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Sample records for alloy-901 incoloy

  1. Intergranular stresses in Incoloy-800

    International Nuclear Information System (INIS)

    Holden, T.M.; Holt, R.A.; Clarke, A.P.

    1997-01-01

    The generation of intergranular residual strains under uniaxial loading conditions in the plastic regime has been measured in detail by neutron diffraction in Incoloy-800. A relatively simple theory, based on the Taylor model, gives a good semiquantitative account of the magnitudes of the strains. The results clarify the interpretation of measurements made earlier on Incoloy-800 steam generator tubes. (author)

  2. Some observations about the Incoloy 800 corrosion

    International Nuclear Information System (INIS)

    Baptista, W.; Sathler, L.; Mattos, O.R.

    1985-01-01

    The chemical and electrochemical characteristics of synthetic solutions similar to those inside the occluded cell corrosion - OCC (pitting, cracks from stress corrosion) of incoloy 800, 25 0 C are studied. (E.G.) [pt

  3. Microstructure effects and modelling of Incoloy 800 behaviour

    International Nuclear Information System (INIS)

    Dumaz, P.; Terriez, J.M.; Regnard, C.; Robert, G.

    1987-01-01

    At 550 0 C Incoloy 800 undergoes γ' hardening. The phase dispersion characteristics, which closely control the strength properties, are time dependent. The behaviour of Incoloy 800 has been studied with mechanical cycle tests. A microstructural study, from which an analysis of the γ' precipitate size effect on the strain phenomena and on rupture at high temperature could be made, was undertaken. A model is proposed which is able to describe the cycle consolidation phenomenon for low cycle fatigue and the fatigue stress relaxation tests, when using the characteristic parameters of the γ' precipitation. (U.K.)

  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. 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.

  6. 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.

  7. Fatigue tests and life estimation of Incoloy alloy 908

    International Nuclear Information System (INIS)

    Feng, J.; Toma, L.S.; Jang, C.H.; Steeves, M.M.

    1997-01-01

    Incoloy reg-sign alloy 908* is a candidate conduit material for Nb 3 Sn cable-in-conduit superconductors. The conduit is expected to experience cyclic loads at 4 K. Fatigue fracture of the conduit is one possible failure mode. So far, fatigue life has been estimated from fatigue crack growth data, which provide conservative results. The more traditional practice of life estimation using S-N curves has not been done for alloy 908 due to a lack of data at room and cryogenic temperatures. This paper presents a series of fatigue test results in response to this need. Tests were performed in reversed bending, rotating bending, and uniaxial fatigue machines. The test matrix included different heat treatments, two load ratios (R=-1 and 0.1), two temperatures (298 and 77 K), and two orientations (longitudinal and transverse). As expected, there is a semi-log linear relation between the applied stress and fatigue life above an applied stress (e.g., 310 MPa for tests at 298 K and R=-1). Below this stress the curves show an endurance limit. The aged and cold-worked materials have longer fatigue lives and higher endurance limits than the others. Different orientations have no apparent effect on life. Cryogenic temperature results in a much high fatigue life than room temperature. A higher tensile mean stress gives shorter fatigue life. It was also found that the fatigue lives of the reversed bending specimens were of the same order as those of the uniaxial test specimens, but were only half the lives of the rotating bending specimens for given stresses. A sample application of the S-N data is discussed

  8. Multi-response optimization of process parameters for TIG welding of Incoloy 800HT by Taguchi grey relational analysis

    OpenAIRE

    Srirangan, Arun; Paulraj, Sathiya

    2017-01-01

    Incoloy 800HT which was selected as one of the prominent material for fourth generation power plant can exhibit appreciable strength, good resistance to corrosion and oxidation in high temperature environment. This study focuses on the multi-objective optimization using grey relational analysis for Incoloy 800HT welded with tungsten inert arc welding process with N82 filler wire of diameter 1.2 mm. The welding input parameters play a vital role in determining desired weld quality. The experim...

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

    Czech Academy of Sciences Publication Activity Database

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

    -, 11 July 2015 (2015), s. 1-6 ISSN 0268-3768 R&D Projects: GA MŠk ED2.1.00/03.0082; GA MŠk(CZ) LO1406 Institutional support: RVO:68145535 Keywords : incoloy alloy 925 * abrasivewater jet turning * traverse speed * surface roughness Subject RIV: JQ - Machines ; Tools Impact factor: 1.568, year: 2015 http://link.springer.com/article/10.1007/s00170-015-7489-0

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

    Czech Academy of Sciences Publication Activity Database

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

    2016-01-01

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

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

    Czech Academy of Sciences Publication Activity Database

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

    2016-01-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

    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. - Highlights: • Trace PbO addition into the high temperature water block the formation of spinel oxides on Incoloy 800HT. • The donor density of oxide film decreases with trace PbO addition. • The current density of potentiodynamic polarization decreases of oxide film with trace PbO addition.

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

    International Nuclear Information System (INIS)

    Krenn, C.R.

    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 {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

  14. Study of Incoloy 800HT alloy tested by heat-cycling

    International Nuclear Information System (INIS)

    Velciu, L.; Meleg, T.; Pantiru, M.; Petrescu, D.; Voicu, F.

    2016-01-01

    This paper investigated Incoloy 800HT (UNS N08811) alloy after some heat-cycling tests. The study continues prior tests realized in INR Pitesti concerning utilization of some nickel-based alloys in the heat exchangers and steam generators construction. The thermal-cycling consist in a successive series of heating and cooling with some rates in a range temperature. Technical parameters of thermal cycling: 50 & 200 cycles, 25 °C/minute heating-cooling rate, temperature range 450-1000°C, and argon working medium. The analysis consisted in metallographic examination (microstructure), Vickers microhardness, and traction tests. The average grain size was determined by linear interception method (ASTM E-112). The micro hardness was calculated by the relationship of the device technical book. On the Strength-Deformation diagrams were obtained: tensile strength and elongation. The tested samples were compared with the ''as received'' material. The results showed a good metallographic and mechanical behaviour of Incoloy 800HT at these thermal-cycling tests. (authors)

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

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    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.

  16. The microstructure of Incoloy 800 H after long-time creep

    International Nuclear Information System (INIS)

    Sheng Zhongqi; Katerbau, K.

    1993-01-01

    The microstructural change of Incoloy 800 H after creep tests with low loads and long rupture time has been investigated. Cavities nucleate at one side of M 23 C 6 carbide particles on grain boundaries. Microcrack propagate by passing through a string of these cavities, M 23 C 6 carbide particles on grain boundaries have a coherent relationship with one of both neighbouring grains, so grain boundaries are strengthened, and the strengthening effect can be estimated for enhanced activation energy. G phase precipitation can be observed on grain boundaries, but no γ' phase particles can be found. Dislocation substructure is different from the typical recovery creep. Dislocation piles appear near M 23 C 6 carbide particles on grain boundaries. Subgrain structure poorly develop and network distribution of dislocation can remain after relative long creep

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

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    Dinu Alice

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Krompholz, K.; Ullrich, G.

    1991-02-01

    Precracked charpy-V-notch specimens of the iron-nickel base alloy Incoloy 800 in the as-received condition and after cold work have been tested using an instrumented impact tester (hammer) in the temperature range 293 ≤ T/K ≤ 1223. The specific impact energies were determined by dial readings, from the integration of the load versus time and the load versus load point displacement diagrams; in all cases the agreement was excellent. The specific impact energies and the impulses are correlated with the test temperature and with the degree of cold work, respectively. The dynamic fracture toughness values were determined following the equivalent energy approach. In all cases a distinct decrease of the mechanical properties in the range between the as-received state and after 5 % cold work was found. The temperature behaviour of the impact energies clearly reveals an increase of its value between room temperature and 673 K. This increase is distinctly reduced after cold work. The dynamic fracture toughness decreases with increasing temperature. The fracture surfaces clearly show elasto-plastic fracture behaviour of the material in the temperature regime investigated. (author) 19 figs., 3 tabs., 7 refs

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

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    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.

  20. Flux Entrapment and Titanium Nitride Defects in Electroslag Remelting of INCOLOY Alloys 800 and 825

    Science.gov (United States)

    Busch, Jonathan D.; deBarbadillo, John J.; Krane, Matthew J. M.

    2013-12-01

    Electroslag remelted (ESR) ingots of INCOLOY alloys 800 and 825 are particularly prone to macroscale slag inclusions and microscale cleanliness issues. Formation of these structures near the ingot surface can cause significant production yield losses (~10 pct) due to the necessity of extensive surface grinding. Slag inclusions from near the outer radius of the toe end of alloy 800 and 825 ingots were found to be approximately 1 to 3 mm in size and have a multiphase microstructure consisting of CaF2, CaTiO3, MgAl2O4, MgO, and some combination of Ca12Al14O32F2 and/or Ca12Al14O33. These inclusions were often surrounded by fields of 1- to 10- μm cuboidal TiN particles. A large number of TiN cuboids were observed in the ESR electrode with similar size and morphology to those observed surrounding slag inclusions in the ESR ingots, suggesting that the TiN particles are relics from the ESR electrode production process. Samples taken sequentially throughout the AOD processes showed that the TiN cuboidals that are found in ESR ingots form between tapping the AOD vessel into the AOD ladle and the casting of ESR electrodes.

  1. 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)

    Fulger, M.; Lucan, D.; Radulescu, M.; Velciu, L.; Demetrescu, I.

    2001-01-01

    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 H 2 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

  2. Stress corrosion cracking initiation of oxidized Incoloy 800 in caustic environment

    International Nuclear Information System (INIS)

    Dinu, A.; Chicinas, I.; Abrudeanu, M.; Velciu, L.; Ionescu, D.; Stanciulescu, M.; Chicinas, I.; Abrudeanu, M.; Dinu, A.

