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.

Phase and group velocities for Lamb waves in DOP-26 iridium alloy sheet

International Nuclear Information System (INIS)

Simpson, W.A.; McGuire, D.J.

1994-07-01

The relatively coarse grain structure of iridium weldments limits the ultrasonic inspection of these structures to frequencies in the low megahertz range. As the material thickness is nominally 0.635 mm for clad vent set capsules, the low frequencies involved necessarily entail the generation of Lamb waves m the specimen. These waves are, of course, dispersive and detailed knowledge of both the phase and group velocities is required in order to determine accurately the location of flaws detected using Lamb waves. Purpose of this study is to elucidate the behavior of Lamb waves propagating in the capsule alloy and to quantify the velocities so that accurate flaw location is ensured. We describe a numerical technique for computing the phase velocities of Lamb waves (or of any other type of guided wave) and derive the group velocities from this information. A frequency-domain method is described for measuring group velocity when multiple Lamb modes are present and mutually interfering in the time domain, and experimental confirmation of the group velocity is presented for the capsule material

Determination of plastic anisotropy of zirconium alloys cladding

International Nuclear Information System (INIS)

Yamshchikov, N.V.; Prasolov, P.F.; Shestak, V.E.

1991-01-01

Method for determining plastic anisotropy of zurconium alloy cladding is described. It is based on consideration of material as a combination of transversal crystallites with known distribution over orientations. Such approach enables to describe cladding resistance to plastic deformation at arbitrary stressed state, using the results of texture investigations and uniaxial tests of samples, cut out of claddings along three directions. Plastic anisotropy of fuel element claddings 9.15 and 13.6 mm in diameter up to several percents of plastic deformation is shown

The effect of surface depletion on the work function of arc-melted dilute solution tungsten-iridium alloys

International Nuclear Information System (INIS)

D'Cruz, L.A.; Bosch, D.R.; Jacobson, D.L.

1991-01-01

The requirements of thermionic electrode materials have emphasized the need for substantial improvements in microstructural stability, strength, and creep resistance at service temperature in excess of 2,500K. The present work extends an earlier study of the effective work function trends of a series of dilute solution tungsten, iridium alloys with iridium contents of 1, 3, and 5 wt%. Since the lifetime of candidate electrode materials is an important consideration, the present work attempts to evaluate the repeatability of the work function trends in these alloys. The effective work function was obtained from measurements of the current emitted from the electrode surface under UHV conditions in the temperature range of 1,800-2,500K using a Vacuum Emission Vehicle (VEV). The data generated in this work have been compared with data obtained in earlier studies performed on these alloys. It was found that the magnitude of the effective work function of these alloys was affected by changes in the subsurface iridium concentration. Furthermore, these alloys exhibited a dependence of the work function on temperature, after prolonged exposure to elevated temperatures. Such a temperature dependence can be explained by diffusion-controlled changes in the coverage of an iridium monolayer on the surface. It is proposed that the significant difference in effective work function trends obtained after prolonged exposure to elevated temperatures is a direct consequence of changes in the coverage of an iridium-rich monolayer on the electrode surface. The constitution of such a surface layer, however, would be governed by composition changes in the subsurface regions of the electrode caused thermally-activated transport processes

Transmission electron microscopy characterization of laser-clad iron-based alloy on Al-Si alloy

International Nuclear Information System (INIS)

Mei, Z.; Wang, W.Y.; Wang, A.H.

2006-01-01

Microstructure characterization is important for controlling the quality of laser cladding. In the present work, a detailed microstructure characterization by transmission electron microscopy was carried out on the iron-based alloy laser-clad on Al-Si alloy and an unambiguous identification of phases in the coating was accomplished. It was found that there is austenite, Cr 7 C 3 and Cr 23 C 6 in the clad region; α-Al, NiAl 3 , Fe 2 Al 5 and FeAl 2 in the interface region; and α-Al and silicon in the heat-affected region. A brief discussion was given for their existence based on both kinetic and thermodynamic principles

Alloy development for high burnup cladding (PWR)

Energy Technology Data Exchange (ETDEWEB)

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

1999-04-01

An overview on current alloy development for high burnup PWR fuel cladding is given. It is mainly based on literature data. First, the reasons for an increase of the current mean discharge burnup from 35 MWd / kg(U) to 70 MWd / kg(U) are outlined. From the material data, it is shown that a batch average burnup of 60-70 MWd / kg(U), as aimed by many fuel vendors, can not be achieved with stand (=ASTM-) Zry-4 cladding tubes without violating accepted design criteria. Specifically criteria which limit maximum oxide scale thickness and maximum hydrogen content, and to a less degree, maximum creep and growth rate, can not be achieved. The development potential of standard Zry-4 is shown. Even when taking advantage of this potential, it is shown that an 'improved' Zry-4 is reaching its limits when it achieves the target burnup. The behavior of some Zr alloys outside the ASTM range is shown, and the advantages and disadvantages of the 3 alloy groups (ZrSn+transition metals, ZrNb, ZrSnNb+transition metals) which are currently considered to have the development potential for high burnup cladding materials are depicted. Finally, conclusions are drawn. (author). 14 refs., 11 tabs., 82 figs.

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

International Nuclear Information System (INIS)

Okauchi, Hironori; Saida, Kazuyoshi; Nishimoto, Kazutoshi

2011-01-01

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)

Corrosion behaviour of cladded nickel base alloys

International Nuclear Information System (INIS)

Brandl, W.; Ruczinski, D.; Nolde, M.; Blum, J.

1995-01-01

As a consequence of the high cost of nickel base alloys their use as surface layers is convenient. In this paper the properties of SA-as well as RES-cladded NiMo 16Cr16Ti and NiCr21Mo14W being produced in single and multi-layer technique are compared and discussed with respect to their corrosion behaviour. Decisive criteria describing the qualities of the claddings are the mass loss, the susceptibility against intergranular corrosion and the pitting corrosion resistance. The results prove that RES cladding is the most suitable technique to produce corrosion resistant nickel base coatings. The corrosion behaviour of a two-layer RES deposition shows a better resistance against pitting than a three layer SAW cladding. 7 refs

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

Directory of Open Access Journals (Sweden)

Bo Cheng

2016-02-01

Full Text Available In severe loss of coolant accidents (LOCA, similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconium alloy fuel cladding materials are rapidly heated due to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident management, an accident tolerant fuel (ATF design may extend coping and recovery time for operators to restore emergency power, and cooling, and achieve safe shutdown. An ATF is required to possess high resistance to steam oxidation to reduce hydrogen generation and sufficient mechanical strength to maintain fuel rod integrity and core coolability. The initiative undertaken by Electric Power Research Institute (EPRI is to demonstrate the feasibility of developing an ATF cladding with capability to maintain its integrity in 1,200–1,500°C steam for at least 24 hours. This ATF cladding utilizes thin-walled Mo-alloys coated with oxidation-resistant surface layers. The basic design consists of a thin-walled Mo alloy structural tube with a metallurgically bonded, oxidation-resistant outer layer. Two options are being investigated: a commercially available iron, chromium, and aluminum alloy with excellent high temperature oxidation resistance, and a Zr alloy with demonstrated corrosion resistance. As these composite claddings will incorporate either no Zr, or thin Zr outer layers, hydrogen generation under severe LOCA conditions will be greatly reduced. Key technical challenges and uncertainties specific to Mo alloy fuel cladding include: economic core design, industrial scale fabricability, radiation embrittlement, and corrosion and oxidation resistance during normal operation, transients, and severe accidents. Progress in each aspect has been made and key results are

Laser Powder Cladding of Ti-6Al-4V α/β Alloy

Science.gov (United States)

Al-Sayed Ali, Samar Reda; Hussein, Abdel Hamid Ahmed; Nofal, Adel Abdel Menam Saleh; Elgazzar, Haytham Abdelrafea; Sabour, Hassan Abdel

2017-01-01

Laser cladding process was performed on a commercial Ti-6Al-4V (α + β) titanium alloy by means of tungsten carbide-nickel based alloy powder blend. Nd:YAG laser with a 2.2-KW continuous wave was used with coaxial jet nozzle coupled with a standard powder feeding system. Four-track deposition of a blended powder consisting of 60 wt % tungsten carbide (WC) and 40 wt % NiCrBSi was successfully made on the alloy. The high content of the hard WC particles is intended to enhance the abrasion resistance of the titanium alloy. The goal was to create a uniform distribution of hard WC particles that is crack-free and nonporous to enhance the wear resistance of such alloy. This was achieved by changing the laser cladding parameters to reach the optimum conditions for favorable mechanical properties. The laser cladding samples were subjected to thorough microstructure examinations, microhardness and abrasion tests. Phase identification was obtained by X-ray diffraction (XRD). The obtained results revealed that the best clad layers were achieved at a specific heat input value of 59.5 J·mm−2. An increase by more than three folds in the microhardness values of the clad layers was achieved and the wear resistance was improved by values reaching 400 times. PMID:29036935

Laser Powder Cladding of Ti-6Al-4V α/β Alloy.

Science.gov (United States)

Al-Sayed Ali, Samar Reda; Hussein, Abdel Hamid Ahmed; Nofal, Adel Abdel Menam Saleh; Hasseb Elnaby, Salah Elden Ibrahim; Elgazzar, Haytham Abdelrafea; Sabour, Hassan Abdel

2017-10-15

Laser cladding process was performed on a commercial Ti-6Al-4V (α + β) titanium alloy by means of tungsten carbide-nickel based alloy powder blend. Nd:YAG laser with a 2.2-KW continuous wave was used with coaxial jet nozzle coupled with a standard powder feeding system. Four-track deposition of a blended powder consisting of 60 wt % tungsten carbide (WC) and 40 wt % NiCrBSi was successfully made on the alloy. The high content of the hard WC particles is intended to enhance the abrasion resistance of the titanium alloy. The goal was to create a uniform distribution of hard WC particles that is crack-free and nonporous to enhance the wear resistance of such alloy. This was achieved by changing the laser cladding parameters to reach the optimum conditions for favorable mechanical properties. The laser cladding samples were subjected to thorough microstructure examinations, microhardness and abrasion tests. Phase identification was obtained by X-ray diffraction (XRD). The obtained results revealed that the best clad layers were achieved at a specific heat input value of 59.5 J·mm -2 . An increase by more than three folds in the microhardness values of the clad layers was achieved and the wear resistance was improved by values reaching 400 times.

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...

Technology readiness level (TRL) assessment of cladding alloys for advanced nuclear fuels

International Nuclear Information System (INIS)

Shepherd, Daniel

2015-01-01

Reliable fuel claddings are essential for the safe, sustainable and economic operation of nuclear stations. This paper presents a worldwide TRL assessment of advanced claddings for Gen III and IV reactors following an extensive literature review. Claddings include austenitic, ferritic/martensitic (F/M), reduced activation (RA) and oxide dispersion strengthened (ODS) steels as well as advanced iron-based alloys (Kanthal alloys). Also assessed are alloys of zirconium, nickel (including Hastelloy R ), titanium, chromium, vanadium and refractory metals (Nb, Mo, Ta and W). Comparison is made with Cf/C and SiCf/SiC composites, MAX phase ceramics, cermets and TRISO fuel particle coatings. The results show in general that the higher the maximum operating temperature of the cladding, the lower the TRL. Advanced claddings were found to have lower TRLs than the corresponding fuel materials, and therefore may be the limiting factor in the deployment of advanced fuels and even possibly the entire reactor in the case of Gen IV. (authors)

Laser cladding of a Mg based Mg–Gd–Y–Zr alloy with Al–Si powders

International Nuclear Information System (INIS)

Chen, Erlei; Zhang, Kemin; Zou, Jianxin

2016-01-01

Graphical abstract: A Mg based Mg–Gd–Y–Zr alloy was treated by laser cladding with Al–Si powders at different laser scanning speeds. The laser clad layer mainly contains Mg_2Si, Mg_1_7Al_1_2 and Al_2(Gd,Y) phases distributed in the Mg matrix. After laser cladding, the corrosion resistance of the Mg alloy was significantly improved together with increased microhardness in the laser clad layers. - Highlights: • A Mg based Mg–Gd–Y–Zr alloy was laser clad with Al–Si powders. • The microstructure and morphology vary with the depth of the clad layer and the laser scanning speed. • Hardness and corrosion resistance were significantly improved after laser cladding. - Abstract: In the present work, a Mg based Mg–Gd–Y–Zr alloy was subjected to laser cladding with Al–Si powders at different laser scanning speeds in order to improve its surface properties. It is observed that the laser clad layer mainly contains Mg_2Si, Mg_1_7Al_1_2 and Al_2(Gd,Y) phases distributed in the Mg matrix. The depth of the laser clad layer increases with decreasing the scanning speed. The clad layer has graded microstructures and compositions. Both the volume fraction and size of Mg_2Si, Mg_1_7Al_1_2 and Al_2(Gd,Y) phases decreases with the increasing depth. Due to the formation of these hardening phases, the hardness of clad layer reached a maximum value of HV440 when the laser scanning speed is 2 mm/s, more than 5 times of the substrate (HV75). Besides, the corrosion properties of the untreated and laser treated samples were all measured in a NaCl (3.5 wt.%) aqueous solution. The corrosion potential was increased from −1.77 V for the untreated alloy to −1.13 V for the laser clad alloy with scanning rate of 2 mm/s, while the corrosion current density was reduced from 2.10 × 10"−"5 A cm"−"2 to 1.64 × 10"−"6 A cm"−"2. The results show that laser cladding is an efficient method to improve surface properties of Mg–Rare earth alloys.

Laser Powder Cladding of Ti-6Al-4V α/β Alloy

Directory of Open Access Journals (Sweden)

Samar Reda Al-Sayed Ali

2017-10-01

Full Text Available Laser cladding process was performed on a commercial Ti-6Al-4V (α + β titanium alloy by means of tungsten carbide-nickel based alloy powder blend. Nd:YAG laser with a 2.2-KW continuous wave was used with coaxial jet nozzle coupled with a standard powder feeding system. Four-track deposition of a blended powder consisting of 60 wt % tungsten carbide (WC and 40 wt % NiCrBSi was successfully made on the alloy. The high content of the hard WC particles is intended to enhance the abrasion resistance of the titanium alloy. The goal was to create a uniform distribution of hard WC particles that is crack-free and nonporous to enhance the wear resistance of such alloy. This was achieved by changing the laser cladding parameters to reach the optimum conditions for favorable mechanical properties. The laser cladding samples were subjected to thorough microstructure examinations, microhardness and abrasion tests. Phase identification was obtained by X-ray diffraction (XRD. The obtained results revealed that the best clad layers were achieved at a specific heat input value of 59.5 J·mm−2. An increase by more than three folds in the microhardness values of the clad layers was achieved and the wear resistance was improved by values reaching 400 times.

Retention and release of tritium in aluminum clad, Al-Li alloys

International Nuclear Information System (INIS)

Louthan, M.R. Jr.

1991-01-01

Tritium retention in and release from aluminum clad, aluminum-lithium alloys is modeled from experimental and operational data developed during the thirty plus years of tritium production at the Savannah River Site. The model assumes that tritium atoms, formed by the 6 Li(n,α) 3 He reaction, are produced in solid solution in the Al-Li alloy. Because of the low solubility of hydrogen isotopes in aluminum alloys, the irradiated Al-Li rapidly becomes supersaturated in tritium. Newly produced tritium atoms are trapped by lithium atoms to form a lithium tritide. The effective tritium pressure required for trap or tritide stability is the equilibrium decomposition pressure of tritium over a lithium tritide-aluminum mixture. The temperature dependence of tritium release is determined by the permeability of the cladding to tritium and the local equilibrium at the trap sites. This model is used to calculate tritium release from aluminum clad, aluminum-lithium alloys. 9 refs., 3 figs

Microstructural and wear characteristics of cobalt free, nickel base intermetallic alloy deposited by laser cladding

International Nuclear Information System (INIS)

Awasthi, Reena; Kumar, Santosh; Viswanadham, C.S.; Srivastava, D.; Dey, G.K.; Limaye, P.K.

2011-01-01

This paper describes the microstructural and wear characteristics of Ni base intermetallic hardfacing alloy (Tribaloy-700) deposited on stainless steel-316 L substrate by laser cladding technique. Cobalt base hardfacing alloys have been most commonly used hardfacing alloys for application involving wear, corrosion and high temperature resistance. However, the high cost and scarcity of cobalt led to the development of cobalt free hardfacing alloys. Further, in the nuclear industry, the use of cobalt base alloys is limited due to the induced activity of long lived radioisotope 60 Co formed. These difficulties led to the development of various nickel and iron base alloys to replace cobalt base hardfacing alloys. In the present study Ni base intermetallic alloy, free of Cobalt was deposited on stainless steel- 316 L substrate by laser cladding technique. Traditionally, welding and thermal spraying are the most commonly employed hardfacing techniques. Laser cladding has been explored for the deposition of less diluted and fusion-bonded Nickel base clad layer on stainless steel substrate with a low heat input. The laser cladding parameters (Laser power density: 200 W/mm 2 , scanning speed: 430 mm/min, and powder feed rate: 14 gm/min) resulted in defect free clad with minimal dilution of the substrate. The microstructure of the clad layer was examined by Optical microscopy, Scanning electron microscopy, with energy dispersive spectroscopy. The phase analysis was performed by X-ray diffraction technique. The clad layer exhibited sharp substrate/clad interface in the order of planar, cellular, and dendritic from the interface upwards. Dilution of clad with Fe from substrate was very low passing from ∼ 15% at the interface (∼ 40 μm) to ∼ 6% in the clad layer. The clad layer was characterized by the presence of hexagonal closed packed (hcp, MgZn 2 type) intermetallic Laves phase dispersed in the eutectic of Laves and face centered cubic (fcc) gamma solid solution. The

Laser cladding of a Mg based Mg–Gd–Y–Zr alloy with Al–Si powders

Energy Technology Data Exchange (ETDEWEB)

Chen, Erlei [School of Materials Engineering, Shanghai University of Engineering Science, Shanghai 201620 (China); Zhang, Kemin, E-mail: zhangkm@sues.edu.cn [School of Materials Engineering, Shanghai University of Engineering Science, Shanghai 201620 (China); Zou, Jianxin [National Engineering Research Center of Light Alloys Net Forming & School of Materials Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240 (China)

2016-03-30

Graphical abstract: A Mg based Mg–Gd–Y–Zr alloy was treated by laser cladding with Al–Si powders at different laser scanning speeds. The laser clad layer mainly contains Mg{sub 2}Si, Mg{sub 17}Al{sub 12} and Al{sub 2}(Gd,Y) phases distributed in the Mg matrix. After laser cladding, the corrosion resistance of the Mg alloy was significantly improved together with increased microhardness in the laser clad layers. - Highlights: • A Mg based Mg–Gd–Y–Zr alloy was laser clad with Al–Si powders. • The microstructure and morphology vary with the depth of the clad layer and the laser scanning speed. • Hardness and corrosion resistance were significantly improved after laser cladding. - Abstract: In the present work, a Mg based Mg–Gd–Y–Zr alloy was subjected to laser cladding with Al–Si powders at different laser scanning speeds in order to improve its surface properties. It is observed that the laser clad layer mainly contains Mg{sub 2}Si, Mg{sub 17}Al{sub 12} and Al{sub 2}(Gd,Y) phases distributed in the Mg matrix. The depth of the laser clad layer increases with decreasing the scanning speed. The clad layer has graded microstructures and compositions. Both the volume fraction and size of Mg{sub 2}Si, Mg{sub 17}Al{sub 12} and Al{sub 2}(Gd,Y) phases decreases with the increasing depth. Due to the formation of these hardening phases, the hardness of clad layer reached a maximum value of HV440 when the laser scanning speed is 2 mm/s, more than 5 times of the substrate (HV75). Besides, the corrosion properties of the untreated and laser treated samples were all measured in a NaCl (3.5 wt.%) aqueous solution. The corrosion potential was increased from −1.77 V for the untreated alloy to −1.13 V for the laser clad alloy with scanning rate of 2 mm/s, while the corrosion current density was reduced from 2.10 × 10{sup −5} A cm{sup −2} to 1.64 × 10{sup −6} A cm{sup −2}. The results show that laser cladding is an efficient method to improve

Laser cladding to select new glassy alloys

International Nuclear Information System (INIS)

Medrano, L.L.O.; Afonso, C.R.M.; Kiminami, C.S.; Gargarella, P.; Ramasco, B.

2016-01-01

A new experimental technique used to analyze the effect of compositional variation and cooling rate in the phase formation in a multicomponent system is the laser cladding. This work have evaluated the use of laser cladding to discover a new bulk metallic glass (BMG) in the Al-Co-Zr system. Coatings with composition variation have made by laser cladding using Al-Co-Zr alloys powders and the samples produced have been characterized by X ray diffraction, microscopy and energy-dispersive X-ray spectroscopy. The results did not show the composition variation as expected, because of incomplete melting during laser process. It was measured a composition variation tendency that allowed the glass forming investigation by the glass formation criterion λ+Δh 1/2 . The results have showed no glass formation in the coating samples, which prove a limited capacity of Zr-Co-Al system to form glass (author)

Electron-beam welding of thorium-doped iridium alloy sheets

International Nuclear Information System (INIS)

David, S.A.; Liu, C.T.; Hudson, J.D.

1979-04-01

Modified iridium alloys containing 100 ppM Th were found to be very susceptible to hot-cracking during gas tungsten-arc and electron-beam welding. However, the electron-beam welding process showed greater promise of success in welding these alloys, in particular Ir--0.3% W doped with 200 ppM Th and 50 ppM Al. The weldability of this particular alloy was extremely sensitive to the welding parameters, such as beam focus condition and welding speed, and the resulting fusion zone structure. At low speed successful electron-beam welds were made over a narrow range of beam focus conditions. However, at high speeds successful welds can be made over an extended range of focus conditions. The fusion zone grain structure is a strong function of welding speed and focus condition, as well. In the welds that showed hot-cracking, a region of positive segregation of thorium was identified at the fusion boundary. This highly thorium-segregated region seems to act as a potential source for the nucleation of a liquation crack, which later grows as a centerline crack

Thermal creep behavior of N36 zirconium alloy cladding tube

International Nuclear Information System (INIS)

Wang, P.; Zhao, W.; Dai, X.

2015-01-01

N36 is an alloy containing Zr, Sn, Nb and Fe that is developed by China as a superior cladding material to meet the performance of PWR fuel assembly at the maximum fuel rod burn-up. The creep characteristics of N36 zirconium alloy cladding tube were investigated at temperature from 593 K to 723 K with stress ranging from 20 MPa to 160 MPa. Transitions in creep mechanisms were noted, showing the distinct three rate-controlled creep mechanisms for the alloy at test conditions. In the region of low stresses with stress exponent n ∼ 1 and activation energy Q ∼ (104±4) kJ.mol -1 , Coble creep, based on diffusion of materials through grain boundaries, is the dominant rate-controlling mechanism, which contributes to the creep deformation. The formation of slip bands acts as an accommodation mechanism. In the region of middle stress with stress exponent n ∼ 3 and activation energy Q ∼ (195±7) kJ.mol -1 , micro-creep, caused by viscous gliding of dislocations due to the interaction of O atoms with dislocations, controls the deformation. In the high stress region with stress exponent n ∼ 5-6 and activation energy Q ∼ (210±10) kJ.mol -1 , two mechanisms of the climb of edge dislocations (EDC) and the motion of jogged screw dislocation (MJS) contribute to rate controlling process. In test conditions N36 alloy cladding tube behaves a type of creep similar to that noted in class-I (A) alloys

Iron-57 and iridium-193 Moessbauer spectroscopic studies of supported iron-iridium catalysts

International Nuclear Information System (INIS)

Berry, F.J.; Jobson, S.

1988-01-01

57 Fe and 193 Ir Moessbauer spectroscopy shows that silica- and alumina-supported iron-iridium catalysts formed by calcination in air contain mixtures of small particle iron(III) oxide and iridium(IV) oxide. The iridium dioxide in both supported catalysts is reduced in hydrogen to metallic iridium. The α-Fe 2 O 3 in the silica supported materials is predominantly reduced in hydrogen to an iron-iridium alloy whilst in the alumina-supported catalyst the iron is stabilised by treatment in hydrogen as iron(II). Treatment of a hydrogen-reduced silica-supported iron catalyst in hydrogen and carbon monoxide is accompanied by the formation of iron carbides. Carbide formation is not observed when the iron-iridium catalysts are treated in similar atmospheres. The results from the bimetallic catalysts are discussed in terms of the hydrogenation of associatively adsorbed carbon monoxide and the selectivity of supported iron-iridium catalysts to methanol formation. (orig.)

Synthesis and Characterization of High-Entropy Alloy AlFeCoNiCuCr by Laser Cladding

Directory of Open Access Journals (Sweden)

Xiaoyang Ye

2011-01-01

Full Text Available High-entropy alloys have been recently found to have novel microstructures and unique properties. In this study, a novel AlFeCoNiCuCr high-entropy alloy was prepared by laser cladding. The microstructure, chemical composition, and constituent phases of the synthesized alloy were characterized by SEM, EDS, XRD, and TEM, respectively. High-temperature hardness was also evaluated. Experimental results demonstrate that the AlFeCoNiCuCr clad layer is composed of only BCC and FCC phases. The clad layers exhibit higher hardness at higher Al atomic content. The AlFeCoNiCuCr clad layer exhibits increased hardness at temperature between 400–700°C.

Alloy development for cladding and duct applications

International Nuclear Information System (INIS)

Straalsund, J.L.; Johnson, G.D.

1981-01-01

Three general classes of materials under development for cladding and ducts are listed. Solid solution strengthened, or austenitic, alloys are Type 316 stainless steel and D9. Precipitation hardened (also austenitic) alloys consist of D21, D66 and D68. These alloys are similar to such commercial alloys as M-813, Inconel 706, Inconel 718 and Nimonic PE-16. The third general class of alloys is composed of ferritic alloys, with current emphasis being placed on HT-9, a tempered martensitic alloy, and D67, a delta-ferritic steel. The program is comprised of three parallel paths. The current reference, or first generation alloy, is 20% cold worked Type 316 stainless steel. Second generation alloys for near-term applications include D9 and HT-9. Third generation materials consist of the precipitation strengthened steels and ferritic alloys, and are being considered for implementation at a later time than the first and second generation alloys. The development of second and third generation materials was initiated in 1974 with the selection of 35 alloys. This program has proceeded to today where there are six advanced alloys being evaluated. These alloys are the developmental alloys D9, D21, D57, D66 and D68, together with the commerical alloy, HT-9. The status of development of these alloys is summarized

Report of Iridium/{sup 238}PuO{sub 2} Compatibility Test

Energy Technology Data Exchange (ETDEWEB)

Taylor, D.H.

2001-08-09

This study indicates that the chemical purity of the fuel used presently to fabricate fueled clad vent sets will not present any special problems to the performance of the fueled clad vent sets as intended. However, cation impurities in the fuel can have a deleterious effect on the iridium cladding and vents and should be minimized as much as practical.

Weldability and mechanical property characterization of weld clad alloy 800H tubesheet forging

International Nuclear Information System (INIS)

King, J.F.; McCoy, H.E.

1984-09-01

The weldability of an alloy 800H forging that simulates a steam generator tubesheet is studied. Weldability was of concern because a wide range of microstructures was present in this forging. The top and portions of the bottom were weld clad with ERNiC-3 weld metal to a thickness of 19 mm similar to that anticipated for HTGR steam generators. Examinations of the clad fusion line in various regions revealed no weldability problems except possibly on the bottom portion, which contained large grains and some as-cast structure. A few microfissures were evident in this region, but no excessive hot cracking tendency was observed. The tensile properties in all areas of the clad forging were reasonable and not influenced greatly by the microstructure. The elevated-temperature tests showed strong tendency for fracture in the heat-affected zone of the alloy 800H. Creep failure at 649 0 C consistently occurred in the heat-affected zone of the alloy 800H, but the creep strength exceeded the expected values for alloy 800H

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

Energy Technology Data Exchange (ETDEWEB)

Zhuang, W., E-mail: wyman.zhuang@dsto.defence.gov.au [Aerospace Division, Defence Science and Technology Organisation, 506 Lorimer Street, Fishermans Bend, Victoria 3207 (Australia); Liu, Q.; Djugum, R.; Sharp, P.K. [Aerospace Division, Defence Science and Technology Organisation, 506 Lorimer Street, Fishermans Bend, Victoria 3207 (Australia); Paradowska, A. [Australian Nuclear Science and Technology Organisation, Lucas Heights, NSW 2232 (Australia)

2014-11-30

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.

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

International Nuclear Information System (INIS)

Zhuang, W.; Liu, Q.; Djugum, R.; Sharp, P.K.; Paradowska, A.

2014-01-01

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

Safety of some fuel cladding materials, alternative to Zr-alloys

International Nuclear Information System (INIS)

Hache, Georges; Clement, Bernard; Barrachin, Marc

2013-01-01

The Fukushima accident underlined the impact of hydrogen production on LWR core melt accident behaviour. New fuel cladding and structural materials are under development by the industry. IRSN performed a bibliographic study on the behaviour of these materials during LWR core melt accidents. Method This presentation is focused on cladding oxidation by steam and more precisely on: - number of H 2 moles produced per cladding length unit at thermochemical equilibrium; - oxidation kinetics; - heat of reaction; - physic-chemical interactions between material or oxidation products and fuel. Silicon carbide (SiC) - During SiC oxidation by steam, nearly 3 times more explosive gases (CO+H 2 ) moles are produced per cladding length unit at thermochemical equilibrium than for Zr-alloys. - SiC oxidation kinetics below 1700 deg. C: According to early tests performed by NASA and ORNL, the oxidation is linear but slow, there is an effective protection by a thin vitreous SiO 2 layer; these tests underlined the importance of the steam pressure and flow rate. Recently, published MIT and ORNL tests confirm that under large break LOCA conditions (∼5 bars) and up to 1200 deg. C, SiC recession is much slower than for Zr-alloys. Tests under small break conditions (3 inches LOCA: ∼40 bars) were not performed or not published. - SiC oxidation kinetics above 1700 deg. C (melting point of SiO 2 ): Molten SiO 2 loses its protective effect; this is known in the literature as 'catastrophic oxidation by molten oxides'. There will be a cliff-edge effect. For un-inerted containments, H 2 recombiners will be saturated, leading to a risk of CO+H 2 explosion in these containments. - During SiC oxidation by steam, the heat of reaction produced per cladding length unit at thermochemical equilibrium is of the same order of magnitude as for Zr alloys. Molten SiO 2 will interact with UO 2 to form molten mixtures at temperatures well below UO 2 melting temperature. - Calculations were

Grain Growth and Precipitation Behavior of Iridium Alloy DOP-26 During Long Term Aging

Energy Technology Data Exchange (ETDEWEB)

Pierce, Dean T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Muralidharan, Govindarajan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Fox, Ethan E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cox, Victoria A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Geer, Tom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-05-01

The influence of long term aging on grain growth and precipitate sizes and spatial distribution in iridium alloy DOP-26 was studied. Samples of DOP-26 were fabricated using the new process, recrystallized for 1 hour (h) at 1375 C, then aged at either 1300, 1400, or 1500 C for times ranging from 50 to 10,000 h. Grain size measurements (vertical and horizontal mean linear intercept and horizontal and vertical projection) and analyses of iridium-thorium precipitates (size and spacing) were made on the longitudinal, transverse, and rolling surfaces of the as-recrystallized and aged specimens from which the two-dimensional spatial distribution and mean sizes of the precipitates were obtained. The results obtained from this study are intended to provide input to grain growth models.

Development of ODS FeCrAl alloys for accident-tolerant fuel cladding

Energy Technology Data Exchange (ETDEWEB)

Dryepondt, Sebastien N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoelzer, David T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Unocic, Kinga A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-09-18

FeCrAl alloys are prime candidates for accident-tolerant fuel cladding due to their excellent oxidation resistance up to 1400 C and good mechanical properties at intermediate temperature. Former commercial oxide dispersion strengthened (ODS) FeCrAl alloys such as PM2000 exhibit significantly better tensile strength than wrought FeCrAl alloys, which would alloy for the fabrication of a very thin (~250 m) ODS FeCrAl cladding and limit the neutronic penalty from the replacement of Zr-based alloys by Fe-based alloys. Several Fe-12-Cr-5Al ODS alloys where therefore fabricated by ball milling FeCrAl powders with Y2O3 and additional oxides such as TiO₂ or ZrO₂. The new Fe-12Cr-5Al ODS alloys showed excellent tensile strength up to 800 C but limited ductility. Good oxidation resistance in steam at 1200 and 1400 C was observed except for one ODS FeCrAl alloy containing Ti. Rolling trials were conducted at 300, 600 C and 800 C to simulate the fabrication of thin tube cladding and a plate thickness of ~0.6mm was reached before the formation of multiple edge cracks. Hardness measurements at different stages of the rolling process, before and after annealing for 1h at 1000 C, showed that a thinner plate thickness could likely be achieved by using a multi-step approach combining warm rolling and high temperature annealing. Finally, new Fe-10-12Cr-5.5-6Al-Z gas atomized powders have been purchased to fabricate the second generation of low-Cr ODS FeCrAl alloys. The main goals are to assess the effect of O, C, N and Zr contents on the ODS FeCrAl microstructure and mechanical properties, and to optimize the fabrication process to improve the ductility of the 2nd gen ODS FeCrAl while maintaining good mechanical strength and oxidation resistance.

Incorporation of iridium into electrodeposited rhenium–nickel alloys

International Nuclear Information System (INIS)

Cohen Sagiv, Maayan; Eliaz, Noam; Gileadi, Eliezer

2013-01-01

Rhenium (Re), a refractory metal that has gained significant recognition as a high performance engineering material, is mostly used in military, aircraft and aerospace applications, as well as for catalysis in the petrochemical industry. However, its performance at high temperature in humid air is limited by the formation of rhenium heptoxide (Re 2 O 7 ), which penetrates the grain boundaries and causes brittleness. Improvement of this is being sought through the incorporation of iridium (Ir) into Re deposits. To this end, suitable plating baths for Re–Ir–Ni coatings were developed. These alloys were deposited from different aqueous solutions on copper substrates under galvanostatic conditions, in a three-electrode cell. The plating bath consisted of iridium tri-chloride, ammonium perrhenate and nickel sulfamate as the electroactive species, and citric acid as the complexing agent. The effects of bath composition and operating conditions on the Faradaic efficiency (FE), partial current densities, as well as on the thickness of the coatings and their composition were studied. Re–Ir–Ni coatings as thick as 18 μm, with Re-content as high as 73 at.% and Ir-content as high as 29 at.%, were obtained, using different plating baths. A mechanism of the electrochemical process was suggested. It was found that both an HCP Ir 0.4 Re 0.6 phase and an HCP Ni phase with nanometric crystallites were formed, possibly together with a hexagonal nickel hydride (Ni 2 H) phase

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

OpenAIRE

Bo Cheng; Young-Jin Kim; Peter Chou

2016-01-01

In severe loss of coolant accidents (LOCA), similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconium alloy fuel cladding materials are rapidly heated due to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident managem...

Failure analysis of fusion clad alloy system AA3003/AA6xxx sheet under bending

Energy Technology Data Exchange (ETDEWEB)

Shi, Y., E-mail: shiyh@mcmaster.ca [Department of Mechanical Engineering, McMaster University, 1280 Main Street West, Hamilton, Ontario, Canada L8S 4L7 (Canada); Jin, H. [Novelis Global Technology Center, P.O. Box 8400, Kingston, Ontario, Canada K7L 5L9 (Canada); Wu, P.D. [Department of Mechanical Engineering, McMaster University, 1280 Main Street West, Hamilton, Ontario, Canada L8S 4L7 (Canada); Lloyd, D.J. [Aluminum Materials Consultants, 106 Nicholsons Point Road, Bath, Ontario, Canada K0H 1G0 (Canada); Embury, D. [Department of Mechanical Engineering, McMaster University, 1280 Main Street West, Hamilton, Ontario, Canada L8S 4L7 (Canada)

2014-07-29

An ingot of AA6xxx Al–Si–Mg–Cu alloy clad with AA3003 Al–Mn alloy was co-cast by Fusion technology. Bending tests and numerical modeling were performed to investigate the potential for sub-surface cracking for this laminate system. To simulate particle-induced crack initiation and growth, both random and stringer particles have been selected to mimic the particle distribution in the tested samples. The morphology of cracking in the model was similar to that observed in clad sheet tested in the Cantilever bend test. The crack initiated in the core close to the clad-core interface where the strain in the core is highest, between particles or near particles and propagates along local shear bands in the core, while the clad layer experiences extreme thinning before failure.

Evaluations of Mo-alloy for light water reactor fuel cladding to enhance accident tolerance

Directory of Open Access Journals (Sweden)

Cheng Bo

2016-01-01

Full Text Available Molybdenum based alloy is selected as a candidate to enhance tolerance of fuel to severe loss of coolant accidents due to its high melting temperature of ∼2600 °C and ability to maintain sufficient mechanical strength at temperatures exceeding 1200 °C. An outer layer of either a Zr-alloy or Al-containing stainless steel is designed to provide corrosion resistance under normal operation and oxidation resistance in steam exceeding 1000 °C for 24 hours under severe loss of coolant accidents. Due to its higher neutron absorption cross-sections, the Mo-alloy cladding is designed to be less than half the thickness of the current Zr-alloy cladding. A feasibility study has been undertaken to demonstrate (1 fabricability of long, thin wall Mo-alloy tubes, (2 formability of a protective outer coating, (3 weldability of Mo tube to endcaps, (4 corrosion resistance in autoclaves with simulated LWR coolant, (5 oxidation resistance to steam at 1000–1500 °C, and (6 sufficient axial and diametral strength and ductility. High purity Mo as well as Mo + La2O3 ODS alloy have been successfully fabricated into ∼2-meter long tubes for the feasibility study. Preliminary results are encouraging, and hence rodlets with Mo-alloy cladding containing fuel pellets have been under preparation for irradiation at the Advanced Test Reactor (ATR in Idaho National Laboratory. Additional efforts are underway to enhance the Mo cladding mechanical properties via process optimization. Oxidation tests to temperatures up to 1500 °C, and burst and creep tests up to 1000 °C are also underway. In addition, some Mo disks in close contact with UO2 from a previous irradiation program (to >100 GWd/MTU at the Halden Reactor have been subjected to post-irradiation examination to evaluate the chemical compatibility of Mo with irradiated UO2 and fission products. This paper will provide an update on results from the feasibility study and discuss the attributes of the

Microstructure and Wear Behavior of CoCrFeMnNbNi High-Entropy Alloy Coating by TIG Cladding

Directory of Open Access Journals (Sweden)

Wen-yi Huo

2015-01-01

Full Text Available Alloy cladding coatings are widely prepared on the surface of tools and machines. High-entropy alloys are potential replacements of nickel-, iron-, and cobalt-base alloys in machining due to their excellent strength and toughness. In this work, CoCrFeMnNbNi HEA coating was produced on AISI 304 steel by tungsten inert gas cladding. The microstructure and wear behavior of the cladding coating were studied by X-ray diffraction, scanning electron microscopy, energy dispersive spectrometer, microhardness tester, pin-on-ring wear tester, and 3D confocal laser scanning microscope. The microstructure showed up as a nanoscale lamellar structure matrix which is a face-centered-cubic solid solution and niobium-rich Laves phase. The microhardness of the cladding coating is greater than the structure. The cladding coating has excellent wear resistance under the condition of dry sliding wear, and the microploughing in the worn cladding coating is shallower and finer than the worn structure, which is related to composition changes caused by forming the nanoscale lamellar structure of Laves phase.

Laser Cladding of γ-TiAl Intermetallic Alloy on Titanium Alloy Substrates

Science.gov (United States)

Maliutina, Iuliia Nikolaevna; Si-Mohand, Hocine; Piolet, Romain; Missemer, Florent; Popelyukh, Albert Igorevich; Belousova, Natalya Sergeevna; Bertrand, Philippe

2016-01-01

The enhancement of titanium and titanium alloy's tribological properties is of major interest in many applications such as the aerospace and automotive industry. Therefore, the current research paper investigates the laser cladding of Ti48Al2Cr2Nb powder onto Ti6242 titanium alloy substrates. The work was carried out in two steps. First, the optimal deposition parameters were defined using the so-called "combined parameters," i.e., the specific energy E specific and powder density G. Thus, the results show that those combined parameters have a significant influence on the geometry, microstructure, and microhardness of titanium aluminide-formed tracks. Then, the formation of dense, homogeneous, and defect-free coatings based on optimal parameters has been investigated. Optical and scanning electron microscopy techniques as well as energy-dispersive spectroscopy and X-ray diffraction analyses have shown that a duplex structure consisting of γ-TiAl and α 2-Ti3Al phases was obtained in the coatings during laser cladding. Moreover, it was shown that produced coatings exhibit higher values of microhardness (477 ± 9 Hv0.3) and wear resistance (average friction coefficient is 0.31 and volume of worn material is 5 mm3 after 400 m) compared to those obtained with bare titanium alloy substrates (353 Hv0.3, average friction coefficient is 0.57 and a volume of worn material after 400 m is 35 mm3).

Static Recovery Modeling of Dislocation Density in a Cold Rolled Clad Aluminum Alloy

Science.gov (United States)

Penlington, Alex

Clad alloys feature one or more different alloys bonded to the outside of a core alloy, with non-equilibrium, interalloy interfaces. There is limited understanding of the recovery and recrystallization behaviour of cold rolled clad aluminum alloys. In order to optimize the properties of such alloys, new heat treatment processes may be required that differ from what is used for the monolithic alloys. This study examines the recovery behaviour of a cold rolled Novelis Fusion(TM) alloy containing an AA6XXX core with an AA3003 cladding on one side. The bond between alloys appears microscopically discrete and continuous, but has a 30 microm wide chemical gradient. The as-deformed structure at the interalloy region consists of pancaked sub-grains with dislocations at the misorientation boundaries and a lower density organized within the more open interiors. X-ray line broadening was used to extract the dislocation density from the interalloy region and an equivalently deformed AA6XXX following static annealing using a modified Williamson-Hall analysis. This analysis assumed that Gaussian broadening contributions in a pseudo-Voigt function corresponded only to strain from dislocations. The kinetics of the dislocation density evolution to recrystallization were studied isothermally at 2 minute intervals, and isochronally at 175 and 205°C. The data fit the Nes model, in which the interalloy region recovered faster than AA6XXX at 175°C, but was slower at 205°C. This was most likely caused by change in texture and chemistry within this region such as over-aging of AA6XXX . Simulation of a continuous annealing and self homogenization process both with and without pre-recovery indicates a detectable, though small change in the texture and grain size in the interalloy region.

Zirconium alloy fuel cladding resistant to PCI crack propagation

International Nuclear Information System (INIS)

Boyle, R.F.; Foster, J.P.

1987-01-01

A nuclear fuel element is described cladding tube comprising: concentric tubular layers of zirconium base alloys; the concentric tubular layers including an inner layer and outer layer; the outer layer metallurgically bonded to the inner layer; the outer layer composed of a first zirconium base alloy characterized by excellent resistance to corrosion caused by exposure to high temperature and pressure aqueous environments; the inner layer composed of a second zirconium base alloy consisting of: about 0.2 to 0.6 wt.% tin, about 0.03 to 0.11 wt.% iron, less than about 0.02 wt.% chromium, up to about 350 ppm oxygen and the remainder being zirconium and incidental impurities, and the inner layer characterized by improved resistance to crack propagation under reactor operating conditions compared to the first zirconium alloy

Oxidation behavior of laser-clad NiAlCrHf alloys

International Nuclear Information System (INIS)

Ribaudo, C.R.

1991-01-01

Laser cladding is the process where a mechanical mixture of powders is rapidly melted and fused to a solid substrate using a CO 2 laser. The effects of laser cladding upon scale retention on NiAlCrHf alloys after cyclic and isothermal exposure to air were investigated. The stress developed in the scale during cooling after exposure was estimated using a thermoelastic model. Additions of up to ∼2 1/2 wt % Hf increasingly promote retention of scales grown at 1,200C. Laser-clad samples containing ∼2 1/2 wt % Hf retained almost-intact scales. The improvement in scale retention is due to improved toughness in scales containing hafnia-rich polycrystallites possibly via microcracking initiated by anisotropic thermal contraction of the hafnia. Laser cladding provides a large concentration of ∼1 μm Hf-rich particles that are precursors of the hafnia in the scale as well as a fine-dendrite spacing that reduced the mean free distance between particles

Iron-chrome-aluminum alloy cladding for increasing safety in nuclear power plants

Science.gov (United States)

Rebak, Raul B.

2017-12-01

After a tsunami caused plant black out at Fukushima, followed by hydrogen explosions, the US Department of Energy partnered with fuel vendors to study safer alternatives to the current UO2-zirconium alloy system. This accident tolerant fuel alternative should better tolerate loss of cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. General electric, Oak ridge national laboratory, and their partners are proposing to replace zirconium alloy cladding in current commercial light water power reactors with an iron-chromium-aluminum (FeCrAl) cladding such as APMT or C26M. Extensive testing and evaluation is being conducted to determine the suitability of FeCrAl under normal operation conditions and under severe accident conditions. Results show that FeCrAl has excellent corrosion resistance under normal operation conditions and FeCrAl is several orders of magnitude more resistant than zirconium alloys to degradation by superheated steam under accident conditions, generating less heat of oxidation and lower amount of combustible hydrogen gas. Higher neutron absorption and tritium release effects can be minimized by design changes. The implementation of FeCrAl cladding is a near term solution to enhance the safety of the current fleet of commercial light water power reactors.

Microstructure and corrosion resistance of TC2 Ti alloy by laser cladding with Ti/TiC/TiB_2 powders

International Nuclear Information System (INIS)

Diao, Yunhua; Zhang, Kemin

2015-01-01

Highlights: • A TiC/TiB_2 composite coating was produced onto a TC2 Ti alloy by laser cladding with Ti/TiC/TiB_2 powders. • A maximum hardness of 1100 HV was achieved in the laser clad TiC/TiB_2 composite layer. • Corrosion resistance of the TC2 alloy in NaCl (3.5 wt%) aqueous solution can be improved after laser cladding. - Abstract: In the present work, a TiC/TiB_2 composite coating was produced onto a TC2 Ti alloy by laser cladding with Ti/TiC/TiB_2 powders. The surface microstructure, phase components and compositions were characterized with methods of optical microscopy (OM), scanning electron microscopy (SEM), X-ray diffractometry (XRD), and energy dispersive spectrometry (EDS). The cladding layer is consisted of Ti, TiC and TiB_2. And the surface microhardness was measured. After laser cladding, a maximum hardness of 1100 HV is achieved in the laser cladding surface layer, which is more three times higher than that of the TC2 substrate (∼300 HV). Due to the formation of TiC and TiB_2 intermetallic compounds in the alloyed region and grain refinement, the microhardness of coating is higher than TC2 Ti alloy. In this paper, the corrosion property of matrix material and treated samples were both measured in NaCl (3.5 wt%) aqueous solution. From the result we can see that the laser cladding specimens’ corrosion property is clearly becoming better than that of the substrate.

Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

International Nuclear Information System (INIS)

Rebak, Raul B.

2014-01-01

The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Rebak, Raul B. [General Electric Global Research, Schnectady, NY (United States)

2014-09-30

The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

Deposition of Co-Ti alloy on mild steel substrate using laser cladding

International Nuclear Information System (INIS)

Alemohammad, Hamidreza; Esmaeili, Shahrzad; Toyserkani, Ehsan

2007-01-01

Laser cladding of a Co-Ti alloy on a mild steel substrate is studied. Premixed powders with the composition of 85 wt% cobalt and 15 wt% titanium are pre-placed on the substrate and a moving laser beam at different velocities is used to produce clad layers well bounded to the substrate. Characteristics of the clad are investigated using optical microscopy, X-ray diffraction (XRD), energy dispersive spectroscopy (EDS) and microhardness tests. The results reveal that the intermetallic phase TiCo 3 and β (i.e. fcc) cobalt are formed in the clad layer. The clad layer can also have major dilution from the substrate depending on the laser scanning velocity. It is observed that a finer microstructure is achievable with higher laser velocities whereas higher hardness is achieved using lower velocities. The latter is due to the formation of a larger fraction of TiCo 3 phase

Deposition of Co-Ti alloy on mild steel substrate using laser cladding

Energy Technology Data Exchange (ETDEWEB)

Alemohammad, Hamidreza [University of Waterloo, Department of Mechanical and Mechatronics Engineering, 200 University Avenue West, Waterloo, ON N2L 3G1 (Canada)], E-mail: shalemoh@engmail.uwaterloo.ca; Esmaeili, Shahrzad [University of Waterloo, Department of Mechanical and Mechatronics Engineering, 200 University Avenue West, Waterloo, ON N2L 3G1 (Canada); Toyserkani, Ehsan [University of Waterloo, Department of Mechanical and Mechatronics Engineering, 200 University Avenue West, Waterloo, ON N2L 3G1 (Canada)

2007-05-15

Laser cladding of a Co-Ti alloy on a mild steel substrate is studied. Premixed powders with the composition of 85 wt% cobalt and 15 wt% titanium are pre-placed on the substrate and a moving laser beam at different velocities is used to produce clad layers well bounded to the substrate. Characteristics of the clad are investigated using optical microscopy, X-ray diffraction (XRD), energy dispersive spectroscopy (EDS) and microhardness tests. The results reveal that the intermetallic phase TiCo{sub 3} and {beta} (i.e. fcc) cobalt are formed in the clad layer. The clad layer can also have major dilution from the substrate depending on the laser scanning velocity. It is observed that a finer microstructure is achievable with higher laser velocities whereas higher hardness is achieved using lower velocities. The latter is due to the formation of a larger fraction of TiCo{sub 3} phase.

Influence of Zirconia on Hydroxyapatite Coating on Ti-Alloy by Laser Cladding

Institute of Scientific and Technical Information of China (English)

杜海燕; 霍伟荣; 高海; 王丽娟; 邱世鹏; 刘家臣

2003-01-01

Coating titanium alloy with the bioceramic material hydroxyapatite(HAP) has been used to improve the poor osteoinductive properties of pure titanium alloy. But in clinical applications, the mechanical failure of HAP-coated titanium alloy implant suffered at the interface of the HAP coatings and titanium alloy substrate will be a potential weakness in prosthesis. Yttria-stablized zirconia (YSZ) is expected to enhance the mechanical properties of the HAP coating and reduce the coefficient of thermal expansion difference between the coated layer and the substrate. These may reinforce the bonding strength between the coatings and the substrate. In this paper, HAP/YSZ composite coatings were cladded by laser. The effects of zirconia on the microstructure, mechanical properties and formation of tricalcium phosphate (TCP, Ca3(PO4)2) of the HAP/YSZ composite coatings were evaluated. XRD, SEM and TEM were used to investigate the phase composition, microstructure and morphology of the coatings. The experimental results showed that adding YSZ in coatings was favorable to the composition and stability of HAP, and to the improvement of the adhesion strength, microhardness and microtoughness. A well uniform, crack-free coating of HAP/YSZ composites was formed on Ti-alloy substrate by laser cladding.

