hafnia protrusions and the

  7. Electrically heated ex-reactor pellet-cladding interaction (PCI) simulations utilizing irradiated Zircaloy cladding

    International Nuclear Information System (INIS)

    In a program sponsored by the Fuel Systems Research Branch of the US Nuclear Regulatory Commission, a series of six electrically heated fuel rod simulation tests were conducted at Pacific Northwest Laboratory. The primary objective of these tests was to determine the susceptibility of irradiated pressurized-water reactor (PWR) Zircaloy-4 cladding to failures caused by pellet-cladding mechanical interaction (PCMI). A secondary objective was to acquire kinetic data (e.g., ridge growth or relaxation rates) that might be helpful in the interpretation of in-reactor performance results and/or the modeling of PCMI. No cladding failures attributable to PCMI occurred during the six tests. This report describes the testing methods, testing apparatus, fuel rod diametral strain-measuring device, and test matrix. Test results are presented and discussed

  8. Analysis of the behaviour of under-clad and surface cracks in cladded components; Untersuchung des Verhaltens von Unterplattierungs- und Oberflaechenrissen in plattierten Komponenten

    Energy Technology Data Exchange (ETDEWEB)

    Hohe, Joerg; Luckow, Sabrina; Siegele, Dieter [Fraunhofer Institut fuer Werkstoffmechanik, Freiburg (Germany)

    2008-07-01

    The issue of the contribution is the characterization of under-clad and surface crack behaviour in ferritic steel components with an austenitic welded cladding. The experimental investigations were performed using large-scale samples. The residual stress field was determined in detail by a numerical simulation of the welding and heat treatment processes. These results were used for the numerical simulation of crack initiation and crack arrest. In all evaluated cases the crack was initiated in the ferritic material, while the cladding stayed intact even in case of a crack jump in the base metal. In the frame of case studies the results were transferred to application relevant geometries. [German] Der Gegenstand des vorliegenden Beitrags ist eine Charakterisierung des Verhaltens von Unterplattierungs- und Oberflaechenfehlern in Komponenten aus ferritischen Staehlen mit einer austenitischen Schweissplattierung. Im Rahmen einer experimentellen Untersuchung wurden hierzu zwei Versuche an Grossproben durchgefuehrt. Das Eigenspannungsfeld wurde detailliert durch eine numerische Simulation der Schweiss- und Waermebehandlungsprozesse bestimmt. Unter Verwendung dieser, an anderer Stelle praesentierten Ergebnisse erfolgte eine numerische Simulation der Bauteilversuche zur genauen Bestimmung der Verhaeltnisse bei Rissinitiierung und Rissarrest. In allen untersuchten Faellen zeigte sich, dass der Bruch im ferritischen Bereich ausgeloest wurde, waehrend die Plattierung auch bei einem Risssprung im Grundwerkstoff intakt blieb. Im Rahmen von Fallstudien wurden die Ergebnisse auf anwendungsrelevante Geometrien uebertragen. (orig.)

  9. Microstructure and Mechanical Properties of Laser Clad and Post-cladding Tempered AISI H13 Tool Steel

    Science.gov (United States)

    Telasang, Gururaj; Dutta Majumdar, Jyotsna; Wasekar, Nitin; Padmanabham, G.; Manna, Indranil

    2015-05-01

    This study reports a detailed investigation of the microstructure and mechanical properties (wear resistance and tensile strength) of hardened and tempered AISI H13 tool steel substrate following laser cladding with AISI H13 tool steel powder in as-clad and after post-cladding conventional bulk isothermal tempering [at 823 K (550 °C) for 2 hours] heat treatment. Laser cladding was carried out on AISI H13 tool steel substrate using a 6 kW continuous wave diode laser coupled with fiber delivering an energy density of 133 J/mm2 and equipped with a co-axial powder feeding nozzle capable of feeding powder at the rate of 13.3 × 10-3 g/mm2. Laser clad zone comprises martensite, retained austenite, and carbides, and measures an average hardness of 600 to 650 VHN. Subsequent isothermal tempering converted the microstructure into one with tempered martensite and uniform dispersion of carbides with a hardness of 550 to 650 VHN. Interestingly, laser cladding introduced residual compressive stress of 670 ± 15 MPa, which reduces to 580 ± 20 MPa following isothermal tempering. Micro-tensile testing with specimens machined from the clad zone across or transverse to cladding direction showed high strength but failure in brittle mode. On the other hand, similar testing with samples sectioned from the clad zone parallel or longitudinal to the direction of laser cladding prior to and after post-cladding tempering recorded lower strength but ductile failure with 4.7 and 8 pct elongation, respectively. Wear resistance of the laser surface clad and post-cladding tempered samples (evaluated by fretting wear testing) registered superior performance as compared to that of conventional hardened and tempered AISI H13 tool steel.

  10. Optimization of Ni-Based WC/Co/Cr Composite Coatings Produced by Multilayer Laser Cladding

    OpenAIRE

    Andrea Angelastro; Sabina L. Campanelli; Giuseppe Casalino; Antonio D. Ludovico

    2013-01-01

    As a surface coating technique, laser cladding (LC) has been developed for improving wear, corrosion, and fatigue properties of mechanical components. The main advantage of this process is the capability of introducing hard particles such as SiC, TiC, and WC as reinforcements in the metallic matrix such as Ni-based alloy, Co-based alloy, and Fe-based alloy to form ceramic-metal composite coatings, which have very high hardness and good wear resistance. In this paper, Ni-based alloy (Colmonoy ...

  11. Patent Analysis of Ferritic/Martensitic Steels for the Fuel Cladding in Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jong Hyuk; Kim, Sung Ho; Kim, Tae Kyu; Kim, Woo Gon; Jang, Jin Sung; Kim, Dae Whan; Han, Chang Hee; Lee, Chan Bock

    2007-09-15

    The Korean, Japanese, U.S. and European patents related to the ferritic/martensitic steels were systematically surveyed to evaluate their patent status, which would be applicable to the fuel cladding materials for the Sodium-cooled Fast Reactor (SFR). From the surveys, totally 38 patents were finally selected for the quantitative and qualitative analysis. Among them, 28 patents (74%) were processed by Japanese companies and Sumitomo Metal industries Ltd. was top-ranked in the number (9) of priority patents. On the basis of these surveys, most patents could be applicable to the fuel cladding materials for SFR and, especially, some useful patents as the cladding were registered by the Russian and the Korean.

  12. Impact of reactor water chemistry on cladding performance

    International Nuclear Information System (INIS)

    Water chemistry may have a major impact on fuel cladding performance in PWRs. If the saturation temperature on the surface of fuel cladding is exceeded, either because of the thermal hydraulics of the system, or because of crud deposition, then LiOH concentration can occur within thick porous oxide films on the cladding. This can degrade the protective film and accelerate the corrosion rate of the cladding. If sufficient boric acid is also present in the coolant then these effects may be mitigated. This is normally the case through most of any reactor fuel cycle. Extensive surface boiling may disrupt this equilibrium because of the volatility of boric acid in steam. Under such conditions severe cladding corrosion can ensue. The potential for such effects on high burnup cladding in CANDU reactors, where bone acid is not present in the primary coolant, is discussed. (author)

  13. Formation of NiCrAlY laser cladding with preplaced method; Funtai tofu reza kuraddingu ho ni yoru NiCrAlY himaku no keisei

    Energy Technology Data Exchange (ETDEWEB)

    Iwamoto, H. [Chiyoda Chemical Engineering and Construction Co. Ltd., Tokyo (Japan)] Sumikawa, T. [Isuzu Motors Ltd., Tokyo (Japan)] Nishida, K. [Ehime Univ., Ehime (Japan). Graduate School; Nishida, M.; Araki, T. [Ehime Univ., Ehime (Japan). Faculty of Engineering

    1997-05-05

    Cladding technique on the surface of structure materials using high corrosion resistance and wear resistance materials has been used in many fields as compared to conventional one. The heat source for cladding was mainly arc welding, however, these recent years, laser cladding where laser having special characteristics like high control, high output density and so forth is used as a heat source, has been studied. In this report, in order to cope with the change in composition of base material, coating layer formation with preplaced method was studied systematically using NiCrAlY powder formed by mixing mechanically commercial metal powder with raw material powder and low power (1200W) CO2 laser. As a result, good NiCrAlY cladded layer with good composition and with no defect was achieved using basic powder formed by mixing mechanically the metallic powder and commercial organic binder. Cladding of broad area was possible by weaving method even at low power (1200W) CO2 laser. Cladded layer with lower dilution ratio and higher cross sectional area was achieved with smaller powder particle diameter. 9 refs., 11 figs., 4 tabs.

  14. Evaluation of residual stress near the weld overlay cladding by welding and post-weld heat treatment

    International Nuclear Information System (INIS)

    Austenitic stainless steel is welded as a cladding on the inner surface of a reactor pressure vessel (RPV) made of low alloy steel. In order to assess the structural integrity of the RPV precisely, the residual stress distribution caused by weld-overlay cladding and post-weld heat treatment (PWHT) is evaluated. Since the cladding layer is very thin compared to vessel wall, it is necessary to evaluate the residual stress distribution around the weld fusion line can be very steep. In this study, cladded specimens were fabricated using different welding methods. Residual stress measurements using both sectioning and deep hole drilling (DHD) methods were then performed to evaluate the residual stress distributions through the weld fusion line. Three-dimensional thermal-elastic-plastic-creep analyses based on finite element method were also conducted to evaluate the residual stress caused by weld-overlay cladding and PWHT. It was shown that analytical results provided reasonable agreements on weld residual stress with experimental results. It was also clarified that the main cause of residual stress due to welding and PWHT was the difference of thermal expansion between weld and base metals. (author)

  15. Optimization of Ni-Based WC/Co/Cr Composite Coatings Produced by Multilayer Laser Cladding

    Directory of Open Access Journals (Sweden)

    Andrea Angelastro

    2013-01-01

    Full Text Available As a surface coating technique, laser cladding (LC has been developed for improving wear, corrosion, and fatigue properties of mechanical components. The main advantage of this process is the capability of introducing hard particles such as SiC, TiC, and WC as reinforcements in the metallic matrix such as Ni-based alloy, Co-based alloy, and Fe-based alloy to form ceramic-metal composite coatings, which have very high hardness and good wear resistance. In this paper, Ni-based alloy (Colmonoy 227-F and Tungsten Carbides/Cobalt/Chromium (WC/Co/Cr composite coatings were fabricated by the multilayer laser cladding technique (MLC. An optimization procedure was implemented to obtain the combination of process parameters that minimizes the porosity and produces good adhesion to a stainless steel substrate. The optimization procedure was worked out with a mathematical model that was supported by an experimental analysis, which studied the shape of the clad track generated by melting coaxially fed powders with a laser. Microstructural and microhardness analysis completed the set of test performed on the coatings.

  16. Experimental results on the interactions between hydrogen and zirconium claddings

    International Nuclear Information System (INIS)

    Experiments were performed with Zr1%Nb and Zircaloy-4 alloys to study the interaction between hydrogen and Zr containing cladding materials. Four main activities are summarised in the report: equilibrium solubility of hydrogen in cladding with oxygen content, escape of hydrogen during steam oxidation, escape of hydrogen during steam oxidation of cladding alloys with H-content, delaying effect of surface oxide layer on the hydrogen absorption from gas phase by the Zr alloys. (author)

  17. Pellet-clad interaction in water reactor fuels

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    The aim of this seminar is was to draw up a comprehensive picture of the pellet clad interaction and its impact on the fuel rod. This document is a detailed abstract of the papers presented during the following five sessions: industrial goals, fuel material behaviour in PCI situation, cladding behaviour relevant to PCI, in pile rod behaviour and modelling of the mechanical interaction between pellet and cladding. (A.L.B.)

  18. Pellet-clad interaction in water reactor fuels

    International Nuclear Information System (INIS)

    The aim of this seminar is was to draw up a comprehensive picture of the pellet clad interaction and its impact on the fuel rod. This document is a detailed abstract of the papers presented during the following five sessions: industrial goals, fuel material behaviour in PCI situation, cladding behaviour relevant to PCI, in pile rod behaviour and modelling of the mechanical interaction between pellet and cladding. (A.L.B.)

  19. Stainless steel clad for light water reactor fuels. Final report

    International Nuclear Information System (INIS)

    Proper reactor operation and design guidelines are necessary to assure fuel integrity. The occurrence of fuel rod failures for operation in compliance with existing guidelines suggests the need for more adequate or applicable operation/design criteria. The intent of this study is to develop such criteria for light water reactor fuel rods with stainless steel clad and to indicate the nature of uncertainties in its development. The performance areas investigated herein are: long term creepdown and fuel swelling effects on clad dimensional changes and on proximity to clad failure; and short term clad failure possibilities during up-power ramps

  20. KALIMER-600-clad Core Fuel Assembly Calculation using MATRA-LMR (V2.0) Code

    International Nuclear Information System (INIS)

    Since the sodium boiling point is very high, maximum cladding and pin temperatures are used for design limit condition in sodium cooled liquid metal reactor. It is necessary to predict accurately the temperature distribution in the core and in the subassemblies to increase the sodium coolant efficiency. Based on the MATRA code, which is developed for PWR analysis, MATRA-LMR has been developed for SFR. The major modifications are: the sodium properties table is implemented as subprogram in the code, Heat transfer coefficients are changed for SFR, te pressure drop correlations are changed for more accurate calculations, which are Novendstern, Chiu-Rohsenow-Todreas, and Cheng-Todreas correlations. This This report describes briefly code structure and equations of MATRA-LMR (Version 2.0), explains input data preparation and shows some calculation results for the KALIMER-600-clad core fuel assembly for which has been performed the conceptual design of the core in the year 2006

  1. EXPERIMENTALLY-STATISTICAL MODEL OF CLADDING LAYER FORMATION PROCESS ON SLIDE-WAYS

    Directory of Open Access Journals (Sweden)

    N. N. Maksimchenko

    2014-11-01

    Full Text Available The developed experimentally-statistical model of the cladding composite layer formation process on slide-ways allows to operate technological modes of cladding by flexible instrument (CFI in order to obtain the set properties of a coating (thickness, continuity, adhesion strength.The established optimum technological modes of CFI process providing formation of continuous, strongly adhered to a basis composite coatings of the required thickness have been used for applying coatings on working surfaces of slide-ways of metal-cutting machine tool beds that has allowed to lower friction factor in coupling on the average by 1.3–1.7-fold and to improve uniformity of slow moving of machine tool units by 1.74-fold in comparison with slide-ways without a coating. 

  2. Analysis of microstructure and properties of multilayer coatings produced by laser cladding

    Science.gov (United States)

    Bykovskiy, D. P.; Petrovskiy, V. N.; Dzhumaev, P. S.; Polskiy, V. I.; Yermachenko, V. M.

    2016-02-01

    Purpose of the work is to prepare multilayer coatings corresponding to specified requirements to recovery and improvement of surface details. Requirements to coatings: providing durable and reliable adhesion base and filler materials, absence of pores, cracks, delaminations, reducing mixing metal base and cladding. We used iron-based PR-10R6M5 and tungsten carbide Hoganas 44712 powders. Experimental determination of the optimal technological mode of application of the single track, the coefficient of overlapping tracks to create a full layer, the angle of the second cladding layer, relative to the first one and, finally, the determination of the optimal additive tungsten carbide to achieve increased durability were produced to fulfill these requirements.

  3. Suppression of Cladding Mode Coupling by B/Ge Codoped Photosensitive Fiber With Photosensitive and Depressed Inner Cladding

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    Excess loss on the short wavelength side of the Bragg resonant wavelength caused by cladding mode coupling limits wide use of grating in the fiber communication system, especially in densed wavelength division multiplexing (DWDM) system.A novel photosensitive fiber design that have depressed cladding and photosensitive inner cladding in the same fiber is proposed, which can suppress cladding mode coupling greatly.Using MCVD method B/Ge codoped fiber with depressed cladding was fabriceted out, which was also doped in boron and germanium and had the photosensitivity.Finally, the transmission spectrum of written grating in this fiber by phase mask method verified its larger photosensitivity and greatly suppression of cladding mode coupling.

  4. Rheological evaluation of pretreated cladding removal waste

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, D.; Chan, M.K.C.; Lokken, R.O.

    1986-01-01

    Cladding removal waste (CRW) contains concentrations of transuranic (TRU) elements in the 80 to 350 nCi/g range. This waste will require pretreatment before it can be disposed of as glass or grout at Hanford. The CRW will be pretreated with a rare earth strike and solids removal by centrifugation to segregate the TRU fraction from the non-TRU fraction of the waste. The centrifuge centrate will be neutralized with sodium hydroxide. This neutralized cladding removal waste (NCRW) is expected to be suitable for grouting. The TRU solids removed by centrifugation will be vitrified. The goal of the Rheological Evaluation of Pretreated Cladding Removal Waste Program was to evaluate those rheological and transport properties critical to assuring successful handling of the NCRW and TRU solids streams and to demonstrate transfers in a semi-prototypic pumping environment. This goal was achieved by a combination of laboratory and pilot-scale evaluations. The results obtained during these evaluations were correlated with classical rheological models and scaled-up to predict the performance that is likely to occur in the full-scale system. The Program used simulated NCRW and TRU solid slurries. Rockwell Hanford Operations (Rockwell) provided 150 gallons of simulated CRW and 5 gallons of simulated TRU solid slurry. The simulated CRW was neutralized by Pacific Northwest Laboratory (PNL). The physical and rheological properties of the NCRW and TRU solid slurries were evaluated in the laboratory. The properties displayed by NCRW allowed it to be classified as a pseudoplastic or yield-pseudoplastic non-Newtonian fluid. The TRU solids slurry contained very few solids. This slurry exhibited the properties associated with a pseudoplastic non-Newtonian fluid.

  5. Rheological evaluation of pretreated cladding removal waste

    International Nuclear Information System (INIS)

    Cladding removal waste (CRW) contains concentrations of transuranic (TRU) elements in the 80 to 350 nCi/g range. This waste will require pretreatment before it can be disposed of as glass or grout at Hanford. The CRW will be pretreated with a rare earth strike and solids removal by centrifugation to segregate the TRU fraction from the non-TRU fraction of the waste. The centrifuge centrate will be neutralized with sodium hydroxide. This neutralized cladding removal waste (NCRW) is expected to be suitable for grouting. The TRU solids removed by centrifugation will be vitrified. The goal of the Rheological Evaluation of Pretreated Cladding Removal Waste Program was to evaluate those rheological and transport properties critical to assuring successful handling of the NCRW and TRU solids streams and to demonstrate transfers in a semi-prototypic pumping environment. This goal was achieved by a combination of laboratory and pilot-scale evaluations. The results obtained during these evaluations were correlated with classical rheological models and scaled-up to predict the performance that is likely to occur in the full-scale system. The Program used simulated NCRW and TRU solid slurries. Rockwell Hanford Operations (Rockwell) provided 150 gallons of simulated CRW and 5 gallons of simulated TRU solid slurry. The simulated CRW was neutralized by Pacific Northwest Laboratory (PNL). The physical and rheological properties of the NCRW and TRU solid slurries were evaluated in the laboratory. The properties displayed by NCRW allowed it to be classified as a pseudoplastic or yield-pseudoplastic non-Newtonian fluid. The TRU solids slurry contained very few solids. This slurry exhibited the properties associated with a pseudoplastic non-Newtonian fluid

  6. Development of SFR Fuel Cladding Tube Materials

    International Nuclear Information System (INIS)

    A R and D program for new materials for SFR cladding tube was started in 2007. The purpose of the R and D program is to develop new cladding materials having a higher creep rupture strength than the Gr.92 steel. For this purpose, the minor alloying elements such as V, Ti, C and N were added into the ferritic/martensitic (FM) steels. 5 new alloys were designed, manufactured and evaluated. Increase of V concentration caused the increase of mass fraction of V-rich MX particles. But high V steel revealed lower yield, tensile and creep rupture strengths. High N and low C steel showed higher tensile strength and lower creep rupture strength than the low N and high C steel. The Zr addition appeared to be more effective than Ti addition in terms of yield, tensile and creep rupture strengths. In order to develop a fabrication process of SFR cladding tube, the effects of the fabrication process parameters such as a tempering temperature, cold rolling and annealing condition on the precipitates and mechanical properties of a normalized FM steel were also evaluated. Nb-rich MX precipitates were found in the specimen tempered at 550''oC while M23C6, Nb- and V-rich MX ones were observed in the specimen tempered at 750''oC. A cold rolling and an annealing at 750''oC of the specimen tempered at 550''oC induced the formation of large inhomogeneous M23C6 carbides, causing a reduced tensile strength. However, the cold rolling of the specimen tempered at 750''oC provided fine precipitates mainly due to a fragmentation of the M23C6 carbides, and an annealing at 700''oC for 30 min was found to be suitable to recover the degraded mechanical properties from a cold working. (author)

  7. Ferrous Alloy Powder for Laser Cladding

    Institute of Scientific and Technical Information of China (English)

    WEN Jialing; NIU Quanfeng; XU Yanmin

    2005-01-01

    This investigation aimed at how to improve the hardness and wear resistance by B, Si and Cr, and how to improve the synthesis property by Re (rare-earth element). Based on the experiment of Fe-based alloys of Fe-Cr-Ni-B-Si-Re, through experiments and a serious of synthesis analysis, including surface quality, spectrum composite, micro-hardness, scanning electron microscopy, as well as the synthesis evaluation,etc, prescriptions were optimized. As a result, a Fe-Cr-Ni-B-Si-Re cladding material with a high property was obtained.

  8. COMPARISON OF CLADDING CREEP RUPTURE MODELS

    International Nuclear Information System (INIS)

    The objective of this calculation is to compare several creep rupture correlations for use in calculating creep strain accrued by the Zircaloy cladding of spent nuclear fuel when it has been emplaced in the repository. These correlations are used to calculate creep strain values that are then compared to a large set of experimentally measured creep strain data, taken from four different research articles, making it possible to determine the best fitting correlation. The scope of the calculation extends to six different creep rupture correlations

  9. Thermomechanical loading applied on the cladding tube during the pellet cladding mechanical interaction phase of a rapid reactivity initiated accident

    Energy Technology Data Exchange (ETDEWEB)

    Hellouin de Menibus, Arthur, E-mail: arthur.hellouin-de-menibus@cea.fr [CEA Saclay/DEN/DANS/DMN/SRMA, 91191 Gif-sur-Yvette (France); Sercombe, Jerome [CEA Cadarache/DEN/CAD/DEC/SESC, 13108 St Paul lez Durance (France); Auzoux, Quentin; Poussard, Christophe [CEA Saclay/DEN/DANS/DMN/SEMI, 91191 Gif-sur-Yvette (France)

    2014-10-15

    Calculations of the CABRI REP-Na5 pulse were performed with the ALCYONE code in order to determine the evolution of the thermomechanical loading applied on the cladding tube during the Pellet–Cladding Mechanical Interaction (PCMI) phase of a rapid Reactivity Initiated Accident (RIA) initiated at 280 °C that lasted 8.8 ms. The evolution of the following parameters are reported: the cladding temperature, heating rate, strain rate and loading biaxiality. The impact of these parameters on the cladding mechanical behavior and fracture are then briefly reviewed.

  10. Research on honeycomb structure explosives and double sided explosive cladding

    International Nuclear Information System (INIS)

    Highlights: • Honeycomb structure explosives are used to ensure the quality of charge. • Double sided explosive cladding can clad two composite plates simultaneously. • The critical thickness of explosives decreased significantly. • The energy efficiency of explosives has been significantly improved. • Experiment results can be better predicted by calculation. - Abstract: In order to resolve the current issues about the backward method of charge and low energy efficiency of explosives, honeycomb structure explosives and double sided explosive cladding were used in the present study. Honeycomb structure explosives are used to ensure the quality of charge. Double sided explosive cladding can clad two composite plates simultaneously. Honeycomb structure explosives and double sided explosive cladding, which significantly reduce the critical thickness of stable detonation of explosives, are used to increase the energy efficiency of explosives and save the amount of explosives. Emulsion explosives with the thickness of 5 mm can be stable detonation. In this paper, the experiment of double sided explosive cladding for two groups of steel of No. 45 with the thickness of 2 mm to steel of Q235 with the thickness of 16 mm and two groups of stainless steel with the thickness of 3 mm to steel of Q235 with the thickness of 16 mm were successfully investigated. Without constraints, the critical diameter of emulsion explosives is 14–16 mm. Compared to the existing explosive cladding method, the consumption of explosives for steel of No. 45 to steel of Q235 and stainless steel to steel of Q235 are reduced by 83% and 77% in the case of cladding the same number of composite plates. The explosive cladding windows and collision velocity of flyer plate were calculated before experiment. It shows that the calculation prefigures exactly the explosive cladding for steel of No. 45 to steel of Q235 and stainless steel to steel of Q235

  11. Formation of intermetallic compound at interface between rare earth elements and ferritic-martensitic steel by fuel cladding chemical interaction

    Institute of Scientific and Technical Information of China (English)

    Jun Hwan Kim; Byoung Oon Lee; Chan Bock Lee; Seung Hyun Jee; Young Soo Yoon

    2012-01-01

    The intermetallic compounds formation at interface between rare earth elements and clad material were investigated to demonstrate the effects of rare earth elements on fuel-cladding chemical interaction (FCCI) behavior.Mischmetal (70Ce-30La) and Nd were prepared as rare earth elements.Diffusion couple testing was performed on the rare earth elements and cladding (9Cr2W steel) near the operation temperature of(sodium-cooled fast reactor) SFR fuel.The performance of a diffusion barrier consisting of Zr and V metallic foil against the rare earth elements was also evaluated.Our results showed that Ce and Nd in the rare earth elements and Fe in the clad material interdiffused and reacted to form intermetallic species according to the parabolic rate law,describing the migration of the rare earth element.The diffusion of Fe limited the reaction progress such that the entire process was governed by the cubic rate law.Rare earth materials could be used as a surrogate for high burnup metallic fuels,and the performance of the barrier material was demonstrated to be effective.

  12. Material Selection for Accident Tolerant Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Snead, Lance Lewis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Alternative cladding materials to Zr-based alloys are being investigated for accident tolerance, which can be defined as > 100X improvement (compared to Zr-based alloys) in oxidation resistance to steam or steam-H2 environments at ≥ 1200°C for short times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases and FeCrAl alloys. Recently reported low mass losses for Mo in steam at 800°C could not be reproduced. Both FeCrAl and MAX phase Ti2AlC form a protective alumina scale in steam. However, commercial Ti2AlC that was not single phase, formed a much thicker oxide at 1200°C in steam and significant TiO2, and therefore Ti2AlC may be challenging to form as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1475°C, while reducing its Cr content to minimize susceptibility to irradiation-assisted α´ formation. The composition effects and critical limits to retaining protective scale formation at > 1400°C are still being evaluated.

  13. Material Selection for Accident Tolerant Fuel Cladding

    Science.gov (United States)

    Pint, B. A.; Terrani, K. A.; Yamamoto, Y.; Snead, L. L.

    2015-09-01

    Alternative cladding materials to Zr-based alloys are being investigated for accident tolerance, which can be defined as >100X improvement (compared to Zr-based alloys) in oxidation resistance to steam or steam-H2 environments at ≥1473 K (1200 °C) for short times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases, and FeCrAl alloys. Recently reported low-mass losses for Mo in steam at 1073 K (800 °C) could not be reproduced. Both FeCrAl and MAX phase Ti2AlC form a protective alumina scale in steam. However, commercial Ti2AlC that was not single phase, formed a much thicker oxide at 1473 K (1200 °C) in steam and significant TiO2, and therefore, Ti2AlC may be challenging to form as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1748 K (1475 °C), while reducing its Cr content to minimize susceptibility to irradiation-assisted α' formation. The composition effects and critical limits to retaining protective scale formation at >1673 K (1400 °C) are still being evaluated.

  14. Instrument for measuring fuel cladding strain

    International Nuclear Information System (INIS)

    Development work to provide instrumentation for the continuous measurement of strain of material specimens such as nuclear fuel cladding has shown that a microwave sensor and associated instrumentation hold promise. The cylindrical sensor body enclosing the specimen results in a coaxial resonator absorbing microwave energy at frequencies dependent upon the diameter of the specimen. Diametral changes of a microinch can be resolved with use of the instrumentation. Very reasonable values of elastic strain were measured at 750F and 10000F for an internally pressurized 20 percent C.W. 316 stainless steel specimen simulating nuclear fuel cladding. The instrument also indicated the creep strain of the same specimen pressurized at 6500 psi and at a temperature of 10000F for a period of 700 hours. Although the indicated strain appears greater than actual, the sensor/specimen unit experienced considerable oxidation even though an inert gas purge persisted throughout the test duration. By monitoring at least two modes of resonance, the measured strain was shown to be nearly independent of sensor temperature. To prevent oxidation, a second test was performed in which the specimen/sensor units were contained in an evacuated enclosure. The strain of the two prepressurized specimens as indicated by the microwave instrumentation agreed very closely with pre- and post-test measurements obtained with use of a laser interferometer

  15. Chemical interaction of fuel and cladding tubes

    International Nuclear Information System (INIS)

    It was attempted to take up the behavior of nuclear fuel in cores and summarize it by the expert committee on the irradiation behavior of nuclear fuel from fiscal 1978 to fiscal 1980 from the following viewpoints. The behavior of nuclear fuel in cores has been treated separately according to each reactor type, accordingly this point is reconsidered. The clearly understood points and the uncertain points are discriminated. It is made more easily understandable for people in other fields of atomic energy. This report is that of the group on the chemical interaction, and the first report of this committee. The chemical interaction as the behavior of fuel in cores is in the unseparable relation to the mechanical interaction, but this relation is not included in this report. The chemical interaction of fuel and cladding tubes under irradiation shows different phenomena in LWRs and FBRs, and is called SCC and FCC, respectively. But this point of causing the difference must be understood to grasp the behavior of fuel. The mutual comparison of oxide fuels for FBRs and LWRs, the stress corrosion cracking of zircaloy tubes, and fuel-cladding chemical interaction in FBRs are reported. (Kako, I.)

  16. Chemical compatibility between U-6wt.%Zr alloy and T91 cladding

    International Nuclear Information System (INIS)

    Full text: Metal fuel based on binary U-Pu and ternary U-Pu-Zr alloys has been considered as a promising advanced fuel for fast reactor in India due to its favorable thermal and neutronic performance, enhanced reactor safety, ease of fabrication and suitability for pyro-reprocessing. Fuel-clad chemical interaction (FCCI) has been recognized as one of the major issues in metallic fuel since it limits life of fuel pin due to formation of low melting eutectic at significantly lower temperature than the melting point of the fuel alloy. As a part of metallic fuel development programme for future FBR's in India, capsule irradiation of metal fuel based on sodium bonded ternary U-Pu-Zr (Zr2 type layer of thickness around 5μm on the clad side, a Zr depleted layer of thickness 20 μm on the fuel side and a thin Zr rich layer of thickness 2-3μm between the above two. The Zr rich layer formed at the interface is known to be fuel-clad diffusion barrier. U at the interface reacts with Fe, Cr of T91 leaving behind Zr which forms a barrier layer at the interface. When the annealing temperature was increased to 750 deg C, a phase with eutectic microstructure was observed. The bright phase is U6Fe phase, the grey and dark phases are U(Fe,Cr)2 and Zr(Fe,Cr)2 respectively. The results indicate that measured eutectic temperature between U-6Zr and T91 is very crucial from reactor safety point of view. Diffusion couple experiments clearly show that Zr rich layer will retard the diffusion of U to the interface below eutectic temperature but above eutectic temperature U diffusion is fast enough to result in eutectic me

  17. Zircaloy-4 hydriding. Hydrogen distribution in PWR's rod cladding

    International Nuclear Information System (INIS)

    In pressurised water reactors, Zircaloy 4 is used as fuel cladding in contact with hot water. The precipitation of hydrides at room temperatures causes mechanical deterioration of the cladding. As the cladding is subjected to a radial temperature gradient, the hydrogen distribution is greatly affected. The image analysis method is used to determine the hydride distribution in the irradiated cladding. To calibrate this method, a device was specially built for the preparation of Zircaloy specimens with known hydrogen contents. The hydriding conditions and hydrogen content determination procedures were fixed. We have successfully realized specimens with various hydrogen contents. With these specimens, a relationship between the parameter Sv (surface density of hydrides) and the hydrogen content was established. This parameter Sv is independent from the Zircaloy 4 metallurgical state (i.e. stress relieved or recrystallized) and from the analysis section (longitudinal or transverse). Study of hydrogen content and hydride distribution in irradiated cladding by means of image analysis showed that the method is limited by its ability of separation between neighbouring hydrides at cladding's periphery where the hydrogen content can reach several thousands ppm. Nevertheless, this method gives us some information about hydride distribution inside the cladding. A model for thermal diffusion was developped to stimulate the migration of hydrogen in Zirconium alloys. This model was used to predict hydrogen distribution in the irradiated cladding. Comparison of model predictions with results of image analysis shows good agreement. (Author). refs., figs., tabs

  18. Embedded cladding surface thermocouples on Zircaloy-sheathed heater rods

    International Nuclear Information System (INIS)

    Titanium-sheathed Type K thermocouples embedded in the cladding wall of zircaloy-sheathed heater rods are described. These thermocouples constitute part of a program intended to characterize the uncertainty of measurements made by surface-mounted cladding thermocouples on nuclear fuel rods. Fabrication and installation detail, and laboratory testing of sample thermocouple installations are included

  19. Qualification of submerged-arc narrow strip cladding process

    International Nuclear Information System (INIS)

    Babcock and Wilcox has developed an unique narrow strip cladding process for use on both plate and forging material for nuclear components. The qualification testing of this low-heat input process for cladding nuclear components is described, including those of SA508 Class 2 material. The theory that explains the acceptable results of these tests is also given

  20. LASER CLADDING WITH COBALT-BASED HARDFACING ALLOYS

    OpenAIRE

    Frenk, A.; WagniÈre, J.-D.

    1991-01-01

    Preliminary results aimed at designing Co-based hardfacing alloys specifically for the laser cladding process are reported. Three alloys, ranging from hypo- to hypereutectic were deposited using scanning velocities between 1.7 and 170 mm/s. The microstructures and the dry sliding wear resistances of the clads were investigated. First trends relating composition to dry sliding wear resistance were deduced.