    2013-01-01

    With the purpose to study the influence of the oxide layer in the first step of the stress corrosion cracking (SCC) mechanism, some oxidized and non-oxidized C-rings cuts from Incoloy 800 tubes were tested in 10% sodium hydroxide solutions (ph=13) at 260 O C and 50 atm, for 57 days. We used in our experiments C-rings because this type of samples have small dimensions and may be stressed using simple methods and in this way they can be exposed in nearly any type of medium. To create some stress concentrations, a mechanical pre-crack of 100.m depth was executed on the external side of the C-rings. The value of the appeared stresses at mechanical crack tip was evaluated using ANSYS code. The oxidized C-rings were obtained by autoclaving for 20 days in demineralised water adjusted with hydrazine (ph=9.7), at 260 O C and 50 atm. By X-rays diffraction there are emphasized iron dichromium oxide, nickel dichromium oxide, magnetite and hematite. Using the scanning electron microscopy we distinguished the presence of a double layer: the inner layer compact and adherent, containing fine grains associated with iron dichromium oxide, nickel dichromium oxide and the outer layer composed of other greater and less uniform particles associated with magnetite and hematite. After the SCC test, it was observed that the presence of the oxide layer led to a transgranulary SCC initiation, direct from a fissure in the oxide layer; the length of the SCC cracks is about 80-100 microns. In the absence of the oxide layer, the SCC cracks start intergranulary, from a local corrosive attack zone and their depth is approximately 200 microns. In both cases the SCC cracks have transgranular propagation. (authors)

  3. Multi-response optimization of process parameters for TIG welding of Incoloy 800HT by Taguchi grey relational analysis

    Directory of Open Access Journals (Sweden)

    Arun Kumar Srirangan

    2016-06-01

    Full Text Available Incoloy 800HT which was selected as one of the prominent material for fourth generation power plant can exhibit appreciable strength, good resistance to corrosion and oxidation in high temperature environment. This study focuses on the multi-objective optimization using grey relational analysis for Incoloy 800HT welded with tungsten inert arc welding process with N82 filler wire of diameter 1.2 mm. The welding input parameters play a vital role in determining desired weld quality. The experiments were conducted according to L9 orthogonal array. The input parameter chosen were the welding current, Voltage and welding speed. The output response for quality targets chosen were the ultimate tensile strength and yield strength (at room temperature, 750 °C and impact toughness. Grey relational analysis was applied to optimize the input parameters simultaneously considering multiple output variables. The optimal parameters combination was determined as A2B1C2 i.e. welding current at 110 A, voltage at 10 V and welding speed at 1.5 mm/s. ANOVA method was used to assess the significance of factors on the overall quality of the weldment. The output of the mechanical properties for best and least grey relational grade was validated by the metallurgical characteristics:

  4. Resistance of Incoloy 800 steam generator tube to pitting corrosion in PWR secondary water at 250°C

    Energy Technology Data Exchange (ETDEWEB)

    Schvartzman, Mônica M.A.M. [Pontifícia Universidade Católica de Minas Gerais (PUC-Minas), Belo Horizonte, MG (Brazil); Albuquerque, Adriana Silva de; Esteves, Luiza; Rabello, Emerson G.; Mansur, Fábio Abud, E-mail: monicacdtn@gmail.com, E-mail: asa@cdtn.br, E-mail: luiza.esteves@cdtn.br, E-mail: egr@cdtn.br, E-mail: fametalurgica@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    The steam generator (SG) is one of the main components of a PWR, so the performance of this type of nuclear power plant depends to a large extent on the trouble-free operation of SGs. Its degradation significantly affects the overall plant performance. Alloy 800NG (Incoloy® 800) is a nickel-iron-chromium alloy used for steam generator tubes in PWRs due to their high strength, good workability and resistance to corrosion. This behavior is attributed to the protective oxide film formed on the metal surface by contact with the high temperature pressurized water. However, chloride is one of major SG impurities that cause the breakdown of the passive film and initiate localized corrosion in passive metals as Alloy 800NG. The aim of this study is to provide information about the pitting corrosion behavior of the Incoloy® 800 steam generator tube under normal secondary circuit parameters (250 deg C and 5 MPa) and abnormal conditions of operation (presence of chloride ions in the secondary water). For this, optical microscopy, XRD and EDS analysis and electrochemical tests have been carried out under simulated PWR secondary water operating conditions. The susceptibility to pitting corrosion was evaluated using electrochemical tests and the oxide layer formed on material was examined by means of scanning electron microscopy (SEM) and energy dispersive X-ray spectrometer (EDS) analyses. (author)

  5. Creep and fatigue properties of Incoloy 800H in a high-temperature gas-cooled reactor (HTGR) helium environment

    International Nuclear Information System (INIS)

    Chow, J.G.Y.; Soo, P.; Epel, L.

    1978-01-01

    A mechanical test program to assess the effects of a simulated HTGR helium environment on the fatigue and creep properties of Incoloy 800H and other primary-circuit metals is described. The emphasis and the objectives of this work are directed toward obtaining information to assess the integrity and safety of an HTGR throughout its service life. The helium test environment selected for study contained 40 μ atm H 2 O, 200 μ atm H 2 , 40 μ atm CO, 10 μ atm CO 2 , and 20 μ atm CH 4 . It is believed that this ''wet'' environment simulates that which could exist in a steam-cycle HTGR containing some leaking steam-generator tubes. A recirculating helium loop operating at about 4 psi in which impurities can be maintained at a constant level, has been constructed to supply the desired environment for fatigue and creep testing

  6. 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)

    Travalloni, A.M.; Sathler, L.; Mattos, O.R.

    1985-01-01

    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) [pt

  7. The effect of borate and phosphate inhibitors on corrosion rate material SS321 and incoloy 800 in chloride containing solution by using potentiodynamic method

    International Nuclear Information System (INIS)

    Febriyanto; Sriyono; Satmoko, Ari

    1998-01-01

    Determination of corrosion rate of steam generator materials (SS 321 and incoloy 800) in chloride containing solution using potentiodynamic method from CMS 100. NaCl 1%, 3% and 5% solution using is used as tested solution. A tested material is grounded by grinding paper on grade 400 600, 800 and 1000, then polished by METADI 1/4 microns paste to get homogeneity. Furthermore, the tested materials is mounted by epoxide resin, so only the surface which contacts to tested solution is open. From the result obtained that borate and phosphate inhibitor can reduce corrosion rate and aggressiveness of chloride ion

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

    International Nuclear Information System (INIS)

    Lucan, D.; Fulger, M.; Savu, G.; Velciu, L.

    2004-01-01

    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

  9. Multi-response optimization of process parameters for GTAW process in dissimilar welding of Incoloy 800HT and P91 steel by using grey relational analysis

    Science.gov (United States)

    vellaichamy, Lakshmanan; Paulraj, Sathiya

    2018-02-01

    The dissimilar welding of Incoloy 800HT and P91 steel using Gas Tungsten arc welding process (GTAW) This material is being used in the Nuclear Power Plant and Aerospace Industry based application because Incoloy 800HT possess good corrosion and oxidation resistance and P91 possess high temperature strength and creep resistance. This work discusses on multi-objective optimization using gray relational analysis (GRA) using 9CrMoV-N filler materials. The experiment conducted L9 orthogonal array. The input parameter are current, voltage, speed. The output response are Tensile strength, Hardness and Toughness. To optimize the input parameter and multiple output variable by using GRA. The optimal parameter is combination was determined as A2B1C1 so given input parameter welding current at 120 A, voltage at 16 V and welding speed at 0.94 mm/s. The output of the mechanical properties for best and least grey relational grade was validated by the metallurgical characteristics.

  10. Corrosion of Inconel-625, Hastelloy-X280 and Incoloy-800 in 550 - 750°C superheated steam. Influence of alloy heat treatment, surface treatment, steam temperature and steam velocity. Part I: Results up to 6000 hours exposure time. RCN Report

    International Nuclear Information System (INIS)

    Tilborg, P.J. van; Linde, A. van der

    1969-10-01

    Sheet samples of Inconel-625, Hastelloy-X280 and Incoloy-800 were tested, in the solution annealed and in the solution annealed + 20% cold worked + 800°C tempered condition, in steam with a velocity of 5 m/sec. at 550, 650 and 750°C and in steam with a volocity of 15 and 85 m/sec. at 550°C. At 550°C and 750°C the samples were tested in the heat treated, annealed or tempered and the heat treated + electropolished condition. At 650°C moreover as heat treated + ground and pickled samples were tested. Post-corrosion sample investigations involved measurement of the adherent oxide thickness, the total amount of corroded metal, the metal loss to system, and the metallographic and microprobe investigation of the adherent oxide film and adjacent diffusion disturbed alloy layer. The results obtained up to 6000 hours exposure time showed that the surface treatment has a decisive influence on the corrosion behaviour of all three alloys tested. The differences in the corrosion data for the two heat treatment conditions are small. The influence of the steam velocity, as tested at 550°C, on the initial corrosion rate was surprisingly high, while the long-term linear corrosion rates are only slightly influenced by the gas velocity. In general the linear corrosion rates were low, 1-5 mg/dm 2 month, and not consistently affected by the test-temperature. The metal loss to system values were 2 <15 mg/dm 2 in the low velocity steam at all three test temperatures and <30 mg/dm 2 in the high velocity steam at 550°C. The metallographic and microprobe examinations revealed no remarkable results, as compared with the results of analogous tests reported in literature. (author)

  11. Mechanisms of Recovering Low Cycle Fatigue Damage in Incoloy 901.

    Science.gov (United States)

    1979-01-01

    IF (KI.L=E.O) -IYPLOT=.F. 004190 IF (HYPLM.oAN.COI1PL3T) O’JALPT=.T. 0C4200 c0 OC4210 PR.INT 6, COMPLOT ,HYPLOT,DUALPT CE4220 6 FORMAT (/ * COMPLOT ~,L3...rCYAX)006450 CALL SCLoIc(c.5) 066 IF ( COMPLOT ) CALL CURVE CXY,IT,-1) 006470 IF (HYPLOT) CALL CURVEXA,YAJT,-1) 0C6480 CALL RES=ET(-8LNKl") Ca6490 CALL...007250 CIL (HYPLOT) CIALt42,1.0,Z.0) ,-I 007160 CALL RENOPI (-!I𔃾Ol-) 007270 CALL RESET (ŗLIGH) 0(7200 04CANLNU 0(PI710)W30 IF ( COMPLOT ) CALL L1)N.INY

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

    NARCIS (Netherlands)

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

    1987-01-01

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

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

    NARCIS (Netherlands)

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

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    Penkalla, H.J.; Nickel, H.; Schubert, F.

    1987-01-01

    This paper is concerned with the description of creep under triaxial loading with the aid of Norton's law and v. Mises' theory. The load combination of internal pressure, tension and torsion is more closely scrutinized. The superposition of cyclic stresses in the tensile threshold range is discussed as well as the instances of partial relaxation. In accordance with the discussed loading modes, experimental results are presented which, until now, have delivered a satisfactory agreement between theory and practice. A more explicit analysis of the creep-induced damage is given regarding the assessment of the expected life until failure under triaxial loading. This delivers information concerning a suitable test stress for the long-term stress assessment. (orig./DG) [de

  15. Electrochemical investigations of the resistance of Inconel 600, Incoloy 800, and Type 347 stainless steel to pitting corrosion in faulted PWR secondary water at 150 to 250 C

    International Nuclear Information System (INIS)

    Hickling, J.; Wieling, N.

