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Sample records for alloy-901 incoloy

  1. Risks and benefits of Incoloy 908

    International Nuclear Information System (INIS)

    The central solenoid and toroidal field coils of ITER are designed with cable-in-conduit Nb3Sn superconducting conductors jacketed with Incoloy 908. This material, developed to limit the critical current degradation of the Nb3Sn strands, shows excellent mechanical properties at 4 K, particularly under cycling. Nevertheless, it is a crack prone material when heat treated in the presence of oxygen, which requires careful control of the atmosphere during manufacture. A programme of extensive mechanical testing of samples and manufacture of a pancake with incoloy conductor jacket was carried out in Europe. The origin of cracks in the incoloy jacket after heat treatment was investigated

  2. 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.

  3. 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.

  4. 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.

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

    Czech Academy of Sciences Publication Activity Database

    Cárach, J.; Hloch, Sergej; Hlaváček, Petr; Vasilko, K.; Lehocká, D.

    Zagreb: Croatian Association of Production Engineering, 2015 - (Abele, E.; Udiljak, T.; Ciglar, D.), s. 77-80 ISBN 978-953-7689-03-2. [CIM 2015 - International Scientific Conference on Production Engineering. Vodice (HR), 10.06.2015-13.06.2015] R&D Projects: GA MŠk(CZ) LO1406; GA MŠk ED2.1.00/03.0082 Institutional support: RVO:68145535 Keywords : incoloy alloy 925 * abrasive water jet turning * traverse speed Subject RIV: JQ - Machines ; Tools

  6. Pitting of Incoloy 800 in presence of CuII

    International Nuclear Information System (INIS)

    The pitting behaviour of Incoloy 800 in presence of CuII ions, at 60 degrees C and 280 degrees C was studied by long term exposition of specimens in aqueous cupric chloride solutions. At 60 degrees C experiments were performed in aerated (7 ppm O2) and deaerated solutions containing 500, 1000, 2000, 10000 and 20000 ppm CuCl2. At 280 degrees C experiments were performed in deaerated 20 ppm, 50 ppm and 100 ppm CuCl2 solutions. During each experiment the open circuit potential of the alloy was measured as a function of time. After corrosion test the specimens were examined by Scanning Electron Microscopy for the presence of pits. In another set of experiments potentiodynamic anodic polarization curves were used to determine the pitting potential of Incoloy 800 in deaerated NaCl solutions at chloride concentrations and pH values corresponding to those possessed by solutions containing 20 ppm to 20000 ppm CuCl2. At 60 degrees C pitting was observed in those solutions where the CuCl2 concentration is higher than 1000 ppm. At 280 degrees C pitting was found in the specimens exposed to those solutions where the CuCl2 concentration was higher than 20 ppm. (author). 3 refs

  7. Study of the susceptibility at SCC of Incoloy-800 in alkaline solution

    International Nuclear Information System (INIS)

    Incoloy-800, used to manufacture the steam generator tubes, has a good general corrosion resistance, but under certain conditions stress corrosion cracking (SCC) is observed. SCC can occur on both the primary and secondary sides of steam generator. Typical location for SCC is at dents, in the roll expansion regions of the tube and in areas containing concentrated impurities. Because under normal conditions the SCC mechanism requires a long time for initiation, it is necessary to use the accelerated corrosion tests to study the susceptibility of Incoloy-800 at caustic SCC. The mode of stressing of Incoloy-800 tubes used in our experiments was C-ring. To evaluate the stress and deformation induced in Incoloy-800 rings the ANSYS code was used. By using the cathodic zone of the potentiodynamic curves, the most important electrochemical parameters were calculated: the corrosion current (Icorr), the corrosion rate (vcorr) and the polarization resistance. The tested samples were examined by using the metallographic method. (author)

  8. 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.

  9. Oxidation behavior of Incoloy 800 under simulated supercritical water conditions

    Energy Technology Data Exchange (ETDEWEB)

    Fulger, M. [Institute for Nuclear Research Pitesti, POB 78, Campului Street, No. 1, 115400 Mioveni (Romania)], E-mail: manuela.fulger@nuclear.ro; Ohai, D.; Mihalache, M.; Pantiru, M. [Institute for Nuclear Research Pitesti, POB 78, Campului Street, No. 1, 115400 Mioveni (Romania); Malinovschi, V. [University of Pitesti, Research Center for Advanced Materials, Targul din Vale Street, No. 1, 110040 Pitesti (Romania)

    2009-03-31

    For a correct design of supercritical water-cooled reactor (SCWR) components, data regarding the behavior of candidate materials in supercritical water are necessary. Corrosion has been identified as a critical problem because the high temperature and the oxidative nature of supercritical water may accelerate the corrosion kinetics. The goal of this paper is to investigate the oxidation behavior of Incoloy 800 exposed in autoclaves under supercritical water conditions for up to 1440 h. The exposure conditions (thermal deaerated water, temperatures of 723, 773, 823 and 873 K and a pressure of 25 MPa) have been selected as relevant for a supercritical power plant concept. To investigate the structural changes of the oxide films, X-ray diffraction (XRD), scanning electron microscopy (SEM), energy dispersive X-ray spectrometry (EDX) and electrochemical impedance spectroscopy (EIS) analyses were used. Results show changes in the oxides chemical composition, microstructure and thickness versus testing conditions (pressure, temperature and time). The oxide films are composed of two layers: an outer layer enriched in Fe oxide and an inner layer enriched in Cr and Ni oxides corresponding to small cavities supposedly due to internal oxidation.

  10. 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.

  11. 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.

  12. 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.

  13. 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.).

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

    Czech Academy of Sciences Publication Activity Database

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

    -, 11 July 2015 (2015), s. 1-6. ISSN 0268-3768 R&D Projects: GA MŠk ED2.1.00/03.0082; GA MŠk(CZ) LO1406 Institutional support: RVO:68145535 Keywords : incoloy alloy 925 * abrasivewater jet turning * traverse speed * surface roughness Subject RIV: JQ - Machines ; Tools Impact factor: 1.458, year: 2014 http://link.springer.com/article/10.1007/s00170-015-7489-0

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

    Czech Academy of Sciences Publication Activity Database

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

    2016-01-01

    Roč. 82, č. 9 (2016), s. 1747-1752. ISSN 0268-3768 R&D Projects: GA MŠk ED2.1.00/03.0082; GA MŠk(CZ) LO1406 Institutional support: RVO:68145535 Keywords : incoloy alloy 925 * abrasive water jet turning * traverse speed Subject RIV: JQ - Machines ; Tools Impact factor: 1.458, year: 2014 http://link.springer.com/article/10.1007%2Fs00170-015-7489-0

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

    International Nuclear Information System (INIS)

    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

  17. 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.

  18. 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

  19. 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

  20. 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.

  1. On the significance of a subsequent ageing after cold working of Incoloy 800 at operational temperatures

    International Nuclear Information System (INIS)

    The influence of cold working and subsequent ageing at operational temperatures on the long-term and short-term mechanical properties of components made from the iron-nickel-chromium base alloy Incoloy 800 are discussed. Long-term properties are time-to-rupture strengths, which are included in the design code, over a lifetime of 300,000 hours. For LWR operating temperatures of 350oC, this is of minor importance. An operating temperature of 550oC is possible for Incoloy 800 with up to 25% cold working and a subsequent solution annealing at 950oC, without loss of time-to-rupture strength compared with the 'as received' state. The short-term mechanical properties are strongly influenced by cold working, in the form of increasing yield strength and rupture strength, and decreasing ductility and consequently loss in impact energies. A subsequent ageing at 550oC leads to a decrease of the yield strength and rupture strength, and an increase of ductility as well as the impact energies. The environmental influence are discussed. (author) 3 figs., 1 tab., 8 refs

  2. 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.

  3. 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)

  4. Dislocation dynamics in the oxide dispersion strengthened alloy INCOLOY MA956

    International Nuclear Information System (INIS)

    In situ straining experiments in a high-voltage electron microscope have been performed on the oxide dispersion strengthened alloy INCOLOY MA 956 at room temperature and between 640 and 1010 C to study the dynamic behaviour of dislocations. Macroscopic compression experiments including stress relaxation tests have been carried out between room temperature and 900 C. The dynamic behaviour of dislocations, the flow stress and its strain rate sensitivity are discussed in the different temperature ranges in terms of long-range dislocation interactions, the Orowan mechanism, the model of the thermally activated detachment of dislocations from oxide particles, solution hardening and a diffusional point defect drag. It is concluded that the latter controls the thermally activated dislocation processes in a wide temperature range around about 700 C. The detachment model may be active only above 900 C. (orig.)

  5. A Study of the Oscillation Marks' Characteristics of Continuously Cast Incoloy Alloy 825 Blooms

    Science.gov (United States)

    Saleem, Saud; Vynnycky, Michael; Fredriksson, Hasse

    2016-06-01

    A comprehensive experimental study of oscillation mark (OM) formation and its characteristics during the solidification of Incoloy alloy 825 in the continuous casting of blooms is investigated by plant trials and metallographic study. The experiments involved two heats with the same casting and mold conditions and sampling at different locations across the strand. The metallographic study combined macro/micro-examinations of OMs and segregation analysis of Cr, Mn, Mo, Ni, and Si by microprobe analysis. The results show that OMs have widely different characteristics, such as mark type, depth, segregation, and accompanying microstructure. Furthermore, the mark pitch can vary considerably even for the similar casting conditions, leading to different conditions for the marks' formation in relation to the mold's cyclic movement. Finally, a mechanism for the OM formation is discussed and proposed. Possible solutions for minimizing the observed defects by optimizing the mold conditions are suggested.

  6. Analyses of oxide films grown on AISI 304L stainless steel and Incoloy 800HT exposed to supercritical water environment

    Science.gov (United States)

    Fulger, Manuela; Mihalache, Maria; Ohai, Dumitru; Fulger, Stefan; Valeca, Serban Constantin

    2011-08-01

    Supercritical water (SCW) is being considered as a cooling medium for the next generation nuclear reactors because it provides high thermal efficiency and plant simplification. However, materials corrosion has been identified as a critical problem due to the oxidative nature of supercritical water. Thus, for safety using of these nuclear reactor systems a systematic study of candidate materials corrosion is needed. As in other high temperature environments, corrosion in SCW occurs by the growth of an oxide layer on the materials surface. The current work aims to evaluate oxidation behavior of AISI 304L SS and Incoloy 800HT in water at supercritical temperatures in the range 723-873 K under a pressure of 25 MPa for up to 1680 h. After exposure to deaerated supercritical water, the samples were investigated using gravimetry, scanning electron microscopy (SEM), energy dispersive X-ray spectroscopy (EDS), X-ray photoelectron spectroscopy (XPS), and electrochemical impedance spectroscopy (EIS). Oxide films grown on these materials have a layered structure with an outer layer consisting of a mixture of iron oxide/iron-nickel spinel oxides and an inner layer consisting of chromium oxide in the case of Incoloy 800HT and nickel-chromium spinel oxide in the case of AISI 304L SS. The mass gains for Incoloy 800HT at all temperatures were small, while comparatively with AISI 304L SS which exhibited higher oxidation rates. In the same time the results obtained by EIS indicate the best corrosion resistance of oxides grown on Incoloy 800HT surface.

  7. Analyses of oxide films grown on AISI 304L stainless steel and Incoloy 800HT exposed to supercritical water environment

    Energy Technology Data Exchange (ETDEWEB)

    Fulger, Manuela, E-mail: manuela.fulger@nuclear.ro [Institute for Nuclear Research Pitesti, POB 78, Campului Street, No. 1, 115400 Mioveni (Romania); Mihalache, Maria; Ohai, Dumitru [Institute for Nuclear Research Pitesti, POB 78, Campului Street, No. 1, 115400 Mioveni (Romania); Fulger, Stefan [University Politechnica Bucharest, Splaiul Independentei Street, No. 313, Bucharest 060042 (Romania); Valeca, Serban Constantin [University of Pitesti, Targul din Vale Street, No. 1, 110040 Pitesti (Romania)

    2011-08-15

    Supercritical water (SCW) is being considered as a cooling medium for the next generation nuclear reactors because it provides high thermal efficiency and plant simplification. However, materials corrosion has been identified as a critical problem due to the oxidative nature of supercritical water. Thus, for safety using of these nuclear reactor systems a systematic study of candidate materials corrosion is needed. As in other high temperature environments, corrosion in SCW occurs by the growth of an oxide layer on the materials surface. The current work aims to evaluate oxidation behavior of AISI 304L SS and Incoloy 800HT in water at supercritical temperatures in the range 723-873 K under a pressure of 25 MPa for up to 1680 h. After exposure to deaerated supercritical water, the samples were investigated using gravimetry, scanning electron microscopy (SEM), energy dispersive X-ray spectroscopy (EDS), X-ray photoelectron spectroscopy (XPS), and electrochemical impedance spectroscopy (EIS). Oxide films grown on these materials have a layered structure with an outer layer consisting of a mixture of iron oxide/iron-nickel spinel oxides and an inner layer consisting of chromium oxide in the case of Incoloy 800HT and nickel-chromium spinel oxide in the case of AISI 304L SS. The mass gains for Incoloy 800HT at all temperatures were small, while comparatively with AISI 304L SS which exhibited higher oxidation rates. In the same time the results obtained by EIS indicate the best corrosion resistance of oxides grown on Incoloy 800HT surface.

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

    Directory of Open Access Journals (Sweden)

    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.

  9. Stress corrosion cracking of 316 SS and Incoloy-800 in high temperature aqueous containing sulfate and chloride

    International Nuclear Information System (INIS)

    The stress corrosion cracking (SCC) susceptibility of 316 stainless steel (SS) which was welded for primary pipe and Incoloy-800 (shotpeening) for steam generator (SG) tube have been investigated by means of a slow strain rate test (SSRT) at a strain rate of 4.2 x 10-6 /s. Tests were conducted at 315 degrees C for 316 SS and 270 degrees C for In-800 in the oxygenated simulated resin intrusion environment (acidic sulfate). Tests of the effect of combination of SO4-2 and Cl- on SCC of Incoloy-800 were also carried out. The results indicate that Incoloy-800 is unsusceptible to SCC either in the environment with SO4-2 (from a few ppm to 1000 ppm, pH 3 ∼ 4) or in the environment of combination of SO4-2 (1000 ppm) and Cl- (from 2 to 1000 ppm). The 316 NG SS is susceptible to transgranular stress corrosion cracking (TGSCC) in the resin intrusion environment with SO4-2 in high temperature water

  10. 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)

  11. Why incoloy 800 was selected as tube material for the new steam generators of Doel 3

    International Nuclear Information System (INIS)

    Incoloy 800 or Inconel 690. This was the basic question for the SG of DOEL 3, to be replaced in 1993. A comparative evaluation was made, focused on the service behaviour, more particularly the corrosion resistance. It was concluded that both materials may be considered for SG replacement. They are indeed substantially superior to alloy 600, but none of them is really better than the other one: - they are both nearly completely immune to primary water stress corrosion cracking; - they are both susceptible to some types of attack from the secondary side. To the 690 credit are the globally better performances in the secondary side corrosion tests. However if the aggressive environments which are unlikely to occur at DOEL 3 are discarded, it appears that the margin between 690 and 800 is small. To the 800 credit is the favourable service experience except the secondary side wastage due to phosphate conditioning. 800 was finally selected not only because of favourable economic arguments, but also because of the important weight attributed to the service behaviour by the Belgian Utilities

  12. Experiments with tubes made of INCOLOY 800H under uniaxial and multiaxial loading conditions

    International Nuclear Information System (INIS)

    In order to verify the transferability of materials data and constitutive equations in the high temperature region, uniaxial and multiaxial creep tests were performed with tubes made of X10 NiCrAlTi 32 20 (INCOLOY 800H). The tubes were loaded by a combination of tension, internal pressure and torsion. The temperature was 9500C. The life-time of the heat exchanger tubes under pure uniaxial tension was found to be longer than the time to rupture of bar material made from the same parent heat. The heat exchanger tubes show a significant creep hardening in the primary creep range and a later onset of the secondary creep range. For the heat exchanger tubes under different multiaxial loading conditions different life times were measured, although in all cases the same deviatoric stress (v. Mises) was used. For the experiments presented in this paper these differences can be explained by constitutive equations. The deformation of the heat exchanger tubes under multiaxial loadings cannot be predicted by the postulated mathematical model. In the experiments with combined loading conditions the tubes failed by leak formation and fracture respectively - a spontaneous failure was never found. (orig.)

  13. 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.

  14. 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.)

  15. Dynamic recrystallization behavior and constitutive analysis of Incoloy 901 under hot working condition

    International Nuclear Information System (INIS)

    Hot deformation behavior of Incoloy 901 Ni-based superalloy was investigated by performing hot compression tests over the temperature range of 950–1100 °C and at strain rates of 0.001–1 s−1. The flow curves at low strain rates i.e. 0.001–0.1 s−1, were characterized by a faint peak or a long plateau after the work hardening region. Only at high strain rates, e.g. 1 s−1, the flow curves exhibited sharp peaks of dynamic recrystallization. The strain associated with the peak points of flow curves firstly increases with strain rate (up to 0.01 s−1) and then unexpectedly decreased at higher strain rates. These anomalous results were attributed to the change in the kinetics of dynamic softening mechanisms. The constitutive analysis of flow stress suggested a change in the hot deformation behavior of the alloy with increasing temperature (over 1050 °C) or strain rate (over 0.01 s−1). Microstructural characterization showed that although dynamic recrystallization would operate over the whole studied ranges of deformation temperature and strain rate, its kinetics was generally sluggish. However, the rate of recrystallization increased remarkably at temperatures beyond 1050 °C or strain rate over 0.01 s−1. Electron backscattered diffraction (EBSD) measurements and optical microscopy observations showed that the nucleation of recrystallization around the grain boundaries or second-phase particles should be due to the gradual evolution of subgrains

  16. 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

  17. Evolution of deformation and annealing textures in Incoloy 800H/HT via different rolling paths and strains

    International Nuclear Information System (INIS)

    In this report, we characterize the deformation and annealing textures of Incoloy 800H/HT, following different rolling conditions that produced different textures in this material. Incoloy 800H/HT is an austenitic Fe–Ni super alloy and is considered to be a candidate material for Gen IV nuclear reactors. Fossil fuel plants have used this alloy for decades; however, as grain structure and texture parameters can strongly affect its physical and mechanical properties in-service, engineers should consider some structural modifications before using this alloy in nuclear reactors. In this study, we used Thermo-Mechanical Processing (TMP) to alter the texture of Incoloy 800H/HT. We applied various thickness reductions (10%, 30%, 50%, 70%, and 90%) to this alloy using two different rolling paths, followed by annealing. Our detailed study of the deformation and annealing texture evolution shows that upon different rolling paths, the final deformation texture and the annealing texture were different. Brass texture was the dominant component for the uni-directional rolled (UDR) samples, while a combination of brass (B) and ND-rotated brass (BT) were the dominant components in the cross-rolled (CR) samples. Annealing textures of UDR samples were mainly Goss, copper, S, recrystallized brass ({236}〈385〉) and minor copper twin {552}〈115〉 components. At lower deformations (<50%), the annealed CR samples showed tilted cube, S, recrystallized brass ({236}〈385〉) and minor Goss twin (113)〈33¯2〉. However, at higher rolling reductions, the B+BT deformation texture was retained for the CR samples

  18. 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

  19. The effect of chlorine on the high temperature corrosion of SiO2-coated Incoloy 800H

    OpenAIRE

    Haanappel, V.; Van Corbach, H.; Hofman, R.; Fransen, T.; Gellings, P.