Microstructural Characterization of Clad Interface in Welds of Ni-Cr-Mo High Strength Low Alloy Steel

Energy Technology Data Exchange (ETDEWEB)

Kim, Hong-Eun; Kim, Min-Chul; Lee, Ho-Jin; Kim, Keong-Ho [KAERI, Daejeon (Korea, Republic of); Lee, Ki-Hyoung [KAIST, Daejeon (Korea, Republic of); Lee, Chang-Hee [Hanyang Univ., Seoul (Korea, Republic of)

2011-08-15

SA508 Gr.4N Ni-Cr-Mo low alloy steel, in which Ni and Cr contents are higher than in commercial SA508 Gr.3 Mn-Mo-Ni low alloy steels, may be a candidate reactor pressure vessel (RPV) material with higher strength and toughness from its tempered martensitic microstructure. The inner surface of the RPV is weld-cladded with stainless steels to prevent corrosion. The goal of this study is to evaluate the microstructural properties of the clad interface between Ni-Cr-Mo low alloy steel and stainless weldment, and the effects of post weld heat treatment (PWHT) on the properties. The properties of the clad interface were compared with those of commercial Mn-Mo-Ni low alloy steel. Multi-layer welding of model alloys with ER308L and ER309L stainless steel by the SAW method was performed, and then PWHT was conducted at 610°C for 30 h. The microstructural changes of the clad interface were analyzed using OM, SEM and TEM, and micro-Vickers hardness tests were performed. Before PWHT, the heat affected zone (HAZ) showed higher hardness than base and weld metals due to formation of martensite after welding in both steels. In addition, the hardness of the HAZ in Ni-Cr-Mo low alloy steel was higher than that in Mn-Mo-Ni low alloy steel due to a comparatively high martensite fraction. The hardness of the HAZ decreased after PWHT in both steels, but the dark region was formed near the fusion line in which the hardness was locally high. In the case of Mn-Mo-Ni low alloy steel, formation of fine Cr-carbides in the weld region near the fusion line by diffusion of C from the base metal resulted in locally high hardness in the dark region. However, the precipitates of the region in the Ni-Cr-Mo low alloy steel were similar to that in the base metal, and the hardness in the region was not greatly different from that in the base metal.

Microstructural Characterization of Clad Interface in Welds of Ni-Cr-Mo High Strength Low Alloy Steel

International Nuclear Information System (INIS)

Kim, Hong-Eun; Kim, Min-Chul; Lee, Ho-Jin; Kim, Keong-Ho; Lee, Ki-Hyoung; Lee, Chang-Hee

2011-01-01

SA508 Gr.4N Ni-Cr-Mo low alloy steel, in which Ni and Cr contents are higher than in commercial SA508 Gr.3 Mn-Mo-Ni low alloy steels, may be a candidate reactor pressure vessel (RPV) material with higher strength and toughness from its tempered martensitic microstructure. The inner surface of the RPV is weld-cladded with stainless steels to prevent corrosion. The goal of this study is to evaluate the microstructural properties of the clad interface between Ni-Cr-Mo low alloy steel and stainless weldment, and the effects of post weld heat treatment (PWHT) on the properties. The properties of the clad interface were compared with those of commercial Mn-Mo-Ni low alloy steel. Multi-layer welding of model alloys with ER308L and ER309L stainless steel by the SAW method was performed, and then PWHT was conducted at 610°C for 30 h. The microstructural changes of the clad interface were analyzed using OM, SEM and TEM, and micro-Vickers hardness tests were performed. Before PWHT, the heat affected zone (HAZ) showed higher hardness than base and weld metals due to formation of martensite after welding in both steels. In addition, the hardness of the HAZ in Ni-Cr-Mo low alloy steel was higher than that in Mn-Mo-Ni low alloy steel due to a comparatively high martensite fraction. The hardness of the HAZ decreased after PWHT in both steels, but the dark region was formed near the fusion line in which the hardness was locally high. In the case of Mn-Mo-Ni low alloy steel, formation of fine Cr-carbides in the weld region near the fusion line by diffusion of C from the base metal resulted in locally high hardness in the dark region. However, the precipitates of the region in the Ni-Cr-Mo low alloy steel were similar to that in the base metal, and the hardness in the region was not greatly different from that in the base metal.

Investigation on wear resistance and corrosion resistance of electron beam cladding co-alloy coating on Inconel617

Science.gov (United States)

Liu, Hailang; Zhang, Guopei; Huang, Yiping; Qi, Zhengwei; Wang, Bo; Yu, Zhibiao; Wang, Dezhi

2018-04-01

To improve surface properties of Inconel 617 alloy (referred to as 617 alloy), co-alloy coating metallurgically bonded to substrate was prepared on the surface of 617 alloy by electron beam cladding. The microstructure, phase composition, microhardness, tribological properties and corrosion resistance of the coatings were investigated. The XRD results of the coatings reinforced by co-alloy (Co800) revealed the presence of γ-Co, CoCx and Cr23C6 phase as matrix and new metastable phases of Cr2Ni3 and Co3Mo2Si. These hypoeutectic structures contain primary dendrites and interdendritic eutectics. The metallurgical bonding forms well between the cladding layer and the matrix of 617 alloy. In most studied conditions, the co-alloy coating displays a better hardness, tribological performance, i.e., lower coefficient of frictions and wear rates, corrosion resistance in 1 mol L‑1 HCl solution, than the 617 alloy.

Air oxidation of Zircaloy-4, M5 (registered) and ZIRLOTM cladding alloys at high temperatures

International Nuclear Information System (INIS)

Steinbrueck, M.; Boettcher, M.

2011-01-01

The paper presents the results of isothermal and transient oxidation experiments of the advanced cladding alloys M5 (registered) and ZIRLO TM in comparison to Zircaloy-4 in air at temperatures from 973 to 1853 K. Generally, oxidation in air leads to a strong degradation of the cladding material. The main mechanism of this process is the formation of zirconium nitride and its re-oxidation. From the point of view of safety, the barrier effect of the fuel cladding is lost much earlier than during accident transients with a steam atmosphere only. Comparison of the three alloys investigated reveals a qualitatively similar, but quantitatively varying oxidation behavior in air. The mainly parabolic oxidation kinetics, where applicable, is comparable for the three alloys. Strong differences of up to 500% in oxidation rates were observed after transition to linear kinetics at temperatures below 1300 K. The paper presents kinetic rate constants as well as critical times and oxide scale thicknesses at the point of transition from parabolic to linear kinetics.

Microstructure and osteoblast response of gradient bioceramic coating on titanium alloy fabricated by laser cladding

International Nuclear Information System (INIS)

Zheng Min; Fan Ding; Li Xiukun; Li Wenfei; Liu Qibin; Zhang Jianbin

2008-01-01

To construct a bioactive interface between metal implant and the surrounding bone tissue, the gradient calcium phosphate bioceramic coating on titanium alloy (Ti-6Al-4V) was designed and fabricated by laser cladding. The results demonstrated that the gradient bioceramic coating was metallurgically bonded to the titanium alloy substrate. The appearance of hydroxyapatite and β-tricalcium phosphate indicated that the bioactive phases were synthesized on the surface of coating. The microhardness gradually decreased from the coating to substrate, which could help stress relaxation between coating and bone tissue. Furthermore, the methyl thiazolyl tetrazolium (MTT) assay of cell proliferation revealed that the laser-cladded bioceramic coating had more favorable osteoblast response compared with the surface of untreated titanium alloy substrate

Microstructure and osteoblast response of gradient bioceramic coating on titanium alloy fabricated by laser cladding

Energy Technology Data Exchange (ETDEWEB)

Zheng Min [State Key Laboratory of Gansu Advanced Non-ferrous Metal Materials, Lanzhou University of Technology, Lanzhou 730050 (China)], E-mail: zhminmin@sina.com; Fan Ding; Li Xiukun [State Key Laboratory of Gansu Advanced Non-ferrous Metal Materials, Lanzhou University of Technology, Lanzhou 730050 (China); Li Wenfei; Liu Qibin [College of Materials Science and Engineering, Guizhou University, Guiyang 550003 (China); Zhang Jianbin [State Key Laboratory of Gansu Advanced Non-ferrous Metal Materials, Lanzhou University of Technology, Lanzhou 730050 (China)

2008-11-15

To construct a bioactive interface between metal implant and the surrounding bone tissue, the gradient calcium phosphate bioceramic coating on titanium alloy (Ti-6Al-4V) was designed and fabricated by laser cladding. The results demonstrated that the gradient bioceramic coating was metallurgically bonded to the titanium alloy substrate. The appearance of hydroxyapatite and {beta}-tricalcium phosphate indicated that the bioactive phases were synthesized on the surface of coating. The microhardness gradually decreased from the coating to substrate, which could help stress relaxation between coating and bone tissue. Furthermore, the methyl thiazolyl tetrazolium (MTT) assay of cell proliferation revealed that the laser-cladded bioceramic coating had more favorable osteoblast response compared with the surface of untreated titanium alloy substrate.

Microstructure and corrosion resistance of TC2 Ti alloy by laser cladding with Ti/TiC/TiB{sub 2} powders

Energy Technology Data Exchange (ETDEWEB)

Diao, Yunhua, E-mail: 990722012@qq.com; Zhang, Kemin, E-mail: zhangkm@sues.edu.cn

2015-10-15

Highlights: • A TiC/TiB{sub 2} composite coating was produced onto a TC2 Ti alloy by laser cladding with Ti/TiC/TiB{sub 2} powders. • A maximum hardness of 1100 HV was achieved in the laser clad TiC/TiB{sub 2} composite layer. • Corrosion resistance of the TC2 alloy in NaCl (3.5 wt%) aqueous solution can be improved after laser cladding. - Abstract: In the present work, a TiC/TiB{sub 2} composite coating was produced onto a TC2 Ti alloy by laser cladding with Ti/TiC/TiB{sub 2} powders. The surface microstructure, phase components and compositions were characterized with methods of optical microscopy (OM), scanning electron microscopy (SEM), X-ray diffractometry (XRD), and energy dispersive spectrometry (EDS). The cladding layer is consisted of Ti, TiC and TiB{sub 2}. And the surface microhardness was measured. After laser cladding, a maximum hardness of 1100 HV is achieved in the laser cladding surface layer, which is more three times higher than that of the TC2 substrate (∼300 HV). Due to the formation of TiC and TiB{sub 2} intermetallic compounds in the alloyed region and grain refinement, the microhardness of coating is higher than TC2 Ti alloy. In this paper, the corrosion property of matrix material and treated samples were both measured in NaCl (3.5 wt%) aqueous solution. From the result we can see that the laser cladding specimens’ corrosion property is clearly becoming better than that of the substrate.

Laser cladding of stainless steel with a copper-silver alloy to generate surfaces of high antimicrobial activity

Science.gov (United States)

Hans, Michael; Támara, Juan Carlos; Mathews, Salima; Bax, Benjamin; Hegetschweiler, Andreas; Kautenburger, Ralf; Solioz, Marc; Mücklich, Frank

2014-11-01

Copper and silver are used as antimicrobial agents in the healthcare sector in an effort to curb infections caused by bacteria resistant to multiple antibiotics. While the bactericidal potential of copper and silver alone are well documented, not much is known about the antimicrobial properties of copper-silver alloys. This study focuses on the antibacterial activity and material aspects of a copper-silver model alloy with 10 wt% Ag. The alloy was generated as a coating with controlled intermixing of copper and silver on stainless steel by a laser cladding process. The microstructure of the clad was found to be two-phased and in thermal equilibrium with minor Cu2O inclusions. Ion release and killing of Escherichia coli under wet conditions were assessed with the alloy, pure silver, pure copper and stainless steel. It was found that the copper-silver alloy, compared to the pure elements, exhibited enhanced killing of E. coli, which correlated with an up to 28-fold increased release of copper ions. The results show that laser cladding with copper and silver allows the generation of surfaces with enhanced antimicrobial properties. The process is particularly attractive since it can be applied to existing surfaces.

Corrosion properties of cladding materials from Zr1Nb alloy

International Nuclear Information System (INIS)

Kloc, K.; Kosler, S.

1975-01-01

The corrosion behaviour was observed of the Zr1Nb alloy in hot water and superheated steam and the effects of impurity content, of the purity of the corrosion environment and of the heat treatment of the alloy were studied on the alloy corrosion resistance. Also studied were the absorption of hydrogen by the alloy and its behaviour in reactor situations. It was ascertained that the alloy has a good corrosion resistance up to a temperature of 350 degC. The corrosion resistance is reduced by the presence of nitrogen above 50 to 70 ppm and of carbon above 50 to 90 ppm. A graphic representation is given of the dependence of corrosion resistance on the temperature of annealing, the nitrogen content of the alloy and the time of the action of hot water or steam, as well as the dependence of the hydrogen content in the alloy on the peripheral tension of the cladding in hot water both in non-active environment and at irradiation with a neutron flux of approximately 10 20 n/cm 2 . (J.B.)

Laser Powder Cladding of Ti-6Al-4V α/β Alloy

OpenAIRE

Samar Reda Al-Sayed Ali; Abdel Hamid Ahmed Hussein; Adel Abdel Menam Saleh Nofal; Salah Elden Ibrahim Hasseb Elnaby; Haytham Abdelrafea Elgazzar; Hassan Abdel Sabour

2017-01-01

Laser cladding process was performed on a commercial Ti-6Al-4V (α + β) titanium alloy by means of tungsten carbide-nickel based alloy powder blend. Nd:YAG laser with a 2.2-KW continuous wave was used with coaxial jet nozzle coupled with a standard powder feeding system. Four-track deposition of a blended powder consisting of 60 wt % tungsten carbide (WC) and 40 wt % NiCrBSi was successfully made on the alloy. The high content of the hard WC particles is intended to enhance the abrasion resist...

Warm hydroforming of iridium + 0.3 wt % tungsten hemishells

International Nuclear Information System (INIS)

Wyder, W.C.

1976-01-01

A technique for the production of iridium alloy hemispherical shells to be used for the primary encapsulation of plutonium-238 oxide spheres is described. The encapsulated spheres, 24 per heat source, provide the thermal heat used to drive thermoelectric converters which supply power for the Department of Defense's LES 8/9 satellites. The technique used a standard production type Hydroform machine converted for hot tooling. The iridium alloy discs were canned in stainless steel waster sheets of a larger size than the iridium discs and separated from the latter by tantalum foil barrier discs. The stainless steel was electron beam welded around the edge to form an envelope. The iridium disc assembly was heated to approximately 900 0 C and the tooling to approximately 500 0 C. After forming, the iridium shell was cut to length while in the waster sheet; and the latter was removed by dissolving in hot aqua regia. After removal of the waster sheets, the hemishells were cleaned and heat treated. Production efficiency reached a level of better than 95 percent and was maintained to produce some 600-odd hemishells

Phase constituents and microstructure of laser cladding Al2O3/Ti3Al reinforced ceramic layer on titanium alloy

International Nuclear Information System (INIS)

Li Jianing; Chen Chuanzhong; Lin Zhaoqing; Squartini, Tiziano

2011-01-01

Research highlights: → In this study, Fe 3 Al has been chosen as cladding powder due to its excellent properties of wear resistance and high strength, etc. → Laser cladding of Fe 3 Al + TiB 2 /Al 2 O 3 pre-placed alloy powder on Ti-6Al-4V alloy substrate can form the Ti 3 Al/Fe 3 Al + TiB 2 /Al 2 O 3 ceramic layer, which can increase wear resistance of substrate. → In cladding process, Al 2 O 3 can react with TiB 2 leading to formation of Ti 3 Al and B. → This principle can be used to improve the Fe 3 Al + TiB 2 laser-cladded coating. - Abstract: Laser cladding of the Fe 3 Al + TiB 2 /Al 2 O 3 pre-placed alloy powder on Ti-6Al-4V alloy can form the Ti 3 Al/Fe 3 Al + TiB 2 /Al 2 O 3 ceramic layer, which can greatly increase wear resistance of titanium alloy. In this study, the Ti 3 Al/Fe 3 Al + TiB 2 /Al 2 O 3 ceramic layer has been researched by means of electron probe, X-ray diffraction, scanning electron microscope and micro-analyzer. In cladding process, Al 2 O 3 can react with TiB 2 leading to formation of amount of Ti 3 Al and B. This principle can be used to improve the Fe 3 Al + TiB 2 laser cladded coating, it was found that with addition of Al 2 O 3 , the microstructure performance and micro-hardness of the coating was obviously improved due to the action of the Al-Ti-B system and hard phases.

Performance testing of refractory alloy-clad fuel elements for space reactors

International Nuclear Information System (INIS)

Dutt, D.S.; Cox, C.M.; Karnesky, R.A.; Millhollen, M.K.

1985-01-01

Two fast reactor irradiation tests, SP-1 and SP-2, provide a unique and self-consistent data set with which to evaluate the technical feasibility of potential fuel systems for the SP-100 space reactor. Fuel pins fabricated with leading cladding candidates (Nb-1Zr, PWC-11, and Mo-13Re) and fuel forms (UN and UO 2 ) are operated at temperatures typical of those expected in the SP-100 design. The first US fast reactor irradiated, refractory alloy clad fuel pins, from the SP-1 test, reached 1 at. % burnup in EBR-II in March 1985. At that time selected pins were discharged for interim examination. These examinations confirmed the excellent performance of the Nb-1Zr clad uranium oxide and uranium nitride fuel elements, which are the baseline fuel systems for two SP-100 reactor concepts

Development of high performance cladding materials

International Nuclear Information System (INIS)

Park, Jeong Yong; Jeong, Y. H.; Park, S. Y.

2010-04-01

The irradiation test for HANA claddings conducted and a series of evaluation for next-HANA claddings as well as their in-pile and out-of pile performances tests were also carried out at Halden research reactor. The 6th irradiation test have been completed successfully in Halden research reactor. As a result, HANA claddings showed high performance, such as corrosion resistance increased by 40% compared to Zircaloy-4. The high performance of HANA claddings in Halden test has enabled lead test rod program as the first step of the commercialization of HANA claddings. DB has been established for thermal and LOCA-related properties. It was confirmed from the thermal shock test that the integrity of HANA claddings was maintained in more expanded region than the criteria regulated by NRC. The manufacturing process of strips was established in order to apply HANA alloys, which were originally developed for the claddings, to the spacer grids. 250 kinds of model alloys for the next-generation claddings were designed and manufactured over 4 times and used to select the preliminary candidate alloys for the next-generation claddings. The selected candidate alloys showed 50% better corrosion resistance and 20% improved high temperature oxidation resistance compared to the foreign advanced claddings. We established the manufacturing condition controlling the performance of the dual-cooled claddings by changing the reduction rate in the cold working steps

A Prediction Study of Aluminum Alloy Oxidation of the Fuel Cladding in Jordan Research and Training Reactor

Energy Technology Data Exchange (ETDEWEB)

Tahk, Y. W.; Oh, J. Y.; Lee, B. H.; Seo, C. G.; Chae, H. T.; Yim, J. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2010-10-15

U{sub 3}Si{sub 2}-Al dispersion fuel with Al cladding will be used for Jordan Research and Training Reactor (JRTR). Aluminum alloy cladding experiences the oxidation layer growth on the surface during the reactor operation. The formation of oxides on the cladding affects fuel performance by increasing fuel temperature. According to the current JRTR fuel management scheme and operation strategy for 5 MW power, a fresh fuel is discharged after 900 effective full power days (EFPD) with 18 cycles of 50 days loading. For the proper prediction of the aluminum oxide thickness of fuel cladding during the long residence time, a reliable model is needed. In this work, several oxide thickness prediction models are compared with the measured data from in-pile test by RERTR program. Moreover, specific parametric studies and a preliminary prediction of the aluminum alloy oxidation using the latest model are performed for JRTR fuel

FUNDAMENTAL MECHANISMS OF CORROSION OF ADVANCED LIGHT WATER REACTOR FUEL CLADDING ALLOYS AT HIGH BURNUP

International Nuclear Information System (INIS)

Lott, Randy G.

2003-01-01

OAK (B204) The corrosion behavior of nuclear fuel cladding is a key factor limiting the performance of nuclear fuel elements, improved cladding alloys, which resist corrosion and radiation damage, will facilitate higher burnup core designs. The objective of this project is to understand the mechanisms by which alloy composition, heat treatment and microstructure affect corrosion rate. This knowledge can be used to predict the behavior of existing alloys outside the current experience base (for example, at high burn-up) and predict the effects of changes in operation conditions on zirconium alloy behavior. Zirconium alloys corrode by the formation f a highly adherent protective oxide layer. The working hypothesis of this project is that alloy composition, microstructure and heat treatment affect corrosion rates through their effect on the protective oxide structure and ion transport properties. The experimental task in this project is to identify these differences and understand how they affect corrosion behavior. To do this, several microstructural examination techniques including transmission electron microscope (TEM), electrochemical impedance spectroscopy (EIS) and a selection of fluorescence and diffraction techniques using synchrotron radiation at the Advanced Photon Source (APS) were employed

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-08-01

In Phase 1 of this project, a variety of developmental and commercial tubing alloys and claddings was 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, RA85H, HR3C, 253MA, Fe{sub 3}Al + 5Cr, 310 modified, 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 air-cooled, retractable corrosion probes, installed in the reheater cavity, controlled to the operating metal temperatures of an existing and advanced-cycle, coal-fired boiler. Samples of each alloy are being exposed for 4,000, 12,000, and 16,000 hours of operation. The present results are for the metallurgical examination of the corrosion probe samples after approximately 4,400 hours of exposure.

Microstructure and mechanical properties of Al–1Mn and Al–10Si alloy circular clad ingot prepared by direct chill casting

International Nuclear Information System (INIS)

Fu, Ying; Jie, Jinchuan; Wu, Li; Park, Joonpyo; Sun, Jianbo; Kim, Jongho; Li, Tingju

2013-01-01

An innovative direct chill casting process to prepare Al–10 wt%Si and Al–1 wt%Mn alloy circular clad ingots has been developed in the present study. The experimental casting parameters were determined by theoretical analysis, numerical simulation and experimental processes. The interface of clad ingots was investigated by methods of metallographic examination, electron probe microanalysis (EPMA) and transmission electron microscopy (TEM). The results showed that excellent metallurgical bonding of two different aluminum alloys could be achieved by direct chill casting. The Al–1Mn alloy which was poured into the mold earlier served as the substrate for heterogeneous nucleation of Al–10Si alloy. Because of diffusion of Si and Mn elements, a diffusion layer with a thickness of about 40 μm on average between the Al–10Si and Al–1Mn alloys could be obtained. The tensile strength of the clad ingot was 106.8 MPa and the fractured position was located in the Al–1Mn alloy side, indicating the strength of the interfacial region is higher than that of Al–1Mn alloy.

Oxidation of aluminum alloy cladding for research and test reactor fuel

Science.gov (United States)

Kim, Yeon Soo; Hofman, G. L.; Robinson, A. B.; Snelgrove, J. L.; Hanan, N.

2008-08-01

The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

Oxidation of aluminum alloy cladding for research and test reactor fuel

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeon Soo [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States)], E-mail: yskim@anl.gov; Hofman, G.L. [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Robinson, A.B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Snelgrove, J.L.; Hanan, N. [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

2008-08-31

The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

Fiber laser cladding of nickel-based alloy on cast iron

Energy Technology Data Exchange (ETDEWEB)

Arias-González, F., E-mail: felipeag@uvigo.es [Applied Physics Dpt., University of Vigo, EEI, Lagoas-Marcosende, Vigo E-36310 (Spain); Val, J. del [Applied Physics Dpt., University of Vigo, EEI, Lagoas-Marcosende, Vigo E-36310 (Spain); Comesaña, R. [Materials Engineering, Applied Mechanics and Construction Dpt., University of Vigo, EEI, Lagoas-Marcosende, Vigo E-36310 (Spain); Penide, J.; Lusquiños, F.; Quintero, F.; Riveiro, A.; Boutinguiza, M.; Pou, J. [Applied Physics Dpt., University of Vigo, EEI, Lagoas-Marcosende, Vigo E-36310 (Spain)

2016-06-30

Highlights: • Fiber laser cladding of Ni-based alloy on cast iron was experimentally studied. • Two different types of cast iron have been analyzed: gray and ductile cast iron. • Suitable processing parameters to generate a Ni-based coating were determined. • Dilution is higher in gray cast iron samples than in ductile cast iron. • Ni-based coating presents higher hardness than cast iron but similar Young's modulus. - Abstract: Gray cast iron is a ferrous alloy characterized by a carbon-rich phase in form of lamellar graphite in an iron matrix while ductile cast iron presents a carbon-rich phase in form of spheroidal graphite. Graphite presents a higher laser beam absorption than iron matrix and its morphology has also a strong influence on thermal conductivity of the material. The laser cladding process of cast iron is complicated by its heterogeneous microstructure which generates non-homogeneous thermal fields. In this research work, a comparison between different types of cast iron substrates (with different graphite morphology) has been carried out to analyze its impact on the process results. A fiber laser was used to generate a NiCrBSi coating over flat substrates of gray cast iron (EN-GJL-250) and nodular cast iron (EN-GJS-400-15). The relationship between processing parameters (laser irradiance and scanning speed) and geometry of a single laser track was examined. Moreover, microstructure and composition were studied by Scanning Electron Microscopy (SEM), Energy Dispersive X-Ray Spectroscopy (EDS) and X-Ray Diffraction (XRD). The hardness and elastic modulus were analyzed by means of micro- and nanoindentation. A hardfacing coating was generated by fiber laser cladding. Suitable processing parameters to generate the Ni-based alloy coating were determined. For the same processing parameters, gray cast iron samples present higher dilution than cast iron samples. The elastic modulus is similar for the coating and the substrate, while the Ni

Microstructure and wear properties of laser cladding Ti-Al-Fe-B coatings on AA2024 aluminum alloy

International Nuclear Information System (INIS)

Xu Jiang; Liu Wenjin; Kan Yide; Zhong Minlin

2006-01-01

In order to improve wear resistance of aluminum alloy, the in situ synthesized TiB 2 and Ti 3 B 4 peritectic composite particulate reinforced metal matrix composite formed on the 2024 aluminum alloy by laser cladding with a powder mixture of Fe coated Boron, Ti and Al was successfully achieved using 3 kW CW CO 2 laser. The laser cladding coating present excellent bonding with aluminum alloy substrate. The chemical composition, microstructure and phase structure of the composite clad coating were analyzed by energy dispersive X-ray spectroscopy (EDX), SEM and XRD. The typical microstructure of composite coating is composed of TiB 2 , Ti 3 B 4 , Al 3 Ti, Al 3 Fe and α-Al. The surface hardness of cladding coating is increased with the amount of added Fe coated B and Ti powder which determines the amount of TiB 2 and Ti 3 B 4 peritectic composite particulate, and obviously higher than that of substrate. The wear tests were carried out using a FALEX-6 type pin-on-disc machine. The test results show that the composite coatings with the in situ synthesized TiB 2 and Ti 3 B 4 peritectic improve wear resistance when compared with the as-received Al substrate

Experience in quality assurance of alloy D9 clad tubes for Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Kapoor, K.; Prahlad, B.

2012-01-01

Stainless Steel Alloy D9 is the material for cladding in various sub-assemblies of Prototype Fast Breeder Reactor (PFBR). The fabrication, inspection, testing and supply of the clad tubes for the first core of PFBR is nearly completed. The paper also compares the specification requirements and the achieved results for some of the critical aspects which is arrived after completing supply against the first core requirement

Microstructure and wear resistance of laser cladded Ni-Cr-Co-Ti-V high-entropy alloy coating after laser remelting processing

Science.gov (United States)

Cai, Zhaobing; Cui, Xiufang; Liu, Zhe; Li, Yang; Dong, Meiling; Jin, Guo

2018-02-01

An attempt, combined with the technologies of laser cladding and laser remelting, has been made to develop a Ni-Cr-Co-Ti-V high entropy alloy coating. The phase composition, microstructure, micro-hardness and wear resistance (rolling friction) were studied in detail. The results show that after laser remelting, the phase composition remains unchanged, that is, as-cladded coating and as-remelted coatings are all composed of (Ni, Co)Ti2 intermetallic compound, Ti-rich phase and BCC solid solution phase. However, after laser remelting, the volume fraction of Ti-rich phase increases significantly. Moreover, the micro-hardness is increased, up to ∼900 HV at the laser remelting parameters: laser power of 1 kW, laser spot diameter of 3 mm, and laser speed of 10 mm/s. Compared to the as-cladded high-entropy alloy coating, the as-remelted high-entropy alloy coatings have high friction coefficient and low wear mass loss, indicating that the wear resistance of as-remelted coatings is improved and suggesting practical applications, like coatings on brake pads for wear protection. The worn surface morphologies show that the worn mechanism of as-cladded and as-remelted high-entropy alloy coatings are adhesive wear.

Investigation on fuel-cladding chemical interaction in metal fuel for FBR. Reaction of rare earth elements with Fe-Cr alloy

International Nuclear Information System (INIS)

Inagaki, Kenta; Ogata, Takanari

2010-01-01

Rare-earth fission product (FP) elements generated in the metal fuel interact with cladding alloy and result in the wastage of the cladding (Fuel-Cladding Chemical Interaction (FCCI)). To evaluate FCCI quantitatively, several influential factors must be considered. They are temperature, temperature gradient, time, composition of the cladding and the behavior of rare-earth FP. In this research, the temperature and time dependencies are investigated with tests in the simplified system. Fe-12wt%Cr was used as stimulant material of cladding and rare-earth alloy 13La -24Ce -12Pr -39Nd -12Sm (RE) as a rare-earth FP. A diffusion couple Fe-Cr/RE was made and annealed at 923K, 853K, 773K or 693K. The structures of reaction layers were analyzed with Electron Probe Micro Analyzer (EPMA) and the details of the structures were clarified. The width of the reaction layer in the Fe-Cr alloy grew in proportion to the square root of time. The reaction rate constants K=(square of the width of reaction layer / time) were evaluated. It was confirmed that the relation between K and the inverse of the temperature showed linearity above 773 K. (author)

Laser cladding of Zr-based coating on AZ91D magnesium alloy for ...

Indian Academy of Sciences (India)

based coating made of Zr powder was fabricated on AZ91D magnesium alloy by laser cladding. The microstructure of the coating was characterized by XRD, SEM and TEM techniques. The wear resistance of the coating was evaluated under dry ...

Laser cladding Co-based alloy/SiCp composite coatings on IF steel

International Nuclear Information System (INIS)

Li Mingxi; He Yizhu; Sun Guoxiong

2004-01-01

Hardfacing coatings, made of Co-Cr-W-Ni-Si alloy + 20% SiCp, deposited by laser cladding on IF steel is introduced. Cross-section of such coatings has been examined to reveal their microstructure using optical microscope, scanning electron microscope (SEM) and X-ray diffractometer (XRD). MM-200 type wear tester is used to examine wear resistance of the coatings. The results showed that SiCp is dissolved completely during laser cladding process under this conditions, the primary phase γ-Co dendrite and Si 2 W, CoWSi, Cr 3 Si, CoSi 2 formed by C, Si reacting with other elements existed in the coatings. There existed some crystallization morphologies in different regions, such as planar (at the interface), followed cellular and dendrite crystallization from interface to the surface. The direction of solidification changes from one direction perpendicular to interface to multi-directions at the central and upper regions of the clad. The results also showed that the wear resistance of the clad improved by adding SiCp

Aluminum alloy for cladding excellent in sacrificial anode property and erosion-corrosion resistance

International Nuclear Information System (INIS)

Imaizumi, S.; Mikami, K.; Yamada, K.

1980-01-01

An aluminum alloy for cladding excellent in sacrificial anode property and erosion-corrosion resistance, which consists essentially of, in weight percentage: zinc - 0.3 to 3.0%, magnesium - 0.2 to 4.0%, manganese - 0.3 to 2.0%, and, the balance aluminum and incidental impurities; said alloy including an aluminum alloy also containing at least one element selected from the group consisting of, in weight percentage: indium - 0.005 to 0.2%, tin - 0.01 to 0.3%, and, bismuth - 0.01 to 0.3%; provided that the total content of indium, tin and bismuth being up to 0.3%

Microstructure and wear properties of laser cladding Ti-Al-Fe-B coatings on AA2024 aluminum alloy

Energy Technology Data Exchange (ETDEWEB)

Xu Jiang [Laser Processing Research Center, Mechanical Engineering Department, Tsinghua University, Beijing 10084 (China)]. E-mail: xujiang73@sina.com.cn; Liu Wenjin [Laser Processing Research Center, Mechanical Engineering Department, Tsinghua University, Beijing 10084 (China); Kan Yide [Laser Processing Research Center, Mechanical Engineering Department, Tsinghua University, Beijing 10084 (China); Zhong Minlin [Laser Processing Research Center, Mechanical Engineering Department, Tsinghua University, Beijing 10084 (China)

2006-07-01

In order to improve wear resistance of aluminum alloy, the in situ synthesized TiB{sub 2} and Ti{sub 3}B{sub 4} peritectic composite particulate reinforced metal matrix composite formed on the 2024 aluminum alloy by laser cladding with a powder mixture of Fe coated Boron, Ti and Al was successfully achieved using 3 kW CW CO{sub 2} laser. The laser cladding coating present excellent bonding with aluminum alloy substrate. The chemical composition, microstructure and phase structure of the composite clad coating were analyzed by energy dispersive X-ray spectroscopy (EDX), SEM and XRD. The typical microstructure of composite coating is composed of TiB{sub 2}, Ti{sub 3}B{sub 4}, Al{sub 3}Ti, Al{sub 3}Fe and {alpha}-Al. The surface hardness of cladding coating is increased with the amount of added Fe coated B and Ti powder which determines the amount of TiB{sub 2} and Ti{sub 3}B{sub 4} peritectic composite particulate, and obviously higher than that of substrate. The wear tests were carried out using a FALEX-6 type pin-on-disc machine. The test results show that the composite coatings with the in situ synthesized TiB{sub 2} and Ti{sub 3}B{sub 4} peritectic improve wear resistance when compared with the as-received Al substrate.

2nd Gen FeCrAl ODS Alloy Development For Accident-Tolerant Fuel Cladding

Energy Technology Data Exchange (ETDEWEB)

Dryepondt, Sebastien N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Massey, Caleb P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Edmondson, Philip D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-08-01

Extensive research at ORNL aims at developing advanced low-Cr high strength FeCrAl alloys for accident tolerant fuel cladding. One task focuses on the fabrication of new low Cr oxide dispersion strengthened (ODS) FeCrAl alloys. The first Fe-12Cr-5Al+Y₂O₃ (+ ZrO₂ or TiO₂) ODS alloys exhibited excellent tensile strength up to 800 C and good oxidation resistance in steam up to 1400 C, but very limited plastic deformation at temperature ranging from room to 800 C. To improve alloy ductility, several fabrication parameters were considered. New Fe-10-12Cr-6Al gas-atomized powders containing 0.15 to 0.5wt% Zr were procured and ball milled for 10h, 20h or 40h with Y2O3. The resulting powder was then extruded at temperature ranging from 900 to 1050 C. Decreasing the ball milling time or increasing the extrusion temperature changed the alloy grain size leading to lower strength but enhanced ductility. Small variations of the Cr, Zr, O and N content did not seem to significantly impact the alloy tensile properties, and, overall, the 2nd gen ODS FeCrAl alloys showed significantly better ductility than the 1st gen alloys. Tube fabrication needed for fuel cladding will require cold or warm working associated with softening heat treatments, work was therefore initiated to assess the effect of these fabrications steps on the alloy microstructure and properties. This report has been submitted as fulfillment of milestone M3FT 16OR020202091 titled, Report on 2nd Gen FeCrAl ODS Alloy Development for the Department of Energy Office of Nuclear Energy, Advanced Fuel Campaign of the Fuel Cycle R&D program.

Microstructure and high temperature oxidation resistance of Ti-Ni gradient coating on TA2 titanium alloy fabricated by laser cladding

Science.gov (United States)

Liu, Fencheng; Mao, Yuqing; Lin, Xin; Zhou, Baosheng; Qian, Tao

2016-09-01

To improve the high temperature oxidation resistance of TA2 titanium alloy, a gradient Ni-Ti coating was laser cladded on the surface of the TA2 titanium alloy substrate, and the microstructure and oxidation behavior of the laser cladded coating were investigated experimentally. The gradient coating with a thickness of about 420-490 μm contains two different layers, e.g. a bright layer with coarse equiaxed grain and a dark layer with fine and columnar dendrites, and a transition layer with a thickness of about 10 μm exists between the substrate and the cladded coating. NiTi, NiTi2 and Ni3Ti intermetallic compounds are the main constructive phases of the laser cladded coating. The appearance of these phases enhances the microhardness, and the dense structure of the coating improves its oxidation resistance. The solidification procedure of the gradient coating is analyzed and different kinds of solidification processes occur due to the heat dissipation during the laser cladding process.

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.

Fuel assembly and fuel cladding tube

International Nuclear Information System (INIS)

Tsutsumi, Shinro; Ito, Ken-ichi; Inagaki, Masatoshi; Nakajima, Junjiro.

1996-01-01

A fuel cladding tube is a zirconium liner tube formed by lining a pure zirconium layer on the inner side of a zirconium alloy tube. The fuel cladding tube is formed by extrusion molding of a composite billet formed by inserting a pure zirconium billet into a zirconium alloy billet. Accordingly, the pure zirconium layer and the zirconium alloy tube are strongly joined by metal bond. The fuel cladding tube has an external oxide film on the outer surface of the zirconium alloy tube and an internal oxide film on the inner side of the pure zirconium layer. The external oxide film has a thickness preferably of about 1μm. The internal oxide film has a thickness of not more than 10μm, preferably, from 1 to 5μm. With such a constitution, flaws to be formed on both inner and outer surfaces of the cladding tube upon assembling a fuel assembly can be reduced thereby enabling to reduce the amount of hydrogen absorbed to the cladding tube. (I.N.)

White Paper Summary of 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding

Energy Technology Data Exchange (ETDEWEB)

Sindelar, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Louthan, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); PNNL, B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-05-29

This white paper recommends that ASTM International develop standards to address the potential impact of hydrides on the long term performance of irradiated zirconium alloys. The need for such standards was apparent during the 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding and Assembly Components, sponsored by ASTM International Committee C26.13 and held on June 10-12, 2014, in Jackson, Wyoming. The potentially adverse impacts of hydrogen and hydrides on the long term performance of irradiated zirconium-alloy cladding on used fuel were shown to depend on multiple factors such as alloy chemistry and processing, irradiation and post irradiation history, residual and applied stresses and stress states, and the service environment. These factors determine the hydrogen content and hydride morphology in the alloy, which, in turn, influence the response of the alloy to the thermo-mechanical conditions imposed (and anticipated) during storage, transport and disposal of used nuclear fuel. Workshop presentations and discussions showed that although hydrogen/hydride induced degradation of zirconium alloys may be of concern, the potential for occurrence and the extent of anticipated degradation vary throughout the nuclear industry because of the variations in hydrogen content, hydride morphology, alloy chemistry and irradiation conditions. The tools and techniques used to characterize hydrides and hydride morphologies and their impacts on material performance also vary. Such variations make site-to-site comparisons of test results and observations difficult. There is no consensus that a single material or system characteristic (e.g., reactor type, burnup, hydrogen content, end-of life stress, alloy type, drying temperature, etc.) is an effective predictor of material response during long term storage or of performance after long term storage. Multi-variable correlations made for one alloy may not represent the behavior of another alloy exposed to

A Prediction Study on Oxidation of Aluminum Alloy Cladding of U{sub 3}Si{sub 2}-Al Fuel Plate

Energy Technology Data Exchange (ETDEWEB)

Tahk, Y.W.; Lee, B.H.; Oh, J.Y.; Park, J.H.; Yim, J.S. [Research Reactor Design and Engineering Div., Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong, Daejeon 305-353 (Korea, Republic of)

2011-07-01

U{sub 3}Si{sub 2}-Al dispersion fuel with aluminum alloy cladding will be used for the Jordan Research and Training Reactor (JRTR). Aluminum alloy cladding undergoes corrosion at slow rates under operational status. This causes thinning of the cladding walls and impairs heat transfer to the coolant. Predictions of the aluminum oxide thickness of the fuel cladding and the maximum temperature difference across the oxide film are needed for reliability evaluation based on the design criteria and limits which prohibit spallation of oxide film. In this work, several oxide thickness prediction models were compared with the measured data of in-pile test results from RERTR program. Moreover, specific parametric studies and a preliminary prediction of the aluminum alloy oxidation using the latest model were performed for JRTR fuel. According to the current JRTR fuel management scheme and operation strategy for 5 MW power, fresh fuel is discharged after 900 effective full power days (EFPD), which is too long a span to predict oxidation properly without an elaborate model. The latest model developed by Kim et al. is in good agreement with the recent in-pile test data as well as with the out-of-pile test data available in the literature, and is one of the best predictors for the oxidation of aluminum alloy cladding in various operating condition. Accordingly, this model was chosen for estimating the oxide film thickness. Through the preliminarily evaluation, water pH level is to be controlled lower than 6.2 for the conservativeness in the case of including the effect of anticipated operational occurrences and the spent fuel residence time in the storage rack after discharging. (author)

Thermal creep properties of alloy D9 stainless steel and 316 stainless steel fuel clad tubes

International Nuclear Information System (INIS)

Latha, S.; Mathew, M.D.; Parameswaran, P.; Bhanu Sankara Rao, K.; Mannan, S.L.

2008-01-01

Uniaxial thermal creep rupture properties of 20% cold worked alloy D9 stainless steel (alloy D9 SS) fuel clad tubes for fast breeder reactors have been evaluated at 973 K in the stress range 125-250 MPa. The rupture lives were in the range 90-8100 h. The results are compared with the properties of 20% cold worked type 316 stainless steel (316 SS) clad tubes. Alloy D9 SS were found to have higher creep rupture strengths, lower creep rates and lower rupture ductility than 316 SS. The deformation and damage processes were related through Monkman Grant relationship and modified Monkman Grant relationship. The creep damage tolerance parameter indicates that creep fracture takes place by intergranular cavitation. Precipitation of titanium carbides in the matrix and chromium carbides on the grain boundaries, dislocation substructure and twins were observed in transmission electron microscopic investigations of alloy D9 SS. The improvement in strength is attributed to the precipitation of fine titanium carbides in the matrix which prevents the recovery and recrystallisation of the cold worked microstructure

Cladding using a 15 kW CO2 laser

International Nuclear Information System (INIS)

Vesely, E.J.; Verma, S.K.

1989-01-01

Laser alloying or cladding differs little in principle from the traditional forms of weld overlays, but lasers as a heat source offer some distinct advantages. With the selective heating attainable using high power lasers, good metallurgical bond of the clad layer, minimal dilution and typically, a very fine homogeneous microstructure can be obtained in the clad layer. This is a review of work in laser cladding using the 15 kW CO 2 laser. The authors discuss the ability of the laser clad surface to increase the high temperature oxidation resistance of a low-alloy carbon steel (4140). Examples of clads subjected to high- temperature thermal cycling of nickel-20% aluminum and TaC + 4140 clad low-alloy steel and straight high-temperature oxidation of Stellite 6-304L cladding on a 4140 substrate are given

Influence of yttrium on microstructure and properties of Ni–Al alloy coatings prepared by laser cladding

Directory of Open Access Journals (Sweden)

Cun-shan Wang

2014-03-01

Full Text Available Ni–Al alloy coatings with different Y additions are prepared on 45# medium steel by laser cladding. The influence of Y contents on the microstructure and properties of Ni–Al alloy coatings is investigated using X-ray diffraction, scanning electron microscopy, electron probe microanalyzer, Vickers hardness tester, friction wear testing machine, and thermal analyzer. The results show that the cladding layers are mainly composed of NiAl dendrites, and the dendrites are gradually refined with the increase in Y additions. The purification effect of Y can effectively prevent Al2O3 oxide from forming. However, when the atomic percent of Y addition exceeds 1.5%, the extra Y addition will react with O to form Y2O3 oxide, even to form Al5Y3O12 oxide, depending on the amount of Y added. The Y addition in a range of 1.5–3.5 at.% reduces the hardness and anti-attrition of cladding layer, but improves obviously its wear and oxidation resistances.

An iron-57 Moessbauer spectroscopic study of titania-supported iron- and iron-iridium catalysts

International Nuclear Information System (INIS)

Berry, F.J.; Jobson, S.

1992-01-01

57 Fe Moessbauer spectroscopy shows that titania-supported iron is reduced by treatment in hydrogen at significantly lower temperatures than corresponding silica- and alumina-supported catalysts. The metallic iron formed under hydrogen at 600deg C is partially converted to carbide by treatment in carbon monoxide and hydrogen. In contrast to its alumina- and silica-supported counterparts, the remainder of the titania-supported iron is unchanged by this gaseous mixture. The 57 Fe Moessbauer spectra of EXAFS show that iron and iridium in the titania-supported iron-iridium catalysts are reduced in hydrogen at even lower temperatures and, after treatment at 600deg C, are predominantly present as the iron-iridium alloy. The treatment of these reduced catalysts in carbon monoxide and hydrogen is shown by Moessbauer spectroscopy and EXAFS to induce the segregation of iron from the iron-iridium alloy and its conversion to iron oxide. (orig.)

Microstructure of bonding zones in laser-clad Ni-alloy-based composite coatings reinforced with various ceramic powders

International Nuclear Information System (INIS)

Pei, Y.T.; Ouyang, J.H.; Lei, T.C.

1996-01-01

Microstructure of the bonding zones (BZs) between laser-clad Ni-alloy-based composite coatings and steel substrates was studied by means of scanning electron microscope (SEM) and transmission electron microscope (TEM) techniques. Observations indicate that for pure Ni-alloy coating the laser parameters selected for good interface fusion have no effect on the microstructure of the BZ except for its thickness. However, the addition of ceramic particles (TiN, SiC, or ZrO 2 ) to the Ni alloy varies the compositional or constitutional undercooling of the melt near the solid/liquid interface and consequently leads to the observed changes of microstructure of the BZs. For TiN/Ni-alloy coating the morphology of γ-Ni solid solution in the BZ changes from dendritic to planar form with increasing scanning speed. A colony structure of eutectic is found in the BZ of SiC/Ni-alloy coating in which complete dissolution of SiC particles takes place during laser cladding. The immiscible melting of ZrO 2 and Ni-alloy powders induces the stratification of ZrO 2 /Ni-alloy coating which consists of a pure ZrO 2 layer fin the upper region and a BZ composed mainly of γ-Ni dendrites adjacent to the substrate. All the BZs studied in this investigation have good metallurgical characteristics between the coatings and the substrates

Phase constituents and microstructure of laser cladding Al{sub 2}O{sub 3}/Ti{sub 3}Al reinforced ceramic layer on titanium alloy

Energy Technology Data Exchange (ETDEWEB)

Li Jianing [Key Laboratory for Liquid-Solid Structural Evolution and Processing of Materials (Ministry of Education), Department of Materials Science, Shandong University, Jing Shi Road 17923, Jinan 250061, Shandong (China); Chen Chuanzhong, E-mail: czchen@sdu.edu.cn [Key Laboratory for Liquid-Solid Structural Evolution and Processing of Materials (Ministry of Education), Department of Materials Science, Shandong University, Jing Shi Road 17923, Jinan 250061, Shandong (China); Lin Zhaoqing [Key Laboratory for Liquid-Solid Structural Evolution and Processing of Materials (Ministry of Education), Department of Materials Science, Shandong University, Jing Shi Road 17923, Jinan 250061, Shandong (China); Squartini, Tiziano [INFM - Department of Physics, Siena University, Siena 53100 (Italy)

2011-04-07

Research highlights: > In this study, Fe{sub 3}Al has been chosen as cladding powder due to its excellent properties of wear resistance and high strength, etc. > Laser cladding of Fe{sub 3}Al + TiB{sub 2}/Al{sub 2}O{sub 3} pre-placed alloy powder on Ti-6Al-4V alloy substrate can form the Ti{sub 3}Al/Fe{sub 3}Al + TiB{sub 2}/Al{sub 2}O{sub 3} ceramic layer, which can increase wear resistance of substrate. > In cladding process, Al{sub 2}O{sub 3} can react with TiB{sub 2} leading to formation of Ti{sub 3}Al and B. > This principle can be used to improve the Fe{sub 3}Al + TiB{sub 2} laser-cladded coating. - Abstract: Laser cladding of the Fe{sub 3}Al + TiB{sub 2}/Al{sub 2}O{sub 3} pre-placed alloy powder on Ti-6Al-4V alloy can form the Ti{sub 3}Al/Fe{sub 3}Al + TiB{sub 2}/Al{sub 2}O{sub 3} ceramic layer, which can greatly increase wear resistance of titanium alloy. In this study, the Ti{sub 3}Al/Fe{sub 3}Al + TiB{sub 2}/Al{sub 2}O{sub 3} ceramic layer has been researched by means of electron probe, X-ray diffraction, scanning electron microscope and micro-analyzer. In cladding process, Al{sub 2}O{sub 3} can react with TiB{sub 2} leading to formation of amount of Ti{sub 3}Al and B. This principle can be used to improve the Fe{sub 3}Al + TiB{sub 2} laser cladded coating, it was found that with addition of Al{sub 2}O{sub 3}, the microstructure performance and micro-hardness of the coating was obviously improved due to the action of the Al-Ti-B system and hard phases.

Laser cladding of nickel base alloy on SS316L for improved wear and corrosion behaviour

International Nuclear Information System (INIS)

Awasthi, Reena; Kushwaha, R.P.; Chandra, Kamlesh; Viswanadham, C.S.; Srivastava, D.; Dey, G.K.; Limaye, P.K.

2013-01-01

Laser cladding by an Nd:YAG laser was employed to deposit Ni base alloy (Ni-Mo-Cr-Si) on stainless steel-316 L substrate. The resulting defect-free clad with minimum dilution of the substrate was characterized by optical microscopy, scanning electron microscopy, X-ray diffraction and Vickers microhardness test. Dry sliding wear of the cladding and the substrate was evaluated using a ball-on-plate reciprocating wear tester against different counter bodies (WC and 52100 Cr steel). The reciprocating sliding wear resistance of the coating was evaluated as a function of the normal load, keeping the sliding amplitude and sliding speed constant. Wear mechanisms were analyzed by observation of wear track morphology using SEM-EDS. The electrochemical corrosion behavior of clad layer was studied in reducing environment (HCl) to estimate the general corrosion resistance of the laser clad layer in comparison with the substrate SS-316L. The clad layer showed higher wear resistance under reducing condition than that of the substrate material stainless steel 316L. (author)

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.

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

International Nuclear Information System (INIS)

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

Martensitic transformation in Cu-2be alloys induced by explosive cladding

Science.gov (United States)

Ganin, E.; Weiss, B. Z.; Komem, Y.

