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Sample records for alloy-802 incoloy

  1. Tangental Turning of Incoloy Alloy 925 Using Abrasive Water Jet

    OpenAIRE

    Cárach, J.; S. Hloch; Hlaváček, P.; K. Vasilko; Lehocká, D.

    2015-01-01

    The paper deals with tangential turning of Incoloy alloy 925 with the diameter 50mm using the abrasive water jet. Experiment was done using the abrasive water jet of pressure p=400MPa and traverse speed at levels of v=1,5;3;4,5;6;7,5;9mm min-1. the abrasive particles were feeded to the water jet in the amount of 400 g min-1. Revolution of Incoloy workpiece during turning was n=34rpm.

  2. Tritium Permeability of Incoloy 800H and Inconel 617

    Energy Technology Data Exchange (ETDEWEB)

    Philip Winston; Pattrick Calderoni; Paul Humrickhouse

    2012-07-01

    Design of the Next Generation Nuclear Plant (NGNP) reactor and its high-temperature components requires information regarding the permeation of fission generated tritium and hydrogen product through candidate heat exchanger alloys. Release of fission-generated tritium to the environment and the potential contamination of the helium coolant by permeation of product hydrogen into the coolant system represent safety basis and product contamination issues. Of the three potential candidates for high-temperature components of the NGNP reactor design, only permeability for Incoloy 800H has been well documented. Hydrogen permeability data have been published for Inconel 617, but only in two literature reports and for partial pressures of hydrogen greater than one atmosphere, far higher than anticipated in the NGNP reactor. To support engineering design of the NGNP reactor components, the tritium permeability of Inconel 617 and Incoloy 800H was determined using a measurement system designed and fabricated at Idaho National Laboratory. The tritium permeability of Incoloy 800H and Inconel 617, was measured in the temperature range 650 to 950°C and at primary concentrations of 1.5 to 6 parts per million volume tritium in helium. (partial pressures of 10-6 atm)—three orders of magnitude lower partial pressures than used in the hydrogen permeation testing. The measured tritium permeability of Incoloy 800H and Inconel 617 deviated substantially from the values measured for hydrogen. This may be due to instrument offset, system absorption, presence of competing quantities of hydrogen, surface oxides, or other phenomena. Due to the challenge of determining the chemical composition of a mixture with such a low hydrogen isotope concentration, no categorical explanation of this offset has been developed.

  3. Tritium Permeability of Incoloy 800H and Inconel 617

    Energy Technology Data Exchange (ETDEWEB)

    Philip Winston; Pattrick Calderoni; Paul Humrickhouse

    2011-09-01

    Design of the Next Generation Nuclear Plant (NGNP) reactor and its high-temperature components requires information regarding the permeation of fission generated tritium and hydrogen product through candidate heat exchanger alloys. Release of fission-generated tritium to the environment and the potential contamination of the helium coolant by permeation of product hydrogen into the coolant system represent safety basis and product contamination issues. Of the three potential candidates for high-temperature components of the NGNP reactor design, only permeability for Incoloy 800H has been well documented. Hydrogen permeability data have been published for Inconel 617, but only in two literature reports and for partial pressures of hydrogen greater than one atmosphere, far higher than anticipated in the NGNP reactor. To support engineering design of the NGNP reactor components, the tritium permeability of Inconel 617 and Incoloy 800H was determined using a measurement system designed and fabricated at Idaho National Laboratory. The tritium permeability of Incoloy 800H and Inconel 617, was measured in the temperature range 650 to 950 C and at primary concentrations of 1.5 to 6 parts per million volume tritium in helium. (partial pressures of 10-6 atm) - three orders of magnitude lower partial pressures than used in the hydrogen permeation testing. The measured tritium permeability of Incoloy 800H and Inconel 617 deviated substantially from the values measured for hydrogen. This may be due to instrument offset, system absorption, presence of competing quantities of hydrogen, surface oxides, or other phenomena. Due to the challenge of determining the chemical composition of a mixture with such a low hydrogen isotope concentration, no categorical explanation of this offset has been developed.

  4. A comparison of the solidification behavior of Incoloy 909 and Inconel 718

    Energy Technology Data Exchange (ETDEWEB)

    Cieslak, M.J.; Knorovsky, G.A. (Process Metallurgy Div., Sandia National Lab., Albuquerque, NM (US)); Headley, T.J. (Electron Optics Div., Sandia National Lab., Albuquerque, NM (US)); Romig, A.D. (Physical Metallurgy Div., Sandia National Lab., Albuquerque, NM (US)); Kollie, T. (Oak Ridge National Lab., TN (USA))

    1990-02-01

    The solidification behavior of two commercial aerospace superalloys, Incoloy 909 and Inconel 718, has been examined. Differential thermal analysis (DTA) revealed that Incoloy 909 terminates solidification with the formation of a single minor constituent at {approx} 1198{degrees}C. Inconel 718 terminates solidification with the formation of two minor constituents at {approx} 1257{degrees}C and {approx} 1185{degrees}C, respectively. Metallography confirmed that a single minor constituent was present in Incoloy 909 while two minor constituents were present in Inconel 718. Samples were also examined by electron probe microanalysis to reveal the patterns of elemental segregation.

  5. MODELLING AND CHARACTERIZATION OF LASER WELDED INCOLOY 800 HT JOINTS

    Directory of Open Access Journals (Sweden)

    Sathiya Paulraj

    2016-06-01

    Full Text Available This study aims at finding the effect of laser welding speed on incoloy 800 HT. This alloy is one of the potential materials for Generation IV nuclear plants. Laser welding has several advantages over arc welding such as low fusion zone, low heat input and concentrated heat intensity. Three different welding speeds were chosen and CO2 laser welding was performed. 2D modeling and simulation were done using ANSYS 15 to find out the temperature distribution at different welding speeds and it was found that an increase in the welding speed decreased the temperature. Mechanical properties such as tensile strength, toughness and hardness were evaluated. The effect of welding speed on metallurgical characteristics was studied using optical microscopy (OM, Scanning Electron Microscopy (SEM with EDS, X-Ray Diffraction (XRD technique and fractographic analysis. From the results it was found that high welding speed (1400 mm/min decreased the joint strength. The M23C6 and Ni3Ti carbides were formed in a discrete chain and in a globular form along the grain boundaries of the weld region which increased the strength of the grain boundaries. Fractographic evaluations of the tested specimens for welding speed (1000 and 1200 mm/min showed deep and wide dimples indicating ductile failures.

  6. Welding Heat Input Effect on Microstructure of Incoloy 28 and S31254

    International Nuclear Information System (INIS)

    This work focuses on studying the effect of welding heat input within the range from one to five kJ/mm on the microstructure of weld metal and heat affected zone (HAZ), micro segregation in the unmixed zone (UMZ) and corrosion resistance of super austenitic stainless steel (SASS) and Incoloy 28. The two materials were welded bead-on-plate with Ni-base electrode ER NiCrMo3. The microstructure, micro segregation in UMZ and corrosion properties of SASS and Incoloy 28 were studied using optical microscope, SEM, EDX, pitting and crevice corrosion tests. No UMZ was observed in Incoloy 28 welded at heat input of one kJ/mm but was observed at three and five kJ/mm. The UMZ appears clearly in SASS welded at all investigated heat inputs. The width of the UMZ increased with increasing heat input. The Mo micro segregation in UMZ within Incoloy 28 increased as the heat input increased from 3 to 5 kJ/mm while that in SASS increased with increasing heat input from 1 to 3 kJ/mm and then decreased slightly with increasing heat input from 3 to 5 kJ/mm. Corrosion resistance was higher after welding with heat input of 1 kJ/mm compared with higher heat inputs.

  7. Hydrogen Permeability of Incoloy 800H, Inconel 617, and Haynes 230 Alloys

    International Nuclear Information System (INIS)

    A potential issue in the design of the NGNP reactor and high-temperature components is the permeation of fission generated tritium and hydrogen product from downstream hydrogen generation through high-temperature components. Such permeation can result in the loss of fission-generated tritium to the environment and the potential contamination of the helium coolant by permeation of product hydrogen into the coolant system. The issue will be addressed in the engineering design phase, and requires knowledge of permeation characteristics of the candidate alloys. Of three potential candidates for high-temperature components of the NGNP reactor design, the hydrogen permeability has been documented well only for Incoloy 800H, but at relatively high partial pressures of hydrogen. Hydrogen permeability data have been published for Inconel 617, but only in two literature reports and for partial pressures of hydrogen greater than one atmosphere, far higher than anticipated in the NGNP reactor. The hydrogen permeability of Haynes 230 has not been published. To support engineering design of the NGNP reactor components, the hydrogen permeability of Inconel 617 and Haynes 230 were determined using a measurement system designed and fabricated at the Idaho National Laboratory. The performance of the system was validated using Incoloy 800H as reference material, for which the permeability has been published in several journal articles. The permeability of Incoloy 800H, Inconel 617 and Haynes 230 was measured in the temperature range 650 to 950 C and at hydrogen partial pressures of 10-3 and 10-2 atm, substantially lower pressures than used in the published reports. The measured hydrogen permeability of Incoloy 800H and Inconel 617 were in good agreement with published values obtained at higher partial pressures of hydrogen. The hydrogen permeability of Inconel 617 and Haynes 230 were similar, about 50% greater than for Incoloy 800H and with similar temperature dependence.

  8. Hydrogen Permeability of Incoloy 800H, Inconel 617, and Haynes 230 Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Pattrick Calderoni

    2010-07-01

    A potential issue in the design of the NGNP reactor and high-temperature components is the permeation of fission generated tritium and hydrogen product from downstream hydrogen generation through high-temperature components. Such permeation can result in the loss of fission-generated tritium to the environment and the potential contamination of the helium coolant by permeation of product hydrogen into the coolant system. The issue will be addressed in the engineering design phase, and requires knowledge of permeation characteristics of the candidate alloys. Of three potential candidates for high-temperature components of the NGNP reactor design, the hydrogen permeability has been documented well only for Incoloy 800H, but at relatively high partial pressures of hydrogen. Hydrogen permeability data have been published for Inconel 617, but only in two literature reports and for partial pressures of hydrogen greater than one atmosphere, far higher than anticipated in the NGNP reactor. The hydrogen permeability of Haynes 230 has not been published. To support engineering design of the NGNP reactor components, the hydrogen permeability of Inconel 617 and Haynes 230 were determined using a measurement system designed and fabricated at the Idaho National Laboratory. The performance of the system was validated using Incoloy 800H as reference material, for which the permeability has been published in several journal articles. The permeability of Incoloy 800H, Inconel 617 and Haynes 230 was measured in the temperature range 650 to 950 °C and at hydrogen partial pressures of 10-3 and 10-2 atm, substantially lower pressures than used in the published reports. The measured hydrogen permeability of Incoloy 800H and Inconel 617 were in good agreement with published values obtained at higher partial pressures of hydrogen. The hydrogen permeability of Inconel 617 and Haynes 230 were similar, about 50% greater than for Incoloy 800H and with similar temperature dependence.

  9. Effects of PbO on the oxide films of incoloy 800HT in simulated primary circuit of PWR

    Science.gov (United States)

    Tan, Yu; Yang, Junhan; Wang, Wanwan; Shi, Rongxue; Liang, Kexin; Zhang, Shenghan

    2016-05-01

    Effects of trace PbO on oxide films of Incoloy 800HT were investigated in simulated primary circuit water chemistry of PWR, also with proper Co addition. The trace PbO addition in high temperature water blocked the protective spinel oxides formation of the oxide films of Incoloy 800HT. XPS results indicated that the lead, added as PbO into the high temperature water, shows not only +2 valance but also +4 and 0 valances in the oxide film of 800HT co-operated with Fe, Cr and Ni to form oxides films. Potentiodynamic polarization results indicated that as PbO concentration increased, the current densities of the less protective oxide films of Incoloy 800HT decreased in a buffer solution tested at room temperature. The capacitance results indicated that the donor densities of oxidation film of Incoloy 800HT decreased as trace PbO addition into the high temperature water.

  10. Effects of grain boundary sliding on the flow properties of Incoloy 800H

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, H.; Korhonen, M.A.; Li Cheyu (Dept. of Materials Science and Enginering, Cornell Univ., Ithaca, NY (United States))

    1992-08-01

    The nature of grain boundary sliding (GBS) is investigated in Incoloy 800H in terms of the effects of stress, temperature and grain size on the flow behavior observed by using the load relaxation test. Flow behaviors are obtained for average grain sizes ranging from 6 to 225 {mu}m at temperatures between 614 and 746degC. The flow behavior of large-grain-size material plotted as stress vs. strain rate in a doubly logarithmic scale, exhibits a sigmoidal shape which has been commonly associated with the effects of GBS on creep deformation. For the materials of smaller grain sizes the deformation properties tend toward those characteristic of structural superplasticity. It is shown that in Incoloy 800H there may exist, as a function of the grain size, a continuous scale of flow properties ranging from the normal creep to superplastic-like behavior. (orig.).

  11. Microstructural characterization of dissimilar welds between Incoloy 800H and 321 Austenitic Stainless Steel

    Energy Technology Data Exchange (ETDEWEB)

    Sayiram, G., E-mail: sayiram.g@vit.ac.in; Arivazhagan, N.

    2015-04-15

    In this work, the microstructural character of dissimilar welds between Incoloy 800H and 321 Stainless Steel has been discussed. The microscopic examination of the base metals, fusion zones and interfaces was characterized using an optical microscope and scanning electron microscopy. The results revealed precipitates of Ti (C, N) in the austenitic matrix along the grain boundaries of the base metals. Migration of grain boundaries in the Inconel 82 weld metal was very extensive when compared to Inconel 617 weldment. Epitaxial growth was observed in the 617 weldment which increases the strength and ductility of the weld metal. Unmixed zone near the fusion line between 321 Stainless Steel and Inconel 82 weld metal was identified. From the results, it has been concluded that Inconel 617 filler metal is a preferable choice for the joint between Incoloy 800H and 321 Stainless Steel. - Highlights: • Failure mechanisms produced by dissimilar welding of Incoloy 800H to AISI 321SS • Influence of filler wire on microstructure properties • Contemplative comparisons of metallurgical aspects of these weldments • Microstructure and chemical studies including metallography, SEM–EDS • EDS-line scan study at interface.

  12. Load relaxation studies of grain boundary sliding in Incoloy 800H

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, H.; Hannula, S.P.; Korhonen, M.A.; Suzuki, H.; Li, C.Y (Cornell Univ., Ithaca, NY (USA))

    Load relaxation tests were performed on Incoloy 800H at elevated temperatures as a function of prior plastic deformation. The log stress vs. log strain rate curves obtained exhibit the typical sigmoidal shape predicted by current theories. A stress enhancement factor with a value near 0.7 can be estimated based on limiting stress values both at the high and low strain rate ends. The results of data analysis yielded long grain boundary stress vs. log grain boundary sliding rate curves. These curves are found to show grain matrix-like characteristics. The significance of these results is discussed in terms of a state variable theory.

  13. The crystallography of fatigue crack initiation in Incoloy-908 and A-286 steel

    Energy Technology Data Exchange (ETDEWEB)

    Krenn, C.R. [Univ. of California, Berkeley, CA (United States). Dept. of Materials Science and Mineral Engineering]|[Lawrence Berkeley National Lab., CA (United States)

    1996-12-01

    Fatigue crack initiation in the austenitic Fe-Ni superalloys Incoloy-908 and A-286 is examined using local crystallographic orientation measurements. Results are consistent with sharp transgranular initiation and propagation occurring almost exclusively on {l_brace}111{r_brace} planes in Incoloy-908 but on a variety of low index planes in A-286. This difference is attributed to the influence of the semicoherent grain boundary {eta} phase in A-286. Initiation in each alloy occurred both intergranularly and transgranularly and was often associated with blocky surface oxide and carbide inclusions. Taylor factor and resolved shear stress and strain crack initiation hypotheses were tested, but despite an inconclusive suggestion of a minimum required {l_brace}111{r_brace} shear stress, none of the hypotheses were found to convincingly describe preferred initiation sites, even within the subsets of transgranular cracks apparently free from the influence of surface inclusions. Subsurface inclusions are thought to play a significant role in crack initiation. These materials have applications for use in structural conduit for high field superconducting magnets designed for fusion energy use.

  14. Effect of molybdenum on grain boundary segregation in Incoloy 901 superalloy

    International Nuclear Information System (INIS)

    Highlights: ► Grain boundary segregation in superalloys can be decreased by controlling of the chemical composition of alloys. ► One of the most effective element on decreasing of grain boundary segregation is Molybdenum. ► The Mo addition up to 6.7 % have a suitable effect on decreasing of grain boundary segregation of other elements.. ► The partitioning coefficients of all elements except Fe and Ni are less than one. - Abstract: In this paper, the effect of molybdenum on the grain boundary segregation of other elements was studied in Incoloy 901 superalloy. Initially, five alloys were prepared with different percentages of Mo by using a vacuum induction furnace. Then, these alloys were remelted by Electro-slag remelting (ESR) process and after homogenizing at 1160 °C for 2 h followed by air cooling, were rolled. The effect of Mo on segregation of elements was evaluated with Scanning Electron Microscopy, Linear Analysis, and the mechanical tests. The results showed that the grain boundary segregations of elements in Incoloy 901 superalloy were decreased by increasing of molybdenum content up to 6.7% and the mechanical properties (tensile and hardness properties) were improved. Also, the segregations of elements were increased by increasing the percentage of Mo from 6.7 to 7.5, and the mechanical properties were reduced

  15. The crystallography of fatigue crack initiation in Incoloy-908 and A-286 steel

    International Nuclear Information System (INIS)

    Fatigue crack initiation in the austenitic Fe-Ni superalloys Incoloy-908 and A-286 is examined using local crystallographic orientation measurements. Results are consistent with sharp transgranular initiation and propagation occurring almost exclusively on {111} planes in Incoloy-908 but on a variety of low index planes in A-286. This difference is attributed to the influence of the semicoherent grain boundary η phase in A-286. Initiation in each alloy occurred both intergranularly and transgranularly and was often associated with blocky surface oxide and carbide inclusions. Taylor factor and resolved shear stress and strain crack initiation hypotheses were tested, but despite an inconclusive suggestion of a minimum required {111} shear stress, none of the hypotheses were found to convincingly describe preferred initiation sites, even within the subsets of transgranular cracks apparently free from the influence of surface inclusions. Subsurface inclusions are thought to play a significant role in crack initiation. These materials have applications for use in structural conduit for high field superconducting magnets designed for fusion energy use

  16. A comparative study of the corrosion resistance of incoloy MA 956 and PM 2000 superalloys

    Directory of Open Access Journals (Sweden)

    Maysa Terada

    2010-12-01

    Full Text Available Austenitic stainless steels, titanium and cobalt alloys are widely used as biomaterials. However, new medical devices require innovative materials with specific properties, depending on their application. The magnetic properties are among the properties of interest for some biomedical applications. However, due to the interaction of magnetic materials with Magnetic Resonance Image equipments they might used only as not fixed implants or for medical devices. The ferromagnetic superalloys, Incoloy MA 956 and PM 2000, produced by mechanical alloying, have similar chemical composition, high corrosion resistance and are used in high temperature applications. In this study, the corrosion resistance of these two ferritic superalloys was compared in a phosphate buffer solution. The electrochemical results showed that both superalloys are passive in this solution and the PM 2000 present a more protective passive film on it associated to higher impedances than the MA 956.