    1993-01-01

    The effect of chlorine in coal gasification atmospheres was investigated on the high temperature corrosion of Incoloy 800H coated with SiO2. The experiments were performed in a representative gas mixture between 450 and 750°C. The weight gain in this environment was significantly decreased with coating thicknesses to about 1.4 µm. Larger thicknesses led to an increased corrosion attack. The coating contained small cracks filled with iron-nickel sulphides. Probably, pre-existing cracks in the ...

  20. Corrosion resistance of incoloy 825 as a structural material for apparatus used in the denitrification of reprocessing waste solutions with formic acid

    International Nuclear Information System (INIS)

    In order to test the corrosion resistance of incoloy 825 as a strural matetial for a denitrification vessel for use in the reprocessing of spent nuclear fuels, corrosion experiments with this material were carried out in boiling formic acid, nitric acid and waste model solutions, using activated material specimens. Corrosion rates in excess of 1 g/m2d were found only in the gaseous phase above formic acid. Provided the filling times with formic acid are kept short, incoloy 825 may be considered a resistant structural material for a denitrification vessel. (orig.)

  1. 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.

  2. 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

  3. The effect of thermo-mechanical processing on grain boundary character distribution in Incoloy 800H/HT

    International Nuclear Information System (INIS)

    In this study, we applied a thermo-mechanical process to alter the grain boundary characteristic distribution (GBCD) with a view to the feasibility of grain boundary engineering in Incoloy 800H/HT. In order to optimize the GBCD through increasing the low Σ coincidence-site lattice (CSL) boundaries, we applied various thickness reductions with two different rolling modes followed by annealing. We used Electron Backscattered Diffraction (EBSD) to analyze the GBCD and CSL boundaries. We found that the coincidence-site lattice boundaries, particularly Σ3 and its variants, increased with the pre-deformation level in cross-rolled (CR) samples. In contrast, the fraction of these CSL boundaries had an optimum in 50% reduction for unidirectionally rolled (UDR) samples. In fact, different Σ3n interactions led to different GBCD in UDR and CR processed samples. Low to medium deformation with UDR and medium to high deformation with CR on the Incoloy 800H/HT samples showed potential for grain boundary engineering

  4. 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

  5. Electrochemical and surface characterisation of oxide films on nano-grain nickel films electrodeposited on INCOLOY-800

    International Nuclear Information System (INIS)

    Nano materials have different properties from the corresponding bulk materials because of fine grain size, large fraction of surface atoms, high surface energy and high grain boundary volume fraction. For similar reasons, the nano-alloy coatings show superior high-temperature corrosion resistance and are generally more resistant to stress corrosion cracking. Hence, it is of interest to know the materials performance, if the structural materials used in nuclear reactors are made of nano-grains. In Indian PHWRs, Incoloy-800 is being used as the steam generator tubing material. It's corrosion resistance property is very important as it forms not only the pressure boundary between the radioactive primary water and non-active secondary water but also from the view point of loss of heavy water, in case of any corrosion damage. In this paper, the corrosion resistance of the oxide films formed on nano-grain nickel film electrodeposited on Incoloy-800 (a) in the presence of saccharine (WS) and (b) in the absence of saccharine (WOS) were compared with that formed on Commercial Ni foil, using electrochemical dc polarization and ac impedance techniques. The surface morphology, elemental analysis and grain size were studied with SEM, EDX and XRD techniques respectively. The nano-grain nickel films were prepared on Incoloy-800 by electrodeposition using Watt's Bath with saccharine sodium as a surfactant. The oxide films were developed by exposing them to LiOH solution (pH-10.0) at 245 deg C for 3 days (A-group) and 7 days (B-group). XRD results showed that the grain size of Ni formed in the absence of saccharine (WOS) was ∼ 60 nm and did not change after being autoclaved. But, for Ni formed in the presence of saccharine (WS), the grain size was ∼ 16 nm which increased to 40-50 nm after being autoclaved. With both A and B-group specimens, the PDAP curves showed an active-passive transition, a passive region and a transpassive region in 2N H2SO4. However, the critical current

  6. 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.

  7. 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)

  8. The Effects of Hf, Y and Zr Additions on the Oxidation Behavior of Ni-Fe based Incoloy 825 at High Temperature

    International Nuclear Information System (INIS)

    Superalloys are protected by making a protective Cr2O3 or Al2O3 layer on the surface when exposed to an aggressive environment at elevated temperature. Recently, studies on the effects of oxidation behavior by the addition of the oxygen active elements in the superalloys have been proceeded, because it improved protectiveness of oxide scales, providing good oxidation resistance to high temperature materials. In this study, the oxidation behavior and morphological features of isothermally and cyclically tested Incoloy 825 with 1 wt% of Hf, 0.5wt% of Zr, and Y additions were investigated in air at 1000 .deg. C and 1100 .deg. C. The OAE added Incoloy 825 alloys showed better oxidation resistance than that without addition. Especially Y added alloy showed the best cyclic oxidation resistance among the alloys tested

  9. Results from investigations with an instrumented impact machine on a molybdenum base alloy, nickel base alloys, and Incoloy 800

    International Nuclear Information System (INIS)

    Experiments were performed on the molybdenum base alloy TZM, the nickel base alloys Nimocast 713 LC, Inconel 625, Nimonic 86, Hastelloy S, and the iron base alloy Incoloy 800 with an instrumented impact machine. The results are discussed in terms of absorbed impact energies and dynamic fracture toughness. In all cases the agreement between the energy determined by the dial reading and the energy determined by the integration of the load vs. load point displacement diagram was excellent. A procedure for the determination of the dynamic fracture toughness for load vs. load point displacement diagrams exhibiting high oscillations using an averaged curve is proposed. Using this procedure a pronounced influence of the experiments with tup and chisel (5.0 m/s and 0.1 m/s respectively) on the dynamic fracture toughness is not detectable. Using half the drop height, i.e. halving the total energy, lowers the dynamic fracture toughness values for these types of alloys. Low absorbed impact energies are often combined with high fracture toughness values. In these cases there is no or only a small reserve in deformation and/or stable crack growth. (Auth.)

  10. 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)

  11. 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.)

  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. Experimental research regarding the corrosion of incoloy-800 and SA 508 cl.2 in the CANDU steam generator

    International Nuclear Information System (INIS)

    Steam generators (SGs) are crucial components of pressurized water reactors. The failure of the steam generator as a result of tube degradation by corrosion has been a major cause of Pressurized Water Reactor (PWR) plant unavailability. Steam generator problems have caused major economic losses in terms of lost electricity production through forced unit outages and, in cases of extreme damage, as additional direct cost for large-scale repair or replacement of steam generators. Steam generator tubes are susceptible to failure by a variety of mechanisms, the vast majority of which are related a corrosion. The feedwater that enters into the steam generators under normal operating conditions is extremely pure, but nevertheless contains low levels (generally in the μg/l concentration range) of impurities such as iron, chloride, sulphate, silicate, etc. When water is converted to steam and exits the steam generator, the non-volatile impurities are left behind. As a result, their concentration in the bulk steam generator water is considerably higher than those in the feedwater. Nevertheless, the concentrations of corrosive impurities are still generally sufficiently low that the bulk water is not significantly aggressive towards steam generator materials. The excellent performance to date of CANDU steam generators can be attributed, in part, to their design and performance characteristics, which typically involve higher recirculation ratios and lower heat fluxes and temperatures. The purpose of this paper consists in assessment of generalized corrosion behaviour of the tubes materials (Incoloy-800) and tubesheet material (carbon steel SA 508 cl.2) at the normal secondary circuit parameters (temperature-260 deg C, pressure-5.1MPa). The testing environment was the demineralized water without impurities, at pH=9.5 regulated with morpholine and ciclohexilamine (all volatile treatment - AVT). The results are presented like micrographies and graphics representing loss of metal

  14. Study of the polarization for Incoloy 800 and for the stainless stell AISI 304 in mixtures of Iron, Nickel and Chromium Chlorides

    International Nuclear Information System (INIS)

    Polarization curves for the Incoloy 800 and for the stainless stell AISI 304 were obtained with static and rotational electrodes. The electrolytes employed showed growing concentrations of mixtures as Iron, Nickel and Chromium Chlorides their proportion being the same as the content of these elements in the respective alloys. The alloys under investigation exhibited a continuous transition behaviour from the passive to the active-passive and to the active conditions. Also, the pH was found the main parameter controlling the anodic behaviour of the alloy. (Author)

  15. Surface residual stress distributions in as-bent Inconel 600 U-bend and Incoloy 800 90-degree bend tubing samples

    International Nuclear Information System (INIS)

    Selected data showing typical macroscopic residual stress distributions in U-bent Inconel 600 and 90 degrees bends in Incoloy 800 are presented. The results indicate regions of both high magnitude tension and compression in the longitudinal direction around the circumference of the bends at the apex. The microscopic residual stress, or percent plastic strain and macroscopic residual distributions in the surface of cross-roll straightened and ground Inconel 600 tubing are described. The results indicate a compressive surface layer accompanied by a yield strength gradient from 90 ksi at the surface to 30 ksi at a depth of 0.003 in

  16. The corrosion behaviour (stress corrosion and intergranular corrosion) of the tube materials inconel 600 and incoloy 800 for nuclear steam generators in treated secondary cycle water during tube burst tests in the temperature range from 270 to 3500C

    International Nuclear Information System (INIS)

    Surface removal, intergranular corrosion and stress corrosion of the materials No. 2.4640 (Inconel 600) and No. 1.4558 (Incoloy 800) were investigated in solutions of NaOH, Na2HPO4 and Na3PO4 whose concentrations were about 10,000 higher than those given in the guidelines. The test parameters varied were the alkalising agent and the pH value and, in addition, temperature (270, 310 and 3500C); the possible influence of low temperature sensitation (for 1000 and 10,000 hours) was also investigated. (orig.)

  17. Deposition of silica coatings on Incoloy 800H substrates using a high power laser

    International Nuclear Information System (INIS)

    Using a 5kW CO2 laser, thin silica coatings have been deposited on steeply inclined Inccoloy 800H substrates traversed beneath the incident laser beam. The small angle of incidence (10deg-15deg) of the beam on the substrate resulting from the large angle of incline gave reduced substrate heating and eliminated melting of the substrate surface, but allowed melting of the silica powder injection into the focus of the beam. The focus was positioned above the substrate and the molten powder was allowed to fall onto the laser-heated substrate below. By making overlapping passes, complete surface coverage was achieved over a large area, the coating thickness being 2-3μm; this overlap filled points of surface roughness to give the component a microscopically smooth outer appearance. The silica coating formed a good bond with the metallic substrate; adhesion appeared to be improved by having a slightly rough finish rather than a highly polished one. Resistance to sulphidation attach was assessed by placing coated samples in a furnace containing a mixture of gases as in a simulated coal gasifer heat exchanger atmosphere at 450 and 750degC. Sulphidation resistance was greatly improved from that of the untreated alloy and the coating did not spall or crack for the period of the test; the few sulphides observed were probably formed at small discontunuities in the coating which may be eliminated by applying a second silica coating. (orig.)

  18. Optimization of INCOLOY alloy 800 mechanical properties for various power plant requirements

    International Nuclear Information System (INIS)

    The AMSE Boiler Code development of design stresses and their optimization for alloys 800 and 800H for conventional and nuclear power plants have coincided with many successful trial installations of these grades of alloy 800. These trial installations, along with laboratory tests, have shown that alloy 800H can be used for long times with a retention of good mechanical properties, including ductility. While gamma prime can be formed, it soon loses its detrimental effect on ductility in service. Sensitization of the alloy also occurs but it, too, has a decreasing effect on corrosion resistance with time in service, especially at elevated temperatures. The authors discuss all of these aspects and conclude that alloy 800 perhaps with low carbon control in the annealed (1800 to 1950degF)(982 to 1066degC) condition gives the optimum combination of properties to 1050degF(566degC) and alloy 800H (.05 to .10%C) solution annealed at 2100 to 2200degF (1149 to 1204degC) has optimum properties above 1050degF (566degC). The effects of aluminium and titanium and the benefit of keeping these within the ASTM limits of .15 to .60% are also stressed. (author)

  19. 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.)

  20. 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

  1. 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

  2. Multiaxial creep of tubes from Incoloy 800 H and Inconel 617 under static and cyclic loading conditions

    International Nuclear Information System (INIS)

    At temperatures above 8000C the material behaviour under mechanical load is determined by creep. The service of heat exchanging components leads to multiaxial loading conditions. For design and inelastic analysis of the component behaviour time dependent design values and suitable constitutive equations are necessary. The present report gives a survey of the approaches to describing creep under multiaxial loading. Norton's law and v. Mises' theory are applied. The load combinations of internal pressure, tensile and torsional stress are studied more closely, cyclic stress superposition in the tensile-pulsating range is discussed and cases of partial relaxation are examined. Experimental results are presented for the loading conditions discussed, and satisfactory agreement between theory and experiment has been found up to now for these results. Regarding lifetime determination under multiaxial creep load, a more precise analysis of creep damage is presented suggesting a suitable deviatoric stress for evaluation in the long-time range. (orig.)

  3. Niobium-3-tin internally cooled cabled superconductor (ICCS) technology I

    International Nuclear Information System (INIS)

    Recent work, using tantalum and Incoloy 903 sheathing has demonstrated that internally cooled, cabled superconductors (ICCS) can be compacted without current degradation. This paper compares Inconel 617 sheathing to previous results. Inconel 617 is a practical engineering material with a thermal contraction lying between stainless steel and Incoloy 903

  4. NIOBIUM-3-TIN INTERNALLY COOLED CABLED SUPERCONDUCTOR (ICCS) TECHNOLOGY I

    OpenAIRE

    Hoenig, M.; Steeves, M.; Cyders, C.

    1984-01-01

    Recent work, using tantalum and Incoloy 903 sheathing has demonstrated that internally cooled, cabled superconductors (ICCS) can be compacted without current degradation. This paper compares Inconel 617 sheathing to previous results. Inconel 617 is a practical engineering material with a thermal contraction lying between stainless steel and Incoloy 903.

  5. 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)

  6. 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.

  7. 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

  8. 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...

  9. Measuring corrosion rates in crevices with a computer-controlled flow loop

    International Nuclear Information System (INIS)

    The corrosion rate produced by galvanic coupling within a crevice is studied using a computer controlled flow loop. The materials studied include: 1018 carbon steel, stainless steel-405, stainless steel-409, stainless steel-347, and Incoloy 800. A description of the test facilities and electronic equipment is included

  10. 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)

  11. Study on resistance to intergranular corrosion of heat transfer incology800H tube

    International Nuclear Information System (INIS)

    By using GB/T15260-94 B method, i.e., copper - copper sulfate -16% sulfuric acid evaluates intergranular corrosion sensitivity of nickel-based alloys, the intergranular corrosion test results of domestic heat transfer Incoloy800H tube of High-Temperature Gas-cool Reactor(HTR) nuclear power plant demonstration project are studied under different test conditions, and compared to the intergranular corrosion performance of imported Incoloy800H alloy tube. The results show that the main factors to influence the alloy resistance to intergranular corrosion are C and Ti contents, and the sensitivity increases with increasing of C content. The effective measure to prevent intergranular corrosion is to add Ti element to reduce the sensitivity to intergranular corrosion. (authors)

  12. 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

  13. 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.)

  14. 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.

  15. Low Cost Silicon Solar Array Project. Feasibility of Low-cost, High-volume Production of Silane and Pyrolysis of Silane to Semiconductor-grade Silicon

    Science.gov (United States)

    Breneman, W. C.; Farrier, E. G.; Morihara, H.

    1978-01-01

    The presence of copper promotes a more rapid approach to the steady stete operating condition and results in a more consistent reactor effluent composition. The average kinetic and equilibrium yield are unchanged. Incoloy has been identified as the preferred choice of material of construction for the hydrogenation reactor although certain metallurgical changes were noted in samples exposed to the H2/HCl atmosphere at 500 C which indicate the need for more testing.

  16. A new experimental facility for the investigation of hydrogen isotopes permeation through heat exchanger tubes with sweep gas intermediate circuit under nuclear conditions

    International Nuclear Information System (INIS)

    In this work a new experimental facility is presented to study the permeation of hydrogen isotopes (tritium and protium) through heat exchanger tubes with sweep gas intermediate circuit under process gas conditions. Beside temperature and pressure dependant measurements to determine the permeation data of Incoloy 800 H double wall tubes further parameter investigations have been made to study the action of a sweep gas circuit against hydrogen permeation. The experimental results show that the obtained impending factor could be destinctly increased. (orig.)