1986-11-01

Formation of a lath-type structure was observed at a distance greater than 100 ώm from the bond interface created by explosive cladding. The laths were found to have a strong deviation from cubic symmetry and to contain numerous internal faults. The electron diffraction patterns do not fit any equilibrium or metastable phase known to exist in a Cu-2Be alloy. Crystallographic analysis based on electron diffraction showed that the laths have an orthorhombic structure. It is postulated that the orthorhombic phase results from a shear (martensitic) transformation which takes place in the a (fcc) phase during cladding. The proposed model assumes that shear occurs on the (111) plane in the [112] direction, and the orientation relationship is suggested to be [100]ORTH(M)∥[110]α and (001)ORTH(M) II (111)α, which is consistent with electron diffraction results. The transformation causes a volume decrease of 1.1 pct. Formation of the new phase was observed only in the solution-treated specimens of Cu-2Be and not in those aged prior to cladding. It is suggested that this may be a result of different stacking fault energies.

Welding iridium heat-source capsules for space missions

International Nuclear Information System (INIS)

Kanne, W.R. Jr.

1982-03-01

A remote computer-controlled welding station was developed to encapsulate radioactive PuO 2 in iridium. Weld quench cracking caused an interruption in production of capsules for upcoming space missions. Hot crack sensitivity of the DOP-26 iridium alloy was associated with low melting constituents in the grain boundaries. The extent of cracking was reduced but could not be eliminated by changes to the welding operation. An ultrasonic test was developed to detect underbead cracks exceeding a threshold size. Production was continued using the ultrasonic test to reject capsules with detectable cracks

Investigations of chemical reactions between U-Zr alloy and FBR cladding materials

International Nuclear Information System (INIS)

Ishii, Tetsuya; Ukai, Shigeharu

2005-07-01

U-Pu-Zr alloys are candidate materials for commercial FBR fuel. However, informations about chemical reactions with cladding materials developed by JNC for commercial FBR have not been well obtained. In this work, the reaction zones formed in four diffusion couples U-10wt.%Zr/PNC-FMS, U-10wt.%Zr/9Cr-ODS, U-10wt.%Zr/12Cr-ODS, and U-10wt.%Zr/Fe at about 1013K have been examined and following results were obtained. 1) At about 1013K, in the U-10wt.%Zr/Fe couple, the liquid phase zones were obtained. In the other couples U-10wt.%Zr/PNC-FMS, U-10wt.%Zr/9Cr-ODS and U-10wt.%Zr/12Cr-ODS, no liquid phase zones were obtained. The obtained chemical reaction zones in the later 3 couples were similar to the reported ones obtained in U-Zr/Fe couples without liquid phase formation. In comparison with the reaction zones obtained in the U-10wt.%Zr/Fe couple, the reaction zones inside cladding materials obtained in the PNC-FMS, 9Cr-ODS, and 12Cr-ODS couples were thin. 2) From the investigations of relationship between the obtained depths of the chemical reaction zones inside cladding materials and composition of the cladding materials, it was considered that the depth of chemical reaction zone would depend on the Cr content of the cladding materials and the depth would decrease with increasing Cr content, resulting in prevention of liquid phase formation. 3) From the investigations of the mechanisms of chemical reactions between U-Pu-Zr/cladding materials, it was considered that the same effect of Cr obtained in the U-Zr/cladding materials would be expected in U-Pu-Zr/cladding materials. Those seemed to indicate that the threshold temperatures of liquid phase formation for U-Pu-Zr/PNC-FMS, U-Pu-Zr/9Cr-ODS, and U-Pu-Zr/12Cr-ODS might be higher than that for U-Pu-Zr/Fe. (author)

Occurence and prediction of sigma phase in fuel cladding alloys for breeder reactors

International Nuclear Information System (INIS)

Anantatmula, R.P.

1982-01-01

In sodium-cooled fast reactor systems, fuel cladding materials will be exposed for several thousand hours to liquid sodium. Satisfactory performance of the materials depends in part on the sodium compatibility and phase stability of the materials. This paper mainly deals with the phase stability aspect, with particular emphasis on sigma phase formation of the cladding materials upon extended exposures to liquid sodium. A new method of predicting sigma phase formation is proposed for austenitic stainless steels and predictions are compared with the experimental results on fuel cladding materials. Excellent agreement is obtained between theory and experiment. The new method is different from the empirical methods suggested for superalloys and does not suffer from the same drawbacks. The present method uses the Fe-Cr-Ni ternary phase diagram for predicting the sigma-forming tendencies and exhibits a wide range of applicability to austenitic stainless steels and heat-resistant Fe-Cr-Ni alloys

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.

Features of single tracks in coaxial laser cladding of a NIbased self-fluxing alloy

Directory of Open Access Journals (Sweden)

Feldshtein Eugene

2017-01-01

Full Text Available In the present paper, the influence of coaxial laser cladding conditions on the dimensions, microstructure, phases and microhardness of Ni-based self-fluxing alloy single tracks is studied. The height and width of single tracks depend on the speed and distance of the laser cladding: increasing the nozzle distance from the deposited surface 1.4 times reduces the width of the track 1.2 - 1.3 times and increases its height 1.2 times. The increase of the laser spot speed 3 times reduces the track width 1.2 - 1.4 times and the height in 1.5 - 1.6 times. At the same time, the increase of the laser spot speed 3 times reduces the track width 1.2 - 1.4 times and the height 1.5 - 1.6 times. Regularities in the formation of single tracks microstructure with different cladding conditions are defined, as well as regularity of distribution of elements over the track depth and in the transient zone. The patterns of microhardness distribution over the track depth for different cladding conditions are found.

Laser cladding of tungsten carbides (Spherotene) hardfacing alloys for the mining and mineral industry

International Nuclear Information System (INIS)

Amado, J.M.; Tobar, M.J.; Alvarez, J.C.; Lamas, J.; Yanez, A.

2009-01-01

The abrasive nature of the mechanical processes involved in mining and mineral industry often causes significant wear to the associated equipment and derives non-negligible economic costs. One of the possible strategies to improve the wear resistance of the various components is the deposition of hardfacing layers on the bulk parts. The use of high power lasers for hardfacing (laser cladding) has attracted a great attention in the last decade as an alternative to other more standard methods (arc welding, oxy-fuel gas welding, thermal spraying). In laser cladding the hardfacing material is used in powder form. For high hardness applications Ni-, Co- or Fe-based alloys containing hard phase carbides at different ratios are commonly used. Tungsten carbides (WC) can provide coating hardness well above 1000 HV (Vickers). In this respect, commercially available WC powders normally contain spherical micro-particles consisting of crushed WC agglomerates. Some years ago, Spherotene powders consisting of spherical-fused monocrystaline WC particles, being extremely hard, between 1800 and 3000 HV, were patented. Very recently, mixtures of Ni-based alloy with Spherotene powders optimized for laser processing were presented (Technolase). These mixtures have been used in our study. Laser cladding tests with these powders were performed on low carbon steel (C25) substrates, and results in terms of microstructure and hardness will be discussed

Laser cladding of tungsten carbides (Spherotene) hardfacing alloys for the mining and mineral industry

Energy Technology Data Exchange (ETDEWEB)

Amado, J.M. [Departamento de Ingenieria Industrial II, Universidade da Coruna, Mendizabal s/n, Ferrol E-15403 (Spain); Tobar, M.J. [Departamento de Ingenieria Industrial II, Universidade da Coruna, Mendizabal s/n, Ferrol E-15403 (Spain)], E-mail: cote@udc.es; Alvarez, J.C.; Lamas, J.; Yanez, A. [Departamento de Ingenieria Industrial II, Universidade da Coruna, Mendizabal s/n, Ferrol E-15403 (Spain)

2009-03-01

The abrasive nature of the mechanical processes involved in mining and mineral industry often causes significant wear to the associated equipment and derives non-negligible economic costs. One of the possible strategies to improve the wear resistance of the various components is the deposition of hardfacing layers on the bulk parts. The use of high power lasers for hardfacing (laser cladding) has attracted a great attention in the last decade as an alternative to other more standard methods (arc welding, oxy-fuel gas welding, thermal spraying). In laser cladding the hardfacing material is used in powder form. For high hardness applications Ni-, Co- or Fe-based alloys containing hard phase carbides at different ratios are commonly used. Tungsten carbides (WC) can provide coating hardness well above 1000 HV (Vickers). In this respect, commercially available WC powders normally contain spherical micro-particles consisting of crushed WC agglomerates. Some years ago, Spherotene powders consisting of spherical-fused monocrystaline WC particles, being extremely hard, between 1800 and 3000 HV, were patented. Very recently, mixtures of Ni-based alloy with Spherotene powders optimized for laser processing were presented (Technolase). These mixtures have been used in our study. Laser cladding tests with these powders were performed on low carbon steel (C25) substrates, and results in terms of microstructure and hardness will be discussed.

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.

Explosion Clad for Upstream Oil and Gas Equipment

International Nuclear Information System (INIS)

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 CO 2 and/or H 2 S 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.

Effect of Cobalt on Microstructure and Wear Resistance of Ni-Based Alloy Coating Fabricated by Laser Cladding

Directory of Open Access Journals (Sweden)

Kaiming Wang

2017-12-01

Full Text Available Ni-based alloy powders with different contents of cobalt (Co have been deposited on a 42CrMo steel substrate surface using a fiber laser. The effects of Co content on the microstructure, composition, hardness, and wear properties of the claddings were studied by scanning electron microscopy (SEM, an electron probe microanalyzer (EPMA, X-ray diffraction (XRD, a hardness tester, and a wear tester. The results show that the phases in the cladding layers are mainly γ, M7(C, B3, M23(C, B6, and M2B. With the increase in Co content, the amounts of M7(C, B3, M23(C, B6, and M2B gradually decrease, and the width of the eutectic structure in the cladding layer also gradually decreases. The microhardness decreases but the wear resistance of the cladding layer gradually improves with the increase of Co content. The wear resistance of the NiCo30 cladding layer is 3.6 times that of the NiCo00 cladding layer. With the increase of Co content, the wear mechanism of the cladding layer is changed from abrasive wear to adhesive wear.

Suppression of dilution in Ni-Cr-Si-B alloy cladding layer by controlling diode laser beam profile

Science.gov (United States)

Tanigawa, Daichi; Funada, Yoshinori; Abe, Nobuyuki; Tsukamoto, Masahiro; Hayashi, Yoshihiko; Yamazaki, Hiroyuki; Tatsumi, Yoshihiro; Yoneyama, Mikio

2018-02-01

A Ni-Cr-Si-B alloy layer was produced on a type 304 stainless steel plate by laser cladding. In order to produce cladding layer with smooth surface and low dilution, influence of laser beam profile on cladding layer was investigated. A laser beam with a constant spatial intensity at the focus spot was used to suppress droplet formation during the cladding layer formation. This line spot, formed with a focussing unit designed by our group, suppressed droplet generation. The layer formed using this line spot with a constant spatial intensity had a much smoother surface compared to a layer formed using a line spot with a Gaussian-like beam. In addition, the dilution of the former layer was much smaller. These results indicated that a line spot with a constant spatial intensity was more effective in producing a cladding layer with smooth surface and low dilution because it suppressed droplet generation.

The wetting of cladding materials and other metals and alloys by sodium

International Nuclear Information System (INIS)

Hodkin, E.N.; Nicholas, M.G.

1976-05-01

The sessile drop technique has been used to investigate the wetting behaviour between sodium and various metals and alloys including FV548, 316L, M316 and PE16. Unoxidised smooth surfaces of these alloys were not wetted by sodium containing 20 ppm of oxygen at temperatures below 300 0 C but were well wetted with advancing contact angles of 20 0 or less at temperatures of 550 0 to 600 0 C. Cold working and surface roughness had little effect on wetting behaviour but other factors exercised significant influences. Chemically or electrolytically polished M316 and PE16 surfaces were less readily wetted than those which had been prepared by mechanical polishing. In general, preoxidation of the alloy surfaces and increased oxygen contamination of the sodium had detrimental effects on wetting behaviour. On the other hand, increasing the chromium content of the alloys, decreasing the oxygen content of the sodium or ion bombarding the alloy sample surfaces had beneficial effects. Auger spectroscopy studies revealed a correlation between the chromium/oxygen ratio of PE16 surfaces and their wettability. The implications of this and other factors on fast reactor coolant/clad wetting behaviour is discussed. (author)

Corrosion fatigue crack growth in clad low-alloy steels: Part 1, medium-sulfur forging steel

International Nuclear Information System (INIS)

James, L.A.; Poskie, T.J.; Auten, T.A.; Cullen, W.H.

1996-01-01

Corrosion fatigue crack propagation tests were conducted on a medium- sulfur ASTM A508-2 forging steel overlaid with weld-deposited Alloy EN82H cladding. The specimens featured semi-elliptical surface cracks penetrating approximately 6.3 mm of cladding into the underlying steel. The initial crack sizes were relatively large with surface lengths of 30.3--38.3 mm, and depths of 13.1--16.8 mm. The experiments were conducted in a quasi-stagnant low-oxygen (O 2 < 10 ppb) aqueous environment at 243 degrees C, under loading conditions (ΔK, R, and cyclic frequency) conductive to environmentally-assisted cracking (EAC) in higher-sulfur steels under quasi-stagnant conditions. Earlier experiments on unclad compact tension specimens of this heat of steel did not exhibit EAC, and the present experiments on semi-elliptical surface cracks penetrating cladding also did not exhibit EAC

Vapor corrosion of aluminum cladding alloys and aluminum-uranium fuel materials in storage environments

International Nuclear Information System (INIS)

Lam, P.; Sindelar, R.L.; Peacock, H.B. Jr.

1997-04-01

An experimental investigation of the effects of vapor environments on the corrosion of aluminum spent nuclear fuel (A1 SNF) has been performed. Aluminum cladding alloys and aluminum-uranium fuel alloys have been exposed to environments of air/water vapor/ionizing radiation and characterized for applications to degradation mode analysis for interim dry and repository storage systems. Models have been developed to allow predictions of the corrosion response under conditions of unlimited corrodant species. Threshold levels of water vapor under which corrosion does not occur have been identified through tests under conditions of limited corrodant species. Coupons of aluminum 1100, 5052, and 6061, the US equivalent of cladding alloys used to manufacture foreign research reactor fuels, and several aluminum-uranium alloys (aluminum-10, 18, and 33 wt% uranium) were exposed to various controlled vapor environments in air within the following ranges of conditions: Temperature -- 80 to 200 C; Relative Humidity -- 0 to 100% using atmospheric condensate water and using added nitric acid to simulate radiolysis effects; and Gamma Radiation -- none and 1.8 x 10 6 R/hr. The results of this work are part of the body of information needed for understanding the degradation of the A1 SNF waste form in a direct disposal system in the federal repository. It will provide the basis for data input to the ongoing performance assessment and criticality safety analyses. Additional testing of uranium-aluminum fuel materials at uranium contents typical of high enriched and low enriched fuels is being initiated to provide the data needed for the development of empirical models

Research Progress on Laser Cladding Amorphous Coatings on Metallic Substrates

Directory of Open Access Journals (Sweden)

CHEN Ming-hui

2017-01-01

Full Text Available The microstructure and property of amorphous alloy as well as the limitations of the traditional manufacturing methods for the bulk amorphous alloy were briefly introduced in this paper.Combined with characteristics of the laser cladding technique,the research status of the laser cladding Fe-based,Zr-based,Ni-based,Cu-based and Al-based amorphous coatings on the metal substrates were mainly summarized.The effects of factors such as laser processing parameter,micro-alloying element type and content and reinforcing phase on the laser cladding amorphous coatings were also involved.Finally,the main problems and the future research directions of the composition design and control of the laser-cladded amorphous coating,the design and optimization of the laser cladding process,and the basic theory of the laser cladding amorphous coatings were also put forward finally.

Microstructure and interfacial evaluation of Co-based alloy coating on copper by pulsed Nd:YAG multilayer laser cladding

International Nuclear Information System (INIS)

Yan Hua; Wang Aihua; Xu Kaidong; Wang Wenyan; Huang Zaowen

2010-01-01

Laser cladding defect-free coatings on copper is rather difficult. The purpose of this study is to fabricate high quality Co-based alloy coating on copper substrate by laser cladding. Powder preplacement with a thickness of 0.7 mm improves the absorptivity of copper substrate to laser effectively and generates defect-free coating. Microstructures, phase constitutions and wear properties are investigated by means of scanning electronic microscopy (SEM) with X-ray energy dispersive microanalysis (EDX), transmission electron microscopy (TEM) and X-ray diffraction (XRD), as well as dry sliding wear test. Experimental results show that α-Co solution, Cr 23 C 6 , Ni 17 W 3 and Cr 4 Ni 15 W are the main phases in the Co-based coating. The Ni-based solid solutions (α-Co, Ni) and (Ni, Cu) are formed at interface, which generate metallurgical bonding by diffusion between Co-based coating and copper substrate. The average microhardness of the coating is 478HV 0.1 . Wear resistance of copper is significantly improved by laser cladding Co-based alloy multilayer coating.

Study of the uniform corrosion of an aluminium alloy used for the fuel cladding of the Jules Horowitz experimental reactor

International Nuclear Information System (INIS)

Wintergerst, M.

2008-01-01

For the Jules Horowitz new material testing reactor, an aluminium base alloy, AlFeNi, will be used for the cladding of the fuel plates. Taking into account the thermal properties of the alloy and of its oxide, the corrosion of the fuel cans presents many problems. The aim of this thesis is to provide a growing kinetic of the oxide layer at the surface of the AlFeNi fuel can in order to predict the life time of fuel element. Thus the mechanism of degradation of the cladding will be describe in order to integrate the different parameters of the operating reactor. (A.L.B.)

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.

The elastic properties of zirconium alloy fuel cladding and pressure tubing materials

International Nuclear Information System (INIS)

Rosinger, H.E.; Northwood, D.O.

1979-01-01

A knowledge of the elastic properties of zirconium alloys is required in the mathematical modelling of cladding and pressure tubing performance. Until recently, little of this type of data was available, particularly at elevated temperatures. The dynamic elastic moduli of zircaloy-2, zircaloy-4, the alloys Zr-1.0 wt%Nb, Zr-2.5 wt%Nb and Marz grade zirconium have therefore been determined over the temperature range 275 to 1000 K. Young's modulus and shear modulus for all the zirconium alloys decrease with temperature and are expressed by empirical relations fitted to the data. The elastic properties are texture dependent and a detailed study has been conducted on the effect of texture on the elastic properties of Zr-1.0 wt% Nb over the temperature range 275 to 775 K. The results are compared with polycrystalline elastic constants computed from single crystal elastic constants, and the effect of texture on the dynamic elastic moduli is discussed in detail. (Auth.)

Optimization of pulsed TIG cladding process of stellite alloy on carbon steel using RSM

Energy Technology Data Exchange (ETDEWEB)

Madadi, F., E-mail: f.madadi@ma.iut.ac.ir [Department of Materials Engineering, Isfahan University of Technology, Isfahan 8415683111 (Iran, Islamic Republic of); Ashrafizadeh, F. [Department of Materials Engineering, Isfahan University of Technology, Isfahan 8415683111 (Iran, Islamic Republic of); Shamanian, M., E-mail: shamanian@cc.iut.ac.ir [Department of Materials Engineering, Isfahan University of Technology, Isfahan 8415683111 (Iran, Islamic Republic of)

2012-01-05

Highlights: > This study is useful to optimize the welding process variables in order to control the heat input and cooling rates such that the hardness and dilution of the clad could be estimated. > Central composite rotatable design technique with five-level, four-factor full-factorial design matrix and mathematical models was used to predict hardness and dilution of pulsed gas tungsten arc weld cladding of stellite6 on carbon steel with high accuracy. > The welding current is an effective parameter affecting heat input and melting. In this regard, it is the most important process parameter which influences the dilution. Increase welding current leads to increase in dilution percentage and vice versa. The effect of percentage on time is less important when compared to the other factors. > The results predicted by mathematical models were close to those obtained by experiments. The confirmation tests also indicated high correlation between the mentioned values. > All of the chosen pulse GTAW parameters were significant and showed a noticeable influence on clad dilution. - Abstract: Stellite 6 is a cobalt-base alloy which is resistant to wear and corrosion and retains these properties at high temperatures. The exceptional wear resistance of Stellite 6 is mainly due to the unique inherent characteristics of the hard carbides dispersed in a Co-Cr alloy matrix. In this study, pulsed tungsten inert gas (TIG) cladding process was carried out to deposit Stellite 6 on plain carbon steel plate. The beneficial effects of this cladding process are low heat input, low distortion, controlled weld bead volume, less hot cracking tendency, less absorption of gases by weld pool and better control of the fusion zone. The dilution effect is a key issue in the quality of cladded layers and, in this regard, the pulsed current tungsten inert gas (PCTIG) was performed to decrease excess heat input and melting of substrate. This paper deals with the investigation of the hardness and

Application of Coating Technology for Accident Tolerant Fuel Cladding

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jeong-Yong; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

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.

General-Purpose Heat Source development: Safety Verification Test Program. Bullet/fragment test series

Energy Technology Data Exchange (ETDEWEB)

George, T.G.; Tate, R.E.; Axler, K.M.

1985-05-01

The radioisotope thermoelectric generator (RTG) that will provide power for space missions contains 18 General-Purpose Heat Source (GPHS) modules. Each module contains four /sup 238/PuO/sub 2/-fueled clads and generates 250 W/sub (t)/. Because a launch-pad or post-launch explosion is always possible, we need to determine the ability of GPHS fueled clads within a module to survive fragment impact. The bullet/fragment test series, part of the Safety Verification Test Plan, was designed to provide information on clad response to impact by a compact, high-energy, aluminum-alloy fragment and to establish a threshold value of fragment energy required to breach the iridium cladding. Test results show that a velocity of 555 m/s (1820 ft/s) with an 18-g bullet is at or near the threshold value of fragment velocity that will cause a clad breach. Results also show that an exothermic Ir/Al reaction occurs if aluminum and hot iridium are in contact, a contact that is possible and most damaging to the clad within a narrow velocity range. The observed reactions between the iridium and the aluminum were studied in the laboratory and are reported in the Appendix.

Development of low-Cr ODS FeCrAl alloys for accident-tolerant fuel cladding

Science.gov (United States)

Dryepondt, Sebastien; Unocic, Kinga A.; Hoelzer, David T.; Massey, Caleb P.; Pint, Bruce A.

2018-04-01

Low-Cr oxide dispersion strengthened (ODS) FeCrAl alloys were developed as accident tolerant fuel cladding because of their excellent oxidation resistance at very high temperature, high strength and improved radiation tolerance. Fe-12Cr-5Al wt.% gas atomized powder was ball milled with Y2O3+FeO, Y2O3+ZrO2 or Y2O3+TiO2, and the resulting powders were extruded at 950 °C. The resulting fine grain structure, particularly for the Ti and Zr containing alloys, led to very high strength but limited ductility. Comparison with variants of commercial PM2000 (Fe-20Cr-5Al) highlighted the significant impact of the powder consolidation step on the alloy grain size and, therefore, on the alloy mechanical properties at T < 500 °C. These low-Cr compositions exhibited good oxidation resistance at 1400 °C in air and steam for 4 h but could not form a protective alumina scale at 1450 °C, similar to observations for fine grained PM2000 alloys. The effect of alloy grain size, Zr and Ti additions, and impurities on the alloy mechanical and oxidation behaviors are discussed.

Clad vent set cup open end (closure weld zone) wall-thickness study

Energy Technology Data Exchange (ETDEWEB)

Ulrich, G.B.; Sherrill, M.W.

1994-09-01

The wall thickness at the open end of Clad Vent Set (CVS) cups is a very important parameter for maintaining control of the fueled CVS closure weld process. Ideally, the wall thickness in the closure weld zone should be constant. The DOP-26 iridium alloy is very difficult to machine; therefore, key dimensional features are established during the two-draw warm-forming operation. Unfortunately, anisotropy in the forming blanks produces four ears at the open end of each cup. Formation of these ears produces axial and circumferential variations in wall thickness. The cup certification requirement is that the wall thickness in the closure weld zone, defined as the 2.5-mm band at the open end of a cup, measure from 0.63 to 0.73 mm. The wall thickness certification data for the open end of the CVS cups have been statistically evaluated. These data show that the cups recently produced for the Cassini mission have well-controlled open-end wall thicknesses.

Performance of IN-706 and PE-16 cladding in mixed-oxide fuel pins

International Nuclear Information System (INIS)

Makenas, B.J.; Lawrence, L.A.; Jensen, B.W.

1982-05-01

Iron-nickel base, precipitation-strengthened alloys, IN-706 and PE-16, advanced alloy cladding considered for breeder reactor applications, were irradiated in mixed-oxide fuel pins in the HEDL-P-60 subassembly in EBR-II. Initial selection of candidate advanced alloys was done using only nonfueled materials test results. However, to establish the performance characteristics of the candidate cladding alloys, i.e., dimensional stability and structural integrity under conditions of high neutron flux, elevated temperature, and applied stress, it was necessary to irradiate fuel pins under typical operating conditions. Fuel pins were clad with solution treated IN-706 and PE-16 and irradiated to peak fluences of 6.1 x 10 22 n/cm 2 (E > .1 MeV) and 8.8 x 10 22 n/cm 2 (E > .1 MeV) respectively. Fabrication and operating parameters for the fuel pins with the advanced cladding alloy candidates are summarized. Irradiation of HEDL-P-60 was interrupted with the breach of a pin with IN-706 cladding at 5.1 at % and the test was terminated with cladding breach in a pin with PE-16 cladding at 7.6 at %

Nonequilibrium synthesis of NbAl3 and Nb-Al-V alloys by laser cladding. II - Oxidation behavior

Science.gov (United States)

Haasch, R. T.; Tewari, S. K.; Sircar, S.; Loxton, C. M.; Mazumder, J.

1992-01-01

Isothermal oxidation behaviors of NbAl3 alloy synthesized by laser cladding were investigated at temperatures between 800 and 1400 C, and the effect of vanadium microalloying on the oxidation of the laser-clad alloy was examined. The oxidation kinetics of the two alloys were monitored using thermal gravimetric weight gain data, and the bulk and surface chemistries were analyzed using XRD and XPS, respectively. It was found that NbAl3 did not form an exclusive layer of protective Al2O3. The oxidation products at 800 C were found to be a mixture of Nb2O5 and Al2O3. At 1200 C, a mixture of NbAlO4, Nb2O5, and Al2O3 formed; and at 1400 C, a mixture of NbAlO4, Al2O3, NbO2, NbO(2.432), and Nb2O5 formed. The addition of V led to a dramatic increase of the oxidation rate, which may be related to the formation of (Nb, V)2O5 and VO2, which grows in preference to protective Al2O3.

High-temperature air oxidation of E110 and Zr-1%Nb alloys claddings with coatings

International Nuclear Information System (INIS)

Kuprin, A.S.; Belous, V.A.; Voyevodin, V.N.; Bryk, V.V.; Vasilenko, R.L.; Ovcharenko, V.D.; Tolmachova, G.N.; V'yugov, P.N.

2014-01-01

Results of experimental study of the influence of protective vacuum-arc claddings on the base of compounds zirconium-chromium and of its nitrides on air oxidation resistance at temperatures 660, 770, 900, 1020, 1100 deg C during 3600 s. of tubes produced of zirconium alloys E110 and Zr-1%Nb (calcium-thermal alloy of Ukrainian production) are presented. Change of hardness, the width of oxide layer and depth of oxygen penetration into alloys from the side of coating and without coating are investigated by the methods of nanoindentation and by scanning electron microscopy. It is shown that the thickness of oxide layer in zirconium alloys at temperatures 1020 and 1100 deg C from the side of the coating doesn't exceed 5 μm, and from the unprotected side reaches the value of ≥ 120 μm with porous and rough structure. Tubes with coatings save their shape completely independently of the type of alloy; tubes without coatings deform with the production of through cracks

Study and development of an Iridium-192 seed for use in ophthalmic cancer

International Nuclear Information System (INIS)

Mattos, Fabio Rodrigues de

2013-01-01

Even ocular tumors are not among the cases with a higher incidence, they affect the population, especially children. The Institute of Energy and Nuclear Research (IPEN-CNEN/SP) in partnership with Escola Paulista de Medicina (UNIFESP), created a project to develop and implement a alternative treatment for ophthalmic cancer that use brachytherapy iridium-192 seeds. The project arose by reason of the Escola Paulista treat many cancer cases within the Unified Health System (SUS) and the research experience of sealed radioactive sources group at IPEN. The methodology was developed from the available infrastructure and the experience of researchers. The prototype seed presents with a core (192-iridium alloy of iridium-platinum) of 3.0 mm long sealed by a capsule of titanium of 0.8 mm outside diameter, 0.05 mm wall thickness and 4,5mm long. This work aims to study and develop a seed of iridium-192 from a platinum-iridium alloy. No study on the fabrication of these seeds was found in available literature. It was created a methodology that involved: characterization of the material used in the core, creation of device for neutron activation irradiation and and seed sealing tests. As a result, proved the feasibility of the method. As a suggestion for future work, studies regarding metrology and dosimetry of these sources and improvement of the methodology should be carried out, for future implementation in national scope. (author)

Nuclear fuel cladding material

International Nuclear Information System (INIS)

Nakahigashi, Shigeo.

1982-01-01

Purpose: To largely improve the durability and the safety of fuel cladding material. Constitution: Diffusion preventive layers, e.g., aluminum or the like are covered on both sides of a zirconium alloy base layer of thin material, and corrosion resistant layers, e.g., copper or the like are covered thereon. This thin plate material is intimately wound in a circularly tubular shape in a plurality of layers to form a fuel cladding tube. With such construction, corrosion of the tube due to fuel and impurity can be prevented by the corrosion resistant layers, and the diffusion of the corrosion resistant material to the zirconium alloy can be prevented by the diffusion preventive layers. Since a plurality of layers are cladded, even if the corrosion resistant layers are damaged or cracked due to stress corrosion, only one layer is damaged or cracked, but the other layers are not affected. (Sekiya, K.)

BISON Fuel Performance Analysis of FeCrAl cladding with updated properties

Energy Technology Data Exchange (ETDEWEB)

Sweet, Ryan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, Nathan M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wirth, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-08-30

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling the integral thermo-mechanical performance of FeCrAl-cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl, and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and

Nuclear-powered pacemaker fuel cladding study

International Nuclear Information System (INIS)

Shoup, R.L.

1976-07-01

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 238 PuO 2 -powered pacemaker could be transformed into a terrorism weapon

Oxidation Behavior of FeCrAl -coated Zirconium Cladding prepared by Laser Coating

Energy Technology Data Exchange (ETDEWEB)

Kim, Il-Hyun; Kim, Hyun-Gil; Choi, Byung-Kwan; Park, Jeong-Yong; Koo, Yang-Hyun; Kim, Jin-Seon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

From the recent research trends, the ATF cladding concepts for enhanced accident tolerance are divided as follows: Mo-Zr cladding to increase the high temperature strength, cladding coating to increase the high temperature oxidation resistance, FeCrAl alloy and SiC/SiCf material to increase the oxidation resistance and strength at high temperature. 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. A laser coating method supplied with FeCrAl powders was developed to decrease the high-temperature oxidation rate in a steam environment through a systematic study for various coating parameters, and a FeCrAl-coated Zircaloy-4 cladding tube of 100 mm in length to the axial direction can be successfully manufactured.

Cladding failure margins for metallic fuel in the integral fast reactor

International Nuclear Information System (INIS)

Bauer, T.H.; Fenske, G.R.; Kramer, J.M.

1987-01-01

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

Mechanical Property and Oxidation Behavior of ATF cladding developed in KAERI

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jung-Hwan; Park, Jeong-Yong; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

To realize the coating cladding, coating material (Cr-based alloy) as well as coating technology (3D laser coating and arc ion plating combined with vacuum annealing) can be developed to meet the fuel cladding criteria. The coated Zr cladding can be produced after the optimization of coating technologies. The coated cladding sample showed the good oxidation/corrosion and adhesion properties without the spalling and/or severe interaction with the Zr alloy cladding from the various tests. Thus, it is known that the mechanical property and oxidation behavior of coated cladding concept developed in KAERI is reasonable for applying the ATF cladding in LWRs. At the present time various ATF concepts have been proposed and developing in many countries. The ATF concepts with potentially improved accident performance can be summarized to the coating cladding, Mo-Zr cladding, FeCrAl cladding, and SiCf/SiC cladding. Regarding the cladding performance, ATF cladding concepts will be evaluated with respect to the accident scenarios and normal operations of LWRs as well as to the fuel cladding fabrication.

A Eutectic Melting Study of Double Wall Cladding Tubes of FeCrAl and Zircaloy-4

Energy Technology Data Exchange (ETDEWEB)

Lim, Woojin; Son, Seongmin; Lee, You Ho; Lee, Jeong Ik; Ryu, Ho Jin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Jeong, Eun [Kyunghee University, Yongin (Korea, Republic of)

2015-10-15

The eutectic melting behavior of FeCrAl/Zircaloy-4 double wall cladding tubes was investigated by annealing at various temperatures ranging from 900 .deg. C to 1300 .deg. C. It was found that significant eutectic melting occurred after annealing at temperatures equal to or higher than 1150 .deg. C. It means that an additional diffusion barrier layer is necessary to limit the eutectic melting between FeCrAl and Zircaloy-4 alloy cladding tubes. Coating of FeCrAl layers on the Zr alloy cladding tube is being investigated for the development of accident tolerant fuel by exploiting of both the oxidation resistance of FeCrAl alloys and the neutronic advantages of Zr alloys. Coating of FeCrAl alloys on Zr alloy cladding tubes can be performed by various techniques including thermal spray, laser cladding, and co-extrusion. Son et al. also reported the fabrication of FeCrAl/Zr ally double wall cladding by the shrink fit method. For the double layered cladding tubes, the thermal expansion mismatch between the dissimilar materials, severe deformation or mechanical failure due to the evolution of thermal stresses can occur when there is a thermal cycling. In addition to the thermal stress problems, chemical compatibilities between the two different alloys should be investigated in order to check the stability and thermal margin of the double wall cladding at a high temperature. Generally, it is considered that Zr alloy cladding will maintain its mechanical integrity up to 1204 .deg. C (2200 .deg. F) to satisfy the acceptance criteria for emergency core cooling systems.

Dry sliding wear behavior of laser clad TiVCrAlSi high entropy alloy coatings on Ti–6Al–4V substrate

International Nuclear Information System (INIS)

Huang, Can; Zhang, Yongzhong; Vilar, Rui; Shen, Jianyun

2012-01-01

Highlights: ► TiVCrAlSi high entropy alloy coatings were obtained on Ti–6Al–4V by laser cladding. ► (Ti,V) 5 Si 3 forms because the formation is accompanied of large variation on enthalpy. ► Wear resistance of Ti–6Al–4V is improved by laser cladding with TiVCrAlSi. ► The wear mechanism is investigated. -- Abstract: Approximately equimolar ratio TiVCrAlSi high entropy alloy coatings has been deposited by laser cladding on Ti–6Al–4V alloy. The analysis of the microstructure by scanning electron microscopy (SEM) shows that the coating is metallurgically bonded to the substrate. X-ray diffraction (XRD) and energy dispersive spectrometer (EDS) analyses show that TiVCrAlSi coating is composed of precipitates of (Ti,V) 5 Si 3 dispersed in a body-centered cubic (BCC) matrix. Intermetallic compound (Ti,V) 5 Si 3 forms because the formation is accompanied by larger variation on enthalpy, which may offset the entropy term. The dry sliding wear tests show that the wear resistance of Ti–6Al–4V is improved by laser cladding with TiVCrAlSi. The enhancement of the wear resistance is explained by the presence of the hard silicide phase dispersed in a relatively ductile BCC matrix, which allows sliding wear to occur in the mild oxidative regime for a wide range of testing conditions.

Microstructural study of the interface in laser-clad Ni-Al bronze on Al alloy AA333 and its relation to cracking

Science.gov (United States)

Liu, Y.; Mazumder, J.; Shibata, K.

1995-06-01

The interface toughness between a laser clad and the substrate determines whether the cladding is useful for engineering application. The objective of this investigation is to correlate the interface properties of laser-clad Ni-AI bronze on Al alloy AA333 with the microstructure and crystal structure of the interface. Scanning electron microscopy (SEM) and transmission electron microscopy (TEM) combined with energy-dispersive X-ray spectroscopy (EDX) are used to examine the interface. In a good clad track, the interface is an irregular curved zone with a varying width (occasionally keyholing structure) from 30 to 150 μm. A compositional transition from the Cu-rich clad (83 wt pct Cu) to the Al-rich substrate (3.2 wt pct Cu) occurs across this interface. Three phases in the interface are identified in TEM: Al solid solution, θ phase, and γ1 phase, as described in the Cu-Al binary phase diagram. In a good clad track, the θ and γ1 phases are distributed in the Al solid solution. In a clad track with cracks, the interface structure spreads to a much larger scale from 300 μm to the whole clad region. Large areas of θ and γ1 phases are observed. The mechanism of cracking at the interface is related to the formation of a twophase region of θ and γ1 phases. To understand the microstructure, a nonequilibrium quasibinary Cu-Al phase diagram is proposed and compared with the equilibrium binary Cu-Al phase diagram. It is found that the occurrence of many phases such as η1η2, ζ1, ζ2, ɛ1, ɛ2, γ0, β0, and β, as described in the equilibrium binary Cu-Al phase diagram, is suppressed by either the cladding process or by the alloying elements. The three identified phases (Al solid solution, θ phase, and γ1, phase) showed significant extension of solubility.

Balancing activity, stability and conductivity of nanoporous core-shell iridium/iridium oxide oxygen evolution catalysts

Energy Technology Data Exchange (ETDEWEB)

Kim, Yong-Tae; Lopes, Pietro Papa; Park, Shin-Ae; Lee, A-Yeong; Lim, Jinkyu; Lee, Hyunjoo; Back, Seoin; Jung, Yousung; Danilovic, Nemanja; Stamenkovic, Vojislav; Erlebacher, Jonah; Snyder, Joshua; Markovic, Nenad M.

2017-11-13

The selection of oxide materials for catalyzing the Oxygen Evolution Reaction in acid-based electrolyzers must be guided by the proper balance between activity, stability and conductivity – a challenging mission of great importance for delivering affordable and environmentally friendly hydrogen. Here we report that the highly conductive nanoporous architecture of an iridium oxide shell on a metallic iridium core, formed through the fast dealloying of osmium from an Ir25Os75 alloy, exhibits an exceptional balance between oxygen evolution activity and stability as quantified by the Activity-Stability FactorASF. Based on this metric, the nanoporous Ir/IrO2 morphology of dealloyed Ir25Os75 shows a factor of ~30 improvement ASFrelative to conventional Ir-based oxide materials and a ~8 times improvement over dealloyed Ir25Os75 nanoparticles due to optimized stability and conductivity, respectively. We propose that the Activity-Stability FactorASF is the key “metric” for determining the technological relevance of oxide-based anodic water electrolyzer catalysts.

Delayed hydride cracking of zirconium alloy fuel cladding

International Nuclear Information System (INIS)

2010-10-01

This report describes the work performed in a coordinated research project on Hydrogen and Hydride Degradation of the Mechanical and Physical Properties of Zirconium Alloys. It is the second in the series. In 2005-2009 that work was extended within a new CRP called Delayed Hydride Cracking in Zirconium Alloy Fuel Cladding. The project consisted of adding hydrogen to samples of Zircaloy-4 claddings representing light water reactors (LWRs), CANDU and Atucha, and measuring the rates of delayed hydride cracking (DHC) under specified conditions. The project was overseen by a supervisory group of experts in the field who provided advice and assistance to participants as required. All of the research work undertaken as part of the CRP is described in this report, which includes details of the experimental procedures that led to a consistent set of data for LWR cladding. The participants and many of their co-workers in the laboratories involved in the CRP contributed results and material used in this report, which compiles the results, their analysis, discussions of their interpretation and conclusions and recommendations for future work. The research was coordinated by an advisor and by representatives in three laboratories in industrialized Member States. Besides the basic goal to transfer the technology of the testing technique from an experienced laboratory to those unfamiliar with the methods, the CRP was set up to harmonize the experimental procedures to produce consistent sets of data, both within a single laboratory and between different laboratories. From the first part of this project it was demonstrated that by following a standard set of experimental protocols, consistent results could be obtained. Thus, experimental vagaries were minimized by careful attention to detail of microstructure, temperature history and stress state in the samples. The underlying idea for the test programme was set out at the end of the first part of the project on pressure tubes. The

Laser Cladding of Ultra-Thin Nickel-Based Superalloy Sheets

Directory of Open Access Journals (Sweden)

Tobias Gabriel

2017-03-01

Full Text Available Laser cladding is a well-established process to apply coatings on metals. However, on substrates considerably thinner than 1 mm it is only rarely described in the literature. In this work 200 µm thin sheets of nickel-based superalloy 718 are coated with a powder of a cobalt-based alloy, Co–28Cr–9W–1.5Si, by laser cladding. The process window is very narrow, therefore, a precisely controlled Yb fiber laser was used. To minimize the input of energy into the substrate, lines were deposited by setting single overlapping points. In a design of experiments (DoE study, the process parameters of laser power, laser spot area, step size, exposure time, and solidification time were varied and optimized by examining the clad width, weld penetration, and alloying depth. The microstructure of the samples was investigated by optical microscope (OM and scanning electron microscopy (SEM, combined with electron backscatter diffraction (EBSD and energy dispersive X-ray spectroscopy (EDX. Similarly to laser cladding of thicker substrates, the laser power shows the highest influence on the resulting clad. With a higher laser power, the clad width and alloying depth increase, and with a larger laser spot area the weld penetration decreases. If the process parameters are controlled precisely, laser cladding of such thin sheets is manageable.

Laser Cladding of Ultra-Thin Nickel-Based Superalloy Sheets.

Science.gov (United States)

Gabriel, Tobias; Rommel, Daniel; Scherm, Florian; Gorywoda, Marek; Glatzel, Uwe

2017-03-10

Laser cladding is a well-established process to apply coatings on metals. However, on substrates considerably thinner than 1 mm it is only rarely described in the literature. In this work 200 µm thin sheets of nickel-based superalloy 718 are coated with a powder of a cobalt-based alloy, Co-28Cr-9W-1.5Si, by laser cladding. The process window is very narrow, therefore, a precisely controlled Yb fiber laser was used. To minimize the input of energy into the substrate, lines were deposited by setting single overlapping points. In a design of experiments (DoE) study, the process parameters of laser power, laser spot area, step size, exposure time, and solidification time were varied and optimized by examining the clad width, weld penetration, and alloying depth. The microstructure of the samples was investigated by optical microscope (OM) and scanning electron microscopy (SEM), combined with electron backscatter diffraction (EBSD) and energy dispersive X-ray spectroscopy (EDX). Similarly to laser cladding of thicker substrates, the laser power shows the highest influence on the resulting clad. With a higher laser power, the clad width and alloying depth increase, and with a larger laser spot area the weld penetration decreases. If the process parameters are controlled precisely, laser cladding of such thin sheets is manageable.

High temperature steam oxidation of Al3Ti-based alloys for the oxidation-resistant surface layer on Zr fuel claddings

International Nuclear Information System (INIS)

Park, Jeong-Yong; Kim, Il-Hyun; Jung, Yang-Il; Kim, Hyun-Gil; Park, Dong-Jun; Choi, Byung-Kwon

2013-01-01

We investigated the feasibility to apply Al 3 Ti-based alloys as the surface layer for improving the oxidation resistance of Zr fuel claddings under accident conditions. Two types of Al 3 Ti-based alloys with the compositions of Al–25Ti–10Cr and Al–21Ti–23Cr in atomic percent were prepared by arc-melting followed by homogenization annealing at 1423 K for 48 h. Al–25Ti–10Cr alloy showed an L1 2 quasi-single phase microstructure with a lot of needle-shaped minor phase and pores. Al–21Ti–23Cr alloy consisted of an L1 2 matrix and Cr 2 Al as the second phase. Al 3 Ti-based alloys showed an extremely low oxidation rate in a 1473 K steam for up to 7200 s when compared to Zircaloy-4. Both alloys exhibited almost the same oxidation rate in the early stage of oxidation, but Al–25Ti–10Cr showed a little lower oxidation rate after 4000 s than Al–21Ti–23Cr. The difference in the oxidation rate between two types of Al 3 Ti-based alloys was too marginal to distinguish the oxidation behavior of each alloy. The resultant oxide exhibited almost the same characteristics in both alloys even though the microstructure was explicitly distinguished from each other. The crystal structure of the oxide formed up to 2000 s was identified as Al 2 O 3 in both alloys. The oxide morphology consisted of columnar grains whose length was almost identical to the average oxide thickness. On the basis of the results obtained, it is considered that Al 3 Ti-based alloy is one of the promising candidates for the oxidation-resistant surface layer on Zr fuel claddings

Influence of processing variables and alloy chemistry on the corrosion behavior of ZIRLO nuclear fuel cladding

International Nuclear Information System (INIS)

Comstock, R.J.; Sabol, G.P.; Schoenberger, G.

1996-01-01

Variations in the thermal heat treatments used during the fabrication of ZIRLO (Zr-1Nb-1Sn-0.1Fe) fuel clad tubing and in ZIRLO alloy chemistry were explored to develop a further understanding of the relationship between processing, microstructure, and cladding corrosion performance. Heat treatment variables included intermediate tube annealing temperatures as well as a beta-phase heat treatment during the latter stages of the tube reduction schedule. Chemistry variables included deviations in niobium and tin content from the nominal composition. The effects of both heat treatment and chemistry on corrosion behavior were assessed by autoclave tests in both pure and lithiated water and high-temperature steam. Analytical electron microscopy demonstrated that the best out-reactor corrosion performance is obtained for microstructures containing a fine distribution of beta-niobium and Zr-Nb-Fe particles. Deviations from this microstructure, such as the presence of beta-zirconium phase, tend to degrade corrosion resistance. ZIRLO fuel cladding was irradiated in four commercial reactors. In all cases, the microstructure in the cladding included beta-niobium and Zr-Nb-Fe particles. ZIRLO fuel cladding processed with a late-stage beta heat treatment to further refine the second-phase particle size exhibited in-reactor corrosion behavior that was similar to reference ZIRLO cladding. Variations of the in-reactor corrosion behavior of ZIRLO were correlated to tin content, with higher oxide thickness observed in the ZIRLO cladding containing higher tin. The results of these studies indicate that optimum corrosion performance of ZIRLO is achieved by maintaining a uniform distribution of fine second-phase particles and controlled levels of tin

Iridium-192 sources production for brachytherapy use

International Nuclear Information System (INIS)

Rostelato, Maria Elisa Chuery Martins

1997-01-01

The incidence of cancer increases every year in Brazil and turns out to be one of the most important causes of mortality. Some of the patients are treated with brachytherapy, a form of lesion treatment which is based on the insertion of sources into tumors, in this particular case, activated iridium wires. During this process, the ionizing radiation efficiently destroys the malignant cells. These iridium wires have a nucleus made out of an iridium-platinum alloy 20-30/70-80 of 0,1 mm in diameter either coated by platinum or encased in a platinum tube. The technique consists in irradiating the wire in the reactor neutron flux in order to produce iridium-192. The linear activity goes from 1 mCi/cm to 4 mCi/cm and the basic characteristic, which is required, is the homogeneity of the activation along the wire. It should not present a dispersion exceeding 5% on a wire measuring 50 cm in length, 0.5 mm or 0.3 mm in diameter. Several experiments were carried out in order to define the activation parameters. Wires from different origins were analyzed. It was concluded that United States of America and France wires were found to be perfectly adequate for brachytherapy purposes and have therefore been sent to specialized hospitals and successfully applied to cancer patients. Considering that the major purpose of this work is to make this product more accessible in Brazil, at a cost reflecting the Brazilian reality, the IPEN is promoting the preparation of iridium-192 sources to be used in brachytherapy, on a national level. (author)

The quest for safe and reliable fuel cladding materials

Energy Technology Data Exchange (ETDEWEB)

Pino, Eddy S.; Abe, Alfredo Y., E-mail: eddypino132@hotmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (POLI/USP), Sao Paulo, SP (Brazil). Lab. de Analise, Avaliacao e Gerenciamento de Risco

2015-07-01

The tragic Fukushima Daiichi Nuclear Plant accident of March, 2011, has brought great unrest and challenge to the nuclear industry, which, in collaboration with universities and nuclear research institutes, is making great efforts to improve the safety in nuclear reactors developing accident tolerant fuels (ATF). This involves the study of different materials to be applied as cladding and, also, the improvement in the fuel properties in order to enhance the fuel performance and safety, specifically under accident conditions. Related to the cladding, iron based alloys and silicon carbide (SiC) materials have been studied as a good alternative. In the case of austenitic stainless steel, there is the advantage that the austenitic stainless steel 304 was used as cladding material in the first PWR (Pressurized Water Reactor) registering a good performance. Then, alternated cladding materials such as iron based alloys (304, 310, 316, 347) should be used to replace the zirconium-based alloys in order to improve safety. In this paper, these cladding materials are evaluated in terms of their physical and chemical properties; among them, strength and creep resistance, thermal conductivity, thermal stability and corrosion resistance. Additionally, these properties are compared with those of conventional zirconium-based alloys, the most used material in actual PWR, to assess the advantages and disadvantages of each material concerning to fuel performance and safety contribution. (author)

The quest for safe and reliable fuel cladding materials

International Nuclear Information System (INIS)

Pino, Eddy S.; Abe, Alfredo Y.; Giovedi, Claudia

2015-01-01

The tragic Fukushima Daiichi Nuclear Plant accident of March, 2011, has brought great unrest and challenge to the nuclear industry, which, in collaboration with universities and nuclear research institutes, is making great efforts to improve the safety in nuclear reactors developing accident tolerant fuels (ATF). This involves the study of different materials to be applied as cladding and, also, the improvement in the fuel properties in order to enhance the fuel performance and safety, specifically under accident conditions. Related to the cladding, iron based alloys and silicon carbide (SiC) materials have been studied as a good alternative. In the case of austenitic stainless steel, there is the advantage that the austenitic stainless steel 304 was used as cladding material in the first PWR (Pressurized Water Reactor) registering a good performance. Then, alternated cladding materials such as iron based alloys (304, 310, 316, 347) should be used to replace the zirconium-based alloys in order to improve safety. In this paper, these cladding materials are evaluated in terms of their physical and chemical properties; among them, strength and creep resistance, thermal conductivity, thermal stability and corrosion resistance. Additionally, these properties are compared with those of conventional zirconium-based alloys, the most used material in actual PWR, to assess the advantages and disadvantages of each material concerning to fuel performance and safety contribution. (author)

Development of an Iridium-192 seed for use in ophthalmic brachytherapy

International Nuclear Information System (INIS)

Mattos, Fabio R.; Rostelato, Maria Elisa C.M.; Zeituni, Carlos; Moura, Joao A.; Costa, Osvaldo L.; Feher, Anselmo; Moura, Eduardo S.; Souza, Carla D.; Peleias Junior, Fernando S.

