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Sample records for alloy primary water

  1. Stress corrosion cracking of nickel base alloys in PWR primary water

    International Nuclear Information System (INIS)

    Guerre, C.; Chaumun, E.; Crepin, J.; De Curieres, I.; Duhamel, C.; Heripre, E.; Herms, E.; Laghoutaris, P.; Molins, R.; Sennour, M.; Vaillant, F.

    2013-01-01

    Stress corrosion cracking (SCC) of nickel base alloys and associated weld metals in primary water is one of the major concerns for pressurized water reactors (PWR). Since the 90's, highly cold-worked stainless steels (non-sensitized) were also found to be susceptible to SCC in PWR primary water ([1], [2], [3]). In the context of the life extension of pressurized water reactors, laboratory studies are performed in order to evaluate the SCC behaviour of components made of nickel base alloys and of stainless steels. Some examples of these laboratory studies performed at CEA will be given in the talk. This presentation deals with both initiation and propagation of stress corrosion cracks. The aims of these studies is, on one hand, to obtain more data regarding initiation time or crack growth rate and, one the other hand, to improve our knowledge of the SCC mechanisms. The aim of these approaches is to model SCC and to predict components life duration. Crack growth rate (CGR) tests on Alloy 82 with and without post weld heat treatment are performed in PWR primary water (Figure 1). The heat treatment seems to be highly beneficial by decreasing the CGR. This result could be explained by the effect of thermal treatment on the grain boundary nano-scopic precipitation in Alloy 82 [4]. The susceptibility to SCC of cold worked austenitic stainless steels is also studied. It is shown that for a given cold-working procedure, SCC susceptibility increases with increasing cold-work ([2], [5]). Despite the fact that the SCC behaviour of Alloy 600 has been widely studied for many years, recent laboratory experiments and analysis ([6], [7], [8]) showed that oxygen diffusion is not a rate-limiting step in the SCC mechanism and that chromium diffusion in the bulk close the crack tip could be a key parameter. (authors)

  2. Metallurgical and mechanical parameters controlling alloy 718 stress corrosion cracking resistance in PWR primary water

    International Nuclear Information System (INIS)

    Deleume, J.

    2007-11-01

    Improving the performance and reliability of the fuel assemblies of the pressurized water reactors requires having a perfect knowledge of the operating margins of both the components and the materials. The choice of alloy 718 as reference material for this study is justified by the industrial will to identify the first order parameters controlling the excellent resistance of this alloy to Stress Corrosion Cracking (SCC). For this purpose, a specific slow strain rate (SSR) crack initiation test using tensile specimen with a V-shaped hump in the middle of the gauge length was developed and modeled. The selectivity of such SSR tests in simulated PWR primary water at 350 C was clearly established by characterizing the SCC resistance of nine alloy 718 thin strip heats. Regardless of their origin and in spite of a similar thermo-mechanical history, they did not exhibit the same susceptibility to SCC crack initiation. All the characterized alloy 718 heats develop oxide scale of similar nature for various exposure times to PWR primary medium in the temperature range [320 C - 360 C]. δ phase precipitation has no impact on alloy 718 SCC initiation behavior when exposed to PWR primary water, contrary to interstitial contents and the triggering of plastic instabilities (PLC phenomenon). (author)

  3. Hydrogen absorption mechanisms and hydrogen interactions - defects: implications to stress corrosion of nickel based alloys in pressurized water reactors primary water

    International Nuclear Information System (INIS)

    Jambon, F.

    2012-01-01

    Since the late 1960's, a special form of stress corrosion cracking (SCC) has been identified for Alloy 600 exposed to pressurized water reactors (PWR) primary water: intergranular cracks develop during the alloy exposure, leading, progressively, to the complete ruin of the structure, and to its replacement. The main goal of this study is therefore to evaluate in which proportions the hydrogen absorbed by the alloy during its exposure to the primary medium can be responsible for SCC crack initiation and propagation. This study is aimed at better understanding of the hydrogen absorption mechanism when a metallic surface is exposed to a passivating PWR primary medium. A second objective is to characterize the interactions of the absorbed hydrogen with the structural defects of the alloy (dislocations, vacancies...) and evaluate to what extent these interactions can have an embrittling effect in relation with SCC phenomenon. Alloy 600-like single-crystals were exposed to a simulated PWR medium where the hydrogen atoms of water or of the pressuring hydrogen gas were isotopically substituted with deuterium, used as a tracer. Secondary ion mass spectrometry depth-profiling of deuterium was performed to characterize the deuterium absorption and localization in the passivated alloy. The results show that the hydrogen absorption during the exposure of the alloy to primary water is associated with the water molecules dissociation during the oxide film build-up. In an other series of experiments, structural defects were created in recrystallized samples, and finely characterized by positron annihilation spectroscopy and transmission electron microscopy, before or after the introduction of cathodic hydrogen. These analyses exhibited a strong hydrogen/defects interaction, evidenced by their structural reorganization under hydrogenation (coalescence, migrations). However, thermal desorption spectroscopy analyses indicated that these interactions are transitory, and dependent on

  4. Low cycle fatigue of Alloy 690 and welds in a simulated PWR primary water environment

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jongdae; Cho, Pyungyeon; Jang, Changheui [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Cho, Pyungyeon [Khalifa Univ., Abu Dhabi (United Arab Emirates); Kim, Tae Soon; Lee, Yong Sung [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2013-05-15

    In this study, environmental fatigue tests for these materials were performed and the new prediction model of fatigue life of Alloy 690 and weld in primary water condition was proposed. To evaluate the fatigue life of Alloy 690 and 52M in a PWR environment, low cycle fatigue tests were performed and revised fatigue life prediction models and environmental factor were proposed. With the revised Fen model for Alloy 690 and 52M, the reliability of the fatigue life prediction has been improved. The reduction of low cycle fatigue life of metallic materials in the primary coolant water environments has been the subject of debate between the utility and regulator since 1980s. It became the significant licensing problem since the issue of RG-1.207 by U. S. NRC. The statistical model for the environmental factor, Fen, specified in RG-1.207 was based on the extensive test results accumulated by the ANL and Japanese national program. Of the materials, the limited fatigue life data of Ni-Cr-Fe alloys were used to develop the Fen for the alloys. Furthermore, test data for Alloy 690 and its weld are limited. Considering that Alloy 690 will be extensively used in the new nuclear power plants, additional effort to validate or improve current Fen model is required.

  5. The effect of prior deformation on stress corrosion cracking growth rates of Alloy 600 materials in a simulated pressurized water reactor primary water

    International Nuclear Information System (INIS)

    Yamazaki, Seiya; Lu Zhanpeng; Ito, Yuzuru; Takeda, Yoichi; Shoji, Tetsuo

    2008-01-01

    The effect of prior deformation on stress corrosion cracking (SCC) growth rates of Alloy 600 materials in a simulated pressurized water reactor primary water environment is studied. The prior deformation was introduced by welding procedure or by cold working. Values of Vickers hardness in the Alloy 600 weld heat-affected zone (HAZ) and in the cold worked (CW) Alloy 600 materials are higher than that in the base metal. The significantly hardened area in the HAZ is within a distance of about 2-3 mm away from the fusion line. Electron backscatter diffraction (EPSD) results show significant amounts of plastic strain in the Alloy 600 HAZ and in the cold worked Alloy 600 materials. Stress corrosion cracking growth rate tests were performed in a simulated pressurized water reactor primary water environment. Extensive intergranular stress corrosion cracking (IGSCC) was found in the Alloy 600 HAZ, 8% and 20% CW Alloy 600 specimens. The crack growth rate in the Alloy 600 HAZ is close to that in the 8% CW base metal, which is significantly lower than that in the 20% CW base metal, but much higher than that in the as-received base metal. Mixed intergranular and transgranular SCC was found in the 40% CW Alloy 600 specimen. The crack growth rate in the 40% CW Alloy 600 was lower than that in the 20% CW Alloy 600. The effect of hardening on crack growth rate can be related to the crack tip mechanics, the sub-microstructure (or subdivision of grain) after cross-rolling, and their interactions with the oxidation kinetics

  6. Primary water stress corrosion cracking resistance of alloy 690 heat affected zones of butt welds

    International Nuclear Information System (INIS)

    Fournier, L.; Calonne, O.; Toloczko, M.B.; Bruemmer, S.M.; Massoud, J.P.; Lemaire, E.; Gerard, R.; Somville, F.; Richnau, A.; Lagerstrom, J.

    2015-01-01

    A wide V-groove butt weld was fabricated from Alloy 690 plates using Alloy 152 filler material, maximum allowable heat input, and very stiff strong-backs. Alloy 690 heat affected zones (HAZ) was characterized in terms of microstructure and plastic strains induced by weld shrinkage. Crack initiation tests were carried out in pure hydrogenated steam at 400 C. degrees for 4000 h. Crack growth rate tests were performed in simulated PWR primary water at a temperature of 360 C. degrees. A maximum plastic strain around 5% was measured in the vicinity of the fusion line, which decreased almost linearly with the distance from the fusion line. Crack initiation tests on Alloy 690 HAZ specimens as well as on 30% cold-rolled Alloy 690 specimens were performed in pure hydrogenated steam at 400 C. degrees (partial pressure of hydrogen = 0.7 bar) for a total of 4000 h using cylindrical notched tensile specimens, reverse U-bends and flat micro-tensile specimens. No crack initiation was detected. Stress corrosion propagation rates revealed extremely low SCC (Stress Corrosion Cracking) growth rates both in the base metal and in the HAZ region whose magnitudes are of no engineering significance. Overall, the results indicated limited plastic strain induced by weld shrinkage in butt weld HAZ, and to no particular susceptibility of primary water stress corrosion cracking. (authors)

  7. Quantitative assessment of intergranular damage due to PWR primary water exposure in structural Ni-based alloys

    International Nuclear Information System (INIS)

    Ter-Ovanessian, Benoît; Deleume, Julien; Cloué, Jean-Marc; Andrieu, Eric

    2013-01-01

    Highlights: ► IG damage occurred on Ni-base alloys during exposure at high temperature water. ► Two characterization methods yield a tomographic analysis of this IG damage. ► Connected or isolated intergranular oxygen/oxide penetrations are quantified. ► Such quantitative description provides information on IGSCC susceptibility. - Abstract: Two nickel-based alloys, alloy 718 and alloy 600, known to have different resistances to IGSCC, were exposed to a simulated PWR primary water environment at 360 °C for 1000 h. The intergranular oxidation damage was analyzed in detail using an original approach involving two characterization methods (Incremental Mechanical Polishing/Microcopy procedure and SIMS imaging) which yielded a tomographic analysis of the damage. Intergranular oxygen/oxide penetrations occurred either as connected or isolated penetrations deep under the external oxide/substrate interface as far as 10 μm for alloy 600 and only 4 μm for alloy 718. Therefore, assessing this damage precisely is essential to interpret IGSCC susceptibility.

  8. Influence of microstructure on stress corrosion cracking susceptibility of alloys 600 and 690 in primary water of pressurized water reactors

    International Nuclear Information System (INIS)

    Kergaravat, J.F.

    1996-01-01

    performed in primary water on the more GBS sensitive materials (Alloy 600 and 690). These specimens showed an important susceptibility to SCC failure. Based on these results, we proposed a new cracking mechanism for Alloy 600 SCC in high temperature primary water. This mechanism takes into account the GBS role during both initiation and propagation steps of failure. (author)

  9. Effect of microstructure on crack growth rate of alloy 690 in primary water

    International Nuclear Information System (INIS)

    Sakakibara, Y.; Hirano, T.; Nakayama, G.

    2015-01-01

    It was reported that the chemical composition and fabrication process of alloy 690 were important for the resistance to SCC in primary water. In this paper, we evaluated crack growth rate (CGR) of commercial thick plates (WT, XT) and forgings (FT, FM) made by some material manufacturers. Specimen WT showed the highest CGR in the thick plates. WT had coarse grain and film-like carbides which were assumed as eutectic M 23 C 6 . Forged alloy 690 MA and TT (FM and FT) showed no CGR. One of alloy 690 plates (XT) was cold rolled by 30% of reduction in our laboratory to investigate the effect of the orientation of the specimen on CGR. The specimens in the S-L and the S-T orientation showed higher CGRs than those in the T-L and the L-S orientation. (authors)

  10. Stress corrosion mechanisms of alloy-600 polycrystals and monocrystals in primary water: effect of hydrogen

    International Nuclear Information System (INIS)

    Foct, F.

    1999-01-01

    The aim of this study is to identify the mechanisms involved in Alloy 600 primary water stress corrosion cracking. Therefore, this work is mainly focussed on the two following points. The first one is to understand the influence of hydrogen on SCC of industrial Alloy 600 and the second one is to study the crack initiation and propagation on polycrystals and single crystals. A cathodic potential applied during slow strain rate tests does not affect crack initiation but increases the slow crack growth rate by a factor 2 to 5. Cathodic polarisation, cold work and 25 cm 3 STP/kg hydrogen content increase the slow CGR so that the K ISCC (and therefore fast CGR) is reached. The influence of hydrogenated primary water has been studied for the first time on Alloy 600 single crystals. Cracks cannot initiate on tensile specimens but they can propagate on pre-cracked specimens. Transgranular cracks present a precise crystallographic aspect which is similar to that of 316 alloy in MgCl 2 solutions. Moreover, the following results improve the description of the cracking conditions. Firstly, the higher the hydrogen partial pressure, the lower the Alloy 600 passivation current transients. Since this result is not correlated with the effect of hydrogen on SCC, cracking is not caused by a direct effect of dissolved hydrogen on dissolution. Secondly, hydrogen embrittlement of Alloy 600 disappears at temperatures above 200 deg.C. Thirdly, grain boundary sliding (GBS) does not directly act on SCC but shows the mechanical weakness of grain boundaries. Regarding the proposed models for Alloy 600 SCC, it is possible to draw the following conclusions. Internal oxidation or absorbed hydrogen effects are the most probable mechanisms for initiation. Dissolution, internal oxidation and global hydrogen embrittlement models cannot explain crack propagation. On the other hand, the Corrosion Enhanced Plasticity Model gives a good description of the SCC propagation. (author)

  11. Corrosion Behavior and Oxide Properties of Zr-Nb-Cu and Zr-Nb-Sn Alloy in High Dissolved Hydrogen Primary Water Chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yun Ju; Kim, Tae Ho; Kim, Ji Hyun [UNIST, Ulsan (Korea, Republic of)

    2016-05-15

    The water-metal interface is regarded as rate-controlling site governing the rapid oxidation transition in high burn-up fuel. And the zirconium oxide is made in water-metal interface and its structure and phase do an important role in terms of oxide properties. During oxidation process, the protective tetragonal oxide layer develops at the interface due to accumulated high stress during oxide growth, and it turns into non-protective monoclinic oxide with increasing oxide thickness, thus decreasing the stress. It has been reported that Nb addition was proven to be very beneficial for increasing the corrosion resistance of the zirconium alloys. From a more recent study, Cu addition in Nb containing Zirconium alloy was reported to be effective for increasing corrosion resistance in water containing B and Li. According to the previous research conducted, Zr-Nb-Cu shows better corrosion resistance than Zircaloy-4. The dissolved hydrogen (DH) concentration is the key issue of primary water chemistry, and the effect of DH concentration on the corrosion rate of nickel based alloy has been researched. However, the effect of DH on the zirconium alloy corrosion mechanism was not fully investigated. In this study, the weight gain measurement, FIB-SEM analysis, and Raman spectroscopic measurement were conducted to investigate the effects of dissolved hydrogen concentration and the chemical composition on the corrosion resistance and oxide phase of Zr-Nb-Cu alloy and Zr-Nb-Sn alloy after oxidizing in a primary water environment for 20 d. The corrosion rate of Zr-Nb-Cu alloy is slow, when it is compared to Zr-Nb-Sn alloy. In SEM images, the oxide thickness of Zr-Nb-Cu alloy is measured to be around 1.06 μm it of Zr-Nb-Sn alloy is measured to be 1.15 μm. It is because of the Segregation made by Sn solute element when Sn solute element oxidized. And according to ex situ Raman spectra, Zr-Nb-Cu alloy oxide has more tetragonal zirconium oxide fraction than Zr-Nb-Sn alloy oxide.

  12. Stress corrosion cracking of Alloy 600 in primary water of PWR: study of chromium diffusion

    International Nuclear Information System (INIS)

    Chetroiu, Bogdan-Adrian

    2015-01-01

    Alloy 600 (Ni-15%Cr-10%Fe) is known to be susceptible to Stress Corrosion Cracking (SCC) in primary water of Pressurized Water Reactors (PWR). Recent studies have shown that chromium diffusion is a controlling rate step in the comprehension of SCC mechanism. In order to improve the understanding and the modelling of SCC of Alloy 600 in PWR primary medium the aim of this study was to collect data on kinetics diffusion of chromium. Volume and grain boundary diffusion of chromium in pure nickel and Alloy 600 (mono and poly-crystals) has been measured in the temperature range 678 K to 1060 K by using Secondary Ions Mass Spectroscopy (SIMS) and Glow Discharge-Optical Spectrometry (GD-OES) techniques. A particular emphasis has been dedicated to the influence of plastic deformation on chromium diffusion in nickel single crystals (orientated <101>) for different metallurgical states. The experimental tests were carried out in order to compare the chromium diffusion coefficients in free lattice (not deformed), in pre-hardening specimens (4% and 20%) and in dynamic deformed tensile specimens at 773 K. It has been found that chromium diffusivity measured in dynamic plastic deformed creep specimens were six orders of magnitude greater than those obtained in not deformed or pre-hardening specimens. The enhancement of chromium diffusivity can be attributed to the presence of moving dislocations generated during plastic deformation. (author)

  13. Effect of Local Strain Distribution of Cold-Rolled Alloy 690 on Primary Water Stress Corrosion Crack Growth Behavior

    Directory of Open Access Journals (Sweden)

    Kim S.-W.

    2017-06-01

    Full Text Available This work aims to study the stress corrosion crack growth behavior of cold-rolled Alloy 690 in the primary water of a pressurized water reactor. Compared with Alloy 600, which shows typical intergranular cracking along high angle grain boundaries, the cold-rolled Alloy 690, with its heterogeneous microstructure, revealed an abnormal crack growth behavior in mixed mode, that is, in transgranular cracking near a banded region, and in intergranular cracking in a matrix region. From local strain distribution analysis based on local mis-orientation, measured along the crack path using the electron back scattered diffraction method, it was suggested that the abnormal behavior was attributable to a heterogeneity of local strain distribution. In the cold-rolled Alloy 690, the stress corrosion crack grew through a highly strained area formed by a prior cold-rolling process in a direction perpendicular to the maximum principal stress applied during a subsequent stress corrosion cracking test.

  14. SCC of Alloy 600 components in PWR primary loop

    International Nuclear Information System (INIS)

    Gomez-Briceno, Dolores; Lapena, Jesus; Castano, M. Luisa; Blazquez, Fernando

    2002-01-01

    Full text: Cracking due to PWSCC in PWR CRDM nozzles and other VHP nozzles fabricated from Alloy 600 is not a new issue. In 1991, a leak was discovered on one CRDM nozzle at Bugey 3 PWR plant in France. The cause of the cracking was identified as primary water stress corrosion cracking. From then, similar cracks have been found in other European and USA PWR plants. The cracks were predominantly axial in orientation and it was accepted that CRDM nozzles and weld cracking in PWR was not a immediate safety concern. However, this consideration has to be reassessed in light of the recent identification of circumferential cracking in CRDM nozzles at Oconee Nuclear Station Unit 2 and 3 along with axial cracking in the Alloy 182 J-groove welds at these two units and at Oconee Nuclear Station 1 and Arkansas Nuclear One Unit 1. Alloy 600 susceptibility in primary water has received an enormous research effort for many years since the Alloy 600 steam generators tube degradation started. A significant amount of information is available to characterise the susceptibility of Alloy 600. However, Alloy 600 susceptibility is strongly dependent on the heat thermomechanical history and both the crack initiation time and the crack growth rate data obtained from representative materials of the VHP nozzles seem to be necessary for the structural integrity assessment of cracking nozzles. An extensive experimental program has been performed at CIEMAT, to study the behaviour of Alloy 600 VHP nozzles in PWR primary conditions. Crack initiation and crack propagation tests have been performed using different types of products (forged bar, tube, plate and steam generator tubing). Long duration crack initiation tests have been carried out, at 330 deg. C and 360 deg. C in water and at 400 deg. C in steam, using ten Alloy 600 heats with yield strength ranging from 291 MPa to 489 MPa. The influence of several parameters (grain boundary carbide distribution, grain size and yield strength) on crack

  15. Effect of Residual Strain on PWSCC Initiation of Alloy 690 in Primary Water of PWR

    International Nuclear Information System (INIS)

    Lee, Eun-Hee; Kim, Sung-Woo; Eom, Ki-Hyeon; Hwang, Seong-Sik

    2016-01-01

    The object of this study is to attempt to correlate the susceptibility to crack initiation of Alloy 690 with the levels of cold work in simulated primary water. Experiments were conducted in high-temperature water to accelerate the cracking behavior and to determine the severity of crack initiation as a function of level of cold work for Alloy 690. Cracking susceptibility was determined by measuring the crack length per unit area and crack density. SCC initiation susceptibility of Alloy 690 increased with increase of the residual strain induced by the prior cold-rolling process. The 20% cold-rolled specimen exhibited mostly IG cracks, while the 30% and 40% cold-rolled specimens revealed significant amounts of IG and TG cracks. Most cracks were observed near the smallest cross-section area of the tapered specimen, where the maximum stress was applied, indicating that the SCC initiation susceptibility was strongly dependent on the applied stress. In order to find the correlation between SCC initiation susceptibility and residual strain induced by cold work, more analyses need to be performed in terms of the crack length per unit area and the crack density, as a strong indicative for SCC initiation susceptibility

  16. Materials Reliability Program Resistance to Primary Water Stress Corrosion Cracking of Alloys 690, 52, and 152 in Pressurized Water Reactors (MRP-111)

    Energy Technology Data Exchange (ETDEWEB)

    Xu, H. [Framatome ANP, Inc., Lynchburg, VA (United States); Fyfitch, S. [Framatome ANP, Inc., Lynchburg, VA (United States); Scott, P. [Framatome ANP, SAS, Paris (France); Foucault, M. [Framatome ANP, SAS, Le Creusot (France); Kilian, R. [Framatome ANP, GmbH, Erlangen (Germany); Winters, M. [Framatome ANP, GmbH, Erlangen (Germany)

    2004-03-01

    Over the last thirty years, stress corrosion cracking in PWR primary water (PWSCC) has been observed in numerous Alloy 600 component items and associated welds, sometimes after relatively long incubation times. Repairs and replacements have generally utilized wrought Alloy 690 material and its compatible weld metals (Alloy 152 and Alloy 52), which have been shown to be very highly resistant to PWSCC in laboratory experiments and have been free from cracking in operating reactors over periods already up to nearly 15 years. It is nevertheless prudent for the PWR industry to attempt to quantify the longevity of these materials with respect to aging degradation by corrosion in order to provide a sound technical basis for the development of future inspection requirements for repaired or replaced component items. This document first reviews numerous laboratory tests, conducted over the last two decades, that were performed with wrought Alloy 690 and Alloy 52 or Alloy 152 weld materials under various test conditions pertinent to corrosion resistance in PWR environments. The main focus of the present review is on PWSCC, but secondary-side conditions are also briefly considered.

  17. Materials Reliability Program Resistance to Primary Water Stress Corrosion Cracking of Alloys 690, 52, and 152 in Pressurized Water Reactors (MRP-111)

    International Nuclear Information System (INIS)

    Xu, H.; Fyfitch, S.; Scott, P.; Foucault, M.; Kilian, R.; Winters, M.

    2004-01-01

    Over the last thirty years, stress corrosion cracking in PWR primary water (PWSCC) has been observed in numerous Alloy 600 component items and associated welds, sometimes after relatively long incubation times. Repairs and replacements have generally utilized wrought Alloy 690 material and its compatible weld metals (Alloy 152 and Alloy 52), which have been shown to be very highly resistant to PWSCC in laboratory experiments and have been free from cracking in operating reactors over periods already up to nearly 15 years. It is nevertheless prudent for the PWR industry to attempt to quantify the longevity of these materials with respect to aging degradation by corrosion in order to provide a sound technical basis for the development of future inspection requirements for repaired or replaced component items. This document first reviews numerous laboratory tests, conducted over the last two decades, that were performed with wrought Alloy 690 and Alloy 52 or Alloy 152 weld materials under various test conditions pertinent to corrosion resistance in PWR environments. The main focus of the present review is on PWSCC, but secondary-side conditions are also briefly considered

  18. Corrosion Control of Alloy 690 by Shot Peening and Electropolishing under Simulated Primary Water Condition of PWRs

    Directory of Open Access Journals (Sweden)

    Kyung Mo Kim

    2015-01-01

    Full Text Available This work clarifies the effect of surface modifications on the corrosion rate of Alloy 690, a nickel-based alloy for steam generator tubes, under the simulated test conditions of the primary water chemistry in nuclear power plants. The surface stress was modified by the shot peening and electropolishing methods. The shot peening treatment was applied using ceramic beads with different intensities by varying the air pressure and projection angle. The corrosion rate was evaluated by gravimetric analysis and the surface was analyzed by scanning electron microscopy (SEM. The corrosion rate of Alloy 690 was evaluated from the influence of the stress state on the metal surface. Based on the observation of the surface after the corrosion test, the oxide composition and its structure were affected by the surface modifications. The corrosion behavior of Alloy 690 was distinguished by the shot peening intensity on the surface, and additional electropolishing was effective at reducing the dissolution of nickel ions from the metal surface.

  19. Effect of Ni and Cr on IGSCC growth rate of Ni-Cr-Fe alloys in PWR primary water

    International Nuclear Information System (INIS)

    Arioka, K.; Yamada, T.; Aoki, M.; Miyamoto, T.

    2015-01-01

    The purpose of this research is to examine the dependence of SCC (Stress Corrosion Crack) growth on nickel and chromium in PWR primary water; the objective is to obtain the basic knowledge to understand SCC behavior of steam generator tubing materials. The second objective is to understand whether accelerated testing at higher temperatures is appropriate for predicting SCC initiation and growth at lower temperatures. For these objectives, SCC growth was measured in PWR primary water at 290, 320, 330, 340, and 360 C. degrees under static load conditions. Tests were performed using 0.5 T compact tension type specimen using 20%CW X%Ni-16%Cr-Fe alloys in the range of nickel concentration between 16 to 60% and laboratory melted nuclear grade 20% cold worked Alloy 800 (USN N08800, CW800NG). Four important patterns were observed. First, significant effect of nickel on IGSCC resistance was observed at 340 and 360 C. degrees. The rate of IGSCC growth decreases with increasing nickel concentration in the range of nickel concentration between 10% to 25% nickel; and then, the rate of IGSCC increases with increasing nickel concentration in the range of Ni content between 50% and 76%. This trend is quite similar to the results reported by Coriou and Staehle tested in deaerated pure water at 350 C. degrees. However, no significant dependence of Ni content on IGSCC in PWR water at 320 and 290 C. degrees was observed. The change in SCC growth dependence on nickel concentration suggested that the main rate limiting processes on IGSCC growth seems to change between 320 and 340 C. degrees. Secondly, significant beneficial effects of chromium in alloys were observed at 320 C. degrees. However, no beneficial effect of chromium addition in alloys was observed at 360 C. degrees. Thirdly, peak temperatures in growth rate of IGSCC were observed in almost all test materials except for 20%CW Alloy 600. Finally, intergranular attack was observed in some alloys at lower temperature, and the

  20. Influence of dissolved hydrogen and temperature on primary water stress corrosion cracking of mill annealed alloy 600

    Energy Technology Data Exchange (ETDEWEB)

    Totsuka, Nobuo; Nishikawa, Yoshito [Inst. of Nuclear Safety System Inc., Mihama, Fukui (Japan); Nakajima, Nobuo

    2002-09-01

    The influence of dissolved hydrogen and temperature on primary water stress corrosion cracking (PWSCC) of alloy 600 was experimentally studied at temperature ranging from 310 to 360degC and hydrogen contents ranging from 0 to 4 ppm using slow strain rate tensile technique (SSRT) and constant load tensile test. As a result, it was revealed that the PWSCC susceptibility of alloy 600 has a maximum near 3 ppm of dissolved hydrogen at 360degC and the peak shifts to 1 ppm at 320degC. The mechanism of the peak shift is not clear yet, however, it is possibly explained by the change of absorbed hydrogen in the metal caused by the change of hydrogen recombination reaction and/or change of the surface film. (author)

  1. Influence of dissolved hydrogen on oxide film and PWSCC of Alloy 600 in PWR primary water

    Energy Technology Data Exchange (ETDEWEB)

    Nakagawa, Tomokazu; Totsuka, Nobuo; Nakajima, Nobuo [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2001-09-01

    In order to investigate the influence of dissolved hydrogen (DH) on the corrosion behavior and PWSCC of Alloy 600 in primary water of PWR under actual operating temperature range, we carried out electrochemical polarization measurement, repassivation test, analysis of the oxide film on the alloy by AES, XPS and PWSCC test. In all cases, the content of DH was changed from 0 to 45 cc/kgH{sub 2}O. The anodic polarization curve reveals that the peak current density increases with increasing DH. The result of the repassivation test shows that the repassivation rate decreases with increasing DH, and the changes of the above two become larger between 11 and 22 cc/kgH{sub 2}O of DH. According to the results of oxide film analysis, it is seen that the oxide films formed below 11 cc/kgH{sub 2}O of DH are relatively thick and rich in Ni, but those formed at higher DH contents are relatively thin and rich in Cr and Fe. The susceptibility of the alloy to PWSCC has a peak at 11 cc/kgH{sub 2}O of DH, which reveals that the property of the oxide film may play important role in PWSCC of alloy. (author)

  2. Kinetics of corrosion products release from nickel-base alloys corroding in primary water conditions. A new modeling of release

    International Nuclear Information System (INIS)

    Carrette, F.; Guinard, L.; Pieraggi, B.

    2002-01-01

    The radioactivity in the primary circuit arises mainly from the activation of corrosion products in the core of pressurised water reactors; corrosion products dissolve from the oxide scales developed on steam generator tubes of alloy 690. The controlling and modelling of this process require a detailed knowledge of the microstructure and chemical composition of oxide scales as well as the kinetics of their corrosion and dissolution. Alloy 690 was studied as tubes and sheets, with three various surface states (as-received, cold-worked, electropolished). Corrosion tests were performed at 325 C and 155 bar in primary water conditions (B/Li - 1000/2 ppm, [H 2 ] 30 cm 3 .kg -1 TPN, [O 2 ] < 5 ppb); test durations ranged between 24 and 2160 hours. Corrosion tests in the TITANE loop provided mainly corrosion and oxidation kinetics, and tests in the BOREAL loop yielded release kinetics. This study revealed asymptotic type kinetics. Characterisation of the oxide scales grown in representative conditions of the primary circuit was performed by several techniques (SEM, TEM, SIMS, XPS, GIXRD). These analyses revealed the essential role of the fine grained cold-worked scale present on as-received and cold-worked materials. This scale controls the corrosion and release phenomena. The kinetic study and the characterisation of the oxide scales contributed to the modelling of the corrosion/release process. A growth/dissolution model was proposed for corrosion product scales grown in non-saturated dynamic fluid. This model provided the temporal evolution of oxide scales and release kinetics for different species (Fe, Ni, Cr). The model was validated for several surface states and several alloys. (authors)

  3. The effect of weld chemistry on the likely primary water stress corrosion cracking susceptibility of alloy 82 dissimilar metal welds

    Energy Technology Data Exchange (ETDEWEB)

    Persaud, S.Y., E-mail: suraj.persaud@mail.utoronto.ca [University of Toronto, Toronto, ON (Canada); Ramamurthy, S. [Surface Science Western, London, ON (Canada); Newman, R.C. [University of Toronto, Toronto, ON (Canada)

    2015-07-01

    Alloy 82 dissimilar weld joints between carbon steel and Alloy 600 were exposed to a simulated primary water environment consisting of hydrogenated steam at 480 {sup o}C and 1 bar. Dilution from the carbon steel to the weld was significant, particularly in the root where Fe was enriched to 35 at. % and Cr was depleted to 10 at. %. The heterogeneous composition of the weld from root to crown resulted in differences in internal and external oxidation tendency. An Fe-rich external surface oxide formed on the weld root which may help to prevent embrittlement and SCC by internal intergranular oxidation. (author)

  4. Stress corrosion cracking of iron-nickel-chromium alloys in primary circuit environment of PWR-type reactors

    International Nuclear Information System (INIS)

    Boursier, Jean-Marie

    1993-01-01

    Stress corrosion cracking of Alloy 600 steam generator tubing is a great concern for pressurized water reactors. The mechanism that controls intergranular stress corrosion cracking of Alloy 600 in primary water (lithiated-borated water) has yet to be clearly identified. A study of stress corrosion cracking behaviour, which can identify the main parameters that control the cracking phenomenon, was so necessary to understand the stress corrosion cracking process. Constant extension rate tests, and constant load tests have evidenced that Alloy 600 stress corrosion cracking involves firstly an initiation period, then a slow propagation stage with crack less than 50 to 80 micrometers, and finally a rapid propagation stage leading to failure. The influence of mechanical parameters have shown the next points: - superficial strain hardening and cold work have a strong effect of stress corrosion cracking resistance (decrease of initiation time and increase of crack growth rate), - strain rate was the most suitable parameter for describing the different stage of propagation. The creep behaviour of alloy 600 has shown an increase of creep rate in primary water compared to air, which implies a local interaction plasticity/corrosion. An assessment of the durations of the initiation and the propagation stages was attempted for the whole uniaxial tensile tests, using the macroscopic strain rate: - the initiation time is less than 100 hours and seems to be an electrochemical process, - the durations of the propagation stage are strongly dependent on the strain rate. The behaviour in high primary water temperature of Alloys 690 and 800, which replace Alloy 600, was studied to appraise their margin, and validate their choice. Then the last chapter has to objective to evaluate the crack tip strain rate, in order to better describe the evolution of the different stages of cracking. (author) [fr

  5. Effects of segregation of primary alloying elements on the creep response in magnesium alloys

    DEFF Research Database (Denmark)

    Huang, Y.D.; Dieringa, H.; Hort, N.

    2008-01-01

    The segregation of primary alloying elements deteriorates the high temperature creep resistance of magnesium alloys. Annealing at high temperatures alleviating their segregations can improve the creep resistance. Present investigation on the effect of segregation of primary alloying elements...... on the creep response may provide some useful information about how to improve the creep resistance of magnesium alloys in the future. (c) 2008 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved....

  6. Stress corrosion cracking growth rate of TT alloy 690 and its weld joint in simulated PWR primary water

    International Nuclear Information System (INIS)

    Yonezawa, T.

    2015-01-01

    Recently, some researchers reported that the SCC growth rate (SCCGR) of cold worked thermally treated (TT) Alloy 690 was significantly different in heat by heat. But, author has hypothesized that these high SCCGRs in cold worked TT Alloy 690 could be due to the metallurgical characteristics of these heats. In order to confirm this hypothesis, this study has been started in the author's laboratory, and the following 4 new evidences were obtained. First, microcracks of carbides and voids were observed in eutectic M 23 C 6 GB carbides (primary carbides) for cold rolled laboratory heat after as cast or lightly forged condition or for chemical composition simulated Bettis'TT Alloy 690 heat, after cold rolling, before SCC test. However, microcracks in primary carbides along grain boundaries and voids were rarely detected in the cold rolled commercial heat of TT Alloy 690 used for CRDM penetrations. Secondly, the SCCGR observed in TT Alloy 690 was different in each hot working process and each heat. Comparing the SCCGRs for all heats of cold worked TT Alloy 690, the SCCGR decreased with increasing of Vickers hardness. However, in same heats of cold worked TT Alloy 690, the SCCGR increased with increasing of Vickers hardness. Thirdly, the SCCGR in cold rolled TT Alloy 690 should be integrated by the effect of hardness or cold working ratio and by the effect of existing ratio of primary M23C6 carbides with cracks and Voids due to chemical composition and the fabrication process of TT Alloy 690. Fourthly, it is argued that the high SCCGRs in highly cold rolled TT Alloy 690 are not representative of the practical situation with TT Alloy 690 in service for CRDM adapter nozzles etc. The high SCCGR of highly cold rolled TT Alloy 690 is not thought to be an accurate tool in predicting the possibility of cracking of TT Alloy 690 for CRDM adapter nozzles. (author)

  7. Visualization of direct contact heat transfer between water and molten alloy

    International Nuclear Information System (INIS)

    Nishi, Yoshihisa; Furuya, Masahiro; Kinoshita, Izumi; Takenaka, Nobuyuki; Matsubayashi, Masahito.

    1996-01-01

    We have been developing an innovative Steam Generator concept of Fast Breeder Reactors by using liquid-liquid direct contact heat transfer. In this concept, the SG shell is filled with a molten alloys, which is heated by primary sodium. Water is fed into the high temperature molten alloy, and evaporates by direct contact heating. In order to obtain the fundamental information to discuss the heat transfer mechanisms of the direct contact between the water and the alloy, this phenomenon was visualized by real-time neutron radiography. JRR-3M real-time thermal neutron radiography in Japan Atomic Energy Research Institute was used. Followings are main results. (1) The vigorous evaporation occurs in the molten alloy. This phenomena is different from the known phenomenon such as the evaporation of refrigerant R-113 in the water. (2) The evaporation in the bubble has finished in a moment due to high heat transfer performance between the liquid and molten alloy. (3) It is confirmed that the velocity of bubble with the rapid evaporation and growth is about 50 cm/s. (author)

  8. Methodologies to assess PWSCC susceptibility of primary component Alloy 600 locations in pressurized water reactors

    International Nuclear Information System (INIS)

    Rao, G.V.

    1993-01-01

    Methodologies to assess susceptibility to Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 600 component locations in the Primary System of Pressurized Water Reactors are presented. The assessment methodologies are presented. The assessment methodologies are based on Relative Susceptibility Index (RSI) and Cumulative Susceptibility Index (CSI) models utilizing key contributing parameters such as service and residual stresses, yield strength, service temperature, material condition and microstructure, and the accumulated service time. To aid in the development of future inspection plans, a method of ranking of the assessed susceptibilities by 'bench marking' with respect to the susceptibility of a reference location of known PWSCC history of a reference location of known PWSCC history is presented. Means of utilizing the susceptibility ranking results in developing a prioritized inspection plan are discussed. A follow-up investigative plan to the initial inspection is proposed, which includes identification of critical sampling locations, sample extraction, sample investigations and testing to ensure that the potentially highest susceptibility locations are free from near term PWSCC and, further, to provide a basis for established schedules for future inspections. Finally, parametric considerations of the contributing factor are presented to help the utility choose suitable option to mitigate the PWSCC issue while minimizing the impact on continued service

  9. Modeling of primary water stress corrosion cracking at control rod drive mechanism nozzles of pressurized water reactors

    International Nuclear Information System (INIS)

    Aly, Omar Fernandes

    2006-01-01

    One of the main failure mechanisms that cause risks to pressurized water reactors is the primary water stress corrosion cracking (PWSCC) occurring in alloys. It can occurs, besides another places, at the control reactor displacement mechanism nozzles. It is caused by the joint effect of tensile stress, temperature, susceptible metallurgical microstructure and environmental conditions of the primary water. These cracks can cause accidents that reduce nuclear safety by blocking the rod's displacement and may cause leakage of primary water, reducing the reactor's life. In this work it is proposed a study of the existing models and a modeling proposal to primary water stress corrosion cracking in these nozzles in a nickel based Alloy 600. It is been superposed electrochemical and fracture mechanics models, and validated using experimental and literature data. The experimental data were obtained at CDTN-Brazilian Nuclear Technology Development Center, in a recent installed slow strain rate testing equipment. In the literature it is found a diagram that indicates a thermodynamic condition for the occurrence of some PWSCC sub modes in Alloy 600: it was used potential x pH diagrams (Pourbaix diagrams), for Alloy 600 in high temperature primary water (300 deg C till 350 deg C). Over it, were located the PWSCC sub modes, using experimental data. It was added a third parameter called 'stress corrosion strength fraction'. However, it is possible to superpose to this diagram, other parameters expressing PWSCC initiation or growth kinetics from other models. Here is the proposition of the original contribution of this work: from an original experimental condition of potential versus pH, it was superposed, an empiric-comparative, a semi-empiric-probabilistic, an initiation time, and a strain rate damage models, to quantify respectively the PWSCC susceptibility, the failure time, and in the two lasts, the initiation time of stress corrosion cracking. It was modeling from our

  10. Effect of surface stress state on dissolution property of Alloy 690 in simulated primary water condition

    International Nuclear Information System (INIS)

    Kim, Kyung Mo; Shim, Hee-Sang; Lee, Eun Hee; Seo, Myung Ji; Han, Jung Ho; Hur, Do Haeng

    2014-01-01

    The dissolution control of nickel is important to reduce the radioactive dose rate and deterioration of fuel performance in the operation of nuclear power plants (PWR). The corrosion properties are affected by the metal surface residual stress introduced in manufacture process such as work hardening. This work studied the effect of surface modification on the release rate of Alloy 690, nickel-base alloy for a steam generator tube, in the test condition of simulated primary water chemistry in PWRs. The surface stress modification was applied by the electro-polishing and shot peening method. Shot peening process was applied using ceramic beads with different intensities through the variation of air pressure. The corrosion release tests performed at 330degC with LiOH 2 ppm and H 3 BO 4 1200 ppm, DH(dissolved hydrogen) 35 cc/kg (STP) and about 20 ppb of DO(dissolved oxygen) condition. The corrosion release rate was evaluated by a gravimetric analysis method and the surface analysed by SEM and optical microscope. The surface residual stress was measured by an X-ray diffractometer, and the distribution of stress state was evaluated by a micro-hardness tester. The metal ion release rate of alloy 690 was evaluated from the influence of the stress state on the metal surface. The oxide property and structure was affected by the residual stress in the oxide layer. (author)

  11. In situ Raman Spectroscopy of Oxide Films on Zirconium Alloy in Simulated PWR Primary Water Condition

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Ho; Choi, Kyoung Joon; Yoo, Seung Chang; Kim, Ji Hyun [UNIST, Ulsan (Korea, Republic of)

    2016-05-15

    The two layered oxide structure is formed in pre-transition oxide for the zirconium alloy in high temperature water environment. It is known that the corrosion rate is related to the volume fraction of zirconium oxide and the pores in the oxides; therefore, the aim of this paper is to investigate the oxidation behavior in the pretransition zirconium oxide in high-temperature water chemistry. In this work, Raman spectroscopy was used for in situ investigations for characterizing the phase of zirconium oxide. In situ Raman spectroscopy is a well-suited technique for investigating in detail the characteristics of oxide films in a high-temperature corrosion environment. In previous studies, an in situ Raman system was developed for investigating the oxides on nickel-based alloys and low alloy steels in high-temperature water environment. Also, the early stage oxidation behavior of zirconium alloy with different dissolved hydrogen concentration environments in high temperature water was treated in the authors' previous study. In this study, a specific zirconium alloy was oxidized and investigated with in situ Raman spectroscopy for 100 d oxidation, which is close to the first transition time of the zirconium alloy oxidation. The ex situ investigation methods such as transmission electron microscopy (TEM) and energy dispersive X-ray spectroscopy (EDS) were used to further characterize the zirconium oxide structure. As oxidation time increased, the Raman peaks of tetragonal zirconium oxide were merged or became weaker. However, the monoclinic zirconium oxide peaks became distinct. The tetragonal zirconium oxide was just found near the O/M interface and this could explain the Raman spectra difference between the 30 d result and others.

  12. Stress corrosion crack initiation of alloy 182 weld metal in primary coolant - Influence of chemical composition

    Energy Technology Data Exchange (ETDEWEB)

    Calonne, O.; Foucault, M.; Steltzlen, F. [AREVA (France); Amzallag, C. [EDF SEPTEN (France)

    2011-07-01

    Nickel-base alloys 182 and 82 have been used extensively for dissimilar metal welds. Typical applications are the J-groove welds of alloy 600 vessel head penetrations, pressurizer penetrations, heater sleeves and bottom mounted instrumented nozzles as well as some safe end butt welds. While the overall performance of these weld metals has been good, during the last decade, an increasing number of cases of stress corrosion cracking of Alloy 182 weld metal have been reported in PWRs. In this context, the role of weld defects has to be examined. Their contribution in the crack initiation mechanism requires laboratory investigations with small scale characterizations. In this study, the influence of both alloy composition and weld defects on PWSCC (Stress Corrosion Cracking in Primary Water) initiation was investigated using U-bend specimens in simulated primary water at 320 C. The main results are the following: -) the chemical compositions of the weld deposits leading to a large propensity to hot cracking are not the most susceptible to PWSCC initiation, -) macroscopically, superficial defects did not evolve during successive exposures. They can be included in large corrosion cracks but their role as 'precursors' is not yet established. (authors)

  13. Hard alloys testing-machine for values of PWR primary coolant circuits

    International Nuclear Information System (INIS)

    Campan, J.L.; Sauze, A.

    1980-01-01

    Testing of valve parts or material used in valve fabrication and particularly seizing conditions in friction of plane surfaces coated with hard alloys of the type stellite. The testing equipment called Marguerite is composed of a hot pressurized water loop in conditions similar to PWR primary coolant circuits (320 0 C, 150 bars) and a testing-machine with measuring instruments. Testing conditions and samples are described [fr

  14. Crack stability analysis of low alloy steel primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T.; Kameyama, M. [Kansai Electric Power Company, Osaka (Japan); Urabe, Y. [Mitsubishi Heavy Industries, Ltd., Takasago (Japan)] [and others

    1997-04-01

    At present, cast duplex stainless steel has been used for the primary coolant piping of PWRs in Japan and joints of dissimilar material have been applied for welding to reactor vessels and steam generators. For the primary coolant piping of the next APWR plants, application of low alloy steel that results in designing main loops with the same material is being studied. It means that there is no need to weld low alloy steel with stainless steel and that makes it possible to reduce the welding length. Attenuation of Ultra Sonic Wave Intensity is lower for low alloy steel than for stainless steel and they have advantageous inspection characteristics. In addition to that, the thermal expansion rate is smaller for low alloy steel than for stainless steel. In consideration of the above features of low alloy steel, the overall reliability of primary coolant piping is expected to be improved. Therefore, for the evaluation of crack stability of low alloy steel piping to be applied for primary loops, elastic-plastic future mechanics analysis was performed by means of a three-dimensioned FEM. The evaluation results for the low alloy steel pipings show that cracks will not grow into unstable fractures under maximum design load conditions, even when such a circumferential crack is assumed to be 6 times the size of the wall thickness.

  15. Precursor Evolution and Stress Corrosion Cracking Initiation of Cold-Worked Alloy 690 in Simulated Pressurized Water Reactor Primary Water

    Energy Technology Data Exchange (ETDEWEB)

    Zhai, Ziqing [Pacific Northwest National Laboratory, 622 Horn Rapids Road, P.O. Box 999, Richland, Washington 99352.; Toloczko, Mychailo [Pacific Northwest National Laboratory, 622 Horn Rapids Road, P.O. Box 999, Richland, Washington 99352.; Kruska, Karen [Pacific Northwest National Laboratory, 622 Horn Rapids Road, P.O. Box 999, Richland, Washington 99352.; Bruemmer, Stephen [Pacific Northwest National Laboratory, 622 Horn Rapids Road, P.O. Box 999, Richland, Washington 99352.

    2017-05-22

    Stress corrosion crack initiation of two thermally-treated, cold-worked (CW) alloy 690 (UNS N06690) materials was investigated in 360oC simulated PWR primary water using constant load tensile (CLT) tests and blunt notch compact tension (BNCT) tests equipped with direct current potential drop (DCPD) for in-situ detection of cracking. SCC initiation was not detected by DCPD for either the 21% or 31%CW CLT specimens loaded at their yield stress after ~9,220 hours, however intergranular (IG) precursor damage and isolated surface cracks were observed on the specimens. The two 31%CW BNCT specimens loaded at moderate stress intensity after several cyclic loading ramps showed DCPD-indicated crack initiation after 10,400 hours of exposure at constant stress intensity, which was resulted from significant growth of IG cracks. The 21%CW BNCT specimens only exhibited isolated small IG surface cracks and showed no apparent DCPD change throughout the test. Post-test cross-section examinations revealed many grain boundary (GB) nano-cavities in the bulk of all the CLT and BNCT specimens particularly for the 31%CW materials. Cavities were also found along GBs extending to the surface suggesting an important role in crack nucleation. This paper provides an overview of the evolution of GB cavities and discusses their effects on crack initiation in CW alloy 690.

  16. In situ Investigation of Oxide Films on Zirconium Alloy in PWR Primary Water Chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Taeho; Choi, Kyoung Joon; Yoo, Seung Chang; Kim, Ji Hyun [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2015-05-15

    Zirconium alloys are used as fuel cladding materials in nuclear power reactors, because these materials have a very low thermal neutron capture cross section as well as desirable mechanical properties. However, the Fukushima accident shows that the oxidation behavior of zirconium alloy is an important issue because the zirconium alloy functions as a shield of nuclear material (i.e., uranium, fission gas), and the degradation on zirconium cladding directly causes severe accident on nuclear power plant. Therefore, to ensure the safety of nuclear power reactors, the performance and sustainability of nuclear fuel should be understood. Currently, the water-metal interface is regarded as the rate-controlling site governing the rapid oxidation transition in high-burn-up fuels. Zirconium oxide is formed at the water-metal interface, and its structure and phase play an important role in determining its mechanical properties. In the early stage of the oxidation process, zirconium oxide with both tetragonal and monoclinic phases is formed. With an increase in the oxidation time to 150 h, the unstable tetragonal phase disappears and the monoclinic phase is dominant and possibly because of the stress relaxation according to previous and present results.

  17. Precursor evolution and SCC initiation of cold-worked alloy 690 in simulated PWR primary water

    Energy Technology Data Exchange (ETDEWEB)

    Zhai, Ziqing; Kruska, Karen; Toloczko, Mychailo B.; Bruemmer, Stephen M.

    2017-03-27

    Stress corrosion crack initiation of two thermally-treated, cold-worked (CW) alloy 690 materials was investigated in 360oC simulated PWR primary water using constant load tensile (CLT) tests and blunt notch compact tension (BNCT) tests equipped with direct current potential drop (DCPD) for in-situ detection of cracking. SCC initiation was not detected by DCPD for the 21% and 31%CW CLT specimens loaded at their yield stress after ~9,220 h, however intergranular (IG) precursor damage and isolated surface cracks were observed on the specimens. The two 31%CW BNCT specimens loaded at moderate stress intensity after several cyclic loading ramps showed DCPD-indicated crack initiation after 10,400h exposure at constant stress intensity, which resulted from significant growth of IG cracks. The 21%CW BNCT specimens only exhibited isolated small IG surface cracks and showed no apparent DCPD change throughout the test. Interestingly, post-test cross-section examinations revealed many grain boundary (GB) nano-cavities in the bulk of all the CLT and BNCT specimens particularly for the 31%CW materials. Cavities were also found along GBs extending to the surface suggesting an important role in crack nucleation. This paper provides an overview of the evolution of GB cavities and will discuss their effects on crack initiation in CW alloy 690.

  18. Visualization of direct contact heat transfer between water and molten alloy by neutron radiography. 1

    International Nuclear Information System (INIS)

    Nishi, Yoshihisa; Furuya, Masahiro; Kinoshita, Izumi; Takenaka, Nobuyuki; Matsubayashi, Masahito.

    1997-01-01

    Design of an innovative Steam Generator (SG) for Liquid Metal Fast Reactors (LMFRs) using liquid-liquid direct contact heat transfer has been developing. In this concept, the SG shell is filled with a molten alloy, which is heated by primary sodium. Water is fed into the high-temperature, molten alloy, and evaporates by direct contact heating. In order to obtain the fundamental information needed to discuss the heat transfer mechanisms of direct contact between the water and molten alloy, this phenomenon was observed by neutron radiography. JRR-3M thermal neutron radiography at the Japan Atomic Energy Research Institute was used. This paper deals with the results of visualization of direct contact heat exchange in the molten alloy. (author)

  19. Separation of primary solid phases from Al-Si alloy melts

    Directory of Open Access Journals (Sweden)

    Ki Young Kim

    2014-07-01

    Full Text Available The iron-rich solids formed during solidification of Al-Si alloys which are known to be detrimental to the mechanical, physical and chemical properties of the alloys should be removed. On the other hand, Al-Si hypereutectic alloys are used to extract the pure primary silicon which is suitable for photovoltaic cells in the solvent refining process. One of the important issues in iron removal and in solvent refining is the effective separation of the crystallized solids from the Al-Si alloy melts. This paper describes the separation methods of the primary solids from Al-Si alloy melts such as sedimentation, draining, filtration, electromagnetic separation and centrifugal separation, focused on the iron removal and on the separation of silicon in the solvent refining process.

  20. Crack growth rates in thick materials of alloy 600 and weld metals of alloy 182 in laboratory primary water comparison with field Experience

    Energy Technology Data Exchange (ETDEWEB)

    Vaillant, F.; Moulart, P.; Boursier, J.M. [Electricite de France (EDF), 75 - Paris (France). Region d' Equipement; Amzallag, C. [Electricite de France (EDF), DIS/SEPTEN, 75 - Paris (France); Daret, J. [CEA Saclay, Dept. de Physico-Chimie DPC/SCCME, 91 - Gif sur Yvette (France)

    2002-07-01

    Since 1991, when a first leakage occurred on the vessel head of Bugey 3 RPV, an important investigation program was undertaken in laboratory in order to assess crack growth rates (CGRs) of vessel head penetrations (VHPs) in alloy 600 and weld metal in alloy 182 in primary environments. SCC (stress corrosion cracking) tests were performed between 290 C and 360 C on pre-cracked specimens under static loading. Alloy 600: On VHPs with YS{sub 20} ranging from 300 MPa to 468 MPa, it was found that the upper bound for CGRs were dependant on (K(T initial)-K(iscc)){sup 0.3}, in accordance with field experience. In laboratory condition, the activation energy was 130 {+-} 20 kJ/mol, the yield stress increased significantly CGRs but some coupling effects were noted with the microstructure. Cold work increased slightly CGRs on a VHP with initial YS = 468 MPa. Additional tests were performed at 290 C and 325 C on rolled bars, rolled plates and forged plates representative of the other components in alloy 600 of the primary circuit: products with low YS and high GBC had low sensitivity to SCC but it could be significantly increased with cold work raising at the level of 468 MPa, the highest YS investigated on VHPs. Stress relief treatment did not significantly modify SCC resistance. On ten products from the various components, the measured CGRs were strongly correlated to the material susceptibility index for SCC initiation. Alloy 182: Some comparisons were performed in laboratory, with different orientations. Similar trends to alloy 600 were found for the influences of K and temperature on CGRs. 10% cold work increased and stress relief treatment decreased CGRs by a factor 2. CGRs of cracks propagating in the direction of dendrites were 2 to 5 times higher than for cracks propagating in the perpendicular direction. For both alloys 600 and 182, a model is proposed to account for the effects of the main parameters on CGRs and the relevance to field experience is discussed

  1. Crack growth rates in thick materials of alloy 600 and weld metals of alloy 182 in laboratory primary water comparison with field Experience

    International Nuclear Information System (INIS)

    Vaillant, F.; Moulart, P.; Boursier, J.M.; Daret, J.

    2002-01-01

    Since 1991, when a first leakage occurred on the vessel head of Bugey 3 RPV, an important investigation program was undertaken in laboratory in order to assess crack growth rates (CGRs) of vessel head penetrations (VHPs) in alloy 600 and weld metal in alloy 182 in primary environments. SCC (stress corrosion cracking) tests were performed between 290 C and 360 C on pre-cracked specimens under static loading. Alloy 600: On VHPs with YS 20 ranging from 300 MPa to 468 MPa, it was found that the upper bound for CGRs were dependant on (K(T initial)-K(iscc)) 0.3 , in accordance with field experience. In laboratory condition, the activation energy was 130 ± 20 kJ/mol, the yield stress increased significantly CGRs but some coupling effects were noted with the microstructure. Cold work increased slightly CGRs on a VHP with initial YS = 468 MPa. Additional tests were performed at 290 C and 325 C on rolled bars, rolled plates and forged plates representative of the other components in alloy 600 of the primary circuit: products with low YS and high GBC had low sensitivity to SCC but it could be significantly increased with cold work raising at the level of 468 MPa, the highest YS investigated on VHPs. Stress relief treatment did not significantly modify SCC resistance. On ten products from the various components, the measured CGRs were strongly correlated to the material susceptibility index for SCC initiation. Alloy 182: Some comparisons were performed in laboratory, with different orientations. Similar trends to alloy 600 were found for the influences of K and temperature on CGRs. 10% cold work increased and stress relief treatment decreased CGRs by a factor 2. CGRs of cracks propagating in the direction of dendrites were 2 to 5 times higher than for cracks propagating in the perpendicular direction. For both alloys 600 and 182, a model is proposed to account for the effects of the main parameters on CGRs and the relevance to field experience is discussed. (authors)

  2. PWSCC Growth Assessment Model Considering Stress Triaxiality Factor for Primary Alloy 600 Components

    Directory of Open Access Journals (Sweden)

    Jong-Sung Kim

    2016-08-01

    Full Text Available We propose a primary water stress corrosion cracking (PWSCC initiation model of Alloy 600 that considers the stress triaxiality factor to apply to finite element analysis. We investigated the correlation between stress triaxiality effects and PWSCC growth behavior in cold-worked Alloy 600 stream generator tubes, and identified an additional stress triaxiality factor that can be added to Garud's PWSCC initiation model. By applying the proposed PWSCC initiation model considering the stress triaxiality factor, PWSCC growth simulations based on the macroscopic phenomenological damage mechanics approach were carried out on the PWSCC growth tests of various cold-worked Alloy 600 steam generator tubes and compact tension specimens. As a result, PWSCC growth behavior results from the finite element prediction are in good agreement with the experimental results.

  3. A detailed TEM and SEM study of Ni-base alloys oxide scales formed in primary conditions of pressurized water reactor

    International Nuclear Information System (INIS)

    Sennour, Mohamed; Marchetti, Loic; Martin, Frantz; Perrin, Stephane; Molins, Regine; Pijolat, Michele

    2010-01-01

    The oxide film formed on nickel-based alloys in pressurized water reactors (PWR) primary coolant conditions (325 o C, aqueous media) is very thin, in the range of 1-100 nm thick, depending on the surface state and on the corrosion test duration. The nature and the structure of this scale have been investigated by Transmission Electron Microscopy (TEM) and Scanning Electron Microscopy (SEM). TEM observations revealed an oxide layer divided in two parts. The internal layer was mainly composed of a continuous spinel layer, identified as a mixed iron and nickel chromite (Ni (1-x) Fe x Cr 2 O 4 ). Moreover, nodules of Cr 2 O 3 , with a size about 5 nm, were present at the interface between this spinel and the alloy. No chromium depletion was observed in the alloy, at the alloy/oxide interface. The external layer is composed of large crystallites corresponding to a spinel structure rich in iron (Ni (1-z) Fe (2+z) O 4 ) resulting from precipitation phenomena. SEM and TEM observations showed a link between the nucleation and/or the growth of crystallites of nickel ferrite and the crystallographic orientation of the substrate. A link between the presence of surface defects and the nucleation of the crystallites was also underlined by SEM observations. Partially hydrated nickel hydroxide, was also observed by TEM in the external scale. Based on these results, some considerations about the mechanism of formation of this oxide layer are discussed.

  4. Structural analysis of surface film on alloy 600 formed under environment of PWR primary water

    Energy Technology Data Exchange (ETDEWEB)

    Terachi, Takumi; Totsuka, Nobuo; Yamada, Takuyo; Nakagawa, Tomokazu [Inst. of Nuclear Safety System Inc., Mihama, Fukui (Japan); Deguchi, Hiroshi [Kansai Electric Power Co., Inc., Osaka (Japan); Horiuchi, Masaki; Oshitani, Masato [Kanden Kako Co., Ltd., Osaka (Japan)

    2002-09-01

    It has been shown by one of the present authors and so forth that PWSCC of alloy 600 relates to dissolved hydrogen concentration (DH) in water and oxide film structure. However, the mechanism of PWSCC has not been clear yet. Therefore, in order to investigate relationship between them, structural analysis of the oxide film formed under the environment of PWR primary water was carried out by using X-ray diffraction, the scanning electron microscope and the transmission electron microscope. Especially, to perform accurate analysis, the synchrotron orbital radiation with SPring-8 was tried to use for thin film X-ray diffraction measurement. From the results, observed are as follows: 1. the oxide film is mainly composed of NiO, under the condition without hydrogen. 2. In the environment of DH 2.75ppm, the oxide film forms thin spinel structures. 3. On the other hand, needlelike oxides are formed at DH 1ppm. For this reason, around 1ppm of DH there would be the boundary that stable NiO and spinel oxide generate, and it agrees with the peak range of the PWSCC susceptibility on hydrogen. From this, it is suggested that the boundary of NiO/spinel oxide affects the SCC susceptibility. (author)

  5. Evolution of Primary Fe-Rich Compounds in Secondary Al-Si-Cu Alloys

    Science.gov (United States)

    Fabrizi, Alberto; Capuzzi, Stefano; Timelli, Giulio

    Although iron is usually added in die cast Al-Si foundry alloys to prevent die soldering, primary Fe-rich particles are generally considered as "hardspot" inclusions which compromise the mechanical properties of the alloy, namely ductility and toughness. As there is no economical methods to remove the Fe excess in secondary Al-Si alloys at this time, the control of solidification process and chemical composition of the alloy is a common industrial practice to overcome the negative effects connected with the presence of Fe-rich particles. In this work, the size and morphology as well as the nucleation density of primary Fe-rich particles have been studied as function of cooling rate and alloy chemical composition for secondary Al-Si-Cu alloys. The solidification experiments were carried out using differential scanning calorimetry whereas morphology investigations were conducted using optical and scanning electron microscopy. Mcrosegregations and chemical composition of primary Fe-rich particles were examined by energy dispersive spectroscopy.

  6. Metallurgical and mechanical parameters controlling alloy 718 stress corrosion cracking resistance in PWR primary water; Facteurs metallurgiques et mecaniques controlant l'amorcage de defauts de corrosion sous contrainte dans l'alliage 718 en milieu primaire des reacteurs a eau sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Deleume, J

    2007-11-15

    Improving the performance and reliability of the fuel assemblies of the pressurized water reactors requires having a perfect knowledge of the operating margins of both the components and the materials. The choice of alloy 718 as reference material for this study is justified by the industrial will to identify the first order parameters controlling the excellent resistance of this alloy to Stress Corrosion Cracking (SCC). For this purpose, a specific slow strain rate (SSR) crack initiation test using tensile specimen with a V-shaped hump in the middle of the gauge length was developed and modeled. The selectivity of such SSR tests in simulated PWR primary water at 350 C was clearly established by characterizing the SCC resistance of nine alloy 718 thin strip heats. Regardless of their origin and in spite of a similar thermo-mechanical history, they did not exhibit the same susceptibility to SCC crack initiation. All the characterized alloy 718 heats develop oxide scale of similar nature for various exposure times to PWR primary medium in the temperature range [320 C - 360 C]. {delta} phase precipitation has no impact on alloy 718 SCC initiation behavior when exposed to PWR primary water, contrary to interstitial contents and the triggering of plastic instabilities (PLC phenomenon). (author)

  7. Replacement of Co-base alloy for radiation exposure reduction in the primary system of PWR

    International Nuclear Information System (INIS)

    Han, Jeong Ho; Nyo, Kye Ho; Lee, Deok Hyun; Lim, Deok Jae; Ahn, Jin Keun; Kim, Sun Jin

    1996-01-01

    Of numerous Co-free alloys developed to replace Co-base stellite used in valve hardfacing material, two iron-base alloys of Armacor M and Tristelle 5183 and one nickel-base alloy of Nucalloy 488 were selected as candidate Co-free alloys, and Stellite 6 was also selected as a standard hardfacing material. These four alloys were welded on 316SS substrate using TIG welding method. The first corrosion test loop of KAERI simulating the water chemistry and operation condition of the primary system of PWR was designed and fabricated. Corrosion behaviors of the above four kinds of alloys were evaluated using this test loop under the condition of 300 deg C, 1500 psi. Microstructures of weldment of these alloys were observed to identify both matrix and secondary phase in each weldment. Hardnesses of weld deposit layer including HAZ and substrate were measured using micro-Vickers hardness tester. The status on the technology of Co-base alloy replacement in valve components was reviewed with respect to the classification of valves to be replaced, the development of Co-free alloys, the application of Co-free alloys and its experiences in foreign NPPs, and the Co reduction program in domestic NPPs and industries. 18 tabs., 20 figs., 22 refs. (Author)

  8. Water chemistry and corrosion control of cladding and primary circuit components. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1999-12-01

    Corrosion is the principal life limiting degradation mechanism in nuclear steam supply systems, especially taking into account the trends to increase fuel burnup, thermal rate and cycle length. Primary circuit components of water cooled power reactors have an impact on Zr-based alloys behaviour due to crud (primary circuit corrosion products) formation, transport and deposition on heat transfer surfaces. Crud deposits influence water chemistry, radiation and thermal hydraulic conditions near cladding surface, and by this way-Zr-based alloy corrosion. During the last decade, significant improvements were achieved in the reduction of the corrosion and dose rates by changing the cladding material for one more resistant to corrosion or by the improvement of water chemistry conditions. However, taking into account the above mentioned tendency for heavier fuel duties, corrosion and water chemistry, control will remain a serious task to work with for nuclear power plant operators and scientists, as well as development of generally accepted corrosion model of Zr-based alloys in a water environment in a new millennium. Upon the recommendation of the International Working Group on Water Reactor Fuel Performance and Technology, water chemistry and corrosion of cladding and primary circuit components are in the focus of the IAEA activities in the area of fuel technology and performance. At present the IAEA performs two co-ordinated research projects (CRPs): on On-line High Temperature Monitoring of Water Chemistry and Corrosion (WACOL) and on Activity Transport in Primary Circuits. Two CRPs deal with hydrogen and hydride degradation of the Zr-based alloys. A state-of-the-art review entitled: 'Waterside Corrosion of Zirconium Alloys in Nuclear Power Plants' was published in 1998. Technical Committee meetings on the subject were held in 1985 (Cadarache, France), 1989 (Portland, USA), 1993 (Rez, Czech Republic). During the last few years extensive exchange of experience in

  9. Effects of metallurgical factors on stress corrosion cracking of Ni-base alloys in high temperature water

    International Nuclear Information System (INIS)

    Yonezawa, T.; Sasaguri, N.; Onimura, K.

    1988-01-01

    Nickel-base Alloy 600 is the principal material used for the steam generator tubes of PWRs. Generally, this alloy has been proven to be satisfactory for this application, however when it is subjected to extremely high stress level in PWR primary water, it may suffer from stress corrosion cracking. The authors have systematically studied the effects of test temperature and such metallurgical factors as cold working, chemical composition and heat treatment on the stress corrosion cracking of Alloy 600 in high temperature water, and also on that of Alloy 690 which is a promising material for the tubes and may provide improved crrosion resistance for steam generators. The test materials, the stress corrosion cracking test and the test results are reported. When the test temperature was raise, the stress corrosion cracking of the nickel-base alloys was accelerated. The time of stress corrosion cracking occurrence decreased with increasing applied stress, and it occurred at the stress level higher than the 0.2 % offset proof stress of Alloy 600. In Alloy 690, stress corrosion cracking was not observed at such stress level. Cold worked Alloy 600 showed higher resistance to stress corrosion cracking than the annealed alloy. (Kako, I.)

  10. Trending analysis of incidents involving primary water stress corrosion cracking on Alloy 600 components at U.S. PWRs

    International Nuclear Information System (INIS)

    Takahara, Shogo; Watanabe, Norio

    2006-01-01

    Primary Water Stress Corrosion Cracking (PWSCC) which occurs on Nickel based alloy (Alloy 600) is a worldwide concern since early 1980's. Recently several significant degradations that originate from PWSCC in the reactor coolant pressure boundary (RCPB) components have been observed at U.S. PWR plants (e.g. Oconee-3, Davis Besse). The United States Nuclear Regulation Commission (NRC) has issued generic communications to address this problem and, in response to the Davis Besse event in 2002, gave the inspection order EA-03-009 for the PWR licensees to implement the inspection of the reactor vessel heads depending upon the effective degradation years. As well, in Japan, PWSCC is considered one of the safety issues, in particular, for aged nuclear power plants and actually, some plants have experienced PWSCC on RCPB components. In the present study, we analyzed the U.S. experience with Alloy 600 degradation by reviewing the licensee event reports from 1999 to 2005 and examined the trend of them mainly focusing on affected components, characteristics of cracking and inspection approaches for detecting the PWSCC. This study indicates that PWSCC is found to be occurred on the RCPB components exposed to the environment with high temperature such as the reactor vessel head, and has the tendency to happen for specific manufactures and material according to the RCPB components. As well, it is shown that for several components, the non-destructive examination is generally needed to detect and/or confirm the PWSCC after the visual inspection and different repair techniques are applied depending on the components affected. (author)

  11. Preliminary study for extension and improvement on modeling of primary water stress corrosion cracking at control rod drive mechanism nozzles of pressurized water reactors

    International Nuclear Information System (INIS)

    Aly, Omar F.; Mattar Neto, Miguel M.; Schvartzman, Monica M.M.A.M.

    2009-01-01

    This study is for to extend, to improve the existing models, and to propose a local approach to assess the primary water stress corrosion cracking in nickel-based components. It is includes a modeling of new data for Alloy 182 and new considerations about initiation and crack growth according a developing method based on EPRI-MRP-115 (2004), and USNRC NUREG/CR-6964 (2008). The experimental data is obtained from CDTN-Brazilian Nuclear Technology Development Center, by tests through slow strain rate test (SSRT) equipment. The model conception assumed is a built diagram which indicates a thermodynamic condition for the occurrence of corrosion submodes in essayed materials, through Pourbaix diagrams, for Nickel Alloys in high temperature primary water. Over them, are superimposed different models, including a semi-empiric-probabilistic one to quantify the primary water stress corrosion cracking susceptibility, and a crack growth model. These constructed models shall be validated with the experimental data. This development aims to extent some of the models obtained to weld metals like the Alloy 182, and to improve the originals obtained according methodologies exposed in above referred reports. These methodologies comprise laboratory testing procedures, data collecting, data screening, modeling procedures, assembling of data from some laboratories in the world, plotting of results, compared analysis and discussion of these results. Preliminary results for Alloy 182 will be presented. (author)

  12. Preliminary study for extension and improvement on modeling of primary water stress corrosion cracking at control rod drive mechanism nozzles of pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Aly, Omar F.; Mattar Neto, Miguel M. [Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), Sao Paulo, SP (Brazil)], e-mail: ofaly@ipen.br, e-mail: mmattar@ipen.br; Schvartzman, Monica M.M.A.M. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil)], e-mail: monicas@cdtn.br

    2009-07-01

    This study is for to extend, to improve the existing models, and to propose a local approach to assess the primary water stress corrosion cracking in nickel-based components. It is includes a modeling of new data for Alloy 182 and new considerations about initiation and crack growth according a developing method based on EPRI-MRP-115 (2004), and USNRC NUREG/CR-6964 (2008). The experimental data is obtained from CDTN-Brazilian Nuclear Technology Development Center, by tests through slow strain rate test (SSRT) equipment. The model conception assumed is a built diagram which indicates a thermodynamic condition for the occurrence of corrosion submodes in essayed materials, through Pourbaix diagrams, for Nickel Alloys in high temperature primary water. Over them, are superimposed different models, including a semi-empiric-probabilistic one to quantify the primary water stress corrosion cracking susceptibility, and a crack growth model. These constructed models shall be validated with the experimental data. This development aims to extent some of the models obtained to weld metals like the Alloy 182, and to improve the originals obtained according methodologies exposed in above referred reports. These methodologies comprise laboratory testing procedures, data collecting, data screening, modeling procedures, assembling of data from some laboratories in the world, plotting of results, compared analysis and discussion of these results. Preliminary results for Alloy 182 will be presented. (author)

  13. Optical modeling of nickel-base alloys oxidized in pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Clair, A. [Laboratoire Interdisciplinaire Carnot de Bourgogne, UMR 6303 CNRS, Universite de Bourgogne, 9 avenue Alain Savary, BP 47870, 21078 Dijon cedex (France); Foucault, M.; Calonne, O. [Areva ANP, Centre Technique Departement Corrosion-Chimie, 30 Bd de l' industrie, BP 181, 71205 Le Creusot (France); Finot, E., E-mail: Eric.Finot@u-bourgogne.fr [Laboratoire Interdisciplinaire Carnot de Bourgogne, UMR 6303 CNRS, Universite de Bourgogne, 9 avenue Alain Savary, BP 47870, 21078 Dijon cedex (France)

    2012-10-01

    The knowledge of the aging process involved in the primary water of pressurized water reactor entails investigating a mixed growth mechanism in the corrosion of nickel-base alloys. A mixed growth induces an anionic inner oxide and a cationic diffusion parallel to a dissolution-precipitation process forms the outer zone. The in situ monitoring of the oxidation kinetics requires the modeling of the oxide layer stratification with the full knowledge of the optical constants related to each component. Here, we report the dielectric constants of the alloys 600 and 690 measured by spectroscopic ellipsometry and fitted to a Drude-Lorentz model. A robust optical stratification model was determined using focused ion beam cross-section of thin foils examined by transmission electron microscopy. Dielectric constants of the inner oxide layer depleted in chromium were assimilated to those of the nickel thin film. The optical constants of both the spinels and extern layer were determined. - Highlights: Black-Right-Pointing-Pointer Spectroscopic ellipsometry of Ni-base alloy oxidation in pressurized water reactor Black-Right-Pointing-Pointer Measurements of the dielectric constants of the alloys Black-Right-Pointing-Pointer Optical simulation of the mixed oxidation process using a three stack model Black-Right-Pointing-Pointer Scattered crystallites cationic outer layer; linear Ni-gradient bottom layer Black-Right-Pointing-Pointer Determination of the refractive index of the spinel and the Cr{sub 2}O{sub 3} layers.

  14. The suppression of dissolution for alloy 690 in high temperature and high pressure water with chromium ion implantation

    International Nuclear Information System (INIS)

    Shibata, Toshio; Fujimoto, Shinji; Ohtani, Saburou; Watanabe, Masanori; Hirao, Kyozo; Okumoto, Masaru; Shibaike, Hiroyuki.

    1994-01-01

    As the material of heat exchanger tubes for PWRs, the nickel alloys such as alloy 690 and alloy 600 have been used, but 58 Ni and 60 Co contained as an impurity elute in primary cooling water, and are radioactivated, in this way, they become the cause of radiation exposure. By increasing chromium concentration, the corrosion resistance of nickel alloys is improved, and for modern heat exchangers, the alloy 690, of which the chromium content is increased up to 30%, has been adopted, and excellent results have been obtained. In this research, aiming at the further reduction of radiation exposure, by increasing the chromium concentration in surface layer using ion implantation technology, the change of the corrosion behavior of alloy 690 in high temperature, high pressure water was investigated. The chemical composition of the alloy 690 used, and the making of plate specimens are shown. The polarization behavior of alloy 690 in 0.1 mol/l sulfuric acid deaerated at normal temperature is reported, and the effect of suppressing dissolution was remarkable in the specimens with much implantation. The electrochemical behavior of alloy 690 in simulated cooling water was investigated. Immobile case has high chromium content and is thin. (K.I.)