  17. Compatibility of heat resistant alloys with boron carbide, 5

    International Nuclear Information System (INIS)

    This paper includes an experimental result of out-of-pile compatibility and capsule design for irradiation test in Japan Materials Testing Reactor (JMTR). The compatibility between sheath material and neutron absorber materials for control rod devices (CRD) was examined for potential use in a very high temperature reactor (VHTR) which is under development at JAERI. The purpose of the compatibility tests are preliminary evaluation of safety prior to irradiation tests. Preliminary compatibility evaluation was concerned with three items as follows : 1) Lithium effects on the penetrating reaction of Incoloy 800H alloy in contact with a mixture of boronated graphite and lithium hydroxide powders, 2) Short term tensile properties of Incoloy 800H and Hastelloy XR alloy reacted with boronated graphite and fracture mode analysis, 3) Reaction behavior of both alloys under transient power conditions of a VHTR. It was clear that the reaction rate constant of the Incoloy 800H alloy was accelerated by doping lithium hydroxide into the boron carbide and graphite powder. The mechanical properties of Incoloy 800H and Hastelloy XR alloy reacted with boronated graphite were decreased. Ultimate tensile strength and tensile ductilities at temperatures over 850 deg C were reduced, but there was no change in the proof (yield) stress. Both alloys exhibited a brittle intergranular fracture mode during transient power conditions of a VHTR and also exhibited severe penetration. Irradiation capsules for compatibility test were designed to simulate three irradiation conditions of VHTR: 1) steady state for VHTR, 2) Transient power condition, 3) Service limited life of CRD. Capsule irradiation experiments have been carried out satisfactorily and thus confirm the validity of the capsule design procedure. (author)

  18. Welding techniques for corrosion-resistant steels and nickel-based alloys

    International Nuclear Information System (INIS)

    The paper briefly describes the welding techniques of TIG welding for large wall thicknesses, submerged arc welding, manual electrode welding and RES strip cladding, and exemplifies specific material-component-related welding tasks by welding performed with material 1.4876, Incoloy 800-H (welding of a finned tube), material 2.4663 (manual electrode welding), NiCr23Co12Mo, and with the materials 1.4876H and 2.4663 (orbital-v-TIG welding of tubes). (DG)

  19. 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

  20. Fracture mechanics analyses with PERMAS for high temperature applications

    International Nuclear Information System (INIS)

    Laterally notched CT specimens and tubes with cracks have been investigated by extensive finite-element analyses in order to determine the decisive fracture mechanics characteristics. Test results on creep crack growth were available for the CT specimens and fatigue crack growth results for tubes. The experiments were carried out on the austenitic alloy X10NiCrAlTi 32 20 (INCOLOY 800H) at temperatures between 550deg C and 800deg C. (orig./MM)

  1. Fracture mechanics analyses with PERMAS for high temperature applications

    Energy Technology Data Exchange (ETDEWEB)

    Altes, J.; Koschmieder, D. (Forschungszentrum Juelich GmbH (Germany). Inst. fuer Nukleare Sicherheitsforschung); Beckers, H. (Gesellschaft fuer Strukturanalyse mbH, Aachen (Germany))

    1990-01-01

    Laterally notched CT specimens and tubes with cracks have been investigated by extensive finite-element analyses in order to determine the decisive fracture mechanics characteristics. Test results on creep crack growth were available for the CT specimens and fatigue crack growth results for tubes. The experiments were carried out on the austenitic alloy X10NiCrAlTi 32 20 (INCOLOY 800H) at temperatures between 550deg C and 800deg C. (orig./MM).

  2. Investigation of Irradiated RPV and ODS Steels by SANS and Residual Stress (Hungary)

    International Nuclear Information System (INIS)

    We have investigated a RPV steel 15H2MFA type irradiated by different fluency by SANS and reactor pipe weld and cladding by residual stress. In cooperation were investigated the aging of other reactor materials as Incoloy 800 and 304L treated at different temperature. Using the SANS technique, the phase separation in the oxide dispersion strengthened (ODS) ferritic steels (PM2000 and MA956) were investigated. (author)

  3. Effect of secondary cycle sulphuric acid injection on steam generator tubes

    International Nuclear Information System (INIS)

    Certain steam generator secondary side tube failures have been attributed to the presence of acid sulphate environment. The presence of sulphates in the secondary loop may be the result of the ingress of cationic resins generating sulphate, sulphur and protons due to decomposition, of the injection of sulphuric acid used as cationic resin regenerative agent, or of the incorporation of small quantities of sulphur-bearing species in the secondary side water, concentrating in the crevices and eventually reaching significant concentrations. The Inconel 690 TT and Incoloy 800 now being used as replacement materials for Inconel 600 in the latest steam generators provide better behaviour than this latter material with respect to the corrosion problems typically encountered on the secondary side of steam generators. In the case on Incoloy 800, this improved behaviour is linked to the design of the steam generators, which avoids zones in which impurities might accumulate. The experimental work described was aimed at gaining insight into the behaviour of the materials replacing Inconel 600, Incoloy 800 SP and Inconel 690 TT, in environment containing sulphates and with acid pH values. (authors). 4 figs., 5 refs

  4. Corrosion of heat exchanger materials under heat transfer conditions

    International Nuclear Information System (INIS)

    Severe pitting has occurred in moderator heat exchangers tubed with Incoloy-800 in Pickering Nuclear Generating Station. The pitting originated on the cooling side (outside) of the tubes and perforation occurred in less than two years. It was known from corrosion testing at CRNL that Incoloy-800 was not susceptible to pitting in Lake Ontario water under isothermal conditions. Corrosion testing with heat transfer across the tube wall was carried out, and it was noted that severe pitting could occur under deposits formed on the tubes in silty Lake Ontario water. Subsequent testing, carried out in co-operation with Ontario Hydro Research Division, investigated the pitting resistance of other candidate tubing alloys: Incoloy-825, 904 L stainless steel, AL-6X, Inconel-625, 70:30 Cu:Ni, titanium, Sanicro-30 and Sanicro-281. Of these, only titanium and Sanicro-28 have not suffered some degree of pitting attack in silt-containing Lake Ontario Water. In the absence of silt, and hence deposits, no pitting took place on any of the alloys tested

  5. Mechanical properties and corrosion behavior of materials exposed to an experimental, atmospheric fluidized-bed combustor

    International Nuclear Information System (INIS)

    A joint materials test program developed by the Institute for Mining and Minerals Research (IMMR) and the Tennessee Valley Authority (TVA) involved the postexposure mechanical properties and corrosion behavior of candidate structural materials in an experimental, atmospheric fluidized-bed combustor (AFBC). This combustor was operated by Accurex Corporation at Research Triangle Park, North Carolina, under the direction of TVA. The materials studied were Type 304, Type 310, and INCOLOY alloy 800 in the form of disc coupons with and without crevice configurations. Type 304 was also used for mechanical property measurements. The alloys were exposed to the combustor environment at about8400C for approximately 330 hours. The ranking in terms of decreasing weight loss was: (1) Type 304, (2) Type 310, and (3) INCOLOY alloy 800. The presence of tight crevices did not enhance the corrosion rate. In addition, the corrosion rates, based on the weight loss (typically 1 to 6 mpy), indicated that the alloys performed reasonably well when considering materials wastage. However, optical microscopy observations showed intergranular corrosion penetration in INCOLOY alloy 800 and Type 304. The mechanical properties of Type 304 were inferior to the unexposed alloy. A comparison of the data obtained from the combustor-exposed 304ss tensile samples with data from control samples exposed in vacuum to a similar thermal history indicated that the chemistry of the AFBC environment did not play a major role in the observed degradation of the mechanical properties

  6. Adhesion of ZrN and SiC thin films on titanium and nickel alloys

    International Nuclear Information System (INIS)

    Chemically inert ceramic coatings are currently being investigated to extend the lifetime of metallic components operating in severe environments. Polycrystalline ZrN and amorphous SiC coatings were deposited by cathodic arc evaporation and by PACVD, respectively, on Titanium Grade 12 and Incoloy 825 metal substrates. The structure of the coatings was verified by SEM and XRD. Residual stress analyses were performed on the ZrN coatings via XRD using the sin2 φ method. Compressive stresses of 3.7 GPa and 2.5 GPa were calculated in the ZrN on the Incoloy and Titanium substrates, respectively. Studies of the interfacial chemistry via AES revealed chemically abrupt interfaces. Scratch tests were employed to assess the critical load for interfacial failure and fracture mechanisms for the various coating systems. Critical loads, characterized by a kidney-shaped crack patterns found in the scratch tracks, occurred at 12 N for ZrN on Titanium and 20 N for ZrN on Incoloy. Interfacial failure of SiC on Titanium was dominated by brittle fracture of the SiC coating at 3N loads

  7. Stress corrosion cracking of candidate materials for nuclear waste containers

    International Nuclear Information System (INIS)

    Types 304L and 316L stainless steel (SS), Incoloy 825, Cu, Cu-30%Ni, and Cu-7%Al have been selected as candidate materials for the containment of high-level nuclear waste at the proposed Yucca Mountain Site in Nevada. The susceptibility of these materials to stress corrosion cracking has been investigated by slow-strain-rate tests (SSRTs) in water which simulates that from well J-13 (J-13 water) and is representative of the groundwater present at the Yucca Mountain site. The SSRTs were performed on specimens exposed to simulated J-13 water at 93 degree C and at a strain rate 10-7 s-1 under crevice conditions and at a strain rate of 10-8 s-1 under both crevice and noncrevice conditions. All the tests were interrupted after nominal elongation strains of 1--4%. Examination by scanning electron microscopy showed some crack initiation in virtually all specimens. Optical microscopy of metallographically prepared transverse sections of Type 304L SS suggests that the crack depths are small (<10 μm). Preliminary results suggest that a lower strain rate increases the severity of cracking of Types 304L and 316L SS, Incoloy 825, and Cu but has virtually no effect on Cu-30%Ni and Cu-7%Al. Differences in susceptibility to cracking were evaluated in terms of a stress ratio, which is defined as the ratio of the increase in stress after local yielding in the environment to the corresponding stress increase in an identical test in air, both computed at the same strain. On the basis of this stress ratio, the ranking of materials in order of increasing resistance to cracking is: Types 304L SS < 316L SS < Incoloy 825 congruent Cu-30%Ni < Cu congruent Cu-7%Al. 9 refs., 12 figs., 7 tabs

  8. Operation experiences of the steam generator repair technique by means of weldless sleeves

    International Nuclear Information System (INIS)

    Field performances of the weldless PLUSS sleeve up to these days establishes it as a reliable technique for use as tube-in-tube repair of damaged tubing of steam generators customarily served by nuclear reactors. For the functioning of the weldless PLUSS Sleeve the proper combination of tube and sleeve material is decisive. It turned out that for steam generator tubing of Inconel Alloy 600 and Inconel Alloy 690 the sleeve out of Incoloy Alloy 800 serves this purpose perfectly. Via hydraulic expansion the weldless sleeve of Incoloy Alloy 800 is permanently fixed against the parent tube of Inconel Alloys 600/690. Upon relieving the exerted hydraulic pressure, the tube springs back more than the sleeve does, the net effect being that there materializes an interference at the interface between sleeve and tube. The PLUSS sleeve, while spanning the defective zone of the tube, takes the advantage of this phenomenon to secure the sleeve within the tube at ambient temperature. Emergence of the unique expansion joint of the PLUSS sleeve is the result. At higher operating temperatures associated with higher system pressures the expansion joint gets even tighter as the interference between the joint members increases owing to the fact that Inconel Alloys 600/690 and Incoloy Alloy 800 possess different coefficients of thermal expansion. The weldless sleeving technique with favourable stress situation and reduced installation steps compared to other kind of sleeving techniques offers a significant improvement in the field application. At present it is probably the most efficient and cost-effective repair technique of all. The paper discusses the working principle of the product and its successful installation in various nuclear power plants worldwide. It further discusses the field application with respect to a particular utility and operational experiences gained so far. Furthermore, the plant performance is dealt with briefly. (author)

  9. 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.).

  10. Influence of hydrogen oxidation kinetics on hydrogen environment embrittlement

    Science.gov (United States)

    Walter, R. J.; Kendig, M. W.; Meisels, A. P.

    1992-01-01

    Results are presented from experiments performed to determine the roles of hydrogen absorption and hydrogen electron transfer on the susceptibility of Fe- and Ni-base alloys to ambient-temperature hydroen embrittlement. An apparent independence is noted between hydrogen environment embrittlement and internal hydrogen embrittlement. The experiments were performed on Inconel 718, Incoloy 903, and A286. The electrochemical results obtained indicate that Inconel 718 either adsorbs hydrogen more rapidly and/or the electrochemical oxidation of the adsorbed hydrogen occurred more rapidly than in the other two materials.

  11. Production and welding technology of some high-temperature nickel alloys in relation to their properties

    International Nuclear Information System (INIS)

    The most effective matching of alloys to the needs of advanced high-temperature gas-cooled reactors requires not only a knowledge of material properties, but also some understanding of the inherent general characteristics of this type of alloy. Some of the characteristic features of high-temperature nickel-based alloys are explored and general guidelines offered for their most effective use. Examples are drawn from three commercial materials: Inconel alloy 617, Incoloy alloy 800H, and Nimonic alloy 86. Such items as hot and cold working, heat treating, welding, and mechanical properties are considered

  12. 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

  13. 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.

  14. 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

  15. Thermal cycling influence on microstructural characterization of alloys with high nickel content

    International Nuclear Information System (INIS)

    The IV nuclear energy generation systems are aimed at making revolutionary improvements in economics, safety and reliability, and sustainability. To achieve these goals, Generation IV systems will operate at higher temperatures and in higher radiation fields. This paper shows the thermal cycling influences on microstructure and hardness of nickel based alloys: Incoloy 800 HT and Inconel 617. These alloys were meekly at a thermal cycling of 25, 50, 75 and 100 cycles. The temperature range of a cycle was between 400OC and 700OC. Nickel base alloys develop their properties by solid solution and/or precipitation strengthening. (authors)

  16. 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.)

  17. Materials in the primary system of water-cooled nuclear power plants

    International Nuclear Information System (INIS)

    The first part of this document deals with the various phenomena of selective corrosion on austenitic CrNi steels, Inconel 600 and Incoloy 800 under the different conditions of light water reactor operation. The paper focuses particularly on the phenomenon of intergranular stress-corrosion cracking. For this purpose the results of the authors' own corrosion tests are presented and discussed in connection with recent results obtained at other research centers and with operating experience accumulated with boiling water reactors and pressurized water reactors. The switch in steam generator tube material from Inconel 600 to Incoloy 800 is described. In the second part of the paper the problems of cracking during the processing of low-alloyed steel are discussed. The motivation for this analysis was the worldwide occurrence of so-called underclad cracks in the steels ASTM A 508 cl. 2 (22 NiMoCr 37) and ASTM A 533 grade B class 1. It is shown that what is involved here is a specialized form of stress relief cracking, for which this type of steel has a certain sensitivity. Several crack tests of this type are described. It is proposed that the concept of relaxation embrittlement be introduced as the cause of these cracks. More detailed information is presented about this problem. 31 refs., 28 figs., 1 tab

  18. Influence of impurity concentration on corrosion behaviour of the pipe-pipesheet joint in the steam generator

    International Nuclear Information System (INIS)

    Corrosion is a major problem affecting the safe operation of the steam generator. The most problems related to corrosion are due to local concentration of the aggressive species and/or impurities in restrained flow zones as, for instance, in the crevices occurring in pipe-pipesheet joints. The effects of buildup in these zones are important in steam generator designing and operation. The aim of this study was to establish the corrosion mechanism and formation of oxide layers on the carbon steel. The results of corrosion testing were obtained on simulation devices, in operation conditions, specific to the NPP secondary circuits (temperature, 260 deg. C, pressure, 5.1 MPa). These crevice simulating devices were made of carbon steel SA 508 cl.2 and Incoloy 800. The testing media were the following: NaCl solutions (pH=10.5) of 25 g/l, 50, 75 and NaCl 75.5 g/l plus Na2SO4 10 g/l solution (pH=10.5). The behaviour of the two materials was studied by a metallographic method. The results are presented as micrographs evidencing the occurrence of pitting corrosion first on the material of pipesheet (SA 508 cl.2) and in the environment of an excessively aggressive medium and, secondly, on the pipe material (Incoloy 800)

  19. Research experience with the secondary side corrosion of CANDU steam generators

    International Nuclear Information System (INIS)

    All steam generator tube failures lead to transfer of the radioactive materials from the primary coolant circuit to the steam generator secondary circuit, and necessitate downtime to locate and plug the failed tubes. For the particular case of the CANDU plants, any steam generator tube failure results in an additional economic penalty through the loss of heavy water. Nearly all of the failures were attributed to secondary side water chemistry conditions and excursions, many of which resulted from condenser cooling water ingress. The investigation of the structural materials corrosion in correlation with the water chemistry, as well as the determination of impurities and corrosion products concentration and deposition and their removing from the CANDU steam generators is a very active field. Both the experimental works and the understanding of the mechanisms involved are submitted to some rapid changes and permanently open to research. To provide information about the corrosion behaviour of the structural materials from CANDU steam generators under normal and abnormal conditions of operation and to identify the failure types produced by corrosion there were performed a lot of corrosion experiments. These experiments consisted in chemical accelerated tests, static autoclaving and electrochemical investigations. The goal of this paper consists in the assessment of the corrosion kinetics for the Incoloy-800 at normal secondary circuit steam generator parameters. The gravimetric method, optical metallographic microscopy, XRD analysis as well as electrochemical measurements have been used to evaluate the corrosion behavior of Incoloy-800. (authors)

  20. Corrosion problems of PWR steam generators

    International Nuclear Information System (INIS)

    Literature data are assessed on corrosion failures of steam generator tubes made of INCONEL 600 or INCOLOY 800. It was found that both alloys with high nickel content showed good stability in a corrosion environment while being sensitive to carbide formation on grain boundaries. The gradual depletion of chromium results from the material and corrosion resistance deteriorates. INCOLOY 800 whose chromium carbide precipitation on grain boundaries in pure water and steam is negligible up to 75O degC and which is not subject to corrosion attacks in the above media and in an oxidizing environment at a temperature to about 700 degC shows the best corrosion resistance. Its favourable properties were tested in long-term operation in the Peach Bottom 1 nuclear power plant where no failures due to corrosion of this material have been recorded since 1967. In view of oxygenic-acid surface corrosion, it is necessary to work in a neutral or slightly basic environment should any one of the two alloys be used for steam generator construction. The results are summed up of an analysis conducted for the Beznau I NOK reactor. Water treatment with ash-free amines can be used as prevention against chemical corrosion mechanisms, although the treatment itself does not ensure corrosion resistance of steam generator key components. (J.B.)