  1. Experimental to study the effect of multiple weld-repairs on microstructure, hardness and residual stress for a stainless steel clad plate

    International Nuclear Information System (INIS)

    Highlights: • Effect of multiple repairs on residual stress, microstructure and hardness for a clad plate has been studied. • A diffusion layer with martensite is generated around weld-base metal interface. • Along the weld–clad metal interface, short ferrite is generated in the fusion zone. • As the repair times increase, the content of short ferrite is increased. • It is proposed that the stainless steel clad plate cannot be repaired more than 2 times. - Abstract: This paper presents an experimental study of multiple repair welds in a stainless steel clad plate. Four weld samples with one, two, three and four repairs in the same area were prepared, respectively, to determine the changes in microstructure, residual stress and micro hardness. Neutron diffraction measurement was used to determine the weld residual stress. The results show that the repair weld contains nine zones with different microstructures. Around weld-base metal interface, a diffusion layer is formed because of the diffusion of C, Cr, Ni and Fe elements. The diffusion layer, which contains martensite with larger hardness than the adjacent metals, should be removed completely before re-repair. Around the weld–clad metal interface, short ferrite is generated along the fusion zone. As the repair number increases, the fusion zone becomes thicker, and the content of short ferrite also increases, which leads to some voids in the third and fourth repair welds. In the fusion zone of the sample with four repairs, massive ferrite is generated because more Cr element is diffused to the fusion zone. Residual stress increases gradually from the weld center and reaches the maximum at heat affected zone (HAZ) and then decreases far away. The residual stress in the sample with four repairs decreases because the hardness is decreased. Based on the comprehensive considerations of microstructure, residual stress and hardness, it proposes that the clad plate cannot be repaired more than 2 times

  2. Toughness properties of end of life reactor pressure vessel cladding

    International Nuclear Information System (INIS)

    The inside surface of reactor pressure vessels is protected against corrosion by a clad overlay made of austenic stainless steel. This cladding, applied by automatic submerged arc welding with strip electrode, is constituted by two layers, the first one in 309L (24 CR - 12 Ni) steel and the second one in 308L (18 Cr - 10 Ni) steel. Safety analysis of reactor pressure vessel (RPV) consists to verify the resistance to fracture of the vessel, assumed containing a small crack just under the cladding. That implies the knowledge of the mechanical properties of the cladding. With the objective to evaluate the mechanical properties of the cladding at the end of life of the RPV, a coordinated French experimental programme has been carried out jointly by CEA, EDF, and Framatome. The first experimental results of this programme are given in this paper. 10 figs, 3 tabs

  3. Cladding rupture detection of tammuz-2 reactor fuel

    International Nuclear Information System (INIS)

    The checking of fuel cladding integrity and uranium contamination level was performed using delayed neutron cladding rupture detection system (DRG). This checking is important for the start up procedure of the reactor. The technique of DRG depends on monitoring the delayed neutron emitters such as Br-87 (t1/2=54.5 sec) and l-137 (t1/2=24 sec) in the coolant of the reactor. As a requirement for the start up two tests for checking fuel cladding integrity were performed. A comparison between the obtained data and the recommended data by the vendor are tabulated. Additional data were obtained at higher power. Accordingly, it can be concluded that the activities detected were due to the uranium contamination of cladding surfaces. Results confirmed that no rupture existed on the claddings

  4. Stress analysis of a cladded medium containing a subsurface crack

    International Nuclear Information System (INIS)

    The stress analysis of a cladded medium containing an underclad subsurface crack is performed, with particular applications to the structural integrity of cladded pressure vessels. To this end, a thin cladding is assumed to be bonded to a substrate through an adhesive bonding agent. By representing the cladding and the adhesive layer as elastic strips with distinct properties, the entire medium takes the form of a 3-layered half-space. Based upon the theory of plane elasticity, a singular integral equation is derived for the current crack problem. The stress intensity factors are defined. As numerical illustrations, the influences of cladding and bonding agent on the values of stress intensity factors are discussed

  5. Mechanical properties of silver halide core/clad IR fibers

    Science.gov (United States)

    Shalem, Shaul; German, Alla; Moser, Frank; Katzir, Abraham

    1996-04-01

    We have developed core/clad polycrystalline silver halide optical fibers with a loss of roughly 0.3 dB/m at 10.6 micrometers. Such fibers, with core diameters 0.3 - 0.6 mm and lengths of 1 to 2 meters are capable of continuously delivering output power densities as high as 14 KW/cm2. The fibers were repetitively bent in the plastic and elastic regimes and the optical transmission monitored during bending. The mechanical properties of the core/clad fibers and of the core only fibers are similar. It was also demonstrated that the 'bending' properties of the core/clad fibers are determined by the cladding material. Our investigations suggest that proper design of the core/clad structure may give significant improvement in mechanical properties such as more cycles to optical failure. This will be very important especially for endoscopic laser surgery and other medical applications.

  6. The ballooning of fuel cladding tubes: theory and experiment

    International Nuclear Information System (INIS)

    Under some conditions, fuel clad ballooning can result in considerable strain before rupture. If ballooning were to occur during a loss-of-coolant accident (LOCA), the resulting substantial blockage of the sub-channel would restrict emergency core cooling. However, circumferential temperature gradients that would occur during a LOCA may significantly limit the average strain at failure. Understandably, the factors that control ballooning and rupture of fuel clad are required for the analysis of a LOCA. Considerable international effort has been spent on studying the deformation of Zircaloy fuel cladding under conditions that would occur during a LOCA. This effort has established a reasonable understanding of the factors that control the ballooning, failure time, and average failure strain of fuel cladding. In this paper, both the experimental and theoretical studies of the fuel clad ballooning are reviewed. (author)

  7. Grounds of VVER-1000 fuel cladding life control

    International Nuclear Information System (INIS)

    Highlights: • CET-method is physically grounded. • Coolant temperature and fuel rearrangement algorithm are key factors. • CET-method allows us to control VVER fuel cladding durability. - Abstract: A VVER-1000 fuel element (FE) cladding failure estimation method based on creep energy theory (CET-method) is physically grounded. Using CET-method, the VVER-1000 regime and fuel design parameters that determine cladding failure conditions are found. It is shown that FE cladding rupture life at normal variable loading operation conditions can be controlled by an optimal assignment of coolant temperature regime and fuel assembly (FA) rearrangement algorithm. Using a FA rearrangement efficiency criterion, it is shown that CET-method allows us to create an automated program-technical complex making control of FE cladding durability and optimization of fuel rearrangements in VVER-1000

  8. Cladding embrittlement during postulated loss-of-coolant accidents.

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.; Yan, Y.; Burtseva, T.; Daum, R.; Nuclear Engineering Division

    2008-07-31

    The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

  9. Application of Coating Technology for Accident Tolerant Fuel Cladding

    International Nuclear Information System (INIS)

    To commercialize the ATF cladding concepts, various factors are considered, such as safety under normal and accident conditions, economy for the fuel cycle, and developing development challenges, and schedule. From the proposed concepts, it is known that the cladding coating, FeCrAl alloy, and Zr-Mo claddings are considered as a near/mid-term application, whereas the SiC material is considered as a long-term application. Among them, the benefit of cladding coating on Zr-based alloys is the fuel cycle economy regarding the manufacturing, neutron cross section, and high tritium permeation characteristics. However, the challenge of cladding coating on Zr-based alloys is the lower oxidation resistance and mechanical strength at high-temperature than other concepts. Another important point is the adhesion property between the Zr-based alloy and coating materials. As an improved coating technology compared to a previous study, a 3D laser coating technology supplied with Cr powders is considered to make a coated cladding because it is possible to make a coated layer on the tubular cladding surface by controlling the 3-diminational axis. We are systematically studying the laser beam power, inert gas flow, cooling of the cladding tube, and powder control as key points to develop 3D laser coating technology. After Cr-coating on the Zr-based cladding, ring compression and ring tensile tests were performed to evaluate the adhesion property between a coated layer and Zr-based alloy tube at room temperature (RT), and a high-temperature oxidation test was conducted to evaluate the oxidation behavior at 1200 .deg. C of the coated tube samples. A 3D laser coating method supplied with Cr powders was developed to decrease the high-temperature oxidation rate in a steam environment through a systematic study for various coating parameters, and a Cr-coated Zircaloy-4 cladding tube of 100 mm in length to the axial direction can be successfully manufactured

  10. Experimental and numerical studies of crack growth in cladded specimens

    International Nuclear Information System (INIS)

    Behaviour of cracks at the inner surface of reactor pressure vessels cladded with a stainless steel layer is judged to be complicated due to differences in the properties of bases and cladding material. The scope of the present study is to form a methodology for analysis of such cracks. The J-integral was selected as a characterizing candidate for initiation and crack growth. The test material was of A533-B steel which clad layered using a commercial strip welding process. Two layers, the first of type 309 and the second of type 308 austenitic stainless steel were applied. In addition, cladding material was provided for fabrication of homogeneous specimens. The fracture resistance properties were developed independently for cladding and base material using homogeneous specimens of each material. Side-grooved bend specimens of type three-points-bending were used in the testing program. It was observed that the cladding was anisotropic with the lowest yield strength in the thickness direction. A fracture toughness between 175 to 184 MPa√m at 60 degree C was obtained for the cladding material. The transferability of fracture results between homogeneous and cladded specimens was studied in single edge notched bend specimens. Some cladded specimens were tested and the experimental data from one test were analyzed with the finite element method. The obtained 3D J-values were the compared with the J-values evaluated by using the measured crack extension in the cladded specimen and the JR-data of the respective material provided from homogeneous specimens. A reasonable good agreement was obtained in this comparison for a small amount of crack growth

  11. CEA studies on advanced nuclear fuel claddings for enhanced accident tolerant LWRs fuel (LOCA and beyond LOCA conditions)

    International Nuclear Information System (INIS)

    This paper gives an overview of CEA studies on advanced nuclear fuel claddings for enhanced Accident Tolerant LWR Fuel in collaboration with industrial partners AREVA and EDF. Two potential solutions were investigated: chromium coated zirconium based claddings and SiC/SiC composite claddings with a metallic liner. Concerning the first solution, the optimization of chromium coatings on Zircaloy-4 substrate has been performed. Thus, it has been demonstrated that, due in particular to their slower oxidation rate, a significant additional 'grace period( can be obtained on high temperature oxidized coated claddings in comparison to the conventional uncoated ones, regarding their residual PQ (Post-Quench) ductility and their ability to survive to the final water quenching in LOCA and, to some extent, beyond LOCA conditions. Concerning the second solution, the innovative 'sandwich' SiC/SiC cladding concept is introduced. Initially designed for the next generation of nuclear reactors, it can be adapted to obtain high safety performance for LWRs in LOCA conditions. The key findings of this work highlight the low sensitivity of SiC/SiC composites under the explored steam oxidation conditions. No signification degradation of the mechanical properties of CVI-HNI SiC/SiC specimen is particularly acknowledged for relatively long duration (beyond 100 h at 1200 Celsius degrees). Despite these very positive preliminary results, significant studies and developments are still necessary to close the technology gap. Qualification for nuclear application requires substantial irradiation testing, additional characterization and the definition of design rules applicable to such a structure. The use of a SiC-based fuel cladding shows promise for the highest temperature accident conditions but remains a long term perspective

  12. Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings - Phase II

    Energy Technology Data Exchange (ETDEWEB)

    Blough, J.L.; Krawchuk, M.T.; Van Weele, S.F. [Foster Wheeler Development Corp., Livingston, NJ (United States)

    1995-08-01

    A number of developmental and commercial tubing alloys and claddings have previously been exposed in Phase I to laboratory fireside corrosion testing simulating a superheater or reheater in a coal-fired boiler. This program is exposing samples of TP 347, RA-85H, HR-3C, 253MA, Fe{sub 3}Al + 5Cr, 310 modified, NF-709, 690 clad, and 671 clad, which showed good corrosion resistance from Phase 1, to the actual operating conditions of a 250-MW, coal-fired boiler. The samples were installed on air-cooled, retractable corrosion probes, installed in the reheater cavity, and are being controlled to the operating metal temperatures of an existing and advanced-cycle coal-fired boiler. The exposure will continue for 4000, 12,000, and 16,000 hours of operation. After the three exposure times, the samples will be metallurgically examined to determine the wastage rates and mode of attack. The probes were commissioned November 16, 1994. The temperatures are being recorded every 15 minutes, and the weighted average temperature calculated for each sample. Each of the alloys is being exposed to a temperature in each of two temperature bands-1150 to 1260{degrees}F and 1260 to 1325{degrees}F. After 2000 hours of exposure, one of the corrosion probes was cleaned and the wall thicknesses were ultrasonically measured. The alloy performance from the field probes will be discussed.

  13. Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings -- Phase 2 field testing

    Energy Technology Data Exchange (ETDEWEB)

    Blough, J.L.; Seitz, W.W.; Girshik, A. [Foster Wheeler Development Corp., Livingston, NJ (United States)

    1998-06-01

    In Phase 1 of this project, laboratory experiments were performed on a variety of developmental and commercial tubing alloys and claddings by exposing them to fireside corrosion tests which simulated a superheater or reheater in a coal-fired boiler. Phase 2 (in situ testing) has exposed samples of 347, RA85H, HR3C, RA253MA, Fe{sub 3}Al + 5Cr, Ta-modified 310, NF 709, 690 clad, 671 clad, and 800HT for up to approximately 16,000 hours to the actual operating conditions of a 250-MW, coal-fired boiler. The samples were installed on air-cooled, retractable corrosion probes, installed in the reheater cavity, and controlled to the operating metal temperatures of an existing and advanced-cycle, coal-fired boiler. Samples of each alloy were exposed for 4,483, 11,348, and 15,883 hours of operation. The present results are for the metallurgical examination of the corrosion probe samples after the full 15,883 hours of exposure. A previous topical report has been issued for the 4,483 hours of exposure.

  14. Novel strip-cast Mg/Al clad sheets with excellent tensile and interfacial bonding properties

    Science.gov (United States)

    Kim, Jung-Su; Lee, Dong Ho; Jung, Seung-Pill; Lee, Kwang Seok; Kim, Ki Jong; Kim, Hyoung Seop; Lee, Byeong-Joo; Chang, Young Won; Yuh, Junhan; Lee, Sunghak

    2016-06-01

    In order to broaden industrial applications of Mg alloys, as lightest-weight metal alloys in practical uses, many efforts have been dedicated to manufacture various clad sheets which can complement inherent shortcomings of Mg alloys. Here, we present a new fabrication method of Mg/Al clad sheets by bonding thin Al alloy sheet on to Mg alloy melt during strip casting. In the as-strip-cast Mg/Al clad sheet, homogeneously distributed equi-axed dendrites existed in the Mg alloy side, and two types of thin reaction layers, i.e., γ (Mg17Al12) and β (Mg2Al3) phases, were formed along the Mg/Al interface. After post-treatments (homogenization, warm rolling, and annealing), the interfacial layers were deformed in a sawtooth shape by forming deformation bands in the Mg alloy and interfacial layers, which favorably led to dramatic improvement in tensile and interfacial bonding properties. This work presents new applications to multi-functional lightweight alloy sheets requiring excellent formability, surface quality, and corrosion resistance as well as tensile and interfacial bonding properties.

  15. Novel strip-cast Mg/Al clad sheets with excellent tensile and interfacial bonding properties.

    Science.gov (United States)

    Kim, Jung-Su; Lee, Dong Ho; Jung, Seung-Pill; Lee, Kwang Seok; Kim, Ki Jong; Kim, Hyoung Seop; Lee, Byeong-Joo; Chang, Young Won; Yuh, Junhan; Lee, Sunghak

    2016-01-01

    In order to broaden industrial applications of Mg alloys, as lightest-weight metal alloys in practical uses, many efforts have been dedicated to manufacture various clad sheets which can complement inherent shortcomings of Mg alloys. Here, we present a new fabrication method of Mg/Al clad sheets by bonding thin Al alloy sheet on to Mg alloy melt during strip casting. In the as-strip-cast Mg/Al clad sheet, homogeneously distributed equi-axed dendrites existed in the Mg alloy side, and two types of thin reaction layers, i.e., γ (Mg17Al12) and β (Mg2Al3) phases, were formed along the Mg/Al interface. After post-treatments (homogenization, warm rolling, and annealing), the interfacial layers were deformed in a sawtooth shape by forming deformation bands in the Mg alloy and interfacial layers, which favorably led to dramatic improvement in tensile and interfacial bonding properties. This work presents new applications to multi-functional lightweight alloy sheets requiring excellent formability, surface quality, and corrosion resistance as well as tensile and interfacial bonding properties. PMID:27245687

  16. Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings -- Phase 2

    Energy Technology Data Exchange (ETDEWEB)

    Blough, J.L.; Seitz, W.W. [Foster Wheeler Development Corp., Livingston, NJ (United States)

    1997-12-01

    In Phase 1 a variety of developmental and commercial tubing alloys and claddings were exposed to laboratory fireside corrosion testing simulating a superheater or reheater in a coal-fired boiler. Phase 2 (in situ testing) has exposed samples of 347 RA-85H, HR3C, 253MA, Fe{sub 3}Al + 5Cr, 310 Ta modified, NF 709, 690 clad, and 671 clad for approximately 4,000, 12,000, and 16,000 hours to the actual operating conditions of a 250-MW coal-fired boiler. The samples were assembled on an air-cooled, retractable corrosion probe, the probe was installed in the reheater activity of the boiler and controlled to the operating metal temperatures of an existing and advanced-cycle coal-fired boiler. The results will be presented for the preliminary metallurgical examination of the corrosion probe samples after 16,000 hours of exposure. Continued metallurgical and interpretive analysis is still on going.

  17. Cladding metallurgy and fracture behavior during reactivity-initiated accidents at high burnup

    International Nuclear Information System (INIS)

    High-burnup fuel failure during a reactivity-initiated accident has been the subject of safety-related concern. Because of wide variations in metallurgical and simulation test conditions, it has been difficult to understand the complex failure behavior from major tests in NSRR and CABRI reactors. In this paper, a failure model based on fracture toughness and microstructural characteristics is proposed in which fracture toughness of high-burnup cladding is assumed to be sensitive to temperature and exhibit ductile-brittle transition phenomena similar to those of irradiated bcc alloys. Significant effects of temperature and shape of the pulse are predicted when a simulated test is conducted near the material's transition temperature. Temperature dependence of fracture toughness is, in turn, sensitive to cladding microstructure such as density, distribution, and orientation of hydrides, oxygen distribution in the metallic phase, and irradiation-induced damage. Because all these factors are strongly influenced by corrosion, the key parameters that influence susceptibility to failure are oxide layer thickness and hydriding behavior. Therefore, fuel failure is predicted to be strongly dependent on cladding axial location as well as on burnup. 10 figs, 21 refs

  18. Tensile Hoop Behavior of Irradiated Zircaloy-4 Nuclear Fuel Cladding

    International Nuclear Information System (INIS)

    A method for evaluating the room temperature ductility behavior of irradiated Zircaloy-4 nuclear fuel cladding has been developed and applied to evaluate tensile hoop strength of material irradiated to different levels. The test utilizes a polyurethane plug fitted within a tubular cladding specimen. A cylindrical punch is used to compress the plug axially, which generates a radial displacement that acts upon the inner diameter of the specimen. Position sensors track the radial displacement of the specimen outer diameter as the compression proceeds. These measurements coupled with ram force data provide a load-displacement characterization of the cladding response to internal pressurization. The development of this simple, cost-effective, highly reproducible test for evaluating tensile hoop strain as a function of internal pressure for irradiated specimens represents a significant advance in the mechanical characterization of irradiated cladding. In this project, nuclear fuel rod assemblies using Zircaloy-4 cladding and two types of mixed uranium-plutonium oxide (MOX) fuel pellets were irradiated to varying levels of burnup. Fuel pellets were manufactured with and without thermally induced gallium removal (TIGR) processing. Fuel pellets manufactured by both methods were contained in fuel rod assemblies and irradiated to burnup levels of 9, 21, 30, 40, and 50 GWd/MT. These levels of fuel burnup correspond to fast (E > 1 MeV) fluences of 0.27, 0.68, 0.98, 1.4 and 1.7 1021 neutrons/cm2, respectively. Following irradiation, fuel rod assemblies were disassembled; fuel pellets were removed from the cladding; and the inner diameter of cladding was cleaned to remove residue materials. Tensile hoop strength of this cladding material was tested using the newly developed method. Unirradiated Zircaloy-4 cladding was also tested. With the goal of determining the effect of the two fuel types and different neutron fluences on clad ductility, tensile hoop strength tests were

  19. Interaction between U-9 wt. % Mo fuel and Zr-1 wt. % Nb cladding alloys

    International Nuclear Information System (INIS)

    A few very recent studies have identified the ability of Zr in acting as a diffusion barrier to reduce the deleterious fuel-clad chemical interaction (FCCI), which restricts nuclear fuel designers around the world to successfully utilize the potential of the γ-phase stabilized U-Mo alloys as reduced enrichment fuels in research and test reactor. Further investigations pertaining to metallurgical interaction between U-Mo and Zr are essential not only to establish Zr as a diffusion barrier in U-Mo fuel but also to envisage Zr-base alloys as cladding as against the currently used Al-alloys. In this work, metallurgical interaction between U-9 wt. % Mo metallic fuel alloy and Zr-1 wt.% Nb clad material has been assessed through scanning electron microscopy (SEM), electron probe microanalysis (EPMA) and transmission electron microscopy (TEM). Interdiffusion of constituent elements across the fuel-clad interface, together with the phase reactions occurring at high temperature and during subsequent cooling, resulted in development of a layered interaction zone where coexistence of a bcc solid solution phase with varying compositions, along with α-U, α-Zr and Mo2Zr phases could be noticed. The instability in the γ-U(Mo,Zr) matrix leading to phase separation into α-U and α-Zr phases and the orientation relationships amongst them were established through microdiffraction and composite selected area electron diffraction (SAED) patterns, respectively. The present study is an endeavor to rationalize these observations, which remain unexplained in literature. (author)

  20. Creep anisotropy of Zircaloy cladding tubes

    International Nuclear Information System (INIS)

    First of all, a survey is given on the texture of Zircaloy cladding tubes obtained depending on the manufacturing conditions, and the state of knowledge on the anisotropy of the mechanical properties of the zirconium alloys connected with the texture is outlined. Theoretical formulations are set up for the phenomenological representation of the anisotropic creep. The results of tension and compression tests and the thus obtained creep site curves exhibit distinct differences with tubes having different textures. Furthermore, on asymmetry regarding compressive tensile stress is found in such a manner that the material under compression stress is more resistant to creep. Finally, discussions follow on the deformation mechanisms and a comparison with flow processes as well as indications on the significance of these creep results within the framework of fuel rod design are given. (IHoe/LH)

  1. Bending of pipes with inconel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Nachpitz, Leonardo; Menezes, Carlos Eduardo B.; Vieira, Carlos R. Tavares [Primus Processamento de Tubos S.A. (PROTUBO), Macae, RJ (Brazil)

    2009-07-01

    The high-frequency induction bending process, using API pipes coated with Inconel 625 reconciled to a mechanical transformation for a higher degree of resistance, was developed through a careful specification and control of the manufacturing parameters and inherent heat treatments. The effects of this technology were investigated by a qualification process consisting of a sequence of tests and acceptance criteria typically required by the offshore industry, and through the obtained results was proved the effectiveness of this entire manufacturing process, without causing interference in the properties and the quality of the inconel cladding, adding a gain of resistance to the base material, guaranteed by the requirements of the API 5L Standard. (author)

  2. Pretreatment of neutralized cladding removal waste sludge

    International Nuclear Information System (INIS)

    This report describes the status of process development for pretreating Hanford neutralized cladding removal waste (NCRW) sludge, of which ∼ 3.3 x 106 L is stored in Tanks 103-AW and 105-AW at the Hanford Site. The initial baseline process chosen for pretreating NCRW sludge is to dissolve the sludge in nitric acid and extract the -transuranic (MU) elements from the dissolved sludge solution with octyl(phenyl)-N,N-diisobutylcarbamoyl methyl phosphine oxide (CNWO). This process converts the NCRW sludge into a relatively large volume of low-level waste (LLW) to be disposed of as grout, leaving only a small volume of high-level waste (HLW) requiring vitrification in the Hanford Waste Vitrification Plant (HWVP)

  3. Correlation of waterside corrosion and cladding microstructure in high-burnup fuel and gadolinia rods

    International Nuclear Information System (INIS)

    Waterside corrosion of the Zircaloy cladding has been examined in high-burnup fuel rods from several BWRs and PWRs, as well as in 3 wt% gadolinia burnable poison rods obtained from a BWR. The corrosion behavior of the high-burnup rods was then correlated with results from a microstructural characterization of the cladding by optical, scanning-electron, and transmission-electron microscopy (OM, SEM, and TEM). OM and SEM examination of the BWR fuel cladding showed both uniform and nodular oxide layers 2 to 45 μm in thickness after burnups of 11 to 30 MWd/kgU. TEM characterization revealed a number of microstructural features that occurred in association with the intermetallic precipitates in the cladding metal, apparently as a result of irradiation-induced or -enhanced processes, i.e., (1) partial amorphization of large Zr-Fe-Cr and Zr-Fe-Ni precipitates (300 to 800 nm in size), (2) virtually complete amorphization of small intermetallic precipitates and subsequent dissolution of the alloying elements, and (3) spinodal-like fluctuation and redistribution of the alloying elements following the amorphization and dissolution. The BWR rods that exhibited white nodular oxides contained large precipitates (300 to 700 nm in size) that were partially amorphized during service, indicating that a distribution of the large intermetallic precipitates is conducive to nodular oxidation. Besides the microstructural features associated with the intermetallic precipitates, TEM stereo electron microscopy revealed microscopic zirconium hydrides (30 to 100 nm in size) that were too small to be resolved by OM or SEM. Stereoscopic examination revealed a tendency for precipitation of the microscopic hydrides with c-component dislocations as the burnup increased. Also, an examination of bright- and dark-field stereopair images revealed three-dimensional distributions of fine cubic-zirconium-oxide precipitates (5 to 10 nm in size) and unidentified ''black-dot'' (5 to 10 nm) and ''white

  4. Applicability of miniature specimen techniques for evaluating the mechanical properties of pressure tube and cladding material

    International Nuclear Information System (INIS)

    The Miniature Specimen Test Techniques (MSTT) namely Small Punch Test (SPT) and Ball Indentation Test (BIT) are commonly employed for evaluating the tensile properties of metallic materials. While discs of 3mm diameter with 0.25mm thickness are utilized for SPT method, the samples for BIT are of any shape with parallel polished top and bottom surfaces having thickness of at least 1mm. The SPT technique is based on driving a ball through clamped miniature disc specimens for deforming till it fractures whereas BIT involves multiple indentations, load and unload cycles at a single indentation point on a polished metallic surface by a spherical indenter. The specimens were fabricated from Zr-2.5%Nb pressure tube (PT) material that is used in pressurised heavy water reactor (PHWR). A suitable die-punch assembly was designed and developed in house to clamp the specimen to carry out the small punch test with the help of a screw-driven universal testing machine. In this work we have utilized miniature tensile specimens for evaluating and comparing the mechanical properties of PT material with that obtained from SPT and BIT. The tensile specimens were prepared using wire cut Electrical Discharge Machining (EDM) as per general guideline of ASTM standard E-8. Tests were carried out at ambient and higher temperatures. The tensile properties obtained from tension test and two MSTTs show that the tensile properties vary with orientation and temperature. In order to evaluate mechanical properties of cladding tube two techniques namely SPT and Ring Tension Test (RTT) have been used. The RTT is another technique, already established for estimation of the mechanical properties of cladding tube material in the transverse direction. Experimental results were generated at ambient and higher temperatures by preparing specimens from the same cladding tube in the form of 3mm discs and rings. The basic fixture that was used for carrying out ring tension test of the cladding tube consists

  5. Reliability of hard plastic clad silica fibers

    Science.gov (United States)

    Skutnik, Bolesh J.; Spaniol, Stefan

    2006-04-01

    New formulations of cladding materials have become available in recent times for Hard Plastic Clad Silica (HPCS) fibers, Initial data showed gains in some properties, particularly dynamic strength, especially for high numerical aperture (NA) fibers. A systematic study has been undertaken to determine the full strength and fatigue behavior of these HPCS fibers and to make comparisons to earlier HPCS fibers. Preliminary results, now confirmed, has shown improved median dynamic strength and higher Weibull slope. Full results are presented below including fatigue behavior and optical properties. These fibers have many applications and benefits in the high power delivery and medical laser uses as highlighted below. High power diode laser systems with their laser diode bars and arrays not only require special fibers to couple directly to the diode emitters, but also require special fibers to couple from the laser to application sites. These latter power delivery fibers are much larger than the internal fibers but still must be flexible, and have not only good strength but also good fatigue behavior. This particularly important industrial systems using robotic arms to apply the high power laser energy at a treatment site. The optical properties of HPCS fibers are well suited for the needs of the delivery of high power from diode laser bars and arrays to an application site. Benefits of strong median dynamic strengths and tighter flaw distributions in such cases will be discussed. Many medical applications, especially endoscopic ones, can benefit from the use of highly flexible, high NA, cost effective, HPCS optical fibers. Benefits of high strength and good fatigue behavior for such fibers in endoscopic procedures, including laser surgery, are discussed briefly including implications for mechanical reliability in medical and industrial settings.

  6. Influence of texture on fracture toughness of zircaloy cladding

    Energy Technology Data Exchange (ETDEWEB)

    Grigoriev, V. [Studsvik Material AB, Nykoeping (Sweden); Andersson, Stefan [Royal Inst. of Tech., Stockholm (Sweden)

    1997-06-01

    The correlation between texture and fracture toughness of Zircaloy 2 cladding has been investigated in connection with axial cracks in fuel rods. The texture of the cladding determines the anisotropy of plasticity of the cladding which, in turn, should influence the strain conditions at the crack-tip. Plastic strains in the cladding under uniaxial tension were characterised by means of the anisotropy constants F, G and H calculated according to Hill`s theory. Test temperatures between 20 and 300 deg C do not influence the F, G and H values. Any significant effect of hydrogen (about 500 wtppm) on the anisotropy constants F, G and H has not been revealed at a test temperature of 300 deg C. The results, obtained for stress-relieved and recrystallized cladding with different texture, show an obvious influence of texture on the fracture toughness of Zircaloy cladding. A higher fracture toughness has been found for cladding with more radial texture. With a 2 page summary in Swedish. 32 refs, 18 figs.

  7. Microstructure, Wear, and Corrosion Characteristics of TiC-Laser Surface Cladding on Low-Carbon Steel

    Science.gov (United States)

    El-Labban, Hashem F.; Mahmoud, Essam Rabea Ibrahim; Algahtani, Ali

    2016-04-01

    Laser cladding was used to produce surface composite layer reinforced with TiC particles on low-carbon steel alloy for improving the wear and corrosion resistances. The cladding process was carried out at powers of 2800, 2000, 1500, and 1000 W, and a fixed traveling speed of 4 mm/s. The produced layers are free from any cracks. Some of the TiC particles were melted and then re-solidified in the form of fine acicular dendrites. The amount of the melted TiC was increased by increasing the laser power. The hardness of the produced layers was improved by about 19 times of the base metal. Decreasing laser power led to hardness increment at the free surface. The improvement in wear resistance was reached to about 25 times (in case of 1500 W) of the base metal. Moreover, the corrosion resistance shows remarkable improvement after the laser treatment.

  8. Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

    Science.gov (United States)

    Karahan, Aydın; Kazimi, Mujid S.

    2013-10-01

    The study evaluates the possible use of graphite foam as the bonding material between U-Pu-Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U-15Pu-6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600-660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors.

  9. Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

    International Nuclear Information System (INIS)

    The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors

  10. Development of dissimilar metal transition joint by hot bond rolling

    International Nuclear Information System (INIS)

    Metallurgically bonded transition joints which enable to connect reprocessing equipments made of superior corrosion resistant valve metals (Ti-5Ta, Zr or Ti) to stainless steel piping are required for nuclear fuel reprocessing plants. The authors have developed dissimilar transition joints made of stainless steel and Ti-5Ta, Zr or Ti with an insert metal of Ta by the hot bond rolling process of clad bars and clad pipes, using a newly developed mill called 'rotary reduction mill'. This report presents the manufacturing process of dissimilar transition joints produced from the clad pipe with three layers by the hot bond rolling. First, the method of hot bond rolling of clad pipe is proposed. Then, the mechanical and corrosion properties of the dissimilar transition joints are evaluated in detail by carrying out various tests. Finally, the rolling properties in the clad pipe method are discussed. (author)

  11. Design of an integrated system to recycle Zircaloy cladding using a hydride–milling–dehydride process

    Energy Technology Data Exchange (ETDEWEB)

    Kelley, Randy, E-mail: rkelley@pitt.edu [Mechanical Engineering Department, 236 Engineering and Science Building, University of Pittsburgh – Johnstown, Johnstown, PA 15904 (United States); McDeavitt, Sean [Texas A and M University, Department of Nuclear Engineering, 327 Zachry Engineering Center, 3133 TAMU, College Station, TX 77843 (United States)

    2013-10-15

    Highlights: • Dehydriding zirconium hydride was studied at relatively low temperatures (<800 °C). • High vacuum pressures decrease dehydriding temperatures. • Specialized equipment was designed, built and demonstrated to process zirconium. • The process hydrided–milled–dehydrided zirconium metal to a fine metal powder. • Two powder samples were analyzed and proved the operation of the machine. -- Abstract: A hydride–dehydride process was evaluated to recover a portion of spent nuclear fuel cladding; a zirconium alloy (Zircaloy), as a metal powder that may be used for advanced nuclear fuel applications. The investigation was part of a broader study that sought to determine the viability of recovering components of used nuclear fuel to for a metal matrix cermet for transuranic burning. The zirconium powder process begins with the conversion of Zircaloy cladding hulls into a brittle zirconium hydride, which is easily pulverized into a powder. The dehydriding process removes hydrogen by heating the powder in a vacuum, resulting in a zirconium metal powder. In support of this, a specialized piece of equipment was designed to demonstrate the entire zirconium conversion process to transform Zircaloy tubes into metal powder without intermediate handling. This was accomplished by building a milling system that rotates inside of controlled atmosphere chamber with an internal heater. The hydriding process was accomplished using an argon–5% hydrogen atmosphere at 500 °C. The process variables for the dehydriding process were determined using a thermogavimetric analysis (TGA) method. It was determined that a rough vacuum (∼0.001 bar) and 800 °C were sufficient to decompose the zirconium hydride. Zirconium metal powder was created using different milling times: 45 min (coarse powder) and 12 h (fine powder). X-ray diffraction (XRD) analysis indicated that the process produced a zirconium metal. Additionally, visual observations of the samples silvery

  12. The Information Sources in Building Cladding Supply Chain

    Directory of Open Access Journals (Sweden)

    Qiang Du

    2012-11-01

    Full Text Available The increasing complexity of the cladding procurement and fragmentation of the supply chain bring challenges for information management. The purpose of this research was to identify the issues concerning the information sources. A questionnaire survey was the key method of this research, while industry meetings and informal interviews were employed to provide in-depth understanding of the communication related issues in the industry. It was found that the participants of the cladding supply chain were experiencing difficulties of identifying information sources and accessing information, and that an industry level third party acting as an independent information source could be accepted by the cladding industry.