  17. Influence of niobium additions on mechanical properties and corrosion of INCOLOY 800 H

    International Nuclear Information System (INIS)

    The studies were carried out with six model alloys of the type INCOLOY alloy 800 H (32 Ni/20 Cr), obtained by variation of the niobium additions with up to 1.55 wt. p.c. of Nb. The mechanical properties and structural characteristics of these samples are listed after treatments as follows: - Aging at 650, 800, and 9000C (Notch bending tests and tensile tests at room temperature). - Carbonisation at 800 and 9000C in PNP standard helium (C-analysis, long-term creep tests at 9000C). Alloys with Nb additions showed constant good strength and ductility after aging, values being better than those for material without Nb additions. The creep tests showed that tensile strengths is improved with increasing niobium content; carbonisation is less than in alloys without Nb. (orig./IHOE)

  18. Electro chemical studies on stress corrosion cracking of Incoloy-800 in caustic solution, part I: As received samples

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    Dinu Alice

    2005-01-01

    Full Text Available Many non-volatile impurities accidentally introduced into the steam generator tend to Concentrate on its surface in restricted flow areas. In this way these impurities can lead to stress corrosion cracking (SCC on stressed tubes of the steam generator. Such impurities can be strong alkaline or acid solutions. To evaluate the effect of alkaline concentrated environments on SCC of steam generator tubes, the tests were con ducted on stressed samples of Incoloy-800 in 10% NaOH solution. To accelerate the SCC process, stressed specimens were anodically polarised in a caustic solution in an electro chemical cell. The method of stressing of Incoloy-800 tubes used in our experiments was the C-ring. Using the cathodic zone of the potentiodynamic curves it was possible to calculate the most important electrochemical parameters: the corrosion current, the corrosion rate, and the polarization resistance. We found that the value of the corrosion potential to initiate the SCC microcracks was -100 mV. The tested samples were examined using the metallographic method. The main experimental results showed that the in crease of the stress state promoted the in crease of the SCC susceptibility of Incoloy-800 samples tested under the same conditions, and that the length of the SCC-type microcracks in creased with the growth of the stress value.

  19. Electrochemical studies on stress corrosion cracking of Incoloy-800 in caustic solution Part I: As received samples

    International Nuclear Information System (INIS)

    Many non-volatile impurities accidentally introduced into the steam generator tend to concentrate on its surface in restricted flow areas. In this way these impurities can lead to stress corrosion cracking (SCC) on stressed tubes of the steam generator. Such impurities can be strong alkaline or acid solutions. To evaluate the effect of alkaline concentrated environments on SCC of steam generator tubes, the tests were conducted on stressed samples of Incoloy-800 in 10% NaOH solution. To accelerate the SCC process, stressed specimens were anodically polarised in a caustic solution in an electrochemical cell. The method of stressing of Incoloy-800 tubes used in our experiments was the C-ring. Using the cathodic zone of the potentiodynamic curves it was possible to calculate the most important electrochemical parameters: the corrosion current, the corrosion rate, and the polarisation resistance. We found that the value of the corrosion potential to initiate the SCC microcracks was -100 mV. The tested samples were examined using the metallographic method. The main experimental results showed that the increase of the stress state promoted the increase of the SCC susceptibility of Incoloy-800 samples tested under the same conditions, and that the length of the SCC-type microcracks increased with the growth of the stress value. (author)

  20. Temperature dependence of the dynamic fracture toughness of the alloy Incoloy 800 after cold work

    International Nuclear Information System (INIS)

    Instrumental impact tests were performed on prefatigued charpy-impact tests specimens of the material Incoloy 800 in the as-received state and after 5%, 10% and 25% cold work in the temperature range 293 ≤ T/K ≤ 1223. The impact energi es were determined with different methods, the agreement was excellent. The temperature dependence of the impact energies exhibit an intermediate maximum at 673 K (400deg C). This maximum is more or less suppressed with increasing cold work. The temperature dependence of the fracture toughness reveals a decrease with increasing temperature. The decrease is most pronounced in the as-received state. The isothermes of the impact energies as a function of cold work show that between the as-received state and 5% cold work a large decrease in the energy can be observed. This is very strongly pronounced at room temperature and at 673 K (400deg C). In this region the usual degree of cold work can be found manufacturing components. From this point of view of weakness of the material with respect to its mechanical properties can be expected. (orig.)

  1. Microstructure and intergranular corrosion of Inconel-600 and Incoloy-800 tubes

    Energy Technology Data Exchange (ETDEWEB)

    Stickler, R.; Weidlich, G.

    1983-04-01

    The precipitation behavior and the resistance against intergranular corrosion was determined for specimens of tubes produced of alloys IN-600 and Incoloy 800. Specimens were sensitized by heat treatments in the time-temperature range between 0.1-100 h and 400-800 C. Light microscopic, scanning electron and transmission microscope methods in addition to X-ray diffraction analyses of bulk residues were applied to characterize microstructure and precipitates. A modified standard procedure was used to determine the susceptibility to intergranular corrosion. The occurrance of grain-boundary attack was evaluated by scanning electron microscope observations. The results indicate that in the time-temperature diagram the susceptibility to intergranular corrosion falls within a typical C-curve, the lower branch of which coincides with the formation of a thin M/sub 23/C/sub 6/ grain boundary film. It is interesting to note that occasionally grain boundary separation could be found after the bend test of tube specimen in the as-received and as-aged condition without any corrosion exposure. These observations may indicate difficulties in the evaluation of the intergranular corrosion attack.

  2. Electrochemical studies on stress corrosion cracking of incoloy-800 in caustic solution. Part II: Precracking samples

    Directory of Open Access Journals (Sweden)

    Dinu Alice

    2006-01-01

    Full Text Available Stress corrosion cracking (SCC in a caustic medium may affect the secondary circuit tubing of a CANDU NPP cooled with river water, due to an accidental formation of a concentrated alkaline environment in the areas with restricted circulation, as a result of a leakage of cooling water from the condenser. To evaluate the susceptibility of Incoloy-800 (used to manufacture steam generator tubes for CANDU NPP to SCC, some accelerated corrosion tests were conducted in an alkaline solution (10% NaOH, pH = 13. These experiments were performed at ambient temperature and 85 °C. We used the potentiodynamic method and the potentiostatic method, simultaneously monitoring the variation of the open circuit potential during a time period (E corr/time curve. The C-ring method was used to stress the samples. In order to create stress concentrations, mechanical precracks with a depth of 100 or 250 μm were made on the outer side of the C-rings. Experimental results showed that the stressed samples were more susceptible to SCC than the unstressed samples whereas the increase in temperature and crack depth lead to an increase in SCC susceptibility. Incipient micro cracks of a depth of 30 μm were detected in the area of the highest peak of the mechanical precrack.

  3. The influence of Incoloy-800 alloy microstructure upon SCC behaviour in the medium of primary circuit of the steam generator

    International Nuclear Information System (INIS)

    Stress cracking corrosion (SCC) in steam generator tubing is one of the major degrading processes appearing from simultaneous mechanical stress and environmental chemical aggression. It can affect the Incoloy-800 tubing in both primary and secondary circuits. The objective of this work was the analysis of behaviour to SCC of the micro structurally modified Incoloy-800 alloy (the material of the steam generator tubing) in hydrogen environment. The microstructural modification of the alloy was achieved by heat treatment in the temperature range 400 deg.C - 800 deg.C, specific to the grain limit of carbide precipitation. Subsequently to heat treatment, occurrence of precipitated carbides is accompanied by occurrence of Cr depleted areas which results in alloy sensitizing. These areas were evidenced by means of potential-dynamical reactivation method. Heat treatment of the samples tested for corrosion is described as well as the intrinsic effect of hydrogen upon Incoloy-800 samples. Incipient intergranular and even intragranular cracks were observed in the stressed and heat treated samples. Occurrence of cracks after 72 h is explained as due to the cathodic polarisation which concentrates the reaction of hydrogen release on alloy in areas with film damages and maximal stresses. The content of absorbed H2 was calculated by means of the electric charge determined by coulometric method in which the samples electrolytically hydrogenated were anodically polarized. The degree of hydrogen embrittlement of samples was determined by comparing the content of hydrogen absorbed by hydrogenated thermo-mechanically treated samples with that of the hydrogen absorbed by heat treated unstressed hydrogenated samples and with the content of hydrogen absorbed in the samples for delivery. In conclusion, the microstructurally modified alloy by heat treatment at 700 deg.C for 1 hour (sensitized) is more susceptible to SCC in primary circuit in the presence of hydrogen while material

  4. The oxidation behavior of three different zones of welded Incoloy 800H alloy

    International Nuclear Information System (INIS)

    Highlights: • The oxidation kinetics of 800H followed the parabolic-rate law in dry-air. • The scales formed on the alloys were composed of Cr2O3 and MCr2O4 (M = Fe, Cr). • Internal-oxidation of Al2O3 and SiO2 dissolved Ti were observed in 800H-SUB and 800H-HAZ • The weight loss behavior of 800H-SUB and 800H-HAZ were observed in wet air. • The mass-loss behavior of 800H-HAZ is more severe than 800H-SUB in wet air. - Abstract: The oxidation behavior of three different zones of welded Incoloy 800H alloys, containing the substrate (800H-SUB), heat-affected zone (800H-HAZ) and the melt zone (800H-MZ) was studied at 950 °C in dry and wet air. The steady-state oxidation rate constants (kp values) were calculated based on the mass-gain data, and the oxidation resistant ability of the alloys followed by the rank of 800H-MZ > 800H-SUB > 800H-HAZ in dry air. The scales formed on the 800H-SUB and 800H-HAZ consisted of a heterophasic mixture of Cr2O3 and FeCr2O4, while a mixture of Cr2O3 and MnCr2O4 was observed on the 800H-MZ. On the other hand, the oxidation kinetics of the alloy, initially followed the parabolic-rate law up to 48 h, while a significant mass-lost kinetics was observed for a prolong exposure in wet air. The detail oxidation mechanisms for the alloys in both environments were investigated

  5. INCOLOY 908, a low coefficient of expansion alloy for high-Strength cryogenic applications: Part I. Physical metallurgy

    Science.gov (United States)

    Morra, M. M.; Ballinger, R. G.; Hwang, I. S.

    1992-12-01

    INCOLOY 908 is a low coefficient of thermal expansion (COE) iron-nickel base superalloy that was developed jointly by The Massachusetts Institute of Technology and the International Nickel Company for cryogenic service. The alloy is stable against phase transformation during prolonged thermal treatments and has a COE compatible with that of Nb3Sn. These properties make the material ideal for use as a structural component in superconducting magnets using Nb3Sn. The evolution of microstructure has been studied as a function of time at temperature over the temperature range of 650 °C to 900 °C for times between 50 and 200 hours. A detailed analysis of precipitated phases has been conducted using X-ray diffraction (XRD), transmission electron microscopy (TEM), and analytical scanning and scanning transmission electron mi- croscopy (STEM) techniques. The primary strengthening phase has been found to be γ ', Ni3(Al, Ti). INCOLOY 908 is stable against overaging, which is defined as the transformation of γ' to η, Ni3Ti, for times to 100 hours at temperatures up to 750 °C. Upon overaging, the strengthening phase transforms to η. A new phase, H x , has been identified and characterized.

  6. Comparison of long and short sample critical currents in incoloy 903 Nb3Sn cable-in-conduit conductors

    International Nuclear Information System (INIS)

    Measurements of critical currents on 330 cm long, 27 strand Nb3Sn ICCS conductors indicate that compaction to 32% helium fraction in Incoloy 903 conduits does not degrade critical current. Short and long sample critical currents at 4.2 K and 1.5 μV/cm for fields from 8 to 12 T were taken to determine how well short samples predict long sample performance. Samples were made from two separate untested 27 strand cables of Oxford Airco bronze matrix Nb3Sn wire encapsulated in Incoloy 903 conduits; the cables were assumed to be identical. Voltage taps were spaced at approximately 2 cm for short samples and 330 cm for long samples. It was found that all samples were at the same state of strain. In two out of three cases, the critical currents of short samples matched those of long samples within 10%. In one case, mechanical damage to a long sample reduced its critical current relative to short samples from the same cable. Samples from the two otherwise identical cables did not yield identical critical currents, a result found to be due to different degrees of cable prereaction. Short samples are concluded to be good predictors of long sample performance as long as both have the same state of strain, the same state of prereaction, and are free of significant mechanical damage

  7. Characterization of oxide films on Incoloy-800, AISI-304 and Monel-400 in ethanolamine by impedance and ESCA techniques

    International Nuclear Information System (INIS)

    The All Volatile Treatment (AVT), using amines, is practiced in the Steam Generator (SG) circuits of Indian Pressurized Heavy Water Reactors (PHWRs) to maintain the pH in the alkaline condition and provide corrosion protection to the structural materials. The extent of corrosion protection is known to depend on the nature of the material, the chemical composition and stability of the surface oxide films formed in the high temperature alkaline conditions. In this context, it is also important to know the type and defect density (n or p type) of the semi-conducting oxide films which will throw light not only on the film's protective nature but also on the composition of the chemical cleaning formulation required for dissolving the steam generator deposits. Since, ethanolamine (ETA) is being used currently as a pH additive in place of morpholine/ammonia due to the beneficial effects of the former, in most of the reactors, a study was undertaken to characterise the films formed on Incoloy-800, Monel-400 and AISI-304 materials with ETA under secondary water chemistry conditions. The coupons of these materials were exposed to 6 ppm of ethanolamine (pH: ∼9.5) in high temperature autoclave at a temperature of 240 deg C for 10 days and the films were characterized by Electrochemical Impedance Spectroscopy, Mott-Schottky and electron spectroscopy for chemical analysis (ESCA) techniques. The impedance analysis showed that the corrosion resistance followed the order: AISI-304 > Incoloy-800 > Monel-400 and the data mostly fitted well to a porous model equivalent circuit. Defect density calculations, made using the capacitance data, showed that Incoloy and Stainless steel had minimum n-type defect densities whereas Monel, a maximum number. The chemical analyses of the surface films on the alloys by ESCA showed that the films were composed of Ni as the dominating element along with Cr, Fe and O. Ni was seen to be around 22-25 at% in the oxides on the alloys. The chemical

  8. Numerical simulation and experimental investigation of temperature distribution in the circumferentially butt GTAW of Incoloy 800H pipes

    International Nuclear Information System (INIS)

    The multi-pass circumferential butt GTAW process of Incoloy 800H pipes was modelled with the FEM in 3D. The element birth and death technique was used for the addition of filler material. Goldak model was used to simulate the distribution of arc heat source. The validation of the simulation model was carried out based on the precise temperature measurements within the HAZ of the welds by thermocouples as well as metallographic characterisation of the cross section of the welds. A good agreement was found between the simulation and experimental results for both thermal field and weld zone shape. The present model showed that increasing the heat input resulted in a wider weld zone as well as a higher HAZ peak temperature. These effects were related to the net heat input and not to either welding current or welding speed, individually. The developed simulation model is a useful tool to investigate the welding thermal regime and the weld pool profile.

  9. Corrosion of superalloy Incoloy 800H/HT as potential component of the waste of salt nuclear reactors

    International Nuclear Information System (INIS)

    The work monitored weight loss of Incoloy 800H/HT (Fe 48%, Ni 29%, Cr: 20%, minority: Al, Ti, Si, Mn, C, Cu) and analyzed its surface by SEM-EDX analysis. Heating was carried out for 8 hours at temperatures of 600 grad C and 900 grad C in the vertical tube furnace gradually with nitrogen (99.99%), argon (99.996%) and air atmosphere. As corrosive medium was used salt with conventional name FLINAK (LiF 46.5% - 11.5% NaF - KF 42%), melting point 454 grad C and a mixture of lithium and potassium chloride in a weight ratio of 1: 1. If no FLINAK was taken, furnace atmosphere was used as corrosive media. Then the samples were cleaned in ultrasound and were weighted again on an analytical balance. (author)

  10. Surface analytical and electrochemical characterization of oxide film layers formed on Incoloy 800 and carbon steel in simulated secondary water chemistry conditions of PHWRs

    Energy Technology Data Exchange (ETDEWEB)

    Rangarajan, Srinivasan; Chandran, Sinu; Balaji, Vadivelu; Narasimhan, Sevilmedu V. [BARC Facilities, Kalpakkam, Tamil Nadu (India). Water and Steam Chemistry Div.

    2011-06-15

    The water chemistry in the steam generator (SG) circuits of Indian pressurized heavy water reactors (PHWRs) is controlled by the all-volatile treatment (AVT) procedure, wherein volatile amines are used to maintain the alkaline pH required for minimizing the corrosion of the structural materials. Earlier, Monel and morpholine were used as the steam generator material and the alkalizing agent respectively. However, currently they have been replaced by Incoloy 800 and ethanolamine (ETA). ETA was chosen because of its beneficial effects due to low pKb and Kd values, loading behavior on the condensate polishing unit (CPU), and also based on cost comparison with other amines. Since we have Incoloy 800 on the tube side and carbon steel (CS) on the shell side in the SG circuits, efforts were taken to study the nature of the oxide films formed on these surfaces and to evaluate the corrosion resistance and electrochemical properties of the same under simulated secondary water chemistry conditions of PHWRs containing different dissolved oxygen (DO) concentrations. In this context, experiments were carried out by exposing finely polished CS and Incoloy 800 coupons to ETA-based medium in the presence and absence of hydrazine (pH: 9.2) at 240 C under two different DO conditions (< 10 {mu}g . L{sup -1} and 300 {mu}g . L{sup -1}) for 24 hours. Oxide films formed under these conditions were characterized using scanning electron microscopy, Raman spectroscopy, electrochemical impedance, polarization and Mott-Schottky techniques. Further, studies at a controlled DO level (< 10 {mu}g . L{sup -1}) were carried out for different time durations, viz., 7 and 30 days. The composition, surface morphology, oxide thickness, resistance, type of semiconductivity and defect density of the oxide films were evaluated and correlated with the DO levels and are discussed elaborately in this paper. (orig.)

  11. Surface analytical and electrochemical characterization of oxide films formed on Incoloy-800 and carbon steel in simulated secondary water chemistry conditions of PHWRs

    International Nuclear Information System (INIS)

    The water chemistry in the Steam Generator (SG) Circuits of Indian Pressurized Heavy Water Reactors (PHWRs) is controlled by the all volatile treatment (AVT) procedure, wherein volatile amines are used to maintain the alkaline pH required for minimizing the corrosion of the structural materials. Earlier, Monel and morpholine were used as the Steam Generator material and the alkalizing agent respectively. However, currently they are replaced by Incoloy-800 and Ethanolamine (ETA). ETA was chosen because of its beneficial effects due to low pKb and Kd values, loading behaviour on condensate polishing unit (CPU) and also on cost comparison with other amines. Since we have Incoloy-800 on the tube side and Carbon steel(CS) on the shell side in the SG circuits, efforts were taken to study the nature of the oxide films formed on these surfaces and to evaluate the corrosion resistance and electrochemical properties of the same, under simulated secondary water chemistry conditions of PHWRs containing different dissolved oxygen (DO) concentration. In this context, experiments were carried out by exposing finely polished CS and Incoloy -800 coupons to ETA based medium in the presence and absence of Hydrazine (pH: 9.2) at 240 oC under two different DO conditions (< 10 ppb and 200 ppb) for 24 hours. Oxide films formed under these conditions were characterized using SEM, Raman spectroscopy, electrochemical impedance, polarization and Mott-Schottky techniques. Further, studies at a controlled DO level ( < 10 ppb) were carried out for different time durations viz., 7- and 30- days. The composition, surface morphology, oxide thickness, resistance, type of semi-conductivity and defect density of the oxide films were evaluated and correlated with the DO levels and discussed elaborately in this paper. (author)

  12. Surface analytical and electrochemical characterization of oxide film layers formed on Incoloy 800 and carbon steel in simulated secondary water chemistry conditions of PHWRs

    International Nuclear Information System (INIS)

    The water chemistry in the steam generator (SG) circuits of Indian pressurized heavy water reactors (PHWRs) is controlled by the all-volatile treatment (AVT) procedure, wherein volatile amines are used to maintain the alkaline pH required for minimizing the corrosion of the structural materials. Earlier, Monel and morpholine were used as the steam generator material and the alkalizing agent respectively. However, currently they have been replaced by Incoloy 800 and ethanolamine (ETA). ETA was chosen because of its beneficial effects due to low pKb and Kd values, loading behavior on the condensate polishing unit (CPU), and also based on cost comparison with other amines. Since we have Incoloy 800 on the tube side and carbon steel (CS) on the shell side in the SG circuits, efforts were taken to study the nature of the oxide films formed on these surfaces and to evaluate the corrosion resistance and electrochemical properties of the same under simulated secondary water chemistry conditions of PHWRs containing different dissolved oxygen (DO) concentrations. In this context, experiments were carried out by exposing finely polished CS and Incoloy 800 coupons to ETA-based medium in the presence and absence of hydrazine (pH: 9.2) at 240 C under two different DO conditions (-1 and 300 μg . L-1) for 24 hours. Oxide films formed under these conditions were characterized using scanning electron microscopy, Raman spectroscopy, electrochemical impedance, polarization and Mott-Schottky techniques. Further, studies at a controlled DO level (-1) were carried out for different time durations, viz., 7 and 30 days. The composition, surface morphology, oxide thickness, resistance, type of semiconductivity and defect density of the oxide films were evaluated and correlated with the DO levels and are discussed elaborately in this paper. (orig.)

