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Sample records for alloy primary water

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

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    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.

  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. 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)

  4. 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)

  5. 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)

  6. 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)

  7. 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)

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

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    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)

  9. 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

  10. 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)

  11. 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)

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

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    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.

  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. 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

  15. 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)

  16. 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

  17. 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.

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

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    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.

  19. 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)

  20. 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.

  1. 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

  2. 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)

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

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    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.

  4. 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

  5. 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.

  6. 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

  7. 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.

  8. 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)

  9. 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)

  10. 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

  11. 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)

  12. 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)

  13. 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.

  14. 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

  15. 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)

  16. 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.

  17. 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.)

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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)

  3. 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....

  4. 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

  5. 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)

  6. 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.

  7. 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

  8. 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)

  9. 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

  10. 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

  11. 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.

  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. 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)

  14. 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.

  15. 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

  16. 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

  17. 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

  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. 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

  20. 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)

  1. 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.)

  2. 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)

  3. 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.

  4. 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)

  5. 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.)

  6. 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.

  7. 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

  8. 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)

  9. 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

  10. 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.

  11. 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.

  12. 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.

  13. 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)

  14. 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

  15. 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.

  16. 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.

  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. 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)

  20. 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

  1. 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

  2. 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)

  3. 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.

  4. 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.

  5. 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

  6. 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)

  7. 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

  8. 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

  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. 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

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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

  16. 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

  17. 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

  18. 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.

  19. 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.

  20. 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.)

  1. 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.)

  2. 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.

  3. 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.

  4. 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)

  5. 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)

  6. 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)

  7. 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.

  8. 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

  9. 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.

  10. 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)

  11. 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)

  12. 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

  13. 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

  14. 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.

  15. 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.

  16. 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 ...

  17. 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)

  18. 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

  19. 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

  20. 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.

  1. 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

  2. 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

  3. 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 ...

  4. 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.

  5. 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

  6. 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

  7. 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.

  8. 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)

  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. 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

  11. 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.

  12. 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)

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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.

  19. 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.)

  20. 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.

  1. 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)

  2. 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

  3. 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.

  4. 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

  5. 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

  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. 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.

  8. 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.

  9. 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.

  10. 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.)

  11. 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

  12. Primary circuit water chemistry during shutdown period at Kalinin NPP

    International Nuclear Information System (INIS)

    Gorbatenko, S.; Otchenashev, G.; Yurmanov, V.

    2005-01-01

    The primary circuit water chemistry feature at Kalinin NPP is using of special up-dated regime during the period of unit shutdown for refueling. The main objective of up-dated regime is removing from the circuit long time living corrosion products on SVO-2 ion exchange filters with the purpose of dose rates reduction from the equipment and in such a way reduction of maintenance personnel overexposure. (N.T.)

  13. 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)

  14. 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)

  15. 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)

  16. 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)

  17. 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

  18. 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

  19. 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.

  20. 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)

  1. 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)

  2. 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

  3. 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)

  4. 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)

  5. 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

  6. 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.

  7. 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.

  8. Reforming water to generate hydrogen using mechanical alloy

    International Nuclear Information System (INIS)

    Pena F, D. L.

    2016-01-01

    The objective of this research was to generate a hydrogen production system by means of mechanical milling, in which 0.1 g of magnesium were weighed using a volume of 300 μL for each water solvent (H_2O) and methanol (CH_3OH) in a container to start mechanical milling for 2, 4 and 6 h. Once the mechanical milling was finished, the hydrogen that was produced every two hours was measured to determine the appropriate milling time in the production, also in each period of time samples of the powders produced during the milling of Mg were taken, in this process we used characterization techniques such as: X-ray diffraction at an angle of 2θi 5 and 2θf 90 degrees and scanning electron microscopy, taking micrographs of 100, 500, 1000 and 5000 magnifications. According to the mechanical milling results hydrogen was obtained when using water, as well as with methanol. In the techniques of X-ray diffraction characterization different results were obtained before and after the milling, since by the diffractogram s is possible to observe how the magnesium to be put in the mechanical milling along with the water and methanol was diminishing to be transformed into hydroxide and magnesium oxide, as well as in the micrographs taken with scanning electron microscopy the change in the magnesium morphology to hydroxide and magnesium oxide is observed. (Author)

  9. The effect of primary recoil spectrum on radiation induced segregation in nickel-silicon alloys

    Science.gov (United States)

    Averback, R. S.; Rehn, L. E.; Wagner, W.; Ehrhart, P.

    1983-08-01

    Segregation of silicon to the surface of Ni-12.7 at% Si alloys during 2.0-MeV He and 3.25-MeV Kr irradiations was measured using Rutherford backscattering spectrometry. For equal calculated defect production rates the Kr irradiation was Ni-Si alloys is presented which qualitatively explains the segregation results. The model assumes that small interstitial-atom-clusters form in individual cascades and that these clusters become trapped at silicon solute atoms. The vacancy thereby becomes the more mobile defect. The model should also have relevance for the observation that void swelling in nickel is suppressed by the addition of silicon solute.

  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. 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)

  12. 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.)

  13. 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.

  14. 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.

  15. 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)

  16. 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)

  17. 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.

  18. Behavior of stainless steels in pressurized water reactor primary circuits

    International Nuclear Information System (INIS)

    Féron, D.; Herms, E.; Tanguy, B.

    2012-01-01

    Stainless steels are widely used in primary circuits of pressurized water reactors (PWRs). Operating experience with the various grades of stainless steels over several decades of years has generally been excellent. Nevertheless, stress corrosion failures have been reported in few cases. Two main factors contributing to SCC susceptibility enhancement are investigated in this study: cold work and irradiation. Irradiation is involved in the stress corrosion cracking and corrosion of in-core reactor components in PWR environment. Irradiated assisted stress corrosion cracking (IASCC) is a complex and multi-physics phenomenon for which a predictive modeling able to describe initiation and/or propagation is not yet achieved. Experimentally, development of initiation smart tests and of in situ instrumentation, also in nuclear reactors, is an important axis in order to gain a better understanding of IASCC kinetics. A strong susceptibility for SCC of heavily cold worked austenitic stainless steels is evidenced in hydrogenated primary water typical of PWRs. It is shown that for a given cold-working procedure, SCC susceptibility of austenitic stainless steels materials increases with increasing cold-work. Results have shown also strong influences of the cold work on the oxide layer composition and of the maximum stress on the time to fracture.

  19. 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)

  20. 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.

  1. 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)

  2. 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.)

  3. 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)

  4. 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)

  5. 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)

  6. 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...

  7. 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.)

  8. 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

  9. 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.

  10. 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

  11. 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

  12. Structure of electron tracks in water. 2. Distribution of primary ionizations and excitations in water radiolysis

    International Nuclear Information System (INIS)

    Pimblott, S.M.; Mozumder, A.

    1991-01-01

    A procedure for the calculation of entity-specific ionization and excitation probabilities for water radiolysis at low linear energy transfer (LET) has been developed. The technique pays due attention to the effects of the ionization threshold and the energy dependence of the ionization efficiency. The numbers of primary ionizations and excitations are not directly proportional to the spur energy. At a given spur energy, ionization follows a binomial distribution subject to an energetically possible maximum. The excitation distribution for a spur of given energy and with a given number of ionizations is given by a geometric series. The occurrence probabilities depend upon the cross sections of ionization, excitation, and other inferior processes. Following the low-LET radiolysis of liquid water the most probable spurs contain one ionization, two ionizations, or one ionization and one excitation, while in water vapor they contain either one ionization or one excitation. In liquid water the most probable outcomes for spurs corresponding to the most probable energy loss (22 eV) and to the mean energy loss (38 eV) are one ionization and one excitation, and two ionizations and one excitation, respectively. In the vapor, the most probable energy loss is 14 eV which results in one ionization or one excitation and the mean energy loss is 34 eV for which the spur of maximum probability contains one ionization and two excitations. The total calculated primary yields for low-LET radiolysis are in approximate agreement with experiment in both phases

  13. 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

  14. 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)

  15. 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

  16. 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.)

  17. 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...

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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)

  3. 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

  4. 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)

  5. 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.

  6. 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.

  7. 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.

  8. 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

  9. 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

  10. 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

  11. UV radiation and primary production in the Antarctic waters

    Digital Repository Service at National Institute of Oceanography (India)

    LokaBharathi, P.A.; Krishnakumari, L.; Bhattathiri, P.M.A.; Chandramohan, D.

    at 683 nm), scalar irradiance (photosynthetically active radiation (PAR), computed primary production (pp), diffuse attenuation coefficient, and UVB (308 and 320 nm) and UVA (340 and 380 nm) radiation and ocean temperature all measured as a function...

  12. Effect of hydrazine on general corrosion of carbon and low-alloyed steels in pressurized water reactor secondary side water

    Energy Technology Data Exchange (ETDEWEB)

    Järvimäki, Sari [Fortum Ltd, Loviisa Power Plant, Loviisa (Finland); Saario, Timo; Sipilä, Konsta [VTT Technical Research Centre of Finland Ltd., Nuclear Safety, P.O. Box 1000, FIN-02044 VTT (Finland); Bojinov, Martin, E-mail: martin@uctm.edu [Department of Physical Chemistry, University of Chemical Technology and Metallurgy, Kl. Ohridski Blvd, 8, 1756 Sofia (Bulgaria)

    2015-12-15

    Highlights: • The effect of hydrazine on the corrosion of steel in secondary side water investigated by in situ and ex situ techniques. • Oxide grown on steel in 100 ppb hydrazine shows weaker protective properties – higher corrosion rates. • Possible explanation of the accelerating effect of higher concentrations of hydrazine on flow assisted corrosion offered. - Abstract: The effect of hydrazine on corrosion rate of low-alloyed steel (LAS) and carbon steel (CS) was studied by in situ and ex situ techniques under pressurized water reactor secondary side water chemistry conditions at T = 228 °C and pH{sub RT} = 9.2 (adjusted by NH{sub 3}). It is found that hydrazine injection to a maximum level of 5.06 μmol l{sup −1} onto surfaces previously oxidized in ammonia does not affect the corrosion rate of LAS or CS. This is confirmed also by plant measurements at Loviisa NPP. On the other hand, hydrazine at the level of 3.1 μmol l{sup −1} decreases markedly the amount and the size of deposited oxide crystals on LAS and CS surface. In addition, the oxide grown in the presence of 3.1 μmol l{sup −1} hydrazine is somewhat less protective and sustains a higher corrosion rate compared to an oxide film grown without hydrazine. These observations could explain the accelerating effect of higher concentrations of hydrazine found in corrosion studies of LAS and CS.

  13. Water quality estimation method for primary coolant circuit

    International Nuclear Information System (INIS)

    Wada, Yoichi; Ibe, Hidefumi.

    1994-01-01

    The present invention is suitable to water quality diagnosis at each of the portions in a reactor upon hydrogen injection for preventing stress corrosion crackings (SCC) of a BWR type reactor. That is, a plurality of simulations are conducted how the water quality at each of the portions in the reactor is changed when hydrogen injection amount is changed depending on the design and operation conditions of the plant. The result of the calculation is stored in a memory device. A water quality distribution in a pressure vessel having a solution which agrees with a value actually measured by a water quality measuring device disposed at the outside of a reactor core is retrieved from the results of the calculation. If no agreeing solution can be found, water quality distribution containing the actually measured value is determined based on the result of the calculation by using interpolation. In the present invention, the result of the calculation obtained by the simulation and the actually measured value at the outside of the reactor core can be utilized, to map the distribution of reactor water ingredients on a screen, which can accurately estimate the water quality at the periphery of the reactor core on real time. As a result, an operational efficiency of a reactor which can control water quality upon hydrogen injection at an optimum condition. (I.S.)

  14. Predicted Variations of Water Chemistry in the Primary Coolant Circuit of a Supercritical Water Reactor

    International Nuclear Information System (INIS)

    Yeh, Tsung-Kuang; Wang, Mei-Ya; Liu, Hong-Ming; Lee, Min

    2012-09-01

    In response to the demand over a higher efficiency for a nuclear power plant, various types of Generation IV nuclear reactors have been proposed. One of the new generation reactors adopts supercritical light water as the reactor coolant. While current in-service light water reactors (LWRs) bear an average thermal efficiency of 33%, the thermal efficiency of a supercritical water reactor (SCWR) could generally reach more than 44%. For LWRs, the coolants are oxidizing due to the presence of hydrogen peroxide and oxygen, and the degradation of structural materials has mainly resulted from stress corrosion cracking. Since oxygen is completely soluble in supercritical water, similar or even worse degradation phenomena are expected to appear in the structural and core components of an SCWR. To ensure proper designs of the structural components and suitable selections of the materials to meet the requirements of operation safety, it would be of great importance for the design engineers of an SCWR to be fully aware of the state of water chemistry in the primary coolant circuit (PCC). Since SCWRs are still in the stage of conceptual design and no practical data are available, a computer model was therefore developed for analyzing water chemistry variation and corrosion behavior of metallic materials in the PCC of a conceptual SCWR. In this study, a U.S. designed SCWR with a rated thermal power of 3575 MW and a coolant flow rate of 1843 kg/s was selected for investigating the variations in redox species concentration in the PCC. Our analyses indicated that the [H 2 ] and [H 2 O 2 ] at the core channel were higher than those at the other regions in the PCC of this SCWR. Due to the self-decomposition of H 2 O 2 , the core channel exhibited a lower [O 2 ] than the upper plenum. Because the middle water rod region was in parallel with the core channel region with relatively high dose rates, the [H 2 ] and [H 2 O 2 ] in this region were higher than those in the other regions

  15. Primary productivity in nearshore waters of Thal, Maharashtra coast

    Digital Repository Service at National Institute of Oceanography (India)

    Varshney, P.K.; Nair, V.R.; Abidi, S.A.H.

    Primary productivity off Thal, Maharashtra, India was evaluated at 3 stations during Feb. 1980 to Jan. 1981. The area was quite turbid and the euphotic zone never exceeded 2.5 m. Column production ranged from 0.69 to 605.21 mg C.m/2.d/2 (av. 78.2 mg...

  16. 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)

  17. The change in the primary production of Danish coastal waters

    International Nuclear Information System (INIS)

    Edelvang, K.; Erichsen, A.; Gustavson, K.; Bundgaard, K.; Dahl-Madsen, K.I.

    2001-01-01

    The background for this study is the development of the 'Farvandsmodel' for the NOVA-2003 programme and the nationally founded research project DECO (Danish Environmental Monitoring of Coastal Waters), which focuses on the use of remote sensing for the monitoring of Danish Coastal waters. Danish national programmes for the monitoring of the marine ecosystem are a relatively new activity, which has grown during the last 20 years. The HAV90 research programme amassed important information to be included in future environmental efforts such as the NOVA-2003 programme, aimed at monitoring the Danish coastal waters. The following is a selection of the topics mentioned in the NOVA-2003 programme (NOVA-2003, 2000) especially relevant to this study: 1) Hydrography. 2) Concentration and spatial distribution of nutrients. 3) Water and nutrient fluxes. 4) Oxygen depletion. As part of this programme, a 3D hydrographic model describing currents and fluxes in Danish waters has been designed by DHI Water and Environment for the Danish Ministry of Energy and Environment. The model is called the 'Farvandsmodel', which is the collective Danish name of this regional 3D hydrodynamic model and its associated database for storage and dissemination of model results and field measurements. The model is planned to be in operation until 2004. It has a great potential within hydrographic modelling in Danish waters, as it is capable of running 5-day prognoses for currents, water levels and stratification. The model is also able to calculate the sensitivity of the present system to changes in various input parameters. In this way the model may be used as a tool for testing the sensitivity of Danish coastal waters to the impact of the green house effects. The nationally funded research programme, DECO (1997-2000), aims to investigate the use of remote sensing for monitoring Danish coastal waters. To support this research, a eutrophication module (EU) was set up for the 'Farvandsmodel'. The

  18. 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

  19. Experimental and numerical investigation on under-water friction stir welding of armour grade AA2519-T87 aluminium alloy

    OpenAIRE

    Sree Sabari, S.; Malarvizhi, S.; Balasubramanian, V.; Madusudhan Reddy, G.

    2016-01-01

    Friction stir welding (FSW) is a promising welding process that can join age hardenable aluminium alloys with high joint efficiency. However, the thermal cycles experienced by the material to be joined during FSW resulted in the deterioration of mechanical properties due to the coarsening and dissolution of strengthening precipitates in the thermo-mechanical affected zone (TMAZ) and heat affected zone (HAZ). Under water friction stir welding (UWFSW) is a variant of FSW process which can maint...

  20. 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

  1. 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.

  2. Modelling the radiolysis of RSG-GAS primary cooling water

    Science.gov (United States)

    Butarbutar, S. L.; Kusumastuti, R.; Subekti, M.; Sunaryo, G. R.

    2018-02-01

    Water chemistry control for light water coolant reactor required a reliable understanding of radiolysis effect in mitigating corrosion and degradation of reactor structure material. It is known that oxidator products can promote the corrosion, cracking and hydrogen pickup both in the core and in the associated piping components of the reactor. The objective of this work is to provide the radiolysis model of RSG GAS cooling water and further more to predict the oxidator concentration which can lead to corrosion of reactor material. Direct observations or measurements of the chemistry in and around the high-flux core region of a nuclear reactor are difficult due to the extreme conditions of high temperature, pressure, and mixed radiation fields. For this reason, chemical models and computer simulations of the radiolysis of water under these conditions are an important route of investigation. FACSIMILE were used to calculate the concentration of O2 formed at relatively long-time by the pure water γ and neutron irradiation (pH=7) at temperature between 25 and 50 °C. This simulation method is based on a complex chemical reaction kinetic. In this present work, 300 MeV-proton were used to mimic γ-rays radiolysis and 2 MeV fast neutrons. Concentration of O2 were calculated at 10-6 - 106 s time scale.

  3. 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

  4. The evaluation of the use of metal alloy fuels in pressurized water reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Lancaster, D.

    1992-10-26

    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.

  5. Corrosion and deuterium uptake of Zr-based alloys in supercritical water

    International Nuclear Information System (INIS)

    Khatamian, D.

    2010-01-01

    To increase the thermodynamic efficiency above 40% in nuclear power plants, the use of supercritical water as the heat transport fluid has been suggested. Zircaloy-2, -4, Zr-Cr-Fe, Zr-1Nb and Zr-2.5Nb were tested as prospective fuel cladding materials in 30 MPa D 2 O at 500 o C. Zircaloy-2 showed the highest rates of corrosion and hydriding. Although Zr-Cr-Fe initially showed a very low corrosion rate, it displayed breakaway corrosion kinetics after 50 h exposure. The best-behaved material both from a corrosion and hydrogen uptake point of view was Zr-2.5Nb. However, the Zr-2.5Nb oxide growth rate was still excessive and beyond the current CANDU design allowance. Similar coupons, coated with Cr, were also tested. The coated layer effectively prevented oxidation of the coupons except on the edges, where the coating was thinner and had some flaws. In addition, the Cr-coated Zr-2.5Nb coupons had the lowest deuterium pickup of all the alloys tested and showed no signs of accelerated or nonuniform corrosion. (author)

  6. Evaluation of crack propagation of alloy 600 tube in high temperature water, (1)

    International Nuclear Information System (INIS)

    Hirano, Hideo; Kawamura, H.; Kawamura, Kohji; Matsubara, Masaaki

    1990-01-01

    This report describes the analysis of stress intensity factors at cracks in alloy 600 steam generator tubes. Based on the results of the analysis, IGA/SCC tests were carried out to examine the effect of stress intensity and water quality on the crack propagation rate. The main test result are as follows: (1) Hoop stress was caused by the pressure difference between the internal and external surface of the steam generator tube. The calculated hoop stress was about 7 kg/mm 2 . In addition, the temperature difference between the internal and external surface caused thermal stress. The thermal stress was about 10 kg/mm 2 at the external surface and the one at the internal surface was about -10 kg/mm 2 . Total stress at the external and internal surface was 17 kg/mm 2 and -3 kg/mm 2 , respectively. (2) The stress intensity factor at the crack tip increased with increasing crack length. For a long crack, the stress intensity factor decreased with increasing crack number. However, for a short crack, the stress intensity factor decreased little with increasing crack number. (3) Under high stress-intensity conditions, i.e. 40∼50 kg·mm -3/2 , the IGA/SCC test showed that IGA/SCC propagated in AVT and AVT/boric-acid solution at 320degC and 350degC. However, the propagation rate was low. (author)

  7. Effect of soluble zinc additions on the SCC performance of nickel alloys in deaerated hydrogenated water

    International Nuclear Information System (INIS)

    Morton, D.S.; Thompson, C.D.; Gladding, D.; Schurman, M.K.

    1997-08-01

    Stress corrosion crack growth rates (SCCGR) of alloy 600, EN82H and X-750 were measured in deaerated hydrogenated water to determine if soluble zinc mitigates SCCGR. Constant load compact tension specimen tests were conducted. Two test strategies were used to discern a possible zinc effect. The first strategy employed separate SCCGR tests in zinc and non-zinc environments and compared the resulting crack growth rates. The second strategy varied zinc levels at the midterm of single specimen SCCGR tests and characterized the resulting crack growth rate effect through an electrical potential drop in-situ crack monitor. Results from the direct comparison and midterm changing chemistry tests did not discern a zinc influence; any apparent zinc influence is within test to test variability (∼1.5x change in crack growth rate). AEM, AUGER and ESCA crack tip fracture surface studies identified that zinc was not incorporated within crack tip oxides. These studies identified nickel rich crack tip oxides and spinel, with incorporated zinc, (∼5 atom percent) bulk surface oxides

  8. 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.

  9. 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

  10. Heterogeneous primary nucleation of ice in water and aqueous solutions

    NARCIS (Netherlands)

    Thijssen, H.A.C.; Vorstman, M.A.G.; Roels, J.A.

    1968-01-01

    The effect of the volume of the liquid sample, the degree of turbulence in the liquid, and the rate of cooling upon the probability of nucleation has been studied for water and aqueous solutions. Nucleation rates were measured for droplets nearly instantaneously cooled to a predetermined

  11. 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.

  12. 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.)

  13. First domestic primary loop recircuration pump for boiling water reactor

    International Nuclear Information System (INIS)

    Fukuda, Minoru; Taka, Shusei; Kato, Hiroyuki

    1981-01-01

    Two primary loop recirculation (PLR) pumps for the second unit of the Fukushima No. 2 Nuclear Power Station of the Tokyo Electric Power Co., Inc., have been manufactured by Ebara Corporation. They are the first domestically produced pumps for commercial power plants and were manufactured under license from Byron Jackson Pump Division of Borg Warner Corporation. This article describes the special features of pump design and stress analysis, and the results of the 700 hours of factory loop tests, which are all essential for the PLR pump. (author)

  14. 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

  15. 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.

  16. 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.

  17. Stress corrosion of low alloy steels used in external bolting on pressurised water reactors

    International Nuclear Information System (INIS)

    Skeldon, P.; Hurst, P.; Smart, N.R.

    1992-01-01

    The stress corrosion cracking (SCC) susceptibility of AISI 4140 and AISI 4340 steels has been evaluated in five environments, three simulating a leaking aqueous boric acid environment and two simulating ambient external conditions ie moist air and salt spray. Both steels were found to be highly susceptible to SCC in all environments at hardnesses of 400 VPN and above. The susceptibility was greatly reduced at hardnesses below 330 VPN but in one environment, viz refluxing PWR primary water, SCC was observed at hardnesses as low as 260VPN. Threshold stress intensities for SCC were frequently lower than those in the literature

  18. 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)

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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)

  4. Relation between the mechanical properties and SCC behavior of the alloys used in high temperature water

    International Nuclear Information System (INIS)

    Tsubota, M.; Katayama, Y.; Kanazawa, Y.

    2007-01-01

    It was shown in the previous reports that carbon and low alloy steels, martensitic stainless steels and cold worked austenitic stainless steels have shown high SCC susceptibility in the highly hardened condition. Those steels had similar critical hardness for SCC (HV300-340), over which the materials showed SCC susceptibility, even though the hardening process was different. Hardening processes applied for the alloys were as follows: (1) Martensitic transformation: Carbon and low alloy steels and martensitic stainless steels. (2) Alpha-prime decomposition (precipitation hardening): martensitic stainless steels. (3) Cold work: austenitic stainless steels. The relationship between the mechanical properties and SCC susceptibility of the alloys is discussed and summarized in the present paper. (author)

  5. 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.

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

    International Nuclear Information System (INIS)

    Lott, Randy G.

    2003-01-01

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

  7. 5. International seminar on primary and secondary side water chemistry of nuclear power plants

    International Nuclear Information System (INIS)

    2001-01-01

    The major subjects of the meetings are: water chemistry of primary and secondary coolant circuits of PWR type reactors (mainly WWER types), corrosion of steam generators, decontamination processes, treatment of radioactive waste waters and related subjects. All the 29 papers were individually indexed and abstracted for the INIS database. (R.P.)

  8. 5. International seminar on primary and secondary side water chemistry of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    The major subjects of the meetings are: water chemistry of primary and secondary coolant circuits of PWR type reactors (mainly WWER types), corrosion of steam generators, decontamination processes, treatment of radioactive waste waters and related subjects. All the 29 papers were individually indexed and abstracted for the INIS database. (R.P.)