  13. Increasing corrosion resistance of carbon steels by surface laser cladding

    Science.gov (United States)

    Polsky, V. I.; Yakushin, V. L.; Dzhumaev, P. S.; Petrovsky, V. N.; Safonov, D. V.

    2016-04-01

    This paper presents results of investigation of the microstructure, elemental composition and corrosion resistance of the samples of low-alloy steel widely used in the engineering, after the application of laser cladding. The level of corrosion damage and the corrosion mechanism of cladded steel samples were established. The corrosion rate and installed discharge observed at the total destruction of cladding were obtained. The regularities of structure formation in the application of different powder compositions were obtained. The optimal powder composition that prevents corrosion of samples of low-carbon low-alloy steel was established.

  14. Fuel compliance model for pellet-cladding mechanical interaction

    International Nuclear Information System (INIS)

    This paper describes two aspects of fuel pellet deformation that play significant roles in determining maximum cladding hoop strains during pellet-cladding mechanical interaction: compliance of fragmented fuel pellets and influence of the pellet end-face design on the transmission of axial compressive force in the fuel stack. The latter aspect affects cladding ridge formation and explains several related observations that cannot be explained by the hourglassing model. An empirical model, called the fuel compliance model and representing the above aspects of fuel deformation, has been developed using the results from two Halden experiments and incorporated into the FRAP-T6 fuel performance code

  15. Postirradiation metallurgical techniques to estimate LOFT peak cladding temperatures

    International Nuclear Information System (INIS)

    A comprehensive review of the zircaloy microstructure and oxidation characteristics was performed to establish the state-of-the-art capability for estimating peak cladding temperatures for LOCE's by postirradiation examination techniques. Cladding oxidation characteristics were compared with microstructure characteristics to determine the best approach. Deficiencies were identified in the low-temperature oxidation characteristics required for LOCE's, and an experimental plan is proposed to obtain low-temperature oxidation data which will expand the current data base and confirm the feasibility of estimating zircaloy cladding temperatures for transient conditions

  16. Study on Cracking Tendency and Mechanism of Gray Cast Iron Laser Cladding

    Institute of Scientific and Technical Information of China (English)

    YE Hong; YAN Zhong-lin; HUANG Qi; YANG Hui

    2004-01-01

    In this paper, NiCrSiB and CoWC35 powder has been used in laser cladding of gray cast iron. The cracking tendency has also been discussed. The cracks have been observed with a scan electron microscopy to analyze the cracking mechanism. The results show that cracks have not appeared in NiCrSiB cladding. Nevertheless, the cracking tendency of CoWC35 cladding is extremely high and there are both cold cracks and hot cracks in the cladding. The cracking tendency of laser cladding depends on physical properties of the cladding material and plasticity and roughness of the cladding.

  17. Study on Cracking Tendency and Mechanism of Gray Cast Iron Laser Cladding

    Institute of Scientific and Technical Information of China (English)

    YEHong; YANZhong-lin; HUANGQi; YANGHui

    2004-01-01

    In this paper, NiCrSiB and COWC35 powder has been used in laser cladding of gray cast iron. The cracking tendency has also been discussed. The cracks have been observed with a scan electron microscopy to analyze the cracking mechanism. The results show that cracks have not appeared in NiCrSiB cladding. Nevertheless, the cracking tendency of CoWC35 cladding is extremely high and there are both cold cracks and hot cracks in the cladding. The cracking tendency of laser cladding depends on physical properties of the cladding material and plasticity and roughness of the cladding.

  18. Characteristics of Ni-based coating layer formed by laser and plasma cladding processes

    Energy Technology Data Exchange (ETDEWEB)

    Xu Guojian [Materials, Physics and Energy Engineering, Graduate School of Engineering, Nagoya University, 1 Furo-cho, Chikusa-ku, Nagoya 464-8603 (Japan)]. E-mail: xuguojian1959@hotmail.com; Kutsuna, Muneharu [Materials, Physics and Energy Engineering, Graduate School of Engineering, Nagoya University, 1 Furo-cho, Chikusa-ku, Nagoya 464-8603 (Japan)]. E-mail: kutsuna@numse.nagoya-u.ac.jp; Liu Zhongjie [Materials, Physics and Energy Engineering, Graduate School of Engineering, Nagoya University, 1 Furo-cho, Chikusa-ku, Nagoya 464-8603 (Japan)]. E-mail: xyliuzhj8@hotmail.com; Zhang Hong [Changchun University of Science and Technology, 7 Weixing Road, Changchun, Jilin Province 130022 (China)]. E-mail: Zhanghongcust@hotmail.com

    2006-02-15

    The clad layers of Ni-based alloy were deposited on the SUS316L stainless plates by CO{sub 2} laser and plasma cladding processes. The smooth clad bead was obtained by CO{sub 2} laser cladding process. The phases of clad layer were investigated by an optical microscope, scanning electron microscopy (SEM), X-ray diffractometer (XRD), electron probe microanalysis (EPMA) and energy-dispersive spectrometer (EDS). The microstructures of clad layers belonged to a hypereutectic structure. Primary phases consist of boride CrB and carbide Cr{sub 7}C{sub 3}. The eutectic structure consists of Ni + CrB or Ni + Cr{sub 7}C{sub 3}. Compared with the plasma cladding, the fine microstructures, low dilutions, high Vickers hardness and excellent wear resistance were obtained by CO{sub 2} laser cladding. All that show the laser cladding process has a higher efficiency and good cladding quality.

  19. Verification of cladding performance analysis models in INFRA

    International Nuclear Information System (INIS)

    Recent trend of PWR fuel development is continuously increasing the burnup to improve the economy as well as the safety. Development of high burnup fuel raised the new issues for the fuel behaviour that was not considered beyond the high burnup. High burnup fuel performance code, INFRA (INtegraed Fuel Rod Analysis), was developed for the prediction of high burnup fuel behavior. Cladding performance models such as creep model, creep-out model, corrosion model and axial irradiation growth model were developed to analyze the performance of high burnup zircaloy-4 cladding. Cladding performance analysis were performed to verify the cladding performance model in INFRA by using the performance data of commercial PWR and halden reactor, etc. INFRA predicted the measured data reasonably well

  20. Manufacturing Technology and Application Trends of Titanium Clad Steel Plates

    Institute of Scientific and Technical Information of China (English)

    Hang SU; Xiao-bing LUO; Feng CHAI; Jun-chang SHEN; Xin-jun SUN; Feng LU

    2015-01-01

    Some of the major manufacturing processes and corresponding mechanical properties of titanium clad steel plates were analyzed, and the consequences of research, manufacturing, and application of titanium clad steel plates in both markets of China and overseas were also summarized. As an economical and environmentally friendly technology, the roll bonding process is ex-pected to become the next-generation mainstream process for the manufacturing of titanium clad steel plate. Some of the crucial and most important technical problems of this particular process, including vacuum sealing technology, surface treatment process technology, application of a transition layer, and rolling process, were discussed along with the advantageous mechanical properties and life-cycle economy of these plates processed by this technology. Finally, the market needs, application trends, and requirements of titanium clad steel plate were also considered from industries of petrochemical, shipbuilding, marine, and electric power.

  1. Chalcogenide optical microwires cladded with fluorine-based CYTOP.

    Science.gov (United States)

    Li, Lizhu; Abdukerim, Nurmemet; Rochette, Martin

    2016-08-22

    We demonstrate optical transmission results of highly nonlinear As2Se3 optical microwires cladded with fluorine-based CYTOP, and compare them with microwires cladded with typical hydrogen-based polymers. In the linear optics regime, the CYTOP-cladded microwire transmits light in the spectral range from 1.3 µm up to >2.5 µm without trace of absorption peaks such as those observed using hydrogen-based polymer claddings. The microwire is also pumped in the nonlinear optics regime, showing multiple-orders of four-wave mixing and supercontinuum generation spanning from 1.0 µm to >4.3 µm. We conclude that with such a broadband transparency and high nonlinearity, the As2Se3-CYTOP microwire is an appealing solution for nonlinear optical processing in the mid-infrared. PMID:27557174

  2. Chemical vapor deposition for silicon cladding on advanced ceramics

    Science.gov (United States)

    Goela, Jitendra S.; Taylor, Raymond L.

    1989-01-01

    Polycrystalline Si was used to clad several advanced ceramic materials such as SiC, Si3N4, sapphire Al2O3, pyrolytic BN, and Si by a CVD process. The thickness of Si cladding ranged from 0.025 to 3.0 mm. CVD Si adhered quite well to all the above materials except Al3O, where the Si cladding was highly stressed and cracked or delaminated. A detailed material characterization of Si-clad SiC samples showed that Si adherence to SiC does not depend much on the substrate surface preparation; that the thermal cycling and polishing of the samples do not cause delamination; and that, in four-point bend tests, the Si-SiC bond remains intact, with the failure occurring in the Si.

  3. High temperature steam oxidation studies on Zircaloy-2 cladding

    International Nuclear Information System (INIS)

    A detailed metallographic characterization and ring compression tests at ambient temperature were carried out on the oxidized clad tube pieces. The results of these experiments are presented in this paper

  4. High Temperature Resistance Claddings for Nuclear Thermal Rockets Project

    Data.gov (United States)

    National Aeronautics and Space Administration — This program will develop a series of nano-/micro-composite coated nuclear reactor facing components using MesoCoat's CermaCladTM process. This proposed SBIR...

  5. Special techniques for tensile tests of irradiated zirconium claddings

    International Nuclear Information System (INIS)

    Irradiated zirconium alloy claddings possessing property anisotropy should be tested in transverse and longitudinal directions. Such mechanical tests can be performed in conditions of large variety of geometric peculiarities of specimens, supports or grips. The objective of the work is the development of the unified complex of updated special techniques that allow investigation of mechanical pre- and post-irradiation properties of VVER claddings including radiation effect of property anisotropy changes in the same way. (author)

  6. First results on the effect of fuel-cladding eccentricity

    International Nuclear Information System (INIS)

    In the traditional fuel-behaviour or hot channel calculations it is assumed that the fuel pellet is centered within the clad. However, in the real life the pellet could be positioned asymmetrically within the clad, which leads to asymmetric gap conductance and therefore it is worthwhile to investigate the magnitude of the effect on maximal fuel temperature and surface heat flux. In this paper our first experiences are presented on this topic. (authors)

  7. Supercontinuum Generation in a Microstructured Fiber with an Irregular Cladding

    Science.gov (United States)

    Minkovich, V. P.; Sotsky, A. B.; Vaca Pereira G., M.; Dzen, I. S.; Sotskaya, L. I.

    2016-05-01

    A broad-band supercontinuum generation was obtained at excitation of a microstructured optical fiber with an irregular cladding by femtosecond laser pulses. To explain the experimental data, calculations of the mode characteristics of microstructured fibers were performed. It was shown that the creation of air channels with different radii in the fiber cladding makes it possible to involve both the fundamental and high fiber modes in the supercontinuum generation that helps to increase the width of the generation spectrum.

  8. Facility for in-reactor creep testing of fuel cladding

    International Nuclear Information System (INIS)

    A biaxial stress creep test facility has been designed and developed for operation in the WR-1 reactor. This report outlines the rationale for its design and describes its construction and the operating experience with it. The equipment is optimized for the determination of creep data on CANDU fuel cladding. Typical results from Zr-2.5 wt% Nb fuel cladding are used to illustrate the accuracy and reliability obtained. (author)

  9. Interim report on the creepdown of Zircaloy fuel cladding

    International Nuclear Information System (INIS)

    This report describes the creepdown phenomenon in Zircaloy fuel cladding and the methods by which it will be measured and analyzed. Instrumentation for monitoring radial deformation in the cladding is described in detail--in terms of theory, design, and stability. The programs that control the microcomputer are listed, both to document the level of sophistication of the instrumentation and to indicate the flexibility of the test equipment

  10. Theoretical analysis of radiation-balanced double clad fiber laser

    Institute of Scientific and Technical Information of China (English)

    CHEN Ji-xin; SUI Zhan; CHEN Fu-shen; LI Ming-zhong; WANG Jian-jun

    2005-01-01

    In this letter,a theoretical model of radiation-balanced double clad fiber laser is presented.The characteristic of the laser with Yb doped double clad fiber is analyzed numerically.It is concluded that high output laser power can be obtained by selecting output coupling mirror with lower reflectivity,improving Yb doped concentration and choosing fiber length. This result can help us to design radiation balanced fiber laser.

  11. Residual stresses in weld-clad reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Cladding of low alloy nuclear reactor pressure vessel steel with austenitic stainless steel introduces in heavy section components high residual stresses which may cause microcrack formation in stress relief heat treatment. In this investigation an attempt is made to contribute to the solution of the stress relief cracking problem by determining quantitatively the magnitude and distribution of the residual stresses after cladding and after subsequent stress relief heat treatment. The distribution of residual stresses was determined on the basis of a combined experimental-mathematical procedure. Heavy section plate specimens of low alloy steel as base material were given an austenitic monolayer-cladding using the techniques of strip electrode and plasma hot wire cladding, respectively. A number of plates was stress relief heat treated. Starting from the cladded surface the thickness of the plates was reduced by subsequent removal of layers of material. The elastic strain reaction to the removal of each layer was measured by strain gauges. From the data obtained the biaxial residual stress distribution was computed as a function of thickness using relations which are derived for this particular case. In summary, lower residual stresses are caused by reduced thickness of the components. As the heat input, is decreased at identical base material thickness, the residual stresses are lowered also. The height of the tensile residual stress peak, however, remains approximataly constant. In stress relief annealed condition the residual stresses in the cladding are in tension; in the base material the residual stresses are negligibly small

  12. Phase transformations at steel/IN626 clad interfaces

    Science.gov (United States)

    Ayer, Raghavan; Mueller, R. R.; Leta, D. P.; Sisak, W. J.

    1989-04-01

    The microstructures of 4130 and 2.25Cr-1Mo steels clad to nickel base IN625 by welding and HIPing were examined by Analytical Electron Microscopy (AEM) and Secondary Ion Mass Spectroscopy (SIMS) to determine the interfacial microstructural characteristics which could affect their mechanical properties and corrosion resistance. The interface microstructures of the clads produced by the two methods were considerably different. The clad produced by welding was characterized by a low density of carbide precipitates confined to a very narrow region (˜1 μm) at the interface of ferrite and austenite. In addition, a thin region of untempered martensite was present at the interface which could affect its resistance to hydrogen embrittlement as well as other mechanical properties. The interface of the HIP clad composite contained several regions of distinct microstructural characteristics with widely varying densities of carbide precipitates. Relative to the clad produced by welding, extensive precipitation was observed both in the steel and in the IN625 at the interface, separated by a region free from precipitation. The extent of precipitation at the interface regions appears to be controlled essentially by the extent of carbon transport across the interface. The article describes the detailed analysis of the interface characteristics, and models are proposed to explain the microstructural evolution at the interface of the HIP and weld clad composites.

  13. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    International Nuclear Information System (INIS)

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10−6 on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure

  14. A cladding oxidation model based on diffusion equations

    International Nuclear Information System (INIS)

    During severe accident in PWRs, the cladding oxidation with steam in the core is very important to the accident process. When oxidation time is long, or oxidation occurs in steam starvation conditions, the parabolic rate correlations based on experiments are restricted, which impacts the prediction of cladding failure, hydrogen production, and temperature. According to Fick's laws, a cladding oxidation model in a wide temperature range based on diffusion equations is developed. The developed oxidation model has a wider applicability than those parabolic rate correlations, and can simulate long-term experiments well. The restricted assumptions of short term oxidation time and enough steam environment in the core implemented by those parabolic rate correlations are removed in the model, therefore this model perfectly fit for long-term and steam starvation conditions which are more realistic during a severe accident. This model also can obtain detailed oxygen distribution in the cladding, which is helpful to simulate the cladding failure in detail and develop advanced cladding failure criteria. (authors)

  15. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyu-Tae, E-mail: ktkim@dongguk.ac.kr

    2013-10-15

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10{sup −6} on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure.

  16. Bragg grating induced cladding mode coupling due to asymmetrical index modulation in depressed cladding fibers

    DEFF Research Database (Denmark)

    Berendt, Martin Ole; Grüne-Nielsen, Lars; Soccolich, C.F.; Bjarklev, Anders Overgaard

    reduce this problem. None of these designs seems to give complete solutions. In particular, the otherwise promising depressed cladding design gives a pronounced coupling to one LP01 mode, this has been referred to as a Ghost grating. To find the modes of the fiber we have established a numerical mode......-solver based on the staircase-approximation method. The Bragg grating causes coupling between the fundamental LP01 mode and higher order LP1p modes that satisfy phase-matching. The coupling strength is determined by the overlap integral of the LP01, the LP1p mode, and the UV-induced index perturbation. For LP0...

  17. Cladding Attachment Over Thick Exterior Insulating Sheathing

    Energy Technology Data Exchange (ETDEWEB)

    Baker, P.; Eng, P.; Lepage, R.

    2014-01-01

    The addition of insulation to the exterior of buildings is an effective means of increasing the thermal resistance of both wood framed walls as well as mass masonry wall assemblies. For thick layers of exterior insulation (levels greater than 1.5 inches), the use of wood furring strips attached through the insulation back to the structure has been used by many contractors and designers as a means to provide a convenient cladding attachment location (Straube and Smegal 2009, Pettit 2009, Joyce 2009, Ueno 2010). The research presented in this report is intended to help develop a better understanding of the system mechanics involved and the potential for environmental exposure induced movement between the furring strip and the framing. BSC sought to address the following research questions: 1. What are the relative roles of the mechanisms and the magnitudes of the force that influence the vertical displacement resistance of the system? 2. Can the capacity at a specified deflection be reliably calculated using mechanics based equations? 3. What are the impacts of environmental exposure on the vertical displacement of furring strips attached directly through insulation back to a wood structure?

  18. Investigating mechanical behavior and radiation resistant of fuel rods clad in nuclear power plant

    International Nuclear Information System (INIS)

    interstitials in metal lattice under irradiation causes increased strength and hardness but decreases ductility in metals.The increase in strength and hardness depends on obstacles that prevent the motion of dislocations. The clustering of point defects are responsible for these changes. Irradiation also induces instabilities in phases due to enhancement of diffusion, solute segregation, precipitate formation, order- disorder transformation and resolution of small precipitates. From the microscopic point of view accumulation of vacancies accompanied by formation of He and H2 gases under irradiation cause an increase in volume which results in swelling and eventually ends up with embrittlement of metals. This subject was described in chapter three Zirconium and its alloys are the best structural materials for fuel cladding of BWR and PWR reactors core. The working condition in the core of nuclear reactor are very serve, respect temperature and radiation dose. It should be realized that, if fuel cladding receive damage and get cracked, the first cooling cycle and the maine equipment will be contaminated with active materials which cause additional environmental problems. Furthermore, replacement of fuel rods are very costly. Therefore, for increasing life time of fuel cladding and minimizing damage, the effect of radiation and heat on Zirconium and its alloys must be investigated. This subject was described in chapter four.The mechanical behavior and radiation resistant of fuel cladding in PWR reactor (specifically WWER ) have been investigated which is described in chapter five. Result, discussion and final conclusion are summarized in last chapter and also several points for improvement have been offered

  19. Bioactivity of fluorapatite/alumina composite coatings deposited on Ti6Al4V substrates by laser cladding

    Science.gov (United States)

    Chien, C. S.; Liu, C. W.; Kuo, T. Y.; Wu, C. C.; Hong, T. F.

    2016-04-01

    Hydroxyapatite (HA) is one of the most commonly used coating materials for metal implants. However, following high-temperature deposition, HA easily decomposes into an unstable phase or forms an amorphous phase, and hence, the long-term stability of the implant is reduced. Accordingly, the present study investigates the use of fluorapatite (FA) fortified with 20 wt% alumina (α-Al2O3) as an alternative biomedical coating material. The coatings are deposited on Ti6Al4V substrates using a Nd:YAG laser cladding process performed with laser powers and travel speeds of 400 W/200 mm/min, 800 W/400 mm/min and 1200 W/600 mm/min, respectively. The results show that for all of the specimens, a strong metallurgical bond is formed at the interface between the coating layer and the transition layer due to melting and diffusion. The XRD analysis results reveal that the cladding layers in all of the specimens consist mainly of FA, β-TCP, CaF2, Ti and θ-Al2O3 phases. In addition, the cladding layers of the specimens prepared using laser powers of 400 and 800 W also contain CaTiO3 and CaAl2O4, while that of the specimen clad using a power of 1200 W contains TTCP and CaO. Following immersion in simulated body fluid for 14 days, all of the specimens precipitate dense bone-like apatite and exhibit excellent bioactivity. However, among all of the specimens, the specimen that is prepared with a laser power of 800 W shows the best biological activity due to the presence of residual FA, apatite-generating CaTiO3 and a rough cladding layer surface.

  20. Laser cladding in-situ carbide particle reinforced Fe-based composite coatings with rare earth oxide addition

    Institute of Scientific and Technical Information of China (English)

    吴朝锋; 马明星; 刘文今; 钟敏霖; 张红军; 张伟明

    2009-01-01

    Particulate reinforced metal matrix composite(PR-MMC) has excellent properties such as good wear resistance,corrosion resistance and high temperature properties.Laser cladding is usually used to form PR-MMC on metal surface with various volume fractions of ceramic particles.Recent literatures showed that laser melting of powder mixture containing carbon and carbide-forming elements,was favorable for the formation of in-situ synthesized carbide particles.In this paper,rare earth oxide(RE2O3) was added into t...

  1. Status of Transuranic Bearing Metallic Fuel Development

    Energy Technology Data Exchange (ETDEWEB)

    Steve Hayes; Bruce Hilton; Heather MacLean; Debbie Utterbeck; Jon Carmack; Kemal Pasamehmetoglu

    2009-09-01

    This paper summarizes the status of the metallic fuel development under the Advanced Fuel Cycle Initiative (AFCI). The metallic fuel development program includes fuel fabrication, characterization, advanced cladding research, irradiation testing and post-irradiation examination (PIE). The focus of this paper is on the recent irradiation experiments conducted in the Advanced Test Reactor and some PIE results from these tests.

  2. Experimental and numerical investigation on cladding of corrosion-erosion resistant materials by a high power direct diode laser

    Science.gov (United States)

    Farahmand, Parisa

    advantages due to creating coating layers with superior properties in terms of purity, homogeneity, low dilution, hardness, bonding, and microstructure. In the development of modern materials for hardfacing applications, the functionality is often improved by combining materials with different properties into composites. Metal Matrix Composite (MMC) coating is a composite material with two constituent parts, i.e., matrix and the reinforcement. This class of composites are addressing improved mechanical properties such as stiffness, strength, toughness, and tribological and chemical resistance. Fabrication of MMCs is to achieve a combination of properties not achievable by any of the materials acting alone. MMCs have attracted significant attention for decades due to their combination of wear-resistivity, corrosion-resistivity, thermal, electrical and magnetic properties. Presently, there is a strong emphasis on the development of advanced functional coatings for corrosion, erosion, and wear protection for different industrial applications. In this research, a laser cladding system equipped with a high power direct diode laser associated with gas driven metal powder delivery system was used to develop advanced MMC coatings. The high power direct diode laser used in this study offers wider beam spot, shorter wavelength and uniform power distribution. These properties make the cladding set-up ideal for coating due to fewer cladding tracks, lower operation cost, higher laser absorption, and improved coating qualities. In order to prevent crack propagation, porosity, and uniform dispersion of carbides in MMC coating, cladding procedure was assisted by an induction heater as a second heat source. The developed defect free MMC coatings were combined with nano-size particles of WC, rare earth (RE) element (La2O3), and Mo as a refractory metal to enhance mechanical properties, chemical composition, and subsequently improve the tribological performance of the coatings. The resistance

  3. Out-of-pile fuel-clad chemical compatibility studies for fast reactors

    International Nuclear Information System (INIS)

    Fuel-clad mechanical and chemical interaction lead to clad corrosion, loss of ductility, embitterment and clad breach limiting the life of a fuel pin. The chemical nature and extent of clad attack depend upon the type of fuel, fuel-clad gap, type of bond between fuel and clad. For FBTR at Kalpakkam, fuel is (U0.3Pu0.7)Cl +x (MKI) and (U0.45Pu0.55)Cl +x (MKII) and the clad is AISI SS316 (20% CW). Extensive work on out-of-pile fuel-clad and fuel-clad-coolant chemical compatibility experiments has been carried out in Radiometallurgy Division. The paper highlights the results of the tests carried out and substantiate it on the basis of the available thermodynamic data. (author)

  4. The prediction of cladding performance in Ultra long Cycle Fast Reactors

    International Nuclear Information System (INIS)

    As a part of R and D activities for the development of advanced fast reactors, HT9 cladding performance of Ultra long Cycle Fast Reactor (UCFR) is evaluated in various cladding peak temperatures and design power levels (1000MWe and 1000MWe). The key design concept of UCFR is a non refueling during 30 to 60 years operation, and this concept may require the maximum cladding temperature of ∼650 .deg. C peak cladding temperature and cladding radiation damage of over 200dpa (displacements pet atom). Therefore, for the design of UCFR, challenges such as thermal creep, irradiation creep and swelling must be quantitatively evaluated. As a cladding material, HT9 shows distinguishably favorable properties for UCFR, In this study, therefore, key design parameters for the cladding performance will be evaluated for UCFR cladding design and resulted the prediction of life time of cladding in UCFR

  5. Development of nuclear fuel for the future -Development of performance improvement of the cladding by ion beam-

    International Nuclear Information System (INIS)

    In this research we analyzed the state of art related to the surface treatment method of nuclear fuel cladding for the development of the surface treatment technique of nuclear fuel cladding by ion beam while investigating major causes of the leakage of fuel rods. Ion implantation simulation code called TRIM-95 was used to decide basic parameters of ion beams and setup an appropriate process for ion implantation. Performance of the ion beam extraction was measured after adding the needed vacuum and cooling system to the existing gas and metal ion implanters. Target system for the ion implantation of fuel cladding improved and a plasma accelerator was installed on the target chamber of the metal ion implanter. The plasma accelerator is used to produce low energy, high current ion beams. The mechanical and chemical properties of the implanted Zircaloy-4 such as micro hardness, wear resistance, fretting wear, friction coefficient and corrosion resistance was measured under the room temperature and atmosphere. A micro structure and composition analysis of Zircaloy-4 sample was performed before and after the implantation to study the cause of the improvement in the mechanical and chemical characteristics. 94 figs, 11 tabs, 51 refs. (Author)

  6. Corrosion Resistance Evaluation of HANA Claddings in Commercial PWR

    International Nuclear Information System (INIS)

    Korea Atomic Energy Research Institute (KAERI) in collaboration with KEPCO Nuclear Fuel (KNF) developed newly-advanced alloy which are named HANA (High-performance Alloy for Nuclear Application) for high burnup PWR nuclear fuel, showed an excellent out-pile corrosion resistance in PWR simulating loop conditions. And in-pile corrosion resistance of HANA claddings, which was examined at the first provisional inspection after -185 FPD of irradiation in the Halden Reactor, and also shown superior to the other references alloy. Also, other researches showed a much better corrosion resistance when compared to the other Zr-based alloy in various corrosion conditions. In this study, the LTA program for newly-developed fuel assembly (HIPER) with the HANA claddings was implemented to justify the performance for 3 cycles of operation schedule in Hanul nuclear power plant. The objective of this study is to compare corrosion properties of reference alloy with HANA claddings loaded in Hanul nuclear power plant.. For the examination procedures, the oxide thickness measurements method and equipment of PSE are described in detail as follow in measurement methods chapter. Finally, based on the above mentioned measurements method, the summarized oxide thickness data obtained from PSE are evaluated for the corrosion resistance in commercial nuclear power plant and some discussion for the corrosion resistance are described. In the past, corrosion resistance of HANA claddings was successfully conducted in test reactor. In this study, the corrosion characteristic of HANA claddings which are applied to HIPER is examined in the commercial nuclear power plant. HANA claddings in the HIPER showed a more improved corrosion resistance than reference alloy claddings and are evaluated well with meeting the oxide thickness criteria

  7. Development of advanced LWR fuel cladding

    International Nuclear Information System (INIS)

    This report describes the results from evaluating the preliminary Zr-based alloys to develop the advanced Zr-based alloys for the nuclear fuel claddings, which should have good corrosion resistance and mechanical properties at high burn-up over 70,000MWD/MTU. It also includes the results from the basic studies for optimizing the processes which are involved in the development of the advanced Zr-based alloys. Ten(10) kinds of candidates for the alloys of which performance is over that of the existing Zircaloy-4 or ZIRLO alloy were selected out of the preliminary alloys of 150 kinds which were newly designed and repeatedly manufactured and evaluated to find out the promising alloys. First of all, the corrosion tests on the preliminary alloys were carried out to evaluate their performance in both pure water and LiOH solution at 360 deg C and in steam at 400 deg C. The tensile tests were performed on the alloys which proved to be good in the corrosion resistance. The creep behaviors were tested at 400 deg C for 10 days with the application of constant load on the samples which showed good performance in the corrosion resistance and tensile properties. The effect of the final heat treatment and A-parameters as well as Sn or Nb on the corrosion resistance, tensile properties, hardness, microstructures of the alloys was evaluated for some alloys interested. The other basic researches on the oxides, electrochemical properties, corrosion mechanism, and the establishment of the phase diagrams of some alloys were also carried out

  8. Development of advanced LWR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yong Hwan; Park, S. Y.; Lee, M. H. [and others

    2000-04-01

    This report describes the results from evaluating the preliminary Zr-based alloys to develop the advanced Zr-based alloys for the nuclear fuel claddings, which should have good corrosion resistance and mechanical properties at high burn-up over 70,000MWD/MTU. It also includes the results from the basic studies for optimizing the processes which are involved in the development of the advanced Zr-based alloys. Ten(10) kinds of candidates for the alloys of which performance is over that of the existing Zircaloy-4 or ZIRLO alloy were selected out of the preliminary alloys of 150 kinds which were newly designed and repeatedly manufactured and evaluated to find out the promising alloys. First of all, the corrosion tests on the preliminary alloys were carried out to evaluate their performance in both pure water and LiOH solution at 360 deg C and in steam at 400 deg C. The tensile tests were performed on the alloys which proved to be good in the corrosion resistance. The creep behaviors were tested at 400 deg C for 10 days with the application of constant load on the samples which showed good performance in the corrosion resistance and tensile properties. The effect of the final heat treatment and A-parameters as well as Sn or Nb on the corrosion resistance, tensile properties, hardness, microstructures of the alloys was evaluated for some alloys interested. The other basic researches on the oxides, electrochemical properties, corrosion mechanism, and the establishment of the phase diagrams of some alloys were also carried out.

  9. Radial thermal expansion of simulated binary alloy metallic fuel pin

    International Nuclear Information System (INIS)

    Metallic alloys based fuels are preferred fuels for future fast breeder reactors in India due to their high breeding ratio, short doubling time, higher fissile atom density, higher thermal conductivity, better neutronic properties as compared to the ceramic fuels like oxide, carbide or nitrides. Sodium bonded U- Pu- Zr ternary fuel with T91 cladding and metallic bonded U-15%Pu binary fuel with T91 cladding having a thin 'Zr' liner inside are the two proposed designs in this regard. In the latter case, 'Zr' layer acts as a barrier layer to prevent fuel clad chemical interaction. Since this fuel composition and the design of the fuel pin differ from the conventional ones, in depth studies are needed both from the reactor engineering and performance point of view. Thermal properties of fuels and cladding material and their mutual interaction (both chemical and mechanical) play a very important role in optimizing many design criteria. The present paper summarizes some of the data generated for thermal expansion of mechanical bonded fuel pin design in the radial direction with 'U' metal having two symmetrical cut design loaded inside the T91 cladding tube with the 'Zr' liner material in-between the fuel and the cladding. The mechanical bonding was carried out by swaging to ensure no gap between the fuel and clad on as fabricated fuel pin

  10. The Influence of the In-Situ Clad Staining on the Corrosion of Zircaloy in PWR Water Environment

    Energy Technology Data Exchange (ETDEWEB)

    Kammenzind, B.F., Eklund, K.L. and Bajaj, R.