    1981-01-01

    Determinations of critical pitting potentials were carried out at temperatures of 150, 200, and 250 C on three corrosion resistant alloys important in the construction of nuclear steam generators. The results are assessed in terms of the effects on pitting resistance of chloride (as NaCl) and phosphate (as Na/sub 2/ HPO/sub 4/) contents of the water and test temperature. A comparison of material behavior was obtained for each set of test conditions. 18 refs

  16. Design and Evaluation of Dual-Expander Aerospike Nozzle Upper Stage Engine

    Science.gov (United States)

    2014-09-18

    Materials Options for DEAN Components Component Materials Chamber Structural Jacket Copper, INCOLOY 909, INCONEL 718, INCONEL 625 , Alloy 188, Aluminum 7075...T6, Aluminum 2024 T6, Oxygen-Free Copper, Titanium, Cobalt Chamber Cooling Jacket Copper, Silicon Carbide, INCOLOY 909, INCONEL 718, INCONEL 625 , Alloy...Carbide, INCOLOY 909, Alloy 188, Cobalt LOX Plumbing Copper, INCOLOY 909, INCONEL 718, INCONEL 625 , Alloy 188, Oxygen-Free Copper, Cobalt LH2 Plumbing

  17. Use of ferritic steels in breeder reactors worldwide

    International Nuclear Information System (INIS)

    Patriarca, P.

    1983-01-01

    The performance of LMFBR reactor steam generator materials is reviewed. Tensile properties of stainless steel-304, stainless steel-316, chromium-molybdenum steels, and Incoloy 800H are presented for elevated temperatures

  18. Optimized Dual Expander Aerospike Rocket

    Science.gov (United States)

    2011-03-01

    O2 plumbing INCONEL ® 625 (Annealed) Aluminum 7075 T6 Not compatible with O2 or H2 / Useable for chamber and aerospike structural jacket as long as...7075 T6 Aluminum 2024 T6 Beryllium- Copper Oxygen- Free Copper Titanium Cobalt Pure Copper Silicon Carbide INCOLOY 909 INCONEL 718 INCONEL 625 Alloy...Aluminum 2024 T6 Beryllium- Copper Oxygen- Free Copper Titanium Cobalt Pure Copper Silicon Carbide INCOLOY 909 INCONEL 718 INCONEL 625 Alloy 188 Aluminum

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

    NARCIS (Netherlands)

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

    1989-01-01

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

  20. Growth kinetics of boride layers formed on 99.0% purity nickel

    Indian Academy of Sciences (India)

    cobalt alloys and most refractory alloys.4–13 The boronizing process is a prominent choice for a wide range of .... borided pure nickel and Nimonic 90 superalloy. In their experimental study, Nimonic 90 superalloy ... 400 and 2500 HV0.1) for borided Nimonic 90 superalloy. Aytekin and Akcin29 have treated the Incoloy 825 ...

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

    NARCIS (Netherlands)

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

    1984-01-01

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

  2. Considerations in selecting tubing materials for CANDU steam generators

    International Nuclear Information System (INIS)

    Hemmings, R.L.

    1978-01-01

    Corrosion resistance is the major consideration in selecting tubing material for CANDU steam generators. Corrosion, and additional considerations, lead to the following steam generator tubing material recommendations: for CANDU-BPHWR's (boiling pressurized heavy water reactors) low-cobalt Incoloy-800; for CANDU-PHWR's (pressurized, non-boiling, heavy water reactors), low-cobalt Monel-400

  3. Vacuum Gas Tungsten Arc Welding

    Science.gov (United States)

    Weeks, J. L.; Todd, D. T.; Wooten, J. R.

    1997-01-01

    A two-year program investigated vacuum gas tungsten arc welding (VGTAW) as a method to modify or improve the weldability of normally difficult-to-weld materials. After a vacuum chamber and GTAW power supply were modified, several difficult-to-weld materials were studied and key parameters developed. Finally, Incoloy 903 weld overlays were produced without microfissures.

  4. Effects of HTGR helium on the high cycle fatigue of structural materials

    International Nuclear Information System (INIS)

    Soo, P.; Sabatini, R.L.; Gerlach, L.

    1982-01-01

    High cycle fatigue tests have been conducted on Incoloy 800H and Hastelloy X in air and in HTGR helium environments containing low and high levels of moisture. For the helium environments, a higher mositure level usually gives a lower fatigue strength. For air, however, the strength is usually much lower than those for helium. For long test times at higher test temperatures, the fatigue strengths for Incoloy 800H often show a large decrease, and the fatigue limits are much lower than those anticipated from low cycle tests. Optical and scanning electron microscope observations were made to correlate fatigue life with surface and bulk microstructural changes in the material during test. Oxide scale cracking and spallation, surface recrystallization and intergranular attack appear to contribute to losses in fatigue strength

  5. Fatigue and creep-fatigue behaviour of high-temperature alloys for HTR-application

    International Nuclear Information System (INIS)

    Meurer, H.-P.; Breitling, H.; Grosser, E.D.

    1981-01-01

    The development of High Temperature Gas Cooled Reactors requires the evaluation of the fatigue behaviour of those alloys which have been taken into account for possible use as structural materials. Comparative fatigue tests of six wrought alloys at 850 0 C revealed differences especially at low strain ranges. The influence of the coolant gas on Incoloy 800 H and Inconel 617 resulted in an increased fatigue life and for Incoloy 800 H in changes of the deformation behaviour. Hold times introduced at maximum tensile strain reduced fatigue life considerably. The hold time data have been evaluated following the rules of ASME Code Case N 47 and design curves for inelastic and elastic analysis are suggested. (Auth.)

  6. Characterization of corrosion deposits on components of candu steam generators

    International Nuclear Information System (INIS)

    Dinu, A.; Mincu, M.; Stanciulescu, M.; Diniasi, D.

    2015-01-01

    X-ray diffraction at small angles of incidence was used to investigate oxidised samples obtained from Incoloy 800, steel 516 and Inconel 600, materials used to manufacture different components of CANDU steam generators. The samples were exposed for 157 days in demineralised water with LiOH (pH=10.5), at 3100C and 5MPa. In the case of 516 steel it was observed that the magnetite is formed at interface metal-oxide, while the external layer is formed preponderant from hematite particles. The external layer present on the surface of Inconel 600 is formed by nickel hydroxide, while the spinelic compounds such as (Cr, Fe)3O4 are distributed in the inner layer. Because the oxide formed on the surface of Incoloy 800 sample is very thin, there is not major difference regarding the composition in the depth profile. The X-rays results were sustained by the SEM results. (authors)

  7. 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)

  8. Characterization and structural integrity tests of ex-service steam generator tubes at Ontario Power Generation

    Energy Technology Data Exchange (ETDEWEB)

    Pagan, Sandra [Ontario Power Generation, 889 Brock Road, Pickering, Ontario (Canada); Duan Xinjian [Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, Ontario (Canada)], E-mail: duanx@aecl.ca; Kozluk, Michael J. [Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, Ontario (Canada); Mills, Brian; Goszczynski, Guylaine [Kinectrics Inc., 800 Kipling Avenue, Toronto, Ontario (Canada)

    2009-03-15

    The Canadian Nuclear Standard CSA N285.4 requires the periodic metallurgical examination of removed ex-service steam generator tubes. This paper describes the practices used for the characterization and structural integrity tests of ex-service steam generator tubes at Ontario Power Generation (OPG). It shows that there is no degradation of mechanical properties of Monel 400 tubes after 7-18 effective full power years (EFPY) of operation and Incoloy 800 tubes after more than 10 EFPY of operation.

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

    International Nuclear Information System (INIS)

    Baba, Shinichi; Kurasawa, Toshimasa; Endow, Taichi; Someya, Hiroyuki; Tanaka, Isao.

    1986-08-01

    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)

  10. Nickel-plating for active metal dissolution resistance in molten fluoride salts

    Energy Technology Data Exchange (ETDEWEB)

    Olson, Luke [Department of Engineering Physics, 1500 Engineering Drive, University of Wisconsin, Madison, WI 53706 (United States); Sridharan, Kumar, E-mail: kumar@engr.wisc.edu [Department of Engineering Physics, 1500 Engineering Drive, University of Wisconsin, Madison, WI 53706 (United States); Anderson, Mark; Allen, Todd [Department of Engineering Physics, 1500 Engineering Drive, University of Wisconsin, Madison, WI 53706 (United States)

    2011-04-15

    Ni electroplating of Incoloy-800H was investigated with the goal of mitigating Cr dissolution from this alloy into molten 46.5%LiF-11.5%NaF-42%KF eutectic salt, commonly referred to as FLiNaK. Tests were conducted in graphite crucibles at a molten salt temperature of 850 deg. C. The crucible material graphite accelerates the corrosion process due to the large activity difference between the graphite and the alloy. For the purposes of providing a baseline for this study, un-plated Incoloy-800H and a nearly pure Ni-alloy, Ni-201 were also tested. Results indicate that Ni-plating has the potential to significantly improve the corrosion resistance of Incoloy-800H in molten fluoride salts. Diffusion of Cr from the alloy through the Ni-plating does occur and if the Ni-plating is thin enough this Cr eventually dissolves into the molten salt. The post-corrosion test microstructure of the Ni-plating, particularly void formation was also observed to depend on the plating thickness. Diffusion anneals in a helium environment of Ni-plated Incoloy-800H and an Fe-Ni-Cr model alloy were also investigated to understand Cr diffusion through the Ni-plating. Further enhancements in the efficacy of the Ni-plating as a protective barrier against Cr dissolution from the alloy into molten fluoride salts can be achieved by thermally forming a Cr{sub 2}O{sub 3} barrier film on the surface of the alloy prior to Ni electroplating.

  11. Creep behavior of materials for high-temperature reactor application

    International Nuclear Information System (INIS)

    Schneider, K.; Hartnagel, W.; Iischner, B.; Schepp, P.