  13. Residual stresses determination in an 8 mm Incoloy 800H weld via neutron diffraction

    International Nuclear Information System (INIS)

    Highlights: • Stress through thickness at 5 mm from weld centerline indicates a “U” distribution. • Declining of tensile stress through thickness occurred at weld centerline. • Residual stress between layers is the lowest. - Abstract: To investigate the distribution of residual stresses, the 8 mm 800H alloy was joined by multi-layer butt TIG process. Residual stresses in the longitudinal, transverse and normal directions were measured via neutron diffraction. These residual stress measurements were taken at a series of points 2 mm below the top surface, covering the fusion zone, heat affected zone (HAZ) and base metal. In addition, two lines of longitudinal residual stress values at the weld centerline and 5 mm from weld centerline through thickness were measured. Results show that both the longitudinal and transverse stresses from the weld centerline to base metal are mainly tensile stresses. The longitudinal residual stress is the largest, with a maximum value of 330 MPa. As for the normal residual stress, the weld zone shows tensile stress, while the HAZ shows compressive stress. The middle of the thickness shows compressive residual stress along the thickness direction. The longitudinal stress at weld centerline through thickness reveals the interlayer heat treat effects leads to a declining of tensile stress. While the stress at 5 mm from weld centerline indicates a “U” distribution due to the mixed microstructure close to fusion line. With the increasing distance from weld seam, the residual stress decreases gradually

  14. Incoloy alloy MA956. Strain rate and temperature effects on the microstructure and ductility

    International Nuclear Information System (INIS)

    MA956 is an iron-base, oxide dispersion strengthened alloy produced by mechanical alloying with the nominal composition Fe-Cr 20-Al 5-Ti 0.5-Y2O3 0.5, which is utilised in applications involving rigorous service conditions. It is ferritic and therefore undergoes a ductile-brittle transition which tends to occur between 40 and 70 C. For this reason, working at elevated temperatures is required. However, the ductility is not a simple function of temperature, strain rate, and grain size. Tensile tests have been carried out at temperatures up to 1000 C, at strain rates of 10-2 to 10-4s-1, and the behaviour of the coarse and fine grained materials is markedly different. Both materials show an increase in elongation around 600 C, but it decreases again with increasing temperature. The elongation continues to decrease in the coarse grained material. However, the fine grained material exhibits an increase in elongation with increasing strain rate at the higher temperatures which peaks around 800 C. The microstructures and fracture surfaces of the materials which have undergone deformation have been studied and provide a basis for understanding the complex mechanical behaviour. (orig.)

  15. Reduction of tritium permeation through Inconel 718 and Incoloy 800 HT by means of natural oxides

    Science.gov (United States)

    Aiello, A.; Utili, M.; Ciampichetti, A.

    2011-10-01

    Chronical releases of tritium from the helium primary coolant into the water secondary coolant is a fundamental safety issue in the design of a fusion reactor steam generator. It is well known that the steam/water circuit of a fusion reactor would be considered not relevant from a radiological point of view, while if a strong permeation of tritium will be present it will be released together with incondensable gases in the condenser. The permeation of hydrogen isotopes through candidate steam generator materials in different conditions was studied in the past. Further experiments demonstrated that nickel alloys of nuclear interest are always covered by a thin and adherent oxide layer able to reduce permeation of orders of magnitude. The major objective of this work is the evaluation of the permeated flux through nickel alloys, when exposed to pure hydrogen and to an oxidant gas stream, to verify the real permeability of these materials in conditions close to those foreseen in the helium side of the steam generator.

  16. On the oxide formation on stainless steels AISI 304 and incoloy 800H investigated with XPS

    NARCIS (Netherlands)

    Langevoort, J.C.; Sutherland, I.; Hanekamp, L.J.; Gellings, P.J.

    1987-01-01

    The influence of cold work on the initially formed oxide layer on the stainless steels AISI 304 and Incology 800H has been studied by XPS. Oxidations were performed at pressures of 10-6-10-4 Pa and temperatures of 300–800 K. All samples showed a similar oxidation behaviour. The oxidation rates of ir

  17. On the kinetics of oxidation of austenitic stainless steels AISI 304 and incoloy 800H

    NARCIS (Netherlands)

    Langevoort, J.C.; Hanekamp, L.J.; Gellings, P.J.

    1987-01-01

    The interaction of oxygen with clean surfaces of stainless steels has been studied by spectroscopic ellipsometry and AES. The reaction involves chemisorption and dissolution of oxygen into the surface of the metal via a place-exchange mechanism. Oxide thickening occurs via cation and anion migration

  18. Incoloy 800 steam generator tubes stubbing by laser and electron beams process

    International Nuclear Information System (INIS)

    The electron beam welding conditions are optimized for different thermal cycles and chemical compositions of the fusion zone. The metallurgical and mechanical properties of the joints are described and compared with the properties of laser and TIG welds

  19. Corrosion aspects of compatible alloys in molten salt (FLiNaK) medium for Indian MSR program in the temperature range of 550-750 °C using electrochemical techniques

    International Nuclear Information System (INIS)

    Corrosion behaviours of different alloys were evaluated in fluoride eutectic FLiNaK in the temperature range of 550-750 °C under static and dynamic conditions. Electrochemical polarization and impedance techniques were used to estimate corrosion rate. The results showed that the corrosion process was controlled by activation and in some cases by formation of passive layer. In static mode, the corrosion rates followed the order : Inconel 625 > Inconel 617 > Inconel 600 > Incoloy 800 > Ni 220 > Hastelloy N > Incoloy 800HT. In dynamic mode, Hastelloy N and Incoloy 800HT showed better corrosion resistance in comparison to other alloys. (author)

  20. Gas-turbine HTGR materials screening test program. Quarterly progress report, July 1, 1976--September 30, 1976. [IN 100; IN 713; MM004; M21; IN 738; RENE 100; MoTZM; Hastelloy X; Inconel 617; MA 753; IN 519, Inconel 706; Inconel 718; A286; 316 SS; Incoloy 800

    Energy Technology Data Exchange (ETDEWEB)

    Rosenwasser, S.N.; Johnson, W.R.

    1976-09-30

    The duration of controlled-impurity creep-screening tests and unstressed aging tests has reached 10,000 hr. Creep and weight change data from testing up to 9,000 hr and results from post-test metallurgical evaluations of several recently returned 3,000-hr specimens, including alloys IN519 and MoTZM, are presented. Preliminary materials requirements for key GT-HTGR 850/sup 0/C (1562/sup 0/F) reactor outlet temperature reference design components are documented.

  1. Long range plan for flexible joint development program

    International Nuclear Information System (INIS)

    Objective is to develop bellows expansion joints into CDS and subsequent LMFBR and liquid metal applications. An assessment was performed on the use of Incoloy 800H and Inconel 718 as bellows materials

  2. The behaviour of amorphous silica coatings at high temperatures in aggressive environments

    OpenAIRE

    Ayres, C. F.; Bennett, M. J.; Gohil, D.D.; Léon, B.; Pérez-Amor, M.; Pou, J.; Saunders, S

    1993-01-01

    Amorphous silica coatings produced by plasma assisted and laser chemical vapour deposition (PACVD and LCVD) on Incoloy 800H and 21/4 Cr 1 Mo ferritic steel were exposed in air and in simulated coal gasification atmospheres (CGA) for periods of up to two years at temperatures between 450°C and 900°C . In some cases interlayers of TiN were used to promote adhesion and to reduce interdiffusion between the coating and substrate. PACVD silica coatings deposited onto Incoloy 800H provided outstandi...

  3. The protective properties of thin alumina films deposited by metal organic chemical vapour deposition against high-temperature corrosion of stainless steels

    NARCIS (Netherlands)

    Morssinkhof, R.W.J.; Fransen, T.; Heusinkveld, M.M.D.; Gellings, P.J.

    1989-01-01

    Coatings of Al2O3 were deposited on Incoloy 800H and AISI 304 by means of metal organic chemical vapour deposition. Diffusion limitation was the rate-determining step above 420 °C. Below this temperature, the activation energy of the reaction appeared to be 30 kJ mol−1. Coating with Al2O3 increases

  4. On the influence of cold work on the oxidation behavior of some austenitic stainless steels: High temperature oxidation

    NARCIS (Netherlands)

    Langevoort, J.C.; Fransen, T.; Gellings, P.J.

    1984-01-01

    AISI 304, 314, 321, and Incoloy 800H have been subjected to several pretreatments: polishing, milling, grinding, and cold drawing. In the temperature range 800–1400 K, cold work improves the oxidation resistance of AISI 304 and 321 slightly, but has a relatively small negative effect on the oxidatio

  5. Residual stress measurement of the jacket material for ITER coil by neutron diffraction

    Energy Technology Data Exchange (ETDEWEB)

    Tsuchiya, Yoshinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Nickel-Iron based super alloy INCOLOY 908 is used for the jacket of a central solenoid coil (CS coil) of the International Thermonuclear Experimental Reactor (ITER). INCOLOY 908, however, has a possibility of fracture due to Stress Accelerated Grain Boundary Oxidation (SAGBO) under a tensile residual stress beyond 200MPa. Therefore it is necessary to measure the residual stress of the jacket to avoid SAGBO. We performed residual stress measurement of the jacket by neutron diffraction using the neutron diffractometer for residual stress analysis (RESA) installed at JRR-3M in JAERI. A sample depth dependence of internal strain was obtained from the (111) plane spacing. A residual stress distribution was calculated from the strain using Young`s modulus and Poisson`s ratio that were evaluated by a tensile test with neutron diffraction. The result shows that the tensile residual stress exceeds 200MPa of the SAGBO condition in some regions inside the jacket. (author)

  6. Nitriding behavior of nickel-based superalloys at 1273 K and above

    International Nuclear Information System (INIS)

    The corrosion behavior of four nickel-based superalloys, Inconel 617, Incoloy 800h, Haynes 230, and Hastelloy X, was studied in various nitrogen-containing atmospheres in the temperature range between 1273 and 1523 K. Inner and outer oxidation, inner nitriding, and decarburization were observed. Aluminum, titanium, and chromium nitrides were found after annealing in pure nitrogen and in N2/Ar/H2 atmosphere. Chromium and tungsten carbides disappeared. Instead of these carbides, a chromium and tungsten-rich phase was formed at the grain boundaries. Under these conditions, Incoloy 800h shows the highest and Haynes 230 the lowest corrosion susceptibility. The type of corrosion products and the kinetics of their formation will be described. The suitability of these alloys for application in the intermediate heat exchanger and in the nitrogen atmospheres of the second cooling circuit will be discussed

  7. Modell experiments to determine the effect of inhibitive oxide layers on metals against hydrogen permeation

    International Nuclear Information System (INIS)

    The coupling of H2-permeation and corrosion has been examined with the high-temperature alloys Incoloy 800 and Incoloy 802. Permeationsrates as well as corrosionsrates have been measured simultanously under H2O-H2 atmospheres in the test-facility HD-PERM. Test parameters have been temperature and oxidationpotential. Parabolic laws for the growth of the oxide scales have been identified and are considered to be highly important for the efficiency of a permeation barrier. A comparison between the temperature dependencies of corrosionsrates and H2-permeationsrates has revealed that permeation and corrosion are coupled only in so far that the permeation barrier is formed by the corrosion reaction. The corrosion data (parabolic rate constant, activation energy) of the oxide scales have given clear indications for the existence of a Cr2O3-layer, which is considered to be responsible for efficient oxide permeation barriers. (orig.)

  8. Identification of Iron Oxides Qualitatively/Quantitatively Formed during the High Temperature Oxidation of Superalloys in Air and Steam Environments

    Institute of Scientific and Technical Information of China (English)

    M.Siddique; N.Hussain; M.Shafi

    2009-01-01

    Mossbauer spectroscopy has been used to study the morphology of iron oxides formed during the oxidation of superalloys, such as SS-304L (1.4306S), Incoloy-800H, Incoloy-825, UBHA-25L, Sanicro-28 and Inconel-690, at 1200℃ exposed in air and steam environments for 400 h. The basic aim was to identify and compare the iron oxides qualitatively and quantitatively, formed during the oxidation of these alloys in two environments. The behaviour of alloy UBHA-25L in high temperature oxidation in both environments indicates that it has good oxidation resistance especially in steam, whereas Sanicro-28 has excellent corrosion resistance in steam environment. In air oxidation of lnconel-690 no iron oxide, with established Mossbauer parameters, was detected.

  9. Bimetallic sleeve for repairing a heat exchanger tube locally damaged and repairing process of a such tube with this sleeve

    International Nuclear Information System (INIS)

    A tubular sleeve comprising two end sections made of the same metal as the heat exchanger tube and connected by a ring of an alloy suitable for welding is inserted into the tube and welded in place from the inside. The tube and end sections may be of Incoloy 800 while the ring is in Ni-Cr Alloy, Inconel 600 or Inconel 690. 3 figs

  10. Exposure of boiler materials at the PRENFLO coal gasification plant: Second series, cooled samples

    Energy Technology Data Exchange (ETDEWEB)

    Leferink, R.G.I.; Kip, J.B.M.; Bouma, J.S. [KEMA Inspecties en Materialen, Arnhem (Netherlands)

    1994-12-31

    A number of alloys, coatings and welds are exposed in the PRENFLO coal gasification test facility to determine their applicability in heat exchangers of coal gasification installations. The corrosion of the materials is analyzed by means of optical microscopy and electron microscopy. The tested materials, arranged according to increasing suitability, are 15Mo3, T91, chromated 10CrMo9.10, chromated/vanadated T22Nb, Incoloy 800LC, SS310 and Sanicro 28

  11. Final Technical Report - High-Performance, Oxide-Dispersion-Strengthened Tubes for Production of Ethylene adn Other Industrial Chemicals

    Energy Technology Data Exchange (ETDEWEB)

    McKimpson, Marvin G.

    2006-04-06

    This project was undertaken by Michigan Technological University and Special Metals Corporation to develop creep-resistant, coking-resistant oxide-dispersion-strengthened (ODS) tubes for use in industrial-scale ethylene pyrolysis and steam methane reforming operations. Ethylene pyrolysis tubes are exposed to some of the most severe service conditions for metallic materials found anywhere in the chemical process industries, including elevated temperatures, oxidizing atmospheres and high carbon potentials. During service, hard deposits of carbon (coke) build up on the inner wall of the tube, reducing heat transfer and restricting the flow of the hydrocarbon feedstocks. About every 20 to 60 days, the reactor must be taken off-line and decoked by burning out the accumulated carbon. This decoking costs on the order of $9 million per year per ethylene plant, accelerates tube degradation, and requires that tubes be replaced about every 5 years. The technology developed under this program seeks to reduce the energy and economic cost of coking by creating novel bimetallic tubes offering a combination of improved coking resistance, creep resistance and fabricability not available in current single-alloy tubes. The inner core of this tube consists of Incoloy(R) MA956, a commercial ferritic Fe-Cr-Al alloy offering a 50% reduction in coke buildup combined with improved carburization resistance. The outer sheath consists of a new material - oxide dispersion strengthened (ODS) Alloy 803(R) developed under the program. This new alloy retains the good fireside environmental resistance of Alloy 803, a commercial wrought alloy currently used for ethylene production, and provides an austenitic casing to alleviate the inherently-limited fabricability of the ferritic Incoloy(R) MA956 core. To provide mechanical compatibility between the two alloys and maximize creep resistance of the bimetallic tube, both the inner Incoloy(R) MA956 and the outer ODS Alloy 803 are oxide dispersion

  12. Fabrication and loading of long-term stress corrosion cracking surveillance specimens for the Dresden 1 decontamination program

    Energy Technology Data Exchange (ETDEWEB)

    Walker, W.L.

    1979-10-01

    Stress-corrosion cracking test specimens were prepared for Dow Nuclear Services for insertion in the Dresden 1 reactor during the chemical decontamination of the primary system, and for subsequent exposure under operating conditions when the station returns to service. The specimens consist of pressurized tubes fabricated from Type-304 and -304L stainless steel, Inconel 600, Incoloy 800, and Zircaloy 2. In addition, constant radius bent-beam specimens of 3/4 hard Type-410 stainless steel were also included. All specimens were stressed to, or slightly above, their respective 0.2% offset yield strengths at the temperatures of interest.

  13. Fabrication and loading of long-term stress corrosion cracking surveillance specimens for the Dresden 1 decontamination program

    International Nuclear Information System (INIS)

    Stress-corrosion cracking test specimens were prepared for Dow Nuclear Services for insertion in the Dresden 1 reactor during the chemical decontamination of the primary system, and for subsequent exposure under operating conditions when the station returns to service. The specimens consist of pressurized tubes fabricated from Type-304 and -304L stainless steel, Inconel 600, Incoloy 800, and Zircaloy 2. In addition, constant radius bent-beam specimens of 3/4 hard Type-410 stainless steel were also included. All specimens were stressed to, or slightly above, their respective 0.2% offset yield strengths at the temperatures of interest

  14. The influence of cold work on the oxidation behaviour of stainless steel

    International Nuclear Information System (INIS)

    In this thesis the study of the interaction of oxygen gas with stainless steel surfaces is described. Thermogravimetry, microscopy and ellipsometry have been used to follow the oxidation in situ, while EDX, AES and XPS have been used to determine the oxide compositions. The aim of this thesis is to reveal the influence on the oxidation behaviour of stainless steel of i) cold work (rolling, drawing, milling, polishing and Ar ion bombardment) ii) the initially formed oxide and iii) the experimental conditions. Two types of stainless steels have been used (AISI 304 (a 18/8 Cr/Ni steel) and Incoloy 800 H (a 20/30 Cr/Ni steel)). (Auth.)

  15. Multifrequency Eddy current testing of heat exchange tubes with a rotating probe

    International Nuclear Information System (INIS)

    Multi-frequency eddy current analyses have been used in France industrially since 1975. In light of the experienced gained during many steam generator inspections, this technique was applied to the examination of sheet and tube heat exchangers featuring tubes in very different materials such as copper, stainless steel and titanium. The principle of multi-frequency Eddy current inspection is first reviewed, using the example of a condenser with nickel alloy tubes (Inconel, Incoloy). This is followed by the description of a specific application of this technique to a condenser with titanium tubes, analyzed with a rotating local probe

  16. Superplastic-like deformation in some solid solution alloys. [FeNiCr; AlMg

    Energy Technology Data Exchange (ETDEWEB)

    Korhonen, M.A.; Wilson, H.; Kuo, R.C.; Li Cheyu (Dept. of Materials Science and Engineering, Cornell Univ., Ithaca, NY (USA))

    1991-05-15

    The stress relaxation and tensile test data of Incoloy 800H and Al-4.6%Mg are described at temperatures where the contribution of grain boundary sliding to flow is significant. It is shown that in the presence of grain boundary sliding the whole range of behavior, from ordinary creep to superplastic-like flow, can be exhibited in the same metals, showing that the strain rate sensitivity increases with decreasing grain size or increasing temperature in the chosen temperature range. The effect of solute hardening on the flow behaviour of solid solution alloys in the presence of grain boundary sliding is also discussed. (orig.).

  17. European cryogenic material testing program for ITER coils and intercoil structures

    Science.gov (United States)

    Nyilas, A.; Portone, A.; Kiesel, H.

    2002-05-01

    The following materials were characterized for the use in the magnet structures of ITER: 1) Type 316LN cast materials having a modified chemistry used for a Model of the TF (Toroidal Field) outer intercoil structure were investigated with respect to tensile, fracture, fatigue crack growth rate (FCGR), and fatigue life behavior between 7 and 4 K. 2) For Type 316LN 80 mm thick plate used for the TFMC (T_oroidal F_ield M_odel C_oil) structure a complete cryogenic mechanical materials characterization was established. 3) For full size coil case mockups, repair weld properties of 240 mm thick narrow-gap welds were investigated to determine their tensile and fracture behavior. 4) For CSMC (C_enter S_olenoid M_odel C_oil) superconductor jackets, the fatigue lives of orbital butt welds made of Incoloy 908 and Type 316LN (aged and unaged) materials were determined up to one million cycles at 7 K. The results reveal to date that the FCGR of aged Type 316LN is inferior to Incoloy 908 material, whilst the fatigue life properties are comparable. However, for Type 316LN jacket structure considerable improvement of FCGR could be achieved by a solution heat treatment process. In addition, tensile and fatigue life tests performed with a new cryogenic mechanical test facility (630 kN capacity) are presented.