    2001-06-21

    Zircaloy cladding tubes strain in-situ during service life in the corrosive environment of a Pressurized Water Reactor for a variety of reasons. First, the tube undergoes stress free growth due to the preferential alignment of irradiation induced vacancy loops on basal planes. Positive strains develop in the textured tubes along prism orientations while negative strains develop along basal orientations (Reference (a)). Second, early in life, free standing tubes will often shrink by creep in the diametrical direction under the external pressure of the water environment, but potentially grow later in life in the diametrical direction once the expanding fuel pellet contacts the cladding inner wall (Reference (b)). Finally, the Zircaloy cladding absorbs hydrogen as a by product of the corrosion reaction (Reference (c)). Once above the solubility limit in Zircaloy, the hydride precipitates as zirconium hydride (References (c) through (j)). Both hydrogen in solid solution and precipitated as Zirconium hydride cause a volume expansion of the Zircaloy metal (Reference (k)). Few studies are reported on that have investigated the influence that in-situ clad straining has on corrosion of Zircaloy. If Zircaloy corrosion rates are governed by diffusion of anions through a thin passivating boundary layer at the oxide-to-metal interface (References (l) through (n)), in-situ straining of the cladding could accelerate the corrosion process by prematurely breaking that passivating oxide boundary layer. References (o) through (q) investigated the influence that an applied tensile stress has on the corrosion resistance of Zircaloy. Knights and Perkins, Reference (o), reported that the applied tensile stress increased corrosion rates above a critical stress level in 400 C and 475 C steam, but not at lower temperatures nor in dry oxygen environments. This latter observation suggested that hydrogen either in the oxide or at the oxide-to-metal interface is involved in the observed stress

  11. Fuel-clad heat transfer coefficient of a defected fuel rod. Transfer coefficient fuel-clad in the presence of water steam

    International Nuclear Information System (INIS)

    Using a special rod built with a stack of UO2 pellets inside a thick Zircaloy clad, the authors report the measurement of the fuel-clad heat transfer coefficient when water vapour in intentionally introduced in the fuel rod at the beginning of its life. They describe the irradiation device, the measurement method (acquired data and mathematical determination of various values: temperature of the inner surface of the cladding, integrated thermal conductivity, fuel surface temperature, fuel-cladding heat transfer coefficient, thermal expansion of the cladding inner radius, UO2 thermal expansion). They finally report the experiment

  12. Nd:YAG laser cladding of marine propeller with hastelloy C-22

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J.D.; Kang, K.H.; Kim, J.N. [Dept. of Mechanical Engineering, Inha University, Yonghyundong 253, Namku, 402-751, Incheon (Korea)

    2004-09-01

    Nd:YAG laser cladding with automatic wire feeding (Hastelloy C-22) has been done to increase the lifetime of marine propellers made of HBsC1. The effects of processing parameters on the quality of clad layer have been investigated and clad layers have analyzed by optical microscopy and Vickers hardness tester. The method to overcome the drop transfer problem during the wire feeding has been introduced. A cladding speed that is too fast or too slow influenced the shape of clad. The good clad layer without cracks and with low dilution has been obtained with the optimum processing parameters. (orig.)

  13. Evaluation of alternative descriptions of PWR cladding corrosion behavior

    International Nuclear Information System (INIS)

    A statistical procedure has been used to evaluate several alternative descriptions of pressurized water reactor (PWR) cladding corrosion behavior, using an extensive database of Improved (low tin) Zr-4 cladding corrosion measurements from fuel irradiated in commercial PWRs. The in-reactor corrosion enhancement factors considered in the model development are based on a comprehensive review of the current literature for PWR cladding corrosion phenomenology and models. In addition, because prediction of PWR cladding corrosion behavior is very sensitive to the values used for the oxide surface temperatures, several models for the forced convection and sub-cooled nucleate boiling (SNB) coolant heat transfer under PWR conditions have also been evaluated. This evaluation determined that the choice of the forced convection heat transfer has the greatest impact on the ability to fit the data. In addition, the SNB heat transfer model used must account for a continuous transition from forced convection conditions to fully developed SNB conditions. With these choices for the heat transfer models, the evaluation determined that the significant in-reactor corrosion enhancement factors are related to the formation of a hydride rim at the cladding outer diameter, the coolant lithium concentration, and the fast neutron fluence (author) (ml)

  14. Quality assurance and quality control in fabrication of cladding tubes

    International Nuclear Information System (INIS)

    Zircaloy 2 and 4 are the most important Zirconium alloys for use as fuel cladding material in light and heavy water reactors. In fast breeder reactors the cladding tubes are of a modified 16/16 - Cr-Ni-type with improved mechanical, long - term creep rate and rupture - life versus temperature properties. Starting with hot-extruded tube shells the fabrication of Zircaloy cladding tubes is done by 3 - 4 cold reduction steps in tube reducers or rolling machines followed by heat treatments in vacuum. To obtain the specified properties a precise combination of final area reduction and final annealing is absolutely necessary. The fabrication route of stainless steel claddings and guide tubes is similar to the Zircaloy production, exceptionally the last cold-forming steps are made on cold-drawing henches, hecause of economic reasons. After each cold reduction the material is annealed at recrystalisation temperatures under protective atmosoheres. For obtaining the same final tube properties for a longer nroduction neriod the implementation of a quality assurance and control system naturally independent of the production is necessary. The application of this system regarding some of the important properties of fuel cladding tubes is reported. (RW)

  15. Cladding Alloys for Fluoride Salt Compatibility Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Muralidharan, Govindarajan [ORNL; Wilson, Dane F [ORNL; Santella, Michael L [ORNL; Holcomb, David Eugene [ORNL

    2011-05-01

    This interim report provides an overview of several candidate technologies for cladding nickel-based corrosion protection layers onto high-temperature structural alloys. The report also provides a brief overview of the welding and weld performance issues associated with joining nickel-clad nickel-based alloys. From the available techniques, two cladding technologies were selected for initial evaluation. The first technique is a line-of-sight method that would be useful for coating large structures such as vessel interiors or large piping. The line-of-sight method is a laser-based surface cladding technique in which a high-purity nickel powder mixed into a polymer binder is first sprayed onto the surface, baked, and then rapidly melted using a high power laser. The second technique is a vapor phase technique based on the nickel-carbonyl process that is suitable for coating inaccessible surfaces such as the interior surfaces of heat exchangers. The final project report will feature an experimental evaluation of the performance of the two selected cladding techniques.

  16. Experience with TRIGA aluminum-clad fuel elements

    International Nuclear Information System (INIS)

    During 8 years of operation the cumulative heat energy produced in the steady-state TRIGA Mark II 250 kW reactor at Ljubljana reached 4683 MWh. The initial core had Al-clad fuel elements only. The reactivity loss due to the burnup has been compensated by fresh fuel elements with SS-cladding and, lately, by FLIP fuel elements, moving the most irradiated Al-clad fuel elements from B and C rings to the F ring and, lately, to the storage rack. The inspection of the fuel elements during the summer of 1973 revealed excessive elongations of some Al-clad fuel elements, up to 36.8 mm. By the neutronography, performed by indirect methods (In, Dy), and also by direct methods (track detector CA 80-15 B) and by special radiographic procedures on the element, the activity of which decayed sufficiently, it has been demonstrated that the growth is due to the elongation of aluminum cladding only. No growth and/or swelling of the ZrH--U fuel or the graphite plugs has been observed within the accuracy of detection. (U.S.)

  17. Study of the solidification of M2 high speed steel Laser Cladding coatings

    Directory of Open Access Journals (Sweden)

    Candel, J. J.

    2013-10-01

    Full Text Available High speed steel laser cladding coatings are complex because cracks appear and the hardness is lower than expected. In this paper AISI M2 tool steel coatings on medium carbon AISI 1045 steel substrate have been manufactured and after Laser Cladding (LC processing it has been applied a tempering heat treatment to reduce the amount of retained austenite and to precipitate secondary carbides. The study of metallurgical transformations by Scanning Electron Microscopy (SEM and Electron Back Scattered Diffraction (EBSD shows that the microstructure is extremely fine and complex, with eutectic transformations and MC, M2C and M6C precipitation. Therefore, after the laser coating is necessary to use post-weld heat treatments.Los recubrimientos de acero rápido por Laser Cladding (LC son complejos porque aparecen fisuras y la dureza es menor a la esperada. En este trabajo se han fabricado recubrimientos de acero AISI M2 sobre acero al carbono AISI 1045 y tras el procesado por láser, se han revenido para reducir la cantidad de austenita retenida y precipitar carburos secundarios. El estudio de las transformaciones metalúrgicas con Microscopía Electrónica de Barrido (MEB y Difracción de Electrones Retrodispersados (EBSD muestra que la microestructura es extremadamente fina y compleja, presenta transformaciones eutécticas y precipitación de carburos MC, M2C y M6C. Por tanto, tras el recubrimiento por láser es necesario recurrir a tratamientos térmicos post-soldeo.

  18. Influences of Cr content and PWHT on microstructure and oxidation behavior of stainless steel weld overlay cladding materials in high temperature water

    Science.gov (United States)

    Cao, X. Y.; Ding, X. F.; Lu, Y. H.; Zhu, P.; Shoji, T.

    2015-12-01

    Influences of Cr content and post weld heat treatment (PWHT) on microstructure and oxidation behavior of stainless steel cladding materials in high temperature water were investigated. The amounts of metal oxidized and dissolved were estimated to compare the oxidation behaviors of cladding materials with different Cr contents and PWHT. The results indicated that higher Cr content led to formation of more ferrite content, and carbides were found along δ/γ phase interface after PWHT. Higher Cr content enhanced the pitting resistance and compactness of the oxide film to reduce metal amount oxidized and dissolved, which mitigated the weight changes and the formation of Fe-rich oxides. PWHT promoted more and deeper pitting holes along the δ/γ phase interface due to formation of carbides, which resulted in an increase in metal amount oxidized and dissolved, and were also responsible for more Fe-rich oxides and higher weight changes.

  19. Cladding and wrapper development for fast breeder reactor high performance

    International Nuclear Information System (INIS)

    In order to ensure economic performance, of both the existing reactors and the future EFR, much recent research has been carried out within the framework of the European R and D agreement to examine the properties of various wrapper and cladding alloys. This paper reviews the status of the European research and development programmes on these steels and highlights the most striking results. For the cladding alloys, results on dimensional stability and tensile properties for fuel pin cladding irradiated in PFR or Phenix will be given. As for wrappers the presently available results of those wrappers irradiated in Phenix and PFR show that both ferritic steels are very good candidates and that on the basis of our present knowledge most of the properties are satisfactory for wrapper applications

  20. Mechanical Property Evaluation of High Burnup PHWR Fuel Clads

    International Nuclear Information System (INIS)

    Assurance of clad integrity is of vital importance for the safe and reliable extension of fuel burnup. In order to study the effect of extended burnup of 15,000 MW∙d/tU on the performance of Pressurised Heavy Water Reactor (PHWR) fuel bundles of 19-element design, a couple of bundles were irradiated in Indian PHWR. The tensile property of irradiated cladding from one such bundle was evaluated using the ring tension test method. Using a similar method, claddings of mixed oxide (MOX) fuel elements irradiated in the pressurized water loop (PWL) of CIRUS to a burnup of 10,000 MW∙d/THM were tested. The tests were carried out both at ambient temperature and at 300°C. The paper will describe the test procedure, results generated and discuss the findings. (author)

  1. A new cladding embrittlement criterion derived from ring compression tests

    Energy Technology Data Exchange (ETDEWEB)

    Herb, Joachim, E-mail: Joachim.Herb@grs.de; Sievers, Jürgen, E-mail: Juergen.Sievers@grs.de; Sonnenburg, Heinz-Günther, E-mail: Heinz-Guenther.Sonnenburg@grs.de

    2014-07-01

    Highlights: • Using FEM it was possible to simulate measured ring compression test data. • The FEM provides burst stresses from Zry-4, M5 and ZIRLO cladding. • The ratio of burst stresses to yield stresses was correlated. • The ratio depends linearly on the state of oxidation and hydriding. • The ratio of stresses at unity can be applied as embrittlement criterion. - Abstract: It is of regulatory interest to prevent the breaking of fuel rods in LOCA transients. In current regulations this is accomplished by limiting the oxidation during LOCA to such an extent that still some residual ductility is preserved in the fuel rod cladding. The current oxidation limit in German as well as in US regulations is set to 17% ECR (Equivalent Cladding Reacted) which aims at maintaining a residual ductility for oxidized claddings. Recent ANL tests have shown that the combination of both oxidation and additionally hydrogen up-take affects the transition to zero-ductility. Furthermore, the oxidation during LOCA transient is accompanied by a significant up-take of hydrogen (secondary hydriding) if the fuel rod bursts during this transient. This secondary hydriding affects the cladding in the vicinity of the burst opening. These findings necessitate a new criterion for preserving cladding's strength. This paper describes a method how to derive a criterion which assures the required residual mechanical strength of the cladding for LOCA transients. This method utilizes the experimental results of 102 ring compression tests (RCT) conducted at ANL and KIT. RCTs of various cladding materials, oxidation levels and hydrogen content were considered. The basic approach was to compare the RCT test data with finite element analyses using the code ADINA. Starting with the cladding oxidation model of Leistikov, both the layer structure of the cladding and the distribution of the oxygen among these layers were determined. The mechanical properties of these layers were taken from

  2. Flux Density through Guides with Microstructured Twisted Clad DB Medium

    Directory of Open Access Journals (Sweden)

    M. A. Baqir

    2014-01-01

    Full Text Available The paper deals with the study of flux density through a newly proposed twisted clad guide containing DB medium. The inner core and the outer clad sections are usual dielectrics, and the introduced twisted windings at the core-clad interface are treated under DB boundary conditions. The pitch angle of twist is supposed to greatly contribute towards the control over the dispersion characteristics of the guide. The eigenvalue equation for the guiding structure is deduced, and the analytical investigations are made to explore the propagation patterns of flux densities corresponding to the sustained low-order hybrid modes under the situation of varying pitch angles. The emphasis has been put on the effects due to the DB twisted pitch on the propagation of energy flux density through the guide.

  3. Modeling alternative clad behavior for accident tolerant systems

    International Nuclear Information System (INIS)

    The US Department of Energy Fuel Cycle Research and Development program has a key goal of helping develop accident tolerant fuels (ATF) through investigating fuel and clad forms. In the current work thermochemical modeling and experiment are being used to assess fuel and clad alternatives. Cladding alternatives that have promise to improve fuel performance under accident conditions include the FeCrAl family of alloys and SiC-based composites. These are high strength and radiation resistant alloys and ceramics that have increased resistance to oxidation as compared to zirconium alloys. Accident modeling codes have indicated substantially increased time to failure and resulting effects. In the current work the thermochemical behavior of these materials are being assessed and the work reported here. (author)

  4. Reidual stresses in weld-clad reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Cladding of low alloy nuclear reactor pressure vessel steel with austenitic stainless steel introduces in heavy section components high residual stresses which may cause microrack formation in stress relief heat treatment. In this investigation an attempt is made to contribute to the solution of the stress relief cracking problem by determining quantitatively the magnitude and distribution of the residual stresses after cladding and after subsequent stress relief heat treatment. Lower residual stresses are caused by reduced thickness of the components. As the heat input is decreased at identical base material thickness, the residual stresses are lowered also. The height of the tensile residual stress peak, however, remains approximately constant. In stress relief annealed condition the residual stresses in the cladding are in tension; in the base material the residual stresses are negligibly small. (Auth.)

  5. Characterization Of Cladding Hull Wastes From Used Nuclear Fuels

    Directory of Open Access Journals (Sweden)

    Kang K.H.

    2015-06-01

    Full Text Available Used cladding hulls from pressurized water reactor (PWR are characterized to provide useful information for the treatment and disposal of cladding hull wastes. The radioactivity and the mass of gamma emitting nuclides increases with an increase in the fuel burn-up and their removal ratios are found to be more than 99 wt.% except Co-60 and Cs-137. In the result of measuring the concentrations of U and Pu included in the cladding hull wastes, most of the residues are remained on the surface and the removal ratio of U and Pu are revealed to be over 99.98 wt.% for the fuel burn-up of 35,000 MWd/tU. An electron probe micro-analyzer (EPMA line scanning shows that radioactive fission products are penetrated into the Zr oxide layer, which is proportional to the fuel burn-up. The oxidative decladding process exhibits more efficient removal ratio of radionuclides.

  6. Interdiffusion between U-Zr-Mo and stainless steel cladding

    International Nuclear Information System (INIS)

    Interdiffusion investigations were carried out at 700 deg C for 200 hours for the diffusion couples assembled with the U-Zr-Mo ternary fuel versus austenitic stainless steel D9 and the U-Zr-Mo ternary fuel versus martensitic stainless steel HT9 respectively to investigate the fuel-cladding compatibility. SEM-EDS analysis was utilized to determine the composition and the penetration depths of the reaction layers. In the case of Fuel/D9 couple, (Fe, Cr, Ni) of the cladding elements formed the precipitates with the Zr, Mo and diminished the U concentration upto 800μ length from the fuel side. Composition of the precipitates was varied with the penetrated elements. In Fuel/HT9 couple, reaction layer was smaller than that of D9 couples and was less affected by cladding elements. The eutectic reaction appeared partially in the Fuel/HT9 diffusion couple

  7. Results of NDE Technique Evaluation of Clad Hydrides

    Energy Technology Data Exchange (ETDEWEB)

    Dennis C. Kunerth

    2014-09-01

    This report fulfills the M4 milestone, M4FT-14IN0805023, Results of NDE Technique Evaluation of Clad Hydrides, under Work Package Number FT-14IN080502. During service, zirconium alloy fuel cladding will degrade via corrosion/oxidation. Hydrogen, a byproduct of the oxidation process, will be absorbed into the cladding and eventually form hydrides due to low hydrogen solubility limits. The hydride phase is detrimental to the mechanical properties of the cladding and therefore it is important to be able to detect and characterize the presence of this constituent within the cladding. Presently, hydrides are evaluated using destructive examination. If nondestructive evaluation techniques can be used to detect and characterize the hydrides, the potential exists to significantly increase test sample coverage while reducing evaluation time and cost. To demonstrate the viability this approach, an initial evaluation of eddy current and ultrasonic techniques were performed to demonstrate the basic ability to these techniques to detect hydrides or their effects on the microstructure. Conventional continuous wave eddy current techniques were applied to zirconium based cladding test samples thermally processed with hydrogen gas to promote the absorption of hydrogen and subsequent formation of hydrides. The results of the evaluation demonstrate that eddy current inspection approaches have the potential to detect both the physical damage induced by hydrides, e.g. blisters and cracking, as well as the combined effects of absorbed hydrogen and hydride precipitates on the electrical properties of the zirconium alloy. Similarly, measurements of ultrasonic wave velocities indicate changes in the elastic properties resulting from the combined effects of absorbed hydrogen and hydride precipitates as well as changes in geometry in regions of severe degradation. However, for both approaches, the signal responses intended to make the desired measurement incorporate a number of contributing

  8. Metallurgical and mechanical behaviours of PWR fuel cladding tube oxidised at high temperature; Comportements metallurqigue et mecanique des materiaux de gainage du combustible REP oxydes a haute temperature

    Energy Technology Data Exchange (ETDEWEB)

    Stern, A

    2007-12-15

    Zirconium alloys are used as cladding materials in Pressurized Water Reactors (PWR). As they are submitted to very extreme conditions, it is necessary to check their behaviour and especially to make sure they meet the safety criteria. They are therefore studied under typical in service-loadings but also under accidental loadings. In one of these accidental scenarios, called Loss of Coolant Accident (LOCA) the cladding temperature may increase above 800 C, in a steam environment, and decrease before a final quench of the cladding. During this temperature transient, the cladding is heavily oxidised, and the metallurgical changes lead to a decrease of the post quench mechanical properties. It is then necessary to correlate this drop in residual ductility to the metallurgical evolutions. This is the problem we want to address in this study: the oxidation of PWR cladding materials at high temperature in a steam environment and its consequences on post quench mechanical properties. As oxygen goes massively into the metallic part - a zirconia layer grows at the same time - during the high temperature oxidation, the claddings tubes microstructure shows three different phases that are the outer oxide layer (zirconia) and the inner metallic phases ({alpha}(O) and 'ex {beta}') - with various mechanical properties. In order to reproduce the behaviour of this multilayered material, the first part of this study consisted in creating samples with different - but homogeneous in thickness - oxygen contents, similar to those observed in the different phases of the real cladding. The study was especially focused on the {beta}-->{alpha} phase transformation upon cooling and on the resulting microstructures. A mechanism was proposed to describe this phase transformation. For instance, we conclude that for our oxygen enriched samples, the phase transformation kinetics upon cooling are ruled by the oxygen partitioning between the two allotropic phases. Then, these materials

  9. Modelling of pellet-cladding interaction in PWR's

    International Nuclear Information System (INIS)

    The pellet-cladding interaction that can occur in a PWR fuel rod design is modelled with the computer codes FRAPCON-1 and ANSYS. The fuel performance code FRAPCON-1 analyses the fuel rod irradiation behavior and generates the initial conditions for the localized fuel rod thermal and mechanical modelling in two and three-dimensional finite elements with ANSYS. In the mechanical modelling, a pellet fragment is placed in the fuel rod gap. Two types of fuel rod cladding materials are considered: Zircaloy and austenitic stainless steel. (author)

  10. Gap conductance in Zircaloy-clad LWR fuel rods

    International Nuclear Information System (INIS)

    This report describes the procedures currently used to calculate fuel-cladding gap conductance in light water reactor fuel rods containing pelleted UO2 in Zircaloy cladding, under both steady-state and transient conditions. The relevant theory is discussed together with some of the approximations usually made in performance modelling codes. The state of the physical property data which are needed for heat transfer calculations is examined and some of the relevant in- and out-of-reactor experimental work on fuel rod conductance is reviewed

  11. Synthesis of clad motion experiments interpretation: codes and validation

    International Nuclear Information System (INIS)

    This communication deals with clad melting and relocation phenomena related to LMFBR safety analysis of loss of flow accidents. We present: - the physical models developed at DSN/CEN Cadarache in single channel and bundle geometry. The interpretation with these models of experiments performed by the STT (CEN Grenoble). It comes out that we have now obtained a good understanding of the involved phenomena in single channel geometry. On the other hand, further studies are necessary for a better knowledge of clad motion phenomena in bundle cases with conditions close to reactor ones

  12. Femtosecond pulse amplification in cladding-pumped fibers

    OpenAIRE

    Minelly, J. D.; Galvanauskas, A.; Fermann, M. E.; Harter, D.; Caplen, J.E.; Chen, Z.J.; Payne, D. N.

    1995-01-01

    Femtosecond pulse amplification in a cladding-pumped fiber amplifier is demonstrated for the first time to our knowledge. Using a cladding-pumped erbium-doped fiber power amplifier and a passively mode-locked fiber seed oscillator in conjunction with an all-fiber chirped-pulse amplification system, we obtain 380-fs near-bandwidth-limited pulses with an average power of 260 mW. The pulse repetition rate is varied between 5 and 50 MHz, and pulse energies as high as 20 nJ are generated.

  13. A COMPREHENSIVE MODEL OF LASER CLADDING BY POWDER FEEDING

    Institute of Scientific and Technical Information of China (English)

    Y.L. Huang; G.Y. Liang; J.Y. Su

    2004-01-01

    A novel model was presented to predict the evolutionary development of cladding layer,and a method based on Lambert-Beer theorem and Mie's theory was adopted to treat the interaction between powder stream and laser beam. By using the continuum model and enthalpy-porosity method, the fluid flow and heat transfer in solid-liquid phase change system were simulated. The commercial software PHOENICS, to which several modules were appended, was used to accomplish the simulation. Numerical computation was performed for Stellite 6 cladding on steel, the obtained results are coincident with those measured in experiment basically.

  14. Laser cladding with wide-band scanning rotative polygon mirror

    International Nuclear Information System (INIS)

    This paper discusses the scanning rotative polygon mirror providing a uniform linear heat source with both amplitude and frequency continuous adjustment that has been developed to produce singlepass widths about 14mm and 13mm, fourpass widths about 43mm and 35mm respectively for NiCrSiB and FeCrSiB alloy cladded on A3 substrate. Bead side angles were 175 degrees and 167 degrees respectively above alloys. A very large smooth area with average roughness Ra = 0.64μm was made by NiCrSiB alloy laser cladded

  15. Low-Stress Silicon Cladding for Surface Finishing Large UVOIR Mirrors Project

    Data.gov (United States)

    National Aeronautics and Space Administration — In this Phase I research, ZeCoat Corporation demonstrated a low-stress silicon cladding process for surface finishing large UVOIR mirrors. A polishable cladding is...

  16. Chemical methods for the use of niobium from pressure vessel cladding as a fast neutron dosimeter

    International Nuclear Information System (INIS)

    the steel samples from the cladding of a pressure vessel of an operating nuclear power reactor were obtained by scraping. The cladding material of the pressure vessel contained about 0.5 % niobium. It was desired to use the niobium as a dosimeter for estimating fast fluences at the pressure vessel. The weak radiation from the reaction product 93mNb cannot be measured in the presence of other elements and interfering activities. A method was developed to separate niobium from other metals present; the concentration and yield of niobium were determined spectrophotometrically. The irradiated niobium was electrodeposited from aqueous solutions on copper discs. The amount of the deposited niobium was determined by a radiochemical method which makes use of its own radioactivity - measured with a liquid scintillation counter - and the known starting mass of niobium. It was possible to determine the deposited niobium masses (5 to 200 microgram) with a desired degree of accuracy. The absolute emission rate of X-rays could then be measured without any self-absorption or interference from other activities. The mass of niobium on each preparate and its X-ray emission rate, later on, were used as basic experimental data for the estimation of last neutron doses at the pressure vessel

  17. Improvement of Ni-Cr-Mo coating performance by laser cladding combined re-melting

    Science.gov (United States)

    Wang, Qin-Ying; Bai, Shu-Lin; Zhang, Yang-Fei; Liu, Zong-De

    2014-07-01

    Although being an efficient technique to produce metallic alloy coating, laser cladding may leave original unmelted particles in the coating. Further treatment is thus necessary to improve the coating quality, and laser re-melting therefore becomes a potential method. In this study, Ni-Cr-Mo alloy coatings were prepared on Q235 steel substrate by laser cladding (coating N1) and then re-melted by laser (coating N2) with the same technic parameters. The initial defect evolution and its effect on hardness and corrosion resistance of coatings were studied. The results show that there are fewer and smaller defects in coating N2 than in coating N1, which is ascribbed to the disappearance and partial melting of Cr/Cr2O3 particles. The nearly unchanged hardness of coatings N1 and N2 is justified by both Vickers tests and nanoindentation combined theoretical calculation. Coating N2 with higher positive corrosion potential and lower corrosion current density exhibits better corrosion resistance than coating N1. Above results prove that laser re-melting can refine the microstructure and improve corrosion resistance of coatings to some degree.

  18. The corrosion of Zircaloy-4 fuel cladding in pressurized water reactors

    International Nuclear Information System (INIS)

    This paper reports on the effects of thermo-mechanical processing of cladding on the corrosion of Zircaloy-4 in commercial PWRs that have been investigated. Visual observations and nondestructive measurements at poolside, augmented by observations in the hot cell, indicate that the initial black oxide transforms into a grey or tan later white oxide layer at a thickness of 10 to 15 μm independent of the thermal processing history of the tubing. At an oxide layer thickness of 60 to 80 μm, the oxide may spall depending somewhat on the particular oxide morphology formed and possibly on the frequency of power and temperature changes of the fuel rods. Because spalling of oxide lowers the metal-to-oxide interface temperature of fuel rods, it reduces the corrosion rate and is beneficial from that point of view. To determine the effect of thermo-mechanical processing on in-reactor corrosion of Zircaloy-4, oxide thickness measurements at poolside and in the hot cell have been analyzed with the MATPRO corrosion model. A calibrated corrosion parameter in this model provides a measure of the corrosion susceptibility of the Zircaloy-4 cladding. It was found necessary to modify the MATPRO equations with a burnup dependent term to obtain a near constant value of the corrosion parameter over a burnup range of approximately 10 to 45 MWd/kgU. Different calculational tests were performed to confirm that the modified model accurately predicts the corrosion behavior of fuel rods

  19. Analysis of the mechanical response of LMFBR fuel clads subjected to in-service property variations

    International Nuclear Information System (INIS)

    Evidence from available literature is presented to illustrate the degradation of 316 type stainless steel considered for use in Liquid Metal Fast Breeder Reactors where sodium is used as the coolant. Nonuniform changes in properties of the steel have been found to occur due to the thermal, thermochemical, and irradiation environment to which it is exposed. Variations in impurity concentrations (such as carbon in the steel) of several orders of magnitude compared to the original values have been observed under controlled sodium exposures. At temperatures relevant to the reactor system migration of impurities by diffusion, compound formations and carburization/decarburization behavior have been observed to occur. Mechanical property measurements such as tensile and yield strengths made under these conditions indicate that thermal and thermochemical influences can result in variations in the above properties quite different from the original material. Modified formulations of the elastic and elastic-plastic analysis of the degraded fuel-clad are presented in two dimensions. The elastic and plastic parameters relating to the properties of the degraded material are represented by spatially varying functions as opposed to being treated as constants which is the conventional case. The changes in the mathematical nature of the constitutive equations are demonstrated by sample illustrations and solutions involving continuous changes in the elastic moduli through-the-thickness of the clad. Recommendations for the establishment of improved Reactor Research Development Standards are made based on the studies

  20. Report on Reactor Physics Assessment of Candidate Accident Tolerant Fuel Cladding Materials in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, Nathan [Univ. of Tennessee, Knoxville, TN (United States); Maldonado, G. Ivan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-28

    This work focuses on ATF concepts being researched at Oak Ridge National Laboratory (ORNL), expanding on previous studies of using alternate cladding materials in pressurized water reactors (PWRs). The neutronic performance of two leading alternate cladding materials were assessed in boiling water reactors (BWRs): iron-chromium-aluminum (FeCrAl) cladding, and silicon carbide (SiC)-based composite cladding. This report fulfills ORNL Milestone M3FT-15OR0202332 within the fiscal year 2015 (FY15)

  1. DEVELOPMENT OF LASER CLADDING WEAR-RESISTANT COATING ON TITANIUM ALLOYS

    OpenAIRE

    RUILIANG BAO; HUIJUN YU; CHUANZHONG CHEN; BIAO QI; LIJIAN ZHANG

    2006-01-01

    Laser cladding is an advanced surface modification technology with broad prospect in making wear-resistant coating on titanium alloys. In this paper, the influences of laser cladding processing parameters on the quality of coating are generalized as well as the selection of cladding materials on titanium alloys. The microstructure characteristics and strengthening mechanism of coating are also analyzed. In addition, the problems and precaution measures in the laser cladding are pointed out.

  2. Effects of the inner mould material on the aluminium–316L stainless steel explosive clad pipe

    International Nuclear Information System (INIS)

    Highlights: ► Different mould materials were adopted to evaluate the effect of the constraint on the clad quality. ► The interface characteristics of clad pipe were analyzed for the different clad pipe. ► The clad pipes possess excellent bonding quality. - Abstract: The clad pipe played an important part in the pipeline system of the nuclear power industry. To prepare the clad pipe with even macrosize and excellent bonding quality, in this work, different mould materials were adopted to evaluate the effect of the constraint on the clad quality of the bimetal pipe prepared by explosive cladding. The experiment results indicated that, the dimension uniformity and bonding interface of clad pipe were poor by using low melting point alloy as mould material; the local bulge or the cracking of the clad pipe existed when the SiC powder was utilized. When the steel mould was adopted, the outer diameter of the clad pipe was uniform from head to tail. In addition, the metallurgical bonding was formed. Furthermore, the results of shear test, bending test and flattening test showed that the bonding quality was excellent. Therefore, the Al–316L SS clad pipe could endure the second plastic forming

  3. Gradient microstructure in laser clad TiC-reinforced Ni-alloy composite coating

    NARCIS (Netherlands)

    Pei, Y.T.; Zuo, T.C.

    1998-01-01

    A gradient TiC–(Ni alloy) composite coating was produced by one step laser cladding with pre-placed mixture powder on a 1045 steel substrate. The clad layers consisted of TiC particles, γ-Ni primary dendrites and interdendritic eutectics. From the bottom to the top of the clad layer produced at 2000

  4. Management of waste cladding hulls. Part II. An assessment of zirconium pyrophoricity and recommendations for handling waste hulls

    Energy Technology Data Exchange (ETDEWEB)

    Kullen, B J; Levitz, N M; Steindler, M J

    1977-11-01

    This report reviews experience and research related to the pyrophoricity of zirconium and zirconium alloys. The results of recent investigations of the behavior of Zircaloy and some observations of industrial handling and treatment of Zircaloy tubing and scrap are also discussed. A model for the management of waste Zircaloy cladding hulls from light water reactor fuel reprocessing is offered, based on an evaluation of the reviewed information. It is concluded that waste Zircaloy cladding hulls do not constitute a pyrophoric hazard if, following the model flow sheet, finely divided metal is oxidized during the management procedure. Steps alternative to the model are described which yield zirconium in deactivated form and also accomplish varying degrees of transuranic decontamination. Information collected into appendixes is (1) a collation of zirconium pyrophoricity data from the literature, (2) calculated radioactivity contents in Zircaloy cladding hulls from spent LWR fuels, and (3) results of a laboratory study on volatilization of zirconium from Zircaloy using HCl or Cl/sub 2/.

  5. Optimization of Hydride Rim Formation in Unirradiated Zr 4 Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Shimskey, Rick W.; Hanson, Brady D.; MacFarlan, Paul J.

