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Sample records for reactor carbon steel

  1. Fracture assessment of Savannah River Reactor carbon steel piping

    International Nuclear Information System (INIS)

    Mertz, G.E.; Stoner, K.J.; Caskey, G.R.; Begley, J.A.

    1991-01-01

    The Savannah River Site (SRS) production reactors have been in operation since the mid-1950's. One postulated failure mechanism for the reactor piping is brittle fracture of the original A285 and A53 carbon steel piping. Material testing of archival piping determined (1) the static and dynamic tensile properties; (2) Charpy impact toughness; and (3) the static and dynamic compact tension fracture toughness properties. The nil-ductility transition temperature (NDTT), determined by Charpy impact test, is above the minimum operating temperature for some of the piping materials. A fracture assessment was performed to demonstrate that potential flaws are stable under upset loading conditions and minimum operating temperatures. A review of potential degradation mechanisms and plant operating history identified weld defects as the most likely crack initiation site for brittle fracture. Piping weld defects, as characterized by radiographic and metallographic examination, and low fracture toughness material properties were postulated at high stress locations in the piping. Normal operating loads, upset loads, and residual stresses were assumed to act on the postulated flaws. Calculated allowable flaw lengths exceed the size of observed weld defects, indicating adequate margins of safety against brittle fracture. Thus, a detailed fracture assessment was able to demonstrate that the piping systems will not fail by brittle fracture, even though the NDTT for some of the piping is above the minimum system operating temperature

  2. Steel slag carbonation in a flow-through reactor system: the role of fluid-flux.

    Science.gov (United States)

    Berryman, Eleanor J; Williams-Jones, Anthony E; Migdisov, Artashes A

    2015-01-01

    Steel production is currently the largest industrial source of atmospheric CO2. As annual steel production continues to grow, the need for effective methods of reducing its carbon footprint increases correspondingly. The carbonation of the calcium-bearing phases in steel slag generated during basic oxygen furnace (BOF) steel production, in particular its major constituent, larnite {Ca2SiO4}, which is a structural analogue of olivine {(MgFe)2SiO4}, the main mineral subjected to natural carbonation in peridotites, offers the potential to offset some of these emissions. However, the controls on the nature and efficiency of steel slag carbonation are yet to be completely understood. Experiments were conducted exposing steel slag grains to a CO2-H2O mixture in both batch and flow-through reactors to investigate the impact of temperature, fluid flux, and reaction gradient on the dissolution and carbonation of steel slag. The results of these experiments show that dissolution and carbonation of BOF steel slag are more efficient in a flow-through reactor than in the batch reactors used in most previous studies. Moreover, they show that fluid flux needs to be optimized in addition to grain size, pressure, and temperature, in order to maximize the efficiency of carbonation. Based on these results, a two-stage reactor consisting of a high and a low fluid-flux chamber is proposed for CO2 sequestration by steel slag carbonation, allowing dissolution of the slag and precipitation of calcium carbonate to occur within a single flow-through system. Copyright © 2014. Published by Elsevier B.V.

  3. Heavy reflector experiments composed of carbon steel and nickel in the IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    Santos, Adimir dos; Silva, Graciete Simoes de Andrade e; Mura, Luis Felipe; Jerez, Rogerio; Mendonca, Arlindo Gilson; Fuga, Rinaldo

    2013-01-01

    The heavy reflector experiments performed in the IPEN/Mb-01 research reactor facility comprise a set of critical configurations employing the standard 28x26-fuel-rod configuration. The heavy reflector either, carbon steel or nickel plates was placed at one of the faces of the IPEN/MB-01 reactor. Criticality is achieved by inserting the control banks BC1 and BC2 to the critical position. 32 plates around 0.3 mm thick were used in all the experiment. The chosen distance between last fuel rod row and the first laminate for all types of laminates was 5.5 mm. Considering initially the carbon steel case, the experimental data reveal that the reactivity decreases up to the fifth plate and after that it increases, becomes nearly zero (which was equivalent to initial zero excess reactivity with zero plates) for the 28 plates case and reaches a value of 42.73 pcm when the whole set of 32 plates are inserted in the reflector. This is a very striking result because it demonstrates that when all 32 plates are inserted in the reflector there is a net gain of reactivity. The reactivity behavior demonstrates all the physics events already mentioned in this work. When the number of plates are small (around 5), the neutron absorption in the plates is more important than the neutron reflection and the reactivity decreases. This condition holds up to a point where the neutron reflection becomes more important than the neutron absorption in the plates and the reactivity increases. The experimental data for the nickel case shows the main features of the carbon steel case, but for the carbon steel case the reactivity gain is small, thus demonstrating that carbon steel or essentially iron has not the reflector capability as the nickel laminates do. The measured data of nickel plates show a higher reactivity gain, thus demonstrating that nickel is a better reflector than iron. The theoretical analysis employing MCNP5 and ENDF/B-VII.0 show that the calculated results have good results up to

  4. Effect of reactor temperature on direct growth of carbon nanomaterials on stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Edzatty, A. N., E-mail: nuredzatty@gmail.com; Syazwan, S. M., E-mail: mdsyazwan.sanusi@gmail.com; Norzilah, A. H., E-mail: norzilah@unimap.edu.my; Jamaludin, S. B., E-mail: sbaharin@unimap.edu.my [Centre of Excellence for Frontier Materials Research, School of Materials Engineering, University Malaysia Perlis (Malaysia)

    2016-07-19

    Currently, carbon nanomaterials (CNMs) are widely used for various applications due to their extraordinary electrical, thermal and mechanical properties. In this work, CNMs were directly grown on the stainless steel (SS316) via chemical vapor deposition (CVD). Acetone was used as a carbon source and argon was used as carrier gas, to transport the acetone vapor into the reactor when the reaction occurred. Different reactor temperature such as 700, 750, 800, 850 and 900 °C were used to study their effect on CNMs growth. The growth time and argon flow rate were fixed at 30 minutes and 200 ml/min, respectively. Characterization of the morphology of the SS316 surface after CNMs growth using Scanning Electron Microscopy (SEM) showed that the diameter of grown-CNMs increased with the reactor temperature. Energy Dispersive X-ray (EDX) was used to analyze the chemical composition of the SS316 before and after CNMs growth, where the results showed that reduction of catalyst elements such as iron (Fe) and nickel (Ni) at high temperature (700 – 900 °C). Atomic Force Microscopy (AFM) analysis showed that the nano-sized hills were in the range from 21 to 80 nm. The best reactor temperature to produce CNMs was at 800 °C.

  5. Microstructure and grain size effects on irradiation hardening of low carbon steel for reactor tanks

    Energy Technology Data Exchange (ETDEWEB)

    Milasin, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1964-05-15

    Irradiation hardening of steel for reactor pressure vessels has been studied extensively during the past few years. A great number of experimental results concerning the behaviour of these steels in the radiation field and several review papers (1,2) have been published. Most of the papers deal with the effects of specific metallurgical factors or irradiation conditions (temperature, flux) on irradiation hardening and embrittlement. In addition, a number of experiments are performed to give evidence on the mechanism of irradiation hardening of these steels. However, this mechanism is still unknown due to the complexity of steel as a system. Among different methods used in radiation damage studies, the changes of mechanical properties have been mainly investigated. By using Hall-Petch's empirical relation, {sigma}{sub y}={sigma}{sub i}+k{sub y} d{sup -1/2} between lower yield stress, {sigma}{sub y}, and grain size, 2d, the information about the effect of irradiation on the parameters {sigma}{sub i} and k{sub y} is obtained. Taking as a base interpretation of {sigma}{sub i} and k{sub y} given by Petch and his co-workers it has been concluded that radiation does not change the stress to start slip but that it increase the friction that opposes the passage of free dislocations across a slip plane. In attempting to apply Hall-Petch's relation to one unirradiated ferritic steel with a carbon content higher than 0.15% some difficulties were encountered. The results obtained indicate that the influence of grain size can not be isolated from other factors introduced by the treatments used to produce different grain sizes. This paper deals with a similar problem in the case of irradiated steel. The results obtained give the changes of the mechanical properties of steel in neutron irradiation field as a function of microstructure and grain size. In addition, the mechanical properties of irradiated steel are measured after annealing at 150 deg C and 450 deg C. On the basis of

  6. Microstructure and grain size effects on irradiation hardening of low carbon steel for reactor tanks

    International Nuclear Information System (INIS)

    Milasin, N.

    1964-05-01

    Irradiation hardening of steel for reactor pressure vessels has been studied extensively during the past few years. A great number of experimental results concerning the behaviour of these steels in the radiation field and several review papers (1,2) have been published. Most of the papers deal with the effects of specific metallurgical factors or irradiation conditions (temperature, flux) on irradiation hardening and embrittlement. In addition, a number of experiments are performed to give evidence on the mechanism of irradiation hardening of these steels. However, this mechanism is still unknown due to the complexity of steel as a system. Among different methods used in radiation damage studies, the changes of mechanical properties have been mainly investigated. By using Hall-Petch's empirical relation, σ y =σ i +k y d -1/2 between lower yield stress, σ y , and grain size, 2d, the information about the effect of irradiation on the parameters σ i and k y is obtained. Taking as a base interpretation of σ i and k y given by Petch and his co-workers it has been concluded that radiation does not change the stress to start slip but that it increase the friction that opposes the passage of free dislocations across a slip plane. In attempting to apply Hall-Petch's relation to one unirradiated ferritic steel with a carbon content higher than 0.15% some difficulties were encountered. The results obtained indicate that the influence of grain size can not be isolated from other factors introduced by the treatments used to produce different grain sizes. This paper deals with a similar problem in the case of irradiated steel. The results obtained give the changes of the mechanical properties of steel in neutron irradiation field as a function of microstructure and grain size. In addition, the mechanical properties of irradiated steel are measured after annealing at 150 deg C and 450 deg C. On the basis of the experimental results obtained the relative microstructure and

  7. Testing of methods for decontamination of stainless steels and carbon steels conformably to demountable equipment of nuclear power plant with WWR type reactor

    International Nuclear Information System (INIS)

    Dergunova, G.M.; Nazarov, V.K.; Ozolin, A.B.; Smirnov, L.M.; Stel'mashuk, V.P.; Yulikov, E.I.; Vlasov, I.N.

    1978-01-01

    Results are given of experiments on decontamination of stainless steel by the oxidation-reduction method and also results of decontamination of carbon steel by means of solutions based on oxalic acid, citric acid and phosphoric acid. Investigations of efficiency of oxidation-reduction treatment were done on samples of stainless steel cut from the pipeline of the primary coolant circuit of reactor. Comparison is given of efficiency of oxidation-reduction methods of contamination of stainless steel in the case of application of different compositions of decontaminating solutions. Dependences are given for decontamination completeness on duration of operations, on temperature and on ratio of volume of decontaminating solutions to surface are of the sample. For carbon steels parameters are given for decontamination process by means of oxalic, citric and phosphoric acid solutions. (I.T.) [ru

  8. Specification for carbon and low alloy steel containment structures for stationary nuclear power reactors. [Now obsolescent (by Amendment No. 1)

    Energy Technology Data Exchange (ETDEWEB)

    1967-01-01

    This British Standard covers the design, construction, inspection and testing of steel reactor containment structures made of carbon and low alloy steel for temperatures not exceeding 300 deg C. Such structures are not in contact with the reactor coolant during normal operation. Pressure-relieved structures are not excluded, provided they are of a form that contains the fission products or ensures their safe disposal. Attachments such as air-locks or piping that is or may become directly connected between the interior of the containment structure and a closure, and may therefore contain radioactive material released during accidents, is considered part of the containment structure.

  9. The reactor vessel steels

    International Nuclear Information System (INIS)

    Bilous, W.; Hajewska, E.; Szteke, W.; Przyborska, M.; Wasiak, J.; Wieczorkowski, M.

    2005-01-01

    In the paper the fundamental steels using in the construction of pressure vessel water reactor are discussed. The properties of these steels as well as the influence of neutron irradiation on its degradation in the time of exploitation are also done. (authors)

  10. Synthesis of Multi-Walled Carbon Nanotubes from Plastic Waste Using a Stainless-Steel CVD Reactor as Catalyst

    Directory of Open Access Journals (Sweden)

    Pranav K. Tripathi

    2017-09-01

    Full Text Available The disposal of non-biodegradable plastic waste without further upgrading/downgrading is not environmentally acceptable and many methods to overcome the problem have been proposed. Herein we indicate a simple method to make high-value nanomaterials from plastic waste as a partial solution to the environmental problem. Laboratory-based waste centrifuge tubes made of polypropylene were chosen as a carbon source to show the process principle. In the process, multi-walled carbon nanotubes (MWCNTs were synthesized from plastic waste in a two-stage stainless steel 316 (SS 316 metal tube that acted as both reactor vessel and catalyst. The steel reactor contains Fe (and Ni, and various alloys, which act as the catalyst for the carbon conversion process. The reaction and products were studied using electron probe microanalysis, thermogravimetric analysis, Raman spectroscopy and transmission electron microscopy and scanning electron microscopy. Optimization studies to determine the effect of different parameters on the process showed that the highest yield and most graphitized MWCNTs were formed at 900 °C under the reaction conditions used (yield 42%; Raman ID/IG ratio = 0.48. The high quality and high yield of the MWCNTs that were produced in a flow reactor from plastic waste using a two stage SS 316 chemical vapor deposition (CVD furnace did not require the use of an added catalyst.

  11. Synthesis of Multi-Walled Carbon Nanotubes from Plastic Waste Using a Stainless-Steel CVD Reactor as Catalyst.

    Science.gov (United States)

    Tripathi, Pranav K; Durbach, Shane; Coville, Neil J

    2017-09-22

    The disposal of non-biodegradable plastic waste without further upgrading/downgrading is not environmentally acceptable and many methods to overcome the problem have been proposed. Herein we indicate a simple method to make high-value nanomaterials from plastic waste as a partial solution to the environmental problem. Laboratory-based waste centrifuge tubes made of polypropylene were chosen as a carbon source to show the process principle. In the process, multi-walled carbon nanotubes (MWCNTs) were synthesized from plastic waste in a two-stage stainless steel 316 (SS 316) metal tube that acted as both reactor vessel and catalyst. The steel reactor contains Fe (and Ni, and various alloys), which act as the catalyst for the carbon conversion process. The reaction and products were studied using electron probe microanalysis, thermogravimetric analysis, Raman spectroscopy and transmission electron microscopy and scanning electron microscopy. Optimization studies to determine the effect of different parameters on the process showed that the highest yield and most graphitized MWCNTs were formed at 900 °C under the reaction conditions used (yield 42%; Raman I D / I G ratio = 0.48). The high quality and high yield of the MWCNTs that were produced in a flow reactor from plastic waste using a two stage SS 316 chemical vapor deposition (CVD) furnace did not require the use of an added catalyst.

  12. Corrosion Behavior of Carbon Steel Coated with Octadecylamine in the Secondary Circuit of a Pressurized Water Reactor

    Science.gov (United States)

    Jäppinen, Essi; Ikäläinen, Tiina; Järvimäki, Sari; Saario, Timo; Sipilä, Konsta; Bojinov, Martin

    2017-12-01

    Corrosion and particle deposition in the secondary circuits of pressurized water reactors can be mitigated by alternative water chemistries featuring film-forming amines. In the present work, the corrosion of carbon steel in secondary side water with or without octadecylamine (ODA) is studied by in situ electrochemical impedance spectroscopy, combined with weight loss/gain measurements, scanning electron microscopy and glow-discharge optical emission spectroscopy. The impedance spectra are interpreted using the mixed-conduction model to extract kinetic parameters of oxide growth and metal dissolution through it. From the experimental results, it can be concluded that ODA addition reduces the corrosion rate of both fresh and pre-oxidized carbon steel in secondary circuit significantly by slowing down both interfacial reactions and transport through the oxide layer.

  13. Reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Van De Velde, J.; Fabry, A.; Van Walle, E.; Chaouuadi, R.

    1998-01-01

    Research and development activities related to reactor pressure vessel steels during 1997 are reported. The objectives of activities of the Belgian Nuclear Research Centre SCK/CEN in this domain are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate a methodology on a broad database; (3) to achieve regulatory acceptance and industrial use

  14. Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Van de Velde, J.; Fabry, A.; Van Walle, E.; Chaoudi, R

    1998-07-01

    SCK-CEN's R and D programme on Reactor Pressure Vessel (RPV) Steels in performed in support of the RVP integrity assessment. Its main objectives are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate the applied methodology on a broad database; (3) to achieve regulatory acceptance and industrial use. Progress and achievements in 1999 are reported.

  15. Phased Array Ultrasonic Examination of Reactor Coolant System (Carbon Steel-to-CASS) Dissimilar Metal Weld Mockup Specimen

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, S. L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cinson, A. D. [US Nuclear Regulatory Commission (NRC), Washington, DC (United States); Diaz, A. A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Anderson, M. T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-11-23

    In the summer of 2009, Pacific Northwest National Laboratory (PNNL) staff traveled to the Electric Power Research Institute (EPRI) NDE Center in Charlotte, North Carolina, to conduct phased-array ultrasonic testing on a large bore, reactor coolant pump nozzle-to-safe-end mockup. This mockup was fabricated by FlawTech, Inc. and the configuration originated from the Port St. Lucie nuclear power plant. These plants are Combustion Engineering-designed reactors. This mockup consists of a carbon steel elbow with stainless steel cladding joined to a cast austenitic stainless steel (CASS) safe-end with a dissimilar metal weld and is owned by Florida Power & Light. The objective of this study, and the data acquisition exercise held at the EPRI NDE Center, were focused on evaluating the capabilities of advanced, low-frequency phased-array ultrasonic testing (PA-UT) examination techniques for detection and characterization of implanted circumferential flaws and machined reflectors in a thick-section CASS dissimilar metal weld component. This work was limited to PA-UT assessments using 500 kHz and 800 kHz probes on circumferential flaws only, and evaluated detection and characterization of these flaws and machined reflectors from the CASS safe-end side only. All data were obtained using spatially encoded, manual scanning techniques. The effects of such factors as line-scan versus raster-scan examination approaches were evaluated, and PA-UT detection and characterization performance as a function of inspection frequency/wavelength, were also assessed. A comparative assessment of the data is provided, using length-sizing root-mean-square-error and position/localization results (flaw start/stop information) as the key criteria for flaw characterization performance. In addition, flaw signal-to-noise ratio was identified as the key criterion for detection performance.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

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

  17. Preparation Femtosecond Laser Prevention for the Cold-Worked Stress Corrosion Crackings on Reactor Grade Low Carbon Stainless Steel

    CERN Document Server

    John Minehara, Eisuke

    2004-01-01

    We report here that the femtosecond lasers like low average power Ti:Sapphire lasers, the JAERI high average power free-electron laser and others could peel off and remove two stress corrosion cracking (SCC) origins of the cold-worked and the cracking susceptible material, and residual tensile stress in hardened and stretched surface of low-carbon stainless steel cubic samples for nuclear reactor internals as a proof of principle experiment except for the third origin of corrosive environment. Because a 143 °C and 43% MgCl2 hot solution SCC test was performed for the samples to simulate the cold-worked SCC phenomena of the internals to show no crack at the laser-peered off strip on the cold-worked side and ten-thousands of cracks at the non-peeled off on the same side, it has been successfully demonstrated that the femtosecond lasers could clearly remove the two SCC origins and could resultantly prevent the cold-worked SCC.

  18. Oxidation suppressing device for steel materials in carbon dioxide cooled reactors

    International Nuclear Information System (INIS)

    Kawakami, Haruo

    1986-01-01

    Purpose: To effectively reduce impurity hydrogens in carbon dioxide. Constitution: At least three gas chambers are arranged serially each by way of a valve in a gas flow channel branched from a primary carbon dioxide coolant circuits. Then, a polymeric partition membrane having higher permeation rate for hydrogen than for carbon dioxide, e.g., made of polytrifluorochloroethylene is disposed between first and second gas chambers and, further, the first and the third gas chambers are connected each by way of a valve to the primary carbon dioxide coolant circuit to constitute a gas recovery channel. Carbon dioxide is caused to flow through the channel by means of a pump disposed between the second and third gas chambers, hydrogen as impurity passed through the partition walls is concentrated and discharged out of the channel, while the carbon dioxide with reduced hydrogen content is returned from the first and the third gas chambers to the circuit. (Sekiya, K.)

  19. Reactor Structural Materials: Reactor Pressure Vessel Steels

    International Nuclear Information System (INIS)

    Chaouadi, R.