  9. 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…

  10. [The prospects for using potable mineral waters as agents for the primary prevention of gastroduodenal ulcers].

    Science.gov (United States)

    Polushina, N D; Frolkov, V K

    1990-01-01

    Primary preventive effects of mineral water Essentuki 17 were investigated on 500 male Wistar rats (body mass 200-250 g). It is demonstrated that oral pretreatment with the above water can prevent the onset of gastroduodenal ulcers. Changes in secretion of gastrin, insulin, glucagon, triiodothyronine and thyroxin support the clinical evidence.

  11. Alternative water chemistry for the primary loop of PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Henzel, N [Siemens AG Unternehmensbereich KWU, Erlangen (Germany)

    1997-02-01

    Advanced fuel element concepts (longer cycles, higher burnup, increased rod power) call for more reactivity binding capacity and, moreover, might produce higher void fractions, particularly in the hot channel. Thus, on the one hand, more alcalizing agent is needed to maintain a high coolant pH according to the approved ``modified boron-lithium mode of operation`` in the presence of more boric acid (chemical shim); on the other hand, increasing enrichment of coolant constituents due to local boiling (higher void fraction), which must not result in accelerated corrosion of fuel cladding and structural materials, imposes enhanced requirements on both, materials technology and water chemistry. At present, the use of boric acid enriched in B10 (the isotope effective in terms of reactivity control) appears to advantageously compromise in capturing more neutrons with less total boron while maintaining or even slightly reducing lithium concentrations at the same time. There is no feasible alternative for boric acid used as the chemical shim and for hydrogen gas as the reducing agent used to suppress oxygen formation by water radiolysis. Systematic screening as performed in phase 1 of a recent project proved potassium hydroxide to be the only potential candidate to favourably replace lithium 7 hydroxide as an alcalizing agent. Unfortunately, the results of pertinent comparative corrosion tests are not unambiguous, and available operational experience with potassium hydroxide in WWER plants is not readily applicable to western world-type PWR plants. Therefore, a switch-over from lithium to potassium can be envisaged only subsequent to a comprehensive qualification program which is planned to be the objective of phase 2 of the project. This program should also comprise zinc addition tests in order to confirm the alleged positive impact of this element on corrosion rates and activity buildup. (Abstract Truncated)

  12. 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)

  13. 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.

  14. 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.

  15. 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

  16. Physico-chemical quality of drinking water in villages of Primary Health Centre, Waghodia, Gujarat (India).

    Science.gov (United States)

    Desai, Gaurav; Vasisth, Smriti; Patel, Maharshi; Mehta, Vaibhav; Bhavsar, Bharat

    2012-07-01

    16 water samples were collected to study the physical and chemical quality of water of main source of drinking water in the villages of Primary Health Centre, Waghodia of Vadodara district of Gujarat. The values recommended by Indian Standard for Drinking Water (IS 10500:1991) were used for comparison of observed values. The study indicates that the contamination problem in these villages is not alarming at present, but Waghodia being industrial town, ground water quality may deteriorate with passage of time, which needs periodical monitoring. The study provides the local area baseline data which may be useful for the comparison of future study.

  17. 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...

  18. 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.

  19. 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.

  20. 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

  1. 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)

  2. 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

  3. 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

  4. 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.

  5. 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

  6. Potable water quality monitoring of primary schools in Magura district, Bangladesh: children's health risk assessment.

    Science.gov (United States)

    Rahman, Aminur; Hashem, Abul; Nur-A-Tomal, Shahruk

    2016-12-01

    Safe potable water is essential for good health. Worldwide, school-aged children especially in the developing countries are suffering from various water-borne diseases. In the study, drinking water supplies for primary school children were monitored at Magura district, Bangladesh, to ensure safe potable water. APHA standard analytical methods were applied for determining the physicochemical parameters of the water samples. For determination of the essential physicochemical parameters, the samples were collected from 20 randomly selected tube wells of primary schools at Magura. The metal contents, especially arsenic (As), iron (Fe), and manganese (Mn), in the water samples were analyzed by atomic absorption spectroscopy. The range of physicochemical parameters found in water samples were as follows: pH 7.05-9.03, electrical conductivity 400-2340 μS/cm, chloride 10-640 mg/L, hardness 200-535 mg/L as CaCO 3 , and total dissolved solids 208-1216 mg/L. The level of metals in the tube well water samples were as follows: As 1 to 55 μg/L, Fe 40 to 9890 μg/L, and Mn 10 to 370 μg/L. Drinking water parameters of Magura district did not meet the requirement of the World Health Organization drinking water quality guideline, or the Drinking Water Quality Standards of Bangladesh.

  7. 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)

  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. 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. Stress corrosion cracking and oxidation of austenitic stainless steel 316 L and model alloy in supercritical water reactor

    International Nuclear Information System (INIS)

    Saez-Maderuelo, A.; Gomez-Briceno, D.; Diego, G.

    2015-01-01

    In this work, an austenitic stainless steel type 316 L was tested in deaerated supercritical water at 400 deg. C and 500 deg. C and 25 MPa to determine how variations in water conditions influence its stress corrosion cracking behaviour and to make progress in the understanding of mechanisms involved in SCC processes in this environment. Moreover, the influence of plastic deformation in the resistance of the material to SCC was also studied at both temperatures. In addition to this, previous oxidation experiments at 400 deg. C and 500 deg. C and at 25 MPa were taken into account to gain some insight in this kind of processes. Furthermore, a cold worked model alloy based on the stainless steel 316 L with some variations in the chemical composition in order to simulate the composition of the grain boundary after irradiation was tested at 400 deg. C and 25 MPa in deaerated supercritical water. (authors)

  11. 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.

  12. 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.

  13. 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

  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. [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.

  16. 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)

  17. 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.

  18. Influence of primary α-phase volume fraction on the mechanical properties of Ti-6Al-4V alloy at different strain rates and temperatures

    Science.gov (United States)

    Ren, Yu; Zhou, Shimeng; Luo, Wenbo; Xue, Zhiyong; Zhang, Yajing

    2018-03-01

    Bimodal microstructures with primary α-phase volume fractions ranging from 14.3% to 57.1% were gained in Ti-6Al-4V (Ti-64) alloy through annealed in two-phase region at various temperatures below the β-transus point. Then the influence of the primary α-phase volume fraction on the mechanical properties of Ti-64 were studied. The results show that, at room temperature and a strain rate of 10‑3 s‑1, the yield stress decreases but the fracture strain augments with added primary α-phase volume fraction. The equiaxed primary α-phase possesses stronger ability to coordinate plastic deformation, leading to the improvement of the ductile as well as degradation of the strength of Ti-64 with higher primary α-phase volume fraction. As the temperature goes up to 473 K, the quasi-static yield stress and ultimate strength decrease first and then increase with the incremental primary α-phase volume fraction, due to the interaction between the work hardening and the softening caused by the DRX and the growth of the primary α-phase. At room temperature and a strain rate of 3×103 s‑1, the varying pattern of strength with the primary α-phase volume fraction resembles that at a quasi-static strain rate. However, the flow stress significantly increases but the strain-hardening rate decreases compared to those at quasi-static strain rate due to the competition between the strain rate hardening and the thermal softening during dynamic compression process.

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

    Directory of Open Access Journals (Sweden)

    Cheng Bo

    2016-01-01

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

  20. Modelling the behaviour of corrosion products in the primary heat transfer circuits of pressurised water reactors

    International Nuclear Information System (INIS)

    Rodliffe, R.S.; Polley, M.V.; Thornton, E.W.

    1985-05-01

    The redistribution of corrosion products from the primary circuit surfaces of a water reactor can result in increased flow resistance, poorer heat transfer performance, fuel failure and radioactive contamination of circuit surfaces. The environment is generally sufficiently well controlled to ensure that the first three effects are not limiting. The last effect is of particular importance since radioactive corrosion products are major contributors to shutdown fields and since it is necessary to ensure that the radiation exposure of personnel is as low as reasonably achievable. This review focusses attention on the principles which must form the basis for any mechanistic model describing the formation, transport and deposition of radioactive corrosion products. It is relevant to all water reactors in which the primary heat transfer medium is predominantly single-phase water and in which steam is generated in a secondary circuit, i.e. including CANDU pressurised heavy water reactors, Sovient VVERs, etc. (author)

  1. Investigative studies on water contamination in Bangladesh. Primary treatment of water samples at the sampling site

    International Nuclear Information System (INIS)

    Sera, K.; Islam, Md. Shafiqul; Takatsuji, T.; Nakamura, T.; Goto, S.; Takahashi, C.; Saitoh, Y.

    2010-01-01

    Arsenic concentration in 13 well waters, 9 pond waters, 10 agricultural waters and a coconut juice taken in Comilla district, Bangladesh, where the problem of arsenic pollution is the most severe, was investigated. High-level arsenic is detected even in the well water which has been kept drinking by the people. Relatively high arsenic concentration was detected for some pond and farm waters even though the sampling was performed just after the rainy season and the waters were expected to be highly diluted. Clear relationship was observed in elemental compositions between the pond water and the coconut juice collected at the edge of the water. These results are expected to become the basic information for evaluating the risk of individual food such as cultured fishes, shrimps and farm products, and for controlling total intakes of arsenic. In order to solve the problem of transportation of water samples internationally, a simple method of target preparation performed at the sampling site was established and its validity was confirmed. All targets were prepared at the sampling sites in this study on the basis of this method. (author)

  2. Primary water chemistry improvement for radiation exposure reduction at Japanese PWR Plants

    Energy Technology Data Exchange (ETDEWEB)

    Nishizawa, Eiichi [Omiya Technical Institute, Saitama-ken (Japan)

    1995-03-01

    Radiation exposure during the refueling outages at Japanese Pressurized Water Reactor (PWR) Plants has been gradually decreased through continuous efforts keeping the radiation dose rates at relatively low level. The improvement of primary water chemistry in respect to reduction of the radiation sources appears as one of the most important contributions to the achieved results and can be classified by the plant operation conditions as follows

  3. 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.

  4. 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)

  5. General and localized corrosion of carbon and low-alloy steels in oxygenated high-temperature water. Final report

    International Nuclear Information System (INIS)

    Macdonald, D.D.; Smialowska, S.; Pednekar, S.

    1983-02-01

    The susceptibilities to stress corrosion cracking (SCC) of two carbon steels, SA106-grB and SA333-gr6, which are used in seamless BWR piping, and a low-alloy pressure vessel steel, A508-C12, were studied in high purity water as a function of oxygen concentration (0.16 to 8 ppM) and temperature (50 to 288 0 C) . The susceptibility to SCC was measured using the slow strain rate technique. The fracture surfaces of the test specimens were also examined using SEM to determine the mode of failure. In water containing 1 and 8 ppM oxygen and at temperatures above 135 0 C, transgranular stress corrosion cracking (TGSCC) was observed to occur in A508-C12, SA333-gr6 and SA106grB steels at very high stresses. The susceptibility to SCC increased with temperature

  6. 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.

  7. Correlations between the electrochemical behaviour and surface film composition of TZM alloy exposed to divertor water coolant environments

    International Nuclear Information System (INIS)

    Maday, M.-F.; Giorgi, R.; Dikonimos-Makris, T.

    1997-01-01

    X-ray photoelectron spectroscopy (XPS) has been carried out on TZM alloy surfaces after short and long immersion tests in high temperature (250 C) aqueous environments simulating possible fusion reactor coolant conditions during operation. Phase identification by XPS was used in connection with the open circuit potential trends to suggest plausible hypotheses about TZM corrosion behaviour in the various chemical environments considered in this study. It was proposed that exposure of TZM to oxidizing water conditions produced poorly protective layers, which consist essentially of low (IV) and intermediate (V) valency Mo oxides/hydroxides. Conversely the results obtained in deaerated and reducing water conditions suggested that barrier films could develop in these environments: the phases exhibit a bilayered structure and consisted of an inner tetravalent Mo oxide/hydroxide and an outer hexavalent Mo oxide. The protective properties of such layers were attributed to the hexavalent Mo species. (orig.)

  8. 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.

  9. Primary collector wall local temperature fluctuations in the area of water-steam phase boundary

    Energy Technology Data Exchange (ETDEWEB)

    Matal, O.; Klinga, J.; Simo, T. [Energovyzkum Ltd., Brno (Switzerland)

    1995-12-31

    A limited number of temperature sensors could be installed at the primary collector surface in the area of water - steam phase boundary. The surface temperatures as well WWER 440 steam generator process data were measured and stored for a long time and off-line evaluated. Selected results are presented in the paper. (orig.). 2 refs.

  10. Primary collector wall local temperature fluctuations in the area of water-steam phase boundary

    Energy Technology Data Exchange (ETDEWEB)

    Matal, O; Klinga, J; Simo, T [Energovyzkum Ltd., Brno (Switzerland)

    1996-12-31

    A limited number of temperature sensors could be installed at the primary collector surface in the area of water - steam phase boundary. The surface temperatures as well WWER 440 steam generator process data were measured and stored for a long time and off-line evaluated. Selected results are presented in the paper. (orig.). 2 refs.

  11. 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...

  12. Primary biodegradation of veterinary antibiotics in aerobic and anaerobic surface water simulation systems

    DEFF Research Database (Denmark)

    Ingerslev, Flemming; Toräng, Lars; Loke, M.-L.

    2001-01-01

    The primary aerobic and anaerobic biodegradability at intermediate concentrations (50-5000 mug/l) of the antibiotics olaquindox (OLA), metronidazole (MET), tylosin (TYL) and oxytetracycline (OTC) was studied in a simple shake flask system simulating the conditions in surface waters. The purpose...

  13. Activity build-up in the primary circuit of pressurized water reactors

    International Nuclear Information System (INIS)

    Sachse, G.; Mittag, I.

    1986-01-01

    Based upon international literature, the following topics are reviewed: research and development efforts; release, transport, and deposition of radioactive corrosion products under primary circuit conditions; experimental results in test and technical systems; possibilities of controlling radiation fields in nuclear power plants by water-chemical measures, decontamination, and high-temperature filtration. (author)

  14. 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.

    -, 11 July 2015 (2015), s. 1-6 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 * abrasivewater jet turning * traverse speed * surface roughness Subject RIV: JQ - Machines ; Tools Impact factor: 1.568, year: 2015 http://link.springer.com/article/10.1007/s00170-015-7489-0

  15. Conventional Treatment of Surface Water Using Moringa Oleifera Seeds Extract as a Primary Coagulant

    Directory of Open Access Journals (Sweden)

    Suleyman A. Muyibi, Ahmed Hissein M Birima, Thamer A. Mohammed

    2012-10-01

    Full Text Available The present study involved the use of a model pilot scale water treatment plant to treat turbid surface water from a stream using processed Moringa oleifera seed with 25 % w/w oil extracted as primary coagulant. The water treatment plant was made up of four unit operations: coagulation, flocculation, sedimentation, and filtration (rapid sand filter. Test runs were carried out for three hours per run over a three-month period with turbidities ranging from 18 to 261 NTU. The turbidity, pH, and alkalinity as well as the filter head loss were measured every 30 minutes during the experimental runs. Average turbidity removal of up to 96 % at an effective doses of 20 and 30 mg/l of oil extracted M. oleifera for low (< 50 NTU and moderate turbidity (< 100 NTU water respectively was observed doses 50 – 80 mg/l for high turbidity (> 100 NTU water. M. oleifera seed extract was found to have no significant effect on pH or alkalinity of the water. The residual turbidities measured during most of the test runs satisfied the Malaysian Guideline for Drinking Water Supplies. Key Words: Moringa oleifera, primary coagulant, coagulation, pilot plant, filtration.

  16. 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

  17. 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

  18. A modeling of elementary passes taking into account the firing angle in abrasive water jet machining of titanium alloy

    Science.gov (United States)

    Bui, Van-Hung; Gilles, Patrick; Cohen, Guillaume; Rubio, Walter

    2018-05-01

    The use of titanium alloys in the aeronautical and high technology domains is widespread. The high strength and the low mass are two outstanding characteristics of titanium alloys which permit to produce parts for these domains. As other hard materials, it is challenging to generate 3D surfaces (e.g. pockets) when using conventional cutting methods. The development of Abrasive Water Jet Machining (AWJM) technology shows the capability to cut any kind of materials and it seems to be a good solution for such titanium materials with low specific force, low deformation of parts and low thermal shocks. Applying this technology for generating 3D surfaces requires to adopt a modelling approach. However, a general methodology results in complex models due to a lot of parameters of the machining process and based on numerous experiments. This study introduces an extended geometry model of an elementary pass when changing the firing angle during machining Ti-6AL-4V titanium alloy with a given machine configuration. Several experiments are conducted to observe the influence of major kinematic operating parameters, i.e. jet inclination angle (α) (perpendicular to the feed direction) and traverse speed (Vf). The material exposure time and the erosion capability of abrasives particles are affected directly by a variation of the traverse speed (Vf) and firing angle (α). These variations lead to different erosion rates along the kerf profile characterized by the depth and width of cut. A comparison demonstrated an efficiency of the proposed model for depth and width of elementary passes. Based on knowledge of the influence of both firing angle and traverse speed on the elementary pass shape, the proposed model allows to develop the simulation of AWJM process and paves a way for milling flat bottom pockets and 3D complex shapes.

  19. Measurement and Elemental Analysis of Muria Sea Water for Primary Water of PWR

    International Nuclear Information System (INIS)

    Dwi Biyantoro; Kris Tri Basuki

    2007-01-01

    Treatment process of water of free mineral and study area of processing of sea water to meet the need of water cooling of PWR. Desalination is a separation process used to reduce the dissolved salt content of saline water. All desalination processes involve three liquid streams: the saline feedwater (brackish water or seawater), low salinity product water, and very saline concentrate. Seawater Muria Jepara is feed type 1 with content salt is low and relative suspension and weight metals is low. There are two types of membrane process used for desalination: reverse osmosis (RO) and electrodialysis (ED). The starting of process is difficult and expensive. However, starting since 1950, desalination process appear to be economically practical for ordinary use. There are 3 key elements significantly affect to the technical performance and long term characteristic type, i.e. 1) energy, 2) corrosivity of seawater, and 3) desalination process technology. These elements closely inter related and important for design improvements and technical performance optimization. From the analysis of sea water of gulf Muria Jepara was found as follows : Fe = 0.176 ± 0.012 ppm, Pb = 0.612 ± 0.017 ppm, Cd = 52.567 ± 0.750 ppm, Cu = 0.044 ± 0.005 ppm, Zn = 0.061 ± 0.003 ppm, Mn = 0.057 ± 0.003 ppm, Ca between = 365.256 - 368.654 ppm, Na between = 9572.000 - 9775.000 ppm, Mg between 759.000 - 779.00 ppm and Ni = 0.524 ± 0.005 ppm. Cost production of process using reverse osmosis as around 0.9 - 1 US$/m3, while using electrodialysis is around 1.2 US$/m3, and by using evaporation process or distillation process is around 1.4 - 1.6 US$/m3. (author)

  20. 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.)

  1. Water consumption in artificial desert oasis based on net primary productivity

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    Analysis of the water consumption is the basis for water allocation in oasis. However, the method of estimating oasis water consumption remains a great challenge. Based on net primary productivity (NPP) and the transpiration coefficient, a vegetation water consumption model was developed to estimate the water consumption in desert oasis in ERDAS environment. Our results demonstrated that the ecosystem in the middle reaches of the Heihe oasis consumed water of 18.41×108-21.9×108 m3 for irrigation. Without taking precipitation into account, the water consumption in farmland accounted for 77.1%-77.8% (or about 13.97×108-16.84×108 m3) of the oasis vegetation water consumption and in the farmland protection system accounting for 22%. The growing period precipitation in desert environments is about 7.02×108 m3, and the total annual precipitation is about 8.29×108 m3. The modeled water consumption of desert vegetation, however, is about 4.57×108 m3, equivalent to only 65% of the growing period precipitation or 55% of the total annual precipitation. The modeled value equals to the cumulative precipitation of greater than 5 mm, which is defined as the effective precipitation in arid desert.

  2. Shape of growing crystals of primary phases in autectic alloys of Fe - Fe2B and Ni - Ni3B systems

    International Nuclear Information System (INIS)

    Tavadze, F.N.; Garibashvili, V.I.; Nakaidze, Sh.G.

    1983-01-01

    Shapes of Fe 2 B and Ni 3 B crystal growth in eutectic Fe-B and Ni-B system alloys are considered. Iron hemiboride primary crystals take the form of a plane-face phase boundary and inherit a tetragonal prismatic lattice. After the crystal attains the critical size the dendritic branching occurs resulting in formation of a typical sceleton dendrite. Comparison of data obtained with entropy of melting for Fe 2 B and Ni 3 B borides shows that FeB crystals during the growth should take the spherical form. It is stated that the shape of growing crystals in Fe-Fe 2 B and Ni-Ni 2 B eutectic colonies is determined by the shape of borides

  3. Experimental and numerical investigation on under-water friction stir welding of armour grade AA2519-T87 aluminium alloy

    Directory of Open Access Journals (Sweden)

    S. Sree Sabari

    2016-08-01

    Full Text Available Friction stir welding (FSW is a promising welding process that can join age hardenable aluminium alloys with high joint efficiency. However, the thermal cycles experienced by the material to be joined during FSW resulted in the deterioration of mechanical properties due to the coarsening and dissolution of strengthening precipitates in the thermo-mechanical affected zone (TMAZ and heat affected zone (HAZ. Under water friction stir welding (UWFSW is a variant of FSW process which can maintain low heat input as well as constant heat input along the weld line. The heat conduction and dissipation during UWFSW controls the width of TMAZ and HAZ and also improves the joint properties. In this investigation, an attempt has been made to evaluate the mechanical properties and microstructural characteristics of AA2519-T87 aluminium alloy joints made by FSW and UWFSW processes. Finite element analysis has been used to estimate the temperature distribution and width of TMAZ region in both the joints and the results have been compared with experimental results and subsequently correlated with mechanical properties.

  4. 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

  5. BWR water chemistry guidelines and PWR primary water chemistry guidelines in Japan – Purpose and technical background

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, Hirotaka, E-mail: kawamuh@criepi.denken.or.jp [Central Research Institute of Electric Power Industry (Japan); Hirano, Hideo [Central Research Institute of Electric Power Industry (Japan); Katsumura, Yousuke [University of Tokyo (Japan); Uchida, Shunsuke [Tohoku University (Japan); Mizuno, Takayuki [Mie University (Japan); Kitajima, Hideaki; Tsuzuki, Yasuo [Japan Nuclear Safety Institute (Japan); Terachi, Takumi [Institute of Nuclear Safety System, Inc. (Japan); Nagase, Makoto; Usui, Naoshi [Hitachi-GE Nuclear Energy, Ltd. (Japan); Takagi, Junichi; Urata, Hidehiro [Toshiba Corporation (Japan); Shoda, Yasuhiko; Nishimura, Takao [Mitsubishi Heavy Industry, Ltd. (Japan)

    2016-12-01

    Highlights: • Framework of BWR/PWR water chemistry Guidelines in Japan are presented. • Guideline necessity, definitions, philosophy and technical background are mentioned. • Some guideline settings for control parameters and recommendations are explaines. • Chemistry strategy is also mentioned. - Abstract: After 40 years of light water reactor (LWR) operations in Japan, the sustainable development of water chemistry technologies has aimed to ensure the highest coolant system component integrity and fuel reliability performance for maintaining LWRs in the world; additionally, it aimed to achieve an excellent dose rate reduction. Although reasonable control and diagnostic parameters are utilized by each boiling water reactor (BWR) and pressurized water reactor (PWR) owner, it is recognized that specific values are not shared among everyone involved. To ensure the reliability of BWR and PWR operation and maintenance, relevant members of the Atomic Energy Society of Japan (AESJ) decided to establish guidelines for water chemistry. The Japanese BWR and PWR water chemistry guidelines provide strategies to improve material and fuel reliability performance as well as to reduce dosing rates. The guidelines also provide reasonable “control values”, “diagnostic values” and “action levels” for multiple parameters, and they stipulate responses when these levels are exceeded. Specifically, “conditioning parameters” are adopted in the Japanese PWR primary water chemistry guidelines. Good practices for operational conditions are also discussed with reference to long-term experience. This paper presents the purpose, technical background and framework of the preliminary water chemistry guidelines for Japanese BWRs and PWRs. It is expected that the guidelines will be helpful as an introduction to achieve safety and reliability during operations.

  6. Analysis of the corrosion products formed on Ti and a Ti-Pd alloy during exposure in hot water

    International Nuclear Information System (INIS)

    Olefjord, I.; Mattsson, H.