    1984-01-01

    Materials for high-temperature gas-cooled reactor (HTGR) application are selected according to their creep behavior. For two alloys--Incoloy-800 used for the live steam tubing of the thorium high-temperature reactor and Inconel-617 evaluated for tubings in advanced HTGRs--creep curves are measured and described by equations. A microstructural interpretation is given. An essential result is that nonstable microstructures determine the creep behavior

  12. Liquid metal fast breeder reactor steam generator: behaviour of heat exchange tubes in face of a through crack resulting in a contact between sodium and water

    International Nuclear Information System (INIS)

    Quinet, J.L.; Lannou, L.

    1978-01-01

    The results of a survey made Electricite de France on the behaviour of cracked tubes under operating conditions of an industrial steam generator are submitted in this communication. A comparison is made of the tube material: INCOLOY 800, 2 1/4 Cr-1 Mo, 9 Cr-2 Mo land to the initial leak. Finally, a description is given of the self-development process of a water leak into sodium. (author)

  13. Corrosion of heat exchanger materials under heat transfer conditions

    International Nuclear Information System (INIS)

    Tapping, R.L.; Lavoie, P.A.; Disney, D.J.

    1986-01-01

    Severe pitting has occurred in moderator heat exchangers tubed with Incoloy-800 in Pickering Nuclear Generating Station. The pitting originated on the cooling water side (outside) of the tubes and perforation occurred in less than two years. It was known from corrosion testing at Chalk River Laboratories 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-28. 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. (author). 3 refs., 4 tabs., 6 figs

  14. Corrosion of heat exchanger materials under heat transfer conditions

    International Nuclear Information System (INIS)

    Tapping, R.L.; Lavoie, P.A.; Disney, D.J.

    1987-01-01

    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-28 1 . 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

  15. Creep properties of superalloys for the HTGR in impure helium environments

    International Nuclear Information System (INIS)

    Kawakami, H.; Nakanishi, T.

    1981-01-01

    This paper describes creep behaviors of two heat resistant alloys, Hastelloy X and Incoloy 800, in helium environments of the HTGR. In impure helium environments, these alloys are susceptible to carburization and oxidization. We have investigated these effects separately, and related them to the creep behaviors of the alloys. Experiments were carried out at 900 0 C both in helium and in air. Carburization results in decrease of secondary creep strain rate and delay of tertiary creep initiation. Oxidization caused decrease in tertiary creep strain rate of Hastelloy X, but did not that of Incoloy 800. Enhancement in tertiary creep strain rate of Hastelloy X in a very weakly oxidizing environment was confirmed in creep crack growth experiment using notched plate specimens. The rupture time of Hastelloy X in helium was short when compared with in air. Stress versus rupture time curves for both environments were parallel up to 5000 hours test, and a ratio of rupture stress in helium to that in air was about 0.9. In case of Incoloy 800, rupture time in helium was markedly prolonged as compared with that in air. (orig.)

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

    International Nuclear Information System (INIS)

    Levy, R.

    1982-01-01

    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 [fr

  17. Properties and application study of Inconel alloy tube made in China

    International Nuclear Information System (INIS)

    Yang Xiang; Su Xingwan; Wen Yan

    1997-01-01

    The mech-physical properties and the corrosion resistance properties of the SG tube of Inconel alloy made in China under any conditions are briefly presented, and the test and research for bending and expending the tubes have been performed. In the process of corrosion experiments the Inconel alloy tubes were compared with that of the same kind of materials made in foreign countries. The Inconel alloy tubes have better stress corrosion resistance cracking prosperities than Inconel 600 and Incoloy 800 when they were in the solutions which contained high concentrated chlorine ion and alkali at high temperature

  18. Superalloy applications in the nuclear field

    International Nuclear Information System (INIS)

    Ramanathan, L.V.; Padilha, A.F.

    1984-01-01

    The process conditions in the areas of nuclear fuel processing, fabrication, utilization, reprocessing and disposal are severe, demanding therefore the use of materials with high temperature mechanical strength and corrosion resistance. A number of refractory metal containing superalloys have found application in the diferrent areas of the nuclear field. The main aspects of the microstructure, strengthening mechanisms and corrosion resistance of 3 superalloys, namely Incoloy 825, Inconel 718 and Hastelloy C have been discussed. The role of the refractory metal elements in influencing the mechanical strength and corrosion resistance of superalloys has been emphasised. (Author) [pt

  19. Fracture mechanics characterization of crack growth under creep and fatigue conditions

    International Nuclear Information System (INIS)

    Hollstein, T.; Kienzler, R.

    1987-01-01

    Based on theoretical considerations and experimental evidence, several concepts have been investigated to correlate crack growth under high-temperature conditions with material parameters of fracture mechanics, e.g., stress-intensity factor K, and line integrals J and C * . It could be shown for different materials that these parameters describe the behavior of cracks independent of geometry and loading conditions within certain limits of validity. The laboratory results then can be transferred to real structures for (residual) lifetime predictions or safety analyses (Incoloy 800H, Thermon 4972, 9% chromium steel). With 45 refs., 3 tabs., 56 figs [de

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

    International Nuclear Information System (INIS)

    Langevoort, J.C.

    1985-01-01

    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.)

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

    International Nuclear Information System (INIS)

    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

  2. Fatigue damage in superalloys determined using Doppler broadening positron annihilation

    Science.gov (United States)

    Hoeckelman, Donald; Leighly, H. P., Jr.

    1990-01-01

    Axial fatigue specimens of three superalloys, Inconel 718, Incoloy 903 and Haynes 188, were machined from solution-heat-treated material and artificially aged. They were subjected to cyclic loading for a selected number of cycles after which the S parameter was determined using Doppler broadening positron annihilation. Initially, the S parameter decreased, followed by a large increase and a subsequent decline leading to fracture. This has been interpreted as the removal of residual vacancies, the introduction of new defects by cyclic loading, and, finally, a clustering of the defects as microcracks which grow to cause failure.

  3. Current CTR-related tritium handling studies at ORNL

    International Nuclear Information System (INIS)

    Watson, J.S.; Bell, J.T.; Clinton, S.D.; Fisher, P.W.; Redman, J.D.; Smith, F.J.; Talbot, J.B.; Tung, C.P.

    1976-01-01

    The Oak Ridge National Laboratory has a comprehensive program concerned with plasma fuel recycle, tritium recovery from blankets, and tritium containment in fusion reactors. Two studies of most current interest are investigations of cryosorption pumping of hydrogen isotopes and measurements of tritium permeation rates through steam generator materials. Cryosorption pumping speeds have been measured for hydrogen, deuterium, and helium at pressures from 10 -8 torr to 3 x 10 -3 torr. Permeation rates through Incoloy 800 have been shown to be drastically reduced when the low pressure side of permeation tubes are exposed to steam. These results will be important considerations in the design of fusion reactor steam generators

  4. Modelling the long-term corrosion behaviour of candidate alloys for Canadian SCWR

    Energy Technology Data Exchange (ETDEWEB)

    Steeves, G.; Cook, W., E-mail: wcook@unb.ca, E-mail: graham.steeves@unb.ca [University of New Brunswick, Department of Chemical Engineering, Fredericton, NB (Canada)

    2015-07-01

    Corrosion behaviour of Inconel 625 and Incoloy 800H, two of the candidate fuel cladding materials for Canadian supercritical water (SCW) reactor designs, were evaluated by exposing the metals to SCW in UNB's SCW flow loop. Individual experiments were conducted over a range of 370{sup o}C and 600{sup o}C. Exposure times were typically intervals of 100, 250, and 500 hours. Experimental data was used to create an empirical kinetic equation for each material. Activation energies for the alloys were determined, and showed a distinct difference between low-temperature electrochemical corrosion mechanism and direct high-temperature chemical oxidation. (author)

  5. Analysis of the factors that impact the reliability of high level waste canister materials

    International Nuclear Information System (INIS)

    Boyd, W.K.; Hall, A.M.

    1977-01-01

    The analysis encompassed identification and analysis of potential threats to canister integrity arising in the course of waste solidification, interim storage at the fuels reprocessing plant, wet and dry shipment, and geologic storage. Fabrication techniques and quality assurance requirements necessary to insure optimum canister reliability were considered taking into account such factors as welding procedure, surface preparation, stress relief, remote weld closure, and inspection methods. Alternative canister materials and canister systems were also considered in terms of optimum reliability in the face of threats to the canister's integrity, ease of fabrication, inspection, handling and cost. If interim storage in air is admissible, the sequence suggested comprises producing a glass-type waste product in a continuous ceramic melter, pouring into a carbon steel or low-alloy steel canister of moderately heavy wall thickness, storing in air upright on a pad and surrounded by a concrete radiation shield, and thereafter placing in geologic storage without overpacking. Should the decision be to store in water during the interim period, then use of either a 304 L stainless steel canister overpacked with a solution-annealed and fast-cooled 304 L container, or a single high-alloy canister, is suggested. The high alloy may be Inconel 600, Incoloy Alloy 800, or Incoloy Alloy 825. In either case, it is suggested that the container be overpacked with a moderately heavy wall carbon steel or low-alloy steel cask for geologic storage to ensure ready retrievability. 19 figs., 5 tables

  6. Tension layer winding of cable-in-conduit conductor

    International Nuclear Information System (INIS)

    Devernoe, A.; Ciancetta, G.; King, M.; Parizh, M.; Painter, T.; Miller, J.

    1996-01-01

    A 710 mm i.d. by 440 mm long, 6 layer Cable-in-Conduit (CIC) coil was precision tension layer wound with Incoloy 908 jacketed conductor to model winding technology that will be used for the Nb 3 Sn outsert coils of the 45 Tesla Hybrid Magnet Project at the US National High Magnetic Field Laboratory. This paper reports on the set up of a new winding facility with unique capabilities for insulating and winding long length CIC conductor and on special procedures which were developed to wind and support layer to layer transitions and to safely form conductor into and out of the winding. Analytical methods used to predict conduit keystoning, springback and back tensioning requirements before winding are reported in comparison to results obtained during winding and actual winding build-up dimensions on a layer by layer basis in comparison to design requirements

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

    International Nuclear Information System (INIS)

    Korzan, M.A.

    1994-05-01

    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

  8. Arc-discharge system for nondestructive detection of flaws in thin ceramic coatings

    International Nuclear Information System (INIS)

    Scott, G.W.; Davis, E.V.