  1. Stress corrosion cracking susceptibility of steam generator tube materials in AVT chemistry contaminated with lead

    International Nuclear Information System (INIS)

    The existence of lead is steam generator detected during analysis of deposits in the damaged areas of tube and in the fracture surfaces of existing cracks, bears out the hypothesis that lead may contribute to the secondary stress corrosion problems of steam generator tubes. The lead present in a steam generator which comes from the turbine, pump seals, valve packing, liners, etc. may form concentration from 0.02% to 0.2% in the sludge (1). The ubiquity of the major sources of lead limits the scope for controlling this contaminant and make difficult to remove it from steam generator. Taking this fact into account, it seems necessary to count on lead being present in new steam generator and to consider the influence of this contaminant on the behaviour of the materials considered as substitutes for Inconel 600. An extensive experimental work has been carried out in order to determine the stress corrosion crackings susceptibility of Inconel 690 TT, Incoloy 800 and Inconel 600 MA and TT in caustic and acidic environments contaminated with lead oxide. The results from this work, in particular for Incoloy 800, raised some concerns about the effect of minor amount of lead oxide in AVT environment. This paper presents the experimental work conducted with a view to determine the influence of lead oxide concentration in AVT conditions on the stress corrosion resistance of nickel alloys used in the fabrication of steam generator tubing. (authors). 3 tabs., 6 refs

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

    International Nuclear Information System (INIS)

    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 (Toroidal Field Model Coil) 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 (Center Solenoid Model Coil) 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

  3. A comparative study of stress corrosion cracking of steam generator tube materials in water at 315 C

    International Nuclear Information System (INIS)

    Stress Corrosion Cracking (SCC) of Type 304 and 304 L Stainless Steels, Inconel-600, Incoloy-800 and Monel-400 has been studied in water at 315 C, with or without 0.5 ppm Pb and 0.05 or 8 ppm O2. In 1600 hours test, results indicated that under mill annealed, coldworked (25%) and stress relieved (675 C, 1 hour) conditions, Type 304 L, Incoloy-800 and Monel-400 were resistant to cracking whereas Inconel-600 cracked intergaranularly, irrespective of the presence of Pb and O2 content in the water. Inconel-600, heat treated at 600 C for 24 hours or more following annealing, was found to be resistant to SCC.This beneficial heat treatment at 600 C has resulted in semi-continuous type precipitates at the grain boundaries. In addition, this heat treatment appeared to have stabilized the microstructure of Inconel-600, as seen indirectly, such as from the changes in mechanical properties, hydrogen uptake and electrochemical behaviour. The effects of these microstructural changes on the SCC behaviour have been discussed. (auth.)

  4. 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.

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

    International Nuclear Information System (INIS)

    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)-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 1/4 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

  6. Neutronic design and comparative characteristics of new pin type control rod for 14-MW TRIGA-SSR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Iorgulis, C.; Truta, C. [Institute for Nuclear Research, 0300 Pitesti (Romania)

    2001-07-01

    This paper presents the structural changes of the control rods (CR-s) which will be done in order to manufacture new ones and therefore to replace the original rods and to increase the present safety features of TRIGA 14 MW reactor in Pitesti. Few years ago two CR-s became inoperable due to the combined effect of welding corrosion, water penetration into the absorbent section of the CR and finally the swelling caused from the high internal pressure of the radiolysis-generated gases (gases which accelerated welding corrosion, water penetration, etc). Although safe reactor operation is not yet affected a decision to redesign a new CR was taken, entirely suitable to the general reactor design. The new CR will use boron carbide (same absorbent as in the original one); this baron carbide will be packed in a different way. A set of 16 Incoloy pins filled with a column of boron carbide pellets, leak-tight welded, will be manufactured. These 16 pins will be mounted in a square array with 5 pins on each side of the square (corner pins being the same far two sides). Neutronic analysis on the preliminary design was done using specific computer codes (WIMS, DFA, and MCNP). Reactivity worth of this new rod would be slight less (8%) than that of the original one, but still with full compensation capacity. Helium release analysis showed that gas pressure in the Incoloy tube after 15 years of normal operation would not exceed 57 ATM. (author)

  7. 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)

  8. Stress corrosion cracking of candidate waste container materials; Final report

    Energy Technology Data Exchange (ETDEWEB)

    Park, J.Y.; Maiya, P.S.; Soppet, W.K.; Diercks, D.R.; Shack, W.J.; Kassner, T.F. [Argonne National Lab., IL (United States)

    1992-06-01

    Six alloys have been selected as candidate container materials for the storage of high-level nuclear waste at the proposed Yucca mountain site in Nevada. These materials are Type 304L stainless steel (SS). Type 316L SS, Incoloy 825, phosphorus-deoxidized Cu, Cu-30%Ni, and Cu-7%Al. The present program has been initiated to determine whether any of these materials can survive for 300 years in the site environment without developing through-wall stress corrosion cracks. and to assess the relative resistance of these materials to stress corrosion cracking (SCC)- A series of slow-strain-rate tests (SSRTs) and fracture-mechanics crack-growth-rate (CGR) tests was performed at 93{degree}C and 1 atm of pressure in simulated J-13 well water. This water is representative, prior to the widespread availability of unsaturated-zone water, of the groundwater present at the Yucca Mountain site. Slow-strain-rate tests were conducted on 6.35-mm-diameter cylindrical specimens at strain rates of 10-{sup {minus}7} and 10{sup {minus}8} s{sup {minus}1} under crevice and noncrevice conditions. All tests were interrupted after nominal elongation strain of 1--4%. Scanning electron microscopy revealed some crack initiation in virtually all the materials, as well as weldments made from these materials. A stress- or strain-ratio cracking index ranks these materials, in order of increasing resistance to SCC, as follows: Type 304 SS < Type 316L SS < Incoloy 825 < Cu-30%Ni < Cu and Cu-7%Al. Fracture-mechanics CGR tests were conducted on 25.4-mm-thick compact tension specimens of Types 304L and 316L stainless steel (SS) and Incoloy 825. Crack-growth rates were measured under various load conditions: load ratios M of 0.5--1.0, frequencies of 10{sup {minus}3}-1 Hz, rise nines of 1--1000s, and peak stress intensities of 25--40 MPa{center_dot}m {sup l/2}.

  9. 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

  10. 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.

  11. A study of the release rate of corrosion products for nuclear SG tubing

    International Nuclear Information System (INIS)

    Out-of-core radiation fields mainly are caused by activated corrosion product released from nuclear heat supply system. It is very important to minimize the generation of radioactive corrosion products which become a radiation source, in order to meet the needs of operation and maintenance. A study was conducted on determining the general corrosion rate and metal release rate for Inconel 690 (Japan), Inconel 690 (China), Inconel 600 and Incoloy 800 tubes on the primary water side of nuclear power plant simulation. The results show that the release rates of corrosion products of Alloy 690 on the primary water side is very low and that they decrease with the increase of chromium content in the alloys. (8 refs, 9 figs., 4 tabs.)

  12. Effect of weld heat input and creep strain on the elevated temperature and crack growth properties of austenitic steels

    International Nuclear Information System (INIS)

    The potential material class for use at 6000C and more, e.g. for steam turbines with improved thermal efficiency, are austenitic steels. Using these steels with welded joints, it is to be considered that, by superposition of weld residual stresses and service stresses, extensive creep strains - and in the worst case crack formation - can occur locally. To assess the influence of these effects on service behaviour, different material states of CrNi-steels and Incoloy 800 were investigated with respect to strength, ductility and, especially, to crack and creep crack growth in the temperature range around 6000C. It is shown that creep embrittlement, not microstructural changes as effected by weld heat input, causes heat affected zone (HAZ)-reheat cracking. Creep embrittlement can be avoided by special design and fabrication rules. (orig.)

  13. 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.

  14. 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.

  15. 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

  16. Behaviour of steam generator tubing in the presence of silicon compounds

    International Nuclear Information System (INIS)

    The chemical reactions that take place between the components of concentrated solutions generate an aggressive environment. The presence of this environment and of the tubesheet crevices lead to localized corrosion and the affected tubes cannot ensure effective heat transfer between the fluids of the primary and secondary circuits. Thus, it becomes necessary to understand the corrosion process that occurs on the CANDU steam generator secondary side. The purpose of this paper is the assessment of the corrosion behaviour of the tube material Incoloy 800 at the normal secondary circuit parameters (temperature = 260 C, pressure = 5.1 MPa). The testing environment was demineralized water containing silicon compounds, at a pH = 9.5 regulated with morpholine and cyclohexylamine. The paper presents the results of metallographic, electronic microscopy and X-ray diffraction examinations, as well as the results of electrochemical measurements. (orig.)

  17. 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)

  18. 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

  19. Fracture mechanics studies on HTR materials

    International Nuclear Information System (INIS)

    Inconel 617, Nimonic 86 and Incoloy 800H were studied with a view to fatigue cracking, creep crack growth and fracture toughness (J-integral R-curve) up to 1273 K (10000C) in order to establish high-temperature toughness concepts and time laws of crack growth for conservative damage analysis. So far, results of creep crack growth tests at 1223 K (9500C) in air and helium are available for Nimonic 86. There are some problems concerning the correlation of creep crack growth results. The energy rate integral Csup(*), promising as it seems, still requires validation. The problems of initiation of steady crack growth in toughness tests and the application of linear-elastic fracture mechanics to fatigue crack growth at high temperatures are discussed. (orig./IHOE)

  20. Effects of specimen geometry on fatigue crack growth in X 10 NiCrAlTi 32 20 und NiCr 22 Co 12 Mo

    International Nuclear Information System (INIS)

    In order to assess the applicability of results obtained with 1''CT specimens, on fracture mechanical parameters of fatigue crack growth at 7000C and 8500C (and also in some tests at room temperature), the highly heat resistant alloys X 10 NiCrAlTi 32 20 (INCOLOY 800 H) and NiCr 22 Co 12 Mo (INCONEL 617) have been chosen for further experiments. The specimen geometries are CT 1'', CT 1/2'', and CCP 1/2'', and thickwalled pipes with circumferential defect at outer surface. The loads applied are of the pulsating tensile stress type (R = 0.05), at a frequency of 5Hz. Results reveal that at all temperatures, the fatigue crack growth is independent of specimen geometries, so that the linear-elastic ΔKI-concept is suitable for describing the crack growth under cyclic load. (orig.)

  1. Application of a passive electrochemical noise technique to localized corrosion of candidate radioactive waste container materials

    International Nuclear Information System (INIS)

    One of the key engineered barriers in the design of the proposed Yucca Mountain repository is the waste canister that encapsulates the spent fuel elements. Current candidate metals for the canisters to be emplaced at Yucca Mountain include cast iron, carbon steel, Incoloy 825 and titanium code-12. This project was designed to evaluate passive electrochemical noise techniques for measuring pitting and corrosion characteristics of candidate materials under prototypical repository conditions. Experimental techniques were also developed and optimized for measurements in a radiation environment. These techniques provide a new method for understanding material response to environmental effects (i.e., gamma radiation, temperature, solution chemistry) through the measurement of electrochemical noise generated during the corrosion of the metal surface. In addition, because of the passive nature of the measurement the technique could offer a means of in-situ monitoring of barrier performance

  2. 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.

  3. Compatibility of certain structural materials of water cooled nuclear power reactors with some decontaminant formulations

    International Nuclear Information System (INIS)

    Compatibility of reactor structural materials with some recent dilute chemical decontamination formulations intended for use or currently in use in water cooled reactors have been evaluated by electrochemical DC polarisation method and compared with that of formulations reported in the literature as useful reagents for decontamination. Citric acid-EDTA and citric acid-ascorbic acid-EDTA mixtures generally yield lower corrosion rates for all these materials as compared to the citric acid-oxalic acid-EDTA formulation. Corrosivity of picolinic acid-ascorbic acid mixture is minimum for carbon steel and zircaloy-2 and maximum for stainless steel-304 and Inconel-600 among the four formulations studied. The corrosion rates of zircaloy-2 and Incoloy-800 are media and temperature independent. (author). 7 refs., 1 tab

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

    Energy Technology Data Exchange (ETDEWEB)

    Mustin, T. [Dept. of Energy, Washington, DC (United States); Stump, R.C. [Dept. of Energy Idaho Operations Office, Idaho Falls, ID (United States); Tyacke, M.J. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States)

    1997-10-09

    This paper reports the activities underway by the US Department of Energy (DOE) for returning Training, Research, Isotope, General Atomics (TRIGA) spent nuclear fuel (SNF) from foreign research reactors (FRR) in four European countries to the Idaho National Engineering and Environmental Laboratory (INEEL). Those countries are Germany, Italy, Romania, and Slovenia. This is part of the ``Nuclear Weapons Nonproliferation Policy`` of returning research reactor SNF containing uranium enriched in the US. This paper describes the results of a pre-assessment trip in September, 1997, to these countries, including: history of the reactors and research being performed; inventory of TRIGA SNF; fuel types (stainless steel, aluminum, or Incoloy) and enrichments; and each country`s plans for returning their TRIGA SNF to the INEEL.

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

    International Nuclear Information System (INIS)

    This paper reports the activities underway by the US Department of Energy (DOE) for returning Training, Research, Isotope, General Atomics (TRIGA) spent nuclear fuel (SNF) from foreign research reactors (FRR) in four European countries to the Idaho National Engineering and Environmental Laboratory (INEEL). Those countries are Germany, Italy, Romania, and Slovenia. This is part of the ''Nuclear Weapons Nonproliferation Policy'' of returning research reactor SNF containing uranium enriched in the US. This paper describes the results of a pre-assessment trip in September, 1997, to these countries, including: history of the reactors and research being performed; inventory of TRIGA SNF; fuel types (stainless steel, aluminum, or Incoloy) and enrichments; and each country's plans for returning their TRIGA SNF to the INEEL

  6. 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

  7. 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.)

  8. Creep rupture behavior of candidate materials for nuclear process heat applications

    International Nuclear Information System (INIS)

    Creep and stress rupture properties are determined for the candidate materials to be used in hightemperature gas-cooled reactor (HTGR) components. The materials and test methods are briefly described based on experimental results of test durations of about20000 h. The medium creep strengths of the alloys Inconel-617, Hastelloy-X, Nimonic-86, Hastelloy-S, Manaurite-36X, IN-519, and Incoloy-800H are compared showing that Inconel-617 has the best creep rupture properties in the temperature range above 8000C. The rupture time of welded joints is in the lower range of the scatterband of the parent metal. The properties determined in different simulated HTGR atmospheres are within the scatterband of the properties obtained in air. Extrapolation methods are discussed and a modified minimum commitment method is favored

  9. 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

  10. 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)

  11. Stress corrosion cracking tests on high-level-waste container materials in simulated tuff repository environments

    International Nuclear Information System (INIS)

    Types 304L, 316L, and 321 austenitic stainless steel and Incoloy 825 are being considered as candidate container materials for emplacing high-level waste in a tuff repository. The stress corrosion cracking susceptibility of these materials under simulated tuff repository conditions was evaluated by using the notched C-ring method. The tests were conducted in boiling synthetic groundwater as well as in the steam/air phase above the boiling solutions. All specimens were in contact with crushed Topopah Spring tuff. The investigation showed that microcracks are frequently observed after testing as a result of stress corrosion cracking or intergranular attack. Results showing changes in water chemistry during test are also presented

  12. Determination of safety margins for creep loaded primary circuit components in case of loss of pressure accidents of a HTR plant (SR 383)

    International Nuclear Information System (INIS)

    The wall thickness of tubes in high temperature plants must be limited in such a way that pressure differences can not produce unadmissible deformations. For HTR (PNP-Plant) the postulated loss of secondary pressure is one of the considered accidents. In that case the tubes of the heat exchangers are loaded by the outher pressure of the primary coolant. So the risk of a creep collapse is given. The report is related to experimental and theoretical work for the creep collapse phenomena. HTR relevant tube geometries of the high temperature alloys NiCr22Co12Mo (INCONEL 617) and X10NiCrAlTi 32 20 (INCOLOY 800) were tested at temperatures of 900 and 950deg C and outer pressure loads in the range 40 bars. The experimental results are compared with theoretically computed values and discussed. The problem of safety margin is treated. Further, simplified procedures are developed for the estimation of the collapse time. (orig.)

  13. 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)

  14. Low concentration NP preoxidation condition for PWR decontamination

    International Nuclear Information System (INIS)

    To use preoxidation condition with low concentration NP (nitric acid permanganate) instead of conventional high concentration AP (alkline permanganate ) for PWR oxidation decontamination (POD) was summarized. Experiments including three parts have been performed. The defilming performance and decontamination factor of preoxidation with low concentration NP, which is 100, 10 times lower than that of AP are better than that with high concentration AP. The reason has been studied with the aid of prefilmed specimens of corrosion potential measuring in NP solution and chromium release in NP and AP solutions. The behaviour of alloy 13 prefilmed specimen in NP preoxidation solution is different from 18-8 ss and Incoloy 800. In the low acidity, the corrosion potential moves toward positive direction as the acidity becomes high

  15. 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.

  16. 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

  17. Tritium permeation through helium-heated steam generators of ceramic breeder blankets for DEMO

    International Nuclear Information System (INIS)

    The specifications of permeation barriers, tritium recovery process maintaining a very low tritium activity in the coolant, and control of the coolant chemistry, required the evaluation of the tritium losses through the steam generators and include the definition of its operating conditions by thermodynamic cycle calculations and its thermal-hydraulic design. For both tasks specific computer tools were developed. The obtained geometry, surface area, and temperature profiles along the heat exchanger tubes were then used to estimate the daily tritium permeation into the steam cycle. Steam oxidized Incoloy 800 austenitic stainless steel was identified as the best suited existing material; in nominal steady-state operation, the tritium escape into the steam cycle could be restricted to less than 10 Ci/d. Tritium permeation during temperature and pressure transients in the steam generator (destruction and possible self-healing of the permeation barrier) is identified to bear a large tritium release potential. Solutions are proposed. (from authors). 4 figs., 1 tab

  18. 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.)

  19. Creep crack growth behavior of several structural alloys

    Science.gov (United States)

    Sadananda, K.; Shahinian, P.

