WorldWideScience

Sample records for hastelloys

  1. Relaxation characteristics of hastelloy X

    International Nuclear Information System (INIS)

    Suzuki, Kazuhiko

    1980-02-01

    Relaxation diagrams of Hastelloy X (relaxation curves, relaxation design diagrams, etc.) were generated from the creep constitutive equation of Hastelloy X, using inelastic stress analysis code TEPICC-J. These data are in good agreement with experimental relaxation data of ORNL-5479. Three typical inelastic stress analyses were performed for various relaxation behaviors of the high-temperature structures. An attempt was also made to predict these relaxation behaviors by the relaxation curves. (author)

  2. Design fatigue curve for Hastelloy-X

    International Nuclear Information System (INIS)

    Nishiguchi, Isoharu; Muto, Yasushi; Tsuji, Hirokazu

    1983-12-01

    In the design of components intended for elevated temperature service as the experimental Very High-Temperature gas-cooled Reactor (VHTR), it is essential to prevent fatigue failure and creep-fatigue failure. The evaluation method which uses design fatigue curves is adopted in the design rules. This report discussed several aspects of these design fatigue curves for Hastelloy-X (-XR) which is considered for use as a heat-resistant alloy in the VHTR. Examination of fatigue data gathered by a literature search including unpublished data showed that Brinkman's equation is suitable for the design curve of Hastelloy-X (-XR), where total strain range Δ epsilon sub(t) is used as independent variable and fatigue life Nsub(f) is transformed into log(log Nsub(f)). (author)

  3. Hastelloy X fuel element creep relaxation and residual effects

    International Nuclear Information System (INIS)

    Castle, R.A.

    1971-01-01

    A worst case, seven element, asymmetric fuel, thermal environment was assumed and a creep relaxation analysis generated. The fuel element clad is .020 inch Hastelloy X. The contact load decreased from 11.6 pounds to 5.87 pounds in 100,000 hours. The residual stresses were then computed for various shutdown times. (U.S.)

  4. Creep properties of Hastelloy X in a carburizing helium environment

    International Nuclear Information System (INIS)

    Nakanishi, T.; Kawakami, H.

    1982-01-01

    In this work, we investigate the environmental effect on the creep behavior of Hastelloy X at 900 0 C in helium and air. Since helium coolant in HTGR is expected to be carburizing and very weakly oxidizing for most metals, testings were focused on the effect of carburizing and slight oxidation. Carburization decreases secondary creep strain rate and delays tertiary creep initiation. On the other hand, the crack growth rate on the specimen surface is enhanced due to very weak oxidation in helium, therefore the tertiary creep strain rate becomes larger than that in air. The rupture time of Hastelloy X was shorter in helium when compared with in air. Stress versus rupture time curves for both environments do not deviate with each other during up to 5000 hours test, and a ratio of rupture stress in helium to that in air was about 0.9

  5. Study on the creep constitutive equation of Hastelloy X, (1)

    International Nuclear Information System (INIS)

    Suzuki, Kazuhiko; Mutoh, Yasushi

    1983-01-01

    In order to carry out the structural design of high temperature pipings, intermediate heat exchangers and isolating valves for a multipurpose high temperature gas-cooled reactor, in which coolant temperature reaches 1000 deg C, the creep characteristics of Hastelloy X used as the heat resistant material must be clarified. In addition to usual creep rupture life and the time to reach a specified creep strain, the dependence of creep strain curves on time, temperature and stress must be determined and expressed with equations. Therefore, using the creep data of Hastelloy X given in the literatures, the creep constitutive equation was made. Since the creep strain curves under the same test condition were different according to heats, the sensitivity analysis of the creep constitutive equation was performed. The form of the creep constitutive equation was determined to be Garofalo type. The result of the sensitivity analysis is reported. (Kako, I.)

  6. High temperature strength of Hastelloy XR under biaxial stress states

    International Nuclear Information System (INIS)

    Muto, Yasushi; Hada, Kazuhiko; Koikegami, Hajime; Ohno, Nobutada.

    1991-01-01

    Biaxial(tension/torsion) creep and creep-fatigue tests were conducted on Hastelloy XR at 950degC in air. Hastelloy XR is a nickel base solution-annealed heat resistant alloy. Thin-walled tubular test specimens were employed. As results of the creep tests, the von Mises' flow rule was revealed to be applicable very well. Under the torsion load, sufficient growth of voids was necessary to initiate the fracture and this resulted in longer life time compared with that under the tension load. Only a few number of small voids could be observed and very long life times were attained under the compression load. The creep-fatigue tests revealed that superposition of constant torsion load on a cyclic axial load reduced the cycles to failure significantly and the amount of reduction was consistent with the prediction by the linear life fraction rule. (author)

  7. Cavitation erosion behavior of Hastelloy C-276 nickel-based alloy

    Energy Technology Data Exchange (ETDEWEB)

    Li, Zhen [State Key Laboratory of Solid Lubrication, Lanzhou Institute of Chemical Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); University of Chinese Academy of Sciences, Beijing 100039 (China); Han, Jiesheng; Lu, Jinjun [State Key Laboratory of Solid Lubrication, Lanzhou Institute of Chemical Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Chen, Jianmin, E-mail: chenjm@lzb.ac.cn [State Key Laboratory of Solid Lubrication, Lanzhou Institute of Chemical Physics, Chinese Academy of Sciences, Lanzhou 730000 (China)

    2015-01-15

    Highlights: • Cavitation erosion behavior of Hastelloy C-276 was studied by ultrasonic apparatus. • The cavitation-induced precipitates formed in the eroded surface for Hastelloy C-276. • The selective cavitation erosion was found in Hastelloy C-276 alloy. - Abstract: The cavitation erosion behavior of Hastelloy C-276 alloy was investigated using an ultrasonic vibratory apparatus and compared with that of 316L stainless steel. The mean depth of erosion (MDE) and erosion rate (ER) curves vs. test time were attained for Hastelloy C-276 alloy. Morphology and microstructure evolution of the eroded surface were observed by scanning electron microscopy (SEM) and field emission scanning electron microscopy (FESEM) and the predominant erosion mechanism was also discussed. The results show that the MDE is about 1/6 times lower than that of the stainless steel after 9 h of testing. The incubation period of Hastelloy C-276 alloy is about 3 times longer than that of 316L stainless steel. The cavitation-induced nanometer-scaled precipitates were found in the local zones of the eroded surface for Hastelloy C-276. The selective cavitation erosion was found in Hastelloy C-276 alloy. The formation of nanometer-scaled precipitates in the eroded surface may play a significant role in the cavitation erosion resistance of Hastelloy C-276.

  8. Cavitation erosion behavior of Hastelloy C-276 nickel-based alloy

    International Nuclear Information System (INIS)

    Li, Zhen; Han, Jiesheng; Lu, Jinjun; Chen, Jianmin

    2015-01-01

    Highlights: • Cavitation erosion behavior of Hastelloy C-276 was studied by ultrasonic apparatus. • The cavitation-induced precipitates formed in the eroded surface for Hastelloy C-276. • The selective cavitation erosion was found in Hastelloy C-276 alloy. - Abstract: The cavitation erosion behavior of Hastelloy C-276 alloy was investigated using an ultrasonic vibratory apparatus and compared with that of 316L stainless steel. The mean depth of erosion (MDE) and erosion rate (ER) curves vs. test time were attained for Hastelloy C-276 alloy. Morphology and microstructure evolution of the eroded surface were observed by scanning electron microscopy (SEM) and field emission scanning electron microscopy (FESEM) and the predominant erosion mechanism was also discussed. The results show that the MDE is about 1/6 times lower than that of the stainless steel after 9 h of testing. The incubation period of Hastelloy C-276 alloy is about 3 times longer than that of 316L stainless steel. The cavitation-induced nanometer-scaled precipitates were found in the local zones of the eroded surface for Hastelloy C-276. The selective cavitation erosion was found in Hastelloy C-276 alloy. The formation of nanometer-scaled precipitates in the eroded surface may play a significant role in the cavitation erosion resistance of Hastelloy C-276

  9. Application of Hastelloy X in gas-cooled reactor systems

    International Nuclear Information System (INIS)

    Brinkman, C.R.; Rittenhouse, P.L.; Corwin, W.R.; Strizak, J.P.; Lystrup, A.; DiStefano, J.R.

    1976-10-01

    Hastelloy X, an Ni--Cr--Fe--Mo alloy, may be an important structural alloy for components of gas-cooled reactor systems. Expected applications of this alloy in the High-Temperature Gas-Cooled Reactor (HTGR) are discussed, and the development of interim mechanical properties and supporting data are reported. Properties of concern include tensile, creep, creep-rupture, fatigue, creep-fatigue interaction, subcritical crack growth, thermal stability, and the influence of helium environments with controlled amounts of impurities on these properties. In order to develop these properties in helium environments that are expected to be prototypic of HTGR operating conditions, it was necessary to construct special environmental test systems. Details of construction and operating parameters are described. Interim results from tests designed to determine the above properties are presented. To date a fairly extensive amount of information has been generated on this material at Oak Ridge National Laboratory and elsewhere concerning behavior in air, which is reviewed. However, only limited data are available from tests conducted in helium. Comparisons of the fatigue and subcritical growth behavior in air between Hastelloy X and a number of other structural alloys are given

  10. Evaluation of creep and relaxation data for hastelloy alloy x sheet

    International Nuclear Information System (INIS)

    Booker, M.K.

    1979-02-01

    Hastelloy alloy X has been a successful high-temperature structural material for more than two decades. Recently, Hastelloy alloy X sheet has been selected as a prime structural material for the proposed Brayton Isotope Power System (BIPS). The material also sees extensive application in the High-Temperature Gas-Cooled Reactor (HTGR). Design of these systems requires a detailed consideration of the high-temperature creep properties of this material. Therefore, available creep, creep-rupture, and relaxation data for Hastelloy alloy X were collected and analyzed to yield mathematical representations of the behavior for design use

  11. Study on the creep constitutive equation of Hastelloy X, (1)

    International Nuclear Information System (INIS)

    Hada, Kazuhiko; Mutoh, Yasushi

    1983-01-01

    A creep constitutive equation of Hastelloy X was obtained from available experimental data. A sensitivity analysis of this creep constitutive equation was carried out. As the result, the following were revealed: (i) Variations in creep behavior with creep constitutive equation are not small. (ii) In a simpler stress change pattern, variations in creep behavior are similar to those in the corresponding fundamental creep characteristics (creep strain curve, stress relaxation curve, etc.). (iii) Cumulative creep damage estimated in accordance with ASME Boiler and Pressure Vessel Code Case N-47 from a stress history predicted by ''the standard creep constitutive equation'' which predicts the average behavior of creep strain curve data is not thought to be on the safe side on account of uncertainties in creep damage caused by variations in creep strain curve. (author)

  12. Creep properties of 20% cold-worked Hastelloy XR

    International Nuclear Information System (INIS)

    Kurata, Y.

    1996-01-01

    The creep properties of Hastelloy XR, in solution-treated and in 20% cold-worked conditions, were studied at 800, 900 and 1000 C. At 800 C, the steady-state creep rate and rupture ductility decrease, while rupture life increases after cold work to 20%. Although the steady-state creep rate and ductility also decrease at 900 C, the beneficial effect on rupture life disappears. Cold work to 20% enhan ces creep resistance of this alloy at 800 and 900 C due to a high density of dislocations introduced by the cold work. Rupture life of the 20% cold-worked alloy becomes shorter and the steady-state creep rate larger at 1000 C during creep of the 20% cold-worked alloy. It is emphasized that these cold work effects should be taken into consideration in design and operation of high-temperature structural components of high-temperature gas-cooled reactors. (orig.)

  13. Corrosion characteristics of Hastelloy N alloy after He+ ion irradiation

    International Nuclear Information System (INIS)

    Lin Jianbo; Yu Xiaohan; Li Aiguo; He Shangming; Cao Xingzhong; Wang Baoyi; Li Zhuoxin

    2014-01-01

    With the goal of understanding the invalidation problem of irradiated Hastelloy N alloy under the condition of intense irradiation and severe corrosion, the corrosion behavior of the alloy after He + ion irradiation was investigated in molten fluoride salt at 700 °C for 500 h. The virgin samples were irradiated by 4.5 MeV He + ions at room temperature. First, the virgin and irradiated samples were studied using positron annihilation lifetime spectroscopy (PALS) to analyze the influence of irradiation dose on the vacancies. The PALS results showed that He + ion irradiation changed the size and concentration of the vacancies which seriously affected the corrosion resistance of the alloy. Second, the corroded samples were analyzed using synchrotron radiation micro-focused X-ray fluorescence, which indicated that the corrosion was mainly due to the dealloying of alloying element Cr in the matrix. Results from weight-loss measurement showed that the corrosion generally correlated with the irradiation dose of the alloy. (author)

  14. Long-term corrosion behaviors of Hastelloy-N and Hastelloy-B3 in moisture-containing molten FLiNaK salt environments

    International Nuclear Information System (INIS)

    Ouyang, Fan-Yi; Chang, Chi-Hung; Kai, Ji-Jung

    2014-01-01

    Highlights: •Corrosion behaviors of Hastelloy-N and -B3 in molten FLiNaK salt at 700 °C. •The alleviated corrosion rate of alloys was observed after long-hour immersion. •Long-term corrosion rate was limited by diffusion from matrix to alloy surface. •Corrosion pattern transferred from intergranular corrosion into general corrosion. •Presence of minor H 2 O did not greatly influence the long-term corrosion behavior. -- Abstract: This study investigated long-term corrosion behaviors of Ni-based Hastelloy-N and Hastelloy-B3 under moisture-containing molten alkali fluoride salt (LiF–NaF–KF: 46.5–11.5–42%) environment at an ambient temperature of 700 °C. The Hastelloy-N and Hastelloy-B3 experienced similar weight losses for tested duration of 100–1000 h, which was caused by aggregate dissolution of Cr and Mo into FLiNaK salts. The corrosion rate of both alloys was high initially, but then reduced during the course of the test. The alleviated corrosion rate was due to the depletion of Cr and Mo near surface of the alloys and thus the long-term corrosion rate was controlled by diffusion of Cr and Mo outward to the alloy surface. The results of microstructural characterization revealed that the corrosion pattern for both alloys tended to be intergranular corrosion at early stage of corrosion test, and then transferred to general corrosion for longer immersion hours

  15. The corrosion behavior of molybdenum and Hastelloy B in sulfur and sodium polysulfides at 623 K

    International Nuclear Information System (INIS)

    Brown, A.P.

    1987-01-01

    An experimental study was completed to determine the corrosion behavior of molybdenum and Hastelloy B, a nickel-based alloy with high molybdenum content, in sulfur and sodium polysulfides (Na/sub 2/S/sub 3/,Na/sub 2/S/sub 4/, Na/sub 2/S/sub 5/) at 623 K. In sulfur, molybdenum corrodes very slowly, with a parabolic rate constant of 3.6 x 10/sup -9/ cm s/sup -1/2/. Hastelloy B shows no measurable corrosion after 100h of exposure to sulfur. The corrosion reaction of molybdenum in Na/sub 2/S/sub 3/ is characterized by the formation of a protective film that effectively eliminates further corrosion after the first 100h of exposure. Hastelloy B, however, corrodes rapidly in Na/sub 2/S/sub 3/, with corrosion rates approaching those of pure nickel under the same conditions. After the first 4h of exposure, the kinetics for the corrosion of Hastelloy B in Na/sub 2/S/sub 3/ follows a linear rate law. The scale morphology has multiple spalled layers of NiS/sub 2/, with some crystallites of NiS/sub 2/ appearing on the leading face of the scale and between the individual scale layers. This spalling causes smaller coupons of the Hastelloy B to corrode faster than larger coupons

  16. Creep properties of EB welded joint on Hastelloy X

    International Nuclear Information System (INIS)

    Arata, Yoshiaki; Susei, Shuzo; Shimizu, Shigeki; Satoh, Keisuke; Nagai, Hiroyoshi.

    1980-01-01

    In order to clarify the creep properties of EB welds on Hastelloy X which is one of the candidate alloys for components of VHTR, creep tests on EB weld metal and welded joint were carried out. The results were discussed in comparison with those of base metal and TIG welds. Further, EB welds were evaluated from the standpoint of high temperature structural design. The results obtained are summarized as follows. 1) Both creep rupture strengths of EB weld metal and EB welded joint are almost equal to that of base metal, but those of TIG welds are lower than base metal. As for the secondary creep rate, EB weld metal is higher and TIG weld metal is lower than base metal. As for the time to onset of tertiary creep, no remarkable difference among base metal, EB weld metal and TIG weld metal is observed. 2) In case of EB weld metal, although anisotropy is slightly observed, the ductility is same or more as compared with base metal. In case of TIG weld metal, on the contrary, anisotropy is not observed and the ductility is essentially low. 3) Such rupture morphology of EB weld metal as appears to have resulted from interconnection of voids which occurred at grain boundary is similar to base metal. In case of TIG weld metal, however, many cracks with sharp tips are observed at grain boundary, and the rupture appears to have occurred in brittle by propagation and connection of the cracks. 4) It can be said from the standpoint of high temperature structural design that EB welding is very suitable to welding for structure where creep effects are significant, because both of the creep ductility and the rupture strength are almost equal to those of base metal. (author)

  17. Applicability of creep damage rules to a nickel-base heat-resistant alloy Hastelloy XR

    International Nuclear Information System (INIS)

    Tsuji, Hirokazu; Nakajima, Najime; Tanabe, Tatsuhiko; Nakasone, Yuji

    1992-01-01

    A series of constant load and temperature creep rupture tests and varying load and/or temperature creep rupture tests was carried out on a nickel-base heat-resistant alloy Hastelloy XR, which was developed for applications in the High-Temperature Engineering Test Reactor, at temperatures ranging from 850 to 1000deg C in order to examine the applicability of the conventional creep damage rules, i.e., the life fraction, the strain fraction and their mixed rules. The life fraction rule showed the best applicability of these three criteria. The good applicability of the rule was considered to result from the fact that the creep strength of Hastelloy XR was not strongly affected by the change of the chemical composition and/or the microstructure during exposure to the high-temperature simulated HTGR helium environment. In conclusion the life fraction rule is applicable in engineering design of high-temperature components made of Hastelloy XR. (orig.)

  18. Status of tellurium--hastelloy N studies in molten fluoride salts

    International Nuclear Information System (INIS)

    Keiser, J.R.

    1977-10-01

    Tellurium, which is a fission product in nuclear reactor fuels, can embrittle the surface grain boundaries of nickel-base structural materials. This report summarizes results of an experimental investigation conducted to understand the mechanism and to develop a means of controlling this embrittlement in the alloy Hastelloy N. The addition of a chromium telluride to salt can be used to provide small partial pressures of tellurium simulating a reactor environment where tellurium appears as a fission product. The intergranular embrittlement produced in Hastelloy N when exposed to this chromium telluride-salt mixture can be reduced by adding niobium to the Hastelloy N or by controlling the oxidation potential of the salt in the reducing range

  19. Creep properties of hastelloy x and their application to the structural design

    International Nuclear Information System (INIS)

    Kiyoshige, Masanori; Murase, Hirokazu; Fujioka, Junzo; Shimizu, Shigeki; Satoh, Keisuke.

    1978-01-01

    In the creep curve of Hastelloy X, it was difficult to divide it into the three stages of creep. However, these stages were made distinguishable by plotting the relationship between creep rates and time in double-logarithmic coordinates. All the creep data of Hastelloy X, except the isochronous stress-strain curves, required for determining the design stress intensities S sub(o) and S sub(t) were arranged through the Larson-Miller parameter. The isochronous stress-strain curves for a heat of Hastelloy X were derived from the constitutive equations obtained from short-term data. A fairly good agreement between the predicted data and the experimental data was obtained. (auth.)

  20. Influence of temperature, environment, and thermal aging on the continuous cycle fatigue behavior of Hastelloy X and Inconel 617

    International Nuclear Information System (INIS)

    Strizak, J.P.; Brinkman, C.R.; Booker, M.K.; Rittenhouse, P.L.

    1982-04-01

    Results are presented for strain-controlled fatigue and tensile tests for two nickel-base, solution-hardened reference structural alloys for use in several High-Temperature Gas-Cooled Reactor (HTGR) concepts. These alloys, Hastelloy X and Inconel 617, were tested from room temperature to 871 0 C in air and impure helium. Materials were tested in both the solution-annealed and the preaged conditios, in which aging consisted of isothermal exposure at one of several temperatures for periods of up to 20,000 h. Comparisons are given between the strain-controlled fatigue lives of these and several other commonly used alloys, all tested at 538 0 C. An analysis is also presented of the continuous cycle fatigue data obtained from room temperature to 427 0 C for Hastelloy G, Hastelloy X, Hastelloy C-276, and Hastelloy C-4, an effort undertaken in support of ASME code development

  1. FINITE ELEMENT ANALYSIS OF HASTELLOY C-22HS IN END MILLING

    Directory of Open Access Journals (Sweden)

    K. Kadirgama

    2011-12-01

    Full Text Available This paper presents a finite element analysis of the stress distribution in the end milling operation of nickel-based superalloy HASTELLOY C-2000. Commercially available finite element software was used to develop the model and analyze the distribution of stress components in the machined surface of HASTELLOY C-22HS following end milling with coated carbide tools. The friction interaction along the tool-chip interface was modeled using the Coulomb friction law. It was found that the stress had lower values under the cut surface and that it increased gradually near the cutting edge.

  2. Effect of temperature upon the fatigue-crack propagation behavior of Hastelloy X-280

    International Nuclear Information System (INIS)

    James, L.A.

    1976-05-01

    The techniques of linear-elastic fracture mechanics were employed to characterize the effect of temperature upon the fatigue-crack propagation behavior of Hastelloy X-280 in an air environment. Also included in this study are survey tests to determine the effects of thermal aging and stress ratio upon crack growth behavior in this alloy

  3. Creep properties of Hastelloy X and their application to structural design

    International Nuclear Information System (INIS)

    Kiyoshige, Masanori; Murase, Koichi; Fujioka, Junzo; Shimizu, Shigeki; Satoh, Keisuke

    1977-01-01

    Creep and stress rupture tests on three heats of Hastelloy X differing in the manufacturing process were carried out at 800 0 C, 900 0 C and 1000 0 C. Interpretation of the observed creep properties was made, and a method for predicting necessary design data from the experimentally obtained results was discussed. The results are as follows. (1) It was difficult to separate the primary, secondary and tertiary creep stages in the creep curve of Hastelloy X of the present tests. However, those were made distinguishable by plotting the results in a double-logarithmic coordinates. From these creep rate curves, the primary and secondary creep rates and the times to the initiation of secondary and tertiary creeps were derived. (2) It is considered that the same stress and temperature dependences between the primary and secondary creep rates exist in the creep behaviour of Hastelloy X of the present tests. (3) All the creep data, except the isochronous stress-strain curve, required for the design such as stress vs. rupture time, stress vs. secondary creep rate and stress vs. time to initiation of tertiary creep could be arranged through the Larson-Miller parameter. On the other hand, the isochronous stress-strain curve was figured out by estimating creep curves. The constitutive equations of creep for a heat of Hastelloy X proposed in this paper and the isochronous stress-strain curves derived from these constitutive equations were consistent with the experimental data obtained for the corresponding material. (auth.)

  4. Oxidation in air of two refractory alloys (Nicral D and Hastelloy X) at 900 and 1100 deg. C

    International Nuclear Information System (INIS)

    Sannier, J.; Dominget, R.; Darras, R.

    1960-01-01

    The oxidation in air of two refractory alloys (Nicral D and Hastelloy X) has been studied at 900 and 1100 deg. C, by means of recording thermo-balances and microscopic cross section examination. At 900 deg. C, the surface oxidation rates of the two alloys are quite similar, but at 1100 deg. C the alloy Nicral D oxidizes faster than the alloy Hastelloy X. On the other hand, after heating at 1100 deg. C for 150 hours, Nicral D shows both intergranular oxidation and a small amount of internal oxidation, whereas Hastelloy X is especially subject to internal oxidation. In addition, two descaling methods were compared: an electrolytic method, in a sodium hydroxide-sodium carbonate bath, and a chemical method using a sodium nitrate-sodium peroxide bath; the latter appears suitable only for Hastelloy X. Reprint of a paper published in Journal of nuclear materials, 3, p. 213-225, 1959 [fr

  5. Composition of eta carbide in Hastelloy N after aging 10,000 hr at 8150C

    International Nuclear Information System (INIS)

    Leitnaker, J.M.; Potter, G.A.; Bradley, D.J.; Franklin, J.C.; Laing, W.R.

    1977-11-01

    The composition of the eta carbide in Hastelloy N containing 0.7 wt percent Si in the alloy approaches M 12 C, rather than M 6 C as indicated in the alloy literature. The silicon content of the eta phase in this case was about 25 at. percent, much higher than has been observed in less highly alloyed material. The data do not permit a definition of the limiting compositions of the phases

  6. Microstructures, mechanical properties and corrosion resistance of Hastelloy C22 coating produced by laser cladding

    International Nuclear Information System (INIS)

    Wang, Qin-Ying; Zhang, Yang-Fei; Bai, Shu-Lin; Liu, Zong-De

    2013-01-01

    Highlights: ► Hastelloy C22 coatings were prepared by diode laser cladding technique. ► Higher laser speed resulted in smaller grain size. ► Size-effect played the key role in the hardness measurements by different ways. ► Coating with higher laser scanning speed displayed higher nano-scratch resistance. ► Small grain size was beneficial for improvement of coating corrosion resistance. -- Abstract: The Hastelloy C22 coatings H1 and H2 were prepared by laser cladding technique with laser scanning speeds of 6 and 12 mm/s, respectively. Their microstructures, mechanical properties and corrosion resistance were investigated. The microstructures and phase compositions were studied by metallurgical microscope, scanning electron microscope and X-ray diffraction analysis. The hardness and scratch resistance were measured by micro-hardness and nanoindentation tests. The polarization curves and electrochemical impedance spectroscopy were tested by electrochemical workstation. Planar, cellular and dendritic solidifications were observed in the coating cross-sections. The coatings metallurgically well-bonded with the substrate are mainly composed of primary phase γ-nickel with solution of Fe, W, Cr and grain boundary precipitate of Mo 6 Ni 6 C. The hardness and corrosion resistance of steel substrate are significantly improved by laser cladding Hastelloy C22 coating. Coating H2 shows higher micro-hardness than that of H1 by 34% and it also exhibits better corrosion resistance. The results indicate that the increase of laser scanning speed improves the microstuctures, mechanical properties and corrosion resistance of Hastelloy C22 coating

  7. Weldability and weld performance of a special grade Hastelloy-X modified for high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Shimizu, S.; Mutoh, Y.

    1984-01-01

    The characteristics of weld defects in the electron beam (EB) welding and the tungsten inert gas (TIG) arc welding for Hastelloy-XR, a modified version of Hastelloy-X, are clarified through the bead-on-plate test and the Trans-Varestraint test. Based on the results, weldabilities on EB and TIG weldings for Hastelloy-XR are discussed and found to be almost the same as Hastelloy-X. The creep rupture behaviors of the welded joints are evaluated by employing data on creep properties of the base and the weld metals. According to the evaluation, the creep rupture strength of the EB-welded joint may be superior to that of the TIG-welded joint. The corrosion test in helium containing certain impurities is conducted for the weld metals. There is no significant difference of such corrosion characteristics as weight gain, internal oxidation, depleted zone, and so on between the base and the weld metals. Those are superior to Hastelloy-X

  8. Corrosion tests of 316L and Hastelloy C-22 in simulated tank waste solutions

    International Nuclear Information System (INIS)

    Danielson, M.J.; Pitman, S.G.

    2000-01-01

    Both the 316L stainless steel and Hastelloy C-22 gave satisfactory corrosion performance in the simulated test environments. They were subjected to 100 day weight loss corrosion tests and electrochemical potentiodynamic evaluation. This activity supports confirmation of the design basis for the materials of construction of process vessels and equipment used to handle the feed to the LAW-melter evaporator. BNFL process and mechanical engineering will use the information derived from this task to select material of construction for process vessels and equipment

  9. ERDA, RBS, TEM and SEM characterization of microstructural evolution in helium-implanted Hastelloy N alloy

    Energy Technology Data Exchange (ETDEWEB)

    Gao, Jie [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); School of Physical Sciences, University of Chinese Academy of Sciences, Beijing 100049 (China); Bao, Liangman [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Huang, Hefei, E-mail: huanghefei@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Li, Yan, E-mail: liyan@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Lei, Qiantao [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Institute of Modern Physics, Fudan University, Shanghai 200433 (China); Deng, Qi [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Liu, Zhe; Yang, Guo [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); School of Physical Sciences, University of Chinese Academy of Sciences, Beijing 100049 (China); Shi, Liqun [Institute of Modern Physics, Fudan University, Shanghai 200433 (China)

    2017-05-15

    Hastelloy N alloy was implanted with 30 keV, 5 × 10{sup 16} ions/cm{sup 2} helium ions at room temperature, and subsequent annealed at 600 °C for 1 h and further annealed at 850 °C for 5 h in vacuum. Using elastic recoil detection analysis (ERDA) and transmission electron microscopy (TEM), the depth profiles of helium concentration and helium bubbles in helium-implanted Hastelloy N alloy were investigated, respectively. The diffusion of helium and molybdenum elements to surface occurred during the vacuum annealing at 850 °C (5 h). It was also observed that bubbles in molybdenum-enriched region were much larger in size than those in deeper region. In addition, it is worth noting that plenty of nano-holes can be observed on the surface of helium-implanted sample after high temperature annealing by scanning electron microscope (SEM). This observation provides the evidence for the occurrence of helium release, which can be also inferred from the results of ERDA and TEM analysis.

  10. Effect of grain size and cold working on high temperature strength of Hastelloy X

    International Nuclear Information System (INIS)

    Fujioka, J.; Murase, H.; Matsuda, S.

    1980-01-01

    Effect of grain size and cold working on creep, creep rupture, low cycle fatigue and tensile strengths of Hastelloy X were studied at temperatures ranging from 800 to 1000 0 C. In order to apply these data to design, the allowable design stresses were estimated by expanding the criteria of ASME Code Case 1592 to such a high temperature range. The allowable design stress increased, on the other hand, the low cycle fatigue life decreased with increasing grain size. Cold working up to a ratio of 5 per cent may not be a serious problem in design, because the allowable design stress and the fatigue life were little affected. The cause of these variations in strength was discussed by examining the initiation and growth of cracks, and the microstructures. (author)

  11. Creep curve formularization at 950degC for Hastelloy XR

    International Nuclear Information System (INIS)

    Kaji, Yoshiyuki; Muto, Yasushi

    1991-03-01

    Creep tests under constant stress were conducted on a nickel-base heat-resistant alloy, Hastelloy XR, in air at 950degC. Minimum creep strain rate, time to the onset of tertiary creep and time to rupture were obtained as a function of applied stress. Then, a creep constitutive equation was made based on the Garofalo formula for primary and secondary creep and based on the Kachanov-Rabotnov formula for tertiary creep, which could represent fairly well the experimental creep deformation curves under the constant stress conditions. The creep deformation under the constant load condition corresponding to the stress increment was analysed using the creep constitutive equation and strain hardening law. Then the calculated creep strain showed slightly higher value than the experimental creep strain, and the calculated life was shorter than the experimental one. (author)

  12. Metallurgical and environmental factors influencing creep behaviour of hastelloy-X

    International Nuclear Information System (INIS)

    Kiuchi, Kiyoshi; Kondo, Tatsuo

    1979-03-01

    Creep and rupture behaviours of Hastelloy-X and its modified version were examined with special reference to the effect of different test environments; i.e. air, high vacuum and the simulated HTR helium coolant. The respective environments showed different effects. The vacuum environment of about 10 -8 torr. gave best reproducible behaviour with essentially no surface-to-volume ratio effect. Such size effect was significant in the other two environments. The simulated HTR environment was characterized in its potentiality of both oxidizing selected alloy constituents and carburization. The observed behaviour was attributed to the depletion of strengthning solute elements caused by the surface reactions and the associated solid state reactions. (author)

  13. Creep curve modeling of hastelloy-X alloy by using the theta projection method

    International Nuclear Information System (INIS)

    Woo Gon, Kim; Woo-Seog, Ryu; Jong-Hwa, Chang; Song-Nan, Yin

    2007-01-01

    To model the creep curves of the Hastelloy-X alloy which is being considered as a candidate material for the VHTR (Very High Temperature gas-cooled Reactor) components, full creep curves were obtained by constant-load creep tests for different stress levels at 950 C degrees. Using the experimental creep data, the creep curves were modeled by applying the Theta projection method. A number of computing processes of a nonlinear least square fitting (NLSF) analysis was carried out to establish the suitably of the four Theta parameters. The results showed that the Θ 1 and Θ 2 parameters could not be optimized well with a large error during the fitting of the full creep curves. On the other hand, the Θ 3 and Θ 4 parameters were optimized well without an error. For this result, to find a suitable cutoff strain criterion, the NLSF analysis was performed with various cutoff strains for all the creep curves. An optimum cutoff strain range for defining the four Theta parameters accurately was found to be a 3% cutoff strain. At the 3% cutoff strain, the predicted curves coincided well with the experimental ones. The variation of the four Theta parameters as the function of a stress showed a good linearity, and the creep curves were modeled well for the low stress levels. Predicted minimum creep rate showed a good agreement with the experimental data. Also, for a design usage of the Hastelloy-X alloy, the plot of the log stress versus log the time to a 1% strain was predicted, and the creep rate curves with time and a cutoff strain at 950 C degrees were constructed numerically for a wide rang of stresses by using the Theta projection method. (authors)

  14. A possibility of enhancing Jc in MgB2 film grown on metallic hastelloy tape with the use of SiC buffer layer

    International Nuclear Information System (INIS)

    Putri, W. B. K.; Kang, B.; Ranot, M.; Lee, J. H.; Kang, W. N.

    2014-01-01

    We have grown MgB 2 on SiC buffer layer by using metallic Hastelloy tape as the substrate. Hastelloy tape was chosen for its potential practical applications, mainly in the power cable industry. SiC buffer layers were deposited on Hastelloy tapes at 400, 500, and 600 degrees C by using a pulsed laser deposition method, and then by using a hybrid physical-chemical vapor deposition technique, MgB 2 films were grown on the three different SiC buffer layers. An enhancement of critical current density values were noticed in the MgB 2 films on SiC/Hastelloy deposited at 500 and 600 degrees C. From the surface analysis, smaller and denser grains of MgB 2 tapes are likely to cause this enhancement. This result infers that the addition of SiC buffer layers may contribute to the improvement of superconducting properties of MgB 2 tapes.

  15. Etude expérimentale du soudage par laser YAG de l'alliage base nickel Hastelloy X Experimental study of YAG laser welding of nickel base alloy Hastelloy X

    Directory of Open Access Journals (Sweden)

    Graneix Jérémie

    2013-11-01

    Full Text Available Le procédé de soudage laser YAG est envisagé pour remplacer le procédé de soudage TIG manuel pour la réalisation de pièces de turboréacteur en alliage nickel-chrome-molybdène Hastelloy X. Cette étude expérimentale a permis de définir un domaine de soudabilité de cet alliage répondant aux critères spécifiques du secteur aéronautique. The YAG laser welding process is contemplated to replace the manual TIG welding process for the production of parts of turbojet in Hastelloy X. This experimental study has identified the field of weldability of this alloy to meet the specific requirements of the aerospace industry.

  16. Creep properties of base metal and welded joint of Hastelloy XR produced for High-Temperature Engineering Test Reactor in simulated primary coolant helium

    International Nuclear Information System (INIS)

    Kurata, Yuji; Tsuji, Hirokazu; Shindo, Masami; Suzuki, Tomio; Tanabe, Tatsuhiko; Mutoh, Isao; Hiraga, Kenjiro

    1999-01-01

    Creep tests of base metal, weld metal and welded joint of Hastelloy XR, which had the same chemical composition as Hastelloy XR produced for an intermediate heat exchanger of the High-Temperature Engineering Test Reactor, were conducted in simulated primary coolant helium. The weld metal and welded joint showed almost equal to or longer rupture time than the base metal of Hastelloy XR at 850 and 900degC, although they gave shorter rupture time at 950degC under low stress and at 1,000degC. The welded joint of Hastelloy XR ruptured at the base metal region at 850 and 900degC. On the other hand, it ruptured at the weld metal region at 950 and 1,000degC. The steady-state creep rate of weld metal of Hastelloy XR was lower than that of base metal at 850, 900 and 950degC. The creep rupture strengths of base metal, weld metal and welded joint of Hastelloy XR obtained in this study were confirmed to be much higher than the design allowable creep-rupture stress (S R ) of the Design Allowable Limits below 950degC. (author)

  17. The anti-corrosion behavior under multi-factor impingement of Hastelloy C22 coating prepared by multilayer laser cladding

    Science.gov (United States)

    Chen, Lin; Bai, Shu-Lin

    2018-04-01

    Hastelloy C22 coating was prepared on substrate of Q235 steel by high power multilayer laser cladding. The microstructure, hardness and anti-corrosion properties of coating were investigated. The corrosion tests in 3.5% NaCl solution were carried out with variation of impingement angle and velocity, and vibration frequency of sample. The microstructure of coating changes from equiaxed grain at the top surface to dendrites oriented at an angle of 60° to the substrate inside the coating. The corrosion rate of coating increases with the increase of impingement angle and velocity, and vibrant frequency of sample. Corrosion mechanisms relate to repassivation and depassivation of coating according to electrochemical measurements. Above results show that multilayer laser cladding can endow Hastelloy C22 coating with fine microstructures, high hardness and good anti-corrosion performances.

  18. Relationship between carburization and zero-applied-stress creep dilation in Alloy 800H and Hastelloy X

    International Nuclear Information System (INIS)

    Inouye, H.; Rittenhouse, P.L.

    1981-01-01

    Typical HTGR candidate alloys can carburize when exposed to simulated service environments. The carbon concentration gradients so formed give rise to internal stresses which could cause dilation. Studies performed with Hastelloy X and Alloy 800H showed that dilations of up to almost 1% can occur at 1000 0 C when carbon pickup is high. Dilation was normally observed only when the carbon increase was >1000 μg/cm 2 and ceased when diffusing carbon reached the center of the specimen. (Auth.)

  19. Comparison of creep behavior under varying load/temperature conditions between Hastelloy XR alloys with different boron content levels

    International Nuclear Information System (INIS)

    Tsuji, Hirokazu; Nakajima, Hajime; Shindo, Masami; Tanabe, Tatsuhiko; Nakasone, Yuji.

    1996-01-01

    In the design of the high-temperature components, it is often required to predict the creep rupture life under the conditions in which the stress and/or temperature may vary by using the data obtained with the constant load and temperature creep rupture tests. Some conventional creep damage rules have been proposed to meet the above-mentioned requirement. Currently only limited data are available on the behavior of Hastelloy XR, which is a developed alloy as the structural material for high-temperature components of the High-Temperature Engineering Test Reactor (HTTR), under varying stress and/or temperature creep conditions. Hence a series of constant load and temperature creep rupture tests as well as varying load and temperature creep rupture tests was carried out on two kinds of Hastelloy XR alloys whose boron content levels are different, i.e., below 10 and 60 mass ppm. The life fraction rule completely fails in the prediction of the creep rupture life of Hastelloy XR with 60 mass ppm boron under varying load and temperature conditions though the rule shows good applicability for Hastelloy XR with below 10 mass ppm boron. The change of boron content level of the material during the tests is the most probable source of impairing the applicability of the life fraction rule to Hastelloy XR whose boron content level is 60 mass ppm. The modified life fraction rule has been proposed based on the dependence of the creep rupture strength on the boron content level of the alloy. The modified rule successfully predicts the creep rupture life under the two stage creep test conditions from 1000 to 900degC. The trend observed in the two stage creep tests from 900 to 1000degC can be qualitatively explained by the mechanism that the oxide film which is formed during the prior exposure to 900degC plays the role of the protective barrier against the boron dissipation into the environment. (J.P.N.)

  20. Weldability of the superalloys Haynes 188 and Hastelloy X by Nd:YAG

    Directory of Open Access Journals (Sweden)

    Graneix Jérémie

    2014-01-01

    Full Text Available The requirements for welded aircraft parts have become increasingly severe, especially in terms of the reproducibility of the geometry and metallurgical grade of the weld bead. Laser welding is a viable method of assembly to meet these new demands, because of automation, to replace the manual TIG welding process. The purpose of this study is to determine the weldability of Hastelloy X and Haynes 188 alloys by the butt welding process with a Nd:YAG laser. To identify the influential parameters of the welding process (laser power, feed rate, focal diameter and flow of gas while streamlining testing, an experimental design was established with the CORICO software using the graphic correlation method. The position of the focal point was fixed at 1/3 of the thickness of the sheet. The gas flow rate and the power of the beam have a major effect on the mechanical properties and geometry of the weld. The strength of the weld is comparable to that of the base metal. However, there is a significant decrease in the elongation at break of approximately 30%. The first observations of the cross section of the weld by scanning electron microscopy coupled with EBSD analysis show a molten zone presenting dendritic large grains compared to the equiaxed grains of the base metals without a heat affected zone.

  1. Ion irradiation-induced swelling and hardening effect of Hastelloy N alloy

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, S.J. [Key Laboratory of Artificial Micro-and Nano-structures of Ministry of Education, School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Li, D.H.; Chen, H.C.; Lei, G.H.; Huang, H.F.; Zhang, W.; Wang, C.B. [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Yan, L., E-mail: yanlong@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Fu, D.J. [Key Laboratory of Artificial Micro-and Nano-structures of Ministry of Education, School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Tang, M. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2017-06-15

    The volumetric swelling and hardening effect of irradiated Hastelloy N alloy were investigated in this paper. 7 MeV and 1 MeV Xe ions irradiations were performed at room temperature (RT) with irradiation dose ranging from 0.5 to 27 dpa. The volumetric swelling increases with increasing irradiation dose, and reaches up to 3.2% at 27 dpa. And the irradiation induced lattice expansion is also observed. The irradiation induced hardening initiates at low ion dose (≤1dpa) then saturates with higher ion dose. The irradiation induced volumetric swelling may be ascribed to excess atomic volume of defects. The irradiation induced hardening may be explained by the pinning effect where the defects can act as obstacles for the free movement of dislocation lines. And the evolution of the defects' size and number density could be responsible for the saturation of hardness. - Highlights: •Irradiation Swelling: The irradiation induced volumetric swelling increases with ion dose. •Irradiation Hardening: The irradiation hardening initiates below 1 dpa, then saturates with higher ion dose (1–10 dpa). •Irradiation Mechanism: The irradiation phenomena are ascribed to the microstructural evolution of the irradiation defects.

  2. Laser texturing of Hastelloy C276 alloy surface for improved hydrophobicity and friction coefficient

    Science.gov (United States)

    Yilbas, B. S.; Ali, H.

    2016-03-01

    Laser treatment of Hastelloy C276 alloy is carried out under the high pressure nitrogen assisting gas environment. Morphological and metallurgical changes in the laser treated layer are examined using the analytical tools including, scanning electron and atomic force microscopes, X-ray diffraction, energy dispersive spectroscopy, and Fourier transform infrared spectroscopy. Microhardness is measured and the residual stress formed in the laser treated surface is determined from the X-ray data. The hydrophibicity of the laser treated surface is assessed using the sessile drop method. Friction coefficient of the laser treated layer is obtained incorporating the micro-tribometer. It is found that closely spaced laser canning tracks create a self-annealing effect in the laser treated layer and lowers the thermal stress levels through modifying the cooling rates at the surface. A dense structure, consisting of fine size grains, enhances the microhardness of the surface. The residual stress formed at the surface is compressive and it is in the order of -800 MPa. Laser treatment improves the surface hydrophobicity significantly because of the formation of surface texture composing of micro/nano-pillars.

  3. Temperature dependence of creep properties of cold-worked Hastelloy XR

    International Nuclear Information System (INIS)

    Kurata, Yuji; Nakajima, Hajime

    1995-01-01

    The creep properties of Hastelloy XR, in a solution treated, 10% or 20% cold-worked condition, were investigated at temperatures from 800 to 1,000degC for the duration of creep tests up to about 2,500 ks. At 800 and 850degC, the steady-state creep rate and rupture ductility decreased and the rupture life increased after cold work of 10% or 20%. Although the rupture life of the 10% cold-worked alloy was longer at 900degC than that of the solution treated one, the rupture lives of the 10% cold-worked and solution treated alloys were almost equal at 950degC, which is the highest helium temperature in an intermediate heat exchanger of the High Temperature Engineering Test Reactor (HTTR). The beneficial effect of 10% cold work on the rupture life and the steady-state creep rate disappeared at 1,000degC. The beneficial effect of 20% cold work disappeared at 950degC because significant dynamic recrystallization occurred during creep. While rupture ductility of this alloy decreased after cold work of 10% or 20%, it recovered to a considerable extend at 1,000degC. It is emphasized that these cold work effects should be taken into consideration in design, operation and residual life estimation of high temperature components of the HTTR. (author)

  4. Creep strength of hastelloy X TIG-welded cylinder under internal pressure at elevated temperature

    International Nuclear Information System (INIS)

    Udoguchi, Teruyoshi; Indo, Hirosato; Isomura, Kazuyuki; Kobatake, Kiyokazu; Nakanishi, Tsuneo.

    1981-01-01

    Creep tests on circumferentially TIG-welded Hastelloy x cylinders were carried out under internal pressure for the investigation of structural behavior of welded components in high temperature environment. The creep rupture strength of TIG-welded cylinders was much lower than that of non-welded cylinders, while such reduction was not found in uniaxial creep tests on TIG-welded bars. It was deduced that the reduction was due to the low ductility (ranging from 1 to 5%) of the weld metal to which enhanced creep was induced by the adjacent base metal whose creep strain rate was much higher than that of the weld metal. Therefore, uniaxial creep tests on bar specimens is not sufficient for proper assessment of the creep rupture strength of welded components. Both creep strain rate and creep ductility should be concerned for the assessment. Creep tests by using components such as cylinder under internal pressure are recommendable for the confirmation of creep strength of welded structures and components. (author)

  5. Effects of product form and boron addition on the creep damage in the modified Hastelloy X alloys in a simulated HTGR helium gas environment

    International Nuclear Information System (INIS)

    Nakasone, Yuji; Tanabe, Tatsuhiko; Tsuji, Hirokazu; Nakajima, Hajime.

    1992-01-01

    The present paper investigates early-stage-creep damage of Hastelloy XR and XR-II alloys, modified versions of Hastelloy X alloy, which have been developed in Japan as most promising candidate structural alloys for Japanese high-temperature gas-cooled reactors (HTGRs). Creep tests were made on Hastelloy XR forging, tube and XR-II tube at 1,123 to 1,273 K in a simulated HTGR helium gas environment. The tests were interrupted at different strain levels of up to 5 % in order to evaluate creep damage via intergranular voids. The void sizes along grain boundaries and the A-parameter, the ratio of the number of damaged grain boundaries, on which one or more voids are found, to that of the total grain boundaries observed are used in order to evaluate creep damage. Statistical analysis of the A-parameter as well as the void sizes reveals that the values of the parameter show wide variations and follow the Weibull distribution, reflecting spatial randomness of the voids. The void sizes along grain boundaries, on the other hand, follow the log-normal distribution. The maximum void size d max and the mean value of the A-parameter A m are calculated and plotted against interruption creep strain ε int . The resultant d max vs. ε int and A m vs. ε int diagrams show that Hastelloy XR forging had suffered more damage than Hastelloy XR tube; nevertheless, the forging has longer interruption life, or the time to reach a given interruption creep strain. The result indicates that grains may have been deformed more easily in Hastelloy XR in the form of tube than in the form of forging. The diagrams also imply that the addition of boron has suppressed the nucleation as well as the growth of voids and thus has brought about longer interruption life of Hastelloy XR-II. (author)

  6. Compatibility of aluminide-coated Hastelloy x and Inconel 617 in a simulated gas-cooled reactor environment

    International Nuclear Information System (INIS)

    Chin, J.; Johnson, W.R.; Chen, K.