    1978-04-01

    The feasibility of nondestructively detecting small cracks or holes in plasma-sprayed ceramic coatings with an electric arc-discharge system was studied. We inspected ZrO 2 coatings 0.46 mm (0.018 in.) thick on Incoloy alloy 800 substrates. Cracks were artificially induced in controlled areas of the specimens by straining the substrates in tension. We designed and built a system to scan the specimen's surface at approximately 50 μm (0.002 in.) clearance with a sharp-pointed metal-tipped probe at high dc potential. The system measures the arc currents occurring at flaws, or plots a map of the scanned area showing points where the arc current exceeds a preset threshold. A theoretical model of the probe-specimen circuit shows constant dc potential to be the best choice for arc-discharge inspection of insulating coatings. Experimental observations and analysis of the data disclosed some potential for flaw description

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

    International Nuclear Information System (INIS)

    Rudelli, M.D.

    1979-04-01

    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.) [es

  10. Measurements of residual stresses and textures by neutron diffraction

    International Nuclear Information System (INIS)

    Hayashi, Makoto

    2008-01-01

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

  11. Theoretical-experimental assessment of the variables affecting fretting of Atucha I nuclear power plant utility steam generators tubes

    International Nuclear Information System (INIS)

    Kulichevsky, Raul M.

    1995-01-01

    Fretting wear of Steam Generator tubes caused by flow induced vibrations generates uncertainty on their integrity. The knowledge of the controlling variables of the wear process may give a criterion to evaluate the tubes residual life. Information on vibratory response and dynamic interaction between tubes and their supports are prerequisites for understanding the relationship between fretting wear and tube vibration. Experimental results of the vibratory response of an Atucha-I nuclear power plant type U-tube, the influence of tube/support clearance on this response and a study of tube/support dynamic interaction, which allow the verification of a finite element model of this type of tubes, are presented in this work. Also wear results for the Incoloy 800/DIN 1.4550 austenitic stainless steel pair of materials and a first evaluation of the wear constant of this pair are presented. (author)

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

    International Nuclear Information System (INIS)

    O'Neal, W.C.; Gregg, D.W.; Hockman, J.N.; Russell, E.W.; Stein, W.

    1984-01-01

    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

  13. Incology alloy 908 data handbook

    Energy Technology Data Exchange (ETDEWEB)

    Toma, L.S.; Steeves, M.M. [Massachusetts Institute of Technology, Cambridge, MA (United States); Reed, R.P. [Cryogenic Materials Inc., Boulder, CO (United States)

    1994-03-01

    This handbook is a compilation of all available properties of Incoloy alloy 908 as of March, 1994. Data included in this paper cover mechanical, elastic, thermal and magnetic characteristics. The mechanical properties include tensile, fracture toughness, fatigue, and stress-rupture for both the base metal and related weld filler metals. Elastic properties listed are Young`s, shear and bulk moduli and Poisson`s ratio. Thermal expansion, thermal conductivity and specific heat and magnetization are also reported. Data presented are summarized in the main body and presented in detail in the supplements. Areas of ongoing research are briefly described, and topics for future research are suggested. The data have been compiled to assist in the design of large-scale superconducting magnets for fusion reactors.

  14. Stress corrosion cracking of steam generator tube and primary pipe in PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Weiguo; Gao Fengqin; Zhou Hongyi

    1993-01-01

    The behavior of stress corrosion cracking (SCC) is studied by slow strain rate test (SSRT), constant load test (CLT) and low frequency cyclic loading test (LFCLT). The purpose of these tests is to get the test data for evaluating the integrity of pressurized boundary of pipes in Qinshan and Guangdong nuclear power plants. Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes in welded heat affected zone (WHAZ) and steam generator tubes, such as Incoloy-800, Inconel-600, Inconel-690 and 321 SS which are used for steam generator in PWR. The effects of material metallurgy, shot-peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior of SCC are investigated

  15. Evaluation of metallic materials for use in engineering barrier systems

    International Nuclear Information System (INIS)

    Pitman, S.G.; Griggs, B.; Elmore, R.P.

    1980-01-01

    Conclusions of this work are as follows: Inconel, Incoloy, Hastelloy C-276, and titanium alloys all had excellent corrosion resistance in all postulated repository environments tested. Further work will be required to evaluate the pertinent enviro-mechanical properties of these materials; the mechanical properties of grade 2 titanium are better than those of grade 12 titanium, except the tensile and yield strengths. These properties include fatigue-crack-growth rate, environmental fatigue-crack-growth rate, fracture toughness, impact toughness, and dynamic fracture toughness; there is no evidence in the current data to indicate that the simulated repository environment is aggressive to grade 2 or grade 12 titanium. This includes data from corrosion-fatigue, crevice corrosion, wedge-loaded cracked specimens, and residual-stress specimens

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

    International Nuclear Information System (INIS)

    Roca, M.; Diaz-Guerra, J.P.

    1978-01-01

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

  17. Stress corrosion cracking of steam generator tube and primary pipe in PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Weiguo; Gao Fengqin; Zhou Hongyi

    1992-03-01

    The behavior of stress corrosion cracking (SCC) was studied by slow strain rate test (SSRT), constant load test (CLT) and low frequency cyclic loading test (LFCLT). The purpose of these tests is to get the test data for evaluating the integrity of pressurized boundary of pipes in Qinshan and Guangdong nuclear power plants (NPPs). Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes in welded heat affected zone (WHAZ) and tubes of heat transfer, such as Incoloy-800, Inconel-600 and 321 SS which are used for steam generator in PWR NPPs. The effects of material metallurgy, shot peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior of SCC were considered

  18. Nuclear waste package design for the Vadose zone in tuff

    International Nuclear Information System (INIS)

    O'Neal, W.C.; Ballou, L.B.; Gregg, D.W.; Russell, E.W.

    1984-02-01

    This report presents an overview of the selection and analysis of conceptual waste package designs that will be used by the Nevada Nuclear Waste Storage Investigations (NNWSI) project for disposal of high-level nuclear waste (HLW) at the proposed Yucca Mountain, Nevada Site. The design requirements that the waste packages are required to meet are listed. Concept drawings for the reference designs and one alternative package design are shown. Four metal alloys; 304L SS, 321 SS, 316L SS and Incoloy 825 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 economic analysis supporting the selection of the conceptual waste package designs is included. Post-closure containment and release rates are not discussed in this paper. 17 references, 2 figures, 2 tables

  19. Initial specifications for nuclear waste package external dimensions and materials

    International Nuclear Information System (INIS)

    Gregg, D.W.; O'Neal, W.C.

    1983-09-01

    Initial specifications of external dimensions and materials for waste package conceptual designs are given for Defense High Level Waste (DHLW), Commercial High Level Waste (CHLW) and Spent Fuel (SF). The designs have been developed for use in a high-level waste repository sited in a tuff media in the unsaturated zone. Drawings for reference and alternative package conceptual designs are presented for each waste form for both vertical and horizontal emplacement configurations. Four metal alloys: 304L SS, 321 SS, 316L SS and Incoloy 825 are considered for the canister or overpack; 1020 carbon steel was selected for horizontal borehole liners, and a preliminary packing material selection is either compressed tuff or compressed tuff containing iron bearing smectite clay as a binder

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-08-04

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

  1. Relaxation and corrosion resistance of alloy 800 used for steam generator tubes of ship borne boilers

    International Nuclear Information System (INIS)

    Corrieu, J.M.; Cortial, F.; Maillard, J.L.; Vernot-Loier, C.; Lebeau, M.

    1994-01-01

    The INCO ''INCOLOY 800'' trademark groups the Fe-Cr-Ni alloys containing 30 to 35% nickel, 19 to 23% chromium, 0,15 to 0,60% aluminium, 0,15 to 0,60% titanium and less than 0,10% carbon contents, used as construction materials for condenser and heat exchanger tubes. In parallel with water chemistry control and studies aimed at reducing the residual stresses resulting from tube expansion, studies have been conducted to a better understanding of this alloy, its metallurgy and its corrosion behaviour under accurately defined fabrication and heat treatment conditions. The purpose of this paper is to present the results of a behaviour study of INDRET alloy 800 concerning isothermal relaxation and effects of the said relaxation heat treatments on alloy microstructure studied with a transmission electron-chemical method to determine the sensitiveness to intergranular corrosion, and by electrochemistry in pressurized hot water. (authors). 4 figs., 5 tabs., 7 refs

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

    International Nuclear Information System (INIS)

    Roach, J.F.; Urban, W.M.; Hartman, D.

    1995-01-01

    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

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

    International Nuclear Information System (INIS)

    Vancon, D.

    1989-11-01

    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 [fr

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

    International Nuclear Information System (INIS)

    Schubert, F.; te Heesen, E.; Bruch, U.; Cook, R.; Diehl, H.; Ennis, P.J.; Jakobeit, W.; Penkalla, H.J.; Ullrich, G.

    1984-01-01

    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 800 0 C. 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

  5. Creep and low cycles fatigue behaviour of inconel 617 and alloy 800H in the temperature range 1073-1223

    International Nuclear Information System (INIS)

    Yun, H.M.

    1984-01-01

    The creep rupture properties of high temperature alloys are being determined as part of the materials programme for the development of the high temperature, gas-cooled reactor (HTGR) as a source of nuclear process heat, especially for the gasification of lignite and coal. INCOLOY 800H AND INCONEL 617 have been tested in the temperature range from 1073 K to 1223 K in air as well as in helium with HTGR specific impurities. The static and dynamic creep behaviour of INCONEL 617 have been determined in constant load creep tests, relaxation tests and stress reduction tests. The results have been interpreted using the internal stress on the applied stress and test temperature was determined. In a few experiments the influence of cold deformation prior to the creep test on the magnitude of the internal stress was also investigated. (Author)

  6. Structural behaviour of a welded superalloy cylinder with internal pressure in a high temperature environment

    International Nuclear Information System (INIS)

    Udoguchi, T.; Nakanishi, T.

    1981-01-01

    Steady and cyclic creep tests with internal pressure were performed at temperatures of 800 to 1000 0 C on Hastelloy X cylinders with and without a circumferential Tungsten Inert Gas (TIG) welding technique. The creep rupture strength of the TIG welded cylinders was much lower than that of the non-welded cylinders whilst creep rupture strength reduction by the TIG technique was not observed in uniaxial creep tests. The reason for the low creep strength of welded cylinders is discussed and it is noted that the creep ductility of weld metal plays an essentially important role. In order to improve the creep strength of the TIG welded cylinder, various welding procedures with assorted weld metals were investigated. Some improvements were obtained by using welding techniques which had either Incoloy 800 or a modified Hastelloy X material as the filler metal. (U.K.)