    1983-07-01

    Creep crack growth behavior of several high temperature alloys, Inconel 600, Inconel 625, Inconel X-750, Hastelloy X, Nimonic PE-16, Incoloy 800, and Haynes 25 (HS-25) was examined at 540, 650, 760, and 870 °C. Crack growth rates were analyzed in terms of both linear elastic stress intensity factor and J*-integral parameter. Among the alloys Inconel 600 and Hastelloy X did not show any observable crack growth. Instead, they deformed at a rapid rate resulting in severe blunting of the crack tip. The other alloys, Inconel 625, Inconel X-750, Incoloy 800, HS-25, and PE-16 showed crack growth at one or two temperatures and deformed continuously at other temperatures. Crack growth rates of the above alloys in terms ofJ* parameter were compared with the growth rates of other alloys published in the literature. Alloys such as Inconel X-750, Alloy 718, and IN-100 show very high growth rates as a result of their sensitivity to an air environment. Based on detailed fracture surface analysis, it is proposed that creep crack growth occurs by the nucleation and growth of wedge-type cracks at triple point junctions due to grain boundary sliding or by the formation and growth of cavities at the boundaries. Crack growth in the above alloys occurs only in some critical range of strain rates or temperatures. Since the service conditions for these alloys usually fall within this critical range, knowledge and understanding of creep crack growth behavior of the structural alloys are important.

  20. 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

  1. 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).

  2. Mechanism of corrosion of structural materials in contact with coal chars in coal gasifier atmospheres. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Douglass, D.L.; Bhide, V.S.; Vineberg, E.

    1980-05-01

    Six alloys, 310 stainless steel, Hastelloy X, Inconel 671, Incoloy 800, Haynes 188, and FeCrAlY (GE1541 and MA956), were corroded in two chars at 1600 and 1800/sup 0/F. The chars, FMC and Husky, contained 2.7 and 0.9% sulfur, respectively. Various parameters were investigated, including char size, cover gas, char quantity, char replenishment period, gas composition, and the use of coatings. The corrosion process was strictly sulfidation when the char was replenished every 24 hours or less. The kinetics of reaction were nearly linear with time. The reaction resulted in thick external sulfide scales with extensive internal sulfidation in the substrate. The kinetics and reaction-product morphologies suggested that diffusion through the sulfide scale played a minor role and that an interfacial reaction was the rate-controlling step. A mathematical model was developed which supported this hypothesis. The reaction rates showed a relatively minor role on alloy composition, depending upon whether the alloys were tested singularly or in combination with others. Inconel 671, the best alloy in CGA environments, consistently corroded the most rapidly of the chromia-former types regardless of char sulfur content or of the temperature. Type 310 stainless was marginally better than Inconel 671. Incoloy 800 was intermediate, whereas, Haynes 188 and Hastelloy X exhibited the best corrosion resistance. The FeCrAlY alloys reacted very rapidly in the absence of preoxidation treatments. All alloys corroded in char at least 1000 times more rapidly than in the CGA (MPC-ITTRI) environment. None of the alloys will be acceptable for use in contact with char unless coatings are applied.

  3. 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

  4. Experimental research on corrosion in constrained circulation zones in CANDU steam generator

    International Nuclear Information System (INIS)

    Corrosion is the major problem affecting safe operation of steam generator (SG). Most of the problems relating to corrosion are due to the local concentration effects of aggressive species and/or impurities in regions with constrained flow such as the cracks at tube-tubular plate joints. The concentration effects are of great importance and present interest in designing and SG operation. In this work we present the results of the corrosion tests performed in operation conditions specific to SG secondary circuit (temperature = 260 deg.C, pressure = 5.1 MPa) on devices simulating cracks in SA 508 cl.2 and Incoloy-800 carbon steel. Testing conditions were: demineralized water (pH = 9.5 controlled with volatile amines), raw Danube River water (pH = 9.5) and NaCl 100 g/l solution (pH 10.5). Investigation of corrosion behavior of the two materials was done metallographically and by X-ray diffraction. The results are presented as micrographs evidencing occurrence of pitting corrosion first on the tubular plate material (SA 508 cl.2) and in an extremely aggressive environment and than on the Incoloy-800 tubing material. A table is given with the geometrical, medium, electrochemical reaction, and alloy composition factors influencing the corrosion resistance in the crack. The following three conclusions are presented: 1. The main source of impurities (which can penetrate in SG) is the cooling water infiltration from condenser (raw Danube River water) as well as water of addition improperly treated. To these, one contribute high concentration factors (105 - 106) of the dissolved substances which can reach the crack and which can constitute extremely aggressive media generated from a medium seemingly non-aggressive. 2. The Incoloy-800 samples checked for 2,400 hours in demineralized and Danube River waters with pH = 9..5 presented an oxide protecting thin layer while in the case of the samples exposed in a solution containing NaCl 100 g/l with pH 10.5 the pitting attack is

  5. Studies relevant to the reactions of gaseous impurities with circuit materials in helium cooled reactors

    International Nuclear Information System (INIS)

    The reactions of water vapour, carbon monoxide and methane with a range of iron and nickel base alloys have been studied with the alloys in the cold worked condition within the temperature range 400 to 8000C. A 1 to 15 μatm H2O, 400 μ atm H2 environment decarburised a mild steel and to a lesser extent a 1.5Cr 1Mo steel at 400 to 5000C. The low alloy steel was also oxidised slightly by the environment. The mild steel at 4000C and the low alloy steel at 5000C did not interact with 4 but did catalyse carbon deposition from low partial pressures of carbon monoxide. The majority of the work was concerned with highly alloyed materials at 700 to 8000C. Their behaviour in complex environments containing 1 to 5 μatm H2O and 100 to 400 μatm CO or 100 μatm CH4 was generally dominated by oxidation by the water vapour. In the complete absence of water vapour the extent of reaction with carbon monoxide was almost independent of carbon monoxide pressure down to 500 μatm and then decreased only slightly down to 10 to 30 μatm. Carburisation by methane was approximately proportional to methane pressure and time of exposure. The presence of water vapour and of oxide films generally inhibited carburisation by both carbon monoxide and methane. The resistance of Incoloy 800, Sandvik 12R72HV and Nimonic 80A to oxidation by water vapour was poor by comparison with that for 316 steel, Esshete 1250 and Hastelloy X. However Hastelloy X and 316 steel together with Sandvik 12R72HV were on occasions susceptible to appreciable reaction with carbon monoxide. The high nickel alloys Hastelloy X, Nimonic 80A and to a lesser extent Incoloy 800 were carburised by methane more readily than 316 steel. Complex behaviour was observed in crevices in environments containing carbon monoxide or methane. It is attributed in part to the influence of water vapour. (author)

  6. 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

  7. Concept and neutronic design on the new pin-type control rods for TRIGA SSR 14 MW-INR Pitesti

    International Nuclear Information System (INIS)

    This paper presents the structural changes of the control rods (CR-s) intended to improve the safety features of TRIGA 14 MW reactor Pitesti. Few years ago two CR-s became inoperable due to the combined effect of welding corrosion, water penetration into the absorbent section of the CR and finally the swelling caused by the high internal pressure of the radiolysis-generated gases (gases which accelerated welding corrosion, water penetration, etc.). Although safe reactor operation is not yet affected, a decision to re-design a new CR was taken, entirely suitable to the general reactor design. The new CR will use boron carbide (same absorbent as in the original one), which will be packed in a different way. A set of 16 Incoloy pins filled with a column of boron carbide pellets, leak-tight welded will be manufactured. These 16 pins will be mounted in a square array with 5 pins on each side. Neutronic analysis on the preliminary design was done using specific computer codes (WIMS, DFA and and MCNP). Reactivity worth of this new rod would be slightly less (8%) than that of the original one, but still with full compensation capacity. Helium leak analysis showed that gas pressure in the Incoloy tube after 15 years of normal operation would not exceed 57 ATM. The new design maintains unchanged all the components of the CR but the shape and the enclosure of the boron carbide absorbent. In conclusion: - same boron carbide is the neutron absorbent as cylindrical pellets of 70% theoretical density; - absorbent pellets are definitely cladded in stainless steel tubes for all the estimated time of use; - the absorbent pins will be disposed in square array, 5 on each side of a standard rod location. So the concept of peripheral absorbent with a central trap is preserved; - all hydraulic and thermal conditions are preserved; - most important, the absorbent structure has to fulfil the major demand of safe operation - shutdown with the most reactive control rod fully withdrawn

  8. 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

  9. 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.

  10. Effects of methane concentration on the controlled-impurity helium corrosion behavior of selected HTGR structural materials

    International Nuclear Information System (INIS)

    The corrosion behavior of candidate structural alloys in a series of three simulated advanced gas-cooled reactor environments at 9000C (16520F), with methane concentration varied, is discussed. The alloys investigated include three wrought alloys, Hastelloy X, Inconel 617, and Incoloy 800H; two cast superalloys, Rene 100 and IN 713; one centrifugally cast alloy, HK 40; and an oxide-dispersion-strengthened alloy, MA 754. Corrosion behavior was found to be strongly dependent upon both the alloy chemistry and the environment. Oxidation, carburization, and/or mixed behavior was observed depending upon the specific conditions. An equilibrium thermodynamics approach has been used to predict alloy behavior and explain observations relevant to the understanding of gas/metal interactions in reactor helium, which inherently contains small amounts of reactive impurity species. Carburization was identified as the primary corrosion phenomenon of concern, and detailed analyses were performed to determine the susceptibility and control of carburization reactions. The presence of alumina scales, containing small amounts of titanium, was found to be particularly effective in inhibiting carburization. Small variations in methane concentration have been shown to have a dramatic effect upon the oxidation potential and subsequent corrosion behavior of the alloy systems

  11. Role of reducing agents in corrosion control of materials relevant to nuclear reactors

    International Nuclear Information System (INIS)

    Materials undergo enhanced corrosion in presence of oxidants in aqueous media. Hydrogen gas or water soluble reducing agents are used for inhibiting corrosion. In the present study feasibility of using alternate reducing agents such as hydrazine, ammonium hydroxide and hydroxylamine, that can stay in liquid phase was investigated. A comparative study of corrosion behavior of the materials carbon steel, stainless steel (SS-304 LN), Monel-400 and Incoloy-800 is presented in this report. In nuclear reactors, radiation adds to corrosion problems. Radiolytic products of water such as oxygen and hydrogen peroxide creates oxidizing environment. Electrochemical investigations at 90°C showed that the reducing agents were very efficient in controlling corrosion of materials. Thermal stability of hydrazine in the temperature range 200 - 285 °C was investigated. The results showed that thermal decomposition of hydrazine followed first order kinetics. As these reducing agents can be used in nuclear reactors, the radiation stability also was studied. The surface morphology of the exposed structural materials at high temperature was characterized by surface techniques such as SEM. (author)

  12. Hydrogen permeation through iron, nickel, and heat resisting alloys at elevated temperatures

    International Nuclear Information System (INIS)

    Hydrogen permeabilities of several metals and alloys were measured over the temperature range of 200 - 10000C and some factors affecting the hydrogen permeability were discussed. Materials studied were iron, nickel, 80Ni-20Cr alloy, 50Fe-30Ni-20Cr alloy, HK 40, Incoloy 800, Hastelloy X, and Inconel 600. The hydrogen permeability of nickel was proportional to the square root of the pressure and inversely proportional to the membrane thickness. The activation energy and pre-exponential factor for the hydrogen permeation through these metals and alloys were derived from the temperature coefficient. The hydrogen permeability of nickel was larger than that of iron (γ), and the permeabilities of the heat resisting alloys were between those of nickel and iron (γ). There was a close correlation between the hydrogen permeability and nickel content in the alloys, that is, the permeability increased with the increase of the nickel content in the alloys. The formation of the oxide film on the alloy surface in wet hydrogen resulted in a remarkable reduction of the hydrogen permeability at elevated temperatures. (auth.)

  13. Solid-state Diffusion Bonding of Candidate Fe-base and Ni-base Alloys for the Application of S-CO2 Cycle Heat Exchanger

    International Nuclear Information System (INIS)

    To achieve efficient heat transfer, compact type heat exchangers, such as printed circuit or plate fin type heat exchanger, are considered for intermediate heat exchangers (IHXs). Solid-state diffusion bonding (DB) is one of key issues for joining the thin metal sheets with flow passages that are either machined or photo-chemically etched. In this study, diffusion bonding was performed for the candidate Fe-base and Ni-base alloys. Tensile properties of the as-bonded were compared with the as-received and characteristics of the aged in high temperature S-CO2 environment were discussed. Studies on diffusion bonding of candidate alloys for the application of super-critical CO2 cycle were carried out. Strength ratios were close to 1 for Fe-base alloys (F91, SS 316H, and SS 347H), while those of Ni-base alloys (Alloy 600, Alloy 690) and Fe-Ni-Cr alloy (Incoloy 800HT) were somewhat decreased to about 0.8 due to the planar grain boundary and precipitates formed along the bond-line. After exposure in high temperature S-CO2 environment for 1000 h, mechanical properties were not changed substantially and the location of the failure was still in the gauge section away from the bond-line for most alloys. Thus, bond-line which plays a role as grain boundary is thought to have superior corrosion and carburization resistance comparable to that of parent matrix

  14. Fracture toughness determination in steam generator tubes

    International Nuclear Information System (INIS)

    The assessment of the structural integrity of steam generator tubes in nuclear power plants deserved increasing attention in the last years due to the negative impact related to their failures. In this context, elastic plastic fracture mechanics (EPFM) methodology appears as a potential tool for the analysis. The application of EPFM requires, necessarily, knowledge of two aspects, i.e., the driving force estimation in terms of an elastic plastic toughness parameter (e.g., J) and the experimental measurement of the fracture toughness of the material (e.g., the material J-resistance curve). The present work describes the development of a non standardized experimental technique aimed to determine J-resistance curves for steam generator tubes with circumferential through wall cracks. The tubes were made of Incoloy 800 (Ni: 30.0-35.0; Cr: 19.0-23.0; Fe: 35.5 min, % in weight). Due to its austenitic microstructure, this alloy shows very high toughness and is widely used in applications where a good corrosion resistance in aqueous environment or an excellent oxidation resistance in high temperature environment is required. Finally, a procedure for the structural integrity analysis of steam generator tubes with crack-like defects, based on a FAD diagram (Failure Assessment Diagram), is briefly described (author)

  15. 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.)

  16. Review of waste package verification tests. Semiannual report, October 1984-March 1985

    Energy Technology Data Exchange (ETDEWEB)

    Soo, P. (ed.)

    1985-07-01

    The potential of WAPPA, a second-generation waste package system code, to meet the needs of the regulatory community is analyzed. The analysis includes an indepth review of WAPPA`s individual process models and a review of WAPPA`s operation. It is concluded that the code is of limited use to the NRC in the present form. Recommendations for future improvement, usage, and implementation of the code are given. This report also describes the results of a testing program undertaken to determine the chemical environment that will be present near a high-level waste package emplaced in a basalt repository. For this purpose, low carbon 1020 steel (a current BWIP reference container material), synthetic basaltic groundwater and a mixture of bentonite and basalt were exposed, in an autoclave, to expected conditions some period after repository sealing (150{sup 0}C, {approx_equal}10.4 MPa). Parameters measured include changes in gas pressure with time and gas composition, variation in dissolved oxygen (DO), pH and certain ionic concentrations of water in the packing material across an imposed thermal gradient, mineralogic alteration of the basalt/bentonite mixture, and carbon steel corrosion behavior. A second testing program was also initiated to check the likelihood of stress corrosion cracking of austenitic stainless steels and Incoloy 825 which are being considered for use as waste container materials in the tuff repository program. 82 refs., 70 figs., 27 tabs.

  17. Study of γ-ray irradiation effects on corrosion resistance of alloys for storage of high-level waste packages, (1)

    International Nuclear Information System (INIS)

    The effects of γ-ray irradiation on corrosion resistance have been studied about candidate alloys as high-level waste canisters, overpacks and storage fasility materials. The purpose of this study is to evaluate integrity of alloys in the view of safety evaluation for interim storage systems such as water and air cooling system. Stress corrosion cracking (SCC) test has been carried out and the double U-bend type specimens were used. The alloys tested were Type 304 ss, Type 304L ss, Type 304EL ss, Type 309s ss, Incoloy 825, Inconel 600, Inconel 625 and SMA 50. Sensitized Type 304 ss, Type 304L ss and Type 309s ss have been found to be susceptible to SCC because of radiolysis in boiling deionized water with γ-ray irradiation but not without irradiation. It was also confirmed that they were susceptible to SCC under the existence of a little chloride ion and dissolved oxygen even without γ-ray irradiation. Fractographic observation revealed that the cracking mode was completely intergranular. The other alloys were not susceptible to SCC, but SMA 50 rusted over its surface. On the other hand, no alloy was susceptible to SCC in atmosphere at room temperature even with γ-ray irradiation. This study was carried out in corporation at JAERI and Kobe Steel Ltd. (author)

  18. Characterization of Alloys with Potential for Application in Cable-in-Conduit Conductors for High-Field Superconducting Magnets

    International Nuclear Information System (INIS)

    Since the introduction of the cable-in-conduit conductor (CICC) concept, a variety of alloys have been proposed for fabricating the jacket. The jacket provides primary containment of the supercritical helium coolant and is typically also the primary structural component for the magnet. These functions create requirements for strength, toughness, weldability, and fabricability in tubular form. When the CICC uses Nb3Sn, there are additional requirements to accommodate the manufacturing and heat-treatment processes for the superconductor as well as its strain-sensitive performance during operation. Both of the present favorite jacket alloys, Incoloy 908 and modified (ultra-low carbon) 316LN, have both demonstrated acceptable functionality as well as a few undesirable features. In this paper, we present data from cryogenic mechanical tests on a group of heat-resistant, high-strength superalloys that appear to offer equal or better mechanical performance (e.g. strength, toughness, and modulus) while mitigating the undesirable aspects (e.g. SAGBO in the case of I908 and thermal-expansion mismatch with Nb3Sn in the case of 316LN). Data are presented for each alloy in the as-received and aged conditions. These alloys are presently being considered as candidates for use in the next-generation hybrid magnet for the NHMFL but may also be of interest to the fusion and energy storage communities

  19. 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.)

  20. 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.

  1. 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

  2. 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

  3. 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

  4. The one-parameter-model - a constitutive equation applied to a heat resistant alloy

    International Nuclear Information System (INIS)

    In the present work a constitutive model earlier developed and used to predict experimental results of hot tests and fatigue tests from creep experiments of metallic materials were modified to comply with the properties of a high temperature resistant material. The improved model accounts for the properties of a material developing a density and a structure of dislocation lines which are capable of interactions with particles (carbides) from a second phase. The time and temperature dependent evolution of the carbide structure has been described by an equation which explains the formation of seeds as well as their growths (Ostwald ripening). The extended model was applied to Incoloy 800H which is known to develop a carbide structure. Therefore hot tensile and fatigue tests, creep and relaxation experiments using the heats ADU and BAK (KFA specifications) at temperature between 800deg C and 900deg C were performed including both solution treated specimens and specimens heat treated for 10, 100 and 1000 hours. As compared with the results from tensile tests where the carbide structures play a subordinated role, alternately, these structures have a decisive influence on the creep properties of specimens during the primary creep phase, i.e. low stresses and high temperatures. (orig.)