    1982-01-01

    This is a preliminary report dealing with the surface analysis of reaction products formed on Ti and a Ti-Pd alloy during their exposure in hot water. The compositions of the aqueous media were varied with respect to the dissolved oxygen and the content of chloride ions. The temperature was 60 0 C and the exposure times were 10 min. and 6 months. Work is in progress in which samples are exposed at 80 0 C and 95 0 C in the aqueous solutions. Surface analysis was also performed on a sample which had been exposed in water-saturated bentonite. It appears from te ESCA spectra that the oxide products formed on the surface consist of TiO 2 . The results also indicate that the thickness of the film formed at 60 0 C in water is in the range 50 A to 100 A. This is somewhat more than that obtained after exposure in water at room temperature. Exposure for 6 months increases the thickness of the oxide two to three times compared to that obtained during the short exposure at 60 0 C. The analyses of the samples that had been embedded in bentonite indicate that the surface reaction products are thinner than those found on the surface after exposure in an open vessel

  7. The chemistry and activity build up in the primary systems of pressurized water nuclear plants

    International Nuclear Information System (INIS)

    Darras, Raymond.

    1980-11-01

    After giving a background information on the present standards for the primary coolant in pressurized water nuclear reactors, the choice of particular chemical additives to the water is presented and their main properties are given; the various radioactivated products that are derived from these additives are also considered. The corrosion products transport through the whole primary circuit is then investigated. Two basically different types of processes, particularly about surface deposits, are characterized: that of suspended solids and that of soluble species, which are both carried by water. The physico-chemical data that rule the variations of solubilities for the more important elements are reviewed with details. From these data, the relation between corrosion products transport and radioactive contamination in primary circuits are examined, and this in the complex physico-chemical conditions of plant operation. Characteristic measurements, from operating power reactors, are also presented to illustrate the preceeding phenomena. Finally a chapter reviews the possible solutions against the radioactive contamination of the circuits and their surroundings: - a more adequate choice of materials, - a search for better surface treatment and application methods, - a better evaluation of the existing water conditioning, - an efficient filtration of the fluid, - the use of decontaminating processes [fr

  8. Flowing-water optical power meter for primary-standard, multi-kilowatt laser power measurements

    Science.gov (United States)

    Williams, P. A.; Hadler, J. A.; Cromer, C.; West, J.; Li, X.; Lehman, J. H.

    2018-06-01

    A primary-standard flowing-water optical power meter for measuring multi-kilowatt laser emission has been built and operated. The design and operational details of this primary standard are described, and a full uncertainty analysis is provided covering the measurement range from 1–10 kW with an expanded uncertainty of 1.2%. Validating measurements at 5 kW and 10 kW show agreement with other measurement techniques to within the measurement uncertainty. This work of the U.S. Government is not subject to U.S. copyright.

  9. 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)

  10. Evaluation of the use of metal alloy fuels in pressurized water reactors

    International Nuclear Information System (INIS)

    1990-01-01

    The project concentrated on model development. Reactor physics modeling involved establishing accurate models with PC versions of COMBINE and VENTURE. Fuel performance analysis will start with METAL- LIFE. In order to justify the change of fuel to metal alloy, large benefits will have to be found; the cost benefit reported is not sufficient. The fuel pin will be annular and contact the clad; the clad thickness will force the fuel to grow toward the central hole. This report reports: design improvements, neutronic model development, COBRA modifications, reactor kinetics model development, RELAP code, and fuel performance

  11. Study of alloy 600'S stress corrosion cracking mechanisms in high temperature water

    International Nuclear Information System (INIS)

    Rios, R.

    1994-06-01

    In order to better understand the mechanisms involved in Alloy 600's stress corrosion cracking in PWR environment, laboratory tests were performed. The influence of parameters pertinent to the mechanisms was studies : hydrogen and oxygen overpressures, local chemical composition, microstructure. The results show that neither hydrogen nor dissolution/oxidation, despite their respective roles in the process, are sufficient to account for experimental facts. SEM observation of micro-cleavage facets on specimens' fracture surfaces leads to pay attention to a new mechanism of corrosion/plasticity interactions. (author). 113 refs., 73 figs., 15 tabs., 4 annexes

  12. Study of alloy 600 (NC15Fe) stress corrosion cracking mechanisms in high temperature water

    International Nuclear Information System (INIS)

    Rios, Richard

    1993-01-01

    In order to better understand the mechanisms involved in Alloy 600's stress corrosion cracking in PWR environment, laboratory tests were performed. The influence of parameters pertinent to the mechanisms was studies: hydrogen and oxygen overpressures, local chemical composition, microstructure. The results show that neither hydrogen nor dissolution/oxidation, despite their respective roles in the process, are sufficient to account for experimental facts. SEM observation of micro-cleavage facets on specimens' fracture surfaces leads to pay attention to a new mechanism of corrosion/plasticity interactions. (author) [fr

  13. Some aspects of primary and secondary water chemistry in CANDU reactors

    International Nuclear Information System (INIS)

    LeSurf, J.E.

    1978-09-01

    A brief review of the water chemistry in various circuits of CANDU reactors is given. Then, five particular aspects of recent work are highlighted: (i) Radiation Field Growth: in-reactor and out-reactor studies have related water chemistry to corrosion product deposition on fuel sheaths and subsequent contamination of out-core surfaces. (ii) Metal Oxide Solubility: novel techniques are being used to measure the solubilities of metal oxides at primary circuit conditions. (iii) Decontamination: the use of heavy water as coolant in CANDU reactors led to the development of a unique decontamination strategy and technique, called CAN-DECON, which has attracted the attention of operators of light-water reactors. (iv) Steam Generator Corrosion: mathematical modelling of the water chemistry in the bulk and crevice regions of nuclear steam generators, supported by chemical experiments, has shown why sea water ingress from leaking condensers can be damaging, and has provided a rapid way to evaluate alternative boiler water chemistries. (v) Automatic Control of Feedwater Chemistry: on-line automatic chemical analysis and computer control of feedwater chemistry provides All Volatile Treatment for normal operation with pure feedwater, and carefully controlled sodium phosphate addition when there is detectable sea-water ingress from leaking condensers. (author)

  14. The Effect of Material Variability on Fatigue Behaviors of Low Alloy Steels in 310 .deg. C Deoxygenated Water

    International Nuclear Information System (INIS)

    Jang, Hun; Jang, Changheui; Kim, Insup; Cho, Hyunchul

    2008-01-01

    As environmental fatigue damage is one of the main crack initiation mechanisms in nuclear power plants (NPPs), it is most important factor to assess the integrity and safety of NPPs. So, based on extensive researches, argon nation laboratory (ANL) suggested the statistical model to predict fatigue life of low alloy steels (LASs) which are widely used as structural material in NPPs. Also, we reported the environmental fatigue behaviors of SA508 Gr.1a LAS. However, from comparison between our experimental fatigue data and ANL's statistical model, our fatigue life data showed poor agreement with the ANL's statistical model. In this regard, the additional low cycle fatigue (LCF) tests were performed in 310 .deg. C deoxygenated water, and compared with ANL's statistical model to evaluate reliability of the data. And then, the effect of material variability on the fatigue life of LASs was investigated through microstructure analysis

  15. Crack growth behaviour of low alloy steels for pressure boundary components under transient light water reactor operating conditions (CASTOC)

    International Nuclear Information System (INIS)

    Foehl, J.; Weissenberg, T.; Gomez-Briceno, D.; Lapena, J.; Ernestova, M.; Zamboch, M.; Seifert, H.P.; Ritter, S.; Roth, A.; Devrient, B.; Ehrnsten, U.

    2004-01-01

    The CASTOC project addresses environmentally assisted cracking (EAC) phenomena in low alloy steels used for pressure boundary components in both Western type boiling water reactors (BWR) and Russian type pressurised water reactors (VVER). It comprises the four work packages (WP): inter-laboratory comparison test (WP1); EAC behaviour under static load (WP2), EAC behaviour under cyclic load and load transients (WP3); evaluation of the results with regard to their relevance for components in practice (WP4). The use of sophisticated test facilities and measurement techniques for the on-line detection of crack advances have provided a more detailed understanding of the mechanisms of environmentally assisted cracking and provided quantitative data of crack growth rates as a function of loading events and time, respectively. The effect of several major parameters controlling EAC was investigated with particular emphasis on the transferability of the results to components in service. The obtained crack growth rate data were reflected on literature data and on commonly applied prediction curves as presented in the appropriate Code. At relevant stress intensity factors it could be shown that immediate cessation of growing cracks occurs after changing from cyclic to static load in high purity oxygenated BWR water and oxygen-free VVER water corresponding to steady state operation conditions. Susceptibility to environmentally assisted cracking under static load was observed for a heat affected zone material in oxygenated high purity water and also in base materials during a chloride transient representing BWR water condition below Action Level 1 of the EPRI Water Chemistry Guidelines according to the lectrical conductivity of the water but in the range of Action Level 2 according to the content of chlorides. Time based crack growth was also observed in one Russian type base material in oxygenated VVER water and in one Western type base material in oxygenated high purity BWR

  16. 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.)

  17. Preliminary study of radionuclide corrosion products in primary cooling water at RSG-GAS

    International Nuclear Information System (INIS)

    Lestari, D.E.; Pudjojanto, M.S.; Subiharto; Budi, S.

    1998-01-01

    Analysis of radionuclides emitting gamma rays at the primary cooling water at RSG-GAS has been carried out. The water coolant samples was performed using a low level background gamma spectrometer unit, including of high resolution of gamma detector HP-Ge Tennelec and Multichannel Analyzer (MCA) ADCAM 100 ORTEC. The result indicated Na-24 and Mn-56 radionuclides that may be as corrosion product and should studied deeply in the future. The expected activity concentration radionuclide for Mn-56 is lower than those written in the Safety Analysis Report (SAR), while for Na-24 is in agreement

  18. 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.

  19. Study on primary coolant system depressurization effect factor in pressurized water reactor

    International Nuclear Information System (INIS)

    Ji Duan; Cao Xuewu

    2006-01-01

    The progression of high-pressure core melting severe accident induced by very small break loss of coolant accident plus the loss of main feed water and auxiliary feed water failure is studied, and the entry condition and modes of primary cooling system depressurization during the severe accident are also estimated. The results show that the temperature below 650 degree C is preferable depressurization input temperature allowing recovery of core cooling, and the available and effective way to depressurize reactor cooling system and to arrest very small break loss of coolant accident sequences is activating pressurizer relief valves initially, then restoring the auxiliary feedwater and opening the steam generator relief valves. It can adequately reduce the primary pressure and keep the capacity loop of long-term core cooling. (authors)

  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. 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.)

  2. 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.

  3. Primary energy consumption of the dwelling with solar hot water system and biomass boiler

    International Nuclear Information System (INIS)

    Berković-Šubić, Mihaela; Rauch, Martina; Dović, Damir; Andrassy, Mladen

    2014-01-01

    Highlights: • Methodology for determing delivered and primary energy is developed. • Conventional and solar hot water system are analyzed. • Influence of system components, heat losses and energy consumption is explored. • Savings when using solar system in delivered energy is 30% and in primary 75%. • Dwelling with higher Q H,nd has 60% shorter payback period. - Abstract: This paper presents a new methodology, based on the energy performance of buildings Directive related European norms. It is developed to overcome ambiguities and incompleteness of these standards in determining the delivered and primary energy. The available procedures from the present “Algorithm for determining the energy demands and efficiency of technical systems in buildings”, normally used for energy performance certification of buildings, also allow detailed analyzes of the influence of particular system components on the overall system energy efficiency. The calculation example is given for a Croatian reference dwelling, equipped with a solar hot water system, backed up with a biomass boiler for space heating and domestic hot water purposes as a part of the dwelling energy performance certification. Calculations were performed for two cases corresponding to different levels of the dwelling thermal insulation with an appropriate heating system capacity, in order to investigate the influence of the building heat losses on the system design and energy consumption. The results are compared against those obtained for the conventional system with a gas boiler in terms of the primary energy consumption as well as of investment and operating costs. These results indicate great reduction in both delivered and primary energy consumption when a solar system with biomass boiler is used instead of the conventional one. Higher savings are obtained in the case of the dwelling with higher energy need for space heating. Such dwellings also have a shorter payback period than the ones with

  4. Evaluation of heat exchange performance for primary pressurized water cooler in HTTR

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Nakagawa, Shigeaki

    2006-01-01

    In High Temperature Engineering Test Reactor (HTTR), the rated thermal power of 30 MW, the generated heat at reactor core is finally dissipated at the air-cooler by way of the heat exchangers of the primary cooling system, such as the primary pressurized water cooler (PPWC) and the intermediate heat exchanger (IHX). The heat exchangers in the primary cooling system are required the heat exchange performance to remove reactor generated heat 30 MW under the condition of reactor coolant outlet temperature 850degC/950degC. Therefore, the heat exchanges are required to satisfy the design criteria of heat exchange performance. In this report, heat exchange performance data of the rise-to-power-up test and the in-service operation for the PPWC in the main cooling system was evaluated. Moreover, the evaluated values were compared with the design values, and it is confirmed that PPWC has the required heat exchange performance in the design. (author)

  5. 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

  6. 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

  7. Improving the submicro determination of vanadium in natural water using primary-secondary wavelength spectrophotometry

    International Nuclear Information System (INIS)

    Khongven Gao

    1999-01-01

    In acidic solution and in the presence of ammonium persulfate, the conventional reaction of vanadium(5) with gallic acid to form an orange complex has been used for the improvement of the determination of trace amounts of vanadium in water by the updated method named primary-secondary wavelength spectrophotometry. The results show that the analytical precision and accuracy were improved and gave higher determination sensitivity than ordinary spectrophotometry. The relative standard deviations were less than 5.6 % [ru

  8. HAZOP-study on heavy water research reactor primary cooling system

    International Nuclear Information System (INIS)

    Hashemi-Tilehnoee, M.; Pazirandeh, A.; Tashakor, S.

    2010-01-01

    By knowledge-based Hazard and Operability (HAZOP) technique, equipment malfunction and deficiencies in the primary cooling system of the generic heavy water research reactor are studied. This technique is used to identify the representative accident scenarios. The related Process Flow Drawing (PFD) is prepared as our study database for this plant. Since this facility is in the design stage, applying the results of HAZOP-study to PFD improves the safety of the plant.

  9. Extensive tissue-specific transcriptomic plasticity in maize primary roots upon water deficit.

    Science.gov (United States)

    Opitz, Nina; Marcon, Caroline; Paschold, Anja; Malik, Waqas Ahmed; Lithio, Andrew; Brandt, Ronny; Piepho, Hans-Peter; Nettleton, Dan; Hochholdinger, Frank

    2016-02-01

    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% of expressed genes were constitutively active in all tissues. In contrast, deficit-responsive genes (1915) were consistently regulated in all tissues, while >75% (1501 genes) were specifically regulated in a single root tissue. Water deficit-responsive genes were most numerous in the cortex of the mature root zone and in the elongation zone. The most prominent functional categories among differentially expressed genes in all tissues were 'transcriptional regulation' and 'hormone metabolism', indicating global reprogramming of cellular metabolism as an adaptation to water deficit. Additionally, the most significant transcriptomic changes in the root tip were associated with cell wall reorganization, leading to continued root growth despite water deficit conditions. This study provides insight into tissue-specific water deficit responses and will be a resource for future genetic analyses and breeding strategies to develop more drought-tolerant maize cultivars. © The Author 2015. Published by Oxford University Press on behalf of the Society for Experimental Biology.

  10. 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)

  11. Declarative virtual water maze learning and emotional fear conditioning in primary insomnia.

    Science.gov (United States)

    Kuhn, Marion; Hertenstein, Elisabeth; Feige, Bernd; Landmann, Nina; Spiegelhalder, Kai; Baglioni, Chiara; Hemmerling, Johanna; Durand, Diana; Frase, Lukas; Klöppel, Stefan; Riemann, Dieter; Nissen, Christoph

    2018-05-02

    Healthy sleep restores the brain's ability to adapt to novel input through memory formation based on activity-dependent refinements of the strength of neural transmission across synapses (synaptic plasticity). In line with this framework, patients with primary insomnia often report subjective memory impairment. However, investigations of memory performance did not produce conclusive results. The aim of this study was to further investigate memory performance in patients with primary insomnia in comparison to healthy controls, using two well-characterized learning tasks, a declarative virtual water maze task and emotional fear conditioning. Twenty patients with primary insomnia according to DSM-IV criteria (17 females, three males, 43.5 ± 13.0 years) and 20 good sleeper controls (17 females, three males, 41.7 ± 12.8 years) were investigated in a parallel-group study. All participants completed a hippocampus-dependent virtual Morris water maze task and amygdala-dependent classical fear conditioning. Patients with insomnia showed significantly delayed memory acquisition in the virtual water maze task, but no significant difference in fear acquisition compared with controls. These findings are consistent with the notion that memory processes that emerge from synaptic refinements in a hippocampal-neocortical network are particularly sensitive to chronic disruptions of sleep, while those in a basic emotional amygdala-dependent network may be more resilient. © 2018 European Sleep Research Society.

  12. 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)

  13. 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.

  14. Effect of cold work hardening on stress corrosion cracking of stainless steels in primary water of pressurized water reactors

    International Nuclear Information System (INIS)

    Raquet, O.; Herms, E.; Vaillant, F.; Couvant, T.; Boursier, J.M.

    2004-01-01

    A R and D program is carried out in CEA and EDF laboratories to investigate separately the effects of factors which could contribute to IASCC mechanism. In the framework of this study, the influence of cold work on SCC of ASSs in primary water is studied to supply additional knowledge concerning the contribution of radiation hardening on IASCC of ASSs. Solution annealed ASSs, essentially of type AISI 304(L) and AISI 316(L), are generally considered very resistant to SCC in nominal primary water. However, Constant Extension Rate Tests (CERTs), performed on cold pressed humped specimens in nominal primary water at 360 deg. C, reveal that these materials can exhibit a high SCC susceptibility: deepest cracks reach 1 mm (mean crack growth rate about 1 μm.h -1 ) and propagation is mainly intergranular for 304L and mainly transgranular for 316L. Indeed, work hardening in conjunction with high localized deformation can promote SCC. The influence of the nature of the cold work (shot peening, reaming, cold rolling, counter sinking, fatigue work hardening and tensile deformation) is investigated by means of screening CERTs performed with smooth specimens in 304L at 360 deg. C. For a given cold work hardening level, the susceptibility to crack initiation strongly depends on the cold working process, and no propagation is observed for a hardness level lower than 300 ±10 HV(0.49N). The propagation of cracks is observed only for dynamic loadings like CERT, traction/relaxation tests and crack growth rate tests performed with CT specimens under trapezoidal loading. Although crack initiation is observed for constant load and constant deformation tests, crack propagation do not seem to occur under these mechanical solicitations for 17000 hours of testing, even for hardness levels higher than 450 HV(0.49N). The mean crack growth rate increases when the hardness increases. An important R and D program is in progress to complement these results and to develop a SCC model for ASSs in

  15. Correlative characterization of primary Al{sub 3}(Sc,Zr) phase in an Al–Zn–Mg based alloy

    Energy Technology Data Exchange (ETDEWEB)

    Li, J.H., E-mail: jie-hua.li@hotmail.com [Institute of Casting Research, Montanuniversität Leoben, A-8700 Leoben (Austria); Wiessner, M. [Materials Center Leoben Forschung GmbH, A-8700 Leoben (Austria); Albu, M. [Institute for Electron Microscopy and Nanoanalysis, Graz University of Technology, Center for Electron Microscopy (Austria); Wurster, S. [Department of Materials Physics, Montanuniversität Leoben, Erich Schmid Institute of Materials Science of the Austrian Academy of Sciences, A-8700 Leoben (Austria); Sartory, B. [Materials Center Leoben Forschung GmbH, A-8700 Leoben (Austria); Hofer, F. [Institute for Electron Microscopy and Nanoanalysis, Graz University of Technology, Center for Electron Microscopy (Austria); Schumacher, P. [Institute of Casting Research, Montanuniversität Leoben, A-8700 Leoben (Austria); Austrian Foundry Research Institute, A-8700 Leoben (Austria)

    2015-04-15

    Three-dimensional electron backscatter diffraction, focused ion beam, transmission electron microscopy and energy filtered transmission electron microscopy were employed to investigate the structural information of primary Al{sub 3}(Sc,Zr) phase, i.e. size, shape, element distribution and orientation relationship with the α-Al matrix. It was found that (i) most primary Al{sub 3}(Sc,Zr) phases have a cubic three-dimensional morphology, with a size of about 6–10 μm, (ii) most primary Al{sub 3}(Sc,Zr) phases are located within the α-Al matrix, and exhibit a cube to cube orientation relationship with the α-Al matrix, and (iii) a layer by layer growth was observed within primary Al{sub 3}(Sc,Zr) phases. Al, Cu, Si and Fe are enriched in the α-Al matrix between the layers of cellular eutectic Al{sub 3}(Sc,Zr) phase, while Sc, Ti and Zr are enriched in small Al{sub 3}(Sc,Zr) phases. A peritectic reaction and subsequent eutectic reaction between Al{sub 3}Sc and Al was proposed to interpret the observed layer by layer growth. This paper demonstrates that the presence of impurities (Fe, Si, Cu, Ti) in the diffusion field surrounding the growing Al{sub 3}(Sc,Zr) particle enhances the heterogeneous nucleation of Al{sub 3}(Sc,Zr) phases. - Highlights: • Most fine cubic primary Al{sub 3}(Sc,Zr) phases were observed within the α-Al matrix. • A layer by layer growth within primary Al{sub 3}(Sc,Zr) phase was observed. • A peritectic and subsequent eutectic reaction between Al{sub 3}Sc and Al was proposed. • Impurities in diffusion fields enhance heterogeneous nucleation of Al{sub 3}(Sc,Zr)

  16. Morphology transition of the primary silicon particles in a hypereutectic A390 alloy in high pressure die casting.

    Science.gov (United States)

    Wang, J; Guo, Z; Song, J L; Hu, W X; Li, J C; Xiong, S M

    2017-11-03

    The microstructure of a high-pressure die-cast hypereutectic A390 alloy, including PSPs, pores, α-Al grains and Cu-rich phases, was characterized using synchrotron X-ray tomography, together with SEM, TEM and EBSD. The Cu-rich phases exhibited a net morphology and distributed at the boundaries of the α-Al grains, which in turn surrounded the PSPs. Statistical analysis of the reconstructed 1000 PSPs showed that both equivalent diameter and shape factor of the PSPs exhibited a unimodal distribution with peaks corresponding to 25 μm and 0.78, respectively.) PSPs morphology with multiple twinning were observed and morphological or growth transition of the PSPs from regular octahedral shape (with a shape factor of 0.85 was mainly caused by the constraint of the Cu-rich phases. In particular, the presence of the Cu-rich phases restricted the growth of the α-Al grains, inducing stress on the internal silicon particles, which caused multiple twinning occurrence with higher growth potential and consequently led to growth transitions of the PSPs.

  17. Development of a higher capacity, lower pressure drop steam/water separator with reduced primary-to-secondary spacing

    International Nuclear Information System (INIS)

    Pruster, W.P.; Kidwell, J.H.; Eaton, A.M.; Wall, J.R.

    1985-01-01

    The goal of this development effort was to double the steam flow capacity of an existing module steam/water separator design without significantly increasing the pressure drop while simultaneously minimizing the vertical distance (spacing) between the primary and secondary separation stages. The development work included extensive air/water and steam/water testing. The steam/water tests were performed at a common pressure of 300 psia (2.1 MPa) with comparable water and steam flows

  18. Pre-service primary school teachers’ abilities in explaining water and air pollution scientifically

    Science.gov (United States)

    Lukmannudin; Sopandi, W.; Sujana, A.; Sukardi, R.

    2018-05-01

    The purpose of this study is to determine the ability of pre-service primary school teachers (PSPST) in explaining the phenomenon of water and air pollution scientifically. The research method used descriptive method of analysis with qualitative approach. The respondents were PSPTP at 4th semester. This study used a four-tier instrument diagnostic test. The number of subjects was 84 PSPTP at Universitas Pendidikan Indonesia, Kampus Daerah Sumedang. The results demonstrate the ability of PSPST in explaining water and air pollution scientifically. The results show that only 6% of PSPST who are able to explain the phenomenon of water pollution and only 4% of PSPST who are able to explain the phenomenon of air pollution. The fact should be attention for PSPST because these understanding are crucial in the process of learning activities in the classroom.

  19. Differential scanning calorimetric study of the binding between native DNA and its primary water of hydration.

    Science.gov (United States)

    Marlowe, R. L.; Lukan, A. M.; Lee, S. A.; Anthony, L.; Rupprecht, A.

    1996-03-01

    Differential scanning calorimetry was used to measure the binding strength between calf-thymus DNA and its primary water of hydration. The specific heat of wet-spun films was found to have a broad endothermic transition near 80 ^oC and a sharp exothermic transition near 250 ^oC. The broad transition is believed to be mainly due to the breaking of the bonds of the strongly bound water of hydration. This transition was found to be reversible, as expected. Kissinger analysis indicates that the activation barrier for breaking the bonds of these water molecules is about 0.6 eV. The sharp transition appeared to be an indication of a thermal decomposition of the DNA. Samples taken above this transition lost mass, showed evidence of having melted, and had turned black in color. This transition is irreversible.