  15. Stress corrosion cracking of alloy 182 weld in a PWR water environment

    International Nuclear Information System (INIS)

    Lima, Luciana Iglesias Lourenco; Schvartzman, Monica Maria de Abreu Mendonca; Quinan, Marco Antonio Dutra; Soares, Antonio Edicleto Gomes; Piva, Stephano P.T.

    2011-01-01

    The weld used to connect two different metals is known as dissimilar metal welds (DMW). In the nuclear power plant, this weld is used to join stainless steel nipples to low alloy carbon steel components on the nuclear pressurized water reactor (PWR). In most cases, nickel alloys are used to joint these materials. These alloys are known to accommodate the differences in composition and thermal expansion of the two materials. The stress corrosion cracking (SCC) is a phenomenon that occurs in nuclear power plants metallic components where susceptibility materials are subjected to the simultaneously effect of mechanical stress and an aggressive media with different compositions. SCC is one of degradation process that gradually introduces damage of components, change their characteristics with the operation time. The nickel alloy 600, and their weld metals (nickel alloys 82 and 182), originally selected due to its high corrosion resistance, it exhibit after long operation period (20 years), susceptibility to the SCC. This study presents a comparative work between the SCC in the Alloy 182 filler metal weld in two different temperatures (303 deg C and 325 deg C) in primary water. The susceptibility to stress corrosion cracking was assessed using the slow strain rate tensile (SSRT) test. The results of the SSRT tests indicated that SCC is a thermally-activated mechanism and that brittle fracture caused by the corrosion process was observed at 325 deg C. (author)

  16. Direct contact heat transfer characteristics between melting alloy and water

    International Nuclear Information System (INIS)

    Kinoshita, Izumi; Nishi, Yoshihisa; Furuya, Masahiro

    1995-01-01

    As a candidate for an innovative steam generator for fast breeder reactors, a heat exchanger with direct contact heat transfer between melting alloy and water was proposed. The evaluation of heat transfer characteristics of this heat exchanger is one of the research subjects for the design and development of the steam generator. In this study, the effect of the pressure on heat transfer characteristics and the required degree of superheating of melting alloy above water saturation temperature are evaluated during the direct contact heat transfer experiment by injecting water into Wood's alloy. In the experiment, the pressure, the temperature of the Wood's alloy, the flow rate of feed water, and the depth of the feed water injection point are varied as parameters. As a result of the experiment, the product of the degree of Wood's alloy superheating above water saturation temperature and the depth of the feed water injection point is constant for each pressure. This constant increases as the pressure rises. (author)

  17. Fracture behavior of nickel-based alloys in water

    Energy Technology Data Exchange (ETDEWEB)

    Mills, W.J.; Brown, C.M.

    1999-08-01

    The cracking resistance of Alloy 600, Alloy 690 and their welds, EN82H and EN52, was characterized by conducting J{sub IC} tests in air and hydrogenated water. All test materials displayed excellent toughness in air and high temperature water, but Alloy 690 and the two welds were severely embrittled in low temperature water. In 54 C water with 150 cc H{sub 2}/kg H{sub 2}O, J{sub IC} values were typically 70% to 95% lower than their air counterparts. The toughness degradation was associated with a fracture mechanism transition from microvoid coalescence to intergranular fracture. Comparison of the cracking response in water with that for hydrogen-precharged specimens tested in air demonstrated that susceptibility to low temperature cracking is due to hydrogen embrittlement of grain boundaries. The effects of water temperature, hydrogen content and loading rate on low temperature crack propagation were studied. In addition, testing of specimens containing natural weld defects and as-machined notches was performed to determine if low temperature cracking can initiate at these features. Unlike the other materials, Alloy 600 is not susceptible to low temperature cracking as the toughness in 54 C water remained high and a microvoid coalescence mechanism was operative in both air and water.

  18. Modeling the initiation of Primary Water Stress Corrosion Cracking in nickel base alloys 182 and 82 of Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Wehbi, Mickael

    2014-01-01

    Nickel base welds are widely used to assemble components of the primary circuit of Pressurized Water Reactors (PWR) plants. International experience shows an increasing number of Stress Corrosion Cracks (SCC) in nickel base welds 182 and 82 which motivates the development of models predicting the time to SCC initiation for these materials. SCC involves several parameters such as materials, mechanics or environment interacting together. The goal of this study is to have a better understanding of the physical mechanisms occurring at grains boundaries involved in SCC. In-situ tensile test carried out on oxidized alloy 182 evidenced dispersion in the susceptibility to corrosion of grain boundaries. Moreover, the correlation between oxidation and cracking coupled with micro-mechanical simulations on synthetic polycrystalline aggregate, allowed to propose a cracking criterion of oxidized grain boundaries which is defined by both critical oxidation depth and local stress level. Due to the key role of intergranular oxidation in SCC and since significant dispersion is observed between grain boundaries, oxidation tests were performed on alloys 182 and 82 in order to model the intergranular oxidation kinetics as a function of chromium carbides precipitation, temperature and dissolved hydrogen content. The model allows statistical analyses and is embedded in a local initiation model. In this model, SCC initiation is defined by the cracking of the intergranular oxide and is followed by slow and fast crack growth until the crack depth reaches a given value. Simplifying assumptions were necessary to identify laws used in the SCC model. However, these laws will be useful to determine experimental conditions of future investigations carried out to improve the calibration used parameters. (author)

  19. Atomic origins of water-vapour-promoted alloy oxidation.

    Science.gov (United States)

    Luo, Langli; Su, Mao; Yan, Pengfei; Zou, Lianfeng; Schreiber, Daniel K; Baer, Donald R; Zhu, Zihua; Zhou, Guangwen; Wang, Yanting; Bruemmer, Stephen M; Xu, Zhijie; Wang, Chongmin

    2018-05-07

    The presence of water vapour, intentional or unavoidable, is crucial to many materials applications, such as in steam generators, turbine engines, fuel cells, catalysts and corrosion 1-4 . Phenomenologically, water vapour has been noted to accelerate oxidation of metals and alloys 5,6 . However, the atomistic mechanisms behind such oxidation remain elusive. Through direct in situ atomic-scale transmission electron microscopy observations and density functional theory calculations, we reveal that water-vapour-enhanced oxidation of a nickel-chromium alloy is associated with proton-dissolution-promoted formation, migration, and clustering of both cation and anion vacancies. Protons derived from water dissociation can occupy interstitial positions in the oxide lattice, consequently lowering vacancy formation energy and decreasing the diffusion barrier of both cations and anions, which leads to enhanced oxidation in moist environments at elevated temperatures. This work provides insights into water-vapour-enhanced alloy oxidation and has significant implications in other material and chemical processes involving water vapour, such as corrosion, heterogeneous catalysis and ionic conduction.

  20. Corrosion Behavior of Nickel-Plated Alloy 600 in High Temperature Water

    International Nuclear Information System (INIS)

    Kim, Ji Hyun; Hwang, Il Soon

    2008-01-01

    In this paper, electrochemical and microstructural characteristics of nickel-plated Alloy 600 wee investigated in order to identify the performance of electroless Ni-plating on Alloy 600 in high-temperature aqueous condition with the comparison of electrolytic nickel-plating. For high temperature corrosion test of nickel-plated Alloy 600, specimens were exposed for 770 hours to typical PWR primary water condition. During the test, open circuit potentials (OCP's) of all specimens were measured using a reference electrode. Also, resistance to flow accelerated corrosion (FAC) test was examined in order to check the durability of plated layers in high-velocity flow environment at high temperature. After exposures to high flow rate aqueous condition, the integrity of surfaces was confirmed by using both scanning electron microscopy (SEM) and energy dispersive spectroscopy (EDS). For the field application, a remote process for electroless nickel-plating was demonstrated using a plate specimen with narrow gap on a laboratory scale. Finally, a practical seal design was suggested for more convenient application

  1. Oxidation behavior of steels and Alloy 800 in supercritical water

    International Nuclear Information System (INIS)

    Olmedo, A.M.; Bordoni, R.; Dominguez, G.; Alvarez, M.G.

    2011-01-01

    The oxidation behavior of a ferritic-martensitic steel T91 and a martensitic steel AISI 403 up to 750 h, and of AISI 316L and Alloy 800 up to 336 h in deaerated supercritical water, 450ºC-25 MPa, was investigated in this paper. After exposure up to 750 h, the weight gain data, for steels T91 and AISI 403, was fitted by ∆W=k t n , were n are similar for both steels and k is a little higher for T91. The oxide films grown in the steels were characterized using gravimetry, scanning electron microscopy/energy dispersive X-ray spectroscopy (SEM/EDS) and X-ray diffraction. The films were adherent and exhibited a low porosity. For this low oxygen content supercritical water exposure, the oxide scale exhibited a typical duplex structure, in which the scale is composed of an outer iron oxide layer of magnetite (Fe 3 O 4 ) and an inner iron/chromium oxide layer of a non-stoichiometric iron chromite (Fe,Cr) 3 O 4 . Preliminary results, with AISI 316L and Alloy 800, for two exposure periods (168 and 336 h), are also reported. The morphology shown for the oxide films grown on both materials up to 336 h of oxidation in supercritical water, resembles that of a duplex layer film like that shown by stainless steels and Alloy 800 oxide films grown in a in a high temperature and pressure (220-350ºC) of a primary or secondary coolant of a plant. (author) [es

  2. Rare earth concentration in the primary Si crystal in rare earth added Al-21 wt. % Si alloy

    Energy Technology Data Exchange (ETDEWEB)

    Chang, J.Y.; Kim, G.H. [Korea Inst. of Science and Technology, Seoul (Korea, Republic of); Moon, I.G.; Choi, C.S. [Yonsei Univ., Seoul (Korea, Republic of). Dept. of Metallurgical Engineering

    1998-07-03

    Al-Si alloys containing more than about 12 wt. % Si exhibit a hypereutectic microstructure, normally consisting of a primary silicon phase in an eutectic matrix. The primary silicon in normal hypereutectic alloys is usually very coarse and thus leads to poor properties to these alloys. Therefore, alloys with a predominantly coarse primary silicon crystal must be modified to ensure adequate mechanical strength and ductility. Further improvement of mechanical properties of these alloys can be achieved by the modification of eutectic microstructure. Therefore, development of a modifier or refiner that can produce both fine primary and eutectic Si is a major factor which can lead to significant enhancement of mechanical properties in hypereutectic Al-Si alloys. Refinement of primary silicon is usually achieved by the addition of phosphor to the melt. On the other hand, it is reported that the rare earth (RE) elements are capable of modifying the eutectic structure of cast Al-Si alloys. According to the literature, Phosphor acts as a heterogeneous nucleation site of Si crystal by forming AlP intermetallic particles at high temperature, i.e., above liquidus temperature of Al-Si alloy. Unlike phosphor, RE was not known to form a stable compound with Al that can act as a nucleation site at high temperature. Therefore, the role of RE as a refiner should be considered by examining the behavior of RE as a solute in the melt. The distribution of RE within the primary Si and in the matrix of the alloy will provide a clue to the role of RE on the modification of primary Si during solidification.

  3. Interactions between drops of a molten aluminum-lithium alloy and liquid water

    International Nuclear Information System (INIS)

    Nelson, L.S.

    1994-01-01

    In certain hypothesized nuclear reactor accident scenarios, 1- to 10-g drops of molten aluminum-lithium alloys might contact liquid water. Because vigorous steam explosions have occurred when large amounts of molten aluminum-lithium alloys were released into water or other coolants, it becomes important to know whether there will be explosions if smaller amounts of these molten alloys similarly come into contact with water. Therefore, the authors released drops of molten Al-3.1 wt pct Li alloy into deionized water at room temperature. The experiments were performed at local atmospheric pressure (0.085 MPa) without pressure transient triggers applied to the water. The absence of these triggers allowed them to (a) investigate whether spontaneous initiation of steam explosions would occur with these drops and (b) study the alloy-water chemical reactions. The drop sizes and melt temperatures were chosen to simulate melt globules that might form during the hypothesized melting of the aluminum-lithium alloy components

  4. Improvement of corrosion resistance of vanadium alloys in high-temperature pressurized water

    International Nuclear Information System (INIS)

    Fujiwara, Mitsuhiro; Sakamoto, Toshiya; Satou, Manabu; Hasegawa, Akira; Abe, Katsunori; Kaiuchi, Kazuo; Furuya, Takemi

    2005-01-01

    Corrosion tests in pressurized and vaporized water were conducted for V-based high Cr and Ti alloys and V-4Cr-4Ti type alloys containing minor elements such as Si, Al and Y. Weight losses were observed for every alloy after corrosion tests in pressurized water. It was apparent that addition of Cr effectively reduced the weight change in pressurized water. The weight loss of V-4Cr-4Ti type alloys in corrosion tests in vaporized water was also reduced as Cr content increased. The V-20Cr-4Ti alloy had a slight weight gain, almost same as that of SUS316, which had the best corrosion properties in the tested alloys. The elongation of alloys with in excess of 10% Cr was reduced as Cr content increased. The elongations of the V-12Cr-4Ti and the V-15Cr-4Ti alloys were significantly reduced by corrosion and cleavage fracture was observed reflecting hydrogen embrittlement. The reduced elongations of the alloys of the alloys were recovered to the same level of as annealed conditions after hydrogen degassing. After corrosion, the V-15Cr-4Ti-0.5Y alloy still kept enough elongation, suggesting that the addition of Y is effective to reduce the hydrogen embrittlement. (author)

  5. The use of hardfacing alloys in nuclear power plants

    International Nuclear Information System (INIS)

    Saarinen, K.; Aaltonen, P.

    1987-08-01

    In this report the structure and applications of cobalt-, nickel- and iron-based hardfacing alloys are reviewed. Cobalt-based hardfacing alloys are widely used in nuclear power plant components due to their good wear and corrosion resistance. However, the wear and corrosion products of the cobalt-containing alloys are released into the primary cooling water and transported to the reactor core where cobalt (Co-59) is transmuted to the radioactive isotope Co-60. It has been estimated that cobalt-based hardfacing alloys are responsible for up to 90% of the total cobalt released to the primary water circuit. The cobalt based hardfacing alloys are used in such components as valves, control blade pins and pumps, etc. In the Finnish nuclear power plants they are not used in in-core components. The replacement of cobalt-containing alloys in primary cooling system components is studied in many laboratories, but substitutes for the cobalt-based alloys in the complete range of nuclear hardfacing applications have so far not been found. However, the modified austenitic stainless steels have showed good resistance to galling wear and are therefore considered substitutes for cobalt-based alloys

  6. Literature Survey on the Stress Corrosion Cracking of Low-Alloy Steels in High Temperature Water

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P

    2002-02-01

    The present report is a summary of a literature survey on the stress corrosion cracking (SCC) behaviour/ mechanisms in low-alloy steels (LAS) in high-temperature water with special emphasis to primary-pressure-boundary components of boiling water reactors (BWR). A brief overview on the current state of knowledge concerning SCC of low-alloy reactor pressure vessel and piping steels under BWR conditions is given. After a short introduction on general aspects of SCC, the main influence parameter and available quantitative literature data concerning SCC of LAS in high-temperature water are discussed on a phenomenological basis followed by a summary of the most popular SCC models for this corrosion system. The BWR operating experience and service cracking incidents are discussed with respect to the existing laboratory data and background knowledge. Finally, the most important open questions and topics for further experimental investigations are outlined. (author)

  7. Primary water stress corrosion cracking resistance of alloy X-750 for guide tube support pins

    International Nuclear Information System (INIS)

    Yonezawa, T.; Onimura, K.; Yonehana, M.; Fujitani, T.

    1990-01-01

    The authors have developed the maintenance free guide tube support pins for PWR, and conducted the three kinds of long time stress corrosion cracking tests in high temperature water, in order to verify the reliability of the maintenance free guide tube support pins. This paper describes the features of our maintenance free support pins and the results of long time stress corrosion cracking test for the maintenance free support pins. After exposure at 320 0 C in the simulated primary water of PWR for about 35,000 hours or at 360 0 C in the same chemistry water as the primary water for about 24,900 hours, no abnormal indication such as cracks was observed in all test support pins exposed 320 0 C and 360 0 C primary water, by ultrasonic inspection, and liquid penetrate test. From the above, it seems that our maintenance free support pins are keepable the soundness up to the end of plant life, in PWR plants

  8. Thermal stability and primary phase of Al-Ni(Cu)-La amorphous alloys

    International Nuclear Information System (INIS)

    Huang Zhenghua; Li Jinfu; Rao Qunli; Zhou Youhe

    2008-01-01

    Thermal stability and primary phase of Al 85+x Ni 9-x La 6 (x = 0-6) and Al 85 Ni 9-x Cu x La 6 (x = 0-9) amorphous alloys were investigated by X-ray diffraction and differential scanning calorimeter. It is revealed that replacing Ni in the Al 85 Ni 9 La 6 alloy by Cu decreases the thermal stability and makes the primary phase change from intermetallic compounds to single fcc-Al as the Cu content reaches and exceeds 4 at.%. When the Ni and La contents are fixed, replacing Al by Cu increases the thermal stability but also promotes the precipitation of single fcc-Al as the primary phase

  9. Effect of surface cold work on corrosion of Alloy 690TT in high temperature high pressure water

    International Nuclear Information System (INIS)

    Wang, J.; Zhang, Z.; Han, E.-H.; Ke, W.

    2009-01-01

    This paper aims to investigate the effect of surface cold work on corrosion of Alloy 690TT. The Alloy 690TT was mechanical ground and electro polished respectively and immersed in primary water at DO = 2 ppm and DH = 2.5ppm respectively. The microstructure of surface and the compositions and morphology of the surface film on Alloy 690TT after immersion test were studied using scanning electron microscopy (SEM), transmission electron microscopy (TEM), Auger electron spectroscopy (AES) and focused ion beam (FIB). The results showed that feather-like oxide with decorated polyhedral oxide formed on ground surface and needle-like oxide with decorated polyhedral oxide formed on electro-polished surface. (author)

  10. Characterization of acoustic cavitation in water and molten aluminum alloy.

    Science.gov (United States)

    Komarov, Sergey; Oda, Kazuhiro; Ishiwata, Yasuo; Dezhkunov, Nikolay

    2013-03-01

    High-intensive ultrasonic vibrations have been recognized as an attractive tool for refining the grain structure of metals in casting technology. However, the practical application of ultrasonics in this area remains rather limited. One of the reasons is a lack of data needed to optimize the ultrasonic treatment conditions, particularly those concerning characteristics of cavitation zone in molten aluminum. The main aim of the present study was to investigate the intensity and spectral characteristics of cavitation noise generated during radiation of ultrasonic waves into water and molten aluminum alloys, and to establish a measure for evaluating the cavitation intensity. The measurements were performed by using a high temperature cavitometer capable of measuring the level of cavitation noise within five frequency bands from 0.01 to 10MHz. The effect of cavitation treatment was verified by applying high-intense ultrasonic vibrations to a DC caster to refine the primary silicon grains of a model Al-17Si alloy. It was found that the level of high frequency noise components is the most adequate parameter for evaluating the cavitation intensity. Based on this finding, it was concluded that implosions of cavitation bubbles play a decisive role in refinement of the alloy structure. Copyright © 2012 Elsevier B.V. All rights reserved.

  11. In situ observation of ultrasonic cavitation-induced fragmentation of the primary crystals formed in Al alloys.

    Science.gov (United States)

    Wang, Feng; Tzanakis, Iakovos; Eskin, Dmitry; Mi, Jiawei; Connolley, Thomas

    2017-11-01

    The cavitation-induced fragmentation of primary crystals formed in Al alloys were investigated for the first time by high-speed imaging using a novel experimental approach. Three representative primary crystal types, Al 3 Ti, Si and Al 3 V with different morphologies and mechanical properties were first extracted by deep etching of the corresponding Al alloys and then subjected to ultrasonic cavitation processing in distilled water. The dynamic interaction between the cavitation bubbles and primary crystals was imaged in situ and in real time. Based on the recorded image sequences, the fragmentation mechanisms of primary crystals were studied. It was found that there are three major mechanisms by which the primary crystals were fragmented by cavitation bubbles. The first one was a slow process via fatigue-type failure. A cyclic pressure exerted by stationary pulsating bubbles caused the propagation of a crack pre-existing in the primary crystal to a critical length which led to fragmentation. The second mechanism was a sudden process due to the collapse of bubbles in a passing cavitation cloud. The pressure produced upon the collapse of the cloud promoted rapid monotonic crack growth and fast fracture in the primary crystals. The third observed mechanism was normal bending fracture as a result of the high pressure arising from the collapse of a bubble cloud and the crack formation at the branch connection points of dendritic primary crystals. The fragmentation of dendrite branches due to the interaction between two freely moving dendritic primary crystals was also observed. A simplified fracture analysis of the observed phenomena was performed. The specific fragmentation mechanism for the primary crystals depended on their morphology and mechanical properties. Copyright © 2017 The Author(s). Published by Elsevier B.V. All rights reserved.

  12. Stress Corrosion Cracking of Ni-Fe-Cr Alloys Relevant to Nuclear Power Plants

    Science.gov (United States)

    Persaud, Suraj

    Stress corrosion cracking (SCC) of Ni-Fe-Cr alloys and weld metals was investigated in simulated environments representative of high temperature water used in the primary and secondary circuits of nuclear power plants. The mechanism of primary water SCC (PWSCC) was studied in Alloys 600, 690, 800 and Alloy 82 dissimilar metal welds using the internal oxidation model as a guide. Initial experiments were carried out in a 480°C hydrogenated steam environment considered to simulate high temperature reducing primary water. Ni alloys underwent classical internal oxidation intragranularly resulting in the expulsion of the solvent metal, Ni, to the surface. Selective intergranular oxidation of Cr in Alloy 600 resulted in embrittlement, while other alloys were resistant owing to their increased Cr contents. Atom probe tomography was used to determine the short-circuit diffusion path used for Ni expulsion at a sub-nanometer scale, which was concluded to be oxide-metal interfaces. Further exposures of Alloys 600 and 800 were done in 315°C simulated primary water and intergranular oxidation tendency was comparable to 480°C hydrogenated steam. Secondary side work involved SCC experiments and electrochemical measurements, which were done at 315°C in acid sulfate solutions. Alloy 800 C-rings were found to undergo acid sulfate SCC (AcSCC) to a depth of up to 300 microm in 0.55 M sulfate solution at pH 4.3. A focused-ion beam was used to extract a crack tip from a C-ring and high resolution analytical electron microscopy revealed a duplex oxide structure and the presence of sulfur. Electrochemical measurements were taken on Ni alloys to complement crack tip analysis; sulfate was concluded to be the aggressive anion in mixed sulfate and chloride systems. Results from electrochemical measurements and crack tip analysis suggested a slip dissolution-type mechanism to explain AcSCC in Ni alloys.

  13. Studies on the growth of oxide films on alloy 800 and alloy 600 in lithiated water at high temperature

    International Nuclear Information System (INIS)

    Olmedo, A.M.; Bordon, R.

    2007-01-01

    In this work, the oxide films grown on Alloy 800 and Alloy 600 in lithiated (pH 25 C d egrees = 10.2-10.4) water at high temperature, with and without hydrogen overpressure (HO) and an initial oxygen dissolved in the water have been studied. The oxide films were grown at different temperatures (220-350 C degrees) and exposure times with HO, and at 315 C degrees without HO in static autoclaves. Some results are also reported for oxide layers grown on Alloy 800 coupons exposed in a high temperature loop during extended exposure times. The average oxide thickness was determined using descaling procedures. The morphology and composition of the oxide films were analyzed with scanning electron microscopy (SEM), EDS and X-ray diffraction (XRD). For both Alloys, at 350 C degrees with HO, the oxide layers were clearly composed of a double layer: an inner one of very small crystallites and an outer layer formed by bigger crystals scattered over the inner one. The analysis by X-ray diffraction indicated the presence of spinel structures like magnetite (Fe 3 O 4 ) and ferrites and/or nickel chromites. In this case the average oxide thickness was around 0.12 to 0.15 μm for both Alloys. Similar values were found at lower temperatures. The morphology of the oxide layer was similar at lower temperatures for Alloy 800, but a different morphology consisting of platelets or needles was found for Alloy 600. The oxide morphology found at 315 C degrees, without HO and with initial dissolved oxygen in the water, was also very different between both Alloys. The oxide film grown on Alloy 600 with an initial dissolved oxygen in the water, showed clusters of platelets forming structures like flowers that were dispersed on an rather homogeneous layer consisting of smaller platelets or needles. The average oxide film grown in this case was around 0.25 μm for Alloy 600 and 0.18 μm for Alloy 800. (author) [es

  14. Comparison of laboratory and field experience of PWSCC in Alloy 182 weld metal

    Energy Technology Data Exchange (ETDEWEB)

    Scott, P.; Meunier, M.-C.; Steltzlen, F. [AREVA NP, Tour AREVA, Paris La Defense (France); Calonne, O.; Foucault, M. [AREVA NP, Centre Technique, Le Creusot Cedex (France); Combrade, P. [ACXCOR, Saint Etienne (France); Amzallag, C. [EDF, SEPTEN, Villeurbanne (France)

    2007-07-01

    Laboratory studies of stress corrosion cracking of the nickel base weld metal, Alloy 182, in simulated PWR primary water suggest similar resistance to crack initiation and somewhat enhanced propagation rates relative to wrought Alloy 600. By contrast, field experience of cracking in the primary circuits of PWRs shows in general much better performance for Alloy 182 relative to Alloy 600 than would be anticipated from laboratory studies. This paper endeavours to resolve this apparent conundrum. It draws on the conclusions of recent research that has focussed on the role of surface finish, particularly cold work and residual stresses resulting from different fabrication processes, on the risk of initiating IGSCC in nickel base alloys in PWR primary water. It also draws on field experience of stress corrosion cracking that highlights the important role of surface finish for crack initiation. (author)

  15. Development of an autoclave with zirconia crystal windows for in-situ observation of sample surface under primary water conditions of pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fukumura, Takuya; Totsuka, Nobuo; Arioka, Koji [Inst. of Nuclear Safety System Inc., Mihama, Fukui (Japan); Nakajima, Nobuo

    2002-09-01

    Elucidating the mechanism for primary water stress corrosion cracking (PWSCC) is important for improving the reliability of structural materials in the primary system of pressurized water reactors (PWR). For this purpose, visualization of corrosion material surface in the primary coolant environment is effective, but it was impossible because of lack of suitable window material. Yttria stabilized zirconia was newly selected as a candidate for in-situ window material in the primary coolant environment of PWR. Its sufficient corrosion resistance was proved by measuring the transmissivity of light after being immersed in the primary coolant environment. A new autoclave with two windows of yttria-stabilized zirconia was developed. The corrosion material surfaces of Alloy600 and SUS304 in the primary coolant environment were clearly observed with this autoclave. Observations of cracks generated on the surface of SUS304 specimen, suggest that its generation time depends on temperature. (author)

  16. Supercritical water corrosion of high Cr steels and Ni-base alloys

    International Nuclear Information System (INIS)

    Jang, Jin Sung; Han, Chang Hee; Hwang, Seong Sik

    2004-01-01

    High Cr steels (9 to 12% Cr) have been widely used for high temperature high pressure components in fossil power plants. Recently the concept of SCWR (supercritical water-cooled reactor) has aroused a keen interest as one of the next generation (Generation IV) reactors. Consequently Ni-base (or high Ni) alloys as well as high Cr steels that have already many experiences in the field are among the potential candidate alloys for the cladding or reactor internals. Tentative inlet and outlet temperatures of the anticipated SCWR are 280 and 510 .deg. C respectively. Among many candidate alloys there are austenitic stainless steels, Ni base alloys, ODS alloys as well as high Cr steels. In this study the corrosion behavior of the high Cr steels and Ni base (or high Ni) alloys in the supercritical water were investigated. The corrosion behavior of the unirradiated base metals could be used in the near future as a guideline for the out-of-pile or in-pile corrosion evaluation tests

  17. High temperature alloys for the primary circuit of a prototype nuclear process heat plant

    International Nuclear Information System (INIS)

    Ennis, P.J.; Schuster, H.

    1979-01-01

    As part of a comprehensive materials test programme for the High Temperature Reactor Project 'Prototype Plant for Nuclear Process Heat' (PNP), high temperature alloys are being investigated for primary circuit components operating at temperatures above 750 0 C. On the basis of important material parameters, in particular corrosion behaviour and mechanical properties in primary coolant helium, the potential of candidate alloys is discussed. By comparing specific PNP materials data with the requirements of PNP and those of conventional plant, the implications for the materials programme and component design are given. (orig.)

  18. Influence of the S/N ratio on the corrosion release of Alloy 690 tubes in a primary coolant

    International Nuclear Information System (INIS)

    Shim, Hee-Sang; Choi, Myung Sik; Kim, Kyung Mo; Seo, Myung Ji; Hur, Do Haeng; Choi, Tack-Sang; Yoo, One

    2014-01-01

    Alloy 690TT is a promising steam generator (SG) tube material of a pressurized water reactor due to its excellent resistance to stress corrosion cracking (SCC) that has caused problems in Alloy 600 as an old SG tube material. The qualities of this material have been managed thoroughly from manufacturing step under various specification regulations as well as in in-service step. For examples, the surface roughness are prescribed as the values less than 1.6 μm for the tube outside and 0.5 μm for the inside, respectively. In addition, the surface state and defect must be qualified through the eddy current test (ECT) and the ultrasonic test (UT) according to the ASME Section III, NB2550. Then, the signal-to-noise (S/N) ratio, which is measured using ECT bobbin probe, is the important criteria to determine the material and it shall be 15 to 1 or higher at the standard frequency for any fixed 0.5 m length of any tube. The corrosion behaviours of the Alloy 690TT under high-temperature pressurized primary water have been studied widely in a point of the SCC but discussed narrowly in a point of the corrosion release. In particular, the effect of the S/N ratio on the corrosion release of this material surface has been rarely investigated. In this work, we evaluate the influence of the S/N ratio on the corrosion release of Alloy 690 SG tubes. The specimens with different S/N ratio were selected through ECT bobbin inspection and a corrosion release test was conducted using a simulated primary circulation loop. The material properties and oxidation behaviours were investigated by surface profiler, scanning electron microscopy, transmission electron microscopy, grazing incidence X-ray diffraction and etc. As a result, the corrosion rate was matched preferably with the MRPC characteristics showing macroscopic surface state rather than with the bobbin S/N ratio results. (author)

  19. Alloy 690 in PWR type reactors; Aleaciones base niquel en condiciones de primario de los reactores tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Briceno, D.; Serrano, M.

    2005-07-01

    Alloy 690, used as replacement of Alloy 600 for vessel head penetration (VHP) nozzles in PWR, coexists in the primary loop with other components of Alloy 600. Alloy 690 shows an excellent resistance to primary water stress corrosion cracking, while Alloy 600 is very susceptible to this degradation mechanisms. This article analyse comparatively the PWSCC behaviour of both Ni-based alloys and associated weld metals 52/152 and 82/182. (Author)

  20. The Effect of Water Mist Cooling of Casting Die on the Solidification, Microstructure and Properties of AlSi20 Alloy

    Directory of Open Access Journals (Sweden)

    Władysiak R.

    2017-03-01

    Full Text Available Unmodified AlSi20 alloy were casted at the research station, allowing for sequential multipoint cooling using a dedicated computer- controlled program. This method allows for the formation of the microstructure of hypereutectic AlSi20 alloy and also increases hardness. Primary silicon dendrites were found in the microstructure of cooled samples. Based on these dendrites, the formation of primary silicon particles is explained. Cooling of casting die with a water mist stream causes changes in solidification, which leads to expansion of the boundary layer with columnar crystals and shrinkage of the core zone with equiaxed crystals. It also causes more regular hardness distribution around pre-eutectic Si crystals, which can lead to tensile strength and machinability improvement.

  1. Effect of cold working on the stress corrosion cracking resistance of nickel-chromium-iron alloys

    International Nuclear Information System (INIS)

    Yonezawa, T.; Onimura, K.

    1987-01-01

    In order to grasp the stress corrosion cracking resistance of cold worked nickel base alloys in PWR primary water, the effect of cold working on the stress corrosion cracking resistance of alloys 600, X-750 and 690, in high temperature water, have been studied. Stress corrosion cracking tests were conducted at 360 0 C (633K) in a simulated PWR primary water for about 12,000 hours (43.2Ms). From the test results, it is concluded that the stress corrosion cracking resistance in the cold worked Alloy 600 at the same applied stress level increases with an increase in cold working ratio, and the cold worked alloys of thermally treated 690 and X-750 have excellent stress corrosion cracking resistance. (Author)

  2. Corrosion of high temperature alloys in the primary circuit helium of high temperature gas cooled reactors. Pt. 2

    International Nuclear Information System (INIS)

    Quadakkers, W.J.

    1985-01-01

    The reactive impurities H 2 O, CO, H 2 and CH 4 which are present in the primary coolant helium of high temperature gas-cooled reactors can cause scale formation, internal oxidation and carburization or decarburization of the high temperature structural alloys. In Part 1 of this contribution a theoretical model was presented, which allows the explanation and prediction of the observed corrosion effects. The model is based on a classical stability diagram for chromium, modified to account for deviations from equilibrium conditions caused by kinetic factors. In this paper it is shown how a stability diagram for a commercial alloy can be constructed and how this can be used to correlate the corrosion results with the main experimental parameters, temperature, gas and alloy composition. Using the theoretical model and the presented experimental results, conditions are derived under which a protective chromia based surface scale will be formed which prevents a rapid transfer of carbon between alloy and gas atmosphere. It is shown that this protective surface oxide can only be formed if the carbon monoxide pressure in the gas exceeds a critical value. Psub(CO), which depends on temperature and alloy composition. Additions of methane only have a limited effect provided that the methane/water ratio is not near to, or greater than, a critical value of around 100/1. The influence of minor alloying additions of strong oxide forming elements, commonly present in high temperature alloys, on the protective properties of the chromia surface scales and the kinetics of carbon transfer is illustrated. (orig.) [de

  3. Electrodeposition of Ni-Mo alloy coatings for water splitting reaction

    Science.gov (United States)

    Shetty, Akshatha R.; Hegde, Ampar Chitharanjan

    2018-04-01

    The present study reports the development of Ni-Mo alloy coatings for water splitting applications, using a citrate bath the inducing effect of Mo (reluctant metal) on electrodeposition, its relationship with their electrocatalytic efficiency were studied. The alkaline water splitting efficiency of Ni-Mo alloy coatings, for both hydrogen evolution reaction (HER) and oxygen evolution reaction were tested using cyclic voltammetry (CV) and chronopotentiometry (CP) techniques. Moreover, the practical utility of these electrode materials were evaluated by measuring the amount of H2 and O2 gas evolved. The variation in electrocatalytic activity with composition, structure, and morphology of the coatings were examined using XRD, SEM, and EDS analyses. The experimental results showed that Ni-Mo alloy coating is the best electrode material for alkaline HER and OER reactions, at lower and higher deposition current densities (c. d.'s) respectively. This behavior is attributed by decreased Mo and increased Ni content of the alloy coating and the number of electroactive centers.

  4. Study on mechanical interaction between molten alloy and water

    International Nuclear Information System (INIS)

    Nishimura, Satoshi; Ueda, Nobuyuki; Nishi, Yoshihisa; Furuya, Masahiro; Kinoshita, Izumi

    1999-01-01

    Simulant experiments using low melting point molten alloy and water have been conducted to observe both fragmentation behavior of molten jet and boiling phenomena of water, and to measure both particle size and shape of fragmented solidified jet, focusing on post-pin-failure molten fuel-coolant interaction (FCl) which was important to evaluate the sequence of the initiating phase for metallic fueled FBR. In addition, characteristics of coolant boiling phenomena on FCIs have been investigated, focusing on the boiling heat transfer in the direct contact heat transfer mode. As a results, it is concluded that the fragmentation of poured molten alloy jet is affected by a degree of boiling of water and is classified into three modes by thermal conditions of both the instantaneous contact interface temperature of two liquids and subcooling of water. In the case of forced convection boiling in direct contact mode, it is found that the heat transfer performance is enhanced by increase of the heat transfer area, due to oscillation of the surface and fragmentation of molten alloy. As a results of preliminary investigation of FCI behavior for metallic fuel core based on these results, it is expected that the ejected molten fuel is fragmented into almost spherical particles due to the developed boiling of sodium. (author)

  5. Proceedings: Primary water stress corrosion cracking: 1989 EPRI remedial measures workshop

    International Nuclear Information System (INIS)

    Gorman, J.A.

    1990-04-01

    A meeting on ''PWSCC Remedial Measures'' was organized to give those working in this area an opportunity to share their results, ideas and plans with regard to development and application of remedial measures directed against the primary water stress corrosion cracking (PWSCC) phenomenon affecting alloy 600 steam generator tubes. Topics discussed included: utility experience and strategies; nondestructive examination (NDE) methods for PWSCC; technical topics ranging from predictive methods for occurrence of PWSCC to results of corrosion tests; and services provided by vendors that can help prevent the occurrence of PWSCC or can help address problems caused by PWSCC once it occurs

  6. XHM-1 alloy as a promising structural material for water-cooled fusion reactor components

    International Nuclear Information System (INIS)

    Solonin, M.I.; Alekseev, A.B.; Kazennov, Yu.I.; Khramtsov, V.F.; Kondrat'ev, V.P.; Krasina, T.A.; Rechitsky, V.N.; Stepankov, V.N.; Votinov, S.N.

    1996-01-01

    Experience gained in utilizing austenitic stainless steel components in water-cooled power reactors indicates that the main cause of their failure is the steel's propensity for corrosion cracking. In search of a material immune to this type of corrosion, different types of austenitic steels and chromium-nickel alloys were investigated and tested at VNIINM. This paper presents the results of studying physical and mechanical properties, irradiation and corrosion resistance in a water coolant at <350 C of the alloy XHM-1 as compared with austenitic stainless steels 00Cr16Ni15Mo3Nb, 00Cr20Ni25Nb and alloy 00Cr20Ni40Mo5Nb. Analysis of the results shows that, as distinct from the stainless steels studied, the XHM-1 alloy is completely immune to corrosion cracking (CC). Not a single induced damage was encountered within 50 to 350 C in water containing different amounts of chlorides and oxygen under tensile stresses up to the yield strength of the material. One more distinctive feature of the alloy compared to steels is that no change in the strength or total elongation is encountered in the alloy specimens irradiated to 32 dpa at 350 C. The XHM-1 alloy has adequate fabricability and high weldability characteristics. As far as its properties are concerned, the XHM-1 alloy is very promising as a material for water-cooled fusion reactor components. (orig.)

  7. Oxidation of Alloy 82 in nominal PWR primary water at 340 deg. C and in hydrogenated steam at 400 deg. C

    International Nuclear Information System (INIS)

    Chaumun, Elizabeth; Guerre Catherine; Duhamel, Cecilie; Sennour, Mohamed; Curieres, Ian-de

    2012-09-01

    Nickel-base weld metals are susceptible to stress corrosion cracking (SCC) in Pressurized Water Reactor (PWR) primary water. As tests in laboratory need to last, in some cases, at least several thousand hours to get stress corrosion crack initiation or propagation in simulated primary water, pure hydrogenated steam at 400 deg. C was used to perform accelerated tests. To confirm that these conditions are still representative of primary water conditions, results of oxidation tests of coupons in hydrogenated steam at 400 deg. C and in primary water at 340 deg. C have been compared. Surface oxide layers have been characterized in order to discuss the influence of the temperature and of the media (water or steam). (authors)

  8. Laser-induced damage thresholds of gold, silver and their alloys in air and water

    Energy Technology Data Exchange (ETDEWEB)

    Starinskiy, Sergey V.; Shukhov, Yuri G.; Bulgakov, Alexander V., E-mail: bulgakov@itp.nsc.ru

    2017-02-28

    Highlights: • Laser damage thresholds of Ag, Au and Ag-Au alloys in air and water are measured. • Alloy thresholds are lower than those of Ag and Au due to low thermal conductivity. • Laser damage thresholds in water are ∼1.5 times higher than those in air. • Light scattering mechanisms responsible for high thresholds in water are suggested. • Light scattering mechanisms are supported by optical reflectance measurements. - Abstract: The nanosecond-laser-induced damage thresholds of gold, silver and gold-silver alloys of various compositions in air and water have been measured for single-shot irradiation conditions. The experimental results are analyzed theoretically by solving the heat flow equation for the samples irradiated in air and in water taking into account vapor nucleation at the solid-water interface. The damage thresholds of Au-Ag alloys are systematically lower than those for pure metals, both in air and water that is explained by lower thermal conductivities of the alloys. The thresholds measured in air agree well with the calculated melting thresholds for all samples. The damage thresholds in water are found to be considerably higher, by a factor of ∼1.5, than the corresponding thresholds in air. This cannot be explained, in the framework of the used model, neither by the conductive heat transfer to water nor by the vapor pressure effect. Possible reasons for the high damage thresholds in water such as scattering of the incident laser light by the vapor-liquid interface and the critical opalescence in the superheated water are suggested. Optical pump-probe measurements have been performed to study the reflectance dynamics of the surface irradiated in air and water. Comparison of the transient reflectance signal with the calculated nucleation dynamics provides evidence that the both suggested scattering mechanisms are likely to occur during metal ablation in water.

  9. Adsorption of methanol, ethanol and water on well-characterized PtSn surface alloys

    Science.gov (United States)

    Panja, Chameli; Saliba, Najat; Koel, Bruce E.

    1998-01-01

    Adsorption and desorption of methanol (CH 3OH), ethanol (C 2H 5OH) and water on Pt(111) and two, ordered, PtSn alloys has been studied primarily using temperature-programmed desorption (TPD) mass spectroscopy. The two alloys studied were the {p(2 × 2) Sn}/{Pt(111) } and (√3 × √3) R30° {Sn}/{Pt(111) } surface alloys prepared by vapor deposition of Sn on Pt(111), with θSn = 0.25 and 0.33, respectively. All three molecules are weakly bonded and reversibly adsorbed under UHV conditions on all three surfaces, molecularly desorbing during TPD without any decomposition. The two PtSn surface alloys were found to chemisorb both methanol and ethanol slightly more weakly than on the Pt(111) surface. The desorption activation energies measured by TPD, and hence the adsorption energies, of both methanol and ethanol progressively decrease as the surface concentration of Sn increases, compared with Pt(111). The decreased binding energy leads one to expect a lower reactivity for these alcohols on the two alloys. The sticking coefficients and the monolayer coverages of these alcohols on the two alloys were identical to that on Pt(111) at 100 K, independent of the amount of Sn present in the surface layer. Alloying Sn in Pt(111) also slightly weakens the adsorption energy of water. Water clusters are formed even at low coverages on all three surfaces, eventually forming a water bilayer prior to the formation of a condensed ice phase. These results are relevant to a molecular-level explanation for the reactivity of Sn-promoted Pt surfaces that have been used in the electro-oxidation of simple organic molecules.

  10. Electroless nickel-plating for the PWSCC mitigation of nickel-base alloys in nuclear power plants

    International Nuclear Information System (INIS)

    Kim, Ji Hyun; Hwang, Il Soon

    2008-01-01

    The feasibility study has been performed as an effort to apply the electroless nickel-plating method for a proposed countermeasure to mitigate primary water stress corrosion cracking (PWSCC) of nickel-base alloys in nuclear power plants. In order to understand the corrosion behavior of nickel-plating at high temperature water, the electrochemical properties of electroless nickel-plated alloy 600 specimens exposed to simulated pressurized water reactor (PWR) primary water were experimentally characterized in high temperature and high pressure water condition. And, the resistance to the flow accelerated corrosion (FAC) test was investigated to check the durability of plated layers in high-velocity water-flowing environment at high temperature. The plated surfaces were examined by using both scanning electron microscopy (SEM) and energy dispersive spectroscopy (EDS) after exposures to the condition. From this study, it is found that the corrosion resistance of electroless nickel-plated Alloy 600 is higher than that of electrolytic plating in 290 deg. C water

  11. Removal of chromium (VI) from water by micro-alloyed aluminium ...

    African Journals Online (AJOL)

    This paper deals with Cr(VI) ion removal from water, by micro-alloyed aluminium composite (MAlC), under flow conditions. In a water environment the MAlC acts as a strong reducing agent. Dissolving it in water is accompanied by the generation of Al(III) ions and reduction of water to H2, with OH- ions. The final product is ...

  12. Effects of Cr and Nb contents on the susceptibility of Alloy 600 type Ni-base alloys to stress-corrosion cracking in a simulated BWR environment

    International Nuclear Information System (INIS)

    Akashi, Masatsune

    1995-01-01

    In order to discuss the effects of chromium and niobium contents on the susceptibility of Alloy 600 type nickel-base alloys to stress-corrosion cracking in the BWR primary coolant environment, a series of creviced bent-beam (CBB) tests were conducted in a high-temperature, high-purity water environment. Chromium, niobium, and titanium as alloying elements improved the resistivity to stress-corrosion cracking, whereas carbon enhanced the susceptibility to it. Alloy-chemistry-based correlations have been defined to predict the relative resistances of alloys to stress-corrosion cracking. A strong correlation was found, for several heats of alloys, between grain-boundary chromium depletion and the susceptibility to stress-corrosion cracking

  13. Standard specification for nuclear-grade silver-indium-cadmium alloy

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2003-01-01

    1.1 This specification covers silver-indium-cadmium alloy for use as a control material in light-water nuclear reactors. 1.2 The scope of this specification excludes the use of this material in applications where material strength of this alloy is a prime requisite. Also, this material must be protected from the primary water by a corrosion and wear resistant cladding. 1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.

  14. 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

  15. Effect of alloying elements on solidification of primary austenite in Ni-Mn-Cu cast iron

    Directory of Open Access Journals (Sweden)

    A. Janus

    2011-04-01

    Full Text Available Within the research, determined were direction and intensity of alloying elements influence on solidification way (directional orvolumetric of primary austenite dendrites in hypoeutectic austenitic cast iron Ni-Mn-Cu. 50 cast shafts dia. 20 mm were analysed.Chemical composition of the alloy was as follows: 1.7 to 3.3 % C, 1.4 to 3.1 % Si, 2.8 to 9.9 % Ni, 0.4 to 7.7 % Mn, 0 to 4.6 % Cu, 0.14 to0.16 % P and 0.03 to 0.04 % S. The discriminant analysis revealed that carbon influences solidification of primary austenite dendrites most intensively. It clearly increases the tendency to volumetric solidification. Influence of the other elements is much weaker. This means that the solidification way of primary austenite dendrites in hypoeutectic austenitic cast iron Ni-Mn-Cu does not differ from that in an unalloyed cast iron.

  16. The effect lead impurities on the corrosion resistance of alloy 600 and alloy 690 in high temperature water

    International Nuclear Information System (INIS)

    Sakai, T.; Nakagomi, N.; Kikuchi, T.; Aoki, K.; Nakayasu, F.; Yamakawa, K.

    1998-01-01

    Degradation of nickel-based alloy steam generator (SG) tubing caused by lead-induced corrosion has been reported recently in some PWR plants. Several laboratory studies also have shown that lead causes intergranular or transgranular stress corrosion cracking (IGSCC or TGSCC) of the tubing materials. Information from previous studies suggests two possible explanations for the mechanism of lead-induced corrosion. One is selective dissolution of tube metal elements, resulting in formation of a lead-containing nickel-depleted oxide film as observed in mildly acidic environments. The other explanation is an increase in potential, as has been observed in lead-contaminated caustic environments, although not in all volatile treatment (AVT) water such as the ammonium-hydrazine water chemistry. These observation suggest that an electrochemical reaction between metal elements and dissolved lead might be the cause of lead-induced corrosion. The present work was undertaken to clarify the lead-induced corrosion mechanism of nickel-based alloys from an electrochemical viewpoint, focusing on mildly acidic and basic environments. These are the probable pH conditions in the crevice region between the tube and tube support plate of the SG where corrosion damage could occur. Measurements of corrosion potential and electrochemical polarization of nickel-based alloys were performed to investigate the effect of lead on electrochemical behavior of the alloys. Then, constant extension rate tests (CERT) were carried out to determine the corrosion susceptibility of the alloys in a lead-contaminated environment. (J.P.N.)

  17. Crack growth rate in the HAZ of alloy 690TT/152

    International Nuclear Information System (INIS)

    Gomez-Briceno, D.; Lapena, J.; Garcia-Redondo, M.; Castro, L.; Perosanz, F.J.; Ahluwalia, K.; Hickling, J.

    2011-01-01

    Crack growth rate (CGR) experiments to obtain data for the HAZ of nickel base alloys using fracture mechanics specimens are a challenge, primarily due to the difficulties of positioning the tip of the notch (or pre-crack) in the desired location within the complex region adjacent to the fusion line that is altered in several ways by the welding process. This paper describes an experimental program carried out to determine the CGR in the HAZ of an Alloy 690 test weld made using Alloy 152. Compact tension (CT) specimens have been tested in simulated PWR primary water at temperatures of 340 and 360 C under cyclic and constant loading (both with and without periodic partial unloading). For the Alloy 690 HAZ tested here, transgranular crack propagation (primarily due to environmentally assisted fatigue) with isolated intergranular secondary cracks was observed and there was no increase of the crack growth rate in comparison with that for Alloy 690 base metal. In both cases, the CGR values at constant load were very low (4*10 -9 mm/s down to effectively zero) and generally comparable with the data found in the literature for intergranular cracking of thermally treated or solution annealed Alloy 690 in simulated primary water. The scarce CGR data for the HAZ of Alloy 690 available to date do not suggest a significant increase in the PWSCC susceptibility of this resistant alloy, but further testing is still required given the expected variability in actual production welds. (authors)

  18. Stress corrosion cracking of uranium--niobium alloys

    International Nuclear Information System (INIS)

    Magnani, N.J.

    1978-03-01

    The stress corrosion cracking behavior of U-2 1 / 4 , 4 1 / 2 , 6 and 8 wt % Nb alloys was evaluated in laboratory air and in aqueous Cl - solutions. Thresholds for crack propagation were obtained in these environments. The data showed that Cl - solutions are more deleterious than air environments. Tests were also conducted in pure gases to identify the species in the air responsible for cracking. These data showed the primary stress corrodent is water vapor for the most reactive alloy, U-2 1 / 4 % Nb, while O 2 is primarily responsible for cracking in the more corrosion resistant alloys, U-6 and 8% Nb. The 4 1 / 2 % alloy was found to be susceptible in both H 2 O and O 2 environments

  19. Evaluating Primary Dendrite Trunk Diameters in Directionally Solidified Al-Si Alloys

    Science.gov (United States)

    Grugel, R. N.; Tewari, S. N.; Poirier, D. R.

    2014-01-01

    The primary dendrite trunk diameters of Al-Si alloys that were directionally solidified over a range of processing conditions have been measured. These data are analyzed with a model based primarily on an assessment of secondary dendrite arm dissolution in the mushy zone. Good fit with the experimental data is seen and it is suggested that the primary dendrite trunk diameter is a useful metric that correlates well with the actual solidification processing parameters. These results are placed in context with the limited results from the aluminium - 7 wt. % silicon samples directionally solidified aboard the International Space Station as part of the MICAST project.

  20. Mechanism study of c.f.c Fe-Ni-Cr alloy corrosion in supercritical water

    International Nuclear Information System (INIS)

    Payet, M.

    2011-01-01

    Supercritical water can be use as a high pressure coolant in order to improve the thermodynamic efficiency of power plants. For nuclear concept, lifetime is an important safety parameter for materials. Thus materials selection criteria concern high temperature yield stress, creep resistance, resistance to irradiation embrittlement and also to both uniform corrosion and stress corrosion cracking.This study aims for supplying a new insight on uniform corrosion mechanism of Fe-Ni-Cr f.c.c. alloys in deaerated supercritical water at 600 C and 25 MPa. Corrosion tests were performed on 316L and 690 alloys as sample autoclaves taking into account the effect of surface finishes. Morphologies, compositions and crystallographic structure of the oxides were determined using FEG scanning electron microscopy, glow discharge spectroscopy and X-ray diffraction. If supercritical water is expected to have a gas-like behaviour in the test conditions, the results show a significant dissolution of the alloy species. Thus the corrosion in supercritical water can be considered similar to corrosion in under-critical water assuming the higher temperature and its effect on the solid state diffusion. For alloy 690, the protective oxide layer formed on polished surface consists of a chromia film topped with an iron and nickel mixed chromite or spinel. The double oxide layer formed on 316L steel seems less protective with an outer porous layer of magnetite and an inhomogeneous Cr-rich inner layer. For each alloy, the study of the inner protective scale growth mechanisms by marker or tracer experiments reveals that diffusion in the oxide scale is governed by an anionic process. However, surface finishes impact deeply the growth mechanisms. Comparisons between the results for the steel suggest that there is a competition between the oxidation of iron and chromium in supercritical water. Sufficient available chromium is required in order to form a thin oxide layer. Highly deformed or ultra fine

  1. Stress Corrosion Cracking of alloy 600 in high temperature water: a study of mechanisms

    International Nuclear Information System (INIS)

    Boursier, J.M.; Bouvier, O. de; Gras, J.M.; Noel, D.; Vaillant, F.; Rios, R.

    1992-12-01

    Investigations of the stress corrosion cracking behaviour of Alloy 600 tubing in high temperature water were performed in order to get a precise knowledge of the different stages of the cracking and their dependence on various parameters. The compatibility of the results with the main mechanisms to be considered was examined. Results showed three stages in the cracking: a true incubation time, a slow-rate propagation period followed by a rapid-propagation stage. Tests separating stress and strain rate contributions show that the strain rate is the main parameter which controls the crack propagation. The hydrogen overpressure was found to increase the crack growth rate up to 1-4 bar, but a strong decrease is observed from 4 to 20 bar. Analysis of the hydrogen ingress in the metal showed that it is neither correlated to the hydrogen overpressure nor to the severity of cracking; so cracking resulting from an hydrogen-model is unlikely. No detrimental effect of oxygen (4 bar) was noticed both in the mill-annealed and the sensitized conditions. Finally, none of the classical mechanisms, neither hydrogen-assisted cracking nor slip-step dissolution, can correctly describe the observed behaviour. Some fractographic examinations, and an influence of primary water on the creep rate of Alloy 600, lead to consider that other recent mechanisms, involving an interaction between dissolution and plasticity, have to be considered

  2. Effect of heat treatment conditions on stress corrosion cracking resistance of alloy X-750 in high temperature water

    International Nuclear Information System (INIS)

    Yonezawa, Toshio; Onimura, Kichiro; Sakamoto, Naruo; Sasaguri, Nobuya; Susukida, Hiroshi; Nakata, Hidenori.

    1984-01-01

    In order to improve the resistance of the Alloy X-750 in high temperature and high purity water, the authors investigated the influence of heat treatment condition on the stress corrosion cracking resistance of the alloy. This paper describes results of the stress corrosion cracking test and some discussion on the mechanism of the stress corrosion cracking of Alloy X-750 in deaerated high temperature water. The following results were obtained. (1) The stress corrosion cracking resistance of Alloy X-750 in deaerated high temperature water remarkably depended upon the heat treatment condition. The materials solution heat treated and aged within temperature ranges from 1065 to 1100 0 C and from 704 to 732 0 C, respectively, have a good resistance to the stress corrosion cracking in deaerated high temperature water. Especially, water cooling after the solution heat treatment gives an excellent resistance to the stress corrosion cracking in deaerated high temperature water. (2) Any correlations were not observed between the stress corrosion cracking susceptibility of Alloy X-750 in deaerated high temperature water and grain boundary chromium depleted zones, precipitate free zones and the grain boundary segregation of impurity elements and so on. It appears that there are good correlations between the stress corrosion cracking resistance of the alloy in the environment and the kinds, morphology and coherency of precipitates along the grain boundaries. (author)

  3. Irradiation assisted stress corrosion cracking of HTH Alloy X-750 and Alloy 625

    International Nuclear Information System (INIS)

    Mills, W.J.; Lebo, M.R.; Bajaj, R.; Kearns, J.J.; Hoffman, R.C.; Korinko, J.J.

    1994-01-01

    In-reactor testing of bolt-loaded precracked compact tension specimens was performed in 360 degree C water to determine effect of irradiation on the SCC behavior of HTH Alloy X-750 and direct aged Alloy 625. Out-of-flux and autoclave control specimens provided baseline data. Primary test variables were stress intensity factor, fluence, chemistry, processing history, prestrain. Results for the first series of experiments were presented at a previous conference. Data from two more recent experiments are compared with previous results; they confirm that high irradiation levels significantly reduce SCC resistance in HTH Alloy X-750. Heat-to-heat differences in IASCC were related to differences in boron content, with low boron heats showing improved SCC resistance. The in-reactor SCC performance of Alloy 625 was superior to that for Alloy X-750, as no cracking was observed in any Alloy 625 specimens even though they were tested at very high K 1 and fluence levels. A preliminary SCC usage model developed for Alloy X-750 indicates that in-reactor creep processes, which relax stresses but also increase crack tip strain rates, and radiolysis effects accelerate SCC. Hence, in-reactor SCC damage under high flux conditions may be more severe than that associated with postirradiation tests. In addition, preliminary mechanism studies were performed to determine the cause of IASCC In Alloy X-750

  4. Substitution of cobalt alloying in PWR primary circuit gate valves

    International Nuclear Information System (INIS)

    Cachon, L.; Sudreau, F.; Brunel, L.

    1995-01-01

    The object of this study is qualify cobalt-free alternative alloys for valve applications. This paper focus on tribological characterization of numerous coatings is done by using the first one, of a classical type. Then tests are performed with the second one which simulates solicitations supported by gate valves in primary circuit of PWR. 35% Ni-Cr - 65% Cr 3 C 2 coating, deposited by detonation gun technology, gives us hope to find a substitute of Stelite 6. (author). 5 refs., 16 figs., 2 tabs

  5. Corrosion Inhibition Study of Al-Cu-Ni Alloy in Simulated Sea-Water ...

    African Journals Online (AJOL)

    A study on the inhibition of Al-Cu-Ni alloy in simulated sea-water environment was investigated using Sodium Chromate as inhibitor. The inhibitor concentration was varied as control, 0.25, 0.5, 1.0, 1.5 and 2.0 Molar. Al-Cu-Ni alloy was sand cast into cylindrical bars of 20 mm x 300 mm dimension. The corrosion of the ...

  6. Study of superficial films and of electrochemical behaviour of some nickel base alloys and titanium base alloys in solution representation of granitic, argillaceous and salted ground waters

    International Nuclear Information System (INIS)

    Quang, K.V.; Da Cunha Belo, M.; Benabed, M.S.; Bourelier, F.; Jallerat, N.; Pari, F.L.

    1985-01-01

    The corrosion behaviour of the stainless steels 304, 316 Ti, 25Cr-20Ni-Mo-Ti, nickel base alloys Hastelloy C4, Inconel 625, Incoloy 800, Ti and Ti-0.2% Pd alloy has been studied in the aerated or deaerated solutions at 20 0 C and 90 0 C whose compositions are representative of interstitial ground waters: granitic or clay waters or salt brine. The electrochemical techniques used are voltametry, polarization resistance and complexe impedance measurements. Electrochemical data show the respective influence of the parameters such as temperature, solution composition and dissolved oxygen, addition of soluble species chloride, fluoride, sulfide and carbonates, on which depend the corrosion current density, the passivation and the pitting potential. The inhibition efficiency of carbonate and bicarbonate activities against pitting corrosion is determined. In clay water at 90 0 C, Ti and Ti-Pd show very high passivation aptitude and a broad passive potential range. Alloying Pd increases cathodic overpotential and also transpassive potential. It makes the alloy less sensitive to the temperature effect. Optical Glow Discharge Spectra show three parts in the composition depth profiles of surface films on alloys. XPS and SIMS spectrometry analyses are also carried out. Electron microscopy observation shows that passive films formed on Ti and Ti-Pd alloy have amorphous structure. Analysis of the alloy constituents dissolved in solutions, by radioactivation in neutrons, gives the order of magnitude of the Ni base alloy corrosion rates in various media. It also points out the preferential dissolution of alloying iron and in certain cases of chromium

  7. Suitability of Co as an alloy material for components of the primary circuit of HTR reactors

    International Nuclear Information System (INIS)

    Iniotakis, N.

    1977-02-01

    For high temperature reactors it is of interest if Co-alloys could be used for the different components of the primary cooling circuit. It has been investigated in detail to what amount the Co-60 created by neutron activation of Co-59 contained in the material of the components could possibly contribute to the contamination of the primary cooling circuit of the reactor. The result of these investigations is compared with the contamination of the cooling circuit by fission and activation products like Co-137, Cs-134, Ag-11om etc. For pebble bed reactors with an OTTO-type fuel management it could be shown that there is no limitation for the use of cobalt in alloys for materials of the components in the primary cooling circuit. The only boundary condition is that the local Thermal Flux at the position of the components should be less than phisub(th) 7 n/cm 2 . sec. (orig.) [de

  8. IGSCC growth behaviors of Alloy 690 in hydrogenated high temperature water

    Energy Technology Data Exchange (ETDEWEB)

    Arioka, K.; Yamada, T.; Miyamoto, T.; Terachi, T. [INSS, (Japan)

    2011-07-01

    The rate of growth of stress corrosion cracking (SCC) was measured for cold worked and thermally treated and solution treated Alloy 690 (UNS N06690, CW TT690, CW ST690) in hydrogenated pressurized water reactor (PWR) primary water under static load condition. Three important patterns were observed: First, Intergranular stress corrosion cracking (IGSCC) was observed on both TT and ST690 even in static load condition if materials were heavily cold worked although the rate of SCC growth was much slower than that of CW mill annealed Alloy 600. Furthermore much rapid SCC growth was recognized in 20% CW TT690 than that of 20% CW ST690. This is quite different result in the literature in high temperature caustic solution. Second, in order to assess the role of creep, rates of creep crack growth were measured in air, argon, and hydrogen gas environments using 20% CW TT690, and 20% CW MA600 in the range of temperatures between 360 and 460 C; intergranular creep cracking (IG creep cracking) was observed on the test materials even in air. Similar slope of 1/T-type temperature dependencies on IGSCC and IG creep crack growth were observed on 20% CW TT690. Similar fracture morphologies and similar 1/T-type temperature dependencies suggest that creep is important in the growth of IGSCC of CW TT690 in high temperature water. Third, cavities and pores were observed at grain boundaries near tips of SCC and creep although the size of the cavities and pores of SCC were much smaller than that of creep cracks. Also the population and size of cavities seem to decrease with decreasing test temperature. These results suggest that the difference in the size and population of cavities might be related with the difference in crack growth rate. And the cavities seem to be formed result from collapse of vacancies at grain boundaries as the crack embryo. This result suggests that diffusion of condensation of vacancies in high stressed fields occurs in high temperature water and gas environments

  9. Laser-induced damage thresholds of gold, silver and their alloys in air and water

    Science.gov (United States)

    Starinskiy, Sergey V.; Shukhov, Yuri G.; Bulgakov, Alexander V.

    2017-02-01

    The nanosecond-laser-induced damage thresholds of gold, silver and gold-silver alloys of various compositions in air and water have been measured for single-shot irradiation conditions. The experimental results are analyzed theoretically by solving the heat flow equation for the samples irradiated in air and in water taking into account vapor nucleation at the solid-water interface. The damage thresholds of Au-Ag alloys are systematically lower than those for pure metals, both in air and water that is explained by lower thermal conductivities of the alloys. The thresholds measured in air agree well with the calculated melting thresholds for all samples. The damage thresholds in water are found to be considerably higher, by a factor of ∼1.5, than the corresponding thresholds in air. This cannot be explained, in the framework of the used model, neither by the conductive heat transfer to water nor by the vapor pressure effect. Possible reasons for the high damage thresholds in water such as scattering of the incident laser light by the vapor-liquid interface and the critical opalescence in the superheated water are suggested. Optical pump-probe measurements have been performed to study the reflectance dynamics of the surface irradiated in air and water. Comparison of the transient reflectance signal with the calculated nucleation dynamics provides evidence that the both suggested scattering mechanisms are likely to occur during metal ablation in water.

  10. Grain boundary selective oxidation and intergranular stress corrosion crack growth of high-purity nickel binary alloys in high-temperature hydrogenated water

    Energy Technology Data Exchange (ETDEWEB)

    Bruemmer, S. M.; Olszta, M. J.; Toloczko, M. B.; Schreiber, D. K.

    2018-02-01

    The effects of alloying elements in Ni-5at%X binary alloys on intergranular (IG) corrosion and stress corrosion cracking (SCC) have been assessed in 300-360°C hydrogenated water at the Ni/NiO stability line. Alloys with Cr or Al additions exhibited grain boundary oxidation and IGSCC, while localized degradation was not observed for pure Ni, Ni-Cu or Ni-Fe alloys. Environment-enhanced crack growth was determined by comparing the response in water and N2 gas. Results demonstrate that selective grain boundary oxidation of Cr and Al promoted IGSCC of these Ni alloys in hydrogenated water.

  11. 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

  12. 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

  13. Effect of temperature and dissolved hydrogen on oxide films formed on Ni and Alloy 182 in simulated PWR water

    International Nuclear Information System (INIS)

    Mendonça, R.; Bosch, R.-W.; Van Renterghem, W.; Vankeerberghen, M.; Araújo Figueiredo, C. de

    2016-01-01

    Alloy 182 is a nickel-based weld metal, which is susceptible to stress corrosion cracking in PWR primary water. It shows a peak in SCC susceptibility at a certain temperature and hydrogen concentration. This peak is related to the electrochemical condition where the Ni to NiO transition takes place. One hypothesis is that the oxide layer at this condition is not properly developed and so the material is not optimally protected against SCC. Therefore the oxide layer formed on Alloy 182 is investigated as a function of the dissolved hydrogen concentration and temperature around this Ni/NiO transition. Exposure tests were performed with Alloy 182 and Ni coupons in a PWR environment at temperatures between 300 °C and 345 °C and dissolved hydrogen concentration between 5 and 35 cc (STP)H 2 /kg. Post-test analysis of the formed oxide layers were carried out by SEM, EDS and XPS. The exposure tests with Ni coupons showed that the Ni/NiO transition curve is at a higher temperature than the curve based on thermodynamic calculations. The exposure tests with Alloy 182 showed that oxide layers were present at all temperatures, but that the morphology changed from spinel crystals to needle like oxides when the Ni/NiO transition curve was approached. Oxide layers were present below the Ni/NiO transition curve i.e. when the Ni coupon was still free of oxides. In addition an evolved slip dissolution model was proposed that could explain the observed experimental results and the peak in SCC susceptibility for Ni-based alloys around the Ni/NiO transition. - Highlights: • Exposure tests with Ni-coupons showed that the Ni/NiO transition curve shifted to more oxidizing conditions. • The Ni specimens tested in PWR water were free of oxides at all temperatures. • The exposure tests with Alloy 182 showed that oxide layers were present at all temperatures. • The Alloy 182 surface morphology changed from spinel crystals to needle like oxides when the Ni/NiO curve was approached

  14. Effect of temperature and dissolved hydrogen on oxide films formed on Ni and Alloy 182 in simulated PWR water

    Energy Technology Data Exchange (ETDEWEB)

    Mendonça, R. [CAPES Foundation, Ministry of Education, Brasilia (Brazil); Bosch, R.-W., E-mail: rbosch@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Van Renterghem, W.; Vankeerberghen, M. [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Araújo Figueiredo, C. de [CDTN/CNEN, Av. Antônio Carlos 6627, 31270-901 Belo Horizonte, MG (Brazil)

    2016-08-15

    Alloy 182 is a nickel-based weld metal, which is susceptible to stress corrosion cracking in PWR primary water. It shows a peak in SCC susceptibility at a certain temperature and hydrogen concentration. This peak is related to the electrochemical condition where the Ni to NiO transition takes place. One hypothesis is that the oxide layer at this condition is not properly developed and so the material is not optimally protected against SCC. Therefore the oxide layer formed on Alloy 182 is investigated as a function of the dissolved hydrogen concentration and temperature around this Ni/NiO transition. Exposure tests were performed with Alloy 182 and Ni coupons in a PWR environment at temperatures between 300 °C and 345 °C and dissolved hydrogen concentration between 5 and 35 cc (STP)H{sub 2}/kg. Post-test analysis of the formed oxide layers were carried out by SEM, EDS and XPS. The exposure tests with Ni coupons showed that the Ni/NiO transition curve is at a higher temperature than the curve based on thermodynamic calculations. The exposure tests with Alloy 182 showed that oxide layers were present at all temperatures, but that the morphology changed from spinel crystals to needle like oxides when the Ni/NiO transition curve was approached. Oxide layers were present below the Ni/NiO transition curve i.e. when the Ni coupon was still free of oxides. In addition an evolved slip dissolution model was proposed that could explain the observed experimental results and the peak in SCC susceptibility for Ni-based alloys around the Ni/NiO transition. - Highlights: • Exposure tests with Ni-coupons showed that the Ni/NiO transition curve shifted to more oxidizing conditions. • The Ni specimens tested in PWR water were free of oxides at all temperatures. • The exposure tests with Alloy 182 showed that oxide layers were present at all temperatures. • The Alloy 182 surface morphology changed from spinel crystals to needle like oxides when the Ni/NiO curve was

  15. Low cycle fatigue behaviors of low alloy steels in 310 .deg. C deoxygenated water

    International Nuclear Information System (INIS)

    Jang, Hun

    2008-02-01

    After low cycle fatigue tests of SA508 Gr.1a low alloy steel in 310 .deg. C deoxygenated water, the fatigue surface and the sectioned area of specimens were observed to understand the effect of the cyclic strain rate on the environmentally assisted cracking behaviors. From the fatigue crack morphologies of the specimen tested at a strain rate of 0.008 %/s, unclear ductile striations and blunt crack tip were observed. So, metal dissolution could be the main cracking mechanism of the material at the strain rate. On the other hand, on the fatigue surface of the specimen tested at strain rates of 0.04 and 0.4 %/s, the brittle cracks and the flat facets, which are the evidence of the hydrogen induced cracking, were observed. Also, the tendency of linkage between the main crack and micro-cracks was observed on the sectioned area. Therefore, the main cracking mechanism at the strain rates of 0.04 and 0.4 %/s could be the hydrogen induced cracking. Additionally, the evidence of the dissolved MnS inclusions was observed on the fatigue surface from energy dispersive x-ray spectrometer analyses. So, despite of the low sulfur content of the test material, the sulfides seem to contribute to environmentally assisted cracking of SA508 Gr.1a low alloy steel in 310 .deg. C deoxygenated water. Additionally, our experimental fatigue life data of SA508 Gr.1a low alloy steel (heat A) showed a consistent difference with statistical model produced in argon national laboratory. So, additional low cycle fatigue tests of other heat SA508 Gr.1a (heat B) and SA508 Gr.3 low alloy steels were performed to investigate the effect of material variability on fatigue behaviors of low alloy steels in 310 .deg. C deoxygenated water. In results, the fatigue lives of three low alloy steels were increased following order: SA508 Gr.1a low alloy steel - heat A, SA508 Gr.3 low alloy steel, and SA508 Gr.1a low alloy steel - heat B. From microstructure observation, the fatigue surface of SA508 Gr.1a low alloy

  16. Spheroidization of primary Mg{sub 2}Si in Al-20Mg{sub 2}Si-4.5Cu alloy modified with Ca and Sb during T6 heat treatment process

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Hong-Chen; Wang, Hui-Yuan, E-mail: wanghuiyuan@jlu.edu.cn; Chen, Lei; Zha, Min, E-mail: minzha@jlu.edu.cn; Wang, Cheng; Li, Chao; Jiang, Qi-Chuan

    2017-02-08

    The morphology evolution of primary Mg{sub 2}Si particles in a Al-20Mg{sub 2}Si-4.5Cu alloy both unmodified and modified with 0.5 wt% Ca-Sb prepared by hot-extrusion followed by T6 heat treatment was investigated in the present study. Interestingly, we found that the combination of hot-extrusion and T6 heat treatment was efficient in transforming truncated octahedral primary Mg{sub 2}Si into sphere in the modified alloy. In contrast, the primary Mg{sub 2}Si particles still kept dentritic in the unmodified alloy. It suggested that the formation of truncated octahedral primary Mg{sub 2}Si particles in as-cast state, the fragmentation of particles by hot-extrusion and the enhanced solid-state diffusion of Si and/or Mg atoms during heat treatment were responsible for the spheroidization of primary Mg{sub 2}Si. Moreover, the existence of fine (~10–20 µm) spherical primary Mg{sub 2}Si played an important role in strengthening the alloy, i.e., the ultimate tensile strength (UTS) increased from ~227 MPa in the unmodified alloy to ~303 MPa in the modified one. It is because the fine spherical primary Mg{sub 2}Si particles can provide a higher fracture stress and strength of the matrix/particle interface. Our study offered a simple methodology to prepare spherical primary Mg{sub 2}Si in an Al-high Mg{sub 2}Si alloy, which is beneficial to design novel light-weight Al-Mg-Si alloys with improved mechanical properties.

  17. Assessment of effects of Fort St. Vrain HTGR primary coolant on Alloy 800. Final report

    International Nuclear Information System (INIS)

    Trester, P.W.; Johnson, W.R.; Simnad, M.T.; Burnette, R.D.; Roberts, D.I.

    1982-08-01

    A comprehensive review was conducted of primary helium coolant chemistry data, based on current and past operating histories of helium-cooled, high-temperature reactors (HTGRs), including the Fort St. Vrain (FSV) HTGR. A reference observed FSV reactor coolant environment was identified. Further, a slightly drier expected FSV coolant chemistry was predicted for reactor operation at 100% of full power. The expected environment was compared with helium test environments used in the US, United Kingdom, Germany, France, and Japan. Based on a comprehensive review and analysis of mechanical property data reported for Alloy 800 tested in controlled-impurity helium environments (and in air when appropriate for comparison), an assessment was made of the effect of FSV expected helium chemistry on material properties of alloy 800, with emphasis on design properties of the Alloy 800 material utilized in the FSV steam generators

  18. The Degradation Interface of Magnesium Based Alloys in Direct Contact with Human Primary Osteoblast Cells.

    Science.gov (United States)

    Ahmad Agha, Nezha; Willumeit-Römer, Regine; Laipple, Daniel; Luthringer, Bérengère; Feyerabend, Frank

    2016-01-01

    Magnesium alloys have been identified as a new generation material of orthopaedic implants. In vitro setups mimicking physiological conditions are promising for material / degradation analysis prior to in vivo studies however the direct influence of cell on the degradation mechanism has never been investigated. For the first time, the direct, active, influence of human primary osteoblasts on magnesium-based materials (pure magnesium, Mg-2Ag and Mg-10Gd alloys) is studied for up to 14 days. Several parameters such as composition of the degradation interface (directly beneath the cells) are analysed with a scanning electron microscope equipped with energy dispersive X-ray and focused ion beam. Furthermore, influence of the materials on cell metabolism is examined via different parameters like active mineralisation process. The results are highlighting the influences of the selected alloying element on the initial cells metabolic activity.