  5. 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.

  6. 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.

  7. Review of waste package verification tests. Semiannual report, October 1984-March 1985

    International Nuclear Information System (INIS)

    The potential of WAPPA, a second-generation waste package system code, to meet the needs of the regulatory community is analyzed. The analysis includes an indepth review of WAPPA's individual process models and a review of WAPPA's operation. It is concluded that the code is of limited use to the NRC in the present form. Recommendations for future improvement, usage, and implementation of the code are given. This report also describes the results of a testing program undertaken to determine the chemical environment that will be present near a high-level waste package emplaced in a basalt repository. For this purpose, low carbon 1020 steel (a current BWIP reference container material), synthetic basaltic groundwater and a mixture of bentonite and basalt were exposed, in an autoclave, to expected conditions some period after repository sealing (1500C, approx. =10.4 MPa). Parameters measured include changes in gas pressure with time and gas composition, variation in dissolved oxygen (DO), pH and certain ionic concentrations of water in the packing material across an imposed thermal gradient, mineralogic alteration of the basalt/bentonite mixture, and carbon steel corrosion behavior. A second testing program was also initiated to check the likelihood of stress corrosion cracking of austenitic stainless steels and Incoloy 825 which are being considered for use as waste container materials in the tuff repository program. 82 refs., 70 figs., 27 tabs

  8. Effects of temperature on the wear behavior of steam generator tube materials

    International Nuclear Information System (INIS)

    Reciprocating sliding wear and rotating wear experiments have been performed in various environmental conditions such as various temperatures, air and water environments to examine the wear properties of Inconel 690 steam generator tubes. For present study, the test rig was designed to examine the reciprocating and rotating wear properties in high temperature. The test rig consists of pressure vessel, electric heater, reciprocating, rotating and loading units. Tests were performed at constant applied load and sliding distance to investigate the effects of temperature and environment. Then, the worn surfaces were observed using scanning electron microscope to investigate the wear mechanism. In water environment, the wear loss increased with temperature between 25 and 150 .deg. C. Beyond 150 .deg. C the loss decreased with increasing temperature. However, in air environment, the loss decreased with increasing temperature in all temperature ranges. The wear loss variation with temperature was due to the formation of oxide layer. It was assumed that the layer was formed by wear debris sintering, cold welding or oxidation mechanisms and these properties were varied with tests temperatures and environments, which was confirmed by identification of the oxide layer by worn surface observation. The wear rate of Incoloy 800 was smaller than other steam generator tube materials. It seemed that these wear rate differences were due to the oxide layer differences. From the EDAX analysis, chemical composition of worn surface was different from each other

  9. Analytical evaluation of the environment effect on creep rupture strength

    International Nuclear Information System (INIS)

    An analytical approach was made in evaluating semi-quantitatively the effect of environment on rupture strength of materials. In the analysis the zone formed in the material by reaction with the environment was assumed to bear the applied load as one of the strength members. In calculations a law of mixtures of creep strength and the linear damage rule were applied. In the modeling of the load bearing by the composite structure of the environment-affected and intact zones, both parallel and series models were considered to formulate the equations. The equation for the parallel-loaded model was properly adopted in explaining semi-quantitatively the case of Incoloy alloy 800 crept in air, which was strengthened with the layer formed by nitrization. The equation for the serially loaded model was more successfully adopted to the evaluation of the rupture strength of dissimilar weld joints. The latter was also considered to be potentially adoptable to the problems of the effect of specimen size and shape on rupture strength, which had been often taken into account in evaluating the environment effect. For application of the developed method, examination was made to the possible decrease in rupture strength of Hastelloy alloy XR in long term tests by the formation of Cr depleted zone due to oxidation in HTGR impure helium, and the results were compared with the values obtained by experiments. (author)

  10. Properties of structural materials for sodium-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    For the selection of the structural materials for superheaters and reheaters of sodium-cooled fast breeder reactors, it is important to grasp the change of strength due to the complex change of material properties, such as the combination of surface corrosion, the decarbonization of 2-1/4 Cr-1Mo steel, the carburization of austenitic stainless steel, the structural change due to heating history, etc. in sodium environment. The stress corrosion cracking of austenitic materials in water must be studied also. The materials taken up in this paper are austenitic stainless steels such as SUS 304, SUS 316, SUS 321, and SUS 347, iron-based superalloy Incoloy 800, and ferritic alloy steel 2-1/4Cr-1Mo steel. The data on the above described properties of the materials are given. Also the tensile strength, creep rupture and fatigue characteristics of the parent materials in the amount of corrosion in sodium. The strength of ferritic alloy steel is lowered owing to the decarbonization in sodium, but the change of strength due to carburization was not observed. There is some possibility that the unstabilized steels such as SUS 304 and SUS 316 become sensitive to stress corrosion cracking, and the stabilized steels such as SUS 321 and SUS 347 become sensitive to it in long hour heating. The tensile strength of welded joints is almost same as that of parent materials, but the elongation decreases by about 10%. (Kako, I.)

  11. Safety research on iodine plate-out during postulated HTGR core heatup events

    International Nuclear Information System (INIS)

    In support of probabilistic risk assessment (PRA) studies on the high-temperature gas-cooled reactor (HTGR), an experimental program was conducted for iodine plateout on HTGR primary circuit metals during core heatup conditions. Metal iodine formation and adsorption characteristics were measured primarily for mild steel and to a limited extent for Incoloy 800 and other alloys. Pseudoisopiestic tests indicated quantitative formation of less volatile and water soluble iodides, FeI2 or CrI2, during core heatup conditions. The rate of formation of FeI2 was limited by mass transfer at temperatures above 5700K and was proportional to the partial pressure of iodine. The rate of iodide formation on chrome-nickel alloys appeared to be temperature sensitive, indicating slower reaction kinetics. The iodides preferentially plated out on surfaces at 520 to 620 K. Plateout tests were also performed for FeI2 in helium carrier gas flowing over mild steel or quartz surfaces over which a temperature gradient was maintained. PADLOC computer program correlations of the plateout profile based on the FeI2 vapor pressure assumed in the PRA studies were in fair agreement. The temperature at which most of the plateout occurred was from 620 to 700 K, depending on the partial pressure of the FeI2 tested. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Soo, P. (ed.)

    1986-01-01

    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.

  13. 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.

  14. Importance of oxide dispersion strengthening for creep detain at high temperatures

    International Nuclear Information System (INIS)

    Dispersion strengthening of the heat-resistant alloys is beneficial for their creep characteristics. The detain of creep is caused by in coherent oxide particles acting as pinning barriers for dislocations. determination of threshold stress for creep demands mechanical testing revealing a dependence of creep rate from the stress at constant temperature. The description of deformation microprocess like Orowan mechanism, climb, thermally activated detachment of dislocations from the particles, based on observations of dislocation-particle interaction in electron microscope. Studies of ferritic alloys (e.g. INCOLOY MA956) show fall-down of the creep resistance in intermediate temperatures due to the thermally activated dislocation recovery. At temperatures of 0.5-0.7 Tm the strain rate can be varied by changing the stress below the critical value. In high stresses are activated dislocation pinning, solid solution hardening and climb over particles. Near the critical stress, a high stress sensitivity of a strain rate can be modelled using Roesler-Arzt equation. Low stresses cause localized deformation, were collective motion of dislocations take place. In our opinion, the smaller stress sensitivity of the strain rate at low stresses is caused by a thermal processes like, the easy glide of dislocations in areas of lower particle density. (author)

  15. Microstructure Characterization Of Oxide Scale Growth On Oxide Dispersion Strengthened (ODS) MA -956 Implanted By Yttrium Ion

    International Nuclear Information System (INIS)

    The microstructure of oxide scale growth on ODS MA-956 implanted by yttrium ion have been studied. The specimen of ferritic Oxide Dispersion Strengthened (ODS) alumina-forming INCOLOY MA-956 with chemical composition (wt.%) of Fe-20Cr-5AI-0.4Ti-0.5Y2O3 used for this investigation. An implantation of certain Y elements applied to the commercial alloys substrate and undertaken by utilizing an acceleration potential of 85 keY at dose of 1x 017 ion/cm2. Result from SEM (Scanning Electron Microscope), have been combined with EPMA (Electron Probe Micro Analysis), EELS (Electron Energy Loss Spectroscopy), and TEMIEDX (Transmission Electron Microscope/Energy Dispersive X-ray-analysis). The results discussed by correlating the microstructural analysis of scale morphology and composition using some microstructural characterization techniques. The results showed the existence of yttrium on the top of the oxide layer, in the substrate and at position 1 μm from the substrate. The existence of aluminum and oxygen K edge only in the oxide layer, the existence of titanium particle in the grain boundary of the oxide layer and in the oxide layer, and the existence of Y and Ti particle in the interface layer between the oxide layer and the substrate

  16. A study on U-Zr-Er metallic alloy

    International Nuclear Information System (INIS)

    The Institute for Nuclear Research (INR) located in Pitesti, Romania operates a Steady State Reactor (SSR), which is a 14-MW TRIGA research reactor. Initially, the core configuration for full power operation used 29 fuel clusters each containing high enriched uranium U 10 wt % (with 93.09% 235U enrichment) - ZrH - Er (2.8 wt %) fuel-moderator rods cladded in Incoloy. The Institute for Nuclear Research makes efforts to develop the fabrication technology of TRIGA low enriched uranium fuel. This paper presents a characterization of the U-Zr-Er metallic rods obtained at INR. It is important to determine the rods elemental composition before the hydration at an H/Zr atom ration of 1.6. The percentages of U, Zr and Er are needed in the determination of the quantity of hydrogen necessary for the hydration process. The TRIGA fuel must contain 53.9% Zr, 45% U (with 20% 235U enrichment) and 1.1% Er. The U-Zr-Er metallic rods obtained are analysed by X-ray fluorescence spectrometry and X-ray diffraction. (authors)

  17. 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.)

  18. 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.

  19. Advances in the development and description of tritium permeation barriers in high-temperature alloys

    International Nuclear Information System (INIS)

    Recent experimental findings on tritium permeation barriers are described with special emphasis on the interpretation in terms of mechanical behavior. Kinetic measurements of the water vapor corrosion reaction with Incoloy-800 have been performed first by determining hydrogen production and permeation rates on line. Growth laws of the oxide scales have been determined indicating that a visually parabolic phase can be attributed to a scale of enhanced impeding effect against permeation. A certain amount of the hydrogen created by the corrosion reaction permeates spontaneously through the metal at a fraction varying between 1 and 10%. A new quality of oxide layer has been identified that can be characterized by enhanced activation energies for hydrogen permeation of about 150 kJ/mol as well as a modified pressure dependence proportional p1 in a limited range. Such scales show improved impeding factors >>100. Moreover, the effect of an additional layer on the opposite side of the tube specimen has been studied that shows a different impeding behavior dependent on the direction of the hydrogen/tritium flow. A model has been discussed describing the impeding effect of oxide scales in terms of surface controlled reaction steps rather than bulk diffusion, as has been the usual procedure hitherto. The model proposed offers a qualitative understanding of experimental findings characterizing high-quality layers

  20. Application of artificial neural networks for modeling localized corrosion

    International Nuclear Information System (INIS)

    Artificial neural networks (ANN) were applied to modeling localized corrosion of Incoloy Alloy 825 in simulated J - 13 well water. ANN as a non linear models can represent accurately localized corrosion phenomena caused by an environment containing chlorides, nitrates, fluorides and sulfates at various temperature ranges. Although the nature of the dependent variable of the ANN models, the visual rating of the localized corrosion is qualitative, a good correspondence between the output of the model and the actual indications is determined. Accurate ANN modeling has been carried out by using the visual inspection of the specimen surface, in contrast to linear modeling where in order to get a sound correlation between the system variables, a complex dependent parameter, having no clear physical meaning has been chosen. It has also been found that one can extrapolate to a certain extent, beyond the ability to interpolate (as with linear models). The ANN model predicted with a low relative error the visual rating of the corrosion rate of records which where part of the testing set of the ANN and belonging to the original full factorial design experiment. Thus, such models can be used for detailed analysis procedures as sensitivity, knowledge acquisition and optimization. (author). 7 refs, 9 figs

  1. Reduction of activation and contamination of reactor circuits

    International Nuclear Information System (INIS)

    After investigation of the metal loss rate of INCOLOY 800 samples by several test series performes at 3420C and 150 bar with different oxygen contents in demineralized water (deionate), the protective gold plating applied on the inner side of the first autoclave system had to be renewed. Within the period of reporting the second autoclave system (same dimensions as the first autoclave system) also provided with an inner gold plating was nearly completed. Also a small sized third autoclave system has been completed which allows to study the metal loss rate to pressurized water of standard stainless steel, the material making up 15% of the inner surface of a primary reactor circuit. Reduction of the oxygen content in deionate by noble gas sweeping, which is much more effective than chemical reduction by hydrazine, was improved in combination with the first autoclave system. Apparently the minimum oxygen content of 0,05 mg/kg required for primary loop water of pressurized reactors can still be lowered by the factor of 20. (orig./RW)

  2. 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.

  3. Corrosion of high nickel alloys in combined fluoride solutions for nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    The chemical reprocessing of fuel from the Pressurized Water Reactor (PWR) Core 2, Seed 1 and 2, requires hydrofluoric acid, nitric acid, and sulfuric acid singly and in mixtures. High nickel alloys tested in scoping tests at the boiling point in PWR process solutions included Hastelloy C-276, Hastelloy C-4, Hastelloy S, Hastelloy G, Inconel 625, Inconel 690 and Incoloy 825. Evaluations of weld performance including some metallographic examinations are included. Different methods of welding and different heats of alloys were investigated. The effect of varying compositions and concentration of process solutions were examined in order to select conditions which would minimize corrosion. Corrosion results are also presented for two Hastelloy C276 corrosion test vessels fabricated from 3-inch welded pipe with welded nozzles. These vessels have been exposed to PWR process solution for several months. They have provided valuable information on the performance of welds, the attack at the vapor-liquid interface, and the mode of attack to be expected in process service. (U.S.)

  4. 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

  5. Acquisition, processing and evaluation of the corrosion data for primary system of CANDU 6 reactor

    International Nuclear Information System (INIS)

    structural materials. For analysis of the corrosive processes, on the basis of the experimental data obtained in-pile and out-of-pile corrosion experiments, on different structural materials, a computer code named GEMXESAN (Gravimetry, Electrochemistry, Metallography, X-Ray Diffraction, Electronic Spectroscopy Analyses) was worked out. It allowed to observe oxidation and hydriding kinetics of Zircaloy-4 and Zr-2.5%Nb sheaths depending on initial surface condition, temperature, and water chemistry (pH, oxygen content, contamination with F- and Cl- ). Also studied were: - nodular corrosion of Zircaloy-4 depending on initial surface condition, temperature and microstructural changes induced by brazing and welding; - stress corrosion cracking (SCC) of Zircaloy-4 sheaths in presence of iodine; - stress corrosion cracking of Incoloy-800 tubes in presence of hydrogen; - embrittlement by hydrogen of Zr-2.5%Nb pressure tubes; - galvanic corrosion of Zr-2.5%Nb pressure tubes ; - crevice corrosion of Zr-2.5%Nb pressure tubes; - corrosion kinetics and corrosion products deposition on carbon steel SA 106 gr.B, martensitic steel 403m and Incoloy-800, as a function of pH, Cl-concentration and dissolved oxygen concentration; - corrosion kinetics, aspect of the corrosive films, deposition and releasing of corrosion products, as well as the specific activity of radionuclides on different structural materials coupons exposed in PHT autoclaves from Cernavoda NPP-Unit 1. The system of the acquisition, processing and evaluation of the corrosion data allows us to obtain: - water chemistry assessment (normal operation procedures and water chemistry criteria renewal); - corrosion assessment (determination of the corrosion rates, deposition and releasing of the corrosion products) after different operation periods; - information about appearance and evolution of some accelerated corrosion processes in primary circuit; - modeling of some oxidation processes. (authors)

  6. Primary chemistry as an action lever to reduce the source term. Review of the experience at Doel 3 and 4

    International Nuclear Information System (INIS)

    Dose rate reduction is of primary concern in order to make nuclear energy part of the future energy mix. This article presents the chemistry related actions undertaken at Doel 3 and 4 units in order to reduce the source term. Doel 3 and 4 are two 3 loops 1000 MWe Pressurized Water Reactors which respectively started operating commercial in 1982 and 1985. Both units have undergone a steam generators replacement: from Inconel 600MA tubing material to Incoloy 800 in 1993 for D3 and to Inconel 690TT in 1996 for D4. Until now, very similar chemistry strategies during operation, startup and shutdown have been applied in both units. They are in good accordance with the international standards. Both units operate under a coordinated chemistry, with a target pHTave=7.2. Dissolved hydrogen concentration is maintained in the range 25-35 cc/kg STD H2O. During shutdown, oxygenation is performed at full loop by means of hydrogen peroxide injections. However, significant differences in dose rates have been observed between the two units, with higher levels in Doel 3. In an attempt to understand these differences and to define new improvements pathways to lower the dose rates, this article assesses the chemistry strategy based on the available radiochemistry data, the RCS oxide layers characterization and the measured dose rates around the primary system. Primary chemistry strategy seems to be suited to maintain low level of radiation fields at Doel 4. However, Doel 3 is characterized by a contamination mainly driven by the Co-60. Specific actions have to be undertaken to minimize dose rates at Doel 3. Regarding the high contribution of Co-60 to dose rates, zinc injection seems to be a promising mitigation technique. (author)

  7. Fundamental studies on electron beam welding of heat-resistant superalloys for nuclear plants, (1)

    International Nuclear Information System (INIS)

    The basic investigation and research on the multi-purpose utilization of nuclear reactors have been carried out as the national project. The equipments for high temperature gas-cooled reactors are exposed to severe conditions in helium atmosphere of 1000 deg C, therefore the use of heat-resistant alloys such as Hastelloy, Inconel and Incoloy has been examined. The electron beam welding recently expanding the fields of application has excellent properties, such as the energy density is very high, the power output can be controlled freely as occasion arises, deep penetration can be obtained with small heat input, welding of high precision is feasible because the width of weld is narrow and the distortion due to the welding is small, and the weld of good quality can be obtained as the welding is carried out in vacuum. However, when the welding conditions are improper, the defects peculiar to electron beam welding arise, such as porosity, cold shut, spike phenomenon, and cracking due to welding. In this study, the characteristics of weld beads of respective heat-resistant alloys, especially the penetration mode and the properties of defects, were investigated by changing the parameters of electron beam welding, and the correlation among these was discussed. The range of proper welding conditions was set up for respective materials. Moreover, the correlation among the cracking susceptibility due to electron beam welding, the high temperature ductility of materials and the results of Trans-Varestraint test was investigated, and these testing methods are very useful for the evaluation of cracking susceptibility. (Kako, I.)