  20. Primary Water Chemistry Control at Units of Paks Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Schunk, J.; Pinter, G. Patek T.; Tilky, P.; Doma, A. [Paks Nuclear Power Plant Co. Ltd., Paks (Hungary); Osz, J. [Budapest University of Technology and Economics, Budapest (Hungary)

    2013-03-15

    The primary water chemistry of the four identical units of Paks Nuclear Power Plant has been developed based on Western type PWR units, taking into consideration some Russian modifications. The political changes in the 1990s have also influenced the water chemistry specifications and directions. At PWR units the transition operational modes have been developed while in case of WWER units - in lack of central uniform regulation - this question has become the competence and responsibility of each individual plant. This problem has resulted in separate water chemistry developments with a considerable time delay. The need for lifetime extensions worldwide has made the development of startup and shutdown chemistry procedures extremely important, since they considerably influence the long term and safe operation of plants. The uniformly structured limit value system, the principles applied for the system development, and the logic schemes for actions to be taken are discussed in the paper, both for normal operation and transition modes. (author)

  1. 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)

  2. Corrosion of aluminium alloy test coupons in water of spent fuel storage pool at RA reactor

    International Nuclear Information System (INIS)

    Pesic, M.; Maksin, T.; Jordanov, G.; Dobrijevic, R.

    2004-12-01

    Study on corrosion of aluminium cladding, of the TVR-S type of enriched uranium spent fuel elements of the research reactor RA in the storage water pool is examined in the framework nr the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) 'Corrosion of Research Reactor Clad-Clad Spent Fuel in Water' since 2002. Standard racks with aluminium coupons are exposed to water in the spent fuel pools of the research reactor RA. After predetermined exposure times along with periodic monitoring of the water parameters, the coupons are examined according to the strategy and the protocol supplied by the IAEA. Description of the standard corrosion racks, experimental protocols, test procedures, water quality monitoring and compilation of results of visual examination of corrosion effects are present in this article. (author)

  3. Determination of rhodium in metallic alloy and water samples using cloud point extraction coupled with spectrophotometric technique

    Science.gov (United States)

    Kassem, Mohammed A.; Amin, Alaa S.

    2015-02-01

    A new method to estimate rhodium in different samples at trace levels had been developed. Rhodium was complexed with 5-(4‧-nitro-2‧,6‧-dichlorophenylazo)-6-hydroxypyrimidine-2,4-dione (NDPHPD) as a complexing agent in an aqueous medium and concentrated by using Triton X-114 as a surfactant. The investigated rhodium complex was preconcentrated with cloud point extraction process using the nonionic surfactant Triton X-114 to extract rhodium complex from aqueous solutions at pH 4.75. After the phase separation at 50 °C, the surfactant-rich phase was heated again at 100 °C to remove water after decantation and the remaining phase was dissolved using 0.5 mL of acetonitrile. Under optimum conditions, the calibration curve was linear for the concentration range of 0.5-75 ng mL-1 and the detection limit was 0.15 ng mL-1 of the original solution. The enhancement factor of 500 was achieved for 250 mL samples containing the analyte and relative standard deviations were ⩽1.50%. The method was found to be highly selective, fairly sensitive, simple, rapid and economical and safely applied for rhodium determination in different complex materials such as synthetic mixture of alloys and environmental water samples.

  4. Effects of Chemistry Parameters of Primary Water affecting Leakage of Steam Generator Tube Cracks

    Energy Technology Data Exchange (ETDEWEB)

    Shin, D. M.; Cho, N. C.; Kang, Y. S.; Lee, K. H. [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    Degradation of steam generator (SG) tubes can affect pressure boundary tightness. As a defense-in-depth measure, primary to secondary leak monitoring program for steam generators is implemented, and operation is allowed under leakage limits in nuclear power plants. Chemistry parameters that affect steam generator tube leakage due to primary water stress corrosion cracking (PWSCC) are investigated in this study. Tube sleeves were installed to inhibit leakage and improve tube integrity as a part of maintenance methods. Steam generators occurred small leak during operation have been replaced with new steam generators according to plant maintenance strategies. The correlations between steam generator leakage and chemistry parameters are presented. Effects of primary water chemistry parameters on leakage from tube cracks were investigated for the steam generators experiencing small leak. Unit A experienced small leakage from steam generator tubes in the end of operation cycle. It was concluded that increased solubility of oxides due to high pHT could make leakage paths, and low boron concentration lead to less blockage in cracks. Increased dissolved hydrogen may retard crack propagations, but it did not reduce leak rate of the leaking steam generator. In order to inhibit and reduce leakage, pH{sub T} was controlled by servicing cation bed operation. The test results of decreasing pHT indicate low pHT can reduce leak rate of PWSCC cracks in the end of cycle.

  5. Effects of Chemistry Parameters of Primary Water affecting Leakage of Steam Generator Tube Cracks

    International Nuclear Information System (INIS)

    Shin, D. M.; Cho, N. C.; Kang, Y. S.; Lee, K. H.

    2016-01-01

    Degradation of steam generator (SG) tubes can affect pressure boundary tightness. As a defense-in-depth measure, primary to secondary leak monitoring program for steam generators is implemented, and operation is allowed under leakage limits in nuclear power plants. Chemistry parameters that affect steam generator tube leakage due to primary water stress corrosion cracking (PWSCC) are investigated in this study. Tube sleeves were installed to inhibit leakage and improve tube integrity as a part of maintenance methods. Steam generators occurred small leak during operation have been replaced with new steam generators according to plant maintenance strategies. The correlations between steam generator leakage and chemistry parameters are presented. Effects of primary water chemistry parameters on leakage from tube cracks were investigated for the steam generators experiencing small leak. Unit A experienced small leakage from steam generator tubes in the end of operation cycle. It was concluded that increased solubility of oxides due to high pHT could make leakage paths, and low boron concentration lead to less blockage in cracks. Increased dissolved hydrogen may retard crack propagations, but it did not reduce leak rate of the leaking steam generator. In order to inhibit and reduce leakage, pH_T was controlled by servicing cation bed operation. The test results of decreasing pHT indicate low pHT can reduce leak rate of PWSCC cracks in the end of cycle

  6. Influence of surface condition on the corrosion resistance of copper alloy condenser tubes in sea water

    Energy Technology Data Exchange (ETDEWEB)

    Sato, S; Nagata, K; Yamauchi, S

    1979-07-01

    Investigation was made on the influence of various surface conditions of aluminum brass tube. The corrosion behavior of aluminum brass tube, with nine kinds of surface conditions, was studied in stagnant 0.1N NaHCo/sub 3/ solution and flowing sea water (natural, Fe/sup + +/ containing and S/sup - -/ containing water). Surface treatments investigated contained bright annealing, special annealing to form carbon film, hot oxidizing and pickling. Anodic polarization measurements in 0.1N NaHCO/sub 3/ solution showed that the oxidized surface was superior and that the pickled surface was inferior. However, relation between these characteristics and corrosion resistance in sea water has not been established. Electrochemical characteristics in flowing sea water were dependent on the surface conditions in the very beginning of immersion time; nobler corrosion potential for the surface with carbon film, higher polarization resistance for the bright annealed and the oxidized surface, and faster decrease of polarization resistance in S/sup - -/ containing sea water for the pickled surface. However, these differences disappeared in the immersion time of only 2 to 7 days. It was revealed, by the statistical analysis on the corrosion depth in corrosion test in flowing sea water and in jet impingement test, that the corrosion behavior was not influenced by surface conditions, but was significantly influenced by quality of sea water and sponge ball cleaning. Sulfide ion of 0.05 ppm caused severe pitting corrosion, and sponge ball cleaning of 5 chances a week caused erosion corrosion. From above results, it was concluded that surface conditions of aluminum brass were not important to sea water corrosion, and that quality of sea water and operating condition such as sponge ball cleaning were more significant.

  7. Study of the contamination level in the primary circuit of a pressurised water reactor

    International Nuclear Information System (INIS)

    Gomit, J.-M.

    1980-07-01

    The purpose of this study is to work out a model which allows to predict the contamination level in the primary circuit of a pressurised water reactor (PWR). We assume that the passage of fission products from fuel (UO 2 ) to water takes place in two stages: a) the fission products created in the fuel diffuse and go out into the gap; b) owing to failure of clad, fission products stored in the gap diffuse into the water. A migration constant will correspond to each of these stages (ν(C) for fuel and ν(J) for gap). We have designed two models: - an empiric model in which constants ν(C) and ν(J) are adjusted from experiments CYRANO and BOUFFON carried out by the CEA; - a theoretical model to describe the physical mechanisms of migration inside the fuel. This leads us to introduce trapping and resolution probabilities. This models lead to a theoretical definition of ν(C). We have undertaken a second qualification of the empiric model using code PROFIP 3. Application to the Tihange reactor enabled us to get a good estimation of activity in the primary circuit and the number of failed rods during the first cycle [fr

  8. 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)

  9. Water use efficiency in a primary subtropical evergreen forest in Southwest China.

    Science.gov (United States)

    Song, Qing-Hai; Fei, Xue-Hai; Zhang, Yi-Ping; Sha, Li-Qing; Liu, Yun-Tong; Zhou, Wen-Jun; Wu, Chuan-Sheng; Lu, Zhi-Yun; Luo, Kang; Gao, Jin-Bo; Liu, Yu-Hong

    2017-02-20

    We calculated water use efficiency (WUE) using measures of gross primary production (GPP) and evapotranspiration (ET) from five years of continuous eddy covariance measurements (2009-2013) obtained over a primary subtropical evergreen broadleaved forest in southwestern China. Annual mean WUE exhibited a decreasing trend from 2009 to 2013, varying from ~2.28 to 2.68 g C kg H 2 O -1 . The multiyear average WUE was 2.48 ± 0.17 (mean ± standard deviation) g C kg H 2 O -1 . WUE increased greatly in the driest year (2009), due to a larger decline in ET than in GPP. At the diurnal scale, WUE in the wet season reached 5.1 g C kg H 2 O -1 in the early morning and 4.6 g C kg H 2 O -1 in the evening. WUE in the dry season reached 3.1 g C kg H 2 O -1 in the early morning and 2.7 g C kg H 2 O -1 in the evening. During the leaf emergence stage, the variation of WUE could be suitably explained by water-related variables (relative humidity (RH), soil water content at 100 cm (SWC_100)), solar radiation and the green index (Sgreen). These results revealed large variation in WUE at different time scales, highlighting the importance of individual site characteristics.

  10. Overview of the current National Primary Drinking Water Regulations and regulation development process

    International Nuclear Information System (INIS)

    Cotruvo, J.A.; Regelski, M.

    1989-01-01

    The promulgation of the National Primary Drinking Water Regulations (NPDWR) follows specific steps. First, the Advance Notice of Proposed Rule Making (ANPRM) is published. Second, the EPA, as mandated by the SDWA Amendments, proposes maximum contaminant levels (MCLs), (enforceable standards) and maximum contaminant level goals (MCLGs) simultaneously. The Office of Drinking Water developed a six-phase schedule that has attempted to parallel the SDWA-specified deadlines: Phase I - Voltile organic chemicals - July 8, 1987; Phase II - Synthetic organic chemicals and inorganic chemicals - June 1989, microbials and surface water treatment - June 1989, and Lead/copper - December 1988; Phase III - Radionuclides - December 1988; Phase IV - Disinfectants and disinfection by-products - June 1989; Phase V - Other inorganic chemicals, synthetic organic chemicals, and pesticides - June 1989; and Phase VI - 25 additional chemicals - January 199. In selecting contaminants for regulation, the most relevant criteria are (1) potential health risk; (2) ability to detect a contaminant in the drinking water; and (3) occurrence or potential occurrence in drinking water. The EPA uses a three category approach for setting maximum contaminant level goals for carcinogens: Category I, strong evidence of carcinogenicity-zero; Category II, equivocal evidence - reference dose (RfD) approach or 0.00001 to 0.000001 cancer risk range; and Category III, inadequate or no evidence from animal studies - RfD approach. 10 refs., 5 tabs

  11. Lead in drinking water: sampling in primary schools and preschools in south central Kansas.

    Science.gov (United States)

    Massey, Anne R; Steele, Janet E

    2012-03-01

    Studies in Philadelphia, New York City, Houston, Washington, DC, and Greenville, North Carolina, have revealed high lead levels in drinking water. Unlike urban areas, lead levels in drinking water in suburban and rural areas have not been adequately studied. In the study described in this article, drinking water in primary schools and preschools in five suburban and rural south central Kansas towns was sampled to determine if any exceeded the U.S. Environmental Protection Agency (U.S. EPA) guidance level for schools and child care facilities of 20 parts per billion (ppb). The results showed a total of 32.1% of the samples had detectable lead levels and 3.6% exceeded the U.S. EPA guidance level for schools and child care providers of 20 ppb. These results indicate that about one-third of the drinking water consumed by children age six and under in the five suburban and rural south central Kansas towns studied has some lead contamination, exposing these children to both short-term and long-term health risks. The authors suggest a need for increased surveillance of children's drinking water in these facilities.

  12. Inhibition of localized attack on the aluminium alloy AA 6351 in glycol/water solutions

    Energy Technology Data Exchange (ETDEWEB)

    Monticelli, C; Brunoro, G; Zucchi, F; Fagioli, F

    1989-06-01

    The objective of this work was to examine the feasibility of enhancing pitting resistance of AA 6351 (nominal composition: 1% Si, 0.6% Mg, 0.3% Mn, balance Al) by adding suitable inhibitors to the solutions. The compounds used were two inorganic salts: sodium molybdate and sodium tungstate and two derivatives of pyrimidine: 2-aminopyrimidine (2AP) and 2-hydroxypyrimidine (2HP). The inhibiting efficiencies of these substances were tested by both short-time electrochemical tests (galvanic coupling tests and polarization curves) and long-time immersions under experimental conditions causing the localized attack. Molybdate, tungstate and, to some extent, also 2AP efficiently inhibit AA 6351 localized corrosion in degraded solutions at 80/sup 0/C and in pure boiling solutions, for long exposure periods. The short-time electrochemical tests suggest that molybdate and tungstate are able to retard the electrochemical processes occurring on both the aluminium alloy and the small copper cathodic area produced by copper deposition. On the other hand, the 2AP efficiency is attributed to some complexing capability of this pyrimidine derivative towards dissolved copper ions, that are stabilized in solution. 2HP does not prevent AA 6351 localized attack. (orig./MM).

  13. Chemical aspects of fission product transport in the primary circuit of a light water reactor

    International Nuclear Information System (INIS)

    Bowsher, B.R.; Dickinson, S.; Nichols, A.L.; Ogden, J.S.; Potter, P.E.

    1985-01-01

    The transport and fission products in the primary circuit of a light water reactor are of fundamental importance in assessing the consequences of severe accidents. Recent experimental studies have concentrated upon the behaviour of simulant fission product species such as caesium iodide, caesium hydroxide and tellurium, in terms of their vapour deposition characteristics onto metals representative of primary circuit materials. An induction furnace has been used to generate high-density/structural materials aerosols for subsequent analysis, and similar equipment has been incorporated into a glove-box to study lightly-irradiated UO/sub 2/ clad in Zircaloy. Analytical techniques are being developed to assist in the identification of fission product chemical species released from the fuel at temperatures from 1000 to 2500 0 C. Matrix isolation-infrared spectroscopy has been used to identify species in the vapour phase, and specific data using this technique are reported

  14. Chemical aspects of fission product transport in the primary circuit of a light water reactor

    International Nuclear Information System (INIS)

    Bowsher, B.R.; Dickinson, S.; Nichols, A.L.; Ogden, J.S.; Potter, P.E.

    1985-01-01

    The transport and deposition of fission products in the primary circuit of a light water reactor are of fundamental importance in assessing the consequences of severe accidents. Recent experimental studies have concentrated upon the behavior of simulant fission product species such as cesium iodide, cesium hydroxide and tellurium, in terms of their vapor deposition characteristics onto metals representative of primary circuit materials. An induction furnace has been used to generate high density/structural materials aerosols for subsequent analysis, and similar equipment has been incorporated into a glove-box to study lightly-irradiated UO 2 clad in Zircaloy. Analytical techniques are being developed to assist in the identification of fission product chemical species released from the fuel at temperatures from 1000 to 2500 0 C. Matrix isolation-infrared spectroscopy has been used to identify species in the vapor phase, and specific data using this technique are reported

  15. 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

  16. 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

  17. Role of electromagnetic filter in limitating radioactivity in the primary circuits of light water reactors

    International Nuclear Information System (INIS)

    Dolle, L.

    1978-01-01

    High temperature electromagnetic filtration of particulate corrosion products can be carried out with discharges up to 5% of the cooling flow rate. It allows efficient extraction of particulate matter which rate constants required for considerable reduction of activable crud deposition in the core. The paper holds a review of the preventing operation in the primary circuit of a PWR, and reports experimental results of efficiency measurments with an electromagnetic filter set in out-of-pile and in-pile pressurized water loops. The notable efficiencies towards radioactive fine grain and colloidal matter justify more extensive reactor scale application experiments. (author)

  18. Role of electromagnetic filter in limitating radioactivity in the primary circuits of light water reactors

    International Nuclear Information System (INIS)

    Dolle, L.

    1978-01-01

    High temperature electromagnetic filtration of particulate corrosion products can be carried out with discharges up to 5% of the cooling flow rate. It allows efficient extraction of particulate matter with rate constants required for considerable reduction of activable crud deposition in the core. The paper holds a review of the preventing operation in the primary circuit of a PWR, and reports experimental results of efficiency measurements with an electromagnetic filter set in out-of-pile and in-pile pressurized water loops. The notable efficiencies towards radioactive fine grain and colloidal matter justify more extensive reactor scale application experiments

  19. Designing a water calorimeter as primary standard of gamma rays at IPEN/CNEN

    International Nuclear Information System (INIS)

    Cintra, Felipe B. de; Caldas, Linda V.E.

    2013-01-01

    This work aims to describe the present stage and the next steps of the development of a water calorimeter of the Calibration Laboratory of IPEN/CNEN. This calorimeter will be used as a primary standard of gamma ray sources at the laboratory. Between the design and the construction step it will be shown how this model was chosen and how it is modeled virtually with computer simulation. The two main codes used, MCNP and Fluent, to characterize the prototype before its construction are presented. (author)

  20. Research and Service Experience with Environmentally-Assisted Cracking in Carbon and Low-Alloy Steels in High-Temperature Water

    International Nuclear Information System (INIS)

    Seifert, Hans-Peter; Ritter, Stefan

    2005-11-01

    The most relevant aspects of research and service experience with environmentally-assisted cracking (EAC) of carbon (C) and low-alloy steels (LAS) in high-temperature (HT) water are reviewed, with special emphasis on the primary pressure boundary components of boiling water reactors (BWRs). The main factors controlling the susceptibility to EAC under light water reactor (LWR) conditions are discussed with respect to crack initiation and crack growth. The adequacy and conservatism of the current BWRVIP-60 stress corrosion cracking (SCC) disposition lines (DLs), ASME III fatigue design curves, and ASME XI reference fatigue crack growth curves, as well as of the GE EAC crack growth model are evaluated in the context of recent research results. The operating experience is summarized and compared to the experimental/mechanistic background knowledge. Finally, open questions and possible topics for further research are identified. Laboratory investigations revealed significant effects of simulated reactor environments on fatigue crack initiation/growth, as well as the possibility of SCC crack growth for certain specific critical combinations of environmental, material and loading parameters. During the last three decades, the major factors of influence and EAC susceptibility conditions have been readily identified. Most parameter effects on EAC initiation and growth are adequately known with acceptable reproducibility and reasonably understood by mechanistic models. Tools for incorporating environmental effects in ASME III fatigue design curves have been developed/qualified and should be applied in spite of the high degree of conservatism in fatigue evaluation procedures. The BWRVIP-60 SCC DLs and ASME XI reference fatigue crack growth curves are usually conservative and adequate under most BWR operation circumstances. The operating experience of C and LAS primary pressure-boundary components in LWRs is very good worldwide. However, isolated instances of EAC have occurred

  1. Research and Service Experience with Environmentally-Assisted Cracking in Carbon and Low-Alloy Steels in High-Temperature Water

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, Hans-Peter; Ritter, Stefan [Paul Scherrer Inst., Laboratory for Materials Behaviour, Villigen (Switzerland). Nuclear Energy and Safety Research Dept.

    2005-11-15

    The most relevant aspects of research and service experience with environmentally-assisted cracking (EAC) of carbon (C) and low-alloy steels (LAS) in high-temperature (HT) water are reviewed, with special emphasis on the primary pressure boundary components of boiling water reactors (BWRs). The main factors controlling the susceptibility to EAC under light water reactor (LWR) conditions are discussed with respect to crack initiation and crack growth. The adequacy and conservatism of the current BWRVIP-60 stress corrosion cracking (SCC) disposition lines (DLs), ASME III fatigue design curves, and ASME XI reference fatigue crack growth curves, as well as of the GE EAC crack growth model are evaluated in the context of recent research results. The operating experience is summarized and compared to the experimental/mechanistic background knowledge. Finally, open questions and possible topics for further research are identified. Laboratory investigations revealed significant effects of simulated reactor environments on fatigue crack initiation/growth, as well as the possibility of SCC crack growth for certain specific critical combinations of environmental, material and loading parameters. During the last three decades, the major factors of influence and EAC susceptibility conditions have been readily identified. Most parameter effects on EAC initiation and growth are adequately known with acceptable reproducibility and reasonably understood by mechanistic models. Tools for incorporating environmental effects in ASME III fatigue design curves have been developed/qualified and should be applied in spite of the high degree of conservatism in fatigue evaluation procedures. The BWRVIP-60 SCC DLs and ASME XI reference fatigue crack growth curves are usually conservative and adequate under most BWR operation circumstances. The operating experience of C and LAS primary pressure-boundary components in LWRs is very good worldwide. However, isolated instances of EAC have occurred

  2. Carbon and oxygen dynamics on the Louisiana continental shelf: role of water column primary production and respiration

    Science.gov (United States)

    We conducted a multi-year study of the Louisiana continental shelf (LCS) to better understand the linkages between water column net metabolism and the formation of hypoxia (dissolved oxygen respiration (R) and primary p...

  3. Primary productivity, phytoplankton standing crop and physico-chemical characteristics of the Antarctic and adjacent central Indian Ocean waters

    Digital Repository Service at National Institute of Oceanography (India)

    JiyalalRam, M.

    Primary productivity, phytoplankton pigments and physico-chemical properties were studied in Antarctic waters and adjoining Indian Ocean between 11 degrees and 67 degrees E longitudes from polynya region (60 degrees S) to equator during the austral...

  4. 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.

  5. 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  

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

    African Journals Online (AJOL)

    driniev

    2004-07-03

    Jul 3, 2004 ... aluminium composite (MAlC) under flow conditions ... Behaviour of the composite in water is under significant influence of pH, which affects its efficacy and .... 1.2 x 1.2 mm, ∅ 0.6) with MAl, by means of a special gas burner.

  7. Characterization and Analysis of Ionic Zinc Alloy Running a Potential Dynamic Polarization in Sea Water

    International Nuclear Information System (INIS)

    Zaifol Samsu; Muhammad Daud; Siti Radiah Mohd Kamarudin

    2011-01-01

    Zinc is an active metal. The reactive nature of zinc allows it to be used for sacrificial anode in cathodic protection systems by electrically coupled to the protected metal. Zinc is especially well suited for cathodic protection on ships that move between salt water and harbors in brackish rivers or estuaries (1). Zinc anodes also are used to protect ballast tank, heat exchangers, and many mechanical components on ships, coastal power plants, and similar structures. Cathodic protective by zinc is used in sea water, brackish water, fresh water, and in some soil. The relative reactivity of zinc and its ability to attract oxidation to itself makes it an efficient sacrificial anode in cathodic protection (2). For example, cathodic protection of a buried pipeline can be achieved by connecting anodes made from zinc to the pipe. Zinc acts as the anode (negative terminus) by slowly corroding away as it passes electric current to the steel pipeline. When exposed to environment containing halide ions, of which the chloride (Cl-) is the most frequently encountered in service, the oxide film breaks down at specific points leading to the formation of pits on the zinc surface (3). (author)

  8. 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.)

  9. TRANP - a computer code for digital simulation of steady - state and transient behavior of a pressurizer water reactor primary circuit

    International Nuclear Information System (INIS)

    Chalhoub, E.S.

    1980-09-01

    A digital computer code TRANP was developed to simulate the steady-state and transient behavior of a pressurizer water reactor primary circuit. The development of this code was based on the combining of three codes already developed for the simulation of a PWR core, a pressurizer, a steam generator and a main coolant pump, representing the primary circuit components. (Author) [pt

  10. Corrosion behavior of AISI 4130 steel alloy in ethylene glycol–water mixture in presence of molybdate

    International Nuclear Information System (INIS)

    Danaee, I.; Niknejad Khomami, M.; Attar, A.A.

    2012-01-01

    The electrochemical behavior of steel alloy in ethylene glycol–water mixture with different concentrations was investigated by polarization curves, AC impedance measurements, current transient and atomic force microscopy. The results obtained showed that corrosion rate was decreased with increasing ethylene glycol concentration. The effect of molybdate as inhibitor was studied and high inhibition efficiency was obtained. It was found that surface passivation was occurred in presence of inhibitor. The inhibiting effect of the molybdate was explained on the basis of the competitive adsorption between the inorganic anions and the aggressive Cl − ions and the adsorption isotherm basically obeys the Langmuir adsorption isotherm. Thermodynamic parameters for steel corrosion and inhibitor adsorption were determined and reveal that the adsorption process is spontaneous. Also phenomenon of both physical and chemical adsorption is proposed. -- Highlights: ► Corrosion rate was decreased with increasing ethylene glycol concentration. ► High inhibition efficiency was obtained for molybdate. ► Surface passivation was occurred in presence of inhibitor. ► The adsorption isotherm basically obeys the Langmuir adsorption isotherm.