  7. Low concentration NP preoxidation condition for PWR decontamination

    International Nuclear Information System (INIS)

    Huang Fuduan; Yu Degui; Lu Jingju; Ding Dejun; Zhao Yukun

    1991-02-01

    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

  8. Comparison of French and German NPP water chemistry programs

    Energy Technology Data Exchange (ETDEWEB)

    Staudt, U. [VGB Powertech (Germany); Odar, S. [Framatome ANP GmbH (Germany); Stutzmann, A. [EDF/GDL (France)

    2002-07-01

    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)

  9. 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

  10. 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

  11. Correlation of substructure with mechanical properties of plastically deformed reactor structural materials. Progress report, January 1, 1974--December 31, 1975

    International Nuclear Information System (INIS)

    Moteff, J.

    1976-01-01

    Ratio of the subgrain boundary dislocations to those contributing to creep deformation was found to be independent of applied stress and creep strain after the steady-state creep stage is reached. The observed cell or subgrain sizes are correlated with flow stress in Type 304 ss, and the deformation rate-stress relation obeys the equation epsilon =β lambda 3 (sigma/sub T//E)/sub n/ exp (-Q/sub c//RT), where lambda = subgrain size, sigma/sub T/ = effective true stress, E = Young modulus, and Q/sub c/ = 85 kcal/mole. Well-developed subgrains were observed in TEM on 304 ss tested in creep at 704 0 C. Role of twin boundary-grain boundary intersections in microcracking behavior of 304 ss deformed in slow tension and creep at 650 0 C was investigated. Grain shape analysis show that intragranular deformation becomes more predominant in the grains with the larger intercept distances, and that grain boundary sliding becomes important as the strain rate decreases. RT mechanical properties of austenitic ss are enhanced by subgrains formed during high-temperature deformation. The substructural development during high-temperature low-cycle fatigue of 304 ss was studied using TEM. Fatigue properties of Incoloy 800 tested in bend and push-pull modes are being compared. Effects of hold time on fatigue substructure and fracture of 304 ss are being studied. 31 figures, 53 references

  12. Cyclic Oxidation Behaviour of Domestic Superalloys at Elevated Temperature

    International Nuclear Information System (INIS)

    Kang, S. C.; Kim, G. M.; Chon, Y. G.

    1991-01-01

    The cyclic oxidation behaviour of commercial superalloys produced in Korea was investigated in air at 1000 .deg. C and 1100 .deg. C. Cyclic oxidation test was carried out by cyclically oxidizing the specimens in an apparatus which periodically removed the specimens from the furnace and reinserted them. The influence of growth stress and thermal stress on the cyclic oxidation was studied by examination of the oxide structures, their morphologies, and EDS line scanning of cross-section of cyclically oxidized specimens. The results showed that Inconel 601 was the best in the cyclic oxidation resistance among the tested alloys, followed by Nimonic 80A, Incoloy 825 and Inconel 718 at 1100 .deg. C. As in the case of the isothermal oxidation, relatively pure Cr 2 O 3 was effective in the beginning of cyclic oxidation experiment. But, later on, other oxidation products as well as Cr 2 O 3 were formed, resulting in the spallation of oxide scales. Especially, Nb and Mo in the alloys were determental to the cyclic oxidation behavior

  13. Choice of materials for the immobilization of 85-krypton in a metallic matrix by combined ion implantation and sputtering

    International Nuclear Information System (INIS)

    Whitmell, D.S.

    1985-01-01

    Immobilization in a metal matrix by combined ion implantation and sputtering promises to offer an ideal method for the containment of krypton-85 arising from the reprocessing of nuclear fuel. A 50 kW inactive pilot plant has been built and operated to prepare a copper deposit 22 mm thick weighing 23 kg and containing over 30 liters of inactive gas. The gas incorporation rate exceeded the design figure of 0.3 liters/hour and the vessel was operated at powers up to 30 kW, which corresponds to that envisaged for the industrial vessel. The power consumption was less than 100 kWh/liter. A full-scale vessel (1 m long, 0.26 m diameter) has also been tested at low power. Samples of alternative candidate materials: stainless steel, incoloy, nickel and nickel-lanthanum have been prepared and tested. Nickel appears to be the most promising since it incorporates gas with an efficiency 70% greater than copper and also retains the gas to a temperature at least 100 0 C higher than copper. Tests are being carried out with 100 Curies of radioactive krypton in order to demonstrate that the process will operate satisfactorily at the high internal β irradiation levels that will exist in an active plant and to prepare samples containing krypton-85 for long term leakage measurements and for assessment of any effects caused by the build-up of the decay product rubidium

  14. Effect of Ovality in Inlet Pigtail Pipe Bends Under Combined Internal Pressure and In-Plane Bending for Ni-Fe-Cr B407 Material

    Directory of Open Access Journals (Sweden)

    Ramaswami P.

    2017-09-01

    Full Text Available The present paper makes an attempt to depict the effect of ovality in the inlet pigtail pipe bend of a reformer under combined internal pressure and in-plane bending. Finite element analysis (FEA and experiments have been used. An incoloy Ni-Fe-Cr B407 alloy material was considered for study and assumed to be elastic-perfectly plastic in behavior. The design of pipe bend is based on ASME B31.3 standard and during manufacturing process, it is challenging to avoid thickening on the inner radius and thinning on the outer radius of pipe bend. This geometrical shape imperfection is known as ovality and its effect needs investigation which is considered for the study. The finite element analysis (ANSYS-workbench results showed that ovality affects the load carrying capacity of the pipe bend and it was varying with bend factor (h. By data fitting of finite element results, an empirical formula for the limit load of inlet pigtail pipe bend with ovality has been proposed, which is validated by experiments.

  15. 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.

  16. Corrosion of container materials under clay repository conditions

    International Nuclear Information System (INIS)

    Debruyn, W.

    1990-01-01

    The work done in Belgium on steels and a number of corrosion-resistant materials is discussed. Laboratory screening tests have been performed to find candidate container materials. Materials of interest have been further tested in surface clays and are being tested in deep clay formations at the Mol site. These tests have concentrated on characterizations of the clay environment under equilibrium and disturbed conditions. The performance of some materials will be monitored for up to 50000 hours in the form of conventional corrosion specimens. Eventually corrosion and performance tests will be performed on full-size or scaled-down containers. The effects of parameters identified as being important based on characterization of the clay environment will be studied further in the laboratory. Electrochemical measurements and experiments on the effects of gamma radiation have been started. The materials that have been tested in clay environments include corrosion allowance materials - carbon steel, unalloyed cast iron, and cast iron alloyed with silicon and nickel - as well as corrosion resistant materials: AISI 304, 316 and 430 stainless steels; aluminum alloys; nickel 200; Inconel 600 and 625; Incoloy 800; Hastelloy C4 and B; and titanium grades 2 and 7

  17. Alloy SCR-3 resistant to stress corrosion cracking

    International Nuclear Information System (INIS)

    Kowaka, Masamichi; Fujikawa, Hisao; Kobayashi, Taiki

    1977-01-01

    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.)

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

    International Nuclear Information System (INIS)

    Ilg, Ulf

    2008-01-01

    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.)

  19. Postirradiation results and evaluation of helium-bonded uranium--plutonium carbide fuel elements irradiated in EBR-II. Interim report

    International Nuclear Information System (INIS)

    Latimer, T.W.; Barner, J.O.; Kerrisk, J.F.; Green, J.L.

    1976-02-01

    An evaluation was made of the performance of 74 helium-bonded uranium-plutonium carbide fuel elements that were irradiated in EBR-II at 38-96 kW/m to 2-12 at. percent burnup. Only 38 of these elements have completed postirradiation examination. The higher failure rate found in fuel elements which contained high-density (greater than 95 percent theoretical density) fuel than those which contained low-density (77-91 percent theoretical density) fuel was attributed to the limited ability of the high-density fuel to swell into the void space provided in the fuel element. Increasing cladding thickness and original fuel-cladding gap size were both found to influence the failure rates for elements containing low-density fuel. Lower cladding strain and higher fission-gas release were found in high-burnup fuel elements having smear densities of less than 81 percent. Fission-gas release was usually less than 5 percent for high-density fuel, but increased with burnup to a maximum of 37 percent in low-density fuel. Maximum carburization in elements attaining 5-10 at. percent burnup and clad in Types 304 or 316 stainless steel and Incoloy 800 ranged from 36-80 μm and 38-52 μm, respectively. Strontium and barium were the fission products most frequently found in contact with the cladding but no penetration of the cladding by uranium, plutonium, or fission products was observed

  20. 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.

  1. Evaluation of the Control Rod Super Alloy Material of HTR-PM

    International Nuclear Information System (INIS)

    Li Pengjun; Yan He; Diao Xingzhong

    2014-01-01

    The control rod drive mechanism (CRDM) system is served as the first reactivity control and shutdown system for the high temperature reactor pebble-bed module (HTR-PM) in Shandong, China. And the control rod, which is pulled up and down by a chain sprocket mechanism of CRDM to realize reactivity control, compensation and shutdown, has to be durable under temperature as high as 550℃ for a long time. Thus the material persistent strength under high temperature is quite important for the reliability of the CRDM. In this paper, a review on material selection of control rod of high temperature gas cooled reactors, including AVR and THTR-300 in Germany, HTTR in Japan, PBMR in South Africa and Dragon in Britain, was summarized. The major parameters of two kinds of high temperature alloy, incoloy 800H and alloy 625, were compared and discussed. According to the ASME NH volume, a design criterion for the control rod was established and applied in the analysis of the chain by using finite element method. The numerical simulations showed that the chain made of alloy 625 could meet the condition and work for a long time under high temperature. (author)

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

    International Nuclear Information System (INIS)

    Soo, P.

    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

  3. Vaccum Gas Tungsten Arc Welding, phase 1

    Science.gov (United States)

    Weeks, J. L.; Krotz, P. D.; Todd, D. T.; Liaw, Y. K.