    1982-03-01

    Commercially prepared aluminide coatings on Hastelloy X and Inconel 617 substrates were exposed to controlled-impurity helium at 850 0 and 950 0 C for 3000 h. Optical and scanning electron (SEM) microscopy, electron microprobe profiles, and SEM X-ray mapping were used to evaluate and compare exposed and unexposed control samples. Four coatings were evaluated: aluminide, aluminide with platinum, aluminide with chromium, and aluminide with rhodium. With extended time at elevated temperature, nickel diffused into the aluminide coatings to form epsilon-phase (Ni 3 Al). This diffusion was the primary cause of porosity formation at the aluminide/alloy interface

  7. Electrochemical impedance spectrometry using 316L steel, hastelloy, maraging, Inconel 600, Elgiloy, carbon steel, TiN and NiCr. Simulation in tritiated water. 2 volumes

    International Nuclear Information System (INIS)

    Bellanger, G.

    1994-03-01

    Polarization and electrochemical impedance spectrometry curves are presented and discussed. These curves make it possible to ascertain the corrosion domains and to compare the slow and fast kinetics (voltammetry) of different stainless steel alloys. These corrosion kinetics, the actual or simulated tritiated water redox potentials, and the corrosion potentials provide a classification of the steels studied here: 316L, Hastelloy, Maraging, Inconel 600, Elgiloy, carbon steel and TiN and NiCr deposits. From the results it can be concluded that Hastelloy and Elgiloy have the best corrosion resistance. (author). 49 refs., 695 figs., tabs

  8. Creep-Rupture Properties and Corrosion Behaviour of 21/4 Cr-1 Mo Steel and Hastelloy X-Alloys in Simulated HTGR Environment

    DEFF Research Database (Denmark)

    Lystrup, Aage; Rittenhouse, P. L.; DiStefano, J. R.

    Hastelloy X and 2/sup 1///sub 4/ Cr-1 Mo steel are being considered as structural alloys for components of a High-Temperature Gas-Cooled Reactor (HTGR) system. Among other mechanical properties, the creep behavior of these materials in HTGR primary coolant helium must be established to form part...

  9. Collect Available Creep-Fatigue Data and Study Existing Creep-Fatigue Evaluation Procedures for Grade 91 and Hastelloy XR

    International Nuclear Information System (INIS)

    Asayama, Tai; Tachibana, Yukio

    2007-01-01

    This report describes the results of investigation on Task 5 of DOE/ASME Materials Project based on a contract between ASME Standards Technology, LLC (ASME ST-LLC) and Japan Atomic Energy Agency (JAEA). Task 5 is to collect available creep-fatigue data and study existing creep-fatigue evaluation procedures for Grade 91 steel and Hastelloy XR. Part I of this report is devoted to Grade 91 steel. Existing creep-fatigue data were collected (Appendix A) and analyzed from the viewpoints of establishing a creep-fatigue procedure for VHTR design. A fair amount of creep-fatigue data has been obtained and creep-fatigue phenomena have been clarified to develop design standards mainly for fast breeder reactors. Following this, existing creep-fatigue procedures were studied and it was clarified that the creep-fatigue evaluation procedure of the ASME-NH has a lot of conservatisms and they were analyzed in detail from the viewpoints of the evaluation of creep damage of material. Based on the above studies, suggestions to improve the ASME-NH procedure along with necessary research and development items were presented. Part II of this report is devoted to Hastelloy XR. Existing creep-fatigue data used for development of the high temperature structural design guideline for High Temperature Gas-cooled Reactor (HTGR) were collected. Creep-fatigue evaluation procedure in the design guideline and its application to design of the intermediate heat exchanger (IHX) for High Temperature Engineering Test Reactor (HTTR) was described. Finally, some necessary research and development items in relation to creep-fatigue evaluation for Gen IV and VHTR reactors were presented.

  10. Electrochemical impedance spectrometry using 316L steel, hastelloy, maraging, Inconel 600, Elgiloy, carbon steel, TiN and NiCr. Simulation in tritiated water. 2 volumes; Spectrometrie d`impedance electrochimique sur acier 316L, hastelloy, maraging inconel 600, elgiloy, acier au carbone, TiN, NiCr. Simulations en eau tritiee. 2 volumes

    Energy Technology Data Exchange (ETDEWEB)

    Bellanger, G.

    1994-03-01

    Polarization and electrochemical impedance spectrometry curves are presented and discussed. These curves make it possible to ascertain the corrosion domains and to compare the slow and fast kinetics (voltammetry) of different stainless steel alloys. These corrosion kinetics, the actual or simulated tritiated water redox potentials, and the corrosion potentials provide a classification of the steels studied here: 316L, Hastelloy, Maraging, Inconel 600, Elgiloy, carbon steel and TiN and NiCr deposits. From the results it can be concluded that Hastelloy and Elgiloy have the best corrosion resistance. (author). 49 refs., 695 figs., tabs.

  11. Evaluation on materials performance of Hastelloy Alloy XR for HTTR uses-5 (Creep properties of base metal and weldment in air)

    International Nuclear Information System (INIS)

    Watanabe, Katsutoshi; Nakajima, Hajime; Koikegami, Hajime; Higuchi, Makoto; Nakanishi, Tsuneo; Saitoh, Teiichiro; Takatsu, Tamao.

    1994-01-01

    Creep properties of weldment made from Hastelloy Alloy XR base metals and filler metals for the High Temperature Engineering Test Reactor (HTTR) components were examined by means of creep and creep rupture tests at 900 and 950degC in air. The results obtained are as follows: creep rupture strength was nearly equal or higher than that of Hastelloy Alloy XR master curve and was much higher than design creep rupture strength [S R ]. Furthermore, creep rupture strength and ductility of the present filler metal was in the data band in comparison with those of the previous filler metals. It is concluded from these reasons that this filler metal has fully favorable properties for HTTR uses. (author)

  12. Response Surface Design Model to Predict Surface Roughness when Machining Hastelloy C-2000 using Uncoated Carbide Insert

    International Nuclear Information System (INIS)

    Razak, N H; Rahman, M M; Kadirgama, K

    2012-01-01

    This paper presents to develop of the response surface design model to predict the surface roughness for end-milling operation of Hastelloy C-2000 using uncoated carbide insert. Mathematical model is developed to study the effect of three input cutting parameters includes the feed rate, axial depth of cut and cutting speed. Design of experiments (DOE) was implemented with the aid of the statistical software package. Analysis of variance (ANOVA) has been performed to verify the fit and adequacy of the developed mathematical model. The result shows that the feed rate gave the more effect on the both prediction values of Ra compared to the cutting speed and axial depth of cut. SEM and EDX analyses were performed in different cutting conditions. It can be concluded that the feed rate and cutting force give the higher impact to influence the machining characteristics of surface roughness. Thus, the optimizing the cutting conditions are essential in order to improve the surface roughness in machining of Hastlelloy C-2000.

  13. Analyzing the effect of cutting parameters on surface roughness and tool wear when machining nickel based hastelloy - 276

    International Nuclear Information System (INIS)

    Khidhir, Basim A; Mohamed, Bashir

    2011-01-01

    Machining parameters has an important factor on tool wear and surface finish, for that the manufacturers need to obtain optimal operating parameters with a minimum set of experiments as well as minimizing the simulations in order to reduce machining set up costs. The cutting speed is one of the most important cutting parameter to evaluate, it clearly most influences on one hand, tool life, tool stability, and cutting process quality, and on the other hand controls production flow. Due to more demanding manufacturing systems, the requirements for reliable technological information have increased. For a reliable analysis in cutting, the cutting zone (tip insert-workpiece-chip system) as the mechanics of cutting in this area are very complicated, the chip is formed in the shear plane (entrance the shear zone) and is shape in the sliding plane. The temperature contributed in the primary shear, chamfer and sticking, sliding zones are expressed as a function of unknown shear angle on the rake face and temperature modified flow stress in each zone. The experiments were carried out on a CNC lathe and surface finish and tool tip wear are measured in process. Machining experiments are conducted. Reasonable agreement is observed under turning with high depth of cut. Results of this research help to guide the design of new cutting tool materials and the studies on evaluation of machining parameters to further advance the productivity of nickel based alloy Hastelloy - 276 machining.

  14. Compatibility studies of type 316 stainless steel and Hastelloy N in KNO3--NaNO2--NaNO3

    International Nuclear Information System (INIS)

    Devan, J.H.; Keiser, J.R.

    1978-01-01

    The nitrate-based fused salt mixture KNO 3 --NaNO 2 --NaNO 3 (44--49--7 mol %) has been widely used as a heat transport fluid and for metallurgical heat-treating. We have measured the corrosion rate of this salt in the presence of a temperature gradient for an iron-base material, type 316 stainless steel, and a nickel-base material, Hastelloy N. Corrosion rates were measured with maximum loop temperatures of 431 and 504 0 C. Measured corrosion rates were in all cases less than 8 μm/year

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

    International Nuclear Information System (INIS)

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

    1979-12-01

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

  16. Influence of the brazing parameters on microstructure and mechanical properties of brazed joints of Hastelloy B2 nickel base alloy; Influencia de los parametros de soldeo fuerte en la microestructura y propiedades mecanicas de la union de la aleacion base niquel Hastelloy B2

    Energy Technology Data Exchange (ETDEWEB)

    Sotelo, J. C.; Gonzalez, M.; Porto, E.

    2014-07-01

    A study of the high vacuum brazing process of solid solution strengthened Hastelloy B2 nickel alloy has been done. A first stage of research has focused on the selection of the most appropriate brazing filler metal to the base material and vacuum furnace brazing process. The influence of welding parameters on joint microstructure constituents, relating the microstructure of the joint to its mechanical properties, has been evaluated. Two gaps of 50 and 200 micrometers, and two dwell times at brazing temperature of 10 and 90 minutes were studied. The braze joint mainly consists of the nickel rich matrix, nickel silicide and ternary compounds. Finally, the results of this study have shown the high bond strength for small gaps and increased dwell times of 90 minutes. (Author)

  17. Application of the Taguchi technique for the optimization of surface roughness and tool life during the milling of Hastelloy C22

    International Nuclear Information System (INIS)

    Kivak, Turgay; Mert, Senol

    2017-01-01

    In this study, the effects of machining parameters on surface roughness (Ra) and tool life (Tl) were investigated in the milling of Hastelloy C22 alloy with TiAlN-coated carbide inserts. A number of milling experiments were conducted using the L_2_7 (3"3) Taguchi orthogonal array on a CNC milling machine under different cutting conditions (dry, compressed air and wet). The cutting condition, cutting speed and feed rate were determined as the essential machining parameters. Analysis of variance (ANOVA) and signal-to-noise (S/N) ratio were employed to evaluate the effects of the machining parameters on Ra and Tl, and prediction models were created using quadratic regression analyses. The results revealed that the feed rate and cutting condition were the most influential factors on surface roughness and flank wear. The maximum tool life was achieved under wet cutting condition using a cutting speed of 30 x min"-"1 and a feed rate of 0.08 mm x rev"-"1, while the minimum surface roughness value was obtained under wet cutting condition using a cutting speed of 50 m x min"-"1 and the same feed rate. Using the optimum cutting parameters for Tl (30 m x min"-"1, 0.08 mm x rev"-"1), increases of 234 % and 67 % in tool life were observed under wet and compressed air cutting conditions, respectively, compared to the dry cutting condition.

  18. Application of the Taguchi technique for the optimization of surface roughness and tool life during the milling of Hastelloy C22

    Energy Technology Data Exchange (ETDEWEB)

    Kivak, Turgay; Mert, Senol [Duezce Univ. (Turkey). Dept. of Manufacturing Engineering

    2017-02-01

    In this study, the effects of machining parameters on surface roughness (Ra) and tool life (Tl) were investigated in the milling of Hastelloy C22 alloy with TiAlN-coated carbide inserts. A number of milling experiments were conducted using the L{sub 27} (3{sup 3}) Taguchi orthogonal array on a CNC milling machine under different cutting conditions (dry, compressed air and wet). The cutting condition, cutting speed and feed rate were determined as the essential machining parameters. Analysis of variance (ANOVA) and signal-to-noise (S/N) ratio were employed to evaluate the effects of the machining parameters on Ra and Tl, and prediction models were created using quadratic regression analyses. The results revealed that the feed rate and cutting condition were the most influential factors on surface roughness and flank wear. The maximum tool life was achieved under wet cutting condition using a cutting speed of 30 x min{sup -1} and a feed rate of 0.08 mm x rev{sup -1}, while the minimum surface roughness value was obtained under wet cutting condition using a cutting speed of 50 m x min{sup -1} and the same feed rate. Using the optimum cutting parameters for Tl (30 m x min{sup -1}, 0.08 mm x rev{sup -1}), increases of 234 % and 67 % in tool life were observed under wet and compressed air cutting conditions, respectively, compared to the dry cutting condition.

  19. One new route to optimize the oxidation resistance of TiC/hastelloy (Ni-based alloy) composites applied for intermediate temperature solid oxide fuel cell interconnect by increasing graphite particle size

    Science.gov (United States)

    Qi, Qian; Liu, Yan; Wang, Lujie; Zhang, Hui; Huang, Jian; Huang, Zhengren

    2017-09-01

    TiC/hastelloy composites with suitable thermal expansion and excellent electrical conductivity are promising candidates for IT-SOFC interconnect. In this paper, the TiC/hastelloy composites are fabricated by in-situ reactive infiltration, and the oxidation resistance of composites is optimized by increasing graphite particle size. Results show that the increase of graphite particles size from 1 μm to 40 μm reduces TiC particle size from 2.68 μm to 2.22 μm by affecting the formation process of TiC. Moreover, the decrease of TiC particles size accelerates the fast formation of dense and continuous TiO2/Cr2O3 oxide layer, which bring down the mass gain (800 °C/100 h) from 2.03 mg cm-2 to 1.18 mg cm-2. Meanwhile, the coefficient of thermal expansion decreases from 11.15 × 10-6 °C-1 to 10.80 × 10-6 °C-1, and electrical conductivity maintains about 5800 S cm-1 at 800 °C. Therefore, the decrease of graphite particle size is one simple and effective route to optimize the oxidation resistance of composites, and meantime keeps suitable thermal expansion and good electrical conductivity.

  20. Material design data of 2.25Cr-1Mo steel and hastelloy-x for the experimental multi-purpose very-high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Kodaira, Tsuneo; Suzuki, Michiaki; Uga, Takeo

    1975-08-01

    The preliminary structural design guidelines for the experimental multi-purpose very-high temperature gas-cooled reactor have recently been prepared. The components of the primary system operating at temperatures of creep dominant range are grouped in those of pressure and temperature boundaries respectively. In the material selection, 2 1/4Cr-1Mo steel is chosen for the former and Hastelloy-X for the latter taking into account of material properties at operating temperature. Deriving from the literature in the field, material design data of the alloys are established in design forms such as Sy, So, Sm, St, 100% of minimum stress to rupture, design fatigue curves, isochronous stress-strain curves, creep-fatigue interaction damage factor and so on, which are defined in ASME Code Section III, Code Case 1592. (auth.)

  1. Influencia de los parámetros de soldeo fuerte en la microestructura y propiedades mecánicas de la unión de la aleación base níquel Hastelloy B2

    Directory of Open Access Journals (Sweden)

    Sotelo, José Carlos

    2014-09-01

    Full Text Available A study of the high vacuum brazing process of solid solution strengthened Hastelloy B2 nickel alloy has been done. A first stage of research has focused on the selection of the most appropriate brazing filler metal to the base material and vacuum furnace brazing process. The influence of welding parameters on joint microstructure constituents, relating the microstructure of the joint to its mechanical properties, has been evaluated. Two gaps of 50 and 200 micrometers, and two dwell times at brazing temperature of 10 and 90 minutes were studied. The braze joint mainly consists of the nickel rich matrix, nickel silicide and ternary compounds. Finally, the results of this study have shown the high bond strength for small gaps and increased dwell times of 90 minutes.Se realizó un estudio pormenorizado del proceso de soldeo fuerte en horno de alto vacío de la aleación base níquel Hastelloy B2 fortalecida por solución sólida. En una primera fase del trabajo se seleccionó el material de aporte acorde al material objeto de unión y a la fuente de calentamiento seleccionada. Posteriormente, se evaluó la influencia del gap (50 y 200 micrómetros y tiempo de permanencia a temperatura de soldeo (10 y 90 minutos sobre los microconstituyentes de la unión, relacionando la microestructura con las propiedades mecánicas de la junta. Los análisis metalográficos mostraron una unión constituida por una matriz rica en níquel, siliciuros de níquel y compuestos ternarios. Finalmente, los resultados de los ensayos mecánicos a esfuerzos cortantes mostraron una elevada resistencia para gaps de 50 micrómetros y tiempos de permanencia de 90 minutos.

  2. 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)

  3. Absorption of molten fluoride salts in glassy carbon, pyrographite and Hastelloy B

    Czech Academy of Sciences Publication Activity Database

    Vacík, Jiří; Naramoto, H.; Červená, Jarmila; Hnatowicz, Vladimír; Peka, I.; Fink, D.

    2001-01-01

    Roč. 289, č. 3 (2001), s. 308-314 ISSN 0022-3115 R&D Projects: GA ČR GA202/96/0077; GA ČR GV202/97/K038; GA AV ČR KSK1010104 Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.366, year: 2001

  4. Fatigue performance of TBCs on Hastelloy X substrate during cyclic bending

    Czech Academy of Sciences Publication Activity Database

    Mušálek, Radek; Kovářík, O.; Tomek, L.; Medřický, Jan; Pala, Zdeněk; Haušild, P.; Čapek, J.; Kolařík, K.; Curry, N.; Bjorklund, S.

    2016-01-01

    Roč. 25, 1-2 (2016), s. 231-243 ISSN 1059-9630. [ITSC 2015: International Thermal Spray Conference and Exposition. Long Beach, California, 11.05.2015-14.05.2015] R&D Projects: GA ČR GB14-36566G Institutional support: RVO:61389021 Keywords : atmospheric plasma spray * failure mechanism * fatigue * HVAF * NiCoCrAlY * thermal barrier coatings * yttria-stabilized zirconia Subject RIV: JK - Corrosion ; Surface Treatment of Materials Impact factor: 1.488, year: 2016 http://link.springer.com/article/10.1007%2Fs11666-015-0321-4

  5. Composite Cu/Fe/MgB{sub 2} superconducting wires and MgB{sub 2}/YSZ/Hastelloy coated conductors for ac and dc applications

    Energy Technology Data Exchange (ETDEWEB)

    Glowacki, B A [Department of Materials Science and Metallurgy, University of Cambridge, Pembroke Street, Cambridge (United Kingdom); Majoros, M [Interdisciplinary Research Centre in Superconductivity, University of Cambridge, Madingley Road, Cambridge (United Kingdom); Vickers, M [Department of Materials Science and Metallurgy, University of Cambridge, Pembroke Street, Cambridge (United Kingdom); Eisterer, M [Atomic Institute of the Austrian Universities, A-1020 Vienna (Austria); Toenies, S [Atomic Institute of the Austrian Universities, A-1020 Vienna (Austria); Weber, H W [Atomic Institute of the Austrian Universities, A-1020 Vienna (Austria); Fukutomi, M [National Institute for Materials Science, Superconducting Materials Center, 1-2-1, Sengen, Ibaraki (Japan); Komori, K [National Institute for Materials Science, Superconducting Materials Center, 1-2-1, Sengen, Ibaraki (Japan); Togano, K [National Institute for Materials Science, Superconducting Materials Center, 1-2-1, Sengen, Ibaraki (Japan)

    2003-02-01

    We discuss the results of a study of MgB{sub 2} multifilamentary conductors and coated conductors from the point of view of their future dc and ac applications. The correlation between the slope of the irreversibility line induced by neutron irradiation defects and in situ structural imperfections and the critical temperature and critical current density is discussed with respect to the conductor performance and applicability. We debate the possible origin of the observed anomalous decrease of ac susceptibility at 50 K in copper clad in situ powder-in-tube MgB{sub 2} wires. Different conductor preparation methods and conductor architectures, and attainable critical current densities are presented. Some numerical results on critical currents, thermal stability and ac losses of future MgB{sub 2} multifilamentary and coated conductors with magnetic cladding of their filaments are also discussed.

  6. Étude du soudage LASER Yb : YAG homogène et hétérogène des superalliages Hastelloy X et Haynes 188

    OpenAIRE

    Graneix, Jérémie

    2015-01-01

    De nos jours, la complexité des pièces mécaniques est de plus en plus grande dans de nombreux secteurs industriels de pointes tels que l’aéronautique, l’aérospatiale ou bien encore le domaine médical. Pour répondre à ces nouvelles exigences, notamment en termes de géométrie, de nouvelles techniques de conception et de fabrication automatisées intègrant également une meilleure traçabilité des pièces, sont mises en place. Ce travail de thèse s’est inscrit dans un projet partenarial entre le Lab...

  7. Thermodynamic Assessment of Hot Corrosion Mechanisms of Superalloys Hastelloy N and Haynes 242 in Eutectic Mixture of Molten Salts KF and ZrF4

    Energy Technology Data Exchange (ETDEWEB)

    Michael V. Glazoff

    2012-02-01

    The KF - ZrF4 system was considered for the application as a heat exchange agent in molten salt nuclear reactors (MSRs) beginning with the work carried out at ORNL in early fifties. Based on a combination of excellent properties such as thermal conductivity, viscosity in the molten state, and other thermo-physical and rheological properties, it was selected as one of possible candidates for the nuclear reactor secondary heat exchanger loop.

  8. Development of plate-fin heat exchanger for intermediate heat exchanger of high-temperature gas cooled reactor. Fabrication process, high-temperature strength and creep-fatigue life prediction of plate-fin structure made of Hastelloy X

    International Nuclear Information System (INIS)

    Mizokami, Yorikata; Igari, Toshihide; Nakashima, Keiichi; Kawashima, Fumiko; Sakakibara, Noriyuki; Kishikawa, Ryouji; Tanihira, Masanori

    2010-01-01

    The helium/helium heat exchanger (i.e., intermediate heat exchanger: IHX) of a high-temperature gas-cooled reactor (HTGR) system with nuclear heat applications is installed between a primary system and a secondary system. IHX is operated at the highest temperature of 950degC and has a high capacity of up to 600 MWt. A plate-fin-type heat exchanger is the most suitable for IHX to improve construction cost. The purpose of this study is to develop an ultrafine plate-fin-type heat exchanger with a finer pitch fin than a conventional technology. In the first step, fabrication conditions of the ultrafine plate fin were optimized by press tests. In the second step, a brazing material was selected from several candidates through brazing tests of rods, and brazing conditions were optimized for plate-fin structures. In the third step, tensile strength, creep rupture, fatigue, and creep-fatigue tests were performed as typical strength tests for plate-fin structures. The obtained data were compared with those of the base metal and plate-fin element fabricated from SUS316. Finally, the accuracy of the creep-fatigue life prediction using both the linear cumulative damage rule and the equivalent homogeneous solid method was confirmed through the evaluation of creep-fatigue test results of plate-fin structures. (author)

  9. Mechanical characterization of superalloys for space reactors

    International Nuclear Information System (INIS)

    Duchesne, J.

    1989-01-01

    The aim of this work is the selection of structural materials that can be used in the temperature range 600-900 0 C for a gas cooled space reactor producing electricity. Superalloys fit best the temperature range required. Five nickel base alloys are chosen for their good mechanical behaviour: HAYNES 230, HASTELLOY S, HASTELLOY X, HASTELLOY XR and PYRAD 38D. Metallography, tensile and hardness tests are realized. Sample contraction is evidenced for some creep tests, under low stress: 20MPa at 800 0 C, on HAYNES 230 and HASTELLOY X, probably related to the structural evolution of these materials corresponding to a decrease of the crystal parameter [fr

  10. Development of Textured Buffer Layer on Metal Tapes for Oxide Superconductors

    National Research Council Canada - National Science Library

    Bhattacharya, Rabi

    2002-01-01

    .... UES, in collaboration with Argonne National Laboratory, has developed a multilayer architecture based on in-plane textured MgO film by inclined substrate deposition technique oristatic and moving Hastelloy substrates...

  11. Manufacture of a heat-resistant alloy with modified specifications for HTGR structural applications

    International Nuclear Information System (INIS)

    Sahira, K.; Kondo, T.; Takeiri, T.

    1984-01-01

    A method of manufacturing a nuclear grade nickel-base heat-resistant alloy in application to heliumcooled reactor primary circuit components has been developed. The Hastelloy-XR alloy, a version of Hastelloy-X, was made available by combining the basic studies of the oxidation behavior of Hastelloy-X and the improvement of manufacturing techniques. In the primary and remelting steps, the choice of appropriate processes was made by performing numerical analyses of the statistical deviation of both chemical composition and the products' mechanical properties. The feasibility of making larger electroslag remelting ingots with reasonable control of macrosegregation was examined by the calculation of a molten metal pool shape during melting. The hot workability of Hastelloy-XR was confirmed to be equivalent to that of Hastelloy-X and the importance of controlling the thermal and mechanical processes more closely was stressed in obtaining a higher level of quality assurance for the nuclear applications. The possibility of enhancing the high-temperature mechanical performance of Hastelloy-XR was suggested based on the preliminary test results with the heats manufactured with controlled boron content

  12. Preliminary lifetime predictions for 304 stainless steel as the LANL ABC blanket material

    International Nuclear Information System (INIS)

    Park, J.J.; Buksa, J.J.; Houts, M.G.; Arthur, E.D.

    1997-11-01

    The prediction of materials lifetime in the preconceptual Los Alamos National Laboratory (LANL) Accelerator-Based Conversion of Plutonium (ABC) is of utmost interest. Because Hastelloy N showed good corrosion resistance to the Oak Ridge National Laboratory Molten Salt Reactor Experiment fuel salt that is similar to the LANL ABC fuel salt, Hastelloy N was originally proposed for the LANL ABC blanket material. In this paper, the possibility of using 304 stainless steel as a replacement for the Hastelloy N is investigated in terms of corrosion issues and fluence-limit considerations. An attempt is made, based on the previous Fast Flux Test Facility design data, to predict the preliminary lifetime estimate of the 304 stainless steel used in the blanket region of the LANL ABC

  13. A life evaluation under creep-fatigue-environment interaction of Ni-base wrought alloys

    International Nuclear Information System (INIS)

    Hattori, Hiroshi; Kitagawa, Masaki; Ohtomo, Akira; Itoh, Mitsuyoshi

    1986-01-01

    In order to determine a failure criteria under cyclic loading and affective environment for HTGR systems, a series of strain controlled low-cycle fatigue tests were carried out at HTGR maximum gas temperatures in air, in vacuum and in HTGR helium environments on two nickel-base wrought alloys, namely Inconel 617 and Hastelloy XR. This paper first describes the creep-fatigue-environment properties of these alloys followed by a proposal of an evaluation method of creep-fatigue-environment interaction based on the experimental data to define the more reasonable design criteria, which is a modification of the linear damage summation rule. Second, the creep-fatigue properties of Hastelloy XR at 900 deg C and the result evaluated by this proposed method are shown. This criterion is successfully applied to the life prediction at 900 deg C. In addition, the creep-fatigue properties of Hastelloy XR-II are discussed. (author)

  14. Mechanical characterization of metallic materials for high-temperature gas-cooled reactors in air and in helium environments

    International Nuclear Information System (INIS)

    Sainfort, G.; Cappelaere, M.; Gregoire, J.; Sannier, J.

    1984-01-01

    In the French R and D program for high-temperature gas-cooled reactors (HTGRs), three metallic alloys were studied: steel Chromesco-3 with 2.25% chromium, alloy 800H, and Hastelloy-X. The Chromesco-3 and alloy 800H creep behavior is the same in air and in HTGR atmosphere (helium). The tensile tests of Hastelloy-X specimens reveal that aging has embrittlement and hardening effects up to 700 0 C, but the creep tests at 800 0 C show opposite effects. This particular behavior could be due to induced precipitation by aging and the depletion of hardening elements from the matrix. Tests show a low influence of cobalt content on mechanical properties of Hastelloy-X

  15. 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

  16. Compatibility of heat resistant alloys with boron carbide, (4)

    International Nuclear Information System (INIS)

    Baba, Sinichi; Saruta, Toru; Ooka, Kiichi; Tanaka, Isao; Aoyama, Isao

    1985-07-01

    This paper relates to the compatibility test of control rod sheath (Hastelloy XR alloy) and neutron absorber (boronated graphite) for the VHTR, which has been researched and developed by JAERI. The irradiation was conducted by using the OGL-1 irradiation facility in the JMTR in order to study reaction behaviour between Hastelloy XR alloy and boronated graphite as well as to determine a reaction barrier performance of refractory metal foils Nb, Mo, W and Re. Irradiation conditions were as follows. Neutron dose : 4.05 x 10 22 m -2 (E 18 m -2 (E > 0.16 pJ, 1 Mev). Helium coolant : Average temperature 855 0 C, Pressure 2.94 MPa, Total impurity concentration 400 kBq/m 3 . Irradiation time : 5.0 Ms (1390 hours). Post-irradiation examinations i.e. visual inspection, dimensional inspection, weight measurement, metallography, hardness test, morphological observations by SEM and analysis of element distributions by EPMA were carried out. In the result, reaction products of Hastelloy XR alloy were observed in the ellipsoidal form locally. These results were same as those of the out-of-pile tests. Obvious irradiation effects were not detectable but a little accelarated increase in reaction depth of Hastelloy XR alloy by heat effect of specimens was observed. The refractory metal foils had a good performance of reaction barrier between Hastelloy XR alloy and boronated graphite. Furthermore, movement of Ni, Fe and Cr in the reaction area of Hastelloy XR alloy, difference in the reaction depth of B and C, irradiation effects on diffusion coefficient, lithium production and heat effect are discussed. (author)

  17. Design and qualification testing of a strontium-90 fluoride heat source

    International Nuclear Information System (INIS)

    Fullam, H.T.

    1981-12-01

    The Strontium Heat Source Development Program began at the Pacific Northwest Laboratory (PNL) in 1972 and is scheduled to be completed by the end of FY-1981. The program is currently funded by the US Department of Energy (DOE) By-Product Utilization Program. The primary objective of the program has been to develop the data and technology required to permit the licensing of power systems for terrestrial applications that utilize 90 SrF 2 -fueled radioisotope heat sources. A secondary objective of the program has been to design and qualification-test a general purpose 90 SrF 2 -fueled heat source. The effort expended in the design and testing of the heat source is described. Detailed information is included on: heat source design, licensing requirements, and qualification test requirements; the qualification test procedures; and the fabrication and testing of capsules of various materials. The results obtained in the qualification tests show that the outer capsule design proposed for the 90 SrF 2 heat source is capable of meeting current licensing requirements when Hastelloy S is used as the outer capsule material. The data also indicate that an outer capsule of Hastelloy C-4 would probably also meet licensing requirements, although Hastelloy S is the preferred material. Therefore, based on the results of this study, the general purpose 90 SrF 2 heat source will consist of a standard WESF Hastelloy C-276 inner capsule filled with 90 SrF 2 and a Hastelloy S outer capsule having a 2.375-in. inner diameter and 0.500-in. wall thickness. The end closures for this study, the general purpose 90 SrF 2 heat a Hastelloy S outer capsule having a 2.375-in. inner diameter and 0.500-in. wall thickness. The end closures for the outer capsule will utilize an interlocking joint design requiring a 0.1-in. penetration closure weld

  18. Outgassing rates before, during and after bake-out for various vacuum and first wall candidate materials of a large tokamak device

    International Nuclear Information System (INIS)

    Yoshikawa, H.; Gomay, J.; Sugiyama, Y.; Mizuno, M.; Komiya, S.; Tazima, T.

    1977-01-01

    Outgassing rates of vacuum wall candidate materials; stainless steel SS-304L and YUS-170, Inconel-625 and Hastelloy-X, and first wall materials; molybdenum, pyrolytic graphite and silicon carbide are measured before, during and after a bake-out at 500 0 C. The outgassing rate from the inside wall of the cylinder made of each material is estimated from the pressure difference between before and after a calibrated orifice. The ultimate outgassing rates of SS-304L and pyrolytic graphite, and YUS-170 Inconel-625, Hastelloy-X and molybdenum are the orders of 10 -10 and 10 -11 Pa.l.s -1 cm -2 , respectively

  19. Swelling in neutron irradiated nickel-base alloys

    International Nuclear Information System (INIS)

    Brager, H.R.; Bell, W.L.

    1972-01-01

    Inconel 625, Incoloy 800 and Hastelloy X were neutron irradiated at 500 to 700 0 C. It was found that of the three alloys investigated, Inconel 625 offers the greatest swelling resistance. The superior swelling resistance of Inconel 625 relative to that of Hastelloy-X is probably related to differences in the concentrations of the minor rather than major alloy constituents, and can involve (a) enhanced recombination of defects in the Inconel 625 and (b) preferential attraction of vacancies to incoherent precipitates. (U.S.)

  20. Compatibility studies of potential molten-salt breeder reactor materials in molten fluoride salts

    International Nuclear Information System (INIS)

    Keiser, J.R.

    1977-05-01

    The molten fluoride salt compatibility studies carried out during the period 1974--76 in support of the Molten-Salt Reactor Program are summarized. Thermal-convection and forced-circulation loops were used to measure the corrosion rate of selected alloys. Results confirmed the relationship of time, initial chromium concentration, and mass loss developed by previous workers. The corrosion rates of Hastelloy N and Hastelloy N modified by the addition of 1--3 wt percent Nb were well within the acceptable range for use in an MSBR. 13 figures, 3 tables

  1. Creep properties of heat-resistant superalloys for nuclear plants in helium

    International Nuclear Information System (INIS)

    Shimizu, Shigeki; Satoh, Keisuke; Matsuda, Shozo; Murase, Hirokazu; Fujioka, Junzo.

    1979-01-01

    In order to estimate the creep and rupture strengths of candidate alloys for the intermediate heat exchanger of VHTR, creep and stress rupture tests in impure helium were conducted on Hastelloy X, Inconel 617, Inconel 625, Incoloy 800 and Incoloy 807 at 900 0 C. The results were discussed in comparison with those in air and the alloys were examined from the point of view of the elevated temperature structural design. The main results obtained are summarized as follows: (1) No appreciable decrease in creep and rupture strengths in helium as compared with those in air is observed on Hastelloy X and Inconel 625. On the contrary, the creep and rupture strengths of Inconel 617 in helium decrease slightly as compared with those in air. In the case of Incoloy 807, the creep strength to cause 1 percent total strain and that to initiate secondary creep increase remarkably in helium as compared with those in air. However, the creep strength to cause initiation of tertiary creep and the rupture strength in helium remarkably decrease as compared with those in air. (2) The order of magnitude of the S 0 value for each material in helium is as follows; Hastelloy X > Inconel 617 > Incoloy 807 > Inconel 625 > Incoloy 800 Meanwhile, that of the S sub(t) value in helium is; Inconel 617 > Hastelloy X > Incoloy 807 > Inconel 625 > Incoloy 800. (author)

  2. Molten salt reactors: chemistry

    International Nuclear Information System (INIS)

    1983-01-01

    This work is a critical analysis of the 1000 MW MSBR project. Behavior of rare gases in the primary coolant circuit, their extraction from helium. Coating of graphite by molybdenum, chemistry of protactinium and niobium produced in the molten salt, continuous reprocessing of the fuel salt and use of stainless steel instead of hastelloy are reviewed [fr

  3. Evaluation of candidate alloys for the construction of metal flex hoses in the STS launch environment

    Science.gov (United States)

    Ontiveros, Cordelia

    1988-01-01

    Various vacuum jacketed cryogenic supply lines at the Shuttle launch site use convoluted flexible expansion joints. The atmosphere at the launch site has a very high salt content, and during a launch, fuel combustion products include hydrochloric acid. This extremely corrosive environment has caused pitting corrosion failure in the flex hoses, which were made of 304L stainless steel. A search was done to find a more corrosion resistant replacement material. This study focused on 19 metal alloys. Tests which were performed include electrochemical corrosion testing, accelerated corrosion testing in a salt fog chamber, long term exposure at the beach corrosion testing site, and pitting corrosion tests in ferric chloride solution. Based on the results of these tests, the most corrosion resistant alloys were found to be (in order) Hastelloy C-22, Inconel 625, Hastelloy C-276, Hastelloy C-4, and Inco Alloy G-3. Of these top five alloys, the Hastelloy C-22 stands out as being the best of those tested for this application.

  4. Heat-source specification 500 watt(e) RTG

    International Nuclear Information System (INIS)

    1983-02-01

    This specification establishes the requirements for a 90 SrF 2 heat source and its fuel capsule for application in a 500 W(e) thermoelectric generator. The specification covers: fuel composition and quantity; the Hastelloy S fuel capsule material and fabrication; and the quality assurance requirements for the assembled heat source

  5. 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.)

  6. Noncontaminating technique for making holes in existing process systems

    Science.gov (United States)

    Hecker, T. P.; Czapor, H. P.; Giordano, S. M.

    1972-01-01

    Technique is developed for making cleanly-contoured holes in assembled process systems without introducing chips or other contaminants into system. Technique uses portable equipment and does not require dismantling of system. Method was tested on Inconel, stainless steel, ASTMA-53, and Hastelloy X in all positions.

  7. Mechanical characterization of superalloys for space reactors

    International Nuclear Information System (INIS)

    Duchesne, J.

    1989-01-01

    The purpose of this work is the choice of materials usable between 600 and 900 0 C for nuclear space reactor structures. The main criterion of selection for these materials is their good creep behaviour. Consequently, macroscopic theories of creep and several extrapolation methods were described. Superalloys seem the best materials for the studied range of temperatures. Five of them, base nickel, ones unusual in nuclear industry were selected for their good mechanical properties. Three of them are industrial alloys: the first, HAYNES 230 is a recent one, HASTELLOY S and X are more standard materials. The last two, HASTELLOY XR and PYRAD 38 D are issued from special fabrications. Creep tests metallographic investigations, hardness and tensile tests were performed. A contraction of samples was observed during some creep tests under a low stress, 20MPa at 800 0 C, for HAYNES 230 and HASTELLOY X. This could be due to a structural evolution of these materials connected to a decrease of the cristalline parameter. In addition, correlations were observed between certain characteristics determined from slow tensile tests and short duration creep tests. These correlations present a large interest because, at the present time, creep tests cannot be executed on irradiated materials in our laboratories. Consequently creep behaviour of irradiated materials seem may be deduced. Further studies are needed to explain and confirm the behaviour of the most interesting materials under low stresses: HAYNES 230 and HASTELLOY XR to anticipate their behaviour in working conditions [fr

  8. Creep properties of heat-resistant superalloys for nuclear plants in helium

    International Nuclear Information System (INIS)

    Shimizu, Shigeki; Satoh, Keisuke; Honda, Yoshio; Matsuda, Shozo; Murase, Hirokazu

    1979-01-01

    Creep properties of candidate superalloys for VHTR components in a helium environment at both temperatures of 800 0 C and 900 0 C were compared with those of the same alloys in the atmospheric condition, and the superalloys were contrasted with each other from the viewpoint of high temperature structural design. At 800 0 C, no significant effect of a helium environment on creep properties of the superalloys is observed. At 900 0 C, however, creep strength of Inconel 617, Incoloy 800 and Incoloy 807 in the helium environment decrease more than in the atmospheric environment. In Hastelloy X and Inconel 625, there is no significant difference between creep strengths in helium and those in the atmospheric condition. Concerning So and St values in helium at 900 0 C, Inconel 617 and Hastelloy X are clearly superior to other superalloys. (author)

  9. Fabrication of three 2500-watt (thermal) strontium-90 heat sources

    International Nuclear Information System (INIS)

    DeVore, J.R.; Haff, K.W.; Tompkins, J.A.

    1986-08-01

    Three 2500-watt (thermal) heat sources were fabricated by the Oak Ridge National Laboratory (ORNL) for the purpose of fueling a 500-watt (electric) thermoelectric generator as part of the US Department of Energy's Byproducts Utilization Program (BUP). Each of the sources, which are the largest ever assembled, consist of hot-pressed pellets of 90 Sr fluoride, doubly encapsulated in three Haynes-25 inner capsules and in a Hastelloy-S outer capsule. The total 90 Sr inventory of all three sources is 1.12 million curies. The sources were fabricated at the ORNL Fission Product Development Laboratory (FPDL), which is a facility that is capable of processing multi-megacurie quantities of radioactive materials, chiefly 137 Cs and 90 Sr. The source was tested to determine compliance with all of the IAEA Safety Series No. 33 requirements. The source fabrication, assembly, and testing are described in the presentation

  10. Examination of several pre-oxidation procedures and their effect as hydrogen permeation-barrier

    International Nuclear Information System (INIS)

    Heimes, E.

    1986-03-01

    Several pre-oxidation procedures have been tested with respect to their effect as a hydrogen permeation barrier at the high temperature alloys Hastelloy X and Inconel 617. By outside coating of Hastelloy X samples with alumina the determined impeding effects were very low. A surface aluminium enrichment by different procedures were accomplished before selective oxidation. The method of Aluminium-Hot-Dipping generated oxide layers with a four- to fivefold higher impeding effect compared to specimens fabricated by a standard procedure. With the aid of a metallographical follow-up examination it was shown that the higher impeding effects are due to an improved adhesion between the oxide layer and the high temperature material, whereby in the cooling period after manufacturing a smaller amount of oxide cracking is obtainable. (orig./PW) [de

  11. 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.)

  12. Degradation modes of nickel-base alternate waste package overpack materials

    International Nuclear Information System (INIS)

    Pitman, S.G.

    1988-07-01

    The suitability of Ti Grade 12 for waste package overpacks has been questioned because of its observed susceptibility to crevice corrosion and hydrogen-assisted crack growth. For this reason, materials have been selected for evaluation as alternatives to Ti Grade 12 for use as waste package overpacks. These alternative materials, which are based on the nickel-chromium-molybdenum (Ni-Cr-Mo) alloy system, are Inconel 625, Hastelloy C-276, and Hastelloy C-22. The degradation modes of the Ni-base alternate materials have been examined at Pacific Northwest Laboratory to determine the suitability of these materials for waste package overpack applications in a salt repository. Degradation modes investigated included general corrosion, crevice corrosion, pitting, stress-corrosion cracking, and hydrogen embrittlement

  13. Materials performance in off-gas systems containing iodine

    International Nuclear Information System (INIS)

    Beavers, J.A.; Berry, W.E.; Griess, J.C.

    1981-11-01

    During the reprocessing of spent reactor fuel elements, iodine is released to gas streams from which it is ultimately removed by conversion to nonvolatile iodic acid. Under some conditions iodine can produce severe corrosion in off-gas lines; in this study these conditions were established. Iron- and nickel-based alloys containing more than 6% molybdenum, such as Hastelloy G (7%), Inconel 625 (9%), and Hastelloy C-276 (16%), as well as titanium and zirconium, remained free of attack under all conditions tested. When the other materials, notably the austenitic stainless steels, were exposed to gas streams containing even only low concentrations of iodine and water vapors at 25 and 40 0 C, a highly corrosive, brownish-green liquid formed on their surfaces. In the complete absence of water vapor, the iodine-containing liquid did not form and all materials remained unaffected. The liquid that formed had a low pH (usually 2 inhibited attack

  14. Neutronics analysis for HYLIFE-II

    International Nuclear Information System (INIS)

    Tobin, M.T.

    1990-01-01

    A preliminary neutronics analysis of the HYLIFE-2 reactor concept gives a tritium breeding ratio of 1.17 and a system energy multiplication factor of 1.14. Modified SS-316 (in which Mn is substituted for Ni) is superior to Hastelloy X and Hastelloy N as a firstwall material considering He generation, dpa-limited lifetime, and shallow-burial index. Since Flibe is corrosive to Mn metals, however, a favorable first-wall material is yet to be decided on. Flibe impurities considered (e.g., inherent impurities and those arising from wall erosion or secondary-coolant leakage) do not increase the hazard to the public over that of pure Flibe. The main issues for HYLIFE-2 are the high shallow-burial index (106) and the requirement to contain some 99.7% of the 18 F inventory to prevent its release to the public 18 refs., 3 figs., 9 tabs

  15. Abrasive wear of WC-NiMoCrFeCo thermally sprayed coatings in dependence on different types of abrasive sands

    Czech Academy of Sciences Publication Activity Database

    Kašparová, M.; Zahálka, F.; Houdková, Š.; Ctibor, Pavel

    2010-01-01

    Roč. 48, č. 1 (2010), s. 75-85 ISSN 0023-432X R&D Projects: GA AV ČR 1QS200430560 Institutional research plan: CEZ:AV0Z20430508 Keywords : WC-Hastelloy * abrasive wear * Al2O3 sand * SiO2 sand * braun size * abrasive efficiency Subject RIV: JG - Metallurgy Impact factor: 0.471, year: 2010 http://kovmat.sav.sk/abstract.php?rr=48&cc=1&ss=73

  16. Two different mechanisms of fatigue damage due to cyclic stress loading at 77 K for MOCVD-YBCO-coated conductors

    International Nuclear Information System (INIS)

    Sugano, M; Yoshida, Y; Hojo, M; Shikimachi, K; Hirano, N; Nagaya, S

    2008-01-01

    Tensile fatigue tests were carried out at 77 K for YBCO-coated conductors fabricated by metal-organic chemical vapor deposition (MOCVD). The S-N relationship, variation of critical current (I c ) during cyclic loading and microscopic fatigue damage were investigated. Fatigue strength at 10 6 cycles was evaluated to be σ max = 1300 MPa and 890 MPa under the stress ratios of 0.5 and 0.1. Two different mechanisms of fatigue damage, depending on the number of stress cycles to failure, were observed. In one of the fracture mechanisms, fatigue behavior is characterized by overall fracture which occurs at 10 4 -10 5 cycles. For these specimens, I c after unloading does not degrade before overall fracture. Although only shallow slip bands were found at the Ag surface, fatigue cracks were found on the Hastelloy C-276 surface of the fractured specimen. These results suggest that overall fracture due to cyclic stress was caused by fatigue of the Hastelloy substrate. In the other fracture mechanism, even though overall fracture did not occur at 10 6 cycles, a slight decrease of I c was detected after 10 5 cycles. No fatigue crack was found on the Hastelloy surface, while deep slip bands corresponding to the initial stage of fatigue crack were observed on the Ag surface. From these results, we concluded that I c degradation at a high cycle number is attributed to the fatigue of the Ag stabilizing layer

  17. Effect of a helium environment on the mechanical properties of HTGR primary system metals

    International Nuclear Information System (INIS)

    Chow, J.G.Y.; Soo, P.; Sabatini, R.L.

    1978-01-01

    Creep and high cycle fatigue tests have been carried out on Incoloy 800H and Hastelloy X in a helium environment containing 40 μ atm of H 2 O, 200 μ atm H 2 , 40 μ atm CO, 20 μ atm CH 4 and 10 μ atm CO 2 . The creep behavior of Incoloy 800H does not appear to show significant differences from that measured in air. However, the Hastelloy X at the maximum test temperature studied (871 0 C, 1600 0 F) shows behavior which is inferior. With respect to high cycle fatigue, the Incoloy 800H is weaker in the helium environment at a test temperature of 649 0 C (1200 0 F). At 760 0 C (1400 0 F) the strength in helium is higher but there is a tendency to lose strength more rapidly than for the air tests as the test time increases. Hastelloy X tested at 871 0 C (1600 0 F) also shows higher strength in helium for short test times but for extended tests the strengths in air and helium become similar. Scanning electron microprobe analyses have been carried out to correlate the strength measurements with surface oxidation characteristics and internal structural changes

  18. Catalytic effect of different reactor materials under subcritical water conditions: decarboxylation of cysteic acid into taurine

    Science.gov (United States)

    Faisal, M.