    2000-01-01

    The objectives of SCK-CEN's R and D programme on Rector Pressure Vessel (RPV) Steels are:(1) to complete the fracture toughness data bank of various reactor pressure vessel steels by using precracked Charpy specimens that were tested statically as well as dynamically; (2) to implement the enhanced surveillance approach in a user-friendly software; (3) to improve the existing reconstitution technology by reducing the input energy (short cycle welding) and modifying the stud geometry. Progress and achievements in 1999 are reported

  20. A mechanistic model for predicting flow-assisted and general corrosion of carbon steel in reactor primary coolants

    Energy Technology Data Exchange (ETDEWEB)

    Lister, D. [University of New Brunswick, Fredericton, NB (Canada). Dept. of Chemical Engineering; Lang, L.C. [Atomic Energy of Canada Ltd., Chalk River Lab., ON (Canada)

    2002-07-01

    Flow-assisted corrosion (FAC) of carbon steel in high-temperature lithiated water can be described with a model that invokes dissolution of the protective oxide film and erosion of oxide particles that are loosened as a result. General corrosion under coolant conditions where oxide is not dissolved is described as well. In the model, the electrochemistry of magnetite dissolution and precipitation and the effect of particle size on solubility move the dependence on film thickness of the diffusion processes (and therefore the corrosion rate) away from reciprocal. Particle erosion under dissolving conditions is treated stochastically and depends upon the fluid shear stress at the surface. The corrosion rate dependence on coolant flow under FAC conditions then becomes somewhat less than that arising purely from fluid shear (proportional to the velocity squared). Under non-dissolving conditions, particle erosion occurs infrequently and general corrosion is almost unaffected by flow For application to a CANDU primary circuit and its feeders, the model was bench-marked against the outlet feeder S08 removed from the Point Lepreau reactor, which furnished one value of film thickness and one of corrosion rate for a computed average coolant velocity. Several constants and parameters in the model had to be assumed or were optimised, since values for them were not available. These uncertainties are no doubt responsible for the rather high values of potential that evolved as steps in the computation. The model predicts film thickness development and corrosion rate for the whole range of coolant velocities in outlet feeders very well. In particular, the detailed modelling of FAC in the complex geometry of one outlet feeder (F11) is in good agreement with measurements. When the particle erosion computations are inserted in the balance equations for the circuit, realistic values of crud level are obtained. The model also predicts low corrosion rates and thick oxide films for inlet

  1. A mechanistic model for predicting flow-assisted and general corrosion of carbon steel in reactor primary coolants

    International Nuclear Information System (INIS)

    Lister, D.

    2002-01-01

    Flow-assisted corrosion (FAC) of carbon steel in high-temperature lithiated water can be described with a model that invokes dissolution of the protective oxide film and erosion of oxide particles that are loosened as a result. General corrosion under coolant conditions where oxide is not dissolved is described as well. In the model, the electrochemistry of magnetite dissolution and precipitation and the effect of particle size on solubility move the dependence on film thickness of the diffusion processes (and therefore the corrosion rate) away from reciprocal. Particle erosion under dissolving conditions is treated stochastically and depends upon the fluid shear stress at the surface. The corrosion rate dependence on coolant flow under FAC conditions then becomes somewhat less than that arising purely from fluid shear (proportional to the velocity squared). Under non-dissolving conditions, particle erosion occurs infrequently and general corrosion is almost unaffected by flow For application to a CANDU primary circuit and its feeders, the model was bench-marked against the outlet feeder S08 removed from the Point Lepreau reactor, which furnished one value of film thickness and one of corrosion rate for a computed average coolant velocity. Several constants and parameters in the model had to be assumed or were optimised, since values for them were not available. These uncertainties are no doubt responsible for the rather high values of potential that evolved as steps in the computation. The model predicts film thickness development and corrosion rate for the whole range of coolant velocities in outlet feeders very well. In particular, the detailed modelling of FAC in the complex geometry of one outlet feeder (F11) is in good agreement with measurements. When the particle erosion computations are inserted in the balance equations for the circuit, realistic values of crud level are obtained. The model also predicts low corrosion rates and thick oxide films for inlet

  2. Low temperature radiation embrittlement for reactor vessel steels

    International Nuclear Information System (INIS)

    Ginding, I.A.; Chirkina, L.A.

    1978-01-01

    General conceptions of cold brittleness of bcc metals are in a review. Considered are experimental data and theoretical representations about the effect of irradiation conditions, chemical composition, phase and structural constitutions, grain size, mechanical and thermomechanical treatments on low-temperature irradiation embrittlement of reactor vessel steels. Presented are the methods for increasing radiation stability of metals (carbon and Cr-Mo steels) used in manufacturing reactor vessels

  3. Thermal embrittlement of reactor vessel steels

    International Nuclear Information System (INIS)

    Corwin, W.R.; Nanstad, R.K.; Alexander, D.J.; Stoller, R.E.; Wang, J.A.; Odette, G.R.

    1995-01-01

    As a result of observations of possible thermal embrittlement from recent studies with welds removed from retired steam generators of the Palisades Nuclear Plant (PNP), an assessment was made of thermal aging of reactor pressure vessel (RPV) steels under nominal reactor operating conditions. Discussions are presented on (1) data from the literature regarding relatively low-temperature thermal embrittlement of RPV steels; (2)relevant data from the US power reactor-embrittlement data base (PR-EDB); and (3)potential mechanisms of thermal embrittlement in low-alloy steels

  4. Development of ferritic steels for fusion reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L.; Maziasz, P.J.; Corwin, W.R.

    1988-08-01

    Chromium-molybdenum ferritic (martensitic) steels are leading candidates for the structural components for future fusion reactors. However, irradiation of such steels in a fusion environment will produce long-lived radioactive isotopes that will lead to difficult waste-disposal problems. Such problems could be reduced by replacing the elements in the steels (i.e., Mo, Nb, Ni, N, and Cu) that lead to long-lived radioactive isotopes. We have proposed the development of ferritic steels analogous to conventional Cr-Mo steels, which contain molybdenum and niobium. It is proposed that molybdenum be replaced by tungsten and niobium be replaced by tantalum. Eight experimental steels were produced. Chromium concentrations of 2.25, 5, 9, and 12% were used (all concentrations are in wt %). Steels with these chromium compositions, each containing 2% W and 0.25% V, were produced. To determine the effect of tungsten and vanadium, 2.25 Cr steels were produced with 2% W and no vanadium and with 0.25% V and O and 1% W. A 9Cr steel containing 2% W, 0.25 V, and 0.07% Ta was also studied. For all alloys, carbon was maintained at 0.1%. Tempering studies on the normalized steels indicated that the tempering behavior of the new Cr-W steels was similar to that of the analogous Cr-Mo steels. Microscopy studies indicated that 2% tungsten was required in the 2.25 Cr steels to produce 100% bainite in 15.9-mm-thick plate during normalization. The 5Cr and 9Cr steels were 100% martensite, but the 12 Cr steel contained about 75% martensite with the balance delta-ferrite. 33 refs., 35 figs., 5 tabs.

  5. Development of ferritic steels for fusion reactor applications

    International Nuclear Information System (INIS)

    Klueh, R.L.; Maziasz, P.J.; Corwin, W.R.

    1988-08-01

    Chromium-molybdenum ferritic (martensitic) steels are leading candidates for the structural components for future fusion reactors. However, irradiation of such steels in a fusion environment will produce long-lived radioactive isotopes that will lead to difficult waste-disposal problems. Such problems could be reduced by replacing the elements in the steels (i.e., Mo, Nb, Ni, N, and Cu) that lead to long-lived radioactive isotopes. We have proposed the development of ferritic steels analogous to conventional Cr-Mo steels, which contain molybdenum and niobium. It is proposed that molybdenum be replaced by tungsten and niobium be replaced by tantalum. Eight experimental steels were produced. Chromium concentrations of 2.25, 5, 9, and 12% were used (all concentrations are in wt %). Steels with these chromium compositions, each containing 2% W and 0.25% V, were produced. To determine the effect of tungsten and vanadium, 2.25 Cr steels were produced with 2% W and no vanadium and with 0.25% V and O and 1% W. A 9Cr steel containing 2% W, 0.25 V, and 0.07% Ta was also studied. For all alloys, carbon was maintained at 0.1%. Tempering studies on the normalized steels indicated that the tempering behavior of the new Cr-W steels was similar to that of the analogous Cr-Mo steels. Microscopy studies indicated that 2% tungsten was required in the 2.25 Cr steels to produce 100% bainite in 15.9-mm-thick plate during normalization. The 5Cr and 9Cr steels were 100% martensite, but the 12 Cr steel contained about 75% martensite with the balance delta-ferrite. 33 refs., 35 figs., 5 tabs

  6. Corrosion of carbon steel welds

    International Nuclear Information System (INIS)

    Daniel, B.

    1988-09-01

    This report assesses the factors which cause preferential attack to occur in carbon steel fusion welds. It was concluded that the main factors were: the inclusion content of the weld metal, the potential of the weld metal being less noble than that of the parent, and the presence of low-temperature transformation products in the heat-affected zone of the weld. These factors should be minimized or eliminated as appropriate so that the corrosion allowances determined for carbon steel waste drums is also adequate for the welds. An experimental/theoretical approach is recommended to evaluate the relative corrosion resistance of welds prepared from BS 4360 grade 43A steel to that of the parent material. (author)

  7. Use of ferritic steels in breeder reactors worldwide

    International Nuclear Information System (INIS)

    Patriarca, P.

    1983-01-01

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

  8. Dissimilar welding in nuclear reactors: review of the main aspects related to dissimilar welding of carbon steel and stainless steel with addition of nickel alloys

    International Nuclear Information System (INIS)

    Ribeiro, Vladimir Soler

    2015-01-01

    This work presents a review of the literature about Stress Corrosion Cracking, which is one of the main damage mechanisms that affect PWR and BWR type nuclear reactor. It deals with issues related to one of the sources of the problem, which are the trative residual stress caused by the thermal cycle of welding. It also addressed measurement techniques of residual stresses, with emphasis on technique the center hole drilling. It is shown that, once found the problem, there are means to mitigate the damage, which allow prolonging the life of the component. (author)

  9. Integrity of Magnox reactor steel pressure vessels

    International Nuclear Information System (INIS)

    Flewitt, P.E.J.; Williams, G.H.; Wright, M.B.

    1992-01-01

    The background to the safety assessment of the steel reactor pressure vessels for Magnox power stations is reviewed. The evolved philosophy adopted for the 1991 safety cases prepared for the continued operation of four Magnox power stations operated by Nuclear Electric plc is described, together with different aspects of the multi-legged integrity argument. The main revisions to the materials mechanical property data are addressed together with the assessment methodology adopted and their implications for the overall integrity argument formulated for the continued safe operation of these reactor pressure vessels. (author)

  10. On high temperature strength of carbon steels

    International Nuclear Information System (INIS)

    Ichinose, Hiroyuki; Tamura, Manabu; Kanero, Takahiro; Ihara, Yoshihito

    1977-01-01

    In the steels for high temperature use, the oxidation resistance is regarded as important, but carbon steels show enough oxidation resistance to be used continuously at the temperature up to 500 deg. C if the strength is left out of consideration, and up to 450 deg. C even when the strength is taken into account. Moreover, the production is easy, the workability and weldability are good, and the price is cheap in carbon steels as compared with alloy steels. In the boilers for large thermal power stations, 0.15-0.30% C steels are used for reheater tubes, main feed water tubes, steam headers, wall water tubes, economizer tubes, bypass pipings and others, and they account for 70% of all steel materials used for the boilers of 350 MW class and 30% in 1000 MW class. The JIS standard for the carbon steels for high temperature use and the related standards in foreign countries are shown. The high temperature strength of carbon steels changes according to the trace elements, melting and heat treatment as well as the main compositions of C, Si and Mn. Al and N affect the high temperature strength largely. The characteristics of carbon steels after the heating for hours, the factors controlling the microstructure and high temperature strength, and the measures to improve the high temperature strength of carbon steels are explained. (Kako, I.)

  11. ICP-AES determination of trace elements in carbon steel

    International Nuclear Information System (INIS)

    Sengupta, Arijit; Rajeswari, B.; Kadam, R.M.; Babu, Y.; Godbole, S.V.

    2010-01-01

    Full text: Carbon steel, a combination of the elements iron and carbon, can be classified into four types as mild, medium, high and very high depending on the carbon content which varies from 0.05% to 2.1%. Carbon steel of different types finds application in medical devices, razor blades, cutlery and spring. In the nuclear industry, it is used in feeder pipes in the reactor. A strict quality control measure is required to monitor the trace elements, which have deleterious effects on the mechanical properties of the carbon steel. Thus, it becomes imperative to check the purity of carbon steel as a quality control measure before it is used in feeder pipes in the reactor. Several methods have been reported in literature for trace elemental determination in high purity iron. Some of these include neutron activation analysis, atomic absorption spectrometry and atomic emission spectrometry. Inductively coupled plasma atomic emission spectrometry (ICP-AES) is widely recognized as a sensitive technique for the determination of trace elements in various matrices, its major advantages being good accuracy and precision, high sensitivity, multi-element capability, large linear dynamic range and relative freedom from matrix effects. The present study mainly deals with the direct determination of trace elements in carbon steel using ICP-AES. An axially viewing ICP spectrometer having a polychromator with 35 fixed analytical channels and limited sequential facility to select any analytical line within 2.2 nm of a polychromator line was used in these studies. Iron, which forms one of the main constituents of carbon steel, has a multi electronic configuration with line rich emission spectrum and, therefore, tends to interfere in the determination of trace impurities in carbon steel matrix. Spectral interference in ICP-AES can be seriously detrimental to the accuracy and reliability of trace element determinations, particularly when they are performed in the presence of high

  12. Connections: Superplasticity, Damascus Steels, Laminated Steels, and Carbon Dating

    Science.gov (United States)

    Wadsworth, Jeffrey

    2016-12-01

    In this paper, a description is given of the connections that evolved from the initial development of a family of superplastic plain carbon steels that came to be known as Ultra-High Carbon Steels (UHCS). It was observed that their very high carbon contents were similar, if not identical, to those of Damascus steels. There followed a series of attempts to rediscover how the famous patterns found on Damascus steels blades were formed. At the same time, in order to improve the toughness at room temperature of the newly-developed UHCS, laminated composites were made of alternating layers of UHCS and mild steel (and subsequently other steels and other metals). This led to a study of ancient laminated composites, the motives for their manufacture, and the plausibility of some of the claims relating to the number of layers in the final blades. One apparently ancient laminated composite, recovered in 1837 from the great pyramid of Giza which was constructed in about 2750 B.C., stimulated a carbon dating study of ancient steels. The modern interest in "Bladesmithing" has connections back to many of these ancient weapons.

  13. Ultrahigh Ductility, High-Carbon Martensitic Steel

    Science.gov (United States)

    Qin, Shengwei; Liu, Yu; Hao, Qingguo; Zuo, Xunwei; Rong, Yonghua; Chen, Nailu

    2016-10-01

    Based on the proposed design idea of the anti-transformation-induced plasticity effect, both the additions of the Nb element and pretreatment of the normalization process as a novel quenching-partitioning-tempering (Q-P-T) were designed for Fe-0.63C-1.52Mn-1.49Si-0.62Cr-0.036Nb hot-rolled steel. This high-carbon Q-P-T martensitic steel exhibits a tensile strength of 1890 MPa and elongation of 29 pct accompanied by the excellent product of tensile and elongation of 55 GPa pct. The origin of ultrahigh ductility for high-carbon Q-P-T martensitic steel is revealed from two aspects: one is the softening of martensitic matrix due to both the depletion of carbon in the matensitic matrix during the Q-P-T process by partitioning of carbon from supersaturated martensite to retained austenite and the reduction of the dislocation density in a martensitic matrix by dislocation absorption by retained austenite effect during deformation, which significantly enhances the deformation ability of martensitic matrix; another is the high mechanical stability of considerable carbon-enriched retained austenite, which effectively reduces the formation of brittle twin-type martensite. This work verifies the correctness of the design idea of the anti-TRIP effect and makes the third-generation advanced high-strength steels extend to the field of high-carbon steels from low- and medium-carbon steels.

  14. Cubic martensite in high carbon steel

    Science.gov (United States)

    Chen, Yulin; Xiao, Wenlong; Jiao, Kun; Ping, Dehai; Xu, Huibin; Zhao, Xinqing; Wang, Yunzhi

    2018-05-01

    A distinguished structural characteristic of martensite in Fe-C steels is its tetragonality originating from carbon atoms occupying only one set of the three available octahedral interstitial sites in the body-centered-cubic (bcc) Fe lattice. Such a body-centered-tetragonal (bct) structure is believed to be thermodynamically stable because of elastic interactions between the interstitial carbon atoms. For such phase stability, however, there has been a lack of direct experimental evidence despite extensive studies of phase transformations in steels over one century. In this Rapid Communication, we report that the martensite formed in a high carbon Fe-8Ni-1.26C (wt%) steel at room temperature induced by applied stress/strain has actually a bcc rather than a bct crystal structure. This finding not only challenges the existing theories on the stability of bcc vs bct martensite in high carbon steels, but also provides insights into the mechanism for martensitic transformation in ferrous alloys.

  15. Measurement of carbon activity in sodium and steel and the behaviour of carbon-bearing species

    International Nuclear Information System (INIS)

    Rajendran Pillai, S.; Ranganathan, R.; Mathews, C.K.

    1988-01-01

    Carburization or decarburization of structural materials in a sodium system depends on the local differences in carbon activity. The behaviour of carbon-bearing species in sodium influences its carbon activity. In order to understand the behaviour of carbon in these systems, an electrochemical carbon meter was fabricated in our laboratory. The original version of this meter was capable of operating in the temperature range of 850-980 K. Studies are carried out to extend this lower limit of temperature. Employing the carbon meter, experiments were carried out to understand the behaviour of carbon-bearing species. Gas equilibration experiments were also carried out with the same view. A new method for measuring the carbon activity in steels are described which employs the carbon meter. A review on these investigations and the conclusions reached on the behaviour of carbon in fast reactor loops are described

  16. Long term integrity of reactor pressure vessel and primary containment vessel after the severe accidents in Fukushima Daiichi Nuclear Power Station. Leaching property of spent oxide fuel segment and corrosion property of a carbon steel under artificial seawater immersion

    International Nuclear Information System (INIS)

    2014-06-01

    Primary containment vessel (PCV), reactor pressure vessel and pedestal in Fukushima Daiichi Nuclear power station units 1 through 3 have been exposed to severe thermal, chemical and mechanical conditions due to core meltdown events and seawater injections for emergent core cooling. These components will be immersed in diluted seawater with dissolved fission products under irradiation until the end of debris removal. Fresh water injected into the cores contacts with debris to cool, dissolves or erodes their constituents, mixed with retained water, and becomes 'accumulated water' with radioactive nuclides. We have focused the leaching of fission products into the accumulated water under lower temperature (323 K). FUGEN spent oxide fuel segments were immersed to determine the leaching factor of fission product and actinide elements. Since PCV made from carbon steel is one of the most important boundaries to prevent from fission products release, corrosion behavior has been paid attention to evaluate their integrity. Carbon steel specimens were immersion- and electrochemical-tested in diluted seawater with simulants of the accumulated water at 323 K in order to evaluate the effect of fission products in particular cesium and radiation. (author)

  17. Microbial-Influenced Corrosion of Corten Steel Compared with Carbon Steel and Stainless Steel in Oily Wastewater by Pseudomonas aeruginosa

    Science.gov (United States)

    Mansouri, Hamidreza; Alavi, Seyed Abolhasan; Fotovat, Meysam

    2015-07-01

    The microbial corrosion behavior of three important steels (carbon steel, stainless steel, and Corten steel) was investigated in semi petroleum medium. This work was done in modified nutrient broth (2 g nutrient broth in 1 L oily wastewater) in the presence of Pseudomonas aeruginosa and mixed culture (as a biotic media) and an abiotic medium for 2 weeks. The behavior of corrosion was analyzed by spectrophotometric and electrochemical methods and at the end was confirmed by scanning electron microscopy. The results show that the degree of corrosion of Corten steel in mixed culture, unlike carbon steel and stainless steel, is less than P. aeruginosa inoculated medium because some bacteria affect Corten steel less than other steels. According to the experiments, carbon steel had less resistance than Corten steel and stainless steel. Furthermore, biofilm inhibits separated particles of those steels to spread to the medium; in other words, particles get trapped between biofilm and steel.

  18. Medium temperature carbon dioxide gas turbine reactor

    International Nuclear Information System (INIS)

    Kato, Yasuyoshi; Nitawaki, Takeshi; Muto, Yasushi

    2004-01-01

    A carbon dioxide (CO 2 ) gas turbine reactor with a partial pre-cooling cycle attains comparable cycle efficiencies of 45.8% at medium temperature of 650 deg. C and pressure of 7 MPa with a typical helium (He) gas turbine reactor of GT-MHR (47.7%) at high temperature of 850 deg. C. This higher efficiency is ascribed to: reduced compression work around the critical point of CO 2 ; and consideration of variation in CO 2 specific heat at constant pressure, C p , with pressure and temperature into cycle configuration. Lowering temperature to 650 deg. C provides flexibility in choosing materials and eases maintenance through the lower diffusion leak rate of fission products from coated particle fuel by about two orders of magnitude. At medium temperature of 650 deg. C, less expensive corrosion resistant materials such as type 316 stainless steel are applicable and their performance in CO 2 have been proven during extensive operation in AGRs. In the previous study, the CO 2 cycle gas turbomachinery weight was estimated to be about one-fifth compared with He cycles. The proposed medium temperature CO 2 gas turbine reactor is expected to be an alternative solution to current high-temperature He gas turbine reactors

  19. EIS Response of MIC on Carbon Steel

    DEFF Research Database (Denmark)

    Hilbert, Lisbeth Rischel; Maahn, Ernst

    1998-01-01

    Abstract Microbially influenced corrosion of carbon steel under sulphate reducing (sulphide-producing) bacterial activity (SRB) results in the formation of both ferrous sulphides as well as biofilm on the metal surface. The electrochemical characteristics of the ferrous sulphide/steel interface...... as compared to the biofilm/ferrous sulphide/steel interface has been studied with EIS, DC polarisations (Tafel, LPR) and a potentiostatic step technique. The electrochemical response is related to a threshold sulphide concentration above which very characteristic changes such as indications of finite...