    2013-09-30

    The purpose of this work is to build on the results reported in the M2 milestone M2FT 13PN0805051, document number FCRD-USED-2013-000151 (Hanson, 2013). In that work, it was demonstrated that unirradiated samples of zircaloy-4 cladding could be pre-hydrided at temperatures below 400°C in pure hydrogen gas and that the growth of hydrides on the surface could be controlled by changing the surface condition of the samples and form a desired hydride rim on the outside diameter of the cladding. The work performed at Pacific Northwest National Laboratory since the issuing of the M2 milestone has focused its efforts to optimize the formation of a hydride rim on available zircaloy-4 cladding samples by controlling temperature variation and gas flow control during pre-hydriding treatments. Surface conditioning of the outside surface was also examined as a variable. The results of test indicate that much of the variability in the hydride thickness is due to temperature variation occurring in the furnaces as well as how hydrogen gas flows across the sample surface. Efforts to examine other alloys, gas concentrations, and different surface conditioning plan to be pursed in the next FY as more cladding samples become available

  6. UK experience on fuel and cladding interaction in oxide fuels

    International Nuclear Information System (INIS)

    The occurrence of fuel cladding interactions in fast reactor fuels has been observed in UK irradiations over a period of years. Chemical incompatibility between fuel and clad represents a potential source of failure and has, on this account, been studied using a variety of techniques. The principal fuel of interest to the UK for fast reactor application is mixed uranium plutonium oxide clad in stainless steel and it is in this field that the majority of work has been concentrated. Some consideration has been given to carbide fuels, because of their application as an advanced fuel. This experience is described in the accompanying paper. Several complementary initiatives have been followed to investigate the interactions in oxide fuel. The principal source of experimental information is from the experimental fuel irradiation programme in the Dounreay Fast Reactor (DFR). Supporting information has been obtained from irradiation programmes in Materials Testing Reactors (MTR). Conditions approaching those in a fast reactor are obtained and the effects of specific variables have been examined in specifically designed experiments. Out-of-reactor experiments have been used to determine the limits of fuel and cladding compatibility and also to give indications of corrosion The observations from all experiments have been examined in the light of thermo-dynamic predictions of fuel behaviour to assess the relative significance of various observations and operating conditions. An experimental programme to control and limit the interactions in oxide fuel is being followed

  7. Plastic deformation of the cladding of Fortissimo fuel elements

    International Nuclear Information System (INIS)

    A study of a large number of standard Fortissimo pins, clad in solution treated 316 steel, shows that the plastic strain depends linearly on the fission gas pressure and the dose (in dpaF). The derived modulus of irradiation creep ranges from 1 to 2 x 10-6 (MPa dpaF)-1 at 4500C and increases steadily with temperature. (author)

  8. Clad failure detection in G 3 - operational feedback

    International Nuclear Information System (INIS)

    After briefly reviewing the role and the principles of clad failure detection, the author describes the working conditions and the conclusions reached after 4 years operation of this installation on the reactor G 3. He mentions also the modifications made to the original installation as well as the tests carried out and the experiments under way. (author)

  9. Fundamental metallurgical aspects of axial splitting in zircaloy cladding

    International Nuclear Information System (INIS)

    Fundamental metallurgical aspects of axial splitting in irradiated Zircaloy cladding have been investigated by microstructural characterization and analytical modeling, with emphasis on application of the results to understand high-burnup fuel failure under RIA situations. Optical microscopy, SEM, and TEM were conducted on BWR and PWR fuel cladding tubes that were irradiated to fluence levels of 3.3 x 1021 n cm-2 to 5.9 x 1021 n cm-2 (E > 1 MeV) and tested in hot cell at 292--325 C in Ar. The morphology, distribution, and habit planes of macroscopic and microscopic hydrides in as-irradiated and posttest cladding were determined by stereo-TEM. The type and magnitude of the residual stress produced in association with oxide-layer growth and dense hydride precipitation, and several synergistic factors that strongly influence axial-splitting behavior were analyzed. The results of the microstructural characterization and stress analyses were then correlated with axial-splitting behavior of high-burnup PWR cladding reported for simulated-RIA conditions. The effects of key test procedures and their implications for the interpretation of RIA test results are discussed

  10. Analysis of PWR cladding transient load under LOCA quench conditions

    International Nuclear Information System (INIS)

    LOCA is a classical design basis accident needed to be considered in all LWR safety analyses. The thermal shock induced by quench during LOCA may cause fracture in the claddings which could lead to core damage. Therefore it is necessary to study the cladding behavior during quench. This paper reports the results of LOCA quench experiments and simulations using the RANNS code for evaluating local mechanical and thermal states of axial load on the cladding. The experimental measurements suggest the rate of load gain decreases with an increasing of the ECR value due to the thicker zirconia layer which serves as a thermal barrier. In addition, the temperature-induced stress on the cladding along the axial direction appears uneven. Therefore, it is found that the LOCA simulation needs multiple elements in the axial direction for obtaining a fairly good prediction of the axial load gain. Finally, the RANNS simulation of the pellet center temperature is validated, and the RANNS code shows the capability in predicting the axial load generated during quench for ECR of lower or equal to 15%. (author)

  11. Burst investigation on zircaloy-4 claddings in inert environment

    International Nuclear Information System (INIS)

    Highlights: • Burst investigations on zircaloy-4 cladding in argon environment. • Clad wall displacement measurement. • Effect of internal overpressure and heating rate on burst parameters. • Semi-empirical correlation for burst stress has been proposed. - Abstract: An extensive burst investigation has been carried out on the zircaloy-4 claddings in an inert environment to simulate clad burst during a postulated loss-of-coolant-accident (LOCA) conditions. The parameters varied during the burst experiments were heating rate and internal overpressure. The temperature, internal overpressure and ballooning data were monitored online and recorded during the heating process of burst specimen. In addition, post-experiment measurements were also conducted on the burst specimen to determine various burst parameters–burst strains and burst stress. A semi-empirical correlation was developed to predict the burst stress for a given burst temperature. A reasonable agreement between the predicted and experimental data has been observed. The proposed correlation was also compared with available established correlation for steam environment

  12. Underwater cladding with laser beam and plasma arc welding

    International Nuclear Information System (INIS)

    Two welding processes, plasma arc (transferred arc) (PTA) and laser beam, were investigated to apply cladding to austenitic stainless steels and Inconel 600. These processes have long been used to apply cladding layers , but the novel feature being reported here is that these cladding layers were applied underwater, with a water pressure equivalent to 24 m (80 ft). Being able to apply the cladding underwater is very important for many applications, including the construction of off-shore oil platforms and the repair of nuclear reactors. In the latter case, being able to weld underwater eliminates the need for draining the reactor and removing the fuel. Welding underwater in reactors presents numerous challenges, but the ability to weld without having to drain the reactor and remove the fuel provides a huge cost savings. Welding underwater in reactors must be done remotely, but because of the radioactive corrosion products and neutron activation of the steels, remote welding would also be required even if the reactor is drained and the fuel removed. In fact, without the shielding of the water, the remote welding required if the reactor is drained might be even more difficult than that required with underwater welds. Furthermore, as shall be shown, the underwater welds that the authors have made were of high quality and exhibit compressive rather than tensile residual stresses

  13. Interfacial adhesion of laser clad functionally graded materials

    NARCIS (Netherlands)

    De Hosson, JTM; Pei, YT; Ocelik, [No Value; Sudarshan, TS; Stiglich, JJ; Jeandin, M

    2002-01-01

    Specially designed samples of laser clad AlSi40 functionally graded materials (FGM) are made for evaluating the interfacial adhesion. To obtain the interfacial bond strength notches are made right at the interface of the FGMs. In-sitit microstructural observations during straining in an FEG-ESEM (fi

  14. Interfacial adhesion of laser clad functionally graded materials

    NARCIS (Netherlands)

    Pei, Y. T.; Ocelik, V.; De Hosson, J. T. M.

    2003-01-01

    Specially designed samples of laser clad AlSi40 functionally graded materials (FGM) are made for evaluating the interfacial adhesion. To obtain the interfacial bond strength notches are made right at the interface of the FGMs. In-situ microstructural observations during straining in a field-emission

  15. Ultrahigh Temperature-Sensitive Silicon MZI with Titania Cladding

    OpenAIRE

    Lee, Jong-Moo

    2015-01-01

    We present a possibility of intensifying temperature sensitivity of a silicon Mach-Zehnder interferometer (MZI) by using a highly negative thermo-optic property of titania (TiO2). Temperature sensitivity of an asymmetric silicon MZI with a titania cladding is experimentally measured from +18 to −340 pm/°C depending on design parameters of MZI.

  16. Foam coating on aluminum alloy with laser cladding

    NARCIS (Netherlands)

    Ocelik, V.; van Heeswijk, V.; de Hosson, J.T.M.; Csach, K.

    2004-01-01

    dThis article concentrates on the creation of a foam layer on an Al-Si substrate with laser technology. The cladding of At-Si powder in the front of a laser track has been separated from the side injection of mixture of Al-Si/TiH2 powder (foaming agent), which allows for fine tuning of the main proc

  17. Delayed hydride cracking of Zircaloy-4 fuel cladding

    International Nuclear Information System (INIS)

    Crack propagation rates, grown by the delayed hydride cracking mechanism, were measured in Zircaloy-4 fuel cladding, according to a Coordinated Research Project (CRP) sponsored by the International Atomic Energy Agency (IAEA). During the first stage of the program a Round Robin Testing was performed on fuel cladding samples provided by Studsvik (Sweden), of the type used in PWR reactors. Crack growth in the axial direction is obtained through the specially developed 'pin load testing' (PLT) device. In these tests, crack propagation rates were determined at 250 C degrees on several samples of the material described above, obtaining a mean value of about 2.5 x 10-8 m s-1. The results were analyzed and compared satisfactorily with those obtained by the other laboratories participating in the CRP. At the present moment, similar tests on CANDU and Atucha I type fuel cladding are being performed. It is thought that the obtained results will give valuable information concerning the analysis of possible failures affecting fuel cladding under reactor operation. (author)

  18. Techniques developed to determine KIH of Zircaloy-4 cladding material

    International Nuclear Information System (INIS)

    Zircaloy-4, used as a fuel cladding material, is known to be susceptible to delayed hydride cracking (DHC). The study of the DHC mechanism and development of an approach to mitigate its occurrence, are of importance to the nuclear industry worldwide. Coordinated by the International Atomic Energy Agency (IAEA), an international research program was established in 2011 with the objective of experimentally determining the critical stress intensity factor (KIH) of DHC for various Zircaloy-4 cladding materials. Representing Canada, AECL Chalk River Laboratories (CRL) participates in this program. During 2011 to 2013, various techniques were developed at CRL with the objective of accurately determining the KIH of Pressurized Heavy Water Reactor (PHWR)-type Zircaloy-4 cladding and other zirconium-based cladding materials. These techniques include: 1) charging hydrogen into thin-wall test specimens with a gaseous approach, 2) determining hydrogen concentration in the specimens using differential scanning calorimetry, 3) fatigue pre-cracking of the specimens, and 4) establishing an empirical relationship between the stress-intensity factor (KI) and crack length of the specimens being studied. This paper describes the working principle of the techniques, and associated experimental results. (author)

  19. Phosphate glass-clad tellurium semiconductor core optical fibers

    International Nuclear Information System (INIS)

    Highlights: • Glass-clad tellurium semiconductor core optical fibers were successfully synthesized. • The core were found to be highly crystalline and phase-pure. • The core in fibers have high transparency at infrared wavelength from 4 to 25 μm. • The tellurium core fibers have great potential utility in Raman fiber amplifiers. - Abstract: For the first time to the best of our knowledge phosphate glass-clad optical fibers comprising tellurium (Te) semiconductor core have been fabricated using a molten core approach. The cores were found to be highly crystalline and phase-pure as evidenced by X-ray diffraction (XRD) and corroborated by Micro-Raman spectrum. Elemental analysis across the core/clad interface suggests that there is some diffusion of oxygen and phosphorus into the core region and, conversely, diffusion of Te into the cladding region. Unfortunately, the propagation loss of the Te core fibers was too high to measure due to the significant scattering from the grain boundaries, oxygen and phosphor precipitates. However, the larger Raman gain, infrared and terahertz transparency of tellurium over silicon and germanium should make these fibers of significant value for fiber-based mid- to long-wave infrared, terahertz waveguides and Raman-shifted infrared light sources

  20. Modelling of ultrasonic nondestructive testing of cracks in claddings

    International Nuclear Information System (INIS)

    Nondestructive testing with ultrasound is a standard procedure in the nuclear power industry. To develop and qualify the methods extensive experimental work with test blocks is usually required. This can be very time-consuming and costly and it also requires a good physical intuition of the situation. A reliable mathematical model of the testing situation can, therefore, be very valuable and cost-effective as it can reduce experimental work significantly. A good mathematical model enhances the physical intuition and is very useful for parametric studies, as a pedagogical tool, and for the qualification of procedures and personnel. The present project has been concerned with the modelling of defects in claddings. A cladding is a layer of material that is put on for corrosion protection, in the nuclear power industry this layer is often an austenitic steel that is welded onto the surface. The cladding is usually anisotropic and to some degree it is most likely also inhomogeneous, particularly in that the direction of the anisotropy is varying. This degree of inhomogeneity is unknown but probably not very pronounced so for modelling purposes it may be a valid assumption to take the cladding to be homogeneous. However, another important complicating factor with claddings is that the interface between the cladding and the base material is often corrugated. This corrugation can have large effects on the transmission of ultrasound through the interface and can thus greatly affect the detectability of defects in the cladding. In the present project the only type of defect that is considered is a planar crack that is situated inside the cladding. The investigations are, furthermore, limited to two dimensions, and the crack is then only a straight line. The crack can be arbitrarily oriented and situated, but it must not intersect the interface to the base material. The crack can be surface-breaking, and this is often the case of most practical interest, but it should then be

  1. Preliminary study of mechanical behavior for Cr coated Zr-4 Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do-Hyoung; Kim, Hak-Sung [Hanyang Univ., Seoul (Korea, Republic of); Kim, Hyo-Chan; Yang, Yong-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    To decrease the oxidation rate of Zr-based alloy components, many concepts of accident tolerant fuel (ATF) such as Mo-Zr cladding, SiC/SiCf cladding and iron-based alloy cladding are under development. One of the promised concept is the coated cladding which can remarkably increase the corrosion and wear resistance. Recently, KAERI is developing the Cr coated Zircaloy cladding as accident tolerance cladding. To coat the Cr powder on the Zircaloy, 3D laser coating technology has been employed because it is possible to make a coated layer on the tubular cladding surface by controlling the 3-diminational axis. Therefore, for this work, the mechanical integrity of Cr coated Zircaloy should be evaluated to predict the safety of fuel cladding during the operating or accident of nuclear reactor. In this work, the mechanical behavior of the Cr coated Zircaloy cladding has been studied by using finite element analysis (FEA). The ring compression test (RCT) of fuel cladding was simulated to evaluate the validity of mechanical properties of Zr-4 and Cr, which were referred from the literatures and experimental reports. In this work, the mechanical behavior of the Cr coated Zircaloy cladding has been studied by using finite element analysis (FEA). The ring compression test (RCT) of fuel cladding was simulated to evaluate the validity of mechanical properties of Zr-4 and Cr. The pellet-clad mechanical interaction (PCMI) properties of Cr coated Zr-4 cladding were investigated by thermo-mechanical finite element analysis (FEA) simulation. The mechanical properties of Zr-4 and Cr was validated by simulation of ring compression test (RCT) of fuel cladding.

  2. Preliminary study of mechanical behavior for Cr coated Zr-4 Fuel Cladding

    International Nuclear Information System (INIS)

    To decrease the oxidation rate of Zr-based alloy components, many concepts of accident tolerant fuel (ATF) such as Mo-Zr cladding, SiC/SiCf cladding and iron-based alloy cladding are under development. One of the promised concept is the coated cladding which can remarkably increase the corrosion and wear resistance. Recently, KAERI is developing the Cr coated Zircaloy cladding as accident tolerance cladding. To coat the Cr powder on the Zircaloy, 3D laser coating technology has been employed because it is possible to make a coated layer on the tubular cladding surface by controlling the 3-diminational axis. Therefore, for this work, the mechanical integrity of Cr coated Zircaloy should be evaluated to predict the safety of fuel cladding during the operating or accident of nuclear reactor. In this work, the mechanical behavior of the Cr coated Zircaloy cladding has been studied by using finite element analysis (FEA). The ring compression test (RCT) of fuel cladding was simulated to evaluate the validity of mechanical properties of Zr-4 and Cr, which were referred from the literatures and experimental reports. In this work, the mechanical behavior of the Cr coated Zircaloy cladding has been studied by using finite element analysis (FEA). The ring compression test (RCT) of fuel cladding was simulated to evaluate the validity of mechanical properties of Zr-4 and Cr. The pellet-clad mechanical interaction (PCMI) properties of Cr coated Zr-4 cladding were investigated by thermo-mechanical finite element analysis (FEA) simulation. The mechanical properties of Zr-4 and Cr was validated by simulation of ring compression test (RCT) of fuel cladding

  3. Seawater immersion tests of irradiated Zircaloy-2 cladding tube

    International Nuclear Information System (INIS)

    In the Fukushima Dai-ichi Nuclear Power Plant accident, seawater was temporarily injected into the spent fuel pools since the electrically powered water cooling and feeding functions had been lost. For fuel assemblies which experienced seawater immersion, surface corrosion due to seawater constituents and the resultant degradation of mechanical properties are of concern. In particular, cladding tubes act as the most important boundary to contain radioactive fission products inside fuel rods. Therefore, in order to assess the integrity of cladding tubes, the effects of seawater immersion on corrosion behavior and mechanical properties for as-received and irradiated Zircaloy-2 cladding tubes were investigated in the present study. As the test materials, as-received and irradiated Zircaloy-2 tube specimens were used. Zircaloy-2 cladding tubes had been irradiated to about 44 GWd/t in the advanced thermal reactor FUGEN, which is similar to the type of BWRs used at the Fukushima plant. Assumptions were made about the environment which the spent fuel pools experienced in the accident, and the immersion test temperature was accordingly set at 70-90 °C. Two kinds of seawater were used for the immersion tests. One was artificial seawater and the other was natural seawater. These were not diluted and the immersion time was up to about 1000 hours. After the immersion tests, metallurgical investigations and ring tensile tests were carried out. As a result, no obvious surface corrosion and no significant degradation in the tensile strength property were observed after both artificial and natural seawater immersion tests for both as-received and irradiated tubes. This suggests that the effects of seawater immersions on corrosion behavior and mechanical properties (especially tensile properties) for as-received and irradiated Zircaloy-2 cladding tubes are probably negligible. (author)

  4. Local strain in cladding tube due to radial pellet cracking

    International Nuclear Information System (INIS)

    A study was made to develop a method for evaluation of the local strain in a cladding tube of the Advanced Thermal Reactor due to radial cracking to a UO2 fuel pellet. Effects of the number of cracks, initial crack width and the friction coefficient of a pellet-clad interface on behaviors of the local strain in a cladding tube were evaluated with a modelized experiment. Analytical evaluation of a membrane strain was also carried out on the basis of a procedure similar to that proposed by J. H. Gittus, Nuclear Engineering and Design 18 (1972) 69-82, in order to follow the experimental results and to extend the model experiment to cladding tube. A Zircaloy-2 ring specimen with inner diameter of 95 mm, height of 25 mm and wall thickness of 5 mm was expanded at room temperature with equally divided peripheral dice of a tool steel set in a specimen. Strains on an external surface of the specimen were measured with 28 wire strain gages with gage length of 0.3 mm. An elastic-plastic analysis was carried out in order to evaluate the membrane hoop strain in the cladding tube on the basis of a simple procedure similar to that proposed by Gittus. The results of analysis showed that the maximum hoop strain occured at a location apart from the dice edge. This was caused by unloading in the crack opening portion. The strain concentration factor obtained from analysis is greater than that obtained from experiment. The difference of concentration factors between analysis and experiment is due to the bending strain. Therefore, the strain concentration factor at the inner surface is evaluated from the experimental concentration factor at the external surface and the analytical concentration factor of a membrane strain. (Auth.)

  5. Hydrogen measurement in cladding material by neutron transmission analysis

    International Nuclear Information System (INIS)

    Hydrogen absorption in light water reactor fuel cladding is one phenomenon that limits the operating life under normal conditions. In failed fuel, the potential for secondary degradation by hydriding and the subsequent radionuclide release in the primary system is a current concern. The characterization of hydrogen or hydrides in fuel cladding supports efforts to safely extend operation of fuel elements at high burnup or after pinhole leaks develop. Destructive techniques for measuring hydrogen concentration of zirconium fuel elements are well developed. A number of nondestructive techniques such as neutron radiography, neutron scattering, and ultrasound have been reported. Neutron transmission analysis is feasible to use as a nondestructive technique for determining hydrogen content in zirconium fuel rod claddings. Two sample geometries with equivalent hydrogen linear density that ranged between 0 and 12 mg/cm2 were studied. Sample A was a mockup of the maximum transmission path length of a typical boiling water reactor (BWR) fuel rod cladding (12.3-mm diameter, 0.76-mm thickness) that can be tested without interference of the fuel pellet. Sample B was a mockup of the transmission path length through the diameter of a BWR fuel rod (i.e., twice the cladding wall thickness and the fuel pellet diameter). The dependence of the mass signals for samples A and B on equivalent hydrogen linear density is shown. The mass signals from samples A and B show the same dependence (i.e., slope) on equivalent hydrogen, which indicates that the thin slab assumptions are valid for this application. The y-intercepts of the mass signals are offset by a constant that corresponds to the differences in attenuation of the fuel and zirconium sections. The standard deviation of the mass signal measurements was ±0.003. This translates to an uncertainty of ±0.122 mg/cm2 in the hydrogen linear density, which is equivalent to ±230 ppm hydrogen for BWR fuel rod dimensions

  6. Cladding hull decontamination and densification process. Part 1. The prototype cladding hull decontamination system

    International Nuclear Information System (INIS)

    A prototype system for decontaminating Zircaloy-4 cladding hulls has been assembled and tested at Pacific Northwest Laboratory. The decontamination process consists of treatment with a gaseous mixture of hydrogen fluoride (HF) and argon (Ar) followed by a dilute aqueous etch of ammonium oxalate, ammonium citrate, ammonium fluoride, and hydrogen peroxide. The continuous cleaning process described in this report successfully descaled small portions of most charges, but was unable to handle the original design capacity of 4 kg/hr because of problems in the following areas: control of HF reactor temperatures, regulation of HF and argon mixtures and flows, isolation of the HF reactor atmosphere from the aqueous washer/rinser atmosphere, regulation of undesirable side reactions, and control over hull transport through the system. Due to the limited time available to solve these problems, the system did not attain fully operational status. The work was performed with unirradiated hulls that simulated irradiated hulls. The system was not built to be remotely operable. The process chemistry and system equipment are described in this report with particular emphasis on critical operating areas. Recommendations for improved system operation are included

  7. Observation of cladding modes spatio-spectral distribution in large mode area photonic crystal fiber

    International Nuclear Information System (INIS)

    We report the observation of spatio-spectral distribution in cladding modes of a single-mode large mode area photonic crystal fiber. The cladding modes excitation was achieved without any external fiber exposure. The optical field patterns of the cladding modes within different pump wavelength are investigated. To the best of knowledge the spatio- spectral distribution in cladding modes of large mode photonic crystal fiber is demonstrated for the first time. The results are of immediate interest in applications demanding devices based on core and cladding mode coupling in photonic crystal fibers

  8. A state of the Art report on Manufacturing technology of high burn-up fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyeong Ho; Nam, Cheol; Baek, Jong Hyuk; Choi, Byung Kwon; Park, Sang Yoon; Lee, Myung Ho; Jeong, Yong Hwan

    1999-09-01

    In order to manufacturing the prototype fuel cladding, overall manufacturing processes and technologies should be thoroughly understood on the manufacturing processes and technologies of foreign cladding tubes. Generally, the important technology related to fuel cladding tube manufacturing processes for PWRs/PHWRs is divided into three stages. The first stage is to produce the zirconium sponge from zirconium sand, the second stage is to produce the zircaloy shell or TREX from zirconium sponge ingot and finally, cladding is produced from TREX or zircaloy shell. Therefore, the manufacturing processes including the first and second stages are described in brief in this technology report in order to understand the whole fuel cladding manufacturing processes. (author)

  9. A state of the Art report on Manufacturing technology of high burn-up fuel cladding

    International Nuclear Information System (INIS)

    In order to manufacturing the prototype fuel cladding, overall manufacturing processes and technologies should be thoroughly understood on the manufacturing processes and technologies of foreign cladding tubes. Generally, the important technology related to fuel cladding tube manufacturing processes for PWRs/PHWRs is divided into three stages. The first stage is to produce the zirconium sponge from zirconium sand, the second stage is to produce the zircaloy shell or TREX from zirconium sponge ingot and finally, cladding is produced from TREX or zircaloy shell. Therefore, the manufacturing processes including the first and second stages are described in brief in this technology report in order to understand the whole fuel cladding manufacturing processes. (author)

  10. Bonding and Integration of C-C Composite to Cu-Clad-Molybdenum for Thermal Management Applications

    Science.gov (United States)

    Asthana, R.; Singh, M.; Shpargel, T.P.

    2008-01-01

    Two- and three-dimensional carbon-carbon composites with either resin-derived matrix or CVI matrix were joined to Cu-clad-Mo using active Ag-Cu braze alloys for thermal management applications. The joint microstructure and composition were examined using Field-Emission Scanning Electron Microscopy and Energy-Dispersive Spectroscopy, and the joint hardness was characterized using the Knoop microhardness testing. Observations on the infiltration of the composite with molten braze, dissolution of metal substrate, and solute segregation at the C-C surface have been discussed. The thermal response of the integrated assembly is also briefly discussed.

  11. Three-dimensional interpretation of cleavage fracture tests of cladded specimens with local approach to cleavage fracture

    International Nuclear Information System (INIS)

    Electricite de France has conducted during these last years an experimental and numerical research programme in order to evaluate fracture mechanics analyses used in nuclear reactor pressure vessels integrity assessment, regarding the risk of brittle fracture. Two cladded specimens made of ferritic steel A508 Cl3 with stainless steel cladding, and containing shallow subclad flaws, have been tested in four point bending at very low temperature to obtain cleavage failure. The crack instability was obtained in base metal by cleavage fracture, without crack arrest. The tests have been interpreted by local approach to cleavage fracture (Beremin model) using three-dimensional finite element computations. After the elastic-plastic computation of stress intensity factor KJ along the crack front, the probability of cleavage failure of each specimen is evaluated using m, σu Beremin model parameters identified on the same material. The failure of two specimens is conservatively predicted by both analyses. The elastic-plastic stress intensity factor KJ in base metal is always greater than base metal fracture toughness K1c. The calculated probabilities of cleavage failure are in agreement with experimental results. The sensitivity of Beremin model to numerical aspects is finally exposed. (orig.)

  12. Study on modes of energy action in laser-induction hybrid cladding

    International Nuclear Information System (INIS)

    The shape and microstructure in laser-induction hybrid cladding were investigated, in which the cladding material was provided by means of three different methods including the powder feeding, cold pre-placed coating (CPPC) and thermal pre-placed coating (TPPC). Moreover, the modes of energy action in laser-induction hybrid cladding were also studied. The results indicate that the cladding material supplying method has an important influence on the shape and microstructure of coating. The influence is decided by the mode of energy action in laser-induction hybrid cladding. During the TPPC hybrid cladding of Ni-based alloy, the laser and induction heating are mainly performed on coating. During the CPPC hybrid cladding of Ni-based alloy, the laser and induction heating are mainly performed on coating and substrate surface, respectively. In powder feeding hybrid cladding, a part of laser is absorbed by the powder particles directly, while the other part of laser penetrating powder cloud radiates on the molten pool. Meanwhile, the induction heating is entirely performed on the substrate. In addition, the wetting property on the interface is improved and the metallurgical bond between the coating and substrate is much easier to form. Therefore, the powder feeding laser-induction hybrid cladding has the highest cladding efficiency and the best bond property among three hybrid cladding methods.

  13. CO2 laser beam test of an actively cooled first-wall element with a graphite-clad SiC armor tile

    International Nuclear Information System (INIS)

    A graphite-clad SiC tile of 29mm dia and 15mm thickness is bonded to a base metal (Cu or 316 SS) with insertion of Cu-35 vol%C composite sheet. The elements are tested under active cooling condition by using 3.5kW CO2 laser beam at heat flux condition of 0.3 ∼ 1.7kW/cm2 and a pulse length of 40s

  14. The effect of oxidation on the emissivity of the fuel cladding of an advanced gas-cooled reactor during a temperature transient

    International Nuclear Information System (INIS)

    Changes in the hemispherical emissivity of stainless steel fuel cans were monitored during oxidation in CO2 as the metal temperature rose from 7000C to temperatures approaching the melting point of the cladding (approx. 13000C). Examination by scanning electron microscope revealed several types of oxide. In some cases, magnetite crystals were formed on top of the oxide layer with a corresponding reduction in the emissivity of about 5%. (author)

  15. Air-clad fibers: pump absorption assisted by chaotic wave dynamics?

    CERN Document Server

    Mortensen, Niels Asger

    2007-01-01

    Wave chaos is a concept which has already proved its practical usefulness in design of double-clad fibers for cladding-pumped fiber lasers and fiber amplifiers. In general, classically chaotic geometries will favor strong pump absorption and we address the extent of chaotic wave dynamics in typical air-clad geometries. While air-clad structures supporting sup-wavelength convex air-glass interfaces (viewed from the high-index side) will promote chaotic dynamics we find guidance of regular whispering-gallery modes in air-clad structures resembling an overall cylindrical symmetry. Highly symmetric air-clad structures may thus suppress the pump-absorption efficiency eta below the ergodic scaling law eta proportional to Ac/Acl, where Ac and Acl are the areas of the rare-earth doped core and the cladding, respectively.

  16. High burnup effects on the burst behavior of Zr based alloy claddings under LOCA conditions

    International Nuclear Information System (INIS)

    A current loss of coolant accident (LOCA) criterion is based on the results obtained from non pressurized claddings specimens under simulated LOCA condition. However, integrity of fuel cladding can be significantly affected by ballooning and rupture that caused by pressure difference between inner and outer cladding during LOCA. Ballooning may cause the fuel relocation or fuel dispersal due to its rupture opening during accidents. In addition, wall thickness of cladding can be reduced and local regions near the rupture open would become heavily oxidized and hydride d. Therefore, integral test that can simulate whole process during LOCA should be carried out for comprehensive safety analysis. Although a number of researches have been conducted, most investigations of them were performed using as received cladding specimens. In this study, burst behavior of several kinds of zirconium based alloys was investigated by integral LOCA test and high burnup effects on the burst behavior of fuel cladding were also examined using H charged cladding sample

  17. Oxide thickness measurement technique for duplex-layer Zircaloy-4 cladding

    Energy Technology Data Exchange (ETDEWEB)

    McClelland, R.G.; O' Leary, P.M. (Siemens Nuclear Power Corp., Richland, WA (United States))

    1992-01-01

    Siemens Nuclear Power Corporation (SNP) is investigating the use of duplex-layer Zircaloy-4 tubing to improve the waterside corrosion resistance of cladding for high-burnup pressurized water reactor (PWR) fuel designs. Standard SNP PWR cladding is typically 0.762-mm (0.030-in.)-thick Zircaloy-4. The SNP duplex cladding is nominally 0.660-mm (0.026-in.)-thick Zircalloy-4 with an [approximately]0.102-mm (0.004-in.) outer layer of another, more corrosion-resistant, zirconium-based alloy. It is common industry practice to monitor the in-reactor corrosion behavior of Zircaloy cladding by using an eddy-current lift-off' technique to measure the oxide thickness on the outer surface of the fuel cladding. The test program evaluated three different cladding samples, all with the same outer diameter and wall thickness: Zircaloy-4 and duplex clad types D2 and D4.

  18. Corrosion of research reactor aluminium clad spent fuel in water

    International Nuclear Information System (INIS)

    A large variety of research reactor spent fuel with different fuel meats, different geometries and different enrichments in 235U are presently stored underwater in basins located around the world. More than 90% of these fuels are clad in aluminium or aluminium based alloys that are notoriously susceptible to corrosion in water of less than optimum quality. Some fuel is stored in the reactor pools themselves, some in auxiliary pools (or basins) close to the reactor and some stored at away-from-reactor pools. Since the early 1990s, when corrosion induced degradation of the fuel cladding was observed in many of the pools, corrosion of research reactor aluminium clad spent nuclear fuel stored in light water filled basins has become a major concern, and programmes were implemented at the sites to improve fuel storage conditions. The IAEA has since then established a number of programmatic activities to address corrosion of research reactor aluminium clad spent nuclear fuel in water. Of special relevance was the Coordinated Research Project (CRP) on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase I) initiated in 1996, whose results were published in IAEA Technical Reports Series No. 418. At the end of this CRP it was considered necessary that a continuation of the CRP should concentrate on fuel storage basins that had demonstrated significant corrosion problems and would therefore provide additional insight into the fundamentals of localized corrosion of aluminium. As a consequence, the IAEA started a new CRP entitled Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase II), to carry out more comprehensive research in some specific areas of corrosion of aluminium clad spent nuclear fuel in water. In addition to this CRP, one of the activities under IAEA's Technical Cooperation Regional Project for Latin America Management of Spent Fuel from Research Reactors (2001-2006) was corrosion monitoring and surveillance of research

  19. Asymptotic Method for Cladding Stress Evaluation in PCMI

    International Nuclear Information System (INIS)

    A PCMI (Pellet Cladding Mechanical Interaction) failure was first reported in the GETR (General Electric Test Reactor) at Vacellitos in 1963, and such failures are still occurring. Since the high stress values in the cladding tube has been of a crucial concern in PCMI studies, there have been many researches on the stress analysis of a cladding tube pressed by a pellet. Typical works can be found in some references. It has often been assumed, however, that the cracks in the pellet were equally spaced and the pellet was a rigid body. In addition, the friction coefficient was arbitrarily chosen so that a slipping between the pellets and cladding tube could not be logically defined. Moreover, the stress intensification due to the sharp edge of a pellet fragment has never been realistically considered. These problems above drove us to launch a framework of a PCMI study particularly on stress analysis technology to improve the present analysis method incorporating the actual PCMI conditions such as the stress intensification, arbitrary distribution of the pellet cracks, material properties (esp. pellet) and slipping behavior of the pellet/cladding interface. As a first step of this work, this paper introduces an asymptotic method that was originally developed for a stress analysis in the vicinity of a sharp notch of a homogeneous body. The intrinsic reason for applying this method is to simulate the stress singularity that is expected to take place at the sharp edge of a pellet fragment due to cracking during irradiation. As a first attempt of this work, an eigenvalue problem is formulated in the case of adhered contact, and the generalized stress intensity factors are defined and evaluated. Although some works obviously remain to be accomplished, for the present framework on the PCMI analysis (e. g., slipping behaviour, contact force etc.), it was addressed that the asymptotic method can produce the stress values that cause the cladding tube failure in PCMI more

  20. Filtration and Leach Testing for PUREX Cladding Sludge and REDOX Cladding Sludge Actual Waste Sample Composites

    Energy Technology Data Exchange (ETDEWEB)

    Shimskey, Rick W.; Billing, Justin M.; Buck, Edgar C.; Casella, Amanda J.; Crum, Jarrod V.; Daniel, Richard C.; Draper, Kathryn E.; Edwards, Matthew K.; Hallen, Richard T.; Kozelisky, Anne E.; MacFarlan, Paul J.; Peterson, Reid A.; Swoboda, Robert G.