  11. Corrosion behaviour of material no. 1. 4539 and nickel based alloys in gas waters. Korrosionsverhalten des Werkstoffs 1. 4539 und von Nickelbasis-Legierungen in Gaswaessern

    Energy Technology Data Exchange (ETDEWEB)

    Rolle, D [Didier Saeurebau GmbH, Koenigswinter (Germany); Buehler, H E [Didier-Werke AG, Anlagentechnik, Wiesbaden (Germany); Kalfa, H

    1993-01-01

    Laboratory tests with synthetic gas waters containing the gases ammonia, carbon dioxide, hydrogen sulphide and hydrogen cyanide were carried out in order to examine the influence of medium components on the corrosion of material No. 1.4539 and nickel based alloys Hastelloy C-4, C-22 and C-276. Hydrogen sulfide was identified as the decisive component for corrosion. For stainless steel corrosion rates of about 2 mm.a[sup -1] were already found at 50deg C in a critical pH-range with sulfide concentrations > 2%. As cyanide stimulates corrosion by dissolving sulfide surface layers by complexation of the iron ions, an increased material loss rate per unit area was found in the critical range with increasing cyanide concentration. The much more stable nickel based alloys only revealed considerable weight losses after being exposed in the autoclave at 100deg C. The graduation of the loss rates C-22 > C-4 > C-276 can be explained by the different contents of high grade alloy elements. The testing of nickel based alloys of the Hastelloy type and of material No. 1.4539 and 1.4571 by means of the dynamic tensile test (CERT-method) revealed no risks of stress corrosion cracking in the tested media. (orig.).

  12. 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.

  13. Modeling and experimental study of oil/water contact angle on biomimetic micro-parallel-patterned self-cleaning surfaces of selected alloys used in water industry

    Energy Technology Data Exchange (ETDEWEB)

    Nickelsen, Simin; Moghadam, Afsaneh Dorri, E-mail: afsaneh@uwm.edu; Ferguson, J.B.; Rohatgi, Pradeep

    2015-10-30

    Graphical abstract: - Highlights: • Wetting behavior of four metallic materials as a function of surface roughness has been studied. • A model to predict the abrasive particle size and water/oil contact angles relationship is proposed. • Active wetting regime is determined in different materials using the proposed model. - Abstract: In the present study, the wetting behavior of surfaces of various common metallic materials used in the water industry including C84400 brass, commercially pure aluminum (99.0% pure), Nickle–Molybdenum alloy (Hastelloy C22), and 316 Stainless Steel prepared by mechanical abrasion and contact angles of several materials after mechanical abrasion were measured. A model to estimate roughness factor, R{sub f}, and fraction of solid/oil interface, ƒ{sub so}, for surfaces prepared by mechanical abrasion is proposed based on the assumption that abrasive particles acting on a metallic surface would result in scratches parallel to each other and each scratch would have a semi-round cross-section. The model geometrically describes the relation between sandpaper particle size and water/oil contact angle predicted by both the Wenzel and Cassie–Baxter contact type, which can then be used for comparison with experimental data to find which regime is active. Results show that brass and Hastelloy followed Cassie–Baxter behavior, aluminum followed Wenzel behavior and stainless steel exhibited a transition from Wenzel to Cassie–Baxter. Microstructural studies have also been done to rule out effects beyond the Wenzel and Cassie–Baxter theories such as size of structural details.

  14. 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.

  15. Genome-wide transcriptome analysis of soybean primary root under varying water-deficit conditions.

    Science.gov (United States)

    Song, Li; Prince, Silvas; Valliyodan, Babu; Joshi, Trupti; Maldonado dos Santos, Joao V; Wang, Jiaojiao; Lin, Li; Wan, Jinrong; Wang, Yongqin; Xu, Dong; Nguyen, Henry T

    2016-01-15

    Soybean is a major crop that provides an important source of protein and oil to humans and animals, but its production can be dramatically decreased by the occurrence of drought stress. Soybeans can survive drought stress if there is a robust and deep root system at the early vegetative growth stage. However, little is known about the genome-wide molecular mechanisms contributing to soybean root system architecture. This study was performed to gain knowledge on transcriptome changes and related molecular mechanisms contributing to soybean root development under water limited conditions. The soybean Williams 82 genotype was subjected to very mild stress (VMS), mild stress (MS) and severe stress (SS) conditions, as well as recovery from the severe stress after re-watering (SR). In total, 6,609 genes in the roots showed differential expression patterns in response to different water-deficit stress levels. Genes involved in hormone (Auxin/Ethylene), carbohydrate, and cell wall-related metabolism (XTH/lipid/flavonoids/lignin) pathways were differentially regulated in the soybean root system. Several transcription factors (TFs) regulating root growth and responses under varying water-deficit conditions were identified and the expression patterns of six TFs were found to be common across the stress levels. Further analysis on the whole plant level led to the finding of tissue-specific or water-deficit levels specific regulation of transcription factors. Analysis of the over-represented motif of different gene groups revealed several new cis-elements associated with different levels of water deficit. The expression patterns of 18 genes were confirmed byquantitative reverse transcription polymerase chain reaction method and demonstrated the accuracy and effectiveness of RNA-Seq. The primary root specific transcriptome in soybean can enable a better understanding of the root response to water deficit conditions. The genes detected in root tissues that were associated with

  16. Influence of thermal aging on primary water stress corrosion cracking of cast duplex stainless steels

    International Nuclear Information System (INIS)

    Yamada, Takuyo; Negishi, Kazuo; Totsuka, Nobuo; Nakajima, Nobuo

    2001-01-01

    In order to evaluate the SCC susceptibility of cast duplex stainless steels which are often used for the main coolant piping of pressurized water reactors (PWRs), the slow strain rate test (SSRT) and the constant load test (CLT) of the materials were performed in simulated primary water at 360degC. The stainless steel contains ferritic phase with ranging from 8 to 23% and its mechanical properties are affected by long time thermal aging. Therefore, we paid attention to the influence of its ferrite content and thermal aging on the SCC susceptibility of this stainless steel and prepared three kinds of specimen with different ferrite contents (23%, 15% and 8%). The reduction in area observed by the SSRT in simulated primary water at 360degC was smaller than that obtained by the tensile test in air at the same temperature. Microcracks were observed on the unaged specimen surfaces and aged ones at 400degC for 10,000 hours after 3,000 hours of the CLT with the load condition of two times of yield strength. The SCC susceptibility was evaluated by reduction ratio defined by the ratio of the reduction in area by the SSRT to that by the tensile test. The reduction ratio was not clear for low ferrite specimens, but apparently decreased with increasing aging time for the specimen with 23% ferrite. This change by aging time can be explained as follows: (1) the brittle fracture in the unaged specimens is mainly caused by quasi-cleavage fracture in austenitic phase. (2) After aging, it becomes a mixture of quasi-cleavage fracture in both austenitic and ferritic phases and phase boundary fracture of both phases. (author)

  17. Influence of thermal aging on primary water stress corrosion cracking of cast duplex stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Yamada, Takuyo; Negishi, Kazuo; Totsuka, Nobuo; Nakajima, Nobuo [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2001-09-01

    In order to evaluate the SCC susceptibility of cast duplex stainless steels which are often used for the main coolant piping of pressurized water reactors (PWRs), the slow strain rate test (SSRT) and the constant load test (CLT) of the materials were performed in simulated primary water at 360degC. The stainless steel contains ferritic phase with ranging from 8 to 23% and its mechanical properties are affected by long time thermal aging. Therefore, we paid attention to the influence of its ferrite content and thermal aging on the SCC susceptibility of this stainless steel and prepared three kinds of specimen with different ferrite contents (23%, 15% and 8%). The reduction in area observed by the SSRT in simulated primary water at 360degC was smaller than that obtained by the tensile test in air at the same temperature. Microcracks were observed on the unaged specimen surfaces and aged ones at 400degC for 10,000 hours after 3,000 hours of the CLT with the load condition of two times of yield strength. The SCC susceptibility was evaluated by reduction ratio defined by the ratio of the reduction in area by the SSRT to that by the tensile test. The reduction ratio was not clear for low ferrite specimens, but apparently decreased with increasing aging time for the specimen with 23% ferrite. This change by aging time can be explained as follows: (1) the brittle fracture in the unaged specimens is mainly caused by quasi-cleavage fracture in austenitic phase. (2) After aging, it becomes a mixture of quasi-cleavage fracture in both austenitic and ferritic phases and phase boundary fracture of both phases. (author)

  18. Primary water chemistry control at units of Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Schunk, J.; Patek, G.; Pinter, T.; Tilky, P.; Doma, A.; Osz, J.

    2010-01-01

    The primary water chemistry of the four identical units of Paks Nuclear Power Plant has been developed based on Western-type PWR units, taking into consideration some Soviet-Russian modifications. The political changes in 90s have also influenced the water chemistry specifications and directions. At PWR units the transition operational modes have been developed while in case of VVER units - in lack of central uniform regulation - this question has become the competence and responsibility of each individual plant. This problem has resulted in separate water chemistry developments with a considerable time delay. The needs for life-time extensions all over the World have made the development of start-up and shut-down chemistry procedures extremely important, since they considerably influence the long term and safe operation of plants. The uniformly structured limit value system, the principles applied for the system development, and the logic schemes for actions to be taken are discussed in the paper, both for normal operation and transition modes. (author)

  19. Water chemical control of the TRIGA IPR-R1 reactor primary cooling system

    International Nuclear Information System (INIS)

    Auler, Lucia M.L.A; Chaves, Renata D.A.; Palmieri, Helena E.L.; Menezes, Maria Angela de B.C.; Oliveira, Paulo F.; Kastner, Geraldo F.; Damazio, Ilza; Fagundes, Oliene dos R.; Cintra, Maria Olivia C.; Andrade, Geraldo V. de; Amaral, Angela M.; Franco, Milton B.; Fortes, Flavio; Gomes, Nilton Carlos; Vidal, Andrea; Maretti Junior, Fausto; Knupp, Eliana A.N.; Souza, Wagner de; Guedes, Joao B.; Furtado, Renato C.S.

    2013-01-01

    The TRIGA Mark I IPR-R1 reactor located at CDTN/CNEN has been in operation and contributed to research and with services to society since 1960. Is has been used in several activities such as nuclear power plant operation, graduate and post-graduate training courses, isotope production, and as an analytical irradiation tool of different types of samples. Among the several structural and operational safety requirements is the chemical quality control of the primary circuit cooling water. The aim of this work was to check the cooling water quality from the pool reactor. A water sampling plan was proposed (May, 2011 - June, 2012) and presents the results obtained in this period. The natural radioactivity level as gross alpha and gross beta activity and other chemical parameters (pH and electric conductivity) of the samples were analyzed. Some instrumental techniques were used: potentiometric methods (pH), conductometric methods (electrical conductivity, EC) and gross α and gross β proportional counting system). (author)

  20. Screening and Quantification of Aliphatic Primary Alkyl Corrosion Inhibitor Amines in Water Samples by Paper Spray Mass Spectrometry.

    Science.gov (United States)

    Jjunju, Fred P M; Maher, Simon; Damon, Deidre E; Barrett, Richard M; Syed, S U; Heeren, Ron M A; Taylor, Stephen; Badu-Tawiah, Abraham K

    2016-01-19

    Direct analysis and identification of long chain aliphatic primary diamine Duomeen O (n-oleyl-1,3-diaminopropane), corrosion inhibitor in raw water samples taken from a large medium pressure water tube boiler plant water samples at low LODs (corrosion inhibitors in an industrial water boiler plant and other related samples in the water treatment industry. This approach was applied for the analysis of three complex water samples including feedwater, condensate water, and boiler water, all collected from large medium pressure (MP) water tube boiler plants, known to be dosed with varying amounts of polyamine and amine corrosion inhibitor components. Polyamine chemistry is widely used for example in large high pressure (HP) boilers operating in municipal waste and recycling facilities to prevent corrosion of metals. The samples used in this study are from such a facility in Coventry waste treatment facility, U.K., which has 3 × 40 tonne/hour boilers operating at 17.5 bar.

  1. Influence of Variously Modified Surface of Aluminium Alloy on the Effect of Pulsating Water Jet

    Czech Academy of Sciences Publication Activity Database

    Klich, Jiří; Klichová, Dagmar; Foldyna, Vladimír; Hlaváček, Petr; Foldyna, Josef

    2017-01-01

    Roč. 63, č. 10 (2017), s. 577-582 ISSN 0039-2480 R&D Projects: GA MPO(CZ) FV10446; GA MŠk(CZ) LO1406; GA MŠk ED2.1.00/03.0082 Institutional support: RVO:68145535 Keywords : pulsating water jet * surface topography * material erosion Subject RIV: JQ - Machines ; Tools OBOR OECD: Mechanical engineering Impact factor: 0.914, year: 2016 http://ojs.sv-jme.eu/index.php/sv-jme/ article /view/sv-jme.2017.4356

  2. Alloy materials

    Energy Technology Data Exchange (ETDEWEB)

    Hans Thieme, Cornelis Leo (Westborough, MA); Thompson, Elliott D. (Coventry, RI); Fritzemeier, Leslie G. (Acton, MA); Cameron, Robert D. (Franklin, MA); Siegal, Edward J. (Malden, MA)

    2002-01-01

    An alloy that contains at least two metals and can be used as a substrate for a superconductor is disclosed. The alloy can contain an oxide former. The alloy can have a biaxial or cube texture. The substrate can be used in a multilayer superconductor, which can further include one or more buffer layers disposed between the substrate and the superconductor material. The alloys can be made a by process that involves first rolling the alloy then annealing the alloy. A relatively large volume percentage of the alloy can be formed of grains having a biaxial or cube texture.

  3. 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.

  4. Access to water and sanitation facilities in primary schools: A neglected educational crisis in Ngamiland district in Botswana

    Science.gov (United States)

    Ngwenya, B. N.; Thakadu, O. T.; Phaladze, N. A.; Bolaane, B.

    2018-06-01

    In developing countries, the sanitation and hygiene provision often receives limited resources compared to the water supply. However, water supply benefits tend to diminish if improved sanitation and hygiene are neglected. This paper presents findings of a situational analysis of water supply, sanitation and hygiene infrastructure and their utilization in three primary schools in north-western Botswana. The overall objective of the paper is to determine access and functionality of water supply, sanitation and hygiene infrastructure in three primary schools. The specific objectives are: a) Learners' perspective of their water and sanitation facilities and b) gendered utilization of sanitation and hygiene facilities. Data were collected through a face-to-face administered social survey tool to 286 learners selected through proportionate stratified random sampling from three purposively selected villages in the middle and lower Okavango Delta. Findings indicate that standpipes provide 96% of potable water supply. However, the majority (65% of leaners) indicated that they 'sometimes' experienced water shortage due to dry/nonfunctioning taps/pumps and leaks/wastage. Overall, schools have relatively sufficient sanitation facilities consisting of both water borne toilets and VIP latrines. The major sanitation gap identified was that 80% flush toilets hardly work, while 77% of VIP toilets were in disrepair. Furthermore, poor water supply compromised hand washing with 65.7% learners "always" washing their hands if school standpipes had water, while the majority did not wash hands if standpipes were dry. The study concluded that availability of sanitation infrastructure does not necessarily translate into utilization in the study area due to multiple problems, such as lack of personal hygiene supplies (regular toilet paper and hand washing detergents), privacy issues and recurring water problems. The chronicity of inadequate water, sanitation and hygiene infrastructure in

  5. Fatigue crack growth threshold of austenitic stainless steels in simulated PWR primary water

    International Nuclear Information System (INIS)

    Tsutsumi, Kazuya; Yamamoto, Kenji; Nitta, Yoshikazu

    2007-01-01

    Many studies have revealed that fatigue crack growth (FCG) rate of austenitic stainless steels is accelerated in light water reactor environment compared to that in air at room temperature. Major driving factors in the acceleration of FCG rate are stress ratio, temperature and stress rise time. Based on this knowledge, FCG curves have been developed considering these factors as parameters. However, there are few data of FCG threshold ΔK th in light water reactor environment. Hence it is necessary to clarify FCG rate under near-threshold condition for more accurate evaluation of fatigue crack growth behavior under cyclic stress with relatively low ΔK. In the present study, therefore, ΔK th was determined for austenitic stainless steels in simulated PWR primary water, and FCG behavior under near-threshold condition was revealed by collecting fatigue crack propagation data. The results are summarized as follows: No propagation of fatigue crack was found in high temperature water, and there was a definite ΔK th . Average ΔK eff,th was 4.3 MPa·m 0.5 at 325degC, 3.3 MPa·m 0.5 at 100degC, and there was no considerable reduction compared to currently known ΔK eff,th in air. Thus, it was revealed tha ambient conditions had minimal effect, on ΔK eff,th , ΔK th increases with increasing temperature and decreasing frequency. As a result of fracture surface observation, oxide-induced-crack-closure was considered to be a cause of the dependency described above. In addition, it was suggested that changes in material properties also had influence on ΔK th, since ΔK eff,th itself increased at elevated temperature. (author)

  6. Influence of deformation on SCC susceptibility of austenitic stainless steel in PWR primary water

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-09-01

    Slow strain rate tests (SSRT) were carried out to evaluate the SCC susceptibility of four types of austenitic stainless steels (SUS304, SUS316, SUS304L and SUS316L) in PWR primary water. The influence of deformation on SCC susceptibility of SUS316 was studied. All types of stainless steel were susceptible to SCC, and the SCC susceptibility varied depending on the steel type. The comparison of the SSRT results and tensile test in air based on the reduction of area measurement showed that the SCC susceptibility increased with increasing the degree of deformation. For explaining the influence of deformation on SCC susceptibility, it is necessary to evaluate both intergranular and transgranular fractures. (author)

  7. Radiation Protection Aspects of Primary Water Chemistry and Source-term Management Report

    International Nuclear Information System (INIS)

    2014-04-01

    Since the beginning of the 1990's, occupational exposures in nuclear power plant has strongly decreased, outlining efforts achieved by worldwide nuclear operators in order to reach and maintain occupational exposure as low as reasonably achievable (ALARA) in accordance with international recommendations and national regulations. These efforts have focused on both technical and organisational aspects. According to many radiation protection experts, one of the key features to reach this goal is the management of the primary system water chemistry and the ability to avoid dissemination of radioactivity within the system. It outlines the importance for radiation protection staff to work closely with chemistry staff (as well as operation staff) and thus to have sufficient knowledge to understand the links between chemistry and the generation of radiation field. This report was prepared with the primary objective to provide such knowledge to 'non-chemist'. The publication primarily focuses on three topics dealing with water chemistry, source term management and remediation techniques. One key objective of the report is to provide current knowledge regarding these topics and to address clearly related radiation protection issues. In that mind, the report prepared by the EGWC was also reviewed by radiation protection experts. In order to address various designs, PWRs, VVERs, PHWRs and BWRs are addressed within the document. Additionally, available information addressing current operating units and lessons learnt is outlined with choices that have been made for the design of new plants. Chapter 3 of this report addresses current practices regarding primary chemistry management for different designs, 'how to limit activity in the primary circuit and to minimise contamination'. General information is provided regarding activation, corrosion and transport of activated materials in the primary circuit (background on radiation field generation). Primary chemistry aspects that

  8. 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.

  9. Determination of primary yields in the alpha radiolysis of alkaline water

    International Nuclear Information System (INIS)

    Auclair, Guy

    2001-01-01

    This work presents a fundamental study of the radiolysis of water within the framework of the management of nuclear waste. During their storage, the packages of cemented radioactive waste are likely to release molecular hydrogen. Indeed, interstitial water undergoes decomposition under irradiation. This phenomenon is called radiolysis. In order to envisage the impact of H 2 de-gasification on the security of the installations, it is necessary to determine the primary radiolytic yields in the cementing medium (characterised by a pH ranging between 12 and 14), which provides a basic simulations thus allowing us to obtain both the quantities of gas and the pressure in the pore. Such data is currently not available in the literature. Studies were undertaken with a beam of accelerated helium ions in order to reproduce the conditions of irradiation on solutions at pH = 13 in order to determine a first complete series of radiolytic yields.A more complete study was undertaken on the effects of LET and pH on the yield of molecular hydrogen. The results seem to show that the yield of this primary product is little influenced by pH. Such results were in good agreement with those obtained by Monte-Carlo simulations. These studies have shown that, contrary to γ irradiations, the irradiations with α-particles do not lead to the same characteristic times. The extrapolation of this data with respect to the problem of the packaging of nuclear waste is delicate due to the limited amount of results in the literature and also the chemical and physical complexity of the concretes. (author) [fr

  10. Fact and fiction in ECP measurement and control in boiling water reactor primary coolant circuits

    International Nuclear Information System (INIS)

    Macdonald, D.D.

    2005-01-01

    A review is presented of various electrochemical potentials, including the electrochemical corrosion potential (ECP), that are used in the mitigation of stress corrosion cracking in the primary coolant circuits of boiling water reactors (BWRs). Attention is paid to carefully defining each potential in terms of fundamental electrochemical concepts, so as to counter the confusion that has arisen due to the misuse of previously accepted terminology. A brief discussion is also included of reference electrodes and it is shown on the basis of experimental data that the use of a platinum redox sensor as a reference electrode in the monitoring of ECP in BWR primary coolant circuits is inappropriate and should be discouraged. If platinum is used as a reference electrode, because of extenuating circumstances (e.g., potential measurements in high dose regions in a reactor core), the onus must be placed on the user to demonstrate quantitatively that the electrode behaves as an equilibrium electrode under the specified conditions and/or that its potential is invariant with changes in the independent variables of the system. Preferably, a means should also be demonstrated of transferring the measured potential to the standard hydrogen electrode (SHE) scale. (orig.)

  11. Model for cobalt 60/58 deposition on primary coolant piping in a boiling water reactor

    International Nuclear Information System (INIS)

    Dehollander, W.R.

    1979-01-01

    A first principles model for deposition of radioactive metals into the corrosion films of primary coolant piping is proposed. It is shown that the predominant mechanism is the inclusion of the radioactive species such as Cobalt 60 into the spinel structure of the corrosion film during the act of active corrosion. This deposition can occupy only a defined fraction of the available plus 2 valence sites of the spinel. For cobalt ions, this ratio is roughly 4.6 x 10 -3 of the total iron sites. Since no distinction is made between Cobalt 60, Cobalt 58, and Cobalt 59 in this process, the radioactivity associated with this inclusion is a function of the ratio of the radioactive species to the nonradioactive species in the water causing the corrosion of the pipe metal. The other controlling parameter is the corrosion rate of the pipe material. This can be a function of time, for example, and it shown that freshly descaled metal when exposed to the cobalt containing water can incorporate as much as 10 x 10 -3 cobalt ions per iron atom in the initial corrosion period. This has implications for the problem of decontaminating nuclear reactor piping. Equations and selected observations are presented without reference to any specifically identified reactor or utility, so as to protect any proprietary interest

  12. EDF program on SCC initiation of cold-worked stainless steels in primary water

    Energy Technology Data Exchange (ETDEWEB)

    Huguenin, P.; Vaillant, F.; Couvant, T. [Electricite de France (EDF/RD), Site des Renardieres, 77 - Moret sur loing (France); Buisse, L. [EDF UTO, 93 - Noisy-Le-Grand (France); Huguenin, P.; Crepin, J.; Duhamel, C.; Proudhon, H. [MINES ParisTech, Centre des Materiaux, 91 - Evry (France); Ilevbare, G. [EPRI California (United States)

    2009-07-01

    A few cases of Intergranular Stress Corrosion Cracking (IGSCC) on cold-worked austenitic stainless steels in primary water have been detected in French Pressurized Water Reactors (PWRs). A previous program launched in the early 2000's identified the required conditions for SCC of cold-worked stainless steels. It was found that a high strain hardening coupled with cyclic loading favoured SCC, whereas cracking under static conditions appeared to be difficult. A propagation model was also proposed. The first available results of the present study demonstrate the strong influence of a trapezoidal cyclic loading on the creep of 304L austenitic stainless steel. While no creep was detected under a pure static loading, the creep rate was increased by a factor 102 under a trapezoidal cyclic loading. The first results of SCC initiation performed on notched specimens under a trapezoidal cyclic loading at low frequency are presented. The present study aims at developing an engineering model for IGSCC initiation of 304L, 316L and weld 308L stainless steels. The effect of the pre-straining on the SCC mechanisms is more specifically studied. Such a model will be based on (i) SCC initiation tests on notched and smooth specimens under 'trapezoidal' cyclic loading and, (ii) constant strain rate SCC initiation tests. The influence of stress level, cold-work level, strain path, surface roughness and temperature is particularly investigated. (authors)

  13. Influence of Thermal Aging on Primary Water Stress Corrosion Cracking of Cast Duplex Stainless Steels

    International Nuclear Information System (INIS)

    Yamada, T.; Totsuka, N.; Nakajima, N.; Arioka, K.; Negishi, K.