    2018-03-01

    In order to understand the influence of reactor materials on the catalytic effect for a particular reaction, the decomposition of cysteic acid from Ni/Fe-based alloy reactors under subcritical water conditions was examined. Experiments were carried out in three batch reactors made of Inconel 625, Hastelloy C-22 and SUS 316 over temperatures of 200 to 300 °C. The highest amount of eluted metals was found for SUS 316. The results demonstrated that reactor materials contribute to the resulting product. Under the tested conditions, cysteic acid decomposes readily with SUS 316. However, the Ni-based materials (Inconel 625 and Hastelloy C-22) show better resistance to metal elution. It was found that among the materials used in this work, SUS 316 gave the highest reaction rate constant of 0.1934 s‑1. The same results were obtained at temperatures of 260 and 300 °C. Investigation of the Arrhenius activation energy revealed that the highest activation energy was for Hastelloy C-22 (109 kJ/mol), followed by Inconel 625 (90 kJ/mol) and SUS 316 (70 kJ/mol). The decomposition rate of cysteic acid was found to follow the results for the trend of the eluted metals. Therefore, it can be concluded that the decomposition of cysteic acid was catalyzed by the elution of heavy metals from the surface of the reactor. The highest amount of taurine from the decarboxylation of cysteic acid was obtained from SUS 316.

  19. Properties of super alloys for high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Izaki, Takashi; Nakai, Yasuo; Shimizu, Shigeki; Murakami, Takashi

    1975-01-01

    The existing data on the properties at high temperature in helium gas of iron base super alloys. Incoloy-800, -802 and -807, nickel base super alloys, Hastelloy-X, Inconel-600, -617 and -625, and a casting alloy HK-40 were collectively evaluated from the viewpoint of the selection of material for HTGRs. These properties include corrosion resistance, strength and toughness, weldability, tube making, formability, radioactivation, etc. Creep strength was specially studied, taking into consideration the data on the creep characteristics in the actual helium gas atmosphere. The necessity of further long run creep data is suggested. Hastelloy-X has completely stable corrosion resistance at high temperature in helium gas. Incoloy 800 and 807 and Inconel 617 are not preferable in view of corrosion resistance. The creep strength of Inconel 617 extraporated to 1,000 deg C for 100,000 hours in air was the greatest rupture strength of 0.6 kg/mm 2 in all above alloys. However, its strength in helium gas began to fall during a relatively short time, so that its creep strength must be re-evaluated in the use for long time. The radioactivation and separation of oxide film in primary construction materials came into question, Inconel 617 and Incoloy 807 showed high induced radioactivity intensity. Generally speaking, in case of nickel base alloys such as Hastelloy-X, oxide film is difficult to break away. (Iwakiri, K.)

  20. Some observations on the mechanism of corrosion to be encountered in nuclear waste repositories located in tuffaceous rock

    International Nuclear Information System (INIS)

    Wilde, M.H.; Wilde, B.E.

    1993-01-01

    Potentiostatic anodic polarization studies have been conducted in a J-13 simulated nuclear waste repository environment, which was allowed to evaporate to dryness followed by rehydration prior to polarization. The behavior of Type 316L stainless steel, AISI 1020 carbon steel, Hastelloy C22 and platinum was compared with that noted previously for a non-baked simulate. The anodic dissolution characteristics of Type 316L stainless steel in environments containing 1000X Cl - J-13 depend markedly on whether the solution is merely a mixture of virgin chemicals or a mixture that has been evaporated to dryness, baked and rehydrated to the same volume. In the non-evaporated environment Type 316L stainless steel pitted severely, and in the evaporated/rehydrated environment a non-corroding type of behavior was observed along with the precipitation of a dense scale. Similar behavior was observed for Hastelloy C22. The polarization curves for carbon steel and platinum were the same as those noted for 316L and Hastelloy C22, when conducted in the evaporated/rehydrated environment. X-ray diffraction studies indicated that the scale produced in all tests conducted on evaporated/rehydrated solutions was calcium carbonate. Based on the qualitatively similar polarization characteristics of materials having such widely differing corrosion properties, it is concluded that the major factor controlling the anodic charge transfer reaction under these conditions is the formation of a calcium carbonate scale. (Author)

  1. Corrosion behaviour of container materials for the disposal of high-level wastes in rock salt formations

    International Nuclear Information System (INIS)

    Smailos, E.; Schwarzkopf, W.; Koester, R.

    1986-01-01

    In 1983-84 extensive laboratory-scale experiments (immersion tests) to evaluate the long-term corrosion behaviour of selected materials in salt brines and first in situ experiments were performed. In the laboratory experiments the materials Ti 99.8-Pd, Hastelloy C4 and hot-rolled low carbon steel (reference materials in the joint European corrosion programme) as well as cast steel, spheoroidal cast iron, Si-cast iron and the Ni-Resists type D2 and D4 were investigated. The investigated parameters were: temperature (90 0 C; 170 0 C, 200 0 C), gamma-radiation (10 5 rad/h) and different compositions of salt brines. The results obtained show that, in addition to Ti 99.8-Pd, also Hastelloy C4 and unalloyed steels are in principle suitable for being used for long-term stable HLW-containers if the gamma dose rate is reduced by suitable shielding. Furthermore, the susceptibility of Hastelloy C4 to crevice corrosion must be taken into account. Further studies will be necessary to provide final evidence of the suitability of the materials examined. These will mainly involve clarification of questions related to hydrogen embrittlement (Ti 99.8-Pd, unalloyed steels) and to the influence of pressure and saline impurities (e.g. antiJ, antiBr) on corrosion

  2. Advanced heat exchanger development for molten salts

    Energy Technology Data Exchange (ETDEWEB)

    Sabharwall, Piyush, E-mail: Piyush.Sabharwall@inl.gov [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Clark, Denis; Glazoff, Michael [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Zheng, Guiqiu; Sridharan, Kumar; Anderson, Mark [University of Wisconsin, Madison (United States)

    2014-12-15

    Highlights: • Hastelloy N and 242, shows corrosion resistance to molten salt at nominal operating temperatures. • Both diffusion welds and sheet material in Hastelloy N were corrosion tested in at 650, 700, and 850 °C for 200, 500, and 1000 h. • Thermal gradients and galvanic couples in the molten salts enhance corrosion rates. • Corrosion rates found were typically <10 mils per year. - Abstract: This study addresses present work concerned with advanced heat exchanger development for molten salt in nuclear and non-nuclear thermal systems. The molten salt systems discussed herein use alloys, such as Hastelloy N and 242, that show good corrosion resistance in molten salt at nominal operating temperatures up to 700 °C. These alloys were diffusion welded, and the corresponding information is presented. Test specimens were prepared for exposing diffusion welds to molten salt environments. Hastelloy N and 242 were found to be weldable by diffusion welding, with ultimate tensile strengths about 90% of base metal values. Both diffusion welds and sheet material in Hastelloy N were corrosion tested in 58 mol% KF and 42 mol% ZrF{sub 4} at 650, 700, and 850 °C for 200, 500, and 1000 h. Corrosion rates were similar between welded and nonwelded materials, typically <100 μm per year after 1000 h of corrosion tests. No catastrophic corrosion was observed in the diffusion welded regions. For materials of construction, nickel-based alloys and alloys with dense nickel coatings are effectively inert to corrosion in fluorides, but not so in chlorides. Hence, additional testing of selected alloys for resistance to intergranular corrosion is needed, as is a determination of corrosion rate as a function of the type of salt impurity and alloy composition, with respect to chromium and carbon, to better define the best conditions for corrosion resistance. Also presented is the division of the nuclear reactor and high-temperature components per American Society of Mechanical

  3. Influence of heating rate on corrosion behavior of Ni-base heat resistant alloys in simulated VHTR helium environment

    International Nuclear Information System (INIS)

    Kurata, Yuji; Kondo, Tatsuo

    1985-04-01

    The influence of heating rate on corrosion and carbon transfer was studied for Ni-base heat resistant alloys exposed to simulated VHTR(very high temperature reactor) coolant environment. Special attention was focused to relationship between oxidation and carburization at early stage of exposure. Tests were conducted on two heats of Hastelloy XR with different boron(B) content and the developmental alloys, 113MA and KSN. Two kinds of heating rates, i.e. 80 0 C/min and 2 0 C/min, were employed. Corrosion tests were carried out at 900 0 C up to 500 h in JAERI Type B helium, one of the simulated VHTR primary coolant specifications. Under higher heating rate, oxidation resistance of both heats of Hastelloy XR(2.8 ppmB and 40 ppmB) were equivalent and among the best, then KSN and 113MA followed in the order. Under lower heating rate only alloy, i.e. Hastelloy XR with 2.8 ppmB, showed some deteriorated oxidation resistance while all others being unaffected by the heating rate. On the other hand the carbon transfer behavior showed strong dependence on the heating rate. In case of higher heating rate, significant carburization occured at early stage of exposure and thereafter the progress of carburization was slow in all the alloys. On the other hand only slow carburization was the case throughout the exposure in case of lower heating rate. The carburization in VHTR helium environment was interpreted as to be affected by oxide film formation in the early stage of exposure. The carbon pick-up was largest in Hastelloy XR with 40 ppmB and it was followed by Hastelloy XR with 2.8 ppmB. 113MA and KSN were carburized only slightly. The observed difference of carbon pick-up among the alloys tested was interpreted to be attributed mainly to the difference of the carbon activity, the carbide precipitation characteristics among the alloys tested. (author)

  4. Summary of Dissimilar Metal Joining Trials Conducted by Edison Welding Institute

    Energy Technology Data Exchange (ETDEWEB)

    MJ Lambert

    2005-11-18

    Under the direction of the NASA-Glenn Research Center, the Edison Welding Institute (EWI) in Columbus, OH performed a series of non-fusion joining experiments to determine the feasibility of joining refractory metals or refractory metal alloys to Ni-based superalloys. Results, as reported by EWI, can be found in the project report for EWI Project 48819GTH (Attachment A, at the end of this document), dated October 10, 2005. The three joining methods used in this investigation were inertia welding, magnetic pulse welding, and electro-spark deposition joining. Five materials were used in these experiments: Mo-47Re, T-111, Hastelloy X, Mar M-247 (coarse-grained, 0.5 mm to several millimeter average grain size), and Mar M-247 (fine-grained, approximately 50 {micro}m average grain size). Several iterative trials of each material combination with each joining method were performed to determine the best practice joining method. Mo-47Re was found to be joined easily to Hastelloy X via inertia welding, but inertia welding of the Mo-alloy to both Mar M-247 alloys resulted in inconsistent joint strength and large reaction layers between the two metals. T-111 was found to join well to Hastelloy X and coarse-grained Mar M-247 via inertia welding, but joining to fine-grained Mar M-247 resulted in low joint strength. Magnetic pulse welding (MPW) was only successful in joining T-111 tubing to Hastelloy X bar stock. The joint integrity and reaction layer between the metals were found to be acceptable. This single joining trial, however, caused damage to the electromagnetic concentrators used in this process. Subsequent design efforts to eliminate the problem resulted in a loss of power imparted to the accelerating work piece, and results could not be reproduced. Welding trials of Mar M-247 to T-111 resulted in catastrophic failure of the bar stock, even at lower power. Electro-spark deposition joining of Mo-47Re, in which the deposited material was Hastelloy X, did not have a

  5. Summary of Dissimilar Metal Joining Trials Conducted by Edison Welding Institute

    International Nuclear Information System (INIS)

    MJ Lambert

    2005-01-01

    Under the direction of the NASA-Glenn Research Center, the Edison Welding Institute (EWI) in Columbus, OH performed a series of non-fusion joining experiments to determine the feasibility of joining refractory metals or refractory metal alloys to Ni-based superalloys. Results, as reported by EWI, can be found in the project report for EWI Project 48819GTH (Attachment A, at the end of this document), dated October 10, 2005. The three joining methods used in this investigation were inertia welding, magnetic pulse welding, and electro-spark deposition joining. Five materials were used in these experiments: Mo-47Re, T-111, Hastelloy X, Mar M-247 (coarse-grained, 0.5 mm to several millimeter average grain size), and Mar M-247 (fine-grained, approximately 50 (micro)m average grain size). Several iterative trials of each material combination with each joining method were performed to determine the best practice joining method. Mo-47Re was found to be joined easily to Hastelloy X via inertia welding, but inertia welding of the Mo-alloy to both Mar M-247 alloys resulted in inconsistent joint strength and large reaction layers between the two metals. T-111 was found to join well to Hastelloy X and coarse-grained Mar M-247 via inertia welding, but joining to fine-grained Mar M-247 resulted in low joint strength. Magnetic pulse welding (MPW) was only successful in joining T-111 tubing to Hastelloy X bar stock. The joint integrity and reaction layer between the metals were found to be acceptable. This single joining trial, however, caused damage to the electromagnetic concentrators used in this process. Subsequent design efforts to eliminate the problem resulted in a loss of power imparted to the accelerating work piece, and results could not be reproduced. Welding trials of Mar M-247 to T-111 resulted in catastrophic failure of the bar stock, even at lower power. Electro-spark deposition joining of Mo-47Re, in which the deposited material was Hastelloy X, did not have a

  6. Corrosion behaviour of selected high-level waste packaging materials under gamma irradiation and in-situ disposal conditions in rock salt

    International Nuclear Information System (INIS)

    Smailos, E.; Schwarzkopf, W.; Koester, R.

    1988-07-01

    Corrosion studies performed until now on a number of materials have shown that unalloyed steels, Hastelloy C4 and Ti 99.8-Pd are the most promising materials for a long-term resistant packaging to be used in high-level waste (HLW) canister disposal in rock salt formations. To characterize their corrosion behaviour in more detail, additional studies have been performed. The influence has been examined which is exerted by the gamma dose rate (1 Gy/h to 100 Gy/h) on the corrosion of three preselected steels and Hastelloy C4 at 90 0 C in a salt brine (Q-brine) rich in MgCl 2 , i.e., conditions relevant to accident scenarios in a repository. In addition, in-situ corrosion experiments have been carried out in the Asse salt mine at elevated temperatures (120 0 C to 210 0 C) in the absence and in the presence of a gamma radiation field of 3 x 10 2 Gy/h, within the framework of the German/US Brine Migration Test. Under the test conditions the gamma radiation did not exert a significant influence on the corrosion of the steels investigated, whereas Hastelloy C4, exposed to dose rates of 10 Gy/h and 100 Gy/h, underwent pitting and crevice corrosion (20 μm/a at the maximum).The low amounts of migrated salt brine (140 ml after 900 days) in the in-situ- experiment did not produce noticeable corrosion of the materials. (orig./RB) [de

  7. In-plane aligned YBCO film on textured YSZ buffer layer deposited on NiCr alloy tape by laser ablation with only O+ ion beam assistance

    International Nuclear Information System (INIS)

    Xin Tang Huang

    2000-01-01

    High critical current density and in-plane aligned YBa 2 Cu 3 O 7-x (YBCO) film on a textured yttria-stabilized zirconia (YSZ) buffer layer deposited on NiCr alloy (Hastelloy c-275) tape by laser ablation with only O + ion beam assistance was fabricated. The values of the x-ray phi-scan full width at half-maximum (FWHM) for YSZ(202) and YBCO(103) are 18 deg. and 11 deg., respectively. The critical current density of YBCO film is 7.9 x 105 A cm -2 at liquid nitrogen temperature and zero field, and its critical temperature is 90 K. (author)

  8. Resonance bending fatigue testing with simultaneous damping measurement and its application on layered coatings

    Czech Academy of Sciences Publication Activity Database

    Kovářík, O.; Haušild, P.; Čapek, J.; Medřický, Jan; Siegl, J.; Mušálek, Radek; Pala, Zdeněk; Curry, N.; Bjorklund, S.

    2016-01-01

    Roč. 82, January (2016), s. 300-309 ISSN 0142-1123. [International Conference on Fatigue Damage of Structural Materials Conference/10./. Massachusetts, 21.09.2014-26.09.2014] R&D Projects: GA ČR GB14-36566G Institutional support: RVO:61389021 Keywords : Crack detection * Damping * Fatigue * Hastelloy-X * Nondestructive test ing Subject RIV: JK - Corrosion ; Surface Treatment of Materials Impact factor: 2.899, year: 2016 http://www.sciencedirect.com/science/article/pii/S0142112315002443

  9. Effect of helium on creep and fatigue (MAT 11)

    International Nuclear Information System (INIS)

    Schroeder, H.

    1991-03-01

    This final report contains experimental results on mechanical properties (creep, fatigue, tensile) and microstructural investigations (SEM, TEM) of pre-implanted samples of steels or alloys. (AISI 316, AISI 316L, DIN 1.4970, JPCA 8206, DIN 1.4914; Incoloy 800H, Hastelloy X, DIN 1.4981, (Fe 0.49 Ni 0.51 ) 3 V, Fe17Ni17Cr, Fe15Ni15Cr, Nimonic PE 16, Ni8Si). Furthermore theoretical aspects and developed models and mechanisms for helium embrittlement are described. This report is presented in the form of an extended summary without figures. (MM)

  10. ECLSS Sustaining Metal Materials Compatibility Final Report, Electrochemical and Crevice Corrosion Test Results

    Science.gov (United States)

    Lee, R. E.

    2015-01-01

    Electrochemical test results are presented for six noble metals evaluated in two acidic test solutions which are representative of waste liquids processed in the Environmental Control and Life Support System (ECLSS) aboard the International Space Station (ISS). The two test solutions consisted of fresh waste liquid which had been modified with a proposed or alternate pretreatment formulation and its associated brine concentrate. The six test metals included three titanium grades, (Commercially Pure, 6Al-4V alloy and 6Al-4V Low Interstitial alloy), two nickel-chromium alloys (Inconel® 625 and Hastelloy® C276), and one high tier stainless steel (Cronidur® 30).

  11. Mechanical structure and problem of thorium molten salt reactor

    International Nuclear Information System (INIS)

    Kamei, Takashi

    2011-01-01

    After Fukushima Daiichi accident, there became great interest in Thorium Molten Salt Reactor (MSR) for the safety as station blackout leading to auto drainage of molten salts with freeze valve. This article described mechanical structure of MSR and problems of materials and pipes. Material corrosion problem by molten salts would be solved using modified Hastelloy N with Ti and Nb added, which should be confirmed by operation of an experimental reactor. Trends in international activities of MSR were also referred including China declaring MSR development in January 2011 to solve thorium contamination issues at rare earth production and India rich in thorium resources. (T. Tanaka)

  12. Metallic materials corrosion problems in molten salt reactors

    International Nuclear Information System (INIS)

    Chauvin, G.; Dixmier, J.; Jarny, P.

    1977-01-01

    The USA forecastings concerning the molten salt reactors are reviewed (mixtures of fluorides containing the fuel, operating between 560 and 700 0 C). Corrosion problems are important in these reactors. The effects of certain characteristic factors on corrosion are analyzed: humidity and metallic impurities in the salts, temperature gradients, speed of circulation of salts, tellurium from fission products, coupling. In the molten fluorides and experimental conditions, the materials with high Ni content are particularly corrosion resistant alloys (hastelloy N). The corrosion of this material is about 2.6 mg.cm -2 at 700 0 C [fr

  13. Corrosion behaviour of container materials for the disposal of high-level waste forms in rock salt formations

    International Nuclear Information System (INIS)

    Smailos, E.; Schwarzkopf, W.; Koester, R.

    1987-05-01

    Extensive laboratory-scale experiments to evaluate the long-term corrosion behaviour of selected materials in brines and first in situ experiments were performed. In the laboratory experiments the materials Ti 99.8-Pd, Hastelloy C4 and hot-rolled low carbon steel as well cast steel, spheroidal cast iron, Si-cast iron and the Ni-Resists type D2 and D4 were investigated. The investigated parameters were: temperature, gamma-radiation and different compositions of salt brines. (orig./PW) [de

  14. Stress Corrosion Evaluation of Various Metallic Materials for the International Space Station Water Recycling System

    Science.gov (United States)

    Torres, P. D.

    2015-01-01

    A stress corrosion evaluation was performed on Inconel 625, Hastelloy C276, titanium commercially pure (TiCP), Ti-6Al-4V, Ti-6Al-4V extra low interstitial, and Cronidur 30 steel as a consequence of a change in formulation of the pretreatment for processing the urine in the International Space Station Environmental Control and Life Support System Urine Processing Assembly from a sulfuric acid-based to a phosphoric acid-based solution. The first five listed were found resistant to stress corrosion in the pretreatment and brine. However, some of the Cronidur 30 specimens experienced reduction in load-carrying ability.

  15. Studies on neutron irradiation effects of iron alloys and nickel-base heat resistant alloys

    International Nuclear Information System (INIS)

    Watanabe, Katsutoshi

    1987-09-01

    The present paper describes the results of neutron irradiation effects on iron alloys and nickel-base heat resistant alloys. As for the iron alloys, irradiation hardening and embrittlement were investigated using internal friction measurement, electron microscopy and tensile testings. The role of alloying elements was also investigated to understand the irradiation behavior of iron alloys. The essential factors affecting irradiation hardening and embrittlement were thus clarified. On the other hand, postirradiation tensile and creep properties were measured of Hastelloy X alloy. Irradiation behavior at elevated temperatures is discussed. (author)

  16. 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

  17. Electrochemical Impedance Spectroscopy Of Metal Alloys

    Science.gov (United States)

    Macdowell, L. G.; Calle, L. M.

    1993-01-01

    Report describes use of electrochemical impedance spectroscopy (EIS) to investigate resistances of 19 alloys to corrosion under conditions similar to those of corrosive, chloride-laden seaside environment of Space Transportation System launch site. Alloys investigated: Hastelloy C-4, C-22, C-276, and B-2; Inconel(R) 600, 625, and 825; Inco(R) G-3; Monel 400; Zirconium 702; Stainless Steel 304L, 304LN, 316L, 317L, and 904L; 20Cb-3; 7Mo+N; ES2205; and Ferralium 255. Results suggest electrochemical impedance spectroscopy used to predict corrosion performances of metal alloys.

  18. Evaluation and Control of Mechanical Degradation of Austenitic Stainless 310S Steel Substrate During Coated Superconductor Processing

    Science.gov (United States)

    Kim, Seung-Gyu; Kim, Najung; Shim, Hyung-Seok; Kwon, Oh Min; Kwon, Dongil

    2018-05-01

    The superconductor industry considers cold-rolled austenitic stainless 310S steel a less expensive substitute for Hastelloy X as a substrate for coated superconductor. However, the mechanical properties of cold-rolled 310S substrate degrade significantly in the superconductor deposition process. To overcome this, we applied hot rolling at 900 °C (or 1000 °C) to the 310S substrate. To check the property changes, a simulated annealing condition equivalent to that used in manufacturing was determined and applied. The effects of the hot rolling on the substrate were evaluated by analyzing its physical properties and texture.

  19. Experimental investigation of the temperature dependence of sound velocity in the structural materials for nuclear power engineering

    International Nuclear Information System (INIS)

    Roshchupkin, V.V.; Pokrasin, M.A.; Chernov, A.I.; Semashko, N.A.; Filonenko, S.F.

    1999-01-01

    The purpose of the study consists in determination of the sound velocity temperature dependence in structural materials for nuclear power engineering. In particular, the Zr-2.5%Nb, Hastelloys-H alloys and X2.5M steel are studied. The facility for studying acoustic parameters of metals and alloys is described. The software makes it possible to obtain the results in various forms with the data stored in the memory for further analysis. The data on the above alloys obtained by use of various methods are presented and analyzed [ru

  20. Galvanic corrosion resistance of welded dissimilar nickel-base alloys

    International Nuclear Information System (INIS)

    Corbett, R.A.; Morrison, W.S.; Snyder, R.J.

    1986-01-01

    A program for evaluating the corrosion resistance of various dissimilar welded nickel-base alloy combinations is outlined. Alloy combinations included ALLCORR, Hastelloy C-276, Inconel 72 and Inconel 690. The GTAW welding process involved both high and minimum heat in-put conditions. Samples were evaluated in the as-welded condition, as well as after having been aged at various condtions of time and temperature. These were judged to be most representative of process upset conditions which might be expected. Corrosion testing evaluated resistance to an oxidizing acid and a severe service environment in which the alloy combinations might be used. Mechanical properties are also discussed

  1. Comparison of Cobalt based Catalysts Supported on MWCNT and SBA-15 Supporters for Fischer-tropsch Synthesis by Using Novel Vortex Type Reactor

    International Nuclear Information System (INIS)

    Yakubov, A.; Shahrun, M.S.; Kutty, M.G.; Hamid, S.B.A.; Piven, V.

    2011-01-01

    10 and 40 wt% Co/ Multi wall Carbon Nano tubes (MWCNT) and 10 and 40 wt% Co/ Santa Barbara Amorphous-15 (SBA) catalysts were prepared via incipient wetness impregnation and characterized by Scanning Electron Microscopy equipped with Energy Dispersive X-ray Spectroscopy (SEM and EDX), N 2 adsorption-desorption (BET), X-ray Diffractometry (XRD), Transmission Electron Microscopy (TEM) and Temperature- Programmed Reduction and H 2 desorption TPD/RO. Co(NO 3 ) 2 * 6H 2 O was used as a cobalt precursor. 200 ml hastelloy autoclave reactor was implemented to see the performance of the catalysts. This report presents details about the catalyst synthesis and reactor study. (author)

  2. Resonance bending fatigue testing with simultaneous damping measurement and its application on layered coatings

    Czech Academy of Sciences Publication Activity Database

    Kovářík, O.; Haušild, P.; Čapek, J.; Medřický, Jan; Siegl, J.; Mušálek, Radek; Pala, Zdeněk; Curry, N.; Bjorklund, S.

    2016-01-01

    Roč. 82, January (2016), s. 300-309 ISSN 0142-1123. [International Conference on Fatigue Damage of Structural Materials Conference/10./. Massachusetts, 21.09.2014-26.09.2014] R&D Projects: GA ČR GB14-36566G Institutional support: RVO:61389021 Keywords : Crack detection * Damping * Fatigue * Hastelloy-X * Nondestructive testing Subject RIV: JK - Corrosion ; Surface Treatment of Materials Impact factor: 2.899, year: 2016 http://www.sciencedirect.com/science/article/pii/S0142112315002443

  3. Testing and analyses of a high temperature duct for gas-cooled reactors

    International Nuclear Information System (INIS)

    Black, W.E.; Roberge, A.; Felten, P.; Bastien, R.

    1979-01-01

    A 0.6 scale model of a steam cycle gas-cooled reactor high temperature duct was tested in a closed loop helium facility. The object of the test series was to determine: 1) the thermal effects of gas permeation within the thermal barrier, 2) the plastic deformation of the metallic components, and 3) the thermal performance of the fibrous insulation. A series of tests was performed with thermal cyclings from 100 0 C to 760 0 C at 50 atmospheres until the system thermal performance had stabilized hence enabling predictions for the reactor life. Additional tests were made to assess permeation by deliberately simulating sealing weld failures thereby allowing gas flow by-pass within the primary thermal barrier. After 100 cycles the entire primary structure was found to have performed without structural failure. Due to high pressures exerted by the insulation on the cover plates and a design oversight, the thin seal sheets were unable to expand in an anticipated manner. Local buckling resulted. Pre and post test metallurgical analyses were conducted on the Hastelloy-X structures and reference specimens. The results gave evidence of aging in the form of noticeable changes in room temperature tensile and reduction in area parameters. The Hastelloy-X welds exhibited greater changes in properties due to thermal aging. The antifriction coating (Cr 3 C 2 ) performed well without spallation or excessive wear. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-01-01

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

  5. Corrosion characteristics of thermal sprayed coating of stainless alloys in chloride solution; Taishoku gokin yosha himaku no enkabutsu yoekichu ni okeru fushoku tokusei

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, T. [Ajinomoto Co. Inc., Tokyo (Japan); Ishikawa, K. [Tokyo Metallikon Co. Ltd., Tokyo (Japan); Kitamura, Y. [Kitamura Technical Consultant Office, Kanagawa (Japan)

    1994-12-15

    With an objective to develop a thermal sprayed coating of environment interruption type that can be sprayed at sites, electrochemical discussions, SEM observation, and EPMA surface analysis were performed on corrosion characteristics in chloride solution of coatings of SUS 304, 316 and Hastelloy C thermally sprayed onto test pieces made of structural steel SS400, as well as the effect of improvement in corrosion resistance by means of a coating reforming treatment. The following conclusions were obtained: the degradation in corrosion resistance of the coatings is attributable to increase in anodic solubility due to appearance of innumerable crevices as a result of deposited particles forming porous structure and due to drop of Cr content in the matrix caused by generation of oxides on the surface of the crevices, by which the corrosion progresses in the form of crevice corrosion; and denseness of the passive coating is lost on the surface of the deposited particles, accelerating the cathodic reaction. A suitable means that could be used practically in chloride solution would be a method to use a material with less crevice susceptibility such as Hastelloy C as a base material, and seal the crevice structure with epoxy resin, etc. 7 refs., 10 figs., 3 tabs.

  6. Creep rupture properties under varying load/temperature conditions on a nickel-base heat-resistant alloy strengthened by boron addition

    International Nuclear Information System (INIS)

    Tsuji, Hirokazu; Tanabe, Tatsuhiko; Nakajima, Hajime

    1994-01-01

    A series of constant load and temperature creep rupture tests and varying load and temperature creep rupture tests was carried out on Hastelloy XR whose boron content level is 60 mass ppm at 900 and 1000 C in order to examine the behavior of the alloy under varying load and temperature conditions. The life fraction rule completely fails in the prediction of the creep rupture life under varying load and temperature conditions though the rule shows good applicability for Hastelloy XR whose boron content level is below 10 mass ppm. The modified life fraction rule has been proposed based on the dependence of the creep rupture strength on the boron content level of the alloy. The modified rule successfully predicts the creep rupture life under the test conditions from 1000 to 900 C. The trend observed in the tests from 900 to 1000 C can be qualitatively explained by the mechanism that the oxide film which is formed during the prior exposure to 900 C plays the role of the protective barrier against the boron dissipation into the environment. (orig.)

  7. Creep rupture properties under varying load/temperature conditions on a nickel-base heat-resistant alloy strengthened by boron addition

    International Nuclear Information System (INIS)

    Tsuji, Hirokazu; Nakajima, Hajime; Tanabe, Tatsuhiko.

    1993-09-01

    A series of constant load and temperature creep rupture tests and varying load and temperature creep rupture tests was carried out on Hastelloy XR whose boron content level is 60 mass ppm at 900 and 1000degC in order to examine the behavior of the alloy under varying load and temperature conditions. The life fraction rule completely fails in the prediction of the creep rupture life under varying load and temperature conditions though the rule shows good applicability for Hastelloy XR whose boron content level is below 10 mass ppm. The modified life fraction rule has been proposed based on the dependence of the creep rupture strength on the born content level of the alloy. The modified rule successfully predicts the creep rupture life under the test conditions from 1000degC to 900degC. The trend observed in the tests from 900degC to 1000degC can be qualitatively explained by the mechanism that the oxide film which is formed during the prior exposure to 900degC plays the role of the protective barrier against the boron dissipation into the environment. (author)

  8. Materials for advanced high temperature reactors

    International Nuclear Information System (INIS)

    Graham, L.W.

    1977-01-01

    Materials are studied in advanced applications of high temperature reactors: helium gas turbine and process heat. Long term creep behavior and corrosion tests are conducted in simulated HTR helium up to 1000 deg C with impurities additions in the furnace atmosphere. Corrosion studies on AISI 321 steels at 800-1000 deg C have shown that the O 2 partial pressure is as low as 10 -24+-3 atm, Ni and Fe cannot be oxidised above about 500 and 600 deg C, Cr cease to oxidise at 800 to 900 deg C and Ti at 900 to 1000 deg C depending on alloy composition γ' strengthened superalloys must depend on a protective corrosion mechanism assisted by the presence of Ti and possibly Cr. Carburisation has been identified metallographically in several high temperature materials: Hastelloy X and M21Z. Alloy TZM appears to be inert in HTR Helium at 900 and 1000 deg C. In alloy 800 and Inconel 625 surface cracks initiation is suppressed but crack propagation is accelerated but this was not apparent in AISI steels, Hastelloy X or fine grain Inconel at 750 deg C

  9. A comparative study on optimization of machining parameters by turning aerospace materials according to Taguchi method

    Directory of Open Access Journals (Sweden)

    Altin Abdullah

    2017-01-01

    Full Text Available The effects of cutting tool coating material and cutting speed on cutting forces and surface roughness were investigated by Taguchi experimental design. Main cutting force, Fz is considered as a criterion. The effects of machining parameters were investigated using Taguchi L18 orthogonal array. Optimal cutting conditions were determined using the signal-to-noise (S/N ratio which is calculated for average surface roughness and cutting force according to the “the smaller is better” approach. Using results of analysis of variance (ANOVA and signal-to-noise (S/N ratio, effects of parameters on both average surface roughness and cutting forces were statistically investigated. It was observed that feed rate and cutting speed had higher effect on cutting force in Hastelloy X, while the feed rate and cutting tool had higher effect on cutting force in Inconel 625. According to average surface roughness the cutting tool and feed rate had higher effect in Hastelloy X and Inconel 625.

  10. Development of the Sixty Watt Heat-Source hardware components

    International Nuclear Information System (INIS)

    McNeil, D.C.; Wyder, W.C.

    1995-01-01

    The Sixty Watt Heat Source is a nonvented heat source designed to provide 60 thermal watts of power. The unit incorporates a plutonium-238 fuel pellet encapsulated in a hot isostatically pressed General Purpose Heat Source (GPHS) iridium clad vent set. A molybdenum liner sleeve and support components isolate the fueled iridium clad from the T-111 strength member. This strength member serves as the pressure vessel and fulfills the impact and hydrostatic strength requirements. The shell is manufactured from Hastelloy S which prevents the internal components from being oxidized. Conventional drawing operations were used to simplify processing and utilize existing equipment. The deep drawing reqirements for the molybdenum, T-111, and Hastelloy S were developed from past heat source hardware fabrication experiences. This resulted in multiple step drawing processes with intermediate heat treatments between forming steps. The molybdenum processing included warm forming operations. This paper describes the fabrication of these components and the multiple draw tooling developed to produce hardware to the desired specifications. copyright 1995 American Institute of Physics

  11. Corrosion susceptibility study of candidate pin materials for ALTC (Active Lithium/Thionyl Chloride) batteries

    Science.gov (United States)

    Bovard, Francine S.; Cieslak, Wendy R.

    1987-09-01

    The corrosion susceptibilities of eight alternate battery pin material candidates for ALTC (Active Lithium/Thionyl Chloride) batteries in 1.5M LiAlCl4/SOCl2 electrolyte have been investigated using ampule exposure and electrochemical tests. The thermal expansion coefficients of these candidate materials are expected to match Sandia-developed Li-corrosion resistant glasses. The corrosion resistances of the candidate materials, which included three stainless steels (15-5 PH, 17-4 PH, and 446), three Fe-Ni glass sealing alloys (Kovar, Alloy 52, and Niromet 426), a Ni-based alloy (Hastelloy B-2) and a zirconium-based alloy (Zircaloy), were compared to the reference materials Ni and 316L SS. All of the candidate materials showed some evidence of corrosion and, therefore, did not perform as well as the reference materials. The Hastelloy B-2 and Zircaloy are clearly unacceptable materials for this application. Of the remaining alternate materials, the 446 SS and Alloy 52 are the most promising candidates.

  12. Corrosion susceptibility study of candidate pin materials for ALTC (active lithium/thionyl chloride) batteries. [Active lithium/thionyl chloride

    Energy Technology Data Exchange (ETDEWEB)

    Bovard, F.S.; Cieslak, W.R.

    1987-09-01

    (ALTC = active lithium/thionyl chloride.) We have investigated the corrosion susceptibilities of eight alternate battery pin materials in 1.5M LiAlCl/sub 4//SOCl/sub 2/ electrolyte using ampule exposure and electrochemical tests. The thermal expansion coefficients of these candidate materials are expected to match Sandia-developed Li-corrosion resistant glasses. The corrosion resistances of the candidate materials, which included three stainless steels (15-5 PH, 17-4 PH, and 446), three Fe-Ni glass sealing alloys (Kovar, Alloy 52, and Niromet 426), a Ni-based alloy (Hastelloy B-2) and a zirconium-based alloy (Zircaloy), were compared to the reference materials Ni and 316L SS. All of the candidate materials showed some evidence of corrosion and, therefore, did not perform as well as the reference materials. The Hastelloy B-2 and Zircaloy are clearly unacceptable materials for this application. Of the remaining alternate materials, the 446 SS and Alloy 52 are the most promising candidates.

  13. Time-dependent high-temperature low-cycle fatigue behavior of nickel-base heat-resistant alloys for HTGR

    International Nuclear Information System (INIS)

    Tsuji, Hirokazu; Kondo, Tatsuo

    1988-06-01

    A series of strain controlled low-cycle fatigue tests at 900 deg C in the simulated HTGR helium environment were conducted on Hastelloy X and its modified version, Hastelloy XR in order to examine time-dependent high-temperature low-cycle fatigue behavior. In the tests with the symmetric triangular strain waveform, decreasing the strain rate led to notable reductions in the fatigue life. In the tests with the trapezoidal strain waveform with different holding types, the fatigue life was found to be reduced most effectively in tensile hold-time experiments. Based on the observations of the crack morphology the strain holding in the compressive side was suggested to play the role of suppressing the initiation and the growth of internal cracks or cavities, and to cause crack branching. When the frequency modified fatigue life method and/or the prediction of life by use of the ductility were applied, both the data obtained with the symmetric triangular strain waveform and those with the tensile hold-time experiments lay on the straight line plots. The data, however, obtained with the compressive and/or both hold-time experiments could not be handled satisfactorily by those methods. When the cumulative damage rule was applied, it was found that the reliability of HTGR components was ensured by limiting the creep-fatigue damage fraction within the value of 1. (author)

  14. Preparation of YBaCuO superconducting tape by RF magnetron sputtering

    Energy Technology Data Exchange (ETDEWEB)

    Fukutomi, Masao; Akutsu, Nakao; Tanaka, Yoshiaki; Asano, Toshihisa; Maeda, Hiroshi (National Research Inst. for Metals, Tsukuba (Japan); Mitsui Mining and Smelting Co., Ltd., Tokyo (Japan))

    1989-04-01

    The effect of buffer layers, conditions of film preparation, and the relation between superconducting characteristics and bombardment of high energy ions on films were discussed in an attempt to fabricate YBaCuO films on metallic substrates by sputtering. Hastelloy-X tapes and Chromel (Ni-10Cr) fine wires were used as metallic substrates, and MgO films as buffer layers, which were provided by sputtering a MgO sintered target and annealing. As a result, superconducting films were favorably obtained on the Hastelloy tapes with the MgO buffer layers, however, counter diffusion at the interface of the film and layer was unavoidable in annealing. C axis-highly oriented film with high zero resistance Tc was obtained in such an arrangement of the target and substrate as to lower the effect of 0{sup {minus}} ion resputtering, resulting in the most favorable Tc=80.4K. YBaCuO superconducting films could be also deposited on a bundle of Chromel fine wires preliminarily. 11 refs., 7 figs.

  15. Corrosion of nickel-base heat resistant alloys in simulated VHTR coolant helium at very high temperatures

    International Nuclear Information System (INIS)

    Shindo, Masami; Kondo, Tatsuo

    1976-01-01

    A comparative evaluation was made on three commercial nickel-base heat resistant alloys exposed to helium-base atmosphere at 1000 0 C, which contained several impurities in simulating the helium cooled very high temperature nuclear reactor (VHTR) environment. The choice of alloys was made so that the effect of elements commonly found in commercial alloys were typically examined. The corrosion in helium at 1000 0 C was characterized by the sharp selection of thermodynamically unstable elements in the oxidizing process and the resultant intergranular penetration and internal oxidation. Ni-Cr-Mo-W type solution hardened alloy such as Hastelloy-X showed comparatively good resistance. The alloy containing Al and Ti such as Inconel-617 suffered adverse effect in contrast to its good resistance to air oxidation. The alloy nominally composed only of noble elements, Ni, Fe and Mo, such as Hastelloy-B showed least apparent corrosion, while suffered internal oxidation due to small amount of active impurities commonly existing in commercial heats. The results were discussed in terms of selection and improvement of alloys for uses in VHTR and the similar systems. (auth.)

  16. In situ corrosion studies on selected high level waste packaging materials under simulated disposal conditions in rock salt

    International Nuclear Information System (INIS)

    Smailos, E.; Schwarzkopf, W.; Koester, R.

    1988-01-01

    In order to qualify corrosion resistant materials for high level waste (HLW) packagings acting as a long-term barrier in a rock salt repository, the corrosion behavior of preselected materials is being investigated in laboratory-scale and in-situ experiments. This work reports about in-situ corrosion experiments on unalloyed steels, Ti 99.8-Pd, Hastelloy C4, and iron-base alloys, as nodular cast iron, Ni-Resist D4 and Si-cast iron, under simulated disposal conditions. The results of the investigations can be summarized as follows: (1) all materials investigated exhibited high resistance to corrosion under the conditions prevailing in the Brine Migration Test; (2) all materials and above all the materials with passivating oxide layers such as Ti 99.8-Pd and Hastelloy C4 which may corrode selectively already in the presence of minor amounts of brine had been resistant with respect to any type of local corrosion attack; the gamma-radiation of 3 · 10 2 Gy/h did not exert an influence on the corrosion behavior of the materials

  17. High temperature oxidation characteristics of developed Ni-Cr-W superalloys in air

    International Nuclear Information System (INIS)

    Suzuki, Tomio; Shindo, Masami

    1996-11-01

    For expanding utilization of the Ni-Cr-W superalloy, which has been developed as one of new high temperature structural materials used in the advanced High Temperature Gas-cooled Reactors (HTGRs), in various engineering fields including the structural material for heat utilization system, the oxidation behavior of this alloy in air as one of high oxidizing environments becomes one of key factors. The oxidation tests for the industrial scale heat of Ni-Cr-W superalloy with the optimized chemical composition and five kinds of experimental Ni-Cr-W alloys with different Cr/W ratio were carried out at high temperatures in the air compared with Hastelloy XR. The conclusions were obtained as follows. (1) The oxidation resistance of the industrial scale heat of Ni-Cr-W superalloy with the optimized chemical composition was superior to that of Hastelloy XR. (2) The most excellent oxidation resistance was obtained in an alloy with 19% Cr of the industrial scale heat of Ni-Cr-W superalloy. (author)

  18. Effect of carburizing helium environment on creep behavior of Ni-base heat-resistant alloys for high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Kurata, Yuji; Ogawa, Yutaka; Nakajima, Hajime

    1988-01-01

    Creep tests were conducted on Ni-base heat-resistant alloys Hastelloy XR and XR-II, i.e. versions of Hastelloy X modified for nuclear applications, at 950degC using four types of helium environment with different impurity compositions, and mainly the effect of carburization was examined. For all the materials tested, the values of creep rupture time obtained under the carburizing conditions were similar to or longer than those in the commonly used, standard test environment (JAERI Type B helium). The difference among the results was interpreted by the counterbalancing effects of the strengthening due to carburization and possible weakening caused under very low oxidizing potential. In the corrosion monitoring specimens pronounced carbon pick-up was observed in the environment with high carbon activity and very low oxidizing potential. Based on the results obtained in the present and the previous works, it is suggested that a moderate control of the impurity chemistry is important rather than simple purification of the coolant in protecting the material from the environment-enhanced degradation. Either condition with high or low extremes in the oxidizing and carburizing potentials may cause enhanced degradation and thus are desirable to be avoided at the elevated temperatures. (author)

  19. Compatibility tests between molten salts and metal materials (2)

    International Nuclear Information System (INIS)

    Shiina, Yasuaki

    2003-08-01

    Latent heat storage technology using molten salts can reduce temperature fluctuations of heat transfer fluid by latent heat for middle and high temperature regions. This enables us to operate several heat utilization systems in cascade connected to High Temperature Gas Cooled Reactors (HTGRs) from high to low temperature range by setting the latent heat storage system after a heat utilization system to reduce thermal load after the heat utilization systems. This latent heat technology is expected to be used for effective use of heat such as equalization of electric load between night and daytime. In the application of the latent heat technology, compatibility between molten salts and metal materials is very important because molten salts are corrosive, and heat transfer pipes and vessels will contact with the molten salts. It will be necessary to prevail the latent heat storage technique that normal metal materials can be used for the pipes and vessels. However, a few studies have been reported of compatibility between molten salts and metals in middle and high temperature ranges. In this study, four molten salts, range of the melting temperature from 490degC to 800degC, are selected and five metals, high temperature and corrosion resistance steels of Alloy600, HastelloyB2, HastelloyC276, SUS310S and pure Nickel are selected for the test with the consideration of metal composition. Test was performed in an electric furnace by setting the molten salts and the metals in melting pots in an atmosphere of nitrogen. Results revealed excellent corrosion resistance of pure Nickel and comparatively low corrosion resistance of nickel base alloys such as Alloy600 and Hastelloys against Li 2 CO 3 . Corrosion resistance of SUS310S was about same as nickel based alloys. Therefore, if some amount of corrosion is permitted, SUS310S would be one of the candidate alloys for structure materials. These results will be used as reference data to select metals in latent heat technology

  20. Fracture and flaking off behavior of coated layer of DyBCO coated conductor under applied tensile strain

    International Nuclear Information System (INIS)

    Arai, T.; Shin, J.K.; Matsubayashi, H.; Ochiai, S.; Okuda, H.; Osamura, K.; Prusseit, W.

    2009-01-01

    The tensile behavior of the DyBa 2 Cu 3 O 7-δ (DyBCO) coated conductor with MgO buffer layer deposited on the Hastelloy C-276 substrate by inclined substrate deposition (ISD) was studied. The tensile stress-strain curve showed a flat region, characterized by the discontinuous yielding of the substrate due to the Lueders band extension from the gripped portions of the sample. In the area where the Lueders band had passed, the coating layer showed severe multiple transverse cracking due to the localized plastic deformation of the substrate. The flaking off of the coating layers took place at high applied strain, due to the buckling fracture of the coated layers in the sample width direction, accompanied by the interfacial debonding.

  1. Self-field ac losses in biaxially aligned Y endash Ba endash Cu endash O tape conductors

    International Nuclear Information System (INIS)

    Iijima, Y.; Hosaka, M.; Sadakata, N.; Saitoh, T.; Kohno, O.; Takeda, K.

    1997-01-01

    Self-field ac losses were measured by the conventional ac four-probe method in biaxially aligned Y endash Ba endash Cu endash O tapes using polycrystalline Hastelloy tapes with textured yttria-stabilized-zirconia buffer layers. The ac losses increased in proportion to the fourth power of transport current in the high J c sample, and agreed well with Norris close-quote equation for thin strip conductors. However, the low J c sample had rather higher losses than Norris close-quote prediction, suggesting excessive magnetic flux penetration caused by percolated current paths. The results confirmed Norris close-quote prediction of the low ac losses for thin strip conductors, and indicated the importance of removing percolated structures of current paths to avoid higher ac losses than the theoretical predictions based on uniform conductors. copyright 1997 American Institute of Physics

  2. Corrosion resistant alloy uses in the power industry

    International Nuclear Information System (INIS)

    Nickerson, J.L.; Hall, F.A.; Asphahani, A.I.