    1995-01-01

    This two year program will investigate Vacuum Gas Tungsten Arc Welding (VGTAW) as a method to modify or improve the weldability of normally difficult-to-weld materials. VGTAW appears to offer a significant improvement in weldability because of the clean environment and lower heat input needed. The overall objective of the program is to develop the VGTAW technology and implement it into a manufacturing environment that will result in lower cost, better quality and higher reliability aerospace components for the space shuttle and other NASA space systems. Phase 1 of this program was aimed at demonstrating the process's ability to weld normally difficult-to-weld materials. Phase 2 will focus on further evaluation, a hardware demonstration and a plan to implement VGTAW technology into a manufacturing environment. During Phase 1, the following tasks were performed: (1) Task 11000 Facility Modification - an existing vacuum chamber was modified and adapted to a GTAW power supply; (2) Task 12000 Materials Selection - four difficult-to-weld materials typically used in the construction of aerospace hardware were chosen for study; (3) Task 13000 VGTAW Experiments - welding experiments were conducted under vacuum using the hollow tungsten electrode and evaluation. As a result of this effort, two materials, NARloy Z and Incoloy 903, were downselected for further characterization in Phase 2; and (4) Task 13100 Aluminum-Lithium Weld Studies - this task was added to the original work statement to investigate the effects of vacuum welding and weld pool vibration on aluminum-lithium alloys.

  4. Preliminary design of the cooling system for a gas-cooled, high-fluence fast pulsed reactor (HFFPR)

    International Nuclear Information System (INIS)

    Monteith, H.C.

    1978-10-01

    The High-Fluence Fast Pulsed Reactor (HFFPR) is a research reactor concept currently being evaluated as a source for weapon effects experimentation and advanced reactor safety experiments. One of the designs under consideration is a gas-cooled design for testing large-scale weapon hardware or large bundles of full-length, fast reactor fuel pins. This report describes a conceptual cooling system design for such a reactor. The primary coolant would be helium and the secondary coolant would be water. The size of the helium-to-water heat exchanger and the water-to-water heat exchanger will be on the order of 0.9 metre (3 feet) in diameter and 3 metres (10 feet) in length. Analysis indicates that the entire cooling system will easily fit into the existing Sandia Engineering Reactor Facility (SERF) building. The alloy Incoloy 800H appears to be the best candidate for the tube material in the helium-to-water heat exchanger. Type 316 stainless steel has been recommended for the shell of this heat exchanger. Estimates place the cost of the helium-to-water heat exchanger at approximately $100,000, the water-to-water heat exchanger at approximately $25,000, and the helium pump at approximately $450,000. The overall cost of the cooling system will approach $2 million

  5. Fracture toughness determination in steam generator tubes

    International Nuclear Information System (INIS)

    Bergant M; Yawny, A; Perez Ipina, J

    2012-01-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Ellingson, W. A.

    1979-01-01

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

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

    International Nuclear Information System (INIS)

    Schwarze, E.; Schuster, H.; Nickel, H.

    1992-01-01

    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.) [de

  8. The corrosion of steam generator surfaces under typical secondary coolant conditions: effects of pH excusions on the alloy surface composition

    Energy Technology Data Exchange (ETDEWEB)

    McIntyre, N.S.; Davidson, R.D.; Walzak, T.L. [University of Western Ontario, London, ON (Canada); Brennenstuhl, A.M.; Gonzalez, F.; Corazza, S. [Ontario Hydro, Toronto, ON (Canada)

    1995-07-01

    Specimens of Inconel 600 (I-600), Inconel 690 (I-690), Incoloy 800 (I-800) and Monel 400 (M-400) were exposed to high temperature steam generator water conditions typical of those which might be found when the secondary coolant pH is allowed to fall into the acid regime. The specimens were analysed by surface and electrochemical techniques, either directly following exposure in the acid medium steam generator coolant or after adjusting the coolant pH and chemistry to near-normal steam generator conditions. The initial acid exposure resulted in the growth of a chromium-rich surface corrosion product film on all alloys (except M-400) and the precipitation of nickel-rich sulphates. Following the return to a high pH, the alloy again had chromium-rich surface oxides but also exhibited sulphide crystallites adhering to the base oxide, particularly for I-600. The tendency to retain these sulphides is attributed to the porosity of the protective oxide through which nickel is transported to the solution. The conversion of sulphate-sulphide is believed to occur as the pH is raised to normal alkaline conditions. In the case of M-400, a chromium oxide layer is not available to restrict the transport of nickel to the solution. As a result, a thick layer of sulphide crystals grows on the surface of the M-400 alloy even when subjected to a mild acid pH excursion. Even trace concentrations of sulphide/sulphate under normal pH control are shown to react with alloys whose oxide films appear permeable to nickel transport. (Author).

  9. Evaluation of nitrogen containing reducing agents for the corrosion control of materials relevant to nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Padma S. [Water and Steam Chemistry Division, BARC Facilities, Kalpakkam, Tamilnadu (India); Mohan, D. [Department of Chemistry, Anna University, Chennai, Tamilnadu (India); Chandran, Sinu; Rajesh, Puspalata; Rangarajan, S. [Water and Steam Chemistry Division, BARC Facilities, Kalpakkam, Tamilnadu (India); Velmurugan, S., E-mail: svelu@igcar.gov.in [Water and Steam Chemistry Division, BARC Facilities, Kalpakkam, Tamilnadu (India)

    2017-02-01

    Materials undergo enhanced corrosion in the presence of oxidants in aqueous media. Usually, hydrogen gas or water soluble reducing agents are used for inhibiting corrosion. In the present study, the feasibility of using alternate reducing agents such as hydrazine, aqueous ammonia, and hydroxylamine that can stay in the liquid phase was investigated. A comparative study of corrosion behavior of the structural materials of the nuclear reactor viz. carbon steel (CS), stainless steel (SS-304 LN), monel-400 and incoloy-800 in the oxidizing and reducing conditions was also made. In nuclear industry, the presence of radiation field adds to the corrosion problems. The radiolysis products of water such as oxygen and hydrogen peroxide create an oxidizing environment that enhances the corrosion. Electrochemical studies at 90 °C showed that the reducing agents investigated were efficient in controlling corrosion processes in the presence of oxygen and hydrogen peroxide. Evaluation of thermal stability of hydrazine and its effect on corrosion potential of SS-304 LN were also investigated in the temperature range of 200–280 °C. The results showed that the thermal decomposition of hydrazine followed a first order kinetics. Besides, a change in electrochemical corrosion potential (ECP) was observed from −0.4 V (Vs SHE) to −0.67 V (Vs SHE) on addition of 5 ppm of hydrazine at 240 °C. Investigations were also made to understand the distribution behavior of hydrogen peroxide and hydrazine in water-steam phases and it was found that both the phases showed identical behavior. - Highlights: • Hydrazine was found to be a promising reducing agent for oxidant control. • In presence of hydrazine corrosion potential of SS304 LN was well below −230 mV. • SS304LN could be protected from IGSCC by hydrazine addition. • Thermal and radiation stability of hydrazine at 285 °C was found satisfactory.

  10. Chemical and X-ray diffraction analysis on selected samples from the TMI-2 reactor core

    International Nuclear Information System (INIS)

    Kleykamp, H.; Pejsa, R.

    1991-05-01

    Selected samples from different positions of the damaged TMI-2 reactor core were investigated by X-ray microanalysis and X-ray diffraction. The measurements yield the following resolidified phases after cooling: Cd and In depleted Ag absorber material, intermetallic Zr-steel compounds, fully oxidized Zircaloy, UO 2 -ZrO 2 solid solutions and their decomposed phases, and Fe-Al-Cr-Zr spinels. The composition of the phases and their lattice parameters as well as the eutectic and monotectic character can serve as indicators of local temperatures of the core. The reaction sequences are estimated from the heterogeneous equilibria of these phases. The main conclusions are: (1) Liquefaction onset is locally possible by Inconel-Zircaloy and steel-Zircaloy reactions of spacers and absorber guide tubes at 930deg C. However, increased rates of dissolution occur above 1200deg C. (2) UO 2 dissolution in the Inconel-steel-Zircaloy melt starts at 1300deg C with increased rates above 1900deg C. (3) Fuel temperatures in the core centre are increased above 2550deg C, liquid (U,Zr)O 2 is generated. (4) Square UO 2 particles are reprecipitated from the Incoloy-steel-Zircaloy-UO 2 melt during cooling, the remaining metallic melt is oxygen poor; two types of intermetallic phases are formed. (5) Oxidized Fe and Zr and Al 2 O 3 from burnable absorber react to spinels which form a low melting eutectic with the fuel at 1500deg C. The spinel acts as lubricant for fuel transport to the lower reactor plenum above 1500deg C. (6) Ruthenium (Ru-106) is dissolved in the steel phase, antimony (Sb-125) in the α-Ag absorber during liquefaction. (7) Oxidation of the Zircaloy-steel phases takes place mainly in the reflood stage 3 of the accident scenario. (orig.) [de

  11. Characterization of the oxide formed in the presence of poly acrylic acid over the steam generator structural materials of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Joshi, Akhilesh C.; Rufus, Appadurai L.; Suresh, Sumathi; Chandramohan, Palogi; Rangarajan, Srinivasan [Water and Steam Chemistry Division, BARC Facilities, Kalpakkam, TN 603 102 (India); Velmurugan, Sankaralingam, E-mail: svelu@igcar.gov.in [Water and Steam Chemistry Division, BARC Facilities, Kalpakkam, TN 603 102 (India)

    2013-06-15

    Graphical abstract: Oxide deposited on carbon steel surface exposed to secondary side water chemistry conditions of nuclear power plant in the (a) absence and (b) presence of poly acrylic acid (PAA), a polymeric dispersant. Highlights: ► PAA cannot prevent the formation of inner oxide arising out of corrosion. ► Inner oxide formed in presence of PAA is protective. ► Crystallites formed in the presence of PAA are smaller in size. ► PAA prevents the formation of outer oxide layer arising out of precipitation. -- Abstract: On-line addition of polymeric dispersants, such as poly acrylic acid (PAA), to the steam generator (SG) results in the formation of a better protective inner oxide layer that reduces subsequent corrosion of structural materials. Its dispersive action inhibits the growth of a secondary oxide layer thereby facilitating their easy removal. This paper discusses the effect of PAA on the nature of oxides formed over the surfaces of SG. In the case of carbon steel, the inner oxide layer (magnetite) formed in the presence of PAA was protective. Electrochemical studies showed a minimum concentration of 350 ppb of PAA was found to be optimum. On the monel surface, in the absence of PAA, nickel ferrite was formed while in the presence of PAA, the oxide formed was a mixture of oxides of copper and nickel. A concentration of 700 ppb of PAA was found to be optimum for monel. In the case of incoloy, the effect of PAA was not discernible except for the size and morphology of the crystallites formed.

  12. Operating experience with steam generators

    International Nuclear Information System (INIS)

    Bouecke, R.