2011-01-01

The Institute for Energy and Nuclear Research (IPEN), in partnership with the School or Medicine (UNIFESP), created a project that aims to develop and implement an ophthalmic therapeutic treatment for cancer with Iridium-192 seeds. The School of Medicine treats many cancer cases in the SUS (Brazilian Public Health System), and brachytherapy group of IPEN has extensive experience in prototype sources. The seed to be manufactured will perform as follows: a core of Iridium-192 is packaged inside small cylindrical seeds consist of a titanium capsule of 0.8 mm outer diameter, 0.05 mm wall thickness and 4 5 mm in length. The core is an alloy of platinum-iridium (20/80) of 3.0 mm in length and 0.3 mm in diameter. Material analysis, neutron activation and activity measurements were carried out. (author)

Modification of tribology and high-temperature behavior of Ti-48Al-2Cr-2Nb intermetallic alloy by laser cladding

International Nuclear Information System (INIS)

Liu Xiubo; Wang Huaming

2006-01-01

In order to improve the tribology and high-temperature oxidation properties of the Ti-48Al-2Cr-2Nb intermetallic alloy simultaneously, mixed NiCr-Cr 3 C 2 precursor powders had been investigated for laser cladding treatment to modify wear and high-temperature oxidation resistance of the material. The alloy samples were pre-placed with NiCr-80, 50 and 20%Cr 3 C 2 (wt.%), respectively, and laser treated at the same parameters, i.e., laser output power 2.8 kW, beam scanning speed 2.0 mm/s, beam dimension 1 mm x 18 mm. The treated samples underwent tests of microhardness, wear and high-temperature oxidation. The results showed that laser cladding with different constitution of mixed precursor NiCr-Cr 3 C 2 powders improved surface hardness in all cases. Laser cladding with NiCr-50%Cr 3 C 2 resulted in the best modification of tribology and high-temperature oxidation behavior. X-ray diffraction (XRD), optical microscope (OM), scanning electron microscopy (SEM) and energy-dispersive spectrometer (EDS) analyses indicated that the formation of reinforced Cr 7 C 3 , TiC and both continuous and dense Al 2 O 3 , Cr 2 O 3 oxide scales were supposed to be responsible for the modification of the relevant properties. As a result, the present work had laid beneficial surface engineering foundation for TiAl alloy applied as future light weight and high-temperature structural candidate materials

Creep behavior under internal pressure of zirconium alloy cladding oxidized in steam at high temperature

International Nuclear Information System (INIS)

Chosson, Raphael

2014-01-01

During hypothetical Loss-Of-Coolant-Accident (LOCA) scenarios, zirconium alloy fuel cladding tubes creep under internal pressure and are oxidized on their outer surface at high temperature (HT). Claddings become stratified materials: zirconia and oxygen-stabilized α phase, called α(O), are formed on the outer surface of the cladding whereas the inner part remains in the β domain. The strengthening effect of oxidation on the cladding creep behavior under internal pressure has been highlighted at HT. In order to model this effect, the creep behavior of each layer had to be determined. This study focused on the characterization of the creep behavior of the α(O) phase at HT, through axial creep tests performed under vacuum on model materials, containing from 2 to 7 wt.% of oxygen and representative of the α(O) phase. For the first time, two creep flow regimes have been observed in this phase. Underlying physical mechanisms and relevant microstructural parameters have been discussed for each regime. The strengthening effect due to oxygen on the α(O) phase creep behavior at HT has been quantified and creep flow equations have been identified. A ductile to brittle transition criterion has been also suggested as a function of temperature and oxygen content. Relevance of the creep flow equations for each layer, identified in this study or from the literature, has been discussed. Then, a finite element model, describing the oxidized cladding as a stratified material, has been built. Based on this model, a fraction of the experimental strengthening during creep is predicted. (author) [fr

High-resolution characterization of oxidation mechanism of zirconium nuclear fuel cladding alloys

International Nuclear Information System (INIS)

Hu, J.; Lozano-Perez, S.; Grovenor, C.

2015-01-01

Full text of publication follows. Zirconium alloys are used extensively as cladding materials in modern light water reactors to separate the uranium dioxide (UO 2 ) fuel rods and the coolant water in order to prevent the escape of radioactive fission products whilst maintaining heat transfer to the coolant. With increasing demand for high burn-up in modern nuclear reactors, environmental degradation of these alloys is now the life limiting factor for fuel assemblies. As part of the MUZIC-2 collaboration studying oxidation and hydrogen pickup in Zr alloys, several high resolution analysis techniques have been used to study the microstructure of a range of commercial and developmental Zr alloys. The sample used for this investigation was prepared from a Westinghouse TM developmental alloy with composition of Zr-0.9Nb-0.01Sn-0.08Fe (wt %) in the recrystallized condition. The sample was oxidised in an autoclave at EDF Energy under simulated PWR water conditions at 360 C. degrees for 360 days. Using Transmission Electron Microscope (TEM), we have studied the development of the equiaxed-columnar-equiaxed grain structure, and observe that the columnar grains are both longer and show a stronger preferred texture in more corrosion-resistant alloys. Fresnel imaging revealed the existence of both parallel interconnected pores and some vertically interconnected pores along the columnar oxide grain boundaries, which become more disconnected near the metal-oxide interface. Electron Energy Loss Spectroscopy (EELS) provided accurate quantitative analysis of the oxygen concentration across the interface, identifying the existence of local regions of stoichiometric ZrO and Zr 3 O 2 with varying thickness. These observations will be discussed in the context of current models for oxidation in zirconium alloys. (authors)

Corrosion characteristics of K-claddings

International Nuclear Information System (INIS)

Park, J. Y.; Choi, B. K.; Jung, Y. H.; Jung, Y. H.

2004-01-01

The Improvement of the corrosion resistance of nuclear fuel claddings is the critical issue for the successful development of the high burn-up fuel. KAERI have developed the K-claddings having a superior corrosion resistance by controlling the alloying element addition and optimizing the manufacturing process. The comparative evaluation of the corrosion resistance for K-claddings and the foreign claddings was performed and the effect of the heat treatment on the corrosion behavior of K-claddings was also examined. Corrosion tests were carried out in the conditions of 360 .deg. C pure water, PWR-simulating loop and 400 .deg. C steam, From the results of the corrosion tests, it was found that the corrosion resistance of K-claddings is superior to those of Zry4 and A claddings and K6 showed a better corrosion resistance than K3. The corrosion behavior of K-cladding was strongly influenced by the final annealing rather than the intermediate annealing, and the corrosion resistance increased with decreasing the final annealing temperature

PCI resistant light water reactor fuel cladding

International Nuclear Information System (INIS)

Foster, J.P.; Sabol, G.P.

1988-01-01

A tubular nuclear fuel element cladding tube is described, the fuel element cladding tube forming the entire fuel element cladding and consisting of: a single continuous wall, the single continuous wall consisting of a single alloy selected from the group consisting of zirconium base alloys, A, B, C, D, and E; the single continuous wall characterized by a cold worked and stress relieved microstructure throughout; wherein the zirconium base alloy A contains 0.2 - 0.6 w/o Sn, 0.03 - 0.11 w/o sum of Fe and Cr, section 600 ppm O and section 1500 ppm total impurities; the zirconium base alloy B contains 0.1 - 0.6 w/oo Sn, 0.04 - 0.24 w/o Fe, 0.05 - 0.15 w/o Cr, section 0.08 w/o Ni, section 600 ppm O and section 1500 ppm total impurities; the zirconium base alloy C contains 1.2 - 1.7 w/o Sn, 0.04 - 0.24 w/o Fe, 0.05 - 0.15 w/o Cr, section 0.08 w/o Ni, section 600 ppm O, and section 1500 ppm total impurities; the zirconium base alloy D contains 0.15 - 0.6 w/o Sn, 0.15 - 0.5 w/o Fe, section 600 ppm O, and section 1500 ppm total impurities; and the zirconium base alloy E contains 0.4 - 0.6 w/o Sn, 0.1 - 0.3 w/o Fe, 0.03 - 0.07 w/o Ni, section 600 ppm O, and section 1500 ppm total impurities

Investigation and basic evaluation for ultra-high burnup fuel cladding material

International Nuclear Information System (INIS)

Ioka, Ikuo; Nagase, Fumihisa; Futakawa, Masatoshi; Kiuchi, Kiyoshi

2001-03-01

In ultra-high burnup of the power reactor, it is an essential problem to develop the cladding with excellent durability. First, development history and approach of the safety assessment of Zircaloy for the high burnup fuel were summarized in the report. Second, the basic evaluation and investigation were carried out on the material with high practicability in order to select the candidate materials for the ultra-high burnup fuel. In addition, the basic research on modification technology of the cladding surface was carried out from the viewpoint of the addition of safety margin as a cladding. From the development history of the zirconium alloy including the Zircaloy, it is hard to estimate the results of in-pile test from those of the conventional corrosion test (out-pile test). Therefore, the development of the new testing technology that can simulate the actual environment and the elucidation of the corrosion-controlling factor of the cladding are desired. In cases of RIA (Reactivity Initiated Accident) and LOCA (Loss of Coolant Accident), it seems that the loss of ductility in zirconium alloys under heavy irradiation and boiling of high temperature water restricts the extension of fuel burnup. From preliminary evaluation on the high corrosion-resistance materials (austenitic stainless steel, iron or nickel base superalloys, titanium alloy, niobium alloy, vanadium alloy and ferritic stainless steel), stabilized austenitic stainless steels with a capability of future improvement and high-purity niobium alloys with a expectation of the good corrosion resistance were selected as candidate materials of ultra-high burnup cladding. (author)

Analyses on Silicide Coating for LOCA Resistant Cladding

Energy Technology Data Exchange (ETDEWEB)

Sweidan, Faris B.; Lee, You Ho; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

2015-10-15

A particular focus of accident-tolerant fuel has been cladding due to the rapid high-temperature oxidation of zirconium-based cladding with the evolution of H2 when steam is a reactant. Some key features of the coated cladding include high-temperature resistance to oxidation, lower processing temperatures, and a high melting point of the coating. Zirconium alloys exhibit a reasonably high melting temperature, so a coating for the cladding is appealing if the coating increases the high-temperature resistance to oxidation. In this case, the cladding is protected from complete oxidation. The cladding coating involves the application of zirconium silicide onto Zr-based cladding. Zirconium silicide coating is expected to produce a glassy layer that becomes more protective at elevated temperature. For this reason, silicide coatings on cladding offer the potential for improved reliability at normal operating temperatures and at the higher transient temperatures encountered during accidents. Although ceramic coatings are brittle and may have weak points to be used as coating materials, several ceramic coatings were successful and showed adherent behavior and high resistance to oxidation. In this study, the oxidation behavior of zirconium silicide and its oxidation kinetics are analyzed. Zirconium silicide is a new suggested material to be used as coatings on existing Zr-based cladding alloys, the aim of this study is to evaluate if zirconium silicide is applicable to be used, so they can be more rapidly developed using existing cladding technology with some modifications. These silicide coatings are an attractive alternative to the use of coatings on zirconium claddings or to the lengthy development of monolithic ceramic or ceramic composite claddings and coatings.

Analyses on Silicide Coating for LOCA Resistant Cladding

International Nuclear Information System (INIS)

Sweidan, Faris B.; Lee, You Ho; Ryu, Ho Jin

2015-01-01

A particular focus of accident-tolerant fuel has been cladding due to the rapid high-temperature oxidation of zirconium-based cladding with the evolution of H2 when steam is a reactant. Some key features of the coated cladding include high-temperature resistance to oxidation, lower processing temperatures, and a high melting point of the coating. Zirconium alloys exhibit a reasonably high melting temperature, so a coating for the cladding is appealing if the coating increases the high-temperature resistance to oxidation. In this case, the cladding is protected from complete oxidation. The cladding coating involves the application of zirconium silicide onto Zr-based cladding. Zirconium silicide coating is expected to produce a glassy layer that becomes more protective at elevated temperature. For this reason, silicide coatings on cladding offer the potential for improved reliability at normal operating temperatures and at the higher transient temperatures encountered during accidents. Although ceramic coatings are brittle and may have weak points to be used as coating materials, several ceramic coatings were successful and showed adherent behavior and high resistance to oxidation. In this study, the oxidation behavior of zirconium silicide and its oxidation kinetics are analyzed. Zirconium silicide is a new suggested material to be used as coatings on existing Zr-based cladding alloys, the aim of this study is to evaluate if zirconium silicide is applicable to be used, so they can be more rapidly developed using existing cladding technology with some modifications. These silicide coatings are an attractive alternative to the use of coatings on zirconium claddings or to the lengthy development of monolithic ceramic or ceramic composite claddings and coatings

The effect of particle size on the heat affected zone during laser cladding of Ni-Cr-Si-B alloy on C45 carbon steel

Science.gov (United States)

Tanigawa, Daichi; Abe, Nobuyuki; Tsukamoto, Masahiro; Hayashi, Yoshihiko; Yamazaki, Hiroyuki; Tatsumi, Yoshihiro; Yoneyama, Mikio

2018-02-01

Laser cladding is one of the most useful surface coating methods for improving the wear and corrosion resistance of material surfaces. Although the heat input associated with laser cladding is small, a heat affected zone (HAZ) is still generated within the substrate because this is a thermal process. In order to reduce the area of the HAZ, the heat input must therefore be reduced. In the present study, we examined the effects of the powdered raw material particle size on the heat input and the extent of the HAZ during powder bed laser cladding. Ni-Cr-Si-B alloy layers were produced on C45 carbon steel substrates in conjunction with alloy powders having average particle sizes of 30, 40 and 55 μm, while measuring the HAZ area by optical microscopy. The heat input required for layer formation was found to decrease as smaller particles were used, such that the HAZ area was also reduced.

Cladding tube manufacturing technology

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

This report gives an overview of the manufacturing routine of PWR cladding tubes. The routine essentially consists of a series of deformation and annealing processes which are necessary to transform the ingot geometry to tube dimensions. By changing shape, microstructure and structure-related properties are altered simultaneously. First, a short overview of the basics of that part of deformation geometry is given which is related to tube reducing operations. Then those processes of the manufacturing routine which change the microstructure are depicted, and the influence of certain process parameters on microstructure and material properties are shown. The influence of the resulting microstructure on material properties is not discussed in detail, since it is described in my previous report 'Alloy Development for High Burnup Cladding.' Because of their paramount importance still up to now, and because manufacturing data and their influence on properties for other alloys are not so well established or published, the descriptions are mostly related to Zry4 tube manufacturing, and are only in short for other alloys. (author). 9 refs., 46 figs.

A preliminary study of factors affecting the calibration stability of the iridium versus iridium-40 percent rhodium thermocouple

Science.gov (United States)

Ahmed, Shaffiq; Germain, Edward F.; Daryabeigi, Kamran; Alderfer, David W.; Wright, Robert E.

1987-01-01

An iridium versus iridium-40% rhodium thermocouple was studied. Problems associated with the use of this thermocouple for high temperature applications (up to 2000 C) were investigated. The metallurgical studies included X-ray, macroscopic, resistance, and metallographic studies. The thermocouples in the as-received condition from the manufacturer revealed large amounts of internal stress caused by cold working during manufacturing. The thermocouples also contained a large amount of inhomogeneities and segregations. No phase transformations were observed in the alloy up to 1100 C. It was found that annealing the thermocouple at 1800 C for two hours, and then at 1400 C for 2 to 3 hours yielded a fine grain structure, relieving some of the strains, and making the wire more ductile. It was also found that the above annealing procedure stabilized the thermal emf behavior of the thermocouple for application below 1800 C (an improvement from + or - 1% to + or - 0.02% within the range of the test parameters used).

Characterisation of AGR fuel cladding alloy using secondary ion mass spectrometry

International Nuclear Information System (INIS)

Allen, G.C.; Sparry, R.P.; Wild, R.K.

1987-08-01

Uranium dioxide fuel used in the Advanced Gas Cooled Reactor (AGR) is contained in a ribbed can of 20wt%Cr/25wt%Ni/Nb stabilised steel. Laboratory circumstances, spall during thermal cycling. To date it has been difficult to identify active material originating from the oxidation product of the cladding alloy in the cooling circuit. In an attempt to solve this problem we have set out to characterise fully a sample of oxide from this source and work is in progress to obtain suitable oxide samples from the surface of a 20%Cr/25%Ni/Nb stainless steel. In view of its high sensitivity and the ability to obtain chemical information from relatively small areas we have sought to use Secondary Ion Mass Spectroscopy (SIMS). (author)

Laser Cladding of Composite Bioceramic Coatings on Titanium Alloy

Science.gov (United States)

Xu, Xiang; Han, Jiege; Wang, Chunming; Huang, Anguo

2016-02-01

In this study, silicon nitride (Si3N4) and calcium phosphate tribasic (TCP) composite bioceramic coatings were fabricated on a Ti6Al4V (TC4) alloy using Nd:YAG pulsed laser, CO2 CW laser, and Semiconductor CW laser. The surface morphology, cross-sectional microstructure, mechanical properties, and biological behavior were carefully investigated. These investigations were conducted employing scanning electron microscope, energy-dispersive x-ray spectroscopy, and other methodologies. The results showed that both Si3N4 and Si3N4/TCP composite coatings were able to form a compact bonding interface between the coating and the substrate by using appropriate laser parameters. The coating layers were dense, demonstrating a good surface appearance. The bioceramic coatings produced by laser cladding have good mechanical properties. Compared with that of the bulk material, microhardness of composite ceramic coatings on the surface significantly increased. In addition, good biological activity could be obtained by adding TCP into the composite coating.

Accident tolerant fuel cladding development: Promise, status, and challenges

Science.gov (United States)

Terrani, Kurt A.

2018-04-01

The motivation for transitioning away from zirconium-based fuel cladding in light water reactors to significantly more oxidation-resistant materials, thereby enhancing safety margins during severe accidents, is laid out. A review of the development status for three accident tolerant fuel cladding technologies, namely coated zirconium-based cladding, ferritic alumina-forming alloy cladding, and silicon carbide fiber-reinforced silicon carbide matrix composite cladding, is offered. Technical challenges and data gaps for each of these cladding technologies are highlighted. Full development towards commercial deployment of these technologies is identified as a high priority for the nuclear industry.

Laser Cladding of Ti-6Al-4V Alloy with Ti-Al2O3 Coating for Biomedical Applications

Science.gov (United States)

Mthisi, A.; Popoola, A. P. I.; Adebiyi, D. I.; Popoola, O. M.

2018-05-01

The indispensable properties of Ti-6Al-4V alloy coupled with poor tribological properties and delayed bioactivity make it a subject of interest to explore in biomedical application. A quite number of numerous coatings have been employed on titanium alloys, with aim to overcome the poor properties exhibited by this alloy. In this work, the possibility of laser cladding different ad-mixed powders (Ti - 5 wt.% Al2O3 and Ti - 8wt.% Al2O3) on Ti-6Al-4V at various laser scan speed (0.6 and 0.8 m/min) were investigated. The microstructure, phase constituents and corrosion of the resultant coatings were characterized by scanning electron microscope (SEM), Optical microscope, X-Ray diffractometer (XRD) and potentiostat respectively. The electrochemical behaviour of the produced coatings was studied in a simulated body fluid (Hanks solution). The microstructural results show that a defect free coating is achieved at low scan speed and ad-mixed of Ti-5 wt. % Al2O3. Cladding of Ti - Al2O3 improved the corrosion resistance of Ti-6Al-4V alloy regardless of varying neither scan speed nor ad-mixed percentage. However, Ti-5 wt.% Al2O3 coating produced at low scan speed revealed the highest corrosion resistance among the coatings due to better quality coating layer. Henceforth, this coating may be suitable for biomedical applications.

Stress-corrosion cracking properties of candidate fuel cladding alloys for the Canadian SCWR: a summary of literature data and recent test results

Energy Technology Data Exchange (ETDEWEB)

Zheng, W.; Zeng, Y., E-mail: Wenyue@NRcan.gc.ca [CanmetMATERIALS, Hamilton, ON (Canada); Luo, J. [Univ. of Alberta, Edmonton, AB (Canada); Novotny, R. [JRC-European Commission, Patten (Netherlands); Li, J.; Amirkhiz, B.S., E-mail: Jian.li@nrcan.gc.ca [CanmetMATERIALS, Hamilton, ON (Canada); Guzonas, D. [Atomic Energy of Canada Limited, Chalk River, ON (Canada); Matchim, M.; Collier, J.; Yang, L., E-mail: lin.yang@nrcan.gc.ca [CanmetMATERIALS, Hamilton, ON (Canada)

2014-07-01

Cracking of fuel claddings is a serious concern when selecting candidate alloys for the development of a next-generation reactor. Whether the cracking is due to an environment-metal interaction such as stress-corrosion, or a pure metallurgical process such as localized plastic deformation along grain boundaries, the final impact is the same: cracking of the cladding can lead to fuel failure. In the course of a review of potential candidate alloys in preparation for further assessment under conditions relevant to the Canadian SCWR concept, relevant cracking studies reported for five short-listed alloys (namely 310S, 347H, 800H, 625 and 214) in the open literature were examined, and the key findings are provided in this paper. Discussions are also made of the recent SCC data from capsule tests and slow-strain rate tests (SSRT) in supercritical water. The data suggest that there is a threshold strain level below which SCC is not developed during SSRT tests. The practical implication of this finding is also discussed. (author)

ZIRCONIUM-CLADDING OF THORIUM

Science.gov (United States)

Beaver, R.J.

1961-11-21

A method of cladding thorium with zirconium is described. The quality of the bond achieved between thorium and zirconium by hot-rolling is improved by inserting and melting a thorium-zirconium alloy foil between the two materials prior to rolling. (AEC)

Adaptation of fuel code for light water reactor with austenitic steel rod cladding

International Nuclear Information System (INIS)

Gomes, Daniel de Souza; Silva, Antonio Teixeira; Giovedi, Claudia

2015-01-01

Light water reactors were used with steel as nuclear fuel cladding from 1960 to 1980. The high performance proved that the use of low-carbon alloys could substitute the current zirconium alloys. Stainless steel is an alternative that can be used as cladding. The zirconium alloys replaced the steel. However, significant experiences in-pile occurred, in commercial units such as Haddam Neck, Indian Point, and Yankee experiences. Stainless Steel Types 347 and 348 can be used as cladding. An advantage of using Stainless Steel was evident in Fukushima when a large number of hydrogens was produced at high temperatures. The steel cladding does not eliminate the problem of accumulating free hydrogen, which can lead to a risk of explosion. In a boiling water reactor, environments easily exist for the attack of intergranular corrosion. The Stainless Steel alloys, Types 321, 347, and 348, are stabilized against attack by the addition of titanium, niobium, or tantalum. The steel Type 348 is composed of niobium, tantalum, and cobalt. Titanium preserves type 321, and niobium additions stabilize type 347. In recent years, research has increased on studying the effects of irradiation by fast neutrons. The impact of radiation includes changes in flow rate limits, deformation, and ductility. The irradiation can convert crystalline lattices into an amorphous structure. New proposals are emerging that suggest using a silicon carbide-based fuel rod cladding or iron-chromium-aluminum alloys. These materials can substitute the classic zirconium alloys. Once the steel Type 348 was chosen, the thermal and mechanical properties were coded in a library of functions. The fuel performance codes contain all features. A comparative analysis of the steel and zirconium alloys was made. The results demonstrate that the austenitic steel alloys are the viable candidates for substituting the zirconium alloys. (author)

Adaptation of fuel code for light water reactor with austenitic steel rod cladding

Energy Technology Data Exchange (ETDEWEB)

Gomes, Daniel de Souza; Silva, Antonio Teixeira, E-mail: dsgomes@ipen.br, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (POLI/USP), Sao Paulo, SP (Brazil). Lab. de Analise, Avaliacao e Gerenciamento de Risco

2015-07-01

Light water reactors were used with steel as nuclear fuel cladding from 1960 to 1980. The high performance proved that the use of low-carbon alloys could substitute the current zirconium alloys. Stainless steel is an alternative that can be used as cladding. The zirconium alloys replaced the steel. However, significant experiences in-pile occurred, in commercial units such as Haddam Neck, Indian Point, and Yankee experiences. Stainless Steel Types 347 and 348 can be used as cladding. An advantage of using Stainless Steel was evident in Fukushima when a large number of hydrogens was produced at high temperatures. The steel cladding does not eliminate the problem of accumulating free hydrogen, which can lead to a risk of explosion. In a boiling water reactor, environments easily exist for the attack of intergranular corrosion. The Stainless Steel alloys, Types 321, 347, and 348, are stabilized against attack by the addition of titanium, niobium, or tantalum. The steel Type 348 is composed of niobium, tantalum, and cobalt. Titanium preserves type 321, and niobium additions stabilize type 347. In recent years, research has increased on studying the effects of irradiation by fast neutrons. The impact of radiation includes changes in flow rate limits, deformation, and ductility. The irradiation can convert crystalline lattices into an amorphous structure. New proposals are emerging that suggest using a silicon carbide-based fuel rod cladding or iron-chromium-aluminum alloys. These materials can substitute the classic zirconium alloys. Once the steel Type 348 was chosen, the thermal and mechanical properties were coded in a library of functions. The fuel performance codes contain all features. A comparative analysis of the steel and zirconium alloys was made. The results demonstrate that the austenitic steel alloys are the viable candidates for substituting the zirconium alloys. (author)

Hydrogenation and high temperature oxidation of Zirconium claddings

International Nuclear Information System (INIS)

Novotny, T.; Perez-Feró, E.; Horváth, M.

2015-01-01

In the last few years a new series of experiments started for supporting the new LOCA criteria, considering the proposals of US NRC. The effects which can cause the embrittlement of VVER fuel claddings were reviewed and evaluated in the framework of the project. The purpose of the work was to determine how the fuel cladding’s hydrogen uptake under normal operating conditions, effect the behavior of the cladding under LOCA conditions. As a first step a gas system equipment with gas valves and pressure gauge was built, in which the zirconium alloy can absorb hydrogen under controlled conditions. In this apparatus E110 (produced by electrolytic method, currently used at Paks NPP) and E110G (produced by a new technology) alloys were hydrogenated to predetermined hydrogen contents. According the results of ring compression tests the E110G alloys lose their ductility above 3200 ppm hydrogen content. This limit can be applied to determine the ductile-brittle transition of the nuclear fuel claddings. After the hydrogenation, high temperature oxidation experiments were carried out on the E110G and E110 samples at 1000 °C and 1200 °C. 16 pieces of E110G and 8 samples of E110 with 300 ppm and 600 ppm hydrogen content were tested. The oxidation of the specimens was performed in steam, under isothermal conditions. Based on the ring compression tests load-displacement curves were recorded. The main objective of the compression tests was to determine the ductile-brittle transition. These results were compared to the results of our previous experiments where the samples did not contain hydrogen. The original claddings showed more ductile behavior than the samples with hydrogen content. The higher hydrogen content resulted in a more brittle mechanical behavior. However no significant difference was observed in the oxidation kinetics of the same cladding types with different hydrogen content. The experiments showed that the normal operating hydrogen uptake of the fuel claddings

Current status of materials development of nuclear fuel cladding tubes for light water reactors

International Nuclear Information System (INIS)

Duan, Zhengang; Yang, Huilong; Satoh, Yuhki; Murakami, Kenta; Kano, Sho; Zhao, Zishou; Shen, Jingjie; Abe, Hiroaki

2017-01-01

Zirconium-based (Zr-based) alloys have been widely used as materials for the key components in light water reactors (LWRs), such as fuel claddings which suffer from waterside corrosion, hydrogen uptakes and strength loss at elevated temperature, especially during accident scenarios like the lost-of-coolant accident (LOCA). For the purpose of providing a safer, nuclear leakage resistant and economically viable LWRs, three general approaches have been proposed so far to develop the accident tolerant fuel (ATF) claddings: optimization of metallurgical composition and processing of Zr-based alloys, coatings on existing Zr-based alloys and replacement of current Zr-based alloys. In this manuscript, an attempt has been made to systematically present the historic development of Zr-based cladding, including the impacts of alloying elements on the material properties. Subsequently, the research investigations on coating layer on the surface of Zr-based claddings, mainly referring coating materials and fabrication methods, have been broadly reviewed. The last section of this review provides the introduction to alternative materials (Non-Zr) to Zr-based alloys for LWRs, such as advanced steels, Mo-based, and SiC-based materials.

Current status of materials development of nuclear fuel cladding tubes for light water reactors

Energy Technology Data Exchange (ETDEWEB)

Duan, Zhengang, E-mail: duan_zg@imr.tohoku.ac.jp [Department of Quantum Science and Energy Engineering, Graduate School of Engineering, Tohoku University, Sendai, Miyagi 980-8577 (Japan); Yang, Huilong [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan); Satoh, Yuhki [Institute for Materials Research, Tohoku University, Sendai, Miyagi 980-8577 (Japan); Murakami, Kenta; Kano, Sho; Zhao, Zishou; Shen, Jingjie [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan); Abe, Hiroaki, E-mail: abe.hiroaki@n.t.u-tokyo.ac.jp [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan)

2017-05-15

Zirconium-based (Zr-based) alloys have been widely used as materials for the key components in light water reactors (LWRs), such as fuel claddings which suffer from waterside corrosion, hydrogen uptakes and strength loss at elevated temperature, especially during accident scenarios like the lost-of-coolant accident (LOCA). For the purpose of providing a safer, nuclear leakage resistant and economically viable LWRs, three general approaches have been proposed so far to develop the accident tolerant fuel (ATF) claddings: optimization of metallurgical composition and processing of Zr-based alloys, coatings on existing Zr-based alloys and replacement of current Zr-based alloys. In this manuscript, an attempt has been made to systematically present the historic development of Zr-based cladding, including the impacts of alloying elements on the material properties. Subsequently, the research investigations on coating layer on the surface of Zr-based claddings, mainly referring coating materials and fabrication methods, have been broadly reviewed. The last section of this review provides the introduction to alternative materials (Non-Zr) to Zr-based alloys for LWRs, such as advanced steels, Mo-based, and SiC-based materials.

Effect of high hydrogen content on metallurgical and mechanical properties of zirconium alloy claddings after heat-treatment at high temperature

International Nuclear Information System (INIS)

Turque, Isabelle

2016-01-01

Under hypothetical loss-of-coolant accident conditions, fuel cladding tubes made of zirconium alloys can be exposed to steam at high temperature (HT, up 1200 C) before being cooled and then quenched in water. In some conditions, after burst occurrence the cladding can rapidly absorb a significant amount of hydrogen (secondary hydriding), up to 3000 wt.ppm locally, during steam exposition at HT. The study deals with the effect, poorly studied up to date, of high contents of hydrogen on the metallurgical and mechanical properties of two zirconium alloys, Zircaloy-4 and M5, during and after cooling from high temperatures, at which zirconium is in its β phase. A specific facility was developed to homogeneously charge in hydrogen up to ∼ 3000 wt.ppm cladding tube samples of several centimeters in length. Phase transformations, chemical element partitioning and hydrogen precipitation during cooling from the β temperature domain of zirconium were studied by using several techniques, for the materials containing up to ∼ 3000 wt.ppm of hydrogen in average: in-situ neutron diffraction upon cooling from 700 C, X-ray diffraction, μ-ERDA, EPMA and electron microscopy in particular. The results were compared to thermodynamic predictions. In order to study the effect of high hydrogen contents on the mechanical behavior of the (prior-)μ phase of zirconium, axial tensile tests were performed at various temperatures between 20 and 700 C upon cooling from the β temperature domain, on samples with mean hydrogen contents up to ∼ 3000 wt.ppm. The results show that metallurgical and mechanical properties of the (prior-)β phase of zirconium alloys strongly depend on temperature and hydrogen content. (author) [fr

BOWIEITE: A NEW RHODIUM-IRIDIUM-PLATINUM SULFIDE IN PLATINUM-ALLOY NUGGETS, GOODNEWS BAY, ALASKA.

Science.gov (United States)

Desborough, George A.; Criddle, Alan J.

1984-01-01

Bowieite (Rh,Ir,Pt)//2S//3, a new mineral species, is found in three nuggets of platinum from Goodnews Bay, Alaska. In linearly polarized reflected light, and compared to the host, higher reflecting white platinum-iridium alloy, bowieite is pale gray to pale gray-brown; neither bireflectance nor reflectance pleochroism is apparent. With polars crossed, its anisotropic rotation tints vary from gray to dark brown. Luminance values (relative to the CIE illuminant C) for R//1 and R//2, computed from full spectral data for the most bireflectant grain, are 45. 8% and 48. 2% in air, and 30. 5% and 33. 0% in oil, respectively. VHN//1//0//0 1288 (858 to 1635). Bowieite is orthorhombic, space group Pnca, with a 8. 454(7) -8. 473(8), b 5. 995(1)-6. 002(7), c 6. 143(1)-6. 121(8) A, Z equals 4. Some grains that are 2. 6 to 3. 8 atomic % metal-deficient occur as an optically coherent rim on bowieite; the rim and the bowieite grain are not optically continuous.

Novel twin-roll-cast Ti/Al clad sheets with excellent tensile properties.

Science.gov (United States)

Kim, Dae Woong; Lee, Dong Ho; Kim, Jung-Su; Sohn, Seok Su; Kim, Hyoung Seop; Lee, Sunghak

2017-08-14

Pure Ti or Ti alloys are recently spot-lighted in construction industries because they have excellent resistance to corrosions, chemicals, and climates as well as various coloring characteristics, but their wide applications are postponed by their expensiveness and poor formability. We present a new fabrication process of Ti/Al clad sheets by bonding a thin Ti sheet on to a 5052 Al alloy melt during vertical-twin-roll casting. This process has merits of reduced production costs as well as improved tensile properties. In the as-twin-roll-cast clad sheet, the homogeneously cast microstructure existed in the Al alloy substrate side, while the Ti/Al interface did not contain any reaction products, pores, cracks, or lateral delamination, which indicated the successful twin-roll casting. When this sheet was annealed at 350 °C~600 °C, the metallurgical bonding was expanded by interfacial diffusion, thereby leading to improvement in tensile properties over those calculated by a rule of mixtures. The ductility was also improved over that of 5052-O Al alloy (25%) or pure Ti (25%) by synergic effect of homogeneous deformation due to excellent Ti/Al bonding. This work provides new applications of Ti/Al clad sheets to lightweight-alloy clad sheets requiring excellent formability and corrosion resistance as well as alloy cost saving.

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

International Nuclear Information System (INIS)

Huang Yongjun; Zeng Xiaoyan

2009-01-01

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.

The Effect of Peak Temperatures and Hoop Stresses on Hydride Reorientations of Zirconium Alloy Cladding Tubes under Interim Dry Storage Condition

International Nuclear Information System (INIS)

Cha, Hyun Jin; Jang, Ki Nam; Kim, Kyu Tae

2016-01-01

In this study, the effect of peak temperatures and hoop tensile stresses on hydride reorientation in cladding was investigated. It was shown that the 250ppm-H specimens generated larger radial hydride fractions and longer radial hydrides than the 500ppm-H ones. The precipitated hydride in radial direction severely degrades mechanical properties of spent fuel rod. Hydride reorientation is related to cladding material, cladding temperature, hydrogen contents, thermal cycling, hoop stress and cooling rate. US NRC established the regulation on cladding temperature during the dry storage, which is the maximum fuel cladding temperature should not exceed 400 .deg. C for all fuel burnups under normal conditions of storage. However, if it is proved that the best estimate cladding hoop stress is equal to or less than 90MPa for the temperature limit proposed, a higher short-term temperature limit is allowed for low burnup fuel. In this study, 250ppm and 500ppm hydrogen-charged Zr-Nb alloy cladding tubes were selected to evaluate the effect of peak temperatures and hoop tensile stresses on the hydride reorientation during the dry storage. In order to evaluate threshold stresses in relation to various peak temperatures, four peak temperatures of 250, 300, 350, and 400 .deg. C and three tensile hoop stresses of 80, 100, 120MPa were selected.

Performance of refractory alloy-clad fuel pins

International Nuclear Information System (INIS)

Dutt, D.S.; Cox, C.M.; Millhollen, M.K.

1984-12-01

This paper discusses objectives and basic design of two fuel-cladding tests being conducted in support of SP-100 technology development. Two of the current space nuclear power concepts use conventional pin type designs, where a coolant removes the heat from the core and transports it to an out-of-core energy conversion system. An extensive irradiation testing program was conducted in the 1950's and 1960's to develop fuel pins for space nuclear reactors. The program emphasized refractory metal clad uranium nitride (UN), uranium carbide (UC), uranium oxide (UO 2 ), and metal matrix fuels (UCZr and BeO-UO 2 ). Based on this earlier work, studies presented here show that UN and UO 2 fuels in conjunction with several refractory metal cladding materials demonstrated high potential for meeting space reactor requirements and that UC could serve as an alternative but higher risk fuel

Characterization of hard coatings produced by laser cladding using laser-induced breakdown spectroscopy technique

Energy Technology Data Exchange (ETDEWEB)

Varela, J.A.; Amado, J.M.; Tobar, M.J.; Mateo, M.P.; Yañez, A.; Nicolas, G., E-mail: gines@udc.es

2015-05-01

Highlights: • Chemical mapping and profiling by laser-induced breakdown spectroscopy (LIBS) of coatings produced by laser cladding. • Production of laser clads using tungsten carbide (WC) and nickel based matrix (NiCrBSi) powders. • Calibration by LIBS of hardfacing alloys with different WC concentrations. - Abstract: Protective coatings with a high abrasive wear resistance can be obtained from powders by laser cladding technique, in order to extend the service life of some industrial components. In this work, laser clad layers of self-fluxing NiCrBSi alloy powder mixed with WC powder have been produced on stainless steel substrates of austenitic type (AISI 304) in a first step and then chemically characterized by laser-induced breakdown spectroscopy (LIBS) technique. With the suitable laser processing parameters (mainly output power, beam scan speed and flow rate) and powders mixture proportions between WC ceramics and NiCrBSi alloys, dense pore free layers have been obtained on single tracks and on large areas with overlapped tracks. The results achieved by LIBS technique and applied for the first time to the analysis of laser clads provided the chemical composition of the tungsten carbides in metal alloy matrix. Different measurement modes (multiple point analyses, depth profiles and chemical maps) have been employed, demonstrating the usefulness of LIBS technique for the characterization of laser clads based on hardfacing alloys. The behavior of hardness can be explained by LIBS maps which evidenced the partial dilution of some WC spheres in the coating.

Experiments for evaluation of corrosion to develop storage criteria for interim dry storage of aluminum-alloy clad spent nuclear fuel

International Nuclear Information System (INIS)

Peacock, H.B.; Sindelar, R.L.; Lam, P.S.; Murphy, T.H.

1994-01-01

The technical bases for specification of limits to environmental exposure conditions to avoid excessive degradation are being developed for storage criteria for dry storage of highly-enriched, aluminum-clad spent nuclear fuels owned by the US Department of Energy. Corrosion of the aluminum cladding is a limiting degradation mechanism (occurs at lowest temperature) for aluminum exposed to an environment containing water vapor. Attendant radiation fields of the fuels can lead to production of nitric acid in the presence of air and water vapor and would exacerbate the corrosion of aluminum by lowering the pH of the water solution. Laboratory-scale specimens are being exposed to various conditions inside an autoclave facility to measure the corrosion of the fuel matrix and cladding materials through weight change measurements and metallurgical analysis. In addition, electrochemical corrosion tests are being performed to supplement the autoclave testing by measuring differences in the general corrosion and pitting corrosion behavior of the aluminum cladding alloys and the aluminum-uranium fuel materials in water solutions

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.

Dynamics of Gradient Bioceramic Composite Coating on Surface of Titanium Alloy by Wide-Band Laser Cladding

Institute of Scientific and Technical Information of China (English)

LIU Qi-bin; ZOU Long-jiang; ZHU Wei-dong; LI Hai-tao; DONG Chuang

2004-01-01

The gradient bioceramic coating was prepared on the surface of titanium alloy using wide-band laser cladding. The dynamics of gradient bioceramic composite coating containing hydroxyapatite (HA) prepared with mixture of CaHPO4*2H2O and CaCO3 under the condition of wide-band laser was studied theoretically. The corresponding mathematical model and its numerical solution were presented. The examination experiment showed that HA bioceramic composite coatings can be obtained by appropriately choosing wide-band laser cladding parameters. The microstructure and surface morphology of HA bioceramic coating were observed by SEM and X-ray diffraction. The experimental results showed that the bioceramic coating is composed of HA, β-TCP, CaO, CaTiO3 and TiO2. The surface of bioceramic coating takes coral-shaped structure or short-rod piled structure, which helps osteoblast grow into bioceramic and improves the biocompatibility.

A study on wear resistance and microcrack of the Ti3Al/TiAl + TiC ceramic layer deposited by laser cladding on Ti-6Al-4V alloy

International Nuclear Information System (INIS)

Li Jianing; Chen Chuanzhong; Squartini, Tiziano; He Qingshan

2010-01-01

Laser cladding of the Al + TiC alloy powder on Ti-6Al-4V alloy can form the Ti 3 Al/TiAl + TiC ceramic layer. In this study, TiC particle-dispersed Ti 3 Al/TiAl matrix ceramic layer on the Ti-6Al-4V alloy by laser cladding has been researched by means of X-ray diffraction, scanning electron microscope, electron probe micro-analyzer, energy dispersive spectrometer. The main difference from the earlier reports is that Ti 3 Al/TiAl has been chosen as the matrix of the composite coating. The wear resistance of the Al + 30 wt.% TiC and the Al + 40 wt.% TiC cladding layer was approximately 2 times greater than that of the Ti-6Al-4V substrate due to the reinforcement of the Ti 3 Al/TiAl + TiC hard phases. However, when the TiC mass percent was above 40 wt.%, the thermal stress value was greater than the materials yield strength limit in the ceramic layer, the microcrack was present and its wear resistance decreased.

Deformation behavior of laser welds in high temperature oxidation resistant Fe-Cr-Al alloys for fuel cladding applications

Science.gov (United States)

Field, Kevin G.; Gussev, Maxim N.; Yamamoto, Yukinori; Snead, Lance L.

2014-11-01

Ferritic-structured Fe-Cr-Al alloys are being developed and show promise as oxidation resistant accident tolerant light water reactor fuel cladding. This study focuses on investigating the weldability and post-weld mechanical behavior of three model alloys in a range of Fe-(13-17.5)Cr-(3-4.4)Al (wt.%) with a minor addition of yttrium using modern laser-welding techniques. A detailed study on the mechanical performance of bead-on-plate welds using sub-sized, flat dog-bone tensile specimens and digital image correlation (DIC) has been carried out to determine the performance of welds as a function of alloy composition. Results indicated a reduction in the yield strength within the fusion zone compared to the base metal. Yield strength reduction was found to be primarily constrained to the fusion zone due to grain coarsening with a less severe reduction in the heat affected zone. For all proposed alloys, laser welding resulted in a defect free weld devoid of cracking or inclusions.

Deformation behavior of laser welds in high temperature oxidation resistant Fe–Cr–Al alloys for fuel cladding applications

Energy Technology Data Exchange (ETDEWEB)

Field, Kevin G., E-mail: fieldkg@ornl.gov; Gussev, Maxim N., E-mail: gussevmn@ornl.gov; Yamamoto, Yukinori, E-mail: yamamotoy@ornl.gov; Snead, Lance L., E-mail: sneadll@ornl.gov

2014-11-15

Ferritic-structured Fe–Cr–Al alloys are being developed and show promise as oxidation resistant accident tolerant light water reactor fuel cladding. This study focuses on investigating the weldability and post-weld mechanical behavior of three model alloys in a range of Fe–(13–17.5)Cr–(3–4.4)Al (wt.%) with a minor addition of yttrium using modern laser-welding techniques. A detailed study on the mechanical performance of bead-on-plate welds using sub-sized, flat dog-bone tensile specimens and digital image correlation (DIC) has been carried out to determine the performance of welds as a function of alloy composition. Results indicated a reduction in the yield strength within the fusion zone compared to the base metal. Yield strength reduction was found to be primarily constrained to the fusion zone due to grain coarsening with a less severe reduction in the heat affected zone. For all proposed alloys, laser welding resulted in a defect free weld devoid of cracking or inclusions.

Influences of precursor constitution and processing speed on microstructure and wear behavior during laser clad composite coatings on γ-TiAl intermetallic alloy

International Nuclear Information System (INIS)

Liu Xiubo; Yu Rongli

2009-01-01

The effects of constitution of precursor mixed powders and scan speed on microstructure and wear properties were designed and investigated during laser clad γ/Cr 7 C 3 /TiC composite coatings on γ-TiAl intermetallic alloy substrates with NiCr-Cr 3 C 2 precursor mixed powders. The results indicate that both the constitution of the precursor mixed powders and the beam scan rate have remarkable influence on microstructure and attendant hardness as well as wear resistance of the formed composite coatings. The wear mechanisms of the original TiAl alloy and laser clad composite coatings were investigated. The composite coating with an optimum compromise between constitution of NiCr-Cr 3 C 2 precursor mixed powders as well as being processed under moderate scan speed exhibits the best wear resistance under dry sliding wear test conditions

Laser cladding to select new glassy alloys; Uso do metodo de revestimento por laser na selecao de novas ligas vitreas

Energy Technology Data Exchange (ETDEWEB)

Medrano, L.L.O.; Afonso, C.R.M.; Kiminami, C.S.; Gargarella, P., E-mail: eomedranos@hotmail.com [Universidade Federal de Sao Carlos (UFSCar), SP (Brazil). Departamento de Engenharia de Materiais; Vilar, R. [Instituto Superior Tecnico, Departamento de Engenharia Quimica, Lisboa (Portugal); Ramasco, B. [Whirlpool Latin America, Rio Claro, SP (Brazil)

2016-07-01

A new experimental technique used to analyze the effect of compositional variation and cooling rate in the phase formation in a multicomponent system is the laser cladding. This work have evaluated the use of laser cladding to discover a new bulk metallic glass (BMG) in the Al-Co-Zr system. Coatings with composition variation have made by laser cladding using Al-Co-Zr alloys powders and the samples produced have been characterized by X ray diffraction, microscopy and energy-dispersive X-ray spectroscopy. The results did not show the composition variation as expected, because of incomplete melting during laser process. It was measured a composition variation tendency that allowed the glass forming investigation by the glass formation criterion λ+Δh{sup 1/2}. The results have showed no glass formation in the coating samples, which prove a limited capacity of Zr-Co-Al system to form glass (author)

Long term immersion test of aluminum alloy AA 6061 used for fuel cladding in MTR type reactors

International Nuclear Information System (INIS)

Linardi, Evelina M.; Rodriguez, Sebastian; Haddad, Roberto; Lanzani, Liliana

2009-01-01

In this work we present the results of long term immersion tests performed in the aluminum alloy AA 6061, used for fuel cladding in MTR type reactors. The tests were performed at open circuit potential in high purity water (ρ = 18.2 MΩ.cm) and in 10 -3 M NaCl solution. Two kinds of assemblies were studied: simple sheets and artificial crevices, immersed during 6, 12 and 18 months at room temperature. In both media and both assemblies, the aluminum hydroxide phases crystalline bayerite and bohemite were identified. It was found that a kind of localized attack named alkaline attack occurs around the iron-rich intermetallics. These particles were confirmed to control the corrosion of the AA 6061 alloy in an aerated medium. Immersion times for up to 18 months did not increase the oxide growth or the alkaline attack on the AA 6061 alloy. (author)

Experimental study and modeling of high-temperature oxidation and phase transformation of cladding-tubes made in zirconium alloy

International Nuclear Information System (INIS)

Mazeres, Benoit

2013-01-01

One of the hypothetical accident studied in the field of the safety studies of Pressurized light Water Reactor (PWR) is the Loss-Of-Coolant-Accident (LOCA). In this scenario, zirconium alloy fuel claddings could undergo an important oxidation at high temperature (T≅ 1200 C) in a steam environment. Cladding tubes constitute the first confinement barrier of radioelements and then it is essential that they keep a certain level of ductility after quenching to ensure their integrity. These properties are directly related to the growth kinetics of both the oxide and the αZr(O) phase and also to the oxygen diffusion profile in the cladding tube after the transient. In this context, this work was dedicated to the understanding and the modeling of the both oxidation phenomenon and oxygen diffusion in zirconium based alloys at high temperature. The numerical tool (EKINOX-Zr) used in this thesis is based on a numerical resolution of a diffusion/reaction problem with equilibrium-conditions on three moving boundaries: gas/oxide, oxide/αZr(O), αZr(O)/βZr. EKINOX-Zr kinetics model is coupled with ThermoCalc software and the Zircobase database to take into account the influence of the alloying elements (Sn, Fe, Cr, Nb) but also the influence of hydrogen on the solubility of oxygen. This study focused on two parts of the LOCA scenario: the influence of a pre-oxide layer (formed in-service) and the effects of hydrogen. Thanks to the link between EKINOX-Zr and the thermodynamic database Zircobase, the hydrogen effects on oxygen solubility limit could be considered in the numerical simulations. Thus, simulations could reproduce the oxygen diffusion profiles measured in pre-hydrided samples. The existence of a thick pre-oxide layer on cladding tubes can induce a reduction of this pre-oxide layer before the growth of a high-temperature one during the high temperature dwell under steam. The first simulations performed using the numerical tool EKINOX-Zr showed that this particular

Diffusion in cladding materials

International Nuclear Information System (INIS)

Anand, M.S.; Pande, B.M.; Agarwala, R.P.

1992-01-01

Aluminium has been used as a cladding material in most research reactors because its low neutron absorption cross section and ease of fabrication. However, it is not suitable for cladding in power reactors and as such zircaloy-2 is normally used as a clad because it can withstand high temperature. It has low neutron absorption cross section, good oxidation, corrosion, creep properties and possesses good mechanical strength. With the passage of time, further development in this branch of science took place and designers started looking for better neutron economy and less hydrogen pickup in PHW reactors. The motion of fission products in the cladding material could pose a problem after long operation. In order to understand their behaviour under reactor environment, it is essential to study first the diffusion under normal conditions. These studies will throw light on the interaction of defects with impurities which would in turn help in understanding the mechanism of diffusion. In this article, it is intended to discuss the diffusion behaviour of impurities in cladding materials.(i.e. aluminium, zircaloy-2, zirconium-niobium alloy etc.). (author). 94 refs., 4 figs., 3 tabs

Formation of anomalous eutectic in Ni-Sn alloy by laser cladding

Science.gov (United States)

Wang, Zhitai; Lin, Xin; Cao, Yongqing; Liu, Fencheng; Huang, Weidong

2018-02-01

Ni-Sn anomalous eutectic is obtained by single track laser cladding with the scanning velocity from 1 mm/s to 10 mm/s using the Ni-32.5 wt.%Sn eutectic powders. The microstructure of the cladding layer and the grain orientations of anomalous eutectic were investigated. It is found that the microstructure is transformed from primary α-Ni dendrites and the interdendritic (α-Ni + Ni3Sn) eutectic at the bottom of the cladding layer to α-Ni and β-Ni3Sn anomalous eutectic at the top of the cladding layer, whether for single layer or multilayer laser cladding. The EBSD maps and pole figures indicate that the spatially structure of α-Ni phase is discontinuous and the Ni3Sn phase is continuous in anomalous eutectic. The transformation from epitaxial growth columnar at bottom of cladding layer to free nucleation equiaxed at the top occurs, i.e., the columnar to equiaxed transition (CET) at the top of cladding layer during laser cladding processing leads to the generation of anomalous eutectic.