  18. Evolution of Microstructure and Texture During Hot Compression of a Ni-Fe-Cr Superalloy

    Science.gov (United States)

    Coryell, S. P.; Findley, K. O.; Mataya, M. C.; Brown, E.

    2012-02-01

    Superalloys are being employed in more extreme conditions requiring higher strength, which requires producers to forge products to finer grain sizes with less grain size variability. To assess grain size, crystallographic texture, and substructure as a function of forging conditions, frictionless uniaxial compression testing characteristic of hot working was performed on INCOLOY 945 (Special Metals Corporation, Huntington, WV), which is a newly developed hybrid of alloys 718 and 925, over a range of temperatures and strain rates. The microstructure and texture were investigated comprehensively using light optical microscopy, electron backscatter diffraction (EBSD), electron channeling contrast imaging (ECCI), and transmission electron microscopy (TEM) to provide detailed insight into microstructure evolution mechanisms. Dynamic recrystallization, nucleated by grain/twin boundary bulging with occasional subgrain rotation, was found to be a dominant mechanism for grain refinement in INCOLOY 945. At higher strain rates, static recrystallization occurred by grain boundary migration. During deformation, duplex slip along {111} planes occurred until a stable fiber compression texture was established. Recrystallization textures were mostly random but shifted toward the compression texture with subsequent deformation. An exception occurred at 1423 K (1150 °C) and 0.001 seconds-1, the condition with the largest fraction of recrystallized grains, where a fiber texture developed, which may be indicative of preferential growth of specific grain orientations.

  19. Correlation of substructure with mechanical properties of plastically deformed reactor structural materials. Progress report, January 1, 1976--June 30, 1977

    Energy Technology Data Exchange (ETDEWEB)

    Moteff, J.

    1977-07-08

    Transmission electron microscopy used to evaluate the deformation (creep, fatigue and tensile) induced microstructure of 304 SS, Incoloy 800, 330 SS and three of the experimental alloys (E19, E23 and E36) obtained from the National Alloy Program clearly shows that the relationship between the subgrain size (lambda) and the applied stress (sigma) obeys the equation lambda = Ab (sigma/E)/sup -1/ where A is a constant of the order of 4, b the Burgers rector and E is Young's modulus. Hot-hardness studies on 304 SS, 316 SS, Incoloy 800, 2 /sup 1///sub 4/ Cr-1 Mo steels, 330 SS, Inconel 718, PE-16, Inconel 706, M-813 and the above three experimental alloys suggests that reasonable effective activation energies for creep may be obtained through the use of the hardness test as a strength microprobe tool. The ordering of the strength levels obtained through hot-hardness follows quite closely that obtained in tensile tests when those data are available.

  20. Stress corrosion cracking susceptibility of steam generator tubing on secondary side in restricted flow areas

    International Nuclear Information System (INIS)

    Nuclear steam generator tubes operate in high temperature water and on the secondary side in restricted flow areas many nonvolatile impurities accidentally introduced into circuit tend to concentrate. The concentration process leads to the formation of highly aggressive alkaline or acid solutions in crevices, and these solutions can cause stress corrosion cracking (SCC) on stressed tube materials. Even though alloy 800 has shown to be highly resistant to general corrosion in high temperature water, it has been found that the steam generator tubes may crack during service from the primary and/or secondary side. Stress corrosion cracking is still a serious problem occurring on outside tubes in operating steam generators. The purpose of this study was to evaluate the environmental factors affecting the stress corrosion cracking of steam generators tubing. The main test method was the exposure for 1000 hours into static autoclaves of plastically stressed C-rings of Incoloy 800 in caustic solutions (10% NaOH) and acidic chloride solutions because such environments may sometimes form accidentally in crevices on secondary side of tubes. Because the kinetics of corrosion of metals is indicated by anodic polarization curves, in this study, some stressed specimens were anodically polarized in caustic solutions in electrochemical cell, and other in chloride acidic solutions. The results presented as micrographs, potentiokinetic curves, and electrochemical parameters have been compared to establish the SCC behavior of Incoloy 800 in such concentrated environments. (authors)

  1. Influence of sulfur on the passivity of inconel 600 in aqueous environment at 3000C. Relationship with stress corrosion

    International Nuclear Information System (INIS)

    Dissolution kinetics and repassivation of inconel 600 in simulated primary coolant circuits of PWR is studied by fast traction experiment under potentiostatic control. The notion of elementary electrochemical transient is introduced. The model of anodic dissolution - film rupture allows the calculation of crack growth in constant deformation rate tests. When sulfur concentration is smaller than 100 micrograms/g the current is low, above the current is high. Calculation of crack growth from high level current are consistent with experimental data. Influence of pH, temperature, solution composition are determined. A Comparative study with nickel, incoloy 690 and a 19% chromium alloy was carried out to understand fast traction phenomena. Chromium plays an important part without pollution a protecting chromium oxide is formed. In polluted environment sulfur prevent nucleation of this compound and chromium hydroxides are precipitated on the surface. With pure nickel there is no passivity in presence of sulfur

  2. Improvement and Micro-mechanism of a Low Expansion Superalloy

    Institute of Scientific and Technical Information of China (English)

    1999-01-01

    Chromium free low expansion superalloys Incoloy 900 series based on the system Fe-Ni-Co-Nb-Ti are precipitation-strengthen austenitic alloys with combination of very low thermal expansion coefficient and high tensile strength. The most important problem of these alloys is the susceptibility to stress accelerated grain boundary oxygen embrittlement (SAGBO) due to the absence of chromium. The oxidation resistance of the alloys after they were exposed for long time at high temperatures and the notch bar rupture strength have been improved with appropriate rare earth (RE) addition. With trace yttrium addition, the platelet precipitates become smaller and denser and the phase of precipitate transforms from ε phase to H phase. The crystal structures of the platelet phases in the conventional and Y-containing superalloys have been investigated and determined. The interface between the H phase and the matrix has also been demonstrated.

  3. Large-scale tests of insulated conduit for the ITER CS coil

    Science.gov (United States)

    Reed, R. P.; Walsh, R. P.; Schutz, J. B.

    Compression-fatigue tests at 77 K were conducted on test modules of insulated Incoloy 908 conduit. To replicate the operating conditions for the ITER central solenoid (CS) full-scale coil, fatigue loads up to 3.6 MN were applied for 10 5 cycles; no mechanical breakdowns occurred. The conduits were insulated with a preimpregnated resin system, a tetraglycidyl diaminodiphenyl methane (TGDM) epoxy cured with DDS aromatic amine. The conduits were joined by vacuum-pressure impregnation with a diglycidyl ether of bisphenol-F epoxy/anhydride-cured resin system. In the 4×4 stacked-conduit test modules, the layer insulation (a high-pressure laminate of TGDM epoxy cured with DDS aromatic amine) was inserted. Periodically during the tests, breakdown voltage was measured across the conduits of both turn and layer insulation; throughout the test, breakdown voltages were at least 46 kV. The addition of a barrier increased structural and electrical reliability.

  4. Cryogenic evaluation of epoxy bond strength

    Science.gov (United States)

    Albritton, N.; Young, W.

    The purpose of the work presented here was to determine methods of optimizing the adhesion of a particular epoxy (CTD-101K, Composite Technology Development Inc.) to a particular nickel-based alloy substrate (Incoloy ® 908, Inco Alloys International) for cryogenic applications. Initial efforts were focused on surface preparation of the substrate material via various mechanical and chemical cleaning techniques. Test samples, fabricated to simulate the conduit-to-insulation interface, were put through a mock heat treat and vacuum/pressure impregnation process. Samples were compression/shear load tested to compare the bond strengths at room temperature and liquid nitrogen temperature. The resulting data indicate that acid etching creates a higher bond strength than the other tested techniques and that the bond formed is stronger at cryogenic temperatures than at room temperature. A description of the experiment along with the resulting data is presented here.

  5. TPX: Contractor preliminary design review. Volume 5, Manufacturing R ampersand D

    International Nuclear Information System (INIS)

    TPX Insulation ampersand Impregnation R ampersand D test results are reported for 1x2 samples designed for screening candidate conduit insulation systems for TPX PF and TF coils. The epoxy/glass insulation system and three proposed alternate insulation systems employing Kapton, was evaluated in as received sample condition and after 10 thermal cycles in liquid nitrogen. Two DGBA impregnation systems, Shell 826 and CTD101K were investigated. Square incoloy 908 and 316 LN stainless hollow conduits were used for 1x2 sample fabrication. Capacitance, dielectric loss, and insulation resistance dielectric characteristics were measured for all samples. Partial discharge performance was measured for samples either in air, under silicon oil, or under liquid nitrogen up to 10kVrms at 60 Hz. Hipot screening was performed at 10 kVdc. The samples were cross sectioned and evaluated for impregnation quality. The implications of the test results on the TPX preliminary design decision are discussed

  6. Preclosure analysis of conceptual waste package designs for a nuclear waste repository in tuff

    International Nuclear Information System (INIS)

    This report discusses the selection and analysis of conceptual waste package developed by the Nevada Nuclear Waste Storage Investigations (NNWSI) project for possible disposal of high-level nuclear waste at a candidate site at Yucca Mountain, Nevada. The design requirements that the waste package must conform to are listed, as are several desirable design considerations. Illustrations of the reference and alternative designs are shown. Four austenitic stainless steels (316L SS, 321 SS, 304L SS and Incoloy 825 high nickel alloy) have been selected for candidate canister/overpack materials, and 1020 carbon steel has been selected as the reference metal for the borehole liners. A summary of the results of technical and ecnonmic analyses supporting the selection of the conceptual waste package designs is included. Postclosure containment and release rates are not analyzed in this report

  7. Comparison of French and German NPP water chemistry programs

    International Nuclear Information System (INIS)

    PWRs in the western hemisphere obey basically the same rules concerning design, choice of material and operational mode. In spite of these basic similarities, the manufacturers of PWRs in different countries developed different solutions in respect to single components in the steam/water cycle. Looking specifically at France and Germany, the difference in the tubing material of the steam generators (Inconel 600/690 chosen by Framatome and Incoloy 800 chosen by the former Siemens KWU) led to specific differences in the respective chemistry programs and in some respect to different 'philosophies' in operating the water/steam cycle. Compared to this, basic differences in operating the reactor coolant system cannot be observed. Nevertheless specific solutions as zinc injection and the use of enriched B-10 are applied in German PWRs. The application of such measures arises from a specific dose rate situation in older PWRs (zinc injection) or from economic reasons mainly (B-10). (authors)

  8. Discussion on RVI Material for the New Generation Nuclear Reactor%新一代反应堆堆内构件材料

    Institute of Scientific and Technical Information of China (English)

    许斌; 王庆田; 李宁; 李峰

    2012-01-01

    根据新一代反应堆堆内构件的环境条件,提出堆内构件选材原则,介绍了堆内材料相关发展动态以及5类材料的应用情况,然后对材料性能进行了综合分析,结果表明,镍基合金Inconel 690、奥氏体不锈钢316LN、铁镍合金LF2和Incoloy 800有希望在新一代反应堆堆内构件中得到应用.

  9. Technical analysis of failure of catalyst support of reformer furnace tube of a hydrogen generation unit%制氢装置转化炉炉管催化剂支托失效分析

    Institute of Scientific and Technical Information of China (English)

    齐庆轩; 冯岩

    2012-01-01

    The damages of catalyst support of reformer furnace tubes of No. 1 hydrogen generation unit in SINOPEC Shijiazhuang Refining & Chemical Co. , Ltd. in two maintenances were introduced. The composition analysis of catalyst support, the study on the macroscopic photo, the analysis of damaged surface of catalyst support of Incoloy800H, the metallographic analysis, energy dispersion spectrum (EDS) analysis, and scanning electron microscope analysis of support' s section area as well as study on the anti-caburization performances of Cr25Ni20 and Incoloy800H materials have concluded the following: The damages of catalyst support was caused by surface carburization of catalyst support material under high temperature in the presence of hydrogen, which led to phase changes of material surface structure, material stratification, loosening and bulging of surface material structure, large amount of micro-crackings in grain boundary at surface area and eduction of large amount of carbides. All these will reduce the ductility and plasticity of material. Therefore, it is difficult for the catalyst support to restore its original state after thermal expansion, which explains why there are some bulges on the tube of failed support. The anti-carburization performance is greatly improved after application of Cr25Ni20 steel material.%对两次检修中所发现的中国石油化工股份有限公司石家庄炼化分公司l号制氢装置转化炉炉管催化剂支托出现损坏的情况做了介绍,并对两种批次的催化剂支托进行了成分分析,对使用Incoloy800H材料、损坏严重的催化剂支托表面宏观照片进行了分析,对损伤支托的横截面进行金相、电镜及能谱检验分析.通过对Cr25Ni20和lncoloy800H两种材料的抗渗碳能力的比较,得出了以下结论:催化剂支托损伤的原因是材料在含氢高温环境下发生了表面渗碳现象,直接导致材料表面组织相变、材料分层、表现材料组织疏松和隆起,

  10. Measurements of residual stresses and textures by neutron diffraction

    International Nuclear Information System (INIS)

    Many measurement methods of residual stress are compared and characteristic properties of neutron diffraction method are described. The penetration depth of neutron, photon radiation and Cu-Kα ray to metals are compared and the values of neutron are larger than others. Two kinds of measurement methods of residual stress by neutron diffraction, the angular scattering and the time of flight method, are explained. The results of measurement of residual stresses of carbon steel and titanium butt weld joint, Wasploy alloy, aluminum alloy and Incoloy 800 tube in stream generator of nuclear power plant are reported. Neutron diffraction profile of SiCp/Al2024-T6 was measured by TOF method. The textures of Zr-2.5% Nb and SUS316 steel were observed. (S.Y.)

  11. Influence of aqueous environment pH on the corrosion behaviour of the CANDU steam generator tubing material

    International Nuclear Information System (INIS)

    The generalized corrosion is an undesirable process because it is accompanied by deposition of the corrosion products which affect the steam generator performances. It is very important to understand the generalized corrosion mechanism in order to evaluate the amounts of corrosion products which exist in the steam generator after a determined period of operation. The purpose of the experimental research consists in the assessment of corrosion behavior of the tube material (Incoloy-800) at normal secondary circuit parameters (temperature - 260 deg. C, pressure - 5.1 MPa). The testing environment was the demineralized water without impurities, at different pH values regulated with morpholine and cycloheyilamine (all volatile treatment). The results are presented as micrographs and graphics representing loss of metal by corrosion, corrosion rate, the total corrosion products, the adherent corrosion product, the released corrosion products and the release of the metal. (authors)

  12. Preselection of Ni-Cr(-Mo) alloys as potential canister materials for vitrified high active nuclear waste by electrochemical testing

    Science.gov (United States)

    Bort, H.; Wolf, I.; Leistikow, S.

    1987-07-01

    Several Ni-Cr(-Mo) alloys (Hastelloy C4, Inconel 625, Sanicro 28, Incoloy 825, Inconel 690) were tested by electrochemical methods to characterize their corrosion behavior in chloride containing solutions at various temperatures and pH-values in respect to their application as canister materials for final radioactive waste storage. Especially, Hastelloy C4 was tested by potentiodynamic, potentiostatic and galvanostic measurements. As electrolytes H 2SO 4 solutions were used, as parameters temperature, chloride content and pH-value were varied. All tested alloys showed a clearly limited resistance against pitting corrosion phenomena; under severe conditions even crevice corrosion phenomena were observed. The best corrosion behavior, however, is shown by Hastelloy C4, which has the lowest passivation current density of all tested alloys and the largest potential region with protection against local corrosion phenomena.

  13. Preselection of Ni-Cr(-Mo) alloys as potential canister materials for vitrified high active nuclear waste by electrochemical testing

    Energy Technology Data Exchange (ETDEWEB)

    Bort, H.; Wolf, I.; Leistikow, S.

    1987-07-01

    Several Ni-Cr(-Mo) alloys (Hastelloy C4, Inconel 625, Sanicro 28, Incoloy 825, Inconel 690) were tested by electrochemical methods to characterize their corrosion behavior in chloride containing solutions at various temperatures and pH-values in respect to their application as canister materials for final radioactive waste storage. Especially, Hastelloy C4 was tested by potentiodynamic, potentiostatic and galvanostic measurements. As electrolytes H/sub 2/SO/sub 4/ solutions were used, as parameters temperature, chloride content and pH-value were varied. All tested alloys showed a clearly limited resistance against pitting corrosion phenomena; under severe conditions even crevice corrosion phenomena were observed. The best corrosion behavior, however, is shown by Hastelloy C4, which has the lowest passivation current density of all tested alloys and the largest potential region with protection against local corrosion phenomena.

  14. Temperature measurement: Development work on noise thermometry and improvement of conventional thermocouples for applications in nuclear process heat (PNP)

    International Nuclear Information System (INIS)

    The behaviour was studied of NiCr-Ni sheathed thermocouples (sheath Inconel 600 or Incoloy 800, insulation MgO) in a helium and carbon atmosphere at temperatures of 950-1150 deg. C. All the thermocouples used retained their functional performance. The insulation resistance tended towards a limit value which is dependent on the temperature and quality of the thermocouple. Temperature measurements were loaded with great uncertainty in the temperature range of 950-1150 deg. C. Recalibrations at the temperature of 950 deg. C showed errors of up to 6%. Measuring sensors were developed which consist of a sheathed double thermocouple with a noise resistor positioned between the two hot junctions. Using the noise thermometer it is possible to recalibrate the thermocouple at any time in situ. A helium system with a high temperature experimental area was developed to test the thermocouples and the combined thermocouple-noise thermometer sensors under true experimental conditions

  15. Power plant VII - Air-air /tube boiler/

    Science.gov (United States)

    Roche, M.

    An attempt to design a solar thermal electric central receiver power plant in the multi-MW size with acceptable efficiencies using air in the power loop is described. The turbine and generator are placed in the tower to reduce heat losses in the superheated gas, and the depleted gas loop is coupled to a low temperature generator powered by boiling water. The receiver cavity is configured to retain a maximum amount of flux and has brick walls. Nickel alloys are indicated for the air tubes in the receiver, with Inconel 601, Incoloy 800, and Inconel 600 considered acceptable. The gas leaving the chamber will be at 950 C to power a high pressure turbine, followed by entrance into a heat exchanger to boil the water for the low-pressure turbine, and is then discharged. Thermodynamic efficiencies between 13.9-20.3 percent for a 4700 kW plant are considered feasible with the design.

  16. Analysis of structural materials for fast-breeder reactors by X-ray fluorescence spectrometry

    International Nuclear Information System (INIS)

    A procedure for the X-ray spectrometric determination of Co, Cr, Cu, Mn, Mo, Nb, Ni, P, S, Si and Ti in stainless steels and some nickel-base alloys, such as incoloy-800, is described. The use of different sets of standards has allowed the calculation of the inter-element influence coefficients for the correction of matrix effects, making the method suited for wide concentration range determinations. The efficiency of X-ray tubes with Cr and W targets has been studied, the former allowing the determination of all the above-named elements. The average relative error is 3.7%, except for P and S, where the determinations are semiquantitative. The use of a programmable spectrometer interfaced with a 16 K computer facilitates considerably the treatment of data with a proper mathematic model and furthermore provides an automatic performance of the analyses. (author)

  17. Steam generator materials and secondary side water chemistry in nuclear power stations

    International Nuclear Information System (INIS)

    The main purpose of this work is to summarize the European and North American experiences regarding the materials used for the construction of the steam generators and their relative corrosion resistance considering the water chemestry control method. Reasons underlying decision for the adoption of Incoloy 800 as the material for the secondary steam generator system for Atucha I Nuclear Power Plant (Atucha Reactor) and Embalse de Rio III Nuclear Power Plant (Cordoba Reactor) are pointed out. Backup information taken into consideration for the decision of utilizing the All Volatil Treatment for the water chemistry control of the Cordoba Reactor is detailed. Also all the reasonswhich justify to continue with the congruent fosfatic method for the Atucha Reactor are analyzed. Some investigation objectives which would eventually permit the revision of the decisions taken on these subjects are proposed. (E.A.C.)