  19. In-situ electrochemical study of Zr1nb alloy corrosion in high temperature Li{sup +} containing water

    Energy Technology Data Exchange (ETDEWEB)

    Krausová, Aneta [University of Chemistry and Technology, Technická 3, 166 28 Prague 6 (Czech Republic); Macák, Jan, E-mail: macakj@vscht.cz [University of Chemistry and Technology, Technická 3, 166 28 Prague 6 (Czech Republic); Sajdl, Petr [University of Chemistry and Technology, Technická 3, 166 28 Prague 6 (Czech Republic); Novotný, Radek [JRC-IET, Westerduinveg 3, 1755 LE Petten (Netherlands); Renčiuková, Veronika [University of Chemistry and Technology, Technická 3, 166 28 Prague 6 (Czech Republic); Vrtílková, Věra [ÚJP a.s., Nad Kamínkou 1345, 156 10 Prague 5 (Czech Republic)

    2015-12-15

    Long-term in-situ corrosion tests were performed in order to evaluate the influence of lithium ions on the corrosion of zirconium alloy. Experiments were carried out in a high-pressure high-temperature loop (280 °C, 8 MPa) in a high concentration water solution of LiOH (70 and 200 ppm Li{sup +}) and in a simulated WWER primary coolant environment. The kinetic parameters characterising the oxidation process have been explored using in-situ electrochemical impedance spectroscopy and slow potentiodynamic polarization. Also, a suitable equivalent circuit was suggested, which would approximate the impedance characteristics of the corrosion of Zr–1Nb alloy. The Mott–Schottky approach was used to determine the semiconducting character of the passive film. - Highlights: • Zr1Nb alloy was tested in WWER coolant and in LiOH solutions at 280 °C. • Corrosion rates were estimated in-situ from electrochemical data. • Electrochemical data agreed well with weight gains and metallography data. • Increase of corrosion rate in LiOH appeared after short exposure (300–500 h). • Very high donor densities (1.1–1.2 × 10{sup 20} cm{sup −3}) of Zr oxide grown in LiOH were found.

  20. Water hammer characteristics of integral pressurized water reactor primary loop

    International Nuclear Information System (INIS)

    Zuo, Qiaolin; Qiu, Suizheng; Lu, Wei; Tian, Wenxi; Su, Guanghui; Xiao, Zejun

    2013-01-01

    Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions

  1. Water hammer characteristics of integral pressurized water reactor primary loop

    Energy Technology Data Exchange (ETDEWEB)

    Zuo, Qiaolin [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Qiu, Suizheng, E-mail: szqiu@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Lu, Wei; Tian, Wenxi; Su, Guanghui [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Xiao, Zejun [Nuclear Power Institute of China, Chengdu, Sichuan 610041 (China)

    2013-08-15

    Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions.

  2. Influence of Localized Plasticity on IASCC Sensitivity of Austenitic Stainless Steels under PWR Primary Water

    Science.gov (United States)

    Cissé, Sarata; Tanguy, Benoit; Laffont, Lydia; Lafont, Marie-Christine; Guerre, Catherine; Andrieu, Eric

    The sensibility of precipitation-strengthened A286 austenitic stainless steel to Stress Corrosion Cracking (SCC) is studied by means of Slow Strain Rate Tests (SSRT). First, alloy cold working by Low Cycle Fatigue (LCF) is investigated. Fatigue tests under plastic strain control are performed at different strain levels (Δ ɛp/2=0.2%, 0.5% and 0.8%) in order to establish correlation between stress softening and deformation microstructure resulting from LCF tests. Deformed microstructures have been identified through TEM investigations. Three states of cyclic behaviour for precipitation-strengthened A286 have been identified: hardening, cyclic softening and finally saturation of softening. It is shown that the A286 alloy cyclic softening is due to microstructural features such as defects — free deformation bands resulting from dislocations motion along family plans , that swept defects or γ' precipitates and lead to deformation localization. In order to quantify effects of plastic localized deformation on intergranular stress corrosion cracking (IGSCC) of the A286 alloy in PWR primary water, slow strain rate tests are conducted. For each cycling conditions, two specimens at a similar stress level are tested: the first containing free precipitate deformation bands, the other not significant of a localized deformation state. SSRT tests are still in progress.

  3. Corrosion testing of NiCrAl(Y) coating alloys in high-temperature and supercritical water

    International Nuclear Information System (INIS)

    Biljan, S.; Huang, X.; Qian, Y.; Guzonas, D.

    2011-01-01

    With the development of Generation IV (Gen IV) nuclear power reactors, materials capable of operating in high-temperature and supercritical water environment are essential. This study focuses on the corrosion behavior of five alloys with compositions of Ni20Cr, Ni5Al, Ni50Cr, Ni20Cr5Al and Ni20Cr10AlY above and below the critical point of water. Corrosion tests were conducted at three different pressures, while the temperature was maintained at 460 o C, in order to examine the effects of water density on the corrosion. From the preliminary test results, it was found that the binary alloys Ni20Cr and Ni50Cr showed weight loss above the critical point (23.7 MPa and 460 o C). The higher Cr content alloy Ni50Cr suffered more weight loss than Ni-20Cr under the same conditions. Accelerated weight gain was observed above the critical point for the binary alloy Ni5Al. The combination of Cr, Al and Y in Ni20Cr10AlY provides stable scale formation under all testing conditions employed in this study. (author)

  4. Impact Fretting Wear Behavior of Alloy 690 Tubes in Dry and Deionized Water Conditions

    Institute of Scientific and Technical Information of China (English)

    Zhen-Bing Cai; Jin-Fang Peng; Hao Qian; Li-Chen Tang; Min-Hao Zhu

    2017-01-01

    The impact fretting wear has largely occurred at nuclear power device induced by the flow-induced vibration,and it will take potential hazards to the service of the equipment.However,the present study focuses on the tangential fretting wear of alloy 690 tubes.Research on impact fretting wear of alloy 690 tubes is limited and the related research is imminent.Therefore,impact fretting wear behavior of alloy 690 tubes against 304 stainless steels is investigated.Deionized water is used to simulate the flow environment of the equipment,and the dry environment is used for comparison.Varied analytical techniques are employed to characterize the wear and tribochemical behavior during impact fretting wear.Characterization results indicate that cracks occur at high impact load in both water and dry equipment;however,the water as a medium can significantly delay the cracking time.The crack propagation behavior shows a jagged shape in the water,but crack extended disorderly in dry equipment because the water changed the stress distribution and retarded the friction heat during the wear process.The SEM and XPS analysis shows that the main failure mechanisms of the tube under impact fretting are fatigue wear and friction oxidation.The effect of medium(water) on fretting wear is revealed,which plays a potential and promising role in the service of nuclear power device and other flow equipments.

  5. Role of hydrogen in the intergranular cracking mechanism by stress corrosion in primary medium of nickel based alloys 600 and 690

    International Nuclear Information System (INIS)

    Odemer, G.; Coudurier, A.; Jambon, F.; Chene, J.; Odemer, G.; Coudurier, A.; Chene, J.

    2007-01-01

    The aim of this work is to characterize the sensitivity to hydrogen embrittlement of alloys 600 and 690 in order to better understand the eventual role of hydrogen in the stress corrosion mechanism which affects these alloys when they are exposed in PWR primary medium. (O.M.)

  6. Pacific Northwest National Laboratory Investigation of the Stress Corrosion Cracking in Nickel-Base Alloys, Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Bruemmer, Stephen M.; Toloczko, Mychailo B.; Olszta, Matthew J.

    2012-03-01

    The objective of this program is to evaluate the primary water stress corrosion cracking (PWSCC) susceptibility of high chromium alloy 690 and its weld metals, establish quantitative measurements of crack-growth rates and determine relationships among cracking susceptibility, environmental conditions and metallurgical characteristics. Stress-corrosion, crack-growth rates have been determined for 12 alloy 690 specimens, 11 alloy 152/52/52M weld metal specimens, 4 alloy 52M/182 overlay specimens and 2 alloy 52M/82 inlay specimens in simulated PWR primary water environments. The alloy 690 test materials included three different heats of extruded control-rod-drive mechanism (CRDM) tubing with variations in the initial material condition and degree of cold work for one heat. Two cold-rolled (CR) alloy 690 plate heats were also obtained and evaluated enabling comparisons to the CR CRDM materials. Weld metal, overlay and inlay specimens were machined from industry mock ups to provide plant-representative materials for testing. Specimens have been tested for one alloy 152 weld, two alloy 52 welds and three alloy 52M welds. The overlay and inlay specimens were prepared to propagate stress-corrosion cracks from the alloy 182 or 82 material into the more resistant alloy 52M. In all cases, crack extension was monitored in situ by direct current potential drop (DCPD) with length resolution of about +1 µm making it possible to measure extremely low growth rates approaching 5x10-10 mm/s. Most SCC tests were performed at 325-360°C with hydrogen concentrations from 11-29 cc/kg; however, environmental conditions were modified during a few experiments to evaluate the influence of temperature, water chemistry or electrochemical potential on propagation rates. In addition, low-temperature (~50°C) cracking behavior was examined for selected alloy 690 and weld metal specimens. Extensive characterizations have been performed on material microstructures and stress-corrosion cracks by

  7. Relationship between stress corrosion cracking and low frequency fatigue-corrosion of alloy 600 in PWR primary water

    International Nuclear Information System (INIS)

    Bosch, C.

    1998-01-01

    Stress corrosion cracking of PWR vessel head adapters is a main problem for nuclear industry. With the aim to better understand the influence of the mechanical parameters on the cracking phenomena (by stress corrosion (SCC) or fatigue corrosion (FC)) of alloy 600 exposed to primary PWR coolant, a parametrical study has been carried out. Crack propagation tests on CT test specimens have been implemented under static loads (stress corrosion tests) or low frequency cyclic loads (fatigue corrosion tests). Results (frequency influence, type of cycles, ratio charge on velocities and propagation modes of cracks) have allowed to characterize the transition domain between the crack phenomena of SCC and FC. With the obtained results, it has been possible too to differentiate the effects due to environmental factors and the effects due to mechanical factors. At last, a quantitative fractographic study and the observations of the microstructure at the tip of crack have led to a better understanding of the transitions of the crack propagation mode between the SCC and the FC. (O.M.)

  8. EPRI PWR primary water chemistry guidelines revision

    International Nuclear Information System (INIS)

    McElrath, Joel; Fruzzetti, Keith

    2014-01-01

    EPRI periodically updates the PWR Primary Water Chemistry Guidelines as new information becomes available and as required by NEI 97-06 (Steam Generator Program Guidelines) and NEI 03-08 (Guideline for the Management of Materials Issues). The last revision of the PWR water chemistry guidelines identified an optimum primary water chemistry program based on then-current understanding of research and field information. This new revision provides further details with regard to primary water stress corrosion cracking (PWSCC), fuel integrity, and shutdown dose rates. A committee of industry experts, including utility specialists, nuclear steam supply system (NSSS) and fuel vendor representatives, Institute of Nuclear Power Operations (INPO) representatives, consultants, and EPRI staff collaborated in reviewing the available data on primary water chemistry, reactor water coolant system materials issues, fuel integrity and performance issues, and radiation dose rate issues. From the data, the committee updated the water chemistry guidelines that all PWR nuclear plants should adopt. The committee revised guidance with regard to optimization to reflect industry experience gained since the publication of Revision 6. Among the changes, the technical information regarding the impact of zinc injection on PWSCC initiation and dose rate reduction has been updated to reflect the current level of knowledge within the industry. Similarly, industry experience with elevated lithium concentrations with regard to fuel performance and radiation dose rates has been updated to reflect data collected to date. Recognizing that each nuclear plant owner has a unique set of design, operating, and corporate concerns, the guidelines committee has retained a method for plant-specific optimization. Revision 7 of the Pressurized Water Reactor Primary Water Chemistry Guidelines provides guidance for PWR primary systems of all manufacture and design. The guidelines continue to emphasize plant

  9. The Degradation Interface of Magnesium Based Alloys in Direct Contact with Human Primary Osteoblast Cells.

    Directory of Open Access Journals (Sweden)

    Nezha Ahmad Agha

    Full Text Available Magnesium alloys have been identified as a new generation material of orthopaedic implants. In vitro setups mimicking physiological conditions are promising for material / degradation analysis prior to in vivo studies however the direct influence of cell on the degradation mechanism has never been investigated. For the first time, the direct, active, influence of human primary osteoblasts on magnesium-based materials (pure magnesium, Mg-2Ag and Mg-10Gd alloys is studied for up to 14 days. Several parameters such as composition of the degradation interface (directly beneath the cells are analysed with a scanning electron microscope equipped with energy dispersive X-ray and focused ion beam. Furthermore, influence of the materials on cell metabolism is examined via different parameters like active mineralisation process. The results are highlighting the influences of the selected alloying element on the initial cells metabolic activity.

  10. The corrosion resistance of Zr-Nb and Zr-Nb-Sn alloys in high-temperature water and steam

    International Nuclear Information System (INIS)

    Dalgaard, S.B.

    1960-03-01

    An alloy of reactor-grade sponge zirconium-2.5 wt. % niobium was exposed to water and steam at high temperature. The corrosion was twice that of Zircaloy-2 while hydrogen pickup was found to be equal to that of Zircaloy-2. Ternary additions of tin to this alloy in the range 0.5-1.5 had no effect on the corrosion resistance in water at 315 o C up to 100 days. At higher temperatures, tin increased the corrosion, the effect varying with temperature. Heat treatment of the alloys was shown to affect corrosion resistance. (author)

  11. The corrosion resistance of Zr-Nb and Zr-Nb-Sn alloys in high-temperature water and steam

    Energy Technology Data Exchange (ETDEWEB)

    Dalgaard, S B

    1960-03-15

    An alloy of reactor-grade sponge zirconium-2.5 wt. % niobium was exposed to water and steam at high temperature. The corrosion was twice that of Zircaloy-2 while hydrogen pickup was found to be equal to that of Zircaloy-2. Ternary additions of tin to this alloy in the range 0.5-1.5 had no effect on the corrosion resistance in water at 315{sup o}C up to 100 days. At higher temperatures, tin increased the corrosion, the effect varying with temperature. Heat treatment of the alloys was shown to affect corrosion resistance. (author)

  12. The effect of temperature on primary defect formation in Ni–Fe alloy

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Chengbin, E-mail: wangchengbin@sinap.ac.cn [Key Laboratory of Interfacial Physics and Technology, Chinese Academy of Sciences, Shanghai 201800 (China); Zhang, Wei; Ren, Cuilan [Key Laboratory of Interfacial Physics and Technology, Chinese Academy of Sciences, Shanghai 201800 (China); Huai, Ping [Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800 (China); Zhu, Zhiyuan [Key Laboratory of Interfacial Physics and Technology, Chinese Academy of Sciences, Shanghai 201800 (China)

    2014-02-15

    Molecular dynamics (MD) simulations have been used to study the influence of temperature on defect generation and evolution in nickel and Ni–Fe alloy (with 15% and 50% Fe content) with a 10-keV primary knock-on atom (PKA) at six different temperatures from 0 to 1500 K. The recently available Ni–Fe potential is used with its repulsive part modified by Vörtler. The temporal evolution and temperature dependence of stable defect formation and in-cascade clustering processes are analysed. The number of stable defect and the interstitial clustering fraction are found to increase with temperature whereas the vacancy clustering fraction decreases with temperature. The alloy composition dependence of the stable defect number is also found for the PKA energy considered here. Additionally, a study of the temperature influence on the cluster size distribution is performed, revealing a systematic change in the cluster size distributions, with higher temperature cascades producing larger interstitial clusters.

  13. Primary and secondary creep in aluminum alloys as a solid state transformation

    Science.gov (United States)

    Fernández, R.; Bruno, G.; González-Doncel, G.

    2016-08-01

    Despite the massive literature and the efforts devoted to understand the creep behavior of aluminum alloys, a full description of this phenomenon on the basis of microstructural parameters and experimental conditions is, at present, still missing. The analysis of creep is typically carried out in terms of the so-called steady or secondary creep regime. The present work offers an alternative view of the creep behavior based on the Orowan dislocation dynamics. Our approach considers primary and secondary creep together as solid state isothermal transformations, similar to recrystallization or precipitation phenomena. In this frame, it is shown that the Johnson-Mehl-Avrami-Kolmogorov equation, typically used to analyze these transformations, can also be employed to explain creep deformation. The description is fully compatible with present (empirical) models of steady state creep. We used creep curves of commercially pure Al and ingot AA6061 alloy at different temperatures and stresses to validate the proposed model.

  14. PWSCC Preventive Maintenance Activities for Alloy 600 in Japanese PWR Plants

    International Nuclear Information System (INIS)

    Yamamoto, K.; Sugimoto, N.; Onishi, K.; Okimura, K.

    2012-01-01

    Because many nuclear plants have been in operation for ages, the importance of preventive maintenance technologies is getting higher. One conspicuous problem found in pressurized water reactor (PWR) plants is the primary water stress corrosion cracking (PWSCC) observed in Alloy 600 (a kind of high nickel based alloy) parts. Alloy 600 was used for butt welds between low alloy steel and stainless steel of nozzles of Reactor Vessel (RV), Steam Generator (SG), and Pressurizer (Pz). As PWSCC occurred at these parts may cause Loss of Coolant Accident (LOCA), preventive maintenance is necessary. PWSCC is considered to be caused by a mixture of three elements: high residual tensile stress on surface, material (Alloy 600) and environment. PWSCC can be prevented by improving one of the elements. MHI has been developing stress improvement methods, for example, Water Jet Peening (WJP), Shot Peening by Ultrasonic vibration (USP), and Laser Stress Improvement Process (L-SIP). According to the situation, appropriate method is applied for each part. WJP has been applied for RV nozzles of a lot of plants in Japan. However PWSCC was observed in RV nozzles during the inspection before WJP in recent years, MHI developed the Advanced INLAY system to improve the material from Alloy 600 to Alloy 690. Alloy 600 on the inner surface of the nozzles is removed and welding with Alloy 690 is performed. In addition, heat treatments for the nozzles are difficult for its structural situation, so ambient temperature temper bead welding technique for RV nozzles was developed to make the heat treatments unnecessary. This paper describes countermeasures against PWSCC and introduces the maintenance activities performed in Japan. (author)

  15. Corrosion of aluminum alloys as a function of alloy composition

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.

    1969-10-01

    A study was initiated which included nineteen aluminum alloys. Tests were conducted in high purity water at 360 0 C and flow tests (approx. 20 ft/sec) in reactor process water at 130 0 C (TF-18 loop tests). High-silicon alloys and AlSi failed completely in the 360 0 C tests. However, coupling of AlSi to 8001 aluminum suppressed the failure. The alloy compositions containing iron and nickel survived tht 360 0 C autoclave exposures. Corrosion rates varied widely as a function of alloy composition, but in directions which were predictable from previous high-temperature autoclave experience. In the TF-18 loop flow tests, corrosion penetrations were similar on all of the alloys and on high-purity aluminum after 105 days. However, certain alloys established relatively low linear corrosion rates: Al-0.9 Ni-0.5 Fe-0.1 Zr, Al-1.0 Ni-0.15 Fe-11.5 Si-0.8 Mg, Al-1.2 Ni-1.8 Fe, and Al-7.0 Ni-4.8 Fe. Electrical polarity measurements between AlSi and 8001 alloys in reactor process water at temperatures up to 150 0 C indicated that AlSi was anodic to 8001 in the static autoclave system above approx. 50 0 C

  16. Study of corrosion of aluminium alloys of nuclear purity in ordinary water, пart one

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2004-01-01

    Full Text Available Effects of corrosion of aluminum alloys of nuclear purity in ordinary water of the spent fuel storage pool of the RA research reactor at VINČA Institute of Nuclear Sciences has been examined in the frame work of the International Atomic Energy Agency Coordinated Research Project "Corrosion of Research Reactor Aluminum-Clad Spent Fuel in Water" since 2002. The study presented in this paper comprises activities on determination and monitoring of chemical parameters and radio activity of water and sludge in the RA spent fuel storage pool and results of the initial study of corrosion effects obtained by visual examinations of surfaces of various coupons made of aluminum alloys of nuclear purity of the test racks exposed to the pool water for a period from six months to six years.

  17. Apatite Formation and Biocompatibility of a Low Young's Modulus Ti-Nb-Sn Alloy Treated with Anodic Oxidation and Hot Water.

    Directory of Open Access Journals (Sweden)

    Hidetatsu Tanaka

    Full Text Available Ti-6Al-4V alloy is widely prevalent as a material for orthopaedic implants because of its good corrosion resistance and biocompatibility. However, the discrepancy in Young's modulus between metal prosthesis and human cortical bone sometimes induces clinical problems, thigh pain and bone atrophy due to stress shielding. We designed a Ti-Nb-Sn alloy with a low Young's modulus to address problems of stress disproportion. In this study, we assessed effects of anodic oxidation with or without hot water treatment on the bone-bonding characteristics of a Ti-Nb-Sn alloy. We examined surface analyses and apatite formation by SEM micrographs, XPS and XRD analyses. We also evaluated biocompatibility in experimental animal models by measuring failure loads with a pull-out test and by quantitative histomorphometric analyses. By SEM, abundant apatite formation was observed on the surface of Ti-Nb-Sn alloy discs treated with anodic oxidation and hot water after incubation in Hank's solution. A strong peak of apatite formation was detected on the surface using XRD analyses. XPS analysis revealed an increase of the H2O fraction in O 1s XPS. Results of the pull-out test showed that the failure loads of Ti-Nb-Sn alloy rods treated with anodic oxidation and hot water was greater than those of untreated rods. Quantitative histomorphometric analyses indicated that anodic oxidation and hot water treatment induced higher new bone formation around the rods. Our findings indicate that Ti-Nb-Sn alloy treated with anodic oxidation and hot water showed greater capacity for apatite formation, stronger bone bonding and higher biocompatibility for osteosynthesis. Ti-Nb-Sn alloy treated with anodic oxidation and hot water treatment is a promising material for orthopaedic implants enabling higher osteosynthesis and lower stress disproportion.

  18. Kinetics of passivation of a nickel-base alloy in high temperature water

    Energy Technology Data Exchange (ETDEWEB)

    Machet, A. [Laboratoire de Physico-Chimie des Surfaces, CNRS-ENSCP (UMR 7045), Ecole Nationale Superieure de Chimie de Paris, Universite Pierre et Marie Curie, F-75231 Paris cedex 05 (France)]|[Framatome ANP, Tour AREVA, F-92084 Paris-la-Defense (France); Galtayries, A.; Zanna, S.; Marcus, P. [Laboratoire de Physico-Chimie des Surfaces, CNRS-ENSCP (UMR 7045), Ecole Nationale Superieure de Chimie de Paris, Universite Pierre et Marie Curie, F-75231 Paris cedex 05 (France); Jolivet, P.; Scott, P. [Framatome ANP, Tour AREVA, F-92084 Paris-la-Defense (France); Foucault, M.; Combrade, P. [Framatome ANP, Centre Technique, F-71205 Le Creusot (France)

    2004-07-01

    The kinetics of passivation and the composition of the surface oxide layer, in high temperature and high pressure water, of a nickel-chromium-iron alloy (Alloy 600) have been investigated by X-ray Photoelectron Spectroscopy (XPS). The samples have been exposed for short (0.4 - 8.2 min) and longer (0 - 400 hours) time periods to high temperature (325 deg. C) and high pressure water (containing boron and lithium) under controlled hydrogen pressure. The experiments were performed in two types of autoclaves: a novel autoclave dedicated to short time periods and a classic static autoclave for the longer exposures. In the initial stage of passivation, a continuous ultra-thin layer of chromium oxide (Cr{sub 2}O{sub 3}) is rapidly formed on the surface with an external layer of chromium hydroxide. For longer times of passivation, the oxide layer is in a duplex form with an internal chromium oxide layer and an external layer of nickel hydroxide. The growth of the internal Cr{sub 2}O{sub 3} oxide layer has been fitted by three classical models (parabolic, logarithmic and inverse logarithmic laws) for the short passivation times, and the growth curves have been extrapolated to longer passivation periods. The comparison with the experimental results reveals that the kinetics of passivation of Alloy 600 in high temperature and high pressure water, for passivation times up to 400 hours, is well fitted by a logarithmic growth law. (authors)

  19. Kinetics of passivation of a nickel-base alloy in high temperature water

    International Nuclear Information System (INIS)

    Machet, A.; Galtayries, A.; Zanna, S.; Marcus, P.; Jolivet, P.; Scott, P.; Foucault, M.; Combrade, P.

    2004-01-01

    The kinetics of passivation and the composition of the surface oxide layer, in high temperature and high pressure water, of a nickel-chromium-iron alloy (Alloy 600) have been investigated by X-ray Photoelectron Spectroscopy (XPS). The samples have been exposed for short (0.4 - 8.2 min) and longer (0 - 400 hours) time periods to high temperature (325 deg. C) and high pressure water (containing boron and lithium) under controlled hydrogen pressure. The experiments were performed in two types of autoclaves: a novel autoclave dedicated to short time periods and a classic static autoclave for the longer exposures. In the initial stage of passivation, a continuous ultra-thin layer of chromium oxide (Cr 2 O 3 ) is rapidly formed on the surface with an external layer of chromium hydroxide. For longer times of passivation, the oxide layer is in a duplex form with an internal chromium oxide layer and an external layer of nickel hydroxide. The growth of the internal Cr 2 O 3 oxide layer has been fitted by three classical models (parabolic, logarithmic and inverse logarithmic laws) for the short passivation times, and the growth curves have been extrapolated to longer passivation periods. The comparison with the experimental results reveals that the kinetics of passivation of Alloy 600 in high temperature and high pressure water, for passivation times up to 400 hours, is well fitted by a logarithmic growth law. (authors)

  20. Preparations and properties of anti-corrosion additives of water-soluble metal working fluids for aluminum alloy materials.

    Science.gov (United States)

    Watanabe, Shoji

    2008-01-01

    This short review describes various types of anti-corrosion additives of water-soluble metal working fluids for aluminum alloy materials. It is concerned with synthetic additives classified according to their functional groups; silicone compounds, carboxylic acids and dibasic acids, esters, Diels-Alder adducts, various polymers, nitrogen compounds, phosphoric esters, phosphonic acids, and others. Testing methods for water-soluble metal working fluids for aluminum alloy materials are described for a practical application in a laboratory.

  1. Corrosion mechanism of a Ni-based alloy in supercritical water: Impact of surface plastic deformation

    International Nuclear Information System (INIS)

    Payet, Mickaël; Marchetti, Loïc; Tabarant, Michel; Chevalier, Jean-Pierre

    2015-01-01

    Highlights: • The dissolution of Ni and Fe cations occurs during corrosion of Ni-based alloys in SCW. • The nature of the oxide layer depends locally on the alloy microstructure. • The corrosion mechanism changes when cold-work increases leading to internal oxidation. - Abstract: Ni–Fe–Cr alloys are expected to be a candidate material for the generation IV nuclear reactors that use supercritical water at temperatures up to 600 °C and pressures of 25 MPa. The corrosion resistance of Alloy 690 in these extreme conditions was studied considering the surface finish of the alloy. The oxide scale could suffer from dissolution or from internal oxidation. The presence of a work-hardened zone reveals the competition between the selective oxidation of chromium with respect to the oxidation of nickel and iron. Finally, corrosion mechanisms for Ni based alloys are proposed considering the effects of plastically deformed surfaces and the dissolution.

  2. 3D microstructural evolution of primary recrystallization and grain growth in cold rolled single-phase aluminum alloys

    Science.gov (United States)

    Adam, Khaled; Zöllner, Dana; Field, David P.

    2018-04-01

    Modeling the microstructural evolution during recrystallization is a powerful tool for the profound understanding of alloy behavior and for use in optimizing engineering properties through annealing. In particular, the mechanical properties of metallic alloys are highly dependent upon evolved microstructure and texture from the softening process. In the present work, a Monte Carlo (MC) Potts model was used to model the primary recrystallization and grain growth in cold rolled single-phase Al alloy. The microstructural representation of two kinds of dislocation densities, statistically stored dislocations and geometrically necessary dislocations were quantified based on the ViscoPlastic Fast Fourier transform method. This representation was then introduced into the MC Potts model to identify the favorable sites for nucleation where orientation gradients and entanglements of dislocations are high. Additionally, in situ observations of non-isothermal microstructure evolution for single-phase aluminum alloy 1100 were made to validate the simulation. The influence of the texture inhomogeneity is analyzed from a theoretical point of view using an orientation distribution function for deformed and evolved texture.

  3. Copper alloys disintegration using pulsating water jet

    Czech Academy of Sciences Publication Activity Database

    Lehocká, D.; Klich, Jiří; Foldyna, Josef; Hloch, Sergej; Królczyk, J. B.; Cárach, J.; Krolczyk, G.

    2016-01-01

    Roč. 82, March 2016 (2016), s. 375-383 ISSN 0263-2241 R&D Projects: GA MŠk(CZ) LO1406; GA MŠk ED2.1.00/03.0082 Institutional support: RVO:68145535 Keywords : pulsating water jet * generation of pulses * disintegration * surface morphology * copper alloys Subject RIV: JQ - Machines ; Tools Impact factor: 2.359, year: 2016 http://ac.els-cdn.com/S0263224116000154/1-s2.0-S0263224116000154-main.pdf?_tid=8f8d1de6-99e9-11e6-afbc-00000aacb362&acdnat=1477314089_59912e52847e91e2030d6a1afd09e7b2

  4. Parametric tests of the effects of water chemistry impurities on corrosion of Zr-alloys under simulated BWR condition

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, S; Ito, K [Nippon Nuclear Fuel Development Co. Ltd., Oarai, Ibaraki (Japan); Lin, C C [GE Nucklear Energy (United States); Cheng, B [Electric Power Research Inst. (United States); Ikeda, T [Toshiba Corp. (Japan); Oguma, M [Hitachi, Ltd (Japan); Takei, T [Tokyo Electric Power Co., Inc. (Japan); Vitanza, C; Karlsen, T M [Institutt for Energiteknikk, Halden (Norway). OECD Halden Reaktor Projekt

    1997-02-01

    The Halden BWR corrosion test loop was constructed to evaluate the impact of water chemistry variables, heat flux and boiling condition on corrosion performance of Zr-alloys in a simulated BWR environment. The loop consists of two in-core rigs, one for testing fuel rod segments and the other for evaluating water chemistry variables utilizing four miniautoclaves. Ten coupon specimens are enclosed in each miniautoclave. The Zr-alloys for the test include Zircaloy-2 having different nodular corrosion resistance and five new alloys. The first and second of the six irradiation tests planned in this program were completed. Post-irradiation examination of those test specimens have shown that the test loop is capable of producing nodular corrosion on the fuel rod cladding tested under the reference chemistry condition. The miniautoclave tests showed that nodular corrosion could be formed without flux and boiling under some water chemistry conditions and the new alloys, generally, had higher corrosion resistance than the Zircaloy in high oxygen environments. (author). 5 refs, 4 figs, 5 tabs.

  5. Corrosion mechanisms of aluminum alloys in waters of low conductivity

    International Nuclear Information System (INIS)

    Haddad, Roberto E; Lanazani, Liliana; Rodriguez, Sebastian

    2006-01-01

    After completing their burn cycle, nuclear fuels in experimental reactors made with aluminum alloys have to remain for long periods in distilled water, in interim storage. While aluminum alloys are resistant to corrosion in pure water, severe deterioration occurs in elements that have been immersed for periods of up to 30 years. Pitting-like surface alterations can even occur in nuclear quality waters (conductivity below 5 μS/cm and dissolved ions content below detection thresholds) in time periods of less than one year. An important factor that could become a potential promoter of this phenomena is the presence of dust particles and others, that could settle on the metallic surface, generating a locally aggressive medium. A simple immersion experiment demonstrates that these points can become initiation sites for pitting with very low concentrations of chlorides (under 10 ppm), especially if the electrochemical potential is increased by contact with another metallic material, even staying below the pitting potential in this medium. There are several corrosion mechanisms acting simultaneously, depending on the nature of the deposits. Pitting under glass particles has been detected, which may be related to a simple crevice corrosion process. In the case of iron oxides, however, the results depend on the type of oxide. Pits more than 100 microns deep have been obtained in 7 day immersion tests, so in spent fuel storage sites these mechanisms could easily cause penetration of the 500 micron aluminum plates during the time covering the interim storage under water, which could be decades, with similar chemical conditions (CW)

  6. Apatite Formation and Biocompatibility of a Low Young’s Modulus Ti-Nb-Sn Alloy Treated with Anodic Oxidation and Hot Water

    Science.gov (United States)

    Tanaka, Hidetatsu; Mori, Yu; Noro, Atsushi; Kogure, Atsushi; Kamimura, Masayuki; Yamada, Norikazu; Hanada, Shuji; Masahashi, Naoya; Itoi, Eiji

    2016-01-01

    Ti-6Al-4V alloy is widely prevalent as a material for orthopaedic implants because of its good corrosion resistance and biocompatibility. However, the discrepancy in Young’s modulus between metal prosthesis and human cortical bone sometimes induces clinical problems, thigh pain and bone atrophy due to stress shielding. We designed a Ti-Nb-Sn alloy with a low Young’s modulus to address problems of stress disproportion. In this study, we assessed effects of anodic oxidation with or without hot water treatment on the bone-bonding characteristics of a Ti-Nb-Sn alloy. We examined surface analyses and apatite formation by SEM micrographs, XPS and XRD analyses. We also evaluated biocompatibility in experimental animal models by measuring failure loads with a pull-out test and by quantitative histomorphometric analyses. By SEM, abundant apatite formation was observed on the surface of Ti-Nb-Sn alloy discs treated with anodic oxidation and hot water after incubation in Hank’s solution. A strong peak of apatite formation was detected on the surface using XRD analyses. XPS analysis revealed an increase of the H2O fraction in O 1s XPS. Results of the pull-out test showed that the failure loads of Ti-Nb-Sn alloy rods treated with anodic oxidation and hot water was greater than those of untreated rods. Quantitative histomorphometric analyses indicated that anodic oxidation and hot water treatment induced higher new bone formation around the rods. Our findings indicate that Ti-Nb-Sn alloy treated with anodic oxidation and hot water showed greater capacity for apatite formation, stronger bone bonding and higher biocompatibility for osteosynthesis. Ti-Nb-Sn alloy treated with anodic oxidation and hot water treatment is a promising material for orthopaedic implants enabling higher osteosynthesis and lower stress disproportion. PMID:26914329

  7. Characteristics of Film Formed on Alloy 600 and Alloy 690 in Water Containing lead

    International Nuclear Information System (INIS)

    Hwang Seong Sik; Lee, Deok Hyun; Kim, Hong Pyo; Kim, Joung Soo; Kim, Ju Yup

    1999-01-01

    Anodic polarization behaviors of Alloy 600 and Alloy 690 have been studied as a function of lead content in the solution of pH 4 and 10 at 90 .deg. C. As the amount of lead in the solution increased, critical current densities and passive current densities of Alloy 600 and Alloy 690 increased, while the breakdown potential of the alloys decreased. The high critical current density in the high lead solution was thought to come from the combination of an enhanced dissolution of constituents on the surface of the alloys by the lead and an anodic dissolution of metallic lead deposited on the surface of the specimens. The morphology of lead precipitated on the specimen after the anodic scan changed with the pH of solution: small irregular particles were precipitated on the surface of the specimen in the solution of pH 4, while the high density of regular sized particles was formed on it in the solution of pH 10.Pb was observed to enhance Cr depletion from the outer surface of Alloy 600 and Alloy 690 and also to increase the ratio of O 2- /OH - in the surface film formed in the high lead solution. The SCC resistance of Alloy 600 and Alloy 690 may have decreased due to the poor quality of the passive film formed and the enhanced oxygen evolution in the solution containing lead

  8. Hydrogen permeation in FeCrAl alloys for LWR cladding application

    Science.gov (United States)

    Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; Snead, Lance L.

    2015-06-01

    FeCrAl, an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In this study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. The total tritium inventory inside the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.

  9. Failure probability analyses for PWSCC in Ni-based alloy welds

    International Nuclear Information System (INIS)

    Udagawa, Makoto; Katsuyama, Jinya; Onizawa, Kunio; Li, Yinsheng

    2015-01-01

    A number of cracks due to primary water stress corrosion cracking (PWSCC) in pressurized water reactors and Ni-based alloy stress corrosion cracking (NiSCC) in boiling water reactors have been detected around Ni-based alloy welds. The causes of crack initiation and growth due to stress corrosion cracking include weld residual stress, operating stress, the materials, and the environment. We have developed the analysis code PASCAL-NP for calculating the failure probability and assessment of the structural integrity of cracked components on the basis of probabilistic fracture mechanics (PFM) considering PWSCC and NiSCC. This PFM analysis code has functions for calculating the incubation time of PWSCC and NiSCC crack initiation, evaluation of crack growth behavior considering certain crack location and orientation patterns, and evaluation of failure behavior near Ni-based alloy welds due to PWSCC and NiSCC in a probabilistic manner. Herein, actual plants affected by PWSCC have been analyzed using PASCAL-NP. Failure probabilities calculated by PASCAL-NP are in reasonable agreement with the detection data. Furthermore, useful knowledge related to leakage due to PWSCC was obtained through parametric studies using this code

  10. Effects of Heat-treatment on the Tensile Properties of Ti-Al-Zr Alloy

    International Nuclear Information System (INIS)

    Kim, Tae Hoon; Kang, Chang Sun; Baek, Jong Hyuk; Choi, Byoung Kwon; Jeong, Yong Hwan

    2006-01-01

    Ti-Al-Zr, titanium alloy, has been well known material as one of the candidates for heat-exchange tubes in steam generators in SMART (System integrated Modular Advanced ReacTor). But the primary circuit with the primary coolant is much different from that of commercial PWRs, i.e., an ammonia is used as a pH raising agent and the heat-exchange tubes are exposed to the primary coolant water at high temperatures and in high-pressure environments. Thus, excellent mechanical properties and corrosion resistance are required for the safe operation during the lifetime. A lot of tests were done to examine the mechanical properties of the Ti-Al-Zr alloy in the room temperature. But the test of this work is done in the more realistic condition from the viewpoint of the system characteristics for SMART design concept. Therefore, the purpose of this study is to evaluate the effects of annealing and cooling rate on the tensile properties of Ti-Al-Zr alloy at the operation temperature

  11. The performance of alloy 625 in the high temperature application of Heavy Water Plants

    International Nuclear Information System (INIS)

    Mitra, J.; Dey, G.K.; Sundararaman, M.; Dubey, J.S.; De, P.K.; Kumar, Niraj

    2006-01-01

    Wrought and centrifugally cast alloy 625 tubes are used in the cracker units of ammonia based Heavy Water Plants (HWP). During the service of about 100,000 h, the ammonia cracker tubes, predictably, have been exposed to temperatures below 600degC to above 765degC and have undergone several hundreds of start-shutdown cycles, producing several ordered phases in the alloy. To understand the effect of the ordered phases on the structure properties, Alloy 625 samples were aged at 540degC, 700degC and 850degC temperatures, for duration up to 1200 h. Results were compared with that of cast and wrought Alloy 625 samples, which aged during the service of 100,000 h and that failed during the service after about 24,000 h along with that of aged samples, which were resolutionised at 1170degC for 2h. (author)

  12. Visualization of steam bubbles with evaporation in molten alloy

    International Nuclear Information System (INIS)

    Nishi, Yoshihisa; Furuya, Masahiro; Kinoshita, Izumi; Takenaka, Nobuyuki; Matsubayashi, Masahito

    1997-01-01

    An innovative Steam Generator concept of Fast Breeder Reactors by using liquid-liquid direct contact heat transfer has been developed. In this concept, the SG shell is filled with a molten alloy heated by primary sodium. Water is fed into the high temperature molten alloy, and evaporates by direct contact heating. In order to obtain the fundamental information to discuss the heat transfer mechanisms of the direct contact between the water and the molten alloy, this phenomenon was visualized by neutron radiography. JRR-3M radiography in Japan Atomic Energy Research Institute was used. Followings are main results. (1) The bubbles with evaporation are risen with vigorous form changing, coalescence and break-up. Because of these vigorous evaporation, this system have the high heat transfer performance. (2) The rising velocities and volumes of bubbles are calculated from pixcel values of images. The velocities of the bubbles with evaporation are about 60 cm/s, which is larger than that of inert gas bubbles in molten alloy (20-40 cm/s). (3) The required heat transfer length of evaporation is calculated from pixcel values of images. The relation between heat transfer length and superheat temperature, obtained through the heat transfer test, is conformed by this calculation. (author)

  13. Corrosion phase formation on container alloys in basalt repository environments

    International Nuclear Information System (INIS)

    Johnston, R.G.; Anantatmula, R.P.; Lutton, J.M.; Rivera, C.L.

    1986-01-01

    The Basalt Waste Isolation Project is evaluating the suitability of basalt in southeastern Washington State as a possible location for a nuclear waste repository. The performance of the waste package, which includes the waste form, container, and surrounding packing material, will be affected by the stability of container alloys in the repository environment. Primary corrosion phases and altered packing material containing metals leached from the container may also influence subsequent reactions between the waste form and repository environment. Copper- and iron-based alloys were tested at 50 0 to 300 0 C in an air/steam environment and in pressure vessels in ground-water-saturated basalt-bentonite packing material. Reaction phases formed on the alloys were identified and corrosion rates were measured. Changes in adhering packing material were also evaluated. The observed reactions and their possible effects on container alloy durability in the repository are discussed

  14. Primary processes during water radiolysis

    International Nuclear Information System (INIS)

    Pikaev, A.K.

    1980-01-01

    Briefly reviewed are investigations of primary process mechanism taking place during radiolysis of water and similar systems, executed by direct and indirect methods. A conclusion is made on the important role of the water structure during radiolysis of aqueous solutions of some substances. A necessity to take account of this factor during consideration of radiolysis theoretical models is pointed out

  15. Zinc-nickel alloy electrodeposits for water electrolysis

    Energy Technology Data Exchange (ETDEWEB)

    Sheela, G.; Pushpavanam, Malathy; Pushpavanam, S. [Central Electrochemical Research Inst., Karaikudi (India)

    2002-06-01

    Electrodeposited zinc-nickel alloys of various compositions were prepared. A suitable electrolyte and conditions to produce alloys of various compositions were identified. Alloys produced on electroformed nickel foils were etched in caustic to leach out zinc and to produce the Raney type, porous electro catalytic surface for hydrogen evolution. The electrodes were examined by polarisation measurements, to evaluate their Tafel parameters, cyclic voltammetry, to test the change in surface properties on repeated cycling, scanning electron microscopy to identify their microstructure and X-ray diffraction. The catalytic activity as well as the life of the electrode produced from 50% zinc alloy was found to be better than others. (Author)

  16. Synthesis and Properties of Water-Soluble Blue-Emitting Mn-Alloyed CdTe Quantum Dots

    Science.gov (United States)

    Tynkevych, Olena; Karavan, Volodymyr; Vorona, Igor; Filonenko, Svitlana; Khalavka, Yuriy

    2018-05-01

    In this work, we prepared CdTe quantum dots, and series of Cd1-xMnxTe-alloyed quantum dots with narrow size distribution by an ion-exchange reaction in water solution. We found that the photoluminescence peaks are shifted to higher energies with the increasing Mn2+ content. So far, this is the first report of blue-emitting CdTe-based quantum dots. By means of cyclic voltammetry, we detected features of electrochemical activity of manganese energy levels formed inside the Cd1-xMnxTe-alloyed quantum dot band gap. This allowed us to estimate their energy position. We also demonstrate paramagnetic behavior for Cd1-xMnxTe-alloyed quantum dots which confirmed the successful ion-exchange reaction.

  17. Synthesis and Properties of Water-Soluble Blue-Emitting Mn-Alloyed CdTe Quantum Dots.

    Science.gov (United States)

    Tynkevych, Olena; Karavan, Volodymyr; Vorona, Igor; Filonenko, Svitlana; Khalavka, Yuriy

    2018-05-02

    In this work, we prepared CdTe quantum dots, and series of Cd 1-x Mn x Te-alloyed quantum dots with narrow size distribution by an ion-exchange reaction in water solution. We found that the photoluminescence peaks are shifted to higher energies with the increasing Mn 2+ content. So far, this is the first report of blue-emitting CdTe-based quantum dots. By means of cyclic voltammetry, we detected features of electrochemical activity of manganese energy levels formed inside the Cd 1-x Mn x Te-alloyed quantum dot band gap. This allowed us to estimate their energy position. We also demonstrate paramagnetic behavior for Cd 1-x Mn x Te-alloyed quantum dots which confirmed the successful ion-exchange reaction.

  18. Effect of technological parameters on formability of semi-solid rheological casting-forging 6061 alloy

    Directory of Open Access Journals (Sweden)

    Jianbo TAN

    2016-02-01

    Full Text Available The 6061 alloy cooling curve is determined by analysis software, and the 6061 semi-solid alloy is prepared by manual paddling process. The primary solid fraction is tested through prepared water quenched samples under different temperature. With H1F100 type servo press and cup type test mold, the forming of the 6061 semi-solid alloy rheological casting-forging is made. The influence of alloy temperature, forming pressure, upper mould temperature and holding time on the formability of 6061 alloy is researched. The results show that within the same set of mold completing casting and forging of the alloy is feasible. Along with the increase of the alloy temperature and the upper mould temperature, the formability of finished products becomes better. Under this experimentation, when the temperature of the semi-solid alloy is amongst 642 ℃ to 645 ℃ and the upper mould preheating temperature is amongst 200 ℃ to 300 ℃, casting defects such as cold insulation will form in the casting-forging sample of semi-solid 6061 alloy with the prolongation of holding time.

  19. Corrosion behavior of alloy 800H (Fe-21Cr-32Ni) in supercritical water

    International Nuclear Information System (INIS)

    Tan, L.; Allen, T.R.; Yang, Y.

    2011-01-01

    Research highlights: → Testing conditions and sample microstructures showed effects on corrosion behaviors. → EBSD and FIB/TEM were used to characterize microstructure in addition to SEM/EDS and XRD. → The formation mechanism of mushroom-shaped oxidation is proposed. → Oxidation thermodynamics and kinetics predict and interpret the corrosion behaviors. - Abstract: The effect of testing conditions (temperature, time, and oxygen content) and material's microstructure (the as-received and the grain boundary engineered conditions) on the corrosion behavior of alloy 800H in high-temperature pressurized water was studied using a variety of characterization techniques. Oxidation was observed as the primary corrosion behavior on the samples. Oxide exfoliation was significantly mitigated on the grain boundary engineered samples compared to the as-received ones. The oxide formation, including some 'mushroom-shaped oxidation', is predicted via a combination of thermodynamics and kinetics influenced by the preferential diffusion of specific species using short-cut diffusion paths.

  20. Photo-Electrochemical Effect of Zinc Addition on the Electrochemical Corrosion Potentials of Stainless Steels and Nickel Alloys in High Temperature Water

    International Nuclear Information System (INIS)

    Lee, Yi-Ching; Fong, Clinton; Fang-Chu, Charles; Chang, Ching

    2012-09-01

    Hydrogen water chemistry (HWC) is one of the main mitigating methods for stress corrosion cracking problem of reactor core stainless steel and nickel based alloy components. Zinc is added to minimize the radiation increase associated with HWC. However, the subsequently formed zinc-containing surface oxides may exhibit p-type semiconducting characteristics. Upon the irradiation of Cherenkov and Gamma ray in the reactor core, the ECP of stainless steels and nickel based alloys may shift in the anodic direction, possibly offsetting the beneficial effect of HWC. This study will evaluate the photo-electrochemical effect of Zinc Water Chemistry on SS304 stainless steel and Alloy 182 nickel based weld metal under simulated irradiated BWR water environments with UV illumination. The experimental results reveal that Alloy 182 nickel-based alloy generally possesses n-type semiconductor characteristics in both oxidizing NWC and reducing HWC conditions with zinc addition. Upon UV irradiation, the ECP of Alloy 182 will shift in the cathodic direction. In most conditions, SS304 will also exhibit n-type semiconducting properties. Only under hydrogen water chemistry, a weak p-type property may emerge. Only a slight upward shift in the anodic direction is detected when SS304 is illuminated with UV light. The potential influence of p-type semiconductor of zinc containing surface oxides is weak and the mitigation effect of HWC on the stress corrosion cracking is not adversely affected. (authors)

  1. Primary Structure and Mechanical Properties of AlSi2 Alloy Continuous Ingots

    Directory of Open Access Journals (Sweden)

    Wróbel T.

    2017-06-01

    Full Text Available The paper presents the research results of horizontal continuous casting of ingots of aluminium alloy containing 2% wt. silicon (AlSi2. Together with the casting velocity (velocity of ingot movement we considered the influence of electromagnetic stirring in the area of the continuous casting mould on refinement of the ingot’s primary structure and their selected mechanical properties, i.e. tensile strength, yield strength, hardness and elongation. The effect of primary structure refinement and mechanical properties obtained by electromagnetic stirring was compared with refinement obtained by using traditional inoculation, which consists in introducing additives, i.e. Ti, B and Sr, to the metal bath. On the basis of the obtained results we confirmed that inoculation done by electromagnetic stirring in the range of the continuous casting mould guarantees improved mechanical properties and also decreases the negative influence of casting velocity, thus increasing the structure of AlSi2 continuous ingots.

  2. On the corrosion behaviour of stainless steel, nickel-chromium and zirconium-alloys in pore water of Portland cement

    International Nuclear Information System (INIS)

    Heitz, E.; Graefen, H.

    1991-12-01

    On the basis of an extensive review of literature and available experience, an evaluation was made of the corrosion of a metallic matrix for radioactive nuclides embedded in porous, water containing Portland cement. As a metallic matrix, austenitic high-alloy steel, nickel-base alloys and zirconium alloys are discussed. Pore waters in Portland cement have low aggressivity. However, through contact with formation water, chloride and sulphate enrichment can occur. Although corrosion is principally possible on the basis of purely thermodynamic considerations, it can be assumed that local corrosion (pitting, stress corrosion cracking, intergranular corrosion) is highly improbable under the given boundary conditions. This is valid for all three groups of alloys and means that only low release rates of corrosion products are to be expected. As a result of the discussion on radiolysis-induced corrosion, additional corrosion activity can be excluded. Final conclusions concerning the stimulation of corrosion processes by microbial action cannot be drawn and, therefore, additional experiments are proposed. The release rates of radioactive products are controlled by a very low dissolution rate of the materials in the passive state. All three groups of alloys show this type of general dissolution. From a survey of literature data it can be concluded that release rates greater than 250 mg/m 2 per day are not exceeded. Since these data were mainly obtained by electrochemical methods, it is proposed that quantitative analytical investigations of the corrosion products in pore water be made. On the whole the release rates determined are far below corrosion rates which are generally technically relevant. (author) 13 figs., 9 tabs., 61 refs

  3. The combined effect of titanic carbide and aluminum phosphide on the refinement of primary silicon in Al-50Si alloy

    Energy Technology Data Exchange (ETDEWEB)

    Dai Hongshang [Key Lab. of Liquid Structure and Heredity of Materials, Ministry of Education, Shandong Univ., Jinan (China); Liu Xiangfa [Key Lab. of Liquid Structure and Heredity of Materials, Ministry of Education, Shandong Univ., Jinan (China); Shandong Binzhou Bohai Piston Co., Ltd., Binzhou, SD (China)

    2008-12-15

    Two refinement methods for Al-50Si alloy are presented in this article: one way is using a newly developed Si-20P alloy at 1573 K: another technique is using the Si-20P alloy in company with Al-TiO{sub 2}-C mixture powder at 1473 K. Compared to the first method, the second one not only has better refinement effect on primary Si but also lower refinement temperature. These results are due to the combined effect of TiC and AlP on the refinement process, and the duplex TiC/AlP nucleus of primary silicon has been demonstrated using electron probe micro-analysis. Moreover, the reaction of Al-TiO{sub 2}-C mixture powder with increasing temperature was investigated using differential scanning calorimetry, which shows that the TiC particles are produced at about 1473 K. AlP particles combine with the in-situ TiC particles in the melt, which is the main reason for the formation of a duplex nucleus, and the disregistry between TiC and AlP in low-index planes is also discussed. (orig.)

  4. Size effect of primary Y{sub 2}O{sub 3} additions on the characteristics of the nanostructured ferritic ODS alloys: Comparing as-milled and as-milled/annealed alloys using S/TEM

    Energy Technology Data Exchange (ETDEWEB)

    Saber, Mostafa, E-mail: msaber@ncsu.edu; Xu, Weizong; Li, Lulu; Zhu, Yuntian; Koch, Carl C.; Scattergood, Ronald O.

    2014-09-15

    The need for providing S/TEM evidence to clarify the mechanisms of nano-scale precipitate formation was the motivation of this investigation. In this study, an Fe–14Cr–0.4Ti alloy was ball-milled with different amounts of Y{sub 2}O{sub 3} content up to 10 wt.%, and then annealed at temperatures up to 1100 °C. Micron-size Y{sub 2}O{sub 3} particles were substituted for the nano-size counterpart to elucidate the mechanism of oxide precipitate formation. The S/TEM studies revealed that the microstructure of the alloy with 10 wt.% yttria contained amorphous undissolved Y{sub 2}O{sub 3} after ball milling, while a small part of the initial oxide particles were dissolved into the solid solution. Consequently, when the amount of yttria was reduced to 1 wt.%, the amorphous phase of the yttria vanished and the whole content of Y{sub 2}O{sub 3} was dissolved into the BCC solid solution. Defect analysis of precipitates on the annealed samples via S/TEM and micro-hardness studies revealed that the use of micron-size primary oxide particles can produce nano-size precipitates, stable up to temperatures as high as 1100 °C, and uniformly distributed throughout the microstructure. This study indicates that the use of high energy ball milling along with micron-size primary oxide particles can lead to nanostructured ferritic ODS alloys without the use of nano-size primary oxide additions.

  5. Primary water stress corrosion cracks in nickel alloy dissimilar metal welds: Detection and sizing using established and emerging nondestructive examination techniques

    International Nuclear Information System (INIS)

    Braatz, B.G.; Doctor, S.R.; Cumblidge, S.E.; Prokofiev, I.G.

    2012-01-01

    The U.S. Nuclear Regulatory Commission has established the Program to Assess the Reliability of Emerging Nondestructive Techniques (PARENT) as a follow-on to the international cooperative Program for the Inspection of Nickel Alloy Components (PINC). The goal of PINC was to evaluate the capabilities of various nondestructive evaluation (NDE) techniques to detect and characterize surface-breaking primary water stress corrosion cracks in dissimilar-metal welds (DMW) in bottom-mounted instrumentation (BMI) penetrations and small-bore (∼400-mm diameter) piping components. A series of international blind round-robin tests were conducted by commercial and university inspection teams. Results from these tests showed that a combination of conventional and phased-array ultrasound techniques provided the highest performance for flaw detection and depth sizing in dissimilar metal piping welds. The effective detection of flaws in BMIs by eddy current and ultrasound shows that it may be possible to reliably inspect these components in the field. The goal of PARENT is to continue the work begun in PINC and apply the lessons learned to a series of open and blind international round-robin tests that will be conducted on a new set of piping components including large-bore (∼900-mm diameter) DMWs, small-bore DMWs, and BMIs. Open round-robin testing will engage universities and industry worldwide to investigate the reliability of emerging NDE techniques to detect and accurately size flaws having a wide range of lengths, depths, orientations, and locations. Blind round-robin testing will invite testing organizations worldwide, whose inspectors and procedures are certified by the standards for the nuclear industry in their respective countries, to investigate the ability of established NDE techniques to detect and size flaws whose characteristics range from easy to very difficult to detect and size. This paper presents highlights of PINC and reports on the plans and progress for

  6. Primary Water Stress Corrosion Cracks in Nickel Alloy Dissimilar Metal Welds: Detection and Sizing Using Established and Emerging Nondestructive Examination Techniques

    Energy Technology Data Exchange (ETDEWEB)

    Braatz, Brett G.; Cumblidge, Stephen E.; Doctor, Steven R.; Prokofiev, Iouri

    2012-12-31

    The U.S. Nuclear Regulatory Commission has established the Program to Assess the Reliability of Emerging Nondestructive Techniques (PARENT) as a follow-on to the international cooperative Program for the Inspection of Nickel Alloy Components (PINC). The goal of PINC was to evaluate the capabilities of various nondestructive evaluation (NDE) techniques to detect and characterize surface-breaking primary water stress corrosion cracks in dissimilar-metal welds (DMW) in bottom-mounted instrumentation (BMI) penetrations and small-bore (≈400-mm diameter) piping components. A series of international blind round-robin tests were conducted by commercial and university inspection teams. Results from these tests showed that a combination of conventional and phased-array ultrasound techniques provided the highest performance for flaw detection and depth sizing in dissimilar metal piping welds. The effective detection of flaws in BMIs by eddy current and ultrasound shows that it may be possible to reliably inspect these components in the field. The goal of PARENT is to continue the work begun in PINC and apply the lessons learned to a series of open and blind international round-robin tests that will be conducted on a new set of piping components including large-bore (≈900-mm diameter) DMWs, small-bore DMWs, and BMIs. Open round-robin testing will engage universities and industry worldwide to investigate the reliability of emerging NDE techniques to detect and accurately size flaws having a wide range of lengths, depths, orientations, and locations. Blind round-robin testing will invite testing organizations worldwide, whose inspectors and procedures are certified by the standards for the nuclear industry in their respective countries, to investigate the ability of established NDE techniques to detect and size flaws whose characteristics range from easy to very difficult to detect and size. This paper presents highlights of PINC and reports on the plans and progress for

  7. Synthetic sea water - An improved stress corrosion test medium for aluminum alloys

    Science.gov (United States)

    Humphries, T. S.; Nelson, E. E.

    1973-01-01

    A major problem in evaluating the stress corrosion cracking resistance of aluminum alloys by alternate immersion in 3.5 percent salt (NaCl) water is excessive pitting corrosion. Several methods were examined to eliminate this problem and to find an improved accelerated test medium. These included the addition of chromate inhibitors, surface treatment of specimens, and immersion in synthetic sea water. The results indicate that alternate immersion in synthetic sea water is a very promising stress corrosion test medium. Neither chromate inhibitors nor surface treatment (anodize and alodine) of the aluminum specimens improved the performance of alternate immersion in 3.5 percent salt water sufficiently to be classified as an effective stress corrosion test method.

  8. Corrosion studies on Cu-Ni alloys and ferritic steel in salt water for desalination service

    International Nuclear Information System (INIS)

    Shibad, P.R.; Balachandra, J.

    1975-01-01

    Corrosion studies on In 838 and In 848 alloys in 3% NaCl solution, synthetic sea water and in 3% NaCl at pH3 and pH10 indicate that the latter alloy is more corrosion resistant than the former at room (28 0 C), and boiling temperature (101 0 C) and at 125 0 C. Ferritic steel is unaffected in boiling synthetic sea water. In boiling 3% NaCl solution at pH3 and pH10, (the pH values adjusted at room temperature) increase in the rate of corrosion of ferritic steel compared to that at room temperature has been observed. A fair correlation between polarization characteristics and dissolution rates in these solutions is seen for all these materials. (author)

  9. Correlation between Ni base alloys surface conditioning and cation release mitigation in primary coolant

    Energy Technology Data Exchange (ETDEWEB)

    Clauzel, M.; Guillodo, M.; Foucault, M. [AREVA NP SAS, Technical Centre, Le Creusot (France); Engler, N.; Chahma, F.; Brun, C. [AREVA NP SAS, Chemistry and Radiochemistry Group, Paris La Defense (France)

    2010-07-01

    The mastering of the reactor coolant system radioactive contamination is a real stake of performance for operating plants and new builds. The reduction of activated corrosion products deposited on RCS surfaces allows minimizing the global dose integrated by workers which supports the ALARA approach. Moreover, the contamination mastering limits the volumic activities in the primary coolant and thus optimizes the reactor shutdown duration and environment releases. The main contamination sources on PWR are due to Co-60 and Co-58 nuclides which come respectively Co-59 and Ni-58, naturally present in alloys used in the RCS. Co is naturally present as an impurity in alloys or as the main component of hardfacing materials (Stellites™). Ni is released mainly by SG tubes which represent the most important surface of the RCS. PWR steam generators (SG), due to the huge wetted surface are the main source of corrosion products release in the primary coolant circuit. As corrosion products may be transported throughout the whole circuit, activated in the core, and redeposited all over circuit surfaces, resulting in an increase of activity buildup, it is of primary importance to gain a better understanding of phenomenon leading to corrosion product release from SG tubes before setting up mitigation measures. Previous studies have shown that SG tubing made of the same material had different release rates. To find the origin of these discrepancies, investigations have been performed on tubes at the as-received state and after exposure to a nominal primary chemistry in titanium recirculating loop. These investigations highlighted the existence of a correlation between the inner surface metallurgical properties and the release of corrosion products in primary coolant. Oxide films formed in nominal primary chemistry are always protective, their morphology and their composition depending strongly on the geometrical, metallurgical and physico-chemical state of the surface on which they

  10. Microstructure and Corrosion Resistance Property of a Zn-AI-Mg Alloy with Different Solidification Processes

    Directory of Open Access Journals (Sweden)

    Jiang Guang-rui

    2017-01-01

    Full Text Available Zn-Al-Mg alloy coating attracted much attention due to its high corrosion resistance properties, especially high anti-corrosion performance at the cut edge. As the Zn-Al-Mg alloy coating was usually produced by hot-dip galvanizing method, solidification process was considered to influence its microstructure and corrosion properties. In this work, a Zn-Al-Mg cast alloy was melted and cooled to room temperature with different solidification processes, including water quench, air cooling and furnace cooling. Microstructure of the alloy with different solidification processes was characterized by scanning electron microscopy (SEM. Result shows that the microstructure of the Zn-Al-Mg alloy are strongly influenced by solidification process. With increasing solidification rate, more Al is remained in the primary crystal. Electrochemical analysis indicates that with lowering solidification rate, the corrosion current density of the Zn-Al-Mg alloy decreases, which means higher corrosion resistance.

  11. Part of the hydrogen in the intergranular crack by stress corrosion in primary circuit for the 600 and 690 nickel base alloys

    International Nuclear Information System (INIS)

    Odemer, G.; Coudurier, A.; Jambon, F.; Chene, J.; Odemer, G.; Coudurier, A.; Chene, J.

    2007-01-01

    The aim of this study is, in a first part, to characterize the hydrogen embrittlement sensitivity of the 600 and 690 based alloys in order to better understand the hydrogen role in the stress corrosion mechanism which appears in theses alloys in the primary circuit of the PWR type reactors. The authors studies how the hydrogen embrittlement is resulting from an interaction between the hydrogen and the plastic deformation. (A.L.B.)

  12. Features investigation of corrosion-electrochemical behaviour of Al-alloys for engineering an effective protection of the water-distillings setups

    International Nuclear Information System (INIS)

    Fokin, M.N.; Lomakina, S.V.; Tselykh, O.G.; Shatova, T.S.; Trubetskaya, L.F.

    1993-01-01

    The problem of aluminium alloy application in distilling setups is studied. Investigation into the features of corrosion and electrochemical behaviour of aluminium alloys under sea water distillation allows one to reveal the main control factors and to propose optimal alloy compositions capable of providing the safe setup operation on their base. Preliminary treatment in tungsten and molybdenum isopolycompound solutions is proposed which reduces sedimentation which in its turn is very important for distilling setups

  13. Generalized corrosion of nickel base alloys in high temperature aqueous media: a contribution to the comprehension of the mechanisms

    International Nuclear Information System (INIS)

    Marchetti-Sillans, L.

    2007-11-01

    In France, nickel base alloys, such as alloy 600 and alloy 690, are the materials constituting steam generators (SG) tubes of pressurized water reactors (PWR). The generalized corrosion resulting from the interaction between these alloys and the PWR primary media leads, on the one hand, to the formation of a thin protective oxide scale (∼ 10 nm), and on the other hand, to the release of cations in the primary circuit, which entails an increase of the global radioactivity of this circuit. The goal of this work is to supply some new comprehension elements about nickel base alloys corrosion phenomena in PWR primary media, taking up with underlining the effects of metallurgical and physico-chemical parameters on the nature and the growth mechanisms of the protective oxide scale. In this context, the passive film formed during the exposition of alloys 600, 690 and Ni-30Cr, in conditions simulating the PWR primary media, has been analyzed by a set of characterization techniques (SEM, TEM, PEC and MPEC, XPS). The coupling of these methods leads to a fine description, in terms of nature and structure, of the multilayered oxide forming during the exposition of nickel base alloys in primary media. Thus, the protective part of the oxide scale is composed of a continuous layer of iron and nickel mixed chromite, and Cr 2 O 3 nodules dispersed at the alloy / mixed chromite interface. The study of protective scale growth mechanisms by tracers and markers experiments reveals that the formation of the mixed chromite is the consequence of an anionic mechanism, resulting from short circuits like grain boundaries diffusion. Besides, the impact of alloy surface defects has also been studied, underlining a double effect of this parameter, which influences the short circuits diffusion density in oxide and the formation rate of Cr 2 O 3 nodules. The sum of these results leads to suggest a description of the nickel base alloys corrosion mechanisms in PWR primary media and to tackle some

  14. X-ray photoelectron spectroscopy characterization of the ω phase in water quenched Ti-5553 alloy

    International Nuclear Information System (INIS)

    Qin, Dongyang; Lu, Yafeng; Zhang, Kong; Liu, Qian; Zhou, Lian

    2012-01-01

    X-ray photoelectron spectroscopy was used to investigate the ω phase in water quenched Ti-5553 alloy with a nominal composition of Ti–5Al–5V–5Mo–3Cr (wt.%), and the ω and the β phase were distinguished by deconvoluting the XPS spectra of Al2p, V2p and Cr2p core level regions. In addition, it is found that the binding energy of core level electron of alloying elements shifts comparing with that of pure metals, and the fact was interpreted by charge redistribution model. X-ray photoelectron spectroscopy technique could be used to characterize the nano-scale ω phase in β alloys. - Highlights: ► We characterize the ω phase in Ti-5553 alloy by XPS. ► Binding energy of Al2p, V2p and Cr2p electron are different in the ω and β phase. ► Structural difference leads to the binding energy gap.

  15. Tangential turning of Incoloy alloy 925 using abrasive water jet technology

    Czech Academy of Sciences Publication Activity Database

    Cárach, J.; Hloch, S.; Hlaváček, Petr; Ščučka, Jiří; Martinec, Petr; Petrů, J.; Zlámal, T.; Zeleňák, Michal; Monka, P.; Lehocká, D.; Krolczyk, J.

    2016-01-01

    Roč. 82, č. 9 (2016), s. 1747-1752 ISSN 0268-3768 R&D Projects: GA MŠk ED2.1.00/03.0082; GA MŠk(CZ) LO1406 Institutional support: RVO:68145535 Keywords : incoloy alloy 925 * abrasive water jet turning * traverse speed Subject RIV: JQ - Machines ; Tools Impact factor: 2.209, year: 2016 http://link.springer.com/article/10.1007%2Fs00170-015-7489-0

  16. Photoelectrochemical Water Splitting Properties of Ti-Ni-Si-O Nanostructures on Ti-Ni-Si Alloy

    Directory of Open Access Journals (Sweden)

    Ting Li

    2017-10-01

    Full Text Available Ti-Ni-Si-O nanostructures were successfully prepared on Ti-1Ni-5Si alloy foils via electrochemical anodization in ethylene glycol/glycerol solutions containing a small amount of water. The Ti-Ni-Si-O nanostructures were characterized by field-emission scanning electron microscopy (FE-SEM, energy dispersive spectroscopy (EDS, X-ray diffraction (XRD, and diffuse reflectance absorption spectra. Furthermore, the photoelectrochemical water splitting properties of the Ti-Ni-Si-O nanostructure films were investigated. It was found that, after anodization, three different kinds of Ti-Ni-Si-O nanostructures formed in the α-Ti phase region, Ti2Ni phase region, and Ti5Si3 phase region of the alloy surface. Both the anatase and rutile phases of Ti-Ni-Si-O oxide appeared after annealing at 500 °C for 2 h. The photocurrent density obtained from the Ti-Ni-Si-O nanostructure photoanodes was 0.45 mA/cm2 at 0 V (vs. Ag/AgCl in 1 M KOH solution. The above findings make it feasible to further explore excellent photoelectrochemical properties of the nanostructure-modified surface of Ti-Ni-Si ternary alloys.

  17. Photoelectrochemical Water Splitting Properties of Ti-Ni-Si-O Nanostructures on Ti-Ni-Si Alloy.

    Science.gov (United States)

    Li, Ting; Ding, Dongyan; Dong, Zhenbiao; Ning, Congqin

    2017-10-31

    Ti-Ni-Si-O nanostructures were successfully prepared on Ti-1Ni-5Si alloy foils via electrochemical anodization in ethylene glycol/glycerol solutions containing a small amount of water. The Ti-Ni-Si-O nanostructures were characterized by field-emission scanning electron microscopy (FE-SEM), energy dispersive spectroscopy (EDS), X-ray diffraction (XRD), and diffuse reflectance absorption spectra. Furthermore, the photoelectrochemical water splitting properties of the Ti-Ni-Si-O nanostructure films were investigated. It was found that, after anodization, three different kinds of Ti-Ni-Si-O nanostructures formed in the α-Ti phase region, Ti₂Ni phase region, and Ti₅Si₃ phase region of the alloy surface. Both the anatase and rutile phases of Ti-Ni-Si-O oxide appeared after annealing at 500 °C for 2 h. The photocurrent density obtained from the Ti-Ni-Si-O nanostructure photoanodes was 0.45 mA/cm² at 0 V (vs. Ag/AgCl) in 1 M KOH solution. The above findings make it feasible to further explore excellent photoelectrochemical properties of the nanostructure-modified surface of Ti-Ni-Si ternary alloys.

  18. Primary water chemistry for NPP with VVER-TOI

    International Nuclear Information System (INIS)

    Susakin, S.N.; Brykov, S.I.; Zadonsky, N.V.; Bystrova, O.S.

    2012-09-01

    Nowadays within the framework of development of the nuclear power industry in Russia the VVER-TOI reactor is under designing (Standard optimized design). The given design provides for improvement of operation safety level, of technical-economic, operational and load-follow characteristics, and for the raise of competitive capacity of reactor plant and NPP as a whole. In VVER-TOI reactor plant design the primary water chemistry has been improved considering operation experience of VVER reactor plants and a possibility of RP operation under load-follow modes from the viewpoint of meeting the following requirements: - suppression of generation of oxidizing radiolytic products under power operation; - assurance of corrosion resistance of structural materials of equipment and pipelines throughout the NPP design service life; - minimization of deposits on surfaces of the reactor core fuel rods and on heat exchange surface of steam generators; - minimization of accumulation of activated corrosion products; - minimization of the amount of radioactive processing waste. In meeting these requirements an important role is devoted to suppression of generation of oxidizing radiolytic products owing to accumulation of hydrogen in the primary coolant. At NPP with VVER-1000 reactor the ammonia-potassium water chemistry is used wherein the hydrogen accumulation is provided at the expense of ammonia proportioning. Usage of ammonia leads to generation of additional amount of radioactive processing waste and to increased irregularity of maintaining the water chemistry under the daily load-follow modes. In VVER TOI design the primary water chemistry is improved by replacing the proportioning of ammonia with the proportioning of gaseous hydrogen. Different process schemes were considered that provide for a possibility of hydrogen accumulation and maintaining owing to direct proportioning of gaseous hydrogen. The obtained results showed that transition to the potassium water chemistry

  19. Stress relaxation study of water atomized Cu-Cr-Zr powder alloys consolidated by inverse warm extrusion

    International Nuclear Information System (INIS)

    Poblano-Salas, C.A.; Barceinas-Sanchez, J.D.O.

    2009-01-01

    Stress relaxation testing in compression at high temperature was performed on Cu-Cr-Zr alloys produced by consolidation of water atomized powders. Precipitation and recrystallization were monitored during stress relaxation experiments carried out at an ageing temperature of 723 K. Pre-straining imposed to the Cu-Cr-Zr samples prior to stress relaxation testing resulted in reduced hardness compared to that reported for conventionally-aged alloys; it also resulted in shorter times for achieving maximum strengthening on ageing.

  20. Oxidation behavior of austenitic iron-base ODS alloy in supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Behnamian, Y.; Dong, Z.; Zahiri, R.; Kohandehghan, A.; Mitlin, D., E-mail: behnamia@ualberta.ca, E-mail: zdong@ualberta.ca, E-mail: kohandeh@ualberta.ca, E-mail: rzahiris@ualberta.ca, E-mail: dave.mitlin@ualberta.ca [Univ. of Alberta, Edmondon, AB (Canada); Zhou, Z., E-mail: zhouzhj@mater.ustb.edu.cn [Univ. of Science and Tech. Beijing, Beijing (China); Chen, W.; Luo, J., E-mail: weixing.chen@ualberta.ca, E-mail: Jingli.luo@ualberta.ca [Univ. of Alberta, Edmonton, AB (Canada); Zheng, W., E-mail: wenyue@nrcan.gc.ca [Natural Resources Canada, Canmet MATERIALS, Hamilton, ON (Canada); Guzonas, D. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    In this study, the effect of exposure time on the corrosion of the 304 stainless steel based oxide dispersion strengthened alloy, SS304ODS, in supercritical water was investigated at 650 {sup o}C with constant dissolved oxygen concentration. The results show that the oxidation of SS304ODS in supercritical water followed a parabolic law at 650 {sup o}C. Discontinuous oxide scale with two distinct layers has formed after 550 hours. The inner layer was chromium-rich while the outer layer was iron-rich (Magnetite). The oxide islands grow with increasing the exposure time. With increasing exposure time, the quantity of oxide islands increased in which major preferential growth along oxide-substrate interface was observed. The possible mechanism of SS304ODS oxidation in supercritical water was also discussed. (author)

  1. Corrosion of alloy 800 in PHWR primary and secondary conditions

    International Nuclear Information System (INIS)

    Maroto, A.J.G.; Blesa, M.A.; Villegas, M.; Olmedo, A.M.; Bordoni, R.; Alvarez, M.G.; Sainz, R.

    1998-01-01

    A hot leg section of a steam generator tubing was removed for destructive examination from one of the steam generators (SG) of the Embalse Nuclear Power Plant. The tube material is Alloy 800 and carbon steel is the tube support plate material. Samples of the deposits were taken at the first tube support plate and at the top, mid-height and bottom of the sludge pile. Transverse sections were taken at several locations along the tube length measuring the oxide thicknesses and studying the morphology of the oxide layer by scanning electron microscopy on the primary and secondary side at each location. Deposit layers on the outer tube surface revealed iron as major component and the presence of calcium, phosphorous, zinc and manganese. The oxide scale thickness at the secondary side in the open area was around 22 to 30 μm. The oxide thickness grown under isothermal conditions on the corrosion test samples installed in the autoclaves facilities of the primary circuit of the plant was measured and compared with that found on the inner surface of the examined tube section. The oxide thickness of the test samples was around 1-2 μm showing the influence of the deposition of corrosion products from the coolant. Deposition and precipitation of oxide was also found in the actual tube, where the common feature was the irregularity of the oxide layer on the primary side and thicknesses values in the range 4 to 10 μm were measured. The autoclave tests and SG tubing examination permit to compare the influence of materials and of operating (flow rate, isothermal vs non-isothermal) conditions on corrosion and deposition. (author)

  2. Development of a copper alloy to beryllium HIP bonding technology for the ITER first wall

    International Nuclear Information System (INIS)

    Sherlock, P.; Peacock, A.T.; Mc Callum, A.D.