    1989-01-01

    Nickel-base alloys have been used as cost-effective measures in a variety of severely corrosive situations in pollution control units for coal-fired power plants. Cost effectiveness and practical answers to corrosion problems are illustrated (specifically the wallpaper concept/metallic lining technique). Numerous cases of successful use of HASTELLOY alloys in Flue Gas Desulfurization (FGD) systems and hazardous waste treatment incineration scrubber systems are listed. In this paper developments in nickel-base alloys and their use in FGD and other segments of the power industry are discussed. In the Ni-Cr-Mo-W alloy family, the C-22 alloy has the best resistance to localized corrosion in halide environments (chloride/fluoride-containing solutions). This alloy is also used effectively as a universal filler metal to weld less-resistant alloys were weld corrosion may be a problem. Field performance of this alloy in the power industry is described

  3. Preliminary design study of an alternate heat source assembly for a Brayton isotope power system

    Science.gov (United States)

    Strumpf, H. J.

    1978-01-01

    Results are presented for a study of the preliminary design of an alternate heat source assembly (HSA) intended for use in the Brayton isotope power system (BIPS). The BIPS converts thermal energy emitted by a radioactive heat source into electrical energy by means of a closed Brayton cycle. A heat source heat exchanger configuration was selected and optimized. The design consists of a 10 turn helically wound Hastelloy X tube. Thermal analyses were performed for various operating conditions to ensure that post impact containment shell (PICS) temperatures remain within specified limits. These limits are essentially satisfied for all modes of operation except for the emergency cooling system for which the PICS temperatures are too high. Neon was found to be the best choice for a fill gas for auxiliary cooling system operation. Low cycle fatigue life, natural frequency, and dynamic loading requirements can be met with minor modifications to the existing HSA.

  4. Effect of long-term aging at 8150C on the tensile properties and microstructural stability of four cobalt- and nickel-base superalloys

    International Nuclear Information System (INIS)

    Hammond, J.P.

    1976-08-01

    Two heats of Haynes alloy 25 and one heat each of Haynes alloy 188, Hastelloy N, and Inconel 625 were tensile tested after aging for 11,000 h at 816 0 C. Yield strength, ultimate tensile strength, and elongation were determined 24, 316, 760, and 982 0 C and compared with typical properties for these materials in the solution annealed condition. Toughness values were determined for these materials from their engineering stress-strain curves. The long-term aging treatment degraded ductility and toughness at room temperature but, contrary to behavior expected for overaging, enhanced them over those for the solution annealed condition in tests at 760 0 C. The tensile properties of the aged superalloys were correlated with mode of fracture and the amounts, identity, and morphology of the precipitates. Aging substantially depleted the hardener tungsten from the matrix in the cobalt-base alloys

  5. Materials for advanced high temperature reactors

    International Nuclear Information System (INIS)

    Graham, L.W.

    1976-01-01

    The results recently obtained from the Dragon program are presented to illustrate materials behavior: (a) effect of temperature on oxidation and carburisation in HTR helium (variation in oxide depth and in C content of AISI 321 after 5000 hours in HTR helium; effect of temperature on surface scale formation in the γ' strengthened alloys Nimonic 80A and 713LC); (b) effect of alloy composition on oxidation and carburisation behavior (influence of Nb and Ti on the corrosion of austenitic steels; influence of Ti and Al in IN-102; weight gain of cast high Ni alloys); (c) effect of environment on creep strength (results of tests for hastelloy X, grade I inconel 625, grade II inconel 625 and inconel 617 in He and air between 750 and 800 0 C)

  6. Development of corrosion and wear resistant coatings by an improved HVOF spraying process

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, Y.; Kawakita, J.; Kuroda, S. [National Inst. for Materials Science, Tsukuba (Japan)

    2005-07-01

    We have developed an improved HVOF spray process called ''Gas-shrouded HVOF'' (GS-HVOF) over the past several years. By using an extension nozzle at the exit of a commercial HVOF spray gun, GS-HVOF is capable of controlling the oxidation of sprayed materials during flight as well as achieving higher velocity of sprayed particles. These features result in extremely dense and clean microstructure of the sprayed coatings. The process has been successfully applied to corrosion resistant alloys such as SUS316L, Hastelloy C, and alloy 625 as well as cermets such as WC-Cr{sub 3}C{sub 2}-Ni. The spray process, coatings microstructure and property evaluation will be discussed with potential industrial applications in the near future. (orig.)

  7. The molten salt reactor adventure

    International Nuclear Information System (INIS)

    MacPherson, H.G.

    1985-01-01

    A personal history of the development of molten salt reactors in the United States is presented. The initial goal was an aircraft propulsion reactor, and a molten fluoride-fueled Aircraft Reactor Experiment was operated at Oak Ridge National Laboratory in 1954. In 1956, the objective shifted to civilian nuclear power, and reactor concepts were developed using a circulating UF 4 -ThF 4 fuel, graphite moderator, and Hastelloy N pressure boundary. The program culminated in the successful operation of the Molten Salt Reactor Experiment in 1965 to 1969. By then the Atomic Energy Commission's goals had shifted to breeder development; the molten salt program supported on-site reprocessing development and study of various reactor arrangements that had potential to breed. Some commercial and foreign interest contributed to the program which, however, was terminated by the government in 1976. The current status of the technology and prospects for revived interest are summarized

  8. Molten salt breeder reactor

    International Nuclear Information System (INIS)

    Furukawa, Kazuo; Tsukada, Kineo; Nakahara, Yasuaki; Oomichi, Toshihiko; Oono, Hideo.

    1982-01-01

    Purpose: To simplify the structure, as well as improve the technical reliability and safety by the elimination of a proton beam entering window. Constitution: The nuclear reactor container main body is made of Hastelloy N and provided at the inner surface with two layers of graphite shields except for openings. An aperture was formed in the upper surface of the container, through which protons accelerated by a linear accelerator are directly entered to the liquid surface of molten salts such as 7LiF-BeF 2 -ThF 4 , 7LiF-NaF-ThF 4 , 7LiF-Rb-UF 4 , NaF-KF-UF 4 and the like. The heated molten salts are introduced by way of a pipeway into a heat exchanger where the heat is transferred to coolant salts and electric generation is conducted by way of heated steams. (Furukawa, Y.)

  9. Study of tertiary creep instability in several elevated-temperature structural materials

    International Nuclear Information System (INIS)

    Booker, M.K.; Sikka, V.K.

    1978-01-01

    Data for a number of common elevated temperature structural materials have been analyzed to yield mathematical predictions for the time and strain to tertiary creep at various rupture lives and temperatures. Materials examined include types 304 and 316 stainless steel, 2 1/4 Cr-1 Mo steel, alloy 800H, alloy 718, Hastelloy alloy X, and ERNiCr--3 weld metal. Data were typically examined over a range of creep temperatures for rupture lives ranging from less than 100 to greater than 10,000 hours. Within a given material, trends in these quantities can be consistently described, but it is difficult to directly relate the onset of tertiary creep to failure-inducing instabilities. A series of discontinued tests for alloy 718 at 649 and 620 0 C showed that the material fails by intergranular cracking but that no significant intergranular cracking occurs until well after the onset of tertiary creep

  10. Stress Corrosion Evaluation of Nitinol 60 for the International Space Station Water Recycling System

    Science.gov (United States)

    Torres, P. D.

    2016-01-01

    A stress corrosion cracking (SCC) evaluation of Nitinol 60 was performed because this alloy is considered a candidate bearing material for the Environmental Control and Life Support System (ECLSS), specifically in the Urine Processing Assembly of the International Space Station. An SCC evaluation that preceded this one during the 2013-2014 timeframe included various alloys: Inconel 625, Hastelloy C-276, titanium (Ti) commercially pure (CP), Ti 6Al-4V, extra-low interstitial (ELI) Ti 6Al-4V, and Cronidur 30. In that evaluation, most specimens were exposed for a year. The results of that evaluation were published in NASA/TM-2015-218206, entitled "Stress Corrosion Evaluation of Various Metallic Materials for the International Space Station Water Recycling System,"1 available at the NASA Scientific and Technical Information program web page: http://www.sti.nasa.gov. Nitinol 60 was added to the test program in 2014.

  11. Effect of glass-ceramic-processing cycle on the metallurgical properties of candidate alloys for actuator housings

    Energy Technology Data Exchange (ETDEWEB)

    Weirick, L.J.

    1982-01-01

    This report summarizes the results from an investigation on the effect of a glass ceramic processing cycle on the metallurgical properties of metal candidates for actuator housings. The cycle consists of a 980/sup 0/C sealing step, a 650/sup 0/C crystallization step and a 475/sup 0/C annealing step. These temperatue excursions are within the same temperature regime as annealing and heat treating processes normally employed for metals. Therefore, the effect of the processing cycle on metallurgical properties of microstructure, strength, hardness and ductility were examined. It was found that metal candidates which are single phase or solid solution alloys (such as 21-6-9, Hastelloy C-276 and Inconel 625) were not affected whereas multiphase or precipitation hardened alloys (such as Inconel 718 and Titanium ..beta..-C) were changed by the processing cycle for the glass ceramic.

  12. High-Temperature Structural Analysis of a Small-Scale Prototype of a Process Heat Exchanger (IV) - Macroscopic High-Temperature Elastic-Plastic Analysis -

    International Nuclear Information System (INIS)

    Song, Kee Nam; Hong, Sung Deok; Park, Hong Yoon

    2011-01-01

    A PHE (Process Heat Exchanger) is a key component required to transfer heat energy of 950 .deg. C generated in a VHTR (Very High Temperature Reactor) to a chemical reaction that yields a large quantity of hydrogen. A small-scale PHE prototype made of Hastelloy-X was scheduled for testing in a small-scale gas loop at the Korea Atomic Energy Research Institute. In this study, as a part of the evaluation of the high-temperature structural integrity of the PHE prototype, high-temperature structural analysis modeling, and macroscopic thermal and elastic-plastic structural analysis of the PHE prototype were carried out under the gas-loop test conditions as a preliminary qwer123$ study before carrying out the performance test in the gas loop. The results obtained in this study will be used to design the performance test setup for the modified PHE prototype

  13. Abiotic nitrate and sulphate reduction by hydrogen: a comparative experimental study

    International Nuclear Information System (INIS)

    Truche, L.; Berger, G.; Albrecht, A.; Giffaut, E.

    2010-01-01

    Document available in extended abstract form only. The bituminous waste which is part of the intermediate level, long-lived waste (MAVL) is characterised, amongst others, by the coexistence of nitrates, sulphates, organic matter, native metals and hydrogen gas in the waste mixture and package. It can be considered as the most complex example that will be used to discuss redox reactions occurring in such waste mixtures. The evaluation of the redox conditions requires quantification of the amount of electron acceptors and donors and definition of the kinetics of redox reaction. The objectives of an experimental study to unravel some of these reaction complexities are: - to investigate nature and rate of sulphate and nitrate reduction by hydrogen in the presence of different catalysts (stainless steel, hastelloy, magnetite and argillite); - to compare sulphate and nitrate as electron acceptors; - to provide a mechanistic model of these reactions. It is well known that reduction of sulphate and nitrate requires high activation energies, usually supplied either by thermal processes or via bacterial and surface catalysis, of which the latter has been investigated in this study. Preliminary experiments performed at 150 deg. C and under H 2 pressure show that sulphate reduction is enhanced in the presence of magnetite, but essentially under the restricted condition of low sulphate concentration and at a pH below the Point of Zero Charge of magnetite. This suggests that sorption of sulphate contributes to the catalysed reaction (at low pH) but provided that the magnetite surface sites are not saturated with respect to aqueous sulphate (low concentration). On the contrary, nitrate reduction is observed whatever the pH and the nitrate concentration in the presence of both magnetite and hastelloy C276 (Ni, Cr, Mo, W, Fe alloy). The effect of temperature on the rate of nitrate reduction (500 ppm KNO 3 solution) is shown by comparing three different experiments conducted in

  14. Embrittlement of nickel-, cobalt-, and iron-base superalloys by exposure to hydrogen

    Science.gov (United States)

    Gray, H. R.

    1975-01-01

    Five nickel-base alloys (Inconel 718, Udimet 700, Rene 41, Hastelloy X, and TD-NiCr), one cobalt-base alloy (L-605), and an iron-base alloy (A-286) were exposed in hydrogen at 0.1 MN/sq m (15 psi) at several temperatures in the range from 430 to 980 C for as long as 1000 hours. These alloys were embrittled to varying degrees by such exposures in hydrogen. Embrittlement was found to be: (1) sensitive to strain rate, (2) reversible, (3) caused by large concentrations of absorbed hydrogen, and (4) not associated with any detectable microstructural changes in the alloys. These observations are consistent with a mechanism of internal reversible hydrogen embrittlement.

  15. Radiolysis of water in the vicinity of passive surfaces

    International Nuclear Information System (INIS)

    Moreau, S.; Fenart, M.; Renault, J.P.

    2014-01-01

    Highlights: • HO° production through water radiolysis is enhanced near metal surfaces. • Hastelloy and Stainless steel surfaces can also produce HO° radicals through hydrogen peroxide activation. • There is a deficit in solvated electron production compared to hydroxyl radicals near metal surfaces. - Abstract: Porous metals were used to describe the water radiolysis in the vicinity of metal surfaces. The hydroxyl radical production under gamma irradiation was measured by benzoate scavenging in water confined in a 200 nm porous Ni base alloy or in Stainless steel. The presence of the metallic surfaces changed drastically the HO° production level and lifetime. The solvated electron production was measured via glycylglycine scavenging for Stainless steel and was found to be significantly smaller than hydroxyl production. These observations imply that interfacial radiolysis may deeply impact the corrosion behavior of the SS and Ni based alloys

  16. Selection of replacement material for the failed surface level gauge wire in Hanford waste tanks

    International Nuclear Information System (INIS)

    Anantatmula, R.P.; Pitman, S.G.; Lund, A.L.

    1995-10-01

    Surface level gauges fabricated from AISI Type 316 stainless steel (316) wire failed after only a few weeks of operation in underground storage tanks at the Hanford Site. The wire failure was determined to be due to chloride ion assisted corrosion of the 316 wire. Radiation-induced breakdown of the polyvinyl chloride (PVC) riser liners is suspected to be the primary source of the chloride ions. An extensive literature search followed by expert concurrence was undertaken to select a replacement material for the wire. Platinum (Pt)-20 % Iridium (Ir) alloy was selected as the replacement material from tile candidate materials, P-20% Ir, Pt-1O% Rhodium (Rh), Pt-20%Rh and Hastelloy C-22. The selection was made on the basis of the alloy's immunity towards acidic and basic environments as well as its adequate tensile properties in the fully annealed state

  17. System design description of forced-convection molten-salt corrosion loops MSR-FCL-3 and MSR-FCL-4

    International Nuclear Information System (INIS)

    Huntley, W.R.; Silverman, M.D.

    1976-11-01

    Molten-salt corrosion loops MSR-FCL-3 and MSR-FCL-4 are high-temperature test facilities designed to evaluate corrosion and mass transfer of modified Hastelloy N alloys for future use in Molten-Salt Breeder Reactors. Salt is circulated by a centrifugal sump pump to evaluate material compatibility with LiF-BeF 2 -ThF 4 -UF 4 fuel salt at velocities up to 6 m/s (20 fps) and at salt temperatures from 566 to 705 0 C (1050 to 1300 0 F). The report presents the design description of the various components and systems that make up each corrosion facility, such as the salt pump, corrosion specimens, salt piping, main heaters, salt coolers, salt sampling equipment, and helium cover-gas system, etc. The electrical systems and instrumentation and controls are described, and operational procedures, system limitations, and maintenance philosophy are discussed

  18. Fundamental studies on electron-beam welding of heat-resistant superalloys for nuclear plants: Report 4. Mechanical properties of welded joints

    International Nuclear Information System (INIS)

    Susei, S.; Shimizu, S.; Aota, T.

    1982-04-01

    In this report, electron-beam (EB) welded joints and TIG welded joints of various superalloys to be used for nuclear plants, such as Hastelloy-type, Inconel-type and Incoloy-type, are systematically evaluated in terms of tensile properties, low-cycle fatigue properties at elevated temperatures, creep and creep-rupture properties. It was fully confirmed as conclusion that the EB welded joints are superior to the TIG welded ones in mechanical properties, especially at high temperature. In the evaluation of creep properties, ductility is one of the most important criteria to represent the resistance against fracture due to creep deformation, and this criterion is very useful in evaluating the properties of welded joints. Therefore, the more comparable to the base metal the electron beam welded joint becomes in terms of ductility, the more resistant is it against fracture. From this point of view, the electron beam welded joint is considerably superior to the TIG welded joint [fr

  19. Corrosion testing of selected packaging materials for disposal of high-level waste glass in rock salt formations

    International Nuclear Information System (INIS)

    Smailos, E.; Schwarzkopf, W.; Koester, R.; Fiehn, B.; Halm, G.

    1990-05-01

    In previous corrosion studies performed in salt brines, unalloyed steels, Ti 99.8-Pd and Hastelloy C4 have proved to be the most promising materials for long-term resistant packagings to be used in heat-generating waste (vitrified HLW, spent fuel) disposal in rock-salt formations. To characterise the corrosion behaviour of these materials in more detail, further in-depth laboratory-scale and in-situ corrosion studies have been performed in the present study. Besides the above-mentioned materials, also some in-situ investigations of the iron-base materials Ni-Resist D2 and D4, cast iron and Si-cast iron have been carried out in order to complete the results available to date. (orig.) [de

  20. Frictional forces in an SOFC stack with sliding seals

    Energy Technology Data Exchange (ETDEWEB)

    Yamazaki, T; Oishi, N; Namikawa, T; Yamazaki, Y [Tokyo Institute of Technology, Tokyo (Japan)

    1996-06-05

    The detrimental thermal stresses in planar SOFC stacks can be reduced using sliding seals. In the proposal planar stack the electrolyte film is sandwiched by YSZ support rings to release the thermal stresses. In order to estimate the strength of the support ring, the frictional forces between heat resistant alloy and YSZ were measured at 900{degree}C. The coefficient of friction between Hastelloy X and YSZ increased when they were measured lifter 144h heating. However, the coefficient of friction between HA-214 and YSZ did not increase. The measurement and a calculation of the stresses in the support rings led the result that a thickness of 0.6mm was necessary for 200mm diameter support rings under a stack pressure of 0.1kgcm{sup -2}. 6 refs., 9 figs., 1 tab.

  1. Recanning of canned motors (Paper No. 6.3)

    International Nuclear Information System (INIS)

    Lahiri, B.N.; Venkatappiah, J.; Srikrishnamurthy, G.

    1992-01-01

    Good performance of any plant necessarily depends on the quality of preventive as well as break-down maintenance management practised along with the adequacy of knowledge and skill of the supporting personnel. Heavy water plants, operating under exceedingly high pressure and varied corrosive atmospheres definitely call for additional care and attention in this regard. Canned motors of different power ratings and shapes find their use in operating heavy water plants. In this paper, complete re-canning procedure of one such canned motor has been discussed giving design details of hydraulic expansion fixture and pulsed-tig welding procedure for longitudinal seam and circumferential edge welding of 0.5 mm thick hastelloy cans. (author). 4 figs

  2. Vessels for elevated temperature service

    International Nuclear Information System (INIS)

    O'Donnell, W.J.; Porowski, J.S.

    1983-01-01

    The subject is covered in chapters, entitled: introduction (background; elevated temperature concerns; design tools); design of pressure vessels for elevated temperature per ASME code; basic elevated temperature failure modes; allowable stresses and strains per ASME code (basic allowable stress limits; ASME code limits for bending; time-fraction summations; strain limits; buckling and instability; negligible creep and stress-rupture effects); combined membrane and bending stresses in creep regime; thermal stress cycles; bounding methods based on elastic core concept (bounds on accumulated strains; more accurate bounds; strain ranges; maximum stresses; strains at discontinuities); elastic follow-up; creep strain concentrations; time-dependent fatigue (combined creep rupture and fatigue damage; limits for inelastic design analyses; limits for elastic design analyses); flaw evaluation techniques; type 316 stainless steel; type 304 stainless steel; steel 2 1/4Cr1Mo; Inconel 718; Incolloy 800; Hastelloy X; detailed inelastic design analyses. (U.K.)

  3. High temperature ductility of austenitic alloys exposed to thermal neutrons

    International Nuclear Information System (INIS)

    Watanabe, K.; Kondo, T.; Ogawa, Y.

    1982-01-01

    Loss of high temperature ductility due to thermal neutron irradiation was examined by slow strain rate test in vacuum up to 1000 0 C. The results on two heats of Hastelloy alloy X with different boron contents were analyzed with respect to the influence of the temperatures of irradiation and tensile tests, neutron fluence and the associated helium production due to nuclear transmutation reaction. The loss of ductility was enhanced by increasing either temperature or neutron fluence. Simple extrapolations yielded the estimated threshold fluence and the end-of-life ductility values at 900 and 1000 0 C in case where the materials were used in near-core regions of VHTR. The observed relationship between Ni content and the ductility loss has suggested a potential utilization of Fe-based alloys for seathing of the neutron absorber materials

  4. Hanford waste encapsulation: strontium and cesium

    International Nuclear Information System (INIS)

    Jackson, R.R.

    1976-06-01

    The strontium and cesium fractions separated from high radiation level wastes at Hanford are converted to the solid strontium fluoride and cesium chloride salts, doubly encapsulated, and stored underwater in the Waste Encapsulation and Storage Facility (WESF). A capsule contains approximately 70,000 Ci of 137 Cs or 70,000 to 140,000 Ci of 90 Sr. Materials for fabrication of process equipment and capsules must withstand a combination of corrosive chemicals, high radiation dosages and frequently, elevated temperatures. The two metals selected for capsules, Hastelloy C-276 for strontium fluoride and 316-L stainless steel for cesium chloride, are adequate for prolonged containment. Additional materials studies are being done both for licensing strontium fluoride as source material and for second generation process equipment

  5. Corrosion studies and recommendation of alloys for an incinerator of glove-boxes wastes

    International Nuclear Information System (INIS)

    Devisme, F.; Garnier, M.H.

    1992-01-01

    In the framework of the development of an incineration process for high chlorinated wastes, commercial alloys have been investigated by means of parametric laboratory tests in HCl containing gas mixtures and also in field tests. Recommendations may be formulated for the three main components i.e. pyrolyser, calciner and cooler. In very low oxygen-potential atmospheres, the alloys Hastelloy C276 and Inconel 625 present the best behaviours. For the calciner, alloy Inconel 601 is more satisfactory than AISI 310 steel. As for the cooler, only the alloy Haynes 214 appears acceptable at 1100 deg C. Because of the very low stress level affecting the components, thermomechanical properties do not modify these recommendations based on corrosion behaviour

  6. The performance of electroless nickel deposits in oil-field environments

    International Nuclear Information System (INIS)

    Mack, R.; Bayes, M.

    1984-01-01

    An experimental study was conducted on an electroless nickel plated (represented by Enplate NI-422) C-90 steel, uncoated C-90 steel, AISI 420, 174 PH, SAF 2205, and HASTELLOY /sup R/ G-3 to determine their corrosion-performance in twelve simulated downhole oil or gas production environments during 28 day exposures. These environments were aqueous brines containing various concentrations of Cl - , H 2 S and/or CO 2 , and over a range of temperatures. The results from this study and oilfield data for electroless nickel plated low alloy steels are presented and discussed. The study demonstrates the feasibility of electroless nickel coated low alloy steels as an economical substitute for some highly alloyed materials in certain oilfield applications; the field data support this

  7. Salt Fog Testing Iron-Based Amorphous Alloys

    International Nuclear Information System (INIS)

    Rebak, Raul B.; Aprigliano, Louis F.; Day, S. Daniel; Farmer, Joseph C.

    2007-01-01

    Iron-based amorphous alloys are hard and highly corrosion resistant, which make them desirable for salt water and other applications. These alloys can be produced as powder and can be deposited as coatings on any surface that needs to be protected from the environment. It was of interest to examine the behavior of these amorphous alloys in the standard salt-fog testing ASTM B 117. Three different amorphous coating compositions were deposited on 316L SS coupons and exposed for many cycles of the salt fog test. Other common engineering alloys such as 1018 carbon steel, 316L SS and Hastelloy C-22 were also tested together with the amorphous coatings. Results show that amorphous coatings are resistant to rusting in salt fog. Partial devitrification may be responsible for isolated rust spots in one of the coatings. (authors)

  8. Corrosion behaviour of container materials for geological disposal of high level radioactive waste

    International Nuclear Information System (INIS)

    Accary, A.

    1985-01-01

    The disposal of high level radioactive waste in geological formations, based on the multibarrier concept, may include the use of a container as one of the engineered barriers. In this report the requirements imposed on this container and the possible degradation processes are reviewed. Further on an overview is given of the research being carried out by various research centres in the European Community on the assessment of the corrosion behaviour of candidate container materials. The results obtained on a number of materials under various testing conditions are summarized and evaluated. As a result, three promising materials have been selected for a detailed joint testing programme. It concerns two highly corrosion resistant alloys, resp. Ti-Pd (0.2 Pd%) and Hastelloy C4 and one consumable material namely a low carbon steel. Finally the possibilities of modelling the corrosion phenomena are discussed

  9. Compatibility of graphite with a martensitic-ferritic steel, an austenitic stainless steel and a Ni-base alloy up to 1250 C

    International Nuclear Information System (INIS)

    Hofmann, P.

    1994-08-01

    To study the chemical interactions between graphite and a martensitic-ferritic steel (1.4914), an austenitic stainless steel (1.4919; AISI 316), and a Ni-base alloy (Hastelloy X) isothermal reaction experiments were performed in the temperature range between 900 and 1250 C. At higher temperatures a rapid and complete liquefaction of the components occurred as a result of eutectic interactions. The chemical interactions are diffusion-controlled processes and can be described by parabolic rate laws. The reaction behavior of the two steels is very similar. The chemical interactions of the steels with graphite are much faster above 1100 C than those for the Ni-base alloy. Below 1000 C the effect is opposite. (orig.) [de

  10. Effect of fission product interactions on the corrosion and mechanical properties of HTGR alloys

    International Nuclear Information System (INIS)

    Aronson, S.; Chow, J.G.Y.; Soo, P.; Friedlander, M.

    1978-01-01

    Preliminary experiments have been carried out to determine how fission product interactions may influence the mechanical integrity of reference HTGR structural metals. In this work Type 304 stainless steel, Incoloy 800 and Hastelloy X were heated to 550 to 650 0 C in the presence of CsI. It was found that no corrosion of the alloys occurred unless air or oxygen was also present. A mechanism for the observed behavior is proposed. A description is also given of some long term exposures of HTGR materials to more prototypic, low concentrations of I 2 , Te 2 and CsI in the presence of low partial pressures of O 2 . These samples are scheduled for mechanical bend tests after exposure to determine the degree of embrittlement

  11. Susceptibility to Stress Corrosion Cracking of 254SMO SS

    Directory of Open Access Journals (Sweden)

    De Micheli Lorenzo

    2002-01-01

    Full Text Available The susceptibility to stress corrosion cracking (SCC of solubilized and sensitized 254SMO SS was studied in sodium chloride, and sodium fluoride solutions at 80 °C and sulfuric acid solutions in presence of sodium chloride at 25 °C. The influence of salt concentration, pH values and the addition of thiosulfate was examined. The susceptibility to SCC was evaluated by Slow Strain Rate Tests (SSRT, at 1.5 x 10-6 s-1 strain rate. The behavior of 254SMO was compared to those of AISI 316L SS and Hastelloy C276. 254SMO showed an excellent resistance to SCC in all conditions, except in the more acidic solutions (pH <= 1 where, in the sensitized conditions, intergranular stress corrosion cracking occurred.

  12. Segregation in welded nickel-base alloys

    International Nuclear Information System (INIS)

    Akhtar, J.I.; Shoaib, K.A.; Ahmad, M.; Shaikh, M.A.

    1990-05-01

    Segregation effects have been investigated in nickel-base alloys monel 400, inconel 625, hastelloy C-276 and incoloy 825, test welded under controlled conditions. Deviations from the normal composition have been observed to varying extents in the welded zone of these alloys. Least effect of this type occurred in Monel 400 where the content of Cu increased in some of the areas. Enhancement of Al and Ti has been found over large areas in the other alloys which has been attributed to the formation of low melting slag. Another common feature is the segregation of Cr, Fe or Ti, most likely in the form of carbides. Enrichment of Al, Ti, Nb, Mb, Mo, etc., to different amounts in some of the areas of these materials is in- terpretted in terms of the formation of gamma prime precipitates or of Laves phases. (author)

  13. 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

  14. Design and fabrication of the Mini-Brayton Recuperator (MBR)

    Science.gov (United States)

    Killackey, J. J.; Graves, R.; Mosinskis, G.

    1978-01-01

    Development of a recuperator for a 2.0 kW closed Brayton space power system is described. The plate-fin heat exchanger is fabricated entirely from Hastelloy X and is designed for 10 years continuous operation at 1000 K (1300 F) with a Xenon-helium working fluid. Special design provisions assure uniform flow distribution, crucial for meeting 0.975 temperature effectiveness. Low-cycle fatigue, resulting from repeated startup and shutdown cycles, was identified as the most critical structural design problem. It is predicted that the unit has a minimum fatigue life of 220 cycles. This is in excess of the BIPS requirement of 100 cycles. Heat transfer performance and thermal cycle testing with air, using a prototype unit, verified that all design objectives can be met.

  15. FRAUD/SABOTAGE Killing Nuclear-Reactors Need Modeling!!!: "Super"alloys GENERIC ENDEMIC Wigner's-Disease/.../IN-stability: Ethics? SHMETHICS!!!

    Science.gov (United States)

    Asphahani, Aziz; Siegel, Sidney; Siegel, Edward

    2010-03-01

    Carbides solid-state chemistry domination of old/new nuclear- reactors/spent-fuel-casks/refineries/jet/missile/rocket-engines in austenitic/FCC Ni/Fe-based(so miscalled)``super"alloys(182/82; Hastelloy-X,600,304/304L-SSs,...,690!!!) GENERIC ENDEMIC EXTANT detrimental(synonyms): Wigner's-diseas(WD)[J.Appl.Phys.17,857 (1946)]/Ostwald-ripening/spinodal-decomposition/overageing- embrittlement/thermal-leading-to-mechanical(TLTM)-INstability: Mayo[Google:``If Leaks Could Kill"; at flickr.com search on ``Giant-Magnotoresistance"; find: Siegel[J.Mag.Mag.Mtls.7,312 (1978)]Politics(1979)]-Hoffman[animatedsoftware.com], what DOE/NRC MISlabels as ``butt-welds" ``stress-corrosion cracking" endpoint's ROOT-CAUSE ULTIMATE-ORIGIN is WD overageing-embritt- lement caused brittle-fracture cracking from early/ongoing AEC/ DOE-n"u"tional-la"v"atories sabotage!!!

  16. Progress in understanding the mechanical behavior of pressure-vessel materials at elevated temperatures

    International Nuclear Information System (INIS)

    Swindeman, R.W.; Brinkman, C.R.

    1981-01-01

    Progress during the 1970's on the production of high-temperature mechanical properties data for pressure vessel materials was reviewed. The direction of the research was toward satisfying new data requirements to implement advances in high-temperature inelastic design methods. To meet these needs, servo-controlled testing machines and high-resolution extensometry were developed to gain more information on the essential behavioral features of high-temperature alloys. The similarities and differences in the mechanical response of various pressure vessel materials were identified. High-temperature pressure vessel materials that have received the most attention included Type 304 stainless steel, Type 316 stainless steel, 2 1/4 Cr-1 Mo steel, alloy 800H, and Hastelloy X

  17. 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

  18. Creep collapse of thick-walled heat transfer tube subjected to external pressure at high temperature

    International Nuclear Information System (INIS)

    Ioka, Ikuo; Kaji, Yoshiyuki; Terunuma, Isao; Nekoya, Shin-ichi; Miyamoto, Yoshiaki

    1994-09-01

    A series of creep collapse tests of thick-walled heat transfer tube were examined experimentally and analytically to confirm an analytical method for creep deformation behavior of a heat transfer tube of an intermediate heat exchanger (IHX) at a depressurization accident of secondary cooling system of HTTR (High Temperature Engineering Test Reactor). The tests were carried out using thick-walled heat transfer tubes made of Hastelloy XR at 950degC in helium gas environment. The predictions of creep collapse time obtained by a general purpose FEM-code ABAQUS were in good agreement with the experimental results. A lot of cracks were observed on the outer surface of the test tubes after the creep collapse. However, the cracks did not pass through the tube wall and, therefore, the leak tightness was maintained regardless of a collapse deformation for all tubes tested. (author)

  19. Research and Development programs for HTGRs in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Nishiguchi, Isoharu; Saito, Sinzo [Department of HTTR Project, Japan Atomic Energy Research Institute (Japan)

    1990-07-01

    Since 1969, JAERI has conducted research and development (R and D) programs for High-Temperature Gas-Cooled Reactors (HTGR). And the High Temperature engineering Test Reactor (HTTR), which will be the first High Temperature Gas Cooled Reactor (HTGR) in Japan, is under licensing process now. In this paper, some of the results of R and D are outlined in the following fields which are closely connected with the HTTR design, that is: i) fuel; ii) nuclear design; iii) thermal-hydraulic design; iv) graphite structure and v) high temperature metal structure. In the field of fuel, extensive investigations have been performed to develop the fabrication technology of coated particle fuel (cpf). In parallel, data of coated fuel particle failure and fission product release in in- and ex-reactor experiments as well as mechanical properties data were obtained and irradiation tests have been done using the Oarai Gas Loop No.1 (OGL-1) to verify the integrity of mass-produced fuel. Concerning the nuclear design, critical experiments were conducted using the Very High-Temperature Reactor Critical Assembly (VHTRC). Also carried out were hydrodynamical and thermal experiments using the Helium Engineering Demonstration Loop (HENDEL). On the graphite structures which compose the reactor internals, design criteria have been developed based on ASME BandPV Code Section III Div.2, subsection CE and design data have been accumulated on a domestic graphite material. High temperature metal structure is also one of major subjects of R and D for HTGRs. Hastelloy XR, which is a modified version of Hastelloy X, was developed and various tests have been conducted which include creep tests, creep-fatigue tests, etc. to establish design criteria and allowables. Component tests of the Intermediate Heat Exchanger (IHX) have been also performed. (author)

  20. 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)

  1. In-situ corrosion studies on selected high-level waste packaging materials under simulated disposal conditions in rock salt formations

    International Nuclear Information System (INIS)

    Schwarzkopf, W.; Smailos, E.; Koester, R.

    1988-01-01

    This work reports about in situ corrosion experiments on unalloyed steels, Ti 99,8-Pd, Hastelloy C4, and iron-base alloys, as modular cast iron, Ni-Resist D4 and Si-cast iron, under simulated disposal conditions. The experiments were carried out in the frame of the German/US Brine Migration Test in heated tubed boreholes in the Asse salt mine at T = 150 0 C to 210 0 C, both in the absence and in the presence of a γ-radiation field of 3x10 2 Gy/h (Co-60 source). In addition, the material used to protect the tubing from corrosion (Inconel 600) as well as the backfill material for the annular gap (Al 2 O 3 spheres) were investigated for possible corrosion attack. All materials investigate exhibited high resistance to corrosion under the conditions prevailing in the Brine Migration Test. All material specimens corroded at much lower rates than determined in the previous laboratory-scale tests. All materials and above all the materials with passivating oxide layers such as Ti 99.8-Pd and Hastelloy C4 which may corrode selectively already in the presence of minor amounts of brine had been resistant with respect to any type of local corrosion attack. The γ-radiation of 3x10 2 Gy/h did not exert an influence on the corrosion behaviour of the materials. No corrosion attacks were observed on the Al 2 O 3 spheres. In the case of Inconel 600 traces of sulphur were detected probably resulting from the reaction of Ni with H 2 S to NiS. Measurable general and local corrosion, however, have not been observed. (orig./IHOE) [de

  2. Molten salt reactor type

    International Nuclear Information System (INIS)

    1977-01-01

    This document is one of the three parts of a first volume devoted to the compilations of American data on the molten salt reactor concept. This part 'CIRCUITS' regroups under a condensed form - in French and using international units - the essential information contained in both basic documents of the American project for a molten-salt breeder power plant. This part is only dealing with things relating to the CEA-EDF workshop 'CIRCUITS'. It is not concerned with information on: the reactor and the moderator replacement, the primary and secondary salts, and the fuel salt reprocessing, that are dealt with in parts 'CORE' and 'CHEMISTRY' respectively. The possible evolutions in the data - and solutions - taken by the American designers for their successive projects (1970 to 1972) are shown. The MSBR power plant comprises three successive heat transfer circuits. The primary circuit (Hastelloy N), radioactive and polluted, containing the fuel salt, includes the reactor, pumps and exchangers. The secondary circuit (pipings made of modified Hastelloy N) contaminated in the exchanger, ensures the separation between the fuel and the fluid operating the turbo-alternator. The water-steam circuit feeds the turbine with steam. This steam is produced in the steam generator flowed by the secondary fluid. Some subsidiary circuits (discharge and storage of the primary and secondary salts, ventilation of the primary circuit ...) complete the three principal circuits which are briefly described. All circuits are enclosed inside the controlled-atmosphere building of the nuclear boiler. This building also ensures the biological protection and the mechanical protection against outer aggressions [fr

  3. The mechanical properties of T-111 at low to intermediate temperatures

    International Nuclear Information System (INIS)

    McCoy, H.E.; DiStefano, J.R.

    1997-01-01

    In the design of the 60-W Isotopic Heat Source (IHS), a tantalum alloy (T-111) strength member serves as the primary containment shell for the IHS during operation (He-gas internal environment and inert gas or vacuum external environment). An outer Hastelloy S clad is used to protect the T-111 from oxidation, and both the Hastelloy S clad and the T-111 strength member are sealed by automatic gas tungsten arc (GTA) welding. The expected life of the IHS is 5 years at about 650 C preceded by up to 5 years of storage at approximately 300 C. For this application, one important concern is failure of the T-111 strength member due to capsule pressurization arising from helium generation as a fuel decay product. To provide specific data on the mechanical behavior of base and solid metal T-111 under conditions appropriate to the IHS use conditions, a testing program was formulated and carried out. Three types of mechanical tests were conducted. Tensile properties were measured over the temperature range of 25 to 1100 C on T-111 base metal and samples with either longitudinal or transverse autogenous welds. Creep tests on base metal and samples with transverse welds were run to failure over the temperature range of 1100 to 850 C. Creep tests were also run on several transverse weld samples over the temperature range of 500 to 900 C at stresses where failure did not occur, and the creep rates were measured. Two prototypical capsules of the T-111 strength member were fabricated by EG and G Mound Applied Technologies (Mound Laboratories). To verify the mechanical properties design data developed above, these were tested to failure (leak) in a vacuum chamber with the inside of the capsule pressurized by either argon or helium

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-30

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

  5. Development of temperature statistical model when machining of aerospace alloy materials

    Directory of Open Access Journals (Sweden)

    Kadirgama Kumaran

    2014-01-01

    Full Text Available This paper presents to develop first-order models for predicting the cutting temperature for end-milling operation of Hastelloy C-22HS by using four different coated carbide cutting tools and two different cutting environments. The first-order equations of cutting temperature are developed using the response surface methodology (RSM. The cutting variables are cutting speed, feed rate, and axial depth. The analyses are carried out with the aid of the statistical software package. It can be seen that the model is suitable to predict the longitudinal component of the cutting temperature close to those readings recorded experimentally with a 95% confident level. The results obtained from the predictive models are also compared with results obtained from finite-element analysis (FEA. The developed first-order equations for the cutting temperature revealed that the feed rate is the most crucial factor, followed by axial depth and cutting speed. The PVD coated cutting tools perform better than the CVD-coated cutting tools in terms of cutting temperature. The cutting tools coated with TiAlN perform better compared with other cutting tools during the machining performance of Hastelloy C-22HS. It followed by TiN/TiCN/TiN and CVD coated with TiN/TiCN/Al2O3 and TiN/TiCN/TiN. From the finite-element analysis, the distribution of the cutting temperature can be discussed. High temperature appears in the lower sliding friction zone and at the cutting tip of the cutting tool. Maximum temperature is developed at the rake face some distance away from the tool nose, however, before the chip lift away.

  6. Research and Development programs for HTGRs in JAERI

    International Nuclear Information System (INIS)

    Nishiguchi, Isoharu; Saito, Sinzo

    1990-01-01

    Since 1969, JAERI has conducted research and development (R and D) programs for High-Temperature Gas-Cooled Reactors (HTGR). And the High Temperature engineering Test Reactor (HTTR), which will be the first High Temperature Gas Cooled Reactor (HTGR) in Japan, is under licensing process now. In this paper, some of the results of R and D are outlined in the following fields which are closely connected with the HTTR design, that is: i) fuel; ii) nuclear design; iii) thermal-hydraulic design; iv) graphite structure and v) high temperature metal structure. In the field of fuel, extensive investigations have been performed to develop the fabrication technology of coated particle fuel (cpf). In parallel, data of coated fuel particle failure and fission product release in in- and ex-reactor experiments as well as mechanical properties data were obtained and irradiation tests have been done using the Oarai Gas Loop No.1 (OGL-1) to verify the integrity of mass-produced fuel. Concerning the nuclear design, critical experiments were conducted using the Very High-Temperature Reactor Critical Assembly (VHTRC). Also carried out were hydrodynamical and thermal experiments using the Helium Engineering Demonstration Loop (HENDEL). On the graphite structures which compose the reactor internals, design criteria have been developed based on ASME BandPV Code Section III Div.2, subsection CE and design data have been accumulated on a domestic graphite material. High temperature metal structure is also one of major subjects of R and D for HTGRs. Hastelloy XR, which is a modified version of Hastelloy X, was developed and various tests have been conducted which include creep tests, creep-fatigue tests, etc. to establish design criteria and allowables. Component tests of the Intermediate Heat Exchanger (IHX) have been also performed. (author)

  7. Current status and future prospects of Japanese national project on coated conductor development and its applications

    Science.gov (United States)

    Shiohara, Y.; Yoshizumi, M.; Izumi, T.; Yamada, Y.

    2008-09-01

    Four years of the current five-year national project since 2003 for development of coated conductors using Y-system superconductors have passed and lots of remarkable results have been achieved. In this paper, the current status and the future prospect of this project are reviewed. The current national project comprises several groups of national laboratories, universities and private companies. The group of high performance tape development, consisting of Fujikura and SRL-NCCC, has worked on the tape by PLD-REBCO superconducting tapes on the PLD-CeO 2/IBAD-GZO buffered substrates. The high product of Ic and L equal to 112,166 A m was achieved in the 368 m-304.8 A GdBCO tape whose Ic value is mostly above 350 A/cm-w. The performance under the magnetic fields was also improved up to 42 A at 3 T in a GdBCO short film with doping of ZrO 2. About 61 m long GdBCO tape with ZrO 2 doping showed a high Ic value of 220 A at self field and 30 A at 3 T. On the other hand, the other group focusing on the low production cost has worked on processes of TFA-MOD and MOCVD, etc. The extremely high Ic value of 735 A/cm-w was attained in the TFA-MOD films on PLD-CeO 2/IBAD-GZO/Hastelloy C276 substrate by means of using the effect of Ba-poor nominal composition. In the efforts for long tape production, 200 m long tapes with high Ic values of 200 A/cm-w and 205 A/cm-w were obtained by MOD-YBCO/PLD-CeO 2/IBAD-GZO/Hastelloy C276 and PLD-HoBCO on buffered NiW substrate, respectively. The Ic × L value of the MOD-derived tape was 40,000 A m, which is the highest value in the world by the MOD process. Based on the above achievements on the coated conductor process development, two new additional goals were set in the project. One is the development for the extremely low cost tape and another is the development of the basic technologies for making the electric power devices including cables, transformers, motors, current-limiters and cryocoolers. Some of the new themes already revealed the

  8. Current status and future prospects of Japanese national project on coated conductor development and its applications

    Energy Technology Data Exchange (ETDEWEB)

    Shiohara, Y. [Superconductivity Research Laboratory, ISTEC, Shinonome 1-10-13, Koto-ku, Tokyo 135-0062 (Japan)], E-mail: shiohara@istec.or.jp; Yoshizumi, M.; Izumi, T. [Superconductivity Research Laboratory, ISTEC, Shinonome 1-10-13, Koto-ku, Tokyo 135-0062 (Japan); Yamada, Y. [Superconductivity Research Laboratory, ISTEC, Nagoya Coated Conductor Center, 2-4-1, Mutsuno, Atsuta-ku, Nagoya 456-8587 (Japan)

    2008-09-15

    Four years of the current five-year national project since 2003 for development of coated conductors using Y-system superconductors have passed and lots of remarkable results have been achieved. In this paper, the current status and the future prospect of this project are reviewed. The current national project comprises several groups of national laboratories, universities and private companies. The group of high performance tape development, consisting of Fujikura and SRL-NCCC, has worked on the tape by PLD-REBCO superconducting tapes on the PLD-CeO{sub 2}/IBAD-GZO buffered substrates. The high product of I{sub c} and L equal to 112,166 A m was achieved in the 368 m-304.8 A GdBCO tape whose I{sub c} value is mostly above 350 A/cm-w. The performance under the magnetic fields was also improved up to 42 A at 3 T in a GdBCO short film with doping of ZrO{sub 2}. About 61 m long GdBCO tape with ZrO{sub 2} doping showed a high I{sub c} value of 220 A at self field and 30 A at 3 T. On the other hand, the other group focusing on the low production cost has worked on processes of TFA-MOD and MOCVD, etc. The extremely high I{sub c} value of 735 A/cm-w was attained in the TFA-MOD films on PLD-CeO{sub 2}/IBAD-GZO/Hastelloy C276 substrate by means of using the effect of Ba-poor nominal composition. In the efforts for long tape production, 200 m long tapes with high I{sub c} values of 200 A/cm-w and 205 A/cm-w were obtained by MOD-YBCO/PLD-CeO{sub 2}/IBAD-GZO/Hastelloy C276 and PLD-HoBCO on buffered NiW substrate, respectively. The I{sub c} x L value of the MOD-derived tape was 40,000 A m, which is the highest value in the world by the MOD process. Based on the above achievements on the coated conductor process development, two new additional goals were set in the project. One is the development for the extremely low cost tape and another is the development of the basic technologies for making the electric power devices including cables, transformers, motors, current

  9. Current status and future prospects of Japanese national project on coated conductor development and its applications

    International Nuclear Information System (INIS)

    Shiohara, Y.; Yoshizumi, M.; Izumi, T.; Yamada, Y.