    2009-03-02

    A testing program evaluating actual tank waste was developed in response to Task 4 from the M-12 External Flowsheet Review Team (EFRT) issue response plan (Barnes and Voke 2006). The test program was subdivided into logical increments. The bulk water-insoluble solid wastes that are anticipated to be delivered to the Hanford Waste Treatment and Immobilization Plant (WTP) were identified according to type such that the actual waste testing could be targeted to the relevant categories. Under test plan TP RPP WTP 467 (Fiskum et al. 2007), eight broad waste groupings were defined. Samples available from the 222S archive were identified and obtained for testing. Under this test plan, a waste testing program was implemented that included: • Homogenizing the archive samples by group as defined in the test plan. • Characterizing the homogenized sample groups. • Performing parametric leaching testing on each group for compounds of interest. • Performing bench-top filtration/leaching tests in the hot cell for each group to simulate filtration and leaching activities if they occurred in the UFP2 vessel of the WTP Pretreatment Facility. This report focuses on a filtration/leaching test performed using two of the eight waste composite samples. The sample groups examined in this report were the plutonium-uranium extraction (PUREX) cladding waste sludge (Group 3, or CWP) and reduction-oxidation (REDOX) cladding waste sludge (Group 4, or CWR). Both the Group 3 and 4 waste composites were anticipated to be high in gibbsite, thus requiring caustic leaching. WTP RPT 167 (Snow et al. 2008) describes the homogenization, characterization, and parametric leaching activities before benchtop filtration/leaching testing of these two waste groups. Characterization and initial parametric data in that report were used to plan a single filtration/leaching test using a blend of both wastes. The test focused on filtration testing of the waste and caustic leaching for aluminum, in the form

  1. Measurement of cladding strain during simulated transient tests

    International Nuclear Information System (INIS)

    A diametral extensometer was developed and employed during temperature ramp tests with the Fuel Cladding Transient Tester (FCTT). Plastic strain measurements were performed using unirradiated 20% cold-worked AISI 316 stainless steel tubing ramped at 5.6 and 1110C/s with internal pressures from 3.4 to 93.1 MPa. Results demonstrated that plastic deformation can occur at stresses well below the conventional 0.2% yield strength and that most deformation in such tests occurs in the final 500C before failure. Postirradiation tests were performed on fuel pin cladding irradiated to 5.8 x 1022 n/cm (E > 0.1 MeV) with irradiation temperatures to 5400C. The tests showed that, for test pressures of 17.2 MPa or less, the stress-strain behavior was unchanged from unirradiated material behavior although the strains at failure were greatly decreased

  2. Modeling of Zircaloy cladding degradation under repository conditions

    International Nuclear Information System (INIS)

    Two potential degradation mechanisms, creep and stress corrosion cracking, of Zircaloy cladding during repository storage of spent nuclear fuel have been investigated. The deformation and fracture map methodology has been used to predict maximum allowable initial storage temperatures to achieve a thousand year life without rupture as a function of spent-fuel history. A stress analysis of fuel rods has been performed. Stresses in the outer zirconium oxide layer and the inner Zircaloy tube have been predicted for typical internal pressurization, oxide layer thickness, volume expansion from formation of the oxide layer and thermal expansion coefficients of the cladding and oxide. Stress relaxation occurring in-reactor has also been taken into account. The calculations indicate that for the anticipated storage conditions investigated, the outer zirconium oxide layer is in a state of compression thus making it unlikely that stress corrosion cracking of the exterior surface will occur. 20 refs., 6 figs., 9 tabs

  3. Chemical interaction at the FBR cladding fuel interfaces

    International Nuclear Information System (INIS)

    Pins containing UO2-30 wt.%PuO2 and/or Caesium and/or Telluriom as doping elements have been irradiated for about 40 days in the BR2 reactor. The effects of two Cs/Te ratios, namely 1.3 and 4 and a wide range of O/M ratios on the inner corrosion of the clad have been investigated. The influence of Tellurium on the attack of the cladding has been pointed out. It may be responsible for the Chromium NS Nickel depletion in the grain boundaries of the steel. It is necessary to measure the effective Ts/Te ratio associated with the local corrosion layers. This local Cs/Te ratio should be more useful than the initial mean Cs/Te ratio in a pin for understanding the corrosion phenomene. (author)

  4. Iodine induced stress corrosion cracking of Zircaloy fuel cladding materials

    International Nuclear Information System (INIS)

    This report documents the work performed by the Co-ordinated Research Project (CRP) on Stress Corrosion Cracking of Zirconium Alloy Fuel Cladding. The project consisted of out-of-pile laboratory measurements of crack propagation rates in Zircaloy sheet specimens in an iodine containing atmosphere. The project was overseen by a supervisory group consisting of experts in the field, who also contributed a state of the art review. This report describes all of the work undertaken as part of the CRP, and includes: a review of the state of the art understanding of stress corrosion cracking behaviour of zirconium alloy cladding material; a description of the experimental equipment, test procedures, material characterizations and test matrix; discussion of the work undertaken by the host laboratory and the specific contributions by each of the four participant laboratories; a compilation of all experimental results obtained; and the supervisory group's analysis and discussion of the results, plus conclusions and recommendations

  5. Corrosion of spent nuclear fuel aluminium cladding in ordinary water

    International Nuclear Information System (INIS)

    Corrosion of aluminium alloy cladding of spent nuclear fuel elements in ordinary water is examined in the spent fuel storage pool of the RA research reactor at the Vinca Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro. Experimental examinations are carried out within framework of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) 'Corrosion of Research Reactor Aluminium-Clad Spent Fuel in Water', Phase II. Racks with coupons made of different aluminium alloys were exposed to water influence for period of six months to six years. The project comprises also activities on monitoring of the water chemistry and radioactivity in the storage pool. Visual and microscopic examinations of surfaces of aluminium coupons of the test racks have been done recently and results were presented in this paper confirming strong influence of water quality and exposition time to corrosion process. (author)

  6. Clad — Automatic Differentiation Using Clang and LLVM

    Science.gov (United States)

    Vassilev, V.; Vassilev, M.; Penev, A.; Moneta, L.; Ilieva, V.

    2015-05-01

    Differentiation is ubiquitous in high energy physics, for instance in minimization algorithms and statistical analysis, in detector alignment and calibration, and in theory. Automatic differentiation (AD) avoids well-known limitations in round-offs and speed, which symbolic and numerical differentiation suffer from, by transforming the source code of functions. We will present how AD can be used to compute the gradient of multi-variate functions and functor objects. We will explain approaches to implement an AD tool. We will show how LLVM, Clang and Cling (ROOT's C++11 interpreter) simplifies creation of such a tool. We describe how the tool could be integrated within any framework. We will demonstrate a simple proof-of-concept prototype, called Clad, which is able to generate n-th order derivatives of C++ functions and other language constructs. We also demonstrate how Clad can offload laborious computations from the CPU using OpenCL.

  7. Full vectorial analysis of multilayer leaky cladding optical fibre

    CERN Document Server

    Labonté, Laurent; Kumar, A; Dussardier, Bernard; Monnom, Gérard

    2010-01-01

    We analyze a multilayer leaky cladding (MLC) fibre using the finite element method and study the effect of the MLC on the bending loss and birefringence of two types of structures: i) a circular-core large-mode area structure and ii) an elliptical-small-core structure. In a large-mode-area structure, we verify that the multi-layer leaky cladding strongly discriminates against higher order modes to achieve single-mode operation, the fibre shows negligible birefringence, and the bending loss of the fibre is low for bending radii larger than 10 cm. In the elliptical-small-core structure we show that the MLC reduces the birefringence of the fibre. This prevents the structure from becoming birefringent in case of any departures from circular geometry. The study should be useful in the designs of MLC fibres for various applications including high-power amplifiers, gain flattening of fibre amplifiers and dispersion compensation.

  8. Processing and properties of Ag-clad BSCCO superconductors

    International Nuclear Information System (INIS)

    Long lengths of mono- and multifilament Ag-clad BSCCO (Bi-Sr-Ca-Cu-O) conductors with critical current densities of >104 A/cm2 at 77 K were fabricated by the powder-in-tube method. Tc magnets were assembled by stacking pancake coils fabricated from long tapes and then tested vs applied magnetic field at various temperatures. A magnet that contained ∼2400 m of Tc conductor generated a field of 3.2 T at 4.2 K. In-situ tensile and bending properties of the Ag-clad conductors were studied. Multilayer Ag/superconductor composites were fabricated by chemical etching. Preliminary results with multilayer tapes show that continuous Ag reinforcement of the BSCCO core improves strain tolerance of the tapes so they can carry 90% of their initial Ic at 1% bend strain desite a higher superconductor/Ag ratio than that of unreinforced tapes

  9. Zircaloy cladding performance under spent fuel disposal conditions

    International Nuclear Information System (INIS)

    The Brookhaven National Laboratory (BNL) Waste Materials and Environment Modeling (WMEM) Program has been assigned the task of helping the DOE formulate and certify analytical tools needed to support and/or strengthen the Waste Package Licensing Strategy. One objective of the WMEM program is to perform qualitative and quantitative analyses of irradiated Zircaloy cladding. This progress report presents the early findings of an on-going literature evaluation and the results of the numerical implementation of two models of Zircaloy creep. The report only addresses cladding degradation modes within intact, dry waste containers. Additional degradation modes will be considered when the study is expanded to include moist environments and partly failed containers. Further updates of the present analyses will also be provided

  10. Spent fuel resistance to internally produced cladding degradation

    International Nuclear Information System (INIS)

    These tests were conducted over a narrow temperature range considerably above anticipated disposal conditions and utilized only one set of rods from a single reactor. The following conclusions are made: The measured cladding strain was sufficiently large so that failure mechanism verification by inducing breaches in unmodified rods heated to elevated temperatures for short periods of time does not appear to be practical based on a stress rupture mechanism. At the elevated test temperatures, though, Blackburn's formulization based on stress rupture gives very conservative estimates of breach times. In addition to the high cladding strain, the fuel exhibited no additional gas release or axial fission product migration at 4820C. The nondestructive examination gave no additional indication of internal deterioration of the fuel rod

  11. Modeling of realistic cladding structures for photonic bandgap fibers

    DEFF Research Database (Denmark)

    Mortensen, Niels Asger; Nielsen, Martin Dybendal

    2004-01-01

    Cladding structures of photonic bandgap fibers often have airholes of noncircular shape, and, typically, close-to-hexagonal airholes with curved corners are observed. We study photonic bandgaps in such structures by aid of a two-parameter representation of the size and curvature. For the fundamen......Cladding structures of photonic bandgap fibers often have airholes of noncircular shape, and, typically, close-to-hexagonal airholes with curved corners are observed. We study photonic bandgaps in such structures by aid of a two-parameter representation of the size and curvature. For the...... fundamental bandgap we find that the bandgap edges (the intersections with the air line) shift toward shorter wavelengths when the air-filling fraction f is increased. The bandgap also broadens, and the relative bandwidth increases exponentially with f2. Compared with recent experiments [Nature 424, 657 (2003...

  12. The characterization of activities associated with irradiated fuel element claddings

    International Nuclear Information System (INIS)

    The object of the present work was to characterise the natures and amounts of the various α and βγ activities associated with cladding hulls. The claddings studied were stainless steel from a Fast Reactor and from an Advanced Gas Reactor and Zircaloy from a Boiling Water Reactor, from a Pressurized Water Reactor and from a Steam Generating Heavy Water Reactor. The hulls were examined by the following methods: alpha spectrometry to identify and quantify the α emitters and to estimate their depths of penetration, partial and complete dissolution of hulls followed by gross α counting, α spectrometry and γ spectrometry, fission track autoradiography to determine the distribution of fissile material associated with hulls, neutron activation to determine the total fissile content of the hulls, chemical separations followed by β counting and chemical treatment with various reagents to examine the ease of decontamination

  13. Evaluation of Tritium Content and Release from Pressurized Water Reactor Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Sharon M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chattin, Marc Rhea [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Giaquinto, Joseph [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jubin, Robert Thomas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    It is expected that tritium pretreatment will be required in future reprocessing plants to prevent the release of tritium to the environment (except for long-cooled fuels). To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified. Tritium in light water reactor (LWR) fuel is dispersed between the fuel matrix and the fuel cladding, and some tritium may be in the plenum, probably as tritium labelled water (THO) or T2O. In a standard processing flowsheet, tritium management would be accomplished by treatment of liquid streams within the plant. Pretreating the fuel prior to dissolution to release the tritium into a single off-gas stream could simplify tritium management, so the removal of tritium in the liquid streams throughout the plant may not be required. The fraction of tritium remaining in the cladding may be reduced as a result of tritium pretreatment. Since Zircaloy® cladding makes up roughly 25% by mass of UNF in the United States, processes are being considered to reduce the volume of reprocessing waste for Zircaloy® clad fuel by recovering the zirconium from the cladding for reuse. These recycle processes could release the tritium in the cladding. For Zircaloy-clad fuels from light water reactors, the tritium produced from ternary fission and other sources is expected to be divided between the fuel, where it is generated, and the cladding. It has been previously documented that a fraction of the tritium produced in uranium oxide fuel from LWRs can migrate and become trapped in the cladding. Estimates of the percentage of tritium in the cladding typically range from 0–96%. There is relatively limited data on how the tritium content of the cladding varies with burnup and fuel history (temperature, power, etc.) and how pretreatment impacts its release. To gain a better understanding of how tritium in cladding

  14. Study of zircaloy-4 cladding air degradation at high temperature

    OpenAIRE

    Lasserre, Marina; COINDREAU, Olivia; Pijolat, Michèle; Peres, Véronique; Mermoux, Michel; Mardon, Jean Paul

    2013-01-01

    Zircaloy cladding, providing the first containment of UO2 fuel in Pressurised Water Reactors, can be exposed to air during accidental situations. This might occur during reactor operation (in case of a core meltdown accident with subsequent reactor pressure vessel breaching), under shutdown conditions with the upper head of the vessel removed, in spent fuel storage pools after accidental loss of cooling or during degraded transport situations. The fuel assemblies inadequately cooled, heat up ...

  15. Modelling nuclear fuel behaviour and cladding viscoelastic response

    OpenAIRE

    Tulkki, Ville

    2015-01-01

    In light water reactors the nuclear fuel is in the form of uranium dioxide pellets stacked inside a thin-walled tube made from Zirconium alloy. The fuel rods provide the first barriers to the release of radioactivity as the isotopes are contained within the fuel matrix and the cladding tubes. Fuel behaviour analysis investigates the state of the fuel at given boundary conditions and irradiation history. The scope of this thesis consists of two main themes. The first is the uncertainty and ...

  16. Failure Characteristic of Laser Cladding Samples on Repeated Impact

    Institute of Scientific and Technical Information of China (English)

    SHI Shi-hong; ZHENG Qi-guang; FU Ge-yan; ZHANG Jin-ping

    2004-01-01

    Using self-made impact fatigue test instruments and related analytic devices,the mechanical components with laser cladding layer have been attempted.It is found that,on repeated impact force,several failure modes of the components include the surface cracks,surface plastic deformation,corrosive pitting and coat collapse,etc.The paper reported the test method and initial analysis conclusions about the unique failure characteristics of the mechanical components on repeated impact load.

  17. Eddy-Current Testing of Finned Fuel Cladding

    International Nuclear Information System (INIS)

    Eddy-current methods of testing reactor-fuel components are well established. The literature, however, mainly describes tests which are applied to simple geometries such as cylindrical rods or tubes. Recent AECL fuel designs have called for cladding with heat transfer or locating fins along the length of the fuel. This paper describes the application of eddy-current techniques to three such designs. The function and geometry of the fins must be considered in the selection of the optimum test parameters and the most suitable test coil geometry. Thus, the presence of fins may limit or restrict the test but they will not prevent a successful test. Where the fin geometry is complex eddy currents may well be the most suitable of the non-destructive methods which can be used for flaw detection. The thickness of aluminium cladding over a uranium core is measured with a small probe coil placed between the fins and shielded from them. Two flaw detection tests are described, one on sintered aluminium product (SAP) tubing using an internal bobbin coil and the other on an aluminium-clad uranium-aluminium alloy rod with an external encircling coil. The instrumentation described is relatively simple. A small portable instrument was designed for the cladding thickness measurement. For flaw detection a standard oscilloscope with a plug-in carrier-amplifier module provides a means of sensing and displaying the test coil impedance variations. This equipment ,although it does not permit sophisticated methods of eliminating unwanted noise is adequate for a variety of testing applications and has been specified for routine fuel testing on a production basis. (author)

  18. Test plan for spent fuel cladding containment credit tests

    International Nuclear Information System (INIS)

    Lawrence Livermore National Laboratory has chosen Westinghouse Hanford Company as a subcontractor to assist them in determining the requirements for successful disposal of spent fuel rods in the proposed Nevada Test Site repository. An initial scoping test, with the objective of determining whether or not the cladding of a breached fuel rod can be given any credit as an effective barrier to radionuclide release, is described in this test plan. 8 references, 2 figures, 4 tables

  19. Core temperature in super-Gaussian pumped air-clad photonic crystal fiber lasers compared with double-clad fiber lasers

    Indian Academy of Sciences (India)

    P Elahi; H Nadgaran; F Kalantarifard

    2007-03-01

    In this paper we investigate the core temperature of air-clad photonic crystal fiber (PCF) lasers pumped by a super-Gaussian (SG) source of order four. The results are compared with conventional double-clad fiber (DCF) lasers pumped by the same super-Gaussian and by top-hat pump profiles.

  20. The clad collapse modelling of Indian PHWR fuel element, an FEM approach

    International Nuclear Information System (INIS)

    The fuel elements for PHWR use a thin, collapsible zircaloy clad design. This design is consistent with essential neutron economy in PHWRs, and also results in better heat transfer between fuel and clad. However, thin clad may give rise to problem of permanent clad collapse under coolant pressure in axial gap and radial gap available during the initial stay of fuel inside the reactor. Present work explores the problem of longitudinal ridges, formed due to permanent circumferential collapse of clad on fuel. The tip of these ridges has the potential to become the site for crack initiation under subsequent cyclic thermal/pressure loading. The collapse behavior of fuel element is studied using FEM modeling of pellet, clad and their contact. This study considers the effects of clad thickness, clad yield strength, clad initial ovality, anisotropy in clad yield strength, and radial gap of fuel element on the collapse behavior. The verification of present model is done for the results of critical buckling pressure required for the longitudinal ridge formation by the available CANDU experimental data, which matched satisfactorily for the yield strength ratio (circumferential YS to longitudinal YS) of 1.5. In addition the longitudinal ridge height and increase in ovality were calculated for the collapse experiments done on the 220 MWe PHWR fuel elements

  1. Microbial biofilm growth on irradiated, spent nuclear fuel cladding

    International Nuclear Information System (INIS)

    A fundamental criticism regarding the potential for microbial influenced corrosion in spent nuclear fuel cladding or storage containers concerns whether the required microorganisms can, in fact, survive radiation fields inherent in these materials. This study was performed to unequivocally answer this critique by addressing the potential for biofilm formation, the precursor to microbial-influenced corrosion, in radiation fields representative of spent nuclear fuel storage environments. This study involved the formation of a microbial biofilm on irradiated spent nuclear fuel cladding within a hot cell environment. This was accomplished by introducing 22 species of bacteria, in nutrient-rich media, to test vessels containing irradiated cladding sections and that was then surrounded by radioactive source material. The overall dose rate exceeded 2 Gy/h gamma/beta radiation with the total dose received by some of the bacteria reaching 5 x 103 Gy. This study provides evidence for the formation of biofilms on spent-fuel materials, and the implication of microbial influenced corrosion in the storage and permanent deposition of spent nuclear fuel in repository environments

  2. Development of advanced cladding material for burnup extension

    International Nuclear Information System (INIS)

    The development of new cladding materials is one of the critical issues on burnup extension. The practical life of Zircaloy would be limited by the growth of oxide films and by the ductility loss due to hydride precipitation, oxygen absorption and radiation damage. In the case of high burnup using MOX fuels, the low neutron adsorption cross section of Zircaloy is not a dominant factor for selecting the cladding material, because MOX fuels can be enriched up to 20%Pu. Austenitic stainless steel, titanium alloy, niobium alloy, ferritic steel and nickel base superalloy are considered as candidate materials. The corrosion resistance, mechanical properties and the irradiation resistance of these materials were examined for evaluating the practical possibility as a cladding material. The austenitic stainless steel with high g phase stability was selected as the primary candidate material. However, it is required to improve the resistance to irradiation associated stress corrosion cracking through the experience in LWR plants. In the JAERI, the austenitic stainless steel with intergranular corrosion resistance has been developed by the adjustment of the chemical composition, the modification of the metallographic structure by thermo-mechanical treatment and the purification by electron beam melting. (author)

  3. Final report on accident tolerant fuel performance analysis of APMT-Steel Clad/UO₂ fuel and APMT-Steel Clad/UN-U₃Si₅ fuel concepts

    Energy Technology Data Exchange (ETDEWEB)

    Unal, Cetin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Galloway, Jack D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-09-12

    In FY2014 our group completed and documented analysis of new Accident Tolerant Fuel (ATF) concepts using BISON. We have modeled the viability of moving from Zircaloy to stainless steel cladding in traditional light water reactors (LWRs). We have explored the reactivity penalty of this change using the MCNP-based burnup code Monteburns, while attempting to minimize this penalty by increasing the fuel pellet radius and decreasing the cladding thickness. Fuel performance simulations using BISON have also been performed to quantify changes to structural integrity resulting from thinner stainless steel claddings. We account for thermal and irradiation creep, fission gas swelling, thermal swelling and fuel relocation in the models for both Zircaloy and stainless steel claddings. Additional models that account for the lower oxidation stainless steel APMT are also invoked where available. Irradiation data for HT9 is used as a fallback in the absence of appropriate models. In this study the isotopic vectors within each natural element are varied to assess potential reactivity gains if advanced enrichment capabilities were levied towards cladding technologies. Recommendations on cladding thicknesses for a robust cladding as well as the constitutive components of a less penalizing composition are provided. In the first section (section 1-3), we present results accepted for publication in the 2014 TOPFUEL conference regarding the APMT/UO₂ ATF concept (J. Galloway & C. Unal, Accident Tolerant and Neutronically Favorable LWR Cladding, Proceedings of WRFPM 2014, Sendai, Japan, Paper No.1000050). Next we discuss our preliminary findings from the thermo-mechanical analysis of UN-U₃Si₅ fuel with APMT clad. In this analysis we used models developed from limited data that need to be updated when the irradiation data from ATF-1 test is available. Initial results indicate a swelling rate less than 1.5% is needed to prevent excessive clad stress.

  4. Power ramped cladding stresses and strains in 3D simulations with burnup-dependent pellet–clad friction

    International Nuclear Information System (INIS)

    Highlights: ► This paper presents 2D plane strain and 3D simulations of pellet–cladding interaction during base irradiation and ramp tests. ► Inverse analysis is used to estimate the evolution of friction at the pellet–clad interface with burnup. ► The number of radial cracks that form in ramped rodlets is the main parameter on which inverse analysis is based. ► Calculations show that the sole evolution of the friction coefficient with burnup is sufficient to capture the radial crack pattern. ► A simple relation between the friction coefficient and the burnup is thus proposed and used in 3D simulations of PCI. - Abstract: This paper presents 2D(r, θ) plane strain and 3D simulations of PCI during base irradiation and ramp tests. Inverse analysis is used to estimate the evolution of friction at the pellet–clad interface with burnup. The number of radial cracks that form during power ramp tests in seventeen UO2-Zy4 rodlets with burnups in the range 20–60 GWd/tU is the main parameter on which inverse analysis is based. It is shown that the sole evolution of the friction coefficient with burnup is sufficient to capture the radial crack pattern of the rodlets after power ramping. A simple relation between the friction coefficient and the burnup variation after initial pellet–clad contact is thus proposed and used in 3D simulations of PCI. The delayed gap closing at mid-pellet level with respect to inter-pellet level leads to an axial variation of the friction coefficient, with maximum values near the pellet ends. The consequences in terms of PCI failure propensity are then discussed.

  5. Hierarchical microstructure of explosive joints: Example of titanium to steel cladding

    International Nuclear Information System (INIS)

    Research highlights: → The cladding interface consists of four hierarchical levels. → FeTi and Fe9.64Ti0.36 inclusions form at wave crest due to partial melting. → Intermetallic inclusions are surrounded by the heat affected zone of parent steel. → Thin reaction layer along the interface ensures good macroscopic bonding of the joints. - Abstract: The microstructure of explosive cladding joints formed among parallel Ti and steel plates was examined by electron microscopy. The bonding interface and the bulk materials around it form pronounced hierarchical microstructures. This hierarchy is characterized by the following features: at the mesoscopic scale of the hierarchy a wavy course of the interface characterizes the interface zone. This microstructure level is formed by heavy plastic shear waves (wavelength ∼ 0.5 mm) which expand within the two metal plates during the explosion parallel to the bonding interface. At the micro-scale range, intermetallic inclusions (size ∼ 100-200 μm) are formed just behind the wave crests on the steel side as a result of partial melting. Electron diffraction revealed FeTi and metastable Fe9.64Ti0.36. Most of the observed phases do not appear in the equilibrium Fe-Ti phase diagram. These intermetallic inclusions are often accompanied by micro-cracks of similar dimension. At the smallest hierarchy level we observe a reaction layer of about 100-300 nm thickness consisting of nano-sized grains formed along the entire bonding interface. Within that complex hierarchical micro- and nanostructure, the mesoscopic regime, more precisely the type and brittleness of the intermetallic zones, seems to play the dominant role for the mechanical behavior of the entire compound.

  6. Analysis and optimization of process parameters in Al-SiCp laser cladding

    Science.gov (United States)

    Riquelme, Ainhoa; Rodrigo, Pilar; Escalera-Rodríguez, María Dolores; Rams, Joaquín

    2016-03-01

    The laser cladding process parameters have great effect on the clad geometry and on dilution in the single and multi-pass aluminum matrix composite reinforced with SiC particles (Al/SiCp) coatings on ZE41 magnesium alloys deposited using a high-power diode laser (HPLD). The influence of the laser power (500-700 W), scan speed (3-17 mm/s) and laser beam focal position (focus, positive and negative defocus) on the shape factor, cladding-bead geometry, cladding-bead microstructure (including the presence of pores and cracks), and hardness has been evaluated. The correlation of these process parameters and their influence on the properties and ultimately, on the feasibility of the cladding process, is demonstrated. The importance of focal position is demonstrated. The different energy distribution of the laser beam cross section in focus plane or in positive and negative defocus plane affect on the cladding-bead properties.

  7. Characterization of SiC-SiC composites for accident tolerant fuel cladding

    Science.gov (United States)

    Deck, C. P.; Jacobsen, G. M.; Sheeder, J.; Gutierrez, O.; Zhang, J.; Stone, J.; Khalifa, H. E.; Back, C. A.

    2015-11-01

    Silicon carbide (SiC) is being investigated for accident tolerant fuel cladding applications due to its high temperature strength, exceptional stability under irradiation, and reduced oxidation compared to Zircaloy under accident conditions. An engineered cladding design combining monolithic SiC and SiC-SiC composite layers could offer a tough, hermetic structure to provide improved performance and safety, with a failure rate comparable to current Zircaloy cladding. Modeling and design efforts require a thorough understanding of the properties and structure of SiC-based cladding. Furthermore, both fabrication and characterization of long, thin-walled SiC-SiC tubes to meet application requirements are challenging. In this work, mechanical and thermal properties of unirradiated, as-fabricated SiC-based cladding structures were measured, and permeability and dimensional control were assessed. In order to account for the tubular geometry of the cladding designs, development and modification of several characterization methods were required.

  8. AgI-coated silver-clad stainless steel hollow waveguides for infrared lightwave transmission and their applications.

    Science.gov (United States)

    Hongo, Akihito; Sato, Shinobu; Hattori, Akio; Iwai, Katsumasa; Takaku, Hiroyuki; Miyagi, Mitsunobu

    2012-01-01

    We fabricated silver iodide (AgI)-coated silver hollow waveguides to transmit a wide range of infrared (IR) light. Silver-clad stainless steel pipes were used as a supporting pipe. Since this type of metallic hollow waveguide has high mechanical strength and heat resistance, it is suitable as a rigid lightwave probe for various applications such as dental or medical laser treatment, IR spectroscopy, thermal radiometry, and laser processing. Considering these applications, we estimated the hollow waveguides with different thicknesses of the AgI layer. By optimizing the AgI layer thickness according to the wavelength of propagating light, we succeeded in efficiently transmitting Er-YAG and CO(2) laser light. We also studied the optical characteristics of a wide range of incoherent light for IR spectroscopy and radiometry applications using these metallic hollow waveguides as lightwave probes. PMID:22270406

  9. Study of reactions between fuel (mixed oxide (UPu)Osub(2-x)) and cladding (stainless-steel) in reactors: influence of iodine compounds

    International Nuclear Information System (INIS)

    The influence of iodine compounds on the development of the oxide-cladding reaction was examined. The action of iodine, cesium and cesium iodide on type 316 stainless was determined in the presence or absence of uranium oxide or mixed uranium-plutonium oxide type fuel in a closed system, isothermal or with a temperature gradient. The study of the stainless steel iodine reactions was developed in particular. These experiments showed that cesium combines with uranium oxide to give cesium uranate Cs2U2O7; it is not unreasonable to suppose that cesium urano-plutonate Cs2(U,Pu)2O7 could be formed inside the pile. It was then shown that cesium iodide in the presence of sufficiently non-stoichiometric mixed oxide could contribute towards the degradation of the stainless steel cladding. Under these conditions the reaction is accompained by a transport of manganese, chromium and iron into the hot parts of the fuel by a Van-Arkel type mechanism. This might explain the presence of metallic precipitates in the fuel, but the role assigned to molybdenum iodide in the same phenomenon is considered unlikely. Finally it is proposed to deposit a thin layer of manganese metal on the inner surface of the cladding in order to minimize the action of fission products (CsI, Te)

  10. Reflood behavior at low initial clad temperature in Slab Core Test Facility Core-II

    International Nuclear Information System (INIS)

    In order to study the reflood behavior with low initial clad temperature, a reflood test was performed using the Slab Core Test Facility (SCTF) with initial clad temperature of 573 K. The test conditions of the test are identical with those of SCTF base case test S2-SH1 (initial clad temperature 1073 K) except the initial clad temperature. Through the comparison of results from these two tests, the following conclusions were obtained. (1) The low initial clad temperature resulted in the low differential pressures through the primary loops due to smaller steam generation in the core. (2) The low initial clad temperature caused the accumulated mass in the core to be increased and the accumulated mass in the downcomer to be decreased in the period of the lower plenum injection with accumulator (before 50s). In the later period of the cold leg injection with LPCI (after 100s), the water accumulation rates in the core and the downcomer were almost the same between both tests. (3) The low initial clad temperature resulted in the increase of the core inlet mass flow rate in the lower plenum injection period. However, the core inlet mass flow rate was almost the same regardless of the initial clad temperature in the later period of the cold leg injection period. (4) The low initial clad temperature resulted in the low turnaround temperature, high temperature rise and fast bottom quench front propagation. (5) In the region apart from the quench front, low initial clad temperature resulted in the lower heat transfer. In the region near the quench front, almost the same heat transfer coefficient was observed between both tests. (6) No flow oscillation with a long period was observed in the SCTF test with low initial clad temperature of 573 K, while it was remarkable in the Cylindrical Core Test Facility (CCTF) test which was performed with the same initial clad temperature. (J.P.N.)

  11. Assessment of thin-walled cladding tube mechanical properties by segmented expanding Mandrel test

    OpenAIRE

    NILSSON Karl-Fredrik

    2013-01-01

    This paper presents the principles of the segmented expanding mandrel test for thin-walled cladding tubes, which can be used as a basic material characterisation test to determine stressstrain curves and ductility or as a test to simulate mechanical pellet-cladding interaction. The paper discusses the strengths and weaknesses of the test method and it illustrates how the test can be used to simulate hydride reorientations in zirconium claddings and quantify how hydride reorientation affect...

  12. Efficient waveguide lasers in femtosecond laser inscribed double-cladding waveguides of Yb:YAG ceramics

    OpenAIRE

    Jia, Yuechen; R. Vázquez de Aldana, Javier; Chen, Feng

    2013-01-01

    We report on the fabrication of depressed double-cladding waveguides in Yb:YAG ceramics by using femtosecond (fs) laser inscription. The double-cladding structures consist of tubular central structures with 30 μm diameter and concentric larger size tubular claddings with diameters of 100-200 μm. Continuous wave laser oscillations at wavelength of 1030 nm have been realized at room temperature through optical pump at 946 nm. The obtained maximum output power of the double-...

  13. Air-clad fibers: pump absorption assisted by chaotic wave dynamics?

    OpenAIRE

    Mortensen, Niels Asger

    2007-01-01

    Wave chaos is a concept which has already proved its practical usefulness in design of double-clad fibers for cladding-pumped fiber lasers and fiber amplifiers. In general, classically chaotic geometries will favor strong pump absorption and we address the extent of chaotic wave dynamics in typical air-clad geometries. While air-clad structures supporting sup-wavelength convex air-glass interfaces (viewed from the high-index side) will promote chaotic dynamics we find guidance of regular whis...