    2005-01-01

    The primary first wall (PFW) panels of the ITER blanket concept comprise a bi-metallic copper alloy/stainless steel water-cooled heatsink faced with a plasma facing material. Precipitation strengthened CuCrZr is one option for the copper alloy of the heatsink; beryllium, in the form of tiles is an option for the plasma facing material. Over recent years, the technology needed to HIP bond the beryllium tiles to CuCrZr alloy has been developed. This paper describes small samples and larger mock-ups produced during the development of this HIP bonding technology and outlines how structural analyses were used to gain an understanding of the bonding process and refine the design

  3. Thermal Aging Effects on Heat Affected Zone of Alloy 600 in Dissimilar Metal Weld

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Jun Hyuk; Choi, Kyoung Joon; Yoo, Seung Chang; Kim, Ji Hyun [UNIST, Ulsan (Korea, Republic of)

    2016-05-15

    Dissimilar metal weld (DMW), consists of Alloy 600, Alloy 182, and A508 Gr.3, is now being widely used as the reactor pressure vessel penetration nozzle and the steam generator tubing material for pressurized water reactors (PWR) because of its mechanical property, thermal expansion coefficient, and corrosion resistance. The heat affected zone (HAZ) on Alloy 600 which is formed by welding process is critical to crack. According to G.A. Young et al. crack growth rates (CGR) in the Alloy 600 HAZ were about 30 times faster than those in the Alloy 600 base metal tested under the same conditions [3]. And according to Z.P. Lu et al. CGR in the Alloy 600 HAZ can be more than 20 times higher than that in its base metal. To predict the life time of components, there is a model which can calculate the effective degradation years (EDYs) of the material as a function of operating temperature. This study was conducted to investigate how thermal aging affects the hardness of dissimilar metal weld from the fusion boundary to Alloy 600 base metal and the residual strain at Alloy 600 heat affected zone. Following conclusions can be drawn from this study. The hardness, measured by Vickers hardness tester, peaked near the fusion boundary between Alloy 182 and Alloy 600, and it decreases as the picked point goes to Alloy 600 base metal. Even though the formation of precipitate such as Cr carbide, thermal aging doesn't affect the value and the tendency of hardness because of reduced residual stress. According to kernel average misorientation mapping, residual strain decreases when the material thermally aged. And finally, in 30 years simulated specimen, the high residual strain almost disappears. Therefore, the influence of residual strain on primary water stress corrosion cracking can be diminished when the material undergoes thermal aging.

  4. Cytotoxicity and ion release of alloy nanoparticles

    International Nuclear Information System (INIS)

    Hahn, Anne; Fuhlrott, Jutta; Loos, Anneke; Barcikowski, Stephan

    2012-01-01

    It is well-known that nanoparticles could cause toxic effects in cells. Alloy nanoparticles with yet unknown health risk may be released from cardiovascular implants made of Nickel–Titanium or Cobalt–Chromium due to abrasion or production failure. We show the bio-response of human primary endothelial and smooth muscle cells exposed to different concentrations of metal and alloy nanoparticles. Nanoparticles having primary particle sizes in the range of 5–250 nm were generated using laser ablation in three different solutions avoiding artificial chemical additives, and giving access to formulations containing nanoparticles only stabilized by biological ligands. Endothelial cells are found to be more sensitive to nanoparticle exposure than smooth muscle cells. Cobalt and Nickel nanoparticles caused the highest cytotoxicity. In contrast, Titanium, Nickel–Iron, and Nickel–Titanium nanoparticles had almost no influence on cells below a nanoparticle concentration of 10 μM. Nanoparticles in cysteine dissolved almost completely, whereas less ions are released when nanoparticles were stabilized in water or citrate solution. Nanoparticles stabilized by cysteine caused less inhibitory effects on cells suggesting cysteine to form metal complexes with bioactive ions in media.

  5. Modeling the long-term evolution of the primary damage in ferritic alloys using coarse-grained methods

    International Nuclear Information System (INIS)

    Becquart, C.S.; Barbu, A.; Bocquet, J.L.; Caturla, M.J.; Domain, C.; Fu, C.-C.; Golubov, S.I.; Hou, M.; Malerba, L.; Ortiz, C.J.; Souidi, A.; Stoller, R.E.

    2010-01-01

    Knowledge of the long-term evolution of the microstructure after introduction of primary damage is an essential ingredient in understanding mechanical property changes that occur during irradiation. Within the European integrated project 'PERFECT,' different techniques have been developed or improved to model microstructure evolution of Fe alloys under irradiation. This review paper aims to present the current state of the art of these techniques, as developed in the project, as well as the main results obtained.

  6. Stress corrosion cracking studies on ferritic low alloy pressure vessel steel - water chemistry and modelling aspects

    International Nuclear Information System (INIS)

    Tipping, P.; Ineichen, U.; Cripps, R.

    1994-01-01

    The susceptibility of low alloy ferritic pressure vessel steels (A533-B type) to stress corrosion cracking (SCC) degradation has been examined using various BWR type coolant chemistries. Fatigue pre-cracked wedge-loaded double cantilever beams and also constantly loaded 25 mm thick compact tension specimens have shown classical SCC attack. The influence of parameters such as dissolved oxygen content, water impurity level and conductivity, material chemical composition (sulphur content) and stress intensity level are discussed. The relevance of SCC as a life-limiting degradation mechanism for low alloy ferritic nuclear power plant PV steel is examined. Some parameters, thought to be relevant for modelling SCC processes in low alloy steels in simulated BWR-type coolant, are discussed. 8 refs., 1 fig., 4 tabs

  7. Effect of Water Chemistry Variations on Corrosion of Zr-Alloys for BWR Applications

    International Nuclear Information System (INIS)

    Kim, Young-Jin; Yang- Lin, Pi; Lutz, Dan; Kucuk, Aylin; Cheng, Bo

    2012-09-01

    Two reference water chemistry conditions (60 ppb Zn and 60 μg/cm 2 Pt/Rh with either 500 ppb O 2 and 500 ppb H 2 O 2 , or 150 ppb H 2 ) were chosen for testing at 300 deg. C in refreshed autoclaves. For each reference water chemistry, the potential effects due to three chemical impurities of interest to BWRs (33 ppm Na, 10 ppm Li, and 10 ppm EHC fluid) were evaluated. Zircaloy-2 and GNF-Ziron (a Zr-based alloy with higher Fe additions than Zircaloy-2) cladding tubes were tested and the effects of tubing process variation and pre-filming were investigated. Tested channel materials included Zircaloy-2, Zircaloy-4, GNF-Ziron and NSF (a Zr-based alloy with Sn, Nb and Fe additions). The corrosion weight gain and hydrogen absorption were measured up to 12 months of exposure for a given water chemistry condition. Tests under 150 ppb H 2 based water chemistry, with or without chemical impurities, generally resulted in greater amounts of corrosion after 12 month exposure compared with 500 ppb O 2 and 500 ppb H 2 O 2 based water chemistries. Of the added chemical impurities, only 33 ppm Na addition produced slightly increased corrosion. Under various test conditions, the presence of a thin pre-film resulted in some initial corrosion benefits, but the benefits were no longer evident after 12 months exposure; however, slight hydrogen benefits remained. For GNF-Ziron cladding, hydrogen absorption was generally lower compared with similarly processed Zircaloy-2 under 150 ppb H 2 based water chemistry, when corrosion was generally higher. Of the channel material tested, NSF developed the lowest level of hydrogen absorption, particularly under 150 ppb H 2 based water chemistries. (authors)

  8. Stress-corrosion cracking of Inconel alloy 600 in high-temperature water: an update

    International Nuclear Information System (INIS)

    Bandy, R.; van Rooyen, D.

    1983-01-01

    Inconel 600 has been tested in high-temperature aqueous media (without oxygen) in several tests. Data are presented to relate failure times to periods of crack initiation and propagation. Quantitative relationships have been developed from tests in which variations were made in temperature, applied load, strain rate, water chemistry, and the condition of the test alloy

  9. Thermally activated low temperature creep and primary water stress corrosion cracking of NiCrFe alloys

    International Nuclear Information System (INIS)

    Hall, M.M. Jr.

    1993-01-01

    A phenomenological SCC-CGR model is developed based on an apriori assumption that the SCC-CGR is controlled by low temperature creep (LTC). This mode of low temperature time dependent deformation occurs at stress levels above the athermal flow stress by a dislocation glide mechanism that is thermally activated and may be environmentally assisted. The SCC-CGR model equations developed contain thermal activation parameters descriptive of the dislocation creep mechanism. Thermal activation parameters are obtained by fitting the CGR model to SCC-CGR data obtained on Alloy 600 and Alloy X-750. These SCC-CGR activation parameters are compared to LTC activation parameters obtained from stress relaxation tests. When the high concentration of hydrogen at the tip of an SCC crack is considered, the SCC-CGR activation energies and rate sensitivities are shown to be quantitatively consistent with hydrogen reducing the activation energy and increasing the strain rate sensitivity in LTC stress relaxation tests. Stress dependence of SCC-CGR activation energy consistent with that found for the LTC activation energy. Comparisons between temperature dependence of the SCC-CGR stress sensitivity and LTC stress sensitivity provide a basis for speculation on effects of hydrogen and solute carbon on SCC crack growth rates

  10. Environmentally assisted fatigue evaluation model of alloy 690 steam generator tube in high temperature water

    International Nuclear Information System (INIS)

    Tan Jibo; Wu Xinqiang; Han Enhou; Wang Xiang; Liu Xiaoqiang; Xu Xuelian

    2015-01-01

    Nickel-based alloy 690 has been widely used as steam generator tube in light water reactor (LWR) nuclear power plants, which may suffer from corrosion fatigue during long-term service. Many researches and operating experience indicated that the effect of LWR environment could significantly reduce the fatigue life of structural materials. However. such an environmental degradation effect was not fully addressed in the current ASME code design fatigue curves. Therefore, the Regulatory Guide 1.207 issued by US NRC required a new NPP have to incorporate the environment effects into fatigue analyses. In the last few decades, researchers in USA and Japan systematically investigated the corrosion fatigue behavior of nuclear-grade structural materials in LWR environment. Then, ANL model and JSME model were proposed, which incorporated environmental effects, including temperature, dissolved oxygen (DO) and strain rate for the nickel-based alloys. Due to lack of experiment data on domestic materials, there is no related environmental fatigue design model in China. In the present work, based on the corrosion fatigue tests of a kind of boat-shaped specimen in borated and lithiated high temperature water, the corrosion fatigue behavior and environmentally assisted cracking mechanism of domestic Alloy 690 steam generator tube have been investigate. An IMR model for the nickel-based alloy was proposed. The environmental fatigue life correction factor (F en ) was established, which addressed the environmental factors, including temperature, strain rate and dissolved oxygen. The method to evaluate environmental fatigue damage of structural materials in NPPs was proposed. (authors)

  11. Effects of Al-Mn-Ti-P-Cu master alloy on microstructure and properties of Al-25Si alloy

    Directory of Open Access Journals (Sweden)

    Xu Chunxiang

    2013-09-01

    Full Text Available To obtain a higher microstructural refining efficiency, and improve the properties and processing ability of hypereutectic Al-25Si alloy, a new environmentally friendly Al-20.6Mn-12Ti-0.9P-6.1Cu (by wt.% master alloy was fabricated; and its modification and strengthening mechanisms on the Al-25Si alloy were studied. The mechanical properties of the unmodified, modified and heat treated alloys were investigated. Results show that the optimal addition amount of the Al-20.6Mn-12Ti-0.9P-6.1Cu master alloy is 4wt.%. In this case, primary Si and eutectic Si as well as メ-Al phase were clearly refined, and this refining effect shows an excellent long residual action as it can be heat-retained for at least 5 h. After being T6 heat treated, the morphology of primary and eutectic Si in the Al-25Si alloys with the addition of 4wt.% Al-20.6Mn-12Ti-0.9P-6.1Cu alloy changes into particles and short rods. The average grain size of the primary and eutectic Si decreases from 250 レm (unmodified to 13.83 レm and 35 レm (unmodified to 7 レm; the メ-Al becomes obviously finer and the distribution of Si phases tends to be uniform and dispersed. Meanwhile, the tensile properties are improved obviously; the tensile strengths at room temperature and 300 ìC reach 241 MPa and 127 MPa, increased by 153.7% and 67.1%, respectively. In addition, the tensile fracture mechanism changes from brittle fracture for the alloy without modification to ductile fracture after modification. Modifying the morphology of Si phase and strengthening the matrix can effectively block the initiation and propagation of cracks, thus improving the strength of the hypereutectic Al-25Si alloy.

  12. Electrochemical evaluation of zinc effect on the corrosion of nickel alloy in PWR solutions with increasing temperature

    International Nuclear Information System (INIS)

    Alvial M, Gaston; Neves, Celia F.C.; Schvartzman, Monica M.A.M.; Quinan, Marco Antonio D.

    2007-01-01

    The main objective for the addition of zinc acetate to the reactor coolant system of PWRs is to effect radiation dose rate reductions. However, zinc is also added as an approach to mitigate the occurrence or severity of primary water stress corrosion cracking of nickel alloy 600. The mechanism by which zinc affects the corrosion of austenitic nickel-base alloys is by incorporation of zinc into the spinel oxide corrosion films. The purpose of this work is to evaluate the influence of zinc on the corrosion behavior of the nickel alloy 600 in PWR chemical environment (1200 ppm B, 2.2 ppm Li, deoxygenated water) with increasing temperature at room pressure. Electrochemical tests (anodic potentiodynamic polarization and electrochemical impedance spectroscopy) were used to characterize the alloy 600. Two conditions were applied: 0 and 100 ppb zinc and the temperature range was 50 - 90 deg C, at ambient pressure. Potentiodynamic polarization was inefficient to present conclusive results. Impedance measurements showed single semicircle in the Nyquist plane suggesting reduction of the charge transference resistance in zinc-containing solutions. This effect is evident at 90 deg C suggesting prejudicial influence of zinc for the alloy 600 at room pressure. (author)

  13. Thermodynamic stability of oxides in the Ni-Cr-Fe system and stress corrosion crack growth kinetics of alloy 600 in primary water

    International Nuclear Information System (INIS)

    Caron, D.; Cassagne, T.; Daret, J.; Santarini, G.; Mazille, H.

    1999-01-01

    In the framework of the study of stress corrosion of alloy-600, a thermodynamical study of stoichiometric simple and mixed oxides of Ni-Cr-Fe system has been performed. This theoretical work shows that the oxidation of alloy-600 is dependent on temperature and on the quantity of dissolved hydrogen

  14. Influence of hydrazine primary water chemistry on corrosion of fuel cladding and primary circuit components

    International Nuclear Information System (INIS)

    Iourmanov, V.; Pashevich, V.; Bogancs, J.; Tilky, P.; Schunk, J.; Pinter, T.

    1999-01-01

    Earlier at Paks 1-4 NPP standard ammonia chemistry was in use. The following station performance indicators were improved when hydrazine primary water chemistry was introduced: occupational radiation exposures of personnel; gamma-radiation dose rates near primary system components during refuelling and maintenance outages. The reduction of radiation exposures and radiation fields were achieved without significant expenses. Recent results of experimental studies allowed to explain the mechanism of hydrazine dosing influence on: corrosion rate of structure materials in primary coolant; behaviour of soluble and insoluble corrosion products including long-life corrosion-induced radionuclides in primary system during steady-state and transient operation modes; radiolytic generation of oxidising radiolytic products in core and its corrosion activity in primary system; radiation situation during refuelling and maintenance outages; foreign material degradation and removal (including corrosion active oxidant species) from primary system during abnormal events. Operational experience and experimental data have shown that hydrazine primary water chemistry allows to reduce corrosion wear and thereby makes it possible to extend the life-time of plant components in primary system. (author)

  15. The effect of carbon distribution on deformation and cracking of Ni-16Cr-9Fe-C alloys

    International Nuclear Information System (INIS)

    Hertzberg, J.L.; Was, G.S.

    1995-01-01

    Constant extension rate tensile (CERT) tests and constant load tensile (CLT) tests were conducted on controlled purity Ni-16Cr-9Fe-C alloys. The amount and form of carbon were varied in order to investigate the roles of carbon in solution and as intergranular (IG) carbides in the deformation and IG cracking behavior in 360 C argon and primary water environments. Results show that the strength, ductility and creep resistance of these alloys are increased with carbon present in solid solution, while IG cracking on the fracture surface is suppressed. Alloys containing carbon in the form of IG carbides, however, exhibit reduced strength and ductility relative to carbon in solution, while maintaining high IG cracking resistance with respect to carbon-free alloys. CERT results of commercial alloy 600 and controlled purity, carbon containing alloys yield comparable failure strains and IG cracking amounts. CLT comparisons with creep tests of alloy 600 suggest that alloys containing IG carbides are more susceptible to creep than those containing all carbon in solid solution

  16. Influence of irradiation and radiolysis on the corrosion rates and mechanisms of zirconium alloys

    International Nuclear Information System (INIS)

    Verlet, Romain

    2015-01-01

    The nuclear fuel of pressurized water reactors (PWR) in the form of uranium oxide UO 2 pellets (or MOX) is confined in a zirconium alloy cladding. This cladding is very important because it represents the first containment barrier against the release of fission products generated by the nuclear reaction to the external environment. Corrosion by the primary medium of zirconium alloys, particularly the Zircaloy-4, is one of the factors limiting the reactor residence time of the fuel rods (UO 2 pellets + cladding). To optimize core management and to extend the lifetime of the fuel rods in reactor, new alloys based on zirconium-niobium (M5) have been developed. However, the corrosion mechanisms of these are not completely understood because of the complexity of these materials, corrosion environment and the presence of radiation from the nuclear fuel. Therefore, this thesis specifically addresses the effects of radiolysis and defects induced by irradiation with ions in the matrix metal and the oxide layer on the corrosion rate of Zircaloy-4 and M5. The goal is to separate the influence of radiation damage to the metal, that relating to defects created in the oxide and that linked to radiolysis of the primary medium on the oxidation rate of zirconium alloys in reactor. 1) Regarding effect of irradiation of the metal on the oxidation rate: type dislocation loops appear and increase the oxidation rate of the two alloys. For M5, in addition to the first effect, a precipitation of fines needles of niobium reduced the solid solution of niobium concentration in the metal and ultimately in the oxide, which strongly reduces the oxidation rate of the alloy. 2) Regarding the effect of irradiation of the oxide layer on the oxidation rate: defects generated by the nuclear cascades in the oxide increase the oxidation rate of the two materials. For M5, germination of niobium enriched zones in irradiated oxide also causes a decrease of the niobium concentration in solid solution

  17. Process of Equiaxed Grains of RE-Al Alloy under Slope Vibration

    International Nuclear Information System (INIS)

    Xie Shikun; Yi Rongxi; Pan Xiaoliang; Zheng Xiaoqiu; Guo Xiuyan

    2010-01-01

    A new technique using slope vibration casting process during heating and isothermal holding period to prepare Al-7Si-2RE alloy has been studied. The small, near-spherical and non-dendritic microstructure with the semi-solid processing requirements has been obtained. Experiments show that the cooling method, pouring process and the convection of melt caused by slope vibration had significant effects on the formation of near-spherical primary gains. The water-cooled copper mold casting with slope vibration at the temperature near liquidus can obtain Al-7Si-2RE alloy with small homogeneous equiaxed grains, the average grain diameter is 48.3 μm, and the average grain roundness is 1.92.

  18. EFFECTS OF IRRADIATION ON THERMAL CONDUCTIVITY OF ALLOY 690 AT LOW NEUTRON FLUENCE

    Directory of Open Access Journals (Sweden)

    WOO SEOG RYU

    2013-04-01

    Full Text Available Alloy 690 has been selected as a steam generator tubing material for SMART owing to a near immunity to primary water stress corrosion cracking. The steam generators of SMART are faced with a neutron flux due to the integrated arrangement inside a reactor vessel, and thus it is important to know the irradiation effects of the thermal conductivity of Alloy 690. Alloy 690 was irradiated at HANARO to fluences of (0.7−28 × 1019n/cm2 (E>0.1MeV at 250°C, and its thermal conductivity was measured using the laser-flash equipment in the IMEF. The thermal conductivity of Alloy 690 was dependent on temperature, and it was a good fit to the Smith-Palmer equation, which modified the Wiedemann-Franz law. The irradiation at 250°C did not degrade the thermal conductivity of Alloy 690, and even showed a small increase (1% at fluences of (0.7∼28 × 1019n/cm2 (E>0.1MeV.

  19. Fretting wear behavior of zirconium alloy in B-Li water at 300 °C

    Science.gov (United States)

    Zhang, Lefu; Lai, Ping; Liu, Qingdong; Zeng, Qifeng; Lu, Junqiang; Guo, Xianglong

    2018-02-01

    The tangential fretting wear of three kinds of zirconium alloys tube mated with 304 stainless steel (SS) plate was investigated. The tests were conducted in an autoclave containing 300 °C pressurized B-Li water for tube-on-plate contact configuration. The worn surfaces were examined with scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS) and 3D microscopy. The cross-section of wear scar was examined with transmission electron microscope (TEM). The results indicated that the dominant wear mechanism of zirconium alloys in this test condition was delamination and oxidation. The oxide layer on the fretted area consists of outer oxide layer composed of iron oxide and zirconium oxide and inner oxide layer composed of zirconium oxide.

  20. EBR-II Primary Tank Wash-Water Alternatives Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Demmer, R. L.; Heintzelman, J. B.; Merservey, R. H.; Squires, L. N.

    2008-05-01

    The EBR-II reactor at Idaho National Laboratory was a liquid sodium metal cooled reactor that operated for 30 years. It was shut down in 1994; the fuel was removed by 1996; and the bulk of sodium metal coolant was removed from the reactor by 2001. Approximately 1100 kg of residual sodium remained in the primary system after draining the bulk sodium. To stabilize the remaining sodium, both the primary and secondary systems were treated with a purge of moist carbon dioxide. Most of the residual sodium reacted with the carbon dioxide and water vapor to form a passivation layer of primarily sodium bicarbonate. The passivation treatment was stopped in 2005 and the primary system is maintained under a blanket of dry carbon dioxide. Approximately 670 kg of sodium metal remains in the primary system in locations that were inaccessible to passivation treatment or in pools of sodium that were too deep for complete penetration of the passivation treatment. The EBR-II reactor was permitted by the Idaho Department of Environmental Quality (DEQ) in 2002 under a RCRA permit that requires removal of all remaining sodium in the primary and secondary systems by 2022. The proposed baseline closure method would remove the large components from the primary tank, fill the primary system with water, react the remaining sodium with the water and dissolve the reaction products in the wash water. This method would generate a minimum of 100,000 gallons of caustic, liquid, low level radioactive, hazardous waste water that must be disposed of in a permitted facility. On February 19-20, 2008, a workshop was held in Idaho Falls, Idaho, to look at alternatives that could meet the RCRA permit clean closure requirements and minimize the quantity of hazardous waste generated by the cleanup process. The workshop convened a panel of national and international sodium cleanup specialists, subject matter experts from the INL, and the EBR-II Wash Water Project team that organized the workshop. The

  1. The resistance to PWSCC of explosively expanded Alloy 600 tube-to-tubesheet joints

    International Nuclear Information System (INIS)

    Gold, R.E.; Pement, F.W.; Tarabek, S.A.; Economy, G.

    1992-01-01

    Experimental evaluations were performed to determine the approximate magnitude of the residual stresses associated with explosively expanded steam generator tubing, and to assess the resistance to primary water stress corrosion cracking (PWSCC) of these expansions. Indexing of residual stresses was performed by means of magnesium chloride exposures of surrogate stainless steel mockups. The PWSCC resistance was evaluated by the testing of pressurized mockups of explosively expanded mill annealed Alloy 600 tubing in a highly accelerated Alloy 600 tubing in a highly accelerated steam test environment. Shot peening of the inside tube surfaces was demonstrated to be effective in modifying the residual stresses, providing additional resistance to PWSCC

  2. An Overview of the EPRI PWR Primary Chemistry Program

    International Nuclear Information System (INIS)

    Perkins, David; Fruzzetti, Keith; Haas, Carey; Wells, Dan

    2012-09-01

    and changed from stainless steel to the various advanced alloys utilized today. The ever present demand to shorten outage durations is compounding the challenges faced in shutdown chemistry. Utilities continue to optimize shutdown and cool down activities for refueling outages supporting outage durations < 30 days and in some cases < 20 days. The industry continues to review plant shutdown strategies including a rapid shutdown and cool down that allows utilities to enter cold shutdown conditions in = 6 hours from hot standby. The EPRI Materials Reliability Program (MRP), Fuel Reliability Program (FRP), Chemistry and Radiation Management (RM) programs have ongoing research related to optimized chemistry supporting industry efforts in dose reduction, mitigation of primary water stress corrosion cracking (PWSCC) in nickel-based alloys and improved fuel reliability. This paper describes the ongoing EPRI pressurized water reactor primary water chemistry research supporting these programs, as well as an update to previous papers presented in the conference series. (authors)

  3. Corrosion behavior in high-temperature pressurized water of Zircaloy-4 joints brazed with Zr-Cu-based amorphous filler alloys

    Science.gov (United States)

    Lee, Jung Gu; Lee, Gyoung-Ja; Park, Jin-Ju; Lee, Min-Ku

    2017-05-01

    The compositional effects of ternary Zr-Cu-X (X: Al, Fe) amorphous filler alloys on galvanic corrosion susceptibility in high-temperature pressurized water were investigated for Zircaloy-4 brazed joints. Through an Al-induced microgalvanic reaction that deteriorated the overall nobility of the joint, application of the Zr-Cu-Al filler alloy caused galvanic coupling to develop readily between the Al-bearing joint and the Al-free base metal, finally leading to massive localized corrosion of the joint. Contrastingly, joints prepared with a Zr-Cu-Fe filler alloy showed excellent corrosion resistance comparable to that of the Zircaloy-4 base metal, since the Cu and Fe elements forming fine intermetallic particles with Zr did not influence the electrochemical stability of the resultant joints. The present results demonstrate that Fe is a more suitable alloying element than Al for brazing filler alloys subjected to high-temperature corrosive environments.

  4. Antibacterial biodegradable Mg-Ag alloys

    Directory of Open Access Journals (Sweden)

    D Tie

    2013-06-01

    Full Text Available The use of magnesium alloys as degradable metals for biomedical applications is a topic of ongoing research and the demand for multifunctional materials is increasing. Hence, binary Mg-Ag alloys were designed as implant materials to combine the favourable properties of magnesium with the well-known antibacterial property of silver. In this study, three Mg-Ag alloys, Mg2Ag, Mg4Ag and Mg6Ag that contain 1.87 %, 3.82 % and 6.00 % silver by weight, respectively, were cast and processed with solution (T4 and aging (T6 heat treatment.The metallurgical analysis and phase identification showed that all alloys contained Mg4Ag as the dominant β phase. After heat treatment, the mechanical properties of all Mg-Ag alloys were significantly improved and the corrosion rate was also significantly reduced, due to presence of silver. Mg(OH2 and MgO present the main magnesium corrosion products, while AgCl was found as the corresponding primary silver corrosion product. Immersion tests, under cell culture conditions, demonstrated that the silver content did not significantly shift the pH and magnesium ion release. In vitro tests, with both primary osteoblasts and cell lines (MG63, RAW 264.7, revealed that Mg-Ag alloys show negligible cytotoxicity and sound cytocompatibility. Antibacterial assays, performed in a dynamic bioreactor system, proved that the alloys reduce the viability of two common pathogenic bacteria, Staphylococcus aureus (DSMZ 20231 and Staphylococcus epidermidis (DSMZ 3269, and the results showed that the killing rate of the alloys against tested bacteria exceeded 90%. In summary, biodegradable Mg-Ag alloys are cytocompatible materials with adjustable mechanical and corrosion properties and show promising antibacterial activity, which indicates their potential as antibacterial biodegradable implant materials.

  5. EBR-II Primary Tank Wash-Water Alternatives Evaluation

    International Nuclear Information System (INIS)

    Demmer, R.; Heintzelman, J.; Squires, L.; Meservey, R.

    2009-01-01

    The EBR-II reactor at Idaho National Laboratory was a liquid sodium metal cooled reactor that operated for 30 years. Approximately 1100 kg of residual sodium remained in the primary system after draining the bulk sodium. To stabilize the remaining sodium, both the primary and secondary systems were treated with a purge of moist carbon dioxide. The passivation treatment was stopped in 2005 and the primary system is maintained under a blanket of dry carbon dioxide. Approximately 670 kg of sodium metal remains in the primary system in locations that were inaccessible to passivation treatment or in pools of sodium that were too deep for complete penetration of the passivation treatment. The EBR-II reactor was permitted by the Idaho Department of Environmental Quality (DEQ) in 2002 under a RCRA permit that requires removal of all remaining sodium. The proposed baseline closure method would remove the large components from the primary tank, fill the primary system with water, react the remaining sodium with the water and dissolve the reaction products in about 100,000 gallons of wash water. On February 19-20, 2008, a workshop was held in Idaho Falls, Idaho, to evaluate alternatives that could meet the RCRA permit clean closure requirements and minimize the quantity of hazardous waste generated by the cleanup process. The workshop convened a panel of national and international sodium cleanup specialists, subject matter experts from the INL, and the EBR-II Wash Water Project team that organized the workshop. The workshop was conducted by a trained facilitator using Value Engineering techniques to elicit the most technically sound solutions from the workshop participants. A brainstorming session was held to identify possible alternative treatment methods that would meet the primary functions and criteria of neutralizing the hazards, maximizing byproduct removal and minimizing waste generation. An initial list of some 20 probable alternatives was evaluated and refined down

  6. Corrosion behavior in high-temperature pressurized water of Zircaloy-4 joints brazed with Zr-Cu-based amorphous filler alloys

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Gu, E-mail: jglee88@ulsan.ac.kr [School of Materials Science and Engineering, University of Ulsan, Ulsan 44610 (Korea, Republic of); Lee, Gyoung-Ja; Park, Jin-Ju [Nuclear Materials Development Division, Korea Atomic Energy Research Institute (KAERI), Yuseong, Daejeon 34057 (Korea, Republic of); Lee, Min-Ku, E-mail: leeminku@kaeri.re.kr [Nuclear Materials Development Division, Korea Atomic Energy Research Institute (KAERI), Yuseong, Daejeon 34057 (Korea, Republic of)

    2017-05-15

    The compositional effects of ternary Zr-Cu-X (X: Al, Fe) amorphous filler alloys on galvanic corrosion susceptibility in high-temperature pressurized water were investigated for Zircaloy-4 brazed joints. Through an Al-induced microgalvanic reaction that deteriorated the overall nobility of the joint, application of the Zr-Cu-Al filler alloy caused galvanic coupling to develop readily between the Al-bearing joint and the Al-free base metal, finally leading to massive localized corrosion of the joint. Contrastingly, joints prepared with a Zr-Cu-Fe filler alloy showed excellent corrosion resistance comparable to that of the Zircaloy-4 base metal, since the Cu and Fe elements forming fine intermetallic particles with Zr did not influence the electrochemical stability of the resultant joints. The present results demonstrate that Fe is a more suitable alloying element than Al for brazing filler alloys subjected to high-temperature corrosive environments. - Highlights: •Corrosion of Zircaloy-4 joints brazed with Zr-Cu-X filler alloys was investigated. •Alloyed Al deteriorated the overall nobility of joints by microgalvanic reaction. •Compositional gradient of Al in joints was the driving force for galvanic corrosion. •Cu and Fe did not influence the electrochemical stability of joints. •Zr-Cu-Fe filler alloy yielded excellent high-temperature corrosion resistance.

  7. Corrosion behavior in high-temperature pressurized water of Zircaloy-4 joints brazed with Zr-Cu-based amorphous filler alloys

    International Nuclear Information System (INIS)

    Lee, Jung Gu; Lee, Gyoung-Ja; Park, Jin-Ju; Lee, Min-Ku

    2017-01-01

    The compositional effects of ternary Zr-Cu-X (X: Al, Fe) amorphous filler alloys on galvanic corrosion susceptibility in high-temperature pressurized water were investigated for Zircaloy-4 brazed joints. Through an Al-induced microgalvanic reaction that deteriorated the overall nobility of the joint, application of the Zr-Cu-Al filler alloy caused galvanic coupling to develop readily between the Al-bearing joint and the Al-free base metal, finally leading to massive localized corrosion of the joint. Contrastingly, joints prepared with a Zr-Cu-Fe filler alloy showed excellent corrosion resistance comparable to that of the Zircaloy-4 base metal, since the Cu and Fe elements forming fine intermetallic particles with Zr did not influence the electrochemical stability of the resultant joints. The present results demonstrate that Fe is a more suitable alloying element than Al for brazing filler alloys subjected to high-temperature corrosive environments. - Highlights: •Corrosion of Zircaloy-4 joints brazed with Zr-Cu-X filler alloys was investigated. •Alloyed Al deteriorated the overall nobility of joints by microgalvanic reaction. •Compositional gradient of Al in joints was the driving force for galvanic corrosion. •Cu and Fe did not influence the electrochemical stability of joints. •Zr-Cu-Fe filler alloy yielded excellent high-temperature corrosion resistance.

  8. The creep and intergranular cracking behavior of Ni-Cr-Fe-C alloys in 360 degree C water

    International Nuclear Information System (INIS)

    Angeliu, T.M.; Paraventi, D.J.; Was, G.S.

    1995-01-01

    Mechanical testing of controlled-purity Ni-xCr-9Fe-yC alloys at 360 C revealed an environmental enhancement in IG cracking and time-dependent deformation in high purity and primary water over that exhibited in argon. Dimples on the IG facets indicate a creep void nucleation and growth failure mode. IG cracking was primarily located at the interior of the specimen and not necessarily linked to direct contact with the environment. Controlled potential CERT experiments showed increases in IG cracking as the applied potential decreased, suggesting that hydrogen is detrimental to the mechanical properties. It is proposed that the environment, through the presence of hydrogen, enhances IG cracking by enhancing the matrix dislocation mobility. This is based on observations that dislocation-controlled creep controls the IG cracking of controlled-purity Ni-xCr-9Fe-yC in argon at 360 C and grain boundary cavitation and sliding results that show the environmental enhancement of the creep rate is primarily due to an increase in matrix plastic deformation. However, controlled potential CLT experiments did not exhibit a change in the creep rate as the applied potential decreased. While this does not clearly support hydrogen assisted creep, the material may already be saturated with hydrogen at these applied potentials and thus no effect was realized. Chromium and carbon decrease the IG cracking in high purity and primary water by increasing the creep resistance. The surface film does not play a significant role in the creep or IG cracking behavior under the conditions investigated

  9. Corrosion and hydriding behaviour of some Zr 2.5 wt% Nb alloys in water, steam and various gases at high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Dalgaard, S. B.

    1962-05-15

    Fuel sheaths and pressure tubes in Canadian power reactors are at present made from Zircaloy-2. Mechanical properties of a suitably heat treated Zr 2.5 wt% Nb alloy are superior to those of Zircaloy-2, but any new alloy must have resistance to corrosion and hydriding by the coolant and by the gas that insulates the pressure tube from the cold moderator. Exposed to water at temperatures up to 325{sup o}C, the Zr 2.5 wt% Nb alloy has corrosion resistance acceptable for power reactors. Resistance to air and carbon dioxide is less favourable. Addition of tin, or iron and chromium, to the base alloy have little effect on the corrosion resistance, but the addition of copper reduces corrosion in water and steam to some extent and in air and carbon dioxide to a greater extent. Studies of the effect of heat treatment suggest that the amount of niobium in a solid-solution controls the rate of oxidation and hydriding and that concentration, size and distribution of second phase is of little importance. Initial results obtained in NRX indicate that a thermal flux of 3-7 x 10{sup 13} n/cm{sup 2}/sec has little or no effect on oxidation and hydriding in high temperature water. (author)

  10. Experiments with the low-melting indium-bismuth alloy system

    International Nuclear Information System (INIS)

    Krepski, R.P.

    1992-01-01

    The following is a laboratory experiment designed to create an interest in and to further understanding of materials science. The primary audience for this material is the junior high school or middle school science student having no previous familiarity with the material, other than some knowledge of temperature and the concepts of atoms, elements, compounds, and chemical reactions. The objective of the experiment is to investigate the indium-bismuth alloy system. Near the eutectic composition, the liquidus is well below the boiling point of water, allowing simple, minimal hazard casting experiments. Such phenomena as metal oxidation, formation of intermetallic compound crystals, and an unusual volume increase during solidification could all be directly observed. A key concept for students to absorb is that properties of an alloy (melting point, mechanical behavior) may not correlate with simple interpolation of properties of the pure components. Discussion of other low melting metals and alloys leads to consideration of environmental and toxicity issues, as well as providing some historical context. Wetting behavior can also be explored

  11. Fatigue crack propagation in aluminum-lithium alloys

    Science.gov (United States)

    Rao, K. T. V.; Ritchie, R. O.; Piascik, R. S.; Gangloff, R. P.

    1989-01-01

    The principal mechanisms which govern the fatigue crack propagation resistance of aluminum-lithium alloys are investigated, with emphasis on their behavior in controlled gaseous and aqueous environments. Extensive data describe the growth kinetics of fatigue cracks in ingot metallurgy Al-Li alloys 2090, 2091, 8090, and 8091 and in powder metallurgy alloys exposed to moist air. Results are compared with data for traditional aluminum alloys 2024, 2124, 2618, 7075, and 7150. Crack growth is found to be dominated by shielding from tortuous crack paths and resultant asperity wedging. Beneficial shielding is minimized for small cracks, for high stress ratios, and for certain loading spectra. While water vapor and aqueous chloride environments enhance crack propagation, Al-Li-Cu alloys behave similarly to 2000-series aluminum alloys. Cracking in water vapor is controlled by hydrogen embrittlement, with surface films having little influence on cyclic plasticity.

  12. Stress corrosion cracking behaviour of low alloy steels in high temperature water: Description and results from modelling

    International Nuclear Information System (INIS)

    Tirbonod, B.

    2001-01-01

    The initiation and growth of a crack by stress and corrosion in the low alloy steels used for the pressure vessels of Boiling Water Reactors may affect the availability and safety of the plant. This paper presents a new model for stress corrosion cracking of the low alloy steels in high temperature water. The model, based on observations, assumes the crack growth mechanism to be based on an anodic dissolution and cleavage. The main results deal with the position of the dissolution cell found at the crack tip, and with the identification of the parameters sensitive to crack growth, among which are the electrolyte composition and the cleavage length. The model is conservative, in qualitative agreement with measurements conducted at PSI, and may be extended to other metal-environment systems. (author)

  13. Corrosion of high-density sintered tungsten alloys

    International Nuclear Information System (INIS)

    Batten, J.J.; Moore, B.T.

    1989-01-01

    In comparative corrosion tests, the corrosion resistance of an Australian tungsten alloy (95% W, 3.5% Ni, 1.5% Fe) was found to be superior to three other tungsten alloys and, under certain conditions, even more corrosion-resistant than pure tungsten. Corrosion resistance was evaluated after immersion in both distilled water and 5% sodium chloride solutions, and in cyclic humidity and salt mist environments. For all but the Australian alloy, the rate of corrosion in sodium chloride solution was markedly less than that in distilated water. In all cases, alloys containing copper had the greatest corrosion rates. Corrosion mechanisms were investigated using a scanning electron microscope, analysis of corrosion products and galvanic corrosion studies. For the alloys, corrosion was attributed primarily to a galvanic reaction. Whether the tungsten or binder phase of the alloy became anodic, and thus was attacked preferentially, depended upon alloy composition and corrosion environment. 16 refs., 4 tabs., 4 figs

  14. Assessment of possibility of primary water stress corrosion cracking occurrence based on residual stress analysis in pressurizer safety nozzle of nuclear power plant

    International Nuclear Information System (INIS)

    Lee, Kyoung Soo; Kim, W.; Lee, Jeong Geun

    2012-01-01

    Primary water stress corrosion cracking (PWSCC) is a major safety concern in the nuclear power industry worldwide. PWSCC is known to initiate only in the condition in which sufficiently high tensile stress is applied to alloy 600 tube material or alloy 82/182 weld material in pressurized water reactor operating environments. However, it is still uncertain how much tensile stress is required to generate PWSCC or what causes such high tensile stress. This study was performed to predict the magnitude of weld residual stress and operating stress and compare it with previous experimental results for PWSCC initiation. For the study, a pressurizer safety nozzle was selected because it is reported to be vulnerable to PWSCC in overseas plants. The assessment was conducted by numerical analysis. Before performing stress analysis for plant conditions, a preliminary mock-up analysis was done. The result of the preliminary analysis was validated by residual stress measurement in the mockup. After verification of the analysis methodology, an analysis under plant conditions was conducted. The analysis results show that the stress level is not high enough to initiate PWSCC. If a plant is properly welded and operated, PWSCC is not likely to occur in the pressurizer safety nozzle.

  15. Dendrite tungsten liquation in molybdenum alloys

    International Nuclear Information System (INIS)

    Kantor, M.M.; Ageeva, E.N.; Kolotinskij, V.N.

    1992-01-01

    A study was made on primary crystallization structure of ingots of Mo-W-B system alloys with electron microscopy were used to establish, that cells and cellular dendrites were the main elements of primary crystallization structure. Method of local X-ray spectral analysis enabled to establish, that intracrystallite liquation at cellular growth developed more intensively, as compared to the case of cellular dendrite formation. Change of boron content in alloys didn't practically affect the degree of development of intracrystallite W liquation in Mo

  16. Corrosion-resistant amorphous alloy ribbons for electromagnetic filtration of iron rusts from water

    International Nuclear Information System (INIS)

    Kawashima, Asahi; Asami, Katsuhiko; Sato, Takeaki; Hashimoto, Koji

    1985-01-01

    An attempt was made to use corrosion-resistant amorphous Fe-9Cr-13P-7C alloy ribbons as an electromagnetic filter material for trapping various iron rusts suspended in water at 40 0 C. The ferrimagnetic Fe 3 O 4 rust was trapped with the 100 % efficiency and paramagnetic rusts such as α-Fe 2 O 3 , α-FeOOH and amorphous ferric oxyhydroxide were trapped with certain efficiencies at the magnetic field strength of 0.5-10 kOe. The regeneration of the filter by back-washing was easy. The trapping capacity of electromagnetic filter was proportional to the edge length of the filter material where the high magnetic field strength existed. Therefore, melt-spun thin and narrow amorphous alloy ribbons having the high corrosion resistance have the potential utility as electromagnetic filter material. (author)

  17. Research and service experience with environmentally assisted cracking of low-alloy steel

    Energy Technology Data Exchange (ETDEWEB)

    Hickling, J. [Electric Power Reasearch Inst., Palo Alto, CA (United States); Seifert, H.P.; Ritter, S. [Paul Scherrer Inst. (Switzerland)

    2005-01-01

    Environmentally assisted cracking (EAC) of carbon and low-alloy steels has been identified as a possible degradation mechanism for pressure vessels and piping in nuclear power plants. Selected aspects of research and service experience with cracking of these materials in high-temperature water are reviewed, with special emphasis on the primary pressure boundary in boiling water reactors. The main factors controlling EAC susceptibility under reactor conditions are discussed with regard to both crack initiation and crack growth. The adequacy and conservatism of the relevant engineering criteria for component design and disposition of detected or postulated flaws are evaluated in the context of recent research results, e.g., on the effects of so-called ''ripple loading'' or of water chemistry transients. Finally, the relevant operating experience over the last 30 years is briefly summarized and compared with the background knowledge which has been accumulated in more recent laboratory experiments. Some of the insights gained in this work may also be of value in improving understanding and prediction of the EAC behavior of carbon and low-alloy steels in certain fossil plant components, if appropriate allowances are made for differences in temperature and water chemistry. (orig.)

  18. PREPARATION OF ACTINIDE-ALUMINUM ALLOYS

    Science.gov (United States)

    Moore, R.H.

    1962-09-01

    BS>A process is given for preparing alloys of aluminum with plutonium, uranium, and/or thorium by chlorinating actinide oxide dissolved in molten alkali metal chloride with hydrochloric acid, chlorine, and/or phosgene, adding aluminum metal, and passing air and/or water vapor through the mass. Actinide metal is formed and alloyed with the aluminum. After cooling to solidification, the alloy is separated from the salt. (AEC)

  19. Strain rate and temperature effects on the stress corrosion cracking of Inconel 600 steam generator tubing in the primary water conditions

    International Nuclear Information System (INIS)

    Kim, U.C.; van Rooyen, D.

    1985-01-01

    A single heat of Inconel Alloy 600 was examined in this work, using slow strain rate tests (SSRT) in simulated primary water at temperatures of 325 0 -345 0 -365 0 C. The best measure of stress corrosion cracking (SCC) was percent SCC present on the fracture surface. Strain rate did not seem to affect crack growth rate significantly, but there is some question about the accuracy of calculating these values in the absence of a direct indication of when a crack initiates. Demarcation was determined between domains of temperature/strain rate where SCC either did, or did not, occur. Slower extension rates were needed to produce SCC as the temperature was lowered. 10 figs

  20. Variant selection of primary, secondary and tertiary twins in a deformed Mg alloy

    International Nuclear Information System (INIS)

    Mu, Sijia; Jonas, John J.; Gottstein, Günter

    2012-01-01

    Samples of magnesium alloy AZ31 were deformed in plane strain compression in a channel die at 100 °C and a strain rate of 5 × 10 −3 s −1 . The initial texture was favorably oriented for extension twinning. At a true strain of ε = −0.11, many primary extension twins were observed to consume their parent grains completely. Furthermore, numerous secondary contraction twins formed within the primary extension twins and some tertiary extension twins grew within the secondary contraction twins. The orientations of the parent grains and all three generations of twins were measured. The twin variants selected during each of the three stages of twinning were determined by electron backscatter diffraction techniques and the absent potential twin variants were also identified. The way in which the selected primary extension twins grow so as to consume the parent grains and contact all the neighboring grains is explained in terms of the accommodation strains imposed on the neighboring grains. The analysis shows that the primary twin selected is not necessarily the variant with the highest Schmid factor but the one that requires the least accommodation work in most of the neighboring grains. The same principle was found to hold for the secondary and tertiary twins. By contrast, potential high Schmid factor twins that required the consumption of appreciable accommodation energy did not form. A Taylor simulation produced similar results and indicates that the accommodation strain concept is consistent with the principle of the minimization of plastic work.

  1. The primary processes by impact of ionizing radiations with water

    International Nuclear Information System (INIS)

    Znamirovschi, V.; Mastan, I.; Cozar, O.

    1976-01-01

    The problem concerning primary processes in radiolysis of water is discussed. The results on the excitation and ionization of water molecule, dissociation of the parent-molecular ion of water and dissociation of excited molecule of water are presented. (author)

  2. Oxidization and stress corrosion cracking initiation of austenitic alloys in supercritical water

    International Nuclear Information System (INIS)

    Behnamian, Y.; Li, M.; Luo, J.L.; Chen, W.X.; Zheng, W.; Guzonas, D.A.

    2012-01-01

    This study determined the stress corrosion cracking behaviour of austenitic alloys in pure supercritical water. Austenitic stainless steels 310S, 316L, and Inconel 625 were tested as static capsule samples at 500 o C for up to 5000 h. After that period, crack initiations were readily observed in all samples, signifying susceptibility to stress corrosion cracking. The microcracks in 316L stainless steel and Inconel 625 were almost intergranular, whereas transgranular microcrack initiation was observed in 310S stainless steel. (author)

  3. Oxidization and stress corrosion cracking initiation of austenitic alloys in supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Behnamian, Y.; Li, M.; Luo, J.L.; Chen, W.X. [Univ. of Alberta, Dept. of Chemical and Materials Engineering, Edmonton, Alberta (Canada); Zheng, W. [Materials Technology Laboratory, NRCan, Ottawa, Ontario (Canada); Guzonas, D.A. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    This study determined the stress corrosion cracking behaviour of austenitic alloys in pure supercritical water. Austenitic stainless steels 310S, 316L, and Inconel 625 were tested as static capsule samples at 500{sup o}C for up to 5000 h. After that period, crack initiations were readily observed in all samples, signifying susceptibility to stress corrosion cracking. The microcracks in 316L stainless steel and Inconel 625 were almost intergranular, whereas transgranular microcrack initiation was observed in 310S stainless steel. (author)

  4. Vaporization Rate Analysis of Primary Cooling Water from Reactor PUSPATI TRIGA (RTP) Tank

    International Nuclear Information System (INIS)

    Tonny Anak Lanyau; Mohd Fazli Zakaria; Yahya Ismail

    2011-01-01

    Primary cooling system consists of pumps, heat exchangers, probes, a nitrogen-16 diffuser and associated valves is connected to the reactor TRIGA PUSPATI (RTP) tank by aluminium pipes. Both the primary cooling system and the reactor tank is filled with demineralized light water (H 2 O), which serves as a coolant, moderator as well as shielding. During reactor operation, vaporization in the reactor tank will reduce the primary water and contribute to the formation of vapor in the reactor hall. The vaporization may influence the function of the water subsequently may affect the safety of the reactor operation. It is essential to know the vaporization rate of the primary water to ensure its functionality. This paper will present the vaporization rate of the primary cooling water from the reactor tank and the influence of temperature of the water in the reactor tank to the vaporization rate. (author)

  5. Rapid Diffusion and Nanosegregation of Hydrogen in Magnesium Alloys from Exposure to Water.

    Science.gov (United States)

    Brady, Michael P; Ievlev, Anton V; Fayek, Mostafa; Leonard, Donovan N; Frith, Matthew G; Meyer, Harry M; Ramirez-Cuesta, Anibal J; Daemen, Luke L; Cheng, Yongqiang; Guo, Wei; Poplawsky, Jonathan D; Ovchinnikova, Olga S; Thomson, Jeffrey; Anovitz, Lawrence M; Rother, Gernot; Shin, Dongwon; Song, Guang-Ling; Davis, Bruce

    2017-11-01

    Hydrogen gas is formed when Mg corrodes in water; however, the manner and extent to which the hydrogen may also enter the Mg metal is poorly understood. Such knowledge is critical as stress corrosion cracking (SCC)/embrittlement phenomena limit many otherwise promising structural and functional uses of Mg. Here, we report via D 2 O/D isotopic tracer and H 2 O exposures with characterization by secondary ion mass spectrometry, inelastic neutron scattering vibrational spectrometry, electron microscopy, and atom probe tomography techniques direct evidence that hydrogen rapidly penetrated tens of micrometers into Mg metal after only 4 h of exposure to water at room temperature. Further, technologically important microalloying additions of mechanical properties of Mg significantly increased the extent of hydrogen ingress, whereas Al additions in the 2-3 wt % range did not. Segregation of hydrogen species was observed at regions of high Mg/Zr/Nd nanoprecipitate density and at Mg(Zr) metastable solid solution microstructural features. We also report evidence that this ingressed hydrogen was unexpectedly present in the alloy as nanoconfined, molecular H 2 . These new insights provide a basis for strategies to design Mg alloys to resist SCC in aqueous environments as well as potentially impact functional uses such as hydrogen storage where increased hydrogen uptake is desired.

  6. Preliminary assessment of stress corrosion cracking of nickel based alloy 182 in nuclear reactor environment

    International Nuclear Information System (INIS)

    Lima, Luciana Iglesias Lourenco; Bracarense, Alexandre Queiroz; Schvartzman, Monica Maria de Abreu Mendonca; Quinan, Marco Antonio Dutra

    2010-01-01

    Stress corrosion crack (SCC) in a primary circuit of a nuclear pressurized water reactor consists of a degradation process in which aggressive media, stress and material susceptibility are present. Over the last thirty years, SCC has been observed in dissimilar metal welds. This study presents a comparative work between the SCC in the alloy 182 filler metal weld in two different hydrogen concentrations (25 e 50 cm 3 H 2 /kg H 2 O) in primary water. The susceptibility to stress corrosion cracking was assessed using the slow strain rate tensile (SSRT) test. The results of the SSRT test indicated that the material is more susceptible to SCC at 25 cm 3 H 2 /kg H 2 O. (author)

  7. Aeronautical Industry Requirements for Titanium Alloys

    Science.gov (United States)

    Bran, D. T.; Elefterie, C. F.; Ghiban, B.

    2017-06-01

    The project presents the requirements imposed for aviation components made from Titanium based alloys. A significant portion of the aircraft pylons are manufactured from Titanium alloys. Strength, weight, and reliability are the primary factors to consider in aircraft structures. These factors determine the requirements to be met by any material used to construct or repair the aircraft. Many forces and structural stresses act on an aircraft when it is flying and when it is static and this thesis describes environmental factors, conditions of external aggression, mechanical characteristics and loadings that must be satisfied simultaneously by a Ti-based alloy, compared to other classes of aviation alloys (as egg. Inconel super alloys, Aluminum alloys).For this alloy class, the requirements are regarding strength to weight ratio, reliability, corrosion resistance, thermal expansion and so on. These characteristics additionally continue to provide new opportunities for advanced manufacturing methods.

  8. Development of water chemistry diagnosis system for BWR primary loop

    International Nuclear Information System (INIS)

    Nagase, Makoto; Asakura, Yamato; Sakagami, Masaharu; Uchida, Shunsuke; Ohsumi, Katsumi.

    1988-01-01

    The prototype of a water chemistry diagnosis system for BWR primary loop has been developed. Its purposes are improvement of water chemistry control and reduction of the work burden on plant chemistry personnel. It has three main features as follows. (1) Intensifying the observation of water chemistry conditions by variable sampling intervals based on the on-line measured data. (2) Early detection of water chemistry data trends using a second order regression curve which is calculated from the measured data, and then searching the cause of anomaly if anything (3) Diagnosis of Fe concentration in feedwater using model simulations, in order to lower the radiation level in the primary system. (author)

  9. Experimental study on forced convection boiling heat transfer on molten alloy

    International Nuclear Information System (INIS)

    Nishimura, Satoshi; Ueda, Nobuyuki; Nishi, Yoshihisa; Furuya, Masahiro; Kinoshita, Izumi

    1999-01-01

    In order to clarify the characteristics of forced convection boiling heat transfer on molten metal, basic experiments have been carried out with subcooled water flowing on molten Wood's alloy pool surface. In these experiments, water flows horizontally in a rectangular duct. A cavity filled with Wood's alloy is present in a portion of the bottom of the duct. Wood's alloy is heated by a copper conductor at the bottom of the cavity. The experiments have been carried out with various velocities and subcoolings of water, and temperature of Wood's alloy. Boiling curves on the molten alloy surface were obtained and compared with that on a solid heat transfer surface. It is observed that the boiling curve on molten alloy is in a lower superheat region than the boiling curve on a solid surface. This indicates that the heat transfer performance of forced convection boiling on molten alloy is enhanced by increase of the heat transfer area, due to oscillation of the surface and fragmentation of molten alloy

  10. Water quality control device and water quality control method for reactor primary coolant system

    International Nuclear Information System (INIS)

    Wada, Yoichi; Ibe, Eishi; Watanabe, Atsushi.

    1995-01-01

    The present invention is suitable for preventing defects due to corrosion of structural materials in a primary coolant system of a BWR type reactor. Namely, a concentration measuring means measures the concentration of oxidative ingredients contained in a reactor water. A reducing electrode is disposed along a reactor water flow channel in the primary coolant system and reduces the oxidative ingredients. A reducing counter electrode is disposed along the reactor water flow channel in the primary coolant system, and electrically connected to the reducing electrode. The reactor structural materials are used as a reference electrode providing a reference potential to the reducing electrode and the reducing counter electrode. A potential control means controls the potential of the reducing electrode relative to the reference potential based on the signals from the concentration measuring means. A stable reference potential in a region where an effective oxygen concentration is stable can be obtained irrespective of the change of operation conditions by using the reactor structural materials disposed to a boiling region in the reactor core as a reference electrode. As a result, the water quality can be controlled at high accuracy. (I.S.)

  11. Water and oil wettability of anodized 6016 aluminum alloy surface

    Science.gov (United States)

    Rodrigues, S. P.; Alves, C. F. Almeida; Cavaleiro, A.; Carvalho, S.

    2017-11-01

    This paper reports on the control of wettability behaviour of a 6000 series aluminum (Al) alloy surface (Al6016-T4), which is widely used in the automotive and aerospace industries. In order to induce the surface micro-nanostructuring of the surface, a combination of prior mechanical polishing steps followed by anodization process with different conditions was used. The surface polishing with sandpaper grit size 1000 promoted aligned grooves on the surface leading to static water contact angle (WCA) of 91° and oil (α-bromonaphthalene) contact angle (OCA) of 32°, indicating a slightly hydrophobic and oleophilic character. H2SO4 and H3PO4 acid electrolytes were used to grow aluminum oxide layers (Al2O3) by anodization, working at 15 V/18° C and 100 V/0 °C, respectively, in one or two-steps configuration. Overall, the anodization results showed that the structured Al surfaces were hydrophilic and oleophilic-like with both WCA and OCA below 90°. The one-step configuration led to a dimple-shaped Al alloy surface with small diameter of around 31 nm, in case of H2SO4, and with larger diameters of around 223 nm in case of H3PO4. The larger dimples achieved with H3PO4 electrolyte allowed to reach a slight hydrophobic surface. The thicker porous Al oxide layers, produced by anodization in two-step configuration, revealed that the liquids can penetrate easily inside the non-ordered porous structures and, thus, the surface wettability tended to superhydrophilic and superoleophilic character (CA OCA. This inversion in favour of the hydrophilic-oleophobic surface behaviour is of great interest either for lubrication of mechanical components or in water-oil separation process.

  12. Salvaging of service exposed cast alloy 625 cracker tubes of ammonia based Heavy Water Plants

    International Nuclear Information System (INIS)

    Kumar, Niraj; Misra, B.; Mahajan, M.P.; Mittra, J.; Sundararaman, M.; Chakravartty, J.K.

    2006-01-01

    In ammonia based heavy water plants, cracking of ammonia vapour, enriched in deuterium is carried out inside a cracker tube, packed with catalyst. These cracker tubes are made of alloy 625 (either wrought or cast) having dimensions of about 12.5 metres long, 88 mm outer diameter and 7.9 mm wall thickness. Seventy such tubes are housed in a typical ammonia cracker unit. The anticipated design life of such tube is 1,00,000 hrs. when operated at 720 degC based on creep as main degradation mechanism. Presently, these tubes are being operated at 680 degC skin temperature. Alloy 625 tubes are costly and normally not manufactured in India and are being imported. The cast alloy 625 cracker tubes have outlived their design life of 100,000 hrs. Therefore it has been decided to salvage the cast cracker tubes and extend the life further as it had already been done for wrought tubes. Similar to the earlier attempt of resolutionising of wrought alloy 625 tubes, efforts are in progress to salvage these cast tubes. In this study, cast tubes samples were subjected to solution-annealing treatment at two different temperatures, 1100degC and 1160degC respectively for two hrs. Mechanical properties along with the microstructure of the samples, which were resolutionized at 1160degC were comparable with that of virgin material. The 12.5 metres long cast alloy 625 cracker tubes will also be shortly solution-annealed in a specially designed resistance heating furnace after completing some more tests. (author)

  13. Prediction of pure water stress corrosion cracking (PWSCC) in nickel base alloys using crack growth rate models

    International Nuclear Information System (INIS)

    Thompson, C.D.; Krasodomski, H.T.; Lewis, N.; Makar, G.L.

    1995-01-01

    The Ford/Andresen slip dissolution SCC model, originally developed for stainless steel components in BWR environments, has been applied to Alloy 600 and Alloy X-750 tested in deaerated pure water chemistry. A method is described whereby the crack growth rates measured in compact tension specimens can be used to estimate crack growth in a component. Good agreement was found between model prediction and measured SCC in X-750 threaded fasteners over a wide range of temperatures, stresses, and material condition. Most data support the basic assumption of this model that cracks initiate early in life. The evidence supporting a particular SCC mechanism is mixed. Electrochemical repassivation data and estimates of oxide fracture strain indicate that the slip dissolution model can account for the observed crack growth rates, provided primary rather than secondary creep rates are used. However, approximately 100 cross-sectional TEM foils of SCC cracks including crack tips reveal no evidence of enhanced plasticity or unique dislocation patterns at the crack tip or along the crack to support a classic slip dissolution mechanism. No voids, hydrides, or microcracks are found in the vicinity of the crack tips creating doubt about classic hydrogen related mechanisms. The bulk oxide films exhibit a surface oxide which is often different than the oxides found within a crack. Although bulk chromium concentration affects the rate of SCC, analytical data indicates the mechanism does not result from chromium depletion at the grain boundaries. The overall findings support a corrosion/dissolution mechanism but not one necessarily related to slip at the crack tip

  14. Modeling of grain boundary stresses in Alloy 600

    Energy Technology Data Exchange (ETDEWEB)

    Kozaczek, K.J. [Oak Ridge National Lab., TN (United States); Sinharoy, A.; Ruud, C.O. [Pennsylvania State Univ., University Park, PA (United States); Mcllree, A.R. [Electric Power Research Inst., Palo Alto, CA (United States)

    1995-04-01

    Corrosive environments combined with high stress levels and susceptible microstructures can cause intergranular stress corrosion cracking (IGSCC) of Alloy 600 components on both primary and secondary sides of pressurized water reactors. One factor affecting the IGSCC is intergranular carbide precipitation controlled by heat treatment of Alloy 600. This study is concerned with analysis of elastic stress fields in vicinity of M{sub 7}C{sub 3} and M{sub 23}C{sub 6} carbides precipitated in the matrix and at a grain boundary triple point. The local stress concentration which can lead to IGSCC initiation was studied using a two-dimensional finite element model. The intergranular precipitates are more effective stress raisers than the intragranular precipitates. The combination of the elastic property mismatch and the precipitate shape can result in a local stress field substantially different than the macroscopic stress. The maximum local stresses in the vicinity of the intergranular precipitate were almost twice as high as the applied stress.

  15. Modelling of zirconium alloys corrosion in LWRs

    International Nuclear Information System (INIS)

    Kritskij, V.G.; Berezina, I.G.; Kritskij, A.V.; Stjagkin, P.S.

    1999-01-01

    Chemical parameters, that exerted effect on Zr+1%Nb alloy corrosion and deserved consideration during reactor operation, were defined and a model was developed to describe the influence of physical and chemical parameters on zirconium alloys corrosion in nuclear power plants. The model is based on the correlation between the zirconium oxide solubility in high-temperature water under the influence of the chemical parameters and the measured values of fuel cladding corrosion under LWR conditions. The intensity of fuel cladding corrosion in the primary circuits depends on the coolant water quality, growth of iron oxide deposits and vaporization portion. Mathematically, the oxidation rate can be expressed as a sum of heat and radiation components. The temperature dependence on the oxidation rate can be described by the Arrenius equation. The radiation component of Zr uniform corrosion equation is a function of several factors such as neutron fluency, the temperature the metallurgical composition and et. We assume that the main factor is the changing of water chemistry and the H 2 O 2 concentration play the determinative role. Probably, the influence of H 2 O 2 is based on the formation of unstable compound ZrO 3 ·nH 2 O and Zr(OH) 4 with high solubility. The validity of the used formulae was confirmed by corrosion measurements on WWER and RBMK fuel cladding. The model can be applied for calculating the reliability of nuclear fuel operation. (author)

  16. Remediation of phosphate-contaminated water by electrocoagulation with aluminium, aluminium alloy and mild steel anodes.

    Science.gov (United States)

    Vasudevan, Subramanyan; Lakshmi, Jothinathan; Jayaraj, Jeganathan; Sozhan, Ganapathy

    2009-05-30

    The present study provides an electrocoagulation process for the remediation of phosphate-contaminated water using aluminium, aluminium alloy and mild steel as the anodes and stainless steel as the cathode. The various parameters like effect of anode materials, effect of pH, concentration of phosphate, current density, temperature and co-existing ions, and so forth, and the adsorption capacity was evaluated using both Freundlich and Langmuir isotherm models. The adsorption of phosphate preferably fitting the Langmuir adsorption isotherm suggests monolayer coverage of adsorbed molecules. The results showed that the maximum removal efficiency of 99% was achieved with aluminium alloy anode at a current density of 0.2 A dm(-2), at a pH of 7.0. The adsorption process follows second-order kinetics.

  17. Radiation corrosion in aluminum alloy bellows

    International Nuclear Information System (INIS)

    Konno, Osamu

    1987-01-01

    Testing was carried out in which materials for vacuum devices (Al, Ti, Cu, SUS) are exposed to electron beams (50 MeV, average current 80 μA) to determine the changes in the quantity, partial pressure and composition of the gases released from the materials. The test appratus used are made of Al alloys alone. During the test, vacuum leak is found in the Al alloy bellows used in the drive device. The leak is found to result from corrosion caused by water. The surface structure is analyzed by SEM, EPMA, ESCA and IMA. It is confirmed that the Al alloy used as material for the bellows if highly resistant to corrosion. It is concluded that it is necessary to use high purity cooling water to prevent the cooling water from causing corrosion. It has been reported that high purity aluminum is very high in resistance to corrosion. Based on these measurements and considerations, it is suggested that when aluminum is to be used as material for vacuum devices in an accelerator, it is required to provide protection film on its surface to prevent corrosion or to use cooling water pipes cladded with pure aluminum and an aluminum alloy. In addition, the temperature of the cooling water should be set after adequately considering the environmental conditions in the room. (Nogami, K.)

  18. Dose rate determining factors of PWR primary water

    International Nuclear Information System (INIS)

    Terachi, Takumi; Kuge, Toshiharu; Nakano, Nobuo

    2014-01-01

    The relationship between dose rate trends and water chemistry has been studied to clarify the determining factors on the dose rates. Therefore dose rate trends and water chemistry of 11 PWR plants of KEPCO (Kansai Electric Power Co., Inc.) were summarized. It is indicated that the chemical composition of the oxide film, behaviour of corrosion products and Co-58/Co-60 ratio in the primary system have effected dose rate trends based on plant operation experiences for over 40 years. According to plant operation experiences, the amount of Co-58 has been decreasing with the increasing duration of SG (Steam Generator) usage. It is indicated that the stable oxide film formation on the inner surface of SG tubing, is a major beneficial factor for radiation sources reduction. On the other hand, the reduction of the amount of Co-60 for the long term has been not clearly observed especially in particular high dose plants. The primary water parameters imply that considering release and purification balance on Co-59 is important to prevent accumulation of source term in primary water. In addition, the effect of zinc injection, which relates to the chemical composition of oxide film, was also assessed. As the results, the amount of radioactive Co has been clearly decreased. The decreasing trend seems to correlate to the half-life of Co-60, because it is considered that the injected zinc prevents the uptake of radioactive Co into the oxide film on the inner surface of the components and piping. In this paper, the influence of water chemistry and the replacement experiences of materials on the dose rates were discussed. (author)

  19. Superconductivity in zirconium-rhodium alloys

    Science.gov (United States)

    Zegler, S. T.

    1969-01-01

    Metallographic studies and transition temperature measurements were made with isothermally annealed and water-quenched zirconium-rhodium alloys. The results clarify both the solid-state phase relations at the Zr-rich end of the Zr-Rh alloy system and the influence upon the superconducting transition temperature of structure and composition.

  20. Corrosion performance of new Zircaloy-2-based alloys

    International Nuclear Information System (INIS)

    Rudling, P.; Mikes-Lindbaeck, M.; Lethinen, B.; Andren, H.O.; Stiller, K.

    1994-01-01

    A material development project was initiated to develop a new zirconium alloy, outside the ASTM specifications for Zircaloy-2 and Zircaloy-4, with optimized hydriding and corrosion properties for both boiling water reactors and pressurized water reactors. A number of different alloys were manufactured. These alloys were long-term corrosion tested in autoclaves at 400 C in steam. Also, a 520 C/24 h steam test was carried out. The zirconium metal microstructure and the chemistry of precipitates were characterized by analytical electron microscopy. The metal matrix chemistry was determined by atom probe analysis. The paper describes the correlations between corrosion material performance and zirconium alloy microstructure

  1. Mechanical and microstructural characterization of the nickel base alloy (Alloy 600) after heat treatment

    International Nuclear Information System (INIS)

    Fernandes, Stela Maria de Carvalho

    1993-01-01

    The characterization of microstructural and mechanical properties of cold rolled and heat treated alloys 600 made in Brazil were investigated. The recovery and recrystallization behavior as well as solubilization and aging have been studied using optical, scanning electron and transmission electron microscopy. Microhardness and tensile testing have been carried out. The recovery process of the cold rolled alloy 600 occurred until 600 deg C and the recrystallization stage was situated between 600 and 850 deg C. The primary recrystallization temperature was obtained at 850 deg C after 1 hour (isochronal heat treatments). The aged alloy 600 shows carbide precipitation on grains bu with ductility maintenance. (author)

  2. SOLUTION TREATMENT EFFECT ON MICROSTRUCTURE AND MECHANICAL PROPERTIES OF AUTOMOTIVE CAST ALLOY

    Directory of Open Access Journals (Sweden)

    Eva Tillová

    2012-02-01

    Full Text Available The contribution describes influence of the heat treatment (solution treatment at temperature 545°C and 565°C with different holding time 2, 4, 8, 16 and 32 hours; than water quenching at 40°C and natural aging at room temperature during 24 hours on mechanical properties (tensile strength and Brinell hardness and microstructure of the secondary AlSi12Cu1Fe automotive cast alloy. Mechanical properties were measured in line with EN ISO. A combination of different analytical techniques (light microscopy, scanning electron microscopy (SEM were therefore been used for study of microstructure. Solution treatment led to changes in microstructure includes the spheroidization and coarsening of eutectic silicon. The dissolution of precipitates and the precipitation of finer hardening phase further increase the hardness and tensile strength of the alloy. Optimal solution treatment (545°C/4 hours most improves mechanical properties and there mechanical properties are comparable with mechanical properties of primary AlSi12Cu1Fe alloy. Solution treatment at 565 °C caused testing samples distortion, local melting process and is not applicable for this secondary alloy with 12.5 % Si.

  3. Creep-rupture behavior of 2-1/4 Cr-1 Mo steel, Alloy 800H and Hastelloy Alloy X in a simulated HTGR helium environment

    International Nuclear Information System (INIS)

    Lai, G.Y.; Wolwowicz, R.J.

    1979-12-01

    Creep-rupture testing was conducted on 1 1/4 Cr-1 Mo steel, Alloy 800H and Hastelloy Alloy X in flowing helium containing nominal concentration of following gases: 1500 μatm H 2 , 450 μatm CO, 50 μatm CH 4 , 50 μatm H 2 O and 5 μatm CO 2 . This environment is believed to represent maximum permissible levels of impurities in the primary coolant for the steam-cycle system of a high-temperature gas-cooled reactor (HTGR) when it is operating continuously with a water and/or steam leak at technical specification limits. Two or three heats of material for each alloy were investigated. Tests were conducted at 482 0 C and 760 0 C (1200 0 F and 1400 0 F) for Alloy 800H, and at 760 0 C and 871 0 C (1400 0 F and 1600 0 F) for Hastelloy Alloy X for times up to 10,000 h. Selected tests were performed on same heat of material in both air and helium environments to make a direct comparison of creep-rupture behaviors between two environments. Metallurgical evaluation was performed on selected post test specimens with respect to gas-metal interactions which included oxidation, carburization and/or decarburization. Correlation between gaseous corrosion and creep-rupture behavior was attempted. Limited tests were also performed to investigate the specimen size effects on creep-rupture behavior in the helium environment

  4. Aluminium alloys containing iron and nickel

    International Nuclear Information System (INIS)

    Coriou, H.; Fournier, R.; Grall, L.; Hure, J.; Herenguel, J.; Lelong, P.

    1958-01-01

    The first part of this report addresses mechanism, kinetics and structure factors of aluminium alloys containing iron and nickel in water and high temperature steam. The studied alloys contain from 0.3 to 0.7 per cent of iron, and 0.2 to 1.0 per cent of nickel. Corrosion resistance and corrosion structure have been studied. The experimental installation, process and samples are presented. Corrosion structures in water at 350 C are identified and discussed (structure of corrosion products, structure of metal-oxide interface), and then in steam at different temperatures (350-395 C). Corrosion kinetics is experimentally studied (weight variation in time) in water at 350 C and in steam at different temperatures. Reactions occurring at over-heated steam (more than 400 C) are studied, and the case of welded alloys is also addressed. The second part addresses the metallurgical mechanism and processes influencing aluminium alloy resistance to corrosion by high temperature water as it appeared that separated phases protect the solid solution through a neighbourhood action. In order to avoid deep local corrosions, it seems necessary to multiply protective phases in an as uniform as possible way. Some processes enabling this result are described. They belong to conventional metallurgy or to powder metallurgy (with sintering and extrusion)

  5. 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.)

  6. The water footprint of energy consumption: an assessment of water requirements of primary energy carriers

    NARCIS (Netherlands)

    Gerbens-Leenes, P.W.; Hoekstra, A.Y.; Van der Meer, T.H.

    2007-01-01

    Gerbens-Leenes, P.W., Hoekstra, A.Y., Van der Meer, T.H., 2007. The water footprint of energy consumption: an assessment of water requirements of primary energy carriers. In: proceedings ‘First World Water Sustainability-Renewable Energy Congress and Exhibition’. 25-28 November 2007, Maastricht, the

  7. Conducting water chemistry of the secondary coolant circuit of VVER-based nuclear power plant units constructed without using copper containing alloys

    Science.gov (United States)

    Tyapkov, V. F.

    2014-07-01

    The secondary coolant circuit water chemistry with metering amines began to be put in use in Russia in 2005, and all nuclear power plant units equipped with VVER-1000 reactors have been shifted to operate with this water chemistry for the past seven years. Owing to the use of water chemistry with metering amines, the amount of products from corrosion of structural materials entering into the volume of steam generators has been reduced, and the flow-accelerated corrosion rate of pipelines and equipment has been slowed down. The article presents data on conducting water chemistry in nuclear power plant units with VVER-1000 reactors for the secondary coolant system equipment made without using copper-containing alloys. Statistical data are presented on conducting ammonia-morpholine and ammonia-ethanolamine water chemistries in new-generation operating power units with VVER-1000 reactors with an increased level of pH. The values of cooling water leaks in turbine condensers the tube system of which is made of stainless steel or titanium alloy are given.

  8. Effect of chloride and sulphate ions on the electrochemical corrosion behavior of alloy 800NG in PWR secondary water environment at 250 deg C

    International Nuclear Information System (INIS)

    Mansur, Fabio A.; Schvartzman, Monica Maria de A.M.; Quinan, Marco A.D.; Soares, Antonio E.G.; Nogueira, Pedro Henrique B.O.