Characteristics of WWER-1000 fuel rod claddings and FA components from E635 alloy at burnups up to 72 MWd/kgU

International Nuclear Information System (INIS)

Nikulin, A.; Novikov, A.; Peregud, M.; Shishov, V.; Shevyakov, A.; Volkova, I.; Novoselov, A.; Kobylyansky, G.

2011-01-01

In this paper operation experience, results of investigated E365 alloy components of Balakovo NPP Unit 1 and Kalinin NPP unit 1 fuel assemblies are presented. Appearance, shape changes and geometric size, corrosion state of guide thimbles, angles and fuel rods, corrosion of fuel claddings are studied. At the end authors concluded that: I) E635 alloy corroborated its high operation reliability as fuel claddings and WWER-1000 FA components during 6 year service to the fuel burnup of 72MWd/kgU; II) Based on the results from the post-irradiation investigations of the fuel rods and other structural elements of WWER-1000 FAA, fabricated from E635 alloy, in terms of the basic operational characteristics, their resources after the 6 year operation cycle have not been exhausted; III) The geometrical parameters, corrosion states, tensile properties of items fabricated from fuel alloy did not attain the values that would prevent their further operation: 1) the elongations of the fuel rods at the mean burnups up to 66.2 MWd/kgU do not exceed 15 mm or 4.9%; 8) the amount of the oxide coat at surface of GT and CT does not exceed 45 μm, the hydrogen content is <0.03% mass; 9) the oxide coat at the surfaces of the frame angles does not exceed 50 μm, the hydrogen content is <0.04% mass

In-pile cladding tests at NRI Rez and PIE capabilities and experience

International Nuclear Information System (INIS)

Zmitko, M.

2002-01-01

In-pile cladding corrosion test facilities and relevant post-irradiation capabilities at NRI Rez plc are overviewed. Basic information about the research rector LVR-15 and in-pile water loops is given. An experience in the field of Zr-alloy cladding corrosion testing and investigation of cladding corrosion behaviour is demonstrated for two experimental programmes conducted at NRI Rez in the past period. The first example describes results obtained at studying of corrosion behaviour of advanced Zr-alloys under PWR conditions with a special concern to a high lithium content and subcooled surface boiling. The second example informs about completion of the experimental programme supported by the IAEA which is focused on investigation of Zircaloy-4 cladding behaviour under VVER water chemistry, thermal-hydraulic and irradiation conditions with the main to obtain experimental data for an assessment of the Zircaloy-4 cladding compatibility with VVER conditions. (author)

Microstructure and high-temperature oxidation resistance of TiN/Ti3Al intermetallic matrix composite coatings on Ti6Al4V alloy surface by laser cladding

Science.gov (United States)

Zhang, Xiaowei; Liu, Hongxi; Wang, Chuanqi; Zeng, Weihua; Jiang, Yehua

2010-11-01

A high-temperature oxidation resistant TiN embedded in Ti3Al intermetallic matrix composite coating was fabricated on titanium alloy Ti6Al4V surface by 6kW transverse-flow CO2 laser apparatus. The composition, morphology and microstructure of the laser clad TiN/Ti3Al intermetallic matrix composite coating were characterized by optical microscopy (OM), scanning electron microscopy (SEM), X-ray diffraction (XRD) and energy dispersive spectrometer (EDS). In order to evaluate the high-temperature oxidation resistance of the composite coatings and the titanium alloy substrate, isothermal oxidation test was performed in a conventional high-temperature resistance furnace at 600°C and 800°C respectively. The result shows that the laser clad intermetallic composite coating has a rapidly solidified fine microstructure consisting of TiN primary phase (granular-like, flake-like, and dendrites), and uniformly distributed in the Ti3Al matrix. It indicates that a physical and chemical reaction between the Ti powder and AlN powder occurred completely under the laser irradiation. In addition, the microhardness of the TiN/Ti3Al intermetallic matrix composite coating is 844HV0.2, 3.4 times higher than that of the titanium alloy substrate. The high-temperature oxidation resistance test reveals that TiN/Ti3Al intermetallic matrix composite coating results in the better modification of high-temperature oxidation behavior than the titanium substrate. The excellent high-temperature oxidation resistance of the laser cladding layer is attributed to the formation of the reinforced phase TiN and Al2O3, TiO2 hybrid oxide. Therefore, the laser cladding TiN/Ti3Al intermetallic matrix composite coating is anticipated to be a promising oxidation resistance surface modification technique for Ti6Al4V alloy.

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.

Development and characterisation of iridium-192 seeds for brachytherapy treatment of ocular tumors

International Nuclear Information System (INIS)

Peleias Jr, F.S.; Zeituni, C.A.; Souza, C.D.; Rostelato, M.E.CM.; Mattos, F.R.; Banega, M.A.G.; Rodrigues, B.T.; Tiezzi, R.; Oliveira, T.B.; Feher, A.; Moura, J.A.; Costa, O.L.

2014-01-01

Even ocular tumors are not amongst the cases with a high incidence, they affect the population, particularly children. The Institute of Energy and Nuclear Research (IPEN-CNEN/SP) in partnership with Escola Paulista de Medicina (UNIFESP), created a project to develop an alternative treatment for ophthalmic cancer that uses iridium-192 seeds in brachytherapy. This work aims to study and develop a seed of iridium-192 from a platinum-iridium alloy The prototype seed has a 3.0 mm long core sealed by a titanium capsule of 0.8 mm of outer diameter, 0.05 mm of wall thickness and 4.5 mm long. We developed a methodology that covered: characterisation of the material used in the core, creation of a device for neutron activation of the cores and leakage tests. The results show that this methodology is feasible. As a suggestion for future work, studies regarding metrology and dosimetry of these sources should be carried out. (authors)

Corrosion and protection of spent Al-clad research reactor fuel during extended wet storage

International Nuclear Information System (INIS)

Ramanathan, Lalgudi V.

2009-01-01

A variety of spent research reactor fuel elements with different fuel meats, geometries and 235 U enrichments are presently stored under water in basins throughout the world. More than 90% of these fuels are clad in aluminum (Al) or its alloy and are susceptible to corrosion. This paper presents an overview of the influence of Al alloy composition, galvanic effects (Al alloy/stainless steel), crevice effects, water parameters and synergism between these parameters as well as settled solids on the corrosion of typical Al alloys used as fuel element cladding. Pitting is the main form of corrosion and is affected by water conductivity, chloride ion content, formation of galvanic couples with rack supports and settled solid particles. The extent to which these parameters influence Al corrosion varies. This paper also presents potential conversion coatings to protect the spent fuel cladding. (author)

Preparation and characterization of laser cladding wollastonite derived bioceramic coating on titanium alloy.

Science.gov (United States)

Li, Huan-cai; Wang, Dian-gang; Chen, Chuan-zhong; Weng, Fei; Shi, Hua

2015-09-25

The bioceramic coating is fabricated on titanium alloy (Ti6Al4V) by laser cladding the preplaced wollastonite (CaSiO3) powders. The coating on Ti6Al4V is characterized by x-ray diffraction, scanning electron microscopy coupled with energy dispersive spectroscopy, and attenuated total reflection Fourier-transform infrared. The interface bonding strength is measured using the stretching method using an RGD-5-type electronic tensile machine. The microhardness distribution of the cross-section is determined using an indentation test. The in vitro bioactivity of the coating on Ti6Al4V is evaluated using the in vitro simulated body fluid (SBF) immersion test. The microstructure of the laser cladding sample is affected by the process parameters. The coating surface is coarse, accidented, and microporous. The cross-section microstructure of the ceramic layer from the bottom to the top gradually changes from cellular crystal, fine cellular-dendrite structure to underdeveloped dendrite crystal. The coating on Ti6Al4V is composed of CaTiO3, CaO, α-Ca2SiO4, SiO2, and TiO2. After soaking in the SBF solution, the calcium phosphate layer is formed on the coating surface.

Duplex stainless steel surface bay laser cladding

International Nuclear Information System (INIS)

Amigo, V.; Pineda, Y.; Segovia, F.; Vicente, A.

2004-01-01

Laser cladding is one of the most promising techniques to restore damaged surfaces and achieve properties similar to those of the base metal. In this work, duplex stainless steels have been cladded by a nickel alloy under different processing conditions. The influence of the beam speed and defocusing variables ha been evaluated in the microstructure both of the cladding and heat affected zone, HAZ. These results have been correlated to mechanical properties by means of microhardness measurements from cladding area to base metal through the interface. This technique has shown to be very appropriate to obtain controlled mechanical properties as they are determined by the solidification microstructure, originated by the transfer of mass and heat in the system. (Author) 21 refs

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.

Study on process of laser cladded nuclear valve parts

International Nuclear Information System (INIS)

Zhang Chunliang

2000-01-01

The microstructure and performances of the Co-base alloy coatings that are formed by laser cladding, plasma spurt welding and arc surfacing on the nuclear valve-sealing surface have been studied and compared. The combination costs of laser cladding, plasma spurt welding and arc, surfacing have been analyzed and compared. The results showed that the laser cladding processing has the advantages of high efficiency, low energy cost, a little machining allowance, high rate of finished products and low combination cost, compared with plasma spurt welding processing and arc surfacing processing. The laser cladding technology can improve the qualities of nuclear valve parts and increase their service life. Therefore, the laser cladding processing is a new technology with developing potential

Development Status of Accident Tolerant Fuel Cladding for LWRs

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jung-Hwan; Yang, Jae-Ho; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Hydrogen explosions and the release of radionuclides are caused by severe damage of current nuclear fuels, which are composed of fuel pellets and fuel cladding, during an accident. To reduce the damage to the public, the fuels have to enhance their integrity under an accident environment. Enhanced accident tolerance fuels (ATFs) can tolerate a loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normal operations as well as operational transients, in comparison with the current UO{sub 2}-Zr alloy system used in the LWR. Surface modified Zr cladding as a new concept was suggested to apply an enhanced ATF cladding. The aim of the partial ODS treatment is to increase the high-temperature strength to suppress the ballooning/rupture behavior of fuel cladding during an accident event. The target of the surface coating is to increase the corrosion resistance during normal operation and increase the oxidation resistance during an accident event. The partial ODS treatment of Zircaloy-4 cladding can be produced using a laser beam scanning method with Y2O3 powder, and the surface Cr-alloy and Cr/FeCrAl coating on Zircaloy-4 cladding can be obtained after the development of 3D laser coating and arc ion plating technologies.

Cladding of Advanced Al Alloys Employing Friction Stir Welding

NARCIS (Netherlands)

van der Stelt, A.A.; Bor, Teunis Cornelis; Geijselaers, Hubertus J.M.; Akkerman, Remko; van den Boogaard, Antonius H.

2013-01-01

In this paper an advanced solid state cladding process, based on Friction Stir Welding, is presented. The Friction Surface Cladding (FSC) technology enables the deposition of a solid-state coating using filler material on a substrate with good metallurgical bonding. A relatively soft AA1050 filler

Corrosion performance of optimised and advanced fuel rod cladding in PWRs at high burnups

International Nuclear Information System (INIS)

Jourdain, P.; Hallstadius, L.; Pati, S.R.; Smith, G.P.; Garde, A.M.

1997-01-01

The corrosion behaviour both in-pile and out-of-pile for a number of cladding alloys developed by ABB to meet the current and future needs for fuel rod cladding with improved corrosion resistance is presented. The cladding materials include: 1) Zircaloy-4 (OPTIN) with optimised composition and processing and Zircaloy-2 optimised for Pressurised Water Reactors (PWR), (Zircaloy-2P), and 2) several alternative zirconium-based alloys with compositions outside the composition range for Zircaloys. The data presented originate from fuel rods irradiated in six PWRs to burnups up to about 66 MWd/kgU and from tests conducted in 360 o water autoclave. Also included are in-pile fuel rod growth measurements on some of the alloys. (UK)

Cladding failure margins for metallic fuel in the integral fast reactor

International Nuclear Information System (INIS)

Bauer, T.H.; Fenske, G.R.; Kramer, J.M.

1987-01-01

The Integral Fast Reactor (IFR) concept being developed at Argonne National Laboratory has prompted a renewed interest in uranium-based metal alloys as a fuel for sodium-cooled fast reactors. In this paper we 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. In the final section of this paper we extend the calculations to consider the failure of IFR ternary fuel under reactor accident conditions. (orig./GL)

Method for automatic filling of nuclear fuel rod cladding tubes

International Nuclear Information System (INIS)

Bezold, H.

1979-01-01

Prior to welding the zirconium alloy cladding tubes with end caps, they are automatically filled with nuclear fuel tablets and ceramic insulating tablets. The tablets are introduced into magazine drums and led through a drying oven to a discharging station. The empty cladding tubes are removed from this discharging station and filled with tablets. A filling stamp pushes out the columns of tablets in the magazine tubes of the magazine drum into the cladding tube. Weight and measurement of length determine the filled state of the cladding tube. The cladding tubes are then led to the welding station via a conveyor belt. (DG) [de

Influence of Zircaloy cladding composition on hydride formation during aqueous hydrogen charging

Energy Technology Data Exchange (ETDEWEB)

Rajasekhara, S. [Intel Corporation, 2501 NW 229th Av., Hillsboro, OR 97124 (United States); Kotula, P.G.; Enos, D.G.; Doyle, B.L. [Sandia National Laboratories, Albuquerque, NM, 87185 (United States); Clark, B.G., E-mail: blyclar@sandia.gov [Sandia National Laboratories, Albuquerque, NM, 87185 (United States)

2017-06-15

Although hydrogen uptake in Zirconium (Zr) based claddings has been a topic of many studies, hydrogen uptake as a function of alloy composition has received little attention. In this work, commercial Zr-based cladding alloys (Zircaloy-2, Zircaloy-4 and ZIRLO™), differing in composition but with similar initial textures, grain sizes, and surface roughness, were aqueously charged with hydrogen for 100, 300, and 1000 s at nominally 90 °C to produce hydride layers of varying thicknesses. Transmission electron microscope characterization following aqueous charging showed hydride phase and orientation relationship were identical in all three alloys. However, elastic recoil detection measurements confirmed that surface hydride layers in Zircaloy-2 and Zircaloy-4 were an order of magnitude thicker relative to ZIRLO™. - Highlights: •Aqueous charging was performed to produce a layer of zirconium hydride for three different Zr-alloy claddings. •Hydride thicknesses were analyzed by elastic recoil detection and transmission electron microscopy. •Zircaloy-2 and Zircaloy-4 formed thicker hydride layers than ZIRLO™ for the same charging durations.

Characterization of hard coatings produced by laser cladding using laser-induced breakdown spectroscopy technique

Science.gov (United States)

Varela, J. A.; Amado, J. M.; Tobar, M. J.; Mateo, M. P.; Yañez, A.; Nicolas, G.

2015-05-01

Protective coatings with a high abrasive wear resistance can be obtained from powders by laser cladding technique, in order to extend the service life of some industrial components. In this work, laser clad layers of self-fluxing NiCrBSi alloy powder mixed with WC powder have been produced on stainless steel substrates of austenitic type (AISI 304) in a first step and then chemically characterized by laser-induced breakdown spectroscopy (LIBS) technique. With the suitable laser processing parameters (mainly output power, beam scan speed and flow rate) and powders mixture proportions between WC ceramics and NiCrBSi alloys, dense pore free layers have been obtained on single tracks and on large areas with overlapped tracks. The results achieved by LIBS technique and applied for the first time to the analysis of laser clads provided the chemical composition of the tungsten carbides in metal alloy matrix. Different measurement modes (multiple point analyses, depth profiles and chemical maps) have been employed, demonstrating the usefulness of LIBS technique for the characterization of laser clads based on hardfacing alloys. The behavior of hardness can be explained by LIBS maps which evidenced the partial dilution of some WC spheres in the coating.

MODELLING OF NUCLEAR FUEL CLADDING TUBES CORROSION

Directory of Open Access Journals (Sweden)

Miroslav Cech

2016-12-01

Full Text Available This paper describes materials made of zirconium-based alloys used for nuclear fuel cladding fabrication. It is focused on corrosion problems their theoretical description and modeling in nuclear engineering.

High-temperature steam oxidation kinetics of the E110G cladding alloy

International Nuclear Information System (INIS)

Király, Márton; Kulacsy, Katalin; Hózer, Zoltán; Perez-Feró, Erzsébet; Novotny, Tamás

2016-01-01

In the course of recent years, several experiments were performed at MTA EK (Centre for Energy Research, Hungarian Academy of Sciences) on the isothermal high-temperature oxidation of the improved Russian cladding alloy E110G in steam/argon atmosphere. Using these data and designing additional supporting experiments, the oxidation kinetics of the E110G alloy was investigated in a wide temperature range, between 600 °C and 1200 °C. For short durations (below 500 s) or high temperatures (above 1065 °C) the oxidation kinetics was found to follow a square-root-of-time dependence, while for longer durations and in the intermediate temperature range (800–1000 °C) it was found to approach a cube-root-of-time dependence rather than a square-root one. Based on the results a new best-estimate and a conservative oxidation kinetics model were created. - Highlights: • Steam oxidation kinetics of E110G was studied at MTA EK based on old and new data. • New best-estimate and conservative steam oxidation kinetics were proposed for E110G. • The exponent of oxidation time changed depending on oxidation temperature. • A simple exponential curve was used instead of Arrhenius-type curve for the factor.

Cladding tube manufacturing technology

International Nuclear Information System (INIS)

Hahn, R.; Jeong, Y. H.; Baek, B. J.; Kim, K. H.; Kim, S. J.; Choi, B. K.; Kim, J. M.

1999-04-01

This report gives an overview of the manufacturing routine of PWR cladding tubes. The routine essentially consists of a series of deformation and annealing processes which are necessary to transform the ingot geometry to tube dimensions. By changing shape, microstructure and structure-related properties are altered simultaneously. First, a short overview of the basics of that part of deformation geometry is given which is related to tube reducing operations. Then those processes of the manufacturing routine which change the microstructure are depicted, and the influence of certain process parameters on microstructure and material properties are shown. The influence of the resulting microstructure on material properties is not discussed in detail, since it is described in my previous report A lloy Development for High Burnup Cladding . Because of their paramount importance still up to now, and because manufacturing data and their influence on properties for other alloys are not so well established or published, the descriptions are mostly related to Zry4 tube manufacturing, and are only in short for other alloys. (author). 9 refs., 46 figs

A study on wear resistance and microcrack of the Ti{sub 3}Al/TiAl + TiC ceramic layer deposited by laser cladding on Ti-6Al-4V alloy

Energy Technology Data Exchange (ETDEWEB)

Li Jianing, E-mail: ljnljn1022@163.com [Key Laboratory for Liquid-Solid Structural Evolution and Processing of Materials (Ministry of Education), Department of Materials Science, Shandong University, Jing Shi Road 17923, Jinan 250061, Shandong (China); Chen Chuanzhong [Key Laboratory for Liquid-Solid Structural Evolution and Processing of Materials, Ministry of Education, Department of Materials Science, Shandong University, Jing Shi Road 17923, Jinan 250061, Shandong (China); Squartini, Tiziano [INFM-Department of Physics, Siena University, Siena 53100 (Italy); He Qingshan [Key Laboratory for Liquid-Solid Structural Evolution and Processing of Materials, Ministry of Education, Department of Materials Science, Shandong University, Jing Shi Road 17923, Jinan 250061, Shandong (China)

2010-12-15

Laser cladding of the Al + TiC alloy powder on Ti-6Al-4V alloy can form the Ti{sub 3}Al/TiAl + TiC ceramic layer. In this study, TiC particle-dispersed Ti{sub 3}Al/TiAl matrix ceramic layer on the Ti-6Al-4V alloy by laser cladding has been researched by means of X-ray diffraction, scanning electron microscope, electron probe micro-analyzer, energy dispersive spectrometer. The main difference from the earlier reports is that Ti{sub 3}Al/TiAl has been chosen as the matrix of the composite coating. The wear resistance of the Al + 30 wt.% TiC and the Al + 40 wt.% TiC cladding layer was approximately 2 times greater than that of the Ti-6Al-4V substrate due to the reinforcement of the Ti{sub 3}Al/TiAl + TiC hard phases. However, when the TiC mass percent was above 40 wt.%, the thermal stress value was greater than the materials yield strength limit in the ceramic layer, the microcrack was present and its wear resistance decreased.

Material Selection for Accident Tolerant Fuel Cladding

International Nuclear Information System (INIS)

Pint, Bruce A.; Terrani, Kurt A.; Yamamoto, Yukinori; Snead, Lance Lewis

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-H 2 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 800°C could not be reproduced. Both FeCrAl and MAX phase Ti 2 AlC form a protective alumina scale in steam. However, commercial Ti 2 AlC that was not single phase, formed a much thicker oxide at 1200°C in steam and significant TiO 2 , and therefore Ti 2 AlC 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.

Microstructure and properties of the low-power-laser clad coatings on magnesium alloy with different amount of rare earth addition

International Nuclear Information System (INIS)

Zhu, Rundong; Li, Zhiyong; Li, Xiaoxi; Sun, Qi

2015-01-01

Highlights: • The low-power-laser was used to obtain the excellent coatings with different amount of Y_2O_3 addition. • The addition of rare earth oxide Y_2O_3 refined and purified the microstructure of the coatings, meanwhile, increased the thickness of the coatings and reduced the dilution of cladding materials from based alloy. • The primary phases in the coatings are Mg_3_2Al_4_7Cu_7, MgCu_6Al_5, Al_2CuMg and Al_1_2Mg_1_7. The A_l_4MgY and MgAl_2O_4 phase can be found in Y_2O_3-modified coatings. • The micro-hardness and the abrasion resistance of the coatings with Y_2O_3 had been improved obviously compared with the coatings without Y_2O_3. • The corrosion resistance of the AZ91D magnesium alloy had been improved by laser cladding. And the effect of Y_2O_3 on the corrosion potential of the coatings was less than the effect of Y_2O_3 on corrosion current density of the coatings. - Abstract: Due to the low-melting-point and high evaporation rate of magnesium at elevated temperature, high power laser clad coating on magnesium always causes subsidence and deterioration in the surface. Low power laser can reduce the evaporation effect while brings problems such as decreased thickness, incomplete fusion and unsatisfied performance. Therefore, low power laser with selected parameters was used in our research work to obtain Al–Cu coatings with Y_2O_3 addition on AZ91D magnesium alloy. The addition of Y_2O_3 obviously increases thickness of the coating and improves the melting efficiency. Furthermore, the effect of Y_2O_3 addition on the microstructure of laser clad Al–Cu coatings was investigated by scanning electron microscopy. The energy-dispersive spectrometer (EDS) and X-ray diffractometer (XRD) were used to examine the elemental and phase compositions of the coatings. The properties were investigated by micro-hardness test, dry wear test and electrochemical corrosion. It was found that the addition of Y_2O_3 refined the microstructure. The micro

Consolidation of cladding hulls from the electrometallurgical treatment of spent fuel

International Nuclear Information System (INIS)

Keiser, D.D. Jr.

1998-01-01

To consolidate metallic waste that is residual from Argonne National Laboratory's electrometallurgical treatment of spent nuclear fuel, waste ingots are currently being cast using an induction furnace located in a hot cell. These ingots, which have been developed to serve as final waste forms destined for repository disposal, are stainless steel (SS)-Zr alloys (the Zr is very near 15 wt.%). The charge for the alloys consists of stainless steel cladding hulls, Zr from the fuel being treated, noble metal fission products, and minor amounts of actinides that are present with the cladding hulls. The actual in-dated cladding hulls have been characterized before they were melted into ingots, and the final as-cast ingots have been characterized to determine the degree of consolidation of the charge material. It has been found that ingots can be effectively cast from irradiated cladding hulls residual from the electrometallurgical treatment process by employing an induction furnace located in a hot cell

Laser cladding of Zr on Mg for improved corrosion properties

International Nuclear Information System (INIS)

Subramanian, R.; Sircar, S.; Mazumder, J.

1989-01-01

This paper reports the results of laser cladding of Mg-2wt%Zr, and Mg-5wt%Zr powder mixture onto magnesium. The microstructure of the laser clad was studied. From the microstructural study, the epitaxial regrowth of the clad region on the underlying substrate was observed. Martensite plates of different size were observed in transmission electron microscope for MG-2wt%Zr and Mg-5wt%Zr laser clad. The corrosion properties of the laser clad were evaluated in sea water (3.5% NaCl). The position of the laser claddings in the galvanic series of metals in sea water, the anodic polarization characteristics of the laser claddings and the protective nature and the stability of the passivating film formed have been determined. The formation of pits on the surface of the laser clad subjected to corrosion is reported. The corrosion properties of the laser claddings are compared with that of the commercially used magnesium alloy AZ91B

Prediction model for oxide thickness on aluminum alloy cladding during irradiation

International Nuclear Information System (INIS)

Kim, Yeon Soo; Hofman, G.L.; Hanan, N.A.; Snelgrove, J.L.

2003-01-01

An empirical model predicting the oxide film thickness on aluminum alloy cladding during irradiation has been developed as a function of irradiation time, temperature, heat flux, pH, and coolant flow rate. The existing models in the literature are neither consistent among themselves nor fit the measured data very well. They also lack versatility for various reactor situations such as a pH other than 5, high coolant flow rates, and fuel life longer than ∼1200 hrs. Particularly, they were not intended for use in irradiation situations. The newly developed model is applicable to these in-reactor situations as well as ex-reactor tests, and has a more accurate prediction capability. The new model demonstrated with consistent predictions to the measured data of UMUS and SIMONE fuel tests performed in the HFR, Petten, tests results from the ORR, and IRIS tests from the OSIRIS and to the data from the out-of-pile tests available in the literature as well. (author)

Development of Cold Spray Coatings for Accident-Tolerant Fuel Cladding in Light Water Reactors

Science.gov (United States)

Maier, Benjamin; Yeom, Hwasung; Johnson, Greg; Dabney, Tyler; Walters, Jorie; Romero, Javier; Shah, Hemant; Xu, Peng; Sridharan, Kumar

2018-02-01

The cold spray coating process has been developed at the University of Wisconsin-Madison for the deposition of oxidation-resistant coatings on zirconium alloy light water reactor fuel cladding with the goal of improving accident tolerance during loss of coolant scenarios. Coatings of metallic (Cr), alloy (FeCrAl), and ceramic (Ti2AlC) materials were successfully deposited on zirconium alloy flats and cladding tube sections by optimizing the powder size, gas preheat temperature, pressure and composition, and other process parameters. The coatings were dense and exhibited excellent adhesion to the substrate. Evaluation of the samples after high-temperature oxidation tests at temperatures up to 1300°C showed that the cold spray coatings significantly mitigate oxidation kinetics because of the formation of thin passive oxide layers on the surface. The results of the study indicate that the cold spray coating process is a viable near-term option for developing accident-tolerant zirconium alloy fuel cladding.

Flaw behavior in mechanically loaded clad plates

International Nuclear Information System (INIS)

Iskander, S.K.; Robinson, G.C.; Oland, C.B.

1989-01-01

A small crack near the inner surface of clad nuclear reactor pressure vessels is an important consideration in the safety assessment of the structural integrity of the vessel. Four-point bend tests on large plate specimens, conforming to ASTM specification for pressure vessel plates, alloy steels, quenched and tempered, Mn-Mo and Mn-Mo-Ni (A533) grade B six clad and two unclad with stainless steels 308, 309 and 312 weld wires, were performed to determine the effect of cladding upon the propagation of small surface cracks subjected to stress states. Results indicated that the tough surface layer composed of cladding and/or heat-affected zone has enhanced the load-bearing capacity of plates under conditions where unclad plates have ruptured. The results are interpreted in terms of fracture mechanics. The behavior of flaws in clad reactor pressure vessels is examined in the light of the test results. 11 refs., 8 figs., 2 tabs

Strengthening of the nuclear valve sealing surface by laser cladding technology

International Nuclear Information System (INIS)

Shi Shihong; Huang Guodong

1998-07-01

A 5 kW laser with CO 2 flow transverse for cladding Co-base alloy or Ni-base alloy coat on the austenite matrix of the nuclear valve sealing surface is introduced. The results show that, after the sealing surface of valve is processed by the laser cladding, the coat of 3.0 mm thick can be made with smooth surface. The test and comparison analysis indicate that the structure and all performance have obvious advantages over that of the plasma spurt welding, bead welding and flame welding processing

Study on microstructure and high temperature wear resistance of laser cladded nuclear valve clack

International Nuclear Information System (INIS)

Zhang Chunliang; Chen Zichen

2002-01-01

Laser cladding of Co-base alloy on the nuclear valve-sealing surface are performed with a 5 kW CO 2 transverse flowing laser. The microstructure and the high temperature impact-slide wear resistance of the laser cladded coating and the plasma cladded coating are studied. The results show that the microstructure, the dilution rate and the high temperature impact-slide wear resistance of the laser cladded coating have obvious advantages over the spurt cladding processing

Method of processing spent fuel cladding tubes

International Nuclear Information System (INIS)

Nakatsuka, Masafumi; Ouchi, Atsuhiro; Imahashi, Hiromichi.

1986-01-01

Purpose: To decrease the residual activity of spent fuel cladding tubes in a short period of time and enable safety storage with simple storage equipments. Constitution: Spent fuel cladding tubes made of zirconium alloys discharged from a nuclear fuel reprocessing step are exposed to a grain boundary embrittling atmosphere to cause grain boundary destruction. This causes grain boundary fractures to the zirconium crystal grains as the matrix of nuclear fuels and then precipitation products precipitated to the grain boundary fractures are removed. The zirconium constituting the nuclear fuel cladding tube and other ingredient elements contained in the precipitation products are separated in this removing step and they are separately stored respectively. As a result, zirconium constituting most part of the composition of the spent nuclear fuel cladding tubes can be stored safely at a low activity level. (Takahashi, M.)

High-temperature oxidation kinetics of sponge-based E110 cladding alloy

Science.gov (United States)

Yan, Yong; Garrison, Benton E.; Howell, Mike; Bell, Gary L.

2018-02-01

Two-sided oxidation experiments were recently conducted at 900°C-1200 °C in flowing steam with samples of sponge-based Zr-1Nb alloy E110. Although the old electrolytic E110 tubing exhibited a high degree of susceptibility to nodular corrosion and experienced breakaway oxidation rates in a relatively short time, the new sponge-based E110 demonstrated steam oxidation behavior comparable to Zircaloy-4. Sample weight gain and oxide layer thickness measurements were performed on oxidized E110 specimens and compared to oxygen pickup and oxide layer thickness calculations using the Cathcart-Pawel correlation. Our study shows that the sponge-based E110 follows the parabolic law at temperatures above 1015 °C. At or below 1015 °C, the oxidation rate was very low when compared to Zircaloy-4 and can be represented by a cubic expression. No breakaway oxidation was observed at 1000 °C for oxidation times up to 10,000 s. Arrhenius expressions are given to describe the parabolic rate constants at temperatures above 1015 °C and cubic rate constants are provided for temperatures below 1015 °C. The weight gains calculated by our equations are in excellent agreement with the measured sample weight gains at all test temperatures. In addition to the as-fabricated E110 cladding sample, prehydrided E110 cladding with hydrogen concentrations in the 100-150 wppm range was also investigated. The effect of hydrogen content on sponge-based E110 oxidation kinetics was minimal. No significant difference was found between as-fabricated and hydrided samples with regard to oxygen pickup and oxide layer thickness for hydrogen contents below 150 wppm.

Zirconium alloy barrier having improved corrosion resistance

International Nuclear Information System (INIS)

Adamson, R.B.; Rosenbaum, H.S.

1983-01-01

A nuclear fuel element for use in the core of a nuclear reactor has a composite cladding container having a substrate and a dilute zirconium alloy liner bonded to the inside surface of the substrate. The dilute zirconium alloy liner forms about 1 to about 20 percent of the thickness of the cladding and is comprised of zirconium and a metal selected from the group consisting of iron, chromium, iron plus chromium, and copper. The dilute zirconium alloy liner shields the substrate from impurities or fission products from the nuclear fuel material and protects the substrate from stress corrosion and stress cracking. The dilute zirconium alloy liner displays greater corrosion resistance, especially to oxidation by hot water or steam than unalloyed zirconium. The substrate material is selected from conventional cladding materials, and preferably is a zirconium alloy. (author)

Magnetostrictive clad steel plates for high-performance vibration energy harvesting

Science.gov (United States)

Yang, Zhenjun; Nakajima, Kenya; Onodera, Ryuichi; Tayama, Tsuyoki; Chiba, Daiki; Narita, Fumio

2018-02-01

Energy harvesting technology is becoming increasingly important with the appearance of the Internet of things. In this study, a magnetostrictive clad steel plate for harvesting vibration energy was proposed. It comprises a cold-rolled FeCo alloy and cold-rolled steel joined together by thermal diffusion bonding. The performances of the magnetostrictive FeCo clad steel plate and conventional FeCo plate cantilevers were compared under bending vibration; the results indicated that the clad steel plate construct exhibits high voltage and power output compared to a single-plate construct. Finite element analysis of the cantilevers under bending provided insights into the magnetic features of a clad steel plate, which is crucial for its high performance. For comparison, the experimental results of a commercial piezoelectric bimorph cantilever were also reported. In addition, the cold-rolled FeCo and Ni alloys were joined by thermal diffusion bonding, which exhibited outstanding energy harvesting performance. The larger the plate volume, the more the energy generated. The results of this study indicated not only a promising application for the magnetostrictive FeCo clad steel plate as an efficient energy harvester, related to small vibrations, but also the notable feasibility for the formation of integrated units to support high-power trains, automobiles, and electric vehicles.

Status Report on the Fabrication of Fuel Cladding Chemical Interaction Test Articles for ATR Irradiations

Energy Technology Data Exchange (ETDEWEB)

Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-09-28

FeCrAl alloys are a promising new class of alloys for light water reactor (LWR) applications due to their superior oxidation and corrosion resistance in high temperature environments. The current R&D efforts have focused on the alloy composition and processing routes to generate nuclear grade FeCrAl alloys with optimized properties for enhanced accident tolerance while maintaining properties needed for normal operation conditions. Therefore, the composition and processing routes must be optimized to maintain the high temperature steam oxidation (typically achieved by increasing the Cr and Al content) while still exhibiting properties conducive to normal operation in a LWR (such as radiation tolerance where reducing Cr content is favorable). Within this balancing act is the addition of understanding the influence on composition and processing routes on the FeCrAl alloys for fuel-cladding chemical interactions (FCCI). Currently, limited knowledge exists on FCCI for the FeCrAl-UO₂ clad-fuel system. To overcome the knowledge gaps on the FCCI for the FeCrAl-UO2 clad-fuel system a series of fueled irradiation tests have been developed for irradiation in the Advanced Test Reactor (ATR) housed at the Idaho National Laboratory (INL). The first series of tests has already been reported. These tests used miniaturized 17x17 PWR fuel geometry rodlets of second-generation FeCrAl alloys fueled with industrial Westinghouse UO₂ fuel. These rodlets were encapsulated within a stainless steel housing.To provide high fidelity experiments and more robust testing, a new series of rodlets have been developed deemed the Accident Tolerant Fuel Experiment #1 Oak Ridge National Laboratory FCCI test (ATF-1 ORNL FCCI). The main driving factor, which is discussed in detail, was to provide a radiation environment where prototypical fuel-clad interface temperatures are met while still maintaining constant contact between industrial fuel and the candidate cladding alloys

Evaluation of long-term creep behaviour on K-cladding tubes

International Nuclear Information System (INIS)

Bang, J. G.; Jeong, Y. H.; Jeong, Y. H.

2003-01-01

KAERI has developed new zirconium alloys for high burnup fuel cladding. To evaluate the performance of these alloys, various out-pile tests are conducting. At high burnup, the creep resistance as well as corrosion resistance is one of the major factors determining nuclear fuel performance. Long-term creep test was performed at 350 .deg. C and 400 .deg. C and 100, 120, 135, and 150 MPa of applied hoop stress to evaluate the creep properties. The creep resistance was strongly affected by the final heat treatment conditions, while there was no effect of intermediate heat treatment. The creep strain of the recrystallized alloys is higher than that of the stress-relieved alloys by a factor of 3. The alloying elements also influenced the creep behaviour. Increase of Sn content enhanced the creep resistance, while Nb decreased the creep resistance. As a result of texture analysis, basal pole was directed to normal direction, while prism pole was to rolling direction. The development of the deformation texture and the ammealing texture showed almost similar process to Zircaloy cladding

Temperature and humidity effects on the corrosion of aluminium-base reactor fuel cladding materials during dry storage

International Nuclear Information System (INIS)

Peacock, H.B.; Sindelar, R.L.; Lam, P.S.

2004-01-01

The effect of temperature and relative humidity on the high temperature (up to 200 deg. C) corrosion of aluminum cladding alloys was investigated for dry storage of spent nuclear fuels. A dependency on alloy type and temperature was determined for saturated water vapor conditions. Models were developed to allow prediction of cladding behaviour of 1100, 5052, and 6061 aluminum alloys for up to 50+ years at 100% relative humidity. Calculations show that for a closed system, corrosion stops after all moisture and oxygen is used up during corrosion reactions with aluminum alloys. (author)

Fuel-cladding mechanical interaction effects in fast reactor mixed oxide fuel

Energy Technology Data Exchange (ETDEWEB)

Boltax, A [Westinghouse Electric Corporation, Advanced Reactor Division, Madison, PA (United States); Biancheria, A

1977-04-01

Thermal and fast reactor irradiation experiments on mixed oxide fuel pins under steady-state and power change conditions reveal evidence for significant fuel-cladding mechanical interaction (FCMI) effects. Analytical studies with the LIFE-III fuel performance code indicate that high cladding stresses can be produced by general and local FCMI effects. Also, evidence is presented to show that local cladding strains can be caused by the accumulation of cesium at the fuel-cladding interface. Although it is apparent that steady-state FCMI effects have not given rise to cladding breaches in current fast reactors, it is anticipated that FCMI may become more important in the future because of interest in: higher fuel burnups; increased power ramp rates; load follow operation; and low swelling cladding alloys. (author)

Fuel-cladding mechanical interaction effects in fast reactor mixed oxide fuel

International Nuclear Information System (INIS)

Boltax, A.; Biancheria, A.

1977-01-01

Thermal and fast reactor irradiation experiments on mixed oxide fuel pins under steady-state and power change conditions reveal evidence for significant fuel-cladding mechanical interaction (FCMI) effects. Analytical studies with the LIFE-III fuel performance code indicate that high cladding stresses can be produced by general and local FCMI effects. Also, evidence is presented to show that local cladding strains can be caused by the accumulation of cesium at the fuel-cladding interface. Although it is apparent that steady-state FCMI effects have not given rise to cladding breaches in current fast reactors, it is anticipated that FCMI may become more important in the future because of interest in: higher fuel burnups; increased power ramp rates; load follow operation; and low swelling cladding alloys. (author)

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

International Nuclear Information System (INIS)

McClelland, R.G.; O'Leary, P.M.

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 ∼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

Study of the aqueous corrosion mechanisms and kinetics of the AlFeNi aluminium based alloy used for the fuel cladding in the Jules Horowitz research reactor

International Nuclear Information System (INIS)

Wintergerst, M.

2009-05-01

For the Jules Horowitz new material-testing reactor (JHR), an aluminium base alloy, called AlFeNi, will be used for the cladding of the fuel plates. This alloy (Al - 1% Fe - 1% Ni - 1 % Mg), which is already used as fuel cladding, was developed for its good corrosion resistance in water at high temperatures. However, few studies dealing with the alteration process in water and the relationships with irradiation effects have been performed on this alloy. The conception of the JHR fuel requires a better knowledge of the corrosion mechanisms. Corrosion tests were performed in autoclaves at 70 C, 165 C and 250 C on AlFeNi plates representative of the fuel cladding. Several techniques were used to characterize the corrosion scale: SEM, TEM, EPMA, XRD, Raman spectroscopy. Our observations show that the corrosion scale is made of two main layers: a dense amorphous scale close to the metal and a porous crystalline scale in contact with the water. More than the morphology, the chemical compositions of both layers are different. This duplex structure results from a mixed growth mechanism: an anionic growth to develop the inner oxide and a cationic diffusion followed by a dissolution-precipitation process to form the outer one. Dynamic experiments at 70 C and corrosion kinetics measurements have demonstrated that the oxide growth process is controlled by a diffusion step associated to a dissolution/precipitation process. A corrosion mechanism of the AlFeNi alloy in aqueous media has been proposed. Then post-irradiation exams performed on irradiated fuel plates were used to investigate the effects of the irradiation on the corrosion behaviour in the reactor core. (author)

Development of Self-Healing Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding

Energy Technology Data Exchange (ETDEWEB)

Sridharan, Kumar [University of Wisconsin-Madison; Mariani, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, Xianming [Idaho National Lab. (INL), Idaho Falls, ID (United States); Xu, Peng [Westinghouse Electric Company; Lahoda, Ed [Westinghouse Electric Company

2018-03-31

Given the long-term goal of developing such coatings for use with nuclear reactor fuel cladding, this work describes results of oxidation and corrosion behavior of bulk zirconium-silicide and fabrication of zirconium-silicide coatings on zirconium-alloy test flats, tube configurations, and SiC test flats. In addition, boiling heat transfer of these modified surfaces (including ZrSi2 coating) during clad quenching experiments is discussed in detail. Oxidation of bulk ZrSi2 was found to be negligible compared to Zircaloy-4 (a common Zr-alloy cladding material) and mechanical integrity of ZrSi2 was superior to that of bulk Zr2Si at high temperatures in ambient air. Very interesting and unique multi-nanolayered composite of ZrO2 and SiO2 were observed. Physical model for the oxidation has been proposed wherein Zr–Si–O mixture undergoes a spinodal phase decomposition into ZrO2 and SiO2, which is manifested as a nanoscale assembly of alternating layer of the two oxides. Steam corrosion at high pressure (10.3 MPa) led to weight loss of ZrSi2 and produced oxide scale with depletion of silicon, possibly attributed to volatile silicon hydroxide, gaseous silicon monoxide, and a solubility of silicon dioxide in water. Only Zircon phase (ZrSiO4) formed during oxidation of ZrSi2 at 1400°C in air, and allowed for immobilization silicon species in oxide scale in the aqueous environments. Zirconium-silicide coatings (on zirconium-alloy substrates) investigated in this study were deposited primarily using magnetron sputter deposition method and slurry method, although powder spray deposition processes cold spray and thermal spray methods were also investigated. The optimized ZrSi2 sputtered coating exhibited a highly protective nature at elevated temperatures in ambient air by mitigating oxygen permeation to the underlying zirconium alloy substrate. The high oxidation resistance of the coating has been shown to be due to nanocrystalline SiO2 and ZrSiO4 phases in the amorphous

WWER water chemistry related to fuel cladding behaviour

Energy Technology Data Exchange (ETDEWEB)

Kysela, J; Zmitko, M [Nuclear Research Inst. plc., Rez (Czech Republic); Vrtilkova, V [Nuclear Fuel Inst., Prague (Czech Republic)

1997-02-01

Operational experience in WWER primary water chemistry and corrosion related to the fuel cladding is reviewed. Insignificant corrosion of fuel cladding was found which is caused by good corrosion resistance of Zr1Nb material and relatively low coolant temperature at WWER-440 reactor units. The differences in water chemistry control is outlined and an attention to the question of compatibility of Zircaloys with WWER water chemistry is given. Some results of research and development in field of zirconium alloy corrosion behaviour are discussed. Experimental facility for in-pile and out-of-pile cladding material corrosion testing is shown. (author). 14 refs, 5 figs, 3 tabs.

Residual stresses in weld-clad reactor pressure vessel steel

International Nuclear Information System (INIS)

Bertram, W.

1975-01-01

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

Inpile (in PWR) testing of cladding materials

International Nuclear Information System (INIS)

Hahn, R.; Jeong, Y. H.; Baek, B. J.; Kim, K. H.; Kim, S. J.; Choi, B. K.; Kim, J. M.

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

Protection of spent aluminum-clad research reactor fuels during extended wet storage

International Nuclear Information System (INIS)

Fernandes, Stela M.C.; Correa, Olandir V.; Souza, Jose A.; Ramanathan, Lalgudi V.; Antunes, Renato A.

2013-01-01

Aluminum-clad spent nuclear fuel from research reactors (RR) is stored in light water filled pools or basins worldwide. Many incidences of pitting corrosion of the fuel cladding has been reported and attributed to synergism in the effect of certain water parameters. Protection of spent Al-clad RR fuel with a conversion coating was proposed in 2008. Preliminary results revealed increased pitting corrosion resistance of cerium oxide coated aluminum alloys AA 1050 and AA 6061, used as RR fuel plate cladding. Further development of conversion coatings for Al alloys was carried out and this paper presents: (a) the preparation and characterization of hydrotalcite (HTC) coatings; (b) the results of laboratory tests in which the corrosion behavior of coated Al alloys in NaCl solutions was determined; (c) the results of field tests in which un-coated, boehmite coated, HTC coated and cerium modified boehmite / HTC coated AA 1050 and AA 6061 coupons were exposed to the IEA-R1 reactor spent fuel basin for extended periods. In these field tests the coupons coated with HTC from a high temperature (HT) bath and subsequently modified with Ce were the most resistant to pitting corrosion. In laboratory tests also, HT- hydrotalcite + Ce coated specimens were the most corrosion resistant in 0.01 M NaCl. The role of cerium in increasing the corrosion resistance imparted by the different conversion coatings of spent Al-clad RR fuel elements is presented. (author)

Analytical functions used for description of the plastic deformation process in Zirconium alloys WWER type fuel rod cladding under designed accident conditions

International Nuclear Information System (INIS)

Fedotov, A.

2003-01-01

The aim of this work was to improve the RAPTA-5 code as applied to the analysis of the thermomechanical behavior of the fuel rod cladding under designed accident conditions. The irreversible process thermodynamics methods were proposed to be used for the description of the plastic deformation process in zirconium alloys under accident conditions. Functions, which describe yielding stress dependence on plastic strain, strain rate and temperature may be successfully used in calculations. On the basis of the experiments made and the existent experimental data the dependence of yielding stress on plastic strain, strain rate, temperature and heating rate for E110 alloy was determined. In future the following research work shall be made: research of dynamic strain ageing in E635 alloy under different strain rates; research of strain rate influence on plastic strain in E635 alloy under test temperature higher than 873 K; research of deformation strengthening of E635 alloy under high temperatures; research of heating rate influence n phase transformation in E110 and E635 alloys

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.

Plasma spheroidization and cladding of powders

Energy Technology Data Exchange (ETDEWEB)

Petrunichev, V.A.; Averin, V.V.; Sorokin, L.M.; Koroleva, E.B.

1987-02-01

With reference to experimental results for nickel and chromium alloys, it is shown that complex alloy powders can be spheroidized in plasma discharges using an argon plasma with hydrogen. The spheroidizing process is accompanied by the reduction of surface oxides, with uniform element distribution within the particles; the granulometric composition of the particles is preserved. It is also shown that plasma technology can be used for producing metal-clad oxide and carbide powders, which improve the performance of cermets and coatings.

Investigation on fuel-cladding chemical interaction in metal fuel for FBR

International Nuclear Information System (INIS)

Inagaki, Kenta; Nakamura, Kinya; Ogata, Takanari; Uwaba, Tomoyuki

2013-01-01

During steady-state irradiation of metallic fuel in fast reactors, rare-earth fission products can react with stainless steel cladding at the fuel-cladding interface. The authors conducted isothermal annealing tests with some diffusion couples to investigate the structure of the wastage layer formed at the interface. Candidate cladding alloys, ferritic-martensitic steel (PNC-FMS) and oxide-dispersion-strengthened (ODS) steel were assembled with rare-earth alloys, RE5 : La-Ce-Pr-Nd-Sm, which simulate the fission yield of rare-earth fission products. The diffusion couples were isothermally annealed in the temperature range of 500-650°C for up to 170 h. In both RE5/ODS-steel and RE5/PNC-FMS couples, the wastage layer of the two-phase region of the (Fe, Cr) 17 RE 2 matrix phase with the precipitation of the (Fe, RE, Cr) phase was formed. The structure was similar to that formed in RE5/Fe-12Cr and RE5/HT9 couples, which implies that the reaction between REs and steel is not significantly influenced by the minor alloying elements within the candidate cladding materials. It was also clarified that the increase in the wastage layer thickness was diffusion-controlled. The temperature dependence of the reaction rate constants were formulated, which can be the basis for the quantification of the wastage layer growth. (author)

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

International Nuclear Information System (INIS)

Xu Guojian; Kutsuna, Muneharu; Liu Zhongjie; Zhang Hong

2006-01-01

The clad layers of Ni-based alloy were deposited on the SUS316L stainless plates by CO 2 laser and plasma cladding processes. The smooth clad bead was obtained by CO 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 7 C 3 . The eutectic structure consists of Ni + CrB or Ni + Cr 7 C 3 . Compared with the plasma cladding, the fine microstructures, low dilutions, high Vickers hardness and excellent wear resistance were obtained by CO 2 laser cladding. All that show the laser cladding process has a higher efficiency and good cladding quality

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.

Irradiation effects on mechanical properties of fuel element cladding from thermal reactors

International Nuclear Information System (INIS)

Chatterjee, S.

2005-01-01

During reactor operation, UO 2 expands more than the cladding tube (Zirconium alloys for thermal reactors), is hotter, cracks and swells. The fuel therefore will interact with the cladding, resulting in straining of the later. To minimize the possibility of rupture of the cladding, ideally it should have good ductility as well as high strength. However, the ductility reduces with increase in fuel element burn-up. Increased burn-up also increases swelling of the fuel, leading to increased contact pressure between the fuel and the cladding tube. This would cause strains to be concentrated over localized regions of the cladding. For fuel elements burnup exceeding 40 GWd/T, the contribution of embrittlement due to hydriding, and the increased possibility of embrittlement due to stress corrosion cracking, also need to be considered. In addition to the tensile properties, the other mechanical properties of interest to the performance of cladding tube in an operating fuel element are creep rate and fatigue endurance. Irradiation is reported to have insignificant effect on high cycle endurance limit, and fatigue from fuel element vibration is most unlikely, to be life limiting. Even though creep rates due to irradiation are reported to increase by an order of magnitude, the cladding creep ductility would be so high that creep type failures in fuel element would be most improbable. Thus, the most important limiting aspect of mechanical performance of fuel element cladding has been recognized as the tensile ductility resulting from the stress conditions experienced by the cladding. Some specific fission products of threshold amount (if) deposited on the cladding, and hydride morphology (e.g. hydride lenses). The presentation will brief about irradiation damage in cladding materials and its significance, background of search for better Zirconium alloys as cladding materials, and elaborate on the types of mechanical tests need to be conducted for the evaluation of claddings

Corrosion of aluminum-clad alloys in wet spent fuel storage

International Nuclear Information System (INIS)

Howell, J.P.