  20. Corrosion of carbon steel in neutral water

    International Nuclear Information System (INIS)

    Kawai, Noboru; Iwahori, Toru; Kurosawa, Tatsuo

    1983-01-01

    The initial corrosion behavior of materials used in the construction of heat exchanger and piping system of BWR nuclear power plants and thermal power plants have been examined in neutral water at 30, 50, 100, 160, 200, and 285 deg C with two concentrations of dissolved oxygen in the water. In air-saturated water, the corrosion rate of carbon steel was so higher than those in deaerated conditions and the maximum corrosion rate was observed at 200 deg C. The corrosion rate in deaerated water gradually increased with increasing the water temperature. Low alloy steel (2.25 Cr, 1Mo) exhibited good corrosion resistance compared with the corrosion of carbon steel under similar testing conditions. Oxide films grown on carbon steel in deaerated water at 50, 100, 160, 200, and 285 deg C for 48 and 240 hrs were attacked by dissolved oxygen in room temperature water respectively. However the oxide films formed higher than about 160 deg C showed more protective. The electrochemical behavior of carbon steel with oxide films was also similar to the effect of temperature on the stability of oxide films. (author)

  1. Use of stainless steel as structural materials in reactor cores

    International Nuclear Information System (INIS)

    Teodoro, C.A.

    1990-01-01

    Austenitic stainless steels are used as structural materials in reactor cores, due to their good mechanical properties at working temperatures and high generalized corrosion resistance in aqueous medium. The objective of this paper is to compare several 300 series austenitic stainless steels related to mechanical properties, localized corrosion resistance (SCC and intergranular) and content of delta ferrite. (author)

  2. Corrosion of carbon steel and low-alloy steel in diluted seawater containing hydrazine under gamma-rays irradiation

    International Nuclear Information System (INIS)

    Nakano, Junichi; Yamamoto, Masahiro; Tsukada, Takashi

    2014-01-01

    Seawater was injected into reactor cores of Units 1, 2, and 3 in the Fukushima Daiichi nuclear power station as an urgent coolant. It is considered that the injected seawater causes corrosion of steels of the reactor pressure vessel and primary containment vessel. To investigate the effects of gamma-rays irradiation on weight loss in carbon steel and low-alloy steel, corrosion tests were performed in diluted seawater at 50°C under gamma-rays irradiation. Specimens were irradiated with dose rates of 4.4 kGy/h and 0.2 kGy/h. To evaluate the effects of hydrazine (N 2 H 4 ) on the reduction of oxygen and hydrogen peroxide, N 2 H 4 was added to the diluted seawater. In the diluted seawater without N 2 H 4 , weight loss in the steels irradiated with 0.2 kGy/h was similar to that in the unirradiated steels, and weight loss in the steels irradiated with 4.4 kGy/h increased to approximate 1.7 times of those in the unirradiated steels. Weight loss in the steels irradiated in the diluted seawater containing N 2 H 4 was similar to that in the diluted seawater without N 2 H 4 . When N 2 was introduced into the gas phase in the flasks during gamma-rays irradiation, weight loss in the steels decreased. (author)

  3. Modeling irradiation embrittlement in reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Odette, G.R.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 10, numerical modeling of irradiation embrittlement in reactor vessel steels are introduced. Physically-based models are developed and their role in advancing the state-of-the-art of predicting irradiation embrittlement of RPV steels is stressed

  4. Microstructural evolution in neutron irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

    English, C.A.; Phythian, W.J.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. The microstructural evolution in neutron irradiated reactor pressure vessel steels is described. The damage mechanisms are elaborated and techniques for examining the microstructure are suggested. The importance of the initial damage event is analysed, and the microstructural evolution in RPV steels is examined

  5. The effect of variations in carbon activity on the carburization of austenitic steels in sodium

    International Nuclear Information System (INIS)

    Gwyther, J.R.; Hobdell, M.R.; Hooper, A.J.

    1978-07-01

    Experience has shown that the liquid sodium coolant of fast breeder reactors is an effective carbon-transport medium; the resulting carburization of thin austenitic stainless steel components (eg IHX and fuel cladding) could adversely affect their mechanical integrity. The degree and nature of steel carburization depend, inter alia, on the carbon activity of the sodium environment. Exploratory tests are described in which specimens of austenitic stainless steel were carburized in sodium, the carbon activity of which was continuously monitored by a BNL electrochemical carbon meter. The sodium carbon activity was initially high, but decreased with time, simulating conditions equivalent to plant start-up or coolant clean-up following accidental oil ingress. The extent and nature of steel carburization was identified by metallography, electron microscopy, X-ray crystallography and chemical analysis. (author)

  6. Effect of heat treatment on carbon steel pipe welds

    International Nuclear Information System (INIS)

    Mohamad Harun.

    1987-01-01

    The heat treatment to improve the altered properties of carbon steel pipe welds is described. Pipe critical components in oil, gasification and nuclear reactor plants require adequate room temperature toughness and high strength at both room and moderately elevated temperatures. Microstructure and microhardness across the welds were changed markedly by the welding process and heat treatment. The presentation of hardness fluctuation in the welds can produce premature failure. A number of heat treatments are suggested to improve the properties of the welds. (author) 8 figs., 5 refs

  7. Dynamical analysis on carbon transfer in liquid metal cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Kataoka, Tadayuki; Matsumoto, Keishi

    1979-01-01

    The dynamical analysis was undertaken on the exchange of carbon taking place between the structural steels and sodium for the case of a bi-metallic secondary system constituted of type 304 stainless and 2 1/4Cr-1Mo steels, representing the secondary system of a liquid sodium cooled fast breeder reactor. The analysis brought to light the effects to be expected on the long terms carbon transfer behavior of: (a) the surface areas of structural steels in contact with flowing sodium, (b) the thickness of the sodium-boundary layer, (c) the initial carbon concentration in the sodium, and (d) the rate of carbon contamination of the sodium. (author)

  8. Adaptation of fuel code for light water reactor with austenitic steel rod cladding

    International Nuclear Information System (INIS)

    Gomes, Daniel de Souza; Silva, Antonio Teixeira; Giovedi, Claudia

    2015-01-01

    Light water reactors were used with steel as nuclear fuel cladding from 1960 to 1980. The high performance proved that the use of low-carbon alloys could substitute the current zirconium alloys. Stainless steel is an alternative that can be used as cladding. The zirconium alloys replaced the steel. However, significant experiences in-pile occurred, in commercial units such as Haddam Neck, Indian Point, and Yankee experiences. Stainless Steel Types 347 and 348 can be used as cladding. An advantage of using Stainless Steel was evident in Fukushima when a large number of hydrogens was produced at high temperatures. The steel cladding does not eliminate the problem of accumulating free hydrogen, which can lead to a risk of explosion. In a boiling water reactor, environments easily exist for the attack of intergranular corrosion. The Stainless Steel alloys, Types 321, 347, and 348, are stabilized against attack by the addition of titanium, niobium, or tantalum. The steel Type 348 is composed of niobium, tantalum, and cobalt. Titanium preserves type 321, and niobium additions stabilize type 347. In recent years, research has increased on studying the effects of irradiation by fast neutrons. The impact of radiation includes changes in flow rate limits, deformation, and ductility. The irradiation can convert crystalline lattices into an amorphous structure. New proposals are emerging that suggest using a silicon carbide-based fuel rod cladding or iron-chromium-aluminum alloys. These materials can substitute the classic zirconium alloys. Once the steel Type 348 was chosen, the thermal and mechanical properties were coded in a library of functions. The fuel performance codes contain all features. A comparative analysis of the steel and zirconium alloys was made. The results demonstrate that the austenitic steel alloys are the viable candidates for substituting the zirconium alloys. (author)

  9. Adaptation of fuel code for light water reactor with austenitic steel rod cladding

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel de Souza; Silva, Antonio Teixeira, E-mail: dsgomes@ipen.br, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (POLI/USP), Sao Paulo, SP (Brazil). Lab. de Analise, Avaliacao e Gerenciamento de Risco

    2015-07-01

    Light water reactors were used with steel as nuclear fuel cladding from 1960 to 1980. The high performance proved that the use of low-carbon alloys could substitute the current zirconium alloys. Stainless steel is an alternative that can be used as cladding. The zirconium alloys replaced the steel. However, significant experiences in-pile occurred, in commercial units such as Haddam Neck, Indian Point, and Yankee experiences. Stainless Steel Types 347 and 348 can be used as cladding. An advantage of using Stainless Steel was evident in Fukushima when a large number of hydrogens was produced at high temperatures. The steel cladding does not eliminate the problem of accumulating free hydrogen, which can lead to a risk of explosion. In a boiling water reactor, environments easily exist for the attack of intergranular corrosion. The Stainless Steel alloys, Types 321, 347, and 348, are stabilized against attack by the addition of titanium, niobium, or tantalum. The steel Type 348 is composed of niobium, tantalum, and cobalt. Titanium preserves type 321, and niobium additions stabilize type 347. In recent years, research has increased on studying the effects of irradiation by fast neutrons. The impact of radiation includes changes in flow rate limits, deformation, and ductility. The irradiation can convert crystalline lattices into an amorphous structure. New proposals are emerging that suggest using a silicon carbide-based fuel rod cladding or iron-chromium-aluminum alloys. These materials can substitute the classic zirconium alloys. Once the steel Type 348 was chosen, the thermal and mechanical properties were coded in a library of functions. The fuel performance codes contain all features. A comparative analysis of the steel and zirconium alloys was made. The results demonstrate that the austenitic steel alloys are the viable candidates for substituting the zirconium alloys. (author)

  10. Interphase and intergranular stress generation in carbon steels

    International Nuclear Information System (INIS)

    Oliver, E.C.; Daymond, M.R.; Withers, P.J.

    2004-01-01

    Neutron diffraction spectra have been acquired during tensile straining of high and low carbon steels, in order to compare the evolution of internal stress in ferritic steel with and without a reinforcing phase. In low carbon steel, the generation of intergranular stresses predominates, while in high carbon steel similar intergranular stresses among ferrite grain families are superposed upon a large redistribution of stress between phases. Comparison is made to calculations using elastoplastic self-consistent and finite element methods

  11. Assessment of martensitic steels for advanced fusion reactors

    International Nuclear Information System (INIS)

    Wareing, J.; Tavassoli, A.A.

    1995-01-01

    Martensitic steels are currently considered in Europe to be prime structural candidate materials for the first wall and breeding blanket of the DEMO fusion reactor. In this design, reactor power and wall loading will be significantly higher than those of an experimental reactor. ITER and will give rise to component operating temperatures in the range 250 to 550 0 C with neutron doses higher than 70 dpa. These conditions render austenitic stainless steel, which will be used in ITER, less favourable. Factors contributing to the promotion of martensitic steels are their excellent resistance to irradiation induced swelling, low thermal expansion and high thermal conductivity allied to advanced industrial maturity, compared to other candidate materials vanadium alloys. This paper described the development and optimisation of the steel and weld metal. Using data design rules generated on modified 9 Cr 1 Mo steel during its qualification as a steam generator material for the European Fast Reactor (EFR), interim design guidelines are formulated. Whilst the merits of the steel are validated, it is shown that irradiation embrittlement at low temperature, allied to the need for prolonged post-weld hat treatment and the long term creep response of welds remain areas of some concern. (author). 18 refs., 6 figs., 2 tabs

  12. Application of high strength steel to nuclear reactor containment vessel

    International Nuclear Information System (INIS)

    Susukida, H.; Sato, M.; Takano, G.; Uebayashi, T.; Yoshida, K.

    1976-01-01

    Nuclear reactor containment vessels are becoming larger in size with the increase in the power generating capacity of nuclear power plants. For example, a containment vessel for a PWR power plant with an output of 1,000 MWe becomes an extremely large one if it is made of the conventional JIS SGV 49 (ASTM A 516 Gr. 70) steel plates less than 38 mm in thickness. In order to design the steel containment vessel within the conventional dimensional range, therefore, it is necessary to use a high strength steel having a higher tensile strength than SGV 49 steel, good weldability and a higher fracture toughness and moreover, possessing satisfactory properties without undergoing post-weld heat treatment. The authors conducted a series of verification tests on high strength steel developed by modifying the ASTM A 543 Grade B Class 1 steel with a view to adopting it as a material for the nuclear reactor containment vessels. As the result of evaluation of the test results from various angles, we confirmed that the high strength steel is quite suitable for the manufacture of nuclear reactor containment vessels. (auth.)

  13. Characteristics of Modified 9Cr-1Mo Steel for Reactor Pressure Vessel of Very High Temperature Gas Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Ho; Ryu, W. S.; Han, Chang Hee; Yoon, J. H.; Chang, Jong Hwa

    2004-11-15

    Many researches and developments have been progressed for the construction of VHTR by 2020 in Korea. Modified 9Cr-1Mo steel has been receiving attention for the application to the reactor pressure vessel material of VHTR. We collected and analyzed the research data for modified 9Cr-1Mo steel in order to understand the characteristics of modified 9Cr-1Mo steel. The modified 9Cr-1Mo steel is a modified alloy system similar to conventional 9Cr-1Mo grade ferritic steel. Modifications include additions of vanadium, niobium, and nitrogen, as well as lower carbon content. In this report, we summarized the change of microstructure and mechanical properties after tempering, thermal aging, and irradiation. Modified 9Cr-1Mo steel has high strength and thermal conductivity, low thermal expansion, and good resistance to corrosion. But the irradiation embrittlement behavior of modified 9Cr-1Mo steel should be evaluated and the evaluation methodology also should be developed. At the same time, the characteristics of weldment which is the weak part in pressure vessel should be evaluated.

  14. Integrating Steel Production with Mineral Carbon Sequestration

    Energy Technology Data Exchange (ETDEWEB)

    Klaus Lackner; Paul Doby; Tuncel Yegulalp; Samuel Krevor; Christopher Graves

    2008-05-01

    The objectives of the project were (i) to develop a combination iron oxide production and carbon sequestration plant that will use serpentine ores as the source of iron and the extraction tailings as the storage element for CO2 disposal, (ii) the identification of locations within the US where this process may be implemented and (iii) to create a standardized process to characterize the serpentine deposits in terms of carbon disposal capacity and iron and steel production capacity. The first objective was not accomplished. The research failed to identify a technique to accelerate direct aqueous mineral carbonation, the limiting step in the integration of steel production and carbon sequestration. Objective (ii) was accomplished. It was found that the sequestration potential of the ultramafic resource surfaces in the US and Puerto Rico is approximately 4,647 Gt of CO2 or over 500 years of current US production of CO2. Lastly, a computer model was developed to investigate the impact of various system parameters (recoveries and efficiencies and capacities of different system components) and serpentinite quality as well as incorporation of CO2 from sources outside the steel industry.

  15. East/west steels for reactor pressure vessels

    International Nuclear Information System (INIS)

    Davies, M.; Kryukov, A.; Nikolaev, Y.; English, C.

    1997-01-01

    The report consist of three parts dealing with comparison of the irradiation behaviour of 'Eastern' and 'Western' steels, mechanisms of irradiation embrittlement and the role of compositional variations on the irradiation sensitivity of pressure vessels. Nickel, copper and phosphorus are the elements rendering the most essential influence on behaviour of pressure vessel steels under irradiation and subsequent thermal annealing. For WWER-440 reactor pressure vessel (RPV) steels in which nickel content does nor exceed 0.3% the main affecting factors are phosphorous and copper. For WWER-1000 RPV welds in which nickel content generally exceed 1.5% the role of nickel in radiation embrittlement is decisive. In 'Western' type steels main influencing elements are nickel and copper. The secondary role of phosphorus in radiation embrittlement of 'Western' steels is caused by lower relative content compared to 'Eastern' steels. The process of how copper, phosphorus and nickel contents affect the irradiation sensitivity of both types of steel seem to be similar. Some distinctions between the observed radiation effects is apparently caused by differences in the irradiation conditions and ratios of the contents of above mentioned elements in both types of steel. For 'Eastern' RPV steels the dependence of the recovery degree of irradiated steels due to postirradiation thermal annealing id obviously dependent on phosphorus contents and the influence of nickel contents on this process is detectable

  16. Microbially induced corrosion of carbon steel in deep groundwater environment

    Directory of Open Access Journals (Sweden)

    Pauliina eRajala

    2015-07-01

    Full Text Available The metallic low and intermediate level radioactive waste generally consists of carbon steel and stainless steels. The corrosion rate of carbon steel in deep groundwater is typically low, unless the water is very acidic or microbial activity in the environment is high. Therefore, the assessment of microbially induced corrosion of carbon steel in deep bedrock environment has become important for evaluating the safety of disposal of radioactive waste. Here we studied the corrosion inducing ability of indigenous microbial community from a deep bedrock aquifer. Carbon steel coupons were exposed to anoxic groundwater from repository site 100 m depth (Olkiluoto, Finland for periods of three and eight months. The experiments were conducted at both in situ temperature and room temperature to investigate the response of microbial population to elevated temperature. Our results demonstrate that microorganisms from the deep bedrock aquifer benefit from carbon steel introduced to the nutrient poor anoxic deep groundwater environment. In the groundwater incubated with carbon steel the planktonic microbial community was more diverse and 100-fold more abundant compared to the environment without carbon steel. The betaproteobacteria were the most dominant bacterial class in all samples where carbon steel was present, whereas in groundwater incubated without carbon steel the microbial community had clearly less diversity. Microorganisms induced pitting corrosion and were found to cluster inside the corrosion pits. Temperature had an effect on the species composition of microbial community and also affected the corrosion deposits layer formed on the surface of carbon steel.

  17. Experimental and numerical simulation of carbon manganese steel ...

    African Journals Online (AJOL)

    Experimental and numerical simulation of carbon manganese steel for cyclic plastic behaviour. J Shit, S Dhar, S Acharyya. Abstract. The paper deals with finite element modeling of saturated low cycle fatigue and the cyclic hardening phenomena of the materials Sa333 grade 6 carbon steel and SS316 stainless steel.

  18. Marine atmospheric corrosion of carbon steels

    Energy Technology Data Exchange (ETDEWEB)

    Morcillo, M.; Alcantara, J.; Diaz, I.; Chico, B.; Simancas, J.; Fuente, D. de la

    2015-07-01

    Basic research on marine atmospheric corrosion of carbon steels is a relatively young scientific field and there continue to be great gaps in this area of knowledge. The presence of akaganeite in the corrosion products that form on steel when it is exposed to marine atmospheres leads to a notable increase in the corrosion rate. This work addresses the following issues: (a) environmental conditions necessary for akaganeite formation; (b) characterisation of akaganeite in the corrosion products formed; (c) corrosion mechanisms of carbon steel in marine atmospheres; (d) exfoliation of rust layers formed in highly aggressive marine atmospheres; (e) long-term corrosion rate prediction; and (f) behaviour of weathering steels. Field research has been carried out at Cabo Vilano wind farm (Camarinas, Galicia) in a wide range of atmospheric salinities and laboratory work involving the use of conventional atmospheric corrosion techniques and near-surface and bulk sensitive analytical techniques: scanning electron microscopy (SEM)/energy dispersive spectrometry (EDS), X-ray diffraction (XRD), Mossbauer spectroscopy and SEM/μRaman spectroscopy. (Author)

  19. Plasticity of low carbon stainless steels

    International Nuclear Information System (INIS)

    Bulat, S.I.; Fel'dgandler, Eh.G.; Kareva, E.N.

    1975-01-01

    In the temperature range 800-1200 0 C and with strain rates of from 10 -3 to 3 s -1 , austenitic (000Kh18N12) and austenitic-ferrite (000Kh26N6) very low carbon stainless steels containing 0.02-0.03% C exhibit no higher resilience than corresponding ordinary steels containing 0.10-0.12% C. However, the plasticity of such steels (particularly two-phase steels) at 900-1100 0 C is appreciably inferior owing to the development of intergranular brittle fracture. Pressure treatment preceded by partial cooling of the surface to 850 0 C yields rolled and forged products with acceptable indices but is inconvenient technically. At the Zlatoustovsk and Ashin metallurgical plants successful tests have been performed involving the forging and rolling of such steels heated to 1280-1300 0 C without partial cooling; it was necessary to improve the killing conditions, correct the chemical composition (increasing the proportion of ferrite) and take measures against heat loss. (author)

  20. Internal friction in martensitic carbon steels

    International Nuclear Information System (INIS)

    Hoyos, J.J.; Ghilarducci, A.A.; Salva, H.R.; Chaves, C.A.; Velez, J.M.

    2009-01-01

    This paper proposes relationships between the internal friction and the microstructure of two steels containing 0.626 and 0.71 wt.% carbon. The steels were annealed at 1093 K for 5 min, quenched into water and tempered for 10 min at 423, 573 and 723 K. Internal friction was measured by using a forced vibration pendulum, in a temperature range from 100 to 450 K. The internal friction spectrum is decomposed into four peaks: P1 at 215 K, P2 at 235 K, P3 at 260 K and P4 at 380 K for 3 Hz. Peak P1 is attributed to the interactions between dislocations and carbon atoms. Peak P2 is related to the interaction between dislocations and carbide. Peak P3 is related to the generations of kink - pairs along edge dislocations. Peak P4 is attributed to epsilon carbide precipitation.