    2013-01-01

    Alloy 800NG (nuclear grade) is used in nuclear steam generators (SG) as the tubing material for pressurized water reactors (PWRs) because of its high corrosion resistance. The corrosion resistance is due to the protective character of the oxide film formed on the tube surface by contact with the high temperature pressurized water. Nevertheless, corrosion has been the major cause of tube failures in nuclear SGs. The existing experience of different nuclear power plants shows that the water chemistry has an important role in maintaining the integrity of the protective oxide films. Many of such problems have been attributed to secondary side water chemistry conditions and excursions, many of which have been resulted from condenser cooling water ingress. Alloy 800 is known to undergo passivity breakdown and pitting in the presence of chloride ions under oxidative water conditions. In this work the effect of chloride and sulphate ions at various concentrations on the corrosion behavior of Alloy 800 tube at 250 deg C was investigated using the potentiodynamic anodic polarization technique. An active-passive transition occurred at 250 deg C in all studied conditions and the oxide film grown on surface showed greater porosity and lower resistance to localised corrosion in all studied conditions. (author)

  9. Passive Corrosion Behavior of Alloy 22

    International Nuclear Information System (INIS)

    R.B. Rebak; J.H. Payer

    2006-01-01

    Alloy 22 (NO6022) was designed to stand the most aggressive industrial applications, including both reducing and oxidizing acids. Even in the most aggressive environments, if the temperature is lower than 150 F (66 C) Alloy 22 would remain in the passive state having particularly low corrosion rates. In multi-ionic solutions that may simulate the behavior of concentrated ground water, even at near boiling temperatures, the corrosion rate of Alloy 22 is only a few nano-meters per year because the alloy is in the complete passive state. The corrosion rate of passive Alloy 22 decreases as the time increases. Immersion corrosion testing also show that the newer generation of Ni-Cr-Mo alloys may offer a better corrosion resistance than Alloy 22 only in some highly aggressive conditions such as in hot acids

  10. Aging management of major light water reactor components

    International Nuclear Information System (INIS)

    Shah, V.N.; Sinha, U.P.; Ware, A.G.

    1992-01-01

    Review of technical literature and field experience has identified stress corrosion cracking as one of the major degradation mechanisms for the major light water reactor components. Three of the stress corrosion cracking mechanisms of current concern are (a) primary water stress corrosion cracking (PWSCC) in pressurized water reactors, and (b) intergranular stress corrosion cracking (IGSCC) and (c) irradiation-assisted stress corrosion cracking (IASCC) in boiling water reactors. Effective aging management of stress corrosion cracking mechanisms includes evaluation of interactions between design, materials, stressors, and environment; identification and ranking of susceptible sites; reliable inspection of any damage; assessment of damage rate; mitigation of damage; and repair and replacement using corrosion-resistant materials. Management of PWSCC includes use of lower operating temperatures, reduction in residual tensile stresses, development of reliable inspection techniques, and use of Alloy 690 as replacement material. Management of IGSCC of nozzle and attachment welds includes use of Alloy 82 as weld material, and potential use of hydrogen water chemistry. Management of IASCC also includes potential use of hydrogen water chemistry

  11. Annex 5 - Fabrication of U-Al alloy

    International Nuclear Information System (INIS)

    Drobnjak, Dj.; Lazarevic, Dj.; Mihajlovic, A.

    1961-01-01

    Alloy U-Al with low content of aluminium is often used for fabrication of fuel elements because it is stable under moderate neutron flux density. Additionally this type of alloys show much better characteristics than pure uranium under reactor operating conditions (temperature, mechanical load, corrosion effect of water). This report contains the analysis of the phase diagram of U-Al alloy with low content of aluminium, applied procedure for alloying and casting with detailed description of equipment. Characteristics of the obtained alloy are described and conclusions about the experiment and procedure are presented [sr

  12. Requirements of titanium alloys for aeronautical industry

    Science.gov (United States)

    Ghiban, Brânduşa; Bran, Dragoş-Teodor; Elefterie, Cornelia Florina

    2018-02-01

    The project presents the requirements imposed for aeronatical components made from Titanium based alloys. Asignificant portion of the aircraft pylons are manufactured from Titanium alloys. Strength, weight, and reliability are the primary factors to consider in aircraft structures. These factors determine the requirements to be met by any material used to construct or repair the aircraft. Many forces and structural stresses act on an aircraft when it is flying and when it is static and this thesis describes environmental factors, conditions of external aggression, mechanical characteristics and loadings that must be satisfied simultaneously by a Ti-based alloy, compared to other classes of aviation alloys (as egg. Inconel super alloys, Aluminum alloys). For this alloy class, the requirements are regarding strength to weight ratio, reliability, corrosion resistance, thermal expansion and so on. These characteristics additionally continue to provide new opportunities for advanced manufacturing methods.

  13. Corrosion problems in light water nuclear reactors

    International Nuclear Information System (INIS)

    Berry, W.E.

    1984-01-01

    The corrosion problems encountered during the author's career are reviewed. Attention is given to the development of Zircaloys and attendant factors that affect corrosion; the caustic and chloride stress corrosion cracking (SCC) of austenitic stainless steel steam generator tubing; the qualification of Inconel Alloy 600 for steam generator tubing and the subsequent corrosion problem of secondary side wastage, caustic SCC, pitting, intergranular attack, denting, and primary side SCC; and SCC in weld and furnace sensitized stainless steel piping and internals in boiling water reactor primary coolants. Also mentioned are corrosion of metallic uranium alloy fuels; corrosion of aluminum and niobium candidate fuel element claddings; crevice corrosion and seizing of stainless steel journal-sleeve combinations; SCC of precipitation hardened and martensitic stainless steels; low temperature SCC of welded austenitic stainless steels by chloride, fluoride, and sulfur oxy-anions; and corrosion problems experienced by condensers

  14. Ni nanotube array-based electrodes by electrochemical alloying and de-alloying for efficient water splitting.

    Science.gov (United States)

    Teng, Xue; Wang, Jianying; Ji, Lvlv; Lv, Yaokang; Chen, Zuofeng

    2018-05-17

    The design of cost-efficient earth-abundant catalysts with superior performance for the hydrogen evolution reaction (HER) and oxygen evolution reaction (OER) is extremely important for future renewable energy production. Herein, we report a facile strategy for constructing Ni nanotube arrays (NTAs) on a Ni foam (NF) substrate through cathodic deposition of NiCu alloy followed by anodic stripping of metallic Cu. Based on Ni NTAs, the as-prepared NiSe2 NTA electrode by NiSe2 electrodeposition and the NiFeOx NTA electrode by dipping in Fe3+ solution exhibit excellent HER and OER performance in alkaline conditions. In these systems, Ni NTAs act as a binder-free multifunctional inner layer to support the electrocatalysts, offer a large specific surface area and serve as a fast electron transport pathway. Moreover, an alkaline electrolyzer has been constructed using NiFeOx NTAs as the anode and NiSe2 NTAs as the cathode, which only demands a cell voltage of 1.78 V to deliver a water-splitting current density of 500 mA cm-2, and demonstrates remarkable stability during long-term electrolysis. This work provides an attractive method for the design and fabrication of nanotube array-based catalyst electrodes for highly efficient water-splitting.

  15. Corrosion of copper alloys in sulphide containing district heting systems

    DEFF Research Database (Denmark)

    Thorarinsdottir, R.I.; Maahn, Ernst Emanuel

    1999-01-01

    Copper and some copper alloys are prone to corrosion in sulphide containing geothermal water analogous to corrosion observed in district heating systems containing sulphide due to sulphate reducing bacteria. In order to study the corrosion of copper alloys under practical conditions a test...... was carried out at four sites in the Reykjavik District Heating System. The geothermal water chemistry is different at each site. The corrosion rate and the amount and chemical composition of deposits on weight loss coupons of six different copper alloys are described after exposure of 12 and 18 months......, respectively. Some major differences in scaling composition and the degree of corrosion attack are observed between alloys and water types....

  16. Tribological Behaviors of Graphene and Graphene Oxide as Water-Based Lubricant Additives for Magnesium Alloy/Steel Contacts

    Directory of Open Access Journals (Sweden)

    Hongmei Xie

    2018-01-01

    Full Text Available The tribological behaviors of graphene and graphene oxide (GO as water-based lubricant additives were evaluated by use of a reciprocating ball-on-plate tribometer for magnesium alloy-steel contacts. Three sets of test conditions were examined to investigate the effect of concentration, the capacity of carrying load and the endurance of the lubrication film, respectively. The results showed that the tribological behaviors of water can be improved by adding the appropriate graphene or GO. Compared with pure deionized water, 0.5 wt.% graphene nanofluids can offer reduction of friction coefficient by 21.9% and reduction of wear rate by 13.5%. Meanwhile, 0.5 wt.% GO nanofluids were found to reduce the friction coefficient and wear rate up to 77.5% and 90%, respectively. Besides this, the positive effect of the GO nanofluids was also more pronounced in terms of the load-carrying capacity and the lubrication film endurance. The wear mechanisms have been tentatively proposed according to the observation of the worn surfaces by field emission scanning electron microscope-energy dispersive spectrometer (FESEM-EDS and Raman spectrum as well as the wettability of the nanofluids on the magnesium alloy surface by goniometer.

  17. Tribological Behaviors of Graphene and Graphene Oxide as Water-Based Lubricant Additives for Magnesium Alloy/Steel Contacts.

    Science.gov (United States)

    Xie, Hongmei; Jiang, Bin; Dai, Jiahong; Peng, Cheng; Li, Chunxia; Li, Quan; Pan, Fusheng

    2018-01-29

    The tribological behaviors of graphene and graphene oxide (GO) as water-based lubricant additives were evaluated by use of a reciprocating ball-on-plate tribometer for magnesium alloy-steel contacts. Three sets of test conditions were examined to investigate the effect of concentration, the capacity of carrying load and the endurance of the lubrication film, respectively. The results showed that the tribological behaviors of water can be improved by adding the appropriate graphene or GO. Compared with pure deionized water, 0.5 wt.% graphene nanofluids can offer reduction of friction coefficient by 21.9% and reduction of wear rate by 13.5%. Meanwhile, 0.5 wt.% GO nanofluids were found to reduce the friction coefficient and wear rate up to 77.5% and 90%, respectively. Besides this, the positive effect of the GO nanofluids was also more pronounced in terms of the load-carrying capacity and the lubrication film endurance. The wear mechanisms have been tentatively proposed according to the observation of the worn surfaces by field emission scanning electron microscope-energy dispersive spectrometer (FESEM-EDS) and Raman spectrum as well as the wettability of the nanofluids on the magnesium alloy surface by goniometer.

  18. Volatilization from PCA steel alloy

    Energy Technology Data Exchange (ETDEWEB)

    Hagrman, D.L.; Smolik, G.R.; McCarthy, K.A.; Petti, D.A.

    1996-08-01

    The mobilizations of key components from Primary Candidate Alloy (PCA) steel alloy have been measured with laboratory-scale experiments. The experiments indicate most of the mobilization from PCA steel is due to oxide formation and spalling but that the spalled particles are large enough to settle rapidly. Based on the experiments, models for the volatization of iron, manganese, and cobalt from PCA steel in steam and molybdenum from PCA steel in air have been derived.

  19. Characterisation of Oxides Formed on the Internal Surface of Steam Generator Tubes in Alloy 690 Corroded in the Primary Environment of Pressurised Water Reactors

    International Nuclear Information System (INIS)

    Carrette, Florence; Leclercq, Stephanie; Legras, Laurent

    2012-09-01

    Since the end of the 1990s, EDF R and D has been studying the phenomenon of corrosion product release from Steam Generator tubes in order to minimize the Source Term of the contamination and radiation exposure during operation and maintenance of Pressurised Water Reactors. With the BOREAL loop, release tests in primary water at 325 deg. C were performed on various Steam Generator tubes made of alloy 690. The experimental conditions of these tests (chemistry, temperature and hydraulics) were the same for all the tests but the results showed various behaviours towards release. For some tubes, the release was weak whereas for others, it was higher; the release rate of the tubes decreased more or less quickly with time. In order to explain these results, the internal surface of the tubes was characterised before and after the tests. Before the tests, various parameters were studied; the main parameters were the roughness, the impurities, the grain size and the cold work. The results demonstrated that it was not easy to quantify the influence of each parameter on release and to differentiate the tubes. A new parameter was proposed to characterise the internal extreme surface of SG tubes: the surface nano-hardness by nano-indentation measurements. The tubes were also observed and analysed by SEM, (X)TEM. Data obtained by (X)TEM revealed differences of the surface state (layer of perturbed microstructure, density of dislocations, grain size, impurities, initial oxide,...). After the tests, the oxides formed on the internal surface and the underlying material of the samples were characterised by SEM, (X)TEM and SIMS. The examinations showed various types of oxides. For some tubes, a duplex oxide scale was identified, for the others, only one oxide scale was observed. For equivalent durations of corrosion, the thickness of the enriched - chromium oxide layer can vary from 5 nm to 100 nm and the chemical composition can be different. The examinations of the underlying

  20. Progressive degradation of alloy 690 and the development of a significant improvement in alloy 800CR

    International Nuclear Information System (INIS)

    Staehle, Roger W.; Arioka, Koji; Tapping, Robert

    2015-01-01

    The present most widely used alloys for tubing in steam generators and structural materials in water cooled reactors are Alloy 690 and Alloy 800. However, both alloys, while improved over Alloy 600 may not meet the needs of longer range applications in the range of 80-100 years. Alloy 690 sustains damage resulting from the formation of cavities at grain boundaries which eventually cover about 50% of the area of the grain boundaries with the remainder covering being covered with carbides. The cavities seem to nucleate on the carbides leaving the grain boundaries a structure of cavities and carbides. Such a structure will lead the Alloy 690 to fail completely. Normal Alloy 800 does not produce such cavities and probably retains a large amount of its corrosion resistance but does sustain progressive SCC at low rate. A new alloy, 800CR, has been developed in a collaboration among Arioka, Tapping, and Staehle. This alloy is based on a Cr composition of 23.5-27% with the remainder retaining the previous Alloy 800 composition. 800CR sustains a crack velocity about 100 times less than Alloy 690 and a negligible rate of initiation. The 800CR, alloy is now seeking a patent. (authors)

  1. Modeling the electrochemistry of the primary circuits of light water reactors

    International Nuclear Information System (INIS)

    Bertuch, A.; Macdonald, D.D.; Pang, J.; Kriksunov, L.; Arioka, K.

    1994-01-01

    To model the corrosion behaviors of the heat transport circuits of light water reactors, a mixed potential model (NTM) has been developed and applied to both boiling water reactors (BWRs) and pressurized water reactors (PWRs). Using the data generated by the GE/UKEA-Harwell radiolysis model, electrochemical potentials (ECPs) have been calculated for the heat transport circuits of eight BWRs operating under hydrogen water chemistry (HWC). By modeling the corrosion behaviors of these reactors, the effectiveness of HWC at limiting IGSCC and IASCC can be determined. For simulating PWR primary circuits, a chemical-radiolysis model (developed by the authors) was used to generate input parameters for the MPM. Corrosion potentials of Type 304 and 316 SSs in PWR primary environments were calculated using the NTM and were found to be in good agreement with the corrosion potentials measured in the laboratory for simulated PWR primary environments

  2. Primary water chemistry monitoring from the point of view of radiation build-up

    International Nuclear Information System (INIS)

    Horvath, G.L.; Civin, V.; Pinter, T.

    1997-01-01

    Basic operational principles of a computer code system calculating the primary circuit corrosion product activities based on actual measured plant chemistry data are presented. The code system consists of two parts: FeSolub.prg: calculates the characteristic iron solubilities based on actual primary water chemistry (H 3 BO 3 KOH, ... etc.) and plant load (MW) data. A developed solubility calculation method has been applied fitted to magnetite solubility data of several authors; RADTRAN.exe: calculates primary circuit water and surface corrosion product activities based on results of FeSolub.prg or planned water chemistry data up to the next shutdown. The computer code system is going to be integrated into a general primary water chemistry monitoring and surveillance system. (author). 15 refs, 4 figs, 3 tabs

  3. Environmental fatigue in aluminum-lithium alloys

    Science.gov (United States)

    Piascik, Robert S.

    1992-01-01

    Aluminum-lithium alloys exhibit similar environmental fatigue crack growth characteristics compared to conventional 2000 series alloys and are more resistant to environmental fatigue compared to 7000 series alloys. The superior fatigue crack growth behavior of Al-Li alloys 2090, 2091, 8090, and 8091 is due to crack closure caused by tortuous crack path morphology and crack surface corrosion products. At high R and reduced closure, chemical environment effects are pronounced resulting in accelerated near threshold da/dN. The beneficial effects of crack closure are minimized for small cracks resulting in rapid growth rates. Limited data suggest that the 'chemically small crack' effect, observed in other alloy system, is not pronounced in Al-Li alloys. Modeling of environmental fatigue in Al-Li-Cu alloys related accelerated fatigue crack growth in moist air and salt water to hydrogen embrittlement.

  4. An Investigation into Water Chemistry in Primary Coolant Circuit of an Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    Wu, Bing-Jhen; Yeh, Tsung-Kuang; Wang, Mei-Ya; Sheu, Rong-Jiun

    2012-09-01

    To ensure operation safety, an optimization on the coolant chemistry in the primary coolant circuit of a nuclear reactor is essential no matter what type or generation the reactor belongs to. For a better understanding toward the water chemistry in an advanced boiling water reactor (ABWR), such as the one being constructed in the northern part of Taiwan, and for a safer operation of this ABWR, we conducted a proactive, thorough water chemistry analysis prior to the completion of this reactor in this study. A numerical simulation model for water chemistry analyses in ABWRs has been developed, based upon the core technology we established in the past. This core technology for water chemistry modeling is basically an integration of water radiolysis, thermal-hydraulics, and reactor physics. The model, by the name of DEMACE - ABWR, is an improved version of the original DEMACE model and was used for radiolysis and water chemistry prediction in the Longmen ABWR in Taiwan. Predicted results pertinent to the water chemistry variation and the corrosion behavior of structure materials in the primary coolant circuit of this ABWR under rated-power operation were reported in this paper. (authors)

  5. Irradiation-assisted stress corrosion cracking in HTH Alloy X-750 and Alloy 625

    International Nuclear Information System (INIS)

    Bajaj, R.; Mills, W.J.; Lebo, M.R.; Hyatt, B.Z.; Burke, M.G.

    1995-01-01

    In-reactor testing of bolt-loaded compact tension specimens was performed in 360 C water to determine the irradiation-assisted stress corrosion cracking (IASCC) behavior of HTH Alloy X-750 and direct-aged Alloy 625. New data confirm previous results showing that high irradiation levels reduce SCC resistance in Alloy X-750. Heat-to-heat variability correlates with boron content, with low boron heats showing improved IASCC properties. Alloy 625 is resistant to IASCC, as no cracking was observed in any Alloy 625 specimens. Microstructural, microchemical and deformation studies were performed to characterize the mechanisms responsible for IASCC in Alloy X-750 and the lack of an effect in Alloy 625. The mechanisms under investigation are: boron transmutation effects, radiation-induced changes in microstructure and deformation characteristics, and radiation-induced segregation. Irradiation of Alloy X-750 caused significant strengthening and ductility loss that was associated with the formation of cavities and dislocation loops. High irradiation levels did not cause significant segregation of alloying or trace elements in Alloy X-750. Irradiation of Alloy 625 resulted in the formation of small dislocation loops and a fine body-centered-orthorhombic phase. The strengthening due to the loops and precipitates was apparently offset by a partial dissolution of γ double-prime precipitates, as Alloy 625 showed no irradiation-induced strengthening or ductility loss. In the nonirradiated condition, an IASCC susceptible HTH heat containing 28 ppm B showed grain boundary segregation of boron, whereas a nonsusceptible HTH heat containing 2 ppm B and Alloy 625 with 20 ppm B did not show significant boron segregation. Transmutation of boron to helium at grain boundaries, coupled with matrix strengthening, is believed to be responsible for IASCC in Alloy X-750, and the absence of these two effects results in the superior IASCC resistance displayed by Alloy 625

  6. The Development of the Low-Cost Titanium Alloy Containing Cr and Mn Alloying Elements

    Science.gov (United States)

    Zhu, Kailiang; Gui, Na; Jiang, Tao; Zhu, Ming; Lu, Xionggang; Zhang, Jieyu; Li, Chonghe

    2014-04-01

    The α + β-type Ti-4.5Al-6.9Cr-2.3Mn alloy has been theoretically designed on the basis of assessment of the Ti-Al-Cr-Mn thermodynamic system and the relationship between the molybdenum equivalent and mechanical properties of titanium alloys. The alloy is successfully prepared by the split water-cooled copper crucible, and its microstructures and mechanical properties at room temperature are investigated using the OM, SEM, and the universal testing machine. The results show that the Ti-4.5Al-6.9Cr-2.3Mn alloy is an α + β-type alloy which is consistent with the expectation, and its fracture strength, yield strength, and elongation reach 1191.3, 928.4 MPa, and 10.7 pct, respectively. Although there is no strong segregation of alloying elements under the condition of as-cast, the segregation of Cr and Mn is obvious at the grain boundary after thermomechanical treatment.

  7. Design criteria of primary coolant chemistry in SMART-P

    International Nuclear Information System (INIS)

    Choi, Byung Seon; Kim, Ah Young; Kim, Seong Hoon; Yoon, Ju Hyeon; Zee, Sung Qunn

    2005-01-01

    SMART-P differs significantly from commercially designed PWRs. Materials inventories used in SMART-P differ from that at PWRs. All surfaces of the primary circuit with the primary coolant are either made from or plated with stainless steel. The material of steam generator (SG) is also different from that of the standard material of the commercially operating PWRs: titanium alloy for the steam generator tubes. Also, SMART-P primary coolant technology differs from that in PWRs: ammonia is used as a pH raising agent and hydrogen formed due to radiolytic processes is kept in specific range by ammonia dosing. Nevertheless, main objectives of the SMART-P primary coolant are the same as at PWRs: to assure primary system pressure boundary integrity, fuel cladding integrity and to minimize out-of-core radiation buildup. The objective of this work is to introduce the design criteria for the primary water chemistry for SMART-P from the viewpoint of the system characteristics and the chemical design concept

  8. United modification of Al-24Si alloy by Al-P and Al-Ti-C master alloys

    Institute of Scientific and Technical Information of China (English)

    韩延峰; 刘相法; 王海梅; 王振卿; 边秀房; 张均艳

    2003-01-01

    The modification effect of a new type of Al-P master alloy on Al-24Si alloys was investigated. It is foundthat excellent modification effect can be obtained by the addition of this new type of A1-P master alloy into Al-24Simelt and the average primary Si grain size is decreased below 47 μm from original 225 μm. It is also found that theTiC particles in the melt coming from Al8Ti2C can improve the modification effect of the Al-P master alloy. Whenthe content of TiC particles in the Al-24Si melt is 0.03 %, the improvement reaches the maximum and keeps steadywith increasing content of TiC particles. Modification effect occurs at 50 min after the addition of the Al-P master al-loy and TiC particles, and keeps stable with prolonging holding time.

  9. 78 FR 10269 - National Primary Drinking Water Regulations: Revisions to the Total Coliform Rule

    Science.gov (United States)

    2013-02-13

    ... Illness CWS--Community Water System DBP--Disinfection Byproduct DWC--Drinking Water Committee EA--Economic... 141 and 142 National Primary Drinking Water Regulations: Revisions to the Total Coliform Rule; Final...-9684-8] RIN 2040-AD94 National Primary Drinking Water Regulations: Revisions to the Total Coliform Rule...

  10. Integrity evaluation of Alloy 600 RV head penetration tubes in Korean PWR plants

    International Nuclear Information System (INIS)

    Kang, Young Hwan; Park, Sung Ho; Hong, Sung Yull; Choi, Kwang Hee

    1995-01-01

    The structural integrity assessment of Alloy 600 RV head penetration tubes has been an important issue for the economical and reliable operation of power plants. In this paper, an overview of the integrity evaluation program for the RV head penetration tubes in Korean nuclear power plants is presented. Since the crack growth mechanism of the penetration tube is due to the primary water stress corrosion cracking (PWSCC) which is mainly related to the stress at the tube, the present paper consists of three primary activities: the stress evaluation, the flaw evaluation, and data generation through material and mechanical tests. (author). 5 refs, 2 figs, 1 tab

  11. Thermal bonding of light water reactor fuel using nonalkaline liquid-metal alloy

    International Nuclear Information System (INIS)

    Wright, R.F.; Tulenko, J.S.; Schoessow, G.J.; Connell, R.G. Jr.; Dubecky, M.A.; Adams, T.

    1996-01-01

    Light water reactor (LWR) fuel performance is limited by thermal and mechanical constraints associated with the design, fabrication, and operation of fuel in a nuclear reactor. A technique is explored that extends fuel performance by thermally bonding LWR fuel with a nonalkaline liquid-metal alloy. Current LWR fuel rod designs consist of enriched uranium oxide fuel pellets enclosed in a zirconium alloy cylindrical clad. The space between the pellets and the clad is filled by an inert gas. Because of the low thermal conductivity of the gas, the gas space thermally insulates the fuel pellets from the reactor coolant outside the fuel rod, elevating the fuel temperatures. Filling the gap between the fuel and clad with a high-conductivity liquid metal thermally bonds the fuel to the cladding and eliminates the large temperature change across the gap while preserving the expansion and pellet-loading capabilities. The application of liquid-bonding techniques to LWR fuel is explored to increase LWR fuel performance and safety. A modified version of the ESCORE fuel performance code (ESBOND) is developed to analyze the in-reactor performance of the liquid-metal-bonded fuel. An assessment of the technical feasibility of this concept for LWR fuel is presented, including the results of research into materials compatibility testing and the predicted lifetime performance of liquid-bonded LWR fuel. The results show that liquid-bonded boiling water reactor peak fuel temperatures are 400 F lower at beginning of life and 200 F lower at end of life compared with conventional fuel

  12. The Integration of a Structural Water Gas Shift Catalyst with a Vanadium Alloy Hydrogen Transport Device

    Energy Technology Data Exchange (ETDEWEB)

    Barton, Thomas; Argyle, Morris; Popa, Tiberiu

    2009-06-30

    This project is in response to a requirement for a system that combines water gas shift technology with separation technology for coal derived synthesis gas. The justification of such a system would be improved efficiency for the overall hydrogen production. By removing hydrogen from the synthesis gas stream, the water gas shift equilibrium would force more carbon monoxide to carbon dioxide and maximize the total hydrogen produced. Additional benefit would derive from the reduction in capital cost of plant by the removal of one step in the process by integrating water gas shift with the membrane separation device. The answer turns out to be that the integration of hydrogen separation and water gas shift catalysis is possible and desirable. There are no significant roadblocks to that combination of technologies. The problem becomes one of design and selection of materials to optimize, or at least maximize performance of the two integrated steps. A goal of the project was to investigate the effects of alloying elements on the performance of vanadium membranes with respect to hydrogen flux and fabricability. Vanadium was chosen as a compromise between performance and cost. It is clear that the vanadium alloys for this application can be produced, but the approach is not simple and the results inconsistent. For any future contracts, large single batches of alloy would be obtained and rolled with larger facilities to produce the most consistent thin foils possible. Brazing was identified as a very likely choice for sealing the membranes to structural components. As alloying was beneficial to hydrogen transport, it became important to identify where those alloying elements might be detrimental to brazing. Cataloging positive and negative alloying effects was a significant portion of the initial project work on vanadium alloying. A water gas shift catalyst with ceramic like structural characteristics was the second large goal of the project. Alumina was added as a

  13. Primary water chemistry monitoring from the point of view of radiation build-up

    Energy Technology Data Exchange (ETDEWEB)

    Horvath, G L [Institute for Electrical Power Research, Budapest (Hungary); Civin, V [Hungarian Electricity Generating Board, Budapest (Hungary); Pinter, T [Nuclear Power Plant PAKS, Budapest (Hungary)

    1997-02-01

    Basic operational principles of a computer code system calculating the primary circuit corrosion product activities based on actual measured plant chemistry data are presented. The code system consists of two parts: FeSolub.prg: calculates the characteristic iron solubilities based on actual primary water chemistry (H{sub 3}BO{sub 3}KOH, ... etc.) and plant load (MW) data. A developed solubility calculation method has been applied fitted to magnetite solubility data of several authors; RADTRAN.exe: calculates primary circuit water and surface corrosion product activities based on results of FeSolub.prg or planned water chemistry data up to the next shutdown. The computer code system is going to be integrated into a general primary water chemistry monitoring and surveillance system. (author). 15 refs, 4 figs, 3 tabs.

  14. VVER operational experience - effect of preconditioning and primary water chemistry on radioactivity build-up

    International Nuclear Information System (INIS)

    Zmitko, M.; Kysela, J.; Dudjakova, K.; Martykan, M.; Janesik, J.; Hanus, V.; Marcinsky, P.

    2004-01-01

    The primary coolant technology approaches currently used in VVER units are reviewed and compared with those used in PWR units. Standard and modified water chemistries differing in boron-potassium control are discussed. Preparation of the VVER Primary Water Chemistry Guidelines in the Czech Republic is noted. Operational experience of some VVER units, operated in the Czech Republic and Slovakia, in the field of the primary water chemistry, and radioactivity transport and build-up are presented. In Mochovce and Temelin units, a surface preconditioning (passivation) procedure has been applied during hot functional tests. The main principles of the controlled primary water chemistry applied during the hot functional tests are reviewed and importance of the water chemistry, technological and other relevant parameters is stressed regarding to the quality of the passive layer formed on the primary system surfaces. The first operational experience obtained in the course of beginning of these units operation is presented mainly with respect to the corrosion products coolant and surface activities. Effect of the initial passivation performed during hot functional tests and the primary water chemistry on corrosion products radioactivity level and radiation situation is discussed. (author)

  15. EELS and electron diffraction studies on possible bonaccordite crystals in pressurized water reactor fuel CRUD and in oxide films of alloy 600 material

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Jiaxin [Studsvik Nuclear AB, Nykoping (Sweden); Lindberg, Fredrik [Swerea KIMAB AB, Kista (Sweden); Wells, Daniel [Electric Power Research Institute, Charlotte (United States); Bengysson, Bernt [Ringhals AB, Ringhalsverket, Varobacka (Sweden)

    2017-06-15

    Experimental verification of boron species in fuel CRUD (Chalk River Unidentified Deposit) would provide essential and important information about the root cause of CRUD-induced power shifts (CIPS). To date, only bonaccordite and elemental boron were reported to exist in fuel CRUD in CIPS-troubled pressurized water reactor (PWR) cores and lithium tetraborate to exist in simulated PWR fuel CRUD from some autoclave tests. We have reevaluated previous analysis of similar threadlike crystals along with examining some similar threadlike crystals from CRUD samples collected from a PWR cycle that had no indications of CIPS. These threadlike crystals have a typical [Ni]/[Fe] atomic ratio of ⁓2 and similar crystal morphology as the one (bonaccordite) reported previously. In addition to electron diffraction study, we have applied electron energy loss spectroscopy to determine boron content in such a crystal and found a good agreement with that of bonaccordite. Surprisingly, such crystals seem to appear also on corroded surfaces of Alloy 600 that was exposed to simulated PWR primary water with a dissolved hydrogen level of 5 mL H{sub 2}/kg H{sub 2}O, but absent when exposed under 75 mL H{sub 2}/kg H{sub 2}O condition. It remains to be verified as to what extent and in which chemical environment this phase would be formed in PWR primary systems.

  16. Corrosion of high-density sintered tungsten alloys. Part 1

    International Nuclear Information System (INIS)

    Batten, J.J.; McDonald, I.G.; Moore, B.T.; Silva, V.M.

    1988-10-01

    The corrosion behaviour of four tungsten alloys has been evaluated through weight loss measurements after total immersion in both distilled water insight into the mechanism of corrosion was afforded by an examination of the and 5% sodium chloride solutions. Some insight the mechanism of corrosion was afforded by using the Scanning Electron Microscopy and through an analysis of the corrosion products. Pure tungsten and all the alloys studied underwent corrosion during the tests, and in each case the rare of corrosion in sodium chloride solution was markedly less than that in distilled water. A 95% W, 3.5% Ni, 1.5% Fe alloy was found to be the most corrosion resistant of the alloys under the experimental conditions. Examination of the data shows that for each of the tests, copper as an alloying element accelerates corrosion of tungsten alloys. 9 refs., 7 tabs., 12 figs

  17. Environmentally Assisted Cracking of Alloys at Temperatures near and above the Critical Temperature of Water

    International Nuclear Information System (INIS)

    Watanabe, Yutaka

    2008-01-01

    Physical properties of water, such as dielectric constant and ionic product, significantly vary with the density of water. In the supercritical conditions, since density of water widely varies with pressure, pressure has a strong influence on physical properties of water. Dielectric constant represents a character of water as a solvent, which determines solubility of an inorganic compound including metal oxides. Dissociation equilibrium of an acid is also strongly dependent on water density. Dissociation constant of acid rises with increased density of water, resulting in drop of pH. Density of water and the density-related physical properties of water, therefore, are the major governing factors of corrosion and environmentally assisted cracking of metals in supercritical aqueous solutions. This paper discusses importance of 'physical properties of water' in understanding corrosion and cracking behavior of alloys in supercritical water environments, based on experimental data and estimated solubility of metal oxides. It has been pointed out that the water density can have significant effects on stress corrosion cracking (SCC) susceptibility of metals in supercritical water, when dissolution of metal plays the key role in the cracking phenomena

  18. Effect of zinc additions on oxide rupture strain and repassivation kinetics of iron-based alloys in 288 C water

    International Nuclear Information System (INIS)

    Angeliu, T.M.; Andresen, P.L.

    1996-01-01

    The effect of Zn water chemistry additions on the mechanism of intergranular stress corrosion cracking (IGSCC) of Fe-based alloys in water at 288 C was evaluated in terms of the slip-dissolution model. In this model, an increase in the oxide film rupture strain or surface film repassivation kinetics improved resistance to IGSCC. The oxide rupture strain of type 304L (UNS S30403) stainless steel (SS) increased up to a factor of two in deaerated and 200 ppb oxygenated, high-purity water ( 300 h of exposure. Repassivation kinetics experiments showed Zn additions of ∼ 100 ppb increased the repassivation rate of an Fe-12% Cr alloys up to a factor of two in various deaerated water environments at 288 C. Life prediction modeling revealed that the combination of a more ductile oxide film and faster repassivation kinetics resulted in a reduction in the overall crack growth rate (CGR) by at least a factor of four. This factor of improvement was consistent with data from compact tension experiments in similar environments where CGR decreased as the Zn addition increased, with a greater decrease in CGR realized at lower pre-Zn CGR

  19. Copper and nickel alloys and titanium for seawater applications

    International Nuclear Information System (INIS)

    Richter, H.

    1977-01-01

    Copper and nickel alloys and titanium have been successfully used for heat exchangers on ships, in power plants and for chemical apparatus and piping systems because of their resistance against corrosion in sea water. Aluminium brass and copper nickel alloys, the standard materials for condensers and coolers, however, may be attacked, the corrosion depending on water quality, water velocity, and structural conditions. The mechanisms of corrosion are discussed. Under severe conditions the use of titanium may be indicated. The use of nickel base alloys is advantageous at elevated temperatures, e.g. for chemical reactions and for evaporation processes. Examples are given for application and for prevention of corrosion. (orig.) [de

  20. Aluminium-nickel-iron alloys resistant to corrosion by water at high temperature. Their basic properties - their improvement

    International Nuclear Information System (INIS)

    Coriou, H.; Fournier, R.; Grall, L.; Hure, J.

    1959-01-01

    The development of the investigations carried out on these alloys is reviewed, showing the establishment of their fundamental, particularly structural, properties. This is followed by studies on: 1 - The penetration process in corrosion. The results of micrographic studies of the metal oxide interface are given for a series of alloys treated in water and steam between 350 and 395 deg. C. The hypothesis of attack by pockets of gas pressure is corroborated, and a second process of deep penetration by islands of intergranular-type corrosion is shown to take place. These patches, distinct from the surface corrosion layer and sometimes forming at a considerable depth inside the metal, would be due to heterogeneities in composition of the solid solution making up the matrix of these alloys. 2 - The role of titanium and zirconium additions on rolled metal. Systematic studies are carried out on a series of alloys with titanium and zirconium contents between 0.05 and 0.15 per cent. The favourable effect of titanium in particular has been demonstrated. Zirconium acts in the same way, but less efficiently. The improvement due to these additions can be compared to their action on the distribution of the second phases, which tend to become more pronounced and more homogeneously distributed. The influence of solder on these alloys has been studied, showing up the part played by the structure gradients introduced by fission. (author) [fr

  1. Analysis of local dislocation densities in cold-rolled alloy 690 using transmission electron microscopy

    International Nuclear Information System (INIS)

    Ahn, Tae-Young; Kim, Sung Woo; Hwang, Seong Sik

    2016-01-01

    Service failure of alloy 690 in NPP has not been reported. However, some research groups reported that primary water stress corrosion cracking (PWSCC) occurred in severely cold-rolled alloy 690. Transgranular craking was also reported in coll-rolled alloy 690 with a banded structure. In order to understand the effect of cold rolling on the cracking of alloy 690, many research groups have focused on the local strain and the cracked carbide induced by cold-rolling, by using electron backscatter diffraction (EBSD). Transmission electron microscopy (TEM) has been widely used to characterize structural materials because this technique has superior spatial resolution and allows for the analysis of crystallographic and chemical information. The aim of the present study is to understand the mechanism of the abnormally high crack growth rate (CGR) in cold-rolled alloy 690 with a banded structure. The local dislocation density was measured by TEM to confirm the effects of local strain on the stress corrosion cracking (SCC) of alloy 690 with a banded structure. The effects of intragranular carbides on the SCC were also evaluated in this study. The local dislocation densities were directly measured using TEM to understand the effect of local strain on the SCC of Ni-based alloy 690 with a banded structure. The dislocation densities in the interior of the grains sharply increased in highly cold-rolled specimens due to intragranular carbide, which acted as a dislocation source

  2. Modelling and numerical simulation of the corrosion product transport in the pressurised water reactor primary circuit

    International Nuclear Information System (INIS)

    Marchetto, C.

    2002-05-01

    During operation of pressurised water reactor, corrosion of the primary circuit alloys leads to the release of metallic species such as iron, nickel and cobalt in the primary fluid. These corrosion products are implicated in different transport phenomena and are activated in the reactor core where they are submitted to neutron flux. The radioactive corrosion products are afterwards present in the out of flux parts of primary circuit where they generate a radiation field. The first part of this study deals with the modelling of the corrosion: product transport phenomena. In particular, considering the current state of the art, corrosion and release mechanisms are described empirically, which allows to take into account the material surface properties. New mass balance equations describing the corrosion product behaviour are thus obtained. The numerical resolution of these equations is implemented in the second part of this work. In order to obtain large time steps, we choose an implicit time scheme. The associated system is linearized from the Newton method and is solved by a preconditioned GMRES method. Moreover, a time step auto-adaptive management based on Newton iterations is performed. Consequently, an efficient resolution has been implemented, allowing to describe not only the quasi-steady evolutions but also the fast transients. In a last step, numerical simulations are carried out in order to validate the new corrosion product transport modelling and to illustrate the capabilities of this modelling. Notably, the numerical results obtained indicate that the code allows to restore the on-site observations underlining the influence of material surface properties on reactor contamination. (author)

  3. The effect of molten salt on high temperature behavior of stainless steel and titanium alloy with the presence of water vapor

    Science.gov (United States)

    Baharum, Azila; Othman, Norinsan Kamil; Salleh, Emee Marina

    2018-04-01

    The high temperature oxidation experiment was conducted to study the behavior of titanium alloy Ti6A14V and stainless steel 316 in Na2SO4-50%NaCl + Ar-20%O2 (molten salt) and Na2SO4-50%NaCl + Ar-20%O2 + 12% H2O (molten salt + water vapor) environment at 900°C for 30 hours using horizontal tube furnace. The sample then was investigated using weight change measurement analysis and X-ray diffraction (XRD) analysis to study the weight gained and the phase oxidation that occurred. The weight gained of the titanium alloy was higher in molten salt environment compared to stainless steel due to the rapid growth in the oxide scale but showed almost no change of weight gained upon addition of water vapor. This is due to the alloy was fully oxidized. Stainless steel showed more protection and better effect in molten salt environment compared to mixed environment showed by slower weight gain and lower oxidation rate. Meanwhile, the phase oxidation test of the samples showed that the titanium alloy consist of multi oxide layer of rutile (TiO2) and Al2O3 on the surface of the exposed sample. While stainless steel show the formation of both protective Cr-rich oxide and non-protective Fe-rich oxide layer. This can be concluded that stainless steel is better compared to Ti alloy due to slow growing of chromia oxide. Therefore it is proven that stainless steel has better self-protection upon high temperature exposure.

  4. Effect of Adding Elements on Microstructure of Mg-3Si Alloy

    Directory of Open Access Journals (Sweden)

    CUI Bin

    2017-03-01

    Full Text Available The microstructure of alloy Mg-3Si(mass fraction/%, same as below after successive additions with different elements of Zn, Nd, Gd and Y was observed and the microstructure evolution was investigated by scanning electron microscopy and X-ray diffraction. The results show the primary Mg2Si particles co-exist with eutectic Mg2Si particles in binary alloy Mg-Si. With minor addition of Zn element, only primary Mg2Si can be found in ternary Mg-3Si-3Zn system while eutectic Mg2Si particles disappear. In quaternary alloy Mg-2.0Nd-3.0Zn-3.0Si, the addition of Nd element can effectively refine the primary Mg2Si particles and form some Mg41Nd5 particles. After continuous adding of Gd and Y elements into quaternary system, Gd5Si3 and YSi particles increase significantly in the alloy Mg-8.0Gd-4.0Y-2.0Nd-3.0Zn-3.0Si, while volume fraction of primary Mg2Si decrease significantly. Thermo-Calc calculation predicts that the Gibbs free energy for primary particles Gd5Si3, YSi is lower, and therefore Gd, Y atom and Si are more likely to form compounds. In Mg-8Gd-4Y-2Nd-3Zn-3Si alloy, room temperature Gibbs free energy for primary particles Mg2Si, Gd5Si3, YSi is -9.56×104, -8.72×104, -2.83×104J/mol, respectively, and the mass fraction of these particles is 8.07%, 5.27%, 1.40% respectively.

  5. Nature of negative microplastic deformation in alloys

    International Nuclear Information System (INIS)

    Palatnik, L.S.; Ivantsov, V.I.; Kagan, Ya.I.; Papirov, I.I.; Fat'yanova, N.B.; AN Ukrainskoj SSR, Kharkov. Fiziko-Tekhnicheskij Inst.)

    1985-01-01

    The paper deals with investigation of microplastic deformation of corrosion resistant aging 40KhNYU alloy and the study of physical nature of negative microdeformation in this alloy under tension. Investigation of microplasticity of 40KhNYU alloy was conducted by the method of mechanostatic hysteresis using resistance strain gauge for measuring stresses and deformations. Microplasticity curves for 40KhNYU alloy were obtained. They represent the result of competition between usual (positive) microdeformation and phase (negative) deformation under tensile effect on the alloy. It was established that the negative microdeformation increment occurs during secondary aging of the phase precipitated from initial supersat urated solid solution (primary decomposition product). This phase decomposes under tension with disperse phase precipitation which promotes decreasing its specific volume and specimen volume as a whole

  6. Elemental volatility of HT-9 fusion reactor alloy

    International Nuclear Information System (INIS)

    Henslee, S.P.; Neilson, R.M. Jr.

    1985-01-01

    The volatility of elemental constituents from HT-9, a ferritic steel, proposed for fusion reactor structures, was investigated. Tests were conducted in flowing air at temperatures from 800 to 1200 0 C for durations of 1 to 20 h. Elemental volatility was calculated in terms of the weight fraction of the element volatilized from the initial alloy; molybdenum, manganese, and nickel were the primary constituents volatilized. Comparisons with elemental volatilities observed for another candidate fusion reactor materials. Primary Candidate Alloy (PCA), an austenitic stainless steel, indicate significant differences between the volatilities of these steels that may impact fusion reactor safety analysis and alloy selection. Scanning electron microscopy and energy dispersive spectrometry were used to investigate the oxide layers formed on HT-9 and to measure elemental contents within these layers

  7. Nickel-based gadolinium alloy for neutron adsorption application in ram packages

    International Nuclear Information System (INIS)

    Robino, C.; McConnell, P.; Mizia, R.

    2004-01-01

    This paper will outline the results of a metallurgical development program that is investigating the alloying of gadolinium into a nickel-chromium-molybdenum alloy matrix. Gadolinium has been chosen as the neutron absorption alloying element due to its high thermal neutron absorption cross section and low solubility in the expected U.S. repository environment. The nickel-chromium-molybdenum alloy family was chosen for its known corrosion performance, mechanical properties, and weldability. The workflow of this program includes chemical composition definition, primary and secondary melting studies, ingot conversion processes, properties testing, and national consensus codes and standards work. The microstructural investigation of these alloys shows that the gadolinium addition is not soluble in the primary austenite metallurgical phase and is present in the alloy as gadolinium-rich second phase. This is similar to what is observed in a stainless steel alloyed with boron. The mechanical strength values are similar to those expected for commercial Ni-Cr-Mo alloys. The alloys have been corrosion tested in simulated Yucca Mountain aqueous chemistries with acceptable results. The initial results of weldability tests have also been acceptable. Neutronic testing in a moderated critical array has generated favorable results. An American Society for Testing and Materials material specification has been issued for the alloy and a Code Case has been submitted to the American Society of Mechanical Engineers for code qualification. The ultimate goal is acceptance of the alloy for use at the Yucca Mountain repository

  8. Assessment of the radiation-induced loss of ductility in V-Cr-Ti alloys

    Energy Technology Data Exchange (ETDEWEB)

    Rowcliffe, A.F.; Zinkle, S.J. [Oak Ridge National Lab., TN (United States)

    1997-04-01

    Alloys based on the V-Cr-Ti system are attractive candidates for structural applications in fusion systems because of their low activation properties, high thermal stress factor (high thermal conductivity, moderate strength, and low coefficient of thermal expansion), and their good compatibility with liquid lithium. The U.S. program has defined a V-4Cr-4Ti (wt %) alloy as a leading candidate alloy based upon evidence from laboratory-scale (30 kg) heats covering the approximate composition range 0-8 wt % Ti and 5 to 15 wt % Cr. A review of the effects of neutron displacement damage, helium, and hydrogen generation on mechanical behavior, and of compatibility with lithium, water, and helium environments was presented at the ICFRM-5 conference at Clearwater in 1991. The results of subsequent optimization studies, focusing on the effects of fast reactor irradiation on tensile and impact properties of a range of alloys, were presented at the ICFRM-6 conference at Stresa in 1993. The primary conclusion of this work was that the V-4Cr-4Ti alloy composition possessed a near-optimal combination of physical and mechanical properties for fusion structural applications. Subsequently, a production-scale (500 kg) heat of V-4Cr-4Ti (Heat No. 832665) was procured from Teledyne Wah-Chang, together with several 15 kg heats of alloys with small variations in Cr and Ti. Further testing has been carried out on these alloys, including neutron irradiation experiments to study swelling and mechanical property changes. This paper discusses ductility measurements from some of these tests which are in disagreement with earlier work.

  9. Chemical degradation of drinking water disinfection byproducts by millimeter-sized particles of iron-silicon and magnesium-aluminum alloys.

    Science.gov (United States)

    Li, Tianyu; Chen, Yongmei; Wan, Pingyu; Fan, Maohong; Yang, X Jin

    2010-03-03

    The candidature of Fe-Si and Mg-Al alloys at millimeter-scale particle sizes for chemical degradation of disinfection byproducts (DBPs) in drinking water systems was substantiated by their enhanced corrosion resistance and catalytic effect on the degradation. The Mg-Al particles supplied electrons for reductive degradation, and the Fe-Si particles acted as a catalyst and provided the sites for the reaction. The alloy particles are obtained by mechanical milling and stable under ambient conditions. The proposed method for chemical degradation of DBPs possesses the advantages of relatively constant degradation performance, long-term durability, no secondary contamination, and ease of handling, storage and maintenance in comparison with nanoparticle systems.

  10. Corrosion behavior of electrodeposited Co-Fe alloys in aerated solutions

    Energy Technology Data Exchange (ETDEWEB)

    Chansena, A. [Research Unit on Corrosion, College of Data Storage Innovation, King Mongkut' s Institute of Technology Ladkrabang, Bangkok 10520 (Thailand); Sutthiruangwong, S., E-mail: sutha.su@kmitl.ac.th [Department of Chemistry, Faculty of Science, King Mongkut' s Institute of Technology Ladkrabang, Bangkok 10520 (Thailand); Research Unit on Corrosion, College of Data Storage Innovation, King Mongkut' s Institute of Technology Ladkrabang, Bangkok 10520 (Thailand)

    2017-05-01

    Co-Fe alloy is an important component for reader-writer in hard disk drive. The surface of the alloy is exposed to the environment both in gas phase and in liquid phase during manufacturing process. The study of corrosion behavior of Co-Fe alloys can provide useful fundamental data for reader-writer production planning especially when corrosion becomes a major problem. The corrosion study of electrodeposited Co-Fe alloys from cyclic galvanodynamic polarization was performed using potentiodynamic polarization technique. The composition of electrodeposited Co-Fe alloys was determined by X-ray fluorescence spectrometry. The patterns from X-ray diffractometer showed that the crystal structure of electrodeposited Co-Fe alloys was body-centered cubic. A vibrating sample magnetometer was used for magnetic measurements. The saturation magnetization (M{sub s}) was increased and the intrinsic coercivity (H{sub ci}) was decreased with increasing Fe content. The corrosion rate study was performed in aerated deionized water and aerated acidic solutions at pH 3, 4 and 5. The corrosion rate diagram for Co-Fe alloys was constructed. It was found that the corrosion rate of Co-Fe alloys was increased with increasing Fe content in both aerated deionized water and aerated acidic solutions. In aerated pH 3 solution, the Co-Fe alloy containing 78.8% Fe showed the highest corrosion rate of 7.7 mm yr{sup −1} with the highest M{sub s} of 32.0 A m{sup 2} kg{sup −1}. The corrosion rate of the alloy with 23.8% Fe was at 1.1 mm yr{sup −1} with M{sub s} of 1.2 A m{sup 2} kg{sup −1}. In aerated deionized water, the alloy with the highest Fe content of 78.5% still showed the highest corrosion rate of 0.0059 mm yr{sup −1} while the alloy with the lowest Fe content of 20.4% gave the lowest corrosion rate of 0.0045 mm yr{sup −1}. - Highlights: • The aeration during corrosion measurement simulates reader-writer head production environment. • The corrosion rate diagram for Co-Fe alloys

  11. Effects of Mn addition on microstructure and hardness of Al-12.6Si alloy

    Science.gov (United States)

    Biswas, Prosanta; Patra, Surajit; Mondal, Manas Kumar

    2018-03-01

    In this work, eutectic Al-12.6Si alloy with and without manganese (Mn) have been developed through gravity casting route. The effect of Mn concentration (0.0 wt.%, 1 wt%, 2 wt% and 3 wt%) on microstructural morphology and hardness property of the alloy has been investigated. The eutectic Al-12.6 Si alloy exhibits the presence of combine plate, needle and rod-like eutectic silicon phase with very sharp corners and coarser primary silicon particles within the α-Al phase. In addition of 1wt.% of Mn in the eutectic Al-12.6Si alloy, sharp corners of the primary Si and needle-like eutectic Si are became blunt and particles size is reduced. Further, increase in Mn concentration (2.0 wt.%) in the Al-12.6Si alloy, irregular plate shape Al6(Mn,Fe) intermetallics are formed inside the α-Al phase, but the primary and eutectic phase morphology is similar to the eutectic Al-12.6Si alloy. The volume fraction of Al6(Mn,Fe) increases and Al6(Mn,Fe) particles appear as like chain structure in the alloy with 3 wt.% Mn. An increase in Mn concentration in the Al-12.6Si alloys result in the increase in bulk hardness of the alloy as an effects of microstructure modification as well as the presence of harder Al6(Mn,Fe) phase in the developed alloy.

  12. Solidification of eutectic system alloys in space (M-19)

    Science.gov (United States)

    Ohno, Atsumi

    1993-01-01

    It is well known that in the liquid state eutectic alloys are theoretically homogeneous under 1 g conditions. However, the homogeneous solidified structure of this alloy is not obtained because thermal convection and non-equilibrium solidification occur. The present investigators have clarified the solidification mechanisms of the eutectic system alloys under 1 g conditions by using the in situ observation method; in particular, the primary crystals of the eutectic system alloys never nucleated in the liquid, but instead did so on the mold wall, and the crystals separated from the mold wall by fluid motion caused by thermal convection. They also found that the equiaxed eutectic grains (eutectic cells) are formed on the primary crystals. In this case, the leading phase of the eutectic must agree with the phase of the primary crystals. In space, no thermal convection occurs so that primary crystals should not move from the mold wall and should not appear inside the solidified structure. Therefore no equiaxed eutectic grains will be formed under microgravity conditions. Past space experiments concerning eutectic alloys were classified into two types of experiments: one with respect to the solidification mechanisms of the eutectic alloys and the other to the unidirectional solidification of this alloy. The former type of experiment has the problem that the solidified structures between microgravity and 1 g conditions show little difference. This is why the flight samples were prepared by the ordinary cast techniques on Earth. Therefore it is impossible to ascertain whether or not the nucleation and growth of primary crystals in the melt occur and if primary crystals influence the formation of the equiaxed eutectic grains. In this experiment, hypo- and hyper-eutectic aluminum copper alloys which are near eutectic point are used. The chemical compositions of the samples are Al-32.4mass%Cu (Hypo-eutectic) and Al-33.5mass%Cu (hyper-eutectic). Long rods for the samples are

  13. Development of heat treated Zr-2.5% Nb alloy tubes for pressure tubes

    International Nuclear Information System (INIS)

    Saibaba, N.; Jha, S.K.; Tonpe, S.

    2011-01-01

    Zr-2.5% Nb alloy is the candidate material for pressure tubes of Pressurized Heavy Water Reactors (PHWR), and are manufactured in cold working condition while heat treated pressure tubes are used in RBMK and FUGEN type of reactors. The diametral creep of these tubes is the life limiting factor. This paper presents the extensive work carried out for the optimization of process parameters to manufacture heat treated Zr-2.5% Nb pressure tubes. Extensive dilactometry study was carried out to establish the transus temperature for the alloy and the effect of soaking temperature and cooling rate on the microstructure was characterized. On the basis of the study, water quenching (at 883 deg C) in the a b region with 20-25% primary a phase was selected, further cold worked, aged and finally autoclaved. Mechanical properties of the finished tubes were found to be comparable to the cold worked route. Large number of full sized tubes of about 700 - 800 mm long was produced to establish the repeatability. (author)

  14. PWR type reactor equipped with a primary circuit loop water level gauge

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro.

    1990-01-01

    The time of lowering a water level to less than the position of high temperature side pipeway nozzle has been rather delayed because of the swelling of mixed water level due to heat generation of the reactor core. Further, there has been a certain restriction for the installation, maintenance and adjustment of a water level gauge since it is at a position under high radiation exposure. Then, a differential pressure type water level gauge with temperature compensation is disposed at a portion below a water level gauge of a pressurizer and between the steam generator exit plenum and the lower end of the loop seal. Further, a similar water level system is disposed to all of the loops of the primary circulation circuits. In a case that the amount of water contained in a reactor container should decreased upon occurrence of loss of coolant accidents caused by small rupture and stoppage of primary circuit pumps, lowering of the water level preceding to the lowering of the water level in the reactor core is detected to ensure the amount of water. Since they are disposed to all of the loops and ensure the excess margin, reliability for the detection of the amount of contained water can be improved by averaging time for the data of the water level and averaging the entire systems, even when there are vibrations in the fluid or pressure in the primary circuit. (N.H.)

  15. Investigation on copper alloy and titanium heat exchanger tubes behaviour in sea water service

    International Nuclear Information System (INIS)

    Casarini, G.; Bianchi, M.; Winkler, L.; Caspani, M.

    1982-01-01

    Because of the contradictory behaviour in service of some copper alloys used in heat exchangers cooled by sea water (Mediterranean Sea - North Africa), a comparative study on the behaviour of some tubular test samples was performed by means of accelerated test run ''in situ'' using two little heat exchangers supplied by Foster Wheeler Italiana. The aim of the investigation was to obtain quick and reliable information on optimizing the choise of the most suitable material for the construction of new heat exchangers

  16. TEM investigations on the effect of chromium content and of stress relief treatment on precipitation in Alloy 82

    International Nuclear Information System (INIS)

    Sennour, M.; Chaumun, E.; Crépin, J.; Duhamel, C.; Gaslain, F.; Guerre, C.; Curières, I. de

    2013-01-01

    Highlights: •Slight change of the Cr content does not affect the microstructure of the butt welds. •Stress relief thermal treatment leads to the intergranular precipitation of Cr 23 C 6 . •The Cr 23 C 6 carbides are supposed to improve the SCC resistance of the butt welds. -- Abstract: Nickel-base alloys are widely used in nuclear Pressurized Water Reactors (PWRs). Most of them have been found susceptible to Stress Corrosion Cracking (SCC) in nominal PWR primary water. The time to initiation depends on the material and is longer for weld metals than for Alloy 600. This study will focus on Alloy 82, which is used in Dissimilar Metal Welds (DMWs). In service, DMWs are either in the as-welded state or have undergone a stress relief treatment. Previous SCC studies showed that the heat treatment reduces significantly the SCC susceptibility of the weld. In this context, this study focuses on the microstructure characterization of the weld in the as-welded state and in the heat-treated state. As chromium content is also a key factor for the SCC susceptibility, welds with low chromium content and medium chromium content were studied. The lower SCC susceptibility of the heat-treated welds was attributed to intergranular Cr 23 C 6 resulting from a combined effect of heat treatment and chromium and carbon contents. These intergranular carbides could explain the better behavior of Alloy 82, compared to other nickel-base alloys

  17. Water chemistry management of research reactor in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Yoshijima, Tetsuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    The JRR-3M cooling system consists of four systems, namely; (1) primary cooling system, (2) heavy water cooling system, (3) helium system and (4) secondary cooling system. The heavy water is used for reflector and pressurized with helium gas. Water chemistry management of the JRR-3M cooling systems is one of the important subject for the safety operation. The main objects are to prevent the corrosion of cooling system and fuel elements, to suppress the plant radiation build-up and to minimize the generation of radioactive waste. All measured values were within the limits of specifications and JRR-3M reactor was operated with safety in 1996. Spent fuels of JRR-3M reactor are stored in the spent fuel pool. This pool water has been analyzed to prevent corrosion of aluminum cladding of spent fuels. Water chemistry of spent fuel pool water is applied to the prevention of corrosion of aluminum alloys including fuel cladding. The JRR-2 reactor was eternally stopped in December 1996 and is now under decommissioning. The JRR-2 reactor is composed of heavy water tank, fuel guide tube and horizontal experimental hole. These are constructed of aluminum alloy and biological shield and upper shield are constructed of concrete. Three types of corrosion of aluminum alloy were observed in the JRR-2. The Alkaline corrosion of aluminum tube occurred in 1972 because of the mechanical damage of the aluminum fuel guide tube which is used for fuel handling. Modification of the reactor top shield was started in 1974 and completed in 1975. (author)

  18. Deformation structure of water atomised and extruded Cu-Cr-Zr alloys

    International Nuclear Information System (INIS)

    Correia, J.B.; Davies, H.A.; Sellars, C.M.

    1994-01-01

    Copper alloy powders containing Cr and/or Zr were produced via water atomisation and consolidated by extrusion. The crystallographic orientations relative to the extrusion axis, both in the as-extruded condition and after heat treatment were assessed by an inverse pole figure texture determination method using X-ray diffraction. Transmission electron microscopy (TEM) was also used for microstructural observations. The as-extruded materials showed a clear double fibre texture ( left angle 111 right angle + left angle 100 right angle ), the left angle 111 right angle component being the stronger. After isochronal 1 hour heat treatments in the temperature range 400-700 C, a clear trend of a decreasing left angle 111 right angle component with increasing temperatures was seen. For the highest heat treatment temperature used (700 C), the alloys approached a random crystallographic distribution. The microstructure of the as-extruded material showed a banded structure with sharp boundaries parallel to the extrusion direction. Also, the left angle 111 right angle or left angle 100 right angle directions were generally parallel to the extrusion direction. TEM observation of longitudinal sections after exposure to increasingly high temperature, notably above 500 C, indicated that boundaries started to lose their sharpness and to spread sideways. When the same materials were observed in transverse sections, the competitive growth of some subgrains was clearly seen. (orig.)

  19. Catalytic Hydrogenation of Levulinic Acid in Water into g-Valerolactone over Bulk Structure of Inexpensive Intermetallic Ni-Sn Alloy Catalysts

    Directory of Open Access Journals (Sweden)

    Rodiansono Rodiansono

    2015-07-01

    Full Text Available A bulk structure of inexpensive intermetallic nickel-tin (Ni-Sn alloys catalysts demonstrated highly selective in the hydrogenation of levulinic acid in water into g-valerolactone. The intermetallic Ni-Sn catalysts were synthesized via a very simple thermochemical method from non-organometallic precursor at low temperature followed by hydrogen treatment at 673 K for 90 min. The molar ratio of nickel salt and tin salt was varied to obtain the corresponding Ni/Sn ratio of 4.0, 3.0, 2.0, 1.5, and 0.75. The formation of Ni-Sn alloy species was mainly depended on the composition and temperature of H2 treatment. Intermetallics Ni-Sn that contain Ni3Sn, Ni3Sn2, and Ni3Sn4 alloy phases are known to be effective heterogeneous catalysts for levulinic acid hydrogenation giving very excellence g-valerolactone yield of >99% at 433 K, initial H2 pressure of 4.0 MPa within 6 h. The effective hydrogenation was obtained in H2O without the formation of by-product. Intermetallic Ni-Sn(1.5 that contains Ni3Sn2 alloy species demonstrated very stable and reusable catalyst without any significant loss of its selectivity. © 2015 BCREC UNDIP. All rights reserved. Received: 26th February 2015; Revised: 16th April 2015; Accepted: 22nd April 2015  How to Cite: Rodiansono, R., Astuti, M.D., Ghofur, A., Sembiring, K.C. (2015. Catalytic Hydrogenation of Levulinic Acid in Water into g-Valerolactone over Bulk Structure of Inexpensive Intermetallic Ni-Sn Alloy Catalysts. Bulletin of Chemical Reaction Engineering & Catalysis, 10 (2: 192-200. (doi:10.9767/bcrec.10.2.8284.192-200Permalink/DOI: http://dx.doi.org/10.9767/bcrec.10.2.8284.192-200  

  20. Vanadium alloys for structural applications in fusion systems: A review of vanadium alloy mechanical and physical properties

    Energy Technology Data Exchange (ETDEWEB)

    Loomis, B.A.; Smith, D.L.

    1991-12-16

    The current knowledge is reviewed on (1) the effects of neutron irradiation on tensile strength and ductility, ductile-brittle transition temperature, creep, fatigue, and swelling of vanadium-base alloys, (2) the compatibility of vanadium-base alloys with liquid lithium, water, and helium environments, and (3) the effects of hydrogen and helium on the physical and mechanical properties of vanadium alloys that are potential candidates for structural materials applications in fusion systems. Also, physical and mechanical properties issues are identified that have not been adequately investigated in order to qualify a vanadium-base alloy for the structural material in experimental fusion devices and/or in fusion reactors.

  1. Vanadium alloys for structural applications in fusion systems: A review of vanadium alloy mechanical and physical properties

    International Nuclear Information System (INIS)

    Loomis, B.A.; Smith, D.L.

    1991-01-01

    The current knowledge is reviewed on (1) the effects of neutron irradiation on tensile strength and ductility, ductile-brittle transition temperature, creep, fatigue, and swelling of vanadium-base alloys, (2) the compatibility of vanadium-base alloys with liquid lithium, water, and helium environments, and (3) the effects of hydrogen and helium on the physical and mechanical properties of vanadium alloys that are potential candidates for structural materials applications in fusion systems. Also, physical and mechanical properties issues are identified that have not been adequately investigated in order to qualify a vanadium-base alloy for the structural material in experimental fusion devices and/or in fusion reactors

  2. dK/da effects on the SCC growth rates of nickel base alloys in high-temperature water

    Science.gov (United States)

    Chen, Kai; Wang, Jiamei; Du, Donghai; Andresen, Peter L.; Zhang, Lefu

    2018-05-01

    The effect of dK/da on crack growth behavior of nickel base alloys has been studied by conducting stress corrosion cracking tests under positive and negative dK/da loading conditions on Alloys 690, 600 and X-750 in high temperature water. Results indicate that positive dK/da accelerates the SCC growth rates, and the accelerating effect increases with dK/da and the initial CGR. The FRI model was found to underestimate the dK/da effect by ∼100X, especially for strain hardening materials, and this underscores the need for improved insight and models for crack tip strain rate. The effect of crack tip strain rate and dK/dt in particular can explain the dK/da accelerating effect.

  3. Effects of thermal aging and stress triaxiality on PWSCC initiation susceptibility of nickel-based Alloy 600

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Seung Chang; Choi, Kyoung Joon; Kim, Tae Ho; Kim, Ji Hyun [Dept. of Nuclear Science and Engineering, School of Mechanical and Nuclear Engineering, Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2016-10-15

    In present study, effects of thermal aging and triaxial stress were investigated in terms of primary water stress corrosion cracking susceptibility. The thermal aging was applied via heat treatment at 400°C and triaxial stress was applied via notched tensile test specimen. The crack initiation time of each specimen were then measured by direct current potential drop method during slow strain rate test at primary water environment. Alloys with 10 years thermal aging exhibited the highest susceptibility to stress corrosion cracking and asreceived specimen shows lowest susceptibility. The trend was different with triaxial stress applied; 20 years thermal aging specimen shows highest susceptibility and as-received specimen shows lowest. It would be owing to change of precipitate morphology during thermal aging and different activated slip system in triaxial stress state.

  4. Microstructure and orientation evolution in unidirectional solidified Al–Zn alloys

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Zhongwei, E-mail: chzw@nwpu.edu.cn; Wang, Enyuan; Hao, Xiaolei

    2016-06-14

    Morphological instability and growth orientation evolution during unidirectional solidification of Al–Zn alloys with different pulling speeds were investigated by X-ray diffraction (XRD) and electron back-scatter diffraction (EBSD) in scanning electron microscope (SEM). The experimental results show that, as the pulling speed increases, the primary dendrite spacing becomes smaller gradually and dendrite trunks incline to the heat flow direction perfectly in unidirectional solidified Al–9.8 wt%Zn and Al–89 wt%Zn alloys. However, regardless of the pulling speed in unidirectional solidified Al–Zn alloys under fixed thermal gradient, the regular dendrites with <100> directions of primary trunks and secondary arms in 9.8 wt% Zn composition are replaced by <110> dendrites of primary trunks and secondary arms in 89 wt% Zn composition. In unidirectional solidified Al–32 wt% Zn alloy, cellular, fractal seaweed, and stabilized seaweed structures were observed at high pulling speeds. At a high pulling speed of 1000 µm/s, seaweed structures transform to the columnar dendrites with <110> trunks and <100> arms. The above orientation evolution can be attributed to low anisotropy of solid-liquid interface energy and the seaweed structure is responsible for isotropy of {111} planes.

  5. Alloy catalyst material

    DEFF Research Database (Denmark)

    2014-01-01

    The present invention relates to a novel alloy catalyst material for use in the synthesis of hydrogen peroxide from oxygen and hydrogen, or from oxygen and water. The present invention also relates to a cathode and an electrochemical cell comprising the novel catalyst material, and the process use...... of the novel catalyst material for synthesising hydrogen peroxide from oxygen and hydrogen, or from oxygen and water....

  6. Synthesis of shape memory alloys using electrodeposition

    Science.gov (United States)

    Hymer, Timothy Roy

    Shape memory alloys are used in a variety of applications. The area of micro-electro-mechanical systems (MEMS) is a developing field for thin film shape memory alloys for making actuators, valves and pumps. Until recently thin film shape memory alloys could only be made by rapid solidification or sputtering techniques which have the disadvantage of being "line of sight". At the University of Missouri-Rolla, electrolytic techniques have been developed that allow the production of shape memory alloys in thin film form. The advantages of this techniques are in-situ, non "line of sight" and the ability to make differing properties of the shape memory alloys from one bath. This research focused on the electrodeposition of In-Cd shape memory alloys. The primary objective was to characterize the electrodeposited shape memory effect for an electrodeposited shape memory alloy. The effect of various operating parameters such as peak current density, temperature, pulsing, substrate and agitation were investigated and discussed. The electrodeposited alloys were characterized by relative shape memory effect, phase transformation, morphology and phases present. Further tests were performed to optimize the shape memory by the use of a statistically designed experiment. An optimized shape memory effect for an In-Cd alloy is reported for the conditions of the experiments.

  7. Modification mechanism of eutectic silicon in Al–6Si–0.3Mg alloy with scandium

    Energy Technology Data Exchange (ETDEWEB)

    Patakham, Ussadawut [Manufacturing and Systems Engineering Program, Department of Production Engineering, Faculty of Engineering, King Mongkut’s University of Technology Thonburi, 126 Pracha-Utid Rd., Bangmod, Tungkhru, Bangkok 10140 (Thailand); Kajornchaiyakul, Julathep [National Metal and Material Technology Center, National Science and Technology Development Agency, 114 Thailand Science Park, Klong Nueng, Klong Luang, Pathumthani 12120 (Thailand); Limmaneevichitr, Chaowalit, E-mail: chaowalit.lim@kmutt.ac.th [Manufacturing and Systems Engineering Program, Department of Production Engineering, Faculty of Engineering, King Mongkut’s University of Technology Thonburi, 126 Pracha-Utid Rd., Bangmod, Tungkhru, Bangkok 10140 (Thailand)

    2013-10-25

    Highlights: •Morphologies and growth of Sc and Sr-modified eutectic silicon resemble those of dendrites. •Crystal orientation of eutectic aluminum depends on growth characteristics of eutectic silicon. •We report strong evidence of the occurrence of an impurity-induced twinning mechanism. -- Abstract: The modification mechanism of eutectic silicon in Al–6Si–0.3Mg alloy with scandium was studied. The crystallographic orientation relationships between primary dendrites and the eutectic phase of unmodified and modified Al–6Si–0.3 Mg alloys were determined using electron backscatter diffraction (EBSD). The orientation of aluminum modified with scandium in the eutectic phase was different from that of the neighboring primary dendrites. This result implies that eutectic aluminum grows epitaxially from the surrounding primary aluminum dendrites in the unmodified alloy and that eutectic aluminum grows competitively from the surrounding primary aluminum dendrites in the modified alloy. The pole figure maps of eutectic Si in the [1 0 0], [1 1 0] and [1 1 1] axes of the unmodified and Sc-modified alloys were different, suggesting that the eutectic Al and Si crystals in modified alloy growth are more isotropic and cover a larger set of directions. The lattice fringes of Si of the alloys with and without Sc modification were different in the TEM results. The lattice fringes of Si in modified alloy were found to be multiple twins. However, this was not observed in the unmodified alloy. The growth characteristic of eutectic Si crystal in modified alloy suggests the occurrence of multiple twinning reactions and the formation of a high density of twins. This modification mechanism by Sc is explained by the results of scanning electron microscopy (SEM), electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM) analysis, which provide strong evidence of the occurrence of the impurity-induced twinning (IIT) mechanism.

  8. Emission properties of aluminium-lithium alloy

    International Nuclear Information System (INIS)

    Bondarenko, G.G.; Shishkov, A.V.

    1995-01-01

    High secondary emission properties at comparatively low operation temperatures were obtained when investigating aluminum-lithium alloy Al - 2.2 mass % Li. The maximal value of the coefficient of secondary electron emission for alloy, activated under optimal conditions, is achieved at comparatively low energy of primary electrons, equal to 600 eV. Low value of the first critical potential (15 ± 2 eV) was obtained. It is important for operation of secondary emission cathodes. 12 refs.; 4 figs

  9. Study on microstructure and properties of Mg-alloy surface alloying layer fabricated by EPC

    Directory of Open Access Journals (Sweden)

    Chen Dongfeng

    2010-02-01

    Full Text Available AZ91D surface alloying was investigated through evaporative pattern casting (EPC technology. Aluminum powder (0.074 to 0.104 mm was used as the alloying element in the experiment. An alloying coating with excellent properties was fabricated, which mainly consisted of adhesive, co-solvent, suspending agent and other ingredients according to desired proportion. Mg-alloy melt was poured under certain temperature and the degree of negative pressure. The microstructure of the surface layer was examined by means of scanning electron microscopy. It has been found that a large volume fraction of network new phases were formed on the Mg-alloy surface, the thickness of the alloying surface layer increased with the alloying coating increasing from 0.3 mm to 0.5 mm, and the microstructure became compact. Energy dispersive X-ray (EDX analysis was used to determine the chemical composition of the new phases. It showed that the new phases mainly consist of β-Mg17Al12, in addition to a small quantity of inter-metallic compounds and oxides. A micro-hardness test and a corrosion experiment to simulate the effect of sea water were performed. The result indicated that the highest micro-hardness of the surface reaches three times that of the matrix. The corrosion rate of alloying samples declines to about a fifth of that of the as-cast AZ91D specimen.

  10. Oxidation performance of V-Cr-Ti alloys

    International Nuclear Information System (INIS)

    Natesan, K.; Uz, M.

    2000-01-01

    Vanadium-base alloys are being considered as candidates for the first wall in advanced V-Li blanket concepts in fusion reactor systems. However, a primary deterrent to the use of these alloys at elevated temperatures is their relatively high affinity for interstitial impurities, i.e., O, N, H, and C. The authors conducted a systematic study to determine the effects of time, temperature, and oxygen partial pressure (pO 2 ) in the exposure environment on O uptake, scaling kinetics, and scale microstructure in V-(4--5) wt.% Cr-(4--5) wt.% Ti alloys. Oxidation experiments were conducted on the alloys at pO 2 in the range of 5 x 10 -6 -760 torr (6.6 x 10 -4 -1 x 10 5 Pa) at several temperatures in the range of 350--700 C. Models that describe the oxidation kinetics, oxide type and thickness, alloy grain size, and depth of O diffusion in the substrate of the two alloys were determined and compared. Weight change data were correlated with time by a parabolic relationship. The parabolic rate constant was calculated for various exposure conditions and the temperature dependence of the constant was described by an Arrhenius relationship. The results showed that the activation energy for the oxidation process is fairly constant at pO 2 levels in the range of 5 x 10 -6 -0.1 torr. The activation energy calculated from data obtained in the air tests was significantly lower, whereas that obtained in pure-O tests (at 760 torr) was substantially higher than the energy obtained under low-pO 2 conditions. The oxide VO 2 was the predominant phase that formed in both alloys when exposed to pO 2 levels of 6.6 x 10 -4 to 0.1 torr. V 2 O 5 was the primary phase in specimens exposed to air and to pure O 2 at 760 torr. The implications of the increased O concentration are increased strength and decreased ductility of the alloy. However, the strength of the alloy was not a strong function of the O concentration of the alloy, but an increase in O concentration did cause a substantial decrease

  11. Influence of alkali metal hydroxides on corrosion of Zr-base alloys

    International Nuclear Information System (INIS)

    Jeong, Yong Hwan

    1996-01-01

    The influence of group-1 alkali hydroxides on different Zr-based alloys have been carried out in static autoclaves at 350 deg C in pressurized water, conditioned in low(0.32 mmol), medium(4.3 mmol) and high(31.5 mmol) equimolar concentration of Li-, Na-, K-, Rb- and Cs-hydroxide. Two types of alloys have been investigated: Zr-Sn-(TRM, Transition metal) and Zr-Sn-Nb-(TRM, Transition metal). From the experiments the cation could be identified as the responsible species for corrosion of Zr alloy in alkalized water. The radius of the cation governs the accelerated corrosion in the pre-transition region of Zr alloy. Incorporation of alkali cation into the zirconium oxide lattice is probably the mechanism which allows the corrosion enhancement for Li and Na and the significant lower effect for the other bases. Nb containing alloys showed lower corrosion resistance than Zr-Sn-TRM alloys in all alkali solutions. Both types of alloys were corroded significantly more in LiOH and NaOH than in the other alkali environments. Lowest corrosive aggressiveness has been found for CsOH followed by KOH. Concluding from the corrosion behavior in the different alkali environments and taking into account the tendency to accelerate the corrosion of Zr alloys, CsOH and KOH are possible alternate alkali for PWR (Pressurized Water Reactor) application. (author)

  12. Modelling of hydrogen assisted cracking of nickel-base Alloy X-750 in water

    International Nuclear Information System (INIS)

    Oka, T.; Ballinger, R.G.; Hwang, I.S.