  18. Evaluation of High-Temperature Tensile Property of Diffusion Bond of Austenitic Alloys for S-CO2 Cycle Heat Exchangers

    International Nuclear Information System (INIS)

    To improve the inherent safety of the sodium-cooled fast reactor (SFR), the supercritical CO2 (S-CO2) Brayton cycle is being considered as an alternative power conversion system to steam the Rankine cycle. In the S-CO2 system, a PCHE (printed circuit heat exchanger) is being considered. In this type of heat exchangers, diffusion bonding is used for joining the thin plates. In this study, the diffusion bonding characteristics of various austenitic alloys were evaluated. The tensile properties were measured at temperatures starting from the room temperature up to 650℃. For the 316H and 347H types of stainless steel, the tensile ductility was well maintained up to 550℃. However, the Incoloy 800HT showed lower strength and ductility at all temperatures. The microstructure near the bond line was examined to understand the reason for the loss of ductility at high temperatures

  19. Oxidation and emittance of superalloys in heat shield applications

    Science.gov (United States)

    Wiedemann, K. E.; Clark, R. K.; Unnam, J.

    1986-01-01

    Recently developed superalloys that form alumina coatings have a high potential for heat shield applications for advanced aerospace vehicles at temperatures above 1095C. Both INCOLOY alloy MA 956 (of the Inco Alloys International, Inc.), an iron-base oxide-dispersion-strengthened alloy, and CABOT alloy No. 214 (of the Cabot Corporation), an alumina-forming nickel-chromium alloy, have good oxidation resistance and good elevated temperature strength. The oxidation resistance of both alloys has been attributed to the formation of a thin alumina layer (alpha-Al2O3) at the surface. Emittance and oxidation data were obtained for simulated Space Shuttle reentry conditions using a hypersonic arc-heated wind tunnel. The surface oxides and substrate alloys were characterized using X-ray diffraction and scanning and transmission electron microscopy with an energy-dispersive X-ray analysis unit. The mass loss and emittance characteristics of the two alloys are discussed.

  20. TPX: Contractor preliminary design review. Volume 5, Manufacturing R&D

    Energy Technology Data Exchange (ETDEWEB)

    Roach, J.F.; Urban, W.M.; Hartman, D. [Everson Electric Co., Bekthlehem, PA (United States)

    1995-08-04

    TPX Insulation & Impregnation R&D test results are reported for 1x2 samples designed for screening candidate conduit insulation systems for TPX PF and TF coils. The epoxy/glass insulation system and three proposed alternate insulation systems employing Kapton, was evaluated in as received sample condition and after 10 thermal cycles in liquid nitrogen. Two DGBA impregnation systems, Shell 826 and CTD101K were investigated. Square incoloy 908 and 316 LN stainless hollow conduits were used for 1x2 sample fabrication. Capacitance, dielectric loss, and insulation resistance dielectric characteristics were measured for all samples. Partial discharge performance was measured for samples either in air, under silicon oil, or under liquid nitrogen up to 10kVrms at 60 Hz. Hipot screening was performed at 10 kVdc. The samples were cross sectioned and evaluated for impregnation quality. The implications of the test results on the TPX preliminary design decision are discussed.

  1. Corrosion-mechanical tests of the materials for steam generator tubes of NPP with WWER. [Kh14MF steel

    Energy Technology Data Exchange (ETDEWEB)

    Glebov, V.P.; Grigor' ev, A.S.; Moskvichev, V.F.; Taratuta, V.A.; Trubachev, V.M.; Ehskin, N.B.

    1982-09-01

    A study is made on stress-corrosion resistance of Kh14MF, 08Kh18N10T steels, inconel-600 and incoloy-800 alloys as applied to operation conditions of once-through steam generators of NPP with WWER type reactor. Tests of tube samples of the above steels and alloys have been conducted under conditions of mechanical stresses and thermal cycling. Ammonia and neutral-oxidizing water-chemical conditions are used. A comparison of test results is performed. A conclusion is drawn that Kh14MF steel is not inferior to inconel alloy, as to stress-corrosion resistance and is a promising material for heating surface tubes a once-through steam generator, seperating under neutral-oxidizing water conditions of the secondary coolant circuit of nuclear power stations with WWER type reactors.

  2. Corrosion-mechanical tests of the materials for steam generator tubes of NPP with WWER

    International Nuclear Information System (INIS)

    A study is made on stress-corrosion resistance of Kh14MF, 08Kh18N10T steels, inconel-600 and incoloy-800 alloys as applied to operation conditions of once-through steam generators of NPP with WWER type reactor. Tests of tube samples of the above steels and alloys have been conducted under conditions of mechanical stresses and thermal cycling. Ammonia and neutral-oxidizing water- chemical conditions are used. A comparison of test results is performed. A conclusion is drawn that Kh14MF steel is not inferior to inconel alloy, as to stress-corrosion resistance and is a promising material for heating surface tubes a once-through steam generator, eperating under neutral-oxidizing water conditions of the secondary coolant circuit of nuclear power stations with WWER type reactors

  3. Comparison of mechanical and corrosion behaviour of Alloy 800 H and the new alloy AC 66

    International Nuclear Information System (INIS)

    In coal gasification plants based on nuclear process heat, materials are subjected to high temperature corrosion in process gas atmosphere at 750 to 900 deg. C. The process gas consists of steam, CO, CH4, CO2 and, depending on the gasified coal, low or high H2S-concentrations. The service problems can be divided as follows: 1. Gas-metal interaction: (a) high temperature corrosion; (b) sulphidation; (c) carburization; (d) internal oxidation or internal sulphidation. 2. Ash (slag)-metal interaction: (a) corrosion in molten salts; (b) erosion. 3. Mechanical loading: (a) embrittlement; (b) thermal fluctuations/strain fluctuations; (c) low cycle fatigue; (d) high temperature creep. Therefore materials for heat exchangers must be resistant to these types of high temperature corrosion and they should also have adequate creep rupture strength. Some commercial alloys and various model alloys were exposed to a process gas atmosphere to determine the corrosion behaviour and also stressed mechanically to investigate the interaction of high temperature creep behaviour and corrosion. The tests were carried out for a total period of 10,000 h and specimens were taken out after periods of 1000, 3000, 5000 and 10,000 h. A programme for the development of alloys was started with the aim of optimizing the chemical composition resulting in a good high temperature corrosion resistance and adequate mechanical properties, particularly high creep strength. However, the material must be such that it can be deformed to tubes. Compared with Incoloy 800, one of the new and optimized model alloys (30-32% Ni, 25-27% Cr, and Ce, Fe-balance) exhibits a very good corrosion resistance even when sulphur rich coal is gasified. The creep rupture strength at 900 deg. C is in the range of the creep strength for Incoloy 800. 29 figs

  4. 超临界水冷堆候选高温合金低周疲劳性能研究%Low-Cycle Fatigue Property of Candidate High Temperature Alloys for Supercritical Water Cooled Reactors

    Institute of Scientific and Technical Information of China (English)

    陈乐; 唐睿; 梁波; 张强; 刘鸿

    2014-01-01

    采用MTS材料试验机研究了作为超临界水冷堆候选材料的Inconel-718、Incoloy-825、Incoloy-800H 3种高温合金,在650℃和室温、±0.5%应变幅的低周疲劳性能,并采用扫描电镜对试验后样品进行了断口分析.结果表明:在两种温度条件下,718的疲劳寿命均最高.温度对3种高温合金的稳态迟滞回线面积和弹性变形量几乎无影响;718的稳态迟滞回线面积远低于825和800H,而弹性变形量几乎达到825和800H的2倍,有利于提高其疲劳寿命.在循环变形过程中,718呈循环软化状态,825和800H呈先循环硬化再循环饱和状态,且在高温下循环硬化效应更明显.在650℃低周疲劳试验后,718样品断口表面的疲劳间距不足1 μm,而对于825和800H则分别达到2.28和2~20 μm,进一步表明了718在3种材料中低周疲劳性能最好.

  5. Examination of four-sodium-water reaction target tubes

    Energy Technology Data Exchange (ETDEWEB)

    Lee, W.T.

    1970-06-26

    Metallurgical examinations and evaluations were performed on a group of four target tubes from Atomic Power Development Associates (APDA) Rig 10. This group included two 2/sup 1///sub 4/ Cr--1 Mo ferritic steel tubes, one Type 304 SS, and one Incoloy 800 tube. These tubes had been exposed to small leaks of water in 600/sup 0/F sodium. Metallographic data indicate that the material loss (wastage) was primarily due to impact (or erosion) of liquid particles. No evidence of corrosion was found in the wasted areas of any target tubes. Minor corrosion was observed on the outside diameter of the austenitic stainless-steel tubes only in areas away from impact; no corrosion was detected on the outside diameter of the ferritic steel tubes. Hardness measurements and microstructural analyses revealed that the maximum temperature may have reached 1600/sup 0/F in the wasted areas. This localized temperature increase appeared to enhance the wastage process for all target tubes regardless of composition. Surface grain elongation and plastic deformation lines were observed in wasted areas, as well as in other areas. However, their cause(s) cannot be ascertained with respect to deformation due to water-jet impact, or deformation due to fabrication and handling, or both. Based on the volumetric measurements of the wastage zones produced, Incoloy 800 (tube No. 21) exhibited greater resistance to water-jet impact/erosion (penetration depth approximately 0.011 in. in 570 sec), while 2/sup 1///sub 4/ Cr--1 Mo ferritic steel displayed the least (depth of penetration 0.064 in. in 21 sec).

  6. Liquid Salts as Media for Process Heat Transfer from VHTR's: Forced Convective Channel Flow Thermal Hydraulics, Materials, and Coating

    Energy Technology Data Exchange (ETDEWEB)

    Sridharan, Kumar; Anderson, Mark; Allen, Todd; Corradini, Michael

    2012-01-30

    The goal of this NERI project was to perform research on high temperature fluoride and chloride molten salts towards the long-term goal of using these salts for transferring process heat from high temperature nuclear reactor to operation of hydrogen production and chemical plants. Specifically, the research focuses on corrosion of materials in molten salts, which continues to be one of the most significant challenges in molten salts systems. Based on the earlier work performed at ORNL on salt properties for heat transfer applications, a eutectic fluoride salt FLiNaK (46.5% LiF-11.5%NaF-42.0%KF, mol.%) and a eutectic chloride salt (32%MgCl2-68%KCl, mole %) were selected for this study. Several high temperature candidate Fe-Ni-Cr and Ni-Cr alloys: Hastelloy-N, Hastelloy-X, Haynes-230, Inconel-617, and Incoloy-800H, were exposed to molten FLiNaK with the goal of understanding corrosion mechanisms and ranking these alloys for their suitability for molten fluoride salt heat exchanger and thermal storage applications. The tests were performed at 850C for 500 h in sealed graphite crucibles under an argon cover gas. Corrosion was noted to occur predominantly from dealloying of Cr from the alloys, an effect that was particularly pronounced at the grain boundaries Alloy weight-loss due to molten fluoride salt exposure correlated with the initial Cr-content of the alloys, and was consistent with the Cr-content measured in the salts after corrosion tests. The alloys weight-loss was also found to correlate to the concentration of carbon present for the nominally 20% Cr containing alloys, due to the formation of chromium carbide phases at the grain boundaries. Experiments involving molten salt exposures of Incoloy-800H in Incoloy-800H crucibles under an argon cover gas showed a significantly lower corrosion for this alloy than when tested in a graphite crucible. Graphite significantly accelerated alloy corrosion due to the reduction of Cr from solution by graphite and formation

  7. Liquid Salts as Media for Process Heat Transfer from VHTR's: Forced Convective Channel Flow Thermal Hydraulics, Materials, and Coating

    Energy Technology Data Exchange (ETDEWEB)

    Sridharan, Kumar; Anderson, Mark; Allen, Todd; Corradini, Michael

    2012-01-30

    The goal of this NERI project was to perform research on high temperature fluoride and chloride molten salts towards the long-term goal of using these salts for transferring process heat from high temperature nuclear reactor to operation of hydrogen production and chemical plants. Specifically, the research focuses on corrosion of materials in molten salts, which continues to be one of the most significant challenges in molten salts systems. Based on the earlier work performed at ORNL on salt properties for heat transfer applications, a eutectic fluoride salt FLiNaK (46.5% LiF-11.5%NaF-42.0%KF, mol.%) and a eutectic chloride salt (32%MgCl2-68%KCl, mole %) were selected for this study. Several high temperature candidate Fe-Ni-Cr and Ni-Cr alloys: Hastelloy-N, Hastelloy-X, Haynes-230, Inconel-617, and Incoloy-800H, were exposed to molten FLiNaK with the goal of understanding corrosion mechanisms and ranking these alloys for their suitability for molten fluoride salt heat exchanger and thermal storage applications. The tests were performed at 850C for 500 h in sealed graphite crucibles under an argon cover gas. Corrosion was noted to occur predominantly from dealloying of Cr from the alloys, an effect that was particularly pronounced at the grain boundaries Alloy weight-loss due to molten fluoride salt exposure correlated with the initial Cr-content of the alloys, and was consistent with the Cr-content measured in the salts after corrosion tests. The alloys weight-loss was also found to correlate to the concentration of carbon present for the nominally 20% Cr containing alloys, due to the formation of chromium carbide phases at the grain boundaries. Experiments involving molten salt exposures of Incoloy-800H in Incoloy-800H crucibles under an argon cover gas showed a significantly lower corrosion for this alloy than when tested in a graphite crucible. Graphite significantly accelerated alloy corrosion due to the reduction of Cr from solution by graphite and formation

  8. Molten Chloride Salts for Heat Transfer in Nuclear Systems

    Science.gov (United States)

    Ambrosek, James Wallace

    2011-12-01

    A forced convection loop was designed and constructed to examine the thermal-hydraulic performance of molten KCl-MgCl2 (68-32 at %) salt for use in nuclear co-generation facilities. As part of this research, methods for prediction of the thermo-physical properties of salt mixtures for selection of the coolant salt were studied. In addition, corrosion studies of 10 different alloys were exposed to the KCl-MgCl2 to determine a suitable construction material for the loop. Using experimental data found in literature for unary and binary salt systems, models were found, or developed to extrapolate the available experimental data to unstudied salt systems. These property models were then used to investigate the thermo-physical properties of the LINO3-NaNO3-KNO 3-Ca(NO3), system used in solar energy applications. Using these models, the density, viscosity, adiabatic compressibility, thermal conductivity, heat capacity, and melting temperatures of higher order systems can be approximated. These models may be applied to other molten salt systems. Coupons of 10 different alloys were exposed to the chloride salt for 100 hours at 850°C was undertaken to help determine with which alloy to construct the loop. Of the alloys exposed, Haynes 230 had the least amount of weight loss per area. Nickel and Hastelloy N performed best based on maximum depth of attack. Inconel 625 and 718 had a nearly uniform depletion of Cr from the surface of the sample. All other alloys tested had depletion of Cr along the grain boundaries. The Nb in Inconel 625 and 718 changed the way the Cr is depleted in these alloys. Grain-boundary engineering (GBE) of Incoloy 800H improved the corrosion resistance (weight loss and maximum depth of attack) by nearly 50% as compared to the as-received Incoloy 800H sample. A high temperature pump, thermal flow meter, and pressure differential device was designed, constructed and tested for use in the loop, The heat transfer of the molten chloride salt was found to

  9. Review of waste package verification tests. Semiannual report, April 1985-September 1985

    International Nuclear Information System (INIS)

    Several studies were completed this period to evaluate experimental and analytical methodologies being used in the DOE waste package program. The first involves a determination of the relevance of the test conditions being used by DOE to characterize waste package component behavior in a salt repository system. Another study focuses on the testing conditions and procedures used to measure radionuclide solubility and colloid formation in repository groundwaters. An attempt was also made to evaluate the adequacy of selected waste package performance codes. However, the latter work was limited by an inability to obtain several codes from DOE. Nevertheless, it was possible to comment briefly on the structures and intents of the codes based on publications in the open literature. The final study involved an experimental program to determine the likelihood of stress-corrosion cracking of austenitic stainless steels and Incoloy 825 in simulated tuff repository environments. Tests for six-month exposure periods in water and air-steam conditions are described. 52 figs., 48 tabs

  10. Status of the TRIGA shipments to the INEEL from Europe

    International Nuclear Information System (INIS)

    During 1999 shipment from 4 European countries, involving the following 4 research reactors was foreseen: ENEA of Italy, ICN of Romania, TRIGA-IJS of Slovenia, and MHH of Germany. The research reactors under consideration are LENA of Italy, IFK and DKFZ of Germany. Unique challenges of this task are: first shipment to the INEEL from the east coast of the United States; Need to identify a transportation route and working with the states, tribes and local governments to ensure that adequate public safety and security planning is done and followed; first shipment to INEEL involving both high-income and less-than-high-income countries in one shipment. There is an opportunity to save a significant amount of money for both DOE and the high-income countries by cooperating and coordinating the shipments together. The First will be the shipment to INEEL of mixed TRIGA SNF and more than one shipping cask type. This shipment will include a mixture of LEU, HEU, aluminum clad, stainless steel clad, and Incoloy clad rods. INEEL will need to prepare the safety documentation, procedures, and make equipment and facility modifications necessary to handle the ifferent fuel and cask types

  11. Structure of strongly underexpanded gas jets submerged in liquids – Application to the wastage of tubes by aggressive jets

    International Nuclear Information System (INIS)

    Highlights: • Underexpanded gas jets submerged in liquids behave similarly to homogeneous gas jets. • The counter rotating vortex pairs of jet produce discrete imprints on the targets. • The shape of hollows made on the targets is explained by the jet structure. • The erosion–corrosion phenomenon well explains the wastage of exchange tubes. - Abstract: Strongly underexpanded gas jets submerged in a liquid at rest behave similarly to underexpanded homogeneous gas jets. The existence of the Taylor-Görtler vortices around the inner zone of the gas jets is demonstrated in free gas jets submerged in water by means of optical probe. In the near field, the same phenomenon produces discrete imprints, approximately distributed in a circle, when underexpanded nitrogen jet submerged in liquid sodium hydroxide and underexpanded water vapour jet submerged in liquid sodium impact onto AU4G-T4 and Incoloy 800® alloy targets respectively. For a jet-target couple, the volume of the hollow is satisfactorily related to the strain energy density of the material and the kinetic energy of the gas jet. However, the comparison between volumes of hollows produced by both jets also indicates strong corrosive action of the medium on targets. This allows better understanding of the mechanism of wastage of tubes employed in steam generators integrated in liquid metal fast breeder reactors

  12. Erosion-Corrosion of Iron and Nickel Alloys at Elevated Temperature in a Combustion Gas Environment

    Energy Technology Data Exchange (ETDEWEB)

    Tylczak, Joseph [NETL

    2014-05-02

    This paper reports on the results of a study that compares the erosion-corrosion behavior of a variety of alloys (Fe- 2¼Cr 1Mo, 304 SS, 310 SS, Incoloy 800, Haynes 230 and a Fe3Al) in a combustion environment. Advanced coal combustion environments, with higher temperatures, are driving re-examination of traditional and examination of new alloys in these hostile environments. In order to simulate conditions in advanced coal combustion boilers, a special erosion apparatus was used to allow for impingement of particles under a low abrasive flux in a gaseous environment comprised of 20 % CO2, 0.05 % HCl, 77 % N2, 3 % O2, and 0.1 % SO2. Tests were conducted at room temperature and 700 °C with ~ 270 μm silica, using an impact velocity of 20 m/s in both air and the simulated combustion gas environment. The erosion-corrosion behavior was characterized by gravimetric measurements and by examination of the degraded surfaces optically and by scanning electron microscopy (SEM). At room temperature most of the alloys had similar loss rates. Not surprisingly, at 700 °C the lower chrome-iron alloy had a very high loss rate. The nickel alloys tended to have higher loss rates than the high chrome austenitic alloys.