    2008-01-01

    Four years of the current five-year national project since 2003 for development of coated conductors using Y-system superconductors have passed and lots of remarkable results have been achieved. In this paper, the current status and the future prospect of this project are reviewed. The current national project comprises several groups of national laboratories, universities and private companies. The group of high performance tape development, consisting of Fujikura and SRL-NCCC, has worked on the tape by PLD-REBCO superconducting tapes on the PLD-CeO 2 /IBAD-GZO buffered substrates. The high product of I c and L equal to 112,166 A m was achieved in the 368 m-304.8 A GdBCO tape whose I c value is mostly above 350 A/cm-w. The performance under the magnetic fields was also improved up to 42 A at 3 T in a GdBCO short film with doping of ZrO 2 . About 61 m long GdBCO tape with ZrO 2 doping showed a high I c value of 220 A at self field and 30 A at 3 T. On the other hand, the other group focusing on the low production cost has worked on processes of TFA-MOD and MOCVD, etc. The extremely high I c value of 735 A/cm-w was attained in the TFA-MOD films on PLD-CeO 2 /IBAD-GZO/Hastelloy C276 substrate by means of using the effect of Ba-poor nominal composition. In the efforts for long tape production, 200 m long tapes with high I c values of 200 A/cm-w and 205 A/cm-w were obtained by MOD-YBCO/PLD-CeO 2 /IBAD-GZO/Hastelloy C276 and PLD-HoBCO on buffered NiW substrate, respectively. The I c x L value of the MOD-derived tape was 40,000 A m, which is the highest value in the world by the MOD process. Based on the above achievements on the coated conductor process development, two new additional goals were set in the project. One is the development for the extremely low cost tape and another is the development of the basic technologies for making the electric power devices including cables, transformers, motors, current-limiters and cryocoolers. Some of the new themes already

  10. Minimização de defeitos em revestimentos de superligas de níquel depositada pelo processo TIG com alimentação de arame frio

    Directory of Open Access Journals (Sweden)

    Cleiton Carvalho Silva

    2014-12-01

    Full Text Available O objetivo do presente trabalho foi avaliar a influência dos parâmetros de soldagem na formação de defeitos na soldagem de revestimentos com ligas à base de níquel, e sua possível eliminação através do correto ajuste dos referidos parâmetros. Para tanto, foram depositados revestimentos com as ligas à base de Ni do tipo Inconel 625, Hastelloy C276 e Inconel 686, sobre um substrato de aço C-Mn, através do processo TIG com alimentação de arame frio. O planejamento dos experimentos foi realizado aplicando-se o método Taguchi. Os fatores de controle avaliados foram a Técnica da energia (TE, o nível de energia de soldagem (E, o tipo de liga (L, o gás de proteção (G e o tipo de tecimento (T. Outros parâmetros foram mantidos constantes, tendo sido investigados previamente. Os resultados mostraram que o tipo de tecimento em espiral, embora contribua significativamente para a redução da diluição, causa uma forte instabilidade ao processo, resultando na maioria dos casos em defeitos superficiais ou defeitos entre passes. A condição ótima para evitar a formação de defeitos entre passes identificada pelo método Taguchi foi constituída pelas seguintes combinação de fatores de controle 2-2-2-3-3, ou seja: TE-I; Emédia; Liga Hastelloy C276; Gás de proteção Ar+He; Tecimento Duplo-8. A condição ótima para a soldagem sem defeitos resulta em alto nível de diluição não sendo indicada para a soldagem de revestimentos resistentes à corrosão.

  11. Characterization of complex carbide–silicide precipitates in a Ni–Cr–Mo–Fe–Si alloy modified by welding

    Energy Technology Data Exchange (ETDEWEB)

    Bhattacharyya, D., E-mail: dhb@ansto.gov.au; Davis, J.; Drew, M.; Harrison, R.P.; Edwards, L.

    2015-07-15

    Nickel based alloys of the type Hastelloy-N™ are ideal candidate materials for molten salt reactors, as well as for applications such as pressure vessels, due to their excellent resistance to creep, oxidation and corrosion. In this work, the authors have attempted to understand the effects of welding on the morphology, chemistry and crystal structure of the precipitates in the heat affected zone (HAZ) and the weld zone of a Ni–Cr–Mo–Fe–Si alloy similar to Hastelloy-N™ in composition, by using characterization techniques such as scanning and transmission electron microscopy. Two plates of a Ni–Cr–Mo–Fe–Si alloy GH-3535 were welded together using a TiG welding process without filler material to achieve a joint with a curved molten zone with dendritic structure. It is evident that the primary precipitates have melted in the HAZ and re-solidified in a eutectic-like morphology, with a chemistry and crystal structure only slightly different from the pre-existing precipitates, while the surrounding matrix grains remained unmelted, except for the zones immediately adjacent to the precipitates. In the molten zone, the primary precipitates were fully melted and dissolved in the matrix, and there was enrichment of Mo and Si in the dendrite boundaries after solidification, and re-precipitation of the complex carbides/silicides at some grain boundaries and triple points. The nature of the precipitates in the molten zone varied according to the local chemical composition. - Graphical abstract: Display Omitted - Highlights: • Ni-based alloy with Cr, Mo, Si, Fe and C was welded, examined with SEM, EBSD, and TEM. • Original Ni{sub 2}(Mo,Cr){sub 4}(Si,C) carbides changed from equiaxed to lamellar shape in HAZ. • Composition and crystal structure remained almost unchanged in HAZ. • Original carbides changed to lamellar Ni{sub 3}(Mo,Cr){sub 3}(Si,C) in some cases in weld metal. • Precipitates were mostly incoherent, but semi-coherent in some cases in weld

  12. Characterization of complex carbide–silicide precipitates in a Ni–Cr–Mo–Fe–Si alloy modified by welding

    International Nuclear Information System (INIS)

    Bhattacharyya, D.; Davis, J.; Drew, M.; Harrison, R.P.; Edwards, L.

    2015-01-01

    Nickel based alloys of the type Hastelloy-N™ are ideal candidate materials for molten salt reactors, as well as for applications such as pressure vessels, due to their excellent resistance to creep, oxidation and corrosion. In this work, the authors have attempted to understand the effects of welding on the morphology, chemistry and crystal structure of the precipitates in the heat affected zone (HAZ) and the weld zone of a Ni–Cr–Mo–Fe–Si alloy similar to Hastelloy-N™ in composition, by using characterization techniques such as scanning and transmission electron microscopy. Two plates of a Ni–Cr–Mo–Fe–Si alloy GH-3535 were welded together using a TiG welding process without filler material to achieve a joint with a curved molten zone with dendritic structure. It is evident that the primary precipitates have melted in the HAZ and re-solidified in a eutectic-like morphology, with a chemistry and crystal structure only slightly different from the pre-existing precipitates, while the surrounding matrix grains remained unmelted, except for the zones immediately adjacent to the precipitates. In the molten zone, the primary precipitates were fully melted and dissolved in the matrix, and there was enrichment of Mo and Si in the dendrite boundaries after solidification, and re-precipitation of the complex carbides/silicides at some grain boundaries and triple points. The nature of the precipitates in the molten zone varied according to the local chemical composition. - Graphical abstract: Display Omitted - Highlights: • Ni-based alloy with Cr, Mo, Si, Fe and C was welded, examined with SEM, EBSD, and TEM. • Original Ni 2 (Mo,Cr) 4 (Si,C) carbides changed from equiaxed to lamellar shape in HAZ. • Composition and crystal structure remained almost unchanged in HAZ. • Original carbides changed to lamellar Ni 3 (Mo,Cr) 3 (Si,C) in some cases in weld metal. • Precipitates were mostly incoherent, but semi-coherent in some cases in weld metal

  13. High-Temperature Structural Analysis of a Small-Scale PHE Prototype under the Test Condition of a Small-Scale Gas Loop

    International Nuclear Information System (INIS)

    Song, K.; Hong, S.; Park, H.

    2012-01-01

    A process heat exchanger (PHE) is a key component for transferring the high-temperature heat generated from a very high-temperature reactor (VHTR) to a chemical reaction for the massive production of hydrogen. The Korea Atomic Energy Research Institute has designed and assembled a small-scale nitrogen gas loop for a performance test on VHTR components and has manufactured a small-scale PHE prototype made of Hastelloy-X alloy. A performance test on the PHE prototype is underway in the gas loop, where different kinds of pipelines connecting to the PHE prototype are tested for reducing the thermal stress under the expansion of the PHE prototype. In this study, to evaluate the high-temperature structural integrity of the PHE prototype under the test condition of the gas loop, a realistic and effective boundary condition imposing the stiffness of the pipelines connected to the PHE prototype was suggested. An equivalent spring stiffness to reduce the thermal stress under the expansion of the PHE prototype was computed from the bending deformation and expansion of the pipelines connected to the PHE. A structural analysis on the PHE prototype was also carried out by imposing the suggested boundary condition. As a result of the analysis, the structural integrity of the PHE prototype seems to be maintained under the test condition of the gas loop.

  14. The reactive element effect of yttrium and yttrium silicon on high temperature oxidation of NiCrAl coating

    Science.gov (United States)

    Ramandhany, S.; Sugiarti, E.; Desiati, R. D.; Martides, E.; Junianto, E.; Prawara, B.; Sukarto, A.; Tjahjono, A.

    2018-03-01

    The microstructure formed on the bond coat affects the oxidation resistance, particularly the formation of a protective oxide layer. The adhesion of bond coat and TGO increased significantly by addition of reactive element. In the present work, the effect of yttrium and yttrium silicon as reactive element (RE) on NiCrAl coating was investigated. The NiCrAl (without RE) and NiCrAlX (X:Y or YSi) bond coating were deposited on Hastelloy C-276 substrate by High Velocity Oxygen Fuel (HVOF) method. Isothermal oxidation was carried out at 1000 °C for 100 hours. The results showed that the addition of RE could prevent the breakaway oxidation. Therefore, the coating with reactive element were more protective against high temperature oxidation. Furthermore, the oxidation rate of NiCrAlY coating was lower than NiCrAlYSi coating with the total mass change was ±2.394 mg/cm2 after 100 hours of oxidation. The thickness of oxide scale was approximately 1.18 μm consisting of duplex oxide scale of spinel NiCr2O4 in outer scale and protective α-Al2O3 in inner scale.

  15. PRE design of a molten salt thorium reactor loop

    International Nuclear Information System (INIS)

    Caire, Jean-Pierre; Roure, Anthony

    2007-01-01

    This study is a contribution to the 2004 PCR-RSF program of the Centre National de la Recherche Scientifique (CNRS) devoted to research on high temperature thorium molten salt reactors. A major issue of high temperature molten salt reactors is the very large heat duty to be transferred from primary to secondary loop of the reactor with minimal thermal losses. A possible inner loop made of a series of conventional graphite filter plate exchangers, pipes and pumps was investigated. The loop was assumed to use two counter current flows of the same LiF, BeF 2 , ZrF 4 , UF 4 molten salt flowing through the reactor. The 3D model used the coupling of k-ε turbulent Navier-Stokes equations and thermal applications of the Heat Transfer module of COMSOL Multiphysics. For a reactor delivering 2700 MWth, the model required a set of 114 identical exchangers. Each one was optimized to limit the heat losses to 2882 W. The pipes made of a succession of graphite, ceramics, Hastelloy-N alloy and insulating Microtherm layers led to a thermal loss limited to 550 W per linear meter. In such conditions, the global thermal losses represent only 0.013% of the reactor thermal power for elements covered with an insulator only 3 cm thick. (author)

  16. Technology Development of an Advanced Small-scale Microchannel-type Process Heat Exchanger (PHE) for Hydrogen Production in Iodine-sulfur Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Sah, Injin; Kim, Chan Soo; Kim, Yong Wan; Park, Jae-Won; Kim, Eung-Seon; Kim, Min-Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, ongoing manufacturing processes of the components employed in an advanced small-scale microchannel-type PHE are presented. The components, such as mechanically machined microchannels and a diffusion-bonded stack are introduced. Also, preliminary studies on surface treatment techniques for improving corrosion resistance from the corrosive sulfuric environment will be covered. Ongoing manufacturing process for an advanced small-size microchannel-type PHE in KAERI is presented. Through the preliminary studies for optimizing diffusion bonding condition of Hastelloy-X, a diffusion-bonded stack, consisting of primary and secondary side layer by layer, is scheduled to be fabricated in a few months. Also, surface treatment for enhancing the corrosion resistance from the sulfuric acid environment is in progress for the plates with microchannels. A massive production of hydrogen with electricity generation is expected in a Process Heat Exchanger (PHE) in a Very High Temperature gas-cooled Reactor (VHTR) system. For the application of hydrogen production, a small-scale gas loop for feasibility testing of a laboratory-scale has constructed and operated in Korea Atomic Energy Research Institute (KAERI) as a precursor to an experimental- and a pilot-scale gas loops.

  17. Unusual behavior in the reactivity of 5-substituted-1H-tetrazoles in a resistively heated microreactor

    Directory of Open Access Journals (Sweden)

    Dominique M. Roberge

    2011-04-01

    Full Text Available The decomposition of 5-benzhydryl-1H-tetrazole in an N-methyl-2-pyrrolidone/acetic acid/water mixture was investigated under a variety of high-temperature reaction conditions. Employing a sealed Pyrex glass vial and batch microwave conditions at 240 °C, the tetrazole is comparatively stable and complete decomposition to diphenylmethane requires more than 8 h. Similar kinetic data were obtained in conductively heated flow devices with either stainless steel or Hastelloy coils in the same temperature region. In contrast, in a flow instrument that utilizes direct electric resistance heating of the reactor coil, tetrazole decomposition was dramatically accelerated with rate constants increased by two orders of magnitude. When 5-benzhydryl-1H-tetrazole was exposed to 220 °C in this type of flow reactor, decomposition to diphenylmethane was complete within 10 min. The mechanism and kinetic parameters of tetrazole decomposition under a variety of reaction conditions were investigated. A number of possible explanations for these highly unusual rate accelerations are presented. In addition, general aspects of reactor degradation, corrosion and contamination effects of importance to continuous flow chemistry are discussed.

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

    International Nuclear Information System (INIS)

    Tamura, Manabu; Ogawa, Yutaka; Kurata, Yuji; Kondo, Tatsuo

    1982-04-01

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

  19. Research and development on chemical reactors made of industrial structural materials and hydriodic acid concentration technique for thermochemical hydrogen production IS process

    International Nuclear Information System (INIS)

    Kubo, Shinji; Iwatsuki, Jin; Takegami, Hiroaki; Kasahara, Seiji; Tanaka, Nobuyuki; Noguchi, Hiroki; Kamiji, Yu; Onuki, Kaoru

    2015-10-01

    Japan Atomic Energy Agency has been conducting a study on IS process for thermochemical hydrogen production in order to develop massive hydrogen production technology for hydrogen society. Integrity of the chemical reactors and concentration technology of hydrogen iodide in HIx solution were studied. In the former study, the chemical reactors were trial-fabricated using industrial materials. A test of 30 times of thermal cycle test under circulating condition of the Bunsen reaction solution showed integrity of the Bunsen reactor made of fluororesin lined steel. Also, 100 hours of reaction tests showed integrity of the sulfuric acid decomposer made of silicon carbide and of the hydrogen iodide decomposer made of Hastelloy C-276. In the latter study, concerning electro-electrodialysis using cation-exchange membrane, sulfuric acid in the anolyte had little influence on the concentration performance. These results suggest the purification system of HIx solution can be simplified. Based on the Nernst-Planck equation and the Smoluchowski equation, proton transport number, water permeance, and IR drop of the cation exchange membrane were formulated. The derived equations enable quantitative estimation for the performance indexes of Nafion ® membrane and, also, of ETFE-St membranes made by radiation-induced graft polymerization method. (author)

  20. Corrosion studies on selected metallic materials for application in nuclear waste disposal containers

    International Nuclear Information System (INIS)

    Smailos, E.; Fiehn, B.; Gago, J.A.; Azkarate, I.

    1994-03-01

    In previous corrosion studies, carbon steels and the alloy Ti 99.8-Pd were identified as promising materials for heat-generating nuclear waste containers acting as a radionuclide barrier in a rock-salt repository. To characterize the long-term corrosion behaviour of these materials in more detail, a research programme including laboratory-scale and in-situ corrosion studies has been undertaken jointly by KfK and ENRESA/INASMET. In the period under review, gamma irradiation corrosion studies of up to about 6 months at 10 Gy/h and stress corrosion cracking studies at slow strain rates (10 -4 -10 -7 s -1 ) were performed on three preselected carbon steels in disposal relevant brines (NaCl-rich, MgCl 2 -rich) at 90 C and 150 C (TStE 355, TStE 460, 15 MnNi 6.3). Moreover, results were obtained from long-term in-situ corrosion studies (maximum test duration 9 years) conducted on carbon steel, Ti 99.8-Pd, Hastelloy C4, Ni-resist D4, and Si-cast iron in boreholes in the Asse salt mine. (orig./MM) [de

  1. Effects of a range of machined and ground surface finishes on the simulated reactor helium corrosion of several candidate structural materials

    International Nuclear Information System (INIS)

    Thompson, L.D.

    1981-02-01

    This report discusses the corrosion behavior of several candidate reactor structural alloys in a simulated advanced high-temperature gas-cooled reactor (HTGR) environment over a range of lathe-machined and centerless-ground surface finishes. The helium environment contained 50 Pa H 2 /5 Pa CO/5 Pa CH 4 / 2 O (500 μatm H 2 /50 μatm CO/50 μatm CH 4 / 2 O) at 900 0 C for a total exposure of 3000 h. The test alloys included two vacuum-cast superalloys (IN 100 and IN 713LC); a centrifugally cast austenitic alloy (HK 40); three wrought high-temperature alloys (Alloy 800H, Hastelloy X, and Inconel 617); and a nickel-base oxide-dispersion-strengthened alloy (Inconel MA 754). Surface finish variations did not affect the simulated advanced-HTGR corrosion behavior of these materials. Under these conditions, the availability of reactant gaseous impurities controls the kinetics of the observed gas-metal interactions. Variations in the near-surface activities and mobilities of reactive solute elements, such as chromium, which might be expected to be affected by changes in surface finish, do not seem to greatly influence corrosion in this simulated advanced HTGR environment. 18 figures, 4 tables

  2. Changes in the properties of superalloys by long term heating

    International Nuclear Information System (INIS)

    Susukida, H.; Tsuji, I.; Kawai, H.

    1976-01-01

    A laboratory study was conducted in order to determine the effect of long term heating (max. 10000h at 850 0 and 950 0 C) on the microstructure, tensile properties, hardness and stress rupture properties of four kinds of superalloys. These superalloys are two kinds of solid solution hardened Ni-base superalloys Hastelloy X and Inconel 617 and two kinds of dispersion strengthened Ni-base superalloys TD-Ni and TD-NiCr. The result of the study can be summarized as follows: (1) Solid solution hardened superalloys: Many precipitates were observed in the grains and on the grain boundaries after 100 hours of heating, and the precipitates became coarse-grained by over 1000 hours of heating. This tendency was remarkable when they were heated at 950 0 C. With the change of their microstructure, their mechanical properties also changed, particularly their tensile ductility decreased remarkably. (2) Dispersion strengthened superalloys: Their microstructure and mechanical properties were almost unchanged by long term heating. (3) The authors proposed ''solid solution hardening value'' in order to grasp quantitatively the solid solution hardening which has been discussed by the content of each element hitherto. (auth.)

  3. Corrosion of several metals in supercritical steam at 5380C

    International Nuclear Information System (INIS)

    McCoy, H.E.; McNabb, B.

    1977-05-01

    The corrosion of several iron- and nickel-base alloys in supercritical steam at 24.1 MPa (3500 psi) and 538 0 C was measured to 7.92 x 10 7 s (22,000 h). The experiments were carried out in TVA's Bull Run Steam Plant. Corrosion was measured almost entirely by weight change and visual appearance; a few samples were evaluated by more descriptive analytical techniques. The corrosion rates of low-alloy ferritic steels containing from 1.1 to 8.7 percent Cr and 0.5 to 1.0 percent Mo differed by less than a factor of 2 in steam. Several modified compositions of Hastelloy N were evaluated and found to corrode at about equivalent rates. Of the alloys studied, the lowest weight gain in 3.6 x 10 7 sec (10,000 hr) was 0.01 mg/cm 2 for Inconel 718 and the highest 10 mg/cm 2 for the low-alloy ferritic steels. 25 figures, 3 tables

  4. Modeling the rubbing contact in honeycomb seals

    Science.gov (United States)

    Fischer, Tim; Welzenbach, Sarah; Meier, Felix; Werner, Ewald; kyzy, Sonun Ulan; Munz, Oliver

    2018-03-01

    Metallic honeycomb labyrinth seals are commonly used as sealing systems in gas turbine engines. Because of their capability to withstand high thermo-mechanical loads and oxidation, polycrystalline nickel-based superalloys, such as Hastelloy X and Haynes 214, are used as sealing material. In addition, these materials must exhibit a tolerance against rubbing between the rotating part and the stationary seal component. The tolerance of the sealing material against rubbing preserves the integrity of the rotating part. In this article, the rubbing behavior at the rotor-stator interface is considered numerically. A simulation model is incorporated into the commercial finite element code ABAQUS/explicit and is utilized to simulate a simplified rubbing process. A user-defined interaction routine between the contact surfaces accounts for the thermal and mechanical interfacial behavior. Furthermore, an elasto-plastic constitutive material law captures the extreme temperature conditions and the damage behavior of the alloys. To validate the model, representative quantities of the rubbing process are determined and compared with experimental data from the literature. The simulation results correctly reproduce the observations made on a test rig with a reference stainless steel material (AISI 304). A parametric study using the nickel-based superalloys reveals a clear dependency of the rubbing behavior on the sliding and incursion velocity. Compared to each other, the two superalloys studied exhibit a different rubbing behavior.

  5. Irreversible thermodynamics models and constitutive equations of the irradiation induced deformation and damage accumulating processes

    International Nuclear Information System (INIS)

    Wassilew, C.

    1989-11-01

    This report gives an overall evaluation of several in-reactor deformation and creep-rupture experiments performed in BR-2, FFTF, and Rapsodie on pressurised tubes of the stabilized austenitic stainless steels 1.4970, 1.4981, 1.4988, and the nickel base alloy Hastelloy-X. The irradiation induced deformation processes observed in the components operating in a neutron environment can be divided into two main groups: 1. volume conserving creep and 2. volumetric swelling. Since the observed deformation as well as damage accumulating phenomena are caused by the same constrained generated and free disposable point defects and helium atoms, it is obvious and advisable to analyze, and to model simultaneously the ensemble of the elementary mechanisms and processes effective at the same time. Phenomenological models based on the thermodynamics of irreversible processes have been developed, with the aim of: 1. grasping the partial relationships between the external variables and the response functions (creep, swelling, creep driven swelling, and time to rupture), 2. fathoming the rate-controlling mechanisms, 3. providing insight into the structural details and changes occurring during the deformation and the damage accumulating processes, 4. integrating the damage accumulating processes comprehensively, and 5. formulating the constitutive equations required to describe the elementary processes that generate plastic deformations as well as damage accumulation. (orig./MM)

  6. A high critical current density MOCVD coated conductor with strong vortex pinning centers suitable for very high field use

    International Nuclear Information System (INIS)

    Chen, Z; Kametani, F; Larbalestier, D C; Chen, Y; Xie, Y; Selvamanickam, V

    2009-01-01

    We have made extensive low temperature and high field evaluations of a recent 2.1 μm thick coated conductor (CC) grown by metal-organic chemical vapor deposition (MOCVD) with a view to its use for high field magnet applications, for which its very strong Hastelloy substrate makes it very suitable. This conductor contains dense three-dimensional (Y,Sm) 2 O 3 nanoprecipitates, which are self-aligned in planes tilted ∼7 deg. from the tape plane. Very strong vortex pinning is evidenced by high critical current density J c values of ∼3.1 MA cm -2 at 77 K and ∼43 MA cm -2 at 4.2 K, and by a strongly enhanced irreversibility field H irr , which reaches that of Nb 3 Sn (∼28 T at 1.5 K) at 60 K, even in the inferior direction of H parallel c axis. At 4.2 K, J c values are ∼15% of the depairing current density J d , much the highest of any superconductor suitable for magnet construction.

  7. A high critical current density MOCVD coated conductor with strong vortex pinning centers suitable for very high field use

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Z; Kametani, F; Larbalestier, D C [National High Magnetic Field Laboratory, Florida State University, Tallahassee, FL 32310 (United States); Chen, Y; Xie, Y; Selvamanickam, V [SuperPower Incorporated, Schenectady, NY 12304 (United States)], E-mail: zhijun@asc.magnet.fsu.edu

    2009-05-15

    We have made extensive low temperature and high field evaluations of a recent 2.1 {mu}m thick coated conductor (CC) grown by metal-organic chemical vapor deposition (MOCVD) with a view to its use for high field magnet applications, for which its very strong Hastelloy substrate makes it very suitable. This conductor contains dense three-dimensional (Y,Sm){sub 2}O{sub 3} nanoprecipitates, which are self-aligned in planes tilted {approx}7 deg. from the tape plane. Very strong vortex pinning is evidenced by high critical current density J{sub c} values of {approx}3.1 MA cm{sup -2} at 77 K and {approx}43 MA cm{sup -2} at 4.2 K, and by a strongly enhanced irreversibility field H{sub irr}, which reaches that of Nb{sub 3}Sn ({approx}28 T at 1.5 K) at 60 K, even in the inferior direction of H parallel c axis. At 4.2 K, J{sub c} values are {approx}15% of the depairing current density J{sub d}, much the highest of any superconductor suitable for magnet construction.

  8. Fission product behavior in high-temperature water: CsI vs MoO4

    Science.gov (United States)

    Kanjana, K.; Silva, K.; Channuie, J.

    2017-09-01

    Fission product behaviors of Cs, a major element released in a severe nuclear accident, still remain unclear. The question frequently addressed is whether Cs released will be in the form of Cs2MoO4 or CsOH. This is a challenging issue since it has been demonstrated that the reaction between Cs2MoO4 and water leading to CsOH production is thermodynamically favored. The present research aims at investigation of CsOH generation through this chemical channel. A high-temperature setup with a flow system based on the cooling system of a water-cooled nuclear reactor has been assembled. The reaction between aqueous solutions of CsI and Na2MoO4 in a high-corrosion-resistant hot cell (Hastelloy) has been studied up to 80 °C in deoxygenated system. The products have been characterized using FTIR and XRD. The results have shown that there is no reaction between CsI and Na2MoO4 under the experimental conditions.

  9. Remarkable weakness against cleavage stress for YBCO-coated conductors and its effect on the YBCO coil performance

    International Nuclear Information System (INIS)

    Yanagisawa, Y.; Nakagome, H.; Takematsu, T.; Takao, T.; Sato, N.; Takahashi, M.; Maeda, H.

    2011-01-01

    Cleavage strength for YBCO-coated conductor is extremely low, typically 0.5 MPa. The remarkable weakness is due to cracks on the slit edge of the conductor. The cleavage stress appears on YBCO double pancake coils impregnated with epoxy. The cleavage stress should be avoided in the coil winding. Cleavage strength for an YBCO-coated conductor at 77 K was investigated with a model experiment. The nominal cleavage strength for an YBCO-coated conductor is extremely low, typically 0.5 MPa. This low nominal cleavage strength is due to stress concentration on a small part of the YBCO-coated conductor in cleavage fracture. Debonding by the cleavage stress occurs at the interface between the buffer layer and the Hastelloy substrate. The nominal cleavage strength for a slit edge of the conductor is 2.5-times lower than that for the original edge of the conductor; cracks and micro-peel existing over the slit edge reduce the cleavage strength for the slit edge. Cleavage stress and peel stress should be avoided in coil winding, as they easily delaminate the YBCO-coated conductor, resulting in substantial degradation of coil performance. These problems are especially important for epoxy impregnated YBCO-coated conductor coils. It appears that effect of cleavage stress and peel stress are mostly negligible for paraffin impregnated YBCO-coated conductor coils or dry wound YBCO-coated conductor coils.

  10. Application of the laser spallation technique to the measurement of the adhesion strength of tungsten carbide coatings on superalloy substrates

    Energy Technology Data Exchange (ETDEWEB)

    Boustie, M.; Aoroux, E.; Romain, J.-P. [Ecole Nationale Superieure de Mecanique et d' Aerotechnique (ENSMA), 86 - Futuroscope (FR). Lab. de Combustion et de Detonique (LCD)

    2000-10-01

    High power pulsed lasers are used to induce shock waves in Hastelloy X targets coated with tungsten carbide of 70 {mu}m and 50 {mu}m thickness. In suitable irradiation conditions, a debonding of the substrate/coating interface due to the generation of tensile stresses is observed. Experimental results are analyzed with the use of numerical simulations yielding the stress history at interface and its dependence on laser pulse intensity up to 600 GW/cm{sup 2} with 1 ns and 3 ns durations under direct irradiation, and 23 ns with water confinement. As a consequence of shock decay during the propagation through the substrate, a strong variation of incident intensity results in a small variation of tensile stress. This allows an accurate determination of the debonding threshold which is found in the range of 1.0 to 1.3 GPa for short laser pulses (1 and 3 ns) and 0.5 to 0.6 GPa for long laser pulses (23 ns confined). (orig.)

  11. Effect of Al added to a NiCrMo alloy on the development of the oxide layer of intermetallic coatings

    International Nuclear Information System (INIS)

    D'Oliveira, A.S.C.M.; Cangue, F.J.R.

    2010-01-01

    Components performance in different environment is strongly dependent on oxides that develop on their surfaces. This study analyzed the oxide layer that develops on coatings processed with mixtures of an atomized Hastelloy C alloy with Al powders. Powder mixtures containing 10, 20 and 30wt%Al were deposited on AISI 1020 and AISI304 steel plates. Coatings were subsequently exposed to 850 deg C for two hours in a low PO 2 environment. X-ray diffraction was used to identify the phases that developed in the coating during processing and Raman analysis and Scanning Electron Microscopy were used to characterize the oxide layers. The results showed that coatings processed with the richer Al mixtures, 30wt%Al, which developed NiAl aluminides, reduced the development of α alumina when processing was done on AISI 304. Coatings processed on AISI 1020 with the three powder mixtures tested developed the different allotropic forms of alumina, as predicted for the tested temperature. (author)

  12. Improved interface growth and enhanced flux pinning in YBCO films deposited on an advanced IBAD-MgO based template

    Science.gov (United States)

    Khan, M. Z.; Zhao, Y.; Wu, X.; Malmivirta, M.; Huhtinen, H.; Paturi, P.

    2018-02-01

    The growth mechanism is studied from the flux pinning point of view in small-scale YBa2Cu3O6+x (YBCO) thin films deposited on a polycrystalline hastelloy with advanced IBAD-MgO based buffer layer architecture. When compared the situation with YBCO films grown on single crystal substrates, the most critical issues that affect the suitable defect formation and thus the optimal vortex pinning landscape, have been studied as a function of the growth temperature and the film thickness evolution. We can conclude that the best critical current property in a wide applied magnetic field range is observed in films grown at relatively low temperature and having intermediate thickness. These phenomena are linked to the combination of the improved interface growth, to the film thickness related crystalline relaxation and to the formation of linear array of edge dislocations that forms the low-angle grain boundaries through the entire film thickness and thus improve the vortex pinning properties. Hence, the optimized buffer layer structure proved to be particularly suitable for new coated conductor solutions.

  13. An experimental study of commercially available alloys for potential use in the CANDU®-SCWR

    Energy Technology Data Exchange (ETDEWEB)

    Cook, W.; Miles, J.; Bradley, C. [Univ. of New Brunswick, Dept. of Chemical Engineering, Fredericton, New Brunswick (Canada); Li, J.; Zheng, W. [CANMET-MTL, Hamilton, Ontario (Canada)

    2010-07-01

    In order to study the corrosion of candidate alloys in conditions simulating those proposed for the CANDU®-SCWR, a dynamic test loop has been constructed in the nuclear laboratories in the Department of Chemical Engineering at UNB. For this review, the following materials were selected for analysis: Inconel 625 (I625), Hastelloy C 276 (HC276) and 304SS. I625 and HC276 are members of a class of 'superalloys' comprised primarily of nickel and chromium, which are suitable for high temperature applications and 304SS is a commonly used construction material in current reactors. Coupons were cut from each material and polished to an 800 grit finish. Experiments were conducted at 400{sup o}C, 500{sup o}C and 550{sup o}C at 25MPa with exposure times of 100, 250, and 500 hours to examine the corrosion rate and oxide formation for each of these materials. Preliminary results indicate minimal weight change for both I625 and HC276, while small weight gains were observed for the 304SS samples. These results are supported with SEM work, which shows minimal oxidation on the I625 and HC276 samples. The 304SS samples also showed minimal oxide growth characterized by localized attack on the coupon surfaces. (author)

  14. Corrosion of stainless steels and nickel-base alloys in solutions of nitric acid and hydrofluoric acid

    International Nuclear Information System (INIS)

    Horn, E.M.; Renner, M.

    1992-01-01

    Reactions involving nitric acid may always result in the contamination of this acid with fluorides. In highly concentrted nitric acid, the presence of small amounts of HF will substantially reduce the corrosion of metallic materials. Mixtures consisting of hydrofluoric acid and hypo-azeotropic nitric acid on the other hand will strongly attack: the metal loss will markedly increase with increasing HNO 3 and HF concentrations as well as with rising temperatures. The investigation covered 12 stainless steel grades and nickel-base alloys. With constant HNO 3 content, corrosion rates will rise linearly when increasing the HF concentration. With constant HF concentration (0,25 M), corrosion rates will increase rapidly with increasing nitric acid concentration (from 0.3 M to 14.8 M). This can best be described by superimposing a linear function and a hyperbolic function that is reflecting the change in the HNO 3 content. Alloys containing as much chromium as possible (up to 46 wt.%) will exhibit the best corrosion resistance. Alloy NiCr30FeMo (Hastelloy alloy G-30) proved to be well suitable in this investigation. (orig.) [de

  15. Molten-salt converter reactors

    International Nuclear Information System (INIS)

    Perry, A.M.

    1975-01-01

    Molten-salt reactors appear to have substantial promise as advanced converters. Conversion ratios of 0.85 to 0.9 should be attainable with favourable fuel cycle costs, with 235 U valued at $12/g. An increase in 235 U value by a factor of two or three ($10 to $30/lb. U 3 O 8 , $75/SWU) would be expected to increase the optimum conversion ratio, but this has not been analyzed in detail. The processing necessary to recover uranium from the fuel salt has been partially demonstrated in the MSRE. The equipment for doing this would be located at the reactor, and there would be no reliance on an established recycle industry. Processing costs are expected to be quite low, and fuel cycle optimization depends primarily on inventory and burnup or replacement costs for the fuel and for the carrier salt. Significant development problems remain to be resolved for molten-salt reactors, notably the control of tritium and the elimination of intergranular cracking of Hastelloy-N in contact with tellurium. However, these problems appear to be amenable to solution. It is appropriate to consider separating the development schedule for molten-salt reactors from that for the processing technology required for breeding. The Molten-Salt Converter Reactor should be a useful reactor in its own right and would be an advance towards the achievement of true breeding in thermal reactors. (author)

  16. Effect of Molybdenum on the Corrosion Behavior of High-Entropy Alloys CoCrFeNi2 and CoCrFeNi2Mo0.25 under Sodium Chloride Aqueous Conditions

    Directory of Open Access Journals (Sweden)

    Alvaro A. Rodriguez

    2018-01-01

    Full Text Available The corrosion behavior of high-entropy alloys (HEAs CoCrFeNi2 and CoCrFeNi2Mo0.25 was investigated in 3.5 wt. percent sodium chloride (NaCl at 25°C by electrochemical methods. Their corrosion parameters were compared to those of HASTELLOY® C-276 (UNS N10276 and stainless steel 316L (UNS 31600 to assess the suitability of HEAs for potential industrial applications in NaCl simulating seawater type environments. The corrosion rates were calculated using corrosion current determined from electrochemical experiments for each of the alloys. In addition, potentiodynamic polarization measurements can indicate active, passive, and transpassive behavior of the metal as well as potential susceptibility to pitting corrosion. Cyclic voltammetry (CV can confirm the alloy susceptibility to pitting corrosion. Electrochemical impedance spectroscopy (EIS elucidates the corrosion mechanism under studied conditions. The results of the electrochemical experiments and scanning electron microscopy (SEM analyses of the corroded surfaces revealed general corrosion on alloy CoCrFeNi2Mo0.25 and HASTELLOY C-276 and pitting corrosion on alloy CoCrFeNi2 and stainless steel 316L.

  17. Some Investigations of the Reaction of Activated Charcoal with Fluorine and Uranium Hexafluoride

    Energy Technology Data Exchange (ETDEWEB)

    Del Cul, G.D.; Fiedor, J.N.; Simmons, D.W.; Toth, L.M.; Trowbridge, L.D.; Williams

    1998-09-01

    The Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory has been shut down since 1969, when the fuel salt was drained from the core into two Hastelloy N drain tanks at the reactor site. Over time, fluorine (F{sub 2}) and uranium hexafluoride (UF{sub 6}) moved from the salt through the gas piping to a charcoal bed, where they reacted with the activated charcoal. Some of the immediate concerns related to the migration of F{sub 2} and UF{sub 6} to the charcoal bed were the possibility of explosive reactions between the charcoal and F{sub 2}, the existence of conditions that could induce a criticality accident, and the removal and recovery of the fissile uranium from the charcoal. This report addresses the reactions and reactivity of species produced by the reaction of fluorine and activated charcoal and between charcoal and F{sub 2}-UF{sub 6} gas mixtures in order to support remediation of the MSRE auxiliary charcoal bed (ACB) and the recovery of the fissile uranium. The chemical identity, stoichiometry, thermochemistry, and potential for explosive decomposition of the primary reaction product, fluorinated charcoal, was determined.

  18. In hydrofluoric acid corrosion-resistant materials

    International Nuclear Information System (INIS)

    Hauffe, K.

    1985-01-01

    Copper, red brass (Cu-15 Zn), special treated carbon steel and chromium-nickel-molybdenum steel represent materials of high resistivity against concentrated hydrofluoric acid ( 2 O 3 ) are employed for windows in the presence of hydrogen fluoride and/or hydrofluoric acid because of their superior optical properties and their excellent corrosion resistance. Polyethylen, polypropylene and polyvinyl chloride (PVC) belong to the cheapest corrosion resistant material for container and for coatings in the presence of hydrofluoric acid. Special polyester resins reinforced by glass or graphite fibers have been successfully employed as material for production units with hydrofluoric acid containing liquids up to 330 K. By carbon reinforced epoxy resin represents a corrosion resistant coating. Because of their excellent friction and corrosion resistance against concentrated hot hydrofluoric acid and HNO 3 -HF-solutions, PTFE and polyvinylidene fluoride are used as material for valves and axles in such environment. The expensive alloys, as for instance hastelloy and monel, are substituted more and more by fiber-reinfored polyolefins, PVC and fluorine containing polymers. (orig.) [de

  19. Compatibility of molten salt and structural materials

    International Nuclear Information System (INIS)

    Kawakami, Masahiro

    1994-01-01

    As the important factors for considering the compatibility of fuel salt and coolant salt with structural materials in molten salt reactors, there are the moisture remaining in molten salt and the fluorine potential in molten salt. In this study, as for the metals which are the main components of corrosion resistant alloys, the corrosion by the moisture remaining in molten salt and the dependence of the corrosion on fluorine potential were examined. As the molten salts, an eutectic molten salt LiF-BeF 2 was mainly used, and LiF-KF was used in combination. As the metallic materials, Cr, Ni and Cu which are the main components of corrosion resistant and heat resistant alloys, Hastelloy and Monel, were used. In the experiment, the metal pieces were immersed in the molten salt, and by sampling the molten salt, the change with time lapse of the concentration of the dissolved metals was examined. Besides, the electrochemical measurement was carried out for Cr, of which the corrosion was remarkable, and the change with time lapse of the dissolved ions was examined. The experimental setup, the experimental method, and the results of the immersion test and the electrochemical test are reported. The experiment on the corrosion of metals depending on fluorine potential is also reported. (K.I.)

  20. Non-ferrous metals, anorganic and organic materials resistent to fluorides

    International Nuclear Information System (INIS)

    Hauffe, K.

    1986-01-01

    Aluminium and its alloys are resistant in fluoride solutions up to 400 K. Aluminium is also a suitable reactor material for the thermal decomposition of acidic fluorides between 750 and 825 K. Brass corrodes at room temperature in a 0,1 m KF solution with and without inhibitors very slowly ( -1 ). Nickel and the nickel alloys Inconel 600, Hastelloy N and Monel 500 are the most resistant materials against fluoride solutions and melts. A similar behavior exhibit zirconium-titanium-iron and zirconium-titanium-molybdenum alloys, respectively. From the inorganic compounds, compressed graphite, Al 2 O 3 and hexaborides of earth and rare earth metals, particularly LaB 6 , are extraordinarily resistant against fluorine ions at high temperatures. If the reaction temperature remains below 370 K, then polymers and resins, e.g. polyolefines, PVC, acrylic and epoxy resins and fluorcarbon resins can be employed as coating or compound material (resin + carbon fibers) resistant against fluorine ions up to 370 K. (orig.) [de

  1. Superconducting properties of iron chalcogenide thin films

    Directory of Open Access Journals (Sweden)

    Paolo Mele

    2012-01-01

    Full Text Available Iron chalcogenides, binary FeSe, FeTe and ternary FeTexSe1−x, FeTexS1−x and FeTe:Ox, are the simplest compounds amongst the recently discovered iron-based superconductors. Thin films of iron chalcogenides present many attractive features that are covered in this review, such as: (i easy fabrication and epitaxial growth on common single-crystal substrates; (ii strong enhancement of superconducting transition temperature with respect to the bulk parent compounds (in FeTe0.5Se0.5, zero-resistance transition temperature Tc0bulk = 13.5 K, but Tc0film = 19 K on LaAlO3 substrate; (iii high critical current density (Jc ~ 0.5 ×106 A cm2 at 4.2 K and 0 T for FeTe0.5Se0.5 film deposited on CaF2, and similar values on flexible metallic substrates (Hastelloy tapes buffered by ion-beam assisted deposition with a weak dependence on magnetic field; (iv high upper critical field (~50 T for FeTe0.5Se0.5, Bc2(0, with a low anisotropy, γ ~ 2. These highlights explain why thin films of iron chalcogenides have been widely studied in recent years and are considered as promising materials for applications requiring high magnetic fields (20–50 T and low temperatures (2–10 K.

  2. Remarkable progress in fabricating RE123 coated conductors by IBAD/PLD technique at Fujikura

    Energy Technology Data Exchange (ETDEWEB)

    Igarashi, M; Kakimoto, K; Hanyu, S; Tashita, C; Hayashida, T; Hanada, Y; Fujita, S; Morita, K; Nakamura, N; Sutoh, Y; Kutami, H; Iijima, Y; Saitoh, T, E-mail: m_igarashi@fujikura.co.j [Fujikura Ltd., 1440, Mutsuzaki, Sakura, Chiba, 285-8550 (Japan)

    2010-06-01

    Increase of production rate and improvement of quality for RE123 coated conductors have been tried. In-plane texturing of MgO was attempted by the IBAD system with the world largest ion source. As a result of optimizing condition in large deposition area, the dramatically high throughput of 1000 m / h was realized to obtain the IBAD-MgO with {Delta}{phi} < 10{sup 0}. Furthermore, simple buffer structure was demonstrated. Well textured CeO{sub 2} layer with {Delta}{phi} of around 4{sup 0} was obtained by directly deposited on IBAD-MgO layer in spite of large lattice mismatch of 28% between CeO{sub 2} and MgO. Several over 100 m buffer substrates with the architecture of / PLD-CeO{sub 2} (60 m / h) / IBAD-MgO (333{approx}1000 m / h) / Y{sub 2}O{sub 3} (500 m / h) / Al{sub 2}O{sub 3} (150 m / h) / Hastelloy / were already prepared. On these production substrates, GdBCO layer was deposited by the large PLD system at high throughput. The 260 m long GdBCO tape with I{sub c} > 600 A except some locations was obtained at the throughput of 15 m / h. In addition to the speed-up, the very high I{sub c} of 1040 A was also achieved by the hot-wall heating PLD system.

  3. Consideration on evaluation of internal pressure creep rupture for tube with circumferential joint

    International Nuclear Information System (INIS)

    Nagato, Kotaro; Satoh, Keisuke

    1983-01-01

    The behavior of internal pressure creep rupture of the thin-walled cylinders with circumferential joints is affected by the combination of creep characteristics of parent materials and weld metals. In particular, the compatibility of the creep strain rate of parent materials and weld metals becomes an important controlling factor. The behavior of internal pressure creep of the welded parts in circumferential joint cylinders can be evaluated simply with the uniaxial creep data of parent materials and weld metals, considering it by approximately substituting with the creep behavior of a uniaxial longitudinal joint. The method of evaluation is, first, to analyze the breaking behavior of uniaxial longitudinal joints using the uniaxial creep characteristic values of parent materials and weld metals, and next, by combining the equation for the relation between the rupture times of uniaxial creep and internal pressure creep with the analyzed breaking behavior of uniaxial joints, the internal pressure creep rupture behavior of the cylinders with circumferential joints can be evaluated. The internal pressure creep behavior of the thin-walled cylinders with circumferential joints, their rupture life and the uniaxial creep rupture life of longitudinal joints, and the examination of Hastelloy X cylinders are reported. (Kako, I.)

  4. Effect of Al added to a NiCrMo alloy on the development of the oxide layer of intermetallic coatings; Efeito do teor de Al adicionado a liga NiCrMo no desenvolvimento dos filmes de oxidos em revestimentos intermetalicos

    Energy Technology Data Exchange (ETDEWEB)

    D' Oliveira, A.S.C.M.; Cangue, F.J.R. [Universidade Federal do Parana (DEM/UFPR), Curitiba, PR (Brazil). Dept. de Engenharia Mecanica; Clark, E.; Levi, C. [University of California, Santa Barbara, CA (United States)

    2010-07-01

    Components performance in different environment is strongly dependent on oxides that develop on their surfaces. This study analyzed the oxide layer that develops on coatings processed with mixtures of an atomized Hastelloy C alloy with Al powders. Powder mixtures containing 10, 20 and 30wt%Al were deposited on AISI 1020 and AISI304 steel plates. Coatings were subsequently exposed to 850 deg C for two hours in a low PO{sub 2} environment. X-ray diffraction was used to identify the phases that developed in the coating during processing and Raman analysis and Scanning Electron Microscopy were used to characterize the oxide layers. The results showed that coatings processed with the richer Al mixtures, 30wt%Al, which developed NiAl aluminides, reduced the development of {alpha} alumina when processing was done on AISI 304. Coatings processed on AISI 1020 with the three powder mixtures tested developed the different allotropic forms of alumina, as predicted for the tested temperature. (author)

  5. Investigation of corrosion and analysis of passive films concerning some nickel alloys and stainless steels in reconstructed geological environments

    International Nuclear Information System (INIS)

    Jallerat, Nelly

    1984-01-01

    This research thesis addresses the corrosion behaviour of materials which might be used for the fabrication of radioactive waste containers. After a bibliographical study on films formed on Fe-Cr-Ni alloys, this research concentrates on passivation and de-passivation phenomena of three nickel-base alloys among the most resistant to corrosion and which also meet processing and economic criteria: Hastelloy C4, Inconel 625 and ZICNDU 25-20. Titanium and Ti-Pd alloy are also studied. Parameters governing pitting corrosion are notably studied. After a recall of knowledge on passive films formed on Fe-Cr-Ni alloys, and a presentation of experimental and technical conditions, the author reports and discussed the results obtained by electrochemical studies, reports the determination of factors governing alloy passivation in geological waters. The influence of some soluble impurities is notably studied. The author reports the analysis by glow discharge optical emission spectrometry to determine the composition of passive films with respect to geological water nature, the immersion duration and the electrode potential. Additional surface analyses are performed by X-ray photoelectron spectrometry (XPS or ESCA) and secondary ion mass spectrometry (SIMS). Finally, the author uses a dosing method by neutron radio-activation of alloy elements to determine dissolution mechanisms [fr

  6. Unusual behavior in the reactivity of 5-substituted-1H-tetrazoles in a resistively heated microreactor

    Science.gov (United States)

    Gutmann, Bernhard; Glasnov, Toma N; Razzaq, Tahseen; Goessler, Walter; Roberge, Dominique M

    2011-01-01

    Summary The decomposition of 5-benzhydryl-1H-tetrazole in an N-methyl-2-pyrrolidone/acetic acid/water mixture was investigated under a variety of high-temperature reaction conditions. Employing a sealed Pyrex glass vial and batch microwave conditions at 240 °C, the tetrazole is comparatively stable and complete decomposition to diphenylmethane requires more than 8 h. Similar kinetic data were obtained in conductively heated flow devices with either stainless steel or Hastelloy coils in the same temperature region. In contrast, in a flow instrument that utilizes direct electric resistance heating of the reactor coil, tetrazole decomposition was dramatically accelerated with rate constants increased by two orders of magnitude. When 5-benzhydryl-1H-tetrazole was exposed to 220 °C in this type of flow reactor, decomposition to diphenylmethane was complete within 10 min. The mechanism and kinetic parameters of tetrazole decomposition under a variety of reaction conditions were investigated. A number of possible explanations for these highly unusual rate accelerations are presented. In addition, general aspects of reactor degradation, corrosion and contamination effects of importance to continuous flow chemistry are discussed. PMID:21647324

  7. Chemistry

    International Nuclear Information System (INIS)

    Ferris, L.M.