    1992-01-01

    A closed-form, semi-empirical, electrochemical model has been developed to rationalize the intergranular corrosion fatigue behavior of alloy X-750 in aqueous electrolytes. The model is based on the assumption that, in the electrolytes investigated and for the microstructures studied, that hydrogen assisted crack growth is the dominant mechanism. Further, it is assumed that the rate of hydrogen reduction is a controlling factor in the magnitude of the environmental component of crack growth. Electrolyte conductivity, dissolution and passivation kinetics of precipitates, grain boundary coverage of precipitates are identified as important environmental and microstructural variables governing the hydrogen reduction rate at the crack tip. The model is compared with experimental data for fatigue crack growth where hydrogen is supplied by external charging and with data where galvanically-generated local hydrogen is responsible for enhanced crack growth. It is shown that predicted results characterize the observed effects of frequency, microstructure, electrolyte conductivity, and stress intensity factor. The agreement between the hydrogen reduction model and measured crack growth rate is believed to support the proposed galvanic corrosion mechanism for the intergranular cracking of alloy X-750 in low temperature water

  13. Solidification microstructures of aluminium-uranium alloys

    International Nuclear Information System (INIS)

    Ambrozio Filho, F.; Vieira, R.R.

    1976-01-01

    The solidification of microstrutures of aluminium-uranium alloys in the range of 4 to 20% uranium is investigated. The solidification was obtained both in ingot molds and under controlled directional solidification. The conditions for the presence of primary crystals and eutectic are discussed and an analysis of the influence of variables (growth rate and thermal gradient in the liquid) on the alloy structure is made. The effect of cooling rate on the alloy structures has been determined. It is found that the resulting structure can be derived from the kinectics concept, as required by the coupled-zone theory. Suggestions on the qualitative intervals of composition and temperatures with eutectic growth are presented [pt

  14. Modification effect of Ni-38 wt.%Si on Al-12 wt.%Si alloy

    International Nuclear Information System (INIS)

    Wu Yuying; Liu Xiangfa; Jiang Binggang; Huang Chuanzhen

    2009-01-01

    Modification effect of Ni-38 wt.%Si on the Al-12 wt.%Si alloy has been studied by differential scanning calorimeter, torsional oscillation viscometer and liquid X-ray diffraction experiments. It is found that there is a modification effect of Ni-38 wt.%Si on Al-12 wt.%Si alloy, i.e. primary Si can precipitate in the microstructure of Al-12 wt.%Si alloy when Ni and Si added in the form of Ni-38 wt.%Si, but not separately. Ni-38 wt.%Si alloy brings 'genetic materials' into the Al-Si melt, which makes the melt to form more ordering structure, promotes the primary Si precipitated. Moreover, the addition of Ni-38 wt.%Si, which decreases the solidification supercooling degree of Al-12 wt.%Si alloy, is identical to the effect of heterogeneous nuclei.

  15. Modification effect of Ni-38 wt.%Si on Al-12 wt.%Si alloy

    Energy Technology Data Exchange (ETDEWEB)

    Wu Yuying [Key Laboratory of Liquid Structure and Heredity of Materials, Ministry of Education, Shandong University, Ji' nan 250061 (China)], E-mail: wyy532001@163.com; Liu Xiangfa [Key Laboratory of Liquid Structure and Heredity of Materials, Ministry of Education, Shandong University, Ji' nan 250061 (China); Shandong Binzhou Bohai Piston Co., Ltd., Binzhou 256602, Shandong (China); Jiang Binggang [Key Laboratory of Liquid Structure and Heredity of Materials, Ministry of Education, Shandong University, Ji' nan 250061 (China); Huang Chuanzhen [School of Mechanical Engineering, Shandong University, Jinan 250061 (China)

    2009-05-27

    Modification effect of Ni-38 wt.%Si on the Al-12 wt.%Si alloy has been studied by differential scanning calorimeter, torsional oscillation viscometer and liquid X-ray diffraction experiments. It is found that there is a modification effect of Ni-38 wt.%Si on Al-12 wt.%Si alloy, i.e. primary Si can precipitate in the microstructure of Al-12 wt.%Si alloy when Ni and Si added in the form of Ni-38 wt.%Si, but not separately. Ni-38 wt.%Si alloy brings 'genetic materials' into the Al-Si melt, which makes the melt to form more ordering structure, promotes the primary Si precipitated. Moreover, the addition of Ni-38 wt.%Si, which decreases the solidification supercooling degree of Al-12 wt.%Si alloy, is identical to the effect of heterogeneous nuclei.

  16. Immersion studies on candidate container alloys for the Tuff Repository

    International Nuclear Information System (INIS)

    Beavers, J.A.; Durr, C.L.

    1991-05-01

    Cortest Columbus Technologies (CC Technologies) is investigating the long-term performance of container materials used for high-level radioactive waste packages. This information is being developed for the Nuclear Regulatory Commission to aid in their assessment of the Department of Energy's application to construct a geologic repository for disposal of high-level radioactive waste. This report summarizes the results of exposure studies performed on two copper-base and two Fe-Cr-Ni alloys in simulated Tuff Repository conditions. Testing was performed at 90 degrees C in three environments; simulated J-13 well water, and two environments that simulated the chemical effects resulting from boiling and irradiation of the groundwater. Creviced specimens and U-bends were exposed to liquid, to vapor above the condensed phase, and to alternate immersion. A rod specimen was used to monitor corrosion at the vapor-liquid interface. The specimens were evaluated by electrochemical, gravimetric, and metallographic techniques following approximately 2000 hours of exposure. Results of the exposure tests indicated that all four alloys exhibited acceptable general corrosion rates in simulated J-13 well water. These rates decreased with time. Incipient pitting was observed under deposits on Alloy 825 and pitting was observed on both Alloy CDA 102 and Alloy CDA 715 in the simulated J-13 well water. No SCC was observed in U-bend specimens of any of the alloys in simulated J-13 well water. 33 refs., 48 figs., 23 tabs

  17. Corrosion and oxidation of vanadium-base alloys

    International Nuclear Information System (INIS)

    Loomis, B.A.; Wiggins, G.

    1983-10-01

    The corrosion of several V-base alloys on exposure at elevated temperatures to helium environments containing hydrogen and/or water vapor are presented. These results are utilized to discuss the consequences of the selection of certain radiation-damage resistant, V-base alloys for structural materials applications in a fusion reactor

  18. Corrosion of metals and alloys in the coastal and deep waters of the Arabian Sea and the Bay of Bengal

    Digital Repository Service at National Institute of Oceanography (India)

    Sawant, S.S.; Venkat, K.; Wagh, A.B.

    Corrosion rate of mild steel (MS), stainless steel (SS), copper, brass and cupro-nickel has been determinEd. by exposing metallic coupons in coastal and oceanic waters of the Arabian Sea and Bay of Bengal. Amongst the metals and alloys under study...

  19. Microstructural stability and thermomechanical processing of boron modified beta titanium alloys

    Science.gov (United States)

    Cherukuri, Balakrishna

    One of the main objectives during primary processing of titanium alloys is to reduce the prior beta grain size. Producing an ingot with smaller prior beta grain size could potentially eliminate some primary processing steps and thus reduce processing cost. Trace additions of boron have been shown to decrease the as-cast grain size in alpha + beta titanium alloys. The primary focus of this dissertation is to investigate the effect of boron on microstructural stability and thermomechanical processing in beta titanium alloys. Two metastable beta titanium alloys: Ti-15Mo-2.6Nb-3Al-0.2Si (Beta21S) and Ti-5Al-5V-5Mo-3Cr (Ti5553) with 0.1 wt% B and without boron additions were used in this investigation. Significant grain refinement of the as-cast microstructure and precipitation of TiB whiskers along the grain boundaries was observed with boron additions. Beta21S and Beta21S-0.1B alloys were annealed above the beta transus temperature for different times to investigate the effect of boron on grain size stability. The TiB precipitates were very effective in restricting the beta grain boundary mobility by Zener pinning. A model has been developed to predict the maximum grain size as a function of TiB size, orientation, and volume fraction. Good agreement was obtained between model predictions and experimental results. Beta21S alloys were solution treated and aged for different times at several temperatures below the beta transus to study the kinetics of alpha precipitation. Though the TiB phase did not provide any additional nucleation sites for alpha precipitation, the grain refinement obtained by boron additions resulted in accelerated aging. An investigation of the thermomechanical processing behavior showed different deformation mechanisms above the beta transus temperature. The non-boron containing alloys showed a non-uniform and fine recrystallized necklace structure at grain boundaries whereas uniform intragranular recrystallization was observed in boron containing

  20. On the hardenability of Nb-modified metastable beta Ti-5553 alloy

    Energy Technology Data Exchange (ETDEWEB)

    Campo, K.N.; Andrade, D.R.; Opini, V.C.; Mello, M.G.; Lopes, E.S.N.; Caram, R., E-mail: caram@fem.unicamp.br

    2016-05-15

    Among the commercially available titanium alloys, the metastable β Ti-5553 alloy (Ti–5Al–5V–5Mo–3Cr–0.5Fe wt.%) is an object of great interest because it is employed in aerospace structural applications, primarily in the replacement of steel components. One of the primary advantages of this alloy is its high hardenability, which allows it to retain the β phase at room temperature, even at low cooling rates, thereby allowing the thermoprocessing of thick parts. The aim of this investigation was to evaluate the effect of the replacement of V with Nb on the hardenability of Ti-5553. Based on the molybdenum equivalent criterion, the Nb-modified Ti-5553 alloy was designed to present 12 wt.% of Nb instead of 5 wt.% of V. Samples of both alloys were prepared by melting them in an arc furnace under an inert atmosphere, heat-treated at high temperatures for 12 h and plastic deformed using swage forging. Finally, these samples were solution heat-treated at temperatures above the β-transus followed by cooling at different rates using water quenching, furnace cooling and a modified Jominy end quench test. Characterization was performed by measuring Vickers hardness, X-ray diffraction, and light optical, scanning electron and transmission electron microscopy. The results obtained indicate that metastable β phase can be retained when the cooling rate is higher than 21 °C/s for both alloys. At lower cooling rates, α phase precipitation was observed, but it appeared to be less evident in the Nb-modified Ti-5553, suggesting that the replacement of V with Nb increased the hardenability of the alloy. - Highlights: • Hardenability of Ti alloys are assessed using a modified Jominy end quench test. • Ti-5553 and Nb-modified Ti-5553 are subjected to continuous cooling experiments. • β phase decomposition kinetics is reduced by replacing V with Nb in Ti-5553. • Nb-modified Ti-5553 features improved hardenability. • Replacement of V with Nb causes the

  1. Crack growth rate in the HAZ of alloy 600/182

    Energy Technology Data Exchange (ETDEWEB)

    Gomez-Briceno, D.; Lapena, J.; Garcia-Redondo, M.; Castro, L.; Perosanz, F.J. [CIEMAT (Spain); Ahluwalia, K. [EPRI, (United States); Hickling, J. [EPRI Consultant (Cyprus)

    2011-07-01

    CGR (Crack Growth Rate) experiments to obtain data for the HAZ (Heat Affected Zone) of nickel base alloys using fracture mechanics specimens are a challenge, primarily due to the difficulties of positioning the tip of the notch (or pre-crack) in the desired location within the complex region adjacent to the fusion line. This paper presents some results obtained in an experimental program carried out to the CGR in the HAZ of several welded Alloy 600 plates. Compact tension (CT) specimens have been tested in simulated PWR primary water at temperatures of 340 and 360 C degrees under cyclic and constant loading (both with and without periodic partial unloading). Satisfactory CGR data were obtained for the HAZ in an Alloy 600 plate (mill annealed at high temperature) welded with Alloy 182 under both environmentally assisted fatigue test conditions (cyclic loading at different frequencies) and during stress corrosion testing (i.e. at predominantly constant load). The CGR values were generally similar to those obtained for the corresponding base metal (with tentative evidence for slightly faster growth in the HAZ under pure constant load). The HAZ specimens showed a higher tendency to crack inter-granularly under cyclic loading. CGR values under predominantly SCC conditions corresponded well (after temperature correction) with the MRP - 55 75. percentile disposition curve for PWSCC in Alloy 600 materials. This contrasts with the behavior observed by other investigators, where the HAZ material was found to exhibit markedly higher CGRs. A possible explanation for this discrepancy is the higher PWSCC susceptibility of the Alloy 600 base metal used to prepare the HAZ specimens in this program. It appears that the strong increase in the HAZ CGR observed elsewhere may take place if the base metal is a heat with inherently low PWSCC susceptibility (i.e. with good microstructure, adequate carbide distribution, etc.). However, if the Alloy 600 base metal already has a susceptible

  2. Crack growth rate in the HAZ of alloy 600/182

    International Nuclear Information System (INIS)

    Gomez-Briceno, D.; Lapena, J.; Garcia-Redondo, M.; Castro, L.; Perosanz, F.J.; Ahluwalia, K.; Hickling, J.

    2011-01-01

    CGR (Crack Growth Rate) experiments to obtain data for the HAZ (Heat Affected Zone) of nickel base alloys using fracture mechanics specimens are a challenge, primarily due to the difficulties of positioning the tip of the notch (or pre-crack) in the desired location within the complex region adjacent to the fusion line. This paper presents some results obtained in an experimental program carried out to the CGR in the HAZ of several welded Alloy 600 plates. Compact tension (CT) specimens have been tested in simulated PWR primary water at temperatures of 340 and 360 C degrees under cyclic and constant loading (both with and without periodic partial unloading). Satisfactory CGR data were obtained for the HAZ in an Alloy 600 plate (mill annealed at high temperature) welded with Alloy 182 under both environmentally assisted fatigue test conditions (cyclic loading at different frequencies) and during stress corrosion testing (i.e. at predominantly constant load). The CGR values were generally similar to those obtained for the corresponding base metal (with tentative evidence for slightly faster growth in the HAZ under pure constant load). The HAZ specimens showed a higher tendency to crack inter-granularly under cyclic loading. CGR values under predominantly SCC conditions corresponded well (after temperature correction) with the MRP - 55 75. percentile disposition curve for PWSCC in Alloy 600 materials. This contrasts with the behavior observed by other investigators, where the HAZ material was found to exhibit markedly higher CGRs. A possible explanation for this discrepancy is the higher PWSCC susceptibility of the Alloy 600 base metal used to prepare the HAZ specimens in this program. It appears that the strong increase in the HAZ CGR observed elsewhere may take place if the base metal is a heat with inherently low PWSCC susceptibility (i.e. with good microstructure, adequate carbide distribution, etc.). However, if the Alloy 600 base metal already has a susceptible

  3. An evaluation of selection criteria on primary water chemistry parameters for SMART

    International Nuclear Information System (INIS)

    Choi, B. S.; Kim, S. H.; Yun, J. H.; Bae, Y. Y.; Gee, S. G.

    2003-01-01

    The selection criteria on the primary water chemistry of SMART by comparing the chemical design features with those of the current operating PWRs is analyzed. The most essential differences in water chemistry between the PWRs and SMART reactor is characterized by the presence of boron in water. SMART is boron free reactor, and the ammonia is used as a pH reagent. In SMART reactor hydrogen gas is not added to the primary coolant, but is normally generated from the radiolysis of ammonia of the coolant passes through the core. Ammonia is added once per shift because SMART reactor has no letdown and charging system during power operation. Because of these competing processes, the concentrations of hydrogen, nitrogen and ammonia in the primary coolant are steady state concentrations, which depend on the decomposition/combination rate of ammonia. Ammonia chemistry in SMART reactor has many advantages in that no hydrogen gas injection is needed to control the dissolved oxygen in primary coolant because of spontaneous generation of hydrogen and nitrogen produced by the reaction of ammonia decomposition

  4. Processing of Refractory Metal Alloys for JOYO Irradiations

    International Nuclear Information System (INIS)

    RF Luther; ME Petrichek

    2006-01-01

    This is a summary of the refractory metal processing experienced by candidate Prometheus materiats as they were fabricated into specimens destined for testing within the JOYO test reactor, ex-reactor testing at Oak Ridge National Laboratory (ORNL), or testing within the NRPCT. The processing is described for each alloy from the point of inception to the point where processing was terminated due to the cancellation of Naval Reactor's involvement in the Prometheus Project. The alloys included three tantalum-base alloys (T-111, Ta-10W, and ASTAR-811C), a niobium-base alloy, (FS-85), and two molybdenum-rhenium alloys, one containing 44.5 w/o rhenium, and the other 47.5 w/o rhenium. Each of these alloys was either a primary candidate or back-up candidate for cladding and structural applications within the space reactor. Their production was intended to serve as a forerunner for large scale production ingots that were to be procured from commercial refractory metal vendors such as Wah Chang

  5. Application of Ultra High Pressure Cavitation Peening to Prevent PWSCC on Primary Plant Components

    Energy Technology Data Exchange (ETDEWEB)

    Poling, G.R.

    2015-07-01

    Primary Water Stress Corrosion Cracking (PWSCC) on Alloy 600/82/182 susceptible materials can lead to increased costs for maintenance and repair/replacement activities on nuclear power plant primary components. A process called Ultra High Pressure (UHP) cavitation peening can be safely and cost effectively applied to the susceptible materials to generate compressive stresses on the surface and prevent PWSCC initiation. AREVA has developed the tooling systems to apply the UHP cavitation peening process on reactor vessel head penetration nozzles, bottom mounted nozzles and primary nozzles. Applying the UHP cavitation peening process before PWSCC initiation will prevent future repairs/replacements, reduce maintenance costs, and provide more effective on-time for the reactor. (Author)

  6. Water chemistry of Atucha II PHWVR. Design concepts and evolution

    International Nuclear Information System (INIS)

    Chocron, Mauricio; Rodriguez, Ivanna; Duca, Jorge; Fernandez, Ricardo; Rico, Jorge

    2007-01-01

    Full text: Atucha II is a pressurized heavy water vessel reactor designed by Siemens-KWU, currently part of AREVA NP, of 745 MWe and similar to Atucha I, which has been in operation over 25 years. The primary heat transport system (PHTS) is composed by vertical channels (277-313 C degrees) that allocate the fuel elements while the moderator circuit is composed by a partially separated circuit (142-173 C degrees). The moderation power is transferred to the feedwater through the moderator heat exchangers (HX). These HXs operate as the last, high pressure water-steam cycle heaters as well. Materials (with exception of fuel channels and fuel sheaths which are made of zirconium alloys) are all austenitic steels while cobalt containing alloys have been all replaced at the design stage. Steam generator and moderator HX tubing are Alloy 800 made. The core is operated without boron except with the first fresh nucleus. The secondary circuit or Balance of plant (BOP) is similar in conception to that of a PWR but the moderator HXs. It is entirely built of ferrous alloys, has a feedwater-deaerator tank and moisture separator. The energy sink is the Rio de la Plata River. The Reactors Chemistry Department, Chemistry Division, National Atomic Energy Commission, in its character of R and D institution has been committed by CNA II-N.A.S.A Project to prepare the water chemistry specifications, water chemistry engineering and manuals, considering the type of reactor, design and construction aspects and operation characteristics, taking into account the current state-of-the art and worldwide standards. This includes conceptual aspects and implementation and operative aspects as well. This documentation will be released after a designer's review as it has been stated in the respective agreement. Respecting the confidentiality agreement between CNEA and NASA and the confidentiality regarding handling original documentation provided by the designer, it is considered illustrative to

  7. Generalized corrosion of nickel base alloys in high temperature aqueous media: a contribution to the comprehension of the mechanisms; Corrosion generalisee des alliages a base nickel en milieu aqueux a haute temperature: apport a la comprehension des mecanismes

    Energy Technology Data Exchange (ETDEWEB)

    Marchetti-Sillans, L

    2007-11-15

    In France, nickel base alloys, such as alloy 600 and alloy 690, are the materials constituting steam generators (SG) tubes of pressurized water reactors (PWR). The generalized corrosion resulting from the interaction between these alloys and the PWR primary media leads, on the one hand, to the formation of a thin protective oxide scale ({approx} 10 nm), and on the other hand, to the release of cations in the primary circuit, which entails an increase of the global radioactivity of this circuit. The goal of this work is to supply some new comprehension elements about nickel base alloys corrosion phenomena in PWR primary media, taking up with underlining the effects of metallurgical and physico-chemical parameters on the nature and the growth mechanisms of the protective oxide scale. In this context, the passive film formed during the exposition of alloys 600, 690 and Ni-30Cr, in conditions simulating the PWR primary media, has been analyzed by a set of characterization techniques (SEM, TEM, PEC and MPEC, XPS). The coupling of these methods leads to a fine description, in terms of nature and structure, of the multilayered oxide forming during the exposition of nickel base alloys in primary media. Thus, the protective part of the oxide scale is composed of a continuous layer of iron and nickel mixed chromite, and Cr{sub 2}O{sub 3} nodules dispersed at the alloy / mixed chromite interface. The study of protective scale growth mechanisms by tracers and markers experiments reveals that the formation of the mixed chromite is the consequence of an anionic mechanism, resulting from short circuits like grain boundaries diffusion. Besides, the impact of alloy surface defects has also been studied, underlining a double effect of this parameter, which influences the short circuits diffusion density in oxide and the formation rate of Cr{sub 2}O{sub 3} nodules. The sum of these results leads to suggest a description of the nickel base alloys corrosion mechanisms in PWR primary

  8. Metal-ceramic alloys in dentistry: a review.

    Science.gov (United States)

    Roberts, Howard W; Berzins, David W; Moore, B Keith; Charlton, David G

    2009-02-01

    The purpose of this article is to review basic information about the alloys used for fabricating metal-ceramic restorations in dentistry. Their compositions, properties, advantages, and disadvantages are presented and compared. In addition to reviewing traditional noble-metal and base-metal metal-ceramic alloys, titanium and gold composite alloys are also discussed. A broad search of the published literature was performed using Medline to identify pertinent current articles on metal-ceramic alloys as well as articles providing a historical background about the development of these alloys. Textbooks, the internet, and manufacturers' literature were also used to supplement this information. The review discusses traditional as well as more recently-developed alloys and technologies used in dentistry for fabricating metal-ceramic restorations. Clear advantages and disadvantages for these alloy types are provided and discussed as well as the role that compositional variations have on the alloys' performance. This information should enable clinicians and technicians to easily identify the important physical properties of each type and their primary clinical indications. A number of alloys and metals are available for metal-ceramic use in dentistry. Each has its advantages and disadvantages, primarily based on its specific composition. Continuing research and development are resulting in the production of new technologies and products, giving clinicians even more choices in designing and fabricating metal-ceramic restorations.

  9. THE BEHAVIOR OF SOLUBLE METALS ELUTED FROM Ni/Fe-BASED ALLOY REACTORS AFTER HIGH-TEMPERATURE AND HIGH-PRESSURE WATER PROCESS

    Directory of Open Access Journals (Sweden)

    M. Faisal

    2012-05-01

    Full Text Available The behavior of heavy metals eluted from the wall of Ni/Fe-based alloy reactors after high-temperature and high-pressure water reaction were studied at temperatures ranging from 250 to 400oC. For this purpose, water and cysteic acid were heated in two reactor materials which are SUS 316 and Inconel 625. Under the tested conditions, the erratic behaviors of soluble metals eluted from the wall of Ni/Fe-based alloy in high temperature water were observed. Results showed that metals could be eluted even at a short contact time. The presence of air also promotes elution at sub-critical conditions. At sub-critical conditions, a significant amount of Cr was extracted from SUS 316, while only traces of Ni, Fe, Mo and Mn were eluted. In contrast, Ni was removed in significant amounts compared to Cr when Inconel 625 was tested. It was observed that eluted metals tend to increased under acidic conditions and most of those metals were over the limit of WHO guideline for drinking water. The results are significant both on the viewpoint of environmental regulation on disposal of wastes containing heavy metals, toxicity of resulting product and catalytic effect on a particular reaction.

  10. Oxidation of uranium and uranium alloys

    International Nuclear Information System (INIS)

    Orman, S.

    1976-01-01

    The corrosion behaviour of uranium in oxygen, water and water + oxygen mixtures is compared and contrasted. A considerable amount of work, much of it conflicting, has been published on the U + H 2 O and U + H 2 O + O 2 systems. An attempt has been made to summarise this data and to explain the reasons for the lack of agreement between the experimental results. The evidence for the mechanism involving OH - ion diffusion as the reacting entity in both the U + H 2 O and U + O 2 + H 2 O reactions is advanced. The more limited corrosion data on some lean uranium alloys and on some higher addition alloys referred to as stainless materials is summarised together with some previously unreported results obtained with these materials at AWRE. The data indicates that in the absence of oxygen the lean alloys behave in a similar manner to uranium and evolve hydrogen in approximately theoretical quantities. But the stainless alloys absorb most of the product hydrogen and assessments of reactivity based on hydrogen evolution would be very inaccurate. The direction that future corrosion work on these materials should take is recommended

  11. Effect of high power ultrasound on mechanical properties of Al-Si alloys

    Science.gov (United States)

    Srivastava, N.; Gupta, R.; Chaudhari, G. P.

    2018-03-01

    Effect of high power ultrasonic treatment on the solidification microstructures of Al-Si alloys containing varying content of solute Si (1, 2, 3 and 5 wt %) is investigated. Large variation in microstructures is seen and refinement of primary α-Al grains is observed. It is observed that increasing the weight percentage of solute along with ultrasonic treatment resulted in finer primary phase. By increasing the solute content from 1% to 5 wt.% in Al-Si alloys, hardness increased by about 38% without and 48% with ultrasonic treatment. Tensile strength of the alloys with ultrasonic treatment is higher as compared to those without ultrasonic treated.

  12. Effect of carbon content on solidification behaviors and morphological characteristics of the constituent phases in Cr-Fe-C alloys

    International Nuclear Information System (INIS)

    Lin, Chi-Ming; Lai, Hsuan-Han; Kuo, Jui-Chao; Wu, Weite

    2011-01-01

    A combination of transmission electron microscopy, electron backscatter diffraction and wavelength dispersive spectrum has been used to identify crystal structure, grain boundary characteristic and chemical composition of the constituent phases in Cr-Fe-C alloys with three different carbon concentrations. Depending on the three different carbon concentrations, the solidification structures are found to consist of primary α-phase and [α + (Cr,Fe) 23 C 6 ] eutectic in Cr-18.4Fe-2.3 C alloy; primary (Cr,Fe) 23 C 6 and [α + (Cr,Fe) 23 C 6 ] eutectic in Cr-24.5Fe-3.8 C alloy and primary (Cr,Fe) 7 C 3 and [α + (Cr,Fe) 7 C 3 ] eutectic in Cr-21.1Fe-5.9 C alloy, respectively. The grain boundary analysis is useful to understand growth mechanism of the primary phase. The morphologies of primary (Cr,Fe) 23 C 6 and (Cr,Fe) 7 C 3 carbides are faceted structures with polygonal shapes, different from primary α-phase with dendritic shape. The primary (Cr,Fe) 23 C 6 and (Cr,Fe) 7 C 3 carbides with strong texture exist a single crystal structure and contain a slight low angle boundary, resulting in the polygonal growth mechanism. Nevertheless, the primary α-phase with relative random orientation exhibits a polycrystalline structure and comprises a massive high-angle boundary, caused by the dendritic growth mechanism. - Highlights: ► Microstructures of the as-clad Cr-based alloys are characterized by TEM. ► EBSD technique has been use to characterize the grain boundary of primary phases. ► We examine transitions in morphology about the primary phases. ► Morphologies of primary carbides are polygonal different from primary α-phase. ► Solidification structures rely on C concentrations in Cr-Fe-C alloy.

  13. Electrical Resistance Alloys and Low-Expansion Alloys

    DEFF Research Database (Denmark)

    Kjer, Torben

    1996-01-01

    The article gives an overview of electrical resistance alloys and alloys with low thermal expansion. The electrical resistance alloys comprise resistance alloys, heating alloys and thermostat alloys. The low expansion alloys comprise alloys with very low expansion coefficients, alloys with very low...... thermoelastic coefficients and age hardenable low expansion alloys....

  14. Optimization of the dissolved hydrogen level in PWR to mitigate stress corrosion cracking of nickel alloys. Bibliographic review, modelling and recommendations

    International Nuclear Information System (INIS)

    Labousse, M.; Deforge, D.; Gressier, F.; Taunier, S.; Le Calvar, M.

    2012-09-01

    Nickel based alloys Stress Corrosion Cracking (SCC) has been a major concern for the Nuclear Power Plants (NPP) utilities since more than 40 years. At EDF, this issue led to the replacement of all upper vessel heads and of most of the steam generators with Alloy 600 MA tubes. Under the scope of plant lifetime extension, there is some concerns about the behaviour of Bottom Mounted Instrumentation Nozzles (BMI) made of Alloy 600 welded with Alloy 182 and a few vessel dissimilar metal welds made of Alloy 82, for only three 1450 MWe plants. It is considered for long that Primary Water Stress Corrosion Cracking (PWSCC) is influenced by the dissolved hydrogen (DH) level in primary coolant. Now, the whole community clearly understands that there is a hydrogen level corresponding to a maximum in terms of SCC susceptibility. Many experimental studies were done worldwide to optimize the hydrogen level in primary water during power operation, both in terms of SCC initiation and propagation. From these studies, most of American plants decided to increase the dissolved hydrogen level in order to mitigate crack propagation. Conversely, in Japan, based on crack initiation data, it is thought that drastically decreasing the hydrogen content would rather be beneficial. In order to consolidate EDF position, a review of laboratory tests data was made. Studies on the influence of hydrogen on nickel alloys 600 and 182 PWSCC were compiled and rationalized. Data were collapsed using a classical Gaussian model, such as initially proposed by Morton et al. An alternative model based on more phenomenological considerations was also proposed. Both models lead to similar results. The maximum susceptibility to SCC cracking appears to be rather consistent with the Ni/NiO transition, which was not taken as an initial hypothesis. Regarding crack initiation, an inverse Gaussian model was proposed. Based on the current hydrogen concentration range during power operation and considering components

  15. Physical and welding metallurgy of Gd-enriched austenitic alloys for spent nuclear fuel applications. Part II, nickel base alloys

    International Nuclear Information System (INIS)

    Mizia, Ronald E.; Michael, Joseph Richard; Williams, David Brian; Dupont, John Neuman; Robino, Charles Victor

    2004-01-01

    The physical and welding a metallurgy of gadolinium- (Gd-) enriched Ni-based alloys has been examined using a combination of differential thermal analysis, hot ductility testing. Varestraint testing, and various microstructural characterization techniques. Three different matrix compositions were chosen that were similar to commercial Ni-Cr-Mo base alloys (UNS N06455, N06022, and N06059). A ternary Ni-Cr-Gd alloy was also examined. The Gd level of each alloy was ∼2 wt-%. All the alloys initiated solidification by formation of primary austenite and terminated solidification by a Liquid γ + Ni 5 Gd eutectic-type reaction at ∼1270 C. The solidification temperature ranges of the alloys varied from ∼100 to 130 C (depending on alloy composition). This is a substantial reduction compared to the solidification temperature range to Gd-enriched stainless steels (360 to 400 C) that terminate solidification by a peritectic reaction at ∼1060 C. The higher-temperature eutectic reaction that occurs in the Ni-based alloys is accompanied by significant improvements in hot ductility and solidification cracking resistance. The results of this research demonstrate that Gd-enriched Ni-based alloys are excellent candidate materials for nuclear criticality control in spent nuclear fuel storage applications that require production and fabrication of large amounts of material through conventional ingot metallurgy and fusion welding techniques

  16. Cloud point extraction of palladium in water samples and alloy mixtures using new synthesized reagent with flame atomic absorption spectrometry (FAAS)

    International Nuclear Information System (INIS)

    Priya, B. Krishna; Subrahmanayam, P.; Suvardhan, K.; Kumar, K. Suresh; Rekha, D.; Rao, A. Venkata; Rao, G.C.; Chiranjeevi, P.

    2007-01-01

    The present paper outlines novel, simple and sensitive method for the determination of palladium by flame atomic absorption spectrometry (FAAS) after separation and preconcentration by cloud point extraction (CPE). The cloud point methodology was successfully applied for palladium determination by using new reagent 4-(2-naphthalenyl)thiozol-2yl azo chromotropic acid (NTACA) and hydrophobic ligand Triton X-114 as chelating agent and nonionic surfactant respectively in the water samples and alloys. The following parameters such as pH, concentration of the reagent and Triton X-114, equilibrating temperature and centrifuging time were evaluated and optimized to enhance the sensitivity and extraction efficiency of the proposed method. The preconcentration factor was found to be (50-fold) for 250 ml of water sample. Under optimum condition the detection limit was found as 0.067 ng ml -1 for palladium in various environmental matrices. The present method was applied for the determination of palladium in various water samples, alloys and the result shows good agreement with reported method and the recoveries are in the range of 96.7-99.4%

  17. Solidified structure of Al-Pb-Cu alloys

    International Nuclear Information System (INIS)

    Ikeda, Tetsuyuki; Nishi, Seiki; Kumeuchi, Hiroyuki; Tatsuta, Yoshinori.

    1986-01-01

    Al-Pb-Cu alloys were cast into bars or plates in different two metal mold casting processes in order to suppress gravity segregation of Pb and to achieve homogeneous dispersion of Pb phase in the alloys. Solidified structures were analyzed by a video-pattern-analyzer. Plate castings 15 to 20 mm in thickness of Al-Pb-1 % Cu alloy containing Pb up to 5 % in which Pb phase particles up to 10 μm disperse are achieved through water cooled metal mold casting. The plates up to 5 mm in thickness containing Pb as much as 8 to 10 % cast in this process have dispersed Pb particles up to 5 μm in diameter in the surface layer. Al-8 % Pb-1 % Cu alloy bars 40 mm in diameter and 180 mm in height in which gravity segregation of Pb is prevented can be cast by movable and water sprayed metal mold casting at casting temperature 920 deg C and mold moving speed 1.0 mm/s. Pb phase particles 10 μm in mean size are dispersed in the bars. (author)

  18. Corrosion and protection of magnesium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Ghali, E. [Laval Univ., Quebec City, PQ (Canada). Dept. of Mining and Metallurgy

    2000-07-01

    The oxide film on magnesium offers considerable surface protection in rural and some industrial environments and the corrosion rate lies between that of aluminum and low carbon steels. Galvanic coupling of magnesium alloys, high impurity content such as Ni, Fe, Cu and surface contamination are detrimental for corrosion resistance of magnesium alloys. Alloying elements can form secondary particles which are noble to the Mg matrix, thereby facilitating corrosion, or enrich the corrosion product thereby possibly inhibiting the corrosion rate. Bimetallic corrosion resistance can be increased by fluxless melt protection, choice of compatible alloys, insulating materials, and new high-purity alloys. Magnesium is relatively insensible to oxygen concentration. Pitting, corrosion in the crevices, filiform corrosion are observed. Granular corrosion of magnesium alloys is possible due to the cathodic grain-boundary constituent. More homogeneous microstructures tend to improve corrosion resistance. Under fatigue loading conditions, microcrack initiation in Mg alloys is related to slip in preferentially oriented grains. Coating that exclude the corrosive environments can provide the primary defense against corrosion fatigue. Magnesium alloys that contain neither aluminum nor zinc are the most SCC resistant. Compressive surface residual stresses as that created by short peening increase SCC resistance. Cathodic polarization or cladding with a SCC resistant sheet alloy are good alternatives. Effective corrosion prevention for magnesium alloy components and assemblies should start at the design stage. Selective surface preparation, chemical treatment and coatings are recommended. Oil application, wax coating, anodizing, electroplating, and painting are possible alternatives. Recently, it is found that a magnesium hydride layer, created on the magnesium surface by cathodic charging in aqueous solution is a good base for painting. (orig.)

  19. Nuclear fuel element containing strips of an alloyed Zr, Ti, and Ni getter material

    International Nuclear Information System (INIS)

    Grossman, L.N.; Packard, D.R.

    1975-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed. The nuclear fuel element has disposed therein an alloy having the essential components of nickel, titanium and zirconium, and the alloy reacts with water, water vapor and reactive gases at reactor ambient temperatures. The alloy is disposed in the plenum of the fuel element in the form of strips and preferably the strips are positioned inside a helical member in the plenum. The position of the alloy strips permits gases and liquids entering the plenum to contact and react with the alloy strips. (U.S.)

  20. Evolution of a novel Si-18Mn-16Ti-11P alloy in Al-Si melt and its influence on microstructure and properties of high-Si Al-Si alloy

    Directory of Open Access Journals (Sweden)

    Xiao-Lu Zhou

    Full Text Available A novel Si-18Mn-16Ti-11P master alloy has been developed to refine primary Si to 14.7 ± 1.3 μm, distributed uniformly in Al-27Si alloy. Comparing with traditional Cu-14P and Al-3P, Si-18Mn-16Ti-11P provided a much better refining effect, with in-situ highly active AlP. The refined Al-27Si alloy exhibited a CTE of 16.25 × 10−6/K which is slightly higher than that of Sip/Al composites fabricated by spray deposition. The UTS and elongation of refined Al-27Si alloy were increased by 106% and 235% comparing with those of unrefined alloy. It indicates that the novel Si-18Mn-16Ti-11P alloy is more suitable for high-Si Al-Si alloys and may be a candidate for refining hypereutectic Al-Si alloy for electronic packaging applications. Moreover, studies showed that TiP is the only P-containing phase in Si-18Mn-16Ti-11P master alloy. A core-shell reaction model was established to reveal mechanism of the transformation of TiP to AlP in Al-Si melts. The transformation is a liquid-solid diffusion reaction driven by chemical potential difference and the reaction rate is controlled by diffusion. It means sufficient holding time is necessary for Si-18Mn-16Ti-11P master alloy to achieve better refining effect. Keywords: Hypereutectic Al-Si alloy, Primary Si, Refinement, AlP, Thermal expansion behavior, Si-18Mn-16Ti-11P master alloy

  1. Material reliability of Ni alloy electrodeposition for steam generator tube repair

    International Nuclear Information System (INIS)

    Kim, Dong Jin; Kim, Myong Jin; Kim, Joung Soo; Kim, Hong Pyo

    2007-01-01

    Due to the occasional occurrences of Stress Corrosion Cracking (SCC) in steam generator tubing (Alloy 600), degraded tubes are removed from service by plugging or are repaired for re-use. Since electrodeposition inside a tube dose not entail parent tube deformation, residual stress in the tube can be minimized. In this work, tube restoration via electrodeposition inside a steam generator tubing was performed after developing the following: an anode probe to be installed inside a tube, a degreasing condition to remove dirt and grease, an activation condition for surface oxide elimination, a tightly adhered strike layer forming condition between the electroforming layer and the Alloy 600 tube, and the condition for an electroforming layer. The reliability of the electrodeposited material, with a variation of material properties, was evaluated as a function of the electrodeposit position in the vertical direction of a tube using the developed anode. It has been noted that the variation of the material properties along the electrodeposit length was acceptable in a process margin. To improve the reliability of a material property, the causes of the variation occurrence were presumed, and an attempt to minimize the variation has been made. A Ni alloy electrodeposition process is suggested as a Primary Water Stress Corrosion Cracking (PWSCC) mitigation method for various components, including steam generator tubes. The Ni alloy electrodeposit formed inside a tube by using the installed assembly shows proper material properties as well as an excellent SCC resistance

  2. Autoclave Testing on Zirconium Alloy Materials

    International Nuclear Information System (INIS)

    Hoffmann, Petra-Britt; Sell, Hans-Juergen; Garzarolli, Friedrich

    2012-09-01

    The corrosion of Zirconium components like fuel rod claddings and spacer grids is limiting lifetime and duty of these components. In Pressurized and Boiling Water Reactors (PWR and BWR), different corrosion phenomena are of interest. Although in-pile experience is the final proof for a material development, significant experience was gained by autoclave tests, trying to simulate in-pile conditions but reducing time for return of experience by increased temperatures. For PWR application, the uniform corrosion is studied in water at up to 370 deg. C and in high pressure steam at 400 deg. C, and for BWR, the nodular corrosion is studied in high pressure steam at 500-520 deg. C. Particular attention has to be given to the corrosion media, because oxidative traces in the water can significantly affect the corrosion response. An extensive air removal is thus important for all corrosion tests. This links to the different water chemistry conditions that have been investigated as separate effects otherwise difficult to separate under in-pile conditions. Uniform corrosion in 350 deg. C water is usually a cyclic process with repeated rate transitions. In addition, at high exposure times an acceleration of corrosion can occur, e.g. for Zr-Sn alloys with a high Sn content. In 400 deg. C steam, corrosion rate decreases somewhat with increasing time. Uniform corrosion rate of Zr alloys depends on their Sn- and Fe+Cr contents as well as on their annealing parameters with a similar trend as in PWR and on their yield strength, however with an opposite trend compared to BWR conditions. Nodular corrosion of BWR alloys depends on the annealing parameter with a similar trend as in PWR and out-of-reactor also significantly on the Fe+Cr content. The hydrogen pickup fraction (HPUF) depends largely on details of the water chemistry and can particularly depend on autoclave degassing and probably also on autoclave contaminations. Thus any HPUF value from out-of- pile corrosion tests is only

  3. Computing elastic anisotropy to discover gum-metal-like structural alloys

    Science.gov (United States)

    Winter, I. S.; de Jong, M.; Asta, M.; Chrzan, D. C.

    2017-08-01

    The computer aided discovery of structural alloys is a burgeoning but still challenging area of research. A primary challenge in the field is to identify computable screening parameters that embody key structural alloy properties. Here, an elastic anisotropy parameter that captures a material's susceptibility to solute solution strengthening is identified. The parameter has many applications in the discovery and optimization of structural materials. As a first example, the parameter is used to identify alloys that might display the super elasticity, super strength, and high ductility of the class of TiNb alloys known as gum metals. In addition, it is noted that the parameter can be used to screen candidate alloys for shape memory response, and potentially aid in the optimization of the mechanical properties of high-entropy alloys.

  4. Surface reactivity of colloidal corrosion product and alloys in PWR conditions

    International Nuclear Information System (INIS)

    Lefevre, Gregory; Leclercq, Stephanie; Cabanas, Bruna-Martin; Delaunay, Sophie; Mansour, Carine; Berger, Gilles

    2012-09-01

    The corrosion of metallic components of water circuits of Pressurized Water Reactors generates colloidal particles. These particles are transported in the circuits, they sorb dissolved species and they can deposit on alloys in given parts of the circuits. Sorption and deposition generate several technical drawbacks in both primary and secondary circuits. According to the DLVO theory, adhesion between two surfaces is controlled by electrostatic and Van der Waals forces. The latter are always attractive and does not depends on solution chemistry. On the contrary, electrostatic forces are connected to the surface charge and depend strongly on the chemical properties of the solids and on the chemistry of the solution. Depending on the relative charge of the surfaces, these forces are attractive or repulsive and can have a major effect on the deposition behavior of particles. According to the surface complexation theory, the surface charge of metallic oxides results from sorption or desorption of protons, leading to positive or negative surface sites, and thus, strongly depends on the solution pH. Dissolved species can sorb on the surface, depending on the ionic charge of these species and on the surface charge. Thus, the knowledge of the surface charge of corrosion particles and alloys, their affinity towards several ions as protons, nickel, cobalt, sulfate, or borate ions has been shown to be useful to predict the transport of the contamination in the primary circuit, or to understand the accumulation of impurities in the steam generator in the secondary circuit. At room temperature, these data can be easily measured, or found in literature. In PWR conditions (high temperature, high pressure), most of the usual protocols and commercial instruments cannot be used. For several years, collaboration between EDF R and D and CNRS has been developed to get information about the surface reactivity of iron oxides, ferrites, and alloys in such conditions. Some of the results

  5. Development the Mechanical Properties of (AL-Li-Cu Alloy

    Directory of Open Access Journals (Sweden)

    Ihsan Kadhom AlNaimi

    2017-11-01

    Full Text Available The aim of this research is to develop mechanical properties of a new aluminium-lithium-copper alloy. This alloy prepared under control atmosphere by casting in a permanent metal mould. The microstructure was examined and mechanical properties were tested before and after heat treatment to study the influence of heat treatment on its mechanical properties including; modulus of elasticity, tensile strength, impact, and fatigue. The results showed that the modulus of elasticity of the prepared alloy is higher than standard alloy about 2%. While the alloy that heat treated for 6 h and cooled in water, then showed a higher ultimate tensile stress comparing with as-cast alloy. The homogenous heat treatment gives best fatigue behaviour comparing with as-cast and other heat treatment alloys. Also, the impact test illustrates that the homogeneous heat treatment alloy gives the highest value.

  6. 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.

  7. Temperature dependence and hysteresis of the initial permeability of the 50%Ni - 50%Fe alloy

    International Nuclear Information System (INIS)

    Kekalo, I.B.; Stolyarov, V.L.; Patsionov, V.A.

    1979-01-01

    Studied has been a temperature dependence of the initial permeability of the 50% Ni - 50% Fe alloy after primary and secondary recrystallization and effect of thermomagnetic treatment upon the dependence. For all the alloys with the structure of primary recrystallization a monotonous increase of initial permeability with temperature and the presence of slight temperature hysteresis are typical. Thermomagnetic treatment, not affecting considerably the temperature dependence of permeability for all the primarily recrystallized alloys, changes to a great extent the character of the dependence in the secondary recrystallized alloys. For 20-200-20 deg C temperature cycle of the alloys with secondary recrystallized structure are characterized after thermomagnetic treatment by the presence of gigantic hysteresis of initial permeability and a maximum on the heating branch of the curve in the vicinity of 130 deg C which are accounted for by peculiarities of temperature hysteresis of domain structure in the given alloy

  8. Microstructural Characterization of Aluminum-Lithium Alloys 1460 and 2195

    Science.gov (United States)

    Wang, Z. M.; Shenoy, R. N.

    1998-01-01

    Transmission electron microscopy (TEM) and differential scanning calorimetry (DSC) techniques were employed to characterize the precipitate distributions in lithium-containing aluminum alloys 1460 and 2195 in the T8 condition. TEM examinations revealed delta prime and T1 as the primary strengthening precipitates in alloys 1460 and 2195 respectively. TEM results showed a close similarity of the Russian alloy 1460 to the U.S. alloy 2090, which has a similar composition and heat treatment schedule. DSC analyses also indicate a comparable delta prime volume fraction. TEM study of a fractured tensile sample of alloy 1460 showed that delta prime precipitates are sheared by dislocations during plastic deformation and that intense stress fields arise at grain boundaries due to planar slip. Differences in fracture toughness of alloys 1460 and 2195 are rationalized on the basis of a literature review and observations from the present study.

  9. Electrochemical corrosion characteristics of aluminium alloy 6061 T6 in demineralized water containing 0.1 % chloride ion

    International Nuclear Information System (INIS)

    Zaifol Samsu; Muhammad Daud; Siti Radiah Mohd Kamarudin; Mohd Saari Ripin; Rusni Rejab; Mohd Shariff Sattar

    2012-01-01

    Direct current electrochemical method is one of the techniques has been used to study the corrosion behaviour of metal/alloy in its environment. This paper attempts to investigate the corrosion behaviour of Al 6061 T6 immersed in Reactor TRIGA Mark II pool water containing about 0.1% NaCl content. The result shown that the corrosion rate value of the aluminium 6061 T6 increased with the presence of 0.1 % Ion Chloride content in the demineralized water reactor pool as compared to normal demineralized water. This is due to aggressiveness of chloride ion attack to metal surface. Beside corrosion rate analysis, the further tests such as corrosion behaviour diagram, cyclic polarization have been carried and the results have been reported. (author)

  10. Phases in lanthanum-nickel-aluminum alloys

    International Nuclear Information System (INIS)

    Mosley, W.C.

    1992-01-01

    Lanthanum-nickel-aluminum (LANA) alloys will be used to pump, store and separate hydrogen isotopes in the Replacement Tritium Facility (RTF). The aluminum content (y) of the primary LaNi 5 -phase is controlled to produce the desired pressure-temperature behavior for adsorption and desorption of hydrogen. However, secondary phases cause decreased capacity and some may cause undesirable retention of tritium. Twenty-three alloys purchased from Ergenics, Inc. for development of RTF processes have been characterized by scanning electron microscopy (SEM) and by electron microprobe analysis (EMPA) to determine the distributions and compositions of constituent phases. This memorandum reports the results of these characterization studies. Knowledge of the structural characteristics of these alloys is a useful first step in selecting materials for specific process development tests and in interpreting results of those tests. Once this information is coupled with data on hydrogen plateau pressures, retention and capacity, secondary phase limits for RTF alloys can be specified

  11. Effect of water-cooling treatment times on properties of friction stir welded joints of 7N01-T4 aluminum alloy

    Science.gov (United States)

    Zhang, T. H.; Wang, Y.; Fang, X. F.; Liang, P.; Zhao, Y.; Li, Y. H.; Liu, X. M.

    2018-02-01

    Due to the deformation caused by residual stress in the welding process, welded components need treatment to reduce welding distortion. In this paper, several different times of flame-heating and water-cooling treatment were subjected to the friction stir welding joints of 15mm thick 7N01P-T4 aluminum alloy sheets to study the microstructure variation of friction stir welding joints of 7N01P-T4 aluminum alloy, and to analyze the effect on micro-hardness, tensile and fracture mechanical properties. This investigation will be helpful to optimize treatment methods and provide instruction on industrial production.

  12. Design Features of the SMART Water Chemistry

    International Nuclear Information System (INIS)

    Byung Seon Choi; Seong Hoon Kim; Juhyeon Yoon; Doo Jeong Lee; Yoon Yeong Bae; Sung Kyun Zee

    2004-01-01

    The design features for the primary water chemistry for the SMART are introduced from the viewpoint of the system characteristics and the chemical design concept. The most essential differences in water chemistry between the commercially operating PWRs and SMART are characterized by the presence of boron in the water and the operating mode of the purification system. SMART is a soluble boron free reactor, and the ammonia is used as a pH reagent. The material for SMART steam generator is also different from the standard material of the commercially operating PWRs: titanium alloy for the steam generator tubes. In SMART hydrogen gas which suppresses a generation of oxidizing species by the radiolysis processes in the reactors is not added to the primary coolant, but is normally generated from the radiolysis of the ammonia as the coolant passes through the core. Ammonia is added once per shift because SMART reactor has no letdown and charging system during power operation. Because of these competing processes, the concentrations of hydrogen, nitrogen and ammonia in the primary coolant are in equilibrium, which depend on the decomposition and/or combination rate of the ammonia. The level of permissible oxygen concentration in the primary coolant can be ensured by both suppression of the water radiolysis through maintaining a high enough hydrogen concentration in the primary coolant and by a restriction of the oxygen ingress into the primary coolant with the makeup water. The ammonia chemistry in SMART reactor eliminates the need for hydrogen injection for the control of the dissolved oxygen in the primary coolant because of spontaneous generation of hydrogen and nitrogen produced by the reaction of the ammonia decomposition. (authors)

  13. Quantifying the properties of low-cost powder metallurgy titanium alloys

    International Nuclear Information System (INIS)

    Bolzoni, L.; Ruiz-Navas, E.M.; Gordo, E.

    2017-01-01

    The extensive industrial employment of titanium is hindered by its high production costs where reduction of these costs can be achieved using cheap alloying elements and appropriate alternative processing techniques. In this work the feasibility of the production of low-cost titanium alloys is addressed by adding steel to pure titanium and processing the alloys by powder metallurgy. In particular, a spherical 4140 LCH steel powder commonly used in metal injection moulding is blended with irregular hydride-dehydride Ti. The new low-cost alloys are cold uniaxially pressed and sintered under high vacuum and show comparable properties to other wrought-equivalent and powder metallurgy titanium alloys. Differential thermal analysis and X-ray diffraction analyses confirm that Ti can tolerate the employment of iron as primary alloying element without forming detrimental TiFe-based intermetallic phases. Thus, the newly designed α+β alloys could be used for cheaper non-critical components.

  14. Quantifying the properties of low-cost powder metallurgy titanium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Bolzoni, L., E-mail: bolzoni.leandro@gmail.com [WaiCAM (Waikato Centre for Advanced Materials), The University of Waikato, Private Bag 3105, 3240 Hamilton (New Zealand); Ruiz-Navas, E.M.; Gordo, E. [Department of Materials Science and Engineering, University Carlos III of Madrid, Avda. de la Universidad, 30, 28911 Leganés, Madrid (Spain)

    2017-02-27

    The extensive industrial employment of titanium is hindered by its high production costs where reduction of these costs can be achieved using cheap alloying elements and appropriate alternative processing techniques. In this work the feasibility of the production of low-cost titanium alloys is addressed by adding steel to pure titanium and processing the alloys by powder metallurgy. In particular, a spherical 4140 LCH steel powder commonly used in metal injection moulding is blended with irregular hydride-dehydride Ti. The new low-cost alloys are cold uniaxially pressed and sintered under high vacuum and show comparable properties to other wrought-equivalent and powder metallurgy titanium alloys. Differential thermal analysis and X-ray diffraction analyses confirm that Ti can tolerate the employment of iron as primary alloying element without forming detrimental TiFe-based intermetallic phases. Thus, the newly designed α+β alloys could be used for cheaper non-critical components.

  15. Unexpected formation of hydrides in heavy rare earth containing magnesium alloys

    Directory of Open Access Journals (Sweden)

    Yuanding Huang

    2016-09-01

    Full Text Available Mg–RE (Dy, Gd, Y alloys show promising for being developed as biodegradable medical applications. It is found that the hydride REH2 could be formed on the surface of samples during their preparations with water cleaning. The amount of formed hydrides in Mg–RE alloys is affected by the content of RE and heat treatments. It increases with the increment of RE content. On the surface of the alloy with T4 treatment the amount of formed hydride REH2 is higher. In contrast, the amount of REH2 is lower on the surfaces of as-cast and T6-treated alloys. Their formation mechanism is attributed to the surface reaction of Mg–RE alloys with water. The part of RE in solid solution in Mg matrix plays an important role in influencing the formation of hydrides.

  16. Oxidation Behavior of Mo-Si-B Alloys in Wet Air; TOPICAL

    International Nuclear Information System (INIS)

    M. Kramer; A. Thom; O. Degirmen; V. Behrani; M. Akinc

    2002-01-01

    Multiphase composite alloys based on the Mo-Si-B system are candidate materials for ultra-high temperature applications. In non load-bearing uses such as thermal barrier coatings or heat exchangers in fossil fuel burners, these materials may be ideally suited. The present work investigated the effect of water vapor on the oxidation behavior of Mo-Si-B phase assemblages. Three alloys were studied: Alloy 1= Mo(sub 5)Si(sub 3)B(sub x) (T1)- MoSi(sub 2)- MoB, Alloy 2= T1- Mo(sub 5)SiB(sub 2) (T2)- Mo(sub 3)Si, and Alloy 3= Mo- T2- Mo(sub 3)Si. Tests were conducted at 1000 and 1100C in controlled atmospheres of dry air and wet air nominally containing 18, 55, and 150 Torr H(sub 2)O. The initial mass loss of each alloy was approximately independent of the test temperature and moisture content of the atmosphere. The magnitude of these initial losses varied according to the Mo content of the alloys. All alloys formed a continuous, external silica scale that protected against further mass change after volatilization of the initially formed MoO(sub 3). All alloys experienced a small steady state mass change, but the calculated rates cannot be quantitatively compared due to statistical uncertainty in the individual mass measurements. Of particular interest is that Alloy 3, which contains a significant volume fraction of Mo metal, formed a protective scale. All alloys formed varying amounts of subscale Mo and MoO(sub 2). This implies that oxygen transport through the external silica scale has been significantly reduced. For all alloys, water vapor accelerated the growth of a multiphase interlayer at the silica scale/unoxidized alloy interface. This interlayer is likely composed of fine Mo and MoO(sub 2) that is dispersed within a thin silica matrix. Alloy 3 was particularly sensitive to water accelerated growth of this interlayer. At 1100 C, the scale thickness after 300 hours increased from about 20 mm in dry air to nearly 100 mm in wet air

  17. SCC Initiation Testing of Alloy 600 in High Temperature Water

    Science.gov (United States)

    Etien, Robert A.; Richey, Edward; Morton, David S.; Eager, Julie

    Stress corrosion cracking (SCC) initiation tests have been conducted on Alloy 600 at temperatures from 304 to 367°C. Tests were conducted with in-situ monitored smooth tensile specimens under a constant load in hydrogenated environments. A reversing direct current electric potential drop (EPD) system was used for all of the tests to detect SCC initiation. Tests were conducted to examine the effects of stress (and strain), coolant hydrogen, and temperature on SCC initiation time. The thermal activation energy of SCC initiation was measured as 103 ± 18 kJ/mol in hydrogenated water, which is similar to the thermal activation energy for SCC growth. Results suggest that the fundamental mechanical parameter which controls SCC initiation is plastic strain not stress. SCC initiation was shown to have a different sensitivity than SCC growth to dissolved hydrogen level. Specifically, SCC initiation time appears to be relatively insensitive to hydrogen level in the nickel stability region.

  18. Processing and properties of Nb-Ti-based alloys

    International Nuclear Information System (INIS)

    Sikka, V.K.; Viswanathan, S.

    1992-01-01

    The processing characteristics, tensile properties, and oxidation response of two Nb-Ti-Al-Cr alloys were investigated. One creep test at 650 C and 172 MPa was conducted on the base alloy which contained 40Nb-40Ti-10Al-10Cr. A second alloy was modified with 0.11 at. % carbon and 0.07 at. % yttrium. Alloys were arc melted in a chamber backfilled with argon, drop cast into a water-cooled copper mold, and cold rolled to obtain a 0.8-mm sheet. The sheet was annealed at 1,100 C for 0.5 h. Longitudinal tensile specimens and oxidation specimens were obtained for both the base alloy and the modified alloy. Tensile properties were obtained for the base alloy at room temperature, 400, 600, 700, 800, 900, and 1,000 C, and for the modified alloy at room temperature, 400, 600, 700, and 800 C. Oxidation tests on the base alloy and modified alloy, as measured by weight change, were carried out at 600, 700, 800, and 900 C. Both the base alloy and the modified alloy were extremely ductile and were cold rolled to the final sheet thickness of 0.8 mm without an intermediate anneal. The modified alloy exhibited some edge cracking during cold during cold rolling. Both alloys recrystallized at the end of a 0.5-h annealing treatment. The alloys exhibited moderate strength and oxidation resistance below 600 C, similar to the results of alloys reported in the literature

  19. The Mechanical Properties of AlSi17Cu5 Cast Alloy after Overheating and Modification of CuP Master Alloy

    Directory of Open Access Journals (Sweden)

    Piątkowski J.

    2013-09-01

    Full Text Available The paper presents the results of studies on the effect of the AlSi17Cu5 alloy overheating to atemperature of 920°C and modification with phosphorus (CuP10 on the resultingmechanical (HB, Rm, R0.2 and plastic (A5 and Z properties. It has been shown that, so-called, "timethermal treatment" (TTT of an alloy in the liquid state, consisting inoverheating the metal to about 250°C above Tliq,holding at this temperature by 30 minutes improvesthe mechanical properties. It has also been found that overheating of alloy above Tliq.enhances the process of modification, resulting in the formation of fine-grain structure. The primary silicon crystals uniformly distributed in the eutectic and characteristics ofthe α(Al solution supersaturated with alloying elements present in the starting alloy composition (Cu, Fe provide not only an increase of strength at ambient temperature but also at elevated temperature (250°C.

  20. Environmentally-Assisted Cracking of Low-Alloy Reactor Pressure Vessel Steels under Boiling Water Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P.; Ritter, S

    2002-02-01

    The present report summarizes the experimental work performed by PSI on the environmentally-assisted cracking (EAC) of low-alloy steels (LAS) in the frame of the RIKORR-project during the period from January 2000 to August 2001. Within this project, the EAC crack growth behaviour of different low-alloy reactor pressure vessel (RPV) steels, weld filler and weld heat-affected zone materials is investigated under simulated transient and steady-state BWR/NWC power operation conditions. The EAC crack growth behaviour of different low-alloy RPV steels was characterized by slow rising load (SRL) / low-frequency corrosion fatigue (LFCF) and constant load tests with pre-cracked fracture mechanics specimens in oxygenated high-temperature water at temperatures of either 288, 250, 200 or 150 C. These tests revealed the following important interim results: Under low-flow and highly oxidizing (ECP >= 100 mV SHE) conditions, the ASME XI 'wet' reference fatigue crack growth curve could be significantly exceeded by cyclic fatigue loading at low frequencies (<0.001 Hz), at high and low load-ratios R, and by ripple loading near to DKth fatigue thresholds. The BWR VIP 60 SCC disposition lines may be significantly or slightly exceeded (even in steels with a low sulphur content) in the case of small load fluctuations at high load ratios (ripple loading) or at intermediate temperatures (200 -250 C) in RPV materials, which show a distinct susceptibility to dynamic strain ageing (DSA). (author)

  1. Ti-Mo alloys employed as biomaterials: effects of composition and aging heat treatment on microstructure and mechanical behavior.

    Science.gov (United States)

    Cardoso, Flavia F; Ferrandini, Peterson L; Lopes, Eder S N; Cremasco, Alessandra; Caram, Rubens

    2014-04-01

    The correlation between the composition, aging heat treatments, microstructural features and mechanical properties of β Ti alloys is of primary significance because it is the foundation for developing and improving new Ti alloys for orthopedic biomaterials. However, in the case of Ti-Mo alloys, this correlation is not fully described in the literature. Therefore, the purpose of this study was to experimentally investigate the effect of composition and aging heat treatments on the microstructure, Vickers hardness and elastic modulus of Ti-Mo alloys. These alloys were solution heat-treated and water-quenched, after which their response to aging heat treatments was investigated. Their microstructure, Vickers hardness and elastic modulus were evaluated, and the results allow us to conclude that stabilization of the β phase is achieved with nearly 10% Mo when a very high cooling rate is applied. Young's modulus was found to be more sensitive to phase variations than hardness. In all of the compositions, the highest hardness values were achieved by aging at 723K, which was attributed to the precipitation of α and ω phases. All of the compositions aged at 573K, 623K and 723K showed overaging within 80h. © 2013 Published by Elsevier Ltd.

  2. Process for the production of hydrogen from water

    Science.gov (United States)

    Miller, William E [Naperville, IL; Maroni, Victor A [Naperville, IL; Willit, James L [Batavia, IL

    2010-05-25

    A method and device for the production of hydrogen from water and electricity using an active metal alloy. The active metal alloy reacts with water producing hydrogen and a metal hydroxide. The metal hydroxide is consumed, restoring the active metal alloy, by applying a voltage between the active metal alloy and the metal hydroxide. As the process is sustainable, only water and electricity is required to sustain the reaction generating hydrogen.

  3. [Knowledge, attitude and practice on drinking water of primary and secondary students in Shenzhen].

    Science.gov (United States)

    Liu, Jiaxin; Hu, Xiaoqi; Zhang, Qian; Du, Songming; Pan, Hui; Dai, Xingbi; Ma, Guansheng

    2014-05-01

    To investigate the status on drinking water related knowledge, attitude and practice of primary and secondary students in Shenzhen. All 832 primary and secondary students from three schools in Shenzhen were selected by using multi-stage random sampling method. The information of drinking water related knowledge, time of drinking water and the type of drink chose in different situations were collected by questionnaires. 87.3% of students considered plain water being the healthiest drink in daily life, and the percent in girls (90.6%) was significantly higher than that in boys (84.4% ) (chi2 = 7.13, P = 0.0089). The awareness percent of the harm of dehydration was 84.5%. The percent in high school students (96.4%) was significantly higher than that in primary (73.9%) and middle school students (94.2%) (chi2 = 73.77, P water was in the morning with an empty stomach, and 46.3% chose when they felt thirsty. However, 63.7% drank water when they felt thirsty, and 50.6% drank water in the morning with an empty stomach. The percent of drinking plain water at school was the highest (83.4%), followed by at home (64.1%) and in public (26.2%). There were 45.2% and 53.3% of students, respectively, choosing sugary drinks as their favorite drink and most frequently drinking in public places. Primary and secondary students in Shenzhen have a good awareness of drinking water, which is inconsistent with their practice. Meanwhile, a considerable proportion of students towards choosing drinks have many misconceptions. The education of healthy drinking water should be strengthened.

  4. 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)

  5. 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)

  6. Silicon Alloying On Aluminium Based Alloy Surface

    International Nuclear Information System (INIS)

    Suryanto

    2002-01-01

    Silicon alloying on surface of aluminium based alloy was carried out using electron beam. This is performed in order to enhance tribological properties of the alloy. Silicon is considered most important alloying element in aluminium alloy, particularly for tribological components. Prior to silicon alloying. aluminium substrate were painted with binder and silicon powder and dried in a furnace. Silicon alloying were carried out in a vacuum chamber. The Silicon alloyed materials were assessed using some techniques. The results show that silicon alloying formed a composite metal-non metal system in which silicon particles are dispersed in the alloyed layer. Silicon content in the alloyed layer is about 40% while in other place is only 10.5 %. The hardness of layer changes significantly. The wear properties of the alloying alloys increase. Silicon surface alloying also reduced the coefficient of friction for sliding against a hardened steel counter face, which could otherwise be higher because of the strong adhesion of aluminium to steel. The hardness of the silicon surface alloyed material dropped when it underwent a heating cycle similar to the ion coating process. Hence, silicon alloying is not a suitable choice for use as an intermediate layer for duplex treatment

  7. Effect of hydrogen addition on the microstructure of TC21 alloy

    International Nuclear Information System (INIS)

    Zhu Tangkui; Li Miaoquan

    2010-01-01

    Research highlights: → The aim of this paper is to study the effect of hydrogen content (0-0.887 wt.%H) on microstructure, phase composition, microhardness and β transus temperature of TC21 alloy. The results show that, with increasing hydrogen content, the β phase increases, the α/β interfaces of lamellar transformed β phase disappear, the lattice parameter of β phase increases and the β transus temperature decreases for the hydrogenated TC21 alloy. In comparison to the as-received TC21 alloy, the contrasts of primary α phase and transformed β phase under optical microscope in the TC21 alloy with high hydrogen content are reversed completely. Furthermore, the γ and δ hydrides are detected in the hydrogenated TC21 alloy. In addition, the variations of phase compositions for the hydrogenated TC21 alloy have influence on microhardness and β transus temperature. → In conclusion, this paper shows some significant rules about the influence of hydrogen on TC21 alloy. - Abstract: TC21 alloy was hydrogenated at 750 deg. C with different hydrogen contents ranging from 0 to 0.873 wt.%H, and its microstructural evolution and phase transformations were investigated by optical microscopy (OM) and X-ray diffraction (XRD). The microhardness and the β transus temperature for the hydrogenated TC21 alloy were determined by microhardness testing and metallographical approach, respectively. The results show that, hydrogen addition has a noticeable influence on microstructure, phase composition, microhardness and β transus temperature of TC21 alloy. With increasing hydrogen content, the β phase increases, the α/β interfaces of lamellar transformed β phase disappear, the lattice parameter of β phase increases and the β transus temperature decreases for the hydrogenated TC21 alloy. In comparison to the as-received TC21 alloy, the contrasts of primary α phase and transformed β phase under optical microscope in the hydrogenated TC21 alloy with high hydrogen

  8. Four decades of working experience of Cirus primary cooling water heat exchangers

    International Nuclear Information System (INIS)

    Dubey, P.K.; Ullas, O.P.; Rao, D.V.H.; Zope, A.K.; Kharpate, A.V.

    2006-01-01

    CIRUS is a 40 MW (Th.) research reactor, commissioned in the year 1960. The reactor has natural uranium fuel rods, heavy water as moderator, demineralised water (DM water) as primary coolant, and seawater as secondary coolant. There are six Heat Exchangers in the primary cooling water (PCW) system. Five of them are required for the normal operation of the reactor and one is kept stand by. DM water flows on the shell side of the heat exchanger in two passes. Seawater is used as coolant on the tube side of the heat exchangers in four passes. Cirus has been in operation for around 41 years excluding refurbishment period. During these four decades of reactor operation, PCW heat exchangers have experienced many failures and undergone many modifications in the circuit for ensuring better performance. This paper tries to capture the essence of working experiences with PCW heat exchangers, various problems faced, remedial measures taken during those four decades of reactor operation. (author)

  9. Parameters of straining-induced corrosion cracking in low-alloy steels in high temperature water

    International Nuclear Information System (INIS)

    Lenz, E.; Liebert, A.; Stellwag, B.; Wieling, N.

    Tensile tests with slow deformation speed determine parameters of corrosion cracking at low strain rates of low-alloy steels in high-temperature water. Besides the strain rate the temperature and oxygen content of the water prove to be important for the deformation behaviour of the investigated steels 17MnMoV64, 20 MnMoNi55 and 15NiCuMoNb 5. Temperatures about 240 0 C, increased oxygen contents in the water and low strain rates cause a decrease of the material ductility as against the behaviour in air. Tests on the number of stress cycles until incipient cracking show that the parameters important for corrosion cracking at low strain velocities apply also to low-frequency cyclic loads with high strain amplitude. In knowledge of these influencing parameters the strain-induced corrosion cracking is counteracted by concerted measures taken in design, construction and operation of nuclear power stations. Essential aims in this matter are to avoid as far as possible inelastic strains and to fix and control suitable media conditions. (orig.) [de

  10. Oxidation behaviour of Zr-Ce alloys. Kinetic and microstructure aspects

    International Nuclear Information System (INIS)

    Rouillon, Ludovic

    1996-01-01

    As Zircaloy alloys are used for fuel rods in pressurized water nuclear reactors, this research thesis aims at studying and improving corrosion resistance of zirconium alloys while maintaining their mechanical properties. It more precisely deals with the kinetic and microstructure aspects of the external corrosion of the cladding by the coolant. In the case of Zircaloys, this corrosion is characterized by a kinetic transition from an initially parabolic to a linear regime. This research aims at intervening on this transition by elaborating zirconium alloys containing an element which stabilizes zirconia, in this case cerium. After having reported a bibliographical study on sheath oxidation, on parameters which influence sheath oxidation kinetics, on zirconia stabilization by doping elements, on the interest of lanthanide oxides, the author reports a feasibility study on the use of cerium (choice and preparation, sintered ceramic characterization, annealing of stabilized zirconia), reports a metallurgical study of Zr-Ce alloys, reports the study of the oxidation behaviour of these alloys (in autoclave, in presence of oxygen, under oxygen and then water) and the characterization of the microstructures of the oxide layers. He finally discusses the relationship between microstructure and oxidation kinetics, the role of cerium in the oxidation process, and the role of water in the oxidation process [fr

  11. Requirements on cast steel for the primary coolant circuit of water cooled reactors

    International Nuclear Information System (INIS)

    The most important requirements placed on the structural components of water cooled nuclear reactors include corrosion resistance and mechanical materials properties. Intercrystalline corrosion resistance was tested using the Strauss Test in compliance with the DIN 50914 Standard. Following sensitization between 600 to 700 degC with a dwell time between 15 minutes and 100 hours, a specimen homogeneously annealed with the casting and rapidly water cooled showed no intercrystalline corrosion. Specimens cooled from 1050 degC at a rate of 100 degC per hour showed no unambiguous tendency for intercrystalline corrosion after sensitization; in some cases, however, an initial attack of intercrystalline corrosion was found. It was found that austenitic Cr-Ni cast steel containing 2.5% Mo and about 15% ferrite showed the sensitive intercrystalline corrosion range at higher temperatures and longer dwell times than rolled Cr-Ni steels. In plating the ferritic cast steel with a corrosion resistant plating material, annealing temperature after welding must not exceed 600 to 620 degC otherwise the resistance of the plated layer against intercrystalline corrosion would not be safeguarded, and following annealing for stress removal at a temperature of 600 to 620 degC all requirements must be satisfied by the weld metal and weld transition placed on the initial material. Martensite materials are used for the manufacture of components which are not used under pressure, such as alloys with 13% Cr and 1% to 6% Ni and alloys with 17% Cr and 4% Ni. Carbon content is maintained below 0.10% to guarantee good weldability and the highest corrosion resistance. Cast steels with 13% Cr and 4% Ni after a dwell of 2500 hours in fully desalinated water without oxygen and with 3600 ppm of boron at a test temperature of 95 to 300 degC showed a surface reduction of 0.005 mm annually. In identical conditions except for the water containing oxygen the reduction in surface was 0.05 mm per year. (J.B.)

  12. Recent developments in advanced aircraft aluminium alloys

    International Nuclear Information System (INIS)

    Dursun, Tolga; Soutis, Costas

    2014-01-01

    Highlights: • To compete with composites, performance of aluminium alloys should be increased. • Al–Li alloys have higher strength, fracture and fatigue/corrosion resistance. • Improvements of aerospace Al alloys are due to optimised solute content and ratios. • In selecting new materials, there should be no reduction in the level of safety. • The use of hybrid materials could provide additional opportunities for Al alloys. - Abstract: Aluminium alloys have been the primary material for the structural parts of aircraft for more than 80 years because of their well known performance, well established design methods, manufacturing and reliable inspection techniques. Nearly for a decade composites have started to be used more widely in large commercial jet airliners for the fuselage, wing as well as other structural components in place of aluminium alloys due their high specific properties, reduced weight, fatigue performance and corrosion resistance. Although the increased use of composite materials reduced the role of aluminium up to some extent, high strength aluminium alloys remain important in airframe construction. Aluminium is a relatively low cost, light weight metal that can be heat treated and loaded to relatively high level of stresses, and it is one of the most easily produced of the high performance materials, which results in lower manufacturing and maintenance costs. There have been important recent advances in aluminium aircraft alloys that can effectively compete with modern composite materials. This study covers latest developments in enhanced mechanical properties of aluminium alloys, and high performance joining techniques. The mechanical properties on newly developed 2000, 7000 series aluminium alloys and new generation Al–Li alloys are compared with the traditional aluminium alloys. The advantages and disadvantages of the joining methods, laser beam welding and friction stir welding, are also discussed

  13. Development of Pb-Free Nanocomposite Solder Alloys

    Directory of Open Access Journals (Sweden)

    Animesh K. Basak

    2018-04-01

    Full Text Available As an alternative to conventional Pb-containing solder material, Sn–Ag–Cu (SAC based alloys are at the forefront despite limitations associated with relatively poor strength and coarsening of grains/intermetallic compounds (IMCs during aging/reflow. Accordingly, this study examines the improvement of properties of SAC alloys by incorporating nanoparticles in it. Two different types of nanoparticles were added in monolithic SAC alloy: (1 Al2O3 or (2 Fe and their effect on microstructure and thermal properties were investigated. Addition of Fe nanoparticles leads to the formation of FeSn2 IMCs alongside Ag3Sn and Cu6Sn5 from monolithic SAC alloy. Addition of Al2O3 nano-particles do not contribute to phase formation, however, remains dispersed along primary β-Sn grain boundaries and act as a grain refiner. As the addition of either Fe or Al2O3 nano-particles do not make any significant effect on thermal behavior, these reinforced nanocomposites are foreseen to provide better mechanical characteristics with respect to conventional monolithic SAC solder alloys.