  13. Chemical analysis of nickel- and iron-base high-temperature alloys for nuclear reactor

    International Nuclear Information System (INIS)

    The Committee studied problems in analysis of alloys used for High-Temperature Gas Cooled Reactor from September 1970 to February 1976. The alloys selected from the standpoint of analytical chemistry are Inconel 600, Incoloy 800, Inconel X750, Inco 713C and Hastelloy X. Nine standard samples (JAERI-R 1 to JAERI-R 9) of the high-temperature alloys were prepared primarily for X-ray fluorescence method. Eighteen research institutions in Japan participated in cooperative analyses of the standard samples for 19 elements (C, Si, Mn, P, S, Ni, Cr, Fe, Mo, Cu, W, V, Co, Ti, Al, B, Nb, Ta, Zr). Prior to analyses of the standard samples, 8 cooperative samples (A-H) were analyzed to develop and evaluate analytical methods. Described in this report are preparation and their characteristics of the standard samples, results of analyses, and 93 analytical methods. The results of the cooperative experiments on atomic absorption spectrophotometry and X-ray fluorescence method are also described. (auth.)

  14. Creep Property Characterization of Potential Brayton Cycle Impeller and Duct Materials

    Science.gov (United States)

    Gabb, Timothy P.; Gayda, John; Garg, Anita

    2007-01-01

    Cast superalloys have potential applications in space as impellers within closed-loop Brayton cycle nuclear power generation systems. Likewise wrought superalloys are good candidates for ducts and heat exchangers transporting the inert working gas in a Brayton-based power plant. Two cast superalloys, Mar-M247LC and IN792, and a NASA GRC powder metallurgy superalloy, LSHR, have been screened to compare their respective capabilities for impeller applications. Mar-M247LC has been selected for additional long term evaluations. Initial tests in helium indicate this inert environment may debit long term creep resistance of this alloy. Several wrought superalloys including Hastelloy® X, Inconel® 617, Inconel® 740, Nimonic® 263, Incoloy® MA956, and Haynes 230 are also being screened to compare their capabilities for duct applications. Haynes 230 has been selected for additional long term evaluations. Initial tests in helium are just underway for this alloy. These proposed applications would require sufficient strength and creep resistance for long term service at temperatures up to 1200 K, with service times to 100,000 h or more. Therefore, long term microstructural stability is also being screened.

  15. Corrosion of high Ni-Cr alloys and Type 304L stainless steel in HNO3-HF

    International Nuclear Information System (INIS)

    Nineteen alloys were evaluated as possible materials of construction for steam heating coils, the dissolver vessel, and the off-gas system of proposed facilities to process thorium and uranium fuels. Commercially available alloys were found that are satisfactory for all applications. With thorium fuel, which requires HNO3-HF for dissolution, the best alloy for service at 1300C when complexing agents for fluoride are used is Inconel 690; with no complexing agents at 1300C, Inconel 671 is best. At 950C, six other alloys tested would be adequate: Haynes 25, Ferralium, Inconel 625, Type 304L stainless steel, Incoloy 825, and Haynes 20 (in order of decreasing preference); based on composition, six untested alloys would also be adequate. The ions most effective in reducing fluoride corrosion were the complexing agents Zr4+ and Th4+; Al3+ was less effective. With uranium fuel, modestly priced Type 304L stainless steel is adequate. Corrosion will be most severe in HNO3-HF used occasionally for flushing and in solutions of HNO3 and corrosion products (ferric and dichromate ions). HF corrosion can be minimized by complexing the fluoride ion and by passivation of the steel with strong nitric acid. Corrosion caused by corrosion products can be minimized by operating at lower temperatures

  16. Materials technology for coal-conversion processes. Seventeenth quarterly report, January-March 1979

    Energy Technology Data Exchange (ETDEWEB)

    Ellingson, W. A.

    1979-01-01

    Studies of slag attack on refractories were continued, utilizing conditions relevant to MHD applications. Addition of 10 wt % K/sub 2/O seed to the slag did not increase its corrosive effect on the refractories tested. A hot gas-stream cleanup erosion-monitoring system using an ANL-developed nondestructive ultrasonic system was installed at the Morgantown Energy Technology Center (METC) during this period and was 75% completed. Characteristic-slope values obtained from broadband and resonant-band acoustic-emission transducers during rapid heating of a 95% Al/sub 2/O/sub 3/ refractory panel are consistent with theory. Corrosion information on type and thickness of corrosion-product layers was obtained on Incoloy 800, 310 stainless steel, Inconel 671 and 871 and 982/sup 0/C. Fluid-bed corrosion studies involving sulfation accelerators have shown that addition of 0.3 mol % CaCl/sub 2/ has no significant effect on corrosion behavior of the alloys studied. However, 0.5 mol % NaCl or 1.9 mol % Na/sub 2/CO/sub 3/ increases the corrosion rates of most materials. Failure analyses were performed on components from the slagging gasifier and liquefaction unit at the Grand Forks Energy Technology Center, and a ball valve from the METC Valve Dynamic Test Unit.

  17. Creep Property Characterization of Potential Brayton Cycle Impeller and Duct Materials

    Science.gov (United States)

    Gabb, Timothy P.; Gayda, john; Garg, Anita

    2007-01-01

    Cast superalloys have potential applications in space as impellers within closed-loop Brayton cycle nuclear power generation systems. Likewise wrought superalloys are good candidates for ducts and heat exchangers transporting the inert working gas in a Brayton-based power plant. Two cast superalloys, Mar-M247LC and IN792, and a NASA GRC powder metallurgy superalloy, LSHR, have been screened to compare their respective capabilities for impeller applications. Mar-M247LC has been selected for additional long term evaluations. Initial tests in helium indicate this inert environment may debit long term creep resistance of this alloy. Several wrought superalloys including Hastelloy(Registered TradeMark) X, Inconel(Registered TradeMark) 617, Inconel(Registered TradeMark) 740, Nimonic(Registered TradeMark) 263, Incoloy(Registered TradeMark) MA956, and Haynes 230 are also being screened to compare their capabilities for duct applications. Haynes 230 has been selected for additional long term evaluations. Initial tests in helium are just underway for this alloy. These proposed applications would require sufficient strength and creep resistance for long term service at temperatures up to 1200 K, with service times to 100,000 h or more. Therefore, long term microstructural stability is also being screened.

  18. Tensile and Creep Property Characterization of Potential Brayton Cycle Impeller and Duct Materials

    Science.gov (United States)

    Gabb, Timothy P.; Gayda, John

    2006-01-01

    This paper represents a status report documenting the work on creep of superalloys performed under Project Prometheus. Cast superalloys have potential applications in space as impellers within closed-loop Brayton cycle nuclear power generation systems. Likewise wrought superalloys are good candidates for ducts and heat exchangers transporting the inert working gas in a Brayton-based power plant. Two cast superalloys, Mar-M247LC and IN792, and a NASA GRC powder metallurgy superalloy, LSHR, are being screened to compare their respective capabilities for impeller applications. Several wrought superalloys including Hastelloy X, (Haynes International, Inc., Kokomo, IN), Inconel 617, Inconel 740, Nimonic 263, and Incoloy MA956 (Special Metals Corporation, Huntington, WV) are also being screened to compare their capabilities for duct applications. These proposed applications would require sufficient strength and creep resistance for long term service at temperatures up to 1200 K, with service times to 100,000 h or more. Conventional tensile and creep tests were performed at temperatures up to 1200 K on specimens extracted from the materials. Initial microstructure evaluations were also undertaken.

  19. Alloy SCR-3 resistant to stress corrosion cracking

    International Nuclear Information System (INIS)

    Austenitic stainless steel is used widely because the corrosion resistance, workability and weldability are excellent, but the main fault is the occurrence of stress corrosion cracking in the environment containing chlorides. Inconel 600, most resistant to stress corrosion cracking, is not necessarily safe under some severe condition. In the heat-affected zone of SUS 304 tubes for BWRs, the cases of stress corrosion cracking have occurred. The conventional testing method of stress corrosion cracking using boiling magnesium chloride solution has been problematical because it is widely different from actual environment. The effects of alloying elements on stress corrosion cracking are remarkably different according to the environment. These effects were investigated systematically in high temperature, high pressure water, and as the result, Alloy SCR-3 with excellent stress corrosion cracking resistance was found. The physical constants and the mechanical properties of the SCR-3 are shown. The states of stress corrosion cracking in high temperature, high pressure water containing chlorides and pure water, polythionic acid, sodium phosphate solution and caustic soda of the SCR-3, SUS 304, Inconel 600 and Incoloy 800 are compared and reported. (Kako, I.)

  20. Oxidation of high-temperature alloys J(superalloys) at elevated temperatures in air.II

    Energy Technology Data Exchange (ETDEWEB)

    Hussain, N.; Shahid, K.A.; Rahman, S. [Pakistan Institute of Nuclear Science and Technology, Islamabad (Pakistan)] [and others

    1995-04-01

    In continuation of previous work, the oxidation behavior of three more alloys, namely UHBA 25L, Sanicro 28, and Inconel 690, was studied at high temperatures (600 to 1200{degrees}C). The oxidation kinetics of UHBA 25L and Sanicro 28 followed the parabolic-rate law at 800 and 1000{degrees}C. At 600{degrees}C oxidation rates were very low, while at 1200{degrees}C the parabolic-rate law was initially observed, followed by {open_quotes}breakaway.{close_quotes} Both alloys suffered extensive spalling at all temperatures, despite their high Cr contents of about 25% and 27%, respectively. This is attributed to significant amounts of Mn present. Sanicro 28, with higher Cr, suffered an early breakaway at 100{degrees}C because of the presence of Mo, which forms a low-melting oxide. Inconel 690 showed mixed behavior at 600 and 800{degrees}C, parabolic at 1000{degrees}C and cubic at 1200{degrees}C. Inconel 690 and Incoloy 800H have the best overall performance for the range of temperature and exposure time under study. This may be ascribed to the absence of Mo, the presence of small amounts of Ti and Al, and relatively small amount of Mn in these alloys.

  1. The materials concept in German light water reactors. A contribution to plant safety, economic performance and damage prevention

    International Nuclear Information System (INIS)

    Major decisions taken as early as in the planning and construction phases of nuclear power plants may influence overall plant life. Component quality at the beginning of plant life is determined very much also by a balanced inclusion of the 'design, choice of materials, manufacturing and inspection' elements. One example of the holistic treatment of design, choice of material, and manufacture of important safety-related components in pressurized water reactors is the reactor pressure vessel (RPV) in which the ferritic compound tubes, with inside claddings, for the control rod drive nozzles are screwed into the vessel top. Also the choice of Incoloy 800 for the steam generator tubes, and the design of the main coolant pipes with inside claddings as seamless pipe bends / straight pipes with integrated nozzles connected to mixed welds with austenitic pipes are other special design features of the Siemens/KWU plants. A demonstrably high quality standard by international comparison to this day has been exhibited by the austenitic RPV internals of boiling water reactors, which were made of a low-carbon Nb-stabilized austenitic steel grade by optimum manufacturing technologies. The same material is used for backfitting austenitic pipes. Reliable and safe operation of German nuclear power plants has been demonstrated for more than 4 decades. One major element in this performance is the materials concept adopted in Germany also in the interest of damage prevention. (orig.)

  2. The material concept in German light water reactors. Contribution to plant safety economic efficiency and failure provision; Das Werkstoffkonzept in deutschen Leichtwasserreaktoren. Beitrag zur Anlagensicherheit, Wirtschaftlichkeit und Schadensvorsorge

    Energy Technology Data Exchange (ETDEWEB)

    Ilg, Ulf [EnBW Kernkraft GmbH (Germany). Kernkraftwerk Philippsburg; Koenig, Guenter [EnBW Kernkraft GmbH (Germany). Kernkraftwerk Neckarwestheim; Erve, Manfred [AREVA NP GmbH, Erlangen (Germany)

    2008-07-01

    In the design and construction stage of nuclear power plants relevant decisions may affect the service life of a component, and thus influence safety and availability of the plant. The German ''basic safety concept'' has an important effect on the quality of the BOL (begin of life) status. Materials selection and qualification are of significant importance for the component lifetime and the profitability of the plant. Examples for the implementation of this concept are demonstrated for the steam generator tubing material Incoloy 800, the inside-plated ferritic compound tubes as control rod drive mechanism nozzle through the RPV head of BWR plants that are not susceptible for corrosion enhanced cracking that was observed for Inconel 600 tubing. A fundamental failure analysis of crack formation in Ti stabilized austenitic pipes of BWR plants found since 1993 were definitely identified as intergranular stress corrosion caused by a local sensitization of the welding process induced overheated structured in the heat affected zone. This allowed target-oriented mitigation measures. The safety culture implemented in German nuclear plants in connection with the break preclusion or integrity concept, respectively, including a continuous actualization with respect to the state-of-the art are the technical prerequisites for damage precaution and possible life time extension.

  3. Use of the slow-strain-rate technique for the evaluation of structural materials for application in high-temperature gaseous environments

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, C.E.; Ugiansky, G.M.

    1981-01-01

    Types 309, 310, 310S, 347 and 446 stainless steels, Incoloy 800, and Inconel 671 were tested at temperatures from 370 to 1040/sup 0/C at strain rates from 10/sup -4/ to 10/sup -7//s in H/sub 2/S plus water, gaseous mixtures of CO, CO/sub 2/, H/sub 2/, CH/sub 4/, H/sub 2/S, and H/sub 2/O, and in nominally inert environments of He and Ar. Type 310 steel showed a marked reduction in mechanical properties at low strain rates (< 10/sup -5//s) in H/sub 2/S/H/sub 2/O at 540/sup 0/C, and this was associated with the occurrence of a large degree of secondary intergranular cracking in addition to the main ductile fracture mode. The occurrence of the secondary cracking was taken as the primary indication of embrittlement in subsequent tests. It occurred to some degree in all alloys tested in the simulated coal-gasification environments at 600/sup 0/C. The mechanism(s) of the embrittlement phenomena remain uncertain; a number of possible causes including creep and several environmentally-induced fracture processes are outlined. It is shown that the overall results of the test program are in good agreement with in-plant experience.

  4. Development of hydrothermal power generation plant. Development of binary cycle power generation plant (development of 10 MW-class plant); 1995 nendo nessui riyo hatsuden plant nado kaihatsu binary cycle hatsuden plant no kaihatsu. 10MW kyu plant no kaihatsu

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-01

    A 10 MW-class binary cycle power generation plant has been developed using a down hole pump (DHP) which exchanges the hydrothermal energy with secondary medium in the heat exchanger. For constructing the plant at Kuju-machi, Oita Prefecture, site preparation works, foundation of cooling tower, reconstruction of roads, and survey on environmental influences were conducted. To investigate installation and removal methods of DHP, a geothermal water pump-up system, current status of the binary cycle power generating system in the USA was surveyed. In this survey, a trailer mounting handling machine was inspected. Based on the survey results, a simple assembled, easy-installation type handling equipment was designed. In addition, the replacement work for motor connector joint of DHP and the strength of coil end were improved. Construction and method allowing reuse of the motor cable were considered by improving the cable and cable end portion. The air tight soundness of incoloy corrugate sheath was confirmed. Finally, a reproduction system for waste oil of DHP bearing oil was investigated. 106 figs., 52 tabs.

  5. About the Welding Process of the Undersea Gas Lined-Pipe%海底输气复合管道焊接工艺

    Institute of Scientific and Technical Information of China (English)

    芦红威; 王喆

    2015-01-01

    针对焊接时管线的焊缝和热影响区的化学成分不均匀性、组织偏析等缺陷,制定相应的试验方案. 开发出新的焊接工艺,运用MIG焊接方法、选取相应的焊丝和混合气体作为保护气体,确定最佳焊接参数. 焊接接头的焊接工艺评定试验表明,以此工艺完成的复合管道焊接符合海底输气管道的理化及机械性能要求.%Aimming at the welding defects of tissue segregation and uneven chemical composition of the welding and heat af-fected zone for the submarine gas pipeline, the corresponding test plan is formulated.According to the welding difficulty of the mechanical clad pipe, the new welding process is developed, in which the MIG welding method is applied, selecting Incoloy625 nickel base welding wire, using the Ar+20%He mixed gas as a protective gas.The optimal welding parameters are determined. The welding procedure qualification test of the composite pipe welding joint showed that the welding can meet the requirements of physical and chemical properties and mechanical properties of the submarine gas pipeline.

  6. Stable plasma start-up in the KSTAR device under various discharge conditions

    Science.gov (United States)

    Kim, Jayhyun; Yoon, S. W.; Jeon, Y. M.; Leuer, J. A.; Eidietis, N. W.; Mueller, D.; Park, S.; Nam, Y. U.; Chung, J.; Lee, K. D.; Hahn, S. H.; Bae, Y. S.; Kim, W. C.; Oh, Y. K.; Yang, H. L.; Park, K. R.; Na, H. K.; KSTAR Team

    2011-08-01

    A time series of static nonlinear ferromagnetic calculations was performed to mimic the time-dependent modelling of plasma start-up by assessing the effects of the ferromagnetic Incoloy 908 used in the superconducting coil jackets of the Korea Superconducting Tokamak Advanced Research (KSTAR) device. Time-series calculations of a two-dimensional axisymmetric circuit model with nonlinear ferromagnetic effects enabled us to find appropriate waveforms for the KSTAR poloidal field coil currents that satisfied various start-up requirements, such as the formation and sustainment of field nulls, a sufficient amount of magnetic flux for further plasma current ramp-up, sufficiently large Et ·Bt/Bbottom > 1 kV m-1 contours for successful breakdown, plasma current toroidal equilibria, etc. In addition to the aforementioned requirements, the results introduced in this report also provided the positional stability of the plasma current channel against radial as well as vertical perturbations by compensating the field deformation originating from the ferromagnetic effects. With the improved positional stability, robust plasma start-up was achieved during the 2010 KSTAR campaign under various discharge conditions such as the recovery process from plasma disruptions.

  7. Formation and reversion of strain induced martensite on Fe-Cr-Ni alloys Formação e reversão da martensita em ligas ferro-cromo-níquel

    Directory of Open Access Journals (Sweden)

    Gabriela Lujan Brollo

    2013-06-01

    Full Text Available Austenitic stainless steels represent a significant portion of the alloys used in the aeronautical, chemical, shipbuilding, food processing and biomechanical industries. They combine good mechanical properties with high corrosion resistance. When subjected to cold deformation, these steels exhibit a metastable phase called: strain induced martensite (ferromagnetic, whose formation increases mechanical strength and formability, allowing for a wide range of applications. Heated from room temperature, the strain induced martensite transforms to austenite (non-magnetic. It is easy to find information in literature about the strain induced martensite for 18Cr/8Ni austenitic steels, but there is no data for high nickel alloys like A286 (26Ni, 15Cr, Incoloy 800 (30-40 Ni, 21Cr and Inconel (50Ni, 19Cr. Therefore, this study aimed to verify the formation of strain induced martensite after cold working in Fe-18Cr base alloys with the addition of up to 60 %Ni. The reversion of this phase to austenite after annealing up to 600 ºC was also studied. Optical microscopy, magnetic characterization tests, and x-ray diffraction were used to analyze the transformations.Os aços inoxidáveis austeníticos são materiais de alto valor agregado, representando uma parcela importante das ligas usadas, principalmente, nas indústrias aeronáutica, química, naval, alimentícia e biomecânica. Apresentam boas propriedades mecânicas aliadas à elevada resistência à corrosão. Quando submetido à deformação a frio, esses aços exibem uma fase metaestável denominada martensita induzida por deformação (ferromagnética, cuja formação aumenta a resistência mecânica e conformabilidade, permitindo sua ampla gama de aplicações. Aquecida acima da temperatura ambiente, a martensita induzida por deformação se transforma em austenita. Existem dados na literatura sobre a formação da martensita induzida por deformação em aços austeníticos 18Cr/8Ni, mas não h

  8. High Temperature Fuel Cladding Chemical Interactions Between TRIGA Fuels and 304 Stainless Steel

    Energy Technology Data Exchange (ETDEWEB)

    Perez, Emmanuel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Keiser, Jr., Dennis D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Forsmann, Bryan [Boise State Univ., ID (United States); Janney, Dawn E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Henley, Jody [Idaho National Lab. (INL), Idaho Falls, ID (United States); Woolstenhulme, Eric C. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-02-01

    High-temperature fuel-cladding chemical interactions (FCCI) between TRIGA (Training, Research, Isotopes, General Atomics) fuel elements and the 304 stainless steel (304SS) are of interest to develop an understanding of the fuel behavior during transient reactor scenarios. TRIGA fuels are composed of uranium (U) particles dispersed in a zirconium-hydride (Zr-H) matrix. In reactor, the fuel is encased in 304-stainless-steel (304SS) or Incoloy 800 clad tubes. At high temperatures, the fuel can readily interact with the cladding, resulting in FCCI. A number of FCCI can take place in this system. Interactions can be expected between the cladding and the Zr-H matrix, and/or between the cladding and the U-particles. Other interactions may be expected between the Zr-H matrix and the U-particles. Furthermore, the fuel contains erbium-oxide (Er-O) additions. Interactions can also be expected between the Er-O, the cladding, the Zr-H and the U-particles. The overall result is that very complex interactions may take place as a result of fuel and cladding exposures to high temperatures. This report discusses the characterization of the baseline fuel microstructure in the as-received state (prior to exposure to high temperature), characterization of the fuel after annealing at 950C for 24 hours and the results from diffusion couple experiments carries out at 1000C for 5 and 24 hours. Characterization was carried out via scanning electron microscopy (SEM) and transmission electron microscopy (TEM) with sample preparation via focused ion beam in situ-liftout-technique.