  8. 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.

  9. Behaviour of the steam generator tubing in water with different pH values

    International Nuclear Information System (INIS)

    Steam generators are crucial components of pressurized water reactors. The failure of the steam generator as a result of tube degradation by corrosion has been a major cause of Pressurized Heavy Water Reactor (PHWR) plant unavailability. Steam generator problems have caused major economic losses in terms of lost electricity production through forced unit outages and, in cases of extreme damage, as additional direct cost for large-scale repair or replacement of steam generators. The excellent performance to date of CANDU steam generators can be attributed, in part, to their design and performance characteristics, which typically involve higher recirculation ratios and lower heat fluxes and temperatures. However, the steam generator tubes are susceptible to failure by a variety of mechanisms, the vast majority of which are related to corrosion. 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 with the purpose of evaluating the quantities 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 behaviour of the tubes material, Incoloy-800, at normal secondary circuit parameters (temperature-260 oC, pressure-5.1 MPa). The testing environment was the demineralised water without impurities, at different pH values regulated with morpholine and cyclohexylamine (all volatile treatment-AVT). The results are presented like micrographics and graphics representing weight loss of metal due to corrosion, corrosion rate, total corrosion products formed, the adherent corrosion products, released corrosion products, release rate of corrosion products and release rate of the metal.

  10. High power test of low enriched UZrH

    International Nuclear Information System (INIS)

    TRIGA-LEU fuel is currently undergoing high power tests in the 30 MW Oak Ridge Reactor. These tests are being funded by the Department of Energy through the RERTR program [Reduced Enrichment Research and Test Reactor program administered by Argonne National Laboratory] and began in mid-December, 1979 on a 16-rod shrouded cluster. The fuel rods are 0.51 in. 0D, clad with 0.16 in. Incoloy and the fuel length is 22 in. It is planned to test the UZrH fuel with 45, 30 and 20 wt-% U (nominal 20% enriched), to burnup values of about 50% of the contained U-235 in the 45 wt-% rods and about 40% and 35% burnup in the 30 wt-%, and 20 wt-% U fuel. It will take about 2 years of irradiation to produce the desired burnup in the 45 wt-% U fuel. Currently being tested are six 45 wt-% U and five 30 wt-% U rods. The remaining 5 rods are stainless steel dummies which were necessary to meet an operational requirement of the ORR which limits the power generation in a fuel rod to a value which would not raise the coolant temperature above the saturation level. Maximum calculated fuel rod powers were 40 kW, which would produce a fuel temperature of about 650 deg. C. The measured temperatures are about 400 deg. C and 350 deg. C for the 45 and 30 wt-% U fuel, respectively. Flow and ΔT measurements show the cluster power generation to be about 250 kW, or about 65% of the design value. Reasons for the lower than expected power are still being evaluated and a proposal has been submitted for rearrangement of the fuel rods within the cluster to raise the powers and temperatures in the TRIGA-LEU fuel rods. (author)

  11. 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)

  12. Corrosion Processes of the CANDU Steam Generator Materials in the Presence of Silicon Compounds

    International Nuclear Information System (INIS)

    The feedwater that enters the steam generators (SG) under normal operating conditions is extremely pure but, however, it contains low levels (generally in the μg/l concentration range) of impurities such as iron, chloride, sulphate, silicate, etc. When water is converted into steam and exits the steam generator, the non-volatile impurities are left behind. As a result of their concentration, the bulk steam generator water is considerably higher than the one in the feedwater. Nevertheless, the concentrations of corrosive impurities are in general sufficiently low so that the bulk water is not significantly aggressive towards steam generator materials. The impurities and corrosion products existing in the steam generator concentrate in the porous deposits on the steam generator tubesheet. The chemical reactions that take place between the components of concentrated solutions generate an aggressive environment. The presence of this environment and of the tubesheet crevices lead to localized corrosion and thus the same tubes cannot ensure the heat transfer between the fluids of the primary and secondary circuits. Thus, it becomes necessary the understanding of the corrosion process that develops into SG secondary side. The purpose of this paper is the assessment of corrosion behavior of the tubes materials (Incoloy-800) at the normal secondary circuit parameters (temperature = 2600 deg C, pressure = 5.1 MPa). The testing environment was demineralized water containing silicon compounds, at a pH=9.5 regulated with morpholine and cyclohexyl-amine (all volatile treatment - AVT). The paper presents the results of metallographic examinations as well as the results of electrochemical measurements. (authors)

  13. 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

  14. Corrosion behavior of the tube - tubular plate joint zone in the presence of sediments

    International Nuclear Information System (INIS)

    The corrosion is a very important problem which concerns the safe operation of steam generators. The predominant part of corrosion problems is related to the local concentration of aggressive species and/or to the impurities from the slow-flow regions, like those created by cracks in tube - tubular plate joint zones. The consequences of such local concentrations are very important and as such entail interest in the design and utilization of steam generators. This study presents the results of the corrosion tests performed under specific operation conditions of the secondary circuit in NPP (temperature, 260 o C; pressure, 5.1 MPa) on a crack simulating device made of carbon steel SA 508 cl.2 (forming the tubular plate) and Incoloy-800 (forming the tubes). The chemical medium of these tests was the following: solution of NaCl, 25g/l (pH=10.5); solution of NaCl, 50 g/l (pH=10.5); solution of NaCl, 75g/l (pH=10.5); solution of NaCl, 75g/l + solution of Na2 SO4, 10 g/l (pH=10.5). The behavior of these two materials to corrosion was studied by metallographic investigations. The results are presented as microphotographs evidencing the occurrence of pitting corrosion first on material of the tubular plate, in the presence of medium particularly aggressive and on the material of the tubes. The aim of this study is to establish the corrosion mechanism as well as the formation of the oxide layer on the carbon steel in crack simulating devices. (authors)

  15. Selection of canister materials: electrochemical corrosion tests of HASTELLOY C4 and other Ni-Cr(-Mo) alloys in chloride containing solutions

    International Nuclear Information System (INIS)

    Several Ni-Cr(-Mo) alloys (HASTELLOY C4, INCONEL 625, SANICRO 28, INCOLOY 825, INCONEL 690) were tested by electrochemical methods to characterize their corrosion behaviour 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 which proved to have the highest corrosion resistance of all tested alloys was tested by the following electrochemical methods: (1) Poteniodynamic measurements to determine the characteristic potentials, passive current densities and critical pitting potentials. (2) Potentiostatic measurements in order to evaluate the duration of the incubation period at various potentials. (3) Galvanostatic measurements in order to characterize critical pitting potentials. As electrolyte 1 m H2SO4 was used, as parameters temperature, chloride content and pH-value were varied. Variation of temperature gives the following results: an increase in temperature leads to an increase of the critical passivation current density, the passive potential bandwidth decreases slightly and the passive current density increases with rising temperature. The addition of different chloride contents to the H2SO4 solution shows the following effects: the critical passivation current density and the passive current density increase with increasing chloride concentration and both, the critical pitting potentials and the pitting nucleation potentials, shift towards negative values. As third parameter the pH-value was varied. As expected, an increase of the pH-value extends the passive region to more negative values, the passive current density decreases. The variation of the pH-value does not affect the critical pitting potential. All tested alloys showed a clearly limited resistance against pitting corrosion phenomena. However, the best corrosion behaviour is shown by HASTELLOY C4, which has of all tested alloys the lowest passivation current density

  16. 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)

  17. 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.

  18. 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

  19. Design of a 16-pin fuel cluster for conversion of MTR-plate type reactors to TRIGA

    International Nuclear Information System (INIS)

    General Atomic is currently designing a 16-rod TRIGA fuel cluster to be used in converting and upgrading MTR-plate type reactor cores. The TRIGA conversion cluster is designed to operate at power levels up to 10 MW, however, the achievable power level will be dependent upon the cooling system available in the converted reactor. A coolant flow rate of about 18,900 L/min (5000 GPM) is needed for 10 MW operation. A cooling system with a flow rate of 8300 L/min (2200 GPM) will allow 5 MW operation. The fuel rods used in the conversion bundle are identical in size and uranium content to the fuel rods to be used in the 14 MW TRIGA core being built for the Romanian Institute for Nuclear Technologies. The 12.95 mm (0.51 in.) OD TRIGA fuel-moderator material is U-Er-ZrH1.6 with 10 wt-% uranium (93% enriched). Erbium is included as a burnable poison and is a major contributor to the prompt negative temperature coefficient - the dominant safety feature of TRIGA fuel. Fuel rods are clad with 13.77 mm (0.542 in.) OD incoloy and have an active fuel height of 558.8 mm (22.0 in.). Fuel rod spacing within the cluster is identical to the 14 MW TRIGA design, with 2.54 mm (0.10 in.) between fuel rods and between rods and the cluster shroud. Two intermediate spacers are used within the cluster to maintain clearances along the length of the fuel rods. Maintaining the rod dimensions and spacing equivalent to the 14 MW TRIGA design allows utilization of existing nuclear, thermal and mechanical design information in developing the 16-pin cluster design. (author)

  20. Lead corrosion and transport in simulated secondary feedwater

    Energy Technology Data Exchange (ETDEWEB)

    McGarvey, G.B. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Ross, K.J.; McDougall, T.E. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Turner, C.W. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    1998-07-01

    The ubiquitous presence of lead at trace levels in secondary feedwater is a concern to all operators of steam generators and has prompted laboratory studies of its interaction with Inconel 600, Inconel 690, Monel 400 and Incoloy 800. Acute exposures of steam generator alloys to high levels of,lead in the laboratory and in the field have accelerated the degradation of these alloys. There is some disagreement over the role of lead when the exposure is to chronic levels. It has been proposed that most of the present degradation of steam generator tubes is due to low levels of lead although few if any failures have been experimentally linked to lead when sub-parts per billion levels are present in the feedwater. One reason for the difficulty in assigning the role of the lead is related to its possible immobilization on the surfaces of corrosion products or iron oxide films in the feedwater system. We have measured lead adsorption profiles on the three principal corrosion products in the secondary feedwater; magnetite, lepidocrocite and hematite. In all cases, essentially complete adsorption of the lead is achieved at pH values less than that of the feedwater (9-10). If lead is maintained in this adsorbed state, it may be more chemically benign than lead that is free to dissolve in the feedwater and subsequently adsorb on steam generator tube surfaces. In this paper, we report on lead adsorption onto simulated corrosion products under simulated feedwater conditions and propose a physical model for the transport and fate of lead under operating conditions. The nature of lead adsorption onto the surfaces of different corrosion products will be discussed. The desorption behaviour of lead from iron oxide surfaces following different treatment conditions will be used to propose a model for tile transport and probable fate of lead in the secondary feedwater system. (author)

  1. Lead corrosion and transport in simulated secondary feedwater

    Energy Technology Data Exchange (ETDEWEB)

    McGarvey, G.B. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Ross, K.J.; McDougall, T.E. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Turner, C.W

    1999-07-01

    The ubiquitous presence of lead at trace levels in secondary feedwater is a concern to all operators of steam generators and has prompted laboratory studies of its interaction with Inconel 600, Inconel 690, Monel 400 and Incoloy 800. Acute exposures of steam generator alloys to high levels of lead in the laboratory and in the field have accelerated the degradation of these alloys. There is some disagreement over the role of lead when the exposure is to chronic levels. It has been proposed that most of the present degradation of steam generator tubes is caused by low levels of lead although few, if any, failures have been experimentally linked to lead when it is present in sub-parts per billion in the feedwater. One reason for the difficulty in assigning the role of the lead is related to its possible immobilization on the surfaces of corrosion products or iron oxide films in the feedwater system. We have measured lead adsorption profiles on the 3 principal corrosion products in the secondary feedwater: magnetite, lepidocrocite and hematite. In all cases, essentially complete adsorption of the lead is achieved at pH values that are lower than the pH of the feedwater (9 to 10). If lead is maintained in this adsorbed state, it may be more chemically benign than lead that is free to dissolve in the feedwater and subsequently adsorb on steam generator tube surfaces. In this paper, we report on lead adsorption onto simulated corrosion products under simulated feedwater conditions and propose a physical model for the transport and fate of lead under operating conditions. The nature of lead adsorption onto the surfaces of different corrosion products will be discussed. The desorption behaviour of lead from iron oxide surfaces after different treatment conditions will be used to propose a model for the transport and probable fate of lead in the secondary feedwater system. (author)

  2. 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)

  3. 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.)

  4. Pitting, galvanic, and long-term corrosion studies on candidate container alloys for the Tuff Repository

    International Nuclear Information System (INIS)

    Contest Columbus Technologies, Inc. (CC Technologies) investigated the long-term performance of container materials for high-level radioactive waste packages as part of the information needed by the Nuclear Regulatory Commission to assess the Department of Energy's application to construct a geologic repository for the high-level radioactive waste. The scope of work focused on the Tuff Repository and employed short-term techniques, such as electrochemical and mechanical techniques to examine a wide range of possible failure modes. Two classes of alloys were evaluated for use as container materials for the Tuff Repository; Fe-Cr-Ni alloys and copper-base alloys. The candidate Fe-Cr-Ni alloys were Type 304L Stainless Steel (Alloy 304L) and Incoloy Alloy 825 (Alloy 825). The candidate copper-base alloys were CDA 102 Copper (Alloy CDA 102) and CDA 715 Copper-3D Nickel (Alloy CDA 715). The corrosion testing was performed in a simulated J-13 well water and in solutions selected from an experimental matrix from Task 2 of the program. This report summarizes the results of Task 4 (Pitting Studies), Task 6 (Other Failure Modes) and Task 7 (Long-Term Exposures) of the program. Pit-initiation studies, performed in Task 4, focused on anomalous Cyclic Potentiodynamic Polarization (CPP) behavior of the copper-base alloys reported in Task 2 of the program. Pit propagation studies were performed on Alloy CDA 102 in Task A of the program. Two types of galvanic corrosion studies were performed in Task 6 of the program; thermogalvanic couples and borehole linear-container interactions. In the thermogalvanic couples tests, the effect of temperature variation on the surface of the container on acceleration of corrosion was evaluated for two alloys; Alloy CDA 102 and Alloy 304L. Long-term immersion tests were conducted in Task 7 of the program

  5. 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.

  6. Tritium permeation through helium-heated steam generators of ceramic breeder blankets for DEMO

    International Nuclear Information System (INIS)

    The potential sources of tritium contamination of the helium coolant of ceramic breeder blankets have previously been evaluated for the specific case of the European BIT DEMO blanket. This confirmed that the control of tritium losses to the steam circuit is a critical issue which demands development concerning (a) permeation barriers, (b) tritium recovery processes maintaining a very low tritium activity in the coolant, and (c) control of the coolant chemistry. The specifications of these developments required the evaluation of the tritium losses through the steam generators, and includes the definition of their operating conditions by thermodynamic cycle calculations, and their thermal-hydraulic design. For both tasks, specific computer tools were developed. The geometry obtained, the surface area and the temperature profiles along the heat-exchanger tubes were then used to estimate the daily tritium permeation into the steam cycle. Steam-oxidized Incoloy 800 austenitic stainless steel was identified as the best-suited existing material. Our results indicate that in nominal steady-state operation the tritium escape into the steam cycle could be restricted to less than 10 Ci per day. The conditions for this are specified, but their feasibility demands, in particular, the resolution of certain gas chemistry problems, and their validation in the more stringent environment of an operating blanket. Tritium permeation during temperature and pressure transients in the steam generator (destruction and possible self-healing of the permeation barrier) was identified as bearing a large tritium release potential. The problems associated with such transients are discussed and possible solutions are proposed. (orig.)

  7. Basalt Waste Isolation Project. Quarterly report, January 1-March 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    Deju, R.A.

    1980-04-01

    This report addresses the technical progress for the Basalt Waste Isolation Project for the second quarter of fiscal year 1980. Seismic design values were developed for preliminary repository design purposes; 0.25 g horizontal and 0.125 g vertical maximum accelerations for surface, zero-period conditions. Preliminary seismic data indicate broad, smooth areas exist in the bedrock surface in the western portion of the Cold Creek syncline and a gently undulating bedrock surface in the eastern portion. Test results indicate hydraulic property values fall within the range previously reported for sedimentary and interflow zones in basalt formations at the Hanford Site. Preliminary results of available hydrochemical data obtained from several borehole sites indicate that little, if any, vertical mixing of groundwaters is taking place across this stratigraphic boundary. Multiple barrier studies indicate that the primary candidate canister/overpack alloys are TiCode-12, Inconel 625, Incoloy 825, and Zircaloy 2. Low-carbon steel and cast iron are among the list of secondary candidate canister alloys. Laboratory tests of borehole plug designs have shown that it is feasible to design a composite plug system that will satisfactorily seal a nuclear waste repository in Columbia River basalt. The National Lead Industries, Inc., NLI-1/2 Universal Spent Fuel Shipping Cask was selected for use in Phase II operations. Creep test results of samples of Umtanum basalt from borehole DC-6 were plotted and show the day-to-day variation in deformation versus time. The concept selection phase of repository conceptual design was completed in March 1980. A test plan for the Exploratory Shaft Test Facility was developed and is scheduled for submittal to the US Department of Energy in May 1980.