    1975-01-01

    The chemical research and development efforts related to the design and ultimate operation of molten-salt breeder reactor systems are concentrated on fuel- and coolant-salt chemistry, including the development of analytical methods for use in these systems. The chemistry of tellurium in fuel salt is being studied to help elucidate the role of this element in the intergranular cracking of Hastelloy N. Studies were continued of the effect of oxygen-containing species on the equilibrium between dissolved UF 3 and dissolved UF 4 , and, in some cases, between the dissolved uranium fluorides and graphite, and the UC 2 . Several aspects of coolant-salt chemistry are under investigation. Hydroxy and oxy compounds that could be formed in molten NaBF 4 are being synthesized and characterized. Studies of the chemistry of chromium (III) compounds in fluoroborate melts were continued as part of a systematic investigation of the corrosion of structural alloys by coolant salt. An in-line voltammetric method for determining U 4+ /U 3+ ratios in fuel salt was tested in a forced-convection loop over a six-month period. (LK)

  8. A thermoelectric-conversion power supply system using a strontium heat source of high-level radioactive nuclear waste

    International Nuclear Information System (INIS)

    Chikazawa, Yoshitaka

    2011-01-01

    A thermoelectric-conversion power supply system with radioactive strontium in high-level radioactive waste has been proposed. A combination of Alkali Metal Thermo-Electric Conversion (AMTEC) and a strontium fluoride heat source can provide a compact and long-lived power supply system. A heat source design with strontium fluoride pin bundles with Hastelloy cladding and intermediate copper has been proposed. This design has taken heat transportation into consideration, and, in this regard, the feasibility has been confirmed by a three-dimensional thermal analysis using Star-CD code. This power supply system with an electric output of 1 MW can be arranged in a space of 50 m 2 and approximately 1.1 m height and can be operated for 15 years without refueling. This compact and long-lived power supply is suitable for powering sources for remote places and middle-sized ships. From the viewpoint of geological disposal of high-level waste, the proposed power supply system provides a financial base for strontium-cesium partitioning. That is, a combination of minor-actinide recycling and strontium-cesium partitioning can eliminate a large part of decay heat in high-level waste and thus can save much space for geological disposal. (author)

  9. Introduction of Artificial Pinning Center into PLD-YBCO Coated Conductor on IBAD and Self-Epitaxial CeO2 Buffered Metal Substrate

    International Nuclear Information System (INIS)

    Kobayashi, H.; Yamada, Y.; Ishida, S.; Takahashi, K.; Konishi, M.; Ibi, A.; Miyata, S.; Kato, T.; Hirayama, T.; Shiohara, Y.

    2006-01-01

    In order to fabricate YBa2Cu3O7-x (YBCO) coated conductors with high critical current density Jc in magnetic fields, we fabricated YBCO coated conductors with artificial pinning centers by the pulsed laser deposition (PLD) method on a self epitaxial PLD-CeO2 layer and ion-beam assisted deposition (IBAD)-Gd2Zr2O7 (GZO) buffered Hastelloy tape. Artificial pinning centers were introduced by the PLD deposition using the yttria-stabilized zirconia (YSZ) oxide target (nano-dot method) and YBCO target including YSZ particles (mixed target method). In the experiments using YSZ oxide target, YSZ nano-dots were observed. They were approximately 15 nm in height and 10 nm to 70 nm in diameter. We found that the density of nano-dots was controlled by the number of laser pulses. These samples exhibited higher Jc than YBCO films in magnetic fields. Furthermore, a similar improvement of Jc was observed in the experiments using YBCO target including YSZ particles. TEM observation revealed that columnar nano-structure made of BaZrO3 was formed during YBCO deposition and it was effective for pinning. We call this new epitaxial nano-structure 'bamboo structure' from its anisotropic growth and morphology

  10. Localized corrosion of metallic materials and γ radiation effects in passive layers under simulated radwaste repository conditions. Final report

    International Nuclear Information System (INIS)

    Schultze, J.W.; Kudelka, S.; Michaelis, A.; Schweinsberg, M.; Thies, A.

    1996-02-01

    The task of the project was to simulate the conditions in a radwaste repository and to perform local analyses in order to detect the critical conditions and material susceptibilities leading to localized corrosion of materials. The information thus obtained was to yield more precise data on the long-term stability of materials for the intended purpose, in order to be able to appropriately select or optimize the materials (Ti, TiO.2Pd, Hastelloy C4, fine-grained structural steel). A major aspect to be examined was natural inhomogeneities of the electrode surfaces, as determined by the grain structure of the selected materials. Thus a laterally inhomogeneous composition in the welded zone induces an inhomogeneous current distribution, and hence strong susceptibility to localized corrosion. This effect was to be quantified, and the localized corrosion processes had to be identified by means of novel, electrochemical methods with a resolution power of μm. The investigations were to be made under conditions as near to practice as possible, for instance by simulating radwaste repository conditions and performing measurements at elevated temperatures (170 C) in an autoclave. Another task was to examine the radiation effects of γ radiation on passive layers, and describe the possible modifications induced by recrystallisation, photocorrosion, or oxide formation. (orig./MM) [de

  11. Preparation of SmBCO layer for the surface optimization of GdYBCO film by MOCVD process based on a simple self-heating technology

    Science.gov (United States)

    Zhao, Ruipeng; Zhang, Fei; Liu, Qing; Xia, Yudong; Lu, Yuming; Cai, Chuanbing; Tao, Bowan; Li, Yanrong

    2018-07-01

    The MOCVD process was adopted to grow the REBa2Cu3O7-δ ((REBCO), RE = rare earth elements) films on the LaMnO3 (LMO) templates. Meanwhile, the LMO-template tapes are heated by the joule effect after applying a heating current through the Hastelloy metal substrates. The surface of GdYBCO films prepared by MOCVD method is prone to form outgrowths. So the surface morphology of GdYBCO film is optimized by depositing the SmBCO layer, which is an important process method for the preparation of high-quality multilayer REBCO films. At last, the GdYBCO/SmBCO/GdYBCO multilayer films were successfully prepared on the LMO templates based on the simple self-heating method. It is demonstrated that the GdYBCO surface was well improved by the characterization analysis of scanning electron microscope. And the Δω of REBCO (005) and Δφ of REBCO (103), which were performed by an X-ray diffraction system, are respectively 1.3° and 3.3° What's more, the critical current density (Jc) has been more than 3 MA/cm2 (77 K, 0 T) and the critical current (Ic) basically shows a trend of good linear increase with the increase of the number of REBCO layers.

  12. Thermochemical properties of media for pyrometallurgical nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Hosoya, Yuji; Terai, Takayuki

    1998-01-01

    Molten chloride/cadmium system is considered to be applied to a solvent in pyrochemical reprocessing of spent nuclear fuel. In this work, phase diagrams for molten chloride systems were constructed, using NdCl 3 as an imitative substance in place of UCl 3 or PuCl 3 . Hastelloy-X (Ni/Cr21/Fe18/Mo9/W) was examined as a structural material for the corrosion-resistance against molten chloride baths containing NdCl 3 . The process of corrosion was thermochemically discussed and the form of the corrosion was illustrated. Rutherford backscattering spectroscopy was successfully applied to determine the elemental distribution profile of specimens tested on the compatibility with molten chloride mixture at elevated temperature. Ferritic steel was also examined as another candidate material for the compatibility with molten cadmium covered with LiCl-KCl eutectic salt. Variation of near-surface composition was observed by comparing the results of Rutherford backscattering spectroscopy obtained before and after the dipping. (author)

  13. Revestimento de Níquel Depositado pela Soldagem MIG e MIG com Arame Frio

    Directory of Open Access Journals (Sweden)

    Carlos Alberto Mendes da Mota

    Full Text Available Resumo: Este artigo apresenta um estudo das características operacionais, geométricas e microestruturais de soldas MIG e MIGAF (MIG com arame frio aplicadas no revestimento de chapa de um aço AISI 1020, com uma superliga de níquel ER NiCrMo-4, tipo Hastelloy 276C. A soldagem foi automatizada, na posição plana, e realizada por uma fonte eletrônica em CC+ com aquisição instantânea dos oscilogramas de corrente e tensão de soldagem. As variáveis de entrada foram às velocidades do arame eletrodo e do arame frio, e a velocidade de soldagem. Foram avaliadas a operacionalidade dos processos e o efeito da energia de soldagem sobre as características geométricas (reforço e largura, diluição, microestrutura e microdureza do revestimento. Os resultados indicaram um bom desempenho operacional, a ausência de defeitos nos passes isolados e nos revestimentos para as soldagens com MIGAF. Além disso, constataram-se menores níveis de diluição.

  14. Heat transfer measurements in a forced convection loop with two molten-fluoride salts: LiF--BeF2--ThF2--UF4 and eutectic NaBF4--NaF

    International Nuclear Information System (INIS)

    Silverman, M.D.; Huntley, W.R.; Robertson, H.E.

    1976-10-01

    Heat transfer coefficients were determined experimentally for two molten-fluoride salts [LiF-BeF 2 -ThF 2 -UF 4 (72-16-12-0.3 mole %) and NaBF 4 -NaF (92-8 mole %] proposed as the fuel salt and coolant salt, respectively, for molten-salt breeder reactors. Information was obtained over a wide range of variables, with salt flowing through 12.7-mm-OD (0.5-in.) Hastelloy N tubing in a forced convection loop (FCL-2b). Satisfactory agreement with the empirical Sieder-Tate correlation was obtained in the fully developed turbulent region at Reynolds moduli above 15,000 and with a modified Hausen equation in the extended transition region (Re approx.2100-15,000). Insufficient data were obtained in the laminar region to allow any conclusions to be drawn. These results indicate that the proposed salts behave as normal heat transfer fluids with an extended transition region

  15. JAERI Material Performance Database (JMPD); outline of the system

    International Nuclear Information System (INIS)

    Yokoyama, Norio; Tsukada, Takashi; Nakajima, Hajime.

    1991-01-01

    JAERI Material Performance Database (JMPD) has been developed since 1986 in JAERI with a view to utilizing the various kinds of characteristic data of nuclear materials efficiently. Management system of relational database, PLANNER was employed and supporting systems for data retrieval and output were expanded. JMPD is currently serving the following data; (1) Data yielded from the research activities of JAERI including fatigue crack growth data of LWR pressure vessel materials as well as creep and fatigue data of the alloy developed for the High Temperature Gas-cooled Reactor (HTGR), Hastelloy XR. (2) Data of environmentally assisted cracking of LWR materials arranged by Electric power Research Institute (EPRI) including fatigue crack growth data (3000 tests), stress corrosion data (500 tests) and Slow Strain Rate Technique (SSRT) data (1000 tests). In order to improve user-friendliness of retrieval system, the menu selection type procedures have been developed where knowledge of system and data structures are not required for end-users. In addition a retrieval via database commands, Structured Query Language (SQL), is supported by the relational database management system. In JMPD the retrieved data can be processed readily through supporting systems for graphical and statistical analyses. The present report outlines JMPD and describes procedures for data retrieval and analyses by utilizing JMPD. (author)

  16. Alfa-Laval plate heat exchangers for the power industries

    International Nuclear Information System (INIS)

    Kitae, Junnosuke; Mtsuura, Kazuyuki

    1979-01-01

    Within power-generating plants, the transfer and conversion of heat energy of very large quantity are carried out in the process of energy conversion, accordingly the importance of heat exchangers is very high. Heretofore, multi-tube heat exchangers have been used mostly, but Alfa-Laval group developed the heat exchanger with very high efficiency to incorporate it effectively into a power-generating plant. In this plate type heat exchanger, the heat transfer efficiency is very high, and the quantity of stagnation is small as it is compact, consequently it is suitable to the secondary cooling for power-generating plant or the heat exchange of high-priced liquid heat media such as heavy water. Originally, plate type heat exchangers were used for food and chemical industries, therefore the prevention of mixing two liquids, sanitary construction, and corrosion resistance were required. Then they were adopted in iron and steel industry, and large thermal load, large heat transfer area and corrosion resistance to sea water were required. They were adopted in a nuclear power plant for the first time in 1964. In this heat exchanger, channels are formed with corrugated metal sheets, and titanium, stainless steels, Incoloy, Hastelloy and others are used as occasion demands. The Alfa-Laval heat exchangers and their features are explained. (Kako, I.)

  17. Evaluation of creep-fatigue/ environment interaction in Ni-base wrought alloys for HTGR application

    International Nuclear Information System (INIS)

    Hattori, Hiroshi; Kitagawa, Masaki; Ohtomo, Akira

    1986-01-01

    High Temperature Gas-cooled Reactor (HTGR) systems should be designed based on the high temperature structural strength design procedures. On the development of design code, the determination of failure criteria under cyclic loading and severe environments is one of the most important items. By using the previous experimental data for Ni-base wrought alloys, Inconel 617 and Hastelloy XR, several evaluation methods for creep-fatigue interaction were examined for their capability to predict their cyclic loading behavior for HTGR application. At first, the strainrange partitioning method, the frequency modified damage function and the linear damage summation rule were discussed. However, these methods were not satisfactory with the above experimental results. Thus, in this paper, a new fracture criterion, which is a modification of the linear damage summation rule, is proposed based on the experimental data. In this criterion, fracture is considered to occur when the sum of the fatigue damage, which is the function of the applied cyclic strain magnitude, and the modified creep damage, which is the function of the applied cyclic stress magnitude (determined as time devided by cyclic creep rupture time reflecting difference of creep damages by tensile creep and compressive creep), reaches a constant value. This criterion was successfully applied to the life prediction of materials at HTGR temperatures. (author)

  18. Materials performance in a high-level radioactive waste vitrification system

    International Nuclear Information System (INIS)

    Imrich, K.J.; Chandler, G.T.

    1996-01-01

    The Defense Waste Processing Facility (DWPF) is a Department of Energy Facility designed to vitrify highly radioactive waste. An extensive materials evaluation program has been completed on key components in the DWPF after twelve months of operation using nonradioactive simulated wastes. Results of the visual inspections of the feed preparation system indicate that the system components, which were fabricated from Hastelloy C-276, should achieve their design lives. Significant erosion was observed on agitator blades that process glass frit slurries; however, design modifications should mitigate the erosion. Visual inspections of the DWPF melter top head and off gas components, which were fabricated from Inconel 690, indicated that varying degrees of degradation occurred. Most of the components will perform satisfactorily for their two year design life. The components that suffered significant attack were the borescopes, primary film cooler brush, and feed tubes. Changes in the operation of the film cooler brush and design modifications to the feed tubes and borescopes is expected to extend their service lives to two years. A program to investigate new high temperature engineered materials and alloys with improved oxidation and high temperature corrosion resistance will be initiated

  19. Hot corrosion behavior of Ni-Cr-W-C alloys in impure helium gas

    International Nuclear Information System (INIS)

    Ohmura, Taizo; Sahira, Kensho; Sakonooka, Akihiko; Yonezawa, Noboru

    1976-01-01

    Influence of the minor alloy constituents such as Al, Mn and Si on the hot corrosion behavior of Ni-20Cr-20W-0.07C alloy was studied in 99.995% helium gas at 1000 0 C, comparing with that behavior of commercial Ni-base superalloys (Hastelloy X and Inconel 617). The low oxidizing potential in the impure helium gas usually causes selective oxidation of these elements and the growth of oxide whiskers on the surface of specimen at elevated temperature. The intergranular attack was caused by selective oxidation of Al, Si and Mn. The spalling of oxide film was restrained by addition of Mn and Si, providing tough spinel type oxide film on the surface and 'Keyes' on the oxide-matrix interface respectively. The amount and the morphology of the oxide whiskers depended on Si and Mn content. More than 0.29% of Si content without Mn always caused the growth of rather thinner whiskers with smooth surface, and the whiskers analyzed by electron diffraction patterns and EPMA to be Cr 2 O 3 containing Si. Mn addition changed the whiskers to thicker ones of spinel type oxide (MnCr 2 O 1 ) with rough surface. On the basis of these results, the optimum content of Al, Mn and Si to minimize the growth of whiskers, the intergranular attack and the spalling of oxide film was discussed. (auth.)

  20. Hot corrosion behavior of Ni-Cr-W-C alloys in impure He gas

    International Nuclear Information System (INIS)

    Ohmura, Taizo; Sahira, Kensho; Sakonooka, Akihiko; Yonezawa, Noboru

    1977-01-01

    Influence of the minor alloy constituents such as Al, Mn and Si on the hot corrosion behavior of Ni-20Cr-20W-0.07C alloy was studied in 99.995%He gas at 1,000 0 C, in comparison with the behavior of commercial Ni-base superalloys (Hastelloy X and Inconel 617). The low oxidizing potential in the impure He gas usually causes selective oxidation of the elements described above and the growth of oxide whiskers on the surface of specimen at elevated temperatures. The intergranular attack was caused by selective oxidation of Al, Si and Mn. The spalling of oxide film was restrained by additions of Mn and Si, providing tough spinel type oxide film on the surface and 'keys' on the oxide-matrix interface respectively. The amount and morphology of the oxide whiskers depended on Si and Mn contents. Si of more than 0.29% without Mn always caused the growth of rather thinner whiskers with smooth surface, and the whiskers analyzed by electron diffraction patterns and EPMA to be Cr 2 O 3 containing Si. Mn addition changes the whiskers to thicker ones of spinel type oxide (MnCr 2 O 4 ) with rough surface. On the basis of these results, the optimum contents of Al, Mn and Si to minimize the growth of whiskers, the intergranular attack, and the spalling of oxide film were discussed. (auth.)

  1. Structural integrity assessment of intermediate heat exchanger in the HTTR. Based on results of rise-to-power test

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Takeshi; Tachibana, Yukio; Nakagawa, Shigeaki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2002-12-01

    A helium/helium intermediate heat exchanger (IHX) in the high temperature engineering test reactor (HTTR) is an essential component for demonstration of future nuclear process heat utilization of high temperature gas-cooled reactor (HTGR). The IHX with a heat capacity of 10 MW has 96 helically-coiled heat transfer tubes. Structural design for the IHX had been conducted through elastic-creep analysis of superalloy Hastelloy XR components such as heat transfer tubes and center pipe. In the HTTR rise-to-power test, it was clarified that temperature of the coolant in the IHX at the reactor scrams changes more rapidly than expected in the design. Effects of the IHX coolant temperature change, at anticipated reactor scram from the full power of 30 MW at high temperature test operation, on structural integrity of the heat transfer tubes and the lower reducer of the center pipe were investigated analytically based on the coolant temperature data obtained from the rise-to-power test. As results of the assessment, it was confirmed that cumulative principal creep strain, cumulative creep and fatigue damage factor of the IHX components during 10{sup 5} h of the HTTR lifetime should be below the allowable limits, which are established in the high-temperature structural design code for the HTGR Class 1 components. (author)

  2. Properties of container and backfill materials for the final disposal of highly radioactive fission products

    International Nuclear Information System (INIS)

    Mirschinka, V.

    1983-11-01

    The qualifications of six metallic alloys to serve as canister materials for an in-can glass smelting process were studied. These alloys are: N 6 1.4864 (X 12NiCrSi3616, Thermax 16/36), No. 2.4816 (NiCr15Fe, Inconel 600), No. 2.4610 (Hastelloy C4), No. 2.4778 (UMCO50), No. 1.5415 (15MO3), No. 1.1005 (ZSH-Spezial). The mechanical properties of any of the six materials at high temperatures were found to be sufficient. The chemical interactions between glass and metal were investigated by glass smelting tests and electron microprobe analyses, showing that chromium as an alloying element of the crucible material may affect the quality of the glass product by causing inhomogeneities and a violent blistering in the glass matrix. The resistance against corrosion by concentrated salt solutions under elevated pressure and temperature similar to final depository conditions was tested showing that the presence of a bentonite suspension in the salt solution reduces the corrosion attack of the metal significantly. Diffusion experiments of salt solutions doted with radioactive isotopes Na-22 and Cl-36 as tracer substances were made to show the retardation behaviour of salt ions in compacted bentonite. However, a long-term barrier effect of the bentonite against salt ion diffusion could not be verified. (orig./HOE)

  3. Study of cutting speed on surface roughness and chip formation when machining nickel-based alloy

    International Nuclear Information System (INIS)

    Khidhir, Basim A.; Mohamed, Bashir

    2010-01-01

    Nickel- based alloy is difficult-to-machine because of its low thermal diffusive property and high strength at higher temperature. The machinability of nickel- based Hastelloy C-276 in turning operations has been carried out using different types of inserts under dry conditions on a computer numerical control (CNC) turning machine at different stages of cutting speed. The effects of cutting speed on surface roughness have been investigated. This study explores the types of wear caused by the effect of cutting speed on coated and uncoated carbide inserts. In addition, the effect of burr formation is investigated. The chip burr is found to have different shapes at lower speeds. Triangles and squares have been noticed for both coated and uncoated tips as well. The conclusion from this study is that the transition from thick continuous chip to wider discontinuous chip is caused by different types of inserts. The chip burr has a significant effect on tool damage starting in the line of depth-of-cut. For the coated insert tips, the burr disappears when the speed increases to above 150 m/min with the improvement of surface roughness; increasing the speed above the same limit for uncoated insert tips increases the chip burr size. The results of this study showed that the surface finish of nickel-based alloy is highly affected by the insert type with respect to cutting speed changes and its effect on chip burr formation and tool failure

  4. Effect of test temperature on tensile and fatigue properties of nickel-base heat-resistant alloys

    International Nuclear Information System (INIS)

    Tsuji, Hirokazu; Nakajima, Hajime

    1987-01-01

    A series of tensile and strain controlled low-cycle fatigue tests were conducted at temperatures ranging from RT to 900 0 C on a nickel-base heat-resistant alloy, Hastelloy XR-II, which is one of the candidate alloys for applications in the process heating high-temperature gas-cooled reactor (HTGR). Fatigue tests at room temperature and all tensile tests were conducted in air, while fatigue tests at and above 400 0 C were conducted in the simulated HTGR helium environment. In those tests the effect of test temperature on tensile and fatigue properties was investigated. The ductility minimum point was observed near 600 0 C, while tensile and fatigue strengths decreased with increasing test temperature. The fatigue lives estimated with the method proposed by Manson were compatible with the experimental results under the given conditions. For the specimens fatigued at and above 700 0 C, the percentage of the intergranular fracture mode gradually increased with increasing test temperature. (orig.)

  5. Mechanical Property and Its Comparison of Superalloys for High Temperature Gas Cooled Reactor

    International Nuclear Information System (INIS)

    Kim, Woo Gon; Kim, D. W.; Ryu, W. S.; Han, C. H.; Yoon, J. H.; Chang, J.

    2005-01-01

    Since structural materials for high temperature gas cooled reactor are used during long period in nuclear environment up to 1000 .deg. C, it is important to have good properties at elevated temperature such as mechanical properties (tensile, creep, fatigue, creep-fatigue), microstructural stability, interaction between metal and gas, friction and wear, hydrogen and tritium permeation, irradiation behavior, corrosion by impurity in He. Thus, in order to select excellent materials for the high temperature gas cooled reactor, it is necessary to understand the material properties and to gather the data for them. In this report, the items related to material properties which are needed for designing the high temperature gas cooled reactor were presented. Mechanical properties; tensile, creep, and fatigue etc. were investigated for Haynes 230, Hastelloy-X, In 617 and Alloy 800H, which can be used as the major structural components, such as intermediate heat exchanger (IHX), hot duct and piping and internals. Effect of He and irradiation on these structural materials was investigated. Also, mechanical properties; physical properties, tensile properties, creep and creep crack growth rate were compared for them, respectively. These results of this report can be used as important data to select superior materials for high temperature gas reactor

  6. Numerical models of delamination behavior in 2G HTS tapes under transverse tension and peel

    Science.gov (United States)

    Duan, Yujie; Ta, Wurui; Gao, Yuanwen

    2018-02-01

    In extreme operating environments, delamination in 2G HTS tapes occurs within and/or near the superconductor layer from high transverse tensile stresses caused by fabrication, Lorentz forces and thermal mismatch, etc. Generally, transverse opening and peeling off are the main delamination modes, and are always studied in anvil and peel tests, respectively. Numerical models of these modes for 2G HTS tape are presented wherein the mixed-mode traction-separation law at the interface of the silver and superconductor layers is considered. Plastic deformations of copper, silver, and Hastelloy® in the HTS tape are taken into account. The results obtained from the transverse opening model show that the maximum average tensile stress is smaller than the delamination tensile strength because delamination is asynchronous in the tape. When a crack appears in the tape, only a small stress ( ≤ 1 MPa) is required to expand the crack to other stress free areas through peeling. Using the peeling model, the dependency of the peel strength on peeling angle is investigated under constant fracture toughness. Peel strength decreases with the peeling angle until the minimum value is reached at 150°, and thereafter increases slightly. Other results indicate that peel strength depends strongly on delamination strength, fracture toughness, and thickness of copper layer. The fracture toughness of the delamination interface, which is difficult to obtain by experiment, can be extracted using the present model.

  7. Development of a procedure for estimating the high cycle fatigue strength of some high temperature structural alloys

    International Nuclear Information System (INIS)

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

    1979-01-01

    The generation of strain controlled fatigue data, for the standard strain rate of 4 x 10 -3 sec -1 , presents a problem when the cycles to failure exceed 10 5 because of the prohibitively long test times involved. In an attempt to circumvent this difficulty an evaluation has been made of a test procedure involving a fast cycling rate (40 Hz) and load controlled conditions. The validity of this procedure for extending current fatigue curves from 10 5 to 10 8 cycles and beyond, hinges upon the selection of an appropriate effective strain value, since the strain usually changes rapidly during the early stage of fatigue. Results from annealed 2 1/4 Cr-1 Mo, type 304 stainless steel, Incoloy 800H and Hastelloy X, tested over a wide range of temperatures, show that the strain measured N/sub f/2 is a reasonable estimate since it gives an excellent correlation between the strain and load controlled tests in the 10 5 cycle range where the data overlap. It seems clear that the differences in cycling rate and early stress-strain history for the two tests do not significantly affect the correlation. It may, therefore, be concluded that such load control test procedures may be used as a valid fast way for extending currently available fatigue curves from 10 5 to 10 8 cycles, and beyond

  8. Corrosion studies on retrievable spent fuel containers: a progress report

    International Nuclear Information System (INIS)

    Ludemann, W.D.; Abrego, L.; McCright, R.D.

    1978-12-01

    Spent fuel canisters stored in halite (NaCl) deposits (salt beds) are subject to a severely corrosive environment when the hot brine inclusions, rich in calcium and magnesium chlorides, migrate to the canister. Since no data base exists on corrosion in halite brines, a survey was made of the corrosion resistance of potential canister materials in other concentrated brine environments. Corrosion-resistant metals include Ta, Ti Code 12, TiPd Alloy, Inconel 625, Hastelloy C-276, and Fe-base 29-4 Alloy. Although carbon steels have cost and availability advantages, they suffer from excessive corrosion rates in brines. Corrosion-resistant nonmetals include carbon, Teflon-type fluorocarbons, epoxide coatings, and polymer cements. While these materials are not suitable for constructing the canister, they could be used as a protective coating on a carbon steel canister. On the basis of this survey, we recommend a coated carbon steel canister, used with cathodic protection. It is important to start a test program to gather a data base on the corrosion of materials in halite brines and to verify the suitability of canister materials

  9. Destruction of chemical agent simulants in a supercritical water oxidation bench-scale reactor

    Energy Technology Data Exchange (ETDEWEB)

    Veriansyah, Bambang [Supercritical Fluid Research Laboratory, Clean Technology Research Center, Korea Institute of Science and Technology (KIST), 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of) and Department of Green Process and System Engineering, University of Science and Technology, 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of)]. E-mail: vaveri@kist.re.kr; Kim, Jae-Duck [Supercritical Fluid Research Laboratory, Clean Technology Research Center, Korea Institute of Science and Technology (KIST), 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of) and Department of Green Process and System Engineering, University of Science and Technology, 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of)]. E-mail: jdkim@kist.re.kr; Lee, Jong-Chol [Agency for Defense Development (ADD), P.O. Box 35-1, Yuseong-gu, Daejeon (Korea, Republic of)]. E-mail: jcleeadd@hanafos.com

    2007-08-17

    A new design of supercritical water oxidation (SCWO) bench-scale reactor has been developed to handle high-risk wastes resulting from munitions demilitarization. The reactor consists of a concentric vertical double wall in which SCWO reaction takes place inside an inner tube (titanium grade 2, non-porous) whereas pressure resistance is ensured by a Hastelloy C-276 external vessel. The performances of this reactor were investigated with two different kinds of chemical warfare agent simulants: OPA (a mixture of isopropyl amine and isopropyl alcohol) as the binary precursor for nerve agent of sarin and thiodiglycol [TDG (HOC{sub 2}H{sub 4}){sub 2}S] as the model organic sulfur heteroatom. High destruction rates based on total organic carbon (TOC) were achieved (>99.99%) without production of chars or undesired gases such as carbon monoxide and methane. The carbon-containing product was carbon dioxide whereas the nitrogen-containing products were nitrogen and nitrous oxide. Sulfur was totally recovered in the aqueous effluent as sulfuric acid. No corrosion was noticed in the reactor after a cumulative operation time of more than 250 h. The titanium tube shielded successfully the pressure vessel from corrosion.

  10. FRAUD/SABOTAGE Killing Nuclear-Reactors Need Modeling!!!: ``Super'' alloys GENERIC ENDEMIC Wigner's-Disease/.../IN-stability: Ethics? SHMETHICS!!!

    Science.gov (United States)

    O'Grady, Joseph; Bument, Arlden; Siegel, Edward

    2011-03-01

    Carbides solid-state chemistry domination of old/new nuclear-reactors/spent-fuel-casks/refineries/jet/missile/rocket-engines is austenitic/FCC Ni/Fe-based (so miscalled)"super"alloys(182/82;Hastelloy-X,600,304/304L-SSs,...690!!!) GENERIC ENDEMIC EXTANT detrimental(synonyms): Wigner's-disease(WD) [J.Appl.Phys.17,857 (46)]/Ostwald-ripening/spinodal-decomposition/overageing-embrittlement/thermal-leading-to-mechanical(TLTM)-INstability: Mayo[Google: fLeaksCouldKill > ; - Siegel [ J . Mag . Mag . Mtls . 7 , 312 (78) = atflickr . comsearchonGiant - Magnotoresistance [Fert" [PRL(1988)]-"Gruenberg"[PRL(1989)] 2007-Nobel]necessitating NRC inspections on 40+25=65 Westin"KL"ouse PWRs(12/2006)]-Lai [Met.Trans.AIME, 9A,827(78)]-Sabol-Stickler[Phys.Stat.Sol.(70)]-Ashpahani[ Intl.Conf. Hydrogen in Metals, Paris(1977]-Russell [Prog.Mtls.Sci.(1983)]-Pollard [last UCS rept.(9/1995)]-Lofaro [BNL/DOE/NRC Repts.]-Pringle [ Nuclear-Power:From Physics to Politics(1979)]-Hoffman [animatedsoftware.com], what DOE/NRC MISlabels as "butt-welds" "stress-corrosion cracking" endpoint's ROOT-CAUSE ULTIMATE-ORIGIN is WD overageing-embrittlement caused brittle-fracture cracking from early/ongoing AEC/DOE-n"u"tional-la"v"atories sabotage!!!

  11. FRAUD/SABOTAGE Killing Nuclear-Reactors!!! ``Super"alloys GENERIC ENDEMIC Wigner's-Disease IN-stability!!!

    Science.gov (United States)

    Asphahani, Aziz; Siegel, Sidney; Siegel, Edward

    2010-03-01

    Siegel [[J.Mag.Mag.Mtls.7,312(78); PSS(a)11,45(72); Semis.& Insuls.5(79)] (at: ORNL, ANS, Westin``KL"ouse, PSEG, IAEA, ABB) warning of old/new nuclear-reactors/spent-fuel-casks/refineries/ jet/missile/rocket-engines austenitic/FCC Ni/Fe-based (so MIS- called)``super"alloys(182/82;Hastelloy-X; 600;304/304L-SSs; 690 !!!) GENERIC ENDEMIC EXTANT detrimental(synonyms): Wigner's- diseas(WD)[J.Appl.Phys.17,857(46)]; Ostwald-ripening; spinodal- decomposition; overageing-embrittlement; thermomechanical- INstability: Mayo[Google: ``If Leaks Could Kill"; at flickr.com search on ``Giant-Magnotoresistance"; find: [SiegelPolitics(79)]; Hoffman[animatedsoftware.com],...what DOE/NRC MISlabels as ``butt-welds" ``stress-corrosion cracking" endpoint's ROOT-CAUSE ULTIMATE-ORIGIN is WD overageing-embrit- tlement caused brittle-fracture cracking from early/ongoing AEC/DOE-n``u''tional-la``v''atories sabotage!!!

  12. Compatibility of molten salt and structural materials

    Energy Technology Data Exchange (ETDEWEB)

    Kawakami, Masahiro [Toyohashi Univ. of Technology, Aichi (Japan)

    1994-12-01

    As the important factors for considering the compatibility of fuel salt and coolant salt with structural materials in molten salt reactors, there are the moisture remaining in molten salt and the fluorine potential in molten salt. In this study, as for the metals which are the main components of corrosion resistant alloys, the corrosion by the moisture remaining in molten salt and the dependence of the corrosion on fluorine potential were examined. As the molten salts, an eutectic molten salt LiF-BeF{sub 2} was mainly used, and LiF-KF was used in combination. As the metallic materials, Cr, Ni and Cu which are the main components of corrosion resistant and heat resistant alloys, Hastelloy and Monel, were used. In the experiment, the metal pieces were immersed in the molten salt, and by sampling the molten salt, the change with time lapse of the concentration of the dissolved metals was examined. Besides, the electrochemical measurement was carried out for Cr, of which the corrosion was remarkable, and the change with time lapse of the dissolved ions was examined. The experimental setup, the experimental method, and the results of the immersion test and the electrochemical test are reported. The experiment on the corrosion of metals depending on fluorine potential is also reported. (K.I.).

  13. Requirements for a cleanable steel HEPA filter derived from a systems analysis

    International Nuclear Information System (INIS)

    Bergman, W.

    1996-06-01

    A systems analysis was conducted to determine customer requirements for a cleanable high efficiency particulate air (HEPA) filter in DOE Environmental Management (EM) facilities. The three principal drivers for cleanable steel HEPA are large cost savings, improved filter reliability, and new regulations; they produce a strong incentive to DOE customers to use cleanable steel HEPA filters. Input for customer requirements were obtained from field trips to EM sites and from discussions. Most existing applications require that cleanable steel HEPA filters meet size/performance requirements of standard glass HEPA filters; applications in new facilities can relax size/weight/pressure drop requirements on a case-by-case basis. We then obtained input from commercial firms on availability of cleanable steel HEPA filters. Systems analysis then showed that currently available technology was only able to meet customer needs in a limited number of cases. Further development is needed to meet requirements of EM customers. For cleanable steel HEPA to be retrofitted into existing systems, pressure drop and weight must be reduced. Pressure drop can be reduced by developing steel fiber media from 0.5 μm dia steel fibers. Weight can be reduced by packaging the steel fiber media in one of the standard HEPA configurations. Although most applications will be able to use standard 304 or 316L alloys, an acid resistant alloy such as Hastelloy or Inconel will be needed for incinerator and other thermal processes

  14. Nuclear rich alpha cellulosic waste management experiments by acid digestion

    International Nuclear Information System (INIS)

    Arnal; Cousinou; Desille; Maigret.

    1985-03-01

    At Cadarache, where the French plutonium fuel fabrication plant is located, the strategy used for the management of rich alpha waste (superior to accepted level for storage) consist in incinerating the wastes, crushed and washed by cryogenic crushing and soda-nitric solutions. Although all ''technological'' wastes could be processed this way, the cellulosic are sorted and treated separately by the sulfuric acid digestion process. This process has definite advantages, particularly since it is specific to cellulosis, which dissolves easily at low temperature, i-e under the boiling point of H 2 SO 4 . Except for this aspect, of great importance for the gaz treatment operations and the resistance of material to corrosion, the process is identical to the one given in the literature: dehydration of cellulosis by H 2 SO 4 72% and carbon oxydation by HNO 3 13N. The apparatus used hold in a small volume (10 m 3 ); the gloves-box in which the dissolver and the filtration treatments (insoluble Pu sulfate for one part, and reaction gas for the other) are placed is in stainless steel coated with corrosion proof paint; the equipments are made of glass (dissolver) teflon (flanges) PVDF (pipes) hastelloy (pompes). A general balance is given for the recuperated nuclear materials, as well as for the mass and volumes of input and output cellulosic wastes

  15. Technology readiness level (TRL) assessment of cladding alloys for advanced nuclear fuels

    International Nuclear Information System (INIS)

    Shepherd, Daniel

    2015-01-01

    Reliable fuel claddings are essential for the safe, sustainable and economic operation of nuclear stations. This paper presents a worldwide TRL assessment of advanced claddings for Gen III and IV reactors following an extensive literature review. Claddings include austenitic, ferritic/martensitic (F/M), reduced activation (RA) and oxide dispersion strengthened (ODS) steels as well as advanced iron-based alloys (Kanthal alloys). Also assessed are alloys of zirconium, nickel (including Hastelloy R ), titanium, chromium, vanadium and refractory metals (Nb, Mo, Ta and W). Comparison is made with Cf/C and SiCf/SiC composites, MAX phase ceramics, cermets and TRISO fuel particle coatings. The results show in general that the higher the maximum operating temperature of the cladding, the lower the TRL. Advanced claddings were found to have lower TRLs than the corresponding fuel materials, and therefore may be the limiting factor in the deployment of advanced fuels and even possibly the entire reactor in the case of Gen IV. (authors)

  16. Some Investigations of the Reaction of Activated Charcoal with Fluorine and Uranium Hexafluoride

    International Nuclear Information System (INIS)

    Del Cul, G.D.; Fiedor, J.N.; Simmons, D.W.; Toth, L.M.; Trowbridge, L.D.; Williams

    1998-01-01

    The Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory has been shut down since 1969, when the fuel salt was drained from the core into two Hastelloy N drain tanks at the reactor site. Over time, fluorine (F 2 ) and uranium hexafluoride (UF 6 ) moved from the salt through the gas piping to a charcoal bed, where they reacted with the activated charcoal. Some of the immediate concerns related to the migration of F 2 and UF 6 to the charcoal bed were the possibility of explosive reactions between the charcoal and F 2 , the existence of conditions that could induce a criticality accident, and the removal and recovery of the fissile uranium from the charcoal. This report addresses the reactions and reactivity of species produced by the reaction of fluorine and activated charcoal and between charcoal and F 2 -UF 6 gas mixtures in order to support remediation of the MSRE auxiliary charcoal bed (ACB) and the recovery of the fissile uranium. The chemical identity, stoichiometry, thermochemistry, and potential for explosive decomposition of the primary reaction product, fluorinated charcoal, was determined

  17. Fundamentals of molten-salt thermal technology

    International Nuclear Information System (INIS)

    1980-08-01

    This book has been published by the Society of Molten-Salt Thermal Technology to publish a part of the achievement of its members. This book is composed of seven chapters. The chapter 1 is Introduction. The chapter 2 explains the physical properties of molten salts, such as thermal behavior, surface tension, viscosity, electrical conductivity and others. The chapter 3 presents the compatibility with construction materials. Corrosion in molten salts, the electrochemical behavior of fluoride ions on carbon electrodes in fluoride melts, the behaviors of hastelloy N and metals in melts are items of this chapter. The equipments and instruments for molten salts are described in chapter 4. The heat transfer in molten salts is discussed in chapter 5. The chapter 6 explains the application of molten salt technology. The molten salt technology can be applied not only to thermal engineering and energy engineering but also to chemical and nuclear engineerings, and the technical fundamentals, current development status, technical problems and the perspective for the future are outlined. The chapter 7 is the summary of this book. The commercialization of molten salt power reactors is discussed at the end of this book. (Kato, T.)

  18. The ''C'' family of Ni-Cr-Mo allloys' partnership with the chemical process industry: the last 70 years

    International Nuclear Information System (INIS)

    Agarwal, D.C.; Herda, W.R.

    1997-01-01

    The ''C'' family of alloys, the original being Hastelloy trademark alloy C (1930's) was an innovative optimization of Ni-Cr alloys having good resistance to oxidizing corrosive media and Ni-Mo alloys with superior resistance to reducing corrosive media. This combination resulted in the most versatile corrosion resistant alloy in the ''Ni-Cr-Mo'' alloy family, with exceptional corrosion resistance in a wide variety of severe corrosive environments typically encountered in CPI and other industries. The alloy also exhibited excellent resistance to pitting and crevice corrosion attack in low pH, high chloride oxidizing environments and had virtual immunity to chloride stress corrosion cracking. These properties allowed this alloy to serve the industrial needs for many years, although it had some limitations. The decades of the 1960's (alloy C-276), 1970's (alloy C-4), 1980's (alloy C-22 and 622) and 1990's (alloy 59, alloy 686 and alloy C-2000) saw newer alloy developments with improvements in corrosion resistance, which not only overcame the limitations of alloy C, but further expanded the horizons of applications as the needs of the CPI became more critical, severe and demanding. Today the originally alloy ''C'' of the 1930's is practically obsolete except for some usage in form of castings. This paper presents a chronology of the various corrosion resistant alloy developments during this century, with special emphasis on the last 70 years evolution in the ''C'' family of Ni-Cr-Mo alloys and their applications. (orig.)

  19. Effect of thermal neutron irradiation on mechanical properties of alloys for HTR core applications

    International Nuclear Information System (INIS)

    Ogawa, Yutaka; Kondo, Tatsuo; Ishimoto, Kiyoshi; Ohtsuka, Tamotsu

    1979-01-01

    An industrial heat of Hastelloy-X containing 2.3 ppm boron was creep-tested at 900 0 C after irradiating thermal neutrons by 6.6 x 10 20 n.cm -2 at temperatures 670 to 880 0 C in JMTR. Significant reduction in rupture life and ductility was observed, and large shift of accelerated deformation stage to short time side was also apparent at comparatively high stresses. Below about 2.2 kg.mm -2 , apparent relief from the degradation was seen. The elongation, however, was found to be due to the formation of numerous intergranular cracks in the premature stage of deformation. Based on the post irradiation tensile properties of several industrial alloys the degree of the ductility loss was found to be nearly dependent on the boron content of the alloys. The post irradiation tensile tests for a special low boron grade heat revealed the means of protecting materials from the effect to be feasible. (author)

  20. Effect of thermal neutron irradiation on mechanical properties of alloys for HTR core applications

    International Nuclear Information System (INIS)

    Ogawa, Yutaka; Kondo, Tatsuo; Ishimoto, Kiyoshi; Ohtsuka, Tamotsu

    1979-02-01

    An industrial heat of Hastelloy-X containing 2.3 ppm boron was creep-tested at 900 0 C after irradiating thermal neutrons by 6.6 x 10 20 n/cm 2 at temperatures 670 to 880 0 C in JMTR. Significant reduction in rupture life and ductility was observed, and large shift of accelerated deformation stage to short time side was also apparent at comparatively high stresses. Below about 2.2 kg/mm 2 , apparent relief from the degradation was seen. The elongation, however, was found to be due to the formation of numerous intergranular cracks in the premature stage of deformation. Based on the post irradiation tensile properties of several industrial alloys the degree of the ductility loss was found to be nearly dependent on the boron content of the alloys. The post irradiation tensile tests for a special low boron grade heat revealed the means of protecting materials from the effect to be feasible. (author)

  1. Corrosion of container and infrastructure materials under clay repository conditions

    International Nuclear Information System (INIS)

    Debruyn, W.; Dresselaers, J.; Vermeiren, P.; Kelchtermans, J.; Tas, H.

    1991-01-01

    With regard to the disposal of high-level radioactive waste, it was recommended in a IAEA Technical Committee meeting to perform tests in realistic environments corresponding with normal and accidental conditions, to qualify and apply corrosion monitoring techniques for corrosion evaluation under real repository conditions and to develop corrosion and near-field evolution models. The actual Belgian experimental programme for the qualification of a container for long-term HLW storage in clay formations complies with these recommendations. The emphasis in the programme is indeed on in situ corrosion testing and monitoring and on in situ control of the near-field chemistry. Initial field experiments were performed in a near-surface clay quarry at Terhaegen. Based on a broad laboratory material screening programme and in agreement with the Commission of the European Communities, three reference materials were chosen for extensive in situ overpack testing. Ti/0.2 Pd and Hastelloy C-4 were chosen as reference corrosion resistant materials and a low-carbon steel as corrosion allowance reference material. This report summarizes progress made in the material qualification programme since the CEC contract of 1983-84. 57 Figs.; 15 Tabs.; 18 Refs

  2. Prediction of inelastic behavior and creep-fatigue life of perforated plates

    International Nuclear Information System (INIS)

    Igari, Toshihide; Yamauchi, Masafumi; Nomura, Shinichi.

    1992-01-01

    Prediction methods of macroscopic and local stress-strain behaviors of perforated plates in plastic and creep regime are proposed in this paper, and are applied to the creep-fatigue life prediction of perforated plates. Both equivalent-solid-plate properties corresponding to the macroscopic behavior and the stress-strain concentration around a hole were obtained by assuming the analogy between plasticity and creep and also by extending the authors' proposal in creep condition. The perforated plates which were made of Hastelloy XR were subjected to the strain-controlled cyclic test at 950degC in air in order to experimentally obtain the macroscopic behavior such as the cyclic stress-strain curve and creep-fatigue life around a hole. The results obtained are summarized as follows. (1) The macroscopic behavior of perforated plates including cyclic stress-strain behavior and relaxation is predictable by using the proposed method in this paper. (2) The creep-fatigue life around a hole can be predicted by using the proposed method for stress-strain concentration around a hole. (author)

  3. Materials evaluation for a transuranic processing facility

    International Nuclear Information System (INIS)

    Barker, S.A.; Schwenk, E.B.; Divine, J.R.

    1990-11-01

    The Westinghouse Hanford Company, with the assistance of the Pacific Northwest Laboratory, is developing a transuranium extraction process for preheating double-shell tank wastes at the Hanford Site to reduce the volume of transuranic waste being sent to a repository. The bench- scale transuranium extraction process development is reaching a stage where a pilot plant design has begun for the construction of a facility in the existing B Plant. Because of the potential corrosivity of neutralized cladding removal waste process streams, existing embedded piping alloys in B Plant are being evaluated and ''new'' alloys are being selected for the full-scale plant screening corrosion tests. Once the waste is acidified with HNO 3 , some of the process streams that are high in F - and low in Al and zr can produce corrosion rates exceeding 30,000 mil/yr in austenitic alloys. Initial results results are reported concerning the applicability of existing plant materials to withstand expected process solutions and conditions to help determine the feasibility of locating the plant at the selected facility. In addition, process changes are presented that should make the process solutions less corrosive to the existing materials. Experimental work confirms that Hastelloy B is unsatisfactory for the expected process solutions; type 304L, 347 and 309S stainless steels are satisfactory for service at room temperature and 60 degrees C, if process stream complexing is performed. Inconel 625 was satisfactory for all solutions. 17 refs., 5 figs., 8 tabs

  4. Investigation of metallic, ceramic, and polymeric materials for engineered barrier applications in nuclear-waste packages

    International Nuclear Information System (INIS)

    Westerman, R.E.