  14. Preparation of high-strength Al-Mg-Si-Cu-Fe alloy via heat treatment and rolling

    Science.gov (United States)

    Liu, Chong-yu; Yu, Peng-fei; Wang, Xiao-ying; Ma, Ming-zhen; Liu, Ri-ping

    2014-07-01

    An Al-Mg-Si-Cu-Fe alloy was solid-solution treated at 560°C for 3 h and then cooled by water quenching or furnace cooling. The alloy samples which underwent cooling by these two methods were rolled at different temperatures. The microstructure and mechanical properties of the rolled alloys were investigated by optical microscopy, scanning electron microscopy, transmission electron microscopy, X-ray diffraction analysis, and tensile testing. For the water-quenched alloys, the peak tensile strength and elongation occurred at a rolling temperature of 180°C. For the furnace-cooled alloys, the tensile strength decreased initially, until the rolling temperature of 420°C, and then increased; the elongation increased consistently with increasing rolling temperature. The effects of grain boundary hardening and dislocation hardening on the mechanical properties of these rolled alloys decreased with increases in rolling temperature. The mechanical properties of the 180°C rolling water-quenched alloy were also improved by the presence of β″ phase. Above 420°C, the effect of solid-solution hardening on the mechanical properties of the rolled alloys increased with increases in rolling temperature.

  15. Perceptions of the Water Cycle among Primary School Children in Botswana.

    Science.gov (United States)

    Taiwo, A. A.; Motswiri, M. J.; Masene, R.

    1999-01-01

    Describes qualitative and quantitative methods used to elucidate the nature of the perception of the water cycle held by Botswana primary-grade pupils in three different geographic areas. Concludes that the students' perception of the water cycle was positively influenced by schooling but negatively impacted upon, to some extent, by the untutored…

  16. Electrochemical corrosion behavior of AZ91D alloy in ethylene glycol

    Energy Technology Data Exchange (ETDEWEB)

    Fekry, A.M. [Chemistry Department, Faculty of Science, Cairo University, Giza 12613 (Egypt)], E-mail: hham4@hotmail.com; Fatayerji, M.Z. [Chemistry Department, Faculty of Science, Cairo University, Giza 12613 (Egypt)

    2009-11-01

    The effect of concentration on the corrosion behavior of Mg-based alloy AZ91D was investigated in ethylene glycol-water solutions using electrochemical techniques i.e. potentiodynamic polarization, electrochemical impedance measurements (EIS) and surface examination via scanning electron microscope (SEM) technique. This can provide a basis for developing new coolants for magnesium alloy engine blocks. Corrosion behavior of AZ91D alloy by coolant is important in the automotive industry. It was found that the corrosion rate of AZ91D alloy decreased with increasing concentration of ethylene glycol. For AZ91D alloy in chloride >0.05 M or fluoride <0.05 M containing 30% ethylene glycol solution, they are more corrosive than the blank (30% ethylene glycol-70% water). However, at concentrations <0.05 for chloride or >0.05 M for fluoride containing ethylene glycol solution, some inhibition effect has been observed. The corrosion of AZ91D alloy in the blank can be effectively inhibited by addition of 0.05 mM paracetamol that reacts with AZ91D alloy and forms a protective film on the surface at this concentration as confirmed by surface examination.

  17. High temperature cathodic charging of hydrogen in zirconium alloys and iron and nickel base alloys

    International Nuclear Information System (INIS)

    John, J.T.; De, P.K.; Gadiyar, H.S.

    1990-01-01

    These investigations lead to the development of a new technique for charging hydrogen into metals and alloys. In this technique a mixture of sulfates and bisulfates of sodium and potassium is kept saturated with water at 250-300degC in an open pyrex glass beaker and electrolysed using platinum anode and the material to be charged as the cathode. Most of the studies were carried out on Zr alloys. It is shown that because of the high hydrogen flux available at the surface and the high diffusivity of hydrogen in metals at these temperatures the materials pick up hydrogen faster and more uniformly than the conventional electrolytic charging at room temperature and high temperature autoclaving in LiOH solutions. Chemical analysis, metallographic examination and XRD studies confirm this. This technique has been used to charge hydrogen into many iron and nickel base austentic alloys, which are very resistant to hydrogen pick up and to H-embrittlement. Since this involved a novel method of electrolysing water, the hydrogen/deuterium isotopic ratio has been studied. At this temperatures the D/H ratio in the evolved hydrogen gas was found to be closer to the value in the liquid water, which means a smaller separation factor. This confirm the earlier observation that separation factor decreases with increase of temperature. (author). 16 refs., 21 fi gs., 6 tabs

  18. PWSCC Mitigation of alloy 182: Testing of various mitigation processes

    International Nuclear Information System (INIS)

    Curieres, I. de; Calonne, O.; Crooker, P.

    2011-01-01

    Since the mid nineties, Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 182 welds has occurred. This affects different components, even ones that are considered to have 'low-susceptibility' due to a low operating temperature such as the 'low operating temperature' reactor pressure vessel (RPV) heads in the global PWR fleet and bottom-mounted instrumentation nozzles, a location where currently there is no ready-to-deploy repair or replacement solution. Hence, there is an incentive to identify effective remedial measures to delay or prevent PWSCC initiation, even at 'low temperature' RPV heads in order to avoid wholesale replacement in the future. Working with EPRI, Areva has assessed the efficiency of various technological processes including brushing, polishing or compressive stress methods to mitigate PWSCC in Alloy 182. A first phase of the program is completed and the results will be presented. The emphasis will be put on the program's different testing phases and the different mitigation processes that were tested. Efficiency of 'chemical' surface treatments is not yet proved. EPRI stabilized chromium had a deleterious effect on crack initiation that should be reproduced and understood before drawing a definitive conclusion. The electropolishing process considered does not seem to be sufficiently reliable on Alloy 182 surfaces but longer exposures are required for a more definitive evaluation of this treatment. All tested 'mechanical' surface treatments i.e. -) GE-RENEW brushing, -) Fiber laser peening (Toshiba), -) Water Jet Peening (Mitsubishi), -) Water Jet Peening (Hitachi), -) Combination of GE-RENEW and Hitachi WJP have successfully inhibited crack initiation even though the surface compressive stresses induced on U-ends are lower than those expected on massive components. Past experience shows that crack initiation occurs in less than 250 h on U-bends with 'heavily ground' reference surfaces. Thus, it can be deduced that the present results show

  19. Molten aluminum alloy fuel fragmentation experiments

    International Nuclear Information System (INIS)

    Gabor, J.D.; Purviance, R.T.; Cassulo, J.C.; Spencer, B.W.

    1992-01-01

    Experiments were conducted in which molten aluminum alloys were injected into a 1.2 m deep pool of water. The parameters varied were (i) injectant material (8001 aluminum alloy and 12.3 wt% U-87.7 wt% Al), (ii) melt superheat (O to 50 K), (iii) water temperature (313, 343 and 373 K) and (iv) size and geometry of the pour stream (5, 10 and 20 mm diameter circular and 57 mm annular). The pour stream fragmentation was dominated by surface tension with large particles (∼30 mm) being formed from varicose wave breakup of the 10-mm circular pours and from the annular flow off a 57 mm diameter tube. The fragments produced by the 5 mm circular et were smaller (∼ mm), and the 20 mm jet which underwent sinuous wave breakup produced ∼100 mm fragments. The fragments froze to form solid particles in 313 K water, and when the water was ≥343 K, the melt fragments did not freeze during their transit through 1.2 m of water

  20. The evaluation of the use of metal alloy fuels in pressurized water reactors

    International Nuclear Information System (INIS)

    Lancaster, D.

    1992-01-01

    The use of metal alloy fuels in a PWR was investigated. It was found that it would be feasible and competitive to design PWRs with metal alloy fuels but that there seemed to be no significant benefits. The new technology would carry with it added economic uncertainty and since no large benefits were found it was determined that metal alloy fuels are not recommended. Initially, a benefit was found for metal alloy fuels but when the oxide core was equally optimized the benefit faded. On review of the optimization of the current generation of ''advanced reactors,'' it became clear that reactor design optimization has been under emphasized. Current ''advanced reactors'' are severely constrained. The AP-600 required the use of a fuel design from the 1970's. In order to find the best metal alloy fuel design, core optimization became a central effort. This work is ongoing

  1. Effect of Ti3+ ion on the Corrosion Behavior of Alloy 600

    International Nuclear Information System (INIS)

    Lee, Chang Bong; Lim, Han Gwi; Kim, Bok Hee; Kim, Ki Ju

    1999-01-01

    Alloy 600 has been widely used as a steam generator tubing material in pressurized water reactors(PWRs) nuclear power plants. Corrosion of steam generator tubing mainly occurs on the secondary water side. The purpose of this work is primarily concerned with examining the effect of Ti 3+ ion concentrations on the corrosion behavior of the Alloy 600 steam generator tubing material. Corrosion behavior of the Alloy 600 steam generator tubing material was studied in aqueous solutions with varying Ti 3+ ion concentration at room temperature. Potentiodynamic and potentiostatic polarization techniques were used to determine the corrosion and pitting potentials for the Alloy 600 test material. The addition of Ti 3+ ion to 1000ppm, showed inhibition effect on the corrosion of Alloy 600. But the corrosion of Alloy 600 was accelerated when the concentration of Ti 3+ ion exceeded 1000ppm, it is assumed that the effect of general corrosion of Alloy 600 is more sensitive than pitting corrosion. It is considered that the passive film which was formed on the Alloy 600 surface in the 100ppm Ti 3+ ion containing solution is mainly consisted of TiO 2

  2. Impact of load follow operation on the chemistry of the primary and secondary circuit of a pressurized water reactor

    International Nuclear Information System (INIS)

    Boettcher, F.; Riehm, S.; Bolz, M.; Speck, A.

    2012-09-01

    Germany decided to abandon nuclear energy and to switch to renewable energy forms. According the renewable energy act renewable energy forms have priority to be fed to the grid. The support of wind and solar energy demands more and more load follow operation of the remaining nuclear power plants to stabilize the grid. This report summarizes first experience with load follow operation in two pressurized water reactors (Philippsburg KKP2 and Neckarwestheim GKNI) with regard to chemistry and radiology. The most important mechanisms of dose rate built up on the primary side are described with Co-60 and Co-58 being the main contributors to dose rate. Goal of the primary side chemistry is to avoid or at least to delay the dose rate built-up as far as achievable. Both reactors are operated according to the modified coordinated B-Li-Chemistry with a pH300 of 7.4 as target value for optimised dose build up delay. By using B-10-enriched boric acid with a boron-10 abundance of 30 at-% (compared to ca. 19.9 at-% in natural boron) the pH 300 target value can be reached earlier in the cycle due to the lower concentration of boric acid required for neutron balancing. In GKNI Zn-injection was started 2005 as a mean of dose reduction. Since 2007 GKNI was operated with load follow operation. In KKP2 load reductions due to wind energy excess are more and more common since 2008. The results of dose rate measurements on the primary side are correlated to primary coolant chemistry and load follow operation. The use of enriched boric acid had a positive (i.e. reducing dose rate) impact on the activity build-up of Co-60 on the loop lines, thus proving the effectiveness of the VGB specifications. After 5 years (one half life time of Co-60) of Zn-injection a positive effect on surface occupancy with nuclides can be determined. The impact of short term deviations from optimal chemical conditions during load follow operation on the activity build up is assessed on the basis of the corrosion

  3. Centrifugally cast Zn-27Al-xMg-ySi alloys and their in situ (Mg2Si + Si)/ZA27 composites

    International Nuclear Information System (INIS)

    Wang Qudong; Chen Yongjun; Chen Wenzhou; Wei Yinhong; Zhai Chunquan; Ding Wenjiang

    2005-01-01

    Effects of composition, mold temperature, rotating rate and modification on microstructure of centrifugally cast Zn-27Al-xMg-ySi alloys have been investigated. In situ composites of Zn-27Al-6.3Mg-3.7Si and Zn-27Al-9.8Mg-5.2Si alloys were fabricated by centrifugal casting using heated permanent mold. These composites consist of three layers: inner layer segregates lots of blocky primary Mg 2 Si and a litter blocky primary Si, middle layer contains without primary Mg 2 Si and primary Si, outer layer contains primary Mg 2 Si and primary Si. The position, quantity and distribution of primary Mg 2 Si and primary Si in the composites are determined jointly by alloy composition, solidification velocity under the effect of centrifugal force and their floating velocity inward. Na salt modifier can refine grain and primary Mg 2 Si and make primary Mg 2 Si distribute more evenly and make primary Si nodular. For centrifugally cast Zn-27Al-3.2Mg-1.8Si alloy, the microstructures of inner layer, middle layer and outer layer are almost similar, single layer materials without primary Mg 2 Si and primary Si are obtained, and their grain sizes increased with the mold temperature increasing

  4. Divorced Eutectic Solidification of Mg-Al Alloys

    Science.gov (United States)

    Monas, Alexander; Shchyglo, Oleg; Kim, Se-Jong; Yim, Chang Dong; Höche, Daniel; Steinbach, Ingo

    2015-08-01

    We present simulations of the nucleation and equiaxed dendritic growth of the primary hexagonal close-packed -Mg phase followed by the nucleation of the -phase in interdendritic regions. A zoomed-in region of a melt channel under eutectic conditions is investigated and compared with experiments. The presented simulations allow prediction of the final properties of an alloy based on process parameters. The obtained results give insight into the solidification processes governing the microstructure formation of Mg-Al alloys, allowing their targeted design for different applications.

  5. 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.

  6. 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.

  7. Study on Microstructure and Mechanical Properties of Hypereutectic Al-18Si Alloy Modified with Al-3B.

    Science.gov (United States)

    Gong, Chunjie; Tu, Hao; Wu, Changjun; Wang, Jianhua; Su, Xuping

    2018-03-20

    An hypereutectic Al-18Si alloy was modified via an Al-3B master alloy. The effect of the added Al-3B and the modification temperature on the microstructure, tensile fracture morphologies, and mechanical properties of the alloy were investigated using an optical microscope, Image-Pro Plus 6.0, a scanning electron microscope, and a universal testing machine. The results show that the size of the primary Si and its fraction decreased at first, and then increased as an additional amount of Al-3B was added. When the added Al-3B reached 0.2 wt %, the fraction of the primary Si in the Al-18Si alloy decreased with an increase in temperature. Compared with the unmodified Al-18Si alloy, the tensile strength and elongation of the alloy modified at 850 °C with 0.2 wt % Al-3B increased by 25% and 81%, respectively. The tensile fracture of the modified Al-18Si alloy exhibited partial ductile fracture characteristics, but there were more areas with ductile characteristics compared with that of the unmodified Al-18Si alloy.

  8. Characterization of electrochemical and passive behaviour of Alloy 59 in acid solution

    International Nuclear Information System (INIS)

    Luo, Hong; Gao, Shujun; Dong, Chaofang; Li, Xiaogang

    2014-01-01

    Highlights: • A considerably thinner n-type passive film is observed on the Alloy-59. • The passive film formed in air was thicker than that formed in acid solution. • Primary constituents of passive film in air and acid solution are (Cr, Ni)-oxides and (Cr, Ni) hydroxides, respectively. - Abstract: The electrochemical behaviour and passive film properties of the Alloy 59 in sulfuric acid solution was evaluated by the potentiodynamic electrochemical measurements, electrochemical impedance spectroscopy, Mott-Schottky approach, and ex situ surface analytical technique as X-ray photoelectron spectroscopy (XPS) and Auger Electronic Spectrometer (AES). The results confirmed that the Alloy 59 exhibits well passive behaviour. A considerably thinner n-type passive film is observed on this type alloy. Based on the evaluations of surface composition analysis, the primary constituents of passive film formed in the air and acid solution are different, with the (Cr, Ni)-oxides and (Cr, Ni) hydroxides, respectively

  9. The application of in situ analytical transmission electron microscopy to the study of preferential intergranular oxidation in Alloy 600

    Energy Technology Data Exchange (ETDEWEB)

    Burke, M.G., E-mail: m.g.burke@manchester.ac.uk; Bertali, G.; Prestat, E.; Scenini, F.; Haigh, S.J.

    2017-05-15

    In situ analytical transmission electron microscopy (TEM) can provide a unique perspective on dynamic reactions in a variety of environments, including liquids and gases. In this study, in situ analytical TEM techniques have been applied to examine the localised oxidation reactions that occur in a Ni-Cr-Fe alloy, Alloy 600, using a gas environmental cell at elevated temperatures. The initial stages of preferential intergranular oxidation, shown to be an important precursor phenomenon for intergranular stress corrosion cracking in pressurized water reactors (PWRs), have been successfully identified using the in situ approach. Furthermore, the detailed observations correspond to the ex situ results obtained from bulk specimens tested in hydrogenated steam and in high temperature PWR primary water. The excellent agreement between the in situ and ex situ oxidation studies demonstrates that this approach can be used to investigate the initial stages of preferential intergranular oxidation relevant to nuclear power systems. - Highlights: • In situ analytical TEM has been performed in 1 bar H{sub 2}-H{sub 2}O vapor at 360–480 °C. • Nanoscale GB migration and solute partitioning correlate with ex situ data for Alloy 600 in H{sub 2}-steam. • This technique can provide new insights into localised reactions associated with localised oxidation.

  10. Corrosion Inhibition Study of Al-Cu-Ni Alloy in Simulated Sea-Water ...

    African Journals Online (AJOL)

    Akorede

    ABSTRACT: A study on the inhibition of Al-Cu-Ni alloy in simulated ... which the percentage of Copper, and Nickel were kept .... proceed based on equation of reaction in eqn (4). Al .... Sodium-Modified A356.0-Type Al-Si-Mg Alloy in Simulated.

  11. The mode of stress corrosion cracking in Ni-base alloys in high temperature water containing lead

    International Nuclear Information System (INIS)

    Hwang, S.S.; Kim, H.P.; Lee, D.H.; Kim, U.C.; Kim, J.S.

    1999-01-01

    The mode of stress corrosion cracking (SCC) in Ni-base alloys in high temperature aqueous solutions containing lead was studied using C-rings and slow strain rate testing (SSRT). The lead concentration, pH and the heat treatment condition of the materials were varied. TEM work was carried out to observe the dislocation behavior in thermally treated (TT) and mill annealed (MA) materials. As a result of the C-ring test in 1M NaOH+5000 ppm lead solution, intergranular stress corrosion cracking (IGSCC) was found in Alloy 600MA, whereas transgranular stress corrosion cracking (TGSCC) was found in Alloy 600TT and Alloy 690TT. In most solutions used, the SCC resistance increased in the sequence Alloy 600MA, Alloy 600TT and Alloy 690TT. The number of cracks that was observed in alloy 690TT was less than in Alloy 600TT. However, the maximum crack length in Alloy 690TT was much longer than in Alloy 600TT. As a result of the SSRT, at a nominal strain rate of 1 x 10 -7 /s, it was found that 100 ppm lead accelerated the SCC in Alloy 600MA (0.01%C) in pH 10 at 340 C. IGSCC was found in a 100 ppm lead condition, and some TGSCC was detected on the fracture surface of Alloy 600MA cracked in the 10000 ppm lead solution. The mode of cracking for Alloy 600 and Alloy 690 changed from IGSCC to TGSCC with increasing grain boundary carbide content in the material and lead concentration in the solution. IGSCC seemed to be retarded by stress relaxation around the grain boundaries, and TGSCC in the TT materials seemed to be a result of the crack blunting at grain boundary carbides and the enhanced Ni dissolution with an increase of the lead concentration. (orig.)

  12. Cellular microstructure of chill block melt spun Ni-Mo alloys

    Science.gov (United States)

    Tewari, S. N.; Glasgow, T. K.

    1987-01-01

    Chill block melt spun ribbons of Ni-Mo binary alloys containing 8.0 to 41.8 wt pct Mo have been prepared under carefully controlled processing conditions. The growth velocity has been determined as a function of distance from the quench surface from the observed ribbon thickness dependence on the melt puddle residence time. Primary arm spacings measured at the midribbon thickness locations show a dependence on growth velocity and alloy composition which is expected from dendritic growth models for binary alloys directionally solidified in a positive temperature gradient. Microsegregation across cells and its variation with distance from the quench surface and alloy composition have been examined and compared with theoretical predictions.

  13. Assessment of the Polyacrylic Acid for an Ammonia Water Treatment and for Alloy 800NG SG Tube Material in Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Lamouroux, Christine; You, Dominique; Plancque, Gabriel; Roy, Marc; Laire, Charles; Schnongs, Philippe

    2012-09-01

    To prevent the Steam Generators (SG) fouling by corrosion products or the Tube Support Plate (TSP) blockage the on-line injection of a dispersant such the Polyacrylic Acid (PAA) could be a relevant water treatment. Long-term trials performed in PWRs have shown that the PAA, injected at the SG inlet, facilitate the evacuation of the iron oxides by the SG blowdown. Given the ammonia treatment of the secondary water of the Belgian PWRs, the R and D program carried out was devoted to: - Verify the innocuousness of the PAA and its degradation products versus Alloy 800NG SCC susceptibility in case of over concentrations and sludge presence, - Assess the potential impact of the PAA and its thermal degradation products on the specific NH 3 water treatment. The main results can be summarized as following: The corrosion tests performed with PAA in case of over concentrations and sludge couldn't point out any negative effect of the dispersant on the SCC susceptibility of tubing materials such as Alloy 800NG. No significant modification of the tube oxide layer has been observed. At the SG operating temperature, the PAA is decomposed and a large spectrum from high to lower molecular weights polymers than the initial PAA arises. The fragmentation of the polymer into low molecular weight polyacrylic acids is obtained within 20 minutes and the average molecular weight is reduced by 50% from the original one. The thermal degradation products, their quantity and their kinetic of appearance, have been determined. The generated acetate concentration during the on-line dispersant application should remain low compared to the current values observed in the SG water. From the numerical simulation based on acetate concentration and on the kinetic law deduced from the experimental work, it can be concluded that in a 2-phase medium, the margin on the water pH compared to the neutral pH remains high. At 180 deg. C, no impact on the water pH is identified, taking into account realistic

  14. Microstructure and Properties of Ti-5553 Alloy for Aerospace Fasteners

    Directory of Open Access Journals (Sweden)

    ZHAO Qing-yun

    2017-10-01

    Full Text Available The effect of heat treatment on microstructure and mechanical properties of Ti-5553 alloy was investigated by scanning electron microscopy (SEM and transmission electron microscopy (TEM. The results show that when the alloy is treated in α+β phase zone, tensile strength decreases with raising solution temperature due to decreasing the content of primary α-phase and increasing the size and volume fraction of β phase. A lot of secondary α-phase precipitates from grain boundary and intragranular with β phase transformation during aging treatment. The size of secondary α-phase has significant influence on tensile strength, secondary α-phase coarsens gradually with the increase of aging temperature, resulting in the decrease of tensile strength. It is suggested that for 1240MPa aerospace fasteners the solution temperature of Ti-5553 should be under Tβ, thus adequate β phase, where a lot of secondary α phase precipitates from, is good for the required high strength. Meanwhile, a certain percentage of primary α-phase is kept for acquiring good ductility and toughness. After solution treatment at 810-820℃ for 1.5h, water quenching plus aging at 510℃ for 10h, Ti-5553 shows a better mechanical property with tensile strength 1500MPa, elongation 14.8% and reduction of cross-section area 38.6%. Lots of dimples can be found in tensile fracture after solution treatment and solution+aging treatment, which demonstrate Ti-5553 with good ductility and toughness.

  15. Seacoast stress corrosion cracking of aluminum alloys

    Science.gov (United States)

    Humphries, T. S.; Nelson, E. E.

    1981-01-01

    The stress corrosion cracking resistance of high strength, wrought aluminum alloys in a seacoast atmosphere was investigated and the results were compared with those obtained in laboratory tests. Round tensile specimens taken from the short transverse grain direction of aluminum plate and stressed up to 100 percent of their yield strengths were exposed to the seacoast and to alternate immersion in salt water and synthetic seawater. Maximum exposure periods of one year at the seacoast, 0.3 or 0.7 of a month for alternate immersion in salt water, and three months for synthetic seawater were indicated for aluminum alloys to avoid false indications of stress corrosion cracking failure resulting from pitting. Correlation of the results was very good among the three test media using the selected exposure periods. It is concluded that either of the laboratory test media is suitable for evaluating the stress corrosion cracking performance of aluminum alloys in seacoast atmosphere.

  16. Absorption of dissolved hydrogen from lithiated water during accelerated corrosion of zirconium-2.5 wt% niobium alloy

    International Nuclear Information System (INIS)

    Manolescu, A.V.; Mayer, P.; Rasile, E.M.; Mummenhoff, J.W.

    1982-01-01

    A series of laboratory experiments was carried out to determine the extent of dissolved hydrogen absorption from lithiated water by zirconium-2.5 wt% niobium alloy during corrosion. The material was exposed at 340 0 C to 1 M LiOH aqueous solution containing 0 to approximately 70 cm 3 /L of dissolved hydrogen. Results indicate that dissolved hydrogen has no effect on the corrosion rate or on the amount of hydrogen absorbed by the material

  17. Predictive calculation of phase formation in Al-rich Al-Zn-Mg-Cu-Sc-Zr alloys using a thermodynamic Mg-alloy database

    International Nuclear Information System (INIS)

    Groebner, J.; Rokhlin, L.L.; Dobatkina, T.V.; Schmid-Fetzer, R.

    2007-01-01

    Three series of Al-rich alloys in the system Al-Zn-Mg-Cu-Sc-Zr and the subsystems Al-Zn-Mg-Cu-Sc and Al-Zn-Mg-Sc were studied by thermodynamic calculations. Phase formation was compared with experimental data obtained by DTA and microstructural analysis. Calculated phase diagrams, phase amount charts and enthalpy charts together with non-equilibrium calculations under Scheil conditions reveal significant details of the complex phase formation. This enables consistent and correct interpretation of thermal analysis data. Especially the interpretation of liquidus temperature and primary phase is prone to be wrong without using this tool of computational thermodynamics. All data are predictions from a thermodynamic database developed for Mg-alloys and not a specialized Al-alloy database. That provides support for a reasonable application of this database for advanced Mg-alloys beyond the conventional composition ranges

  18. Predictive calculation of phase formation in Al-rich Al-Zn-Mg-Cu-Sc-Zr alloys using a thermodynamic Mg-alloy database

    Energy Technology Data Exchange (ETDEWEB)

    Groebner, J. [Institute of Metallurgy, Clausthal University of Technology, Robert-Koch Strasse 42, D-38678 Clausthal-Zellerfeld (Germany); Rokhlin, L.L. [Baikov Institute of Metallurgy and Materials Science, Leninsky prosp. 49, 119991 GSP-1, Moscow (Russian Federation); Dobatkina, T.V. [Baikov Institute of Metallurgy and Materials Science, Leninsky prosp. 49, 119991 GSP-1, Moscow (Russian Federation); Schmid-Fetzer, R. [Institute of Metallurgy, Clausthal University of Technology, Robert-Koch Strasse 42, D-38678 Clausthal-Zellerfeld (Germany)]. E-mail: schmid-fetzer@tu-clausthal.de

    2007-05-16

    Three series of Al-rich alloys in the system Al-Zn-Mg-Cu-Sc-Zr and the subsystems Al-Zn-Mg-Cu-Sc and Al-Zn-Mg-Sc were studied by thermodynamic calculations. Phase formation was compared with experimental data obtained by DTA and microstructural analysis. Calculated phase diagrams, phase amount charts and enthalpy charts together with non-equilibrium calculations under Scheil conditions reveal significant details of the complex phase formation. This enables consistent and correct interpretation of thermal analysis data. Especially the interpretation of liquidus temperature and primary phase is prone to be wrong without using this tool of computational thermodynamics. All data are predictions from a thermodynamic database developed for Mg-alloys and not a specialized Al-alloy database. That provides support for a reasonable application of this database for advanced Mg-alloys beyond the conventional composition ranges.

  19. Study of corrosion kinetics of fuel element tubes from calcium-thermal zirconium alloy Zr1Nb in water at 350 degree C and in vapour at 400 and 500 degree C

    International Nuclear Information System (INIS)

    Petel'guzov, I.A.

    2002-01-01

    In the report brought results of corrosion process studies in water medium of pipe samples for fuel element shells from Zr1Nb alloy (earlier KTZ-110),made from the calcium-thermal zirconium alloys developed in the Ukraine of technology and,for the comparison,samples of pipes from the staff alloy E110, applicable in fuel elements acting reactors of type WWER. Tests were conducted under the working temperature of fuel shells in the reactor (350 degree C) in during of 14000 hours and under increased temperatures (400 degree C) within a time acordinly 4000 hours. Samples from the alloy Zr1Nb had more high contents of oxygen (before 0,12%...0,16%), than staff alloy Eh110 (0,08%O). Studies have shown sufficiently high corrosion stability of experimental alloy Zr1Nb, close to stability of alloy E110.Discovered signs of corrosion 'breakway' or 'transition' on kinetic corrosion curves of Zr1Nb alloys and E110 alloy, characterisating zircaloy type of alloy. Considered mechanism of influence of oxygen on the corrosion process of zirconium alloys with the additive a niobium

  20. Ammonia role in WWER primary circuit water chemistry optimization

    International Nuclear Information System (INIS)

    Kritskij, V.G.; Stjagkin, P.S.; Chvedova, M.N.; Slobodov, A.A.

    1999-01-01

    Ammonia influence on iron crud's solubility at 300 deg. C and different relations of boric acid and alkaline cation sum are considered. Reduction of dose rate on WWER-440 steam generators at average ammonia concentration increasing is empirically explained. Practical recommendations on optimization of WWER primary circuit water chemistry are given. (author)

  1. Improving tribological properties of Ti-5Zr-3Sn-5Mo-15Nb alloy by double glow plasma surface alloying

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Lili; Qin, Lin, E-mail: qinlin@tyut.edu.cn; Kong, Fanyou; Yi, Hong; Tang, Bin

    2016-12-01

    Highlights: • The Mo alloyed layers were successfully prepared on TLM surface by DG-PSA. • The surface microhardness of TLM is remarkably enhanced by Mo alloying. • The TLM samples after Mo alloying exhibit good wettability. • The Mo alloyed TLM samples show excellent tribological properties. - Abstract: Molybdenum, an alloying element, was deposited and diffused on Ti-5Zr-3Sn-5Mo-15Nb (TLM) substrate by double glow plasma surface alloying technology at 900, 950 and 1000 °C. The microstructure, composition distribution and micro-hardness of the Mo modified layers were analyzed. Contact angles on deionized water and wear behaviors of the samples against corundum balls in simulated human body fluids were investigated. Results show that the surface microhardness is significantly enhanced after alloying and increases with treated temperature rising, and the contact angles are lowered to some extent. More importantly, compared to as-received TLM alloy, the Mo modified samples, especially the one treated at 1000 °C, exhibit the significant improvement of tribological properties in reciprocating wear tests, with lower specific wear rate and friction coefficient. To conclude, Mo alloying treatment is an effective approach to obtain excellent comprehensive properties including optimal wear resistance and improved wettability, which ensure the lasting and safety application for titanium alloys as the biomedical implants.

  2. Stress corrosion cracking behaviour of Alloy 600 in high temperature water

    International Nuclear Information System (INIS)

    Webb, G.L.; Burke, M.G.

    1995-01-01

    The stress corrosion cracking (SCC) susceptibility of Alloy 600 in deaerated water at 360 deg. C, as measured with statistically-loaded U-bend specimens, is dependent upon microstructure and whether the material was cold-worked and annealed (CWA) or hot-worked and annealed (HWA). All cracking was intergranular, and materials lacking grain boundary carbides were most susceptible to SCC initiation. CWA tubing materials are more susceptible to SCC initiation than HWA ring-rolled forging materials with similar microstructures, as determined by light optical metallography (LOM). In CWA tubing materials one crack dominated and grew to a large size that was observable by visual inspection. HWA materials with a low hot-working finishing temperature (below 925 deg. C) and final anneals at temperatures ranging from 1010 deg. C to 1065 deg. C developed both large cracks, similar to those found in CWA materials, and also small intergranular microcracks, which are detectable only by destructive metallographic examination. HWA materials with a high hot-working finishing temperature (above 980 deg. C) and high-temperature final anneal (above 1040 deg. C), with grain boundaries that are fully decorated, developed only microcracks, which were observed in all specimens examined. These materials developed no large, visually detectable cracks, even after more than 300 weeks exposure. A low-temperature thermal treatment (610 deg. C for 7h), which reduced or eliminates SCC in Alloy 600, did not eliminate microcrack formation in the high temperature processed HWA materials. Detailed microstructural characterization using conventional metallographic and analytical electron microscopy (AEM) techniques was performed on selected materials to identify the factors responsible for the observed differences in cracking behaviour. 11 refs, 12 figs, 3 tabs

  3. Thermal ageing and short-range ordering of Alloy 690 between 350 and 550 °C

    Energy Technology Data Exchange (ETDEWEB)

    Mouginot, Roman, E-mail: roman.mouginot@aalto.fi [Aalto University School of Engineering, Department of Mechanical Engineering, Otakaari 4, 02150 Espoo (Finland); Sarikka, Teemu [Aalto University School of Engineering, Department of Mechanical Engineering, Otakaari 4, 02150 Espoo (Finland); Heikkilä, Mikko [University of Helsinki, Laboratory of Inorganic Chemistry, A.I.Virtasen Aukio 1, 00560 Helsinki (Finland); Ivanchenko, Mykola; Ehrnstén, Ulla [VTT Technical Research Centre of Finland LTD, Kemistintie 3, 02150 Espoo (Finland); Kim, Young Suk; Kim, Sung Soo [Korea Atomic Energy Research Institute, Daedeok-Daero, 989-111, Yuseong, Daejeon, 34057 (Korea, Republic of); Hänninen, Hannu [Aalto University School of Engineering, Department of Mechanical Engineering, Otakaari 4, 02150 Espoo (Finland)

    2017-03-15

    Thermal ageing of Alloy 690 triggers an intergranular (IG) carbide precipitation and is known to promote an ordering reaction causing lattice contraction. It may affect the long-term primary water stress corrosion cracking (PWSCC) resistance of pressurized water reactor (PWR) components. Four conditions of Alloy 690 (solution annealed, cold-rolled and/or heat-treated) were aged between 350 and 550 °C for 10 000 h and characterized. Although no direct observation of ordering was made, variations in hardness and lattice parameter were attributed to the formation of short-range ordering (SRO) in all conditions with a peak level at 420 °C, consistent with the literature. Prior heat treatment induced ordering before thermal ageing. At higher temperatures, stress relaxation, recrystallization and α-Cr precipitation were observed in the cold-worked samples, while a disordering reaction was inferred in all samples based on a decrease in hardness. IG precipitation of M{sub 23}C{sub 6} carbides increased with increasing ageing temperature in all conditions, as well as diffusion-induced grain boundary migration (DIGM). - Highlights: • SRO was suggested in Alloy 690 with 9.18 wt% Fe after thermal ageing at 350, 420 and 475 °C. • Prior thermal treatment promoted SRO before ageing. • Cold work led to recrystallization and precipitation of α-Cr upon ageing at 550 °C. • Thermal ageing promoted IG precipitation of Cr-rich M{sub 23}C{sub 6} carbides and DIGM.

  4. Generation of copper, nickel, and CuNi alloy nanoparticles by spark discharge

    International Nuclear Information System (INIS)

    Muntean, Alex; Wagner, Moritz; Meyer, Jörg; Seipenbusch, Martin

    2016-01-01

    The generation of copper, nickel, and copper-nickel alloy nanoparticles by spark discharge was studied, using different bespoke alloy feedstocks. Roughly spherical particles with a primary particle Feret diameter of 2–10 nm were produced and collected in agglomerate form. The copper-to-nickel ratios determined by Inductively coupled plasma mass spectrometry (ICP-MS), and therefore averaged over a large number of particles, matched the nominal copper content quite well. Further investigations showed that the electrode compositions influenced the evaporation rate and the primary particle size. The evaporation rate decreased with increasing copper content, which was found to be in good accordance with the Llewellyn-Jones model. However, the particle diameter was increasing with an increasing copper content, caused by a decrease in melting temperature due to the lower melting point of copper. Furthermore, the alloy compositions on the nanoscale were investigated via EDX. The nanoparticles exhibited almost the same composition as the used alloy feedstock, with a deviation of less than 7 percentage points. Therefore, no segregation could be detected, indicating the presence of a true alloy even on the nanoscale.

  5. Water-driven micromotors.

    Science.gov (United States)

    Gao, Wei; Pei, Allen; Wang, Joseph

    2012-09-25

    We demonstrate the first example of a water-driven bubble-propelled micromotor that eliminates the requirement for the common hydrogen peroxide fuel. The new water-driven Janus micromotor is composed of a partially coated Al-Ga binary alloy microsphere prepared via microcontact mixing of aluminum microparticles and liquid gallium. The ejection of hydrogen bubbles from the exposed Al-Ga alloy hemisphere side, upon its contact with water, provides a powerful directional propulsion thrust. Such spontaneous generation of hydrogen bubbles reflects the rapid reaction between the aluminum alloy and water. The resulting water-driven spherical motors can move at remarkable speeds of 3 mm s(-1) (i.e., 150 body length s(-1)), while exerting large forces exceeding 500 pN. Factors influencing the efficiency of the aluminum-water reaction and the resulting propulsion behavior and motor lifetime, including the ionic strength and environmental pH, are investigated. The resulting water-propelled Al-Ga/Ti motors move efficiently in different biological media (e.g., human serum) and hold considerable promise for diverse biomedical or industrial applications.

  6. Changes in water chemistry and primary productivity of a reactor cooling reservoir (Par Pond)

    International Nuclear Information System (INIS)

    Tilly, L.J.

    1975-01-01

    Water chemistry and primary productivity of a reactor cooling reservoir have been studied for 8 years. Initially the primary productivity increased sixfold, and the dissolved solids doubled. The dissolved-solids increase appears to have been caused by additions of makeup water from the Savannah River and by evaporative concentration during the cooling process. As the dissolved-solids concentrations and the conductivity of makeup water leveled off, the primary productivity stabilized. Major cation and anion concentrations generally followed total dissolved solids through the increase and plateau; however, silica concentrations declined steadily during the initial period of increased plankton productivity. Standing crops of net seston and centrifuge seston did not increase during this initial period. The collective data show the effects of thermal input to a cooling reservoir, illustrate the need for limnological studies before reactor siting, and suggest the possibility of using makeup-water additions to power reactor cooling basins as a reservoir management tool

  7. Microstructural development in equiatomic multicomponent alloys

    International Nuclear Information System (INIS)

    Cantor, B.; Chang, I.T.H.; Knight, P.; Vincent, A.J.B.

    2004-01-01

    Multicomponent alloys containing several components in equal atomic proportions have been manufactured by casting and melt spinning, and their microstructures and properties have been investigated by a combination of optical microscopy, scanning electron microscopy, electron probe microanalysis, X-ray diffractrometry and microhardness measurements. Alloys containing 16 and 20 components in equal proportions are multiphase, crystalline and brittle both as-cast and after melt spinning. A five component Fe 20 Cr 20 Mn 20 Ni 20 Co 20 alloy forms a single fcc solid solution which solidifies dendritically. A wide range of other six to nine component late transition metal rich multicomponent alloys exhibit the same majority fcc primary dendritic phase, which can dissolve substantial amounts of other transition metals such as Nb, Ti and V. More electronegative elements such as Cu and Ge are less stable in the fcc dendrites and are rejected into the interdendritic regions. The total number of phases is always well below the maximum equilibrium number allowed by the Gibbs phase rule, and even further below the maximum number allowed under non-equilibrium solidification conditions. Glassy structures are not formed by casting or melt spinning of late transition metal rich multicomponent alloys, indicating that the confusion principle does not apply, and other factors are more important in promoting glass formation

  8. Estimating residual life of alloy 600 RPV penetrations

    International Nuclear Information System (INIS)

    Hunt, E.S.; White, G.A.; Pathania, R.; Arey, M.L.; Whitaker, D.E.

    1996-01-01

    Primary water stress corrosion cracking (PWSCC) of Alloy 600 penetrations PWR in reactor pressure vessel (RPV) heads has become a significant economic concern worldwide. PWSCC of these penetrations has led to extended maintenance outages, expensive inspections and repairs, and in some cases, replacement of the entire vessel head. This paper describes methodology developed to predict the remaining life of Alloy 600 penetrations in reactor vessel heads. Predictions of remaining life are an important input to planning models used by utilities to select a strategy for responding to the PWSCC issue at the lowest life cycle cost with an acceptably low risk of leakage. The remaining life of RPV penetrations is determined using the results of inspections of penetrations and statistical methods to predict future degradation. The analysis takes into account the effects of material properties, welding residual stresses, and operating temperature on PWSCC initiation and growth. The probability of developing cracks of various depths is assessed using Monte Carlo methods which provide for uncertainties in the input assumptions. For plants which have not yet performed inspections, remaining life predictions are based on inspection results from similar plants which have performed inspections with corrections made for known differences in design details, material properties and operating conditions

  9. Primary Water Chemistry Control during a Planned Outage at Bruce Power

    International Nuclear Information System (INIS)

    Ma, Guoping; Nashiem, Rod; Matheson, Shane; Yabar, Berman; Harper, Bill; Roberts, John G.

    2012-09-01

    Bruce Power has developed a comprehensive outage water chemistry program, which includes both primary and secondary chemistry requirements during planned outages. The purpose of the program is to emphasize the chemistry requirements during outages and subsequent start-ups in order to maintain the integrity of the systems, minimise activity transport and radiation fields, reduce the Carbon-14 release, and to ensure that the requirements are integrated with the outage management program. Prior to a planned outage, Station Chemical Technical Sections identify outage chemistry requirements to Operations and Outage Planning and ensure that work necessary to correct system chemistry issues is within outage work scope. The outage water chemistry program provides direction for establishing alternative sampling locations as demanded by the system configuration during the outage and identifies outage prerequisites for nuclear system purification capabilities. These requirements are contained in an outage checklist. The paper mainly highlights the primary water chemistry issues and chemistry control strategies during planned outages and discusses challenges and successes. (authors)

  10. VGB primary and secondary side water chemistry guidelines for PWR plants

    International Nuclear Information System (INIS)

    Neder, H.; Wolter, D.; Staudt, U.

    2007-01-01

    The recent revision of the VGB Water Chemistry Guidelines was issued in 2005 and published in the second half of 2006. These guidelines are based on the primary and secondary side operating chemistry experience with all Siemens designed pressurized water reactors gained since the beginning of the 1980s. These guidelines cover For the primary side chemistry Modified lithium boron chemistry, Zinc chemistry for dose rate reduction, Enriched boric acid (EBA) chemistry for high duty core design For the secondary side chemistry High all-volatile treatment (AVT) chemistry (high pH operation) Oxygen injection in the secondary side Especially for the secondary side chemistry, compared with the water chemistry guidelines of other organizations worldwide, these Guidelines are less stringent, providing more operational flexibility to the plant operation, and can be applied for all new designs of steam generators with egg-crates or broached hole tube supports and with I 690TT or I 800 tubing materials. This paper gives an overview of the 2006 revision of the VGB Water Chemistry Guidelines for PWR plants and describes the fundamental goals of water chemistry operation strategies. In addition, the reasons for the selected control parameters and action levels, to achieve an adequate plant performance, are presented based on the operating experience. (orig.)

  11. Primary Dendrite Arm Spacings in Al-7Si Alloy Directionally Solidified on the International Space Station

    Science.gov (United States)

    Angart, Samuel; Lauer, Mark; Poirier, David; Tewari, Surendra; Rajamure, Ravi; Grugel, Richard

    2015-01-01

    Samples from directionally solidified Al- 7 wt. % Si have been analyzed for primary dendrite arm spacing (lambda) and radial macrosegregation. The alloy was directionally solidified (DS) aboard the ISS to determine the effect of mitigating convection on lambda and macrosegregation. Samples from terrestrial DS-experiments thermal histories are discussed for comparison. In some experiments, lambda was measured in microstructures that developed during the transition from one speed to another. To represent DS in the presence of no convection, the Hunt-Lu model was used to represent diffusion controlled growth under steady-state conditions. By sectioning cross-sections throughout the entire length of a solidified sample, lambda was measured and calculated using the model. During steady-state, there was reasonable agreement between the measured and calculated lambda's in the space-grown samples. In terrestrial samples, the differences between measured and calculated lambda's indicated that the dendritic growth was influenced by convection.

  12. The corrosion behaviour of Zr3Al-based alloys

    International Nuclear Information System (INIS)

    Murphy, E.V.; Wieler, R.

    1977-07-01

    The corrosion resistance of several zirconium-aluminum alloys with aluminum contents ranging from 7.6 to 9.6 wt% was examined in 300 deg C and 325 deg C water, 350 deg C and 400 deg C steam and in air and wet CO 2 at 325 deg C and 400 deg C. In the transformed alloys there are three phases present, αZr, Zr 2 Al and Zr 3 Al of which the αZr phase is the least corrosion resistant. The most important factor controlling the corrosion behaviour of these alloys was found to be the size, distribution and amount of the αZr phase in the transformed alloys, which in turn was dependent upon the microstructural scale of the untransformed alloys

  13. Database on Performance of Neutron Irradiated FeCrAl Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Briggs, Samuel A. [Univ. of Wisconsin, Madison, WI (United States); Littrell, Ken [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Parish, Chad M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-01

    The present report summarizes and discusses the database on radiation tolerance for Generation I, Generation II, and commercial FeCrAl alloys. This database has been built upon mechanical testing and microstructural characterization on selected alloys irradiated within the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) up to doses of 13.8 dpa at temperatures ranging from 200°C to 550°C. The structure and performance of these irradiated alloys were characterized using advanced microstructural characterization techniques and mechanical testing. The primary objective of developing this database is to enhance the rapid development of a mechanistic understanding on the radiation tolerance of FeCrAl alloys, thereby enabling informed decisions on the optimization of composition and microstructure of FeCrAl alloys for application as an accident tolerant fuel (ATF) cladding. This report is structured to provide a brief summary of critical results related to the database on radiation tolerance of FeCrAl alloys.

  14. An assessment of corporate entrepreneurship in the alloy mining environment / S.S. Scheepers

    OpenAIRE

    Scheepers, Sarah Susanna

    2009-01-01

    This study highlighted the influence of the 13 corporate entrepreneurial constructs on the entrepreneurial climate in corporate organisations. The primary objective of this study was to assess the level of corporate entrepreneurship in the South African alloy mining environment, with specific reference to Xstrata Alloys and to make recommendations on the encouragement and promoting of a climate conducive to corporate entrepreneurship in Xstrata South Africa (Pty) Ltd - Alloys. The empiri...

  15. Development of phased array UT procedure for crack depth sizing on nickel based alloy weld

    International Nuclear Information System (INIS)

    Hirasawa, Taiji; Okada, Hisao; Fukutomi, Hiroyuki

    2012-01-01

    Recently, it is reported that the primary water stress corrosion cracking (PWSCC) has been occurred at the nickel based alloy weld components such as steam generator safe end weld, reactor vessel safe end weld, and so on, in PWR. Defect detection and sizing is important in order to ensure the reliable operation and life extension of nuclear power plants. In the reactor vessel safe end weld, it was impossible to measure crack depth of PWSCC. The crack was detected in the axial direction of the safe end weld. Furthermore, the crack had some features such as shallow, large aspect ratio (ratio of crack depth and length), sharp geometry of crack tip, and so on. Therefore, development and improvement of defect detection and sizing capabilities for ultrasonic inspection technique is required. Phased array UT technique was applied to nickel based alloy weld specimen with SCC cracks. From the experimental results, good accuracy of crack depth sizing by phased array UT for the inside inspection was shown. From these results, UT procedure for crack depth sizing was verified. Therefore, effectiveness of phased array UT for crack depth sizing in the nickel based alloy welds was shown. (author)

  16. Crack initiation behavior of neutron irradiated model and commercial stainless steels in high temperature water

    Energy Technology Data Exchange (ETDEWEB)

    Stephenson, Kale J., E-mail: kalejs@umich.edu; Was, Gary S.

    2014-01-15

    Highlights: • Environmental constant extension rate tensile tests were performed on neutron irradiated steel. • Percentage of intergranular cracking quantified the cracking susceptibility. • Cracking susceptibility varied with test environment, solute addition, and cold work. • No singular microstructural change could explain increases in cracking susceptibility with irradiation dose. • The increment of yield strength due to irradiation correlated well with cracking susceptibility. -- Abstract: The objective of this study was to isolate key factors affecting the irradiation-assisted stress corrosion cracking (IASCC) susceptibility of eleven neutron-irradiated austenitic stainless steel alloys. Four commercial purity and seven high purity stainless steels were fabricated with specific changes in composition and microstructure, and irradiated in a fast reactor spectrum at 320 °C to doses between 4.4 and 47.5 dpa. Constant extension rate tensile (CERT) tests were performed in normal water chemistry (NWC), hydrogen water chemistry (HWC), or primary water (PW) environments to isolate the effects of environment, elemental solute addition, alloy purity, alloy heat, alloy type, cold work, and irradiation dose. The irradiated alloys showed a wide variation in IASCC susceptibility, as measured by the relative changes in mechanical properties and crack morphology. Cracking susceptibility measured by %IG was enhanced in oxidizing environments, although testing in the lowest potential environment caused an increase in surface crack density. Alloys containing solute addition of Ni or Ni + Cr exhibited no IASCC. Susceptibility was reduced in materials cold worked prior to irradiation, and increased with increasing irradiation dose. Irradiation-induced hardening was accounted for by the dislocation loop microstructure, however no relation between crack initiation and radiation hardening was found.

  17. Effects of alloy composition and flow condition on the flow accelerated corrosion in neutral water condition

    International Nuclear Information System (INIS)

    Satoh, Tomonori; Ugachi, Hirokazu; Tsukada, Takashi; Uchida, Shunsuke

    2008-01-01

    The major mechanism of Flow accelerated corrosion (FAC) is the dissolution of the protective oxide on carbon steel, which is enhanced by mass transfer and erosion under high flow velocity conditions. In this study, the effects of alloy composition and flow velocity on FAC of carbon steel were evaluated by measuring FAC rate of tube type carbon steel specimens in the neutral water condition at 150degC. Obtained results are summarized in follows. 1) High FAC rate was depended upon the v 1.2 in the tube type specimen made of the standard alloy. 2) FAC was mitigated for the carbon steel with more than 0.03% of Cr content. 3) FAC rate decreased as Ni content increased in more than 0.1% of Ni content. 4) The difference in chemical composition of oxide film between Ni added carbon steel and Cr added one was confirmed. The hematite rich oxide was observed for Ni added carbon steel. 5) The effects of Cu on FAC rate was not observed up to 0.1% of Cu content. (author)

  18. The modification of some properties of Al-2%Mg alloy by Ti &Li alloying elements

    Directory of Open Access Journals (Sweden)

    Talib Abdulameer Jasim

    2017-11-01

    Full Text Available Aluminium-Magnisium alloys are light, high strength with resistance to corrosion and good weldability. When the content of magnesium  exceeds 3% there is a tendency to stress corrosion . This work is an attempt is to prepare low density alloy with up to approximately 2.54 g / cm3 by adding different contents of Ti, and lithium to aluminum-2%Magnisium alloy. The lithium is added in two aspects, lithium chloride and pure metal. The casting performed using conventional casting method. Moreover, solution heat treatment (SHT at 520 ºC for 4 hrs, quenching in cold water, and aging at 50ºC for 4 days were done to get better mechanical properties of all samples. Microstructure was inspected by light optical microscope before and after SHT. Alloy3 which contains 1.5%Ti was tested by SEM and EDS spectrometer to exhibit the shape and micro chemical analysis of Al3Ti phase. Hardness, ultimate tensile strength, and modulus of elasticity were tested for all alloys. The results indicated that Al3Ti phase precipitates in alloys contain 0.5%T, 1%Ti, And 1.5%Ti.  The phases Al3Li as well as Al3Ti were precipitated in alloy4 which contains 2%Ti, and 2.24%Li. Mechanical properties test results also showed that the alloy4 has achieved good results, the modulus of elasticity chanced from 310.65GPa before SHT to 521.672GPa, after SHT and aging, the ultimate tensile strength was changed from 365MPa before SHT to 469MPa, after SHT and aging,  and hardness was increased from 128 to 220HV.

  19. Development of the advanced nuclear materials -Development of Inconel alloys-

    International Nuclear Information System (INIS)

    Kuk, Il Hyun; Chang, Jin Sung; Lee, Chang Kyu; Park, Soon Dong; Kim, Woo Kon; Jeong, Man Kyo; Woo, Yoon Myung; Han, Chang Hee

    1995-07-01

    The performance and the integrity of the steam generator U-tubes directly affects the efficiency and economics of nuclear power plant because they are closely interrelated with the maintenance and repair. Also the steam generator U-tubes have been one of world-wide hot issues in nuclear power plants for long time because of their continuing corrosion-related degradation. Right after stress corrosion cracking of Alloy 600 tubes are reported at primary side, in which the environment is believed to be tightly controlled all the time, in mid 80's, alloy 690 has started to replace alloy 600. Alloy 690 is basically same with alloy 600 except more Cr content. Firstly minor elements in alloy 690 (C, B, N, Y, Mo) were added or controlled to improve hot workability and corrosion resistance. It would be much more desirable if the mechanism or basic understanding of the degradation phenomena of steam generator U-tubes in operation conditions can be illuminated through the alloy modification research. Alloy 600 tubes which were preproduced in cooperation with Sammi Special Steel were evaluated, being compared with imported one. Also alloy 600 and alloy 690 tubes were produced from Inconel 600 and 690 INCO- forged bar. These will be closely evaluated with purely Korean-made alloy 600 and 690 tubes. 22 tabs., 93 figs., 14 refs. (Author)

  20. Crevice Corrosion on Ni-Cr-Mo Alloys

    International Nuclear Information System (INIS)

    P. Jakupi; D. Zagidulin; J.J. Noel; D.W. Shoesmith

    2006-01-01

    Ni-Cr-Mo alloys were developed for their exceptional corrosion resistance in a variety of extreme corrosive environments. An alloy from this series, Alloy-22, has been selected as the reference material for the fabrication of nuclear waste containers in the proposed Yucca Mountain repository located in Nevada (US). A possible localized corrosion process under the anticipated conditions at this location is crevice corrosion. therefore, it is necessary to assess how this process may, or may not, propagate if the use of this alloy is to be justified. Consequently, the primary objective is the development of a crevice corrosion damage function that can be used to assess the evolution of material penetration rates. They have been using various electrochemical methods such as potentiostatic, galvanostatic and galvanic coupling techniques. Corrosion damage patterns have been investigated using surface analysis techniques such as scanning electron microscopy (SEM) and optical microscopy. All crevice corrosion experiments were performed at 120 C in 5M NaCl solution. Initiating crevice corrosion on these alloys has proven to be difficult; therefore, they have forced it to occur under either potentiostatic or galvanostatic conditions

  1. Initial stages of solidification of eutectic alloys

    International Nuclear Information System (INIS)

    Lemaignan, Clement

    1980-01-01

    The study of the various initial stages of eutectic solidification - i.e. primary nucleation, eutectic structure formation and stable growth conditions - was undertaken with various techniques including low angle neutron diffusion, in-situ electron microscopy on solidifying alloys and classical metallography. The results obtained allow to discuss the effect of metastable states during primary nucleation, of surface dendrite during eutectic nucleation and also of the crystallographic anisotropy during growth. (author) [fr

  2. The characteristics of corrosion, radiation degradation and dissolution of titanium alloys

    International Nuclear Information System (INIS)

    Sung, K. W.; Na, J. W.; Choi, B. S.; Lee, D. J.; Chang, M. H.

    2001-12-01

    In order to establish the technical bases of water chemistry design requirement related titanium alloys, we investigated the characteristics of corrosion, activation, radiation degradation, radiation hydrogen embrittlement of titanium alloys and dissolution of titanium dioxide. Titanium alloys generally have high corrosion resistance. Corrosion product release from PT-7M and PT-3V titanium alloy surface for 18 months of operation is negligible, and the corrosion penetration for about 30 years is about 1 μm, while the corrosion rates is not higher than one third of that of austenitic steel. Titanium only converts into Sc-46 with 85 day halflife after neutron irradiation, and its radioactivity is not higher than one thousandth of that produced from nickel. Therefore, under the condition without any neutron irradiation, the radiation damage of titanium alloys would have no problem. Titanium dioxide, that protects the metals from the corrosion, has retrograde solubility in neutral solutions. It does not form any complexes with ligands such as ammonia, but Ti(IV) gets more stable by complexing with water molecules. In conclusion, it is estimated that titanium alloys such as PT-7M would be applicable to steam generator materials

  3. Corrosion of aluminum alloys in ocean thermal energy conversion seawaters

    International Nuclear Information System (INIS)

    Larsen-Basse, J.

    1984-01-01

    Aluminum alloys 5052, 3004, and Alclad 3003 and 3004 were exposed to flowing seawater at 2.44 m/s (8 fps) at the Seacoast Test Facility on Hawaii. One year data for warm surface water and three mouth data for cold water from 600 m depth are reported for free fouling, chlorinated and sponge ball cleaned conditions. All alloys pit in deep seawater, but show no pitting in warm surface water. Uniform corrosion in the warm water is initially rapid, but after 25 to 30 days the rate becomes slower and extrapolated 30 year material losses are in the 125 to 215 μm range. Chlorination at a level of 0.05 ppm for one hour per day has only a minor effect on corrosion rates, while sponge ball cleaning leads to erosion-corrosion of the Alclad surfaces and has no effect on alloy 5052. The need for additional testing in tropical seawater is discussed, as is the need for an improved understanding of the formation of inorganic scale films, their properties, and their effect on corrosion rates and heat transfer

  4. Microstructure and mechanical properties of lost foam cast 356 alloys

    Directory of Open Access Journals (Sweden)

    Qi-gui Wang

    2015-05-01

    Full Text Available Microstructure and mechanical properties of lost foam cast aluminum alloys have been investigated in both primary A356 (0.13% Fe and secondary 356 (0.47%. As expected, secondary 356 shows much higher content of Fe-rich intermetallic phases, and in particular the porosity in comparison with primary A356. The average area percent and size (length of Fe-rich intermetallics change from about 0.5% and 6 祄 in A356 to 2% and 25 祄 in 356 alloy. The average area percent and maximum size of porosity also increase from about 0.4% and 420 祄 to 1.4% and 600 祄, respectively. As a result, tensile ductility decreases about 60% and ultimate tensile strength declines about 8%. Lower fatigue strength was also experienced in the secondary 356 alloy. Low cycle fatigue (LCF strength decreased from 187 MPa in A356 to 159 MPa in 356 and high cycle fatigue (HCF strength also declined slightly from 68 MPa to 64 MPa.

  5. Thermal simulation of quenching uranium-0.75% titanium alloy in water

    International Nuclear Information System (INIS)

    Siman-Tov, M.; Llewellyn, G.H.; Childs, K.W.; Ludtka, G.M.; Aramayo, G.A.

    1985-01-01

    A computer model, The Quench Simulator, has been developed to simulate and predict in detail the behavior of U-0.75 Ti alloy when quenched at high temperature (about 850 0 C) in cold water. The code allows one to determine the time- and space-dependent distributions of temperature, residual stress, distortion, and microstructure that evolve during the quenching process. The nonlinear temperature- and microstructure-dependent properties, as well as the cooling rate-dependent heats of transformation, are incorporated into the model. The complex boiling heat transfer with its various regimes and other thermal boundary conditions are simulated. Experiments have been performed and incorporated into the model. Both sudden submersion and gradual controlled immersion can be applied. A parametric and sensitivity study has been performed demonstrating the importance of the thermal boundary conditions applied for achieving certain product characteristics. The thermal aspects of the model and its applications are discussed and demonstrated

  6. Effect of dissolved oxygen on IGSCC of Alloy 600

    International Nuclear Information System (INIS)

    Maeng, W.Y.; Choi, M.S.; Kim, U.C.

    2002-01-01

    The effect of dissolved oxygen on the SCC of Alloy 600 was studied by the slow strain rate test(SSRT) method. The SSRT tests were carried out in aerated and in deaerated pure water at 360 C at the strain rate of 2.5 x 10 -7 /s. Hump specimens were used to shorten test time. The SCC susceptibility was higher in the deaerated water environment than in aerated water environments. The shape of load-deformation curves of the tests in those two environments indicates that oxygen content in water significantly influences the SCC susceptibility of Alloy 600. It was considered that the increase of SCC resistance in aerated water is due to the high corrosion potential of the metal surface, and the according decrease of corrosion current due to the formation of a protective oxide layer. (authors)

  7. A study on the microstructural characteristics of rapidly solidified Al-Fe alloys(I)

    International Nuclear Information System (INIS)

    Kim, D.H.; Lee, H.I.

    1991-01-01

    Solidification microstructures and phases in rapidly solidified Al-5, 10wt% Fe alloys have been investigated by TEM bright field and dark field imaging techniques and electron and x-ray diffraction techniques. Rapid solidification of Al-5, 10wt%Fe alloys produces various metastable and stable phases, such as Al m Fe, Al 6 Fe and Al 13 Fe 4 . In addition to these phases, clusters of randomly oriented few nm scale particles exist in the form of fine cellular network with α-Al or primary spherical particles. Solidification microstructures of the rapidly solidified Al-5, 10wt%Fe alloys consist of various combination of primary phases such as Al 13 Fe 4 , Al m Fe and cluster of nm scale particles, and cellular/dendritic structures such as fine cellular network structure of nm scale particle clusters and α-Al and cellular structure of Al m Fe and α-Al, depending upon alloy compositions and local cooling rates. (Author)

  8. The role of time-dependent deformation in intergranular crack initiation of alloy 600 steam generator tubing material

    International Nuclear Information System (INIS)

    Was, G.S.; Lian, K.

    1998-03-01

    Intergranular stress corrosion cracking (IGSCC) of two commercial alloy 600 conditions (600LT, 600HT) and controlled- purity Ni-18Cr-9Fe alloys (CDMA, CDTT) were investigated using constant extension rate tensile (CERT) tests in primary water (0.01M LiOH+0.01M H 3 BO 3 ) with 1 bar hydrogen overpressure at 360 degrees C and 320 degrees C. Heat treatments produced two types of microstructures in both commercial and controlled-purity alloys: one dominated by grain boundary carbides (600HT and CDTT) and one dominated by intragranular carbides (600LT and CDMA). CERT tests were conducted over a range of strain rates and at two temperatures with interruptions at specific strains to determine the crack depth distributions. Results show that in all samples, IGSCC was the dominant failure mode. For both the commercial alloy and the controlled-purity alloys, the microstructure with grain boundary carbides showed delayed crack initiation and shallower crack depths than did the intragranular carbide microstructure under all experimental conditions. This data indicates that a grain boundary carbide microstructure is more resistant to IGSCC than an intragranular carbide microstructure. Observations support both the film rupture/slip dissolution mechanism and enhanced localized plasticity. The advantage of these results over previous studies is that the different carbide distributions were obtained in the same commercial alloy using different heat treatments, and in the other case, in nearly identical controlled-purity alloys. Therefore, observations of the effects of carbide distribution on IGSCC can more confidently be attributed to the carbide distribution alone rather than other potentially significant differences in microstructure or composition

  9. Grain refinement of AZ91D magnesium alloy by a new Mg–50%Al4C3 master alloy

    International Nuclear Information System (INIS)

    Liu, Shengfa; Chen, Yang; Han, Hui

    2015-01-01

    A novel and simple method for preparing Mg–50%Al 4 C 3 (hereafter in wt.%) master alloy has been developed by powder in-situ synthesis process under argon atmosphere. X-ray diffraction (XRD), scanning electron microscope (SEM) and energy dispersive X-ray spectroscopy (EDS) results show the existence of Al 4 C 3 particles in this master alloy. After adding 1.8% Mg–50%Al 4 C 3 master alloy, the average grain size of α-Mg decreased from 360 μm to 154 μm. Based on the DTA test results and calculation of the planar disregistry between Al 4 C 3 and α-Mg, Al 4 C 3 particles located in the central regions of magnesium grains can act as the heterogeneous nucleus of primary α-Mg phase

  10. Corrosion Characteristics of Ti-xTa Alloys with Ta contents

    International Nuclear Information System (INIS)

    Kim, H. J.; Choe, H. C.

    2013-01-01

    The purpose of this study was to investigate corrosion characteristics of Ti-xTa alloys with Ta contents. Ti-xTa alloys used as samples (x=30, 40%) were arc-melted under argon atmosphere of 99.9% purity. Ti-xTa alloys were homogenized for 12hr at 1000 .deg. C and then water quenched. The surface characteristics of Ti-xTa alloys were investigated using optical microscopy (OM) and X-ray diffractometer (XRD). The anodic corrosion behaviors of the specimens were examined through potentiodynamic, potentiostatic and galvanostatic test in 0.9 % NaCl solution at 36.5 ± 1 .deg. C. After corrosion test, the surface characteristics of Ti-xTa alloys were investigated using OM. The microstructure of Ti-Ta alloy showed the beta structure with Ta content. The corrosion resistance of Ti alloy was improved by increasing Ta content and the corrosion morphology of Ti-Ta alloy showed that the site attacked by chloride ion decreased from the active to passive region with Ta content. Potential of Ti-40Ta alloy increased as time increased, whereas, current density of Ti-40Ta alloy decreased as time increased compared to Ti-30 alloy

  11. Mechanical and corrosion properties of Ni-Cr-Fe Alloy 600 related to primary side SCC

    International Nuclear Information System (INIS)

    Begley, J.A.; Jacko, R.J.; Gold, R.E.

    1987-01-01

    The two-fold objective of the program is to provide the mechanical property data required for the development of a strain rate damage model for environmentally assisted cracking of Inconel 600 and to evaluate critical damage model parameters in primary water environments by conducting a series of stress corrosion tests. The test program includes mechanical property tests at 20 0 C, 316 0 C and strain rate tests to determine critical strain rate SCC parameters in primary water environments. Data are presented from slow strain rate tensile tests, stress relaxation tests and creep tests. A short discussion of the Gerber-Garud Strain Rate Damage Model is included to provide the background rationale for the test program. Utilitarian aspects of the Strain Rate Damage Model and the test program data are presented. Analysis of accelerated stress corrosion testing at high temperatures, and the contribution of thermally activated inelastic deformation to apparent activation energies for stress corrosion cracking is emphasized

  12. Modeling the transport of hydrogen in the primary coolant of pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Subramanian, H.; Velmurugan, S.; Narasimhan, S.V.; Jain, A.K.; Dash, S.C.

    2008-01-01

    Heavy water (D 2 O) is used in primary heat transport systems of PHWRs. To suppress the radiolysis of heavy water and to control oxygen, hydrogen is added at regular intervals to the primary heat transport system. The added hydrogen finds it way to the heavy water storage tank after passing through the bleed condenser. Owing to the different temperatures and two phase region present in these systems, hydrogen gets redistributed. It is important to know the concentration of dissolved hydrogen in these regions in order to ensure a steady state dissolved hydrogen concentration in the primary system. Different power stations report variations in the frequency and quantity of hydrogen added to achieve the prescribed steady state level. This paper makes an attempt to account for the inventory of hydrogen and model its transport in PHT system. (author)

  13. Corrosion behaviour of non-ferrous metals in sea water

    Energy Technology Data Exchange (ETDEWEB)

    Birn, Jerzy; Skalski, Igor [Ship Design and Research Centre, Al. Rzeczypospolitej 8, 80-369 Gdansk (Poland)

    2004-07-01

    The most typical kinds of corrosion of brasses are selective corrosion (dezincification) and stress corrosion. Prevention against these kinds of corrosion lies in application of arsenic alloy addition and appropriate heat treatment removing internal stresses as well as in maintaining the arsenic and phosphorus contents on a proper level. The most typical corrosion of cupronickels is the local corrosion. Selective corrosion occurs less often and corrosion cracking caused by stress corrosion in sea water does not usually occur. Crevice corrosion is found especially in places of an heterogeneous oxidation of the surface under inorganic deposits or under bio-film. Common corrosive phenomena for brasses and cupronickels are the effects caused by sea water flow and most often the impingement attack. Alloy additions improve resistance to the action of intensive sea water flow but situation in this field requires further improvement, especially if the cheaper kinds of alloys are concerned. Contaminants of sea water such as ammonia and hydrogen sulphide are also the cause of common corrosion processes for all copper alloys. Corrosion of copper alloys may be caused also by sulphate reducing bacteria (SRB). Galvanic corrosion caused by a contact with titanium alloys e.g. in plate heat exchangers may cause corrosion of both kinds copper alloys. Bronzes belong to copper alloys of the highest corrosion resistance. Failures that sometimes occur are caused most often by the cavitation erosion, by an incorrect chemical composition of alloys or at last by their inadequate structure. The main problems of aluminium alloys service in sea water are following phenomena: local corrosion (pitting and crevice corrosion), galvanic corrosion, exfoliation and corrosion in the presence of OH- ions. The cause of local corrosion are caused by presence of passive film on the alloy's surface and presence of chlorides in sea water which are able to damage the passive film. Galvanic corrosion is

  14. Integrated Guidelines for Management of Alloy 600 Locations

    Energy Technology Data Exchange (ETDEWEB)

    Na, Kyung-Hwan; Chung, Hansub; Yang, Jun-Seog; Lee, Kyoung-Soo [KHNP-Central Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The locations experiencing PWSCC include steam generator tubes, pressurizer instrumental nozzles, control rod driving mechanism(CRDM) penetration nozzles, reactor outlet nozzles, and bottom mounted instrumental(BMI) nozzles. Korea Hydro and Nuclear Power Co.(KHNP) has developed integrated guidelines for management of alloy 600 locations and the guidelines are under review by the regulator. The guidelines consist of alloy 600 location database, inspection program, maintenance/preventive maintenance method, and finally water chemistry management for PWSCC mitigation. In this paper, the detailed contents are presented. The integrated guidelines collected all relevant information on the management of alloy 600 locations. This information may be useful for establishing the most effective preventive maintenance strategies by prioritization in addition to maintenance strategies. Table II summarize maintenance strategies for alloy 600 locations.

  15. Atmospheric corrosion of uranium-carbon alloys

    International Nuclear Information System (INIS)

    Rousset, P.; Accary, A.

    1965-01-01

    The authors study the corrosion of uranium-carbon alloys having compositions close to that of the mono-carbide; they show that the extent of the observed corrosion effects increases with the water vapour content of the surrounding gas and they conclude that the atmospheric corrosion of these alloys is due essentially to the humidity of the air, the effect of the oxygen being very slight at room temperature. They show that the optimum conditions for preserving U-C alloys are either a vacuum or a perfectly dry argon atmosphere. The authors have also established that the type of corrosion involved is a corrosion which 'cracks under stress' and is transgranular (it can also be intergranular in the case of sub-stoichiometric alloys). They propose, finally, two hypotheses for explaining this mechanism, one of which is illustrated by the existence, at the fissure interface, of corrosion products which can play the role of 'corners' in the mono-carbide grains. (authors) [fr

  16. Relationships Between Solidification Parameters in A319 Aluminum Alloy

    Science.gov (United States)

    Vandersluis, E.; Ravindran, C.

    2018-03-01

    The design of high-performance materials depends on a comprehensive understanding of the alloy-specific relationships between solidification and properties. However, the inconsistent use of a particular solidification parameter for presenting materials characterization in the literature impedes inter-study comparability and the interpretation of findings. Therefore, there is a need for accurate expressions relating the solidification parameters for each alloy. In this study, A319 aluminum alloy castings were produced in a permanent mold with various preheating temperatures in order to control metal cooling. Analysis of the cooling curve for each casting enabled the identification of its liquidus, Al-Si eutectic, and solidus temperatures and times. These values led to the calculation of the primary solidification rate, total solidification rate, primary solidification time, and local solidification time for each casting, which were related to each other as well as to the average casting SDAS and material hardness. Expressions for each of their correlations have been presented with high coefficients of determination, which will aid in microstructural prediction and casting design.

  17. Galvanic Corrosion between Alloy 690 and Magnetite in Alkaline Aqueous Solutions

    Directory of Open Access Journals (Sweden)

    Soon-Hyeok Jeon

    2015-12-01

    Full Text Available The galvanic corrosion behavior of Alloy 690 coupled with magnetite has been investigated in an alkaline solution at 30 °C and 60 °C using a potentiodynamic polarization method and a zero resistance ammeter. The positive current values were recorded in the galvanic couple and the corrosion potential of Alloy 690 was relatively lower. These results indicate that Alloy 690 behaves as the anode of the pair. The galvanic coupling between Alloy 690 and magnetite increased the corrosion rate of Alloy 690. The temperature increase led to an increase in the extent of galvanic effect and a decrease in the stability of passive film. Galvanic effect between Alloy 690 and magnetite is proposed as an additional factor accelerating the corrosion rate of Alloy 690 steam generator tubing in secondary water.

  18. Precipitation Behavior of Magnesium Alloys Containing Neodymium and Yttrium

    Science.gov (United States)

    Solomon, Ellen L. S.

    Magnesium is the lightest of the structural metals and has great potential for reducing the weight of transportation systems, which in turn reduces harmful emissions and improves fuel economy. Due to the inherent softness of Mg, other elements are typically added in order to form a fine distribution of precipitates during aging, which improves the strength by acting as barriers to moving dislocations. Mg-RE alloys are unique among other Mg alloys because they form precipitates that lie parallel to the prismatic planes of the Mg matrix, which is an ideal orientation to hinder dislocation slip. However, RE elements are expensive and impractical for many commercial applications, motivating the rapid design of alternative alloy compositions with comparable mechanical properties. Yet in order to design new alloys reproducing some of the beneficial properties of Mg-RE alloys, we must first fully understand precipitation in these systems. Therefore, the main objectives of this thesis are to identify the roles of specific RE elements (Nd and Y) on precipitation and to relate the precipitate microstructure to the alloy strength. The alloys investigated in this thesis are the Mg-Nd, Mg-Y, and Mg-Y-Nd systems, which contain the main alloying elements of commercial WE series alloys (Y and Nd). In all three alloy systems, a sequence of metastable phases forms upon aging. Precipitate composition, atomic structure, morphology, and spatial distribution are strongly controlled by the elastic strain energy originating from the misfitting coherent precipitates. The dominating role that strain energy plays in these alloy systems gives rise to very unique microstructures. The evolution of the hardness and precipitate microstructure with aging revealed that metastable phases are the primary strengthening phases of these alloys, and interact with dislocations by shearing. Our understanding of precipitation mechanisms and commonalities among the Mg-RE alloys provide future avenues to

  19. Microstructure, corrosion behavior and cytotoxicity of biodegradable Mg-Sn implant alloys prepared by sub-rapid solidification.