  14. Simulation of a pellet-clad mechanical interaction with ABAQUS and its verification

    International Nuclear Information System (INIS)

    Pellet-clad mechanical interaction (PCMI) during power transients for MOX fuel is modelled by a FE method. The PCMI model predicts well clad elongation during power ramp and relaxation during power hold except the fuel behaviour during a power decrease. Higher fiction factor results in the earlier occurrence of PCMI and more enhanced clad elongation. The relaxation is dependent on the irradiation creep rate of the pellet and axial compressive force. Verification of the PCMI model was done using recent MOX experimental data. Temperature and clad elongation for the fuel rod can be evaluated in a reasonable way

  15. Review of experimental data for modelling LWR fuel cladding behaviour under loss of coolant accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park (Sweden)

    2007-02-15

    Extensive range of experiments has been conducted in the past to quantitatively identify and understand the behaviour of fuel rod under loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs). The obtained experimental data provide the basis for the current emergency core cooling system acceptance criteria under LOCA conditions for LWRs. The results of recent experiments indicate that the cladding alloy composition and high burnup effects influence LOCA acceptance criteria margins. In this report, we review some past important and recent experimental results. We first discuss the background to acceptance criteria for LOCA, namely, clad embrittlement phenomenology, clad embrittlement criteria (limitations on maximum clad oxidation and peak clad temperature) and the experimental bases for the criteria. Two broad kinds of test have been carried out under LOCA conditions: (i) Separate effect tests to study clad oxidation, clad deformation and rupture, and zirconium alloy allotropic phase transition during LOCA. (ii) Integral LOCA tests, in which the entire LOCA sequence is simulated on a single rod or a multi-rod array in a fuel bundle, in laboratory or in a tests and results are discussed and empirical correlations deduced from these tests and quantitative models are conferred. In particular, the impact of niobium in zirconium base clad and hydrogen content of the clad on allotropic phase transformation during LOCA and also the burst stress are discussed. We review some recent LOCA integral test results with emphasis on thermal shock tests. Finally, suggestions for modelling and further evaluation of certain experimental results are made.

  16. Development of data base with mechanical properties of un- and pre-irradiated VVER cladding

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V.; Yegorova, L.; Kaplar, E.; Lioutov, K. [Nuclear Safety Inst. of Russian Research Centre, Moscow (Russian Federation). Kurchatov Inst.; Smirnov, V.; Prokhorov, V.; Goryachev, A. [State Research Centre, Dimitrovgrad (Russian Federation). Research Inst. of Atomic Reactors

    1998-03-01

    Analysis of recent RIA test with PWR and VVER high burnup fuel, performed at CABRI, NSRR, IGR reactors has shown that the data base with mechanical properties of the preirradiated cladding is necessary to interpret the obtained results. During 1997 the corresponding cycle of investigations for VVER clad material was performed by specialists of NSI RRC KI and RIAR in cooperation with NRC (USA), IPSN (France) in two directions: measurements of mechanical properties of Zr-1%Nb preirradiated cladding versus temperature and strain rate; measurements of failure parameters for gas pressurized cladding tubes. Preliminary results of these investigations are presented in this paper.

  17. Design of intrinsically single-mode double clad crystalline fiber waveguides for high power lasers

    Science.gov (United States)

    Li, Da; Hong, Pengda; Meissner, Stephanie K.; Meissner, Helmuth E.

    2016-03-01

    Recently, double-clad crystalline fiber waveguides (CFWs), consisting of single crystalline or ceramic RE3+:YAG cores of square cross section and inner claddings of either undoped or laser-inactive-ion-doped YAG and outer claddings of sapphire, have been successfully demonstrated. These waveguides, manufactured by an Adhesive-Free Bonding (AFB®) technique, can be precisely engineered and fabricated with predictable beam propagation behavior. In this work, with high power laser designs in mind, minimum thicknesses for inner cladding are derived for different core cross sections and refractive index differences between the core and inner cladding and sapphire as outer cladding material for common laser core dopants such as Nd3+, Yb3+, Er3+, Tm3+ and Ho3+. All designs are intended to use high NA high power laser diode pumping to obtain high power intrinsically single transverse mode laser output. The obtained data are applicable to any crystalline fiber waveguide design, regardless of fabrication technique. As an example, a CFW with 40 μm × 40 μm 4% Tm:YAG core, 5% Yb:YAG inner cladding, and sapphire outer cladding was calculated to be intrinsically single transverse mode, with the minimum inner cladding width of 21.7 μm determined by the effective index technique [1].

  18. Impact of thicker cladding on the nuclear parameters of the NPP Krsko fuel

    International Nuclear Information System (INIS)

    To make fuel rods more resistant to grid-to-rod fretting or other cladding penetration failures, the cladding thickness could be increased or strengthened. Implementation of thicker fuel rod cladding was evaluated for the NPP Krsko that uses 16 x 16 fuel design. Cladding thickness of the Westinghouse standard fuel design (STD) and optimized fuel design (OFA) is increased. The reactivity effect during the fuel burnup is determined. To obtain a complete realistic view of the fuel behaviour a typical, near equilibrium, 18-month fuel cycle is investigated. The most important nuclear core parameters such as critical boron concentrations, isothermal temperature coefficient and rod worth are determined and compared.

  19. Air-clad fibers: pump absorption assisted by chaotic wave dynamics?

    DEFF Research Database (Denmark)

    Mortensen, Niels Asger

    2007-01-01

    Wave chaos is a concept which has already proved its practical usefulness in design of double-clad fibers for cladding-pumped fiber lasers and fiber amplifiers. In general, classically chaotic geometries will favor strong pump absorption and we address the extent of chaotic wave dynamics in typical...... air-clad structures may thus suppress the pump-absorption efficiency η below the ergodic scaling law η∞ Ac/Acl, where Ac and Acl are the areas of the rare-earth doped core and the cladding, respectively....

  20. Review of experimental data for modelling LWR fuel cladding behaviour under loss of coolant accident conditions

    International Nuclear Information System (INIS)

    Extensive range of experiments has been conducted in the past to quantitatively identify and understand the behaviour of fuel rod under loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs). The obtained experimental data provide the basis for the current emergency core cooling system acceptance criteria under LOCA conditions for LWRs. The results of recent experiments indicate that the cladding alloy composition and high burnup effects influence LOCA acceptance criteria margins. In this report, we review some past important and recent experimental results. We first discuss the background to acceptance criteria for LOCA, namely, clad embrittlement phenomenology, clad embrittlement criteria (limitations on maximum clad oxidation and peak clad temperature) and the experimental bases for the criteria. Two broad kinds of test have been carried out under LOCA conditions: (i) Separate effect tests to study clad oxidation, clad deformation and rupture, and zirconium alloy allotropic phase transition during LOCA. (ii) Integral LOCA tests, in which the entire LOCA sequence is simulated on a single rod or a multi-rod array in a fuel bundle, in laboratory or in a tests and results are discussed and empirical correlations deduced from these tests and quantitative models are conferred. In particular, the impact of niobium in zirconium base clad and hydrogen content of the clad on allotropic phase transformation during LOCA and also the burst stress are discussed. We review some recent LOCA integral test results with emphasis on thermal shock tests. Finally, suggestions for modelling and further evaluation of certain experimental results are made

  1. Path planning and kinematics simulation of surfacing cladding for hot forging die

    Directory of Open Access Journals (Sweden)

    Wang Huajun

    2015-01-01

    Full Text Available During the course of their work, a variety of damage and failure of hot forging die occurs and seriously affect the service life. Multi-layer metal hot forging die with functionally graded material structure can effectively extend the service life. In this paper, According to the needs of strengthening forging cavity, the CAD model of surfacing forming center was designed. Based on technological requirements of surface cladding for die cavity, the coupled movement equation of weld torch was established, and the trajectory of welding positioner and Cartesian robot kinematics was solved. The weld torch path was planned, according to the typical methods used in plane welding, and the surfacing path data was extracted by the secondary development of UG/OPEN. Then the kinematics solver program, which can output the control function of motion simulation, was written in MATLAB to solve the kinematics equation. Finally, in UG NX7.5, the kinematics simulation model was built to verify the correctness of mathematical model and the rationality of welding path planning. The above studies can provide a technical support for the die repair and manufacturing of a multilayer metal forging die.

  2. Oxide properties of autoclaved zircaloy cladding tubes investigated by the photoelectric polarization method

    International Nuclear Information System (INIS)

    Corrosion of zirconium alloys is an important lifetime limiting factor for the nuclear reactor fuel. The corrosion resistance of a metal is highly dependent on the ability of the surface metal oxide to transport electrons and ions, which is related to the stoichiometry of the oxide and the oxide defect concentration. The Photoelectric Polarization Method (PEP) is a structure sensitive method which earlier has been investigated as a possible method to study the defect structure in zirconium oxides. The purpose of the following work is, by using more optimized experimental equipment, to verify if the PEP method is a suitable method to study the defect structure in zirconium oxides and to predict the corrosion resistance for different zirconium alloys. The conclusions from the experiments are as follows: - The modifications of the experimental setup by means of a new source of light (deuterium lamp) and a new oscilloscope with an amplifier gave distinct Vpep signals. - The photoresponse is negative for all types of cladding and under all kind of oxidation regimes and hence the oxide is a n-type semiconductor with deficiency of oxygen. - The method needs to be verified by testing semiconductors with a known defect concentration

  3. Studies on stability of U-UO2/PuO2 Cermet fuel for fast reactor and its interaction with T91 cladding

    International Nuclear Information System (INIS)

    Cermet is an advanced fuel concept for fast reactor systems. The fuel consists of ceramic fissile material dispersed in a metallic matrix. The advantages of this fuel include high thermal conductivity, ability to retain fission products etc. The cermet fuel can be either (U, PuO2) or (enriched U, UO2). In the present study chemical compatibility of (U, Nat UO2) cermet fuel with T91 cladding has been carried out by diffusion couple experiment. U-UO2 cermet fuel has been fabricated by a powder metallurgy route. The thermal stability of U-UO2 cermet fuel was studied by heating the fuel at 1223 K for 1000 h. SEM results reveal no change in the microstructure indicating good stability of the fuel at 1223 K. The results of the diffusion couple experiments indicate that Zr liner was effective in preventing fuel-clad chemical interaction. (author)

  4. Alloy waste forms for metal fission products and actinides isolated by spent nuclear fuel treatment

    International Nuclear Information System (INIS)

    Waste form alloys are being developed at Argonne National Laboratory for the disposal of remnant metallic wastes from an electrometallurgical process developed to treat spent nuclear fuel. This metal waste form consists of the fuel cladding (stainless steel or Zircaloy), noble metal fission products (e.g., Ru, Pd, Mo and Tc), and other metallic wastes. The main constituents of the metal waste stream are the cladding hulls (85 to 90 wt%); using the hulls as the dominant alloying component minimizes the overall waste volume as compared to vitrification or metal encapsulation. Two nominal compositions for the waste form are being developed: (1) stainless steel-15 wt% zirconium for stainless steel-clad fuels and (2) zirconium-8 wt% stainless steel for Zircaloy-clad fuels. The noble metal fission products are the primary source of radiation in the metal waste form. However, inclusion of actinides in the metal waste form is being investigated as an option for interim or ultimate storage. Simulated waste form alloys were prepared and analyzed to determine the baseline alloy microstructures and the microstructural distribution of noble metals and actinides. Corrosion tests of the metal waste form alloys indicate that they are highly resistant to corrosion

  5. Numerical and experimental analysis of the residual stress field in cladded components; Numerische und experimentelle Bestimmung des Eigenspannungszustands in plattierten Komponenten

    Energy Technology Data Exchange (ETDEWEB)

    Siegele, Dieter; Brand, Marcus; Hohe, Joerg [Fraunhofer Inst. fuer Werkstoffmechanik, Freiburg (Germany)

    2008-07-01

    The inner surface of a ferritic reactor pressure vessel is protected against corrosion by an austenitic cladding that is usually performed by a double-pass welding in order to avoid under-clad defects und to improve the microstructural properties of the cladding material. On the other hand the different thermal expansion coefficients of the cladding and the base metal induce a complex residual stress field. This has a non-negligible influence on the fracture mechanical assessment of postulated flaws within or under the cladding. The determination of the residual stress field was achieved by numerical simulation of the cladding process. The calibration of the used equivalent heat sources for the modelling of the heat input within the simulation was performed using measured data of the temperature field in a KTA compliant cladding process of test plates made of plant-representative materials. The simulation of the welding process used the temperature dependent material characteristics taking into account the transformation behaviour of the ferritic base metal. The resulting residual stress field shows significant tensile stresses within the cladding with a subsequent compressive stress field under the cladding. The calculated residual stress field is in good agreement with the experimental data. A comparison of the calculated residual stress field using the process simulation with the results of a simplified modelling assuming an increased stress-free temperature in the range of the operation temperature shows also a good agreement, esp. using the materials characteristics of KTA. [German] Die Innenoberflaeche von Reaktordruckbehaeltern aus ferritischen Werkstoffen wird zum Schutz gegen Korrosion mit einer austenitischen Plattierung versehen. Diese wird ueblicherweise als zweilagige Schweissplattierung ausgefuehrt, um die Bildung von Unterplattierungsfehlern zu vermeiden und die mikrostrukturellen Eigenschaften des Plattierungswerkstoffs zu verbessern. Auf der

  6. Improved LWR Cladding Performance by EPD Surface Modification Technique

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael; Sridharan, Kumar

    2012-11-26

    This project will utilize the electro-phoretic deposition technique (EPD) in conjunction with nanofluids to deposit oxide coatings on prototypic zirconium alloy cladding surfaces. After demonstrating that this surface modification is reproducible and robust, the team will subject the modified surface to boiling and corrosion tests to characterize the improved nucleate boiling behavior and superior corrosion performance. The scope of work consists of the following three tasks: The first task will employ the EPD surface modification technique to coat the surface of a prototypic set of zirconium alloy cladding tube materials (e.g. Zircaloy and advanced alloys such as M5) with a micron-thick layer of zirconium oxide nanoparticles. The team will characterize the modified surface for uniformity using optical microscopy and scanning-electron microscopy, and for robustness using standard hardness measurements. After zirconium alloy cladding samples have been prepared and characterized using the EPD technique, the team will begin a set of boiling experiments to measure the heat transfer coefficient and critical heat flux (CHF) limit for each prepared sample and its control sample. This work will provide a relative comparison of the heat transfer performance for each alloy and the surface modification technique employed. As the boiling heat transfer experiments begin, the team will also begin corrosion tests for these zirconium alloy samples using a water corrosion test loop that can mimic light water reactor (LWR) operational environments. They will perform extended corrosion tests on the surface-modified zirconium alloy samples and control samples to examine the robustness of the modified surface, as well as the effect on surface oxidation

  7. A cumulative damage fraction design approach for LMFBR metallic fuel elements

    International Nuclear Information System (INIS)

    The cumulative damage fraction (CDF) analytical technique is currently being used to analyze the performance of metallic fuel elements for proliferation-resistant LMFBRs. In this technique, the fraction of the total time to rupture of the cladding is calculated as a function of the thermal, stress, and neutronic history. Cladding breach or rupture is implied by CDF = 1. Cladding wastage, caused by interactions with both the fuel and sodium coolant, is assumed to uniformly thin the cladding wall. The irradiation experience of the EBR-II Mark-II driver fuel with solution-annealed Type 316 stainless steel cladding provides an excellent data base for testing the applicability of the CDF technique to metallic fuel. Mark-II drivers exhibit negligible fuel-cladding mechanical interaction, so the main source of stress is fission-gas pressure, which has been well characterized. The fuel-cladding chemical interaction zone, which has a uniform front and exhibits Arrhenius time-temperature behavior, is considered as cladding wastage in determining the hoop stress. The Mark-II lifetimes provide a lower bound for comparison with the CDF calculations, since the elements breach in a restrainer dimple and not in the cladding proper. At 5900C, the measured lower bound on lifetime, 10.0 at.%, is lower than the calculated CDF lifetime of 14.5 at.%. The measured lifetime at this temperature might be consistent with the CDF calculations if the cold-worked dimple were not present, since it acts as a stress riser. At 6700C, all elements have exceeded 8 at.% burnup, which is greater than the calculated CDF lifetime of 7 at.%. Thus, CDF calculations are conservative at 6700C. With this in mind, the calculations have been extended to other types of metal fuels. (orig.)

  8. Results of fuel, cladding and reactor pressure vessel computation

    International Nuclear Information System (INIS)

    The cause of the pellet-clad interaction failures are the combined effects of differential thermal expansion driven localized stress, and aggressive fission products, primarily iodine (pitting corrosion). Reactor pressure vessel is also being exposed to high mechanical loads and to an intensive neutron flux irradiation. This leads to a gradual decrease in resistance to brittle fracture. The finite element method has been selected for the solution of the above problems, and results are presented in form of isobars and isotherms respectively in the meridional cross section. (author)

  9. Widely tunable femtosecond solitonic radiation in photonic crystal fiber cladding

    DEFF Research Database (Denmark)

    Peng, J. H.; Sokolov, A. V.; Benabid, F.;

    2010-01-01

    We report on a means to generate tunable ultrashort optical pulses. We demonstrate that dispersive waves generated by solitons within the small-core features of a photonic crystal fiber cladding can be used to obtain femtosecond pulses tunable over an octave-wide spectral range. The generation...... process is highly efficient and occurs at the relatively low laser powers available from a simple Ti:sapphire laser oscillator. The described phenomenon is general and will play an important role in other systems where solitons are known to exist....

  10. Standard specification for architectural flat glass clad polycarbonate

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This specification covers the quality requirements for cut sizes of glass clad polycarbonate (GCP) for use in buildings as security, detention, hurricane/cyclic wind-resistant, and blast and ballistic-resistant glazing applications. 1.2 The values stated in inch-pound units are to be regarded as the standard. The values given in parentheses are for information only. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  11. Screening of advanced cladding materials and UN–U3Si5 fuel

    International Nuclear Information System (INIS)

    Highlights: • Screening methodology for advanced fuel and cladding. • Cladding candidates, except for silicon carbide, exhibit reactivity penalty versus zirconium alloy. • UN–U3Si5 fuels have the potential to exhibit reactor physics and fuel management performance similar to UO2. • Harder spectrum in the UN ceramic composite fuel increases transuranic build-up. • Fuel and cladding properties assumed in these assessments are preliminary. - Abstract: In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO2) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO2 fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO2–Zr fuel–cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN–U3Si5 fuels with Kanthal AF or APMT cladding. The objective of the U3Si5 phase in the UN–U3Si5 fuel concept is to shield the nitride phase from water. It was shown that UN–U3Si5 fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO2–Zr fuel–cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to 14N content in UN ceramic composites is high. Analysis of the rim effect due to self-shielding in the fuel shows that the UN-based ceramic fuels

  12. Direct Metal Deposition of H13 Tool Steel on Copper Alloy Substrate: Parametric Investigation

    Science.gov (United States)

    Imran, M. Khalid; Masood, S. H.; Brandt, Milan

    2015-12-01

    Over the past decade, researchers have demonstrated interest in tribology and prototyping by the laser aided material deposition process. Laser aided direct metal deposition (DMD) enables the formation of a uniform clad by melting the powder to form desired component from metal powder materials. In this research H13 tool steel has been used to clad on a copper alloy substrate using DMD. The effects of laser parameters on the quality of DMD deposited clad have been investigated and acceptable processing parameters have been determined largely through trial-and-error approaches. The relationships between DMD process parameters and the product characteristics such as porosity, micro-cracks and microhardness have been analysed using scanning electron microscope (SEM), image analysis software (ImageJ) and microhardness tester. It has been found that DMD parameters such as laser power, powder mass flow rate, feed rate and focus size have an important role in clad quality and crack formation.

  13. Ceramic facade cladding as an element of sustainable development

    Directory of Open Access Journals (Sweden)

    Topličić-Ćurčić Gordana

    2015-01-01

    Full Text Available Building in harmony with nature has a small impact on the environment, while meeting the basic needs of the population. Green architecture is a branch of architecture including planning, designing and building of various kinds of buildings, with a low impact on the environment. Construction of the so-called “green structures” is in accord with the concept of sustainability and it attempts to balance environmental, economical and social needs. Environmentally appropriate materials are used in construction of this type of structures, which during their production, application and distribution pollute as little as possible the water, soil and air in the environment. The more sustainable the building materials used for construction are, the more sustainable is the structure and its operation with renewable energy sources. The paper considers ceramic facade elements, i.e. cladding. By using ceramic facade cladding, one achieves a better preception of an urban environment, which enriches our lives for new sensual and visual quality, while observing the green building requirements. [Projekat Ministarstva nauke Republike Srbije, br. TR 36017: Utilization of by - products and recycled waste materials in concrete composites in the scope of sustainable construction development in Serbia: investigation and environmental assessment of possible applications

  14. Laser cladding device for nuclear power plant pipeline

    International Nuclear Information System (INIS)

    The device of the present invention concerns a device for applying a laser cladding treatment to a pipe which is welded passing through a bottom of a reactor pressure vessel for preventing stress corrosion crackings. Even if the distance from the axial center of a condensing lens to a coated surface by distortion of the pipe, temperature on the coated surface is made constant and uniform during cladding. That is, equipments shown below are contained in a rotational cylinder vertically movable in the pipe. (1) a condensing lens, (2) an optical fiber cable for entering laser beams to the condensing lens, (3) a reflection mirror for irradiating laser beams collected by the condensing lens to the coated membrane on the inner surface of the pipe. Then, the position at the end of the optical fiber cable is made movable in an axial direction in the rotational cylinder. With such a constitution, the focus position relative to the coated surface is changed by controlling the movement of the end of the optical fiber so as to make the condensing lens closer or apart. Accordingly, intensity of the energy per unit area of the coated surface can be made controlled constant. (I.S.)

  15. Microbial sampling of aluminum-clad spent nuclear fuel

    International Nuclear Information System (INIS)

    A microbial sampling program was initiated at the Idaho National Engineering and Environmental Laboratory (INEEL) to ascertain the effect of microbial activity on the corrosion of aluminum clad spent nuclear fuel (SNF) stored in wet and dry conditions. In the newest fuel storage pool at the INEEL (CPP-666) pitting corrosion has been observed on aluminum corrosion coupons that can not be explained by the excellent water chemistry. Pitting corrosion of the aluminum-clad SNF and corrosion coupons has been observed in the older fuel storage pool (CPP-603). Therefore a microbial assessment of the bulk water, and basin surfaces of both fuel pools was conducted. The results of this microbial enumeration show that a viable and active microbial population does exist in planktonic form. Sampling of aluminum corrosion coupons placed next to stored fuel elements show that microbial attachment has occurred and a biofilm has formed. The sampling program was then extended to the surfaces of wet and dry stored fuel elements. Viable cells or spores were found on the surfaces of the ATR fuel elements that were stored under wet and dry conditions. This paper discusses the methodology of sampling the surfaces of SNF stored under wet conditions for the presence of microorganisms and the types of organisms found

  16. Reactor water chemistry relevant to coolant-cladding interaction

    International Nuclear Information System (INIS)

    The report is a summary of the work performed in a frame of a Coordinated Research Program organized by the IAEA and carried out from 1981 till 1986. It consists of a survey on our knowledge on coolant-cladding interaction: the basic phenomena, the relevant parameters, their control and the modelling techniques implemented for their assessment. Based upon the results of this Coordinated Research Program, the following topics are reviewed on the report: role of water chemistry in reliable operation of nuclear power plants; water chemistry specifications and their control; behaviour of fuel cladding materials; corrosion product behaviour and crud build-up in reactor circuits; modelling of corrosion product behaviour. This report should be of interest to water chemistry supervisors at the power plants, to experts in utility engineering departments, to fuel designers, to R and D institutes active in the field and to the consultants of these organizations. A separate abstract was prepared for each of the 3 papers included in the Annex of this document. Refs, figs, tabs

  17. Cladding and Structural Materials for Advanced Nuclear Energy Systems

    Energy Technology Data Exchange (ETDEWEB)

    Was, G S; Allen, T R; Ila, D; C,; Levi,; Morgan, D; Motta, A; Wang, L; Wirth, B

    2011-06-30

    The goal of this consortium is to address key materials issues in the most promising advanced reactor concepts that have yet to be resolved or that are beyond the existing experience base of dose or burnup. The research program consists of three major thrusts: 1) high-dose radiation stability of advanced fast reactor fuel cladding alloys, 2) irradiation creep at high temperature, and 3) innovative cladding concepts embodying functionally-graded barrier materials. This NERI-Consortium final report represents the collective efforts of a large number of individuals over a period of three and a half years and included 9 PIs, 4 scientists, 3 post-docs and 12 students from the seven participating institutions and 8 partners from 5 national laboratories and 3 industrial institutions (see table). University participants met semi-annually and participants and partners met annually for meetings lasting 2-3 days and designed to disseminate and discuss results, update partners, address outstanding issues and maintain focus and direction toward achieving the objectives of the program. The participants felt that this was a highly successful program to address broader issues that can only be done by the assembly of a range of talent and capabilities at a more substantial funding level than the traditional NERI or NEUP grant. As evidence of the success, this group, collectively, has published 20 articles in archival journals and made 57 presentations at international conferences on the results of this consortium.

  18. Sol–gel coassembly of macroporous cylinders cladded optical fibers

    International Nuclear Information System (INIS)

    Highlights: • We develop a novel sol–gel cooperatively assembly method. • 3D microstructure fibers can be fabricated in a single step by sol–gel co-assembly. • Cylindrical inverse opals cladded optical fibers show face-centered cubic structure. • SEM images of transverse and longitudinal cross section are observed. • Transmission spectra show deep photonic band gaps up to 50%. - Abstract: In this paper, we provide a facile way to fabricate a microstructure fiber by coating a standard optical fiber with a silica inverse opal through a sol–gel coassembly method. Polystyrene (PS) colloidal suspension of microspheres and a hydrolyzed silicate precursor were added to the solvent together. With the evaporation of the solvent, the assembly of a PS colloidal template and the infiltration of voids of the spheres with silica gel were executed simultaneously to form a colloidal composite in a single step. After removal of the sacrificial colloidal template, a cylindrical inverse opal (CIO) cladded an optical fiber was obtained. Structural properties characterized by optical and scanning electron microscopy (SEM) and unique transmission spectra with photonic band gaps reveal the high quality of the silica CIOs, which can be used as fiber Bragg grating for optical communications

  19. Nuclear fuel cladding tube and method of manufacturing the same

    International Nuclear Information System (INIS)

    A cladding tube main body made of a zirconium alloy and an end plug are joined by welding. Tensile stresses at the weld heat-affected portion between the cladding tube main body and the end plug are removed, so that compression stresses of 0 MPa or more but less than the endurance strength of the zirconium alloy is applied on the weld heat affected portion. As the zirconium alloy, a zircaloy-2 or zircaloy-4 is preferable since it is excellent in the corrosion resistance and strength. The zirconium alloy may preferably be used also to the material of the end plug. The treatment for the removal of the tensile stresses includes a method of applying annealing to the weld heat-affected portion or a method of applying compression stresses thereto by applying external force such as a shot peening treatment. This can suppress occurrence of nodular corrosion and white homogeneous corrosion caused in the vicinity of the welded portion. (I.N.)

  20. Irradiation effect on fatigue behaviour of zircaloy-4 cladding tubes

    International Nuclear Information System (INIS)

    Since nuclear electricity has a predominant share in French generating capacity, PWR's are required to fit grid load following and frequency control operating conditions. Consequently cyclic stresses appear in the fuel element cladding. In order to characterize the possible resulting clad damage, fatigue tests were performed at 350 deg C on unirradiated material or irradiated stress relieved Zircaloy-4 tube portions, using a special device for tube fatigue by repeated pressurization. It appears that, for high stress levels, the material fatigue life is not affected by irradiation. But the endurance fatigue limit undergoes a decrease from the 350 MPa value for unirradiated material to the 210 MPa value for the material irradiated for four cycles in a PWR. However, this effect seems to saturate with irradiation dose: no difference could be detected between the two cycles results and the corresponding four cycles results. The corrosion effect and the load following influence were also investigated: they do not appear to modify the fatigue behaviour in our experimental conditions

  1. Cladding Attachment Over Thick Exterior Insulating Sheathing (Fact Sheet)

    Energy Technology Data Exchange (ETDEWEB)

    2013-11-01

    The addition of insulation to the exterior of buildings is an effective means of increasing the thermal resistance of wood-framed walls and mass masonry wall assemblies. The location of the insulation on the exterior of the structure has many direct benefits, including better effective R-value from reduced thermal bridging, better condensation resistance, reduced thermal stress on the structure, as well as other commonly associated improvements such as increased airtightness and improved water management. For thick layers of exterior insulation (more than 1.5 in.), the use of wood furring strips attached through the insulation back to the structure has been used by many contractors and designers as a means to provide a convenient cladding attachment location. Although the approach has proven effective, there is significant resistance to its widespread implementation due to a lack of research and understanding of the mechanisms involved in the development of the vertical displacement resistance capacity. In addition, the long-term in-service performance of the system has been questioned due to potential creep effects of the assembly under the sustained dead load of the cladding and effects of varying environmental conditions. In addition, the current International Building Code (IBC) and International Residential Code (IRC) do not have a provision that specifically allows this assembly.

  2. Fracture properties of hydrided Zircaloy-4 cladding in recrystallization and stress-relief anneal conditions

    International Nuclear Information System (INIS)

    In this work, the stress-relieved (SRA) and recrystallized (RXA) Zircaloy-4 cladding specimens were hydrogen-charged to the target concentration of 300 wppm and then manufactured into X-specimens for fracture toughness test. The hydrogen embrittlement susceptibility of Zircaloy-4 cladding specimens in both SRA and RXA conditions were investigated. At the hydrogen concentration level of 300 wppm, J-integral values for RXA cladding were higher than those for SRA cladding at both 25 °C and 300 °C. The formation of brittle zirconium hydrides had a significant impact on the fracture toughness of Zircaloy-4 cladding in both SRA and RXA states, especially at 25 °C. Among all the tests, SRA cladding tested at 25 °C exhibited a great loss of the fracture toughness. The micrographic and fractographic observations further demonstrated that the fracture toughness of Zircaloy-4 cladding would be improved by the coarse grains in RXA cladding, but degraded by zirconium hydrides precipitated along the grain boundary.

  3. Mechanical Properties of Cladding and Wrapper Materials for the ASTRID Fast-Reactor Project

    International Nuclear Information System (INIS)

    Requirements for the cladding and wrapper in ASTRID: – AIM1 cladding material: • AIM1 swelling behavior: the Oliphant 1bis irradiation; • Mechanical properties on fresh material. – EM10 wrapper material: • mechanical properties on irradiated material: BOITIX9; • high temperature mechanical behaviour. Ongoing research on AIM1 and EM10

  4. Ukrainian WWER-type NPP units. Methodological basement, results of cladding tightness inspection

    International Nuclear Information System (INIS)

    In the overview report the generalized results of cladding tightness inspection are reviewed for all Ukrainian WWER-type NPP units. Brief analysis of cladding tightness inspection methodology is drawn. Approaches of Ukrainian NPPs are generalized from the viewpoint of use of widened inspection sample analysis. (author)

  5. Mechanical interaction between fuel and cladding during normal transient operating conditions

    International Nuclear Information System (INIS)

    In order to avoid pin failure, it is necessary to be sure that the operating conditions will not induce in the cladding some damaging form of deformation. Stresses and strains taking place in the cladding have different origins. 1. The cladding deformation due to fission gas pressure occurs by irradiation creep and this is not a damaging strain. 2. The thermal stresses induced in the cladding at the beginning of life by the thermal gradient relax rapidly by thermal and irradiation creep. The resulting damage is negligible. 3. Stresses induced by the swelling gradient in the thickness of the cladding probably relax by irradiation creep. But we cannot be sure that the swelling laws deduced from axial profile still apply in the cladding thickness. In particular the influence of stress on swelling is not yet well known. 4. Analysis of fuel-cladding mechanical interaction (FCHI) shows that during steady-state operation it is very unlikely a damaging form of strain due to FCHI to take place in the cladding. It is only during non-steady state operation at power increase that such a strain can occur. Therefore two types of operating conditions may be feared the increase to full power after continued operation at reduced power and the load follow power cycling

  6. Raman probes based on optically-poled double-clad fiber and coupler

    DEFF Research Database (Denmark)

    Brunetti, Anna Chiara; Margulis, Walter; Rottwitt, Karsten

    2012-01-01

    sample of dimethyl sulfoxide (DMSO), when illuminating the waveguide with 1064nm laser light. The Raman signal is collected in the inner cladding, from which it is retrieved with either a bulk dichroic mirror or a double-clad fiber coupler. The coupler allows for a substantial reduction of the fiber...