    1991-01-01

    In contrast to steam generator tube degradation problems that have been widely encountered worldwide, steam generators of the Siemens/KWU design have proven by operating experience that they are very efficient in minimizing tube corrosion or any other SG related problems. The paper will substantiate this statement by addressing the performance characteristics of nearly 20 years of operation experience. Emphasis is put on evaluations comparing the heat transfer capacity of Incoloy 800 with that of Inconel 690 TT. Various tube support designs are discussed with respect to hide-out behaviour. Recent evaluations confirm the superiority of grid type tube support designs compared to tube support plates. Anti vibration bars acc. to Siemens/KWU design allow proper support even of the innermost U-tubes, by which excessive vibration induced tube failures due to fatigue are ruled-out. Steam generators are key components which can heavily effect plant safety and availability. This was recognized by Siemens/KWU at a very early stage of the SG design. A multi level concept was developed and consequently applied, the characteristics of which can be highlighted as follows: - Implementation of specific design features after careful experimental and/or analytical verification; - Material selection based on profound validation tests; - Stringent inspection requirements regarding control of manufacturing; - Deliberate specification and control of water chemistry guidelines; - Close feedback of operational problems to be readily considered in design improvements or remedial actions to be taken. A strict application of this concept has reached a stage which allows the following summarizing statements: - All main design features of the Siemens/KWU SG are in use - basically unchanged - for roughly 20 years; - All of them are verified by excellent operating experience and available for implementation in any replacement or new SG design; - No need for replacement of Siemens/KWU SG is

  13. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ellingson, W A

    1978-01-01

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

  15. Irradiation effects on reactor structural materials. Semi-annual progress report, August 1974--February 1975

    International Nuclear Information System (INIS)

    Claudson, T.T.

    1975-03-01

    Data are reported on: effects of cold work on creep-fatigue of irradiated 304 and 316 stainless steel (ss); swelling of 304 and 316 ss irradiated with protons and fast neutrons; effects of hold time on fatigue crack propagation in neutron-irradiated 20 percent cold-worked 316 ss; radiation resistance of 0.03 percent Cu A533-B steel; microstructure of irradiated Inconel 718, Incoloy 800, PH13-8Mo, Mo, and Nb; dose dependence of 2.8-MeV Ni + ion damage (swelling) in Ni; notch ductility and strength of 316 ss submerged arc weld deposits; effects of microstructure of 316 ss on its irradiation response; in-reactor deformation of 20 percent cold-worked 316 ss; microstructure of HFIR-irradiated 316 ss; void microstructures of V bombarded by 46-MeV Ni 6+ ions (with and without preinjected helium) or 7.5-MeV Ta 3+ ions; swelling of Mo, Mo--0.5 Ti, Nb, Nb--1 Zr, W, and W--25 Re after fast neutron irradiation; swelling of V ion-irradiated Mo; creep of 20 percent cold-worked 316 ss at 850, 1000, and 1100 0 F; effects of fast neutrons on mechanical properties of 20 percent cold-worked 316 ss; notch effects in tensile behavior of irradiated, annealed 304 ss (EBR-II duct thimbles); equations for thermal creep in pressurized tubes of 20 percent cold-worked 316 ss; irradiation creep in cold-worked 316 ss; helium production cross sections in neutron-irradiated elements; and radiation effects on various alloys. (U.S.)

  16. Corrosion Behavior Of Potential Structural Materials For Use In Nitrate Salts Based Solar Thermal Power Plants

    Science.gov (United States)

    Summers, Kodi

    The increasing global demand for electricity is straining current resources of fossil fuels and placing increased pressure on the environment. The implementation of alternative sources of energy is paramount to satisfying global electricity demand while reducing reliance on fossil fuels and lessen the impact on the environment. Concentrated solar power (CSP) plants have the ability to harness solar energy at an efficiency not yet achieved by other technologies designed to convert solar energy to electricity. The problem of intermittency in power production seen with other renewable technologies can be virtually eliminated with the use of molten salt as a heat transfer fluid in CSP plants. Commercial and economic success of CSP plants requires operating at maximum efficiency and capacity which requires high temperature and material reliability. This study investigates the corrosion behavior of structural alloys and electrochemical testing in molten nitrate salts at three temperatures common to CSP plants. Corrosion behavior was evaluated using gravimetric and inductively-coupled plasma optical emission spectroscopy (ICP-OES) analysis. Surface morphology was studied using scanning electron microscopy. Surface oxide structure and chemistry was characterized using X-ray diffraction, Raman spectroscopy, energy dispersive spectroscopy, and X-ray photoelectron spectroscopy. Electrochemical behavior of candidate structural alloys Alloy 4130, austenitic stainless steel 316, and super-austenitic Incoloy 800H was evaluated using potentiodynamic polarization characteristics. It was observed that electrochemical evaluation of these candidate materials correlates well with the corrosion behavior observed from gravimetric and ICP-OES analysis. This study identifies that all three alloys exhibited acceptable corrosion in 300°C molten salt while elevated salt temperatures require the more corrosion resistant alloys, stainless steel 316 and 800H. Characterization of the sample

  17. Application of Corrosion Test for Austenitic SS 304 in PWR with Electrochemical Quartz Crystal Microbalance (EQCM)

    International Nuclear Information System (INIS)

    Park, Jeong Seok; Shin, Sang Hun; Kim, Jong Jin; Kim, Ji Hyun

    2011-01-01

    The secondary system of a pressurized water reactor (PWR) is a loop composed of: a) steam generator, b) steam transport piping, c) steam turbine and d) condenser and some optional component. The materials of these components are mainly carbon steel, different grades of stainless steel, high nickel alloys (Inconel-600 and Incoloy-800), Copper alloys. Among these materials, austenitic stainless steel 304 is widely used in tubing for large-surface condenser exposed to seawater as coolant of steam condenser. Stainless steels show much greater resistance to erosion-corrosion than copper alloys used in steam condenser tubing and are also immune to ammonia attack, ammonia induced stress corrosion cracking. Also, the chloride induced stress corrosion cracking has not been reported in stainless steel tubing in power plant condenser. Nevertheless, it is well known that stainless steels are susceptible to pitting and crevice corrosion in the seawater environment including chloride ions. Metallic materials exposed to untreated seawater suffer the well-known phenomenon of fouling consisting of the formation of an unwanted deposit that covers the surfaces in contact with the water. Five types of fouling are usually considered as biological, corrosion, particulate, chemical and precipitation fouling. Pitting corrosion is aggravated by stagnant or low flow conditions in which sludge such as corrosion of copper alloys in the secondary system can form in the tube surface. From the operational maintenance viewpoint, pitting attack has destructive effect of structural materials on operation of NPP. Countermeasures such as copper alloys replacement, water chemistry control and chemical cleaning were implemented to mitigate the pitting. Although the many studies are performed macroscopically to reduce pitting corrosion in the secondary system, little work has been done to understand mechanism of fouling formation and to confirm the quantitative analysis of the fouling between the metal

  18. Results for the Aboveground Configuration of the Boiling Water Reactor Dry Cask Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Advanced Nuclear Fuel Cycle Technologies; Lindgren, Eric Richard [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Advanced Nuclear Fuel Cycle Technologies

    2016-09-01

    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 also by increasing the internal convection through greater canister helium pressure. These same 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 belowground 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 aboveground and belowground canistered dry cask systems. The purpose of the current investigation was to produce data sets that can be used to test the validity of the assumptions associated with the calculations used to determine steady-state cladding temperatures in modern dry casks that utilize elevated helium pressure in the sealed canister in an aboveground configuration. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly was deployed inside of a representative storage basket and cylindrical pressure vessel that represents a vertical canister system. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. The arrangement of ducting was used to mimic conditions for an aboveground storage configuration in a vertical, dry cask

  19. 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

  20. 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

  1. Thermal-Hydraulic Results 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)

    2017-09-01

    The thermal performance of commercial nuclear spent fuel dry storage casks is 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 canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both aboveground and belowground 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 aboveground and belowground canistered dry cask systems. The purpose of this investigation was to produce validation-quality data that can be used to test the validity of the modeling presently used to determine cladding temperatures in modern vertical dry casks. These cladding temperatures are critical to evaluate cladding integrity throughout the storage cycle. To produce these data sets under well-controlled boundary conditions, the dry cask simulator (DCS) was built to study the thermal-hydraulic response of fuel under a variety of heat loads, internal vessel pressures, and external configurations. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly was deployed inside of a representative storage basket and cylindrical pressure vessel that represents a vertical canister system. The symmetric

  2. 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

  3. 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.

  4. 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.

  5. Zinc injection in German PWR plants

    International Nuclear Information System (INIS)

    Streit, K.

    2004-01-01

    Operating experience acquired at PWR NNPs shows that zinc injection at low concentrations of 5 ppb is a very effective source term reduction measure. This method does not lead to any operating restrictions or other negative effects on plant systems and components. The nuclear industry has been very successful in reducing radiation exposures within the past two decades. Annual exposures could be significantly decreased and are now at a level of around 1 man-Sv per plant and year. This great success can mainly be attributed to the general commitment of plant operators to maintaining radiation exposures of workers in the controlled access area as low as reasonably achievable (ALARA principle). The ALARA principle, of course, also implies evaluation of the economic benefit of radiation protection measures. Radiation source term reduction has drawn increasing attention of plant operators in recent years. For the new PWRs cobalt-based alloys in the primary system have successively been eliminated already at the design and construction phase within the last decade. Use of wear-resistant cobalt-free substitute materials in combination with the general use of advanced alloys for the steam generator tubing of PWRs resulted in low values for the two most common sources of plant radiation fields, namely 58 Co and 60 Co. Investigations showed that the beneficial effect of zinc can be related to its high affinity for mixed spinel oxide phases, resulting in the following two basic effects: -Zinc is incorporated preferentially into the oxide layer on primary system surfaces and thus reduces pickup of 58 Co and 60 Co and - Zinc can displace cobalt isotopes from existing oxide layers. In German PWRs with Incoloy 800 steam generator tubing material (Ni-content -32%), the observed reductions correspond to a decrease in dose rates of around 10 to 15% per year and thus follow, as predicted, the half-life time of 60 Co. Overall reductions in high radiation areas are now in the range of

  6. Embalse steam generators - status in 2009

    International Nuclear Information System (INIS)

    Luna, P.; Yetisir, M.; Roy, S.; MacEacheron, R.

    2009-01-01

    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 plan 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

  7. Flux entrapment and Titanium Nitride defects during electroslag remelting

    Science.gov (United States)

    Busch, Jonathan D.

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