    Science.gov (United States)

    Zhao, Chaoyong; Pan, Fusheng; Zhao, Shuang; Pan, Hucheng; Song, Kai; Tang, Aitao

    2015-09-01

    In this study, biodegradable Mg-Sn alloys were fabricated by sub-rapid solidification, and their microstructure, corrosion behavior and cytotoxicity were investigated by using optical microscopy, scanning electron microscopy equipped with an energy dispersive X-ray spectroscopy, X-ray diffraction, immersion test, potentiodynamic polarization test and cytotoxicity test. The results showed that the microstructure of Mg-1Sn alloy was almost equiaxed grain, while the Mg-Sn alloys with higher Sn content (Sn≥3 wt.%) displayed α-Mg dendrites, and the secondary dendrite arm spacing of the primary α-Mg decreased significantly with increasing Sn content. The Mg-Sn alloys consisted of primary α-Mg matrix, Sn-rich segregation and Mg2Sn phase, and the amount of Mg2Sn phases increased with increasing Sn content. Potentiodynamic polarization and immersion tests revealed that the corrosion rates of Mg-Sn alloys increased with increasing Sn content. Cytotoxicity test showed that Mg-1Sn and Mg-3Sn alloys were harmless to MG63 cells. These results of the present study indicated that Mg-1Sn and Mg-3Sn alloys were promising to be used as biodegradable implants. Copyright © 2015 Elsevier B.V. All rights reserved.

  20. Salt Fog Testing Iron-Based Amorphous Alloys

    International Nuclear Information System (INIS)

    Rebak, Raul B.; Aprigliano, Louis F.; Day, S. Daniel; Farmer, Joseph C.

    2007-01-01

    Iron-based amorphous alloys are hard and highly corrosion resistant, which make them desirable for salt water and other applications. These alloys can be produced as powder and can be deposited as coatings on any surface that needs to be protected from the environment. It was of interest to examine the behavior of these amorphous alloys in the standard salt-fog testing ASTM B 117. Three different amorphous coating compositions were deposited on 316L SS coupons and exposed for many cycles of the salt fog test. Other common engineering alloys such as 1018 carbon steel, 316L SS and Hastelloy C-22 were also tested together with the amorphous coatings. Results show that amorphous coatings are resistant to rusting in salt fog. Partial devitrification may be responsible for isolated rust spots in one of the coatings. (authors)

  1. In vitro profiling of epigenetic modifications underlying heavy metal toxicity of tungsten-alloy and its components

    International Nuclear Information System (INIS)

    Verma, Ranjana; Xu, Xiufen; Jaiswal, Manoj K.; Olsen, Cara; Mears, David; Caretti, Giuseppina; Galdzicki, Zygmunt

    2011-01-01

    Tungsten-alloy has carcinogenic potential as demonstrated by cancer development in rats with intramuscular implanted tungsten-alloy pellets. This suggests a potential involvement of epigenetic events previously implicated as environmental triggers of cancer. Here, we tested metal induced cytotoxicity and epigenetic modifications including H3 acetylation, H3-Ser10 phosphorylation and H3-K4 trimethylation. We exposed human embryonic kidney (HEK293), human neuroepithelioma (SKNMC), and mouse myoblast (C2C12) cultures for 1-day and hippocampal primary neuronal cultures for 1-week to 50-200 μg/ml of tungsten-alloy (91% tungsten/6% nickel/3% cobalt), tungsten, nickel, and cobalt. We also examined the potential role of intracellular calcium in metal mediated histone modifications by addition of calcium channel blockers/chelators to the metal solutions. Tungsten and its alloy showed cytotoxicity at concentrations > 50 μg/ml, while we found significant toxicity with cobalt and nickel for most tested concentrations. Diverse cell-specific toxic effects were observed, with C2C12 being relatively resistant to tungsten-alloy mediated toxic impact. Tungsten-alloy, but not tungsten, caused almost complete dephosphorylation of H3-Ser10 in C2C12 and hippocampal primary neuronal cultures with H3-hypoacetylation in C2C12. Dramatic H3-Ser10 dephosphorylation was found in all cobalt treated cultures with a decrease in H3 pan-acetylation in C2C12, SKNMC and HEK293. Trimethylation of H3-K4 was not affected. Both tungsten-alloy and cobalt mediated H3-Ser10 dephosphorylation were reversed with BAPTA-AM, highlighting the role of intracellular calcium, confirmed with 2-photon calcium imaging. In summary, our results for the first time reveal epigenetic modifications triggered by tungsten-alloy exposure in C2C12 and hippocampal primary neuronal cultures suggesting the underlying synergistic effects of tungsten, nickel and cobalt mediated by changes in intracellular calcium homeostasis and

  2. Laser surface alloying of aluminium-transition metal alloys

    International Nuclear Information System (INIS)

    Almeida, A.; Vilar, R.

    1998-01-01

    Laser surface alloying has been used as a tool to produce hard and corrosion resistant Al-transition metal (TM) alloys. Cr and Mo are particularly interesting alloying elements to produce stable high-strength alloys because they present low diffusion coefficients and solid solubility in Al. To produce Al-TM surface alloys a two-step laser process was developed: firstly, the material is alloyed using low scanning speed and secondly, the microstructure is modified by a refinement step. This process was used in the production of Al-Cr, Al-Mo and Al-Mo and Al-Nb surface alloys by alloying Cr, Mo or Nb powder into an Al and 7175 Al alloy substrate using a CO 2 laser . This paper presents a review of the work that has been developed at Instituto Superior Tecnico on laser alloying of Al-TM alloy, over the last years. (Author) 16 refs

  3. Effects of primary dicarboxylic acids on microstructure and mechanical properties of sub-microcrystalline Ni-Co alloys

    International Nuclear Information System (INIS)

    Vijayakumar, J.; Mohan, S.; Yadav, S. Sunil

    2011-01-01

    Highlights: → The electrodeposited Ni-Co alloys are mostly used in magnetic sensors and it has good mechanical and corrosion resistance properties. → The effect of dicarboxylic acid leads to preferred (2 0 0) crystalline orientation, this may improve magnetic properties dicarboxylic acid can alter the elemental composition of Ni-Co alloy. → Dicarboxylic acid acts as a good brightner. - Abstract: Nickel-cobalt alloys were deposited from sulfate electrolyte with oxalic, malonic and succinic acids as additives and their microstructure and mechanical properties were studied. The crystal structure, surface morphologies, and chemical composition of coatings were investigated using X-ray diffraction, scanning electron microscope, and energy dispersive spectroscopy. The crystal structure and surface morphology analysis showed that the addition of dicarboxylic acid leads to (2 0 0) crystal face and the surface were more compact and uniform due to the grain refining. Ni 60 -Co 40 alloy was achieved when succinic acid is used as additive.

  4. Influence of Chromium and Molybdenum on the Corrosion of Nickel Based Alloys

    International Nuclear Information System (INIS)

    Hayes, J R; Gray, J; Szmodis, A W; Orme, C A

    2005-01-01

    The addition of chromium and molybdenum to nickel creates alloys with exceptional corrosion resistance in a diverse range of environments. This study examines the complementary roles of Cr and Mo in Ni alloy passivation. Four nickel alloys with varying amounts of chromium and molybdenum were studied in 1 molar salt solutions over a broad pH range. The passive corrosion and breakdown behavior of the alloys suggests that chromium is the primary element influencing general corrosion resistance. The breakdown potential was nearly independent of molybdenum content, while the repassivation potential is strongly dependant on the molybdenum content. This indicates that chromium plays a strong role in maintaining the passivity of the alloy, while molybdenum acts to stabilize the passive film after a localized breakdown event

  5. Investigation of water content in primary upper shield of high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Sumita, Junya; Sawa, Kazuhiro; Mogi, Haruyoshi; Itahashi, Shuuji; Kitami, Toshiyuki; Akutu, Youichi; Fuchita, Yasuhiro; Kawaguchi, Toru; Moriya, Masahiro

    1999-09-01

    A primary upper shield of the High Temperature Engineering Test Reactor (HTTR) is composed of concrete (grout) which is packed into iron frames. The main function of the primary upper shield is to attenuate neutron and gamma ray from the core, that leads to satisfy dose equivalent rate limit of operating floor and stand-pipe room. Water content in the concrete is one of the most important things because it strongly affects neutron-shielding ability. Then, we carried out out-of-pile experiments to investigate relationship between temperature and water content in the concrete. Based on the experimental results, a hydrolysis-diffusion model was developed to investigate water release behavior from the concrete. The model showed that water content used for shielding design in the primary upper shield of the HTTR will be maintained if temperature during operating life is under 110degC. (author)

  6. Vacuum Arc Melting Processes for Biomedical Ni-Ti Shape Memory Alloy

    Directory of Open Access Journals (Sweden)

    Tsai De-Chang

    2015-01-01

    Full Text Available This study primarily involved using a vacuum arc remelting (VAR process to prepare a nitinol shape-memory alloy with distinct ratios of alloy components (nitinol: 54.5 wt% to 57 wt%. An advantage of using the VAR process is the adoption of a water-cooled copper crucible, which effectively prevents crucible pollution and impurity infiltration. Optimising the melting production process enables control of the alloy component and facilitates a uniformly mixed compound during subsequent processing. This study involved purifying nickel and titanium and examining the characteristics of nitinol alloy after alloy melt, including its microstructure, mechanical properties, phase transition temperature, and chemical components.

  7. Fatigue cracking of alloy 600 in simulated steam generator crevice environment

    International Nuclear Information System (INIS)

    Ogundele, G.; Lepik, O.

    1998-01-01

    Investigations were carried out to generate fatigue life (S-N) and near-threshold fatigue crack propagation (da/dN) data to determine the environmental influence on fatigue behavior for Alloy 600 in air, deionized water and in simulated Bruce Nuclear Generating Station 'A' crevice environments under appropriate loading conditions. In the low cycle fatigue regime, the simulated crevice environment did not affect the fatigue life of Alloy 600 under the applied loading conditions. The near-threshold fatigue crack growth rates of Alloy 600 in the simulated crevice environment were significantly lower compared to either pure water or air environments and is believed to be the result of higher crack closure in the crevice environment. (author)

  8. The Primary Origin of Dose Rate Effects on Microstructural Evolution of Austenitic Alloys During Neutron Irradiation

    International Nuclear Information System (INIS)

    Okita, Taira; Sato, Toshihiko; Sekimura, Naoto; Garner, Francis A.; Greenwood, Lawrence R.

    2002-01-01

    The effect of dose rate on neutron-induced microstructural evolution was experimentally estimated. Solution-annealed austenitic model alloys were irradiated at approximately 400 degrees C with fast neutrons at seven different dose rates that vary more than two orders difference in magnitude, and two different doses were achieved at each dose rate. Both cavity nucleation and growth were found to be enhanced at lower dose rate. The net vacancy flux is calculated from the growth rate of cavities that had already nucleated during the first cycle of irradiation and grown during the second cycle. The net vacancy flux was found to be proportional to (dpa/sec) exp (1/2) up to 28.8 dpa and 8.4 x 10 exp (-7) dpa/sec. This implies that mutual recombination dominates point defect annihilation, in this experiment even though point defect sinks such as cavities and dislocations were well developed. Thus, mutual recombination is thought to be the primary origin of the effect of dose rate on microstructural evolution

  9. Finite Element Analysis of Warpage in Laminated Aluminium Alloy Plates for Machining of Primary Aeronautic Parts

    International Nuclear Information System (INIS)

    Reis, A. C.; Moreira Filho, L. A.; Menezes, M. A.

    2007-01-01

    The aim of this paper consists in presenting a method of simulating the warpage in 7xxx series aluminium alloy plates. To perform this simulation finite element software MSC.Patran and MSC.Marc were used. Another result of this analysis will be the influence on material residual stresses induced on the raw material during the rolling process upon the warpage of primary aeronautic parts, fabricated through machining (milling) at Embraer. The method used to determinate the aluminium plate residual stress was Layer Removal Test. The numerical algorithm Modified Flavenot Method was used to convert layer removal and beam deflection in stress level. With such information about the level and profile of residual stresses become possible, during the step that anticipate the manufacturing to incorporate these values in the finite-element approach for modelling warpage parts. Based on that warpage parameter surely the products are manufactured with low relative vulnerability propitiating competitiveness and price

  10. Influence of Cr on the nucleation of primary Al and formation of twinned dendrites in Al–Zn–Cr alloys: Can icosahedral solid clusters play a role?

    International Nuclear Information System (INIS)

    Kurtuldu, Güven; Jarry, Philippe; Rappaz, Michel

    2013-01-01

    The equiaxed solidification of Al–20 wt.% Zn alloys revealed an unexpectedly large number of fine grains which are in a twin, or near-twin, relationship with their nearest neighbors when minute amounts of Cr (1000 ppm) are added to the melt. Several occurrences of neighboring grains sharing a nearly common 〈1 1 0〉 direction with a fivefold symmetry multi-twinning relationship have been found. These findings are a very strong indication that the primary face-centered cubic Al phase forms on either icosahedron quasicrystals or nuclei of the parent stable Al 45 Cr 7 phase, which exhibits several fivefold symmetry building blocks in its large monoclinic unit cell. They are further supported by thermodynamic calculations and by grains sometimes exhibiting orientations compatible with the so-called interlocked icosahedron. These results are important, not only because they provide an explanation of the nucleation of twinned dendrites in Al alloys, a topic that has remained unclear over the past 60 years despite several recent investigations, but also because they identify a so far neglected nucleation mechanism in aluminum alloys, which could also apply to other metallic systems

  11. Extensive tissue-specific transcriptomic plasticity in maize primary roots upon water deficit

    OpenAIRE

    Opitz, Nina; Marcon, Caroline; Paschold, Anja; Malik, Waqas Ahmed; Lithio, Andrew; Brandt, Ronny; Piepho, Hans-Peter; Nettleton, Dan; Hochholdinger, Frank

    2015-01-01

    Water deficit is the most important environmental constraint severely limiting global crop growth and productivity. This study investigated early transcriptome changes in maize (Zea mays L.) primary root tissues in response to moderate water deficit conditions by RNA-Sequencing. Differential gene expression analyses revealed a high degree of plasticity of the water deficit response. The activity status of genes (active/inactive) was determined by a Bayesian hierarchical model. In total, 70% o...

  12. Microstructure and texture development of 7075 alloy during homogenisation

    Science.gov (United States)

    Ghosh, Abhishek; Ghosh, Manojit

    2018-06-01

    The microstructure evolution of Al-Zn-Mg-Cu alloy during homogenisation was studied by optical microscope, field emission scanning electron microscope, energy dispersive X-ray Spectroscopy, differential scanning calorimetry and X-ray diffraction in detailed. It has been found that primary cast structure consisted of primary α (Al), lamellar eutectic structure η Mg(Zn, Cu, Al)2 and a small amount of θ (Al2Cu) phase. A transformation of primary eutectic phase from η Mg(Zn, Cu, Al)2 to S (Al2CuMg) was observed after 6 h of homogenisation treatment. The volume fraction of dendrite network structure and intermetallic phase was decreased with increase in holding time and finally disappeared after 96 h of homogenisation, which is consistent with the results of homogenisation kinetic analysis. Crystallographic texture of this alloy after casting and 96 h of homogenisation was also studied. It was found that casting process led the development of strong Goss, Brass, P and CuT components, while after homogenisation Cube, S and Copper components became predominant. Mechanical tests revealed higher hardness, yield strength and tensile strength for cast materials compared to homogenised alloys due to the presence of coarse micro-segregation of MgZn2 phase. The significant improvement of ductility was observed after 96-h homogenisation, which was attributed to dissolution of second phase particles and grain coarsening. Fracture surfaces of the cast samples indicated the presence of shrinkage porosity and consequently failure occurred in the interdendritic regions or grain boundaries with brittle mode, while homogenised alloys failed under ductile mode as evident by the presence of fine dimple surfaces.

  13. Enhanced low-temperature oxidation of zirconium alloys under irradiation

    International Nuclear Information System (INIS)

    Cox, B.; Fidleris, V.

    1989-01-01

    The linear growth of relatively thick (>300 nm) interference-colored oxide films on zirconium alloy specimens exposed in the Advanced Test Reactor (ATR) coolant at ≤55 o C was unexpected. Initial ideas were that this was a photoconduction effect. Experiments to study photoconduction in thin anodic zirconium oxide (ZrO 2 ) films in the laboratory were initiated to provide background data. It was found that, in the laboratory, provided a high electric field was maintained across the oxide during ultraviolet (UV) irradiation, enhanced growth of oxide occurred in the irradiated area. Similarly enhanced growth could be obtained on thin thermally formed oxide films that were immersed in an electrolyte with a high electric field superimposed. This enhanced growth was found to be caused by the development of porosity in the barrier oxide layer by an enhanced local dissolution and reprecipitation process during UV irradiation. Similar porosity was observed in the oxide films on the ATR specimens. Since it is not thought that a high electric field could have been present in this instance, localized dissolution of fast-neutron primary recoil tracks may be the operative mechanism. In all instances, the specimens attempt to maintain the normal barrier-layer oxide thickness, which causes the additional oxide growth. Similar mechanisms may have operated during the formation of thick loosely adherent, porous oxides in homogeneous reactor solutions under irradiation, and may be the cause of enhanced oxidation of zirconium alloys in high-temperature water-cooled reactors in some water chemistries. (author)

  14. Characterization of cylinder liners produced with hypereutectic Al-Si alloys and investigation of corrosion behaviour in synthetic automotive condensed solution

    International Nuclear Information System (INIS)

    Santos, Hamilta de Oliveira

    2006-01-01

    In the present study four hypereutectic Al-Si alloys, three produced by spray forming and one by casting, were characterized for microhardness, roughness, microstructure, texture and corrosion resistance in a synthetic automotive condensed solution (SACS). Two of the spray formed alloys tested were obtained from cylinder liners and the other was laboratory made. Spray forming involves alloy atomization and droplets deposition on a substrate, previous to the solidification of all of the droplets. This process favours the production of materials with a fine microstructure free of macrosegregation that is related to improved hot workability. The microstructure characterization of the four alloys revealed the presence of porosities in the laboratory made alloy. All the three alloys produced by spray forming showed a homogeneous distribution of primary precipitates. The microstructure of one of the alloys showed eutectic microstructure, indicating that this alloy was fabricated by casting. In the cylinder liners, the surface roughness was measured and the microhardness of all the alloys was also evaluated. Furthermore, the laboratory made alloy was hot and cold rolled. Texture determinations were carried out to investigate the correlation between the alloy type and their fabrication process. The texture investigation indicated that the fine distribution of primary silicon phase in the alloy hindered the development of texture typical of aluminium alloys deformation, even after severe mechanical work, such as those used in the conversion of pre-formed in cylinder liners. The surface roughness results indicated typical characteristics of the surface finishing used, honing or chemical etching. The microhardness results were dependent on the fabrication process used, with higher microhardness associated to the eutectic alloy comparatively to the spray formed ones. All hypereutectic alloys were tested for corrosion resistance using electrochemical impedance spectroscopy in

  15. Tensile and toughness assessment of the procured advanced alloys

    Energy Technology Data Exchange (ETDEWEB)

    Tan, Lizhen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sokolov, Mikhail A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoelzer, David T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Busby, Jeremy T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-11

    Life extension of the existing nuclear reactors imposes irradiation of high fluences to structural materials, resulting in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) to initiate the Advanced Radiation Resistant Materials (ARRM) program, aiming to develop and test degradation resistant alloys from current commercial alloy specifications by 2021 to a new advanced alloy with superior degradation resistance by 2024 in light water reactor (LWR)-relevant environments

  16. Fatigue crack growth characteristics of nitrogen-alloyed type 347 stainless under operating conditions of a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Min, Ki Deuk; Hong, Seok Min; Kim, Dae Whan; Lee, Bong Sang [Korea Atomic Energy Research Institute, Nuclear Materials Safety Research Division, Daejeon (Korea, Republic of); Kim, Seon Jin [Hanyang University, Division of materials science and engineering, Seoul (Korea, Republic of)

    2017-06-15

    The fatigue crack growth behavior of Type 347 (S347) and Type 347N (S347N) stainless steel was evaluated under the operating conditions of a pressurized water reactor (PWR). These two materials showed different fatigue crack growth rates (FCGRs) according to the changes in dissolved oxygen content and frequency. Under the simulated PWR conditions for normal operation, the FCGR of S347N was lower than that of S347 and insensitive to the changes in PWR water conditions. The higher yield strength and better corrosion resistance of the nitrogen-alloyed Type 347 stainless steel might be a main cause of slower FCGR and more stable properties against changes in environmental conditions.

  17. Study on Modified Water Glass Used in High Temperature Protective Glass Coating for Ti-6Al-4V Titanium Alloy

    Directory of Open Access Journals (Sweden)

    Shuang Yang

    2018-04-01

    Full Text Available Sodium silicate water glass was modified with sodium polyacrylate as the binder, the composite slurry used for high-temperature oxidation-resistant coating was prepared by mixing glass powder with good lubrication properties in the binder. The properties of the modified binder and high-temperature oxidation resistance of Ti-6Al-4V titanium alloy coated with composite glass coating were studied by XRD, SEM, EDS, TG-DSC and so on. Results showed that sodium polyacrylate modified water glass could obviously improve the suspension stability of the binder, the pyrolytic carbon in the binder at high temperature could increase the surface tension in the molten glass system, and the composite glass coating could be smooth and dense after heating. Pyrolytic carbon diffused and combined with oxygen in the coating under the heating process to protect the titanium alloy from oxidation. The thickness of the oxide layer was reduced 51% after applying the high-temperature oxidation-resistant coating. The coating also showed a nearly 30% reduction in friction coefficient due to the boundary lubricant regime. During cooling, the coating could be peeled off easily because of the mismatched CTE between the coating and substrate.

  18. Searching for Next Single-Phase High-Entropy Alloy Compositions

    Directory of Open Access Journals (Sweden)

    David E. Alman

    2013-10-01

    Full Text Available There has been considerable technological interest in high-entropy alloys (HEAs since the initial publications on the topic appeared in 2004. However, only several of the alloys investigated are truly single-phase solid solution compositions. These include the FCC alloys CoCrFeNi and CoCrFeMnNi based on 3d transition metals elements and BCC alloys NbMoTaW, NbMoTaVW, and HfNbTaTiZr based on refractory metals. The search for new single-phase HEAs compositions has been hindered by a lack of an effective scientific strategy for alloy design. This report shows that the chemical interactions and atomic diffusivities predicted from ab initio molecular dynamics simulations which are closely related to primary crystallization during solidification can be used to assist in identifying single phase high-entropy solid solution compositions. Further, combining these simulations with phase diagram calculations via the CALPHAD method and inspection of existing phase diagrams is an effective strategy to accelerate the discovery of new single-phase HEAs. This methodology was used to predict new single-phase HEA compositions. These are FCC alloys comprised of CoFeMnNi, CuNiPdPt and CuNiPdPtRh, and HCP alloys of CoOsReRu.

  19. Nuclear fuel element containing particles of an alloyed Zr, Ti, and Ni getter material

    International Nuclear Information System (INIS)

    Grossman, L.N.; Levin, H.A.

    1975-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed. The nuclear fuel element has disposed therein an alloy having the essential components of nickel, titanium and zirconium, and the alloy reacts with water, water vapor and reactive gases at reactor ambient temperatures. The alloy is disposed in the plenum of the fuel element in the form of particles in a hollow gas permeable container having a multiplicity of openings of size smaller than the size of the particles. The openings permit gases and liquids entering the plenum to contact the particles of alloy. The container is preferably held in the spring in the plenum of the fuel element. (Official Gazette)

  20. Effect of various dopant elements on primary graphite growth

    International Nuclear Information System (INIS)

    Valle, N; Theuwissen, K; Lacaze, J; Sertucha, J

    2012-01-01

    Five spheroidal graphite cast irons were investigated, a usual ferritic grade and four pearlitic alloys containing Cu and doped with Sb, Sn and Ti. These alloys were remelted in a graphite crucible, leading to volatilization of the magnesium added for spheroidization and to carbon saturation of the liquid. The alloys were then cooled down and maintained at a temperature above the eutectic temperature. During this step, primary graphite could develop showing various features depending on the doping elements added. The largest effects were that of Ti which greatly reduces graphite nucleation and growth, and that of Sb which leads to rounded agglomerates instead of lamellar graphite. The samples have been investigated with secondary ion mass spectrometry to enlighten distribution of elements in primary graphite. SIMS analysis showed almost even distribution of elements, including Mg and Al (from the inoculant) in the ferritic grade, while uneven distribution was evident in all doped alloys. Investigations are going on to clarify if the uneven distribution is associated with structural defects in the graphite precipitates.

  1. Some recent trends in the use of zirconium alloys for nuclear service

    International Nuclear Information System (INIS)

    Balaramamoorthy, K.

    1992-01-01

    Without any exception nuclear power reactors particularly the water cooled ones, operating in the World use natural or slightly enriched uranium oxide fuel pellets with zirconium alloy cladding. While the zirconium alloys have proven to be successful in their designed usage, a desire for longer lifetimes of core components and increased duty cycle puts more demand on materials performance. This demand has led to more in depth studies of phenomena associated with zirconium alloy corrosion mechanism, fine tuning of the zirconium alloy composition, development of fabrication techniques and to the evaluation of newer zirconium alloys for critical applications. (author). 5 refs., 32 figs

  2. Corrosion behaviour of Mg/Al alloys in high humidity atmospheres

    Energy Technology Data Exchange (ETDEWEB)

    Arrabal, R.; Pardo, A.; Merino, M.C.; Mohedano, M.; Casajus, P. [Facultad de Quimicas, Departamento de Ciencia de Materiales, Universidad Complutense, 28040 Madrid (Spain); Merino, S. [Departamento de Tecnologia Industrial, Universidad Alfonso X El Sabio, Villanueva de la Canada, 28691 Madrid (Spain)

    2011-04-15

    The influence of relative humidity (80-90-98% RH) and temperature (25 and 50 C) on the corrosion behaviour of AZ31, AZ80 and AZ91D magnesium alloys was evaluated using gravimetric measurements. The results were compared with the data obtained for the same alloys immersed in Madrid tap water. The corrosion rates of AZ alloys increased with the RH and temperature and were influenced by the aluminium content and alloy microstructure for RH values above 90%. The initiation of corrosion was localised around the Al-Mn inclusions in the AZ31 alloy and at the centre of the {alpha}-Mg phase in the AZ80 and AZ91D alloys. The {beta}-Mg{sub 17}Al{sub 12} phase acted as a barrier against corrosion. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  3. Special features of nickel-molybdenum alloy electrodeposition onto screen-type cathodes

    International Nuclear Information System (INIS)

    Aleksandrova, G.S.; Varypaev, V.N.

    1982-01-01

    Electrolytic nickel-molybdenum alloy, which has a rather low hydrogen overpotential and high corrosion resistance, is of interest as cathode material in industrial electrolysis. Screen-type electrodes with a nickel-molybdenum coating can be used as nonconsumable cathodes in water-activated magnesium-alloy batteries

  4. Palladium alloys for hydrogen diffusion

    International Nuclear Information System (INIS)

    1977-01-01

    A palladium-base alloy with tin and/or a silicon addition and its use in the production of hydrogen from water via a cycle of chemical reactions, of which the decomposition of HI into H 2 and I 2 is the most important, is described

  5. Comparison of mechanical properties for several electrical spring contact alloys

    International Nuclear Information System (INIS)

    Nordstrom, T.V.

    1976-06-01

    Work was conducted to determine whether beryllium-nickel alloy 440 had mechanical properties which made it suitable as a substitute for the presently used precious metal contact alloys Paliney 7 and Neyoro G, in certain electrical contact applications. Possible areas of applicability for the alloy were where extremely low contact resistance was not necessary or in components encountering elevated temperatures above those presently seen in weapons applications. Evaluation of the alloy involved three major experimental areas: 1) measurement of the room temperature microplastic (epsilon approximately 10 -6 ) and macroplastic (epsilon approximately 10 -3 ) behavior of alloy 440 in various age hardening conditions, 2) determination of applied stress effects on stress relaxation or contact force loss and 3) measurement of elevated temperature mechanical properties and stress relaxation behavior. Similar measurements were also made on Neyoro G and Paliney 7 for comparison. The primary results of the study show that beryllium-nickel alloy 440 is from a mechanical properties standpoint, equal or superior to the presently used Paliney 7 and Neyoro G for normal Sandia requirements. For elevated temperature applications, alloy 440 has clearly superior mechanical properties

  6. Porous Nb-Ti based alloy produced from plasma spheroidized powder

    Directory of Open Access Journals (Sweden)

    Qijun Li

    Full Text Available Spherical Nb-Ti based alloy powder was prepared by the combination of plasma spheroidization and mechanical alloying. Phase constituents, microstructure and surface state of the powder, and pore characteristics of the resulting porous alloy were investigated. The results show that the undissolved W and V in the mechanically alloyed powder is fully alloyed after spheroidization, and single β phase is achieved. Particle size of the spheroidized powder is in the range of 20–110 μm. With the decrease of particle size, a transformation from typical dendrite solidification structure to fine cell microstructure occurs. The surface of the spheroidized powder is coated by a layer of oxides consisting mainly of TiO2 and Nb2O5. Probabilities of sinter-neck formation and particle coalescence increases with increasing sintering temperature. Porous skeleton with relatively homogeneous pore distribution and open pore channel is formed after vacuum sintering at 1700 °C, and the porosity is 32%. The sintering kinetic analysis indicates that grain boundary diffusion is the primary mass transport mechanism during sintering process. Keywords: Powder metallurgy, Nb-Ti based alloy, Porous material, Mechanical alloying, Plasma spheroidizing, Solidification microstructure

  7. Age hardening in mechanically alloyed Al-Mg-Li-C-O

    Energy Technology Data Exchange (ETDEWEB)

    Papazian, J.M. (Corporate Research Center, Grumman Corporation, Bethpage, NY (USA)); Gilman, P. (Allied-Signal Inc., Morristown, NJ (USA))

    1990-05-01

    The age-hardening behavior of a series of mechanically alloyed Al-Mg-Li-C-O alloys containing 3.0-4.0 wt.% Mg and 1.3-1.75 wt.% Li was studied using hardness tests, differential scanning calorimetry (DSC) and transmission electron microscopy. The hardness tests showed an increased hardness after 100degC aging in all the alloys containing at least 1.5 at.% Li. Likewise, the calorimetry results showed the presence of pronounced precipitate dissolution peaks in these same alloys after 100degC aging. The volume fraction of precipitates formed (as measured by the dissolution enthalpies of the DSC peaks) increased systematically with increasing solute content. Transmission electron microscopy after 100 and 190degC aging showed images and diffraction spots similar to those of {delta}' (Al{sub 3}Li). Comparison of the DSC results with results from binary Al-Li and Al-Mg alloys indicated that the precipitates formed in the Al-Mg-Li-C-O alloys were similar to those formed in binary Al-Li alloys, and that the primary role of the magnesium was to lower the solid solubility of lithium. (orig.).

  8. Nodular Corrosion Characteristics of Zirconium Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun Gil; Jeong, Y. H.; Park, S. Y.; Lee, D. J

    2003-01-15

    This study was reported the effect of the nodular corrosion on the nuclear reactor environmental along with metallurgical influence, also suggested experimental scheme related to evaluate nodular corrosion characteristics of Zr-1 Nb alloy. Remedial strategies against the nodular corrosion should firstly develop plan to assess the effect of the water quality condition (Oxygen, Hydrogen) as well as the boiling on the nodular corrosion, secondarily establish plan to control heat treatment process to keep a good resistance on nodular corrosion in Zr-1Nb alloy as former western reactor did.

  9. High energy beam thermal processing of alpha zirconium alloys and the resulting articles

    International Nuclear Information System (INIS)

    Sabol, G.P.; McDonald, S.G.; Nurminen, J.I.

    1983-01-01

    Alpha zirconium alloy fabrication methods and resultant products exhibiting improved high temperature, high pressure steam corrosion resistance. The process, according to one aspect of this invention, utilizes a high energy beam thermal treatment to provide a layer of beta treated microstructure on an alpha zirconium alloy intermediate product. The treated product is then alpha worked to final size. According to another aspect of the invention, high energy beam thermal treatment is used to produce an alpha annealed microstructure in a Zircaloy alloy intermediate size or final size component. The resultant products are suitable for use in pressurized water and boiling water reactors

  10. Characterization of oxide scales grown on alloy 310S stainless steel after long term exposure to supercritical water at 500 °C

    Energy Technology Data Exchange (ETDEWEB)

    Behnamian, Yashar, E-mail: behnamia@ualberta.ca [Department of Chemical and Materials Engineering, University of Alberta, Edmonton, Alberta T6G 1H9 (Canada); Mostafaei, Amir [Department of Mechanical Engineering and Materials Science, University of Pittsburgh, Pittsburgh, PA 15261 (United States); Kohandehghan, Alireza [Department of Chemical and Materials Engineering, University of Alberta, Edmonton, Alberta T6G 1H9 (Canada); Amirkhiz, Babak Shalchi [Canmet MATERIALS, Natural Resources Canada, Hamilton, Ontario L8P 0A5 (Canada); Serate, Daniel [Department of Chemical and Materials Engineering, University of Alberta, Edmonton, Alberta T6G 1H9 (Canada); Zheng, Wenyue [Canmet MATERIALS, Natural Resources Canada, Hamilton, Ontario L8P 0A5 (Canada); Guzonas, David [Canadian Nuclear Laboratories, Chalk River Laboratories, Chalk River, Ontario K0J 1J0 (Canada); Chmielus, Markus [Department of Mechanical Engineering and Materials Science, University of Pittsburgh, Pittsburgh, PA 15261 (United States); Chen, Weixing, E-mail: Weixing@ualberta.ca [Department of Chemical and Materials Engineering, University of Alberta, Edmonton, Alberta T6G 1H9 (Canada); Luo, Jing Li, E-mail: Jingli.luo@ualberta.ca [Department of Chemical and Materials Engineering, University of Alberta, Edmonton, Alberta T6G 1H9 (Canada)

    2016-10-15

    The oxide scale grown of static capsules made of alloy 310S stainless steel was investigated by exposure to the supercritical water at 500 °C 25 MPa for various exposure times up to 20,000 h. Characterization techniques such as X-ray diffraction, scanning/transmission electron microscopy, energy dispersive spectroscopy, and fast Fourier transformation were employed on the oxide scales. The elemental and phase analyses indicated that long term exposure to the SCW resulted in the formation of scales identified as Fe{sub 3}O{sub 4} (outer layer), Fe-Cr spinel (inner layer), Cr{sub 2}O{sub 3} (transition layer) on the substrate, and Ni-enrichment (chrome depleted region) in the alloy 310S. It was found that the layer thickness and weight gain vs. exposure time followed parabolic law. The oxidation mechanism and scales grown on the alloy 310S stainless steel exposed to SCW are discussed. - Highlights: •Oxidation of alloy 310S stainless steel exposed to SCW (500 °C/25 MPa) •The layer thickness and weight gain vs. exposure time followed parabolic law. •Oxide layers including Fe{sub 3}O{sub 4} (outer), Fe-Cr spinel (inner) and Cr{sub 2}O{sub 3} (transition) •Ni element is segregated by the selective oxidation of Cr.

  11. Wintertime Arctic Ocean sea water properties and primary marine aerosol concentrations

    Directory of Open Access Journals (Sweden)

    J. Zábori

    2012-11-01

    Full Text Available Sea spray aerosols are an important part of the climate system through their direct and indirect effects. Due to the diminishing sea ice, the Arctic Ocean is one of the most rapidly changing sea spray aerosol source areas. However, the influence of these changes on primary particle production is not known.

    In laboratory experiments we examined the influence of Arctic Ocean water temperature, salinity, and oxygen saturation on primary particle concentration characteristics. Sea water temperature was identified as the most important of these parameters. A strong decrease in sea spray aerosol production with increasing water temperature was observed for water temperatures between −1°C and 9°C. Aerosol number concentrations decreased from at least 1400 cm−3 to 350 cm−3. In general, the aerosol number size distribution exhibited a robust shape with one mode close to dry diameter Dp 0.2 μm with approximately 45% of particles at smaller sizes. Changes in sea water temperature did not result in pronounced change of the shape of the aerosol size distribution, only in the magnitude of the concentrations. Our experiments indicate that changes in aerosol emissions are most likely linked to changes of the physical properties of sea water at low temperatures. The observed strong dependence of sea spray aerosol concentrations on sea water temperature, with a large fraction of the emitted particles in the typical cloud condensation nuclei size range, provide strong arguments for a more careful consideration of this effect in climate models.

  12. Production of titanium alloys with uniform distribution of heat resisting metals

    International Nuclear Information System (INIS)

    Reznichenko, V.A.; Goncharenko, T.V.; Khalimov, F.B.; Vojtechova, E.A.

    1976-01-01

    Consideration is given to the process of the formation of a titanium sponge alloyed with niobium or tantalum, in the joint metallic reduction of titanium, niobium and tantanum chlorides. A percentage composition of the phases observed and the structure of the alloyed sponge have been studied. It is shown that after one remelting operation of the alloyed sponge the alloys of titanium with niobium and tantalum have a uniform component distribution. At the stage of chloride reduction there appear solid solutions based on titanium and an alloying component. The stage of vacuum separation of the reaction mass is associated with a mutual dissolution of the primary phases and the formation of the solid solutions of the alloyed titanium sponge, which, by their composition, are close to the desired alloy composition. The principal features of the formation of a titanium sponge alloyed with niobium and tantalum are in a perfect agreemet with those typical of Ti-Mo and Ti-W sponges, therefore it can be assumed that these features will be also common to the other cases of the metallic reduction of titanium and refractory metals chlorides

  13. Production of titanium alloys with uniform distribution of heat resisting metals

    Energy Technology Data Exchange (ETDEWEB)

    Reznichenko, V A; Goncharenko, T V; Khalimov, F B; Voitechova, E A

    1976-01-01

    Consideration is given to the process of the formation of a titanium sponge alloyed with niobium or tantalum, in the joint metallic reduction of titanium, niobium and tantanum chlorides. A percentage composition of the phases observed and the structure of the alloyed sponge have been studied. It is shown that after one remelting operation of the alloyed sponge the alloys of titanium with niobium and tantalum have a uniform component distribution. At the stage of chloride reduction there appear solid solutions based on titanium and an alloying component. The stage of vacuum separation of the reaction mass is associated with a mutual dissolution of the primary phases and the formation of the solid solutions of the alloyed titanium sponge, which, by their composition, are close to the desired alloy composition. The principal features of the formation of a titanium sponge alloyed with niobium and tantalum are in a perfect agreemet with those typical of Ti-Mo and Ti-W sponges, therefore it can be assumed that these features will be also common to the other cases of the metallic reduction of titanium and refractory metals chlorides.

  14. Primary water chemistry of VVERs-operating experience

    International Nuclear Information System (INIS)

    Kysela, Jan; Zmitko, Milan; Petrecky, Igor

    1998-01-01

    VVER units are operated in mixed boron-potassium-ammonia water chemistry. Several modifications of the water chemistry, differing in boron-potassium co-ordination and in the way how hydrogen concentration is produced and maintain in the coolant, is used. From the operational experience point of view VVER units do not show any significant problems connected with the primary coolant chemistry. The latest results indicate that dose rate levels are slowly returning to the former ones. An improvement of the radiation situation observed last two years is supported by the surface activity measurements. However, the final conclusion on the radiation situation can be made only after evaluation of the several following cycles. Further investigation is also needed to clarify a possible effect of modified water chemistry and shut-down chemistry on radioactivity build-up and dose rate level at Dukovany units. Structure materials composition has a significant effect on radiation situation in the units. It concerns mainly of cobalt content in SG material. There is no clear evidence of possible effect of the SG shut-down regimes on the radiation situation in the units even if the dose rate and surface activity data show wide spread for the individual reactor loops. (S.Y.)

  15. Fatigue of carbon and low-alloy steels in LWR environments

    International Nuclear Information System (INIS)

    Chopra, O.K.; Michaud, W.F.; Shack, W.J.

    1994-01-01

    Fatigue tests have been conducted on A106-Gr B carbon steel and A533-Gr B low-alloy steel to evaluate the effects of an oxygenated-water environment on the fatigue life of these steels. For both steels, environmental effects are modest in PWR water at all strain rates. Fatigue data in oxygenated water confirm the strong dependence of fatigue life on dissolved oxygen (DO) and strain rate. The effect of strain rate on fatigue life saturates at some low value, e.g., between 0.0004 and 0.001%/s in oxygenated water with ∼0.8 ppm DO. The data suggest that the saturation value of strain rate may vary with DO and sulfur content of the steel. Although the cyclic stress-strain and cyclic-hardening behavior of carbon and low-alloy steels is distinctly different, the degradation of fatigue life of these two steels with comparable sulfur levels is similar. The carbon steel exhibits pronounced dynamic strain aging, whereas strain-aging effects are modest in the low-alloy steel. Environmental effects on nucleation of fatigue crack have also been investigated. The results suggest that the high-temperature oxygenated water has little or not effect on crack nucleation

  16. Vacuum Arc Melting Processes for Biomedical Ni-Ti Shape Memory Alloy

    OpenAIRE

    Tsai De-Chang; Chiang Chen-Hsueh

    2015-01-01

    This study primarily involved using a vacuum arc remelting (VAR) process to prepare a nitinol shape-memory alloy with distinct ratios of alloy components (nitinol: 54.5 wt% to 57 wt%). An advantage of using the VAR process is the adoption of a water-cooled copper crucible, which effectively prevents crucible pollution and impurity infiltration. Optimising the melting production process enables control of the alloy component and facilitates a uniformly mixed compound during subsequent processi...

  17. Numerical evaluation of weld overlay applied to a pressurized water reactor nozzle mock-up

    Energy Technology Data Exchange (ETDEWEB)

    Rabello, Emerson G.; Silva, Luiz L.; Gomes, Paulo T.V., E-mail: egr@cdtn.b, E-mail: silvall@cdtn.b, E-mail: gomespt@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Integridade Estrutural

    2011-07-01

    The primary water stress corrosion cracking (PWSCC) is a major mechanism of failure in the primary circuit of PWR type nuclear power plants. The PWSCC is associated with the presence of corrosive environment, the susceptibility to corrosion cracking of the materials involved and the tensile stresses presence. Residual stresses generated during dissimilar materials welding can contribute to PWSCC. An alternative to the PWSCC mitigation is the application of external weld layers in the regions of greatest susceptibility to corrosion cracking. This process, called Weld Overlay (WOL), has been widely used in regions of dissimilar weld (low alloy steel and stainless steel with nickel alloy addition) of nozzles and pipes on the primary circuit in order to promote internal compressive stresses on the wall of these components. This paper presents the steps required to the numerical stress evaluation (by finite element method) during the dissimilar materials welding as well as application of Weld Overlay process in a nozzle mock-up. Thus, one can evaluate the effectiveness of the application of weld overlay process to internal compressive stress generation on the wall nozzle. (author)

  18. Numerical evaluation of weld overlay applied to a pressurized water reactor nozzle mock-up

    International Nuclear Information System (INIS)

    Rabello, Emerson G.; Silva, Luiz L.; Gomes, Paulo T.V.

    2011-01-01

    The primary water stress corrosion cracking (PWSCC) is a major mechanism of failure in the primary circuit of PWR type nuclear power plants. The PWSCC is associated with the presence of corrosive environment, the susceptibility to corrosion cracking of the materials involved and the tensile stresses presence. Residual stresses generated during dissimilar materials welding can contribute to PWSCC. An alternative to the PWSCC mitigation is the application of external weld layers in the regions of greatest susceptibility to corrosion cracking. This process, called Weld Overlay (WOL), has been widely used in regions of dissimilar weld (low alloy steel and stainless steel with nickel alloy addition) of nozzles and pipes on the primary circuit in order to promote internal compressive stresses on the wall of these components. This paper presents the steps required to the numerical stress evaluation (by finite element method) during the dissimilar materials welding as well as application of Weld Overlay process in a nozzle mock-up. Thus, one can evaluate the effectiveness of the application of weld overlay process to internal compressive stress generation on the wall nozzle. (author)

  19. Effect of two-stage sintering process on microstructure and mechanical properties of ODS tungsten heavy alloy

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyong H. [Department of Materials Science and Engineering, Korea Advanced Institute of Science and Technology, 373-1 Kusong-dong, Yusong-gu, Taejon 305-701 (Korea, Republic of); Cha, Seung I. [International Center for Young Scientists, National Institute for Materials Science 1-1, Namiki, Tsukuba 305-0044 (Japan); Ryu, Ho J. [DUPIC, Korea Atomic Energy Research Institute, 150 Deokjin-dong, Yusong-gu, Taejon 305-353 (Korea, Republic of); Hong, Soon H. [Department of Materials Science and Engineering, Korea Advanced Institute of Science and Technology, 373-1 Kusong-dong, Yusong-gu, Taejon 305-701 (Korea, Republic of)], E-mail: shhong@kaist.ac.kr

    2007-06-15

    Oxide dispersion strengthened (ODS) tungsten heavy alloys have been considered as promising candidates for advanced kinetic energy penetrator due to their characteristic fracture mode compared to conventional tungsten heavy alloy. In order to obtain high relative density, the ODS tungsten heavy alloy needs to be sintered at higher temperature for longer time, however, induces growth of tungsten grains. Therefore, it is very difficult to obtain controlled microstructure of ODS tungsten heavy alloy having fine tungsten grains with full densification. In this study, two-stage sintering process, consisted of primary solid-state sintering and followed by secondary liquid phase sintering, was introduced for ODS tungsten heavy alloys. The mechanically alloyed 94W-4.56Ni-1.14Fe-0.3Y{sub 2}O{sub 3} powders are solid-state sintered at 1300-1450 deg. C for 1 h in hydrogen atmosphere, and followed by liquid phase sintering temperature at 1465-1485 deg. C for 0-60 min. The microstructure of ODS tungsten heavy alloys showed high relative density above 97%, with contiguous tungsten grains after primary solid-state sintering. The microstructure of solid-state sintered ODS tungsten heavy alloy was changed into spherical tungsten grains embedded in W-Ni-Fe matrix during secondary liquid phase sintering. The two-stage sintered ODS tungsten heavy alloy from mechanically alloyed powders showed finer microstructure and higher mechanical properties than conventional liquid phase sintered alloy. The mechanical properties of ODS tungsten heavy alloys are dependent on the microstructural parameters such as tungsten grain size, matrix volume fraction and tungsten/tungsten contiguity, which can be controlled through the two-stage sintering process.

  20. Dissolved stable noble gas measurements from primary water of Paks NPP

    International Nuclear Information System (INIS)

    Palcsu, L.; Molnar, M.; Szanto, Zs.; Svingor, E.; Futo, I.; Pinter, T.

    2001-01-01

    A sampling and measuring method of noble gases from the primary water circuit of a VVER type NPP was developed to provide relevant information about the kilter of heating rods and detailed additional information about some working parameters. The helium concentrations and 3 He/ 4 He ratios was used to estimate the content of tritium and alpha emitting isotopes of the primary water. By argon content measurements the air penetration and the required hydrazine amount for the oxygen absorption could be estimated with high accuracy. Continuous monitoring of the concentration and isotope ratios of Xe and Kr in the dissolved gas is proved to be a good tool for high sensitivity detection of small leakage of fuel elements. In case of block-3 xenon surplus was detected. The results indicate possible leakage of fuel rods.(author)

  1. Imparting passivity to vapor deposited magnesium alloys

    Science.gov (United States)

    Wolfe, Ryan C.

    Magnesium has the lowest density of all structural metals. Utilization of low density materials is advantageous from a design standpoint, because lower weight translates into improved performance of engineered products (i.e., notebook computers are more portable, vehicles achieve better gas mileage, and aircraft can carry more payload). Despite their low density and high strength to weight ratio, however, the widespread implementation of magnesium alloys is currently hindered by their relatively poor corrosion resistance. The objective of this research dissertation is to develop a scientific basis for the creation of a corrosion resistant magnesium alloy. The corrosion resistance of magnesium alloys is affected by several interrelated factors. Among these are alloying, microstructure, impurities, galvanic corrosion effects, and service conditions, among others. Alloying and modification of the microstructure are primary approaches to controlling corrosion. Furthermore, nonequilibrium alloying of magnesium via physical vapor deposition allows for the formation of single-phase magnesium alloys with supersaturated concentrations of passivity-enhancing elements. The microstructure and surface morphology is also modifiable during physical vapor deposition through the variation of evaporation power, pressure, temperature, ion bombardment, and the source-to-substrate distance. Aluminum, titanium, yttrium, and zirconium were initially chosen as candidates likely to impart passivity on vapor deposited magnesium alloys. Prior to this research, alloys of this type have never before been produced, much less studied. All of these metals were observed to afford some degree of corrosion resistance to magnesium. Due to the especially promising results from nonequilibrium alloying of magnesium with yttrium and titanium, the ternary magnesium-yttrium-titanium system was investigated in depth. While all of the alloys are lustrous, surface morphology is observed under the scanning

  2. Intergranular stress corrosion cracking of low alloy and carbon steels in high temperature pure water

    International Nuclear Information System (INIS)

    Tsubota, M.; Sakamoto, H.; Tsuzuki, R.

    1993-01-01

    Stress corrosion cracking (SCC) behavior of low alloy steels (A508 and SNCM630) and a carbon steel (SGV480) in high temperature water has been examined with relation to the heat treatment condition, including a long time aging, and the mechanical properties. Intergranular stress corrosion cracking (IGSCC) as observed in the highly hardened specimens, and there was observed in the highly hardened specimens, and there was observed in the highly hardened specimens, and there was observed a close relationship between hardness and SCC susceptibility. From the engineering point of view, it was concluded that adequate SR (stress relief) or tempering heat treatment is necessary to avoid the IGSCC of the welded structures made of low alloy and carbon steels. A508 heat treated with specified quench and temper did not show the SCC susceptibility, even after aging 10000 hours at 350, 400 and 450 degrees C. Tensile properties corresponding to the critical hardness for SSC susceptibility coincided with the values at the 'necking point' in the true stress-strain curve. Ductile-brittle transition observed in the fracture toughness test also occurred at around the critical hardness for SCC susceptibility. Therefore, it was conjectured that the limitation of plasticity was an absolute cause for the SCC susceptibility of the steels

  3. High strength tungsten heavy alloys with molybdenum additions

    International Nuclear Information System (INIS)

    Bose, A.; Sims, D.M.; German, R.M.

    1987-01-01

    Tungsten heavy alloys are candidates for numerous applications based on the unique combination of high density, high strength, and high ductility coupled with excellent machinability. Though there has been considerable research on heavy alloys, the primary focus has been on the ductility. These alloys are well suited for ballistic uses due to their high densities and it is expected that for superior ballistic performance, a high hardness, high strength and moderate ductility alloy would be ideal. The major goal of this investigation was to obtain heavy alloys with hardness greater than HRA 72. It is evident from the phase diagrams that molybdenum, which goes into solution in tungsten, nickel and iron, could act as a potential strengthening addition. With this in view, tungsten heavy alloys with molybdenum additions were fabricated from mixed elemental powders. A baseline composition of 90W-7Ni-3Fe was chosen to its good elongation and moderate strength. The molybdenum additions were made by replacing the tungsten. Compared to the baseline properties with no molybdenum addition, the strength and hardness showed a continuous increase with molybdenum addition. The ductility of the alloy continued to decrease with increasing molybdenum content, but even with 16% wt. % molybdenum of the elongation was still around 6%. An interesting facet of these alloying additions is the grain refinement that is brought about by adding to molybdenum to the system. The grain refinement is related to the lower solubility of tunbsten in the matrix due to partial displacement by molybdenum

  4. Phase-field simulations of dendrite morphologies and selected evolution of primary {alpha}-Mg phases during the solidification of Mg-rich Mg-Al-based alloys

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Mingyue [Key Laboratory for Advanced Materials Processing Technology, Ministry of Education, Department of Mechanical Engineering, Tsinghua University, Beijing 100084 (China); Jing, Tao; Liu, Baicheng [Key Laboratory for Advanced Materials Processing Technology, Ministry of Education, Department of Mechanical Engineering, Tsinghua University, Beijing 100084 (China)

    2009-10-15

    A formulation of solid-liquid interfacial thermodynamic and kinetic anisotropic characteristics for hexagonal close-packed metals is proposed. The two- and three-dimensional dendritic growth of primary Mg in undercooled Mg-Al alloy melts is modeled using the phase-field method, based on a combination of crystallographic lattice symmetry and experimental observations. The morphologies of three-dimensional dendrites are obtained and the calculated results show intricately hierarchical branched structures. The excess free energy of the solution system is based on the Redlich-Kister model.

  5. Phase-field simulations of dendrite morphologies and selected evolution of primary α-Mg phases during the solidification of Mg-rich Mg-Al-based alloys

    International Nuclear Information System (INIS)

    Wang, Mingyue; Jing, Tao; Liu, Baicheng

    2009-01-01

    A formulation of solid-liquid interfacial thermodynamic and kinetic anisotropic characteristics for hexagonal close-packed metals is proposed. The two- and three-dimensional dendritic growth of primary Mg in undercooled Mg-Al alloy melts is modeled using the phase-field method, based on a combination of crystallographic lattice symmetry and experimental observations. The morphologies of three-dimensional dendrites are obtained and the calculated results show intricately hierarchical branched structures. The excess free energy of the solution system is based on the Redlich-Kister model.

  6. Contribution of the low cycle fatigue on ultra high purity Ni-Cr-Fe alloys and on Ni monocrystals to the understanding of the hydrogen role in stress corrosion cracking for the alloys 600 and 690

    International Nuclear Information System (INIS)

    Renaudot, N.

    1999-06-01

    We discuss the role of hydrogen in cracking of Ni base alloys used for pressurised water reactor (PWR) primary tubes (alloy 600 and 690). Cracking can be explained by a Stress Corrosion Cracking (SCC) phenomenon. For this purpose, Low cycle fatigue (R = - 1) under cathodic charging at room temperature is conducted to study hydrogen effects on propagation of cracks mechanically initiated by the formation of Persistent Slip Bands (PSB). Low cycle fatigue on Ultra High Purity specimens (Ni, alloy 600 and 690) reveals the very important hydrogen effect on crack propagation rate, whatever the Cr content in the Ni base alloy. If Cr seems to have an effect over-hydrogen penetration in specimens (by a protective film formation), it have no beneficial effect when hydrogen have diffused ahead of a crack tip. Propagation rates (transgranular or intergranular) are highly increased, no matter of the absence of impurities like sulphur. Then, in PWR, the difference in the behaviour of alloy 600 and 690 could be due to a slower microcrack propagation rate for alloy 690. Protective films could play an important role in this difference, which is to study. Low cycle fatigue on Ni single crystals oriented for single slip shows, for the first time on bulk specimen, a macroscopic softening which can be explained. by hydrogen-dislocation interactions. Moreover, a simple quantitative model based on these interactions results in the same softening as the one observed experimentally. These results allow to validate experimentally one of the most important steps in the 'Corrosion Enhanced Plasticity (CEP) model', i.e. the softening ahead of a stress corrosion crack tip by hydrogen dislocation interactions. This is of importance because this model can explain cracking in numerous FCC materials-environment couple. (author)

  7. Hydrothermal treatment of titanium alloys for the enhancement of osteoconductivity

    Energy Technology Data Exchange (ETDEWEB)

    Zuldesmi, Mansjur, E-mail: mzuldesmi@yahoo.com [Department of Materials Science & Engineering, Graduate School of Engineering, Nagoya University, Nagoya (Japan); Department of Mechanical Engineering, Manad State University (UNIMA) (Indonesia); Waki, Atsushi [Department of Materials Science & Engineering, Graduate School of Engineering, Nagoya University, Nagoya (Japan); Kuroda, Kensuke; Okido, Masazumi [EcoTopia Science Institute, Nagoya University, Nagoya (Japan)

    2015-04-01

    The surface wettability of implants is a crucial factor in their osteoconductivity because it influences the adsorption of cell-attached proteins onto the surface. In this study, a single-step hydrothermal surface treatment using distilled water at a temperature of 180 °C for 3 h was applied to titanium (Ti) and its alloys (Ti–6Al–4V, Ti–6Al–7Nb, Ti–29Nb–13Ta–4.6Zr, Ti–13Cr–1Fe–3Al; mass%) and compared with as-polished Ti implants and with implants produced by anodizing Ti in 0.1 M of H{sub 3}PO{sub 4} with applied voltages from 0 V to 150 V at a scanning rate of 0.1 V s{sup −1}. The surface-treated samples were stored in a five time phosphate buffered saline (× 5 PBS(−)) solution to prevent increasing the water contact angle (WCA) with time. The surface characteristics were evaluated using scanning electron microscopy, X-ray diffraction, X-ray photoelectron spectroscopy, Auger electron spectroscopy, surface roughness, and contact angle measurement using a 2 μL droplet of distilled water. The relationship between WCA and osteoconductivity at various surface modifications was examined using in vivo tests. The results showed that a superhydrophilic surface with a WCA ≤ 10° and a high osteoconductivity (R{sub B–I}) of up to 50% in the cortical bone part, about four times higher than the as-polished Ti and Ti alloys, were provided by the combination of the hydrothermal surface treatment and storage in × 5 of PBS(−). - Highlights: • Hydrothermal treatment in distilled water was applied to titanium alloys. • Surface characteristics and osteoconductivity by in vivo test were evaluated. • Water contact angles of titanium alloys were decreased by hydrothermal treatment. • Osteoconductivity of titanium alloys improved notably by hydrothermal treatment after stored in × 5 of PBS (−)

  8. A new model for the in-reactor corrosion of zirconium alloys

    International Nuclear Information System (INIS)

    Cox, B.

    1997-01-01

    Previous models for the in-reactor corrosion of zirconium alloys have assumed that the mechanism is a completely solid-state diffusion process, determined by the growth and breakdown of the protective oxide film. In-reactor kinetics have been related to out-reactor kinetics with the oxide-metal interface temperature calculated from effects of heat flux. Recent experimental results have suggested that oxide dissolution and reprecipitation may be a major process leading to the formation of thick porous oxide films in-reactor. The model described here is based on the dissolution of primary recoil tracks in the oxide as the primary process distinguishing in-reactor from out-reactor corrosion. The consequences of such a model would be a very different microscopic morphology of in-reactor and out-reactor thick films, a significant irradiation effect on non-heat transfer surfaces, and a change in the kinetics of the overall process. This model should be equally applicable to PWR and BWR water chemistries because of the amphoteric nature of ZrO 2 , and the effects of LiOH should operate by an essentially identical mechanism. A reciprocal rate equation should fit these processes and with additive terms seems capable of accommodating all water chemistry effects, except for discontinuous processes such as nodular corrosion. (author). 60 refs, 6 figs

  9. UV radiation and natural fluorescence linked primary production in Antarctic waters

    Digital Repository Service at National Institute of Oceanography (India)

    LokaBharathi, P.A.; KrishnaKumari, L.; Bhattathiri, P.M.A.; Chandramohan, D.

    Primary productivity and chlorophyll values have been measured using an underwater profiling radiometer for the first time in the waters around Indian Antarctic Station (70°46'S & 11°44'E) in the summer of 1994. The profiles include natural...

  10. Production of Magnesium and Aluminum-Magnesium Alloys from Recycled Secondary Aluminum Scrap Melts

    Science.gov (United States)

    Gesing, Adam J.; Das, Subodh K.; Loutfy, Raouf O.

    2016-02-01

    An experimental proof of concept was demonstrated for a patent-pending and trademark-pending RE12™ process for extracting a desired amount of Mg from recycled scrap secondary Al melts. Mg was extracted by electrorefining, producing a Mg product suitable as a Mg alloying hardener additive to primary-grade Al alloys. This efficient electrorefining process operates at high current efficiency, high Mg recovery and low energy consumption. The Mg electrorefining product can meet all the impurity specifications with subsequent melt treatment for removing alkali contaminants. All technical results obtained in the RE12™ project indicate that the electrorefining process for extraction of Mg from Al melt is technically feasible. A techno-economic analysis indicates high potential profitability for applications in Al foundry alloys as well as beverage—can and automotive—sheet alloys. The combination of technical feasibility and potential market profitability completes a successful proof of concept. This economical, environmentally-friendly and chlorine-free RE12™ process could be disruptive and transformational for the Mg production industry by enabling the recycling of 30,000 tonnes of primary-quality Mg annually.

  11. Sliding wear and friction behavior of zirconium alloy with heat-treated Inconel718

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J.H., E-mail: kimjhoon@cnu.ac.kr [Dept. of Mechanical Design Engineering, Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon 305-764 (Korea, Republic of); Park, J.M. [Dept. of Mechanical Design Engineering, Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon 305-764 (Korea, Republic of); Park, J.K.; Jeon, K.L. [Nuclear Fuel Technology Department, Korea Nuclear Fuel, 1047 Daedukdae-ro, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2014-04-01

    In water-cooled nuclear reactors, the sliding of fuel rod can lead to severe wear and it is an important issue to sustain the structural integrity of nuclear reactor. In the present study, sliding wear behavior of zirconium alloy in dry and water environment using Pin-On-Disk sliding wear tester was investigated. Wear resistance of zirconium alloy against heat-treated Inconel718 pin was examined at room temperature. Sliding wear tests were carried out at different sliding distance, axial load and sliding speed based on ASTM (G99-05). The results of these experiments were verified with specific wear rate and coefficient of friction. The micro-mechanisms responsible for wear in zirconium alloy were identified to be microcutting and microcracking in dry environment. Moreover, micropitting and delamination were observed in water environment.

  12. Thermodynamic analysis of binary Fe{sub 85}B{sub 15} to quinary Fe{sub 85}Si{sub 2}B{sub 8}P{sub 4}Cu{sub 1} alloys for primary crystallizations of α-Fe in nanocrystalline soft magnetic alloys

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, A., E-mail: takeuchi@imr.tohoku.ac.jp; Zhang, Y.; Takenaka, K.; Makino, A. [Institute for Materials Research, Tohoku University, Sendai 980-8577 (Japan)

    2015-05-07

    Fe-based Fe{sub 85}B{sub 15}, Fe{sub 84}B{sub 15}Cu{sub 1}, Fe{sub 82}Si{sub 2}B{sub 15}Cu{sub 1}, Fe{sub 85}Si{sub 2}B{sub 12}Cu{sub 1}, and Fe{sub 85}Si{sub 2}B{sub 8}P{sub 4}Cu{sub 1} (NANOMET{sup ®}) alloys were experimental and computational analyzed to clarify the features of NANOMET that exhibits high saturation magnetic flux density (B{sub s}) nearly 1.9 T and low core loss than conventional nanocrystalline soft magnetic alloys. The X-ray diffraction analysis for ribbon specimens produced experimentally by melt spinning from melts revealed that the samples were almost formed into an amorphous single phase. Then, the as-quenched samples were analyzed with differential scanning calorimeter (DSC) experimentally for exothermic enthalpies of the primary and secondary crystallizations (ΔH{sub x1} and ΔH{sub x2}) and their crystallization temperatures (T{sub x1} and T{sub x2}), respectively. The ratio ΔH{sub x1}/ΔH{sub x2} measured by DSC experimentally tended to be extremely high for the Fe{sub 85}Si{sub 2}B{sub 8}P{sub 4}Cu{sub 1} alloy, and this tendency was reproduced by the analysis with commercial software, Thermo-Calc, with database for Fe-based alloys, TCFE7 for Gibbs free energy (G) assessments. The calculations exhibit that a volume fraction (V{sub f}) of α-Fe tends to increase from 0.56 for the Fe{sub 85}B{sub 15} to 0.75 for the Fe{sub 85}Si{sub 2}B{sub 8}P{sub 4}Cu{sub 1} alloy. The computational analysis of the alloys for G of α-Fe and amorphous phases (G{sub α-Fe} and G{sub amor}) shows that a relationship G{sub α-Fe} ∼ G{sub amor} holds for the Fe{sub 85}Si{sub 2}B{sub 12}Cu{sub 1}, whereas G{sub α-Fe} < G{sub amor} for the Fe{sub 85}Si{sub 2}B{sub 8}P{sub 4}Cu{sub 1} alloy at T{sub x1} and that an extremely high V{sub f} = 0.75 was achieved for the Fe{sub 85}Si{sub 2}B{sub 8}P{sub 4}Cu{sub 1} alloy by including 2.8 at. % Si and 4.5 at. % P into α-Fe. These computational results indicate that the Fe{sub 85}Si{sub 2}B

  13. Advanced Corrosion-Resistant Zr Alloys for High Burnup and Generation IV Application

    International Nuclear Information System (INIS)

    Jeong, Y. H.; Park, S. Y.; Lee, M. H.; Choi, B. K.; Baek, J. H.; Park, J. Y.; Kim, J. H.; Kim, H. G.; Jung, Y. H.; Bang, B. G.

    2006-08-01

    The systematic study was performed to develop the advanced corrosion-resistant Zr alloys for high burnup and Gen IV application. The corrosion behavior was significantly changed with the alloy composition and the corrosion environment. In general, the model alloys with a higher alloying elements showed a higher corrosion resistance. Among the model alloys tested in this study, Zr-10Cr-0.2Fe showed the best corrosion resistance regardless of the corrosion condition. The oxide on the higher corrosion-resistant alloy such as Zr-1.0Cr-0.2Fe consisted of mainly columnar grains, and it have a higher tetragonal phase stability. In comparison with other alloys being considered for the SCWR, the Zr alloys showed a lower corrosion rate than ferritic-martensitic steels. The results of this study imply that, at least from a corrosion standpoint, Zr alloys deserve consideration as potential cladding or structural materials in supercritical water cooled reactors

  14. XPS and STM study of the growth and structure of passive films in high temperature water on a nickel-base alloy

    Energy Technology Data Exchange (ETDEWEB)

    Machet, A.; Galtayries, A.; Zanna, S.; Klein, L.; Maurice, V.; Jolivet, P.; Foucault, M.; Combrade, P.; Scott, P.; Marcus, P

    2004-09-15

    The early stages of passivation in high temperature water of a nickel-chromium-iron alloy (Alloy 600) have been investigated by X-ray Photoelectron Spectroscopy (XPS) and Scanning Tunneling Microscopy (STM). The samples (polycrystal Ni-16Cr-9Fe (wt. %) and single crystal Ni-17Cr-7Fe (1 1 1)) have been exposed for short time periods (0.4-8.2 min) to high temperature (325 deg. C) and high pressure water, under controlled hydrogen pressure, in a microautoclave designed to transfer the samples from and to the XPS spectrometer without air exposure. In the early stages of oxidation of the alloy (0.4-4 min), an ultra-thin oxide layer (about 1 nm) is formed, which consists of chromium oxide (Cr{sub 2}O{sub 3}), according to the Cr 2p{sub 3/2} core level spectrum. An outer layer of Cr(OH){sub 3} with a very small amount of Ni(OH){sub 2} is also revealed by the Cr 2p{sub 3/2}, Ni 2p{sub 3/2}, and O 1s core level spectra. At this early stage, there is a temporary blocking of the growth of Cr{sub 2}O{sub 3}. For longer exposures (4-8 min), the Cr{sub 2}O{sub 3} inner layer becomes thicker, at the expense of the outer Cr(OH){sub 3} layer. This implies the transport of Cr and Ni through the oxide layer, and release of Ni{sup 2+} in the solution. The structure of the ultra-thin oxide film formed on a single crystal Ni-17Cr-7Fe(1 1 1) alloy was analysed by STM in the constant current mode; STM images reveal that, in the early stages of oxidation, the oxide is crystalline, and the observed structure is consistent with the hexagonal structure of the oxygen sub-lattice in the basal plane (0 0 0 1) of {alpha}-Cr{sub 2}O{sub 3}.

  15. Effect of water purity on intergranular stress corrosion cracking of stainless steel and nickel alloys in BWRs

    Energy Technology Data Exchange (ETDEWEB)

    Gordon, B. [Structural Integrity Associates (United States); Garcia, S. [Electric Power Research Institute (United States)

    2011-07-01

    Boiling water reactors (BWRs) operate with very high purity water. While even the utilization of a very low conductivity water (e.g., 0.06 {mu}S/cm) coolant cannot prevent intergranular stress corrosion cracking (IGSCC) of sensitized stainless steel and nickel alloys under oxygenated conditions, the presence of certain impurities in the coolant can dramatically increase the probability of this most insidious form of corrosion. The goal of this paper is to present the effect of effect of only a few ionic impurities plus zinc on the IGSCC propensities of BWR stainless steel piping and reactor internals under both oxygenated, i.e., normal water chemistry (NWC) and deoxygenated, i.e., hydrogen water chemistry (HWC) conditions. More specifically, of the numerous impurities identified in the BWR coolant (e.g., lithium, sodium, potassium, silica, borate, chromate, phosphate, sulphate, chloride, nitrate, cuprous, cupric, ferrous, etc.) only strong acid anions sulfate and chloride that are stable in the highly reducing crack tip environment rather than the bulk water conductivity will be discussed in detail. Nitrate will be briefly discussed as representing a species that is not thermodynamically stable in the crack while the effects of zinc is discussed as a deliberate additive to the BWR environment. (authors)

  16. Effect of Cooling Rate and Chemical Modification on the Tensile Properties of Mg-5wt% Si Alloy

    Science.gov (United States)

    Mirshahi, Farshid; Meratian, Mahmood; Zahrani, Mohsen Mohammadi; Zahrani, Ehsan Mohammadi

    Hypereutectic Mg-Si alloys are a new class of light materials usable for aerospace and other advanced engineering applications. In this study, the effects of both cooling rate and bismuth modification on the micro structure and tensile properties of hypereutectic Mg-5wt% Si alloy were investigated. It was found that the addition of 0.5% Bi, altered the morphology of primary Mg2Si particles from bulky to polygonal shape and reduced their mean size from more than 70 μm to about 30 (am. Also, the tensile strength and elongation of the modified alloy increased about 10% and 20%, respectively, which should be ascribed to the modification of Mg2Si morphology and more uniform distribution of the primary particles. Moreover, an increase in tensile strength value with increase in cooling rate were observed which is attributed to finer micro structure of alloy in higher cooling rates. It was observed that Bi addition is significantly more effective in refining the morphology of primary Mg2Si particles than applying faster cooling rates.

  17. Characterization of a copper-modified Zn-Al eutectoid alloy

    International Nuclear Information System (INIS)

    Sandoval Jimenez, A.R.

    1992-01-01

    This work presents the results of studies performed on an eutectoid Zn-Al alloy with small additions of Cu. It is well known that the microstructure and mechanical properties of an alloy depend on its thermal and mechanical history. This alloy was subjected to different heat treatments and rolling at 250 o C. The microstructure was analyzed by scanning electron microscopy, the composition of the phases present was specified by microprobe and the phase transformation temperatures were determined by DSC. Mechanical tests, rate-of-corrosion tests with sea water and X-ray diffractometry were also performed. With reference to eutectoid Zn-Al alloys with less Cu, the mechanical resistance increases, the phase transformation temperatures are different and the τ 'phase appears after a longer annealing time (96 hs). The microstructures are characteristic of the thermomechanical treatments performed. The alloy show improved corrosion resistance (3 MPY) (Author)

  18. Comparative study of water chemistry and surface oxide composition on alloy 600 steam generator tubing

    International Nuclear Information System (INIS)

    Bjoernkvist, L.; Norring, K.; Nyborg, L.

    1993-01-01

    The Ringhals 3 steam generators experience secondary IGSCC on the tubes at support plate locations. Its sister unit Ringhals 4 is so far without IGSCC. Extensive work has been carried out in order to determine the local chemistry in crevices and the composition of deposits and oxide films on the tubes. Hot soaks of the SG:s at zero power has been performed and the water chemistry in occluded crevices of the SGs was predicted to be alkaline, pH 300degreesC = 10. In addition to eddy current testing, a large number of tubes have been pulled and destructively examined. These analysis include SEM/EDS characterization of TSP crevice deposits and Auger electron spectroscopy (AES) with depth profiling to reveal the composition of the tube OD oxide film. The AES analysis show an outer oxide rich in Fe 3 O 4 , mostly deposited. The actual Alloy 600 oxide is found below the magnetite and is 1-2 μm thick. The composition profile of the oxide exhibits a Cr-depletion relative to Ni in the outer part of the oxide, whereas an enrichment is found in depth. In order to correlate the water chemistry to the oxide composition profiles and deposits on pulled tubes, reference samples were prepared in an autoclave. The environments were chosen similar to the predicted Ringhals 3 and 4 crevice chemistry. Exposure both in an alkaline (pH 320degreesC∼ 9.9) and an acidic (pH 320degreesC ∼4.3) environment, containing sodium, chloride and sulphate, was studied. Some samples were also found on the Alloy 600 samples exposed to alkaline environment. Thus the prediction of alkaline chemistry was verified. The enrichment of chromium relative to nickel was shown to be potential and time dependent resulting in an increased Cr/Ni ratio at Cr-max with increasing potential and time

  19. The novel eutectic microstructures of Si-Mn-P ternary alloy

    International Nuclear Information System (INIS)

    Wu Yaping; Liu Xiangfa

    2010-01-01

    The microstructures of Si-Mn-P alloy manufactured by the technique of combining phosphorus transportation and alloy melting were investigated using electron probe micro-analyzer (EPMA). The phase compositions were determined by energy spectrum and the varieties of eutectic morphologies were discussed. It is found that there is no ternary compound but Si, MnP and MnSi 1.75-x could appear when the Si-Mn-P alloy's composition is proper. Microstructure is greatly refined by rapid solidification technique and the amount of eutectic phases change with faster cooling rates. Moreover, primary Si or MnP are surrounded firstly by the binary eutectic (Si + MnP) and then the ternary eutectic (Si + MnSi 1.75-x + MnP) which also exhibit binary structures due to divorced eutectic determined by the particularity of some Si-Mn-P alloys.

  20. Scientific basis for the choice of primary/secondary water chemistry

    International Nuclear Information System (INIS)

    Garnsey, R.

    1988-01-01

    The purpose of this paper is to illustrate the common scientific basis for the chemistry control strategies which have been developed. The evolution of chemistry control philosophies in some plant designs are outlined as examples. The essential requirement of water chemistry control is to preserve integrity of the circuit under all the environmental conditions experienced within that circuit. There may be specific additional requirements, as in the case of a PWR primary circuit, where boron concentration is used to control reactivity. The crucial requirement or concern can vary. In the primary circuit of a light water reactor the crucial requirement is to supress the activation and transportation of corrosion products and so minimize radiation fields around the circuit. On the secondary side of recirculating steam generators the critical requirement has been to preserve the integrity of generator tubing. In once-through steam generators the critical requirement may be the control of pressure losses associated with corrosion product deposits in the steam generator and the integrity of the turbine in addition to boiler integrity. (Nogami, K.)