1995-09-01

Large quantities of Defense related spent nuclear fuels are being stored in water basins around the United States. Under the non-proliferation policy, there has been no processing since the late 1980's and these fuels are caught in the pipeline awaiting processing or other disposition. At the Savannah River Site, over 200 metric tons of aluminum clad fuel are being stored in four water filled basins. Some of this fuel has experienced significant pitting corrosion. An intensive effort is underway at SRS to understand the corrosion problems and to improve the basin storage conditions for extended storage requirements. Significant improvements have been accomplished during 1993-1995, but the ultimate solution is to remove the fuel from the basins and to process it to a more stable form using existing and proven technology. This report presents a discussion of the fundamentals of aluminum alloy corrosion as it pertains to the wet storage of spent nuclear fuel. It examines the effects of variables on corrosion in the storage environment and presents the results of corrosion surveillance testing activities at SRS, as well as other fuel storage basins within the Department of Energy production sites

Influence of microstructure modification on the circumferential creep of Zr–Nb–Sn–Fe cladding tubes

International Nuclear Information System (INIS)

Jeong, Gu Beom; Kim, In Won; Hong, Sun Ig

2016-01-01

Out-of-reactor, non-irradiated thermal creep performances and lives of annealed and stress-relieved Zr-1.02Nb-0.69Sn-0.12Fe cladding tubes were studied and compared. The creep rates of annealed Zr-1.02Nb-0.69Sn-0.12Fe cladding tubes were appreciably slower than those of stress-relieved annealed counterpart. The stress exponent increased slightly from 5.1 to 6.1 in the stress-relieved cladding to 5.3–6.3 in the annealed cladding. The creep activation energy of the annealed Zr-1.02Nb-0.69Sn-0.12Fe alloy (300–330 kJ/mol) was larger compared to that of the stress-relieved alloy (210–260 kJ/mol). The creep activation energy of annealed alloy is close to that of self-diffusion in α-Zr (336 kJ/mol). The smaller activation energy in the stress-relieved alloy is attributed to the increasing contribution of faster diffusion path such as grain boundaries and dislocations. The presence of dislocation arrays with higher dislocation density and smaller grain size in the stress-relived alloy was confirmed by TEM analysis. The creep rupture time increased dramatically in the annealed Zr–1Nb- 0.7Sn-0.1Fe alloy compared to that of stress-relieved alloy, supporting the decrease of creep rate by annealing. The creep life of Zr-1.02Nb-0.69Sn-0.12Fe claddings can be extended through microstructure modification by annealing at intermediate temperatures in which dislocation creep dominates. - Highlights: • Effect of microstructure modification on creep in Zr–Nb–Sn–Fe tubes was studied. • Creep activation energy in annealed tubes was larger than in stress-relieved tubes. • Lower dislocation density in lager grains was observed after creep in annealed tubes. • Larson–Miller parameter of annealed tube was larger than that of stress-relieved one. • Creep life of tubes was extended through microstructure modification by annealing.

Microstructure and properties of the low-power-laser clad coatings on magnesium alloy with different amount of rare earth addition

Energy Technology Data Exchange (ETDEWEB)

Zhu, Rundong; Li, Zhiyong, E-mail: lizhiyong@nuc.edu.cn; Li, Xiaoxi; Sun, Qi

2015-10-30

Highlights: • The low-power-laser was used to obtain the excellent coatings with different amount of Y{sub 2}O{sub 3} addition. • The addition of rare earth oxide Y{sub 2}O{sub 3} refined and purified the microstructure of the coatings, meanwhile, increased the thickness of the coatings and reduced the dilution of cladding materials from based alloy. • The primary phases in the coatings are Mg{sub 32}Al{sub 47}Cu{sub 7}, MgCu{sub 6}Al{sub 5}, Al{sub 2}CuMg and Al{sub 12}Mg{sub 17}. The A{sub l4}MgY and MgAl{sub 2}O{sub 4} phase can be found in Y{sub 2}O{sub 3}-modified coatings. • The micro-hardness and the abrasion resistance of the coatings with Y{sub 2}O{sub 3} had been improved obviously compared with the coatings without Y{sub 2}O{sub 3}. • The corrosion resistance of the AZ91D magnesium alloy had been improved by laser cladding. And the effect of Y{sub 2}O{sub 3} on the corrosion potential of the coatings was less than the effect of Y{sub 2}O{sub 3} on corrosion current density of the coatings. - Abstract: Due to the low-melting-point and high evaporation rate of magnesium at elevated temperature, high power laser clad coating on magnesium always causes subsidence and deterioration in the surface. Low power laser can reduce the evaporation effect while brings problems such as decreased thickness, incomplete fusion and unsatisfied performance. Therefore, low power laser with selected parameters was used in our research work to obtain Al–Cu coatings with Y{sub 2}O{sub 3} addition on AZ91D magnesium alloy. The addition of Y{sub 2}O{sub 3} obviously increases thickness of the coating and improves the melting efficiency. Furthermore, the effect of Y{sub 2}O{sub 3} addition on the microstructure of laser clad Al–Cu coatings was investigated by scanning electron microscopy. The energy-dispersive spectrometer (EDS) and X-ray diffractometer (XRD) were used to examine the elemental and phase compositions of the coatings. The properties were investigated

Fabrication and use of zircaloy/tantalum-sheathed cladding thermocouples and molybdenum/rhenium-sheathed fuel centerline thermocouples

International Nuclear Information System (INIS)

Wilkins, S.C.; Sepold, L.K.

1985-01-01

The thermocouples described in this report are zircaloy/tantalum-sheathed and molybdenum/rhenium alloy-sheathed instruments intended for fuel rod cladding and fuel centerline temperature measurements, respectively. Both types incorporate beryllium oxide insulation and tungsten/rhenium alloy thermoelements. These thermocouples, operated at temperatures of 2000 0 C and above, were developed for use in the internationally sponsored Severe Fuel Damage test series in the Power Burst Facility. The fabrication steps for both thermocouple types are described in detail. A laser-welding attachment technique for the cladding-type thermocouple is presented, and experience with alternate materials for cladding and fuel therocouples is discussed

ODS Alloys for Nuclear Applications

International Nuclear Information System (INIS)

Jang, Jin Sung

2006-01-01

ODS (oxide dispersion strengthening) alloy is one of the potential candidate alloys for the cladding or in reactor components of Generation IV reactors and for the structural material even for fusion reactors. It is widely accepted as very resistant material to neutron irradiation as well as strong material at high temperature due to its finely distributed and stable oxide particles. Among Generation IV reactors SFR and SCWR are anticipated in general to run in the temperature range between 300 and 550 .deg. C, and the peak cladding temperature is supposed to reach at about 620 .deg. C during the normal operation. Therefore Zr.base alloys, which have been widely known and adopted for the cladding material due to their excellent neutron economics, are no more adequate at these operating conditions. Fe-base ODS alloys in general has a good high temperature strength at the above high temperature as well as the neutron resistance. In this study a range of commercial grade ODS alloys and their applications are reviewed, including an investigation of the stability of a commercial grade 20% Cr Fe-base ODS alloy(MA956). The alloy was evaluated in terms of the fracture toughness change along with the aging treatment. Also an attempt of the development of 9% Cr Fe-base ODS alloys is introduced

Capture of Tritium Released from Cladding in the Zirconium Recycle Process

Energy Technology Data Exchange (ETDEWEB)

Spencer, Barry B [ORNL; Bruffey, Stephanie H [ORNL; DelCul, Guillermo Daniel [ORNL; Walker, Trenton Baird [ORNL

2016-08-31

Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using nonradioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.

Capture of Tritium Released from Cladding in the Zirconium Recycle Process

Energy Technology Data Exchange (ETDEWEB)

Bruffey, Stephanie H [ORNL; Spencer, Barry B [ORNL; DelCul, Guillermo Daniel [ORNL

2016-08-31

This report is issued as the first revision to FCRD-MRWFD-2016-000297. Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using non-radioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.

Characterization of Cassini GPHS fueled clad production girth welds

International Nuclear Information System (INIS)

Franco-Ferreira, E.A.; Moyer, M.W.; Reimus, M.A.H.; Placr, A.; Howard, B.D.

2000-01-01

Fueled clads for radioisotope power systems are produced by encapsulating 238 PuO 2 in iridium alloy cups, which are joined at their equators by gas tungsten arc welding. Cracking problems at the girth weld tie-in area during production of the Galileo/Ulysses GPHS capsules led to the development of a first-generation ultrasonic test for girth weld inspection at the Savannah River Plant. A second-generation test and equipment with significantly improved sensitivity and accuracy were jointly developed by the Oak Ridge Y-12 Plant and Westinghouse Savannah River Company for use during the production of Cassini GPHS capsules by the Los Alamos National Laboratory. The test consisted of Lamb wave ultrasonic scanning of the entire girth weld from each end of the capsule combined with a time-of-flight evaluation to aid in characterizing nonrelevant indications. Tangential radiography was also used as a supplementary test for further evaluation of reflector geometry. Each of the 317 fueled GP HS capsules, which were girth welded for the Cassini Program, was subjected to a series of nondestructive tests that included visual, dimensional, helium leak rate, and ultrasonic testing. Thirty-three capsules were rejected prior to ultrasonic testing. Of the 44 capsules rejected by the standard ultrasonic test, 22 were upgraded to flight quality through supplementary testing for an overall process acceptance rate of 82.6%. No confirmed instances of weld cracking were found

Experimental assessment of fuel-cladding interactions

Energy Technology Data Exchange (ETDEWEB)

Wood, Elizabeth Sooby [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-06-29

A range of fuel concepts designed to better tolerate accident scenarios and reactor transients are currently undergoing fundamental development at national laboratories as well as university and industrial partners. Pellet-clad mechanical and chemical interaction can be expected to affect fuel failure rates experienced during steady state operation, as well as dramatically impact the response of the fuel form under loss of coolant and other accident scenarios. The importance of this aspect of fuel design prompted research initiated by AFC in FY14 to begin exploratory efforts to characterize this phenomenon for candidate fuelcladding systems of immediate interest. Continued efforts in FY15 and FY17 aimed to better understand and simulate initial pellet-clad interaction with little-to-no pressure on the pellet-clad interface. Reported here are the results from 1000 h heat treatments at 400, 500, and 600°C of diffusion couples pairing UN with a FeCrAl alloy, SiC, and Zr-based cladding candidate sealed in evacuated quartz ampoules. No gross reactions were observed, though trace elemental contaminants were identified.

Development of metallic fuel fabrication - A study on the interdiffusion behavior between ternary metallic fuel and cladding materials

Energy Technology Data Exchange (ETDEWEB)

Lee, Byung Soo; Seol, Kyung Won; Shon, In Jin [Chonbuk National University, Chonju (Korea)

1999-04-01

To study a new ternary metallic fuel for liquid metal reactor, various U-Zr-X alloys have been made by induction melting. The specimens were prepared for thermal stability tests at 630 deg. C upto 5000 hours in order to estimate the decomposition of the lamellar structure. Interdiffusion studies were carried out at 700 deg. C for 200 hours for the diffusion couples assembled with U-Zr-X ternary fuel versus austenitic stainless steel D9 and martensitic stainless steel HT9, respectively, to investigate the fuel-cladding compatibility. The ternary alloy, especially U-Zr-Mo and U-Zr-Nb alloys showed relatively good thermal stability as long as 5000hrs at 630 deg. C. From the composition profiles of the interdiffusion study, Fe penetrated deeper to the fuel side than other cladding elements such as Ni and Cr, whereas U did to the cladding side of fuel elements in the fuel/D9 couples. On the contrary, the reaction layers of Fuel/HT9 couple were thinner than that of Fuel/D9 couples and were less affected by cladding element, which was believed to be due to Zr rich layer between the fuel-cladding interface. HT9 is considered to be superior to D9 and a favorable choice as a cladding material in terms of fuel-cladding compatibility. 21 refs., 24 figs., 7 tabs. (Author)

Evaluation of Corrosion of Aluminum Based Reactor Fuel Cladding Materials During Dry Storage

International Nuclear Information System (INIS)

Peacock, H.B. Jr.

1999-01-01

This report provides an evaluation of the corrosion behavior of aluminum cladding alloys and aluminum-uranium alloys at conditions relevant to dry storage. The details of the corrosion program are described and the results to date are discussed

Irradiation experience with HT9-clad metallic fuel

International Nuclear Information System (INIS)

Pahl, R.G.; Lahm, C.E.; Tsai, H.; Billone, M.C.

1991-01-01

The safe and reliable performance of metallic fuel is currently under study and demonstration in the Integral Fast Reactor program. In-reactor tests of HT9-clad metallic fuel have now reached maturity and have all shown good performance characteristics to burnups exceeding 17.5 at. % in the lead assembly. Because this low-swelling tempered martensitic alloy is the cladding of choice for high fluence applications, the experimental observations and performance modelling efforts reported in this paper play an important role in demonstrating reliability

Electrochemical profiling of multi-clad aluminium sheets used in automotive heat exchangers

DEFF Research Database (Denmark)

Bordo, Kirill; Ambat, Rajan; Peguet, Lionel

2014-01-01

The objective of the present study is to understand the mechanisms of corrosion propagation across the multi-clad structure of Al alloys sheets as a function of local alloy composition and microstructure, with and without brazing treatment. Electro-chemical behaviour at different depths was profi...

Meeting the challenge of extremely corrosive service: A primer on clad oilfield equipment

International Nuclear Information System (INIS)

Pendley, M.R.

1993-01-01

Extremely corrosive environments, such as those often encountered in deep, hot, sour oil and gas wells, are usually characterized by the presence of hydrogen sulfide (H 2 S), carbon dioxide (CO 2 ), chlorides, and other corrosive species coupled with high temperatures (> 400 F/204 C) and high pressures (up to 20,000 psi/138 MPa). Most low alloy and stainless steel materials are not suitable for such environments. Extremely corrosive service conditions dictate the use of a corrosion-resistant alloy (CRA) in areas which are exposed to the hostile environment. However, it is often cost-prohibitive to make an entire component out of CRA material. An alternative strategy is to use a low alloy steel for the bulk of the component and clad critical surfaces with a corrosion-resistant material. Clad equipment can provide excellent corrosion resistance in hostile environments at a fraction of the cost of 100% CRA components. This paper will detail the problems posed by extremely corrosive environments and discuss how clad equipment provides a cost-effective solution

Microstructure of laser cladded martensitic stainless steel

CSIR Research Space (South Africa)

Van Rooyen, C

2006-08-01

Full Text Available and martensite with 10% ferrite for Material B. Table 7 - Proposed martensitic stainless steel alloys for laser cladding Material C* Cr Ni Mn Si Mo Co Ms (ºC)* Cr eq Ni eq Material A 0.4 13 - 1 0.5 2.5 5.5 120 16.5 12.5 Material B 0.2 15 2 1 0.7 2.5 5.5 117... dilution, low heat input, less distortion, increased mechanical and corrosion properties excellent repeatability and control of process parameters. Solidification of laser cladded martensitic stainless steel is primarily austenitic. Microstructures...

Zr-rich layers electrodeposited onto stainless steel cladding during the electrorefining of EBR-II fuel

International Nuclear Information System (INIS)

Keiser, D.D. Jr.; Mariani, R.D.

1999-01-01

Argonne National Laboratory is developing an electrometallurgical treatment for spent nuclear fuels. The initial demonstration of this process is being conducted on U-Zr alloy fuel elements irradiated in the experimental breeder reactor II (EBR-II). We report the first metallographic characterization of cladding hull remains for the electrometallurgical treatment of spent metallic fuel. During the electrorefining process, Zr-rich layers, with some U, deposit on all exposed surfaces of irradiated cladding segments (hulls) that originally contained the fuel alloy that was being treated. In some cases, not only was residual Zr (and U) found inside the cladding hulls, but a Zr-rind was also observed near the interior cladding hull surface. The Zr-rind was originally formed during the fuel casting process on the fuel slug. The observation of Zr deposits on all exposed cladding surfaces is explained with thermodynamic principles, when two conditions are met. These conditions are partial oxidation of Zr and the presence of residual uranium in the hulls when the electrorefining experiment is terminated. Comparisons are made between the structure of the initial irradiated fuel before electrorefining and the morphology of the material remaining in the cladding hulls after electrorefining. (orig.)

Influence of laser radiation on structure and properties of steels and alloys

International Nuclear Information System (INIS)

Tarasova, T; Popova, E

2013-01-01

In present study, and laser alloying of different steels and laser cladding of Ti and SiC powders mixtures was carried out, and microstructure, as well as microhardness profile and wear properties were examined. Research of the influence of lasers alloying modes on the elastic and plastic characteristics of the surface was conducted. As a result of chemical reactions in the cladded layer, a new phase (TiC) was synthesized during cladding process. The results showed that, in the clad layer, TiC was solidified to form dendrites in the clad zone. Produced coatings have high microhardness values in the upper and middle clad areas, about two time higher than clad matrix microhardness.

Thermal stress intensity factor for an axial crack in a clad cylinder

International Nuclear Information System (INIS)

Kuo, An Yu; Deardorf, A.F.; Riccardella, P.C.

1993-01-01

Many clad pressure vessels have been found to have cracks running through the inside surface cladding and into the base material. Although Young's moduli and Poisson's ratios of the clad and base materials are about the same for most of the industrial applications, coefficients of thermal expansion of the two dissimilar materials, clad and base materials, are usually quite different. For example, low alloy ferritic steel is a common base material for reactor pressure vessels (RPV) and the vessels are usually clad with austenitic stainless steel. Young's moduli for the low alloy steel and stainless steel at 350 F are 29,000 ksi and 28,000 ksi, respectively, while their coefficients of thermal expansion are 7.47x10 -6 in/in and 9.50x10 -6 in/in-degree F, respectively. The mismatch in coefficients of thermal expansion will cause high residual thermal stress even when the entire vessel is at a uniform temperature. This residual stress is one of the primary reasons why so many cracks have been found in the cladded components. In performing reactor pressure vessel integrity evaluation, such as computing probability of brittle fracture of the RPV, it is necessary to calculate stress intensity factors for cracks, which initiate from the clad material and run into the base metal. This paper presents a convenient method of calculating stress intensity factor for an axial crack emanating from the inside surface of a cladded cylinder under thermal loading. A J-integral like line integral was derived and used to calculate the stress intensity factors from finite element stress solutions of the problem

Research on Microstructure and Property of TiC-Co Composite Material Made by Laser Cladding

Science.gov (United States)

Zhang, Wei

The experiment of laser cladding on the surface of 2Cr13 steel was made. Titanium carbide (TiC) powder and Co-base alloy powder were used as cladding material. The microstructure and property of laser cladding layer were tested. The research showed that laser cladding layer had better properties such as minute crystals, deeper layer, higher hardness and good metallurgical bonding with base metal. The structure of cladding was supersaturated solid solution with dispersed titanium carbide. The average hardness of cladding zone was 660HV0.2. 2Cr13 steel was widely used in the field of turbine blades. Using laser cladding, the good wear layer would greatly increase the useful life of turbine blades.

Study and development of an Iridium-192 seed for use in ophthalmic cancer; Estudo e desenvolvimento de uma semente de iridio-192 para aplicacao em cancer oftalmico

Energy Technology Data Exchange (ETDEWEB)

Mattos, Fabio Rodrigues de

2013-07-01

Even ocular tumors are not among the cases with a higher incidence, they affect the population, especially children. The Institute of Energy and Nuclear Research (IPEN-CNEN/SP) in partnership with Escola Paulista de Medicina (UNIFESP), created a project to develop and implement a alternative treatment for ophthalmic cancer that use brachytherapy iridium-192 seeds. The project arose by reason of the Escola Paulista treat many cancer cases within the Unified Health System (SUS) and the research experience of sealed radioactive sources group at IPEN. The methodology was developed from the available infrastructure and the experience of researchers. The prototype seed presents with a core (192-iridium alloy of iridium-platinum) of 3.0 mm long sealed by a capsule of titanium of 0.8 mm outside diameter, 0.05 mm wall thickness and 4,5mm long. This work aims to study and develop a seed of iridium-192 from a platinum-iridium alloy. No study on the fabrication of these seeds was found in available literature. It was created a methodology that involved: characterization of the material used in the core, creation of device for neutron activation irradiation and and seed sealing tests. As a result, proved the feasibility of the method. As a suggestion for future work, studies regarding metrology and dosimetry of these sources and improvement of the methodology should be carried out, for future implementation in national scope. (author)

Corrosion fatigue crack growth in clad low-alloy steels. Part 2: Water flow rate effects in high-sulfur plate steel

International Nuclear Information System (INIS)

James, L.A.; Lee, H.B.; Wire, G.L.; Novak, S.R.; Cullen, W.H.

1997-01-01

Corrosion fatigue crack propagation tests were conducted on a high-sulfur ASTM A302-B plate steel overlaid with weld-deposited Alloy EN82H cladding. The specimens featured semi-elliptical surface cracks penetrating approximately 6.3 mm of cladding into the underlying steel. The initial crack sizes were relatively large with surface lengths of 22.8--27.3 mm, and depths of 10.5--14.1 mm. The experiments were initiated in a quasi-stagnant low-oxygen (O 2 < 10 ppb) aqueous environment at 243 C, under loading conditions (ΔK, R, cyclic frequency) conducive to environmentally assisted cracking (EAC) under quasi-stagnant conditions. Following fatigue testing under quasi-stagnant conditions where EAC was observed, the specimens were then fatigue tested under conditions where active water flow of either 1.7 m/s or 4.7 m/s was applied parallel to the crack. Earlier experiments on unclad surface-cracked specimens of the same steel exhibited EAC under quasi-stagnant conditions, but water flow rates at 1.7 m/s and 5.0 m/s parallel to the crack mitigated EAC. In the present experiments on clad specimens, water flow at approximately the same as the lower of these velocities did not mitigate EAC, and a free stream velocity approximately the same as the higher of these velocities resulted in sluggish mitigation of EAC. The lack of robust EAC mitigation was attributed to the greater crack surface roughness in the cladding interfering with flow induced within the crack cavity. An analysis employing the computational fluid dynamics code, FIDAP, confirmed that frictional forces associated with the cladding crack surface roughness reduced the interaction between the free stream and the crack cavity

Scientific basis for storage criteria for interim dry storage of aluminum-clad fuels

International Nuclear Information System (INIS)

Sindelar, R.L.; Peacock, H.B. Jr.; Lam, P.S.; Iyer, N.C.; Louthan, M.R. Jr.; Murphy, J.R.

1996-01-01

An engineered system for dry storage of aluminum-clad foreign and domestic research reactor spent fuel owned by the US Department of Energy is being considered to store the fuel up to a nominal period of 40 years prior to ultimate disposition. Scientifically-based criteria for environmental limits to drying and storing the fuels for this system are being developed to avoid excessive degradation in sealed and non-sealed (open to air) dry storage systems. These limits are based on consideration of degradation modes that can cause loss of net section of the cladding, embrittlement of the cladding, distortion of the fuel, or release of fuel and fission products from the fuel/clad system. Potential degradation mechanisms include corrosion mechanisms from exposure to air and/or sources of humidity, hydrogen blistering of the aluminum cladding, distortion of the fuel due to creep, and interdiffusion of the fuel and fission products with the cladding. The aluminum-clad research reactor fuels are predominantly highly-enriched aluminum uranium alloy fuel which is clad with aluminum alloys similar to 1100, 5052, and 6061 aluminum. In the absence of corrodant species, degradation due to creep and diffusion mechanisms limit the maximum fuel storage temperature to 200 C. The results of laboratory scale corrosion tests indicate that this fuel could be stored under air up to 200 C at low relative humidity levels (< 20%) to limit corrosion of the cladding and fuel (exposed to the storage environment through assumed pre-existing pits in the cladding). Excessive degradation of fuels with uranium metal up to 200 C can be avoided if the fuel is sufficiently dried and contained in a sealed system; open storage can be achieved if the temperature is controlled to avoid excessive corrosion even in dry air

Electrochemical behaviour of laser-clad Ti6Al4V with CP Ti in 0.1 M oxalic acid solution

Energy Technology Data Exchange (ETDEWEB)

Obadele, Babatunde Abiodun, E-mail: obadele4@gmail.com [Institute for NanoEngineering Research, Department of Chemical, Metallurgical and Materials Engineering, Tshwane University of Technology, Pretoria (South Africa); Olubambi, Peter A. [Institute for NanoEngineering Research, Department of Chemical, Metallurgical and Materials Engineering, Tshwane University of Technology, Pretoria (South Africa); Andrews, Anthony [Institute for NanoEngineering Research, Department of Chemical, Metallurgical and Materials Engineering, Tshwane University of Technology, Pretoria (South Africa); Department of Materials Engineering, Kwame Nkrumah University of Science and Technology, Kumasi (Ghana); Pityana, Sisa [Institute for NanoEngineering Research, Department of Chemical, Metallurgical and Materials Engineering, Tshwane University of Technology, Pretoria (South Africa); National Laser Center, Council for Scientific and Industrial Research, Pretoria (South Africa); Mathew, Mathew T. [Institute for NanoEngineering Research, Department of Chemical, Metallurgical and Materials Engineering, Tshwane University of Technology, Pretoria (South Africa); Department of Orthopedic Surgery, Rush University Medical Center, Chicago, IL 60612 (United States)

2015-10-15

The relationship between the microstructure and corrosion behaviour of Ti6Al4V alloy and laser-clad commercially pure (CP) Ti coating was investigated. The microstructure, phases and properties of the clad layers were investigated by X-ray diffractometry (XRD), scanning electron microscopy (SEM) and energy dispersive spectrometry (EDS). Electrochemical measurement techniques including open circuit potential (OCP) and potentiodynamic polarisation were used to evaluate the corrosion behaviour of Ti6Al4V alloy in 0.1 M oxalic acid solution and the results compared to the behaviour of laser-clad CP Ti at varying laser scan speed. Results showed that laser-clad CP Ti at scan speed of 0.4 m/min formed a good cladding layer without defects such as cracks and pores. The phase present in the cladding layer was mostly α′-Ti. The microstructures of the clad layer were needle like acicular/widmanstätten α. An improvement in the microhardness values was also recorded. Although the corrosion potentials of the laser-clad samples were less noble than Ti6Al4V alloy, the polarisation measurement showed that the anodic current density was lower and also increases with increasing laser scanning speed. - Highlights: • The microstructure and corrosion behaviour of laser-clad CP Ti was investigated. • Laser-clad CP Ti 0.4 m/min scan speed gave a good coating without cracks and pores. • The phase present in the clad layer was mostly α′-Ti. • An improvement in the microhardness values was also recorded. • Anodic current density for coatings increases with increasing laser scan speed.

Electrochemical behaviour of laser-clad Ti6Al4V with CP Ti in 0.1 M oxalic acid solution

International Nuclear Information System (INIS)

Obadele, Babatunde Abiodun; Olubambi, Peter A.; Andrews, Anthony; Pityana, Sisa; Mathew, Mathew T.

2015-01-01

The relationship between the microstructure and corrosion behaviour of Ti6Al4V alloy and laser-clad commercially pure (CP) Ti coating was investigated. The microstructure, phases and properties of the clad layers were investigated by X-ray diffractometry (XRD), scanning electron microscopy (SEM) and energy dispersive spectrometry (EDS). Electrochemical measurement techniques including open circuit potential (OCP) and potentiodynamic polarisation were used to evaluate the corrosion behaviour of Ti6Al4V alloy in 0.1 M oxalic acid solution and the results compared to the behaviour of laser-clad CP Ti at varying laser scan speed. Results showed that laser-clad CP Ti at scan speed of 0.4 m/min formed a good cladding layer without defects such as cracks and pores. The phase present in the cladding layer was mostly α′-Ti. The microstructures of the clad layer were needle like acicular/widmanstätten α. An improvement in the microhardness values was also recorded. Although the corrosion potentials of the laser-clad samples were less noble than Ti6Al4V alloy, the polarisation measurement showed that the anodic current density was lower and also increases with increasing laser scanning speed. - Highlights: • The microstructure and corrosion behaviour of laser-clad CP Ti was investigated. • Laser-clad CP Ti 0.4 m/min scan speed gave a good coating without cracks and pores. • The phase present in the clad layer was mostly α′-Ti. • An improvement in the microhardness values was also recorded. • Anodic current density for coatings increases with increasing laser scan speed

A Novel Method of Modeling the Deformation Resistance for Clad Sheet

International Nuclear Information System (INIS)

Hu Jianliang; Yi Youping; Xie Mantang

2011-01-01

Because of the excellent thermal conductivity, the clad sheet (3003/4004/3003) of aluminum alloy is extensively used in various heat exchangers, such as radiator, motorcar air conditioning, evaporator, and so on. The deformation resistance model plays an important role in designing the process parameters of hot continuous rolling. However, the complex behaviors of the plastic deformation of the clad sheet make the modeling very difficult. In this work, a novel method for modeling the deformation resistance of clad sheet was proposed by combining the finite element analysis with experiments. The deformation resistance model of aluminum 3003 and 4004 was proposed through hot compression test on the Gleeble-1500 thermo-simulation machine. And the deformation resistance model of clad sheet was proposed through finite element analysis using DEFORM-2D software. The relationship between cladding ratio and the deformation resistance was discussed in detail. The results of hot compression simulation demonstrate that the cladding ratio has great effects on the resistance of the clad sheet. Taking the cladding ratio into consideration, the mathematical model of the deformation resistance for clad sheet has been proved to have perfect forecasting precision of different cladding ratio. Therefore, the presented model can be used to predict the rolling force of clad sheet during the hot continuous rolling process.

U. S. programs on reference and advanced cladding/duct materials

International Nuclear Information System (INIS)

Bennett, J.W.; Holmes, J.J.; Laidler, J.J.

1977-05-01

Two coordinated national programs are presently in place in the United States for development of reference and advanced cladding and duct alloys for near-term and long-term LMFBR applications. A number of government, industrial and university laboratories are active participants in these two ERDA-sponsored programs. The programs are administered by ERDA through a task group organization, with each task group representing a particular technical activity and the membership of the task group drawn from among the laboratories with active involvement in that activity. Technical coordination of the two programs is provided by the Hanford Engineering Development Laboratory. The National Reference Cladding and Duct Program is charged with the responsibility for development of the required technology to permit full utilization of the reference material, 20 percent cold-worked Type 316 stainless steel, in early LMFBR core applications. The current schedule calls for full evaluation of FFTF-related design base data prior to full-power operation of FFTF in early 1980, followed by a confirmation in early 1983 of reference material performance capabilities for initial-core CRBRP applications. Comprehensive evaluation of reference material performance to commercial plant goal fluence levels will be complete by 1985. The National Advanced Alloy Development Program was instituted in 1974 with the objective to develop, by 1986, advanced cladding and duct materials compatible with advanced fuel systems having peak burnup capabilities up to 150 MWD/kg and doubling times of 15 years or less. Screening of a large number of potential alloys was completed in mid-1975, and there are presently 16 candidate alloys under active investigation

Microstructure and wear-resistance of laser clad TiC particle-reinforced coating

NARCIS (Netherlands)

Lei, T.C.; Ouyang, J.H.; Pei, Y.T.; Zhou, Y.

A TiC-Ni alloy composite coating was clad to 1045 steel substrate using a 2kW CO2 laser. The microstructural constituents of the clad layer are found to be gamma-Ni and TiCp in the dendrites, and a fine eutectic of gamma-Ni plus (Fe, Cr)(23)C-6 in the interdendritic areas. Partial dissolution and

Laser and Pressure Resistance Weld of Thin-Wall Cladding for LWR Accident-Tolerant Fuels

Science.gov (United States)

Gan, J.; Jerred, N.; Perez, E.; Haggard, D. C.

2018-02-01

FeCrAl alloy with typical composition of approximately Fe-15Cr-5Al is considered a primary candidate cladding material for light water reactor accident-tolerant fuel because of its superior resistance to oxidation in high-temperature steam compared with Zircaloy cladding. Thin-walled FeCrAl cladding at 350 μm wall thickness is required, and techniques for joining endplug to cladding need to be developed. Fusion-based laser weld and solid-state joining with pressure resistance weld were investigated in this study. The results of microstructural characterization, mechanical property evaluation by tensile testing, and hydraulic pressure burst testing of the welds for the cladding-endplug specimen are discussed.

Iridium: failures & successes

Science.gov (United States)

Christensen, CarissaBryce; Beard, Suzette

2001-03-01

This paper will provide an overview of the Iridium business venture in terms of the challenges faced, the successes achieved, and the causes of the ultimate failure of the venture — bankruptcy and system de-orbit. The paper will address technical, business, and policy issues. The intent of the paper is to provide a balanced and accurate overview of the Iridium experience, to aid future decision-making by policy makers, the business community, and technical experts. Key topics will include the history of the program, the objectives and decision-making of Motorola, the market research and analysis conducted, partnering strategies and their impact, consumer equipment availability, and technical issues — target performance, performance achieved, technical accomplishments, and expected and unexpected technical challenges. The paper will use as sources trade media and business articles on the Iridium program, technical papers and conference presentations, Wall Street analyst's reports, and, where possible, interviews with participants and close observers.

Activity of iridium-ruthenium and iridium-rhodium adsorption catalysts in decomposition of hydrogen peroxide

Energy Technology Data Exchange (ETDEWEB)

Zubovich, I A; Mikhaylov, V A; Migulina, N N [Yaroslavskij Politekhnicheskij Inst. (USSR)

1976-06-01

Experimental data for the activities of iridium-ruthenium and iridium-rhodium adsorption catalysts in the decomposition of hydrogen peroxide are considered and the results of magnetic susceptibility measurements are presented. It is concluded that surface structures (complexes) may be formed and that micro-electronic feaures play a role in heterogeneous catalysis.

Effect of Mo and nano-Nd{sub 2}O{sub 3} on the microstructure and wear resistance of laser cladding Ni-based alloy coatings

Energy Technology Data Exchange (ETDEWEB)

Ding, Lin; Hu, Shengsun; Shen, Junqi [Tianjin University, Tianjin Key Laboratory of Advanced Joining Technology, School of Materials Science and Engineering, Tianjin (China); Quan, Xiumin [Lu' an Vocation Technology College, School of Automobile and Mechanical and Electrical Engineering, Lu' an (China)

2016-04-15

Three kinds of coatings were successfully prepared on Q235 steel by laser cladding technique through adulterating with Mo and nano-Nd{sub 2}O{sub 3} into Ni-based alloys. The effect of Mo and nano-Nd{sub 2}O{sub 3} on the microstructure and properties of Ni-based coatings was investigated systematically by means of optical microscopy, X-ray diffraction, scanning electron microscopy, energy-dispersive spectroscopy, and microhardness testing and wear testing. The results indicated a certain amount of fine grains and polygonal equiaxed grains synthesized after adding Mo and nano-Nd{sub 2}O{sub 3}. Both the microhardness and wear resistance of Ni-based coatings improved greatly with a moderate additional amount of Mo and nano-Nd{sub 2}O{sub 3}. The largest improvement in microhardness was 31.9 and 14.7 %, and the largest reduction in loss was 45.0 and 30.7 %, respectively, for 5.0 wt% Mo powders and 1.0 wt% nano-Nd{sub 2}O{sub 3}. The effect of Mo on microhardness and wear resistance of laser cladding Ni-based alloy coatings is greater than the effect of nano-Nd{sub 2}O{sub 3}. (orig.)

Effect of CeO2 on Microstructure and Wear Resistance of TiC Bioinert Coatings on Ti6Al4V Alloy by Laser Cladding

OpenAIRE

Chen, Tao; Liu, Defu; Wu, Fan; Wang, Haojun

2017-01-01

To solve the lack of wear resistance of titanium alloys for use in biological applications, various prepared coatings on titanium alloys are often used as wear-resistant materials. In this paper, TiC bioinert coatings were fabricated on Ti6Al4V by laser cladding using mixed TiC and ZrO2 powders as the basic pre-placed materials. A certain amount of CeO2 powder was also added to the pre-placed powders to further improve the properties of the TiC coatings. The effects of CeO2 additive on the ph...

FeCrAl/Zr dual layer fuel cladding for improved safety margin under accident scenario

International Nuclear Information System (INIS)

Park, D.J.; Park, J.H.; Jung, Y.I.; Kim, H.G.; Park, J.Y.; Koo, Y.H.

2014-01-01

For application of advanced steel as a cladding material in light water reactor (LWR), FeCrAl/Zr dual layer tube was manufactured by using a hot isostatic pressing (HIP) method. To optimize HIP condition for joining both FeCrAl and Zr alloys, HIP was carried out under various temperature conditions. Tensile test and 3-point bend test performed for measuring mechanical properties of HIPed sample. To better understand microstructural characteristics in interface region between two alloys, SEM and TEM study were conducted by using HIPed sample with different process conditions. Based on this optimization study and analyzed results, optimized HIP condition was determined and FeCrAl/Zr dual layer fuel cladding having same wall thickness with current LWR fuel cladding was manufactured. Simulated loss-of-coolant accident test was carried out using FeCrAl/Zr dual layer cladding sample and fuel integrity was measured by mechanical test. (authors)

Influence of impurities and ion surface alloying on the corrosion resistance of E110 alloy

International Nuclear Information System (INIS)

Kalin, B. A.; Volkov, N. V.; Valikov, R. A.; Novikov, V. V.; Markelov, V. A.; Pimenov, Yu. V.

2013-01-01

The corrosion resistance of zirconium alloys depends on their structural-phase state, the type of core coolant and operating factors. The formation of a protective oxide film on the zirconium alloys is sensitive to the content of impurity atoms present in the charge base of alloys and accumulating in them in the manufacture of products. The impurity composition of the initial zirconium is determined by the method of its manufacture and generally remains unchanged in the products, deter-mining their properties, including their corrosion resistance. An increased content of impurities (C, N, Al, Mo, Fe) both individually and in their combination negatively affects the corrosion resistance of zirconium and its alloys. One of the potentially effective methods to increase the protective properties of oxide films on zirconium alloys is a surface alloying using the regime of mixing the atoms of a film, preliminarily coated on the surface, and the atoms of a target. This method makes it possible to form a given structural-phase state in the thin surface layer with unique physicochemical properties and thus to in-crease the corrosion resistance and wear resistance of fuel claddings. In this context, the object of investigation was samples of cladding tubes from alloy E110 with various content of impurity elements (nitrogen, aluminum, and carbon) with the aim to reduce the negative influence of impurities on the corrosion resistance by changing the structural-phase state of the surface layer of fuel claddings and fuel assembly components with alloying in the regime of ion mixing of atoms

In-pile test results of HANA claddings in Halden research reactor

Energy Technology Data Exchange (ETDEWEB)

Baek, Jong Hyuk; Choi, Byoung Kwon; Jeong, Yong Hwan; Jung, Yun Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2005-07-01

It is a kind of facing tasks in the nuclear industry to develop advanced claddings for high burn-up fuel which is safer and more economical than the existing conventional ones. Since 1997, taking an initiative in KAERI, the Zr cladding development team has carried out the R and D activities for the development of the advanced claddings to be used in the high burn-up fuel (>70,000 MWD.MTU). The team had produced the advanced claddings (HANA, High-performance Alloy for Nuclear Application) from the patented composition and manufacturing process in the international collaboration with U.S. and Japan. Now, the HANA claddings have being demonstrated their good performances from the out-of-pile tests including the corrosion, creep, burst, tensile, microstructures LOCA, RIA, wear, and so on. In parallel to the out-of-pile performance tests, the HANA claddings are being undertaken to evaluate their in-pile properties in Halden research reactor. In this study, it is included the test overviews, conditions, and results of the HANA claddings in the Halden reactor.

Electron beam welding of iridium heat source capsules

International Nuclear Information System (INIS)

Mustaleski, T.M.; Yearwood, J.C.; Burgan, C.E.; Green, L.A.

1991-01-01

The development of the welding procedures for the production of DOP-26 iridium alloy cups for heat source encapsulation is described. All the final assembly welds were made using the electron beam welding process. The welding of the 0.13-mm weld shield required the use of computer controlled X-Y table and a run-off tab. Welding of the frit vent to the cup required that a laser weld be made to hold the frit assembly edges together for the final electron beam weld. Great care is required in tooling design and beam placement to achieve acceptable results. Unsuccessful attempts to use laser beam welding for heat shield butt weld are discussed

A comparative study on the fretting wear properties of advanced zirconium fuel cladding materials

International Nuclear Information System (INIS)

Lee, Young Ho; Kim, Hyung Kyu; Park, Jeong Yong; Kim, Jun Hwan

2005-06-01

Fretting wear tests were carried out in room and high temperature water in order to evaluate the wear properties of new zirconium nuclear fuel claddings (K2∼K6) and the commercial claddings (M5, zirlo and zircaloy-4). The objective is to compare the wear resistance of K2∼K6 claddings with that of the commercial ones at the same test condition. After the wear tests, the average wear volume and the maximum wear depth were evaluated and compared at each test condition. As a result, it is difficult to select the most wear-resistant cladding between the K2∼K6 claddings and the commercial ones. This is because the average wear volume and maximum depth of each cladding included between the scattering range of measured results. However, wear resistance of the tested claddings based on the average wear volume and maximum wear depth could be summarized as follows: K5 > zircaloy-4 > (K2,K3) > (K4,M5) > K6 > zirlo at room temperature, zircaloy-4 > K5 > (K3,K4,zirlo) > (K2,K6) > M5 at high temperature and pressure. Therefore, it is concluded that K5 cladding among the tested new zirconium alloys has relatively higher wear-resistance in room and high temperature condition. In order to examine the wear mechanism, it is necessary to systematically study with the consideration of the alloying element effect and test environment. In this report, the wear test procedure and the wear evaluation method are described in detail

Effect of CeO2 and Y2O3 on microstructure, bioactivity and degradability of laser cladding CaO-SiO2 coating on titanium alloy.

Science.gov (United States)

Li, H C; Wang, D G; Chen, C Z; Weng, F

2015-03-01

To solve the lack of strength of bulk biomaterials for load-bearing applications and improve the bioactivity of titanium alloy (Ti-6Al-4V), CaO-SiO2 coatings on titanium alloy were fabricated by laser cladding technique. The effect of CeO2 and Y2O3 on microstructure and properties of laser cladding coating was analyzed. The cross-section microstructure of ceramic layer from top to bottom gradually changes from cellular-dendrite structure to compact cellular crystal. The addition of CeO2 or Y2O3 refines the microstructure of the ceramic layer in the upper and middle regions. The refining effect on the grain is related to the kinds of additives and their content. The coating is mainly composed of CaTiO3, CaO, α-Ca2(SiO4), SiO2 and TiO2. Y2O3 inhibits the formation of CaO. After soaking in simulated body fluid (SBF), the calcium phosphate layer is formed on the coating surface, indicating the coating has bioactivity. After soaking in Tris-HCl solution, the samples doped with CeO2 or Y2O3 present a lower weight loss, indicating the addition of CeO2 or Y2O3 improves the degradability of laser cladding sample. Copyright © 2015 Elsevier B.V. All rights reserved.

Effects of spacers on blockage of coolant channels in clad melting accidents

Energy Technology Data Exchange (ETDEWEB)

Eggen, D. T.; Scale, T.; Hsieh, S. [Northwestern Univ., Evanston, IL (United States). The Technological Inst.

1977-07-01

The elements and configuration of these assemblies are representative of the current design for a GCFR. The fuel elements are stainless-steel clad, mixed-oxide spaced by a grid structure on 250 mm centers with a pitch of 9.5 mm, diameter, 7.2 mm, and cladding thickness, 0.5 m. Three series of experiments have been conducted to study the flow and disposition of molten cladding metal into a lower powered blanket region of the reactor following a loss of flow situation. The first two series used a simulant fuel-element bundle to simplify the experimental procedure and make visual observation possible. The 'fuel' was simulated by mullite rods 6.4 mm in diameter and 610 mm long. These were clad with a 50 Pb/50 Sn alloy tubing which was drawn onto the 'fuel'. The first series used cast spacers with webs of about 0.5-0.55 mm thickness placed 175 and 425 mm from the top end of the assembly. The second series used grid spacers fabricated of 0.25 mm alloy strips. This provided a more accurate representation of the hydraulic diameter. The bundle was encased in a hexagonal glass tube. The bundle was at 22/sup 0/C and the molten alloy was poured at a temperature of 260/sup 0/C (35/sup 0/C superheat). Motion pictures recorded the experiments and the bundle was sectioned for observation. The third set of experiments was done with a stainless steel bundle of 37 elements fabricated of mullite rods, 7.14 mm diameter. The stainless steel cladding had an O.D. of 8.41 mm. The element pitch was 11.1 mm. The grid spacers were prototypic. The experiment was conducted in an inert-gas tube furnace. The 'core fuel' cladding was melted in an induction furnace and the molten liquid flowed through the center seven element channels. X-ray pictures were taken after the tests and the bundle was sectioned for further study.

Lanthanide based conversion coatings for long term wet storage of aluminium-clad spent fuel

International Nuclear Information System (INIS)

Fernandes, S.M.C.; Correa, O.V.; De Souza, J.A.; Ramanathan, L.V.

2010-01-01

Spent fuels from research reactors are stored in basins with water of less than desirable quality at many facilities around the world and instances of cladding failure caused by pitting corrosion have been reported. Conversion coatings have been used in many industries to protect different metals, including aluminium alloys. This paper presents the results of an ongoing investigation in which the corrosion resistance of lanthanide (cerium, lanthanum and praseodymium) based conversion coated RR fuel cladding alloys has been studied. Electrochemical tests in the laboratory revealed higher corrosion resistance of CeO 2 , La 2 O 3 and Pr 2 O 3 coated AA 1100 and AA 6061 alloys in NaCl solutions. Uncoated and CeO 2 coated coupons of these alloys exposed for 50 days to the spent fuel basin of the IEA-R1 research reactor in IPEN, Brazil, revealed marked reductions in the extent of pitting corrosion. (author)

Iridium Interfacial Stack (IRIS)

Science.gov (United States)

Spry, David James (Inventor)

2015-01-01

An iridium interfacial stack ("IrIS") and a method for producing the same are provided. The IrIS may include ordered layers of TaSi.sub.2, platinum, iridium, and platinum, and may be placed on top of a titanium layer and a silicon carbide layer. The IrIS may prevent, reduce, or mitigate against diffusion of elements such as oxygen, platinum, and gold through at least some of its layers.

Accident tolerant clad material modeling by MELCOR: Benchmark for SURRY Short Term Station Black Out

International Nuclear Information System (INIS)

Wang, Jun; Mccabe, Mckinleigh; Wu, Lei; Dong, Xiaomeng; Wang, Xianmao; Haskin, Troy Christopher; Corradini, Michael L.

2017-01-01

Highlights: • Thermo-physical and oxidation kinetics properties calculation and analysis of FeCrAl. • Properties modelling of FeCrAl in MELCOR. • Benchmark calculation of Surry nuclear power plant. - Abstract: Accident tolerant fuel and cladding materials are being investigated to provide a greater resistance to fuel degradation, oxidation and melting if long-term cooling is lost in a Light Water Reactor (LWR) following an accident such as a Station Blackout (SBO) or Loss of Coolant Accident (LOCA). Researchers at UW-Madison are analyzing an SBO sequence and examining the effect of a loss of auxiliary feedwater (AFW) with the MELCOR systems code. Our research work considers accident tolerant cladding materials (e.g., FeCrAl alloy) and their effect on the accident behavior. We first gathered the physical properties of this alternative cladding material via literature review and compared it to the usual zirconium alloys used in LWRs. We then developed a model for the Surry reactor for a Short-term SBO sequence and examined the effect of replacing FeCrAl for Zircaloy cladding. The analysis uses MELCOR, Version 1.8.6 YR, which is developed by Idaho National Laboratory in collaboration with MELCOR developers at Sandia National Laboratories. This version allows the user to alter the cladding material considered, and our study examines the behavior of the FeCrAl alloy as a substitute for Zircaloy. Our benchmark comparisons with the Sandia National Laboratory’s analysis of Surry using MELCOR 1.8.6 and the more recent MELCOR 2.1 indicate good overall agreement through the early phases of the accident progression. When FeCrAl is substituted for Zircaloy to examine its performance, we confirmed that FeCrAl slows the accident progression and reduce the amount of hydrogen generated. Our analyses also show that this special version of MELCOR can be used to evaluate other potential ATF cladding materials, e.g., SiC as well as innovative coatings on zirconium cladding

Accident tolerant clad material modeling by MELCOR: Benchmark for SURRY Short Term Station Black Out

Energy Technology Data Exchange (ETDEWEB)

Wang, Jun, E-mail: jwang564@wisc.edu [College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States); Mccabe, Mckinleigh [College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States); Wu, Lei [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Dong, Xiaomeng [College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States); Wang, Xianmao [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Haskin, Troy Christopher [College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States); Corradini, Michael L., E-mail: corradini@engr.wisc.edu [College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States)

2017-03-15

Highlights: • Thermo-physical and oxidation kinetics properties calculation and analysis of FeCrAl. • Properties modelling of FeCrAl in MELCOR. • Benchmark calculation of Surry nuclear power plant. - Abstract: Accident tolerant fuel and cladding materials are being investigated to provide a greater resistance to fuel degradation, oxidation and melting if long-term cooling is lost in a Light Water Reactor (LWR) following an accident such as a Station Blackout (SBO) or Loss of Coolant Accident (LOCA). Researchers at UW-Madison are analyzing an SBO sequence and examining the effect of a loss of auxiliary feedwater (AFW) with the MELCOR systems code. Our research work considers accident tolerant cladding materials (e.g., FeCrAl alloy) and their effect on the accident behavior. We first gathered the physical properties of this alternative cladding material via literature review and compared it to the usual zirconium alloys used in LWRs. We then developed a model for the Surry reactor for a Short-term SBO sequence and examined the effect of replacing FeCrAl for Zircaloy cladding. The analysis uses MELCOR, Version 1.8.6 YR, which is developed by Idaho National Laboratory in collaboration with MELCOR developers at Sandia National Laboratories. This version allows the user to alter the cladding material considered, and our study examines the behavior of the FeCrAl alloy as a substitute for Zircaloy. Our benchmark comparisons with the Sandia National Laboratory’s analysis of Surry using MELCOR 1.8.6 and the more recent MELCOR 2.1 indicate good overall agreement through the early phases of the accident progression. When FeCrAl is substituted for Zircaloy to examine its performance, we confirmed that FeCrAl slows the accident progression and reduce the amount of hydrogen generated. Our analyses also show that this special version of MELCOR can be used to evaluate other potential ATF cladding materials, e.g., SiC as well as innovative coatings on zirconium cladding

Water chemistry regimes for VVER-440 units: water chemistry influence on fuel cladding behaviour

International Nuclear Information System (INIS)

Zmitko, M.