    1980-10-01

    An effort to develop licensable engineered barrier systems for the long-term (about 1000 yr) containment of nuclear wastes under conditions of deep continental geologic disposal has been underway at Pacific Northwest Laboratory since January 1979, under the auspices of the High-Level Waste Immobilization Program. In the present work, the barrier system comprises the hard or structural elements of the package: the canister, the overpack(s), and the hole sleeve. A number of candidate metallic, ceramic, and polymeric materials were put through mechanical, corrosion, and leaching screening tests to determine their potential usefulness in barrier-system applications. Materials demonstrating adequate properties in the screening tests will be subjected to more detailed property tests, and, eventually, cost/benefit analyses, to determine their ultimate applicability to barrier-system design concepts. The following materials were investigated: two titanium alloys of Grade 2 and Grade 12; 300 and 400 series stainless steels, Inconels, Hastelloy C-276, titanium, Zircoloy, copper-nickel alloys and cast irons; total of 14 ceramic materials, including two grades of alumina, plus graphite and basalt; and polymers such as polyamide-imide, polyarylene, polyimide, polyolefin, polyphenylene sulfide, polysulfone, fluoropolymer, epoxy, furan, silicone, and ethylene-propylene terpolymer (EPDM) rubber. The most promising candidates for further study and potential use in engineered barrier systems were found to be rubber, filled polyphenylene sulfide, fluoropolymer, and furan derivatives

  5. High Integrity Can Design Interfaces

    International Nuclear Information System (INIS)

    Shaber, E.L.

    1998-01-01

    The National Spent Nuclear Fuel Program is chartered with facilitating the disposition of DOE-owned spent nuclear fuel to allow disposal at a geologic repository. This is done through coordination with the repository program and by assisting DOE Site owners of SNF with needed information, standardized requirements, packaging approaches, etc. The High Integrity Can (HIC) will be manufactured to provide a substitute or barrier enhancement for normal fuel geometry and cladding. The can would be nested inside the DOE standardized canister which is designed to interface with the repository waste package. The HIC approach may provide the following benefits over typical canning approaches for DOE SNF. (a) It allows ready calculation and management of criticality issues for miscellaneous. (b) It segments and further isolates damaged or otherwise problem materials from normal SNF in the repository package. (c) It provides a very long term corrosion barrier. (d) It provides an extra internal pressure barrier for particulates, gaseous fission products, hydrogen, and water vapor. (e) It delays any potential release of fission products to the repository environment. (f) It maintains an additional level of fuel geometry control during design basis accidents, rock-fall, and seismic events. (g) When seal welded, it could provide the additional containment required for shipments involving plutonium content in excess of 20 Ci. (10 CFR 71.63.b) if integrated with an appropriate cask design. Long term corrosion protection is central to the HIC concept. The material selected for the HIC (Hastelloy C-22) has undergone extensive testing for repository service. The most severe theoretical interactions between iron, repository water containing chlorides and other repository construction materials have been tested. These expected chemical species have not been shown capable of corroding the selected HIC material. Therefore, the HIC should provide a significant barrier to DOE SNF dispersal

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-12-15

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

  7. Behaviour of metals and alloys in molten fluoride media

    International Nuclear Information System (INIS)

    Fabre, St.

    2009-01-01

    Fluoride salts are contemplated for Generation IV nuclear systems which structural materials need to resist corrosion at high temperatures. Corrosion of metals in molten fluorides has been investigated in support of the Molten Salt Reactor's development and led to an optimized alloy, Hastelloy-N, but it lacked fundamentals data for the comprehension of materials' degradation mechanisms. The main objective of this work is then to help with the understanding of the corrosion behaviour of nickel and its alloys in fluoride salts. An experimental method was built up using electrochemical techniques and enabled to investigate the thermochemical conditions of the media and the influence of different parameters (media, temperature and quantity of impurities) on the behaviour of the materials. Most tests were performed in LiF-NaF mixtures between 800 and 1000 C. Pure metals can be classified as follows: Cr ≤ Fe ≤ Ni ≤ Mo ≤ W in increasing stability order and two specific behaviours were evidenced: Cr and Fe corrode in the melt, whereas Ni, Mo and W are stable, underlining the significance level of the redox couple controlling the reactions in the mixture. Moreover, corrosion current densities increase with temperature, fluoro-acidity and the quantity of dissolved oxide in the melt. Binary Ni-Cr alloys were also tested; selective attack of Cr is first observed before both elements are oxidized. Combining thermochemical calculations and experimental results enables to propose an approach to establish an optimized composition for a stable alloy. Immersion tests were finally achieved in addition to the electrochemical tests: interpretations of both methods were compared and completed. (author)

  8. Electrochemical, Polarization, and Crevice Corrosion Testing of Nitinol 60, A Supplement to the ECLSS Sustaining Materials Compatibility Study

    Science.gov (United States)

    Lee, R. E.

    2016-01-01

    In earlier trials, electrochemical test results were presented for six noble metals evaluated in test solutions representative of waste liquids processed in the Environmental Control and Life Support System (ECLSS) aboard the International Space Station (ISS). Subsequently, a seventh metal, Nitinol 60, was added for evaluation and subjected to the same test routines, data analysis, and theoretical methodologies. The previous six test metals included three titanium grades, (commercially pure, 6Al-4V alloy and 6Al-4V low interstitial alloy), two nickel-chromium alloys (Inconel(RegisteredTrademark) 625 and Hastelloy(RegisteredTrademark) C276), and one high-tier stainless steel (Cronidur(RegisteredTrademark) 30). The three titanium alloys gave the best results of all the metals, indicating superior corrosive nobility and galvanic protection properties. For this current effort, the results have clearly shown that Nitinol 60 is almost as noble as titanium, being very corrosion-resistant and galvanically compatible with the other six metals electrochemically and during long-term exposure. is also quite noble as it is very corrosion resistant and galvanically compatible with the other six metals from both an electrochemical perspective and long-term crevice corrosion scenario. This was clearly demonstrated utilizing the same techniques for linear, Tafel and cyclic polarization, and galvanic coupling of the metal candidate as was done for the previous study. The high nobility and low corrosion susceptibility for Nitinol 60 appear to be intermediate to the nickel/chromium alloys and the titanium metals with indications that are more reflective of the titanium metals in terms of general corrosion and pitting behavior.

  9. Corrosion of 316 stainless steel in high temperature molten Li{sub 2}BeF{sub 4} (FLiBe) salt

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, Guiqiu, E-mail: guiqiuzheng@gmail.com; Kelleher, Brian; Cao, Guoping; Anderson, Mark; Allen, Todd; Sridharan, Kumar

    2015-06-15

    In support of structural material development for the fluoride-salt-cooled high-temperature reactor (FHR), corrosion tests of 316 stainless steel were performed in the potential primary coolant, molten Li{sub 2}BeF{sub 4} (FLiBe) at 700 °C for an exposure duration up to 3000 h. Tests were performed in both 316 stainless steel and graphite capsules. Corrosion in both capsule materials occurred by the dissolution of chromium from the stainless steel into the salt which led to the depletion of chromium predominantly along the grain boundaries of the test samples. The samples tested in graphite capsules showed a factor of two greater depth of corrosion attack as measured in terms of chromium depletion, compared to those tested in 316 stainless steel capsules. The samples tested in graphite capsules showed the formation of Cr{sub 7}C{sub 3} particulate phases throughout the depth of the corrosion layer. Samples tested in both types of capsule materials showed the formation of MoSi{sub 2} phase due to increased activity of Mo and Si as a result of Cr depletion, and furthermore corrosion promoted the formation of a α-ferrite phase in the near-surface regions of the 316 stainless steel. Based on the corrosion tests, the corrosion attack depth in FLiBe salt was predicted as 17.1 μm/year and 31.2 μm/year for 316 stainless steel tested in 316 stainless steel and in graphite capsules respectively. It is in an acceptable range compared to the Hastelloy-N corrosion in the Molten Salt Reactor Experiment (MSRE) fuel salt.

  10. Preliminary safety analysis of molten salt breeder reactor

    International Nuclear Information System (INIS)

    Cheng Maosong; Dai Zhimin

    2013-01-01

    Background: The molten salt reactor is one of the six advanced reactor concepts identified by the Generation IV International Forum as a candidate for cooperative development, which is characterized by remarkable advantages in inherent safety, fuel cycle, miniaturization, effective utilization of nuclear resources and proliferation resistance. ORNL finished the conceptual design of Molten Salt Breeder Reactor (MSBR) based on the design, building and operation of Molten Salt Reactor Experiment (MSRE). Purpose: We attempt to implement the preliminary safety analysis of MSBR in order to provide a reference for the design and optimization of MSBR in the future. Methods: According to the conceptual design of MSBR, a model of safety analysis using point kinetics coupled with the simplified heat transfer mechanism is presented. The model is applied to simulate the transient phenomena of MSBR initiated by an abnormal step reactivity addition and an abnormal ramp reactivity addition at full-power equilibrium condition. Results: The thermal power in the core increases rapidly at the beginning and is accompanied by a rise of the fuel and graphite temperatures after 100, 300, 500 and 600 pcm reactivity addition. The maximum outlet temperature of the fuel in the core is at 1250℃ in 500 pcm reactivity addition, but up to 1350℃ in 600 pcm reactivity addition. The maximum of the power and the temperature are delayed and lower in the ramp reactivity addition rather than in the step reactivity addition. Conclusions: Based on the results, when the reactivity inserted is less than 500 pcm in maximum at full power equilibrium condition, the structural material in Hastelloy-N is not melted and can keep integrity without external control action. And it is necessary to try to avoid inserting a reactivity at short time. (authors)

  11. High production rate of IBAD-MgO buffered substrate

    Energy Technology Data Exchange (ETDEWEB)

    Yoshizumi, M., E-mail: myoshizumi@istec.or.j [Superconductivity Research Laboratory, ISTEC, Shinonome 1-10-13, Koto-ku, Tokyo 135-0062 (Japan); Miyata, S.; Ibi, A.; Fukushima, H.; Yamada, Y.; Izumi, T.; Shiohara, Y. [Superconductivity Research Laboratory, ISTEC, Shinonome 1-10-13, Koto-ku, Tokyo 135-0062 (Japan)

    2009-10-15

    The conventional IBAD (Ion Beam Assisted Deposition) process using fluorite materials yields low production rates, resulting in high production cost, which reduces the motivation for practical application in spite of its high quality. The IBAD process using rock salt materials, e.g. MgO, is well known as a strong candidate of practical application due to its potential of high production rate and high in-plane grain alignment. In this work, the IBAD-MgO process was investigated for a newly developed architecture of PLD (Pulsed Laser Deposition)-CeO{sub 2}/sputter-LMO (LaMnO{sub 3})/IBAD-MgO/sputter-GZO (Gd{sub 2}Zr{sub 2}O{sub 7})/Hastelloy{sup TM} to make long buffered metal tapes with high properties and a high production rate. The 50 m-long IBAD-MgO substrates with about 4 deg. of DELTAphiCeO{sub 2} in an XRD phi scan could be fabricated repeatedly. A GdBCO (GdBa{sub 2}Cu{sub 3}O{sub x}) layer deposited on the buffered substrate showed the minimum I{sub c} value of 325 A/cm-w in a 41 m-long tape. Almost of the tape showed 500-600 A/cm-w of I{sub c} value. The deposition time for the IBAD-MgO layer was 60 s which was about 2 orders of magnitude shorter than the conventional IBAD process. The production rate of 24 m/h was realized at the IBAD-MgO process to fabricate the GdBCO coated conductor with high J{sub c} and I{sub c} properties.

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  13. The Effects of Gd-Free Impurity Phase on the Aging Behavior for the Microwave Surface Resistance of Ag-coated GdBa2Cu3O7-δ at Cryogenic Temperatures

    Science.gov (United States)

    Lee, Sungho; Yang, Woo Il; Jung, Ho Sang; Oh, Won-Jae; Jang, Jiyeong; Lee, Jae-Hun; Kang, Kihyeok; Moon, Seung-Hyun; Yoo, Sang-Im; Lee, Sang Young

    2018-05-01

    High-T C GdBa2Cu3O7-δ (GdBCO) superconductor has been popular for making superconductive tapes that have much potential for various fields of large-scale applications. We investigated aging effects on the microwave surface resistance (R S) of Ag-coated GdBCO layer on Hastelloy substrate, so called GdBCO coated conductors (CCs), and Ag-coated GdBCO films on LaAlO3 (LAO) single-crystal substrates at cryogenic temperatures and compared them with each other. Unlike the R S of Ag-coated GdBCO films showing significant degradation in 4 weeks, no significant aging effects were found in our Ag-coated GdBCO CCs aged 85 weeks. The reactive co-evaporation deposition and reaction (RCE-DR) method was used for preparing the Ag-coated GdBCO CCs. Such durability of the Ag-coated GdBCO CCs in terms of the R S could be explained by existence of a protective impurity phase, i.e., Gd-free Ba-Cu-O phase as confirmed by transmission electron microscopy study combined with the energy-dispersive X-ray spectroscopy measurements. Although the scope of this study is limited to the Ag-coated GdBCO CCs prepared by using the RCE-DR method, our results suggest that a solution for preventing the aging effects on transport properties of other kinds of Ag-coated GdBCO CCs could be realized by means of an artificially-grown protective impurity layer.

  14. Research and development activities at INE concerning corrosion of final repository container materials; F and E-Arbeiten zur Korrosion von Endlager-Behaelterwerkstoffen im INE

    Energy Technology Data Exchange (ETDEWEB)

    Kienzler, Bernhard

    2017-10-01

    The present work provides a historical overview of the research and development activities carried out at the (Nuclear) Research Center Karlsruhe (today KIT) since the beginning of the 1980s on the corrosion of materials which might be suitable for construction of containers for highly radioactive wastes. The report relates almost exclusively to the work performed by Dr. Emmanuel Smailos, who elaborated the corrosion of various materials at the Institute for Nuclear Waste Disposal (INE). The requirements for the containers and materials, which were subject to changes in time, are presented. The changes were strongly influenced by the changed perception of the use of nuclear energy. The selection of the materials under investigations, the boundary conditions for the corrosion experiments and the analytical methods are described. Results of the corrosion of the materials such as finegrained steel, Hastelloy C4, nodular cast iron, titanium-palladium and copper or copper-nickel alloys in typical salt solutions are summarized. The findings of special investigations, e.g. corrosion under irradiation or the influence of sulfide on the corrosion rates are shown. For construction of disposal canisters, experiments were conducted to determine the contact corrosion, the influence of the hydrogen embrittlement of Ti-Pd and fine-grained steels on the corrosion behavior as well as the corrosion behavior of welding and the influence of different welding processes with the resulting heat-affected zones on the corrosion behavior. The work was contributed to several European research programs and was well recognized in the USA. Investigations on the corrosion of steels in non-saline solutions and corrosion under interim storage conditions as well as under the expected conditions of the Konrad repository for low-level radioactive wastes are also described. In addition, the experiments on ceramic materials are presented and the results of the corrosion of Al{sub 2}O{sub 3} and Zr

  15. Advanced-ORIENT cycle, its scientific progress and prospect for engineering feasibility

    International Nuclear Information System (INIS)

    Koyama, Shin-ichi; Yamagishi, Isao; Fujii, Yasuhiko; Suzuki, Tatsuya; Ozawa, Masaki; Fujita, Reiko; Okada, Ken; Tatenuma, Katsuyoshi; Mimura, Hitoshi

    2011-01-01

    For the ultimate minimization of the ecological risks originated in nuclear fuel recycling, a new fuel cycle paradigm was proposed and the basic researches have been carried out as a first phase under the Adv.-ORIENT (Advanced Optimization by Recycling Instructive Elements) Cycle project. In this paradigm, effective separation of actinide (An) and long lived-fission product (LLFP), transmutation of An, utilization of separated nuclides, such as lanthanides (Lns) and platinum group metals (PGM), were the main directions. In such directions, tertiary pyridine-type resin (TPR) enabled to separate minor actinide (MA)/Ln and then Am/Cm precisely from spent fuel, provided permitting to use HCl as well as and HNO 3 media. Recovery of very pure Am and Cm products could be done in this phase. The PGM and Tc separation; Catalytic electrolytic extraction (CEE) method could effectively separate the light PGM ,Tc from HCl and HNO 3 media, especially by HCl media. The PGM and Tc utilization; Mixed deposit obtained from the CEE experiments, Ru/Rh/Pd/Tc(Re)-Pt electrodes indicated the highest catalytic reactivity on electrolytic production of hydrogen in an alkali solution. Recovery of Cs from simulated spent fuel solution by silica gel loaded with ammonium molybdophosphate (AMP) was carried out, and the uptake rate achieved more than 90%. Separated Cs is expected to utilize as a heat source element. As basic engineering research efforts, some candidate metals, such as Ta, Nb, Zr and Hastelloy-B (Ni-28Mo), were examined to confirm an anti-corrosive property in wide HCl environment. Gram scale experiment to identify a thermo-chemical stability of TPR and TBP (as a reference) was also performed experimentally, and process safety conditions could be found out for its practical use. In this paper, study for each integrant technology was concluded as first trial of Adv.-ORIENT Cycle project, and the perspective for next phase was proposed. (author)

  16. Fundamental studies on electron beam welding on heat resistant superalloys for nuclear plants, 6

    International Nuclear Information System (INIS)

    Susei, Syuzo; Shimizu, Sigeki; Nagai, Hiroyoshi; Aota, Toshikazu; Satoh, Keisuke

    1980-01-01

    In this report, base metal of superalloys for nuclear plants, its electron beam and TIG weld joints were compared with each other in the mechanical properties. Obtained conclusions are summarized as follows: 1) TIG weld joint is superior to electron beam weld joint and base metal in 0.2% proof stress irrespective of the material, and electron beam weld joint is also superior to base metal. There is an appreciable difference in tensile stress between base metal and weld joint regardless of the materials. Meanwhile, electron beam weld joint is superior to TIG weld joint in both elongation and reduction of area. 2) Electron beam weld joint has considerably higher low-cycle fatigue properties at elevated temperatures than TIG weld joint, and it is usually as high as base metal. 3) In the secondary creep rate, base metal of Hastelloy X (HAEM) has higher one than its weld joints. However, electron beam weld joint is nearly comparable to the base metal. 4) There is hardly any appreciable difference between base metal and weld joint in the creep rupture strength without distinction of the material. In the ductility, base metal is much superior and is followed by electron beam weld joint and TIG weld joint in the order of high ductility. However, electron beam weld joint is rather comparable to base metal. 5) In consideration of welded pipe with a circumferential joint, the weld joint should be evaluated in terms of secondary creep rate, elongation and rupture strength. As the weld joint of high creep rupture strength approaches the base metal in the secondary creep rate and the elongation, it seems to be more resistant against the fracture due to creep deformation. In this point of view, electron beam weld joint is far superior to TIG weld joint and nearly comparable to the base metal. (author)

  17. Characterization of amorphous yttria layers deposited by aqueous solutions of Y-chelate alkoxides complex

    Science.gov (United States)

    Kim, Young-Soon; Lee, Yu-Ri; Kim, Byeong-Joo; Lee, Jae-Hun; Moon, Seung-Hyun; Lee, Hunju

    2015-01-01

    Crack-free amorphous yttria layers were deposited by dip coating in solutions of different Y-chelate alkoxides complex. Three Y-chelate solutions of different concentrations were prepared using yttrium acetate tetrahydrate, yttrium stearic acid as Y source materials. PEG, diethanolamine were used as chelating agents, while ethanol, methanol and tetradecane were used as solvent. Three different combinations of chelating and solvents were used to prepare solutions for Y2O3 dip coating on SUS, electropolished and non-electropolished Hastelloy C-276 substrates. The thickness of the films was varied by changing the number of dipping cycles. At an optimized condition, the substrate surface roughness (rms) value was reduced from ∼50 nm to ∼1 nm over a 10 × 10 μm2 area. After Y2O3 deposition, MgO was deposited using ion-beam assisted deposition (IBAD), then LaMnO3 (LMO) was deposited using sputtering and GdBCO was deposited using reactive co-evaporation by deposition and reaction (RCE-DR). Detailed X-ray study indicates that LMO/MgO/Y2O3 and GdBCO/LMO/MgO/Y2O3 stack films have good out-of-plane and in-plane textures with strong c-axis alignment. The critical current (Ic) of GdBCO/LMO/MgO/Y2O3 multilayer structure varied from 190 to 420 A/cm with different solutions, when measured at 77 K. These results demonstrated that amorphous yttria can be easily deposited by dip coating using Y-chelates complex as a diffusion barrier and nucleation layer.

  18. Metallic substrate materials for thin film oxygen transport membranes for application in a fossil power plant

    Energy Technology Data Exchange (ETDEWEB)

    Xing, Y.; Baumann, S.; Sebold, D.; Meulenberg, W.A.; Stoever, D. [Forschungszentrum Juelich GmbH (DE). Inst. fuer Energieforschung (IEF) - IEF-1 Materials Synthesis and Processing

    2010-07-01

    La{sub 0.58}Sr{sub 0.4}CO{sub 0.2}Fe{sub 0.8}O{sub 3-{delta}} (LSCF58428) and Ba{sub 0.5}Sr{sub 0.5}CO{sub 0.8}Fe{sub 3-{delta}} (BSCF5582) exhibit high oxygen permeability due to their high ionic and electronic conductivity. For this reason they are under discussion for application in oxygen transport membranes (OTMs) in zero-emission power plants using oxyfuel technology. A thin film membrane which can increase the oxygen flux is beneficial and a structural substrate is required. Two types of Ni-base alloys were studied as substrate material candidates with a number of advantages, such as high strength, high temperature stability, easy joining and similar thermal expansion coefficient to the selected perovskite materials. Chemical compositions and thermal expansion coefficients of Ni-base alloys were measured in this study. LSCF58428 and BSCF5582 layers were screen printed on Ni-based alloys and co-fired at high temperature in air. The microstructure and element analysis of samples were characterized by scanning electron microscopy (SEM and EDX). A Ni-base alloy, MCrAlY, with a high Al content was the most suitable substrate material, and showed better chemical compatibility with perovskite materials at high temperature than Hastelloy X, which is a chromia-forming Ni-base alloy. A reaction occurred between Sr in the perovskite and the alumina surface layers on MCr-AlY. However, the reaction zone did not increase in thickness during medium-term annealing at 800 C in air. Hence, it is expected that this reaction will not prevent the application of MCr-AlY as a substrate material. (orig.)

  19. Research and development activities at INE concerning corrosion of final repository container materials

    International Nuclear Information System (INIS)

    Kienzler, Bernhard

    2017-01-01

    The present work provides a historical overview of the research and development activities carried out at the (Nuclear) Research Center Karlsruhe (today KIT) since the beginning of the 1980s on the corrosion of materials which might be suitable for construction of containers for highly radioactive wastes. The report relates almost exclusively to the work performed by Dr. Emmanuel Smailos, who elaborated the corrosion of various materials at the Institute for Nuclear Waste Disposal (INE). The requirements for the containers and materials, which were subject to changes in time, are presented. The changes were strongly influenced by the changed perception of the use of nuclear energy. The selection of the materials under investigations, the boundary conditions for the corrosion experiments and the analytical methods are described. Results of the corrosion of the materials such as finegrained steel, Hastelloy C4, nodular cast iron, titanium-palladium and copper or copper-nickel alloys in typical salt solutions are summarized. The findings of special investigations, e.g. corrosion under irradiation or the influence of sulfide on the corrosion rates are shown. For construction of disposal canisters, experiments were conducted to determine the contact corrosion, the influence of the hydrogen embrittlement of Ti-Pd and fine-grained steels on the corrosion behavior as well as the corrosion behavior of welding and the influence of different welding processes with the resulting heat-affected zones on the corrosion behavior. The work was contributed to several European research programs and was well recognized in the USA. Investigations on the corrosion of steels in non-saline solutions and corrosion under interim storage conditions as well as under the expected conditions of the Konrad repository for low-level radioactive wastes are also described. In addition, the experiments on ceramic materials are presented and the results of the corrosion of Al 2 O 3 and ZrO 2 ceramics

  20. Chemistry

    International Nuclear Information System (INIS)

    Ferris, L.M.

    1975-01-01

    Research and development activities dealing with the chemical problems related to design and ultimate operation of molten-salt reactor systems are described. An experimental test stand was constructed to expose metallurgical test specimens to Te 2 vapor at defined temperatures and deposition rates. To better define the chemistry of fluoroborate coolant, several aspects are being investigated. The behavior of hydroxy and oxy compounds in molten NaBF 4 is being investigated to define reactions and compounds that may be involved in corrosion and/or could be involved in methods for trapping tritium. Two corrosion products of Hastelloy N, Na 3 CrF 6 and Na 5 Cr 3 F 14 , were identified from fluoroborate systems. The evaluation of fluoroborate and alternate coolants continued. Research on the behavior of hydrogen and its isotopes is summarized. The solubilities of hydrogen, deuterium, and helium in Li 2 BeF 4 are very low. The sorption of tritium on graphite was found to be significant (a few milligrams of tritium per kilogram of graphite), possibly providing a means of sequestering a portion of the tritium produced. Development of analytical methods continued with emphasis on voltammetric and spectrophotometric techniques for the in-line analysis of corrosion products such as Fe 2+ and Cr 3+ and the determination of the U 3+ /U 4+ ratio in MSBR fuel salt. Similar studies were conducted with the NaBF 4 --NaF coolant salt. Information developed during the previous operation of the CSTF has been assessed and used to formulate plans for evaluation of in-line analytical methods in future CSTF operations. Electroanalytical and spectrophotometric research suggests that an electroactive protonic species is present in molten NaBF 4 --NaF, and that this species rapidly equilibrates with a volatile proton-containing species. Data obtained from the CSTF indicated that tritium was concentrated in the volatile species. (JGB)

  1. Summary of INCO corrosion tests in power plant flue gas scrubbing processes

    International Nuclear Information System (INIS)

    Hoxie, E.C.; Tuffnell, G.W.

    1976-01-01

    Corrosion tests in a number of flue-gas desulfurization units have shown that carbon steel, low alloy steels, and Type 304L stainless steel are inadequate in the wet portions of the scrubbers. Type 316L stainless steel is sometimes subject to localized corrosive attack in scrubber environments with certain combinations of pH and chloride content. A corollary is that corrosion of Type 316L stainless steel might be controlled by control of scrubbing media pH and chloride content. Although an attempt was made to correlate the pitting and crevice corrosion obtained on the Type 316 stainless steel test samples with chloride and pH measurements, relatively wide scatter in the data indicated only a modest correlation. This is attributed to variations in local conditions, especially beneath deposits, that differ from the liquor samples obtained for analysis, to processing upsets, to temperature differences, and to some extent to inaccuracies in measurement of pH and chloride levels. The data do show, however, that molybdenum as an alloying element in stainless steels and high nickel alloys was very beneficial in conferring resistance to localized attack in scrubber environments. High nickel alloys containing appreciable amounts of molybdenum such as Hastelloy alloy C-276 and Inconel alloy 625 can be used for critical components. Chloride stress corrosion cracking (SCC) of austenitic stainless steels has generally not been a problem in FGD scrubbers, apparently because operating temperatures are comparatively low. An exception is reheater tubing where some failures have occurred because of elevated temperatures in conjunction with condensate that forms during shut-down periods or carryover of chloride laden mist from the scrubber. This problem can be overcome by proper alloy selection or maintaining dry conditions

  2. Erosion resistance comparison of alternative surface treatments

    Science.gov (United States)

    Česánek, Z.; Schubert, J.; Houdková, Š.

    2017-05-01

    Erosion is a process characterized by the particle separation and the damage of component functional surfaces. Thermal spraying technology HP/HVOF (High Pressure / High Velocity Oxygen Fuel) is commonly used for protection of component surfaces against erosive wear. Alloy as well as cermet based coatings meet the requirements for high erosion resistance. Wear resistance is in many cases the determining property of required component functioning. The application suitability of coating materials is particularly influenced by different hardness. This paper therefore presents an erosion resistance comparison of alloy and cermet based coatings. The coatings were applied on steel substrates and were subjected to the erosive test using the device for evaluation of material erosion resistance working on the principle of centrifugal erodent flow. Abrasive sand Al2O3 with grain size 212-250 μm was selected as an erosive material. For this purpose, the specimens were prepared by thermal spraying technology HP/HVOF using commercially available powders Stellite 6, NiCrBSi, Cr3C2-25%NiCr, Cr3C2-25%CoNiCrAlY, Hastelloy C-276 and experimental coating TiMoCN-29% Ni. Erosion resistance of evaluated coatings was compared with erosive resistance of 1.4923 high alloyed steel without nitridation and in nitrided state and further with surface treatment using technology PVD. According to the evaluation, the resulting erosive resistance depends not only on the selected erodent and surface protection, but also on the erodent impact angle.

  3. A High Integrity Can Design for Degraded Nuclear Fuel

    International Nuclear Information System (INIS)

    Holmes, P.A.

    1999-01-01

    A high integrity can (HIC), designed to meet the ASME Boiler and Pressure Vessel Code (Section III, Div. 3, static conditions) is proposed for the interim storage and repository disposal of Department of Energy (DOE) spent nuclear fuel. The HIC will be approximately 5 3/8 inches (134.38mm) in outside diameter with 1/4 inch (6.35mm) thick walls, and have a removable lid with a metallic seal that is capable of being welded shut. The opening of the can is approximately 4 3/8 inches (111.13mm). The HIC is primarily designed to contain items in the DOE SNF inventory that do not meet acceptance standards for direct disposal in a geologic repository. This includes fuel in the form of particulate dusts, sectioned pieces of fuel, core rubble, melted or degraded (non-intact) fuel elements, unclad uranium alloys, metallurgical specimens, and chemically reactive fuel components. The HIC is intended to act as a substitute cladding for the spent nuclear fuel, further isolate problematic materials, provide a long-term corrosion barrier, and add an extra internal pressure barrier to the waste package. The HIC will also delay potential fission product release and maintain geometry control for extended periods of time. For the entire disposal package to be licensed by the Nuclear Regulatory Commission, a HIC must effectively eliminate the disposal problems associated with problem SNF including the release of radioactive and/or reactive material and over pressurization of the HIC due to chemical reactions within the can. Two HICs were analyzed to envelop a range of can lengths between 42 and 101 inches. Using Abacus software, the HIC's were analyzed for end, side, and corner drops. Hastelloy C-22 was chosen based upon structural integrity, corrosion resistance, and neutron adsorption properties

  4. Development of Plasma-Sprayed Molybdenum Carbide-Based Anode Layers with Various Metal Oxides for SOFC

    Science.gov (United States)

    Faisal, N. H.; Ahmed, R.; Katikaneni, S. P.; Souentie, S.; Goosen, M. F. A.

    2015-12-01

    Air plasma-sprayed (APS) coatings provide an ability to deposit a range of novel fuel cell materials at competitive costs. This work develops three separate types of composite anodes (Mo-Mo2C/Al2O3, Mo-Mo2C/ZrO2, Mo-Mo2C/TiO2) using a combination of APS process parameters on Hastelloy®X for application in intermediate temperature proton-conducting solid oxide fuel cells. Commercially available carbide of molybdenum powder catalyst (Mo-Mo2C) and three metal oxides (Al2O3, ZrO2, TiO2) was used to prepare three separate composite feedstock powders to fabricate three different anodes. Each of the modified composition anode feedstock powders included a stoichiometric weight ratio of 0.8:0.2. The coatings were characterized by scanning electron microscopy, energy dispersive spectroscopy, x-ray diffraction, nanoindentation, and conductivity. We report herein that three optimized anode layers of thicknesses between 200 and 300 µm and porosity as high as 20% for Mo-Mo2C/Al2O3 (250-µm thick) and Mo-Mo2C/TiO2 (300 µm thick) and 17% for Mo-Mo2C/ZrO2 (220-µm thick), controllable by a selection of the APS process parameters with no addition of sacrificial pore-forming material. The nanohardness results indicate the upper layers of the coatings have higher values than the subsurface layers in coatings with some effect of the deposition on the substrate. Mo-Mo2C/ZrO2 shows high electrical conductivity.

  5. Ultra-large current transport in thick SmBa{sub 2}Cu{sub 3}O{sub 7−x} films grown by reactive co-evaporation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, G.; Jin, H.J. [Department of Physics, Ewha Womans University, Seoul 120-750 (Korea, Republic of); Jo, W., E-mail: wmjo@ewha.ac.kr [Department of Physics, Ewha Womans University, Seoul 120-750 (Korea, Republic of); Nam, D.H.; Cheong, H. [Department of Physics, Sogang University, Seoul 121-742 (Korea, Republic of); Kim, H.S.; Oh, S.S.; Ko, R.K.; Jo, Y.S.; Ha, D.W. [The Korea Electrotechnology Research Institute (KERI), Changwon 641-120 (Korea, Republic of)

    2015-06-15

    Highlights: • Transport properties of 5 μm thick SmBa{sub 2}Cu{sub 3}O{sub 7−x} thin films were investigated. • Laser scanning microscopy was used to demonstrate local transport properties. • Temperature variable laser scanning microscopy shows correlation between structural and transport properties. • Optical measurements described nature of current transport properties in the coated conductors. - Abstract: Structural and transport properties of high performance SmBa{sub 2}Cu{sub 3}O{sub 7−x} coated conductors produced by a dual-chamber co-evaporation are presented. The 5 μm-thick SmBCO coated conductors grown on IBAD-MgO based Hastelloy metal templates show critical currents larger than 1020–1560 A/cm at 77 K and self-field. The current transport characteristics of the conductors are investigated by room-temperature thermoelectric microscopy and low-temperature bolometric microscopy. The local thermoelectric images show the tilted grains, grain boundaries, and microstructural defects on the surface of the coated conductors. The bias current-dependent bolometric response at low temperature displays the current of the local flux flow dissipation as an increasing bias. Furthermore, we measured micro-Raman scattering microscopic imaging on oxygen-related peaks of the conductors. Comparing the Raman signal images with the low temperature optical scanning maps, it is remarkable that the structural disorders represented by oxygen-related Raman peaks are closely related to the low temperature bolometric abnormalities. From this result, a nature of the dissipative current distribution in coated conductors is revealed. The scanning optical microscopic study will provide a promising method for quality assurance of coated conductors.

  6. Device for film deposition and implantation of ions inside pipes of low diameter

    International Nuclear Information System (INIS)

    Pogrebnjak, A.D.; Perekrjostov, V.I.; Tyurin, Yu.N.; Wood, B.P.

    2002-01-01

    Two principally new devices, which can be applied to deposit coatings inside the pipes of low diameter, have been developed. The thickness of coatings and films can be varied. To deposit coatings of a low thickness (about 2 nm) on inside pipe walls using a vacuum-arc source and a sputtering device, which is composed of the pipe applied for anode cooling, the constant magnet, the magnetic circuit, the anode, the cathode, the pipe subjected for coating deposition, the cathode holder, etc. Using this device, we have deposited TiC, Ta, Cr, TiN coatings of various thickness ranging from scores of nano-meters to several micro-meters and with very good adhesion to the substrate. To increase adhesion, we applied 10 to 20 kV voltage during ion implantation to the substrate. To study element and structure composition, we applied RBS, TEM, SEM, XRD analyses, micro-hardness, wear resistance tests and also those for corrosion resistance in acid media. Another version of the source was based on the pulsed plasma-detonation technology and applied an evaporating electrode (for implantation) and a powder, which was injected into a plasma jet. The jet velocity reached several kilometers per second. Current of several kilo-amps passed through the plasma jet and increased its energy. The produced in this way coating thickness reached 30 to 400 micro-meter. Application of the vacuum-arc source for subsequent coating deposition allowed us to improve the servicing characteristics of surface layers. We have deposited NiAl, CoAl, A1 2 O 3 , WC-Co, Hastelloy and stainless steel SS316L

  7. Engineering Database of Liquid Salt Thermophysical and Thermochemical Properties

    Energy Technology Data Exchange (ETDEWEB)

    Manohar S. Sohal; Matthias A. Ebner; Piyush Sabharwall; Phil Sharpe

    2010-03-01

    The purpose of this report is to provide a review of thermodynamic and thermophysical properties of candidate molten salt coolants, which may be used as a primary coolant within a nuclear reactor or heat transport medium from the Next Generation Nuclear Plant (NGNP) to a processing plant, for example, a hydrogen-production plant. Thermodynamic properties of four types of molten salts, including LiF-BeF2 (67 and 33 mol%, respectively; also known as FLiBe), LiF-NaF-KF (46.5, 11.5, and 52 mol%, also known as FLiNaK), and KCl-MgCl2 (67 and 33 mol%), and sodium nitrate-sodium nitrite-potassium nitrate (NaNO3–NaNO2–KNO3, (7-49-44 or 7-40-53 mol%) have been investigated. Limitations of existing correlations to predict density, viscosity, specific heat capacity, surface tension, and thermal conductivity, were identified. The impact of thermodynamic properties on the heat transfer, especially Nusselt number was also discussed. Stability of the molten salts with structural alloys and their compatibility with the structural alloys was studied. Nickel and alloys with dense Ni coatings are effectively inert to corrosion in fluorides but not so in chlorides. Of the chromium containing alloys, Hastelloy N appears to have the best corrosion resistance in fluorides, while Haynes 230 was most resistant in chloride. In general, alloys with increasing carbon and chromium content are increasingly subject to corrosion by the fluoride salts FLiBe and FLiNaK, due to attack and dissolution of the intergranular chromium carbide. Future research to obtain needed information was identified.

  8. Otimização de Parâmetros do Processo de Soldagem Arco Submerso para Revestimentos Anticorrosivos

    Directory of Open Access Journals (Sweden)

    Marcos Mesquita da Silva

    2016-03-01

    Full Text Available Resumo Neste trabalho foi avaliada a influência das variáveis de soldagem em cordões de solda, aplicados pelo processo Arco Submerso (SAW com corrente convencional, visando futura aplicação em revestimentos metálicos contra corrosão. Segmentos de tubo de um aço API 5L Gr B foram utilizados como substrato, além da utilização de um metal de adição de liga de níquel, diâmetro de 1,13mm, classificação AWS ERNiCrMo-4 (Hastelloy C-276, e um fluxo do tipo neutro, básico e aglomerado (EN 760: SA AF 2 DC. Foram analisadas as variáveis tensão, velocidade de alimentação de arame e distância do bico de contato à peça (DBCP – com o restante dos parâmetros constantes – através de um planejamento fatorial completo em dois níveis e pontos centrais. Os resultados mostraram modelos matemáticos estatisticamente significativos e preditivos para as respostas diluição e corrente média, IM. Porém, para a resposta Reforço/Largura (R/L, o modelo se caracterizou como estatisticamente significativo, não preditivo e contendo uma falta de ajuste. A DBCP, por sua vez, foi a variável que se constituiu como a de maior significância na redução da diluição.

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

    International Nuclear Information System (INIS)

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

    2006-12-01

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

  10. 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

  11. Production of Strontium-90 Thermal Power Sources; Fabrication de sources d'energie thermique au strontium-90; Proizvodstvo istochnikov ''teplovoj ehnergii iz Sr''9''0; Preparacion de fuentes de energia termica con estroncio-90

    Energy Technology Data Exchange (ETDEWEB)

    Cochran, J. S.; Bloom, J. L.; Schneider, A. [Martin Company, Nuclear Division, Baltimore 3, MD (United States)

    1963-11-15

    One of the most attractive fields for utilization of large quantities of waste fission products is the field of direct-conversion power supplies for remote locations. Strontium-90 is being given the greatest exploitation because of its availability, nuclear properties, and the relative ease with which it can be fabricated into compact heat sources. Strontium-90 fuelled generators are being used to power automatic weather stations and navigational aids, and consideration is being given to the use of strontium-90 as a power source for space vehicles. Evaluation of several potentially useful strontium compounds led to the selection of the titanate as exhibiting overall properties most desirable for this purpose. Strontium-90, separated from crude fission product streams and purified to the requisite degree by the USAEC's Hanford Works, is shipped in the form of the carbonate to a hot cell facility operated by the Martin Company, where it is converted to titanate pellets. This process is an adaption to remote operation of conventional chemical and ceramic techniques. The pellets are encapsulated in Hastelloy C containers for use in thermoelectric power supplies. Unusual operational problems are encountered because the large quantities of strontium-90 handled (potentially millions of curies per year) represent formidable radiation and contamination hazards. Details of the facility, equipment, process, and safety criteria are given. The operational experience gained during the recent processing of the first 250 000 curies of strontium-90 into fuel for a SNAP-7 generator is described. Encapsulation, calorimetry, decontamination, and waste disposal procedures are also outlined. (author) [French] L'une des utilisations les plus interessantes des produits de fission en grande quantite consiste a les employer comme sources d'energie par combustion directe pour des installations geographiquement isolees. C'est le strontium-90 qui est l e plus utilise parce qu'on en dispose en

  12. Present status and future prospect of coated conductor development and its application in Japan

    Science.gov (United States)

    Shiohara, Y.; Yoshizumi, M.; Izumi, T.; Yamada, Y.

    2008-03-01

    The current national project on coated conductors using Y-system superconductors has been carried out over the project period (FY2003-FY2007). In this paper, the current status and the future prospect of this project are reviewed. The high performance tape development group, consisting of Fujikura and SRL-NCCC, has worked on the tape by PLD-REBCO superconducting layers on PLD-CeO2/IBAD-GZO buffered substrates. A high product of Ic and L, higher than 112 166 A m, was achieved in a 368 m-304.8 A GdBCO tape whose Ic value is mostly above 350 A/cm in width. The performance under magnetic field was also improved up to 42 A at 3 T in a GdBCO short film with doping of ZrO2. 61 m long GdBCO tape with ZrO2 doping showed a high Ic value of 220 A at self field and 30 A at 3 T. On the other hand, another group focusing on low production cost has worked on TFA-MOD and MOCVD processes. The extremely high Ic value of 735 A/cm-w was obtained in TFA-MOD films on PLD-CeO2/IBAD-GZO/Hastelloy substrate due to the effect of Ba-poor nominal composition. In efforts towards long tape production by the SWCC group, a 200 m long tape with a high Ic value of 200 A/cm-w was obtained using a batch-type furnace. The Ic × L value of this tape was 40 000 A m, which is the highest value in the world obtained by the TFA-MOD process. Based on the above achievements in coated conductor process development, two new additional goals were set in the project. One is the development of extremely low cost tape and the other is the development of the basic technologies for making electric power devices of cables, transformers, motors, current-limiters and cryocoolers. Some of the new investigations have already revealed marvellous results, such as a 15 kW motor, low AC loss coils, low AC loss cables, etc.

  13. Present status and future prospect of coated conductor development and its application in Japan

    International Nuclear Information System (INIS)

    Shiohara, Y; Yoshizumi, M; Izumi, T; Yamada, Y

    2008-01-01

    The current national project on coated conductors using Y-system superconductors has been carried out over the project period (FY2003-FY2007). In this paper, the current status and the future prospect of this project are reviewed. The high performance tape development group, consisting of Fujikura and SRL-NCCC, has worked on the tape by PLD-REBCO superconducting layers on PLD-CeO 2 /IBAD-GZO buffered substrates. A high product of I c and L, higher than 112 166 A m, was achieved in a 368 m-304.8 A GdBCO tape whose I c value is mostly above 350 A/cm in width. The performance under magnetic field was also improved up to 42 A at 3 T in a GdBCO short film with doping of ZrO 2 . 61 m long GdBCO tape with ZrO 2 doping showed a high I c value of 220 A at self field and 30 A at 3 T. On the other hand, another group focusing on low production cost has worked on TFA-MOD and MOCVD processes. The extremely high I c value of 735 A/cm-w was obtained in TFA-MOD films on PLD-CeO 2 /IBAD-GZO/Hastelloy substrate due to the effect of Ba-poor nominal composition. In efforts towards long tape production by the SWCC group, a 200 m long tape with a high I c value of 200 A/cm-w was obtained using a batch-type furnace. The I c x L value of this tape was 40 000 A m, which is the highest value in the world obtained by the TFA-MOD process. Based on the above achievements in coated conductor process development, two new additional goals were set in the project. One is the development of extremely low cost tape and the other is the development of the basic technologies for making electric power devices of cables, transformers, motors, current-limiters and cryocoolers. Some of the new investigations have already revealed marvellous results, such as a 15 kW motor, low AC loss coils, low AC loss cables, etc

  14. Present status and future prospect of coated conductor development and its application in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Shiohara, Y; Yoshizumi, M; Izumi, T [Superconductivity Research Laboratory-ISTEC, 10-13 Shinonome 1-Chome, Koto-ku, Tokyo 135-0062 (Japan); Yamada, Y [Nagoya Coated Conductor Center, Superconductivity Research Laboratory-ISTEC, (c-o) Japan Fine Ceramics Center, 4-1 Mutsuno 2-Chome, Atsuta-ku, Nagoya 456-8587 (Japan)], E-mail: shiohara@istec.or.jp

    2008-03-01

    The current national project on coated conductors using Y-system superconductors has been carried out over the project period (FY2003-FY2007). In this paper, the current status and the future prospect of this project are reviewed. The high performance tape development group, consisting of Fujikura and SRL-NCCC, has worked on the tape by PLD-REBCO superconducting layers on PLD-CeO{sub 2}/IBAD-GZO buffered substrates. A high product of I{sub c} and L, higher than 112 166 A m, was achieved in a 368 m-304.8 A GdBCO tape whose I{sub c} value is mostly above 350 A/cm in width. The performance under magnetic field was also improved up to 42 A at 3 T in a GdBCO short film with doping of ZrO{sub 2}. 61 m long GdBCO tape with ZrO{sub 2} doping showed a high I{sub c} value of 220 A at self field and 30 A at 3 T. On the other hand, another group focusing on low production cost has worked on TFA-MOD and MOCVD processes. The extremely high I{sub c} value of 735 A/cm-w was obtained in TFA-MOD films on PLD-CeO{sub 2}/IBAD-GZO/Hastelloy substrate due to the effect of Ba-poor nominal composition. In efforts towards long tape production by the SWCC group, a 200 m long tape with a high I{sub c} value of 200 A/cm-w was obtained using a batch-type furnace. The I{sub c} x L value of this tape was 40 000 A m, which is the highest value in the world obtained by the TFA-MOD process. Based on the above achievements in coated conductor process development, two new additional goals were set in the project. One is the development of extremely low cost tape and the other is the development of the basic technologies for making electric power devices of cables, transformers, motors, current-limiters and cryocoolers. Some of the new investigations have already revealed marvellous results, such as a 15 kW motor, low AC loss coils, low AC loss cables, etc.

  15. Characterization of amorphous yttria layers deposited by aqueous solutions of Y-chelate alkoxides complex

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young-Soon, E-mail: kyscjb@i-sunam.com; Lee, Yu-Ri; Kim, Byeong-Joo; Lee, Jae-Hun; Moon, Seung-Hyun; Lee, Hunju

    2015-01-15

    Highlights: • Economical method for crack-free amorphous yttria layer deposition by dip coating. • Simpler process for planar yttria film as a diffusion barrier and nucleation layer. • Easy control over the film properties with better characteristics. • Easy control over the thickness of the deposited films. • A feasible process that can be easily adopted by HTSCC industries. - Abstract: Crack-free amorphous yttria layers were deposited by dip coating in solutions of different Y-chelate alkoxides complex. Three Y-chelate solutions of different concentrations were prepared using yttrium acetate tetrahydrate, yttrium stearic acid as Y source materials. PEG, diethanolamine were used as chelating agents, while ethanol, methanol and tetradecane were used as solvent. Three different combinations of chelating and solvents were used to prepare solutions for Y{sub 2}O{sub 3} dip coating on SUS, electropolished and non-electropolished Hastelloy C-276 substrates. The thickness of the films was varied by changing the number of dipping cycles. At an optimized condition, the substrate surface roughness (rms) value was reduced from ∼50 nm to ∼1 nm over a 10 × 10 μm{sup 2} area. After Y{sub 2}O{sub 3} deposition, MgO was deposited using ion-beam assisted deposition (IBAD), then LaMnO{sub 3} (LMO) was deposited using sputtering and GdBCO was deposited using reactive co-evaporation by deposition and reaction (RCE-DR). Detailed X-ray study indicates that LMO/MgO/Y{sub 2}O{sub 3} and GdBCO/LMO/MgO/Y{sub 2}O{sub 3} stack films have good out-of-plane and in-plane textures with strong c-axis alignment. The critical current (Ic) of GdBCO/LMO/MgO/Y{sub 2}O{sub 3} multilayer structure varied from 190 to 420 A/cm with different solutions, when measured at 77 K. These results demonstrated that amorphous yttria can be easily deposited by dip coating using Y-chelates complex as a diffusion barrier and nucleation layer.