  1. Stainless steels in boiling water reactors. Corrosion problems and possible solutions

    International Nuclear Information System (INIS)

    Combrade, P.; Desestret, A.; Leroy, F.; Coriou, H.

    1977-01-01

    In boiling water reactors, the heat-carrying water may have an up to 0.1 or even 0.2 ppm oxygen content, which can make it highly agressive at operating temperature for stainless steels subject to high physical stresses. Several metallurgical solutions can be considered, and in particular the use of stainless steels having a mixed austenitic-ferritic structure or of standard austenitic steels (18.10 or 18.10 Mo, such as AISI 304 and 316) with carefully controlled carbon and alloy element contents. The behavior of these steels during prolonged tests in water at 288 0 C with a 30 and even 100 ppm oxygen content turned out to be quite satisfactory [fr

  2. Defects investigation in neutron irradiated reactor steels by positron annihilation

    International Nuclear Information System (INIS)

    Slugen, V.

    2003-01-01

    Positron annihilation spectroscopy (PAS) based on positron lifetime measurements using the Pulsed Low Energy Positron System (PLEPS) was applied to the investigation of defects of irradiated and thermally treated reactor pressure vessel (RPV) steels. PLEPS results showed that the changes in microstructure of the RPV-steel properties caused by neutron irradiation and post-irradiation heat treatment can be well detected. From the lifetime measurements in the near-surface region (20-550 nm) the defect density in Russian types of RPV-steels was calculated using the diffusion trapping model. The post-irradiation heat treatment studies performed on non-irradiated specimens are also presented. (author)

  3. Structure of steel reactor building and construction method therefor

    International Nuclear Information System (INIS)

    Yamakawa, Toshikimi.

    1997-01-01

    The building of the present invention contains a reactor pressure vessel, and has double steel plate walls endurable to elevation of inner pressure and keeping airtightness, and shielding concretes are filled between the double steel plate walls. It also has empty double steel plate walls not filled with concretes and has pipelines, vent ducts, wirings and a support structures for attaching them between the double steel plate walls. It is endurable to a great inner pressure satisfactory and keeps airtightness by the two spaced steel plates. It can be greatly reduced in the weight, and can be manufactured efficiently with high quality in a plant by so called module construction, and the dimension of the entire of the reactor building can be reduced. It is constructed in a dock, transported on the sea while having the space between the two steel plate walls as a ballast tanks, placed in the site, and shielding concretes are filled between the double steel plate walls. The term for the construction can be reduced, and the cost for the construction can be saved. (N.H.)

  4. Reactor pressure vessel steels ASTM A533B and A508 Cl.2

    International Nuclear Information System (INIS)

    Pelli, R.; Kemppainen, M.; Toerroenen, K.

    1979-11-01

    This report presents the tensile test results of steels ASTM A533B and A508 Cl.2 obtained in connection with a programme initiated to gather and create information needed for the assessment of the structural integrity of the reactor pressure vessels. The tensile properties were studied between -196 and 300 degC varying austenitizing and tempering temperatures and having two different carbon contents for the heats of A533B. (author)

  5. Radiation embrittlement of Spanish nuclear reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Bros, J.; Ballesteros, A.; Lopez, A.

    1993-01-01

    Commercial pressurized water reactor (PWR) and boiling water reactor (BWR) nuclear power plants contain a series of pressure vessel steel surveillance capsules as the principal means of monitoring radiation effects on the pressure vessel. Changes in fracture toughness are more severe in surveillance capsules than in reactor vessel materials because of their proximity of the reactor core. Therefore, it is possible to predict changes in fracture toughness of the reactor vessel materials. This paper describes the status of the reactor vessel surveillance program relating to Spanish nuclear power plants. To date, twelve capsules have been removed and analyzed from seven of the nine Spanish reactors in operation. The results obtained from the analysis of these capsules are compared with the predictions of the Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Rev. 2, by means of measured and expected increase of the nil-ductility transition reference temperature (RT NDT ). The comparison is made considering the different variables normally included in the studies of radiation response of reactor pressure vessel materials, such as copper content of steel, level of neutron fluence above 1 MeV, base metal or weld metal, and so forth. The surveillance data have been used for determining the adjusted reference temperatures and upper shelf energies at any time. The results have shown that the seven pressure vessels are in excellent condition to continue operating with safety against brittle fracture beyond the design life, without the need to recuperate the degraded properties of the materials by annealing of the vessel

  6. Irradiation embrittlement of reactor vessel steels

    International Nuclear Information System (INIS)

    Bros, J.

    2000-01-01

    From the historical decision of closing the Yankee Rowe NPP because of the uncertainties on the level of reactor pressure vessel neutron embrittlement, this paper reviews the technical-scientist bases of the degradation phenomena, and refers to the evolution of reactor pressure vessel radiation surveillance programs. (Author)

  7. Austenitic stainless steel bulk property considerations for fusion reactors

    International Nuclear Information System (INIS)

    Mattas, R.F.

    1979-04-01

    The bulk properties of annealed 304, 316, and 20% cold-worked 316 stainless steels are evaluated for the temperature and radiation conditions expected in a near-term fusion reactor. Of interest are the thermophysical properties, void swelling produced by neutron radiaion, and the tensile, creep, and fatigue properties before and after irradiation

  8. Irradiation proposition of ferritic steels in a russian reactor

    International Nuclear Information System (INIS)

    Seran, J.L.; Decours, J.; Levy, L.

    1987-04-01

    Using the low temperatures of russian reactors, a sample irradiation is proposed to study mechanical properties and swelling of martensitic steels (EM10, T91, 1.4914, HT9), ferrito-martensitic (EM12) and ferritic (F17), at temperatures lower than 400 0 C [fr

  9. Study on corrosion of carbon steel in DEA aqueous solutions

    Science.gov (United States)

    Yang, Jun Han; Xie, Jia Lin; Zhang, Li

    2018-02-01

    Corrosion of carbon steel in the CO2 capture process using diethanolamine (DEA) aqueous solutions was investigated. The effects of the mass concentrations of DEA, solution temperature and CO2 loading on the corrosion rate of carbon steel were demonstrated. The experimental results provided comprehensive information on the appropriate concentration range of DEA aqueous solutions under which low corrosion of carbon steel can be achieved.

  10. Microstructure and embrittlement of VVER 440 reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Hennion, A.

    1999-03-01

    27 VVER 440 pressurised water reactors operate in former Soviet Union and in Eastern Europe. The pressure vessel, is made of Cr-Mo-V steel. It contains a circumferential arc weld in front of the nuclear core. This weld undergoes a high neutron flux and contains large amounts of copper and phosphorus, elements well known for their embrittlement potency under irradiation. The embrittlement kinetic of the steel is accelerated, reducing the lifetime of the reactor. In order to get informations on the microstructure and mechanical properties of these steels, base metals, HAZ, and weld metals have been characterized. The high amount of phosphorus in weld metals promotes the reverse temper embrittlement that occurs during post-weld heat treatment. The radiation damage structure has been identified by small angle neutron scattering, atomic probe, and transmission electron microscopy. Nanometer-sized clusters of solute atoms, rich in copper with almost the same characteristics as in western pressure vessels steels, and an evolution of the size distribution of vanadium carbides, which are present on dislocation structure, are observed. These defects disappear during post-irradiation tempering. As in western steels, the embrittlement is due to both hardening and reduction of interphase cohesion. The radiation damage specificity of VVER steels arises from their high amount of phosphorus and from their significant density of fine vanadium carbides. (author)

  11. Microstructural evolution in reactor pressure vessel steel under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ohno, Katsumi; Fukuya, Koji [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)

    2000-09-01

    Understanding microstructural changes in reactor pressure vessel steels is important in order to evaluate radiation-induced embrittlement, one of the major aging phenomena affecting the extension of plant life. In this study, actual surveillance test specimens and samples of rector vessel low-alloy steel (A533B steel) irradiated in a research reactor were examined using state-of-the-art techniques to clarify the neutron flux effect on the microstructural changes. These techniques included small angle neutron scattering and atom probes. Microstructural changes which are considered to be the main factors affecting embrittlement, including the production of copper-rich precipitates and the segregation of impurity elements, were confirmed by the results of the study. In addition, the mechanical properties were predicted based on the obtained quantitative data such as the diameters of precipitates. Consequently, the hardening due to irradiation was almost simulated. (author)

  12. Preparation and characterization of 304 stainless steel/Q235 carbon steel composite material

    Science.gov (United States)

    Shen, Wenning; Feng, Lajun; Feng, Hui; Cao, Ying; Liu, Lei; Cao, Mo; Ge, Yanfeng

    The composite material of 304 stainless steel reinforced Q235 carbon steel has been prepared by modified hot-rolling process. The resulted material was characterized by scanning electron microscope, three-electrode method, fault current impact method, electrochemical potentiodynamic polarization curve measurement and electrochemical impedance spectroscopy. The results showed that metallurgical bond between the stainless steel layer and carbon steel substrate has been formed. The composite material exhibited good electrical conductivity and thermal stability. The average grounding resistance of the composite material was about 13/20 of dip galvanized steel. There has no surface crack and bubbling formed after fault current impact. The composite material led to a significant decrease in the corrosion current density in soil solution, compared with that of hot dip galvanized steel and bare carbon steel. On the basis polarization curve and EIS analyses, it can be concluded that the composite material showed improved anti-corrosion property than hot-dip galvanized steel.

  13. Passivation condition of carbon steel in bentonite/sand mixture

    International Nuclear Information System (INIS)

    Taniguchi, Naoki; Kawakami, Susumu

    2002-03-01

    It is essential to understand the corrosion type of carbon steel under the repository conditions for the lifetime assessment of carbon steel overpack used for geological isolation of high-level radioactive waste. According to the previous study, carbon steel is hard to passivate in buffer material assuming a chemical condition range of groundwater in Japan. However, concrete support will be constructed around the overpack in the case of repository in the soft rock system and groundwater having a higher pH may infiltrate to buffer material. There is a possibility that the corrosion type of carbon steel will be influenced by the rise of the pH in groundwater. In this study, anodic polarization experiments were performed to understand the passivation condition of carbon steel in buffer material saturated with water contacted with concrete. An ordinary concrete an a low-alkalinity concrete were used in the experiment. The results of the experiments showed that the carbon steel can passivate under the condition that water having pH > 13 infiltrate to the buffer material assuming present property of buffer material. If the low-alkalinity concrete is selected as the support material, passivation can not occur on carbon steel overpack. The effect of the factors of buffer material such as dry density and mixing ratio of sand on the passivation of carbon steel was also studied. The results of the study showed that the present property of buffer material is enough to prevent passivation of carbon steel. (author)

  14. Studies of fragileness in steels of vessels of BWR reactors

    International Nuclear Information System (INIS)

    Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

    2003-01-01

    The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA MARK lll reactor and separately with Ni +3 ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A 2 . (Author)

  15. Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-01

    The embrittlement trend curve development project for HFIR reactor pressure vessel (RPV) steels was carried out with three major tasks. Which are (1) data collection to match that used in HFIR steel embrittlement trend published in 1994 Journal Nuclear Material by Remec et. al, (2) new embrittlement data of A212B steel that are not included in earlier HFIR RPV trend curve, and (3) the adjustment of nil-ductility-transition temperature (NDTT) shift data with the consideration of the irradiation temperature effect. An updated HFIR RPV steel embrittlement trend curve was developed, as described below. NDTT( C) = 23.85 log(x) + 203.3 log (x) + 434.7, with 2- uncertainty of 34.6 C, where parameter x is referred to total dpa. The developed update HFIR RPV embrittlement trend curve has higher embrittlement rate compared to that of the trend curve developed in 1994.

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

    International Nuclear Information System (INIS)

    Féron, D.; Herms, E.; Tanguy, B.

    2012-01-01

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

  17. Low carbon manganese-nickel-niobium steel

    International Nuclear Information System (INIS)

    Heisterkamp, F.; Hulka, K.

    1983-11-01

    Experimental heats of a low carbon-manganese-0.5% nickel-0.15% niobium steel have been rolled to plates between 13.5 and 50 mm thickness and to a 16 mm hot strip. Various combinations of soaking temperatures form 1100 0 C to 1300 0 C and of finish rolling temperatures between 710 0 C and 930 0 C have been investigated. From mechanical properties obtained, one can conclude that the investigated steel composition provides very good properties e.g. for pipe steels X65 to X75. In particular, the toughness at low temperature is outstanding despite relaxed rolling conditions. Metalographic and special investigations such as electron microscopy, texture evaluation and chemical extraction, correlated with applied rolling schedules and the mechanical properties obtained resulted in a comprehensive understanding about the benefits of high niobium metallurgy combined with nickel addition. All practically applied welding processes generated mechanical properties, in particular toughness of the weldment, that meet arctic specifications.(Author) [pt

  18. Thermoplastic liners for carbon steel pipelines

    Energy Technology Data Exchange (ETDEWEB)

    Mehdi, Mauyed S.; AlDossary, Abdullah K. [Saudi Aramco, Dhahran (Saudi Arabia)

    2009-12-19

    Materials selection for pipe and fittings used to convey corrosive fluids has often been a challenge. Traditionally, exotic Corrosion Resistant Alloys (CRA) have been used in corrosive environments despite their high cost. Plastic lined carbon steel piping offers a cost effective alternative to the use of CRAs by eliminating corrosion, significantly reducing the use of toxic chemicals and the heavy metal usually present in CRAs. Thermoplastic Liners offer the combination of corrosion resistance and mechanical strength, which are unachievable with singular materials. Under pressure conditions, the liner is fully supported by the metalwork, while under vacuum conditions, the liner must be thick enough along with venting system to withstand the collapsing forces created by the negative pressure. Plastic liners have been used successfully to line and protect metallic pipelines for many years and have become an indispensable requirement of the oil and gas industry particularly with water injection and hydrocarbon services. In the case of internally corroded pipes, the use of thermoplastic liners for rehabilitation is an option to extend the lifetime of companies' assets, reduce maintenance cost and increase intervals between T and Is. For new construction, plastic liners in carbon steel pipes can compete technically and economically with pipelines of CRA materials and other corrosion inhibition systems. This paper describes various design features, installations of thermoplastic liners in comparison to other corrosion inhibition methods. (author)

  19. Mechanistic studies of carbon steel corrosion inhibition by cashew ...

    African Journals Online (AJOL)

    The phenoxide, R-Ar-O- ions from the CNSL inhibitor were found to be responsible for the reduction of the corrosion rate of the carbon steel. Also, it was observed that the surface charge of the carbon steel electrodes was positive with respect to the solutions containing CNSL inhibitor. It is likely that the mechanism of the ...

  20. Improvement in the long term creep rupture strength of SUS 316 steel for fast breeder reactors by nitrogen addition

    International Nuclear Information System (INIS)

    Nakazawa, Takanori; Abo, Hideo; Tanino, Mitsuru; Komatsu, Hazime; Tashimo, Masanori; Nishida, Takashi.

    1989-01-01

    Improvement of creep fatigue property of structural materials for fast breeder reactors. In order to improve the resistance to creep fatigue of SUS 316 steels, the effects of nitrogen, carbon, and molybdenum on creep properties have been investigated, under the concept that creep fatigue endurance is correspond to creep rupture ductility. Creep rupture tests and slow strain rate tensile tests were conducted at 550degC and extensive microstructural works were performed. The strengthening by nitrogen is much greater than carbon. Moreover, while carbon reduces rupture ductility, nitrogen does not change it. The addition of carbon results in coarse carbide formation on grain boundaries during creep, but with nitrogen very fine Fe 2 Mo particles precipitate on grain boundaries. The difference between the effects of nitrogen and carbon on creep properties is arise from the different morphology of precipitation. Strengthening by molybdenum brings about a slight decrease in rupture ductility. On the basis of these results, 0.01%C-0.07%N-11%Ni-16.5%Cr-2%Mo steel is selected as a promising material for fast breeder reactors. This steel has higher rupture ductility and strength than SUS 316 steel. It is also confirmed that this steel has a higher resistance to creep fatigue. (author)

  1. Stainless steel clad for light water reactor fuels. Final report

    International Nuclear Information System (INIS)

    Rivera, J.E.; Meyer, J.E.

    1980-07-01

    Proper reactor operation and design guidelines are necessary to assure fuel integrity. The occurrence of fuel rod failures for operation in compliance with existing guidelines suggests the need for more adequate or applicable operation/design criteria. The intent of this study is to develop such criteria for light water reactor fuel rods with stainless steel clad and to indicate the nature of uncertainties in its development. The performance areas investigated herein are: long term creepdown and fuel swelling effects on clad dimensional changes and on proximity to clad failure; and short term clad failure possibilities during up-power ramps

  2. Perspective steels for generation IV and fusion reactors

    International Nuclear Information System (INIS)

    Bartosva, I.; Cizek, J.

    2013-01-01

    In this study we focus on the F/M steel Eurofer, the European candidate material for the future fusion reactor and for the strengthening we consider oxides of yttrium. The oxides of yttrium and complex yttrium titanium oxides reinforce the material by forming more or less stable obstacles to dislocations, and by promoting grain refinement by pinning grain boundaries. It appears that part of the yttrium titanium oxides particles dissolves from about 600 grad C while pure Yttria particles are stable at least to 1000 grad C in the steel. The aims of this study are the following: 1) Prove the positive effect of strengthening by yttrium oxides. 2) Measure the hardness of base Eurofer and ODS version by Vickers hardness test (HV). 3) Investigate the behaviour of steels at different annealing temperatures and the changes in strength. 4) Assess defects in microstructure by Coincidence Doppler Broadening (CDB) and Positron Annihilation Lifetime Spectroscopy (PALS) at chosen annealing temperatures. (authors)

  3. Development of austenitic stainless steel tubes for nuclear reactor cladding

    International Nuclear Information System (INIS)

    Padilha, A.F.; Ferreira, P.I.; Andrade, P.I.; Andrade, A.H.P. de; Meyerhof, S.; Mauricio, J.

    1984-01-01

    In the development of thin wall tubes for nuclear reactor fuel cladding applications, a great number of activities, related to the fabrication process as the qualification are involved. A test program was envisaged to verify the quality of seam welded stainless steel tubes (AISI 304), obtained as a result of an effort by the IPEN-CNEN/SP and the brazilian industry. The relevant aspects involved in the preparation of the tubes and some preliminary test results are presented. (Author) [pt

  4. Preparation and characterization of 304 stainless steel/Q235 carbon steel composite material

    Directory of Open Access Journals (Sweden)

    Wenning Shen

    Full Text Available The composite material of 304 stainless steel reinforced Q235 carbon steel has been prepared by modified hot-rolling process. The resulted material was characterized by scanning electron microscope, three-electrode method, fault current impact method, electrochemical potentiodynamic polarization curve measurement and electrochemical impedance spectroscopy. The results showed that metallurgical bond between the stainless steel layer and carbon steel substrate has been formed. The composite material exhibited good electrical conductivity and thermal stability. The average grounding resistance of the composite material was about 13/20 of dip galvanized steel. There has no surface crack and bubbling formed after fault current impact. The composite material led to a significant decrease in the corrosion current density in soil solution, compared with that of hot dip galvanized steel and bare carbon steel. On the basis polarization curve and EIS analyses, it can be concluded that the composite material showed improved anti-corrosion property than hot-dip galvanized steel. Keywords: Stainless steel, Carbon steel, Anti-corrosion, Conductivity, Electrochemical, EIS

  5. Basic studies on carbon steel decontamination

    International Nuclear Information System (INIS)

    Pavarotti, M.; Rizzi, R.; Ronchetti, C.

    1982-01-01

    The dissolution of magnetite films grown in autoclave at high temperature on carbon steel has been performed in a dynamic loop in ammoniated citric and oxalic acid solutions at two different temperatures and constant pH. The dissolution process seems to be affected by the dual-layer oxide morphology depending on the growth conditions in the autoclave. The open-circuit potential of the specimens and the corrosion rate measured by the linear polarization method have been monitored. To this aim a particular corrosion cell and a suitable reference electrode have been set up at CISE. Polarization curves have been performed to check the electrochemical processes involved in the anodic and cathodic area. At last the effect of a corrosion inhibitor, of a complexing and a reducing agent and of temperature has also been studied. The work was carried out in the frame of a CNEN research programme for the development of the CIRENE prototype

  6. Medium carbon vanadium micro alloyed steels for drop forging

    International Nuclear Information System (INIS)

    Jeszensky, Gabor; Plaut, Ronald Lesley

    1992-01-01

    Growing competitiveness of alternative manufacturing routes requires cost minimization in the production of drop forged components. The authors analyse the potential of medium carbon, vanadium microalloyed steels for drop forging. Laboratory and industrial experiments have been carried out emphasizing deformation and temperature cycles, strain rates and dwell times showing a typical processing path, associated mechanical properties and corresponding microstructures. The steels the required levels of mechanical properties on cooling after forging, eliminating subsequent heat treatment. The machinability of V-microalloyed steels is also improved when compared with plain medium carbon steels. (author)

  7. Mixed structures in continuously cooled low-carbon automotive steels

    International Nuclear Information System (INIS)

    Khalid, F.A.; Edmonds, D.V.