  8. Behaviour of the steam generator tubing in water with different pH values

    Energy Technology Data Exchange (ETDEWEB)

    Lucan, Dumitra, E-mail: dumitra.lucan@nuclear.r [Department of Corrosion and Circuits Chemistry, Institute for Nuclear Research, POB 78, Pitesti (Romania)

    2011-04-15

    Steam generators are crucial components of pressurized water reactors. The failure of the steam generator as a result of tube degradation by corrosion has been a major cause of Pressurized Heavy Water Reactor (PHWR) plant unavailability. Steam generator problems have caused major economic losses in terms of lost electricity production through forced unit outages and, in cases of extreme damage, as additional direct cost for large-scale repair or replacement of steam generators. The excellent performance to date of CANDU steam generators can be attributed, in part, to their design and performance characteristics, which typically involve higher recirculation ratios and lower heat fluxes and temperatures. However, the steam generator tubes are susceptible to failure by a variety of mechanisms, the vast majority of which are related to corrosion. 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 with the purpose of evaluating the quantities 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 behaviour of the tubes material, Incoloy-800, at normal secondary circuit parameters (temperature-260 {sup o}C, pressure-5.1 MPa). The testing environment was the demineralised water without impurities, at different pH values regulated with morpholine and cyclohexylamine (all volatile treatment-AVT). The results are presented like micrographics and graphics representing weight loss of metal due to corrosion, corrosion rate, total corrosion products formed, the adherent corrosion products, released corrosion products, release rate of corrosion products and release rate of the metal.

  9. Nuclear safety enhancement by structural changes of the control rods

    Energy Technology Data Exchange (ETDEWEB)

    Ciocanescu, M.; Preda, M.; Iorgulis, C.; Truta, C. (Institute for Nuclear Research, Pitesti (Romania))

    1999-12-15

    This paper presents the modification of structure of the control rods which are intended to be done in order to improve the nuclear safety of the TRIGA-SSR-14 MW Pitesti, Romania. In 1996 and 1997 a number of 2 control rod got inoperable due to high corrosion of the welds, water penetration into the absorbent followed by the swelling generated by internal pressure. Although the safe operation of the reactor is not yet affected, it was decided that a more reliable control rod has to be designed and manufactured. Basically, the new control rod will use, as the old ones, boron carbide as absorbent, but boron carbide will be assembled in a different way. There will be manufactured a set of 16 incoloy tubes for each control rod filled with boron carbide pellets and leak tight sealed. These tubes will be mounted in a square array, on the perimeter of the control rod, 5 tubes on each side of the square. The preliminary design on which neutronic analysis on this structure has been performed using specific computation codes. The result of this analyze shows that new control rod reactivity is slightly lower (with about 10%) compared to the older one, but assures a full compensation capacity. Helium release analysis done here reveals that the Helium pressure in the tube at estimated end-of -life would not exceed 57 atm. The advantages of this new concept: it is avoided the problem of helium or radiolysis gases generated into the control rod, and which causes, as we shown, control rod swelling. (orig.)

  10. 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.

  11. Lead corrosion and transport in simulated secondary feedwater

    International Nuclear Information System (INIS)

    The ubiquitous presence of lead at trace levels in secondary feedwater is a concern to all operators of steam generators and has prompted laboratory studies of its interaction with Inconel 600, Inconel 690, Monel 400 and Incoloy 800. Acute exposures of steam generator alloys to high levels of lead in the laboratory and in the field have accelerated the degradation of these alloys. There is some disagreement over the role of lead when the exposure is to chronic levels. It has been proposed that most of the present degradation of steam generator tubes is caused by low levels of lead although few, if any, failures have been experimentally linked to lead when it is present in sub-parts per billion in the feedwater. One reason for the difficulty in assigning the role of the lead is related to its possible immobilization on the surfaces of corrosion products or iron oxide films in the feedwater system. We have measured lead adsorption profiles on the 3 principal corrosion products in the secondary feedwater: magnetite, lepidocrocite and hematite. In all cases, essentially complete adsorption of the lead is achieved at pH values that are lower than the pH of the feedwater (9 to 10). If lead is maintained in this adsorbed state, it may be more chemically benign than lead that is free to dissolve in the feedwater and subsequently adsorb on steam generator tube surfaces. In this paper, we report on lead adsorption onto simulated corrosion products under simulated feedwater conditions and propose a physical model for the transport and fate of lead under operating conditions. The nature of lead adsorption onto the surfaces of different corrosion products will be discussed. The desorption behaviour of lead from iron oxide surfaces after different treatment conditions will be used to propose a model for the transport and probable fate of lead in the secondary feedwater system. (author)

  12. Lead corrosion and transport in simulated secondary feedwater

    International Nuclear Information System (INIS)

    The ubiquitous presence of lead at trace levels in secondary feedwater is a concern to all operators of steam generators and has prompted laboratory studies of its interaction with Inconel 600, Inconel 690, Monel 400 and Incoloy 800. Acute exposures of steam generator alloys to high levels of,lead in the laboratory and in the field have accelerated the degradation of these alloys. There is some disagreement over the role of lead when the exposure is to chronic levels. It has been proposed that most of the present degradation of steam generator tubes is due to low levels of lead although few if any failures have been experimentally linked to lead when sub-parts per billion levels are present in the feedwater. One reason for the difficulty in assigning the role of the lead is related to its possible immobilization on the surfaces of corrosion products or iron oxide films in the feedwater system. We have measured lead adsorption profiles on the three principal corrosion products in the secondary feedwater; magnetite, lepidocrocite and hematite. In all cases, essentially complete adsorption of the lead is achieved at pH values less than that of the feedwater (9-10). If lead is maintained in this adsorbed state, it may be more chemically benign than lead that is free to dissolve in the feedwater and subsequently adsorb on steam generator tube surfaces. In this paper, we report on lead adsorption onto simulated corrosion products under simulated feedwater conditions and propose a physical model for the transport and fate of lead under operating conditions. The nature of lead adsorption onto the surfaces of different corrosion products will be discussed. The desorption behaviour of lead from iron oxide surfaces following different treatment conditions will be used to propose a model for tile transport and probable fate of lead in the secondary feedwater system. (author)

  13. 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)

  14. International cooperation in converting the Philippine Research Reactor PRR-1 and the impacts of its upgraded facilities

    International Nuclear Information System (INIS)

    The paper describes the institutional and physical considerations in the fuel conversion and upgrade in power of the Philippine Research Reactor PRR-1. Various purposes for the conversion and upgrade discussed in the paper can be summarized as production of short-lived isotopes, neutron activation analysis enhancement for an in-house research and analytical service to customers, delayed neutron activation analysis of geological samples from uranium exploration, enhancement of beam tube fluxes for physics experiments, and training. The fuel conversion was from a highly enriched uranium (HEU, 93%), aluminum clad MTR plate type fuel to low enriched (LEU, 20%). Incoloy 800 clad TRIGA uranium-zirconium hydride (UZrH) four-rod fuel clusters. The UZrH clusters were designed to directly replace the plate elements in the original grid plate. Thus no core structure modifications were required. The upgrade in power was from 1 MW to 3 MW. This required an increase in coolant pumping power and flow and an increase in the capacity of the heat exchanger and cooling tower system to reject the additional heat load. The various reactor instrumentation systems were also replaced with an integrated system for reactor monitoring and control. International assistance and cooperation from other countries and the International Atomic Energy Agency have aided the overall facility operation and expertise and the conversion and upgrade through funding grants, expert help and fellowships. The new reactor capabilities include in-core experimental facilities with thermal flux values nearly six times greater than that of the original core, pulsing operation, and increased performance and enhanced safety generally. (author). 3 figs, 1 tab

  15. 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.

  16. 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

  17. 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

  18. Cast-to-cast variation in end-plug welds for TRIGA fuel elements

    International Nuclear Information System (INIS)

    In the Institute for Nuclear Research (INR) Pitesti - TRIGA Reactor Department there are under development activities for assembling TRIGA-LEU fuel elements locally manufactured, through autogenous Tungsten-Inert-Gas (TIG) welding. Due to specific problems occurring in welding Ni alloys, namely the dissimilar joint between Inconel 600 and Inconel 800 at the end-plug weld, weldability tests on Inconel 600 under various conditions were performed. The tests had been carried out in two stages: basic tests, on simple turned rods of Inconel 600; confirmation tests, on real (actual) end plug –to – clad welding. The basic tests had been done on simple rods machined (turned) at 13.8 mm (main diameter of the plugs) on which there have been made simple semicircular weldings ( no joint involved). Confirmation tests were done on the plug-clad assembly (dissimilar welding Incoloy-Inconel), with the welding parameters resulted from the preliminary conclusions of the basic tests. After welding, the samples were transversally sectioned, prepared for metallographic examination according to the specific procedure. The samples were examined at the metallographic microscope, and photo records for each sectioned welding bead have been taken . Measurements have been made on the recorded photos resulting the essential characteristics of the penetration: width W, depth d and ratio W/d. From the obtained results the following conclusions can be formulated: the penetration depth of the end-plug weld at the TRIGA fuel element varies substantially depending on the material cast of which the plug is produced; the optimization tests had covered the whole range of parameters in which do not appear systematic defects in welds that are specific to the alloys of Nickel ( porosity, hot cracking); for 2011-2012 casts higher energy (640 As) is required compared to the welding energy used for the 2009 batch, but to be sure that the manufacturing requirements are fulfilled, it is necessary to carry

  19. 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

  20. Challenges in structural materials for thermal reactors

    International Nuclear Information System (INIS)

    The safe, reliable and economic operation of nuclear power plants is critically dependent on good performance of materials. Material selection is one of the most important steps in the design of the engineering components. Functional requirements and operating conditions such as stress, temperature and environment dictate the choice of materials and their properties. Criteria for selection of materials for non-nuclear components include mechanical properties (strength-ductility-toughness), corrosion properties (resistance to uniform and localized corrosion), fabricability (forming-welding-heat treatment) and cost. Additional considerations for nuclear components include neutron absorption, induced radioactivity and resistance to radiation damage. Nuclear core components are continuously subjected to bombardment by fast neutrons. Important effects of fast neutron irradiation on properties of materials are : (i) change in mechanical properties by irradiation hardening and irradiation embrittlement (ii) dimensional changes by irradiation growth, irradiation creep and void swelling and (iii) change in corrosion properties like increase in corrosion rate and introduction of new corrosion mode. There are 2 types of thermal power reactors viz., Pressure Vessel Type (BWRs and PWRs) and Pressure Tube Type (PHWRs). A very wide range of materials are used for nuclear reactor components. These include carbon and low alloy steels, stainless steels (austenitic, martensitic, precipitation hardenable), nickel-base alloys (inconels, incoloys), zirconium base alloys, etc. Three most important classes of materials are : (1) Low alloy steels for Reactor Pressure Vessel : the main concerns are increase in Ductile-Brittle transition temperature (DBTT) and decrease in toughness due to neutron irradiation. (2) Zirconium Alloys for Reactor Core Components : main concerns are Delayed Hydride cracking (DHC) and change in dimensions due to Irradiation Creep and Growth. (3) Stainless Steels

  1. 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

  2. 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

  3. Safety analysis calculations for research and test reactors

    International Nuclear Information System (INIS)

    The goal of the RERTR (Reduced Enrichment in Research and Test Reactor) Program at ANL is to provide technical means for conversion of research and test reactors from HEU (High-Enrichment Uranium) to LEU (Low-Enrichment Uranium) fuels. In exploring the feasibility of conversion, safety considerations are a prime concern; therefore, safety analyses must be performed for reactors undergoing the conversion. This requires thorough knowledge of the important safety parameters for different types of reactors for both HEU and LEU fuel. Appropriate computer codes are needed to predict transient reactor behavior under postulated accident conditions. In this discussion, safety issues for the two general types of reactors i.e., the plate-type (MTR-type) reactor and the rod-type (TRIGA-type) reactor, resulting from the changes associated with LEU vs. HEU fuels, are explored. The plate-type fuels are typically uranium aluminide (UAlx) compounds dispersed in aluminum and clad with aluminum. Moderation is provided by the water coolant. Self shut-down reactivity coefficients with EU fuel are entirely a result of coolant heating, whereas with LEU fuel there is an additional shut down contribution provided by the direct heating of the fuel due to the Doppler coefficient. In contrast, the rod-type (TRIGA) fuels are mixtures of zirconium hydride, uranium, and erbium. This fuel mixture is formed into rods ( ∼ 1 cm diameter) and clad with stainless steel or Incoloy. In the TRIGA fuel the self-shutdown reactivity is more complex, depending on heating of the fuel rather than the coolant. The two most important mechanisms in providing this feedback are: spectral hardening due to neutron interaction with the ZrH moderator as it is heated and Doppler broadening of resonances in erbium and U-238. Since these phenomena result directly from heating of the fuel, and do not depend on heat transfer to the moderator/coolant, the coefficients are prompt acting. Results of transient calculations

  4. 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

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

    International Nuclear Information System (INIS)

    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

  6. Failure of a TRIGA low enriched uranium (LEU) fuel during long term burnup testing in the Oak Ridge Reactor (ORR)

    International Nuclear Information System (INIS)

    The development of higher loaded LEU uranium zirconium-hydride fuel culminated in long term burn-up testing in the Oak Ridge Research Reactor (ORR). A sixteen rod cluster was designed as the test article containing ER-U-ZrH1.6 fuel pins clad with Incoloy 800. The nominal fuel pin geometry is 0.542 in. (13.77 mm) outside diameter, 22.0 in (559 mm) active length, with a cladding thickness of 0.016 in. (0.406 mm). Irradiation of the TRIGA-LEU cluster in the ORR began on 13 December 1979 and is still in progress. A total of 19 test rods have been included in the irradiation campaign; nine with 45 wt-% uranium (3.7 gms/cc), six with 30 wt-% and four with 20 wt-% - all are enriched to 19.7 percent in uranium 235. The irradiation of the 20 and 30 wt-% rods was terminated in May 1982 after the targeted burn-up levels of 35 and 40% respectively were successfully attained. One of the 45 wt-% rods (not the highest burnup rod) failed in November 1983 as a result of internal pressure causing the clad to split in a typical ductile failure. Of the two possible modes of failure - excess steam pressure or excess hydrogen pressure - steam pressure is the only really probable means of failure. Fuel temperatures would have had to exceed 1150 deg C to produce the necessary hydrogen pressure to cause failure and temperatures that high are extremely improbable without film blanketing at the clad surface. There was no evidence of film blanketing which would have caused obvious clad discoloration. The best explanation for the failure is that a very small clad leak developed in the region where the major failure subsequently occurred. The clad leak was likely the result of a clad imperfection manifesting itself after four years of irradiation. Fission gas release from the small leak resulted in a decision to shut down the ORR. The test fuel rod became water-logged during this shutdown period and during the subsequent startup of the reactor, steam pressure built up. As the power was

  7. 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)

  8. The effects of ID magnetite deposits on steam generator tube eddy current signals

    International Nuclear Information System (INIS)

    The presence of primary-side magnetite deposits in CANDU steam generator tubes poses a significant challenge to the analysis of eddy current inspection data. In principle, magnetite shields the tubing from the eddy current probe, diminishing its sensitivity to flaws. Since a probe's flaw detection and sizing performance is typically assessed with the use of clean laboratory samples, the shielding effect can become significant enough to affect both the probability of detection (POD) and sizing accuracy of the probe. Hence, there is a need to understand the relationship between inside diameter (ID) magnetite fouling and the actual flaw signals from typical flaws, such as pitting corrosion and fretting wear. The studies presented in this paper were performed with bobbin and X-probe data in both Inconel 600 and Incoloy 800 tubing. In comparison with clean tubes, ID deposits have a definite impact on both the detection and sizing capabilities of each probe. In all cases, the flaw signal amplitudes tend to decrease with increasing amounts of ID magnetite fouling. This also causes the probes to undersize the flaws in the fouled tubes. The resulting effect can be expressed in terms of a reduction factor versus a measured quantity with the bobbin probe, called Vshield, which can be used to determine the correction factor for sizing flaws. Given the difficulty in controlling the error sources in laboratory experiments on field-pulled tubes with deposits, the resulting scatter in the data makes the extraction of useful relationships between fouling and flaw signals difficult. This paper, therefore, presents an approach that consists of a combination of computer modelling, laboratory experiments on pulled tube samples, and field-data analysis to allow trends and relationships to be developed for a range of ID magnetite loading and thickness beyond that available in pulled tubes. A 3D electromagnetic finite element model was developed for studying the effects of ID magnetite

  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. 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

  11. 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.

  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. 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

  14. Embalse steam generators - status in 2009

    International Nuclear Information System (INIS)

    The Embalse Nuclear Generating Station (ENGS) is a CANDU 6, a pressurized heavy water plant, with a net capacity of 648 MW. The primary heat transport system at Embalse includes four Steam Generators (SGs) manufactured by Babcock and Wilcox Canada (B and W). These steam generators are vertical recirculating heat exchangers with Incoloy 800 inverted U-tubes and an integral preheater. Embalse SGs performed very well until the late 1990s, when an increase in tube fretting was noticed in the U-bend region. In-service inspection in 2002 and 2004 confirmed that the cause of the tube fretting was flow accelerated corrosion (FAC) damage of scallop bar supports in the U-bend region. The straight leg tube support plates (TSPs) have also been degrading. Degradation was worst at the top support plates, and it was in the form of material loss on the cold leg. The hot leg TSPs were heavily fouled with deposits and flow areas were blocked. Visual inspections and subsequent studies showed that the cause of the TSP degradation was also FAC. The Embalse SGs have carbon steel supports that make them susceptible to FAC. To mitigate the effects of degraded tube support structures, three additional sets of anti-vibration bars were installed in the U-bend regions of all four steam generators in 2004. In 2007, an improved secondary-side chemistry specification was implemented to reduce the FAC rate and the hot leg TSPs was waterlanced. A root cause analysis and condition assessment was performed for the tube supports in 2007. Fitness for Service (FFS) evaluation was completed using the Canadian Industry Guidelines for steam generator tubes. The steam generators were returned to service and the plant has operated without another forced outage to date. The FAC degradation of the carbon steel U-bend tube support systems has had the most significant impact on the plant operation causing a number of forced outages. The discovery of the extent of TSP degradation and difficulties to repair TSPs

  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

  16. Steam generator replacement as a part of a general problem management process

    International Nuclear Information System (INIS)

    effective strategy. The integrated Belgian approach to SG defect management results in a global optimization of the plant operation in terms of safety, reliability and economy. Applying this approach to the particular problem of PWSCC required a number of original developments -in such fields as non destructive testing and, probabilistic modelling of SG behaviour. It allowed after performance of a long term study to conclude that the most cost-effective strategy for Doel 3 was the immediate replacement of the SG combined with an up-rating of the plant. This study included a thorough comparison of the 2 candidate alloys for the tube material of the new SG: Inconel alloy 690 and Incoloy alloy 800. Finally the latter was selected in the peened condition because of its larger service experience and of favourable economic arguments