    2002-01-01

    In order to evaluate the SCC (stress corrosion cracking) susceptibility of cast duplex stainless steels which are used for the main coolant piping material of pressurized water reactors (PWRs), the slow strain rate test (SSRT) and the constant load test (CLT) were performed in simulated PWR primary water at 360 C. The main coolant piping materials contain ferrite phase with ranging from 8 to 23 % and its mechanical properties are affected by long time thermal aging. The 23% ferrite material was prepared for test as the maximum ferrite content of main coolant pipes in Japanese PWRs. The brittle fracture in the non-aged materials after SSRT is mainly caused by quasi-cleavage fracture in austenitic phase. On the other hand, a mixture of quasi-cleavage fracture in austenite and ferrite phases was observed on long time aged material. Also on CLT, (2 times σ y ), after 3,000 hours exposure, microcracks were observed on the surface of non-aged and aged for 10,000 hours at 400 C materials. The crack initiation site of CLT is similar to that of SSRT. The SCC susceptibility of the materials increases with aging time. It is suggested that the ferrite hardening with aging affect SCC susceptibility of cast duplex stainless steels. (authors)

  14. 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.

  15. 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.

  16. Factors leading to poor water sanitation hygiene among primary school going children in Chitungwiza

    Directory of Open Access Journals (Sweden)

    Blessing Dube

    2012-03-01

    Full Text Available Although the world has progressed in the area of water and sanitation, more than 2.3 billion people still live without access to sanitation facilities and some are unable to practice basic hygiene. Access to water and basic sanitation has deteriorated in Chitungwiza and children are at risk of developing illness and missing school due to the deterioration. We sought to investigate the predisposing, enabling and reinforcing factors that are causally related to water- and sanitation- related hygiene practices among school going children. A random sample of 400 primary school children (196 males, 204 females in four schools in Chitungwiza town, Zimbabwe was interviewed. Behavioural factors were assessed through cross examination of the PROCEED PRECEDE Model. The respondents had been stratified through the random sampling where strata were classes. A structured observation checklist was also administered to assess hygiene enabling facilities for each school. Children’s knowledge and perceptions were inconsistent with hygienic behaviour. The family institution seemed to play a more important role in life skills training and positive reinforcement compared to the school (50% vs 27.3%. There was no association between a child’s sex, age and parents’ occupation with any of the factors assessed (P=0.646. Schools did not provide a hygiene enabling environment as there were no learning materials, policy and resources on hygiene and health. The challenges lay in the provision of hygiene enabling facilities, particularly, the lack of access to sanitation for the maturing girl child and a school curriculum that provides positive reinforcement and practical life skills training approach.

  17. Modelling of Transport of Radioactive Substances in the Primary Circuit of Water Cooled Reactors

    International Nuclear Information System (INIS)

    2012-03-01

    coordinated research project (CRP) was proposed to determine the accuracy of existing computer codes and to identify how they could be improved through application of this body of work. Specifically, the CRP was expected to: - Build a database for selected pressurized water reactor (PWR) plants that would contain the design information suitable for their description within a computer code, as well as give the operating history of the plant, which would include the water chemistry data over several refuelling cycles; - Show the contamination of selected out-of-core surfaces such as circulating loops and steam generator channel heads versus operating history and compare the prediction of surface contamination versus time from modern radioactivity transport codes with actual plant data in a blind benchmarking exercise; - Determine how current codes, as well as new ones, could be improved and encourage the development of accurate new codes in Member States using the recommendations from the present work. This report uses as its basis the results of this CRP on 'Modelling of Transport of Radioactive Substances in the Primary Circuit of Water Cooled Reactors', which was conducted over the period 1996-2001 for PWR type reactors. The report also describes the significant progress demonstrated in this field in the period that followed.

  18. Characterization of interfacial reactions and oxide films on 316L stainless steel in various simulated PWR primary water environments

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Junjie; Xiao, Qian [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); State Key Laboratory of Advanced Special Steels, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Lu, Zhanpeng, E-mail: zplu@t.shu.edu.cn [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); State Key Laboratory of Advanced Special Steels, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Shanghai Key Laboratory of Advanced Ferrometallurgy, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Ru, Xiangkun; Peng, Hao; Xiong, Qi; Li, Hongjuan [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China)

    2017-06-15

    The effect of water chemistry on the electrochemical and oxidizing behaviors of 316L SS was investigated in hydrogenated, deaerated and oxygenated PWR primary water at 310 °C. Water chemistry significantly influenced the electrochemical impedance spectroscopy parameters. The highest charge-transfer resistance and oxide-film resistance occurred in oxygenated water. The highest electric double-layer capacitance and constant phase element of the oxide film were in hydrogenated water. The oxide films formed in deaerated and hydrogenated environments were similar in composition but different in morphology. An oxide film with spinel outer particles and a compact and Cr-rich inner layer was formed in both hydrogenated and deaerated water. Larger and more loosely distributed outer oxide particles were formed in deaerated water. In oxygenated water, an oxide film with hematite outer particles and a porous and Ni-rich inner layer was formed. The reaction kinetics parameters obtained by electrochemical impedance spectroscopy measurements and oxidation film properties relating to the steady or quasi-steady state conditions in the time-period of measurements could provide fundamental information for understanding stress corrosion cracking processes and controlling parameters. - Highlights: •Long-term EIS measurements of 316L SS in simulated PWR primary water. •Highest charge-transfer resistance and oxide film resistance in oxygenated water. •Highest electric double-layer capacitance and oxide film CPE in hydrogenated water. •Similar compositions, different shapes of oxides in deaerated/hydrogenated water. •Inner layer Cr-rich in hydrogenated/deaerated water, Ni-rich in oxygenated water.

  19. Seasonal Shifts in Primary Water Source Type: A Comparison of Largely Pastoral Communities in Uganda and Tanzania

    Directory of Open Access Journals (Sweden)

    Amber L. Pearson

    2016-01-01

    Full Text Available Many water-related illnesses show an increase during the wet season. This is often due to fecal contamination from runoff, yet, it is unknown whether seasonal changes in water availability may also play a role in increased illness via changes in the type of primary water source used by households. Very little is known about the dynamic aspects of access to water and changes in source type across seasons, particularly in semi-arid regions with annual water scarcity. The research questions in this study were: (1 To what degree do households in Uganda (UG and Tanzania (TZ change primary water source type between wet and dry seasons?; and (2 How might seasonal changes relate to water quality and health? Using spatial survey data from 92 households each in UG and TZ this study found that, from wet to dry season, 26% (UG and 9% (TZ of households switched from a source with higher risk of contamination to a source with lower risk. By comparison, only 20% (UG and 0% (TZ of households switched from a source with lower risk of contamination to a source with higher risk of contamination. This research suggests that one pathway through which water-related disease prevalence may differ across seasons is the use of water sources with higher risk contamination, and that households with access to sources with lower risks of contamination sometimes choose to use more contaminated sources.

  20. Development of In-Service Inspection system for heat transfer tubes in the primary pressurized water cooler in the HTTR

    Energy Technology Data Exchange (ETDEWEB)

    Shinozaki, Masayuki; Furusawa, Takayuki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Wada, Shigeyuki

    1999-08-01

    The ISI (In-Service Inspection) system has been developed so as to maintain the structural integrity of heat transfer tubes in the primary pressurized water cooler in the HTTR (High Temperature Engineering Test Reactor). This system consists of eddy current probes, ultra-sonic probes, insertion and extraction units, positioning unit and so on. Verification and performance tests of the developed ISI system were carried out using mock-up heat transfer tubes in the primary pressurized water cooler. The constitution of the system, R and D results of the inspection probes, and verification and performance test results of the ISI system for heat transfer tubes are described in this paper. (author)

  1. Effect of supplementation of tender coconut water on blood pressure of primary hypertensive subjects

    Directory of Open Access Journals (Sweden)

    Gullapalli HS, Avinash P Tekade, Namrata H Gullapalli

    2013-04-01

    Full Text Available Background: Hypertension is a major health problem worldwide. Increased vascular resistance, sodium retention & sympathetic over activity contributes to the blood pressure elevation. Plant foods may be beneficial in decreasing blood pressure (BP. Recently much attention has been focused on plant foods that may be beneficial in preventing Hypertension, metabolic syndrome and possibly reduce the risk of various diseases. This clinical study was conducted to test the effectiveness of a structured intervention on BP of primary hypertensive subjects. Aim: To study the effect of Tender Coconut Water (TCW on BP of Primary hypertensive subjects. Methods and Material: 70 subjects were selected randomly sample for 6 weeks of the intervention program. Among them 40 subjects were selected as the experimental group and 30 300ml/day for 6 weeks whereas the control group was instructed to follow the same routine without modifications. One initial, two mid intervention (after every 15 days and one final (post intervention BP recorded for both the groups. The obtained data was statistically analyzed. Results: The mean systolic BP of experimental group and control group were decreased from 145.8 mm Hg and 141mm of Hg to 135.3 mm of Hg and 140 mm of Hg respectively. The mean diastolic BP of experimental group and control group were decreased from 93.7 mm H g and 90.9 mmHg to 86.9 mm of Hg and 89.7 mm of Hg respectively. Conclusion: Irrespective of cause of hypertension TCW has beneficial effect on BP. TCW contains high amount of potassium which causes vasodilatation and also improve the endothelial function.

  2. Compliance Determination for Inactivation Requirements of the National Primary Drinking Water Regulations when a Public Water Systems Uses Dichlor and Trichlor for Primary Disinfection

    Science.gov (United States)

    This memorandum has been developed to assist SDWA primacy agencies (EPA Regions, states and territories) when considering inactivation/disinfection compliance requirements for those water systems that choose to use Dichlor or Trichlor.

  3. Initiation of stress corrosion cracking in pre-stained austenitic stainless steels exposed to primary water

    International Nuclear Information System (INIS)

    Huguenin, P.

    2012-01-01

    Austenitic stainless steels are widely used in primary circuits of Pressurized Water Reactors (PWR) plants. However, a limited number of cases of Intergranular Stress Corrosion Cracking (IGSCC) has been detected in cold-worked (CW) areas of non-sensitized austenitic stainless steel components in French PWRs. A previous program launched in the early 2000's identified the required conditions for SCC of cold-worked stainless steels. It was found that a high strain hardening coupled with a cyclic loading favoured SCC. The present study aims at better understanding the role of pre-straining on crack initiation and at developing an engineering model for IGSCC initiation of 304L and 316L stainless steels in primary water. Such model will be based on SCC initiation tests on notched (not pre-cracked) specimens under 'trapezoidal' cyclic loading. The effects of pre-straining (tensile versus cold rolling), cold-work level and strain path on the SCC mechanisms are investigated. Experimental results demonstrate the dominating effect of strain path on SCC susceptibility for all pre-straining levels. Initiation can be understood as crack density and crack depth. A global criterion has been proposed to integrate both aspects of initiation. Maps of SCC initiation susceptibility have been proposed. A critical crack depth between 10 and 20 μm has been demonstrated to define transition between slow propagation and fast propagation for rolled materials. For tensile pre-straining, the critical crack depth is in the range 20 - 50 μm. Experimental evidences support the notion of a KISCC threshold, whose value depends on materials, pre-straining ant load applied. The initiation time has been found to depend on the applied loading as a function of (σ max max/YV) 11,5 . The effect of both strain path and surface hardening is indirectly taken into account via the yield stress. In this study, material differences rely on strain path effect on mechanical properties. As a result, a stress

  4. The growing rate and the type of corrosion products of aluminium alloy AA 5052 in deionized water at temperature up to 3000C

    International Nuclear Information System (INIS)

    Ferreira, E.G.

    1980-01-01

    The process of corrosion concerning the aluminum alloy AA5052 in deionized water at temperatures of 40 0 C, 80 0 C, 90 0 C, 140 0 C, 200 0 C and 280 0 C is studied. The following methods are used: periodic weighting of the test samples; analysis by neutronic activation of the corrosion products dissolved in water; thermogravimetric and thermodiferential analysis; analysis through X-ray diffraction and from metalografic observations of the crystals produced in the corrosion process; an optical microscope using polarized and normal light and a scanning electronic microscope. The activation energies are calculated for the corrosion film formation, and for the dissolution of the corrosion products in the deionized water. (ARHC) [pt

  5. Experimental investigations concerning the possible effect of dynamic strain ageing on environmentally-assisted cracking of low alloy steels in oxygenated high-temperature water

    International Nuclear Information System (INIS)

    Roth, A.; Devrient, B.; Haenninen, H.; Bruemmer, G.; Ilg, U.; Widera, M.; Hofmann, H.; Wachter, O.

    2003-01-01

    Service experience has revealed cracks due to environmentally-assisted cracking (EAC) in welds of the feedwater piping system of a boiling water reactor (BWR). Two slightly different low alloy steel (LAS) weld filler metals were used in the system of concern, however, only one of them was affected by cracking. To achieve an improved understanding, a laboratory study was initiated to investigate the crack growth behavior of the two relevant weld filler metals in an oxygenated high-temperature water (HTW) environment representing BWR normal water chemistry (NWC) under sequences of cyclic and constant load. Despite the basic similarities in the nominal chemical composition of both weld filler alloys, the crack growth behaviors revealed significant differences. This could not be explained based on the material's sulphur content, which is known to have a pronounced effect on EAC. To elucidate the observed behavior, studies concerning dynamic strain aging (DSA) have been initiated. DSA has been recently suspected to be another parameter that may influence EAC of LAS in HTW. A reasonable coincidence was observed between the susceptibility to DSA exhibited by slow strain rate tensile tests (SSRT) in air and by internal friction measurements with measured free nitrogen contents on the one hand and with the EAC behavior observed in service and in laboratory experiments on the other hand. (orig.)

  6. 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%

  7. Evaluation of short-term effects of rare earth and other elements used in magnesium alloys on primary cells and cell lines.

    Science.gov (United States)

    Feyerabend, Frank; Fischer, Janine; Holtz, Jakob; Witte, Frank; Willumeit, Regine; Drücker, Heiko; Vogt, Carla; Hort, Norbert

    2010-05-01

    Degradable magnesium alloys for biomedical application are on the verge of being used clinically. Rare earth elements (REEs) are used to improve the mechanical properties of the alloys, but in more or less undefined mixtures. For some elements of this group, data on toxicity and influence on cells are sparse. Therefore in this study the in vitro cytotoxicity of the elements yttrium (Y), neodymium (Nd), dysprosium (Dy), praseodymium (Pr), gadolinium (Gd), lanthanum (La), cerium (Ce), europium (Eu), lithium (Li) and zirconium (Zr) was evaluated by incubation with the chlorides (10-2000 microM); magnesium (Mg) and calcium (Ca) were tested at higher concentrations (200 and 50mM, respectively). The influence on viability of human osteosarcoma cell line MG63, human umbilical cord perivascular (HUCPV) cells and mouse macrophages (RAW 264.7) was determined, as well as the induction of apoptosis and the expression of inflammatory factors (TNF-alpha, IL-1alpha). Significant differences between the applied cells could be observed. RAW exhibited the highest and HUCPV the lowest sensitivity. La and Ce showed the highest cytotoxicity of the analysed elements. Of the elements with high solubility in magnesium alloys, Gd and Dy seem to be more suitable than Y. The focus of magnesium alloy development for biomedical applications should include most defined alloy compositions with well-known tissue-specific and systemic effects. Copyright (c) 2009 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

  8. Effects of sewage water on bio-optical properties and primary production of coastal systems in West Australia

    DEFF Research Database (Denmark)

    Stæhr, Peter Anton; Waite, A. M.; Markager, S.

    2008-01-01

    . Both systems were significantly nitrogen limited. However, differences in wastewater treatment (primary vs secondary) and sewage dilution (50%) between the two systems caused a greater difference between systems than locally around the outflows. For both systems, water at the outlet had significantly...... lower water transparency caused by a 20% higher absorption by coloured dissolved organic matter. Nutrient concentrations were also elevated, gradually decreasing with distance north (governing current) of the outflows, causing higher abundance of nano-sized phytoplankton, higher content of carotenoid...

  9. Effect of dynamic strain ageing on the environmentally assisted cracking of low-alloy steels oxygenated high-temperature water

    International Nuclear Information System (INIS)

    Devrient, B.; Roth, A.; Kuester, K.; Ilg, U.; Widera, M.

    2007-01-01

    The plastic deformation behavior of low-alloy steels (LAS) is significantly influenced by their individual susceptibility to dynamic strain ageing (DSA). Interstitial atoms of nitrogen (N) or carbon (C) in the steel matrix can change the mechanical properties like ductility and strength by interaction with moving dislocations during plastic deformation. The degree of DSA is depending on temperature and strain rate during plastic deformation. Under critical parameter combinations strength increases while ductility decreases. Furthermore, the interaction of dislocations and interstitial atoms can lead to a localization of plastic deformation, which results in planar gliding processes. Shear bands in LAS types with a high susceptibility to DSA show significantly higher slip steps during plastic deformation as compared to heats with low susceptibility to DSA. Since the basic mechanism of environmentally-assisted cracking (EAC) of LAS in high-temperature water (HTW) environment is slip-step-dissolution, slip behavior is of crucial nature for the kinetics of crack initiation and crack growth. Therefore, a program concerning deformation behavior, slip characterization regarding distribution and size, and behavior in oxygenated HTW environment was performed. Analysis of slip steps by advanced techniques for surface morphology investigation showed that the maximum height of slip steps is in the range of freshly formed magnetite layers on LAS in oxygenated HTW environment. This supports the active effect of localized deformation on EAC in LAS types of high susceptibility to DSA. The exposure to oxygenated HTW environment with additional mechanical loading under critical combinations of temperature and strain rate of different LAS types with high, intermediate and low susceptibility to DSA in Slow Strain Rate Tensile-tests (SSRT) showed preferential crack initiation in the areas of coarse shear bands due to localized deformation. Furthermore, a continuous transition of the

  10. SCC growth behaviors of austenitic stainless steels in simulated PWR primary water

    Science.gov (United States)

    Terachi, T.; Yamada, T.; Miyamoto, T.; Arioka, K.

    2012-07-01

    The rates of SCC growth were measured under simulated PWR primary water conditions (500 ppm B + 2 ppm Li + 30 cm3/kg-H2O-STP DH2) using cold worked 316SS and 304SS. The direct current potential drop method was applied to measure the crack growth rates for 53 specimens. Dependence of the major engineering factors, such as yield strength, temperature and stress intensity was systematically examined. The rates of crack growth were proportional to the 2.9 power of yield strength, and directly proportional to the apparent yield strength. The estimated apparent activation energy was 84 kJ/mol. No significant differences in the SCC growth rates and behaviors were identified between 316SS and 304SS. Based on the measured results, an empirical equation for crack growth rate was proposed for engineering applications. Although there were deviations, 92.8% of the measured crack growth rates did not exceed twice the value calculated by the empirical equation.

  11. Assessment of Field Experience Related to Pressurized Water Reactor Primary System Leaks

    International Nuclear Information System (INIS)

    Ware, A.G.; Hsu, C.; Atwood, C.L.; Sattison, M.B.; Hartley, R.S.; Shah, V.N.

    1999-01-01

    This paper presents our assessment of field experience related to pressurized water reactor (PWR) primary system leaks in terms of their number and rates, how aging affects frequency of leak events, the safety significance of such leaks, industry efforts to reduce leaks, and effectiveness of current leak detection systems. We have reviewed the licensee event reports to identify the events that took place during 1985 to the third quarter of 1996, and reviewed related technical literature and visited PWR plants to analyze these events. Our assessment shows that USNRC licensees have taken effective actions to reduce the number of leak events. One main reason for this decreasing trend was the elimination or reportable leakages from valve stem packing after 1991. Our review of leak events related to vibratory fatigue reveals a statistically significant decreasing trend with age (years of operation), but not in calendar time. Our assessment of worldwide data on leakage caused by thermal fatigue cracking is that the fatigue of aging piping is a safety significant issue. Our review of leak events has identified several susceptible sites in piping having high safety significance; but the inspection of some of these sites is not required by the ASME Code. These sites may be included in the risk-informed inspection programs

  12. Assessment of Field Experience Related to Pressurized Water Reactor Primary System Leaks

    International Nuclear Information System (INIS)

    Shah, Vikram Naginbhai; Ware, Arthur Gates; Atwood, Corwin Lee; Sattison, Martin Blaine; Hartley, Robert Scott; Hsu, C.

    1999-01-01

    This paper presents our assessment of field experience related to pressurized water reactor (PWR) primary system leaks in terms of their number of rates, how aging affects frequency of leak events, the safety significance of such leaks, industry efforts to reduce leaks, and effectiveness of current leak detection systems. We have reviewed the licensee event reports to identify the events that took place during 1985 to the third quarter of 1996, and reviewed related technical literature and visited PWR plants to analyze these events. Our assessment shows that USNRC licensees have taken effective actions to reduce the number of leak events. One main reason for this decreasing trend was the elimination or reportable leakages from valve stem packing after 1991. Our review of leak events related to vibratory fatigue reveals a statistically significant decreasing trend with age (years of operation), but not in calendar time. Our assessment of worldwide data on leakage caused by thermal fatigue cracking is that the fatigue of aging piping is a safety significant issue. Our review of leak events has identified several susceptible sites in piping having high safety significance; but the inspection of some of these sites is not required by the ASME Code. These sites may be included in the risk-informed inspection programs

  13. The Life-Cycle Costs of School Water, Sanitation and Hygiene Access in Kenyan Primary Schools.

    Science.gov (United States)

    Alexander, Kelly T; Mwaki, Alex; Adhiambo, Dorothy; Cheney-Coker, Malaika; Muga, Richard; Freeman, Matthew C

    2016-06-27

    Water, Sanitation and Hygiene (WASH) programs in schools can increase the health, dignity and comfort of students and teachers. Understanding the costs of WASH facilities and services in schools is one essential piece for policy makers to utilize when budgeting for schools and helping to make WASH programs more sustainable. In this study we collected data from NGO and government offices, local hardware shops and 89 rural primary schools across three Kenyan counties. Current expenditures on WASH, from school and external (NGO, government, parent) sources, averaged 1.83 USD per student per year. After reviewing current expenditures, estimated costs of operations and maintenance for bringing schools up to basic WASH standards, were calculated to be 3.03 USD per student per year. This includes recurrent costs, but not the cost of installing or setting up WASH infrastructure, which was 18,916 USD per school, for a school of 400 students (4.92 USD per student, per year). These findings demonstrate the need for increases in allocations to schools in Kenya, and stricter guidance on how money should be spent on WASH inputs to enable all schools to provide basic WASH for all students.

  14. The influence of ppb levels of chloride impurities on the stress corrosion crack growth behaviour of low-alloy steels under simulated boiling water reactor conditions

    International Nuclear Information System (INIS)

    Seifert, H.P.; Ritter, S.

    2016-01-01

    Highlights: • Chloride effects on SCC crack growth in RPV steels under boiling water reactor conditions. • ppb-levels of chloride may result in fast SCC in normal water chemistry environment. • Much higher chloride tolerance for SCC in hydrogen water chemistry environment. • Potential long-term (memory) effects after severe and prolonged temporary chloride transients. - Abstract: The effect of chloride on the stress corrosion crack (SCC) growth behaviour in low-alloy reactor pressure vessel steels was evaluated under simulated boiling water reactor conditions. In normal water chemistry environment, ppb-levels of chloride may result in fast SCC after rather short incubation periods of few hours. After moderate and short-term chloride transients, the SCC crack growth rates return to the same very low high-purity water values within few 100 h. Potential long-term (memory) effects on SCC crack growth cannot be excluded after severe and prolonged chloride transients. The chloride tolerance for SCC in hydrogen water chemistry environment is much higher.

  15. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part I: BWR/NWC conditions

    International Nuclear Information System (INIS)

    Ritter, S.; Seifert, H.P.; Devrient, B.; Roth, A.; Ehrnsten, U.; Ernestova, M.; Zamboch, M.; Foehl, J.; Weissenberg, T.; Gomez-Briceno, D.; Lapena, J.