    1993-01-01

    Mixed microstructures have been studied in low- carbon microalloyed steels suitable for automotive applications, after continuous cooling from the hot-rolled condition. Microstructural features such as polygonal ferrite, bainitic and acicular ferrite and microphase constituent are identified using transmission electron microscopy. The influence of these mixed structures on the tensile strength, impact toughness and fracture behaviour is examined. It is found that improvements in impact toughness as compared with microalloyed medium- carbon ferrite/pearlite steels can be achieved from these predominantly acicular structures developed by controlling alloy composition and continuous cooling of these lower carbon steels. (orig.)

  8. Estimation of fast neutron fluence in steel specimens type Laguna Verde in TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Galicia A, J.; Francois L, J. L.; Aguilar H, F.

    2015-09-01

    The main purpose of this work is to obtain the fluence of fast neutrons recorded within four specimens of carbon steel, similar to the material having the vessels of the BWR reactors of the nuclear power plant of Laguna Verde when subjected to neutron flux in a experimental facility of the TRIGA Mark III reactor, calculating an irradiation time to age the material so accelerated. For the calculation of the neutron flux in the specimens was used the Monte Carlo code MCNP5. In an initial stage, three sheets of natural molybdenum and molybdenum trioxide (MoO 3 ) were incorporated into a model developed of the TRIGA reactor operating at 1 M Wth, to calculate the resulting activity by setting a certain time of irradiation. The results obtained were compared with experimentally measured activities in these same materials to validate the calculated neutron flux in the model used. Subsequently, the fast neutron flux received by the steel specimens to incorporate them in the experimental facility E-16 of the reactor core model operating at nominal maximum power in steady-state was calculated, already from these calculations the irradiation time required was obtained for values of the neutron flux in the range of 10 18 n/cm 2 , which is estimated for the case of Laguna Verde after 32 years of effective operation at maximum power. (Author)

  9. Radiation damage structure in irradiated and annealed 440 WWER-Type reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Kocik, J.; Keilova, E.

    1993-01-01

    A review of irradiation damages in WWER-type RPV steels based on conventional Transmission Electron Microscopy investigations in a power reactor and a research reactor, is presented; the samples consist in Cr-Mo-V ferritic steel (15Kh2MFA type). The visible part of radiation-induced defects consists of very fine vanadium carbide precipitates, small dislocation loops and black dots (presumably corresponding to clusters and particle embryos formed from vacancies and solute-atoms (vanadium, copper, phosphorus) and carbon associated with vanadium. Radiation-induced defects are concentrated at dislocation substructure during irradiation in a power reactor, revealing the role of radiation-enhanced diffusion in damage structure forming process. Contrarily, the distribution of defects resulting from annealing of specimens irradiated in the research reactor is pre-determined by an homogenous distribution of radiation-induced defects prior to annealing. Increasing the number of re-irradiation and annealing cycles, the amount of dislocation loops among all defects seems to be growing. Simultaneously, the dislocation substructure recovers considerably. (authors). 14 refs., 11 figs., 3 tabs

  10. Radiation damage structure in irradiated and annealed 440 WWER-Type reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Kocik, J; Keilova, E [Czech Nuclear Society, Prague (Czech Republic)

    1994-12-31

    A review of irradiation damages in WWER-type RPV steels based on conventional Transmission Electron Microscopy investigations in a power reactor and a research reactor, is presented; the samples consist in Cr-Mo-V ferritic steel (15Kh2MFA type). The visible part of radiation-induced defects consists of very fine vanadium carbide precipitates, small dislocation loops and black dots (presumably corresponding) to clusters and particle embryos formed from vacancies and solute-atoms (vanadium, copper, phosphorus) and carbon associated with vanadium. Radiation-induced defects are concentrated at dislocation substructure during irradiation in a power reactor, revealing the role of radiation-enhanced diffusion in damage structure forming process. Contrarily, the distribution of defects resulting from annealing of specimens irradiated in the research reactor is pre-determined by an homogenous distribution of radiation-induced defects prior to annealing. Increasing the number of re-irradiation and annealing cycles, the amount of dislocation loops among all defects seems to be growing. Simultaneously, the dislocation substructure recovers considerably. (authors). 14 refs., 11 figs., 3 tabs.

  11. Comparison between instrumented precracked Charpy and compact specimen tests of carbon steels

    International Nuclear Information System (INIS)

    Nanstad, R.K.

    1980-01-01

    The General Atomic Company High Temperature Gas-Cooled Reactor (HTGR) is housed within a prestressed concrete reactor vessel (PCRV). Various carbon steel structural members serve as closures at penetrations in the vessel. A program of testing and evaluation is underway to determine the need for reference fracture toughness (K/sub IR/) and indexing procedures for these materials as described in Appendix G to Section III, ASME Code for light water reactor steels. The materials of interest are carbon steel forgings (SA508, Class 1) and plates (SA537, Classes 1 and 2) as well as weldments of these steels. The fracture toughness behavior is characterized with instrumented precracked Charpy V-votch specimens (PCVN) - slow-bend and dynamic - and compact specimens (10-mm and 25-mm thicknesses) using both linear elastic (ASTM E399) and elastic-plastic (equivalent Energy and J-Integral) analytical procedures. For the dynamic PCVN tests, force-time traces are analyzed according to the procedures of the Pressure Vessel Research Council (PVRC)/Metal Properties Council (MPC). Testing and analytical procedures are discussed and PCVN results are compared to those obtained with compact specimens

  12. Marine atmospheric corrosion of carbon steels

    Directory of Open Access Journals (Sweden)

    Morcillo, Manuel

    2015-06-01

    Full Text Available Basic research on marine atmospheric corrosion of carbon steels is a relatively young scientific field and there continue to be great gaps in this area of knowledge. The presence of akaganeite in the corrosion products that form on steel when it is exposed to marine atmospheres leads to a notable increase in the corrosion rate. This work addresses the following issues: (a environmental conditions necessary for akaganeite formation; (b characterisation of akaganeite in the corrosion products formed; (c corrosion mechanisms of carbon steel in marine atmospheres; (d exfoliation of rust layers formed in highly aggressive marine atmospheres; (e long-term corrosion rate prediction; and (f behaviour of weathering steels. Field research has been carried out at Cabo Vilano wind farm (Camariñas, Galicia in a wide range of atmospheric salinities and laboratory work involving the use of conventional atmospheric corrosion techniques and near-surface and bulk sensitive analytical techniques: scanning electron microscopy (SEM/energy dispersive spectrometry (EDS, X-ray diffraction (XRD, Mössbauer spectroscopy and SEM/μRaman spectroscopy.La investigación fundamental en corrosión atmosférica marina de aceros al carbono es un campo científico relativamente joven que presenta grandes lagunas de conocimiento. La formación de akaganeíta en los productos de corrosión que se forman sobre el acero cuando se expone a atmósferas marinas conduce a un incremento notable de la velocidad de corrosión. En el trabajo se abordan las siguientes cuestiones: (a condiciones ambientales necesarias para la formación de akaganeíta, (b caracterización de la akaganeíta en los productos de corrosión formados, (c mecanismos de corrosión del acero al carbono en atmósferas marinas, (d exfoliación de las capas de herrumbre formadas en atmósferas marinas muy agresivas, (e predicción de la velocidad de corrosión a largo plazo, y (f comportamiento de aceros patinables. La

  13. Corrosion of austenitic and ferritic-martensitic steels exposed to supercritical carbon dioxide

    International Nuclear Information System (INIS)

    Tan, L.; Anderson, M.; Taylor, D.; Allen, T.R.

    2011-01-01

    Highlights: → Oxidation is the primary corrosion phenomenon for the steels exposed to S-CO 2 . → The austenitic steels showed significantly better corrosion resistance than the ferritic-martensitic steels. → Alloying elements (e.g., Mo and Al) showed distinct effects on oxidation behavior. - Abstract: Supercritical carbon dioxide (S-CO 2 ) is a potential coolant for advanced nuclear reactors. The corrosion behavior of austenitic steels (alloys 800H and AL-6XN) and ferritic-martensitic (FM) steels (F91 and HCM12A) exposed to S-CO 2 at 650 deg. C and 20.7 MPa is presented in this work. Oxidation was identified as the primary corrosion phenomenon. Alloy 800H had oxidation resistance superior to AL-6XN. The FM steels were less corrosion resistant than the austenitic steels, which developed thick oxide scales that tended to exfoliate. Detailed microstructure characterization suggests the effect of alloying elements such as Al, Mo, Cr, and Ni on the oxidation of the steels.

  14. Thermal stability study for candidate stainless steels of GEN IV reactors

    International Nuclear Information System (INIS)

    Simeg Veternikova, J.; Degmova, J.; Pekarcikova, M.; Simko, F.; Petriska, M.; Skarba, M.; Mikula, P.; Pupala, M.

    2016-01-01

    Highlights: • Thermal resistance of advanced stainless steels were observed at 1000 °C. • GEN IV candidate steels were confronted to classic AISI steels. • ODS AISI 316 has weaker thermal resistance than classic AISI steel. • Ferritic ODS steels and NF 709 has better thermal resistance than AISI steels. - Abstract: Candidate stainless steels for GEN IV reactors were investigated in term of thermal and corrosion stability at high temperatures. New austenitic steel (NF 709), austenitic ODS steel (ODS 316) and two ferritic ODS steels (MA 956 and MA 957) were exposed to around 1000 °C in inert argon atmosphere at pressure of ∼8 MPa. The steels were further studied in a light of vacancy defects presence by positron annihilation spectroscopy and their thermal resistance was confronted to classic AISI steels. The thermal strain supported a creation of oxide layers observed by scanning electron microscopy (SEM).

  15. Thermal stability study for candidate stainless steels of GEN IV reactors

    Energy Technology Data Exchange (ETDEWEB)

    Simeg Veternikova, J., E-mail: jana.veternikova@stuba.sk [Institute of Nuclear and Physical Engineering, Faculty of Electrical and Information Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia); Degmova, J. [Institute of Nuclear and Physical Engineering, Faculty of Electrical and Information Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia); Pekarcikova, M. [Institute of Materials Science, Faculty of Materials Science and Technology, Slovak University of Technology, Paulinska 16, 917 24 Trnava (Slovakia); Simko, F. [Department of Molten Salts, Institute of Inorganic Chemistry, Slovak Academy of Sciences, Dubravska cesta 9, 845 36 Bratislava (Slovakia); Petriska, M. [Institute of Nuclear and Physical Engineering, Faculty of Electrical and Information Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia); Skarba, M. [Slovak University of Technology, Vazovova 5, 812 43 Bratislava (Slovakia); Mikula, P. [Institute of Nuclear and Physical Engineering, Faculty of Electrical and Information Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia); Pupala, M. [Department of Molten Salts, Institute of Inorganic Chemistry, Slovak Academy of Sciences, Dubravska cesta 9, 845 36 Bratislava (Slovakia)

    2016-11-30

    Highlights: • Thermal resistance of advanced stainless steels were observed at 1000 °C. • GEN IV candidate steels were confronted to classic AISI steels. • ODS AISI 316 has weaker thermal resistance than classic AISI steel. • Ferritic ODS steels and NF 709 has better thermal resistance than AISI steels. - Abstract: Candidate stainless steels for GEN IV reactors were investigated in term of thermal and corrosion stability at high temperatures. New austenitic steel (NF 709), austenitic ODS steel (ODS 316) and two ferritic ODS steels (MA 956 and MA 957) were exposed to around 1000 °C in inert argon atmosphere at pressure of ∼8 MPa. The steels were further studied in a light of vacancy defects presence by positron annihilation spectroscopy and their thermal resistance was confronted to classic AISI steels. The thermal strain supported a creation of oxide layers observed by scanning electron microscopy (SEM).

  16. In-reactor deformation and fracture of austenitic stainless steels

    International Nuclear Information System (INIS)

    Bloom, E.E.; Wolfer, W.G.

    1978-01-01

    An experimental technique for determining in-reactor fracture strain was developed and demonstrated. Differential swelling between a sample holder and a test specimen with a lower swelling rate produced uniaxial deformation. In-reactor deformations of 0.7 to 2.1% were achieved in type 304 stainless steel previously irradiated to fluences up to 8.8 x 10 26 n/m 2 without fracture. These strains are significantly higher than found in postirradiation creep-rupture tests on similar samples. From the measured strain values and published irradiation creep data and correlations, the stress levels during the irradiation were calculated. On the basis of previous postirradiation creep-rupture results, many of the samples that did not fail would be predicted to fail. Thus we conclude that the in-reactor rupture life is longer than predicted by postirradiation tests. Strain in a fractured sample was estimated to be less than 3.8%, and the in-reactor fractures were intergranular--the same fracture mode as found in postirradiation tests. Irradiation creep may relax stresses at crack tips and sliding boundaries, thus retarding the initiation and/or growth of cracks and leading to longer rupture lives in-reactor. However, the very high ductility or superplastic behavior predicted by the strain rate sensitivity of irradiation creep is not achieved because of the eventual interruption of the deformation process by grain boundary fracture

  17. Medium carbon vanadium steels for closed die forging

    International Nuclear Information System (INIS)

    Jeszensky, Gabor; Plaut, Ronald Lesley

    1993-01-01

    This work analyses the medium carbon micro alloyed vanadium potential for closed die forged production. The steels reach the mechanical resistance requests during cooling after forging, eliminating the subsequent thermal treatment. Those steels also present good fatigue resistance and machinability. The industrial scale experiments are also reported

  18. corrosion response of low carbon steel in tropical road mud

    African Journals Online (AJOL)

    Dr Obe

    Corrosion Mitigation efforts using readily available anti- corrosion coatings to protect low carbon steel test coupons against the ... The following protective coating devices were effective: ..... 2 West, J.M (1986): Basic Corrosion and Oxidation,.

  19. An Evaluation of Carbon Steel Corrosion Under Stagnant Seawater Conditions

    National Research Council Canada - National Science Library

    Lee, Jason

    2004-01-01

    Corrosion, of 1020 carbon steel coupons in, natural seawater over a six-month period was more aggressive under stagnant anaerobic conditions than stagnant aerobic conditions as measured by weight loss...

  20. Electrochemical performances of diamond-like carbon coatings on carbon steel, stainless steel, and brass

    International Nuclear Information System (INIS)

    Hadinata, Samuel-Sudibyo; Lee, Ming-Tsung; Pan, Szu-Jung; Tsai, Wen-Ta; Tai, Chen-Yi; Shih, Chuan-Feng

    2013-01-01

    Diamond-like carbon (DLC) coatings have been deposited onto stainless steel, carbon steel and brass by plasma-enhanced chemical vapor deposition, respectively. Atomic arrangement, chemical structure, surface morphology and cross-section microstructure of the DLC coatings were examined by X-ray diffraction, Raman scattering spectroscopy and scanning electron microscopy. The electrochemical behaviors of the DLC coatings in 3.5 wt.% NaCl solution were investigated by performing an open circuit potential (OCP) measurement and a potentiodynamic polarization test. The experimental results showed that properly deposited DLC coatings could cause an increase of OCP by hundreds of millivolts and a reduction of anodic current density by several orders of magnitude as compared to that of the substrate. The results also demonstrated that electrochemical techniques could be used as tools to detect the soundness of the DLC coating by examining OCP and polarization curve, which varied with the form of defect and depended on the type of substrate. - Highlights: ► The substrate could affect the quality of diamond-like carbon (DLC) coating. ► Defect-free DLC coating exhibited extremely low anodic current density. ► The quality of DLC coating on metal could be evaluated by electrochemical test

  1. Electrochemical performances of diamond-like carbon coatings on carbon steel, stainless steel, and brass

    Energy Technology Data Exchange (ETDEWEB)

    Hadinata, Samuel-Sudibyo; Lee, Ming-Tsung [Department of Materials Science and Engineering, National Cheng Kung University, 1, Ta-Hsueh Road, Tainan 701, Taiwan (China); Pan, Szu-Jung [Ocean Energy Research Center, Tainan Hydraulics Laboratory, National Cheng Kung University, 1, Ta-Hsueh Road, Tainan 701, Taiwan (China); Tsai, Wen-Ta, E-mail: wttsai@mail.ncku.edu.tw [Department of Materials Science and Engineering, National Cheng Kung University, 1, Ta-Hsueh Road, Tainan 701, Taiwan (China); Ocean Energy Research Center, Tainan Hydraulics Laboratory, National Cheng Kung University, 1, Ta-Hsueh Road, Tainan 701, Taiwan (China); Tai, Chen-Yi [Ocean Energy Research Center, Tainan Hydraulics Laboratory, National Cheng Kung University, 1, Ta-Hsueh Road, Tainan 701, Taiwan (China); Shih, Chuan-Feng [Ocean Energy Research Center, Tainan Hydraulics Laboratory, National Cheng Kung University, 1, Ta-Hsueh Road, Tainan 701, Taiwan (China); Department of Electrical Engineering, National Cheng Kung University, 1, Ta-Hsueh Road, Tainan 701, Taiwan (China)

    2013-02-01

    Diamond-like carbon (DLC) coatings have been deposited onto stainless steel, carbon steel and brass by plasma-enhanced chemical vapor deposition, respectively. Atomic arrangement, chemical structure, surface morphology and cross-section microstructure of the DLC coatings were examined by X-ray diffraction, Raman scattering spectroscopy and scanning electron microscopy. The electrochemical behaviors of the DLC coatings in 3.5 wt.% NaCl solution were investigated by performing an open circuit potential (OCP) measurement and a potentiodynamic polarization test. The experimental results showed that properly deposited DLC coatings could cause an increase of OCP by hundreds of millivolts and a reduction of anodic current density by several orders of magnitude as compared to that of the substrate. The results also demonstrated that electrochemical techniques could be used as tools to detect the soundness of the DLC coating by examining OCP and polarization curve, which varied with the form of defect and depended on the type of substrate. - Highlights: ► The substrate could affect the quality of diamond-like carbon (DLC) coating. ► Defect-free DLC coating exhibited extremely low anodic current density. ► The quality of DLC coating on metal could be evaluated by electrochemical test.

  2. Application of positron annihilation spectroscopy for investigation of reactor steels

    International Nuclear Information System (INIS)

    Sojak, S.; Slugen, V.; Petriska, M.; Stacho, M.; Veternikova, J.; Sabelova, V.; Egger, W.; Ravelli, L.

    2013-01-01

    Our work is focused on the study of radiation damage simulated by ion implantations and thermal treatment evaluation of RAFM steels in the form of binary Fe-Cr model alloys. In order to study the microstructure recovery after ion irradiation, we applied an approach for restoration of initial physical and mechanical characteristics of structural materials in the form of thermal annealing, with the goal to decrease the size and amount of accumulated defects. The experimental analysis of material damage at microstructural level was performed by the pulsed low energy positron system (PLEPS) [1] at the high intensity positron source NEPOMUC at the Munich research reactor FRM-II. (authors)

  3. Neutron irradiation embrittlement of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Steele, L.E.

    1975-01-01

    The reliability of nuclear power plants depends on the proper functioning of complex components over the whole life on the plant. Particular concern for reliability is directed to the primary pressure boundary. This report focuses on the portion of the primary system exposed to and significantly affected by neutron radiation. Experimental evidence from research programmes and from reactor surveillance programmes has indicated radiation embrittlement of a magnitude sufficient to raise doubts about reactor pressure vessel integrity. The crucial nature of the primary vessel function heightens the need to be alert to this problem, to which, fortunately, there are positive aspects: for example, steels have been developed which are relatively immune to radiation embrittlement. Further, awareness of such embrittlement has led to designs which can accomodate this factor. The nature of nuclear reactors, of the steels used in their construction, and of the procedures for interpreting embrittlement and minimizing the effects are reviewed with reference to the reactors that are expected to play a major role in electric power production from now to about the turn of the century. The report is intended as a manual or guidebook; the aim has been to make each chapter or major sub-division sufficiently comprehensive and self-contained for it to be understood and read independently of the rest of the book. At the same time, it is hoped that the whole is unified enough to make a complete reading useful and interesting to the several classes of reader that are involved with only specific aspects of the topic

  4. Iron cycling at corroding carbon steel surfaces

    Science.gov (United States)

    Lee, Jason S.; McBeth, Joyce M.; Ray, Richard I.; Little, Brenda J.; Emerson, David

    2013-01-01

    Surfaces of carbon steel (CS) exposed to mixed cultures of iron-oxidizing bacteria (FeOB) and dissimilatory iron-reducing bacteria (FeRB) in seawater media under aerobic conditions were rougher than surfaces of CS exposed to pure cultures of either type of microorganism. The roughened surface, demonstrated by profilometry, is an indication of loss of metal from the surface. In the presence of CS, aerobically grown FeOB produced tight, twisted helical stalks encrusted with iron oxides. When CS was exposed anaerobically in the presence of FeRB, some surface oxides were removed. However, when the same FeOB and FeRB were grown together in an aerobic medium, FeOB stalks were less encrusted with iron oxides and appeared less tightly coiled. These observations suggest that iron oxides on the stalks were reduced and solubilized by the FeRB. Roughened surfaces of CS and denuded stalks were replicated with three culture combinations of different species of FeOB and FeRB under three experimental conditions. Measurements of electrochemical polarization resistance established different rates of corrosion of CS in aerobic and anaerobic media, but could not differentiate rate differences between sterile controls and inoculated exposures for a given bulk concentration of dissolved oxygen. Similarly, total iron in the electrolyte could not be used to differentiate treatments. The experiments demonstrate the potential for iron cycling (oxidation and reduction) on corroding CS in aerobic seawater media. PMID:24093730

  5. Steel

    International Nuclear Information System (INIS)

    Zorev, N.N.; Astafiev, A.A.; Loboda, A.S.; Savukov, V.P.; Runov, A.E.; Belov, V.A.; Sobolev, J.V.; Sobolev, V.V.; Pavlov, N.M.; Paton, B.E.