1999-01-01

In this lecture next problems of water chemistry influence on fuel cladding behaviour for VVER-440 units are presented: primary coolant technologies; water chemistry specification and control; fuel integrity considerations; zirconium alloys cladding corrosion (corrosion versus burn-up; water chemistry effect; crud deposition; hydrogen absorption; axial offset anomaly); alternatives for the primary coolant regimes

Annealing Effect on Corrosion Behavior of the Beta-Quenched HANA Alloy

International Nuclear Information System (INIS)

Kim, Hyun Gil; Kim, Il Hyun; Choi, Byung Kwan; Park, Sang Yoon; Park, Jeong Yong; Jeong, Yong Hwan

2009-01-01

The advanced fuel cladding materials named as HANA cladding have been developed at KAERI for application of high burn-up and that cladding showed an improved performance in both in-pile and out-of-pile conditions. However, the cladding performance could be changed by the annealing conditions during the tube manufacturing process. Especially, the corrosion resistance is considerably sensitive to their microstructure which is determined by a manufacturing process in the high Nb-containing zirconium alloys. They reported that the corrosion properties of the Nb-containing Zr alloys were considerably affected by the microstructure conditions such as the Nb concentration in the matrix and the second phase types. Therefore, the corrosion behavior of HANA cladding having the high Nb could be considerably affected by the annealing time and temperatures. The purpose of this study is focused on the annealing effect of the beta quenched HANA alloy to obtain the optimum annealing conditions

Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-01

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designs allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.

Iridium complexes for electrocatalysis

Science.gov (United States)

Sheehan, Stafford Wheeler; Hintermair, Ulrich; Thomsen, Julianne M; Brudvig, Gary W; Crabtree, Robert H

2017-10-17

Solution-phase (e.g., homogeneous) or surface-immobilized (e.g., heterogeneous) electrode-driven oxidation catalysts based on iridium coordination compounds which self-assemble upon chemical or electrochemical oxidation of suitable precursors and methods of making and using thereof are. Iridium species such as {[Ir(LX).sub.x(H.sub.2O).sub.y(.mu.-O)].sub.z.sup.m+}.sub.n wherein x, y, m are integers from 0-4, z and n from 1-4 and LX is an oxidation-resistant chelate ligand or ligands, such as such as 2(2-pyridyl)-2-propanolate, form upon oxidation of various molecular iridium complexes, for instance [Cp*Ir(LX)OH] or [(cod)Ir(LX)] (Cp*=pentamethylcyclopentadienyl, cod=cis-cis,1,5-cyclooctadiene) when exposed to oxidative conditions, such as sodium periodate (NaIO.sub.4) in aqueous solution at ambient conditions.

Property Investigation of Laser Cladded, Laser Melted and Electron Beam Melted Ti-Al6-V4

Science.gov (United States)

2006-05-01

UNCLASSIFIED/UNLIMITED UNCLASSIFIED/UNLIMITED Figure 3: Examples of electron beam melted net shape parts; powder bed [3]. 1.4 Laser Cladding ...description, www.arcam.com. [4] K.-H. Hermann, S. Orban, S. Nowotny, Laser Cladding of Titanium Alloy Ti6242 to Restore Damaged Blades, Proceedings...Property Investigation of Laser Cladded , Laser Melted and Electron Beam Melted Ti-Al6-V4 Johannes Vlcek EADS Deutschland GmbH Corporate Research

Method for the protection of the cladding tubes of fuel rods

International Nuclear Information System (INIS)

Steinberg, E.

1978-01-01

To present stress crack corrosion and to protect the cladding tubes of the fuel rods made of a circonium alloy from attack by iodine, the inward surfaces are provided with protective coatings. Therefore the casting tubes already filled with fuel element pellets are put under over-pressure at a temperature range between 300 and 500 0 C, until almost yield-point is reached. A small amount of H 2 O or H 2 O 2 , filled in, reacts with the cladding tube material to form the Zr-O 2 protective coating. Afterwards comes a pressure relief, and the cladding tube reaches its original dimensions. (DG) [de

Corrosion issues in the long term storage of aluminum-clad spent nuclear fuels

International Nuclear Information System (INIS)

Louthan, M.R. Jr.; Peacock, H.B. Jr.; Sindelar, R.L.; Iyer, N.C.

1996-01-01

Approximately 8% of the spent nuclear fuel owned by the US Department of Energy is clad with aluminum alloys. The spent fuel must be either reprocessed or temporarily stored in wet or dry storage systems until a decision is made on final disposition in a repository. There are corrosion issues associated with the aluminum cladding regardless of the disposition pathway selected. This paper discusses those issues and provides data and analysis to demonstrate that control of corrosion induced degradation in aluminum clad spent fuels can be achieved through relatively simple engineering practices

Reactor fuel cladding tube with excellent corrosion resistance and method of manufacturing the same

International Nuclear Information System (INIS)

Okuda, Takanari; Kanehara, Mitsuo; Abe, Katsuhiro; Nishimura, Takashi.

1995-01-01

The present invention provides a fuel cladding tube having an excellent corrosion resistance and thus a long life, and a suitable manufacturing method therefor. Namely, in the fuel cladding tube, the outer circumference of an inner layer made of a zirconium base alloy is coated with an outer layer made of a metal more corrosion resistant than the zirconium base alloy. Ti or a titanium alloy is suitable for the corrosion resistant metal. In addition, the outer layer can be coated by a method such as vapor deposition or plating, not limited to joining of the inner layer material and the outer layer material. Specifically, a composite material having an inner layer made of a zirconium alloy coated by the outer material made of a titanium alloy is applied with hot fabrication at a temperature within a range of from 500 to 850degC and at a fabrication rate of not less than 5%. The fabrication method includes any of extrusion, rolling, drawing, and casting. As the titanium-base alloy, a Ti-Al alloy or a Ti-Nb alloy containing Al of not more than 20wt%, or Nb of not more than 20wt% is preferred. (I.S.)

Effect of Sr addition on the characteristics of as-cast and rolled 3003/4004 clad aluminum

Energy Technology Data Exchange (ETDEWEB)

Yan, Guangyuan; Mao, Feng; Jie, Jinchuan; Cao, Zhiqiang, E-mail: caozq@dlut.edu.cn; Li, Tingju; Wang, Tongmin, E-mail: tmwang@dlut.edu.cn

2016-09-05

This paper examines the effects of Sr addition on the microstructure, composition distribution and Vickers hardness in the interfacial region of the as-cast and rolled 3003/4004 clad aluminum. The results reveal that the optimum adding amount of Sr on the as-cast Al-1.2Mn/Al−10Si-xSr clad is 0.08 wt%. With Sr content increasing from 0 to 0.08 wt%, the average length and number of the primary α-Al phase growing from the diffusion layer significantly decreased and whose morphology appears in columar dendritic crystals, the celluar dendrite crystals, deep celluar crystals, fine celluar crystals and planar crystals, while the dendritic-crystal primary α-Al phase nucleating and growing from inner Al−Si alloy side also show obvious decease in secondary dendrite spacing; meanwhile, eutectic Si phases were gradually modified from coarse plates, coralloid-plates mixed structure to fine branchy coralloid structure in three-dimensional morphology. After rolling, the diffusion layer thickness of the Al-1.2Mn/Al−10Si−0.08Sr clad is decreased by 66.7%, compared to that of unmodified clad alloy. This decreased diffusion layer thickness may be determined by augmented plastic strain and restraining diffusion of Si atoms in diffusion layer. Morever, average Vickers hardness on interface and Al−Si side of the Al-1.2Mn/Al−10Si−0.08Sr clad showed slight increase and more uniform distribution than that of unmodified clad alloy. This uniform distribution and improved hardness primarily attribute to presence of fine branchy coralloid silicon phase and its stronger dispersion strengthening as well as solution strengthening caused by interdiffusion of Si, Mn and Sr elements. - Highlights: • 3003/4004 clad aluminum was firstly modified by various Sr addition levels. • The optimum adding amount of Sr on the Al−1.2Mn/Al−10Si−xSr clad is 0.08 wt%. • Sr can refine primary α-Al and eutectic silicon phase of the clad simultaneously. • The Sr-modified rolled clad has

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.

Corrosion behaviour of E110- and E635- type zirconium alloys under PWR irradiation simulating conditions

International Nuclear Information System (INIS)

Markelov, V.A.; Novikov, V.V.; Kon'kov, V.F.; Tselishchev, A.V.; Dologov, A.B.; Zmitko, M.; Maserik, V.; Kocik, J.

2008-01-01

As structural materials for VVER 1000 fuel rod claddings and FA components use is made of zirconium alloys E110 (Zr 1Nb) and E635 (Zr 1.2Sn 1Nb 0.35Fe) that meet the design parameters of operation. Nonetheless, the work is in progress to perfect those alloys to reach higher corrosion and shape change resistance. At VNIINM updated E110M and E635M alloys have been developed on E110 and E635 bases. To assess the corrosion behaviour of the updated alloys in comparison to the base alloys their cladding samples were tested in RVS 3 loop of LWR 15 reactor (NRI, Rez) in PWR water chemistry with coolant surface and volume boiling. The data are presented on the influence effected by in pile irradiation for up to 324 days on oxide coat thickness and microstructure of fuel claddings produced from the four tested alloys. It has been revealed that E110 alloy its updated version E110M and E635M alloy compared to E635 have higher corrosion resistances. The paper discusses th+e results on the in pile corrosion of cladding samples from the alloys under study in comparison to the results acquired for similar samples tested in LWR 15 inactive channel and under autoclave conditions. Using methods of TEM, EDX analyses of extraction replicas dislocation structure and phase composition changes were studied in samples of all four alloy claddings LWR 15 reactor irradiated to the material damage dose of 1.5 dpa. The interrelation is discussed between irradiation effected strengthening and corrosion of fuel claddings made of E110 and E635 type zirconium alloys and the evolution of their structure and phase states

Dilution and Ferrite Number Prediction in Pulsed Current Cladding of Super-Duplex Stainless Steel Using RSM

Science.gov (United States)

Eghlimi, Abbas; Shamanian, Morteza; Raeissi, Keyvan

2013-12-01

Super-duplex stainless steels have an excellent combination of mechanical properties and corrosion resistance at relatively low temperatures and can be used as a coating to improve the corrosion and wear resistance of low carbon and low alloy steels. Such coatings can be produced using weld cladding. In this study, pulsed current gas tungsten arc cladding process was utilized to deposit super-duplex stainless steel on high strength low alloy steel substrates. In such claddings, it is essential to understand how the dilution affects the composition and ferrite number of super-duplex stainless steel layer in order to be able to estimate its corrosion resistance and mechanical properties. In the current study, the effect of pulsed current gas tungsten arc cladding process parameters on the dilution and ferrite number of super-duplex stainless steel clad layer was investigated by applying response surface methodology. The validity of the proposed models was investigated by using quadratic regression models and analysis of variance. The results showed an inverse relationship between dilution and ferrite number. They also showed that increasing the heat input decreases the ferrite number. The proposed mathematical models are useful for predicting and controlling the ferrite number within an acceptable range for super-duplex stainless steel cladding.

Surface improvement for inside surface of small diameter pipes by laser cladding technique

International Nuclear Information System (INIS)

Irisawa, Toshio; Morishige, Norio; Umemoto, Tadahiro; Ono, Kazumichi; Hamaoka, Tadashi; Tanaka, Atsushi

1991-01-01

A laser cladding technique has been used for surface improvement in controlling the composition of a metal surface. Recent high power YAG laser development gives an opportunity to use this laser cladding technique for various applications. A YAG laser beam can be transmitted through an optical fiber for a long distance and through narrow spaces. YAG laser cladding was studied for developing alloy steel to prevent stress corrosion cracking in austenitic stainless steel piping. In order to make a cladding layer, mixed metal powder was on the inside surface of the piping using an organic binder. Subsequently the powder beds were melted with a YAG laser beam transmitted through an optical fiber. This paper introduces the Laser cladding technique for surface improvement for the inside surface of a small diameter pipe. (author)

Microscopic Analysis and Electrochemical Behavior of Fe-Based Coating Produced by Laser Cladding

Directory of Open Access Journals (Sweden)

Jinlin Chen

2017-10-01

Full Text Available The effect of laser cladding on the surface microstructure and corrosion properties of coated/uncoated specimens were investigated. Fe-based alloy coating was produced on 35CrMo steel by laser cladding. The phase composition, microstructure, interface element distribution, microhardness and corrosion resistance of the cladding coating were measured. The results show that the cladding layer is mainly composed of α-Fe phases, the microstructure presents a gradient distribution, and a good metallurgical bond is formed at the boundary with the substrate. Microhardness profiles show that the average microhardness of the cladding coating is about 2.1 times higher than that of the uncoated specimen. In addition, the electrochemical results show that the coated specimen exhibits far better corrosion resistance than to the uncoated specimen.

Cladding of Ni superalloy powders on AISI 4140 steel with concentrated solar energy

Energy Technology Data Exchange (ETDEWEB)

Fernandez, B.J.; Lopez, V.; Vazquez, A.J. [Centro Nacional de Investigaciones Metalurgicas, CENIM-CSIC, Madrid (Spain); Martinez, D. [Plataforma Solar de Almeria, Tabernas Almeria (Spain)

1998-05-12

The present work deals with Ni alloy cladding on AISI 4140 steel samples made with high power density concentrated solar beams. The quality of the cladding is high concerning the adherence, low dilution and high hardness of the coating. Some considerations are presented concerning the future of high power density beams related to SUrface Modification of Metallic mAterials with SOLar Energy (SUMMA cum SOLE)

Advanced LMFBR fuel cladding susceptability to stress corrosion due to reprocessing impurities

International Nuclear Information System (INIS)

Henslee, S.P.

1987-03-01

The potential degradation of LMFBR fuel cladding alloys by chlorides, when used in metallic fuel systems, was evaluated. The alloys tested were D-9 and HT-9 stainless steels, austenitic and ferritic alloys respectively. These two alloys were tested in parallel with and their performance compared to the austenitic stainless steel Type 316. All alloys were tested for 7400 hours in a stress rupture environment with chloride exposure at either 550/degree/C 650/degree/C. None of the alloys tested were found to exhibit any degradation in time-to-rupture by the presence of chlorides under the conditions imposed during testing. 8 refs., 4 figs., 2 tabs

A Critical Review of High Entropy Alloys and Related Concepts (Postprint)

Science.gov (United States)

2016-10-21

1301e1305. [306] C. Huang, Y.Z. Zhang, R. Vilar, J.Y. Shen, Dry sliding wear behavior of laser clad TiVCrAlSi high entropy alloy coatings on Ti-6Al...Res. 652e654 (2013) 1115e1118. [313] C. Huang, Y. Zhang, J. Shen, R. Vilar, Thermal stability and oxidation resis- tance of laser clad TiVCrAlSi high...Section 7.1.3). Finally, it is not always true that SS alloys are ductile. Well-known examples include b- titanium alloys, a- titanium alloys with small

High performance fuel technology development : Development of high performance cladding materials

International Nuclear Information System (INIS)

Park, Jeongyong; Jeong, Y. H.; Park, S. Y.

2012-04-01

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

Producing titanium-niobium alloy by high energy beam

Energy Technology Data Exchange (ETDEWEB)

Sharkeev, Yu. P., E-mail: sharkeev@ispms.tsc.ru [Institute of Strength Physics and Materials Science, SB RAS, 2/4 Akademicheski Prosp., Tomsk, 634055 (Russian Federation); National Research Tomsk Polytechnic University, 30 Lenin Av., Tomsk, 634050 (Russian Federation); Golkovski, M. G., E-mail: golkoski@mail.ru [Budker Institute of Nuclear Physics, 11 Akademika Lavrentiev Prosp., Novosibirsk, 630090 (Russian Federation); Glukhov, I. A., E-mail: gia@ispms.tsc.ru; Eroshenko, A. Yu., E-mail: eroshenko@ispms.tsc.ru; Fortuna, S. V., E-mail: s-fortuna@mail.ru [Institute of Strength Physics and Materials Science, SB RAS, 2/4 Akademicheski Prosp., Tomsk, 634055 (Russian Federation); Bataev, V. A., E-mail: bataev@vadm.ustu.ru [Novosibirsk State Technical University, 20 K. Marx Prosp., Novosibirsk, 630073 (Russian Federation)

2016-01-15

The research is involved in producing a Ti-Nb alloy surface layer on titanium substrate by high energy beam method, as well as in examining their structures and mechanical properties. Applying electron-beam cladding it was possible to produce a Ti-Nb alloy surface layer of several millimeters, where the niobium concentration was up to 40% at. and the structure itself could be related to martensite quenching structure. At the same time, a significant microhardness increase of 3200-3400 MPa was observed, which, in its turn, is connected with the formation of martensite structure. Cladding material of Ti-Nb composition could be the source in producing alloys of homogeneous microhardness and desired concentration of alloying niobium element.

Underwater laser beam welding of Alloy 690

International Nuclear Information System (INIS)

Hino, Takehisa; Tamura, Masataka; Kono, Wataru; Kawano, Shohei; Yoda, Masaki

2009-01-01

Stress Corrosion Clacking (SCC) has been reported at Alloy 600 welds between nozzles and safe-end in Pressurized Water Reactor (PWR) plant. Alloy 690, which has higher chromium content than Alloy 600, has been applied for cladding on Alloy 600 welds for repairing damaged SCC area. Toshiba has developed Underwater Laser Beam Welding technique. This method can be conducted without draining, so that the repairing period and the radiation exposure during the repair can be dramatically decreased. In some old PWRs, high-sulfur stainless steel is used as the materials for this section. It has a high susceptibility of weld cracks. Therefore, the optimum welding condition of Alloy 690 on the high-sulfur stainless steel was investigated with our Underwater Laser Beam Welding unit. Good cladding layer, without any crack, porosity or lack of fusion, could be obtained. (author)

Effect of chemical composition on corrosion resistance of Zircaloy fuel cladding tube for BWR

International Nuclear Information System (INIS)

Inagaki, Masahisa; Akahori, Kimihiko; Kuniya, Jirou; Masaoka, Isao; Suwa, Masateru; Maru, Akira; Yasuda, Teturou; Maki, Hideo.

1990-01-01

Effects of Fe and Ni contents on nodular corrosion susceptibility and hydrogen pick-up of Zircaloy were investigated. Total number of 31 Zr alloys having different chemical compositions; five Zr-Sn-Fe-Cr alloys, eight Zr-Sn-Fe-Ni alloys and eighteen Zr-Sn-Fe-Ni-Cr alloys, were melted and processed to thin plates for the corrosion tests in the environments of a high temperature (510degC) steam and a high temperature (288degC) water. In addition, four 450 kg ingots of Zr-Sn-Fe-Ni-Cr alloys were industrially melted and BWR fuel cladding tubes were manufactured through a current material processing sequence to study their producibility, tensile properties and corrosion resistance. Nodular corrosion susceptibility decreased with increasing Fe and Ni contents of Zircaloys. It was seen that the improved Zircaloys having Fe and Ni contents in the range of 0.30 [Ni]+0.15[Fe]≥0.045 (w%) showed no susceptibility to nodular corrosion. An increase of Fe content resulted in a decrease of hydrogen pick-up fraction in both steam and water environments. An increase of Fe and Ni content of Zircaloys in the range of Fe≤0.25 w% and Ni≤0.1 w% did not cause the changes in tensile properties and fabricabilities of fuel cladding tube. The fuel cladding tube of improved Zircaloy, containing more amount of Fe and Ni than the upper limit of Zircaloy-2 specification showed no susceptibility to nodular corrosion even in the 530degC steam test. (author)

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)

Park, Jeong-Yong; KIM, Hyun-Gil; JUNG, Yang-Il; PARK, Dong-Jun; KOO, Yang-Hyun

2013-01-01

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. Al 3 Ti-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

Duplex-cladding: Siemens answer to the requirements of extended burnup in PWRs

International Nuclear Information System (INIS)

Van Swam, L.F.; Sell, H.J.; Eberle, R.; Seibold, A.

1994-01-01

One important goal of nuclear fuel development is to increase the cost-effectiveness of the nuclear fuel cycle by burnup extension. A prerequisite for this goal is a cladding tube with high resistance to corrosion under the operating conditions of modern PWRs. Therefore, in the early eighties Siemens started to investigate the material behaviour of Zirconium based alloys also outside the composition range of Zry-4. The examination included out-of-pile corrosion testing in water and steam, with and without chemical addition, such as LiOH, in-pile testing of path finder fuel rods in a hot PWR up to 80 MWd/kgU and the investigation of mechanical behaviour, growth and creep under normal and the postulated conditions of a loss-of-coolant accident (LOCA). The evaluation of in-pile and out-of-pile experiments on alternative Zr-alloys revealed that improvements in corrosion resistance are frequently accompanied by undesirable changes in material properties which affect mechanical design and LOCA behaviour. To fulfill all requirements - the mechanical and corrosion related ones - and to retain the large experience base with Zry-4, a DUPLEX cladding was selected. The selected ELS DUPLEX cladding consists of a Zircaloy-4 tubing with a thin outer layer of an Extra Low tin (Sn) Zr-alloy. The ELS layer improves the stability against LiOH and allows operation with voided coolant. This advanced product has been engineered for use in highly enriched fuel assemblies in high efficiency plants operating with low neutron leakage core management and high coolant temperatures. It has become the accepted fuel rod cladding for many plants in Germany, Spain and Switzerland. (authors). 6 figs., 2 refs

The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

Energy Technology Data Exchange (ETDEWEB)

Burkes, Douglas E., E-mail: Douglas.Burkes@pnnl.gov; Casella, Amanda J.; Casella, Andrew M.

2017-04-01

The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world's highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding during fabrication and are enhanced during irradiation. One aspect of fuel development and qualification is to demonstrate an appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 °C). The mechanisms responsible for fission gas release events are discussed. - Highlights: •Complementary fission gas release events are reported for U-Mo fuel with and without cladding. •Exothermic reaction between Zr diffusion layer and cladding influences fission gas release. •Mechanisms responsible for fission gas release are similar, but with varying timing and magnitude. •Behavior of samples is similar after 800 °C signaling the onset of superlattice destabilization.

A Comparative Physics Study of Commercial PWR Cores using Metallic Micro-cell UO{sub 2}-Cr (or Mo) Pellets with Cr-based Cladding Coating

Energy Technology Data Exchange (ETDEWEB)

Hwang, Dae Hee; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of); In, Wang Kee [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In this work, a comparative neutronic analysis of the cores using ATFs which include metallic micro-cell UO{sub 2}-Cr, UO{sub 2}-Mo pellets and Cr-based alloy coating on cladding was performed to show the effects of the ATF fuels on the core performance. In this study, the cores having different ATFs use the same initial uranium enrichments. The ATF concepts studied in this work are the metallic microcell UO{sub 2} pellets containing Cr or Mo with cladding outer coating composed of Cr-based alloy which have been suggested as the ATF concepts in KAERI (Korea Atomic Energy Research Institute). The metallic micro-cell pellets and Cr-based alloy coating can enhance thermal conductivity of fuel and reduce the production of hydrogen from the reaction of cladding with coolant, respectively. The objective of this work is to compare neutronic characteristics of commercial PWR equilibrium cores utilizing the different variations of metallic micro-cell UO{sub 2} pellets with cladding coating composed of Cr-based alloy. The results showed that the cores using UO{sub 2}-Cr and UO{sub 2}-Mo pellets with Cr-based alloy coating on cladding have reduced cycle lengths by 60 and 106 EFPDs, respectively, in comparison with the reference UO{sub 2} fueled core due to the reduced heavy metal inventories and large thermal absorption cross section but they do not have any significant differences in the core performances parameters. However, it is notable that the core fueled the micro-cell UO{sub 2}-Mo pellet and Cr-based alloy coating has considerably more negative MTC and slightly more negative FTC than the other cases. These characteristics of the core using micro-cell UO{sub 2}-Mo pellet and Cr-based alloy coating is due to the hard neutron spectrum and large capture resonance cross section of Mo isotopes.

Dimensional changes in FFTF [Fast Flux Test Facility] austenitic cladding and ducts

International Nuclear Information System (INIS)

Makenas, B.J.; Chastain, S.A.; Gneiting, B.C.

1990-11-01

As the standard cladding and duct material for the Fast Flux Test Facility driver fuel, 20% cold-worked 316 stainless steel has provided good service up to a fast fluence of 16 x 10 22 n/cm 2 in extreme cases. The titanium-stabilized variant of 316 SS, called D9, has extended the useful life of the austenitic alloys by increasing the incubation fluence necessary for the onset of volumetric swelling. Duct flat-to-flat, length and bow, pin bundle distortion, fuel pin diameter and length, as well as cladding volumetric swelling have been examined for high fluence components representing both alloys. These data emphasize the importance of the swelling process, the superiority of D9, and the interrelation between deformations in the duct, bundle, and individual pins. 8 refs., 10 figs

Inconel alloy 625 clad steel for application in wet scrubber systems

International Nuclear Information System (INIS)

Morse, S.L.; Shoemaker, L.E.

1984-01-01

Test panels from INCONEL 625 clad plate were successfully installed in two wet flue gas scrubber systems. In one system INCONEL 625 clad plate was located in the roof section of the absorber just ahead of the outlet ducting. The test plates, including weld seams, showed no signs to corrosion after six months of exposure. In the other scrubber test plates located in the outlet duct of an I.D. fan house, in the stack lining, and in the absorber quench area were unattacked after nine months

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.

Development of advanced LWR fuel cladding

International Nuclear Information System (INIS)

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

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

Oxidation resistant chromium coating on Zircaloy-4 for accident tolerant fuel cladding

International Nuclear Information System (INIS)

Park, Jung-Hwan; Kim, Eui-Jung; Jung, Yang-Il; Park, Dong-Jun; Kim, Hyun-Gil; Park, Jeong-Yong; Koo, Yang-Hyun

2015-01-01

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

Oxidation resistant chromium coating on Zircaloy-4 for accident tolerant fuel cladding

Energy Technology Data Exchange (ETDEWEB)

Park, Jung-Hwan; Kim, Eui-Jung; Jung, Yang-Il; Park, Dong-Jun; Kim, Hyun-Gil; Park, Jeong-Yong; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

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.

Proposal of 99.99%-aluminum/7N01-Aluminum clad beam tube for high energy booster of Superconducting Super Collider

International Nuclear Information System (INIS)

Ishimaru, Hajime

1994-01-01

Proposal of 99.99% pure aluminum/7N01 aluminum alloy clad beam tube for high energy booster in Superconducting Super Collider is described. This aluminum clad beam tube has many good performances, but a eddy current effect is large in superconducting magnet quench collapse. The quench test result for aluminum clad beam tube is basically no problem against magnet quench collapse. (author)

Development of a laser multi-layer cladding technology for damage mitigation of fuel spacers in Hanaro reactor

International Nuclear Information System (INIS)

Kim, J. S.; Lee, D. H.; Hwang, S. S.; Suh, J. H.

2002-01-01

A laser multi-layer cladding technology was developed to mitigate the fretting wear damages occurred at fuel spacers in Hanaro reactor. The detailed experimental results are as follows. 1) Analyses of fretting wear damages and fabrication process of fuel spacers 2) Development and analysis of spherical Al 6061 T-6 alloy powders for the laser cladding 3) Analysis of parameter effects on laser cladding process for clad bids, and optimization of laser cladding process 4) Analysis on the changes of cladding layers due to overlapping factor change 5) Microstructural observation and phase analysis 6) Characterization of materials properties (hardness and wear tests) 7) Manufacture of prototype fuel spacers 8) Development of a vision system and revision of its related softwares

Corrosion behaviour of laser clad stainless steels

International Nuclear Information System (INIS)

Damborenea, J.J. de; Weerasinghe, V.M.; West, D.R.F.

1993-01-01

The present paper is focussed in the study of the properties of a clad layer of stainless steel on a mild steel. By blowing powder of the alloy into a melt pool generated by a laser of 2 KW, an homogeneous layer of 316 stainless steel can be obtained. Structure, composition and corrosion behaviour are similar to those of a stainless steel in as-received condition. (Author)

Nuclear-powered pacemaker fuel cladding study

International Nuclear Information System (INIS)

Shoup, R.L.

1976-01-01

The composite of metals and alloys used in the fabrication of 238 Pu cardiac pacemaker fuel capsules resists the effects of high temperatures, high mechanical forces, and chemical corrosives and provides more than adequate protection to the fuel pellet even from deliberate attempts to dissolve the cladding in inorganic acids. This does not imply that opening a pacemaker fuel capsule by inorganic acids is impossible but that it would not be a wise choice

Performance of HT9 clad metallic fuel at high temperature

International Nuclear Information System (INIS)

Pahl, R.G.; Lahm, C.E.; Hayes, S.L.

1992-01-01

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

Analysis of mechanical tensile properties of irradiated and annealed RPV weld overlay cladding

Energy Technology Data Exchange (ETDEWEB)

Novak, J [Czech Nuclear Society, Prague (Czech Republic)

1994-12-31

Mechanical tensile properties of irradiated and annealed outer layer of reactor pressure vessel weld overlay cladding, composed of Cr19Ni10Nb alloy, have been experimentally determined by conventional tensile testing and indentation testing. The constitutive properties of weld overlay cladding are then modelled with two homogenization models of the constitutive properties of elastic-plastic matrix-inclusion composites; numerical and experimental results are then compared. 10 refs., 4 figs., 4 tabs.

Analysis of mechanical tensile properties of irradiated and annealed RPV weld overlay cladding

International Nuclear Information System (INIS)

Novak, J.

1993-01-01

Mechanical tensile properties of irradiated and annealed outer layer of reactor pressure vessel weld overlay cladding, composed of Cr19Ni10Nb alloy, have been experimentally determined by conventional tensile testing and indentation testing. The constitutive properties of weld overlay cladding are then modelled with two homogenization models of the constitutive properties of elastic-plastic matrix-inclusion composites; numerical and experimental results are then compared. 10 refs., 4 figs., 4 tabs

Crack resistance curves determination of tube cladding material

Energy Technology Data Exchange (ETDEWEB)

Bertsch, J. [Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland)]. E-mail: johannes.bertsch@psi.ch; Hoffelner, W. [Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland)

2006-06-30

Zirconium based alloys have been in use as fuel cladding material in light water reactors since many years. As claddings change their mechanical properties during service, it is essential for the assessment of mechanical integrity to provide parameters for potential rupture behaviour. Usually, fracture mechanics parameters like the fracture toughness K {sub IC} or, for high plastic strains, the J-integral based elastic-plastic fracture toughness J {sub IC} are employed. In claddings with a very small wall thickness the determination of toughness needs the extension of the J-concept beyond limits of standards. In the paper a new method based on the traditional J approach is presented. Crack resistance curves (J-R curves) were created for unirradiated thin walled Zircaloy-4 and aluminium cladding tube pieces at room temperature using the single sample method. The procedure of creating sharp fatigue starter cracks with respect to optical recording was optimized. It is shown that the chosen test method is appropriate for the determination of complete J-R curves including the values J {sub 0.2} (J at 0.2 mm crack length), J {sub m} (J corresponding to the maximum load) and the slope of the curve.

Introduction program of M5TM cladding in Japan

International Nuclear Information System (INIS)

Mardon, Jean Paul; Kaneko, Nori

2008-01-01

Experience from irradiation in PWR has confirmed that M5 TM possesses all the properties required for upgraded operation including new fuel management approaches and high duty reactor operation. Specifically, the alloy M5 TM has demonstrated impressive improvements over Zircaloy-4 for fuel rod cladding and fuel assembly structural components. Moreover, several irradiation campaigns have been worldwide performed in order to confirm the excellent M5 TM in-pile behavior in very demanding PWR irradiation conditions (high void fraction, heat flux, temperature, lithium content and Zinc injection). Regarding licensing, the authorization for loading M5 TM alloy has been granted by US, UK, South Korean, German, Chinese, South-African, Swedish and Belgian Safety Authorities. Also the French Nuclear Safety Authority has given individually its authorization to load all-M5 TM fuel assembly batches in 1300MWe plants and a generic license to load all-M5 TM fuel in EDF N4 reactors and M5 TM fuel clad in 900MWe reactors for MOX parity fuel management. Licensing is also now underway in Switzerland, Finland, Brazil and Spain. The M5 TM alloy has demonstrated its superiority at burn-ups beyond current licensing limits, through operations in PWR at fuel rod burn-ups exceeding 71GWd/tU in the United States and 78GWd/tU in Europe. The Japanese nuclear industry has planned a stepwise approach to increase the burn-up of the fuel. Step-I fuel (48GWd/tU Fuel Assembly maximum burn-up) which was introduced in the late 80s. In the 90s started the licensing of the Step-II fuel (55GWd/tU Fuel Assembly maximum burn-up). Because the extension of the burn-up is important to reduce discharge fuel and cycle cost, the Japanese industry has plans to further extend the burn-up. In such burn-up region, fuel cladding with even better corrosion properties and very low hydrogen pick-up shall be necessary. M5 TM alloy, with high anticorrosion/hydriding properties, is suitable for not only the Step-II fuel

Chemical Dissolution of Simulant FCA Cladding and Plates

Energy Technology Data Exchange (ETDEWEB)

Daniel, G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Pierce, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-11-08

The Savannah River Site (SRS) has received some fast critical assembly (FCA) fuel from the Japan Atomic Energy Agency (JAEA) for disposition. Among the JAEA FCA fuel are approximately 7090 rectangular Stainless Steel clad fuel elements. Each element has an internal Pu-10.6Al alloy metal wafer. The thickness of each element is either 1/16 inch or 1/32 inch. The dimensions of each element ranges from 2 inches x 1 inch to 2 inches x 4 inches. This report discusses the potential chemical dissolution of the FCA clad material or stainless steel. This technology uses nitric acid-potassium fluoride (HNO₃-KF) flowsheets of H-Canyon to dissolve the FCA elements from a rack of materials. Historically, dissolution flowsheets have aimed to maximize Pu dissolution rates while minimizing stainless steel dissolution (corrosion) rates. Because the FCA cladding is made of stainless steel, this work sought to accelerate stainless steel dissolution.

Improved Accident Tolerance of Austenitic Stainless Steel Cladding through Colossal Supersaturation with Interstitial Solutes

International Nuclear Information System (INIS)

Ernst, Frank

2016-01-01

We proposed a program-supporting research project in the area of fuel-cycle R&D, specifically on the topic of advanced fuels. Our goal was to investigate whether SECIS (surface engineering by concentrated interstitial solute - carbon, nitrogen) can improve the properties of austenitic stainless steels and related structural alloys such that they can be used for nuclear fuel cladding in LWRs (light-water reactors) and significantly excel currently used alloys with regard to performance, safety, service life, and accident tolerance. We intended to demonstrate that SECIS can be adapted for post-processing of clad tubing to significantly enhance mechanical properties (hardness, wear resistance, and fatigue life), corrosion resistance, resistance to stress-corrosion cracking (hydrogen-induced embrittlement), and - potentially - radiation resistance (against electron-, neutron-, or ion-radiation damage). To test this hypothesis, we measured various relevant properties of the surface-engineered alloys and compared them with corresponding properties of the non-treated, as-received alloys. In particular, we studied the impact of heat exposure corresponding to BWR (boiling-water reactor) working and accident (loss-of-coolant) conditions and the effect of ion irradiation.

Task Group E: fuel-cladding interface reactions. Second quarterly report

International Nuclear Information System (INIS)

Kangilaski, M.; Adamson, M.G.

1974-01-01

An interim assessment of possible interactions and their consequences in the various fuel systems was completed. The assessment discusses the interactions of advanced cladding alloys with: (1) helium bonded mixed oxides; (2) helium and sodium bonded mixed carbides; and (3) helium and sodium bonded mixed nitrides

An integrated approach to selecting materials for fuel cladding in advanced high-temperature reactors

Energy Technology Data Exchange (ETDEWEB)

Rangacharyulu, C., E-mail: chary.r@usask.ca [Univ. of Saskatchewan, Saskatoon, SK (Canada); Guzonas, D.A.; Pencer, J.; Nava-Dominguez, A.; Leung, L.K.H. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2014-07-01

An integrated approach has been developed for selection of fuel cladding materials for advanced high-temperature reactors. Reactor physics, thermalhydraulic and material analyses are being integrated in a systematic study comparing various candidate fuel-cladding alloys. The analyses established the axial and radial neutron fluxes, power distributions, axial and radial temperature distributions, rates of defect formation and helium production using AECL analytical toolsets and experimentally measured corrosion rates to optimize the material composition for fuel cladding. The project has just been initiated at University of Saskatchewan. Some preliminary results of the analyses are presented together with the path forward for the project. (author)

High Strain Rate Testing of Welded DOP-26 Iridium

Energy Technology Data Exchange (ETDEWEB)

Schneibel, J. H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Miller, R. G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carmichael, C. A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Fox, E. E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ulrich, G. B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, E. P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-08-01

The iridium alloy DOP-26 is used to produce Clad Vent Set cups that protect the radioactive fuel in radioisotope thermoelectric generators (RTGs) which provide electric power for spacecraft and rovers. In a previous study, the tensile properties of DOP-26 were measured over a wide range of strain rates and temperatures and reported in ORNL/TM-2007/81. While that study established the properties of the base material, the fabrication of the heat sources requires welding, and the mechanical properties of welded DOP-26 have not been extensively characterized in the past. Therefore, this study was undertaken to determine the mechanical properties of DOP-26 specimens containing a transverse weld in the center of their gage sections. Tensile tests were performed at room temperature, 750, 900, and 1090°C and engineering strain rates of 1×10^-3 and 10 s^-1. Room temperature testing was performed in air, while testing at elevated temperatures was performed in a vacuum better than 1×10^-4 Torr. The welded specimens had a significantly higher yield stress, by up to a factor of ~2, than the non-welded base material. The yield stress did not depend on the strain rate except at 1090°C, where it was slightly higher for the faster strain rate. The ultimate tensile stress, on the other hand, was significantly higher for the faster strain rate at temperatures of 750°C and above. At 750°C and above, the specimens deformed at 1×10^-3 s^-1 showed pronounced necking resulting sometimes in perfect chisel-edge fracture. The specimens deformed at 10 s^-1 exhibited this fracture behavior only at the highest test temperature, 1090°C. Fracture occurred usually in the fusion zone of the weld and was, in most cases, primarily intergranular.

Evaluation of integrally finned cladding for LMFBR fuel pins

International Nuclear Information System (INIS)

Cantley, D.A.; Sutherland, W.H.

1975-01-01

An integral fin design effectively reduces the coolant temperature gradients within an LMFBR subassembly by redistributing coolant flow so as to reduce the maximum cladding temperature and increase the duct wall temperature. The reduced cladding temperatures are offset by strain concentrations resulting from the fin geometry, so there is little net effect on predicted fuel pin performance. The increased duct wall temperatures, however, significantly reduce the duct design lifetime so that the final conclusion is that the integral fin design is inferior to the standard wire wrap design. This result, however, is dependent upon the material correlations used. Advanced alloys with improved irradiation properties could alter this conclusion

Laser cladding of Al-Si/SiC composite coatings : Microstructure and abrasive wear behavior

NARCIS (Netherlands)

Anandkumar, R.; Almeida, A.; Vilar, R.; Ocelik, V.; De Hosson, J.Th.M.

2007-01-01

Surface coatings of an Al-Si-SiC composite were produced on UNS A03560 cast Al-alloy substrates by laser cladding using a mixture of powders of Al-12 wt.% Si alloy and SiC. The microstructure of the coatings depends considerably on the processing parameters. For a specific energy of 26 MJ/m2 the

Critical cladding radius for hybrid cladding modes

Science.gov (United States)

Guyard, Romain; Leduc, Dominique; Lupi, Cyril; Lecieux, Yann

2018-05-01

In this article we explore some properties of the cladding modes guided by a step-index optical fiber. We show that the hybrid modes can be grouped by pairs and that it exists a critical cladding radius for which the modes of a pair share the same electromagnetic structure. We propose a robust method to determine the critical cladding radius and use it to perform a statistical study on the influence of the characteristics of the fiber on the critical cladding radius. Finally we show the importance of the critical cladding radius with respect to the coupling coefficient between the core mode and the cladding modes inside a long period grating.

Allowable peak heat-up cladding temperature for spent fuel integrity during interim-dry storage

Directory of Open Access Journals (Sweden)

Ki-Nam Jang

2017-12-01

Full Text Available To investigate allowable peak cladding temperature and hoop stress for maintenance of cladding integrity during interim-dry storage and subsequent transport, zirconium alloy cladding tubes were hydrogen-charged to generate 250 ppm and 500 ppm hydrogen contents, simulating spent nuclear fuel degradation. The hydrogen-charged specimens were heated to four peak temperatures of 250°C, 300°C, 350°C, and 400°C, and then cooled to room temperature at cooling rates of 0.3 °C/min under three tensile hoop stresses of 80 MPa, 100 MPa, and 120 MPa. The cool-down specimens showed that high peak heat-up temperature led to lower hydrogen content and that larger tensile hoop stress generated larger radial hydride fraction and consequently lower plastic elongation. Based on these out-of-pile cladding tube test results only, it may be said that peak cladding temperature should be limited to a level < 250°C, regardless of the cladding hoop stress, to ensure cladding integrity during interim-dry storage and subsequent transport.

Study of the uniform corrosion of an aluminium alloy used for the fuel cladding of the Jules Horowitz experimental reactor; Etude de la corrosion uniforme d'un alliage d'aluminium utilise comme gainage du combustible nucleaire du reacteur experimental Jules Horowitz

Energy Technology Data Exchange (ETDEWEB)

Wintergerst, M. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SEMI), 91 - Gif-sur-Yvette (France)

2008-07-01

For the Jules Horowitz new material testing reactor, an aluminium base alloy, AlFeNi, will be used for the cladding of the fuel plates. Taking into account the thermal properties of the alloy and of its oxide, the corrosion of the fuel cans presents many problems. The aim of this thesis is to provide a growing kinetic of the oxide layer at the surface of the AlFeNi fuel can in order to predict the life time of fuel element. Thus the mechanism of degradation of the cladding will be describe in order to integrate the different parameters of the operating reactor. (A.L.B.)

Evaluation of the applicability of cladding deformation model in RELAP5/MOD3.2 code for VVER-1000 fuel

International Nuclear Information System (INIS)

Vorob'ev, Yu.; Zhabin, O.

2015-01-01

Applicability of cladding deformation model in RELAP5/MOD3.2 code is analyzed for VVER-1000 fuel cladding from Zr+1%Nb alloy. Experimental data and calculation model of fuel assembly channel of the core are used for this purpose. The model applicability is tested for the cladding temperature range from 600 to 1200 deg C and pressure range from 1 to 12 MPa. Evaluation results demonstrate limited applicability of built-in RELAP5/MOD3.2 cladding deformation model to the estimation of Zr+1%Nb cladding rupture conditions. The limitations found shall be considered in application of RELAP5/MOD3.2 cladding deformation model in the design-basis accident analysis of VVER reactors

Flat-Cladding Fiber Bragg Grating Sensors for Large Strain Amplitude Fatigue Tests

Directory of Open Access Journals (Sweden)

Xijia Gu

2010-08-01

Full Text Available We have successfully developed a flat-cladding fiber Bragg grating sensor for large cyclic strain amplitude tests of up to ±8,000 με. The increased contact area between the flat-cladding fiber and substrate, together with the application of a new bonding process, has significantly increased the bonding strength. In the push-pull fatigue tests of an aluminum alloy, the plastic strain amplitudes measured by three optical fiber sensors differ only by 0.43% at a cyclic strain amplitude of ±7,000 με and 1.9% at a cyclic strain amplitude of ±8,000 με. We also applied the sensor on an extruded magnesium alloy for evaluating the peculiar asymmetric hysteresis loops. The results obtained were in good agreement with those measured from the extensometer, a further validation of the sensor.

Microstructure and Wear Resistance of Laser-Clad (Co, Ni61.2B26.2Si7.8Ta4.8 Coatings

Directory of Open Access Journals (Sweden)

Luan Zhang

2017-10-01

Full Text Available It has been reported that a quaternary Co61.2B26.2Si7.8Ta4.8 alloy is a good glass former and can be laser-clad to an amorphous composite coating with superior hardness and wear resistance. In this paper, alloys with varying Ni contents to substitute for Co are coated on the surface of #45 carbon steel using a 5-kW CO2 laser source for the purpose of obtaining protective coatings. In contrast to the quaternary case, the clad layers are characterized by a matrix of α-(Fe, Co, Ni solid solution plus CoB, Co3B, and Co3Ta types of precipitates. The cladding layer is divided into four regions: Near-surface dendrites, α-(Fe, Co, Ni solid solution plus dispersed particles in the middle zone, columnar bonding zone, and heat-affected area that consists of martensite. The hardness gradually decreases with increasing Ni content, and the maximum hardness occurs in the middle zone. Both the friction coefficient and wear volume are minimized in the alloy containing 12.2% Ni. Compared with the previous cobalt-based quaternary alloy Co61.2B26.2Si7.8Ta4.8, the addition of the Ni element reduces the glass-forming ability and henceforth the hardness and wear resistance of the clad layers.

Quantification of the distribution of hydrogen by nuclear microprobe at the Laboratory Pierre Sue in the width of zirconium alloy fuel clad of PWR reactors

International Nuclear Information System (INIS)

Raepsaet, C.; Bossis, Ph.; Hamon, D.; Bechade, J.L.; Brachet, J.C.

2007-01-01

Among the analysis techniques by ions beams, the micro ERDA (Elastic Detection Analysis) is an interesting technique which allows the quantitative distribution of the hydrogen in materials. In particular, this analysis has been used for hydride zirconium alloys, with the nuclear microprobe of the Laboratory Pierre Sue. This probe allows the characterization of radioactive materials. The technique principles are recalled and then two examples are provided to illustrate the fuel clad behavior in PWR reactors. (A.L.B.)

Corrosion of aluminum alloys in a reactor disassembly basin

International Nuclear Information System (INIS)

Howell, J.P.; Zapp, P.E.; Nelson, D.Z.

1992-01-01

This document discusses storage of aluminum clad fuel and target tubes of the Mark 22 assembly takes place in the concrete-lined, light-water-filled, disassembly basins located within each reactor area at the Savannah River Site (SRS). A corrosion test program has been conducted in the K-Reactor disassembly basin to assess the storage performance of the assemblies and other aluminum clad components in the current basin environment. Aluminum clad alloys cut from the ends of actual fuel and target tubes were originally placed in the disassembly water basin in December 1991. After time intervals varying from 45--182 days, the components were removed from the basin, photographed, and evaluated metallographically for corrosion performance. Results indicated that pitting of the 8001 aluminum fuel clad alloy exceeded the 30-mil (0.076 cm) cladding thickness within the 45-day exposure period. Pitting of the 1100 aluminum target clad alloy exceeded the 30-mil (0.076 cm) clad thickness in 107--182 days exposure. The existing basin water chemistry is within limits established during early site operations. Impurities such as Cl - , NO 3 - and SO 4 - are controlled to the parts per million level and basin water conductivity is currently 170--190 μmho/cm. The test program has demonstrated that the basin water is aggressive to the aluminum components at these levels. Other storage basins at SRS and around the US have successfully stored aluminum components for greater than ten years without pitting corrosion. These basins have impurity levels controlled to the parts per billion level (1000X lower) and conductivity less than 1.0 μmho/cm

Laser cladding: repairing and manufacturing metal parts and tools

Science.gov (United States)

Sexton, Leo

2003-03-01

Laser cladding is presently used to repair high volume aerospace, automotive, marine, rail or general engineering components where excessive wear has occurred. It can also be used if a one-off high value component is either required or has been accidentally over-machined. The ultimate application of laser cladding is to build components up from nothing, using a laser cladding system and a 3D CAD drawing of the component. It is thus emerging that laser cladding can be classified as a special case of Rapid Prototyping (RP). Up to this point in time RP was seen, and is still seen, as in intermediately step between the design stage of a component and a finished working product. This can now be extended so that laser cladding makes RP a one-stop shop and the finished component is made from tool-steel or some alloy-base material. The marriage of laser cladding with RP is an interesting one and offers an alternative to traditional tool builders, re-manufacturers and injection mould design/repair industries. The aim of this paper is to discuss the emergence of this new technology, along with the transference of the process out of the laboratory and into the industrial workplace and show it is finding its rightful place in the manufacturing/repair sector. It will be shown that it can be used as a cost cutting, strategic material saver and consequently a green technology.

Effective Area and Charge Density of Iridium Oxide Neural Electrodes

International Nuclear Information System (INIS)

Harris, Alexander R.; Paolini, Antonio G.; Wallace, Gordon G.

2017-01-01

The effective electrode area and charge density of iridium metal and anodically activated iridium has been measured by optical and electrochemical techniques. The degree of electrode activation could be assessed by changes in electrode colour. The reduction charge, activation charge, number of activation pulses and charge density were all strongly correlated. Activated iridium showed slow electron transfer kinetics for reduction of a dissolved redox species. At fast voltammetric scan rates the linear diffusion electroactive area was unaffected by iridium activation. At slow voltammetric scan rates, the steady state diffusion electroactive area was reduced by iridium activation. The steady state current was consistent with a ring electrode geometry, with lateral resistance reducing the electrode area. Slow electron transfer on activated iridium would require a larger overpotential to reduce or oxidise dissolved species in tissue, limiting the electrodes charge capacity but also reducing the likelihood of generating toxic species in vivo.

The role of cladding material for performance of LWR control assemblies

International Nuclear Information System (INIS)

Dewes, P.; Roppelt, A.

2000-01-01

The lifetime of control assemblies in LWRs can be limited presently by mechanical failure of the absorber cladding. The major cause of failure is mechanical interaction of the absorber with the cladding due to irradiation induced dimensional changes such as absorber swelling and cladding creep, resulting in cracking of the clad. Such failures occurred in both BWRs and PWRs. Experience and in-reactor tests revealed that cracking can be avoided principally by two ways: First, if strain rates and hence, stresses in the cladding are kept low (well below the yield strength), significant strains can be tolerated. This is the case for the cladding of PWR control assemblies with slowly swelling Ag-In-Cd absorber. Recent examinations of highly exposed PWR control assemblies confirmed the design correlation up to the presently used strain limit. Second, in such cases where strongly swelling absorber material like boron carbide is still preferred, materials which are resistant against irradiation assisted stress corrosion cracking (IASCC) can be used. The influence of material composition and condition on IASCC was studied in-reactor using tubular samples of various stainless steels and Ni-base alloys stressed by swelling mandrels. In several programme steps high purity materials with special features had been identified as resistant to IASCC. Another process of cladding damage which may occur in PWRs is wear caused by friction of the control rods in the surrounding guide structure. For replacement control assemblies this problem is solved by coating of the cladding. There exists meanwhile excellent experience of up to 18 operation cycles with coated claddings. (author)

Microstructure and wear resistance of laser cladded composite coatings prepared from pre-alloyed WC-NiCrMo powder with different laser spots

Science.gov (United States)

Yao, Jianhua; Zhang, Jie; Wu, Guolong; Wang, Liang; Zhang, Qunli; Liu, Rong

2018-05-01

The distribution of WC particles in laser cladded composite coatings can significantly affect the wear resistance of the coatings under aggressive environments. In this study, pre-alloyed WC-NiCrMo powder is deposited on SS316L via laser cladding with circular spot and wide-band spot, respectively. The microstructure and WC distribution of the coatings are investigated with optical microscope (OM), scanning electron microscopy (SEM), energy dispersive spectrometer (EDS), and X-ray diffraction (XRD). The wear behavior of the coatings is investigated under dry sliding-wear test. The experimental results show that the partially dissolved WC particles are uniformly distributed in both coatings produced with circular spot and wide-band spot, respectively, and the microstructures consist of WC and M23C6 carbides and γ-(Ni, Fe) solid solution matrix. However, due to Fe dilution, the two coatings have different microstructural characteristics, resulting in different hardness and wear resistance. The wide-band spot laser prepared coating shows better performance than the circular spot laser prepared coating.