  9. Design and Fabrication Technique of the Key Components for Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Jin; Song, Ki Nam; Kim, Yong Wan

    2006-12-15

    The gas outlet temperature of Very High Temperature Reactor (VHTR) may be beyond the capability of conventional metallic materials. The requirement of the gas outlet temperature of 950 .deg. C will result in operating temperatures for metallic core components that will approach very high temperature on some cases. The materials that are capable of withstanding this temperature should be prepared, or nonmetallic materials will be required for limited components. The Ni-base alloys such as Alloy 617, Hastelloy X, XR, Incoloy 800H, and Haynes 230 are being investigated to apply them on components operated in high temperature. Currently available national and international codes and procedures are needed reviewed to design the components for HTGR/VHTR. Seven codes and procedures, including five ASME Codes and Code cases, one French code (RCC-MR), and on British Procedure (R5) were reviewed. The scope of the code and code cases needs to be expanded to include the materials with allowable temperatures of 950 .deg. C and higher. The selection of compact heat exchangers technology depends on the operating conditions such as pressure, flow rates, temperature, but also on other parameters such as fouling, corrosion, compactness, weight, maintenance and reliability. Welding, brazing, and diffusion bonding are considered proper joining processes for the heat exchanger operating in the high temperature and high pressure conditions without leakage. Because VHTRs require high temperature operations, various controlled materials, thick vessels, dissimilar metal joints, and precise controls of microstructure in weldment, the more advanced joining processes are needed than PWRs. The improved solid joining techniques are considered for the IHX fabrication. The weldability for Alloy 617 and Haynes 230 using GTAW and SMAW processes was investigated by CEA.

  10. Corrosion of alloys in a chloride molten salt (NaCl-LiCl) for solar thermal technologies

    Energy Technology Data Exchange (ETDEWEB)

    Gomez-Vidal, Judith C.; Tirawat, Robert

    2016-12-01

    Next-generation solar power conversion systems in concentrating solar power (CSP) applications require high-temperature advanced fluids in the range of 600-800 degrees C. Current commercial CSP plants use molten nitrate salt mixtures as the heat transfer fluid and the thermal energy storage (TES) media while operating with multiple hours of energy capacity and at temperatures lower than 565 degrees C. At higher temperatures, the nitrates cannot be used because they decompose. Molten chloride salts are candidates for CSP applications because of their high decomposition temperatures and good thermal properties; but they can be corrosive to common alloys used in vessels, heat exchangers, and piping at these elevated temperatures. In this article, we present the results of the corrosion evaluations of several alloys in eutectic 34.42 wt% NaCl - 65.58 wt% LiCl at 650-700 degrees C in nitrogen atmosphere. Electrochemical evaluations were performed using open-circuit potential followed by a potentiodynamic polarization sweep. Corrosion rates were determined using Tafel slopes and Faraday's law. A temperature increase of as little as 50 degrees C more than doubled the corrosion rate of AISI stainless steel 310 and Incoloy 800H compared to the initial 650 degrees C test. These alloys exhibited localized corrosion. Inconel 625 was the most corrosion-resistant alloy with a corrosion rate of 2.80+/-0.38 mm/year. For TES applications, corrosion rates with magnitudes of a few millimeters per year are not acceptable because of economic considerations. Additionally, localized corrosion (intergranular or pitting) can be catastrophic. Thus, corrosion-mitigation approaches are required for advanced CSP plants to be commercially viable.

  11. Corrosion Evaluation of Alloys and MCrAlX Coatings in Molten Carbonates for Thermal Solar Applications

    Energy Technology Data Exchange (ETDEWEB)

    Gomez-Vidal, Judith C.; Noel, John; Weber, Jacob

    2016-12-01

    Stainless steels (SS) 310, 321, 347, Incoloy 800H (In800H), alumina-forming austenitic (AFA-OC6), Ni superalloy Inconel 625 (IN625), and MCrAlX (M: Ni, and/or Co; X: Y, Hf, Si, and/or Ta) coatings were corroded in molten carbonates in N2 and bone-dry CO2 atmospheres. Electrochemical tests in molten eutectics K2CO3-Na2CO3 and Na2CO3-K2CO3-Li2CO3 at temperatures higher than 600 degrees C were evaluated using an open-circuit potential followed by a potentiodynamic polarization sweep to determine the corrosion rates. Because the best-performing alloys at 750 degrees C were In800H followed by SS310, these two alloys were selected as the substrate material for the MCrAlX coatings. The coatings were able to mitigate corrosion in molten carbonates environments. The corrosion of substrates SS310 and In800H was reduced from ~2500 um/year to 34 um/year when coated with high-velocity oxyfuel (HVOF) NiCoCrAlHfSiY and pre-oxidized (air, 900 degrees C, 24 h, 0.5 degrees C/min) before molten carbonate exposure at 700 degrees C in bone-dry CO2 atmosphere. Metallographic characterization of the corroded surfaces showed that the formation of a uniform alumina scale during the pre-oxidation seems to protect the alloy from the molten carbonate attack.

  12. 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)

    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 200C and 900C 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 900C, 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

  13. Design and Fabrication Technique of the Key Components for Very High Temperature Reactor

    International Nuclear Information System (INIS)

    The gas outlet temperature of Very High Temperature Reactor (VHTR) may be beyond the capability of conventional metallic materials. The requirement of the gas outlet temperature of 950 .deg. C will result in operating temperatures for metallic core components that will approach very high temperature on some cases. The materials that are capable of withstanding this temperature should be prepared, or nonmetallic materials will be required for limited components. The Ni-base alloys such as Alloy 617, Hastelloy X, XR, Incoloy 800H, and Haynes 230 are being investigated to apply them on components operated in high temperature. Currently available national and international codes and procedures are needed reviewed to design the components for HTGR/VHTR. Seven codes and procedures, including five ASME Codes and Code cases, one French code (RCC-MR), and on British Procedure (R5) were reviewed. The scope of the code and code cases needs to be expanded to include the materials with allowable temperatures of 950 .deg. C and higher. The selection of compact heat exchangers technology depends on the operating conditions such as pressure, flow rates, temperature, but also on other parameters such as fouling, corrosion, compactness, weight, maintenance and reliability. Welding, brazing, and diffusion bonding are considered proper joining processes for the heat exchanger operating in the high temperature and high pressure conditions without leakage. Because VHTRs require high temperature operations, various controlled materials, thick vessels, dissimilar metal joints, and precise controls of microstructure in weldment, the more advanced joining processes are needed than PWRs. The improved solid joining techniques are considered for the IHX fabrication. The weldability for Alloy 617 and Haynes 230 using GTAW and SMAW processes was investigated by CEA

  14. Role of maintenance in nuclear fuel fabrication - NFC's perspective

    International Nuclear Information System (INIS)

    Globally, India is ranked 7th growing economy. To sustain growth rate and for continuous improvement, sustained energy security is the key. For developing countries like India, the most reliable long-term option is nuclear energy. For energy security, India is relying on its unique three stage nuclear power program, architectured by Homi J. Bhabha, the father of India's nuclear power program. Nuclear Fuel Complex, Hyderabad, is playing a pivotal role in India's nuclear power program for fabrication of various nuclear fuels and structurals of zircaloy, D9, incoloy, SS and other strategic materials. Also, NFC is gearing up for ever increasing targets through expansion of existing plants and establishing new ones. Successful completion of every year's annual targets of NFC is inevitable for running all reactors. In this success, NFC Maintenance is playing a very critical role in incessantly running NFC. Global players like L and T, Sandvik-Asia and many others are entrusting NFC with new and 'first time in India' orders for which innumerable modifications have to be carried out in the existing infrastructure for effective execution of the same. For in-time completion of mammoth projects of NFC and to keep its flag high, role of NFC maintenance is expressly of profound importance. The paper discusses contribution of NFC for India's three stage nuclear power program and role of maintenance in accomplishing NFC targets. The paper highlights major contributions of maintenance in smooth functioning of NFC. These include upkeep of machines, improvement of productivity, saving in cost of new equipment's by undertaking revamping, modernization and automation. NFC implements latest design analysis techniques for root cause analysis, maintaining product accuracy of all machines/equipment/units. At NFC, maintenance is also contributing to energy conservation mission by implementing energy audits and energy efficient techniques. (author)

  15. Corrosion studies and mechanical tests on metallic materials for the design of packagings for vitrified high level wastes (HLW)

    International Nuclear Information System (INIS)

    Selective corrosion studies and mechanical tests were performed on various metallic materials in order to find out appropriate materials for the design of HLW packagings serving as a barrier in the repository. Besides the Cr-Ni steels to be used as the canister material according to the previous concept, three nickel-base alloys were investigated as well as a mild steel and the materials Ti 99.8-Pd and the steel X 1 CrMoTi 182 (ELA-ferrite). Since within the framework of accident studies for a repository in a rock salt, salt solutions are being considered as a potential corrosion medium, the corrosion tests were made under selecting conditions in a quinary salt solution at 1700C and 1 bar. The mechanical tests have shown that the nickel-base alloys and most of the Cr-Ni steels undergo slight changes of their mechanical properties as a result of welding and annealing treatment, respectively. This provided the evidence for the suitability of these materials as canister materials. Owing to their high susceptibility to local corrosion attacks, above all to stress corrosion, Cr-Ni steels cannot take over the function of a barrier against brines. The materials Inconel 625, Incoloy 825, ELA ferrite and Ti-Pd resisted stress corrosion, but these materials are susceptible to pitting corrosion. High resistance to local corrosion attacks was exhibited by the materials Hastelloy C 4 and mild steel. After a period of testing of 432 days Hastelloy C 4 showed a uniform corrosion rate of less than 0.03 mm/a. The corrosion rate for mild steel 1.0566 was less than 0.25 mm/a in case of the initial material and less than 0.5 mm/a in case of the annealed material. (orig.)

  16. Welding Techniques for In-Pile Instrumentation at INR Pitesti

    International Nuclear Information System (INIS)

    Welding is widely involved in developing in-pile instrumentation, regarding both attaching sensors and joining components of the in-pile experimental devices. Although new methods were not developed, we combined existing techniques, aiming to master them and to ensure reproducibility, sensitivity, robustness, and fast response (for sensor welding). In this paper we present results regarding thermocouple welded on cladding, and dissimilar joints. Techniques for welding K-type thermocouples on cladding were designated to ensure robust joints able to resist during pressure and temperature transients, to minimize the perturbation on temperature field and on the cladding material structure, to reproduce the joint shape and dimension, and to ensure fast response. Welded or brazed joints between materials with properties suitable to nuclear applications are often required for in-pile instrumentation to allow coupling to other components of the experimental devices. The paper describes work related to Zircaloy-to-Stainless Steel joint through eutectic vacuum brazing. The eutectic was produced by thermal diffusion of the elements in the base materials at their interface, through formation of liquid phases of Zr-Fe and Zr-Ni systems at the Zy-SS interface. Clean, reduced roughness, oxide-free contact interface, and precise temperature control were aimed. The optimal interface was the joint on perfectly matching truncated cone surfaces. Heating to 1030oC was obtained through vacuum induction at 10 kHz frequency, with vacuum between 10-5 and 10-6 mbar. Other dissimilar joints were micro-TIG welds on SS capillary connected to Inconel-Incoloy capsules, for high pressure measurement lines with small internal volume. To evaluate the joints burst tests (for thermocouple welding only) polarized light metallography, macrography, backscattered electron tests, Helium leak test, tensile test, and thermal cycling tests were performed. (author)

  17. The effect of test atmosphere on the formation and propagation of creep cracks in commercial high temperature alloys

    International Nuclear Information System (INIS)

    The influence of surface cracks caused by corrosion on the creep rupture properties of the alloys INCOLOY 800H and INCONEL 617 has been investigated for temperatures in the range 1073 K to 1223 K. The test environments were air and impure helium simulating the primary coolant gas of a high temperature reactor (HTR helium). The depths of surface cracks in creep test specimens were measured metallographically and a characteristic crack depth, a90, was derived. a90 is defined so that 90% of the cracks present have depths below a90. The dependence of a90 on test time, creep strain and stress was examined. The growth of creep cracks at the specimen surface as a function of the creep strain was described analytically. This allowed the stress increase due to loss of specimen cross section by surface crack formation to be estimated. It was shown that the surface cracks resulting from corrosion lead to an increase in the creep rate at creep strains above 5%, but the increases were similar in both atmospheres. Rupture of the specimens occurred when the surface cracks and voids developed inside the specimen due to the creep damage processes. This is the main reason for the similar creep rupture properties of the alloys in the two test environments. Finally, a method has been developed to allow the plotting of the depth of surface cracks caused by corrosion with the stress-rupture curves. In this type of diagram, the damage resulting from surface cracks can be related to the creep rupture data to indicate whether corrosion effects need to be considered in the derivation of design stresses. (orig./IHOE)

  18. Determination of an instability temperature for alloys in the cooling gas of a high temperature reactor

    International Nuclear Information System (INIS)

    High temperature alloys designed to be used for components in the primary circuit of a helium cooled high temperature nuclear reactor show massive CO production above a certain temperature, called the instability temperature T/sub i/, which increases with increasing partial pressure of CO in the cooling gas. At p/sub CO/ = 15 microbar, T/sub i/ lies between 900 and 950 degrees C for the four alloys under investigation: T/sub i/ is lowest for the iron base alloy Incoloy 800 H and increases for the nickel base alloys in the order Inconel 617, HDA 230 and Nimonic 86. Measurements of T/sub i/ made at 3 different laboratories were compared and shown to agree for p/sub CO/25 microbar, compatible with CO production by a reaction of Cr2O3 with carbides. Some measurements of T/sub i/ on HDA 230 and Nimonic 86 were performed in the course of simulated reactor disturbances. They showed that the oxide layer looses its protective properties above T/sub i/. A highlight of the examinations was the detection of eta-carbides (M6C) with unusual properties. M6C is the only type of carbide occuring in HDA 230. An eta-carbide with a lattice constant of 1088.8 pm had developed at the surface of Nimonic 86 during pre-oxidation before the disturbance simulation. Its composition is estimated at Ni3SiMo2C. Eta-carbides containing Si and especially eta-carbides with lattice constants as low as 1088.8 pm have been described only rarely until now. (author)

  19. High Temperature Fuel Cladding Chemical Interactions Between TRIGA Fuels and 304 Stainless Steel

    Energy Technology Data Exchange (ETDEWEB)

    Perez, Emmanuel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Keiser, Jr., Dennis D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Forsmann, Bryan [Boise State Univ., ID (United States); Janney, Dawn E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Henley, Jody [Idaho National Lab. (INL), Idaho Falls, ID (United States); Woolstenhulme, Eric C. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-02-01

    High temperature fuel cladding chemical interactions (FCCI) between TRIGA (Training, Research, Isotopes, General Atomics) fuel elements and the 304 stainless steel (304SS) are of interest to develop an understanding of the fuel behavior during transient reactor scenarios. TRIGA fuels are composed of uranium (U) particles dispersed in a zirconium-hydride (Zr-H) matrix. In reactor, the fuel is encased in 304-stainless-steel (304SS) or Incoloy 800 clad tubes. At high temperatures, the fuel can readily interact with the cladding, resulting in FCCI. A number of FCCI can take place in this system. Interactions can be expected between the cladding and the Zr-H matrix, and/or between the cladding and the U-particles. Other interactions may be expected between the Zr-H matrix and the U-particles. Furthermore, the fuel contains erbium-oxide (Er-O) additions. Interactions can also be expected between the Er-O, the cladding, the Zr-H and the U-particles. The overall result is that very complex interactions may take place may as a result of fuel and cladding exposures to high temperatures. This report discusses the characterization of the baseline fuel microstructure in the as-received state (prior to exposure to high temperature), characterization of the fuel after annealing at 950C for 24 hours and the results from diffusion couple experiments carries out at 1000C for 5 and 24 hours. Characterization was carried out via scanning electron microscopy (SEM) and transmission electron microscopy (TEM) with sample preparation via focused ion beam in situ-liftout-technique.

  20. Materials technology for coal-conversion processes. Sixteenth quarterly report, October--December 1978

    Energy Technology Data Exchange (ETDEWEB)

    Ellingson, W A

    1978-01-01

    Refractories for slag containment, nondestructive evaluation methods, corrosion, erosion, and component failures were studied. Analysis of coal slags reveal ferritic contents of 18 to 61%, suggesting a partial pressure of 0/sub 2/ in the slagging zone of approx. 10/sup -2/ to 10/sup -4/ Pa. A second field test of the high-temperature ultrasonic erosion-monitoring system was completed. Ultrasonic inspecton of the HYGAS cyclone separator shows a reduced erosive-wear rate at 5000 h in the stellite region. The acoustic leak-detection system for valves was field tested using a 150-mm-dia. valve with a range of pressures from 0.34 to 4.05 MPa. Results suggest a linear relation between detected rms levels and leak rates. Studies on acoustic emissions from refractory concrete continued with further development of a real-time data acquisition system. Corrosion studies were conducted on Incoloy 800, Type 310 stainless steel, Inconel 671 and U.S. Steel Alloy 18-18-2 (as-received, thermally aged, and preexposed for 3.6 Ms to multicomponent gas mixtures). Results suggest a decrease in ultimate tensile strength and flow stress after preexposure. Examination of commercial iron- and nickel-base alloys after 100-h exposures in atmospheric-pressure fluidized-bed combustors suggests that the addition of 0.3 mole % CaCl/sub 2/ to the fluidized bed has no effect on the corrosion behavior of these materials; however, 0.5 mole % NaCl increased the corrosion rate of all materials. Failure-analysis activities included (1) the design and assembly of thermowells (Haynes Alloy 188 and slurry-coated Type 310 stainless steel) and (2) examination of components from the Synthane boiler explosion, the IGT Steam--Iron Pilot Plant, the HYGAS Ash Agglomerating Gasifier, and the Westinghouse Coal Gasification PDU.

  1. Evaluation of candidate Stirling engine heater tube alloys after 3500 hours exposure to high pressure doped hydrogen or helium. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Misencik, J.A.; Titran, R.H.

    1984-10-01

    Sixteen commercial tubing alloys were endurance tested at 820/sup 0/ C, 15 MPa in a diesel-fuel fired Stirling engine simulator materials test rig: iron-base N-155, A-286, Incoloy 800, 19-9DL, CG-27, W-545, 12RN72, 253MA, Sanicro 31H and Sanicro 32; nickel-base Inconel 601, Inconel 625, Inconel 718, Inconel 750 and Pyromet 901; and cobalt-base HS-188. The iron-nickel alloys CG-27 and Pyromet 901 exhibited superior oxidation/corrosion resistance to the diesel-fuel combustion products and surpassed the design criterias' 3500 h creep-rupture endurance life. Three other alloys, Inconel 625, W-545, and 12RN72, had creep-rupture failures after 2856, 2777, and 1598 h, respectively. Hydrogen permeability coefficients determined after 250 h of rig exposure show that Pyromet 901 had the lowest Phi value, 0.064x10/sup -6/ cm/sup 2//s MPa/sup 1///sup 2/. The next five hairpin tubes, CG-27, Inconel 601, Inconel 718(wd), Inconel 750, and 12RN72(cw) all had Phi values below 0.2x10/sup -6/ more than a decade lower than the design criteria. Based upon its measured high strength and low hydrogen permeation, CG-27 was selected for 3500 h endurance testing at 21 MPa gas pressure and 820/sup 0/C. Results of the high pressure, 21 MPa, CG-27 endurance test demonstrated that the 1.0 vol % C0/sub 2/ dopant is an effective deterrent to hydrogen permeation. The 21 MPa hydrogen gas pressure apparent permeability coefficient at 820/sup 0/C approached 0.1x10/sup -6/ cm/sup 2/sec MPa/sup 1///sup 2/ after 500 hr, the same as the 15 MPa test. Even at this higher gas pressure and comparable permeation rate, CG-27 passed the 3500 hr endurance test without creep-rupture failures. It is concluded that the CG-27 alloy, in the form of thin wall tubing is suitable for Stirling engine applications at 820/sup 0/C and gas pressures up to 21 MPa.