    1977-01-01

    Steels also containing Al, N and arsenic, are suitable for the construction of large components for high-power nuclear reactors due to their good mechanical properties such as good through-hardening, sufficiently low brittleness conversion temperature and slight displacement of the latter with neutron irradiation. Defined steels and their properties are described. (IHOE) [de

  6. Thermonuclear reactor materials composed of glassy carbons

    International Nuclear Information System (INIS)

    Kazumata, Yukio.

    1979-01-01

    Purpose: To improve the durability to plasma radiation by the use of glassy carbon as the structural materials for the first wall and the blanket in thermonuclear devices. Constitution: The glassy carbon (glass-like carbon) is obtained by forming specific organic substances into a predetermined configuration and carbonizing them by heat decomposition under special conditions. They are impermeable carbon material of 1.40 - 1.70 specific gravity, less graphitizable and being almost in isotropic crystal forms in which isotropic structure such as in graphite is scarcely observed. They have an extremely high hardness, are less likely to be damaged when exposed to radiation and have great strength and corrosion resistance. Accordingly, the service life of the reactor walls and the likes can remarkably be increased by using the materials. (Horiuchi, T.)

  7. Characterization of D2 tool steel friction surfaced coatings over low carbon steel

    International Nuclear Information System (INIS)

    Sekharbabu, R.; Rafi, H. Khalid; Rao, K. Prasad

    2013-01-01

    Highlights: • Solid state coating by friction surfacing method. • D2 tool steel is coated over relatively softer low carbon steel. • Defect free interface between tool steel coating and low carbon steel substrate. • D2 coatings exhibited higher hardness and good wear resistance. • Highly refined martensitic microstructure in the coating. - Abstract: In this work D2 tool steel coating is produced over a low carbon steel substrate using friction surfacing process. The process parameters are optimized to get a defect free coating. Microstructural characterization is carried out using optical microscopy, scanning electron microscopy and X-ray diffraction. Infrared thermography is used to measure the thermal profile during friction surfacing of D2 steel. Wear performance of the coating is studied using Pin-on-Disk wear tests. A lower rotational speed of the consumable rod and higher translational speed of the substrate is found to result in thinner coatings. Friction surfaced D2 steel coating showed fine-grained martensitic microstructure compared to the as-received consumable rod which showed predominantly ferrite microstructure. Refinement of carbides in the coating is observed due to the stirring action of the process. The infrared thermography studies showed the peak temperature attained by the D2 coating to be about 1200 °C. The combined effect of martensitic microstructure and refined carbides resulted in higher hardness and wear resistance of the coating

  8. Monitoring Techniques for Microbially Influenced Corrosion of Carbon Steel

    DEFF Research Database (Denmark)

    Hilbert, Lisbeth Rischel

    2000-01-01

    corrosion rates, when biofilm and corrosion products cover the steel surface. However, EIS might be used for detection of MIC. EN is a suitable technique to characterise the type of corrosion attack, but is unsuitable for corrosion rate estimation. The concentric electrodes galvanic probe arrangement......Abstract Monitoring Techniques for Microbially Influenced Corrosion of Carbon Steel Microbially influenced corrosion (MIC) of carbon steel may occur in media with microbiological activity of especially sulphate-reducing bacteria, e.g. on pipelines buried in soil and on marine structures. MIC...... of carbon steel must be monitored on-line in order to provide an efficient protection and control the corrosion. A number of monitoring techniques is industrially used today, and the applicability and reliability of these for monitoring MIC is evaluated. Coupons and ER are recommended as necessary basic...

  9. Archaeologic analogues: Microstructural changes by natural ageing in carbon steels

    International Nuclear Information System (INIS)

    Munoz, Esther Bravo; Fernandez, Jorge Chamon; Arasanz, Javier Guzman; Peces, Raquel Arevalo; Criado, Antonio Javier; Dietz, Christian; Martinez, Juan Antonio; Criado Portal, Antonio Jose

    2006-01-01

    When discussing the container material for highly active radionuclear waste, carbon steel is one of the materials most frequently proposed by the international scientific community. Evidently, security with respect to the container behaviour into deep geological deposits is fundamental. Among other parameters, knowledge about material mechanical properties is essential when designing the container. Time ageing of carbon steel, apart from possible alterations of the chemical composition (e.g. corrosion) involves important microstructural changes, at the scale of centuries and millenniums. The latter may cause variations of the mechanical properties of carbon steel storage containers, with the corresponding risk of possible leakage. In order to properly estimate such risk and to adjust the corresponding mathematical models to reality, the microstructural changes observed in this study on archaeologic samples are evaluated, comparing ancient and modern steels of similar chemical composition and fabrication processes

  10. In situ 3D monitoring of corrosion on carbon steel and ferritic stainless steel embedded in cement paste

    International Nuclear Information System (INIS)

    Itty, Pierre-Adrien; Serdar, Marijana; Meral, Cagla; Parkinson, Dula; MacDowell, Alastair A.; Bjegović, Dubravka; Monteiro, Paulo J.M.

    2014-01-01

    Highlights: • The morphology of the corrosion of steel in cement paste was studied in situ. • During galvanostatic corrosion, carbon steel reinforcement corroded homogeneously. • On ferritic stainless steel, deep corrosion pits formed and caused wider cracks. • The measured rate of steel loss correlated well with Faraday’s law of electrolysis. - Abstract: In a X-ray microcomputed tomography study, active corrosion was induced by galvanostatically corroding steel embedded in cement paste. The results give insight into corrosion product build up, crack formation, leaching of products into the cracks and voids, and differences in morphology of corrosion attack in the case of carbon steel or stainless steel reinforcement. Carbon steel was homogeneously etched away with a homogeneous layer of corrosion products forming at the steel/cement paste interface. For ferritic stainless steel, pits were forming, concentrating the corrosion products locally, which led to more extensive damage on the cement paste cover

  11. Residual stresses in weld-clad reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Bertram, W.

    1975-01-01

    Cladding of low alloy nuclear reactor pressure vessel steel with austenitic stainless steel introduces in heavy section components high residual stresses which may cause microcrack formation in stress relief heat treatment. In this investigation an attempt is made to contribute to the solution of the stress relief cracking problem by determining quantitatively the magnitude and distribution of the residual stresses after cladding and after subsequent stress relief heat treatment. The distribution of residual stresses was determined on the basis of a combined experimental-mathematical procedure. Heavy section plate specimens of low alloy steel as base material were given an austenitic monolayer-cladding using the techniques of strip electrode and plasma hot wire cladding, respectively. A number of plates was stress relief heat treated. Starting from the cladded surface the thickness of the plates was reduced by subsequent removal of layers of material. The elastic strain reaction to the removal of each layer was measured by strain gauges. From the data obtained the biaxial residual stress distribution was computed as a function of thickness using relations which are derived for this particular case. In summary, lower residual stresses are caused by reduced thickness of the components. As the heat input, is decreased at identical base material thickness, the residual stresses are lowered also. The height of the tensile residual stress peak, however, remains approximataly constant. In stress relief annealed condition the residual stresses in the cladding are in tension; in the base material the residual stresses are negligibly small

  12. Damascus steels: history, processing, properties and carbon dating

    International Nuclear Information System (INIS)

    Wadsworth, J.

    2007-01-01

    In the mid-1970s, a class of steels containing high levels of carbon (∼ 1-2 wt% C) was developed for superplastic characteristics - that is, the ability to plastically deform to an extraordinary degree in tension at intermediate temperatures. Because these steels also had excellent room temperature properties, they were developed for their commercial potential. In the late 1970s, we became aware of the striking compositional similarities between these modern steels and the ancient steels of Damascus. This observation led us to revisit the history and metallurgy of Damascus steels and related steels. The legends and origins of Damascus steel date back to the time of Alexander the Great (323 BC) and the medieval Crusades (11th and 12th century AD), and this material has also been the subject of scrutiny by famous scientist in Europe, including Michael Faraday. Modern attempts to reproduce the legendary surface patterns which famously characterized Damascus steels are described. The extend to which the characteristics of Damascus steels are unusual is discussed. Finally, a program on radiocarbon dating was initiated to directly determine the age of about 50 ancient steels, including a Damascus knife, and the results are summarized. (author)

  13. Corrosion of carbon steel in contact with bentonite

    International Nuclear Information System (INIS)

    Dobrev, D.; Vokal, A.; Bruha, P.

    2010-01-01

    Document available in extended abstract form only. Carbon steel canisters were chosen in a number of disposal concepts as reference material for disposal canisters. The corrosion rates of carbon steels in water solution both in aerobic and anaerobic conditions are well known, but only scarce data are available for corrosion behaviour of carbon steels in contact with bentonite. A special apparatus, which enables to measure corrosion rate of carbon steels under conditions simulating conditions in a repository, namely in contact with bentonite under high pressure and elevated temperatures was therefore prepared to study: - Corrosion rate of carbon steels in direct contact with bentonite in comparison with corrosion rate of carbon steels in synthetic bentonite pore water. - Influence of corrosion products on bentonite. The apparatus is composed of corrosion chamber containing a carbon steel disc in direct contact with compacted bentonite. Synthetic granitic water is above compacted bentonite under high pressure (50 - 100 bar) to simulate hydrostatic pressure in a repository. The experiments can be carried out under various temperatures. Bentonites used for experiments were Na-type of bentonite Volclay KWK 80 - 20 and Ca-Mg Czech bentonite from deposit Rokle. Before adding water into corrosion system the corrosion chamber was purged by nitrogen gas. The saturation of bentonite and corrosion rate were monitored by measuring consumption of water, pressure increase caused by swelling pressure of bentonite and by generation of hydrogen. Corrosion rate was also determined after corrosion experiments from weight loss of samples. The results of experiments show that the corrosion behaviour of carbon steels in contact with bentonite is very different from corrosion of carbon steels in water simulating bentonite pore water solution. The corrosion rates of carbon steel in contact with bentonite reached after 30 days of corrosion the values approaching 40 mm/yr contrary to values

  14. Research to sustain cases for Magnox-reactor steel pressure vessels

    International Nuclear Information System (INIS)

    Graham, W.J.

    1997-01-01

    Britain's Magnox Electric plc owns and operates six power stations, each of which has twin gas-cooled reactors of the Magnox-fuel type. The older group of four power stations has steel pressure-circuits. The reactor cores are housed within spherical, steel vessels. This article describes some of the research which is undertaken to sustain the safety cases for these steel vessels which have now been in operation for just over 30 years. (author) 2 figs., 4 refs

  15. Super-Hydrophobic Green Corrosion Inhibitor On Carbon Steel

    Science.gov (United States)

    Hassan, H.; Ismail, A.; Ahmad, S.; Soon, C. F.

    2017-06-01

    There are many examples of organic coatings used for corrosion protection. In particular, hydrophobic and super-hydrophobic coatings are shown to give good protection because of their enhanced ability to slow down transport of water and ions through the coating. The purpose of this research is to develop water repellent coating to avoid direct contact between metal and environment corrosive and mitigate corrosion attack at pipeline system. This water repellent characteristic on super-hydrophobic coating was coated by electrodeposition method. Wettability of carbon steel with super-hydrophobic coating (cerium chloride and myristic acid) and oxidized surface was investigated through contact angle and inhibitor performance test. The inhibitor performance was studied in 25% tannin acid corrosion test at 30°C and 3.5% sodium chloride (NaCl). The water contact angle test was determined by placing a 4-μL water droplet of distilled water. It shows that the wettability of contact angle super-hydrophobic with an angle of 151.60° at zero minute can be classified as super-hydrophobic characteristic. By added tannin acid as inhibitor the corrosion protection on carbon steel becomes more consistent. This reveals that the ability of the coating to withstand with the corrosion attack in the seawater at different period of immersions. The results elucidate that the weight loss increased as the time of exposure increased. However, the corrosion rates for uncoated carbon steel is high compared to coated carbon steel. As a conclusion, from both samples it can be seen that the coated carbon steel has less corrosion rated compared to uncoated carbon steel and addition of inhibitor to the seawater provides more protection to resist corrosion attack on carbon steel.

  16. intercritical heat treatments effects on low carbon steels quenched

    African Journals Online (AJOL)

    DR B. A. EZEKOYE

    Department of Physics and Astronomy, University of Nigeria, Nsukka. 2. E-mail: benjamin.ezekoye@unn.edu.ng; bezekoye@yahoo.com. ABSTRACT. Six low carbon steels containing carbon in the range 0.13-0.18wt%C were studied after intercritical quenching, intercritical quenching with low temperature tempering, ...

  17. Ultra low carbon bainitic (ULCB) steels after quenching and tempering

    International Nuclear Information System (INIS)

    Lis, A.K.; Lis, J.; Kolan, C.; Jeziorski, L.

    1998-01-01

    The mechanical and Charpy V impact strength properties of new advanced ultra low carbon bainitic (ULBC) steels after water quenching and tempering (WQT) have been investigated. Their chemical compositions are given. The nine continuous cooling transformation diagrams (CCT) of the new ULCB steel grades have been established. The CCT diagrams for ULCB N i steels containing 9% Ni - grade 10N9 and 5% Ni - grade HN5MVNb are given. The comparison between CCT diagrams of 3.5%Ni + 1.5%Cu containing steels grade HSLA 100 and HN3MCu is shown. The effect of the increase in carbon and titanium contents in the chemical composition of ULCB M n steels 04G3Ti, 06G3Ti and 09G3Ti on the kinetics of phase transformations during continuous cooling is presented by the shifting CCT diagrams. The Charpy V impact strength and brittle fracture occurence curves are shown. The effect of tempering temperature on tensile properties of WQT HN3MCu steel is shown and Charpy V impact strength curves after different tempering conditions are shown. The optimum tempering temperatures region of HN3MCu steel for high Charpy V impact toughness at law temperatures - 80 o C(193 K) and -120 o C(153 K) is estimated. The effect of tempering temperature on mechanical properties of HN5MVNb steel is given. The low temperature impact Charpy V toughness of HN5MVNb steel is shown. The optimum range of tempering temperature during 1 hour for high toughness of WQT HN5MVNb steel is given. HN3MCu and HN5MVNb steels after WQT have high yield strength YS≥690 MPa and high Charpy V impact toughness KV≥80 J at -100 o C (173K) and KCV≥50 J/cm 2 at - 120 o C (153K) so they may be used for cryogenic applications

  18. Refurbishment of Pakistan research reactor (PARR-1) for stainless steel lining of the reactor pool

    International Nuclear Information System (INIS)

    Salahuddin, A.; Israr, M.; Hussain, M.

    2002-01-01

    Pakistan Research Reactor-1 (PARR-1) is a pool-type research reactor. Reactor aging has resulted in the increase of water seepage from the concrete walls of the reactor pool. To stop the seepage, it was decided to augment the existing pool walls with an inner lining of stainless steel. This could be achieved only if the pool walls could be accessed unhindered and without excessive radiation doses. For this purpose a partial decommissioning was done by removing all active core components including standard/control fuel elements, reflector elements, beam tubes, thermal shield, core support structure, grid plate and the pool's ceramic tiles, etc. An overall decommissioning program was devised which included procedures specific to each item. This led to the development of a fuel transport cask for transportation, and an interim fuel storage bay for temporary storage of fuel elements (until final disposal). The safety of workers and the environment was ensured by the use of specially designed remote handling tools, appropriate shielding and pre-planned exposure reduction procedures based on the ALARA principle. During the implementation of this program, liquid and solid wastes generated were legally disposed of. It is felt that the experience gained during the refurbishment of PARR-1 to install the stainless steel liner will prove useful and better planning and execution for the future decommissioning of PARR-1, in particular, and for other research reactors like PARR-2 (27 kW MNSR), in general. Furthermore, due to the worldwide activities on decommissioning, especially those communicated through the IAEA CRP on 'Decommissioning Techniques for Research Reactors', the importance of early planning has been well recognized. This has made possible the implementation of some early steps like better record keeping, rehiring of trained manpower, and creation of interim and final waste storage. (author)

  19. Analysis of heat transfer in plain carbon steels

    International Nuclear Information System (INIS)

    Han, Heung Nam; Lee, Kyung Jong

    1999-01-01

    During cooling of steels, the heat transfer was controlled by radiation, convection, conduction and heat evolution from phase transformation. To analyze the heat transfer during cooling precisely, the material constants such as density, heat capacity and the heat evolved during transformation were obtained as functions of temperature and chemical composition for each phase observed in plain carbon steel using a thermodynamic analysis based on the sublattice model of Fe-C-Mn system. The results were applied to 0.049 wt% and 0.155 wt% carbon steels with an austenitic stainless steel as reference by developing a proper heat transfer governing equation. The equation was solved using the lumped system method. In addition, using a transformation dilatometer with adequate experimental conditions to clarify the individual heat transfer effect, the transformation heat evolved during cooling and the transformation behavior as well as the temperature change were observed. The predicted temperature profiles during cooling were well agreed with the measured ones

  20. Friction Welding of Titanium and Carbon Steel

    OpenAIRE

    Atsushi, HASUI; Yoichi, KIRA; Faculty of Science and Technology, Keio University; Ishikawajima-Harima Heavy Industries, Co., Ltd.

    1985-01-01

    Titanium-steel is a combination of dissimilar materials, which are difficult to weld in general, owing to inevitable formation of brittle intermetallic compounds. A prominent feature of friction welding process is ability to weld dissimilar materials in many kinds of combinations. This report deals with friction weldabilily of pure titanium and S25C steel, which are 12 mm in diameter. Main results are summarized as follows; (1) Suitable welding conditions to obtain a sound weld, which has a j...

  1. The anaerobic corrosion of carbon steel in concrete

    International Nuclear Information System (INIS)

    Naish, C.C.; Balkwill, P.H.; O'Brien, T.M.; Taylor, K.J.; Marsh, G.P.

    1991-01-01

    This is the final report of a 2 year programme aimed at (1) determining the rate of anaerobic corrosion of steel in concrete, (2) investigating the nature of the corrosion products formed on carbon steel embedded in cementitious material under anaerobic conditions and (3) evaluating the effect of hydrogen over-pressures on the rate of anaerobic corrosion. All experiments have been carried out at temperatures in the range 20-30 0 C, ie ambient conditions. 4 refs.; 19 figs.; 6 tabs

  2. ESTIMATION OF IRREVERSIBLE DAMAGEABILITY AT FATIGUE OF CARBON STEEL

    Directory of Open Access Journals (Sweden)

    I. O. Vakulenko

    2014-04-01

    Full Text Available Purpose. Damageability estimation of carbon steel in the conditions of cyclic loading. Methodology. The steel fragments of railway wheel rim and rail head served as material for research with chemical composition 0.65 % С, 0.67 % Mn, 0.3 % Si, 0.027 % P, 0.028 % S и 0.7 % C, 0.82 % Mn, 0.56 % Si, 0.025 % P, 0.029 % S accordingly. The microstructure of tested steels corresponded to the state of metal after a hot plastic deformation. The fatigue research was conducted in the conditions of symmetric bend using the proof-of-concept machine of type «Saturn-10». Full Wohler diagrams and the lines corresponding to forming of sub-and micro cracks were constructed. The distribution analysis of internal stresses in the metal under cyclic loading was carried out using the microhardness tester of PMT-3 type.Findings. On the basis of fatigue curves for high-carbon steels analysis the positions of borders dividing the areas of convertible and irreversible damages were determined. The article shows that with the growth of carbon concentration in the steel at invariability of the structural state an increase of fatigue limit is observed. At the same time the acceleration of processes, which determine transition terms from the stage of forming of submicrocracks to the microcracks occurs. The research of microhardness distribution in the metal after destruction confirmed the nature of carbon amount influence on the carbon steel characteristics. Originality. Regardless on the stages of breakdown site forming the carbon steels behavior at a fatigue is determined by the ration between the processes of strengthening and softening. At a cyclic loading the heterogeneity of internal stresses distribution decreases with the increase of distance from the destruction surface. Analysis of metal internal restructuring processes at fatigue loading made it possible to determine that at the stages prior to incubation period in the metal microvolumes the cells are already

  3. Carbon-14 speciation during anoxic corrosion of activated steel in a repository environment

    Energy Technology Data Exchange (ETDEWEB)

    Wieland, E.; Cvetkovic, B.Z.; Kunz, D. [Paul Scherrer Institute, Villigen (Switzerland). Lab. for Waste Management; Salazar, G.; Szidat, S. [Bern Univ. (Switzerland). Dept. of Chemistry and Biochemistry and Oeschger Centre for Climate Change Research

    2018-01-15

    Radioactive waste contains significant amounts of {sup 14}C which has been identified a key radionuclide in safety assessments. In Switzerland, the {sup 14}C inventory of a cement-based repository for low- and intermediate-level radioactive waste (L/ILW) is mainly associated with activated steel (∝85 %). {sup 14}C is produced by {sup 14}N activation in steel parts exposed to thermal neutron flux in light water reactors. Release of {sup 14}C occurs in the near field of a deep geological repository due to anoxic corrosion of activated steel. Although the {sup 14}C inventory of the L/ILW repository and the sources of {sup 14}C are well known, the formation of {sup 14}C species during steel corrosion is only poorly understood. The aim of the present study was to identify and quantify the {sup 14}C-bearing carbon species formed during the anoxic corrosion of iron and steel and further to determine the {sup 14}C speciation in a corrosion experiment with activated steel. All experiments were conducted in conditions similar to those anticipated in the near field of a cement-based repository.