  7. Solution recommended by the CEA to minimize fuel-cladding interaction problems

    International Nuclear Information System (INIS)

    To improve behavior of fuel elements under fast variation of power, the CEA developed a fracture resistant fuel: UO2 DCI and a cladding with low creep under irradiation: strong zircaloy 4. This paper gives main characteristics of these new materials and shows at what extent a better behavior of fuel pins is obtained concerning fuel-cladding interaction

  8. High performance fuel technology development : Development of high performance cladding materials

    International Nuclear Information System (INIS)

    The superior in-pile performance of the HANA claddings have been verified by the successful irradiation test and in the Halden research reactor up to the high burn-up of 67GWD/MTU. The in-pile corrosion and creep resistances of HANA claddings were improved by 40% and 50%, respectively, over Zircaloy-4. HANA claddings have been also irradiated in the commercial reactor up to 2 reactor cycles, showing the corrosion resistance 40% better than that of ZIRLO in the same fuel assembly. Long-term out-of-pile performance tests for the candidates of the next generation cladding materials have produced the highly reliable test results. The final candidate alloys were selected and they showed the corrosion resistance 50% better than the foreign advanced claddings, which is beyond the original target. The LOCA-related properties were also improved by 20% over the foreign advanced claddings. In order to establish the optimal manufacturing process for the inner and outer claddings of the dual-cooled fuel, 18 different kinds of specimens were fabricated with various cold working and annealing conditions. Based on the performance tests and various out-of-pile test results obtained from the specimens, the optimal manufacturing process was established for the inner and outer cladding tubes of the dual-cooled fuel

  9. Study on transient temperature measurement at fuel clad surface in NSRR experiments

    International Nuclear Information System (INIS)

    In NSRR experiments, evolution of fuel clad temperature is measured by thermocouples welded on the clad surface. This report describes the studies performed with the CASTEM code in order to evaluate the measurement error, that is, temperature difference between the thermocouple welding spot and the clad surface far from the spot. The studies show that the welded thermocouple slightly underestimates the clad surface temperature when the clad zirconia thickness is below 30 μm, and slightly overestimates it for thicker zirconia layer. Furthermore, two distinct phases have been identified in all cases. A transient capacitive phase occurs up to 0.1-0.3s while clad temperature remains below 400degC. The temperature error reaches the maximum in this phase; -100degC (underestimation) without zirconia layer and +150degC (overestimation) with 100 μm zirconia layer. A 'fin effect' phase starts when the clad temperature exceeds 400degC and the film boiling regime is clearly established, during which the error stabilizes between -20degC without zirconia layer and +50degC with 100 μm zirconia layer. The influence of the thermocouple is limited to its very vicinity (radius of about 0.5 mm). A transfer function was determined from the calculation results in order to estimate the accurate clad outer temperature from the thermocouple recording. (author)

  10. An Examination of Collaborative Learning Assessment through Dialogue (CLAD) in Traditional and Hybrid Human Development Courses

    Science.gov (United States)

    McCarthy, Wanda C.; Green, Peter J.; Fitch, Trey

    2010-01-01

    This investigation assessed the effectiveness of using Collaborative Learning Assessment through Dialogue (CLAD) (Fitch & Hulgin, 2007) with students in undergraduate human development courses. The key parts of CLAD are student collaboration, active learning, and altering the role of the instructor to a guide who enhances learning opportunities.…

  11. Ballooning and rupture behavior of Zircaloy-4 cladding under transient-heating conditions

    International Nuclear Information System (INIS)

    Phenomena of fuel fragmentation, relocation and dispersal (FFRD) have been observed in several experiments on very-high-burnup fuels under simulated loss-of-coolant-accident (LOCA) conditions using a test reactor. In order to improve the prediction of the phenomena, ballooning and rupture behaviors of cladding under simulated LOCA conditions were investigated by performing laboratory-scale experiments in which internally pressurized non-irradiated Zircaloy-4 (Zry-4) claddings were heated to burst. The maximum circumferential strains of the ballooned claddings were strongly dependent on burst temperature and the trends seemed to depend on the heating rate in the experiment. Values of the maximum circumferential strain were normalized by dividing them by engineering hoop stress at the time of rupture. A correlation between the normalized values and the burst temperatures suggests that the fraction of β phase in Zry-4 cladding affects the extent of the strain of cladding ballooning and the embrittlement of cladding due to oxidation suppresses the ballooning of cladding. The length and width of rupture opening had the peak at ∼1073 K and decreased as the temperature increased from ∼1073 K in the case of the low heating rate while no specific trends were observed in the case of the high heating rate. These results suggest that the embrittlement of cladding due to oxidation affects the rupture behavior and results in small rupture openings. (author)

  12. Simulation and Analysis on Hoop Strength Test of Multilayered SiC Composite Fuel Cladding

    International Nuclear Information System (INIS)

    Silicon carbide-based ceramics and their composites have been studied for fusion and advanced fission energy systems. For fission reactors, SiCf/SiC composites can be applied to core structural materials. Multi-layered SiC composite fuel cladding, which consists of monolith inner/outer layer and intermediate SiCf/SiC composite layer is one of candidates for a replacement for the zirconium alloy cladding, owing to the superior high temperature strength and low hydrogen generation under severe accident conditions. The SiC composite cladding has to retain the mechanical properties and its structure from the inner pressure caused by fission products to apply a cladding of fission reactor. The inner pressure caused by fission products induces hoop stress in a circumferential direction. Hoop strength test using expandable polyurethane plug is designed for evaluating the mechanical properties of fuel cladding. In this paper, hoop strength test of the multilayered cladding was simulated in order to evaluate hoop stress and shear stress at the cladding and the fracture of the cladding was analyzed

  13. Cladding modes in photonic crystal fiber: characteristics and sensitivity to surrounding refractive index

    Science.gov (United States)

    Jiang, Xiuli; Gu, Zhengtian; Zheng, Li

    2016-01-01

    Characteristics of cladding modes in a photonic crystal fiber (PCF) with triangular air-hole lattice in the cladding are numerically analyzed using a finite element method. The transition for LP11 cladding mode to core mode with variation of the normalized wavelength has been shown. The transition of the LP01 cladding mode to the outer silica mode and reorganization of the LP0m cladding modes caused by varying the fiber radius has been investigated. By choosing the optimized fiber radius, which is located in the cladding modes' reorganization region, the sensitivity of the coupled wavelength between the core mode LP01 and cladding mode LP03 to surrounding refractive index is increased by a factor of five and reaches to 2660 nm/refractive index unit over the range of 1.40 to 1.42. The sensitivity is competitive with that of long-period grating in PCF in response to changes in refractive indices of the medium contained in the cladding air channels.

  14. Cladding flaw detection and sizing by horizontally polarized shear wave ultrasonic EMAT transducers

    International Nuclear Information System (INIS)

    Further experimental work was done within the framework of the current research contract on the employment of the EMAT technique in the ultrasonic inspection of reactor vessel cladding. This year's activity focused on the study of the space distribution of the ultrasonic beam generated by the flexible transducers developed during the course of the previous year, and on the inspection of the cladding of the JRC-ISPRA PWR vessel 1:5 scale model. Two transducer pairs were used to make laboratory measurements on the clad and unclad test block sides for the purpose of studying ultrasonic beam distribution. It emerged that the cladding tended to confine the beam. If however the wavelength was equal to or greater than the cladding thickness the confinement was not complete and became less and less evident with increasing wavelength. It was consequently possible to pick up echoes produced by flaws located both within the cladding and in the underlaying layers. The PWR vessel model cladding was then inspected in the neighbourhood of the welds and a large number of flaws was found. The EMAT technique has proved to be suitable for the detection and rough location of flaws but less so for their sizing, although in some cases it was possible to assess the distance between the flaw and the cladding top

  15. Oxidation resistant chromium coating on Zircaloy-4 for accident tolerant fuel cladding

    International Nuclear Information System (INIS)

    The attributes of such a fuel are approved reaction kinetics with steam, a slower hydrogen generation rate, and good cladding thermo-mechanical properties. Many researchers have tried to modify zirconium alloys to improve their oxidation resistance in the early stages of the ATF development. Corrosion resistant coating on cladding is one of the candidate technologies to improve the oxidation resistance of zirconium cladding. By applying coating technology to zirconium cladding, it is easy to obtain corrosion resistance without a change in the base materials. Among the surface coating methods, arc ion plating (AIP) is a coating technology to improve the adhesion owing to good throwing power, and a dense deposit (Fig. 1). Owing to these advantages, AIP has been widely used to efficiently form protective coatings on cutting tools, dies, bearings, etc. In this study, The AIP technique for the protection of zirconium claddings from the oxidation in a high-temperature steam environment was studied. The homogeneous Cr film with a high adhesive ability to the cladding was deposited by AIP and acted as a protection layer to enhance the corrosion resistance of the zirconium cladding. It was concluded that the AIP technology is effective for coating a protective layer on claddings

  16. Yb-Er co-doped phosphate fiber with hexagonal inner cladding

    Science.gov (United States)

    Wen, Lei; Wang, Longfei; He, Dongbing; Chen, Danping; Hu, Lili

    2016-04-01

    An Yb-Er co-doped phosphate glass double-clad fiber with hexagonal inner cladding was fabricated by stack-and-draw method. Output power of 4.9 W was extracted with slope efficiency of 30 % from the fiber with 55 cm in length.

  17. Method of joining nuclear fuel rod end caps and nuclear fuel rod cladding tubes

    International Nuclear Information System (INIS)

    A method of joining fuel rod end caps and cladding tubes by resistance pressure welding within a welding chamber is described. A welding device is brought into engagement with an end portion of a rigidly mounted cladding tube. An opening chuck as well as a divided welding electrode, both of which are mounted at one side of the welding chamber, are shifted along a predetermined length of the cladding tube end portion. The chuck and the divided welding electrode are brought into contact with the cladding tube end portion. Another welding electrode carrying an end cap is thrust into the welding chamber from the other side thereof so that the end cap is fed to the open end of the cladding tube end portion. The welding chamber is sealed by sealing members sealingly engaging the cladding tube end portion and the other welding electrode and then the interior of the welding chamber is evacuated and filled with protective gas. The end cap is pressed onto the open end of the cladding tube end portion. A welding current is passed through the welding electrodes so as to weld the end cap to the end of the cladding tube end portion

  18. Irradiation Performance of Oxide Dispersion Strengthened (ODS) Ferritic Steel Claddings for Fast Reactor Fuels

    International Nuclear Information System (INIS)

    Irradiation tests in Joyo and BOR-60 for the ODS claddings developed by JAEA were carried out in order to confirm the irradiation performance of the ODS claddings and thus judge their applicability to high burnup and high temperature fast reactor fuels. The main points of the tests are summarized as follows. 1) Valuable data indicating application prospects of the ODS claddings for high burnup fuels were obtained regarding superior dimensional stability and integrity of the upper end-plug welded by the PRW method. 2) No significant irradiation effect on mechanical properties of the ODS claddings was confirmed within the irradiation conditions in the Joyo material irradiation tests. The oxide particles and microstructures of ODS claddings were confirmed to be stable during neutron irradiation. 3) FCCI data for the ODS claddings were acquired within the irradiation conditions in the BOR-60 fuel pin irradiation tests, and it was shown that FCCI could be reduced by lowering oxygen potential in the fuel, even for low Cr content claddings such as 9Cr-ODS steel. 4) The manufacturing technology development applied to the full pre-alloy process to improve homogeneity of the ODS cladding has already started, and the expected results are being obtained

  19. Sertification of fuel cladding and grids materials in out of pile conditions

    International Nuclear Information System (INIS)

    The basic standard specifications for fuel rod cladding and bundle materials, are selected. In this paper the standard specifications of material for Zircaloy and plugs and stainless steel springs of fuel rod cladding are presented. The material specification for a Zircaloy fuel bundle assembly Cgrids) is also given. (author)

  20. Advanced oxidation-resistant iron-based alloys for LWR fuel cladding

    Science.gov (United States)

    Terrani, K. A.; Zinkle, S. J.; Snead, L. L.

    2014-05-01

    Application of advanced oxidation-resistant iron alloys as light water reactor fuel cladding is proposed. The motivations are based on specific limitations associated with zirconium alloys, currently used as fuel cladding, under design-basis and beyond-design-basis accident scenarios. Using a simplified methodology, gains in safety margins under severe accidents upon transition to advanced oxidation-resistant iron alloys as fuel cladding are showcased. Oxidation behavior, mechanical properties, and irradiation effects of advanced iron alloys are briefly reviewed and compared to zirconium alloys as well as historic austenitic stainless steel cladding materials. Neutronic characteristics of iron-alloy-clad fuel bundles are determined and fed into a simple economic model to estimate the impact on nuclear electricity production cost. Prior experience with steel cladding is combined with the current understanding of the mechanical properties and irradiation behavior of advanced iron alloys to identify a combination of cladding thickness reduction and fuel enrichment increase (∼0.5%) as an efficient route to offset any penalties in cycle length, due to higher neutron absorption in the iron alloy cladding, with modest impact on the economics.

  1. Analysis of the stress raising action of flaws in laser clad deposits

    International Nuclear Information System (INIS)

    Highlights: ► Laser clad defects are 0D-pores/inclusions, 1D-clad waviness or 2D-planar defects. ► Surface pore of laser clad bar initiates fatigue cracks. ► Side edge surface pores are more critical than in-clad surface pores. ► Smaller notch radius and angle of as-laser clad surface raises stress significantly. ► Planar inner defects grow faster towards surface. - Abstract: Fatigue cracking of laser clad cylindrical and square section bars depends upon a variety of factors. This paper presents Finite Element Analysis (FEA) of the different macro stress fields generated as well as stress raisers created by laser cladding defects for four different fatigue load conditions. As important as the defect types are their locations and orientations, categorized into zero-, one- and two-dimensional defects. Pores and inclusions become critical close to surfaces. The performance of as-clad surfaces can be governed by the sharpness of surface notches and planar defects like hot cracks or lack-of-fusion (LOF) are most critical if oriented vertically, transverse to the bar axis. The combination of the macro stress field with the defect type and its position and orientation determines whether it is the most critical stress raiser. Based on calculated cases, quantitative and qualitative charts were developed as guidelines to visualize the trends of different combinations

  2. Hydrogen permeation in FeCrAl alloys for LWR cladding application

    Science.gov (United States)

    Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; Snead, Lance L.

    2015-06-01

    FeCrAl, an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In this study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. The total tritium inventory inside the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.

  3. Influence of oxide layers on the cladding material in a liquid lead environment: a comparison between MATRA and TRACE

    International Nuclear Information System (INIS)

    The liquid lead-cooled fast reactor (LFR) is one design proposal for further investigations in the frame of the Generation IV project. Due to the high melting point of lead (327 deg. C), the temperature of operation is in the range of 400 to 500 deg. C. This requires a cladding material, which sustains high temperatures over long time periods. The proposed material will be the ferritic-martensitic steel T-91. One drawback of working in a liquid lead environment is the high solubility of metals in it. A protective oxide layer is one option to face the corrosion issue but has the disadvantage of a low thermal conductivity, which increases the cladding temperature. The layout of the fuel assembly (FA) of the LFR is determined to be either an open quadratic array or a hexagonal design with a wrapper. Both designs influence the thermohydraulic of the core since the heat transfer depends on the geometry of the cooling channel. Investigations of the influence of the oxide layer and of the impact of the heat transfer were done for an open quadratic FA. For the assessment of the above mentioned challenges, the subchannel code MATRA and the system code TRACE were extended and applied. Corresponding results are here compared. (authors)

  4. Influence of oxide layers on the cladding material in a liquid lead environment: a comparison between MATRA and TRACE

    Energy Technology Data Exchange (ETDEWEB)

    Jaeger, Wadim; Sanchez Espinoza, Victor H. [Forschungszentrum Karlsruhe GmbH, Institute for Reactor Safety, Eggenstein-Leopoldshafe (Germany)

    2008-07-01

    The liquid lead-cooled fast reactor (LFR) is one design proposal for further investigations in the frame of the Generation IV project. Due to the high melting point of lead (327 deg. C), the temperature of operation is in the range of 400 to 500 deg. C. This requires a cladding material, which sustains high temperatures over long time periods. The proposed material will be the ferritic-martensitic steel T-91. One drawback of working in a liquid lead environment is the high solubility of metals in it. A protective oxide layer is one option to face the corrosion issue but has the disadvantage of a low thermal conductivity, which increases the cladding temperature. The layout of the fuel assembly (FA) of the LFR is determined to be either an open quadratic array or a hexagonal design with a wrapper. Both designs influence the thermohydraulic of the core since the heat transfer depends on the geometry of the cooling channel. Investigations of the influence of the oxide layer and of the impact of the heat transfer were done for an open quadratic FA. For the assessment of the above mentioned challenges, the subchannel code MATRA and the system code TRACE were extended and applied. Corresponding results are here compared. (authors)

  5. Iodine induced stress corrosion of zirconium and zircaloy-4 mechanisms. Translation to pellet cladding interaction conditions in PWR type reactors

    International Nuclear Information System (INIS)

    This thesis is linked to the study of the Pellet-Cladding Interaction (PCI) phenomenon in Pressurized Water Reactors, which can lead to cladding rupture by iodine Stress Corrosion Cracking (SCC) of Zircaloy-4. Results are obtained through slow tensile tests performed in iodine methanol and in iodine vapour, on reference material, neutron irradiated material, and iodine zirconium implanted material. They allow to propose an explanation of the rapidity of the ruptures observed during PCI loadings, and to make the link between laboratory SCC tests and power ramp tests. Indeed, neutron irradiation facilitates the initiation and the transgranular propagation steps of the SCC cracks, due to strain localization and hardening associated to the presence of irradiation defects. On the other hand, recoiled iodine does not seem to affect SCC susceptibility of the material. On a chemical point of view, thermally released iodine is then responsible for PCI/SCC ruptures. A detailed calculation of iodine amount created by fission and released in the gap during irradiation makes it possible to show that local iodine concentration facing the pellet-to-pellet and the radial pellet cracks regions is sufficient for SCC cracks to develop in the metal. Finally, a competition between re-passivation and cracking is underscored. This results are in good agreement with the occurrence of an iodine adsorption mechanism. Adsorption could be assisted by a corrosion-deformation interaction phenomenon, and/or by the formation of solid and gaseous zirconium iodides. (author)

  6. Mechanisms of damage to the oxide layer of cladding of fuel rods under accident conditions like RI

    International Nuclear Information System (INIS)

    During reactivity initiated accident, the importance of cladding tube oxidation on its thermomechanical behavior has been investigated. After RIA tests in experimental reactors oxide damage including radial cracking and spallation of the outer oxide layer has been evidenced. This work aims at better understanding the key mechanisms controlling these phenomena. Laboratory air-oxidation of Zircaloy-4 cladding tubes has been performed at 470 C. SEM micrographs show that radial cracks are initiated from the outer surface of the oxide layer and propagated radially towards the oxide-metal interface. A model predicting the stress evolution within the oxide and the depth of crack has been developed and validated on literature tests and tests of this study. Ring compression tests were used for the experimental study of the oxide degradation under mechanical loading. Experimental data revealed three mechanisms: densification of the radial crack network, propagation of these radial cracks, branching and spallation of oxide fragments. The influence of the circumferential cracks, periodically distributed in the oxide layer, on the stress distribution in oxide fragments has been analysed using finite element modelling. The determining influence of these cracks on the maximum stress oxide fragments has been demonstrated. (author)

  7. Study of oxide and α-Zr(O) growth kinetics from high temperature steam oxidation of Zircaloy-4 cladding

    Science.gov (United States)

    Sawarn, Tapan K.; Banerjee, Suparna; Samanta, Akanksha; Rath, B. N.; Kumar, Sunil

    2015-12-01

    Oxidation kinetics of Zircaloy-4 cladding of fuel pins of Indian pressurized heavy water reactors (IPHWRs) under a simulated loss of coolant accident (LOCA) condition was investigated. The kinetic rate constants for the oxide and oxygen stabilized α-Zr phase growth were established from the isothermal metal-steam reaction at high temperatures (900-1200 °C) with soaking periods in the range of 60-900 s. Oxide and α-Zr(O) layer thickness were measured to derive the respective growth rates. The observed rates obeyed a parabolic law and Arrhenius expressions of rate constants were established. Percentage equivalent clad reacted (%ECR) was calculated using Baker-Just equation. Hydrogen estimation was carried out on the oxidized samples using inert gas fusion technique. The hydrogen pick up was found to be in the range 10-30 ppm. The measured values of oxide and α-Zr(O) layer thickness were compared with the results obtained using OXYCON, an indigenously developed model. The model predicts the oxide growth reasonably well but under predicts the α-Zr(O) growth significantly at thickness values higher than 80 μm.

  8. Microstructure evolution and mechanical properties of multiple-layer laser cladding coating of 308L stainless steel

    International Nuclear Information System (INIS)

    Highlights: • Grain morphology transformations of 308L stainless steel multiple-layer are studied. • The cladding metals solidify in AF mode and consist of austenite and about 10.48% δ ferrite. • The ferrite content distributes into an increasing trend as the number of layers increase. • The distribution of hardness from the substrate to the coating is relatively uniform. • The cladding tensile sample shows good tensile properties, and the fracture mode is the ductile fracture. - Abstract: Multiple-layer laser cladding of 308L stainless steel was obtained by a fiber laser using a way of wire feeding to repair the surface scrapped or erosive parts of 316L stainless steel. The microstructure of the coating was measured by a metallographic microscope, and phase composition was determined by X-ray diffraction. The results show that good metallurgical bonding can be obtained between the 308L stainless steel coating and 316L stainless steel substrate. The coating is mainly composed of columnar dendrites, and there are also a few planar crystals and cellular dendrites distributed in the bonding zone. Meanwhile, some equiaxed grains and steering dendrites are distributed in the apex of the coating. Grains incorporate in epitaxial columnar dendrite's growth between different layers and tracks. It has been proved using XRD that the coating basically consists of austenite and a small amount of δ ferrite. The coating solidifies in FA mode according to the Creq/Nieq ratio and metallurgical analysis results. The average content of δ ferrite is about 10.48% and morphologies of the ferrite are mostly vermicular, skeletal and lathy. Due to heat treatment and different cooling rate, the δ ferrite content generally increases as the number of laser cladding layers increases. The coating and the substrate have equivalent microhardness, and softening zone does not appear in the heat affected zone. The tensile strength and elongation of the coating are 548 MPa and 40

  9. Microstructure evolution and mechanical properties of multiple-layer laser cladding coating of 308L stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Li, Kaibin; Li, Dong, E-mail: lid@sues.edu.cn; Liu, Dongyu; Pei, Guangyu; Sun, Lei

    2015-06-15

    Highlights: • Grain morphology transformations of 308L stainless steel multiple-layer are studied. • The cladding metals solidify in AF mode and consist of austenite and about 10.48% δ ferrite. • The ferrite content distributes into an increasing trend as the number of layers increase. • The distribution of hardness from the substrate to the coating is relatively uniform. • The cladding tensile sample shows good tensile properties, and the fracture mode is the ductile fracture. - Abstract: Multiple-layer laser cladding of 308L stainless steel was obtained by a fiber laser using a way of wire feeding to repair the surface scrapped or erosive parts of 316L stainless steel. The microstructure of the coating was measured by a metallographic microscope, and phase composition was determined by X-ray diffraction. The results show that good metallurgical bonding can be obtained between the 308L stainless steel coating and 316L stainless steel substrate. The coating is mainly composed of columnar dendrites, and there are also a few planar crystals and cellular dendrites distributed in the bonding zone. Meanwhile, some equiaxed grains and steering dendrites are distributed in the apex of the coating. Grains incorporate in epitaxial columnar dendrite's growth between different layers and tracks. It has been proved using XRD that the coating basically consists of austenite and a small amount of δ ferrite. The coating solidifies in FA mode according to the Creq/Nieq ratio and metallurgical analysis results. The average content of δ ferrite is about 10.48% and morphologies of the ferrite are mostly vermicular, skeletal and lathy. Due to heat treatment and different cooling rate, the δ ferrite content generally increases as the number of laser cladding layers increases. The coating and the substrate have equivalent microhardness, and softening zone does not appear in the heat affected zone. The tensile strength and elongation of the coating are 548 MPa and 40

  10. Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors

    International Nuclear Information System (INIS)

    Highlights: • Replaced zirconium-alloy with alternate cladding concepts: 310SS, FeCrAl, 304SS, APMT and SiC. • Performed parametric study to match reactivity lifetime requirements of Zircaloy base case. • Analyzed reactivity coefficients, spectral hardening, fission power profiles and Pu inventory. • Assessed fuel cost changes when replacing Zircaloy cladding. - Abstract: A study analyzed the neutronics of alternate cladding materials in a pressurized water reactor (PWR) environment. Austenitic type 310 (310SS) and 304 stainless steels, ferritic Fe-20Cr-5Al (FeCrAl) and APMT™ alloys, and silicon carbide (SiC)-based materials were considered and compared with Zircaloy-4. SCALE 6.1 was used to analyze the associated neutronics penalty/advantage, changes in reactivity coefficients, and spectral variations once a transition in the cladding was made. In the cases examined, materials containing higher absorbing isotopes invoked a reduction in reactivity due to an increase in neutron absorption in the cladding. Higher absorbing materials produced a harder neutron spectrum in the fuel pellet, leading to a slight increase in plutonium production. A parametric study determined the geometric conditions required to match cycle length requirements for each alternate cladding material in a PWR. A method for estimating the end of cycle reactivity was implemented to compare each model to that of standard Zircaloy-4 cladding. By using a thinner cladding of 350 μm and keeping a constant outer diameter, austenitic stainless steels require an increase of no more than 0.5 wt% enriched 235U to match fuel cycle requirements, while the required increase for FeCrAl was about 0.1%. When modeling SiC (with slightly lower thermal absorption properties than that of Zircaloy), a standard cladding thickness could be implemented with marginally less enriched uranium (∼0.1%). Moderator temperature and void coefficients were calculated throughout the depletion cycle. Nearly identical

  11. Characterization of hydrogen, nitrogen, oxygen, carbon and sulfur in nuclear fuel (UO2) and cladding nuclear rod materials

    International Nuclear Information System (INIS)

    The importance of Hydrogen, Nitrogen, Oxygen, Carbon and Sulfur gases analysis in nuclear fuels such as UO2, U3O8, U3Si2 and in the fuel cladding such as Zircaloy, is a well known as a quality control in nuclear industry. In UO2 pellets, the Hydrogen molecule fragilizes the metal lattice causing the material cracking. In Zircaloy material the H2 molecules cause the boiling of the cladding. Other gases like Nitrogen, Oxygen, Carbon and Sulfur affect in the lattice structure change. In this way these chemical compounds have to be measure within specify parameters, these measurement are part of the quality control of the nuclear industry. The analytical procedure has to be well established by a convention of the quality assurance. Therefore, the Oxygen, Carbon, Sulfur and Hydrogen are measured by infrared absorption (IR) and the nitrogen will be measured by thermal conductivity (TC). The gas/metal analyzer made by LECO Co. model TCHEN-600 is Hydrogen, Oxygen and Nitrogen analyzer in a variety of metals, refractory and other inorganic materials, using the principle of fusion by inert gas, infrared and thermo-coupled detector. The Carbon and Sulfur compounds are measure by LECO Co. model CS-400. A sample is first weighed and placed in a high purity graphite crucible and is casted on a stream of helium gas, enough to release the oxygen, nitrogen and hydrogen. During the fusion, the oxygen present in the sample combines with the carbon crucible to form carbon monoxide. Then, the nitrogen present in the sample is analyzed and released as molecular nitrogen and the hydrogen is released as gas. The hydrogen gas is measured by infrared absorption, and the sample gases pass through a trap of copper oxide which converts CO to CO2 and hydrogen into water. The gases enter the cell where infrared water content is then converted making the measurement of total hydrogen present in the sample. The Hydrogen detection limits for the nuclear fuel is 1 μg/g for the Nitrogen and Oxygen

  12. Microstructure and Tribological Properties of In Situ Synthesized TiN Reinforced Ni/Ti Alloy Clad Layer Prepared by Plasma Cladding Technique

    Science.gov (United States)

    Jin, Guo; Li, Yang; Cui, Huawei; Cui, Xiufang; Cai, Zhaobing

    2016-06-01

    A Ni/Ti composite coating enhanced by an in situ synthesized TiN phase was fabricated on FV520B steel by plasma cladding technology. The in situ formation of the TiN phase was confirmed by XRD, SEM, and TEM. The cladding layer consisted of three regions on going from the top to the bottom, namely, columnar grain regions, columnar dendritic regions, and fine grain regions. The cladding layer was composed of Ni3Ti, TiN, (Fe, Ni), and Ti phases. The dendritic and columnar regions were mainly composed of the Ni3Ti and (Fe, Ni) phases. The Ti phase was observed at the branches of dendrite crystals and columnar grains. The volume fraction of the TiN phase in the cladding layer was about 3.2%. The maximum micro-hardness value of the in situ formed coating (760 HV0.2) was higher than that of the pure coating (537 HV0.2). The cladding layer had a small amount of scratch and wear debris when a load of 20 N was used. As the test load increased, the wear debris in the cladding layer also increased and the massive furrows were not observed.

  13. Influence of cladding on the linear elastic RPV analysis during a loss-of-coolant accident

    International Nuclear Information System (INIS)

    The results of the fictitious linear elastic analysis of an inside cooling with the computer program OCA-1 taking the cladding in the reactor pressure vessel into account differ from the calculations containing no cladding effects as follows: The temperature drop in the vessel wall is retarded, with a higher temperature gradient developing in the cladding layer. In addition to the thermal stresses caused by the temperature gradient more stresses develop as a result of the different coefficients of thermal expansion of austenitic cladding and ferritic basic material. The K1-values for crack depths up to a depths of wall of about twice the width of the cladding layer rise considerably during the whole transient time. (orig.)

  14. Simulation of accident and normal fuel rod work with Zr-cladding

    International Nuclear Information System (INIS)

    The technique of simulation of heat-physics, strength and safety characteristics of reactor RBMK and WWER rods under steady-state, transient and accident conditions is presented. That technique is used in mechanic and heat physics codes PULSAR-2 and STALACTITE. Simulation in both full scale and the most stress-loading part of cladding statement under accident conditions are considered. In this zone local swelling and cladding failure are possible. The accident simulation is based on the mechanical creep-plasticity problem solution in three-dimensional approach. The local cladding swelling is initiated with determining of little hot spot on the clad with several degrees temperature departure from average value. Mechanical problem is solved by finite elements method. Interaction of Zr with steam is taken in to account. Fuel and cladding melting, shortness and dispersion formation processes are simulated under subsequent rods warming up. (author). 2 refs., 6 figs

  15. Sensitivity analysis of a PWR fuel element using zircaloy and silicon carbide claddings

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Rochkhudson B. de; Cardoso, Fabiano; Salome, Jean A.D.; Pereira, Claubia; Fortini, Angela, E-mail: rochkhudson@ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Escola de Engenharia. Departamento de Engenharia Nuclear

    2015-07-01

    The alloy composed of zirconium has been used effectively for over 50 years in claddings of nuclear fuel, especially for PWR type reactors. However, to increase fuel enrichment with the aim of raising the burning and maintaining the safety of nuclear plants is of great relevance the study of new materials that can replace safely and efficiently zircaloy cladding. Among several proposed material, silicon carbide (SiC) has a potential to replace zircaloy as fuel cladding material due to its high-temperature tolerance, chemical stability and low neutron affinity. In this paper, the goal is to expand the study with silicon carbide cladding, checking its behavior when submitted to an environment with boron, burnable poison rods, and temperature variations. Sensitivity calculation and the impact in multiplication factor to both claddings, zircaloy and silicon carbide, were performed during the burnup. The neutronic analysis was made using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation) code. (author)

  16. Introduction of an Innovative Cladding Panel System for Multi-Story Buildings

    Directory of Open Access Journals (Sweden)

    Hathairat Maneetes

    2014-08-01

    Full Text Available An Energy Dissipating Cladding System has been developed for use in buildings designed based on the concept of damage-controlled structure in seismic design. This innovative cladding panel system is capable of functioning both as a structural brace, as well as a source of energy dissipation, without demanding inelastic action and ductility from the basic lateral force resisting system. The structural systems of many modern buildings typically have large openings to accommodate glazing systems, and a popular type of construction uses spandrel precast cladding panels at each floor level that supports strip window systems. The present study focuses on developing spandrel type precast concrete cladding panels as supplementary energy dissipating devices that are added to the basic structural system. Through a series of analytical studies, the result of evaluating the ability of the proposed Energy Dissipating Cladding system to improve the earthquake resistance of the buildings is presented here.

  17. A coating to protect spent aluminium-clad research reactor fuel assemblies during extended wet storage

    International Nuclear Information System (INIS)

    Pitting corrosion of aluminium (Al) alloy clad research reactor (RR) fuel in wet storage facilities can be reduced to a large extent by maintaining water parameters within specified limits. However, factors like bimetallic contact, settled solids and synergistic effects of many storage basin water parameters provoke cladding corrosion. Increase in corrosion resistance of spent Al-clad RR fuels can be achieved through the use of conversion coatings. This paper presents: (a) details about the formation of cerium dioxide as a conversion coating on Al alloys used as RR fuel cladding; (b) the corrosion resistance of cerium dioxide coated Al alloy specimens exposed to NaCl solutions. Marked improvements in corrosion resistance of cerium dioxide coated Al specimens were observed. This paper also presents details of a Latin American Project to develop conversion coatings for long term safe wet storage of spent Al-clad RR spent fuel assemblies. (author)

  18. Core-clad silver halide fibers for CO2 laser power transmission

    Science.gov (United States)

    Paiss, Idan; Moser, Frank; Katzir, Abraham

    1991-07-01

    Core-clad optical fibers with efficient IR power delivery are essential components in the development of laser ensoscope systems for surgical applications. The fabrication of such clad fibers of high quality is still an unsolved technical problem. We have investigated parameters of the fabrication of core-clad polyscrystalline silver halide optical fibers and found conditions that yield fibers with relatively good transmittance at 10.6 micrometers (about 3 dB/meter loss) and capable of delivering output power densities up to 3 kwatt/cm2 in CW operation. This performance is lower than what we achieved in core-only silver halide fibers, but the advantage of the protection provided by the clad and a subsequent plastic overcoat, make these core-clad fibers useful in a number of CO2 power transmission applications in laser surgery.

  19. In-reactor degradation of fuel and cladding in fuel pins operated with weld defects

    Energy Technology Data Exchange (ETDEWEB)

    Mishra, Prerna, E-mail: prernam@barc.gov.in [Post Irradiation Examination Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Jathar, V.P.; Singh, J.L.; Sah, D.N.; Shah, Priti K.; Anantharaman, S. [Post Irradiation Examination Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2013-08-15

    A PHWR fuel pin having an incomplete fusion defect in the end plug weld operated in the reactor in the failed condition for a period of 710 days and accumulated a burnup of 4400 MWd/tU. Detailed non destructive and destructive PIE was carried out on the fuel pin to understand the nature and extent of degradation of fuel and cladding in this fuel pin. PIE included visual examination, ultrasonic testing, metallography of a large number of transverse sections of the fuel pin, beta gamma autoradiography, microhardness measurement of the cladding, ring tension test on the cladding. Examination showed secondary hydriding failures, significant rise in fuel temperature resulting in extensive fuel restructuring, severe oxidation of fuel and cladding by steam providing hydrogen for hydride (deuteride) blister formation. Increase in hardness and loss of ductility was also observed in the cladding.

  20. Ultrasonic testing of nuclear fuel rod welds and clad (LWBR Development Program)

    International Nuclear Information System (INIS)

    Ultrasonic techniques were developed utilizing commercially available equipment as a part of the work required in the LWBR Core Manufacturing program for assurance of fuel rod weld integrity and for measurement of fuel rod clad thickness and clad thickness eccentricity. The need for the highest possible resolution and the undesirability of transducer to rod contact dictated the use of a water immersion technique with pulse-echo instrumentation for both weld and clad thickness ultrasonic tests. For the weld test, both longitudinal wave and shear wave inspections were employed with the transducers shuttled together back and forth longitudinally across the weld zone as the fuel rod was rotated. For the clad thickness test, only the longitudinal wave inspection was used with a helical scan pattern along the full length of the clad