  16. 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

  17. Molten Salt Fuel Version of Laser Inertial Fusion Fission Energy (LIFE)

    International Nuclear Information System (INIS)

    Moir, R.W.; Shaw, H.F.; Caro, A.; Kaufman, L.; Latkowski, J.F.; Powers, J.; Turchi, P.A.

    2008-01-01

    Molten salt with dissolved uranium is being considered for the Laser Inertial Confinement Fusion Fission Energy (LIFE) fission blanket as a backup in case a solid-fuel version cannot meet the performance objectives, for example because of radiation damage of the solid materials. Molten salt is not damaged by radiation and therefore could likely achieve the desired high burnup (>99%) of heavy atoms of 238 U. A perceived disadvantage is the possibility that the circulating molten salt could lend itself to misuse (proliferation) by making separation of fissile material easier than for the solid-fuel case. The molten salt composition being considered is the eutectic mixture of 73 mol% LiF and 27 mol% UF 4 , whose melting point is 490 C. The use of 232 Th as a fuel is also being studied. ( 232 Th does not produce Pu under neutron irradiation.) The temperature of the molten salt would be ∼550 C at the inlet (60 C above the solidus temperature) and ∼650 C at the outlet. Mixtures of U and Th are being considered. To minimize corrosion of structural materials, the molten salt would also contain a small amount (∼1 mol%) of UF 3 . The same beryllium neutron multiplier could be used as in the solid fuel case; alternatively, a liquid lithium or liquid lead multiplier could be used. Insuring that the solubility of Pu 3+ in the melt is not exceeded is a design criterion. To mitigate corrosion of the steel, a refractory coating such as tungsten similar to the first wall facing the fusion source is suggested in the high-neutron-flux regions; and in low-neutron-flux regions, including the piping and heat exchangers, a nickel alloy, Hastelloy, would be used. These material choices parallel those made for the Molten Salt Reactor Experiment (MSRE) at ORNL. The nuclear performance is better than the solid fuel case. At the beginning of life, the tritium breeding ratio is unity and the plutonium plus 233 U production rate is ∼0.6 atoms per 14.1 MeV neutron

  18. Standard partial molar heat capacities and enthalpies of formation of aqueous aluminate under hydrothermal conditions from integral heat of solution measurements

    International Nuclear Information System (INIS)

    Coulier, Yohann; Tremaine, Peter R.

    2014-01-01

    Highlights: • Heats of solution of NaAlO 2 (s) were measured at five temperatures up to 250 °C. • Standard molar enthalpies of solution were determined from the measured heats of solution. • Standard molar enthalpies of solution were correlated with the density model. • The density model allows us to determine the standard molar heat capacities of reaction. - Abstract: Heats of solution of sodium aluminum oxide, NaAlO 2 (s), were measured in aqueous sodium hydroxide solutions using a Tian–Calvet heat-flow calorimeter (Setaram, Model C80) with high pressure “batch cells” made of hastelloy C-276, at five temperatures from (373.15 to 523.15) K, steam saturation pressure, and concentrations from (0.02 to 0.09) mol · kg −1 . Standard molar enthalpies of solution, Δ soln H ∘ , and relative standard molar enthalpies, [H ∘ (T) − H ∘ (298.15 K)], of NaAl(OH) 4 (aq) were determined from the measured heats of solution. The results were fitted with the “density” model. The temperature dependence of Δ soln H ∘ from the model yielded the standard molar heat capacities of reaction, Δ soln C p ∘ , from which standard partial molar heat capacities for aqueous aluminate, C p ∘ [A1(OH) 4 − ,aq], were calculated. Standard partial molar enthalpies of formation, Δ f H ∘ , and entropies, S ∘ , of A1(OH) 4 − (aq) were also determined. The values for C p ∘ [A1(OH) 4 − ,aq] agree with literature data determined up to T = 413 K from enthalpy of solution and heat capacity measurements to within the combined experimental uncertainties. They are consistent with differential heat capacity measurements up to T = 573 K from Schrödle et al. (2010) [29] using the same calorimeter, but this method has the advantage that measurements could be made at much lower concentrations in the presence of an excess concentration of ligand. To our knowledge, these are the first standard partial molar heat capacities measured under hydrothermal conditions by the

  19. Additive Manufacturing of Fuel Injectors

    Energy Technology Data Exchange (ETDEWEB)

    Sadek Tadros, Dr. Alber Alphonse [Edison Welding Institute, Inc., Columbus, OH (United States); Ritter, Dr. George W. [Edison Welding Institute, Inc., Columbus, OH (United States); Drews, Charles Donald [Edison Welding Institute, Inc., Columbus, OH (United States); Ryan, Daniel [Solar Turbines Inc., San Diego, CA (United States)

    2017-10-24

    Additive manufacturing (AM), also known as 3D-printing, has been shifting from a novelty prototyping paradigm to a legitimate manufacturing tool capable of creating components for highly complex engineered products. An emerging AM technology for producing metal parts is the laser powder bed fusion (L-PBF) process; however, industry manufacturing specifications and component design practices for L-PBF have not yet been established. Solar Turbines Incorporated (Solar), an industrial gas turbine manufacturer, has been evaluating AM technology for development and production applications with the desire to enable accelerated product development cycle times, overall turbine efficiency improvements, and supply chain flexibility relative to conventional manufacturing processes (casting, brazing, welding). Accordingly, Solar teamed with EWI on a joint two-and-a-half-year project with the goal of developing a production L-PBF AM process capable of consistently producing high-nickel alloy material suitable for high temperature gas turbine engine fuel injector components. The project plan tasks were designed to understand the interaction of the process variables and their combined impact on the resultant AM material quality. The composition of the high-nickel alloy powders selected for this program met the conventional cast Hastelloy X compositional limits and were commercially available in different particle size distributions (PSD) from two suppliers. Solar produced all the test articles and both EWI and Solar shared responsibility for analyzing them. The effects of powder metal input stock, laser parameters, heat treatments, and post-finishing methods were evaluated. This process knowledge was then used to generate tensile, fatigue, and creep material properties data curves suitable for component design activities. The key process controls for ensuring consistent material properties were documented in AM powder and process specifications. The basic components of the project

  20. Mode-I Fracture Toughness Testing and Coupled Cohesive Zone Modeling at In Situ P, T, and Chemical (H2O-CO2-NaCl) Conditions

    Science.gov (United States)

    Dewers, T. A.; Choens, R. C., II; Regueiro, R. A.; Eichhubl, P.; Bryan, C. R.; Rinehart, A. J.; Su, J. C.; Heath, J. E.

    2017-12-01

    Propagation of mode I cracks is fundamental to subsurface engineering endeavors, but the majority of fracture toughness measurements are performed at ambient conditions. A novel testing apparatus was used to quantify the relationship between supercritical carbon dioxide (scCO2), water vapor, and fracture toughness in analogs for reservoir rock and caprock lithologies at temperature and pressure conditions relevant to geologic carbon storage. Samples of Boise Sandstone and Marcellus Shale were subject to fracture propagation via a novel short rod fracture toughness tester composed of titanium and Hastelloy® and designed to fit inside a pressure vessel. The tester is controlled by a hydraulically-driven ram and instrumented with a LVDT to monitor displacement. We measure fracture toughness under conditions of dry supercritical CO2 (scCO2), scCO2-saturated brine, and scCO2 with varying water content ( 25%, 90%, and 100% humidity) at 13.8 MPa and 70oC. Water film development as a function of humidity is determined in situ during the experiments with a quartz crystal microbalance. Two orientations of the Marcellus are included in the testing matrix. Dry CO2 has a negligible to slightly strengthening effect compared to a control, however hydrous scCO2 can decrease the fracture toughness, and the effect increases with increasing humidity, which likely is due to capillary condensation of reactive water films at nascent crack tips and associated subcritical weakening. A 2D poromechanical finite element model with cohesive surface elements (CSEs) and a chemo-plasticity phenomenology is being used to describe the chemical weakening/softening effects observed in the testing. The reductions in fracture toughness seen in this study could be important in considerations of borehole stability, in situ stress measurements, changes in fracture gradient, and reservoir caprock integrity during CO2 injection and storage. Sandia National Laboratories is a multimission laboratory managed

  1. Assessment of Embrittlement of VHTR Structural Alloys in Impure Helium Environments

    Energy Technology Data Exchange (ETDEWEB)

    Crone, Wendy; Cao, Guoping; Sridhara, Kumar

    2013-05-31

    The helium coolant in high-temperature reactors inevitably contains low levels of impurities during steady-state operation, primarily consisting of small amounts of H{sub 2}, H{sub 2}O, CH{sub 4}, CO, CO{sub 2}, and N{sub 2} from a variety of sources in the reactor circuit. These impurities are problematic because they can cause significant long-term corrosion in the structural alloys used in the heat exchangers at elevated temperatures. Currently, the primary candidate materials for intermediate heat exchangers are Alloy 617, Haynes 230, Alloy 800H, and Hastelloy X. This project will evaluate the role of impurities in helium coolant on the stress-assisted grain boundary oxidation and creep crack growth in candidate alloys at elevated temperatures. The project team will: • Evaluate stress-assisted grain boundary oxidation and creep crack initiation and crack growth in the temperature range of 500-850°C in a prototypical helium environment. • Evaluate the effects of oxygen partial pressure on stress-assisted grain boundary oxidation and creep crack growth in impure helium at 500°C, 700°C, and 850°C respectively. • Characterize the microstructure of candidate alloys after long-term exposure to an impure helium environment in order to understand the correlation between stress-assisted grain boundary oxidation, creep crack growth, material composition, and impurities in the helium coolant. • Evaluate grain boundary engineering as a method to mitigate stress-assisted grain boundary oxidation and creep crack growth of candidate alloys in impure helium. The maximum primary helium coolant temperature in the high-temperature reactor is expected to be 850-1,000°C.Corrosion may involve oxidation, carburization, or decarburization mechanisms depending on the temperature, oxygen partial pressure, carbon activity, and alloy composition. These corrosion reactions can substantially affect long-term mechanical properties such as crack- growth rate and fracture

  2. Passivation of fluorinated activated charcoal

    International Nuclear Information System (INIS)

    Del Cul, G.D.; Trowbridge, L.D.; Simmons, D.W.; Williams, D.F.; Toth, L.M.

    1997-10-01

    The Molten Salt Reactor Experiment (MSRE), at the Oak Ridge National Laboratory has been shut down since 1969 when the fuel salt was drained from the core into two Hastelloy N tanks at the reactor site. In 1995, a multiyear project was launched to remediate the potentially hazardous conditions generated by the movement of fissile material and reactive gases from the storage tanks into the piping system and an auxiliary charcoal bed (ACB). The top 12 in. of the ACB is known by gamma scan and thermal analysis to contain about 2.6 kg U-233. According to the laboratory tests, a few feet of fluorinated charcoal are believed to extend beyond the uranium front. The remainder of the ACB should consist of unreacted charcoal. Fluorinated charcoal, when subjected to rapid heating, can decompose generating gaseous products. Under confined conditions, the sudden exothermic decomposition can produce high temperatures and pressures of near-explosive characteristics. Since it will be necessary to drill and tap the ACB to allow installation of piping and instrumentation for remediation and recovery activities, it is necessary to chemically convert the reactive fluorinated charcoal into a more stable material. Ammonia can be administered to the ACB as a volatile denaturing agent that results in the conversion of the C x F to carbon and ammonium fluoride, NH 4 F. The charcoal laden with NH 4 F can then be heated without risking any sudden decomposition. The only consequence of heating the treated material will be the volatilization of NH 4 F as a mixture of NH 3 and HF, which would primarily recombine as NH 4 F on surfaces below 200 C. The planned scheme for the ACB denaturing is to flow diluted ammonia gas in steps of increasing NH 3 concentration, 2% to 50%, followed by the injection of pure ammonia. This report summarizes the planned passivation treatment scheme to stabilize the ACB and remove the potential hazards. It also includes basic information, results of laboratory tests

  3. HTS current lead units prepared by the TFA-MOD processed YBCO coated conductors

    International Nuclear Information System (INIS)

    Shiohara, K.; Sakai, S.; Ishii, Y.; Yamada, Y.; Tachikawa, K.; Koizumi, T.; Aoki, Y.; Hasegawa, T.; Tamura, H.; Mito, T.

    2010-01-01

    Two superconducting current lead units have been prepared using ten coated conductors of the Tri-Fluoro-Acetate - Metal Organic Deposition (TFA-MOD) processed Y 1 Ba 2 Cu 3 O 7-δ (YBCO) coated conductors with critical current (I c ) of about 170 A at 77 K in self-field. The coated conductors are 5 mm in width, 190 mm in length and about 120 μm in overall thickness. The 1.5 μm thick superconducting YBCO layer was synthesized through the TFA-MOD process on Hastelloy TM C-276 substrate tape with two buffer oxide layers of Gd 2 Zr 2 O 7 and CeO 2 . The five YBCO coated conductors are attached on a 1 mm thick Glass Fiber Reinforced Plastics (GFRP) board and soldered to Cu caps at the both ends. We prepared two 500 A-class current lead units. The DC transport current of 800 A was stably applied at 77 K without any voltage generation in all coated conductors. The voltage between both Cu caps linearly increased with increasing the applied current, and was about 350 μV at 500 A in both current lead units. According to the estimated values of the heat leakage from 77 K to 4.2 K, the heat leakage for the current lead unit was 46.5 mW. We successfully attained reduction of the heat leakage because of improvement of the transport current performance (I c ), a thinner Ag layer of YBCO coated conductor and usage of the GFRP board for reinforcement instead of a stainless steel board used in the previous study. The DC transport current of 1400 A was stably applied when the two current lead units were joined in parallel. The sum of the heat leakages from 77 K to 4.2 K for the combined the current lead units was 93 mW. In comparison with the conventional Cu current leads by gas-cooling, it could be noted that the heat leakage of the current lead is about one order of magnitude smaller than that of the Cu current lead.

  4. Structural concept of angle type of hot isolation valve and its test program at an out-of-pile test facility

    Energy Technology Data Exchange (ETDEWEB)

    Hada, Kazuhiko; Fujisaki, Katsuo; Shibata, Taijyu; Inagaki, Yoshiyuki; Hino, Ryutaro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Koiso, Hiroshi

    1997-02-01

    The Japanese safety regulation generally requires to set an isolation valve at the penetration of the reactor containment vessel on the secondary helium piping system which connects a steam reforming hydrogen production system, located outside the reactor building, to an intermediate heat exchanger (IHX) in the HTTR reactor system. The hot secondary helium which is heated up to the high temperature of 905degC and at the high pressure of 4.1MPa is passing through the isolation valve. So far, such a hot isolation valve has not been industrialized. The present report presents a proposal of a structural design concept of an angle valve as a promising candidate of the hot isolation valve, and a proposal on a test program for demonstrating the technological feasibility of the concept at an out-of-pile test facility before installing at the HTTR. A closing time and a leak rate at a valve seat are the key design parameters for developing the design concept. To set a reasonable value to each parameter, safety requirements on the isolation valve were discussed at first. The target closing time and the acceptable design limit of leak rate at the valve seat for meeting the requirements were specified 30 seconds and 10 STP cm{sup 3}/s, respectively. A nickel-base superalloy Hastelloy XR is feasible as such a valve seat material as to withstand the internal/external pressure of 4.1MPa at the high temperature of 905degC, the severest loading conditions of the valve seat at the accident of secondary helium pipe rupture. Correlation of leak rate at the ambient temperature to that at an operating temperature (900degC) is one of key test subjects of test program at an out-of-pile test facility. Leak rate at the operating temperature is the real parameter to be checked but only the leak rate at the ambient temperature is measured at regulatory examination in service. A test method to develop such correlation was proposed. (author)

  5. Evaluation of electrochemical techniques for measurement of fireside corrosion in thermal power plants; Utvaerdering av elektrokemiska tekniker foer maetning av roekgassidig hoegtemperaturkorrosion i pannor

    Energy Technology Data Exchange (ETDEWEB)

    Hjorrnhede, Anders

    2007-12-15

    The possibility to measure the corrosion rate on-line, in situ, is getting more and more interesting due to increased corrosion. The worsening may be a result of lower fuel quality or increased steam data due to raised efficiency demands. Also the use of inhibitors can be improved and the lifetime of important components can be increased. Today, virtually all corrosion measurements of materials used for waterwalls, superheaters, economisers and other heat-transferring surfaces are based on in-situ probe tests of coupons or rings. The aim of the project is to evaluate the practicability of commercial or semicommercial on-line in-situ corrosion probes for use in waste fired boilers. The target groups are owners of boilers, operators, service personnel but also boiler manufacturer and material producers. Since the use of on-line, in situ, corrosion probes is substantial, some of the most promising corrosion probes have been tested in a waste fired boiler in Hamburg, Germany. Tests in waste fired boiler have never before been performed. The MECO CB, a Linear Polarisation Resistance (LPR) corrosion probe from Coresto Oy, Finland was tested and from Lehrstuhl fuer Umweltverfahrenstechnik und Anlagentechnik (LUAT) der Universitaet Duisburg-Essen, Germany an Electrochemical Noise (EN) - probe was tested. From Pepperl + Fuchs both EN- and LPR-CorrTran corrosion sensors were tested. The test periods were lasting from 1050h to 3750h. The test materials were the low alloyed steel 15Mo3, the Ni-based super alloys Sanicro 63 and Haynes Hastelloy C-2000. The fluegas temperatures were 635 deg C or 520 deg C. The material temperatures were 440 deg C, 420 deg, 350 deg C and was swept from 400 deg C to 300 deg C. All probes are measuring a signal which has a correlation to the corrosion process, but the quantification procedure is not working well. The results achieved from the corrosion probes must be calibrated against corrosion rates measured by means of conventional corrosion

  6. Carburization of austenitic alloys by gaseous impurities in helium

    International Nuclear Information System (INIS)

    Lai, G.Y.; Johnson, W.R.

    1980-03-01

    The carburization behavior of Alloy 800H, Inconel Alloy 617 and Hastelloy Alloy X in helium containing various amounts of H 2 , CO, CH 4 , H 2 O and CO 2 was studied. Corrosion tests were conducted in a temperature range from 649 to 1000 0 C (1200 to 1832 0 F) for exposure time up to 10,000 h. Four different helium environments, identified as A, B, C, and D, were investigated. Concentrations of gaseous impurities were 1500 μatm H 2 , 450 μatm CO, 50 μatm CH 4 and 50 μatm H 2 O for Environment A; 200 μatm H 2 , 100 μatm CO, 20 μatm CH 4 , 50 μatm H 2 O and 5 μatm CO 2 for Environment B; 500 μatm H 2 , 50 μatm CO, 50 μatm CH 4 and 2 O for Environment C; and 500 μatm H 2 , 50 μatm CO, 50 μatm CH 4 and 1.5 μatm H 2 O for Environment D. Environments A and B were characteristic of high-oxygen potential, while C and D were characteristic of low-oxygen potential. The results showed that the carburization kinetics in low-oxygen potential environments (C and D) were significantly higher, approximately an order of magnitude higher at high temperatures, than those in high-oxygen potential environments (A and B) for all three alloys. Thermodynamic analyses indicated no significant differences in the thermodynamic carburization potential between low- and high-oxygen potential environments. It is thus believed that the enhanced carburization kinetics observed in the low-oxygen potential environments were related to kinetic effects. A qualitatively mechanistic model was proposed to explain the enhanced kinetics. The present results further suggest that controlling the oxygen potential of the service environment can be an effective means of reducing carburization of alloys

  7. Development of thin film oxygen transport membranes on metallic supports

    Energy Technology Data Exchange (ETDEWEB)

    Xing, Ye

    2012-04-25

    Asymmetric membrane structure has an attractive potential in the application of O{sub 2}/N{sub 2} gas separation membrane for the future membrane-based fossil fuel power plant using oxyfuel technology, which will reduce the carbon dioxide emission. The aim of this study is the development of a metal supported multi-layer membrane structure with a thin film top membrane layer and porous ceramic interlayers. Four perovskite materials were studied as candidate membrane materials. Material properties of these perovskite materials were investigated and compared. La{sub 0.58}Sr{sub 0.4}Co{sub 0.2}Fe{sub 0.8}O{sub 3-{delta}} (LSCF58428) showed sufficient oxygen permeability, an acceptable thermal expansion coefficient and a moderate sintering temperature. Alternatively, Ba{sub 0.5}Sr{sub 0.5}Co{sub 0.8}Fe{sub 0.2}O{sub 3-{delta}} (BSCF5582) is considered obtaining very high oxygen permeability but a higher thermal expansion and a lower thermal stability than LSCF58428. Four different Ni-based alloys were studied as candidate substrate materials in the asymmetric membrane structure. The chromia-scale alloys (Hastelloy X, Inconel 600 and Haynes 214) caused Cr poisoning of the membrane layer material LSCF58428 during high-temperature co-firing in air. NiCoCrAlY with a high Al content (12.7 wt%) was found to be the most promising substrate material. It showed a good chemical compatibility with perovskite materials at high temperatures. In order to bridge the highly porous substrate and the thin top membrane layer interlayers were developed. Two interlayers were coated by screen printing on the porous NiCoCrAlY substrate which was sintered at 1225 C in flowing H{sub 2} atmosphere. Screen printing pastes were optimized by investigating various solvent and binder combinations and various ceramic powder contents. The first interlayer significantly improved the surface quality and the surface pore size has been reduced from 30-50{mu}m on the substrate to few {mu}m on the first

  8. Engineering Model Propellant Feed System Development for an Iodine Hall Thruster Demonstration Mission

    Science.gov (United States)

    Polzin, Kurt A.

    2016-01-01

    CUBESATS are relatively new spacecraft platforms that are typically deployed from a launch vehicle as a secondary payload, providing low-cost access to space for a wide range of end-users. These satellites are comprised of building blocks having dimensions of 10x10x10 cu cm and a mass of 1.33 kg (a 1-U size). While providing low-cost access to space, a major operational limitation is the lack of a propulsion system that can fit within a CubeSat and is capable of executing high (Delta)v maneuvers. This makes it difficult to use CubeSats on missions requiring certain types of maneuvers (i.e. formation flying, spacecraft rendezvous). Recently, work has been performed investigating the use of iodine as a propellant for Hall-effect thrusters (HETs) 2 that could subsequently be used to provide a high specific impulse path to CubeSat propulsion. 3, 4 Iodine stores as a dense solid at very low pressures, making it acceptable as a propellant on a secondary payload. It has exceptionally high ?Isp (density times specific impulse), making it an enabling technology for small satellite near-term applications and providing the potential for systems-level advantages over mid-term high power electric propulsion options. Iodine flow can also be thermally regulated, subliming at relatively low temperature (engineering model propellant feed system for iSAT (see Fig. 1). The feed system is based around an iodine propellant reservoir and two proportional control valves (PFCVs) that meter the iodine flow to the cathode and anode. The flow is split upstream of the PFCVs to both components can be fed from a common reservoir. Testing of the reservoir is reported to demonstrate that the design is capable of delivering the required propellant flow rates to operate the thruster. The tubing and reservoir are fabricated from hastelloy to resist corrosion by the heated gaseous iodine propellant. The reservoir, tubing, and PFCVs are heated to ensure the sublimed propellant will not re

  9. Passivation of fluorinated activated charcoal

    Energy Technology Data Exchange (ETDEWEB)

    Del Cul, G.D.; Trowbridge, L.D.; Simmons, D.W.; Williams, D.F.; Toth, L.M.

    1997-10-01

    The Molten Salt Reactor Experiment (MSRE), at the Oak Ridge National Laboratory has been shut down since 1969 when the fuel salt was drained from the core into two Hastelloy N tanks at the reactor site. In 1995, a multiyear project was launched to remediate the potentially hazardous conditions generated by the movement of fissile material and reactive gases from the storage tanks into the piping system and an auxiliary charcoal bed (ACB). The top 12 in. of the ACB is known by gamma scan and thermal analysis to contain about 2.6 kg U-233. According to the laboratory tests, a few feet of fluorinated charcoal are believed to extend beyond the uranium front. The remainder of the ACB should consist of unreacted charcoal. Fluorinated charcoal, when subjected to rapid heating, can decompose generating gaseous products. Under confined conditions, the sudden exothermic decomposition can produce high temperatures and pressures of near-explosive characteristics. Since it will be necessary to drill and tap the ACB to allow installation of piping and instrumentation for remediation and recovery activities, it is necessary to chemically convert the reactive fluorinated charcoal into a more stable material. Ammonia can be administered to the ACB as a volatile denaturing agent that results in the conversion of the C{sub x}F to carbon and ammonium fluoride, NH{sub 4}F. The charcoal laden with NH{sub 4}F can then be heated without risking any sudden decomposition. The only consequence of heating the treated material will be the volatilization of NH{sub 4}F as a mixture of NH{sub 3} and HF, which would primarily recombine as NH{sub 4}F on surfaces below 200 C. The planned scheme for the ACB denaturing is to flow diluted ammonia gas in steps of increasing NH{sub 3} concentration, 2% to 50%, followed by the injection of pure ammonia. This report summarizes the planned passivation treatment scheme to stabilize the ACB and remove the potential hazards. It also includes basic information

  10. General Corrosion studies of a Titanium and Incoloy based alloys under ammoniacal medium and at 290 deg. C

    International Nuclear Information System (INIS)

    Gokhale, B.K.; Keny, S.J.; Kumbhar, A.G.; Rangarajan, S.; Bera, S.; Nuwad Jitendra; Kumar, Sanjukta A.; Wagh, D. N.; Pradhan, S.

    2012-09-01

    For their use in future PWR applications, the general corrosion behaviors of two modified alloys of titanium and Incoloy were studied at high temperature and high pressure (290 deg. C, 7400 Kpa) under ammoniated atmosphere and compared. Coupons were exposed to solutions of varying ammonia concentrations of (10, 50 and 100 ppm ) at 290 deg. C under non-deaerated conditions in a static autoclave for 20 days. Surface characteristics of exposed coupons were studied using XRF, SEM, EDAX and XPS. The solution in the autoclave was analyzed for its specific conductivity, pH and for the elemental concentrations leached from alloy. The exposed titanium based alloy showed deposition of white crystalline material (300-1000 nm size) on the surface. Depletion of Ti and increase in the oxygen concentration on the exposed surface was observed. This indicated dissolution of Ti in solution from surface at high temperature and pressure and its reaction with oxygen in solution to form oxide and its redeposition on surface. The oxide film compositions were found to change drastically between 10 and 50 ppm ammoniated solution. Ti was found to be enriched in the oxide film when the solution contained 50 ppm of ammonia whereas the opposite effect was observed at 10 ppm of ammonia. The presence Ti 4+ in oxide environment and traces of Cr 3+ were observed but no nitrogen or Zr was detected. Specific conductivity of the exposed solution was found to increase by 30 μS/cm and pH to decrease by 1.5 units. Slight leaching of Ti was observed in solution. No Zr was found in the leached solution. Presence of other elements like Al, Cr, Ni in the exposed solution indicated leaching of autoclave construction material (hastelloy). This alloy showed good resistance for corrosion under the experimental conditions. The exposed surface of Incoloy based alloy showed Ni, Cr, Cu and Mn on the surface with deposition of crystalline particles (200-300 nm size). The exposed surface also showed a decrease in Cr

  11. Characterization of PZT thin films on metal substrates; Charakterisierung von PZT-Duennschichten auf Metallsubstraten

    Energy Technology Data Exchange (ETDEWEB)

    Dutschke, A.

    2008-02-02

    Lead zirconate titanate (PbZr{sub x}Ti{sub 1-x}O{sub 3},PZT) is one of the most applied ceramic materials because of its distinctive piezo- and ferroelectric properties. Prepared as thin films on flexible, metallic substrates it can be used for various applications as strain gauges, key switches, vibration dampers, microactuators and ultrasonic transducers. The aim of this work is to analyze the microstructure and the phase-content of PZT-thin films deposited on temperature- und acid-resistant hastelloy-sheets, to correlate the results with the ferroelectric and dielectric properties. It is demonstrated, that the specific variation of the microstructure can be achieved by different thermal treatments and the selective addition of Neodymium as dopant. Nd-doping leads to a shift of the maximum nucleation rate towards reduced temperatures and a decrease in the rate of growth compared to undoped films. The PZT-films are prepared by a sol-gel-process in fourfold multilayers with a composition near the morphotropic phase boundary, where the tetragonal und rhombohedral perovskite-phases coexist. The crystallisation in Nd-doped and undoped films takes place heterogeneously, preferentially at the interfaces and on the surface of the multilayered films as well as on the inner surface of pores within the films. For the first time, the Zr:Ti fluctuation phenomena emerging in sol-gel derived PZT films is related to the microstructure and the local phase content on a nanometer scale. In this connection it is proved, that long-distance Zr:Ti gradients arise preferentially before and during the crystallisation of the pyrochlore phase. During the following crystallisation of the perovskite phase, the crystallites grow across these gradients without modifying them. It is pointed out that the fluctuation in the Zr:Ti ratio has only minor influence on the amount of the tetragonal or rhombohedral distortion of the crystallites after the transition from the para- to the ferroelectric

  12. 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.

  13. 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.

  14. Characterization of PZT thin films on metal substrates

    International Nuclear Information System (INIS)

    Dutschke, A.

    2008-01-01

    Lead zirconate titanate (PbZr x Ti 1-x O 3 ,PZT) is one of the most applied ceramic materials because of its distinctive piezo- and ferroelectric properties. Prepared as thin films on flexible, metallic substrates it can be used for various applications as strain gauges, key switches, vibration dampers, microactuators and ultrasonic transducers. The aim of this work is to analyze the microstructure and the phase-content of PZT-thin films deposited on temperature- und acid-resistant hastelloy-sheets, to correlate the results with the ferroelectric and dielectric properties. It is demonstrated, that the specific variation of the microstructure can be achieved by different thermal treatments and the selective addition of Neodymium as dopant. Nd-doping leads to a shift of the maximum nucleation rate towards reduced temperatures and a decrease in the rate of growth compared to undoped films. The PZT-films are prepared by a sol-gel-process in fourfold multilayers with a composition near the morphotropic phase boundary, where the tetragonal und rhombohedral perovskite-phases coexist. The crystallisation in Nd-doped and undoped films takes place heterogeneously, preferentially at the interfaces and on the surface of the multilayered films as well as on the inner surface of pores within the films. For the first time, the Zr:Ti fluctuation phenomena emerging in sol-gel derived PZT films is related to the microstructure and the local phase content on a nanometer scale. In this connection it is proved, that long-distance Zr:Ti gradients arise preferentially before and during the crystallisation of the pyrochlore phase. During the following crystallisation of the perovskite phase, the crystallites grow across these gradients without modifying them. It is pointed out that the fluctuation in the Zr:Ti ratio has only minor influence on the amount of the tetragonal or rhombohedral distortion of the crystallites after the transition from the para- to the ferroelectric state due to

  15. Corrosion behaviour of steels and CRA in sour gas environments

    Energy Technology Data Exchange (ETDEWEB)

    Lara, M. Alvarez de; Lancha, A.M.; Hernandez, F.; Gomez-Briceno, D. [CIEMAT, Avenida Complutense 22, 28040 Madrid (Spain); Coca, P. [ELCOGAS, S.A., C.T. GICC Puertollano, Carretera de Calzada de Calatrava a Puertollano, km 27, 13500 Puertollano, Ciudad Real (Spain)

    2004-07-01

    The ELCOGAS power plant in Puertollano (Spain), with 335 MWe (ISO conditions), is an Integrated Gasification Combined Cycle (IGCC) plant built to demonstrate both the technical and economic feasibility of this alternative for clean generation of electricity from coal. IGCC technology is based on a coal gasification process, namely the conversion of coal into combustible gas, which is then subjected to an exhaustive cleaning process. The result is a synthetic gas, virtually free of pollutants that can be burned with a high efficiency in a combined cycle electricity-generating unit. Basically, the ELCOGAS plant consists of three islands jointly designed and integrated into the process: gasification island, air separation island and combined cycle island. In the gasification island, the gas from the gasifier is cleaned (de-dusted and washed) and desulfurized before being sent to the combined cycle island. The washing system consists of a Venturi scrubber with a separator where halogens and alkalis (NH{sub 3}, HCl, HF) are removed from the previously de-dusted gas by means of the wash water. The halogens and alkalis removed are then stripped from the wash water as stripped gas, which is a sour gas. The coal-gas coming from the separator proceeds to sulphur removal in a MDEA system and then, the clean gas (mainly CO, H{sub 2}) is sent to the combined cycle plant. As COS is a significant part of the sulphur containing gases in the coal gas, hydrolysis of the COS to H{sub 2}S takes place before the desulfurization stage, since MDEA is a selective amine for H{sub 2}S. There are many important areas related to materials corrosion within the gas cleaning system. In the ELCOGAS plant carbon steels, austenitic stainless steels and nickel based alloys, such as AISI 316Ti, AISI 904L and Hastelloy C276, are used in the Venturi, the water separator and the strippers. AISI 316Ti is used for the gas piping from the separator to the COS hydrolysis system. Laboratory tests to evaluate

  16. Corrosion resistance of Ni-Cr-Mo alloys. Chemical composition and metallurgical condition's effects

    International Nuclear Information System (INIS)

    Zadorozne, N.S.; Rebak, Raul B.

    2009-01-01

    Ni-Cr-Mo alloys offer an outstanding corrosion resistance in a wide variety of highly-corrosive environments. This versatility is due to the excellent performance of nickel in hot alkaline solutions and the beneficial effect of chromium and molybdenum in oxidizing and reducing conditions, respectively. Alloy C-22 (22 % Cr-13 % Mo-3% W) is a well known versatile member of this family. Due to its excellent corrosion resistance in a wide variety of environments, Alloy C-22 has been selected for the fabrication of the corrosion-resistant outer shell of the high-level nuclear waste container. The increasing demand of the industry for corrosion resistant alloys with particular properties of corrosion and mechanical resistance has led to the development of new alloys. Alloy C-22HS (Ni-21 % Cr-17 % Mo) is a new high-strength corrosion resistant material recently developed and introduced into the market. This alloy provides a corrosion resistance comparable with that of other C-type alloys, and it can also be age hardened to effectively double its yield strength. HASTELLOY HYBRID-BC1 (Ni-22 % Mo-15 % Cr) is a new development intended for filling the gap between Ni-Mo and Ni-Cr-Mo alloys. This novel alloy is able to withstand HCl and H 2 SO 4 , even in the presence of dissolved oxygen and other oxidizing species. Its resistance to chloride-induced pitting corrosion, crevice corrosion and stress corrosion cracking is also remarkable. Thermal aging of Ni-Cr-Mo alloys leads to microstructure changes depending on the temperature range and exposure time at temperature. A Long Range Ordering (LRO) reaction can occur in the range of 350 C degrees to 600 C degrees, producing an ordered Ni 2 (Cr,Mo) phase. This ordering reaction does not seem to affect the corrosion resistance and produces only a slight loss in ductility. LRO transformation is homogeneous and has proven to be useful to fabricate the age-hard enable Alloy C22-HS. Tetrahedral Close Packed (TCP) phases, like μ, σ and

  17. Corrosion and alteration of materials from the nuclear industry; La Corrosion et l'alteration des materiaux du nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Feron, D.; Guerin, Y.; Latge, C.; Limoge, Y.; Madic, C.; Santarini, G.; Seiler, J.M.; Vernaz, E.; Richet, C.

    2010-07-01

    , testing means, experimental techniques, internal corrosion of zircaloy sheath - the iodine effect, stress corrosion of nickel alloys - hydrogen influence, stress corrosion of stainless steels; C - wear corrosion: a coupled phenomenon, research in the framework of service life extension of the French electronuclear park; 3 - Corrosion in future reactors: A - corrosion in gas reactors: corrosion by helium impurities, oxidation resistance of silicon carbide, corrosion of graphite and carbon-carbon composites; B - corrosion in liquid metal reactors: sodium FBRs, lead and lead alloys reactors; C- corrosion in molten salt reactors: corrosion of Hastelloy N-type nickel alloys by molten fluorides, mass transfer in aniso-thermal fluoride systems, tellurium embrittlement, electrochemical study of pure metals corrosion in molten fluorides; 4 - Materials corrosion and alteration in the back-end of the fuel cycle: A - corrosion in concentrated nitric environment: materials behaviour, self-catalytic mechanism of nitric acid reduction; B - corrosion in unsaturated aqueous environment: metallic corrosion in unsaturated environment - application to the storage of waste containers, bitumens alteration, reinforced concrete behaviour and iron framework corrosion, concrete behaviour in severe thermal environment; C - Corrosion in saturated aqueous environment: metals corrosion in clayey environment, long-term behaviour of glasses, ceramics alteration, underwater concrete durability, clays transformation; D - materials biodegradation: microorganisms and nuclear wastes, biodegradation of bitumen, concretes and steels; 5 - Conclusion, glossary

  18. Corrosion and alteration of materials from the nuclear industry

    International Nuclear Information System (INIS)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Feron, D.; Guerin, Y.; Latge, C.; Limoge, Y.; Madic, C.; Santarini, G.; Seiler, J.M.; Vernaz, E.; Richet, C.

    2010-01-01

    , testing means, experimental techniques, internal corrosion of zircaloy sheath - the iodine effect, stress corrosion of nickel alloys - hydrogen influence, stress corrosion of stainless steels; C - wear corrosion: a coupled phenomenon, research in the framework of service life extension of the French electronuclear park; 3 - Corrosion in future reactors: A - corrosion in gas reactors: corrosion by helium impurities, oxidation resistance of silicon carbide, corrosion of graphite and carbon-carbon composites; B - corrosion in liquid metal reactors: sodium FBRs, lead and lead alloys reactors; C- corrosion in molten salt reactors: corrosion of Hastelloy N-type nickel alloys by molten fluorides, mass transfer in aniso-thermal fluoride systems, tellurium embrittlement, electrochemical study of pure metals corrosion in molten fluorides; 4 - Materials corrosion and alteration in the back-end of the fuel cycle: A - corrosion in concentrated nitric environment: materials behaviour, self-catalytic mechanism of nitric acid reduction; B - corrosion in unsaturated aqueous environment: metallic corrosion in unsaturated environment - application to the storage of waste containers, bitumens alteration, reinforced concrete behaviour and iron framework corrosion, concrete behaviour in severe thermal environment; C - Corrosion in saturated aqueous environment: metals corrosion in clayey environment, long-term behaviour of glasses, ceramics alteration, underwater concrete durability, clays transformation; D - materials biodegradation: microorganisms and nuclear wastes, biodegradation of bitumen, concretes and steels; 5 - Conclusion, glossary

  19. Economical and neutronic performance of HYLIFE-II with mixture of 90% flibe + 10% UF4 (or ThF4)

    International Nuclear Information System (INIS)

    Uenalan, Sebahattin

    2004-01-01

    This work investigated the neutronics behavior and the economics of the HYLIFE-II reactor with ThF 4 and UF 4 , which produces an electrical power of 1 GW from the fusion power of 2.857 GW during the operation period of 30 years. The use of ThF 4 and UF 4 is realized by a mixture zone consisted of 90% flibe (Li 2 BeF 4 ) and 10% fuel, instead of 100% flibe coolant. The mixture compositions are selected as 90% flibe + 10% UF 4 , 90% flibe + 10% ThF 4 and 90% flibe + 5% UF 4 + 5% ThF 4 . The capacity factor of the reactor is 0.75. The mixtures, with zone thickness of 65 cm were circulated with periods of 20.22, 19.89 and 20.11 s during the operation period of 30 years, respectively. In addition, for flibe + UF 4 , power stabilization by means of plutonium separation from the mixture was applied. The use of fuel materials in the HYLIFE-II reactor resulted in high energy production, sufficient tritium breeding, significant fissile fuel breeding and low radiation damage in the first wall. The average values of tritium breeding ratio over 30 years are between 1.08 and 1.12, higher than 1.0 indicating sufficient tritium breeding. Generally, the mixtures with ThF 4 show better performance than the mixture with UF 4 in terms of more energy production and significant fissile fuel breeding. The neutronic performance of the reactor increases with the operational period. However, the stabilization process performed after operation for 6 years causes all neutronic values to remain nearly constant during the followed operation time. At the 6th year of operation, the power production, which is ∼1540 MW(electric) at startup, reached the electrical power of 2 GW for flibe + UF 4 . The power production without the separation process reached ∼3500 MW(electric) for the mixtures with ThF 4 and ∼3000 MW(electric) for the mixture with UF 4 . At the end of the operation period, helium production values in the first wall, made of Hastelloy, are calculated as 590 ppm without the

  20. Low Temperature Surface Carburization of Stainless Steels

    Energy Technology Data Exchange (ETDEWEB)

    Collins, Sunniva R; Heuer, Arthur H; Sikka, Vinod K

    2007-12-07

    Al-4V and Hastelloy C22. Cavitation tests showed significant increases in cavitation resistance for treated materials as compared to the non-treated materials. Standard ASTM pin-on-disk sliding friction and reciprocating friction wear tests also indicate significant enhancement in wear properties. Fatigue testing showed an order of magnitude improvement for treated versus non-treated Type 316 at the same maximum stress level (R = -1). The maximum stress at 107 cycles and the endurance stress for infinite life, improved by approximately 50%, from 30 to 45 ksi. The energy savings from this project is estimated at 21.8 trillion Btu/year by 2020. This energy savings will be associated with a CO2 reduction of 1.3 million ton/year. One application of this technology in a sludge pump of a cardboard recycling plant during the course of this project has resulted in an energy savings of 84. 106 Btu and cost savings of $900.

  1. Thin coatings for heavy industry: Advanced coatings for pipes and valves

    Science.gov (United States)

    Vernhes, Luc

    Pipes and valves are pressure vessels that regulate the flow of materials (liquids, gases, and slurries) by controlling the passageways. To optimize processes, reduce costs, and comply with government regulations, original equipment manufacturers (OEMs) must maintain their products in state-of-the-art condition. The first valves were invented over 3,000 years ago to supply water to farms and cities. They were made with bronze alloys, providing good corrosion resistance and acceptable tribological performance. The industrial revolution drove manufacturers to develop new and improved tribological materials. In the 20th century, innovative alloys such as Monel copper-nickel and Stellite cobalt-chrome as well as hard chrome plating were introduced to better control tribological properties and maximize in-service life. Since then, new materials have been regularly introduced to extend the range of applications for valves. For example, Teflon fluoropolymers are used in corrosive chemical and petrochemical processes, the nickel-based superalloys Hastelloy and Inconel for petrochemical applications, and creep-resistant chromium-rich F91 steel for supercritical power plants. Recently, the valve industry has embraced the use of hard thermal sprayed coatings for the most demanding applications, and is investing heavily in research to develop the most suitable coatings for specific uses. There is increasing evidence that the optimal solution to erosive, corrosive, and fretting wear problems lies in the design and manufacture of multi-layer, graded, and/or nanostructured coatings and coating systems that combine controlled hardness with high elastic modulus, high toughness, and good adhesion. The overall objectives of this thesis were 1) to report on advances in the development of structurally controlled hard protective coatings with tailored mechanical, elastoplastic, and thermal properties; and 2) to describe enhanced wear-, erosion-, and corrosion-resistance and other

  2. Hydrofluoric Acid Corrosion Study of High-Alloy Materials

    International Nuclear Information System (INIS)

    Osborne, P.E.

    2002-01-01

    A corrosion study involving high-alloy materials and concentrated hydrofluoric acid (HF) was conducted in support of the Molten Salt Reactor Experiment Conversion Project (CP). The purpose of the test was to obtain a greater understanding of the corrosion rates of materials of construction currently used in the CP vs those of proposed replacement parts. Results of the study will help formulate a change-out schedule for CP parts. The CP will convert slightly less than 40 kg of 233 U from a gas (UF 6 ) sorbed on sodium fluoride pellets to a more stable oxide (U 3 O 8 ). One by-product of the conversion is the formation of concentrated HF. Six moles of highly corrosive HF are produced for each mole of UF 6 converted. This acid is particularly corrosive to most metals, elastomers, and silica-containing materials. A common impurity found in 233 U is 232 U. This impurity isotope has several daughters that make the handling of the 233 U difficult. Traps of 233 U may have radiation fields of up to 400 R at contact, a situation that makes the process of changing valves or working on the CP more challenging. It is also for this reason that a comprehensive part change-out schedule must be established. Laboratory experiments involving the repeated transfer of HF through 1/2-in. metal tubing and valves have proven difficult due to the corrosivity of the HF upon contact with all wetted parts. Each batch of HF is approximately 1.5 L of 33 wt% HF and is transferred most often as a vapor under vacuum and at temperatures of up to 250 C. Materials used in the HF side of the CP include Hastelloy C-276 and Monel 400 tubing, Haynes 230 and alloy C-276 vessels, and alloy 400 valve bodies with Inconel (alloy 600) bellows. The chemical compositions of the metals discussed in this report are displayed in Table 1. Of particular concern are the almost 30 vendor-supplied UG valves that have the potential for exposure to HF. These valves have been proven to have a finite life due to failure

  3. Hydrofluoric Acid Corrosion Study of High-Alloy Materials

    Energy Technology Data Exchange (ETDEWEB)

    Osborne, P.E.

    2002-09-11

    A corrosion study involving high-alloy materials and concentrated hydrofluoric acid (HF) was conducted in support of the Molten Salt Reactor Experiment Conversion Project (CP). The purpose of the test was to obtain a greater understanding of the corrosion rates of materials of construction currently used in the CP vs those of proposed replacement parts. Results of the study will help formulate a change-out schedule for CP parts. The CP will convert slightly less than 40 kg of {sup 233}U from a gas (UF{sub 6}) sorbed on sodium fluoride pellets to a more stable oxide (U{sub 3}O{sub 8}). One by-product of the conversion is the formation of concentrated HF. Six moles of highly corrosive HF are produced for each mole of UF{sub 6} converted. This acid is particularly corrosive to most metals, elastomers, and silica-containing materials. A common impurity found in {sup 233}U is {sup 232}U. This impurity isotope has several daughters that make the handling of the {sup 233}U difficult. Traps of {sup 233}U may have radiation fields of up to 400 R at contact, a situation that makes the process of changing valves or working on the CP more challenging. It is also for this reason that a comprehensive part change-out schedule must be established. Laboratory experiments involving the repeated transfer of HF through 1/2-in. metal tubing and valves have proven difficult due to the corrosivity of the HF upon contact with all wetted parts. Each batch of HF is approximately 1.5 L of 33 wt% HF and is transferred most often as a vapor under vacuum and at temperatures of up to 250 C. Materials used in the HF side of the CP include Hastelloy C-276 and Monel 400 tubing, Haynes 230 and alloy C-276 vessels, and alloy 400 valve bodies with Inconel (alloy 600) bellows. The chemical compositions of the metals discussed in this report are displayed in Table 1. Of particular concern are the almost 30 vendor-supplied UG valves that have the potential for exposure to HF. These valves have been