  4. Gamma-radiation effect on diamond and steel during their irradiation in WWER type reactors

    International Nuclear Information System (INIS)

    Nikolaenko, V.A.; Karpukhin, V.I.; Amaev, A.D.; Vikhrov, V.I.; Korolev, Yu.N.; Krasikov, E.A.

    1996-01-01

    A study is made into the influence of reactor gamma radiation on expansion of crystal lattice in diamond. The data obtained are compared to those on radiation embrittlement of reactor vessel steels. The necessity of taking into consideration gamma radiation effects on WWER reactor vessel radiation resistance during long-term operation is shown [ru

  5. In situ 3D monitoring of corrosion on carbon steel and ferritic stainless steel embedded in cement paste

    KAUST Repository

    Itty, Pierre-Adrien; Serdar, Marijana; Meral, Cagla; Parkinson, Dula; MacDowell, Alastair A.; Bjegović, Dubravka; Monteiro, Paulo J.M.

    2014-01-01

    In a X-ray microcomputed tomography study, active corrosion was induced by galvanostatically corroding steel embedded in cement paste. The results give insight into corrosion product build up, crack formation, leaching of products into the cracks and voids, and differences in morphology of corrosion attack in the case of carbon steel or stainless steel reinforcement. Carbon steel was homogeneously etched away with a homogeneous layer of corrosion products forming at the steel/cement paste interface. For ferritic stainless steel, pits were forming, concentrating the corrosion products locally, which led to more extensive damage on the cement paste cover. © 2014 Elsevier Ltd.

  6. In situ 3D monitoring of corrosion on carbon steel and ferritic stainless steel embedded in cement paste

    KAUST Repository

    Itty, Pierre-Adrien

    2014-06-01

    In a X-ray microcomputed tomography study, active corrosion was induced by galvanostatically corroding steel embedded in cement paste. The results give insight into corrosion product build up, crack formation, leaching of products into the cracks and voids, and differences in morphology of corrosion attack in the case of carbon steel or stainless steel reinforcement. Carbon steel was homogeneously etched away with a homogeneous layer of corrosion products forming at the steel/cement paste interface. For ferritic stainless steel, pits were forming, concentrating the corrosion products locally, which led to more extensive damage on the cement paste cover. © 2014 Elsevier Ltd.

  7. Content of nitrogen in waste petroleum carbon for steel industries

    International Nuclear Information System (INIS)

    Rios, R.O; Jimenez, A.F; Szieber, C.W; Banchik, A.D

    2004-01-01

    Steel industries use refined carbon as an alloy for steel production. This alloy is produced from waste carbon from the distillation of the petroleum. The refined carbon, called recarburizer, is obtained by calcination at high temperature. Under these thermal conditions the organic molecules decompose and a fraction of the N 2 , S and H 2 , volatile material and moisture are released; while the carbon tends to develop a crystalline structure similar to graphite's. The right combination of calcinations temperature and time in the furnace can optimize the quality of the resulting product. The content of S and N 2 has to be minimized for the use of calcined carbon in the steel industry. Nitrogen content should be reduced by two orders of magnitude, from 1% - 2% down to hundreds of ppm by weight. This work describes the activities undertaken to obtain calcined coke from petroleum from crude oil carbon that satisfies the requirements of the Mercosur standard 02:00-169 (Pending) for use as a carborizer in steels industries. To satisfy the requirements of the Mercosur standards NM 236:00 IRAM-IAS-NM so that graphite is used as a carburizer a content of 300 ppm maximum weight of nitrogen has to be obtained. So the first stage in this development is to define a production process for supplying calcined coke in the range of nitrogen concentrations required by the Mercosur standards (CW)

  8. Nuclear reactors sited deep underground in steel containment vessels

    Energy Technology Data Exchange (ETDEWEB)

    Bourque, Robert [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)

    2006-07-01

    Although nuclear power plants are certainly very safe, they are not perceived as safe by the general populace. Also, there are concerns about overland transport of spent fuel rods and other irradiated components. It is hereby proposed that the nuclear components of nuclear power plants be placed in deep underground steel vessels with secondary coolant fed from them to turbines at or near the surface. All irradiated components, including spent fuel, would remain in the chamber indefinitely. This general concept was suggested by the late Edward Teller, generated some activity 20-25 years ago and appears to be recently reviving in interest. Previous work dealt with issues of geologic stability of underground, possibly reinforced, caverns. This paper presents another approach that makes siting independent of geology by placing the reactor components in a robust steel vessel capable of resisting full overburden pressure as well as pressures resulting from accident scenarios. Structural analysis of the two vessel concepts and approximate estimated costs are presented. This work clears the way for the extensive discussions required to evaluate the advantages of this concept. (author)

  9. Shallow-crack toughness results for reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Theiss, T.J.; Shum, D.K.M.; Rolfe, S.T.

    1992-01-01

    The Heavy Section Steel Technology Program (HSST) is investigating the influence of flaw depth on the fracture toughness of reactor pressure vessel (RPV) steel. To complete this investigation, techniques were developed to determine the fracture toughness from shallow-crack specimens. A total of 38 deep and shallow-crack tests have been performed on beam specimens about 100 mm deep loaded in 3-point bending. Two crack depths (a ∼ 50 and 9 mm) and three beam thicknesses (B ∼ 50, 100, and 150 mm) have been tested. Techniques were developed to estimate the toughness in terms of both the J-integral and crack-tip opening displacement (CTOD). Analytical J-integral results were consistent with experimental J-integral results, confirming the validity of the J-estimation schemes used and the effect of flaw depth on fracture toughness. Test results indicate a significant increase in the fracture toughness associated with the shallow flaw specimens in the lower transition region compared to the deep-crack fracture toughness. There is, however, little or no difference in toughness on the lower shelf where linear-elastic conditions exist for specimens with either deep or shallow flaws. The increase in shallow-flaw toughness compared with deep-flaw results appears to be well characterized by a temperature shift of 35 degree C

  10. Fatigue behaviour of friction welded medium carbon steel and austenitic stainless steel dissimilar joints

    International Nuclear Information System (INIS)

    Paventhan, R.; Lakshminarayanan, P.R.; Balasubramanian, V.

    2011-01-01

    Research highlights: → Fusion welding of dissimilar metals is a problem due to difference in properties. → Solid state welding process such as friction welding is a solution for the above problem. → Fatigue life of friction welded carbon steel and stainless steel joints are evaluated. → Effect of notch on the fatigue life of friction welded dissimilar joints is reported. → Formation of intermetallic is responsible for reduction in fatigue life of dissimilar joints. -- Abstract: This paper reports the fatigue behaviour of friction welded medium carbon steel-austenitic stainless steel (MCS-ASS) dissimilar joints. Commercial grade medium carbon steel rods of 12 mm diameter and AISI 304 grade austenitic stainless steel rods of 12 mm diameter were used to fabricate the joints. A constant speed, continuous drive friction welding machine was used to fabricate the joints. Fatigue life of the joints was evaluated conducting the experiments using rotary bending fatigue testing machine (R = -1). Applied stress vs. number of cycles to failure (S-N) curve was plotted for unnotched and notched specimens. Basquin constants, fatigue strength, fatigue notch factor and notch sensitivity factor were evaluated for the dissimilar joints. Fatigue strength of the joints is correlated with microstructure, microhardness and tensile properties of the joints.

  11. Trial manufacturing of titanium-carbon steel composite overpack

    International Nuclear Information System (INIS)

    Honma, Nobuyuki; Chiba, Takahiko; Tanai, Kenji

    1999-11-01

    This paper reports the results of design analysis and trial manufacturing of full-scale titanium-carbon steel composite overpacks. The overpack is one of the key components of the engineered barrier system, hence, it is necessary to confirm the applicability of current technique in their manufacture. The required thickness was calculated according to mechanical resistance analysis, based on models used in current nuclear facilities. The Adequacy of the calculated dimensions was confirmed by finite-element methods. To investigate the necessity of a radiation shielding function of the overpack, the irradiation from vitrified waste has been calculated. As a result, it was shown that shielding on handling and transport equipment is a more reasonable and practical approach than to increase thickness of overpack to attain a self-shielding capability. After the above investigation, trial manufacturing of full-scale model of titanium-carbon steel composite overpack has been carried out. For corrosion-resistant material, ASTM Grade-2 titanium was selected. The titanium layer was bonded individually to a cylindrical shell and fiat cover plates (top and bottom) made of carbon steel. For the cylindrical shell portion, a cylindrically formed titanium layer was fitted to the inner carbon steel vessel by shrinkage. For the flat cover plates (top and bottom), titanium plate material was coated by explosive bonding. Electron beam welding and gas metal arc welding were combined to weld of the cover plates to the body. No significant failure was evident from inspections of the fabrication process, and the applicability of current technology for manufacturing titanium-carbon steel composite overpack was confirmed. Future research and development items regarding titanium-carbon steel composite overpacks are also discussed. (author)

  12. Irradiation Microstructure of Austenitic Steels and Cast Steels Irradiated in the BOR-60 Reactor at 320°C

    Science.gov (United States)

    Yang, Yong; Chen, Yiren; Huang, Yina; Allen, Todd; Rao, Appajosula

    Reactor internal components are subjected to neutron irradiation in light water reactors, and with the aging of nuclear power plants around the world, irradiation-induced material degradations are of concern for reactor internals. Irradiation-induced defects resulting from displacement damage are critical for understanding degradation in structural materials. In the present work, microstructural changes due to irradiation in austenitic stainless steels and cast steels were characterized using transmission electron microscopy. The specimens were irradiated in the BOR-60 reactor, a fast breeder reactor, up to 40 dpa at 320°C. The dose rate was approximately 9.4x10-7 dpa/s. Void swelling and irradiation defects were analyzed for these specimens. A high density of faulted loops dominated the irradiated-altered microstructures. Along with previous TEM results, a dose dependence of the defect structure was established at 320°C.

  13. Estimation of residual stresses in reactor pressure vessel steel specimens clad by stainless steel strip electrodes

    International Nuclear Information System (INIS)

    Schimmoeller, H.A.; Ruge, J.L.

    1978-01-01

    The equations to determine a two-dimensional state of residual stress in flat laminated plates are well known from an earlier work by one of the authors. The derivation of these equations leads to a linear, inhomogeneous system of Volterra's integral equations of the second kind. To ascertain the unknown residual stresses from these equations it is necessary to cut down the thickness of the test plate layer by layer. This results in two-dimensional deformation reactions in the rest of the test plate, which can be measured, e.g. by a strain gauge rosette applied to the opposite side of the plate. The above-mentioned stress analysis has been transferred to 86mm thick reactor pressure vessel steel specimens (Type 22NiMoCr 37, DIN-No. 1.6751, similar to ASTM A508, Class 2) double-run clad by austenitic stainless steel strip electrodes (first layer 24/13 Cr-Ni steel, second layer 21/10 Cr-Ni steel). The overall dimensions of the clad specimens investigated amounted to 200 x 200 x (86+4.5+4.5)mm. At the surface of the austenitic cladding there is a two-dimensional tensile normal stress state of about 200N/mm 2 parallel, and about 300N/mm 2 transverse, to the welding direction. The maximum tensile stress was 8mm below the interface (fusion line, material transition) in the parent material. The stress distributions of the specimens investigated, determined on the basis of the above-mentioned combined experimental mathematical procedure, are presented graphically for the as-welded (as-delivered) and annealed (600 0 C/12hr) conditions. (author)

  14. Electrochemical noise from corroding carbon steel and aluminium

    International Nuclear Information System (INIS)

    Singh, P.R.; Gaonkar, K.B.; De, P.K.; Banerjee, S.

    1997-05-01

    Electrochemical noise measurements were conducted on carbon steel and aluminium in sodium chloride solutions. Noise parameters like standard deviation of potential and current, noise resistance, pitting index, noise power were studied for the purpose of measuring corrosion rate. These parameters compared well with the corrosion rate. Pitting index was not very reliable. Current noise was more close to the corrosion rates. General corrosion gave rise to white noise type of power spectrum while flicker noise type of spectrum was obtained from pitting attack. Sodium nitrite is shown to inhibit the corrosion of carbon steel. Aluminium corrodes in the early period of exposure and passivates during long exposure

  15. Carbon-14 production in nuclear reactors

    International Nuclear Information System (INIS)

    Davis, W. Jr.

    1977-01-01

    The radioactive nuclide 14 C is formed in all nuclear reactors due to absorption of neutrons by carbon, nitrogen, or oxygen. These may be present as components of the fuel, moderator, or structural hardware, or they may be present as impurities. Most of the 14 C formed in the fuels or in the graphite of HTGRs will be converted to a gaseous form at the fuel reprocessing plant, primarily as carbon dioxide; this will be released to the environment unless special equipment is installed to collect it and convert it to a solid for essentially permanent storage. If the 14 C is released as carbon dioxide or in any other chemical form, it will enter the biosphere, be inhaled or ingested as food by nearly all living organisms including man, and will thus contribute to the radiation burden of these organisms. Detailed estimates are presented of the amounts of 14 C formed in LWRs, HTGR, and LMFBR with emphasis on those pathways that are likely to lead to the release of this nuclide, either at the reactor site or at the fuel reprocessing plant. 83 references

  16. Strength of low-carbon rotor steel

    International Nuclear Information System (INIS)

    Voropaev, V.I.; Filimonov, O.V.; Borisov, I.A.

    1988-01-01

    The results of studying the effect of chemical composition and thermal treatment regimes on the structural strength of steels of the 25KhN3MFA type are presented. It is shown that alloying with niobium from 0.01 to 0.08% steels with the increased nickel content (4.2-4.5%) contributes to the increase of structural strength and reduction of semibrittleness temperature. To obtain high values of strength and plastic properties cooling with the rate of 10 3 -10 5 K/hr is recommended

  17. Effects of LWR coolant environments on fatigue design curves of carbon and low-alloy steels

    International Nuclear Information System (INIS)

    Chopra, O.K.; Shack, W.J.

    1998-03-01

    The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the code specify fatigue design curves for structural materials. While effects of reactor coolant environments are not explicitly addressed by the design curves, test data indicate that the Code fatigue curves may not always be adequate in coolant environments. This report summarizes work performed by Argonne National Laboratory on fatigue of carbon and low-alloy steels in light water reactor (LWR) environments. The existing fatigue S-N data have been evaluated to establish the effects of various material and loading variables such as steel type, dissolved oxygen level, strain range, strain rate, temperature, orientation, and sulfur content on the fatigue life of these steels. Statistical models have been developed for estimating the fatigue S-N curves as a function of material, loading, and environmental variables. The results have been used to estimate the probability of fatigue cracking of reactor components. The different methods for incorporating the effects of LWR coolant environments on the ASME Code fatigue design curves are presented

  18. OF PLAIN CARBON AND LOW ALLOY STEELS

    African Journals Online (AJOL)

    Two steels En 3 and En 39 were given a TiC-TiN. CVD coating in the carburized and uncarburized conditions. The continuity of the coatings and their adherance to the substrate were examined. The thickness of the deposited coatings were also measured, their adherence to the substrate and their thickness was off ected by ...

  19. Aluminide Coating on Stainless Steel for Nuclear Reactor Application: A Preliminary Study

    International Nuclear Information System (INIS)

    Hishamuddin Husain; Zaifol Samsu; Yusof Abdullah; Muhamad Daud

    2015-01-01

    Stainless steels have been used as structural materials in the nuclear reactor since its first generation. Stainless steels type 304 and 316 are commonly used in structural components. Since the first generation materials, improvements were made on Stainless steels. This includes addition of stabilizing elements and by modification of metallurgical structure. This study investigates the formation of aluminide coating on Stainless steels by diffusion to help improve corrosion resistance. Stainless steels type 304 and 316 substrates were immersed in molten aluminium at 750 degree Celsius for 5 minutes. Interaction between molten aluminium and solid to form the outer aluminide coating by hot dipped aluminizing is studied. (Author)

  20. Surface protection of austenitic steels by carbon nanotube coatings

    Science.gov (United States)

    MacLucas, T.; Schütz, S.; Suarez, S.; Mücklich, F.

    2018-03-01

    In the present study, surface protection properties of multiwall carbon nanotubes (CNTs) deposited on polished austenitic stainless steel are evaluated. Electrophoretic deposition is used as a coating technique. Contact angle measurements reveal hydrophilic as well as hydrophobic wetting characteristics of the carbon nanotube coating depending on the additive used for the deposition. Tribological properties of carbon nanotube coatings on steel substrate are determined with a ball-on-disc tribometer. Effective lubrication can be achieved by adding magnesium nitrate as an additive due to the formation of a holding layer detaining CNTs in the contact area. Furthermore, wear track analysis reveals minimal wear on the coated substrate as well as carbon residues providing lubrication. Energy dispersive x-ray spectroscopy is used to qualitatively analyse the elemental composition of the coating and the underlying substrate. The results explain the observed wetting characteristics of each coating. Finally, merely minimal oxidation is detected on the CNT-coated substrate as opposed to the uncoated sample.

  1. Study of crack propagation velocity in steel tanks of PWR type reactor

    International Nuclear Information System (INIS)

    Amzallac, C.; Bernard, J.L.; Slama, G.

    1983-05-01

    Description and results of a serie of tests carried out on crack propagation velocity of steels in PWR environment (pressurized high temperature water), in order to examine the effects of metallurgical parameters such as chemical composition of steel, especially sulfur and carbon content, and steel type (laminate or forged steels), effects of mechanical parameters such as loading ratio, cycle form, frequency and application mode of loads and of chemical parameters (anodal dissolution or fatigue with hydrogen) [fr

  2. Carbon in condensed hydrocarbon phases, steels and cast irons

    Directory of Open Access Journals (Sweden)

    GAFAROVA Victoria Alexandrovna

    2017-11-01

    Full Text Available The article presents a review of studies carried out mainly by the researchers of the Ufa State Petroleum Technological University, which are aimed at detection of new properties of carbon in such condensed media as petroleum and coal pitches, steels and cast irons. Carbon plays an important role in the industry of construction materials being a component of road and roof bitumen and setting the main mechanical properties of steels. It was determined that crystal-like structures appear in classical glass-like substances – pitches which contain several thousands of individual hydrocarbons of various compositions. That significantly extends the concept of crystallinity. In structures of pitches, the control parameter of the staged structuring process is paramagnetism of condensed aromatic hydrocarbons. Fullerenes were detected in steels and cast irons and identified by various methods of spectrometry and microscopy. Fullerene С60, which contains 60 carbon atoms, has diameter of 0,7 nm and is referred to the nanoscale objects, which have a significant influence on the formation of steel and cast iron properties. It was shown that fullerenes appear at all stages of manufacture of cast irons; they are formed during introduction of carbon from the outside, during crystallization of metal in welded joints. Creation of modified fullerene layers in steels makes it possible to improve anticorrosion and tribological properties of structural materials. At the same time, outside diffusion of carbon from the carbon deposits on the metal surface also leads to formation of additional amount of fullerenes. This creates conditions for occurrence of local microdistortions of the structure, which lead to occurrence of cracks. Distribution of fullerenes in iron matrix is difficult to study as the method is labor-intensive, it requires dissolution of the matrix in the hydrofluoric acid and stage fullerene separation with further identification by spectral methods.

  3. Carbon Contamination During Ion Irradiation - Accurate Detection and Characterization of its Effect on Microstructure of Ferritic/Martensitic Steels

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jing; Toloczko, Mychailo B.; Kruska, Karen; Schreiber, Daniel K.; Edwards, Danny J.; Zhu, Zihua; Zhang, Jiandong

    2017-11-17

    Accelerator-based ion beam techniques have been used to study radiation effects in materials for decades. Although carbon contamination induced by ion beam in target materials is a well-known issue, it has not been fully characterized nor quantified for studies in ferritic/martensitic (F/M) steels that are candidate materials for applications such as core structural components in advanced nuclear reactors. It is an especially important issue for this class of material because of the effect of carbon level on precipitate formation. In this paper, the ability to quantify carbon contamination using three common techniques, namely time-of-flight secondary ion mass spectroscopy (ToF-SIMS), atom probe tomography (APT) and transmission electron microscopy (TEM) is compared. Their effectiveness and short-comings in determining carbon contamination will be presented and discussed. The corresponding microstructural changes related to carbon contamination in ion irradiated F/M steels are also presented and briefly discussed.

  4. Nonmetallic inclusions in carbon steel smelted in plasma furnace

    Energy Technology Data Exchange (ETDEWEB)

    Shengelaya, I B; Kostyakov, V N; Nodiy, T K; Imerlishvili, V G; Gavisiani, A G [AN Gruzinskoj SSR, Tbilisi. Inst. Metallurgii

    1979-01-01

    A complex investigation on nonmetallic inclusions in carbon cast iron, smelted in plasma furnace in argon atmosphere and cast partly in the air and partly in argon atmosphere, has been carried out. As compared to open-hearth furnace carbon steel, the test metal was found to contain more oxide inclusions and nitrides; besides, in chromium-containing metal, chromium nitrides form the larger part of nitrides.