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

International Nuclear Information System (INIS)

Guo, Xunzhong; Tao, Jie; Wang, Wentao; Li, Huaguan; Wang, Chen

2013-01-01

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

Improved Accident Tolerance of Austenitic Stainless Steel Cladding through Colossal Supersaturation with Interstitial Solutes

Energy Technology Data Exchange (ETDEWEB)

Ernst, Frank [Case Western Reserve Univ., Cleveland, OH (United States)

2016-10-13

We proposed a program-supporting research project in the area of fuel-cycle R&D, specifically on the topic of advanced fuels. Our goal was to investigate whether SECIS (surface engineering by concentrated interstitial solute – carbon, nitrogen) can improve the properties of austenitic stainless steels and related structural alloys such that they can be used for nuclear fuel cladding in LWRs (light-water reactors) and significantly excel currently used alloys with regard to performance, safety, service life, and accident tolerance. We intended to demonstrate that SECIS can be adapted for post-processing of clad tubing to significantly enhance mechanical properties (hardness, wear resistance, and fatigue life), corrosion resistance, resistance to stress–corrosion cracking (hydrogen-induced embrittlement), and – potentially – radiation resistance (against electron-, neutron-, or ion-radiation damage). To test this hypothesis, we measured various relevant properties of the surface-engineered alloys and compared them with corresponding properties of the non–treated, as-received alloys. In particular, we studied the impact of heat exposure corresponding to BWR (boiling-water reactor) working and accident (loss-of-coolant) conditions and the effect of ion irradiation.

Derivative effect of laser cladding on interface stability of YSZ@Ni coating on GH4169 alloy: An experimental and theoretical study

Science.gov (United States)

Zheng, Haizhong; Li, Bingtian; Tan, Yong; Li, Guifa; Shu, Xiaoyong; Peng, Ping

2018-01-01

Yttria-stabilized zirconia YSZ@Ni core-shell nanoparticles were used to prepare a thermal barrier coating (TBC) on a GH4169 alloy by laser cladding. Microstructural analysis showed that the TBC was composed of two parts: a ceramic and a bonding layer. In places where the ZrO2/Al2O3 eutectic structure was present in the ceramic layer, the Ni atoms diffused into the bonding layer, as confirmed by energy-dispersive X-ray spectroscopy (EDS). The derivative effect of laser cladding results in the original YSZ@Ni core-shell nanoparticles being translated into the Al2O3 crystal, activating the YSZ. The mechanism of ceramic/metal interface cohesion was studied in depth via first-principles and molecular dynamics simulation. The results show that the trend in the diffusion coefficients of Ni, Fe, Al, and Ti is DNi > DFe > DTi > DAl in the melting or solidification process of the material. The enthalpy of formation for Al2O3 is less than that of TiO2, resulting in a thermally grown oxide (TGO) Al2O3 phase transformation. With regard to the electronic structure, the trend in Mulliken population is QO-Ni > QZr-O > QO-Al. Although the bonding is slightly weakened between ZrO2/Al2O3 (QZr-O = 0.158 matrix. Thus, by comparing the connective and diffusive processes, our findings lay the groundwork for detailed and comprehensive studies of the laser cladding process for the production of composite materials.

Nickel alloys and high-alloyed special stainless steels. Properties, manufacturing, applications. 4. compl. rev. ed.

International Nuclear Information System (INIS)

Heubner, Ulrich; Kloewer, Jutta; Alves, Helena; Behrens, Rainer; Schindler, Claudius; Wahl, Volker; Wolf, Martin

2012-01-01

This book contains the following eight topics: 1. Nickel alloys and high-alloy special stainless steels - Material overview and metallurgical principles (U. Heubner); 2. Corrosion resistance of nickel alloys and high-alloy special stainless steels (U. Heubner); 3. Welding of nickel alloys and high-alloy special stainless steels (T. Hoffmann, M. Wolf); 4. High-temperature materials for industrial plant construction (J. Kloewer); 5. Nickel alloys and high-alloy special stainless steels as hot roll clad composites-a cost-effective alternative (C. Schindler); 6. Selected examples of the use of nickel alloys and high-alloy special stainless steels in chemical plants (H. Alves); 7. The use of nickel alloys and stainless steels in environmental engineering (V. Wahl); 8: Nickel alloys and high-alloy special stainless steels for the oil and gas industry (R. Behrens).

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 in the pr......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...... to analyze, compare, and discuss how these various construction solutions point out strategies for development based on fundamentally different mindsets. The research questions address the following issues: How to learn from traditional construction principles: When do we see limitations of tectonic maneuver......, to ask for more restrictive building codes. As an example, in Denmark there are series of increasing demands in the current building legislations that are focused at enhancing the energy performance of buildings, which consequently foster rigid insulation standards and ask for improvement of air...

Solid-liquid phase equilibria of Fe-Cr-Al alloys and spinels

Science.gov (United States)

McMurray, J. W.; Hu, R.; Ushakov, S. V.; Shin, D.; Pint, B. A.; Terrani, K. A.; Navrotsky, A.

2017-08-01

Ferritic FeCrAl alloys are candidate accident tolerant cladding materials. There is a paucity of data concerning the melting behavior for FeCrAl and its oxides. Analysis tools have therefore had to utilize assumptions for simulations using FeCrAl cladding. The focus of this study is to examine in some detail the solid-liquid phase equilibria of FeCrAl alloys and spinels with the aim of improving the accuracy of severe accident scenario computational studies.

Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

Energy Technology Data Exchange (ETDEWEB)

Pasqualini, E.E. [Laboratorio de Nanotecnología Nuclear, Centro Atómico Constituyentes, Comisión Nacional de Energía Atómica, Av. General Paz 1499, B1650KNA, San Martín, Prov. Buenos Aires (Argentina); Robinson, A.B. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Porter, D.L., E-mail: Douglas.Porter@inl.gov [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Wachs, D.M. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Finlay, M.R. [Australian Nuclear Science and Technology Organisation, PMB 1, Menai, NSW, 2234 (Australia)

2016-10-15

Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U–(7–10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry–4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction—either from fabrication or in-reactor testing—and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm{sup 3}, 3.8E+21 (peak).

Investigation of Hot Rolling Influence on the Explosive-Welded Clad Plate

Directory of Open Access Journals (Sweden)

Guanghui ZHAO

2016-11-01

Full Text Available The microstructure, the shear strength and tensile strength of stainless steel explosive-welded clad plate at different rolling reduction were studied. The mechanical properties of the explosive-welded and explosive-rolled clad plates were experimentally measured. Simultaneously, the microstructures of the clad plate were investigated by the Ultra deep microscope and the tensile fracture surface were observed by the scan electron microscope (SEM. It was observed that the tensile strength has been increased considerably, whereas the elongation percentage has been reduced with the increase of hot rolling reduction. In the tensile shear test, the bond strength is higher than the strength of the ferritic stainless steel layer and meets the relevant known standard criterion. Microstructural evaluations showed that the grain of the stainless steel and steel refined with the increase of thickness reduction. Examination of the tensile fracture surfaces reveal that, after hot rolling, the fracture in the low alloy steel and ferritic stainless steel clad plates is of the ductile type.DOI: http://dx.doi.org/10.5755/j01.ms.22.4.12409

Development of weldable, corrosion-resistant iron-aluminide alloys

Energy Technology Data Exchange (ETDEWEB)

Maziasz, P.J.; Goodwin, G.M.; Wang, X.L. [Oak Ridge National Laboratory, TN (United States)

1995-05-01

Corrosion-resistant, weldable FeAl alloys have been developed with improved high-temperature strength industrial applications. Previous processing difficulties with these alloys led to their evaluation as weld-overlay claddings on conventional structural steels to take advantage of their good properties now. Simplified and better processing methods for monolithic FeAl components are also currently being developed so that components for industrial testing can be made. Other avenues for producing FeAl coatings are currently being explored. Neutron scattering experiments residual stress distributions in the FeAl weld-overlay cladding began in FY 1993 and continued this year.

Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels

International Nuclear Information System (INIS)

Lu, Hongbing; Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott

2014-01-01

Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear

Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels

Energy Technology Data Exchange (ETDEWEB)

Lu, Hongbing [Univ. of Texas, Austin, TX (United States); Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott

2014-01-09

Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear

Development of New Heats of Advanced Ferritic/Martensitic Alloys

Energy Technology Data Exchange (ETDEWEB)

Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Pestovich, Kimberly Shay [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Anderoglu, Osman [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Aydogan, Eda [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-06-23

The Fuel Cycle Research and Development program is investigating methods of transmuting minor actinides in various fuel cycle options. To achieve this goal, new fuels and cladding materials must be developed and tested to high burnup levels (e.g. >20%) requiring cladding to withstand very high doses (greater than 200 dpa) while in contact with the coolant and the fuel. To develop and qualify materials to a total fluence greater than 200 dpa requires development of advanced alloys and irradiations in fast reactors to test these alloys. Recent results from testing numerous ferritic/martensitic steels at low temperatures suggest that improvements in low temperature radiation tolerance can be achieved through carefully controlling the nitrogen content in these alloys. Thus, four new heats of HT-9 were produced with controlled nitrogen content: two by Metalwerks and two by Sophisticated Alloys. Initial results on these new alloys are presented including microstructural analysis and hardness testing. Future testing will include irradiation testing with ions and in reactor.

Automatic welding and cladding in heavy fabrication

International Nuclear Information System (INIS)

Altamer, A. de

1980-01-01

A description is given of the automatic welding processes used by an Italian fabricator of pressure vessels for petrochemical and nuclear plant. The automatic submerged arc welding, submerged arc strip cladding, pulsed TIG, hot wire TIG and MIG welding processes have proved satisfactory in terms of process reliability, metal deposition rate, and cost effectiveness for low alloy and carbon steels. An example shows sequences required during automatic butt welding, including heat treatments. Factors which govern satisfactory automatic welding include automatic anti-drift rotator device, electrode guidance and bead programming system, the capability of single and dual head operation, flux recovery and slag removal systems, operator environment and controls, maintaining continuity of welding and automatic reverse side grinding. Automatic welding is used for: joining vessel sections; joining tubes to tubeplate; cladding of vessel rings and tubes, dished ends and extruded nozzles; nozzle to shell and butt welds, including narrow gap welding. (author)

Surface protection of light metals by one-step laser cladding with oxide ceramics

Science.gov (United States)

Nowotny, S.; Richter, A.; Tangermann, K.

1999-06-01

Today, intricate problems of surface treatment can be solved through precision cladding using advanced laser technology. Metallic and carbide coatings have been produced with high-power lasers for years, and current investigations show that laser cladding is also a promising technique for the production of dense and precisely localized ceramic layers. In the present work, powders based on Al2O3 and ZrO2 were used to clad aluminum and titanium light alloys. The compact layers are up to 1 mm thick and show a nonporous cast structure as well as a homogeneous network of vertical cracks. The high adhesive strength is due to several chemical and mechanical bonding mechanisms and can exceed that of plasmasprayed coatings. Compared to thermal spray techniques, the material deposition is strictly focused onto small functional areas of the workpiece. Thus, being a precision technique, laser cladding is not recommended for large-area coatings. Examples of applications are turbine components and filigree parts of pump casings.

Development and fabrication of seamless Aluminium finned clad tubes for metallic uranium fuel rods for research reactor

International Nuclear Information System (INIS)

Singh, A.K.; Hussain, M.M.; Jayachandran, N.K.; Abdulla, K.K.

2012-01-01

Natural uranium metal or its alloy is used as fuel in nuclear reactors. Usually fuel is clad with compatible material to prevent its direct contact with coolant which prevents spread of activity. One of the methods of producing fuel for nuclear reactor is by co-drawing finished uranium rods with aluminum clad tube to develop intimate contact for effective heat removal during reactor operation. Presently seam welded Aluminium tubes are used as clad for Research Reactor fuel. The paper will highlight entire fabrication process followed for the fabrication of seamless Aluminium finned tubes along with relevant characterisation results

Microstructure and Wear Resistance of Laser Cladding TiC Coat on Titanium Alloy%钛合金表面激光熔覆TiC涂层显微结构和耐磨性

Institute of Scientific and Technical Information of China (English)

王慧萍; 李军; 李芳; 李曼萍; 奚文龙

2012-01-01

采用HL-5000型横流CO2激光加工机在TC4钛合金表面激光熔覆TiC+ Ti和TiC+Ti+ F102复合涂层.通过SEM、EDAX、XRD、HXD-1000TMC型显微硬度计,HT-600型高温摩擦磨损试验机,分析了熔覆层的显微组织、成分、物相,测试了激光熔覆层的显微硬度和滑动摩擦磨损性能.结果表明,激光熔覆制备的TiC复合涂层与基体呈冶金结合,在TiC+ Ti激光熔覆层中,熔覆层的组织是在Ti基体上分布着TiC树枝晶;在TiC +Ti+ F102激光熔覆层中,TiC颗粒发生了部分溶解,熔覆层的组织是在Ti基和γ-Ni基的基体上分布着细小的TiC颗粒和TiC树枝晶.TiC+ Ti激光熔覆层的硬度约为700 HV0.1,TiC+Ti+ F102激光熔覆层的硬度约为800 HV0.1,两种复合涂层耐磨性均比TC4钛合金显著提高.%The laser cladding TiC + Ti and TiC + Ti + F102 composite coaling on the surfact of TC4 alloy was obtained with 5.0 Kw continuous wave CO2 laser. The microstructure,composition and phase of the coating were investigated by means of SEM,EDAX,XRD,HXD-1000TMC Microhardness Tester, HT-600 wear machine Moreover, the microhardness and friction wear properties was measured. The results indicate that the laser cladding TiC composite coating is well bonded with the matrix alloy. The microstructures of TiC dendrites in Ti matrix in the clad layer of TiC + Ti laser clad coating. For TiC + Ti + F102 laser clad coating, parts of TiC particles are dissolved to form a microstructures of TiC particles and fine TiC dendrites in the matrix of Ti and y-Ni in the clad layer. The microhardness of TiC + Ti coating is 700 HV0.1. The microhardness of TiC + Ti + F102 coating is 800 HV0.1 , and the coating greatly enhances the wear resistantce of TC4 titanium alloy.

An internal conical mandrel technique for fracture toughness measurements on nuclear fuel cladding

Energy Technology Data Exchange (ETDEWEB)

Sainte Catherine, C.; Le Boulch, D.; Carassou, S. [CEA Saclay, DEN/DMN, Bldg 625 P, Gif-Sur-Yvette, F-91191 (France); Lemaignan, C. [CEA Grenoble, 17 rue des Martyrs, Grenoble, F-38054 (France); Ramasubramanian, N. [ECCATEC Inc., 92 Deburn Drive, Toronto, Ontario (Canada)

2006-07-01

An understanding of the limiting stress level for crack initiation and propagation in a fuel cladding material is a fundamental requirement for the development of water reactor clad materials. Conventional tests, in use to evaluate fracture properties, are of limited help, because they are adapted from ASTM standards designed for thick materials, which differ significantly from fuel cladding geometry (small diameter thin-walled tubing). The Internal Conical Mandrel (ICM) test described here is designed to simulate the effect of fuel pellet diametrical increase on a cladding with an existing axial through-wall crack. It consists in forcing a cone, having a tapered increase in diameter, inside the Zircaloy cladding with an initial axial crack. The aim of this work is to quantify the crack initiation and propagation criteria for fuel cladding material. The crack propagation is monitored by a video system for obtaining crack extension {delta}a. A finite-element (FE) simulation of the ICM test is performed in order to derive J integrals. A node release technique is applied during the FE simulation for crack propagation and the J-resistance curves (J-{delta}a) are generated. This paper presents the test methodology, the J computation validation, and results for cold-worked stress relieved Zircaloy-4 cladding at 20 deg. and 300 deg. C and also for Al 7050-T7651 aluminum alloy tubing at 20 deg. C. (authors)

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

International Nuclear Information System (INIS)

Johnson, A.B. Jr.; Gilbert, E.R.

1983-09-01

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 570 0 C. 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 270 0 C). 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 380 0 C 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 400 0 C 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

Fabrication of Tungsten-Rhenium Cladding materials via Spark Plasma Sintering for Ultra High Temperature Reactor Applications

Energy Technology Data Exchange (ETDEWEB)

Charit, Indrajit; Butt, Darryl; Frary, Megan; Carroll, Mark

2012-11-05

This research will develop an optimized, cost-effective method for producing high-purity tungsten-rhenium alloyed fuel clad forms that are crucial for the development of a very high-temperature nuclear reactor. The study will provide critical insight into the fundamental behavior (processing-microstructure- property correlations) of W-Re alloys made using this new fabrication process comprising high-energy ball milling (HEBM) and spark plasma sintering (SPS). A broader goal is to re-establish the U.S. lead in the research field of refractory alloys, such as W-Re systems, with potential applications in very high-temperature nuclear reactors. An essential long-term goal for nuclear power is to develop the capability of operating nuclear reactors at temperatures in excess of 1,000K. This capability has applications in space exploration and some special terrestrial uses where high temperatures are needed in certain chemical or reforming processes. Refractory alloys have been identified as being capable of withstanding temperatures in excess of 1,000K and are considered critical for the development of ultra hightemperature reactors. Tungsten alloys are known to possess extraordinary properties, such as excellent high-temperature capability, including the ability to resist leakage of fissile materials when used as a fuel clad. However, there are difficulties with the development of refractory alloys: 1) lack of basic experimental data on thermodynamics and mechanical and physical properties, and 2) challenges associated with processing these alloys.

Summary on out-of-pile and in-pile properties of M5 alloy

International Nuclear Information System (INIS)

Zhao Wenjin

2001-01-01

The out-of-pile and in-pile corrosion, mechanical properties, microstructure,hydrogen absorption, creep and growth resistances of M5 alloy using as PWR fuel rod cladding materials developed by FRAMATOME in France has been summarized with reference to the literatures. The results obtained from in-pile irradiation tests show that the corrosion and hydrogen absorption resistances, creep and irradiation growth resistances of M5 alloy cladding are superior to that of the optimized Zircaloy-4. It could be estimated that the M5 alloy enables rod burnups close to 65GWd/tU to be reached

Annual Technical Progress Report of Radioisotope Power System Materials Production and Technical Program Tasks for October 1, 2005 through September 30, 2006

Energy Technology Data Exchange (ETDEWEB)

None

2007-04-02

The Office of Space and Defense Power Systems of the Department of Energy (DOE) provides Radioisotope Power Systems (RPS) for applications where conventional power systems are not feasible. For example, radioisotope thermoelectric generators were supplied by the DOE to the National Aeronautics and Space Administration for deep space missions including the Cassini Mission launched in October of 1997 to study the planet Saturn. For the Cassini Mission, ORNL produced carbon-bonded carbon fiber (CBCF) insulator sets, iridium alloy blanks and foil, and clad vent sets (CVS) used in the generators. The Oak Ridge National Laboratory (ORNL) has been involved in developing materials and technology and producing components for the DOE for more than three decades. This report reflects program guidance from the Office of Space and Defense Power Systems for fiscal year (FY) 2006. Production activities for prime quality (prime) CBCF insulator sets, iridium alloy blanks and foil, and CVS are summarized in this report. Technology activities are also reported that were conducted to improve the manufacturing processes, characterize materials, or to develop information for new radioisotope power systems.

Annual Technical Progress Report of Radioisotope Power System Materials Production and Technology Programs Tasks for October 1, 2005, through September 30, 2006

Energy Technology Data Exchange (ETDEWEB)

None

2006-09-30

The Office of Space and Defense Power Systems of the Department of Energy (DOE) provides Radioisotope Power Systems (RPS) for applications where conventional power systems are not feasible. For example, radioisotope thermoelectric generators were supplied by the DOE to the National Aeronautics and Space Administration for deep space missions including the Cassini Mission launched in October of 1997 to study the planet Saturn. For the Cassini Mission, ORNL produced carbon-bonded carbon fiber (CBCF) insulator sets, iridium alloy blanks and foil, and clad vent sets (CVS) used in the generators. The Oak Ridge National Laboratory (ORNL) has been involved in developing materials and technology and producing components for the DOE for more than three decades. This report reflects program guidance from the Office of Space and Defense Power Systems for fiscal year (FY) 2006. Production activities for prime quality (prime) CBCF insulator sets, iridium alloy blanks and foil, and CVS are summarized in this report. Technology activities are also reported that were conducted to improve the manufacturing processes, characterize materials, or to develop information for new radioisotope power systems.

Annual Technical Progress Report of Radioisotope Power Systems Materials Production and Technology Program Tasks for October 1, 2007 Through September 30,2008

Energy Technology Data Exchange (ETDEWEB)

King, James F [ORNL

2009-04-01

The Office of Radioisotope Power Systems (RPS) of the Department of Energy (DOE) provides RPS for applications where conventional power systems are not feasible. For example, radioisotope thermoelectric generators were supplied by the DOE to the National Aeronautics and Space Administration (NASA) for deep space missions including the Cassini Mission launched in October of 1997 to study the planet Saturn. For the Cassini Mission, ORNL produced carbon-bonded carbon fiber (CBCF) insulator sets, iridium alloy blanks and foil, and clad vent sets (CVS) used in the generators. The Oak Ridge National Laboratory (ORNL) has been involved in developing materials and technology and producing components for the DOE for more than three decades. This report reflects program guidance from the Office of RPS for fiscal year (FY) 2008. Production activities for prime quality (prime) CBCF insulator sets, iridium alloy blanks and foil, and CVS are summarized in this report. Technology activities are also reported that were conducted to improve the manufacturing processes, characterize materials, or to develop information for new RPS.

ANNUAL TECHNICAL PROGRESS REPORT OF RADIOISOTOPE POWER SYSTEM MATERIALS PRODUCTION AND TECHNOLOGY PROGRAM TASKS FOR OCTOBER 1, 2005 THROUGH SEPTEMBER 30, 2006

Energy Technology Data Exchange (ETDEWEB)

King, James F [ORNL

2007-04-01

The Office of Space and Defense Power Systems of the Department of Energy (DOE) provides Radioisotope Power Systems (RPS) for applications where conventional power systems are not feasible. For example, radioisotope thermoelectric generators were supplied by the DOE to the National Aeronautics and Space Administration for deep space missions including the Cassini Mission launched in October of 1997 to study the planet Saturn. For the Cassini Mission, ORNL produced carbon-bonded carbon fiber (CBCF) insulator sets, iridium alloy blanks and foil, and clad vent sets (CVS) used in the generators. The Oak Ridge National Laboratory (ORNL) has been involved in developing materials and technology and producing components for the DOE for more than three decades. This report reflects program guidance from the Office of Space and Defense Power Systems for fiscal year (FY) 2006. Production activities for prime quality (prime) CBCF insulator sets, iridium alloy blanks and foil, and CVS are summarized in this report. Technology activities are also reported that were conducted to improve the manufacturing processes, characterize materials, or to develop information for new radioisotope power systems.

Annual Technical Progress Report of Radioisotope Power Systems Materials Production and Technology Program Tasks for October 1, 2006 Through September 30, 2007

Energy Technology Data Exchange (ETDEWEB)

King, James F [ORNL

2008-04-01

The Office of Radioisotope Power Systems of the Department of Energy (DOE) provides Radioisotope Power Systems (RPS) for applications where conventional power systems are not feasible. For example, radioisotope thermoelectric generators were supplied by the DOE to the National Aeronautics and Space Administration for deep space missions including the Cassini Mission launched in October of 1997 to study the planet Saturn. For the Cassini Mission, ORNL produced carbon-bonded carbon fiber (CBCF) insulator sets, iridium alloy blanks and foil, and clad vent sets (CVS) used in the generators. The Oak Ridge National Laboratory (ORNL) has been involved in developing materials and technology and producing components for the DOE for more than three decades. This report reflects program guidance from the Office of Radioisotope Power Systems for fiscal year (FY) 2007. Production activities for prime quality (prime) CBCF insulator sets, iridium alloy blanks and foil, and CVS are summarized in this report. Technology activities are also reported that were conducted to improve the manufacturing processes, characterize materials, or to develop information for new radioisotope power systems.

Evaluation of a Ductility after High Temperature Oxidation with the Three-Point Bend Test in Zirconium Alloys

International Nuclear Information System (INIS)

Jung, Yang Il; Park, Sang Yoon; Park, Jeong Yong; Jeong, Yong Hwan

2010-01-01

In a light water reactor, the fuel cladding play an important role of preventing leakage of radioactive materials into the coolant, and thus the mechanical integrity of the cladding should be guaranteed under the conditions of normal and transient operation. In the case of a loss of coolant accident (LOCA), the cladding is subjected to a high temperature oxidation which is finally quenched because of an emergency coolant reflooding into the core. In this situation, the current LOCA criteria consist of five separate requirements: i) peak cladding temperature, ii) maximum cladding oxidation, iii) maximum hydrogen generation, iv) coolable geometry, and v) long-term cooling. The claddings lose their ductility due to the microstructural phase transformation from beta to martensite alpha-prime. and hydrogen up-take after LOCA. Since the reduction in ductility can induce embrittlement of claddings, post-quench ductility is one of the major concerns in transient operation circumstances. For the analysis, usually ring compression test are performed on ring samples cut from the tube to examine the oxidized cladding ductility. However, the test would not be applicable to the platelet samples which are general form of a specimen for developing alloys. As a high burn-up fuel cladding materials, Zircaloys are being replaced by modern zirconium alloys such as ZIRLO, and M5. Korea has also developed a new fuel cladding material HANA (high performance alloy for nuclear application) by the Korea Atomic Energy Research Institute. Because of the different composition of the newer claddings in comparison with the conventional Zircaloy-4, the high temperature oxidation behavior and the ductility after the oxidation would be different, and the properties should be evaluated how much the newer claddings were improved

Gel structure of the corrosion layer on cladding pipes of nuclear fuels

Czech Academy of Sciences Publication Activity Database

Medek, Jiří; Weishauptová, Zuzana

2009-01-01

Roč. 393, č. 2 (2009), s. 306-310 ISSN 0022-3115 R&D Projects: GA ČR GA106/04/0043 Institutional research plan: CEZ:AV0Z30460519 Keywords : cladding pipes of nuclear fuels * corrosion layer * zirconium alloys Subject RIV: JF - Nuclear Energetics Impact factor: 1.933, year: 2009

Formation of nano iridium oxide: material properties and neural cell culture

International Nuclear Information System (INIS)

Lee, In-Seop; Whang, Chung-Nam; Lee, Young-Hee; Hwan Lee, Gun; Park, Bong-Joo; Park, Jong-Chul; Seo, Won-Seon; Cui Fuzhai

2005-01-01

Iridium film with the thickness of 30 and 60 nm were formed on both Si wafer and commercially pure (CP) Ti by electron beam evaporation. The thin iridium film showed the identical charge injection capability with the bulk Ir. However, the charge injection value of iridium film was decreased with continuous potential cycling when the deposited iridium became depleted due to the formation of oxide. The number of cycles at which the charge injection value decreased was 800 and 1600 cycles for the 30- and 60-nm-thick Ir film, respectively. FE-SEM observations on the cross section of Ir film clearly showed the thicker iridium oxide was formed with the more potential cycling. Ar ion beam etching to substrates before deposition certainly improved the adhesion strength of Ir film enough to resist to the strain induced by the larger volume occupation of iridium oxide. Swiss 3T3 fibroblasts culture on Ir and Ir oxide showed no cytotoxicity. Also, embryonic cortical neural cell culture on electrode indicated neurons adhered and survived by the formation of neurofilament

Effects of La2O3 on microstructure and wear properties of laser clad γ/Cr7C3/TiC composite coatings on TiAl intermatallic alloy

International Nuclear Information System (INIS)

Liu Xiubo; Yu Rongli

2007-01-01

The effects of La 2 O 3 addition on the microstructure and wear properties of laser clad γ/Cr 7 C 3 /TiC composite coatings on γ-TiAl intermetallic alloy substrates with NiCr-Cr 3 C 2 precursor mixed powders have been investigated by optical microscopy (OM), scanning electron microscopy (SEM), X-ray diffraction (XRD) and energy-dispersive spectrometer (EDS) and block-on-ring wear tests. The responding wear mechanisms are discussed in detail. The results are compared with that for composite coating without La 2 O 3 . The comparison indicates that no evident new crystallographic phases are formed except a rapidly solidified microstructure consisting of the primary hard Cr 7 C 3 and TiC carbides and the γ/Cr 7 C 3 eutectics distributed in the tough γ nickel solid solution matrix. Good finishing coatings can be achieved under a proper amount of La 2 O 3 -addition and a suitable laser processing parameters. The additions of rare-earth oxide La 2 O 3 can refine and purify the microstructure of coatings, relatively decrease the volume fraction of primary blocky Cr 7 C 3 to Cr 7 C 3 /γ eutectics, reduce the dilution of clad material from base alloy and increase the microhardness of the coatings. When the addition of La 2 O 3 is approximately 4 wt.%, the laser clad composite coating possesses the highest hardness and toughness. The composite coating with 4 wt.%La 2 O 3 addition can result the best enhancement of wear resistance of about 30%. However, too less or excessive addition amount of La 2 O 3 have no better influence on wear resistance of the composite coating

High-power laser and arc welding of thorium-doped iridium alloys

International Nuclear Information System (INIS)

David, S.A.; Liu, C.T.

1980-05-01

The arc and laser weldabilities of two Ir-0.3% W alloys containing 60 and 200 wt ppM Th have been investigated. The Ir-.03% W alloy containing 200 wt ppM Th is severely prone to hot cracking during gas tungsten-arc welding. Weld metal cracking results from the combined effects of heat-affected zone liquation cracking and solidification cracking. Scanning electron microscopic analysis of the fractured surface revealed patches of low-melting eutectic. The cracking is influenced to a great extent by the fusion zone microstructure and thorium content. The alloy has been welded with a continuous-wave high-power CO 2 laser system with beam power ranging from 5 to 10 kW and welding speeds of 8 to 25 mm/s. Successful laser welds without hot cracking have been obtained in this particular alloy. This is attributable to the highly concentrated heat source available in the laser beam and the refinement in fusion zone microstructure obtained during laser welding. Efforts to refine the fusion zone structure during gas tungsten-arc welding of Ir-0.3 % W alloy containing 60 wt ppM Th were partially successful. Here transverse arc oscillation during gas tungsten-arc welding refines the fusion zone structure to a certain extent. However, microstructural analysis of this alloy's laser welds indicates further refinement in the fusion zone microstructure than in that from the gas tungsten-arc process using arc oscillations. The fusion zone structure of the laser weld is a strong function of welding speed

Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

Energy Technology Data Exchange (ETDEWEB)

Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Burtseva, T. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-08-30

The purpose of this research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures (PCT) and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile-to-brittle transition temperature (DBTT).

Quarterly Technical Progress Report of Radioisotope Power System Materials Production and Technology Program tasks for January 2000 through March 2000

International Nuclear Information System (INIS)

Moore, J.P.

2000-01-01

The Office of Space and Defense Power Systems (OSDPS) of the Department of Energy (DOE) provides radioisotope Power Systems (BPS) for applications where conventional power systems are not feasible. For example, radioisotope thermoelectric generators were supplied by the DOE to the National Aeronautics and Space Administration for deep space missions including the Cassini Mission launched in October of .I 997 to study the planet Saturn. The Oak Ridge National Laboratory (ORNL) has been involved in developing materials and technology and producing components for the DOE for more than three decades. For the Cassini Mission, for example, ORNL was involved in the production of carbon-bonded carbon fiber (CBCF) insulator sets, iridium alloy blanks and foil, and clad vent sets (CVSs) and weld shields (WSs). This quarterly report has been divided into three sections to reflect program guidance from OSDPS for fiscal year (FY) 2000. The first section deals primarily with maintenance of the capability to produce flight quality carbon-bonded carbon fiber (CBCF) insulator sets, iridium alloy blanks and foil, clad vent sets (CVSs), and weld shields (WSs). In all three cases, production maintenance is assured by the manufacture of limited quantities of flight quality (FQ) components. The second section deals with several technology activities to improve the manufacturing processes, characterize materials, or to develop technologies for two new RPS. The last section is dedicated to studies of the potential for the production of 238Pu at OBNL

Results of post-irradiation examination of WWER fuel assembly structural components made of E110 and E635 alloys

International Nuclear Information System (INIS)

Smirnov, A.; Markov, D.; Smirnov, V.; Polenok, V.; Ivashchenko, A.; Strozhuk, A.

2006-01-01

The paper presents the main examination results on the condition of fuel rods claddings, guide tubes and spacer grids of the WWER FA made of E110 and E635 alloys operated under standard operating conditions. The paper is based on the data obtained during the examination of 28 WWER-1000 FA and 12 WWER-400 FA. E110 alloy is shown to be suitable material for the WWER fuel rod claddings under the normal operating conditions. E635 alloy is attractive to manufacturing of the skeleton components. The currently used combination (E110 as a material of fuel rods claddings and E635 - as a material of the skeleton components) is the optimal solution for the WWER fuel assembly because the advantages of the both alloys are used. (authors)

Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario

Energy Technology Data Exchange (ETDEWEB)

Giovedi, Claudia; Martins, Marcelo Ramos, E-mail: claudia.giovedi@labrisco.usp.br, E-mail: mrmartin@usp.br [Laboratorio de Analise, Avaliacao e Gerenciamento de Risco (LabRisco/POLI/USP), São Paulo, SP (Brazil); Abe, Alfredo; Muniz, Rafael O.R.; Gomes, Daniel de Souza; Silva, Antonio Teixeira e, E-mail: ayabe@ipen.br, E-mail: dsgomes@ipen.br, E-mail: teixiera@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

2017-07-01

Accident tolerant fuels (ATF) has been studied since the Fukushima Daiichi accident in the research efforts to develop new materials which under accident scenarios could maintain the fuel rod integrity for a longer period compared to the cladding and fuel system usually utilized in Pressurized Water Reactors (PWR). The efforts have been focused on new materials applied as cladding, then iron-base alloys appear as a possible candidate. The aim of this paper is to implement modifications in a fuel performance code to evaluate the behavior of iron based alloys under Loss-of-Coolant Accident (LOCA) scenario. For this, initially the properties related to the thermal and mechanical behavior of iron-based alloys were obtained from the literature, appropriately adapted and introduced in the fuel performance code subroutines. The adopted approach was step by step modifications, where different versions of the code were created. The assessment of the implemented modification was carried out simulating an experiment available in the open literature (IFA-650.5) related to zirconium-based alloy fuel rods submitted to LOCA conditions. The obtained results for the iron-based alloy were compared to those obtained using the regular version of the fuel performance code for zircaloy-4. The obtained results have shown that the most important properties to be changed are those from the subroutines related to the mechanical properties of the cladding. The results obtained have shown that the burst is observed at a longer time for fuel rods with iron-based alloy, indicating the potentiality of this material to be used as cladding with ATF purposes. (author)

Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario

International Nuclear Information System (INIS)

Giovedi, Claudia; Martins, Marcelo Ramos; Abe, Alfredo; Muniz, Rafael O.R.; Gomes, Daniel de Souza; Silva, Antonio Teixeira e

2017-01-01

Accident tolerant fuels (ATF) has been studied since the Fukushima Daiichi accident in the research efforts to develop new materials which under accident scenarios could maintain the fuel rod integrity for a longer period compared to the cladding and fuel system usually utilized in Pressurized Water Reactors (PWR). The efforts have been focused on new materials applied as cladding, then iron-base alloys appear as a possible candidate. The aim of this paper is to implement modifications in a fuel performance code to evaluate the behavior of iron based alloys under Loss-of-Coolant Accident (LOCA) scenario. For this, initially the properties related to the thermal and mechanical behavior of iron-based alloys were obtained from the literature, appropriately adapted and introduced in the fuel performance code subroutines. The adopted approach was step by step modifications, where different versions of the code were created. The assessment of the implemented modification was carried out simulating an experiment available in the open literature (IFA-650.5) related to zirconium-based alloy fuel rods submitted to LOCA conditions. The obtained results for the iron-based alloy were compared to those obtained using the regular version of the fuel performance code for zircaloy-4. The obtained results have shown that the most important properties to be changed are those from the subroutines related to the mechanical properties of the cladding. The results obtained have shown that the burst is observed at a longer time for fuel rods with iron-based alloy, indicating the potentiality of this material to be used as cladding with ATF purposes. (author)

Microdiffraction and CBED crystal structure determination of the Si-rich phase in laser-clad Ni alloy FP-5

International Nuclear Information System (INIS)

Liu, Y.; Mazumder, J.

1995-01-01

This paper demonstrates an example of using kinematical microdiffraction to solve an unknown Si-rich phase of micrometer size in a laser-clad Ni alloy FP-5 on Al alloy AA333. The composition of the Si-rich phase obtained by energy-dispersice X-ray spectroscopy (EDX) analysis in a transmission electron microscope is approximately 0.7wt%Al, 71wt%Si, 3.3wt%Cr, 0.8wt%Fe, 21wt%Ni and 2.8wt%Cu. The point group was identified by the standard convergent-beam symmetry analysis to be P6 3 /mmc (No. 194). Structure analysies by microdiffraction (MD) indicates that the Si-rich phase is a close-packed structure.The intensity distribution in the microdiffraction pattern of the [1120] zone axis taken with a very thin area showed a close match with kinematical calculation. A close-packed-structure model specified as ABCACB was deduced from the [1120] zone axis MD pattern. The randomly distributed atoms of all the elements in the unit cell are at 2/3, 1/3-1/12; 1/3, 1/12; 0, 0, 3/12, 1/3, 2/3, 5/12, 2/3, 1/3, 7/12; 0,0, 9/12. The model was checked by comparison with a simulated diffraction pattern map and with a simulated [0001] zone-axis CBED pattern, which showed complete agreement with the proposed model. (orig.)

Laser cladding of bioactive glass coatings.

Science.gov (United States)

Comesaña, R; Quintero, F; Lusquiños, F; Pascual, M J; Boutinguiza, M; Durán, A; Pou, J

2010-03-01

Laser cladding by powder injection has been used to produce bioactive glass coatings on titanium alloy (Ti6Al4V) substrates. Bioactive glass compositions alternative to 45S5 Bioglass were demonstrated to exhibit a gradual wetting angle-temperature evolution and therefore a more homogeneous deposition of the coating over the substrate was achieved. Among the different compositions studied, the S520 bioactive glass showed smoother wetting angle-temperature behavior and was successfully used as precursor material to produce bioactive coatings. Coatings processed using a Nd:YAG laser presented calcium silicate crystallization at the surface, with a uniform composition along the coating cross-section, and no significant dilution of the titanium alloy was observed. These coatings maintain similar bioactivity to that of the precursor material as demonstrated by immersion in simulated body fluid. Copyright 2009 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

Synthesis of benzimidazoles via iridium-catalyzed acceptorless dehydrogenative coupling.

Science.gov (United States)

Sun, Xiang; Lv, Xiao-Hui; Ye, Lin-Miao; Hu, Yu; Chen, Yan-Yan; Zhang, Xue-Jing; Yan, Ming

2015-07-21

Iridium-catalyzed acceptorless dehydrogenative coupling of tertiary amines and arylamines has been developed. A number of benzimidazoles were prepared in good yields. An iridium-mediated C-H activation mechanism is suggested. This finding represents a novel strategy for the synthesis of benzimidazoles.

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

International Nuclear Information System (INIS)

Lee, Changmin; Park, Hyungkwon; Yoo, Jaehong; Lee, Changhee; Woo, WanChuck; Park, Sunhong

2015-01-01

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

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

Physico Chemistry of the Chlorination of Aluminum Claddings in the Framework of HALOX Project

International Nuclear Information System (INIS)

Alvarez, Fabiola; De Micco, Georgina; Bohe, Ana; Pasquevich, Daniel

2003-01-01

The conditioning of spent nuclear fuels from test and research reactors requires a previous physicochemical treatment to stabilize them chemically.A possible way of processing is through what was called in CNEA as Process HALOX (Halogenation and Oxidation).It consists of the selective separation of cladding by halogenation and the subsequent oxidation of the core, previously to insert it into a vitreous matrix.The halogenation aim is to transform the constituents of the 6061aluminum alloy into volatile halides.In this work we present preliminary results of the chlorination of two aluminum alloys: AA 6061 and a type of CuZnAl alloy

Arisings of cladding wastes from nuclear fuel in the European Community

International Nuclear Information System (INIS)

Cottone, G.

1978-01-01

An inquiry has been made in the member states on composition, activation and amounts of cladding wastes arising in the European Community until 1990 from the following reactor types: BWR, PWR, SGHWR, AGR and FBR. The elaborated results of this inquiry are given in this report. On the basis of forecasted reprocessing capacities the cumulative amount of cladding waste in the Community was estimated to reach in 1985 and 1990, respectively, about 2,100 and 7,300 metric tons. This waste will mainly consist of zircaloy and of smaller amounts of stainless steel and nickel alloy. Assuming that 0.5% of the spent fuel remains with the cladding, the contamination has been estimated for cooling times varying from 1 to 1000 years. In the first centuries activation is prevailing, but contamination determines the long-term radioactivity; consequently better decontamination, removing the alpha-bearing compounds, would be beneficial in reducing the long term hazard

Experimental creep behaviour determination of cladding tube materials under multi-axial loadings

International Nuclear Information System (INIS)

Grosjean, Catherine; Poquillon, Dominique; Salabura, Jean-Claude; Cloue, Jean-Marc

2009-01-01

Cladding tubes are structural parts of nuclear plants, submitted to complex thermomechanical loadings. Thus, it is necessary to know and predict their behaviour to preserve their integrity and to enhance their lifetime. Therefore, a new experimental device has been developed to control the load path under multi-axial load conditions. The apparatus is designed to determine the thermomechanical behaviour of zirconium alloys used for cladding tubes. First results are presented. Creep tests with different biaxial loadings were performed. Results are analysed in terms of thermal expansion and of creep strain. The anisotropy of the material is revealed and iso-creep strain curves are given.

Effect of Temperature and Sheet Temper on Isothermal Solidification Kinetics in Clad Aluminum Brazing Sheet

Science.gov (United States)

Benoit, Michael J.; Whitney, Mark A.; Wells, Mary A.; Winkler, Sooky

2016-09-01

Isothermal solidification (IS) is a phenomenon observed in clad aluminum brazing sheets, wherein the amount of liquid clad metal is reduced by penetration of the liquid clad into the core. The objective of the current investigation is to quantify the rate of IS through the use of a previously derived parameter, the Interface Rate Constant (IRC). The effect of peak temperature and initial sheet temper on IS kinetics were investigated. The results demonstrated that IS is due to the diffusion of silicon (Si) from the liquid clad layer into the solid core. Reduced amounts of liquid clad at long liquid duration times, a roughened sheet surface, and differences in resolidified clad layer morphology between sheet tempers were observed. Increased IS kinetics were predicted at higher temperatures by an IRC model as well as by experimentally determined IRC values; however, the magnitudes of these values are not in good agreement due to deficiencies in the model when applied to alloys. IS kinetics were found to be higher for sheets in the fully annealed condition when compared with work-hardened sheets, due to the influence of core grain boundaries providing high diffusivity pathways for Si diffusion, resulting in more rapid liquid clad penetration.

Additive Manufacturing of High-Entropy Alloys by Laser Processing

NARCIS (Netherlands)

Ocelik, V.; Janssen, Niels; Smith, Stefan; De Hosson, J. Th M.

This contribution concentrates on the possibilities of additive manufacturing of high-entropy clad layers by laser processing. In particular, the effects of the laser surface processing parameters on the microstructure and hardness of high-entropy alloys (HEAs) were examined. AlCoCrFeNi alloys with

Weldability of general purpose heat source new-process iridium

International Nuclear Information System (INIS)

Kanne, W.R.

1987-01-01

Weldability tests on General Purpose Heat Source (GPHS) iridium capsules showed that a new iridium fabrication process reduced susceptibility to underbead cracking. Seventeen capsules were welded (a total of 255 welds) in four categories and the number of cracks in each weld was measured

Synthesis and characterization of low-valence actinide phosphide tellurides and ternary selenium-halide iridium complexes; Synthese und Charakterisierung niedervalenter Actinoidphosphidtelluride und ternaerer Selen-Halogenid-Komplexe des Iridiums

Energy Technology Data Exchange (ETDEWEB)

Stolze, Karoline

2016-04-07

The thesis on the synthesis and characterization of low-valence actinide phosphide tellurides and ternary selenium-halide iridium complexes includes two parts: a description of the experimental synthesis of UPTe and U2PTe2O and ThPTe and the synthesis of selenium-chloride iridium complexes and selenium-bromide iridium complexes. The characterization included X-ray diffraction and phase studies.

Nonequilibrium synthesis of Nb-Al alloys by laser processing

International Nuclear Information System (INIS)

Tewari, S.K.; Mazumder, J.

1993-01-01

The technique of laser surface modification provides a unique means of synthesizing novel nonequilibrium materials in near net shape. Claddings of several NbAl 3 alloys with Ti, B and Hf as a ternary alloy addition were prepared using a CW CO 2 laser. Isothermal oxidation behavior of the clads were examined in air. Oxidation tests at 800, 1,200 and 1,400 C. Alternating layers of alumina and NbAlO 4 were not observed in any of the samples as reported in literature for conventionally processed NbAl 3 oxidized under similar conditions. The parabolic rate constants for all the alloys, except 0 B, were comparable to that for isothermal oxidation of β-NiAl, at 1,200 and 1,400 C in 0.1 atm oxygen, which is a known alumina former. Ternary alloying additions for improved oxidation resistance at 1,400 C accompanied with improved ductility were identified

Modeling of High Temperature Oxidation Behavior of FeCrAl Alloy by using Artificial Neural Network

Energy Technology Data Exchange (ETDEWEB)

Kim, Jae Joon; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

2016-10-15

Refractory alloys are candidate materials for replacing current zirconium-base cladding of light water reactors and they retain significant creep resistance and mechanical strength at high temperatures up to 1500 ℃ due to their high melting temperature. Thermal neutron cross sections of refractory metals are higher than that of zirconium, however the loss of neutron can be overcome by reducing cladding thickness which can be facilitated with enhanced mechanical properties. However, most refractory metals show the poor oxidation resistance at a high temperature. Oxidation behaviors of the various compositions of FeCrAl alloys in high temperature conditions were modeled by using Bayesian neural network. The automatic relevance determination (ARD) technique represented the influence of the composition of alloying elements on the oxidation resistance of FeCrAl alloys. This model can be utilized to understand the tendency of oxidation behavior along the composition of each element and prove the applicability of neural network modeling for the development of new cladding material of light water reactors.

Laser welding parameters for manufacturing iridium-192 (Ir-192) source

International Nuclear Information System (INIS)

Anung Pujiyanto; Moch Subechi; Hotman Lubis; Diandono KY

2013-01-01

Number of cervical cancer patients in Indonesia is growing every year. One of cervical cancer treatment was fairly effective use brachytherapy treatment with radioisotope sources of iridium-192. Manufacturing of iridium sources for brachytherapy can be done by incorporating the iridium-192 into stainless steel microcapsules then welding using laser welder which the quality of the welding of iridium source (Ir-192) was determined by the welding parameters such as full power, energy frequency, average power and speed. Based on the result of leakage test using pressure -20 inch Hg and tensile test 2.5 bar showed the welding parameters III and IV did not have leakage and damaged. So that parameters III and IV are recommended to be applied to Ir-192 HDR's source. (author)

Irradiation Performance of U-Mo Alloy Based ‘Monolithic’ Plate-Type Fuel – Design Selection

Energy Technology Data Exchange (ETDEWEB)

A. B. Robinson; G. S. Chang; D. D. Keiser, Jr.; D. M. Wachs; D. L. Porter

2009-08-01

A down-selection process has been applied to the U-Mo fuel alloy based monolithic plate fuel design, supported by irradiation testing of small fuel plates containing various design parameters. The irradiation testing provided data on fuel performance issues such as swelling, fuel-cladding interaction (interdiffusion), blister formation at elevated temperatures, and fuel/cladding bond quality and effectiveness. U-10Mo (wt%) was selected as the fuel alloy of choice, accepting a somewhat lower uranium density for the benefits of phase stability. U-7Mo could be used, with a barrier, where the trade-off for uranium density is critical to nuclear performance. A zirconium foil barrier between fuel and cladding was chosen to provide a predictable, well-bonded, fuel-cladding interface, allowing little or no fuel-cladding interaction. The fuel plate testing conducted to inform this selection was based on the use of U-10Mo foils fabricated by hot co-rolling with a Zr foil. The foils were subsequently bonded to Al-6061 cladding by hot isostatic pressing or friction stir bonding.

Welding of iridium heat source capsule components

International Nuclear Information System (INIS)

Mustaleski, T.M.; Yearwood, J.C.; Burgan, C.E.; Green, L.A.

1991-01-01

Interplanetary spacecraft have long used radioisotope thermoelectric generators (RTG) to produce power for instrumentation. These RTG produce electrical energy from the heat generated through the radioactive decay of plutonium-238. The plutonium is present as a ceramic pellet of plutonium oxide. The pellet is encapsulated in a containment shell of iridium. Iridium is the material of choice for these capsules because of its compatibility with the plutonium dioxide. The high-energy beam welding (electron beam and laser) processes used in the fabrication of the capsules has not been published. These welding procedures were originally developed at the Mound Laboratories and have been adapted for use at the Oak Ridge Y-12 Plant. The work involves joining of thin material in small sizes to exacting tolerances. There are four different electron beam welds on each capsule, with one procedure being used in three locations. There is also a laser weld used to seal the edges of a sintered frit assembly. An additional electron beam weld is also performed to seal each of the iridium blanks in a stainless steel waster sheet prior to forming. In the transfer of these welding procedures from one facility to another, a number of modifications were necessary. These modifications are discussed in detail, as well as the inherent problems in making welds in material which is only 0.005 in. thick. In summary, the paper discusses the welding of thin components of iridium using the high energy beam processes. While the peculiarities of iridium are pertinent to the discussion, much of the information is of general interest to the users of these processes. This is especially true of applications involving thin materials and high-precision assemblies