  2. Analytical and experimental investigation of passive thermal sensors for core catcher of AHWR

    International Nuclear Information System (INIS)

    In a core melt accident of nuclear reactor, the molten core must be cooled to avoid damage to load-bearing structures of the containment and activity spread to the soil, water and environment. To deal with such accidents a robust passive Core Catcher System (CCS) is being designed for Advanced Heavy Water Reactor (AHWR). In post Fukushima scenario, incorporation of core catcher system in all advanced nuclear reactors has become mandatory. The main objectives of the core catcher system are retention of the melt in the cavity, reduction of the volumetric heat generation by adding sacrificial material, quenching the molten mass in shortest possible time and to provide long term cooling for stabilization of the melt. To make CCS, fully passive safety system, an innovative Passive Water Injection (PAWAN) system has been recently designed and developed. The PAWAN senses the core melt accident and injects water from Gravity Driven Water Pool (GDWP) passively into the core catcher. A mesh of Passive Thermal Sensors (PTS) will be deployed for sensing the core melt condition. This paper primarily deals with the design, development and testing of PTS, its time response analysis and layout of effective sensor placement in case of partial or full core melt accident. The passive thermal sensor works on the principle of thermal expansion of fluid. It consists of a temperature sensing bulb, pressurized filled fluid system and a connecting capillary tube to the passive valve actuator. The fluid pressure inside the thermal sensing bulb changes with change in surrounding temperature. A helium gas filled passive thermal sensor made of INCOLOY-800H material has been designed for the high temperature and radioactive environment. The passive thermal sensors were developed in-house and have been successfully tested in furnace at 600℃. The repeatability of these sensors has been checked at 600℃ and found to be very satisfactory.The PAWAN is designed to be robust and sensitive enough to

  3. Dosimetric impact evaluation of primary coolant chemistry of the internal tritium breeding cycle of a fusion reactor DEMO

    International Nuclear Information System (INIS)

    Tritium will be responsible for a large fraction of the environmental impact of the first generation of DT fusion reactors. Today, the efforts of conceptual development of the tritium cycle for DEMO are mainly centred in the so called Inner Breeding Tritium Cycle, conceived as guarantee of reactor fuel self-sufficiency. The EU Fusion Programme develops for the short term of fusion power technology two breeding blanket conceptual designs both helium cooled. One uses Li-ceramic material (HCPB, Helium-Cooled Pebble Bed) and the other a liquid metal eutectic alloy (Pb15.7Li) (HCLL, Helium-Cooled Lithium Lead). Both are Li-6 enriched materials. At a proper scale designs will be tested as Test Blanket Modules in ITER. The tritium cycles linked to both blanket concepts are similar, with some different characteristics. The tritium is recovered from the He purge gas in the case of HCPB, and directly from the breeding alloy through a carrier gas in HCLL. For a 3 GWth self-sufficient fusion reactor the tritium breeding need is few hundred grams of tritium per day. Safety and environmental impact are today the top priority design criteria. Dose impact limits should determine the key margins and parameters in its conception. Today, transfer from the cycle to the environment is conservatively assumed to be operating in a 1-enclosure scheme through the tritium plant power conversion system (intermediate heat exchangers and helium blowers). Tritium loss is caused by HT and T2 permeation and simultaneous primary coolant leakage through steam generators. Primary coolant chemistry appears to be the most natural way to control tritium permeation from the breeder into primary coolant and from primary coolant through SG by H2 tritium flux isotopic swamping or steel (EUROFER/INCOLOY) oxidation. A primary coolant chemistry optimization is proposed. Dynamic flow process diagrams of tritium fluxes are developed ad-hoc and coupled with tritiated effluents dose impact evaluations. Dose

  4. Cleaning the magnesium oxide contaminated stainless steel system using a high temperature decontamination process

    International Nuclear Information System (INIS)

    nickel came into solution in the proportion in which it exists in stainless steel whereas the amount of chromium was low. This was attributed to the difficulty in maintaining a good reducing coolant chemistry conditions during the two months of operation of the system. The results of the visual and microscopic examination of the structural materials viz. carbon steel, Incoloy-800, SS-316, zircaloy and monel-400 exposed to NTA-N2H4 formulation are described in this paper. The application of this process for the decontamination of stainless steel and other chromium containing systems is also described. (authors)

  5. Dosimetric impact evaluation of primary coolant chemistry of the internal tritium breeding cycle of a fusion reactor DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Velarde, M. [Instituto de Fusion Nuclear (DENIM), ETSII, Universidad Politecnica Madrid UPM, J. Gutierrez Abascal 2, Madrid 28006 (Spain); Sedano, L. A. [Asociacion Euratom-Ciematpara Fusion, Av. Complutense 22, 28040 Madrid (Spain); Perlado, J. M. [Instituto de Fusion Nuclear (DENIM), ETSII, Universidad Politecnica Madrid UPM, J. Gutierrez Abascal 2, Madrid 28006 (Spain)

    2008-07-15

    Tritium will be responsible for a large fraction of the environmental impact of the first generation of DT fusion reactors. Today, the efforts of conceptual development of the tritium cycle for DEMO are mainly centred in the so called Inner Breeding Tritium Cycle, conceived as guarantee of reactor fuel self-sufficiency. The EU Fusion Programme develops for the short term of fusion power technology two breeding blanket conceptual designs both helium cooled. One uses Li-ceramic material (HCPB, Helium-Cooled Pebble Bed) and the other a liquid metal eutectic alloy (Pb15.7Li) (HCLL, Helium-Cooled Lithium Lead). Both are Li-6 enriched materials. At a proper scale designs will be tested as Test Blanket Modules in ITER. The tritium cycles linked to both blanket concepts are similar, with some different characteristics. The tritium is recovered from the He purge gas in the case of HCPB, and directly from the breeding alloy through a carrier gas in HCLL. For a 3 GWth self-sufficient fusion reactor the tritium breeding need is few hundred grams of tritium per day. Safety and environmental impact are today the top priority design criteria. Dose impact limits should determine the key margins and parameters in its conception. Today, transfer from the cycle to the environment is conservatively assumed to be operating in a 1-enclosure scheme through the tritium plant power conversion system (intermediate heat exchangers and helium blowers). Tritium loss is caused by HT and T{sub 2} permeation and simultaneous primary coolant leakage through steam generators. Primary coolant chemistry appears to be the most natural way to control tritium permeation from the breeder into primary coolant and from primary coolant through SG by H{sub 2} tritium flux isotopic swamping or steel (EUROFER/INCOLOY) oxidation. A primary coolant chemistry optimization is proposed. Dynamic flow process diagrams of tritium fluxes are developed ad-hoc and coupled with tritiated effluents dose impact evaluations

  6. Status update of the BWR cask simulator

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    The performance of commercial nuclear spent fuel dry storage casks are typically evaluated through detailed numerical analysis of the system's thermal performance. These modeling efforts are performed by the vendor to demonstrate the performance and regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have moved the storage location from above ground to below ground and significantly increased the maximum thermal load allowed in a cask in part by increasing the canister helium pressure. Previous cask performance validation testing did not capture these parameters. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern dry casks. These modern cask designs utilize elevated helium pressure in the sealed canister or are intended for subsurface storage. The BWR cask simulator (BCS) has been designed in detail for both the above ground and below ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below ground storage configurations

  7. Test Plan for the Boiling Water Reactor Dry Cask Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-11-01

    The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the

  8. Liquid Salt Heat Exchanger Technology for VHTR Based Applications

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, Mark; Sridhara, Kumar; Allen, Todd; Peterson, Per

    2012-10-11

    The objective of this research is to evaluate performance of liquid salt fluids for use as a heat carrier for transferring high-temperature process heat from the very high-temperature reactor (VHTR) to chemical process plants. Currently, helium is being considered as the heat transfer fluid; however, the tube size requirements and the power associated with pumping helium may not be economical. Recent work on liquid salts has shown tremendous potential to transport high-temperature heat efficiently at low pressures over long distances. This project has two broad objectives: To investigate the compatibility of Incoloy 617 and coated and uncoated SiC ceramic composite with MgCl2-KCl molten salt to determine component lifetimes and aid in the design of heat exchangers and piping; and, To conduct the necessary research on the development of metallic and ceramic heat exchangers, which are needed for both the helium-to-salt side and salt-to-process side, with the goal of making these heat exchangers technologically viable. The research will consist of three separate tasks. The first task deals with material compatibility issues with liquid salt and the development of techniques for on-line measurement of corrosion products, which can be used to measure material loss in heat exchangers. Researchers will examine static corrosion of candidate materials in specific high-temperature heat transfer salt systems and develop an in situ electrochemical probe to measure metallic species concentrations dissolved in the liquid salt. The second task deals with the design of both the intermediate and process side heat exchanger systems. Researchers will optimize heat exchanger design and study issues related to corrosion, fabrication, and thermal stresses using commercial and in-house codes. The third task focuses integral testing of flowing liquid salts in a heat transfer/materials loop to determine potential issues of using the salts and to capture realistic behavior of the salts in a

  9. TRIGA - LEU cluster with 36 fuel elements

    International Nuclear Information System (INIS)

    Designing the TRIGA - LEU fuel cluster is part of the mechanical design of TRIGA reactor core. The latter is supported by a square frame (11 x 12 132 meshes) accommodating the 35 fuel clusters. The TRIGA fuel cluster is designed to incorporate 36 fuel elements with 3/8 inch diameter allowing the pins to be arranged into a 6 x 6 matrix. The final mechanical design of reactor zone resulted into a cluster of squared cross section with 87.5 mm side and 88.9 mm separation between the centers of the clusters. This cluster was designed by preserving the dimensions and configuration of fuel clusters with 25 elements. By the positioning of the pins inside the cluster one obtains: - a fuel element protection by reducing the failure risks; - delimitation of fixed channel of the cooling flow for each cluster; - a convenient means of manipulation; - a correct water flow for cooling the pins in a fixed channel by preserving the surface of cooling channels from the 25 fuel element cluster. The cluster has the following principal components: - casing; - bottom plug or adapter; - upper plug for maneuvering; - spacer for fuel elements. The cluster casing is made of aluminium with square cross section of 87.5 mm side and is provided at the lower part with an aluminium adapter allowing its insertion in the reactor core frame. This piece is designed to support the ends of the 36 fuel elements in a blocked position. The fuel elements are subject to asymmetric temperature distribution flux conditions, hence an asymmetric temperature distribution results concomitantly with a symmetrical (about 0.8 mm) swelling of the Incoloy 800 can. Also bending of the fuel element occurs which will be limited by the intermediate spacer. At the casing upper part an aluminium upper plug or handle is mounted allowing cluster maneuvering by means of a special tool. The cluster is provided with lateral holes in its upper part ensuring the necessary cooling water flow in case the upper part of the cluster

  10. Corrosion in Supercritical carbon Dioxide: Materials, Environmental Purity, Surface Treatments, and Flow Issues

    Energy Technology Data Exchange (ETDEWEB)

    Sridharan, Kumar; Anderson, Mark

    2013-12-10

    The supercritical CO{sub 2} Brayton cycle is gaining importance for power conversion in the Generation IV fast reactor system because of its high conversion efficiencies. When used in conjunction with a sodium fast reactor, the supercritical CO{sub 2} cycle offers additional safety advantages by eliminating potential sodium-water interactions that may occur in a steam cycle. In power conversion systems for Generation IV fast reactors, supercritical CO{sub 2} temperatures could be in the range of 30°C to 650°C, depending on the specific component in the system. Materials corrosion primarily at high temperatures will be an important issue. Therefore, the corrosion performance limits for materials at various temperatures must be established. The proposed research will have four objectives centered on addressing corrosion issues in a high-temperature supercritical CO{sub 2} environment: Task 1: Evaluation of corrosion performance of candidate alloys in high-purity supercritical CO{sub 2}: The following alloys will be tested: Ferritic-martensitic Steels NF616 and HCM12A, austenitic alloys Incoloy 800H and 347 stainless steel, and two advanced concept alloys, AFA (alumina forming austenitic) steel and MA754. Supercritical CO{sub 2} testing will be performed at 450°C, 550°C, and 650°C at a pressure of 20 MPa, in a test facility that is already in place at the proposing university. High purity CO{sub 2} (99.9998%) will be used for these tests. Task 2: Investigation of the effects of CO, H{sub 2}O, and O{sub 2} impurities in supercritical CO{sub 2} on corrosion: Impurities that will inevitably present in the CO{sub 2} will play a critical role in dictating the extent of corrosion and corrosion mechanisms. These effects must be understood to identify the level of CO{sub 2} chemistry control needed to maintain sufficient levels of purity to manage corrosion. The individual effects of important impurities CO, H{sub 2}O, and O{sub 2} will be investigated by adding them

  11. Flux entrapment and Titanium Nitride defects during electroslag remelting

    Science.gov (United States)

    Busch, Jonathan D.

    Electroslag remelted (ESR) ingots of INCOLOY alloys 800 and 825 are particularly prone to macroscale slag inclusions and microscale cleanliness issues. Formation of these structures near the ingot surface can cause significant production yield losses (˜10%) due to the necessity of extensive surface grinding. Slag inclusions from near the outer radius of the toe end of alloy 800 and 825 ingots were found to be approximately 1 to 3 mm in size and have a multiphase microstructure consisting of CaF2, CaTiO3, MgAl 2O4, MgO and some combination of Ca12Al14 O32F2 and/or Ca12Al14O 33. These inclusions were often surrounded by fields of 1 to 10 μm cuboidal TiN particles. A large number of TiN cuboids were observed in the ESR electrode with similar size and morphology to those observed surrounding slag inclusions in the ESR ingots, suggesting that the TiN particles are relics from ESR electrode production process. Samples taken sequentially throughout the EAF-AOD processes showed that the TiN cuboidals that are found in ESR ingots form between tapping the AOD vessel into the AOD ladle and the casting of ESR electrodes. Analysis of slag skin at various heights of alloy 825 ingots revealed that the phase fraction of CaF2 decreased, whereas TiCaO 3 and Ca12Al14O32F2 increased, from toe to head. The observed increase in TiO2 content suggests that at most a two-fold increase in viscosity of the slag would be expected. Similar analysis of alloy 800 ingots did not reveal significant trends in slag skin composition, possibly due to differences in ingot geometry or the presence of Al toe additions during the remelting of alloy 800. Directional solidification experiments were conducted to determine the solidification sequences of two common ESR slags: Code 316 (33% CaF2, 33% CaO, and 33% Al2O3) and Code 59 (50% CaF2, 20% CaO, 22% Al2O3, 5% MgO, and 3% TiO2). In both cases the changes in slag phase fraction as a function of solidification time were not as significant as predicted

  12. Next Generation Nuclear Plant Intermediate Heat Exchanger Materials Research and Development Plan (PLN-2804)

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright

    2008-04-01

    application in heat exchangers and core internals for the NGNP. The primary candidates are Inconel 617, Haynes 230, Incoloy 800H and Hastelloy XR. Based on the technical maturity, availability in required product forms, experience base, and high temperature mechanical properties all of the vendor pre-conceptual design studies have specified Alloy 617 as the material of choice for heat exchangers. Also a draft code case for Alloy 617 was developed previously. Although action was suspended before the code case was accepted by ASME, this draft code case provides a significant head start for achieving codification of the material. Similarly, Alloy 800H is the material of choice for control rod sleeves. In addition to the above listed considerations, Alloy 800H is already listed in the nuclear section of the ASME Code; although the maximum use temperature and time need to be increased.

  13. Next Generation Nuclear Plant Steam Generator and Intermediate Heat Exchanger Materials Research and Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright

    2010-09-01

    application in heat exchangers and core internals for the NGNP. The primary candidates are Inconel 617, Haynes 230, Incoloy 800H and Hastelloy XR. Based on the technical maturity, availability in required product forms, experience base, and high temperature mechanical properties all of the vendor pre-conceptual design studies have specified Alloy 617 as the material of choice for heat exchangers. Also a draft code case for Alloy 617 was developed previously. Although action was suspended before the code case was accepted by ASME, this draft code case provides a significant head start for achieving codification of the material. Similarly, Alloy 800H is the material of choice for control rod sleeves. In addition to the above listed considerations, Alloy 800H is already listed in the nuclear section of the ASME Code; although the maximum use temperature and time need to be increased.

  14. Performance evaluation and characterization of metallic bipolar plates in a proton exchange membrane (PEM) fuel cell

    Science.gov (United States)

    Hung, Yue

    Bipolar plate and membrane electrode assembly (MEA) are the two most repeated components of a proton exchange membrane (PEM) fuel cell stack. Bipolar plates comprise more than 60% of the weight and account for 30% of the total cost of a fuel cell stack. The bipolar plates perform as current conductors between cells, provide conduits for reactant gases, facilitate water and thermal management through the cell, and constitute the backbone of a power stack. In addition, bipolar plates must have excellent corrosion resistance to withstand the highly corrosive environment inside the fuel cell, and they must maintain low interfacial contact resistance throughout the operation to achieve optimum power density output. Currently, commercial bipolar plates are made of graphite composites because of their relatively low interfacial contact resistance (ICR) and high corrosion resistance. However, graphite composite's manufacturability, permeability, and durability for shock and vibration are unfavorable in comparison to metals. Therefore, metals have been considered as a replacement material for graphite composite bipolar plates. Since bipolar plates must possess the combined advantages of both metals and graphite composites in the fuel cell technology, various methods and techniques are being developed to combat metallic corrosion and eliminate the passive layer formed on the metal surface that causes unacceptable power reduction and possible fouling of the catalyst and the electrolyte. The main objective of this study was to explore the possibility of producing efficient, cost-effective and durable metallic bipolar plates that were capable of functioning in the highly corrosive fuel cell environment. Bulk materials such as Poco graphite, graphite composite, SS310, SS316, incoloy 800, titanium carbide and zirconium carbide were investigated as potential bipolar plate materials. In this work, different alloys and compositions of chromium carbide coatings on aluminum and SS316

  15. Performance evaluation and characterization of metallic bipolar plates in a proton exchange membrane (PEM) fuel cell

    Science.gov (United States)

    Hung, Yue

    Bipolar plate and membrane electrode assembly (MEA) are the two most repeated components of a proton exchange membrane (PEM) fuel cell stack. Bipolar plates comprise more than 60% of the weight and account for 30% of the total cost of a fuel cell stack. The bipolar plates perform as current conductors between cells, provide conduits for reactant gases, facilitate water and thermal management through the cell, and constitute the backbone of a power stack. In addition, bipolar plates must have excellent corrosion resistance to withstand the highly corrosive environment inside the fuel cell, and they must maintain low interfacial contact resistance throughout the operation to achieve optimum power density output. Currently, commercial bipolar plates are made of graphite composites because of their relatively low interfacial contact resistance (ICR) and high corrosion resistance. However, graphite composite's manufacturability, permeability, and durability for shock and vibration are unfavorable in comparison to metals. Therefore, metals have been considered as a replacement material for graphite composite bipolar plates. Since bipolar plates must possess the combined advantages of both metals and graphite composites in the fuel cell technology, various methods and techniques are being developed to combat metallic corrosion and eliminate the passive layer formed on the metal surface that causes unacceptable power reduction and possible fouling of the catalyst and the electrolyte. The main objective of this study was to explore the possibility of producing efficient, cost-effective and durable metallic bipolar plates that were capable of functioning in the highly corrosive fuel cell environment. Bulk materials such as Poco graphite, graphite composite, SS310, SS316, incoloy 800, titanium carbide and zirconium carbide were investigated as potential bipolar plate materials. In this work, different alloys and compositions of chromium carbide coatings on aluminum and SS316