    2004-01-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWR) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It was focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of EAC crack growth behaviour/mechanism of LAS in high-temperature water under steady-state power operation (constant load) and transient operating conditions (e.g., start-up/shut-down, transients in water chemistry and load). Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurised water reactor (VVER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (VVER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarises the most important crack growth results obtained under simulated BWR/NWC conditions. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  16. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part I: BWR/NWC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H.P. [Paul Scherrer Institute, PSI, Villigen (Switzerland); Devrient, B.; Roth, A. [Framatome ANP GmbH, Erlangen (Germany); Ehrnsten, U. [VTT Industrial Systems, Espoo (Finland); Ernestova, M.; Zamboch, M. [Nuclear Research Institute, NRI, Rez (Czech Republic); Foehl, J.; Weissenberg, T. [Staatliche Materialpruefungsanstalt, MPA, Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, CIEMAT, Madrid (Spain)

    2004-07-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWR) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It was focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of EAC crack growth behaviour/mechanism of LAS in high-temperature water under steady-state power operation (constant load) and transient operating conditions (e.g., start-up/shut-down, transients in water chemistry and load). Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurised water reactor (VVER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (VVER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarises the most important crack growth results obtained under simulated BWR/NWC conditions. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  17. Development of new zirconium based alloys for burn-up extension of light water reactor fuels, (1)

    International Nuclear Information System (INIS)

    Isobe, Takeshi; Matsuo, Yutaka

    1992-01-01

    Steam corrosion tests and tensile were conducted to investigate the effects of alloying elements such as Sn, Nb, Fe, Cr, Mo and V, and the mechanical properties of Nb-containing Zr-base alloys. The corrosion resistance of Zr-base alloys in comparison to Zr'y-4 was significantly improved by the reduction of the Sn content by 0.5 wt% and by a small addition of Nb (about 0.05 to 0.2 wt%). However, the decrease in solute Sn atoms degraded mechanical properties. The increase of the total content of Fe and Cr from 0.3 to 0.7 wt% improved the mechanical properties without affecting the corrosion resistance. The decrease of the Fe/Cr ratio from 6.0 to 0.5 increased the corrosion resistance. Small addition of Mo and/or V resulted in a further improvement of mechanical properties. Based on these experiments, three Nb-containing Zr-base alloys with equivalent mechanical properties and superior corrosion resistance to Zr'y-4 were developed. (author)

  18. AN INVESTIGATION OF THE IMPACT OF ALLOY COMPOSITION AND PH ON THE CORROSION OF BRASS IN DRINKING WATER

    Science.gov (United States)

    A better understanding of brass corrosion may provide information and guidance on the use of the safest materials for the production of plumbing fixtures, and optimization of corrosion control treatments. The effect of alloy composition and pH on the metal leached from six differ...

  19. Extraction Tools for Identification of Chemical Contaminants in Estuarine and Coastal Waters to Determine Toxic Pressure on Primary Producers

    NARCIS (Netherlands)

    Booij, P; Sjollema, S.B.; Leonards, P.E.G.; de Voogt, P.; Stroomberg, G.J.; Vethaak, A.D.; Lamoree, M.H.

    2013-01-01

    The extent to which chemical stressors affect primary producers in estuarine and coastal waters is largely unknown. However, given the large number of legacy pollutants and chemicals of emerging concern present in the environment, this is an important and relevant issue that requires further study.

  20. Absorption-based algorithm of primary production for total and size-fractionated phytoplankton in coastal waters

    NARCIS (Netherlands)

    Barnes, M.K.; Tilstone, G.H.; Smyth, T.J.; Suggett, D.J.; Astoreca, R.; Lancelot, C.; Kromkamp, J.C.

    2014-01-01

    Most satellite models of production have been designed and calibrated for use in the open ocean. Coastal waters are optically more complex, and the use of chlorophyll a (chl a) as a first-order predictor of primary production may lead to substantial errors due to significant quantities

  1. Effects of omnivorous tilapia on water turbidity and primary production dynamics in shallow lakes: implications for ecosystem management

    NARCIS (Netherlands)

    Zhang, Xiufeng; Mei, Xueying; Gulati, Ramesh D.

    2017-01-01

    The introduction of omnivorous tilapia into a variety of aquatic systems worldwide has led to a number of serious ecological problems. One of the main issues is an increase in water turbidity, which affects not only light penetration but also primary production and the distribution of phytoplankton

  2. 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.

  3. Differential Scanning Calorimetric Study and Potential Model of the Binding of the Primary Water of Hydration to K-Hyaluronate

    Science.gov (United States)

    Whitson, K. B.; Marlowe, R. L.; Lukan, A. M.; Lee, S. A.; Anthony, L.; Rupprecht, A.

    1997-11-01

    DSC was performed on samples of K-hyaluronate (KHA) through a temperature range of 25-180^oC. A transition peak was observed which is due to the desorption of the primary water of hydration. The maximum position of the peak was observed to change with different scan rates. The average energy of activation, E_A, and enthalpy for desorption of the primary water of hydration was determined to be 0.62 and 0.17 eV per water molecule, respectively. Analysis of Mossbauer data(G. Albanese et al., Hyperfine Int.,) 95, 97 (1995) allowed us to determine the effective force constant, k_eff, of the water bound to KHA to be approximately 19.4 eV/nm^2. The parameters E_A, ΔH,and k_eff allow us to construct a potential model for the primary water of hydration of KHA. Comparison of these parameters with the same parameters for HA and DNA with different counterions reveal that the energy of activation is similar, as well as the enthalpy change.

  4. 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)

  5. Part of the hydrogen in the intergranular crack by stress corrosion in primary circuit for the 600 and 690 nickel base alloys; Role de l'hydrogene dans le mecanisme de fissuration intergranulaire par corrosion sous contrainte en milieu primaire des alliages base nickel 600 et 690

    Energy Technology Data Exchange (ETDEWEB)

    Odemer, G.; Coudurier, A.; Jambon, F.; Chene, J. [CEA Saclay, Dept. de Physico-Chimie (DPC/SCCME/LECA), 91 - Gif sur Yvette (France); Odemer, G.; Coudurier, A.; Chene, J. [Evry Univ., UMR 8587 CNRS / CEA, LAMBE, 91 (France)

    2007-07-01

    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.)

  6. Primary processes in radiation chemistry. LET (Linear Energy Transfer) effect in water radiolysis

    International Nuclear Information System (INIS)

    Trupin-Wasselin, V.

    2000-01-01

    The effect of ionizing radiations on aqueous solutions leads to water ionization and then to the formation of radical species and molecular products (e - aq , H . , OH . , H 2 O 2 , H 2 ). It has been shown that the stopping power, characterized by the LET value (Linear Energy Transfer) becomes different when the nature of the ionizing radiations is different. Few data are nowadays available for high LET radiations such as protons and high energy heavy ions. These particles have been used to better understand the primary processes in radiation chemistry. The yield of a chemical dosimeter (the Fricke dosimeter) and those of the hydrogen peroxide have been determined for different LET. The effect of the dose rate on the Fricke dosimeter yield and on the H 2 O 2 yield has been studied too. When the dose rate increases, an increase of the molecular products yield is observed. At very high dose rate, this yield decreases on account of the attack of the molecular products by radicals. The H 2 O 2 yield in alkaline medium decreases when the pH reaches 12. This decrease can be explained by a slowing down of the H 2 O 2 formation velocity in alkaline medium. Superoxide radical has also been studied in this work. A new detection method: the time-resolved chemiluminescence has been perfected for this radical. This technique is more sensitive than the absorption spectroscopy. Experiments with heavy ions have allowed to determine the O 2 .- yield directly in the irradiation cell. The experimental results have been compared with those obtained with a Monte Carlo simulation code. (O.M.)

  7. Phytoplankton absorption predicts patterns in primary productivity in Australian coastal shelf waters

    Science.gov (United States)

    Robinson, C. M.; Cherukuru, N.; Hardman-Mountford, N. J.; Everett, J. D.; McLaughlin, M. J.; Davies, K. P.; Van Dongen-Vogels, V.; Ralph, P. J.; Doblin, M. A.

    2017-06-01

    The phytoplankton absorption coefficient (aPHY) has been suggested as a suitable alternate first order predictor of net primary productivity (NPP). We compiled a dataset of surface bio-optical properties and phytoplankton NPP measurements in coastal waters around Australia to examine the utility of an in-situ absorption model to estimate NPP. The magnitude of surface NPP (0.20-19.3 mmol C m-3 d-1) across sites was largely driven by phytoplankton biomass, with higher rates being attributed to the microplankton (>20 μm) size class. The phytoplankton absorption coefficient aPHY for PAR (photosynthetically active radiation; āPHY)) ranged from 0.003 to 0.073 m-1, influenced by changes in phytoplankton community composition, physiology and environmental conditions. The aPHY coefficient also reflected changes in NPP and the absorption model-derived NPP could explain 73% of the variability in measured surface NPP (n = 41; RMSE = 2.49). The absorption model was applied to two contrasting coastal locations to examine NPP dynamics: a high chlorophyll-high variation (HCHV; Port Hacking National Reference Station) and moderate chlorophyll-low variation (MCLV; Yongala National Reference Station) location in eastern Australia using the GIOP-DC satellite aPHY product. Mean daily NPP rates between 2003 and 2015 were higher at the HCHV site (1.71 ± 0.03 mmol C m-3 d-1) with the annual maximum NPP occurring during the austral winter. In contrast, the MCLV site annual NPP peak occurred during the austral wet season and had lower mean daily NPP (1.43 ± 0.03 mmol C m-3 d-1) across the time-series. An absorption-based model to estimate NPP is a promising approach for exploring the spatio-temporal dynamics in phytoplankton NPP around the Australian continental shelf.

  8. Magnetite solubility studies under simulated PWR primary-side conditions, using lithiated, hydrogenated water

    International Nuclear Information System (INIS)

    Hewett, John; Morrison, Jonathan; Cooper, Christopher; Ponton, Clive; Connolly, Brian; Dickinson, Shirley; Henshaw, Jim

    2014-01-01

    As software for modelling dissolution, precipitation, and transport of metallic species and subsequent CRUD deposition within nuclear plant becomes more advanced, there is an increasing need for accurate and reliable thermodynamic data. The solubility behaviour of magnetite is an example of such data, and is central to any treatment of CRUD solubility due to the prevalence of magnetite and nickel ferrites in CRUD. Several workers have shown the most consistent solubility data comes from once-through flowing systems. However, despite a strong consensus between the results in acidic to mildly alkaline solutions, there is disagreement between the results at approximately pH 25C 9 and higher. A programme of experimental work is on-going at the University of Birmingham, focusing on solubility of metal oxides (e.g., magnetite) in conditions relevant to PWR primary coolant. One objective of this programme is to calculate thermodynamic constants from the data obtained. Magnetite solubility from 200 to 300°C, in lithiated, hydrogenated water of pH 25C 9–11 is being studied using a once-through rig constructed of 316L stainless steel. The feedwater is pumped at 100 bar pressure through a heated bed of magnetite granules, and the output solution is collected and analysed for iron and several other metals by ICP-MS. This paper presents results from preliminary tests without magnetite granules, in which the corroding surface of the rig itself was used as the sole source of soluble iron and of dissolved hydrogen. Levels of iron were generally within an order of magnitude of literature solubility values. Comparison of results at different flow rates and temperatures, in conjunction with conclusions drawn from the published literature, suggests that this is likely due to the presence of particulate matter in a greatly under-saturated solution, compensating for the low surface area of oxide in contact with the solution. (author)

  9. Assessment of water consumptions in small mediterranean islands' primary schools by means of a long-term online monitoring

    Science.gov (United States)

    Ferraris, Marco; De Gisi, Sabino; Farina, Roberto

    2017-10-01

    A key challenge of our society is improving schools through the sustainable use of resources especially in countries at risk of desertification. The estimation of water consumption is the starting point for the correct dimensioning of water recovery systems. To date, unlike the energy sector, there is a lack of scientific information regarding water consumption in school buildings. Available data refer roughly to indirect estimates by means of utility bills and therefore no information on the role of water leakage in the internal network of the school is provided. In this context, the aim of the work was to define and implement an on-line monitoring system for the assessment of water consumptions in a small Mediterranean island primary school to achieve the following sub-goals: (1) definition of water consumption profile considering teaching activities and secretarial work; (2) direct assessment of water consumptions and leakages and, (3) quantification of the behaviour parameters. The installed monitoring system consisted of 33 water metres (3.24 persons per water metre) equipped with sensors set on 1-L impulse signal and connected to a data logging system. Results showed consumptions in the range 13.6-14.2 L/student/day and leakage equal to 54.8 % of the total water consumptions. Considering the behavioural parameters, the consumptions related to toilet flushing, personal, and building cleaning were, respectively, 54, 43 and 3 % of the total water ones. Finally, the obtained results could be used for dimensioning the most suitable water recovery strategies at school level such as grey water or rainwater recovery systems.

  10. 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}.

  11. Analysis of water hammer-structure interaction in piping system for a loss of coolant accident in primary loop of pressurized water reactor

    International Nuclear Information System (INIS)

    Zhang Xiwen; Yang Jinglong; He Feng; Wang Xuefang

    2000-01-01

    The conventional analysis of water hammer and dynamics response of structure in piping system is divided into two parts, and the interaction between them is neglected. The mechanism of fluid-structure interaction under the double-end break pipe in piping system is analyzed. Using the characteristics method, the numerical simulation of water hammer-structure interaction in piping system is completed based on 14 parameters and 14 partial differential equations of fluid-piping cell. The calculated results for a loss of coolant accident (LOCA) in primary loop of pressurized water reactor show that the waveform and values of pressure and force with time in piping system are different from that of non-interaction between water hammer and structure in piping system, and the former is less than the later

  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. An exploration of factors that influence the regular consumption of water by Irish primary school children.

    LENUS (Irish Health Repository)

    Molloy, C Johnston

    2008-10-01

    Inadequate hydration has been linked to many factors that may impact on children\\'s education and health. Teachers play an important role in the education and behaviour of children. Previous research has demonstrated low water intake amongst children and negative teachers\\' attitudes to water in the classroom. The present study aimed to explore teachers\\' knowledge about water and the perceived barriers to allowing children access to water during lesson time.

  14. VANADIUM ALLOYS

    Science.gov (United States)

    Smith, K.F.; Van Thyne, R.J.

    1959-05-12

    This patent deals with vanadium based ternary alloys useful as fuel element jackets. According to the invention the ternary vanadium alloys, prepared in an arc furnace, contain from 2.5 to 15% by weight titanium and from 0.5 to 10% by weight niobium. Characteristics of these alloys are good thermal conductivity, low neutron capture cross section, good corrosion resistance, good welding and fabricating properties, low expansion coefficient, and high strength.

  15. Corrosion inhibition measures in primary cooling water system during refurbishment of Cirus, re-commissioning and subsequent operation

    International Nuclear Information System (INIS)

    Rai, K.K.; Ramesh, N.; Sharma, R.C.

    2008-01-01

    Cirus is a 40 MWth, heavy water moderated, demineralized light water cooled, natural uranium fuelled research reactor. Reactor was commissioned in year 1960 and operated satisfactorily till 1990. After that availability factor started decreasing mainly due to equipment outage exhibiting signs of ageing. Based upon systematic ageing studies and assessment of condition of systems, structures and components, a refurbishment plan including safety upgrades was drawn up. Reactor was shut down in October 1997 for execution of jobs. After completion of refurbishment jobs reactor was started back in October 2002 and power operation was achieved in 2003. Primary cooling water (PCW) system consists of re-circulating pumps, heat exchangers, expansion tank, piping, valves, emergency storage reservoir (Ball Tank) and other components. Normally the fission heat from fuel is removed by re-circulating coolant in closed loop and transferred to seawater via heat exchangers. In case of outage of pumps, shut down cooling is provided by flow of water from Ball Tank under gravity to the underground dump tanks. The dissolved oxygen is maintained below 2 ppm and pH is maintained neutral to minimize corrosion of fuel cladding (Aluminum). This paper highlights the experience gained during segmentation of primary cooling water pipelines for pressure testing, measures taken to corrosion inhibition of primary cooling water lines to permit execution of refurbishment jobs, inspections and actions taken to repair/replace the corroded PCW pipe line segments, observations regarding corrosion related failures, re-commissioning of the system after refurbishment, assessment for safe reactor operation and experience during power operation. (author)

  16. Role of the water-drinking test in medically treated primary open angle glaucoma patients.

    Science.gov (United States)

    Salcedo, H; Arciniega, D; Mayorga, M; Wu, L

    2018-05-01

    The water-drinking test (WDT) has recently re-emerged as a possible way to determine the competency of the trabecular meshwork. We performed a prospective interventional study to test the hypothesis that the WDT could be useful in assessing fluctuations in patients undergoing treatment for primary open angle glaucoma (POAG). We included 122 patients; 62 on medical treatment for POAG (n=123 eyes) and 60 controls (n=120 eyes). The study group had been on intraocular pressures (IOP) lowering treatment continuously for at least 3months with stable IOP. The WDT was performed during fasting and was considered positive if it fluctuated ≥6mmHg. The patients on medical treatment had a mean age of 50.56±18.45 years vs. 51.35±11.22 for the controls (P=0.34); with 71% being female in the study group and 77% in the control group. In the study group; 52% were on beta blockers (n=64), 27% combination of two or more medications (n=33), 19% prostaglandin analogues (n=24) and 2% alpha agonists (n=2). The WDT was positive in 17.07% (n=21) in the study group and 2.5% (n=3) in the control group (P=0.0001). The mean fluctuation was 7.14±2.15mmHg in the study group and 6.00±0mmHg in the controls (P=0.33). A positive WDT was found in 33.33% (n=11) of those on combination therapy; 12.5% (n=3) prostaglandin analogues and 10.94% (n=7) beta blockers (P=0.03). Combination therapy had the highest positive WDT fluctuation (7.54±2.87) followed by prostaglandin analogues (7.00±1.00) and beta blockers (6.57±0.78) with a P value of 0.44. The WDT can identify significant fluctuations in eyes with POAG that are medically treated. Copyright © 2018 Elsevier Masson SAS. All rights reserved.

  17. Calculation model for predicting concentrations of radioactive corrostion products in the primary coolant of boiling water reactors

    International Nuclear Information System (INIS)

    Uchida, S.; Kikuchi, M.; Asakura, Y.; Yusa, H.; Ohsumi, K.

    1978-01-01

    A calculation model was developed to predict the shutdown dose rate around the recirculation pipes and their components in boiling water reactors (BWRs) by simulating the corrosion product transport in primary cooling water. The model is characterized by separating cobalt species in the water into soluble and insoluble materials and then calculating each concentration using the following considerations: (1) Insoluble cobalt (designated as crud cobalt is deposited directly on the fuel surface, while soluble cobalt (designated as ionic cobalt) is adsorbed on iron oxide deposits on the fuel surface. (2) Cobalt-60 activated on the fuel surface is dissolved in the water in an ionic form, and some is released with iron oxide as crud. The model can follow the reduction of 60 Co in the primary cooling water caused by the control of the iron feed rate into the reactor, which decreases the iron oxide deposits on the fuel surface and then reduces the cobalt adsorption rate. The calculated results agree satisfactorily with the measurements in several BWR plants

  18. Microstructural evolution and mechanical properties of differently heat-treated binder jet printed samples from gas- and water-atomized alloy 625 powders

    International Nuclear Information System (INIS)

    Mostafaei, Amir; Toman, Jakub; Stevens, Erica L.; Hughes, Eamonn T.; Krimer, Yuval L.; Chmielus, Markus

    2017-01-01

    In this study, we investigate the effect of powders resulting from different atomization methods on properties of binder jet printed and heat-treated samples. Air-melted gas atomized (GA) and water atomized (WA) nickel-based alloy 625 powders were used to binder jet print samples for a detailed comparative study on microstructural evolution and mechanical properties. GA printed samples achieved higher sintering density (99.2%) than WA samples (95.0%) due to differences in powder morphology and chemistry. Grain sizes of GA and WA samples at their highest density were 89 ± 21 μm and 88 ± 26 μm, respectively. Mechanical tests were conducted on optimally sintered samples and sintered plus aged samples; aging further improved microstructure and mechanical properties. This study shows that microstructural evolution (densification, and carbide, oxide and intermetallic phase formation) is very different for GA and WA binder jet printed and heat-treated samples. This difference in microstructural evolution results in different mechanical properties with the superior sintered and aged GA specimen reaching a hardness of 327 ± 7 HV_0_._1, yield strength of 394 ± 15 MPa, and ultimate tensile strength of 718 ± 14 MPa which are higher than cast alloy 625 values.

  19. Effects of water plasma immersion ion implantation on surface electrochemical behavior of NiTi shape memory alloys in simulated body fluids

    International Nuclear Information System (INIS)

    Liu, X.M.; Wu, S.L.; Chu, Paul K.; Chung, C.Y.; Chu, C.L.; Yeung, K.W.K.; Lu, W.W.; Cheung, K.M.C.; Luk, K.D.K.

    2007-01-01

    Water plasma immersion ion implantation (PIII) was conducted on orthopedic NiTi shape memory alloy to enhance the surface electrochemical characteristics. The surface composition of the NiTi alloy before and after H 2 O-PIII was determined by X-ray photoelectron spectroscopy (XPS) and atomic force microscopy (AFM) was utilized to determine the roughness and morphology of the NiTi samples. Potentiodynamic polarization tests and electrochemical impedance spectroscopy (EIS) were carried out to investigate the surface electrochemical behavior of the control and H 2 O-PIII NiTi samples in simulated body fluids (SBF) at 37 deg. C as well as the mechanism. The H 2 O-PIII NiTi sample showed a higher breakdown potential (E b ) than the control sample. Based on the AFM results, two different physical models with related equivalent electrical circuits were obtained to fit the EIS data and explain the surface electrochemical behavior of NiTi in SBF. The simulation results demonstrate that the higher resistance of the oxide layer produced by H 2 O-PIII is primarily responsible for the improvement in the surface corrosion resistance

  20. Development of precipitation strengthened brass with Ti and Sn alloying elements additives by using water atomized powder via powder metallurgy route

    Energy Technology Data Exchange (ETDEWEB)

    Li, Shufeng, E-mail: shufengli@hotmail.com [Joining and Welding Research Institute, Osaka University, Osaka (Japan); Imai, Hisashi; Kondoh, Katsuyoshi [Joining and Welding Research Institute, Osaka University, Osaka (Japan); Kojima, Akimichi; Kosaka, Yoshiharu [San-Etsu Metals Co. LTD., 1892 OHTA, Tonami, Toyama (Japan); Yamamoto, Koji; Takahashi, Motoi [Nippon Atomized Metal Powders Corporation, 87-16, Nishi-Sangao, Noda, Chiba (Japan)

    2012-08-15

    Effect of Ti and Sn alloying elements on microstructure and mechanical properties of 60/40 brass has been studied via the powder metallurgy (P/M) route. The water-atomized BS40-0.6Sn1.0Ti (Cu40wt%Zn-0.6wt%Sn1.0wt%Ti) pre-alloyed powder was consolidated at various temperatures within range of 400-600 Degree-Sign C using spark plasma sintering (SPS) and hot extrusion was carried out at 500 Degree-Sign C. Effects of extrusion temperature on microstructure and tensile strength were investigated by employing SEM-EDS/EBSD, TEM, XRD and tensile test. Results indicated that super-saturated solid solution Ti and Sn elements created high chemical potential for a precipitate reaction in rapidly solidified brass powder, which showed significant strengthening effects on the extruded sample consolidated at lower temperature. Solid solubility of Ti in brass matrix decreased with increasing of sintering temperature, thus resulted in degradation of mechanical properties. Consequently, lower hot processing temperature is necessary to obtain excellent mechanical properties for BS40-0.6Sn1.0Ti during sintering and extrusion. An yield strength of 398 MPa and ultimate tensile strength of 615 MPa were achieved, they respectively showed 31.3% and 22.9% higher values than those of extruded Cu40Zn brass. -- Graphical abstract: The Ti and Sn alloying elements additions showed significant grain refinement on Cu40Zn-0.6Sn1.0Ti brass (b) as comparing with that of the conventional Cu40Zn brass (a), detected by electron backscatter diffraction (EBSD) technique. The grain boundaries maps of (a) BS40 (b) BS40-0.6Sn1.0Ti SPS compact sintered at 400 Degree-Sign C reveals by electron backscatter diffraction (EBSD) technique. Highlights: Black-Right-Pointing-Pointer Alloying elements Ti and Sn are proposed as additives in 60/40 brass. Black-Right-Pointing-Pointer Super-saturated Ti in powder creates high chemical potential for precipitation. Black-Right-Pointing-Pointer CuSn{sub 3}Ti{sub 5

  1. Soil Water Retention and Gross Primary Productivity in the Zábrod area in the Šumava Mts

    Czech Academy of Sciences Publication Activity Database

    Šír, Miloslav; Lichner, Ľ.; Tesař, Miroslav; Krejča, M.; Váchal, J.

    roč. 3, s. 1 (2008), s130-s138 ISSN 1801-5395 R&D Projects: GA AV ČR 1QS200420562; GA ČR GA205/05/2312; GA ČR GA205/06/0375; GA MŽP(CZ) SP/1A6/151/07 Institutional research plan: CEZ:AV0Z20600510 Keywords : hydrologic cycle * evapotranspiration * gross primary productivity * entropy production * soil water retention Subject RIV: DA - Hydrology ; Limnology

  2. Surface and microstructure modifications of Ti-6Al-4V titanium alloy cutting by a water jet/high power laser converging coupling

    Science.gov (United States)

    Weiss, Laurent; Tazibt, Abdel; Aillerie, Michel; Tidu, Albert

    2018-01-01

    The metallurgical evolution of the Ti-6Al-4V samples is analyzed after an appropriate cutting using a converging water jet/high power laser system. New surface microstructures are obtained on the cutting edge as a result of thermo-mechanical effects of such hybrid fluid-jet-laser tool on the targeted material. The laser beam allows to melt and the water-jet to cool down and to evacuate the material upstream according to a controlled cutting process. The experimental results have shown that a rutile layer can be generated on the surface near the cutting zone. The recorded metallurgical effect is attributed to the chemical reaction between water molecules and titanium, where the laser thermal energy brought onto the surface plays the role of reaction activator. The width of the oxidized zone was found proportional to the cutting speed. During the reaction, hydrogen gas H2 is formed and is absorbed by the metal. The hydrogen atoms trapped into the alloy change the metastable phase formation developing pure β circular grains as a skin at the kerf surface. This result is original so it would lead to innovative converging laser water jet process that could be used to increase the material properties especially for surface treatment, a key value of surface engineering and manufacturing chains.

  3. 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

  4. 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

  5. Location of leakages in the fresh water primary network distribution in the Nuclear Center

    International Nuclear Information System (INIS)

    Rodriguez, A.G.; Perez, G.V.; Rodriguez, A.F.

    1992-01-01

    In the hydraulic net of Nuclear Centre in Salazar, Mexico was necessary the application of radioactive traces after very much efforts for locating the situation of leaks of water and not obtain satisfactory results for other methods. It was injected a small quantity of 24 Na 2 CO 3 in aqueous solution in the tank discharge which stores the fresh water. After the running water movement was followed for gamma radiation detection omitted by the 24 Na in grave digged each 100 meters along the pipes. In this work is presented the methodology used to locate two water leaks and the corresponding safety radiological considerations. (Author)

  6. Using polyatomic primary ions to probe an amino acid and a nucleic base in water ice

    Energy Technology Data Exchange (ETDEWEB)

    Conlan, X.A. [Surface Analysis Research Centre, School of Chemical Engineering and Analytical Science, University of Manchester, P.O. Box 88, Manchester M60 1QD (United Kingdom)]. E-mail: x.conlan@postgrad.manchester.ac.uk; Biddulph, G.X. [Surface Analysis Research Centre, School of Chemical Engineering and Analytical Science, University of Manchester, P.O. Box 88, Manchester M60 1QD (United Kingdom)]. E-mail: G.Biddulph@postgrad.manchester.ac.uk; Lockyer, N.P. [Surface Analysis Research Centre, School of Chemical Engineering and Analytical Science, University of Manchester, P.O. Box 88, Manchester M60 1QD (United Kingdom); Vickerman, J.C. [Surface Analysis Research Centre, School of Chemical Engineering and Analytical Science, University of Manchester, P.O. Box 88, Manchester M60 1QD (United Kingdom)]. E-mail: John.Vickerman@manchester.ac.uk

    2006-07-30

    In this study on pure water ice, we show that protonated water species [H{sub 2}O] {sub n}H{sup +} are more prevalent than (H{sub 2}O) {sub n} {sup +} ions after bombardment by Au{sup +} monoatomic and Au{sub 3} {sup +} and C{sub 60} {sup +} polyatomic projectiles. This data also reveals significant differences in water cluster yields under bombardment by these three projectiles. The amino acid alanine and the nucleic base adenine in solution have been studied and have been shown to have an effect on the water cluster ion yields observed using an Au{sub 3} {sup +} ion beam.

  7. Development status of nuclear power in China and fundamental research progress on PWR primary water chemistry in China

    International Nuclear Information System (INIS)

    Wu, Xinqiang; Liu, Xiahe; Han, En-Hou; Ke, Wei; Xu, Yuming

    2015-01-01

    China's non-fossil fuels are expected to reach 20% in primary energy ratio by 2030. It is urgent for China to speed up the development of nuclear power to increase energy supply, reduce gas emissions and optimize resource allocation. Chinese government slowed down the approval of new nuclear power plant (NPP) projects after Fukushima accident in 2011. At the end of 2012, the State Council approved the nuclear safety program and adjusted long-term nuclear power development plan (2011-2020), the new NPP's projects have been restarted. In June 2015, there are 23 operating units in mainland in China with total installed capacity of about 21.386 GWe; another 26 units are under construction with total installed capacity of 28.5 GWe. The main type of reactors in operation and under construction in China is pressurized water reactor (PWR), including the first AP1000 NPPs in the world (units 1 in Sanmen) and China self-developed Hualong one NPPs (units 5 and 6 in Fuqing). Currently, China's nuclear power development is facing historic opportunities and also a series of challenges. One of the most important is the safety and economy of nuclear power. The optimization of primary water chemistry is one of the most effective ways to minimize radiation field, mitigate material degradation and maintain fuel performance in PWR NPPs, which is also a preferred path to achieve both safety and economy for operating NPPs. In recent years, an increased attention has been paid to fundamental research and engineering application of PWR primary water chemistry in China. The present talk mainly consists of four parts: (1) development status of China's nuclear power industry; (2) safety of nuclear power and operating water chemistry; (3) fundamental research progress on Zn-injected water chemistry in China; (4) summary and future. (author)

  8. 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....

  9. 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

  10. Nonswelling alloy

    Science.gov (United States)

    Harkness, S.D.

    1975-12-23

    An aluminum alloy containing one weight percent copper has been found to be resistant to void formation and thus is useful in all nuclear applications which currently use aluminum or other aluminum alloys in reactor positions which are subjected to high neutron doses.

  11. Nonswelling alloy

    International Nuclear Information System (INIS)

    Harkness, S.D.

    1975-01-01

    An aluminum alloy containing one weight percent copper has been found to be resistant to void formation and thus is useful in all nuclear applications which currently use aluminum or other aluminum alloys in reactor positions which are subjected to high neutron doses

  12. Unsupported NiPt alloy metal catalysts prepared by water-in-oil (W/O) microemulsion method for methane cracking

    KAUST Repository

    Zhou, Lu

    2016-05-18

    Unsupported NiPt metal catalyst with Ni/Pt molar ratio of 88/12 is prepared by water-in-oil (W/O) microemulsion method in this study. Compared to monometallic Ni and Pt catalysts, the NiPt catalyst exhibits superior activity and stability for methane cracking. By XRD (X-ray powder diffraction), XPS (X-ray photoelectron spectroscopy) and TEM (Transmission electron microscopy) analyses, the formation of Ni(0)Pt(0) alloy is believed to be the main reason for the reactivity improvement of this catalyst. Carbon nano tube (CNT) with Ni(0)Pt(0) particles anchored on the top of tube are found for the NiPt catalyst. © 2016 Elsevier Ltd.

  13. Stress corrosion cracking of alloy 600 in water at high temperature: contribution to a phenomenological approach to the understanding of mechanisms

    International Nuclear Information System (INIS)

    Abadie, Pascale

    1998-01-01

    This research thesis aims at being a contribution to the understanding of mechanisms of stress corrosion cracking of an alloy 600 in water at high temperature. More precisely, it aimed at determining, by using quantitative data characterizing cracking phenomenology, which mechanism(s) is (are) able to explain crack initiation and crack growth. These data concern quantitative characterization of crack initiation, of crack growth and of the influence of two cracking parameters (strain rate, medium hydrogen content). They have been obtained by quantifying cracking through the application of a morphological model. More precisely, these data are: evolution of crack density during a tensile test at slow rate, value of initial crack width with respect to grain boundary length, and relationship between crack density and medium hydrogen content. It appears that hydrogen absorption seems to be involved in the crack initiation mechanism. Crack growth mechanisms and crack growth rates are also discussed [fr

  14. Study of the bipolar electrolysis of the tritiated water applied to the hydrogen isotopes separation by electrochemical permeation threw Pd-Ag alloy membranes

    International Nuclear Information System (INIS)

    Heinze, S.

    2000-01-01

    The objective of the study is to enrich waters of poor tritium concentration, by electrolysis in the same time of an hydrogen emission of low activity. In this framework the hydrogen electrochemical permeation threw Pd-Ag alloy membranes has been used. The first part of the study concerns the hydrogen and the deuterium diffusion threw these membranes. The activation and the thermal treatments influence have been studied. A relation between the membrane microstructure and the diffusion mechanism has been proposed. The second part of the study is devoted to the hydrogen gate mechanism determination in the membrane by impedance spectroscopy. The last part concerns the determination of the isotopic separation factor hydrogen-deuterium. Experimental results agree the calculated theoretical data. The operation of an operational membrane cell has been simulated and the process feasibility has been proved. (A.L.B.)

  15. Evaluation Of Radioactivity Concentration In The Primary Cooling Water System Of The RSG-GAS During Operation With 30% Silicide Fuels

    International Nuclear Information System (INIS)

    Hartoyo, Unggul; Udiyani, P.M.; Setiawanto, Anto

    2001-01-01

    The evaluating radioactivity concentration in the primary cooling water of the RSG-GAS during operation with 30% silicide fuels has been performed. The method of the research is sampling of primary cooling water during operation of the reactor and calculation of its radioactivity concentration. Based on the data obtained from calculation, the identified nuclides in the water are, Mn-56, Sb-124, Sb-122 and Na-24, under the limit of safety value

  16. Responses of growth and primary metabolism of water-stressed barley roots to rehydration

    Science.gov (United States)

    Barley seedlings [Hordeum vulgare L. Brant] were grown in pots in controlled environment chambers and drought treatments were imposed 11 days after sowing. Soil water content decreased from 92% to 10% after an additional 14 days of water stress. Shoot and root growth ceased after 4 and 9 days of wat...

  17. Fact sheet: National primary drinking water regulations for lead and copper

    International Nuclear Information System (INIS)

    1991-05-01

    The Fact Sheet contains a summary of what the regulations will do, establish, and provide; regulatory impact in regards to benefits and costs; treatment technique requirements; tap water monitoring for lead and copper; water quality monitoring (other than lead and copper); monitoring schedules, regulatory schedules for large, medium-sized, and small systems

  18. 75 FR 40925 - National Primary Drinking Water Regulations: Revisions to the Total Coliform Rule

    Science.gov (United States)

    2010-07-14

    ... Technology and Cost US United States UV Ultraviolet Radiation WRF Water Research Foundation Table of Contents..., in soil, and on vegetation. Coliform bacteria may be transported to surface water by run-off or to... of this preamble for detailed discussions of the routine monitoring and repeat sampling requirements...

  19. 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.

  20. 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

  1. Experimental evaluation of tritium permeation through stainless steel tubes of heat exchanger from primary to secondary water in ITER

    International Nuclear Information System (INIS)

    Nakamura, Hirofumi; Nishi, Masataka

    2004-01-01

    Tritium permeation through heat exchanger from primary cooling water to secondary cooling water has been investigated experimentally with SS316L heat exchanger under simulated ITER (international thermonuclear experimental reactor) operation condition in order to establish the tritium permeation evaluation method through the heat exchanger. As the result, the permeation rate of aqueous tritium was found to be about three orders smaller than that of the gaseous tritium. Tritium permeation through the heat exchanger in ITER was then evaluated, and it was revealed that total tritium permeation amount based on obtained aqueous permeability was about one order less than that with the former method with the gaseous permeability and putting the permeation reduction factor as 1000. Evaluated tritium permeation amount into secondary water during 20 years was quite small, which could be considered as negligible from the safety viewpoint

  2. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part II: WWER conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ernestova, M.; Zamboch, M. [Nuclear Research Institute, NRI, Rez (Czech Republic); Devrient, B.; Roth, A. [Framatome ANP GmbH, Erlangen (Germany); Ehrnsten, U. [VTT Industrial Systems, Espoo (Finland); Foehl, J.; Weissenberg, T. [Staatliche Materialpruefungsanstalt, MPA, Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, CIEMAT, Madrid (Spain); Ritter, S.; Seifert, H.P. [Paul Scherrer Institute, PSI, Villigen (Switzerland)

    2004-07-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWRs) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of crack growth behavior of LAS in high-temperature water due to EAC under constant load (steady-state power operation), to study the effect of transient conditions (during operation or start-up/shut-down of a plant) using their impact on time-based and cycle-based crack growth rates and to a more detailed understanding of the acting mechanisms. Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurized water reactor (WWER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (WWER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarizes the most important crack growth results obtained under simulated WWER conditions. The influence of oxygen content and the effect of specimen size (C(T)25 versus C(T)50 specimens) on the crack growth rates are shown. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  3. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part II: WWER conditions

    International Nuclear Information System (INIS)

    Ernestova, M.; Zamboch, M.; Devrient, B.; Roth, A.; Ehrnsten, U.; Foehl, J.; Weissenberg, T.; Gomez-Briceno, D.; Lapena, J.; Ritter, S.; Seifert, H.P.

    2004-01-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWRs) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of crack growth behavior of LAS in high-temperature water due to EAC under constant load (steady-state power operation), to study the effect of transient conditions (during operation or start-up/shut-down of a plant) using their impact on time-based and cycle-based crack growth rates and to a more detailed understanding of the acting mechanisms. Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurized water reactor (WWER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (WWER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarizes the most important crack growth results obtained under simulated WWER conditions. The influence of oxygen content and the effect of specimen size (C(T)25 versus C(T)50 specimens) on the crack growth rates are shown. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  4. Primary production in a shallow water lake with special reference to a reed swamp

    International Nuclear Information System (INIS)

    Andersen, F.Oe.

    1976-01-01

    Phytoplankton gross primary production ( 14 C method) in the shallow, eutrophic Danish Lake Arresoe in 1973 was 980 g C m -2 . Calculated net primary production was near zero. Macrophyte net primary production was measured by harvesting the maximum biomass, and above ground values were between 420 and 1325 g ash free dry wt m -2 , while below ground values were between 2480 and 8570 g ash free dry wt m -2 . The reed swamps were mapped on aerial photographs, and the composition of the macrophyte vegetation was determined. A comparison of macrophyte vegetation in 1944 and 1972 showed a reduction in species diversity, especially of submerged species. The seasonal variations in physical and chemical data indicated strong eutrophication in Arresoe. (author)

  5. [The primary medical sanitary care and characteristics of drinking water supply of population].

    Science.gov (United States)

    Nechaev, V S; Saurina, O S

    2016-01-01

    The article considers characteristics of organization ofprimary medical sanitary care on territory with carcinogenic risks related to drinking water supply as exemplified by the Orlovskaia oblast. The importance of registration by local health authorities the sources of permanent chemical pollution of drinking water. The analysis of the State program of the Orlovskaia oblast “The development of health care in the Orlovskaia oblast in 2013-2020". The necessity of additional inclusion of issue related to healthy drinking water supply of population to prevent development of malignant neoplasms and prevalence of oncologic morbidity on oblast territory.

  6. Standard and hydrazine water chemistry in primary circuit of VVER 440 units

    International Nuclear Information System (INIS)

    Burclova, J.

    1992-01-01

    Standard ammonia-potassium-boron water chemistry of 8 units with VVER 440 in CSFR is discussed as well as the corrosion product activity in the coolant during steady state and shut-down period and surface activity, dose rate build-up and occupational radiation exposure. Available data on hydrazine application (USSR, Hungary) indicate the possibility of the radiation field decreasing. Nevertheless the detailed analysis of 55 cycles of operation under standard water chemistry in Czechoslovakia allows to expect the comparable results for both water chemistries. (author)

  7. 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.

  8. 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.

  9. 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)

  10. The effect of Laser Shock Peening on Fatigue Life Using Pure Water and Hydrofluoric Acid As a Confining Layer of Al – Alloy 7075-T6

    Directory of Open Access Journals (Sweden)

    Shaker Sakran Hassan

    2018-01-01

    Full Text Available Laser shock peening (LSP is deemed as a deep-rooted technology for stimulating compressive residual stresses below the surface of metallic elements. As a result, fatigue lifespan is improved, and the substance properties become further resistant to wear and corrosion. The LSP provides more unfailing surface treatment and a potential decrease in microstructural damage. Laser shock peening is a well-organized method measured up to the mechanical shoot peening. This kind of surface handling can be fulfilled via an intense laser pulse focused on a substantial surface in extremely shorter intervals. In this work, Hydrofluoric Acid (HF and pure water as a coating layer were utilized as a new technique to improve the properties and to harden the treated surface of the Al -alloy 7075-T6. Fatigue life by means of laser peened workpieces was improved to 154.3%, 9.78%, respectively, for Hydrofluoric (HF and pure water compared to un-peened specimens. And the outcomes of Vickers hardness test for laser shock peening with acid and pure water as well as un-peened specimens were 165.2HV30, 143.95HV30 and 134.7HV30, respectively showed a significant improvement in the hardness property.

  11. Tiniest primary producers in the marine environment: An appraisal from the context of waters around India

    Digital Repository Service at National Institute of Oceanography (India)

    Mitbavkar, S.; Anil, A.C.

    Phytoplankton (0.2 mm - 2 mm) are the major primary producers in the marine environment thereby forming a basic link in the marine food web. They are categorized into different groups depending on their size range. Cells in the size range of 0...

  12. 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

  13. [A hepatitis A outbreak caused by contaminated well water in a primary school of Jiangxi province, China, 2009].

    Science.gov (United States)

    Chen, Jing; Cheng, Hui-jian; Zhang, Li-jie; Zong, Jun; Ma, Hui-lai; Zhu, Bao-ping

    2011-10-01

    A hepatitis A outbreak in a primary school was reported by Gan County Center for Disease Control and Province (CDC) and an investigation was conducted to identify the possible source of infection and risk factors for transmission. A probable case was defined as having onset of jaundice (yellow urine, sclera or skin) or a 2-fold increase in Alanine aminotransferase with 2 or more, of the followings symptoms: anorexia, disgust of oil, abdominal pain, nausea, fatigue, vomiting, in students and staff of the primary school between 1 November 2008 and 14 February 2009. A confirmed case was IgM positive for hepatitis A, added on a probable case. We searched for cases through reviewing medical records in the township hospital and village clinics and conducting symptom screening among students or teachers. We also conducted a case-control study to compare the exposure histories of 19 cases and 53 anti-HAV-IgM negative controls randomly selected from those asymptomatic students in the same grade. 21 cases from all the students was identified, with the attack rate as 3.5%. The epidemic curve showed the two peaks of the outbreak were 28 days apart, both indicating that they were related to the exposure of the source of origin. 74% of the case-students drank the unboiled Well B water, compared to 42% of control-students (OR = 4.0, 95%CI: 1.1 - 15). The total bacterial count was 600 cfu/ml and the total coliform was 23 MPN/100 ml in one sample collected from the well water. This hepatitis A outbreak was caused by drinking contaminated water in Well B. We recommended that all the schools should use chlorinated municipal pipe water. Public health authorities should strengthen the supervision of quality of water in schools.

  14. 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)

  15. 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....

  16. The binding of the primary water of hydration to nucleosides, CsDNA and potassium hyaluronate

    Science.gov (United States)

    Lukan, A. M.; Cavanaugh, D.; Whitson, K. B.; Marlowe, R. L.; Lee, S. A.; Anthony, L.; Rupprecht, A.; Mohan, V.

    1998-03-01

    Differential scanning calorimetry (DSC) has been used to study the eight nucleosides, CsDNA and KHA hydrated at 59% relative humidity. Thermograms were measured between 25 and 180 ^oC for scan rates of 1, 2, 5, 10 and 20 K/min. A broad endothermic transition (due to the desorption of the water) near 80 ^oC was observed for all runs. The average enthalpy of desorption per water molecule was evaluated from the area under the peak. A Kissinger analysis of these data yielded the net activation energy for desorption. Both parameters were very similar for the two biopolymers. Rayleigh scattering of Mossbauer radiation (RSMR) data(G. Albanese et al. ) Hyperfine Int. 95, 97 (1995) were analyzed via a simple harmonic oscillator model to evaluate the effective force constant of the water bound to the biopolymer. This analysis suggests that the effective force constant of water bound to HA is much larger (about 5 times) than for water bound to DNA.

  17. Concerted changes in N and C primary metabolism in alfalfa (Medicago sativa) under water restriction.

    Science.gov (United States)

    Aranjuelo, Iker; Tcherkez, Guillaume; Molero, Gemma; Gilard, Françoise; Avice, Jean-Christophe; Nogués, Salvador

    2013-02-01

    Although the mechanisms of nodule N(2) fixation in legumes are now well documented, some uncertainty remains on the metabolic consequences of water deficit. In most cases, little consideration is given to other organs and, therefore, the coordinated changes in metabolism in leaves, roots, and nodules are not well known. Here, the effect of water restriction on exclusively N(2)-fixing alfalfa (Medicago sativa L.) plants was investigated, and proteomic, metabolomic, and physiological analyses were carried out. It is shown that the inhibition of nitrogenase activity caused by water restriction was accompanied by concerted alterations in metabolic pathways in nodules, leaves, and roots. The data suggest that nodule metabolism and metabolic exchange between plant organs nearly reached homeostasis in asparagine synthesis and partitioning, as well as the N demand from leaves. Typically, there was (i) a stimulation of the anaplerotic pathway to sustain the provision of C skeletons for amino acid (e.g. glutamate and proline) synthesis; (ii) re-allocation of glycolytic products to alanine and serine/glycine; and (iii) subtle changes in redox metabolites suggesting the implication of a slight oxidative stress. Furthermore, water restriction caused little change in both photosynthetic efficiency and respiratory cost of N(2) fixation by nodules. In other words, the results suggest that under water stress, nodule metabolism follows a compromise between physiological imperatives (N demand, oxidative stress) and the lower input to sustain catabolism.

  18. Q-factor of coolant flow in the primary circuit of NPP with pressurised water reactors

    International Nuclear Information System (INIS)

    Proskuryakov, K.N.; Belikov, S.O.; Novikov, K.S.

    2011-01-01

    Systems of preoperational vibration dynamic monitoring in of WWER are presented. The results of measurements during commission of NPP with WWER are presented. The paper provides the result of the research, that estimation of coolant fluctuations caused by pulse perturbation of pressure in the primary circuit NPP. It is shown that results could be received at known value of a Q - factor of acoustical oscillatory system only. The research demonstrates the results of dependence of the sound speed from the mass steam content in the coolant flow thru reactor core. The worked out results can be used for identification of the reasons of abnormal growth of level of vibrations of fuel assembly, fuel rod, equipment and internals, and for forecasting the operation conditions which provide of vibration - acoustical resonances in the primary loop equipment. (author)

  19. Dose rates modeling of pressurized water reactor primary loop components with SCALE6.0

    International Nuclear Information System (INIS)

    Matijević, Mario; Pevec, Dubravko; Trontl, Krešimir

    2015-01-01

    Highlights: • Shielding analysis of the typical PWR primary loop components was performed. • FW-CADIS methodology was thoroughly investigated using SCALE6.0 code package. • Versatile ability of SCALE6.0/FW-CADIS for deep penetration models was proved. • The adjoint source with focus on specific material can improve MC modeling. - Abstract: The SCALE6.0 simulation model of a typical PWR primary loop components for effective dose rates calculation based on hybrid deterministic–stochastic methodology was created. The criticality sequence CSAS6/KENO-VI of the SCALE6.0 code package, which includes KENO-VI Monte Carlo code, was used for criticality calculations, while neutron and gamma dose rates distributions were determined by MAVRIC/Monaco shielding sequence. A detailed model of a combinatorial geometry, materials and characteristics of a generic two loop PWR facility is based on best available input data. The sources of ionizing radiation in PWR primary loop components included neutrons and photons originating from critical core and photons from activated coolant in two primary loops. Detailed calculations of the reactor pressure vessel and the upper reactor head have been performed. The efficiency of particle transport for obtaining global Monte Carlo dose rates was further examined and quantified with a flexible adjoint source positioning in phase-space. It was demonstrated that generation of an accurate importance map (VR parameters) is a paramount step which enabled obtaining Monaco dose rates with fairly uniform uncertainties. Computer memory consumption by the S N part of hybrid methodology represents main obstacle when using meshes with large number of cells together with high S N /P N parameters. Detailed voxelization (homogenization) process in Denovo together with high S N /P N parameters is essential for precise VR parameters generation which will result in optimized MC distributions. Shielding calculations were also performed for the reduced PWR

  20. Linear titration plot for the determination of boron in the primary coolant of a pressurized water reactor

    International Nuclear Information System (INIS)

    Midgley, D.; Gatford, C.

    1992-01-01

    A linear titration plot method has been devised for the determination of boron as boric acid in partly neutralized solution, such as occurs in the primary coolant of pressurized water reactors. The total boron and the alkali in the sample are determined simultaneously. Although it is not essential to add mannitol in this method, it is more accurate when the solution is saturated with mannitol. Comparisons are made with other modes of titration: Gran plots, first and second differential potentiometric titrations and indicator titrations. None of these gives the total boron directly in partly neutralized solutions. (author)

  1. Nonlinear dynamic response analysis in piping system for a loss of coolant accident in primary loop of pressurized water reactor

    International Nuclear Information System (INIS)

    Zhang Xiwen; He Feng; Hao Pengfei; Wang Xuefang

    2000-01-01

    Based on the elaborate force and moment analysis with characteristics method and control-volume integrating method for the piping system of primary loop under pressurized water reactor' loss of coolant accident (LOCA) conditions, the nonlinear dynamic response of this system is calculated by the updated Lagrangian formulation (ADINA code). The piping system and virtual underpinning are specially processed, the move displacement of the broken pipe with time is accurately acquired, which is very important and useful for the design of piping system and virtual underpinning

  2. The Design and Development of a Simulation to Teach Water Conservation to Primary School Students

    Science.gov (United States)

    Campbell, Lee

    2004-01-01

    Information and Communications Technology (ICT) plays a dominant role in enhancing teaching and learning. Similar advances have been made in the use of multimedia in the classroom. These advances are coupled with newer developmental tools and techniques. This paper examines the design and development of a simulation on water conservation. Science…

  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, chemica