  5. Oxidation of ultra low carbon and silicon bearing steels

    Energy Technology Data Exchange (ETDEWEB)

    Suarez, Lucia [CTM - Technologic Centre, Materials Technology Area, Manresa, Barcelona (Spain)], E-mail: lucia.suarez@ctm.com.es; Rodriguez-Calvillo, Pablo [CTM - Technologic Centre, Materials Technology Area, Manresa, Barcelona (Spain)], E-mail: pablo.rodriguez@ctm.com.es; Houbaert, Yvan [Department of Materials Science and Engineering, University of Ghent (Belgium)], E-mail: Yvan.Houbaert@UGent.be; Colas, Rafael [Facultad de Ingenieria Mecanica y Electrica, Universidad Autonoma de Nuevo Leon (Mexico)], E-mail: rcolas@mail.uanl.mx

    2010-06-15

    Oxidation tests were carried out in samples from an ultra low carbon and two silicon bearing steels to determine the distribution and morphology of the oxide species present. The ultra low carbon steel was oxidized for short periods of time within a chamber designed to obtain thin oxide layers by controlling the atmosphere, and for longer times in an electric furnace; the silicon steels were reheated only in the electric furnace. The chamber was constructed to study the behaviour encountered during the short period of time between descaling and rolling in modern continuous mills. It was found that the oxide layers formed on the samples reheated in the electric furnace were made of different oxide species. The specimens treated in the chamber had layers made almost exclusively of wustite. Selected oxide samples were studied by scanning electron microscopy to obtain electron backscattered diffraction patterns, which were used to identify the oxide species in the layer.

  6. A Review on the Potential Use of Austenitic Stainless Steels in Nuclear Fusion Reactors

    Science.gov (United States)

    Şahin, Sümer; Übeyli, Mustafa

    2008-12-01

    Various engineering materials; austenitic stainless steels, ferritic/martensitic steels, vanadium alloys, refractory metals and composites have been suggested as candidate structural materials for nuclear fusion reactors. Among these structural materials, austenitic steels have an advantage of extensive technological database and lower cost compared to other non-ferrous candidates. Furthermore, they have also advantages of very good mechanical properties and fission operation experience. Moreover, modified austenitic stainless (Ni and Mo free) have relatively low residual radioactivity. Nevertheless, they can't withstand high neutron wall load which is required to get high power density in fusion reactors. On the other hand, a protective flowing liquid wall between plasma and solid first wall in these reactors can eliminate this restriction. This study presents an overview of austenitic stainless steels considered to be used in fusion reactors.

  7. Microbial corrosion of carbon steel by sulfate-reducing bacteria:

    DEFF Research Database (Denmark)

    Nielsen, Lars Vendelbo; Hilbert, Lisbeth Rischel

    1997-01-01

    Electrochemical measurements (EIS and DC-polarisation curves) have been conducted on carbon steel coupons exposed in SRB-active environments. Results from EIS measurements show that very large interfacial capacities are found in such systems, and consequently high capacitive currents are to be ex...

  8. Carbon distribution in bainitic steel subjected to deformation

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, Yu. F., E-mail: yufi55@mail.ru [Institute of High Current Electronics SB RAS, Tomsk, 634055 (Russian Federation); National Research Tomsk State University, Tomsk, 634050 (Russian Federation); Nikitina, E. N., E-mail: Nikitina-EN@mail.ru; Gromov, V. E., E-mail: gromov@physics.sibsiu.ru [Siberian State Industrial University, Novokuznetsk, 654007 (Russian Federation)

    2015-10-27

    Analysis of the formation and evolution of carbide phase in medium carbon steel with a bainitic structure during compressive deformation was performed by means of transmission electron diffraction microscopy. Qualitative transformations in carbide phase medium size particles, their density and volume concentration depended on the degree of deformation.

  9. Vanadium Effect on a Medium Carbon Forging Steel

    Directory of Open Access Journals (Sweden)

    Carlos Garcia-Mateo

    2016-05-01

    Full Text Available In the present work the influence of vanadium on the hardenability and the bainitic transformation of a medium carbon steel is analyzed. While V in solid solution enhances the former, it hardly affects bainitic transformation. The results also reveal an unexpected result, an increase of the prior austenite grain size as the V content increases.

  10. Welding of stainless steel clad fuel rods for nuclear reactors

    International Nuclear Information System (INIS)

    Neves, Mauricio David Martins das

    1986-01-01

    This work describes the obtainment of austenitic stainless steel clad fuel rods for nuclear reactors. Two aspects have been emphasized: (a) obtainment and qualification of AISI 304 and 304 L stainless steel tubes; b) the circumferential welding of pipe ends to end plugs of the same alloy followed by qualification of the welds. Tubes with special and characteristic dimensions were obtained by set mandrel drawing. Both, seamed and seamless tubes of 304 and 304 L were obtained.The dimensional accuracy, surface roughness, mechanical properties and microstructural characteristics of the tubes were found to be adequate. The differences in the properties of the tubes with and without seams were found to be insignificant. The TIG process of welding was used. The influence of various welding parameters were studied: shielding gas (argon and helium), welding current, tube rotation speed, arc length, electrode position and gas flow. An inert gas welding chamber was developed and constructed with the aim of reducing surface oxidation and the heat affected zone. The welds were evaluated with the aid of destructive tests (burst-test, microhardness profile determination and metallographic analysis) and non destructive tests (visual inspection, dimensional examination, radiography and helium leak detection). As a function of the results obtained, two different welding cycles have been suggested; one for argon and another for helium. The changes in the microstructure caused by welding have been studied in greater detail. The utilization of work hardened tubes, permitted the identification by optical microscopy and microhardness measurements, of the different zones: weld zone; heat affected zone (region of grain growth, region of total and partial recrystallization) and finally, the zone not affected by heat. Some correlations between the welding parameters and metallurgical phenomena such as: solidification, recovery, recrystallization, grain growth and precipitation that occurred

  11. Aircraft-crash-protected steel reactor building roof structure for the European market

    International Nuclear Information System (INIS)

    Posta, B.A.; Kadar, I.; Rao, A.S.

    1996-01-01

    This paper recommends the use of all steel roof structures for the reactor building of European Boiling Water Reactor (BWR) plants. This change would make the advanced US BWR designs more compatible with European requirements. Replacement of the existing concrete roof slab with a sufficiently thick steel plate would eliminate the concrete spelling resulting from a postulated aircraft crash, potentially damaging the drywell head or the spent fuel pool

  12. Effect of radiation damage on operating safety of steel pressure vessels of nuclear reactors

    International Nuclear Information System (INIS)

    Vacek, M.; Havel, S.; Stoces, B.; Brumovsky, M.

    1980-01-01

    The effects are assessed of the environment upon mechanical properties of steel used generally for pressure vessels of light water nuclear reactors. Changes caused by radiation affect the reliability of vessels. Deterioration of steel properties is mainly due to neutron radiation. The article deals with factors bearing upon damage and with methods allowing to evaluate the reliability of vessels and predict their service life. Operating reliability of vessels is very unfavourably affected by planned and accidental reactor transients. (author)

  13. Carbon Equivalence and Weldability of Microalloyed Steels

    Science.gov (United States)

    1989-10-25

    being considered for Navy applications. Wilson et al [181 in their review of copper-adde. steels have shown that in the CCT diagram for modified versions...almost linearly with austenite grain size (Fig. 4.2) [211. Therefore, the net effect of the addition of micro-alloying elements is to shift the CCT ... diagram to shorter times and thereby decrease the amount of martensite in the HAZ. Recently, computation of austenite grain growth in the HAZ’s of micro

  14. Corrosion behaviour of carbon steel in the Tournemire clay

    International Nuclear Information System (INIS)

    Foct, F.; Dridi, W.; Cabrera, J.; Savoye, S.

    2004-01-01

    Carbon steels are possible materials for the fabrication of nuclear waste containers for long term geological disposal in argillaceous environments. Experimental studies of the corrosion behaviour of such materials has been conducted in various conditions. Concerning the numerous laboratory experiments, these conditions (water and clay mixture or compacted clay) mainly concern the bentonite clay that would be used for the engineered barrier. On the opposite, only few in-situ experiments has been conducted directly in the local clay of the repository site (such as Boom clay, etc.). In order to better estimate the corrosion behaviour of carbon steels in natural clay site conditions, an experimental study has been conducted jointly by EDF and IRSN in the argillaceous French site of Tournemire. In this study, A42 carbon steel specimens have been exposed in 3 different zones of the Tournemire clay formation. The first type of environmental conditions concerns a zone where the clay has not been affected by the excavation (EDZ) of the main tunnel neither by the main fracture zone of the clay formation. The second and third ones are located in the EDZ of the tunnel. In the second zone, an additional aerated water flows from the tunnel, whereas it does not in the third place. Some carbon steel specimens have been extracted after several years of exposure to these conditions. The average corrosion rate has been measured by the weight loss technique and the pitting corrosion depth has been evaluated under an optical microscope. Corrosion products have also been characterised by scanning electron microscopy and X-ray diffraction technique. Results are then discussed regarding the surrounding environmental conditions. Calculations of the oxygen transport from the tunnel through the clay and of the clay re-saturation can explain, in a first approach, the corrosion behaviour of the carbon steel in the different tested zones. (authors)

  15. Mechanical properties and fatigue strength of high manganese non-magnetic steel/carbon steel welded joints

    International Nuclear Information System (INIS)

    Nakaji, Eiji; Ikeda, Soichi; Kim, You-Chul; Nakatsuji, Yoshihiro; Horikawa, Kosuke.

    1997-01-01

    The dissimilar materials welded joints of high manganese non-magnetic steel/carbon steel (hereafter referred to as DMW joints), in which weld defects such as hot crack or blowhole are not found, were the good quality. Tensile strength of DMW joints was 10% higher than that of the base metal of carbon steel. In the bend tests, the DMW joints showed the good ductility without crack. Charpy absorbed energy at 0(degC) of the DMW joints was over 120(J) in the bond where it seems to be the lowest. Large hardening or softening was not detected in the heat affected zone. Fatigue strength of the DMW joints is almost the same with that of the welded joints of carbon steel/carbon steel. As the fatigue strength of the DMW joints exceeds the fatigue design standard curve of JSSC for carbon steel welded joints, the DMW joints can be treated the same as the welded joints of carbon steel/carbon steel of which strength is lower than that of high manganese non-magnetic steel, from the viewpoint of the fatigue design. (author)

  16. Current Status of Development of High Nickel Low Alloy Steels for Commercial Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Chul; Lee, B. S.; Park, S. G.; Lee, K. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-12-15

    SA508 Gr.3 Mn-Mo-Ni low alloy steels have been used for nuclear reactor pressure vessel steels up to now. Currently, the design goal of nuclear power plant is focusing at larger capacity and longer lifetime. Requirements of much bigger pressure vessels may cause critical problems in the manufacturing stage as well as for the welding stage. Application of higher strength steel may be required to overcome the technical problems. It is known that a higher strength and fracture toughness of low alloy steels such as SA508 Gr.4N low alloy steel could be achieved by increasing the Ni and Cr contents. Therefore, SA508 Gr.4N low alloy steel is very attractive as eligible RPV steel for the next generation PWR systems. In this report, we propose the possibility of SA508 Gr.4N low alloy steel for an application of next generation commercial RPV, based on the literature research result about development history of the RPV steels and SA508 specification. In addition, we have surveyed the research result of HSLA(High Strength Low Alloy steel), which has similar chemical compositions with SA508 Gr.4N, to understand the problems and the way of improvement of SA508 Gr.4N low alloy steel. And also, we have investigated eastern RPV steel(WWER-1000), which has higher Ni contents compared to western RPV steel.

  17. Topic 1. Steels for light water reactor pressure vessels

    International Nuclear Information System (INIS)

    Brumovsky, M.; Brynda, J.; Kepka, M.; Barackova, L.; Vacek, M.; Havel, S.; Cukr, B.; Protiva, K.; Petrman, I.; Tvrdy, M.; Hyspecka, L.; Mazanec, K.; Kupca, L.; Brezina, M.

    1980-01-01

    Part 1 of the Proceedings consists of papers on the criteria for the selection and comparison of the properties of steel for pressure vessels and on the metallurgy of the said steels, the selection of suitable material for internal tubing systems, the manufacture of high-alloy steels for WWER components, the mechanical and metallurgical properties of steel 22K for WWER 440 pressure components, and of steel 10MnNi2Mo for the WWER primary coolant circuit, and the metallographic assessment of steel 0Kh18N10T. (J.P.)

  18. Galvanic corrosion between carbon steel 1018 and Alloy 600 in crevice with boric acid solution

    International Nuclear Information System (INIS)

    Kim, Dong Jin; Kim, Hong Pyo; Kim, Joung Soo; Machonald, Digby D.

    2005-01-01

    This work dealt with the evaluation of galvanic corrosion rate in a corrosion cell having annular gap of 0.5 mm between carbon steel 1018 and alloy 600 as a function of temperature and boron concentration. Temperature and boron concentration were ranged from 110 to 300 .deg. C and 2000∼10000 ppm, respectively. After the operating temperature of the corrosion cell where the electrolyte was injected was attained at setting temperature, galvanic coupling was made and at the same time galvanic current was measured. The galvanic corrosion rate decreased with time, which was described by corrosion product such as protective film as well as boric acid deposit formed on the carbon steel with time. From the galvanic current obtained as a function of temperature and boron concentration, it was found that the galvanic corrosion rate decreased with temperature while the corrosion rate increased with boron concentration. The experimental results obtained from galvanic corrosion measurement were explained by adhesive property of corrosion product such as protective film, boric acid deposit formed on the carbon steel wall and dehydration of boric acid to be slightly soluble boric acid phase. Moreover the galvanic corrosion rate calculated using initial galvanic coupling current instead of steady state coupling current was remarked, which could give us relatively closer galvanic corrosion rate to real pressurized water reactor

  19. Morphological and microstructural studies on aluminizing coating of carbon steel

    Energy Technology Data Exchange (ETDEWEB)

    Samsu, Zaifol; Othman, Norinsan Kamil; Daud, Abd Razak; Hussein, Hishammuddin [School of Applied Physics, Faculty of Science and Technology, Universiti Kebangsaan Malaysia, 43600 Bangi, Selangor (Malaysia)

    2013-11-27

    Hot dip aluminizing is one of the most effective methods of surface protection for steels and is gradually gaining popularity. The morphology and microstructure of an inter-metallic layer form on the surface of low carbon steel by hot dip aluminization treatment had been studied in detail. This effect has been investigated using optical and scanning electron microscopy, and X-ray diffraction. The result shows that the reaction between the steel and the molten aluminium leads to the formation of Fe–Al inter-metallic compounds on the steel surface. X-ray diffraction and electron microscopic studies showed that a two layer coating was formed consisting of an external Al layer and a (Fe{sub 2}Al{sub 5}) inter metallic on top of the substrate after hot dip aluminizing process. The inter-metallic layer is ‘thick’ and exhibits a finger-like growth into the steel. Microhardness testing shown that the intermetallic layer has high hardness followed by steel substrate and the lowest hardness was Al layer.

  20. Novel sintered ceramic materials incorporated with EAF carbon steel slag

    Science.gov (United States)

    Karayannis, V.; Ntampegliotis, K.; Lamprakopoulos, S.; Papapolymerou, G.; Spiliotis, X.

    2017-01-01

    In the present research, novel sintered clay-based ceramic materials containing electric arc furnace carbon steel slag (EAFC) as a useful admixture were developed and characterized. The environmentally safe management of steel industry waste by-products and their valorization as secondary resources into value-added materials towards circular economy have attracted much attention in the last years. EAF Carbon steel slag in particular, is generated during the manufacture of carbon steel. It is a solid residue mainly composed of rich-in- Fe, Ca and Si compounds. The experimental results show that the beneficial incorporation of lower percentages of EAFC up to 6%wt. into ceramics sintered at 950 °C is attained without significant variations in sintering behavior and physico-mechanical properties. Further heating up to 1100 °C strongly enhances the densification of the ceramic microstructures, thus reducing the porosity and strengthening their mechanical performance. On the other side, in terms of thermal insulation behavior as well as energy consumption savings and production cost alleviation, the optimum sintering temperature appears to be 950 °C.

  1. Microstructural investigations of 0.2% carbon content steel

    Science.gov (United States)

    Tollabimazraehno, Sajjad; Hingerl, Kurt

    2011-10-01

    The effect of thermal annealing to get different phases on low carbon steel was investigated. Steel sheets (0.2 wt. % C) of 900 μm thickness were heat treated to produce different structures. All the samples have the same starting point, transformation to coarse austenite at 900 degree Celsius. The nano indentation results revealed that samples have different hadness. By making conventional SEM micrographs, focus ion beam maps, and Electron backscatter diffraction (EBSD) the microstructural development and grain boundary variation of transformed phases martensite, biainte, tempered martensite and different combination of these phases were studied.

  2. Aerosol measurements from plasma torch cuts on stainless steel, carbon steel, and aluminum

    International Nuclear Information System (INIS)

    Novick, V.J.; Brodrick, C.J.; Crawford, S.; Nasiatka, J.; Pierucci, K.; Reyes, V.; Sambrook, J.; Wrobel, S.; Yeary, J.

    1996-01-01

    The main purpose of this project is to quantify aerosol particle size and generation rates produced by a plasma torch whencutting stainless steel, carbon steel and aluminum. the plasma torch is a common cutting tool used in the dismantling of nuclear facilities. Eventually, other cutting tools will be characterized and the information will be compiled in a user guide to aid in theplanning of both D ampersand D and other cutting operations. The data will be taken from controlled laboratory experiments on uncontaminated metals and field samples taken during D ampersand D operations at ANL nuclear facilities. The plasma torch data was collected from laboratory cutting tests conducted inside of a closed, filtered chamber. The particle size distributions were determined by isokinetically sampling the exhaust duct using a cascade impactor. Cuts on different thicknesses showed there was no observable dependence of the aerosol quantity produced as a function of material thickness for carbon steel. However, data for both stainless steel and aluminum revealed that the aerosol mass produced for these materials appear to have some dependance on thickness, with thinner materials producing tmore aerosols. The results of the laboratory cutting tests show that most measured particle size distributions are bimodal with one mode at about 0.2 μm and the other at about 10 μm. The average Mass Median Aerodynamic Diameters (MMAD's) for these tests are 0.36 ±0.08 μm for stainless steel, 0.48 ±0.17μm for aluminum and 0.52±0.12 μm for carbon steel

  3. Mechanical properties of reactor pressure vessel steels studied by static and dynamic torsion tests

    International Nuclear Information System (INIS)

    Munier, A.; Maamouri, M.; Schaller, R.; Mercier, O.

    1993-01-01

    Internal friction measurements and torsional plastic deformation tests have been performed in reactor pressure vessel steels (unirradiated, irradiated and irradiated/annealed specimens). The results of these experiments have been interpreted with help of transmission electron microscopy observations (conventional and in situ). It is shown how the interactions between screw dislocations and obstacles (Peierls valleys, impurities and precipitates) could explain the low temperature hardening and the irradiation embrittlement of ferritic steels. In addition, it appears that the nondestructive internal friction technique could be used advantageously to follow the evolution of the material properties under irradiation, as for instance the irradiation embrittlement of the reactor pressure vessel steels. (orig.)

  4. Irradiation effects in low-alloy reactor pressure vessel steels (Heavy-Section Steel Technology program series 4 and 5)

    International Nuclear Information System (INIS)

    McGowan, J.J.; Nanstad, R.K.; Thoms, K.R.; Menke, B.H.

    1985-01-01

    This report presents studies on the irradiation effects in low-alloy reactor pressure vessel steels. The Fourth Heavy-Section Steel Technology (HSST) Irradiation Series, almost completed, was aimed at elastic-plastic and fully plastic fracture toughness of low-copper weldments (''current practice welds''). A typical nuclear pressure vessel plate steel was included for statistical purposes. The Fifth HSST Irradiation Series, now in progress, is aimed at determining the shape of the K/sub IR/ curve after significant radiation-induced shift of the transition temperatures. This series includes irradiated test specimens of thicknesses up to 100 mm and weldment compositions typical of early nuclear power reactor pressure vessel welds. 27 refs., 22 figs

  5. Swelling, mechanical properties, and microstructure of Type 316 stainless steel at fusion reactor damage levels

    International Nuclear Information System (INIS)

    Horak, J.A.; Bloom, E.E.; Grossbeck, M.L.; Maziasz, P.J.; Stiegler, J.O.; Wiffen, F.W.

    1979-01-01

    Alloys such as AISI 316 stainless steel exhibit more swelling and larger decreases in ductility when irradiated to produce fusion reactor He and dpa levels than at fast reactor He and dpa levels. For T approx. 0 C to ensure adequate ductility for long-term service

  6. Heat exchange performance of stainless steel and carbon foams modified with carbon nano fibers

    NARCIS (Netherlands)

    Tuzovskaya, I.; Pacheco Benito, Sergio; Chinthaginjala, J.K.; Reed, C.P.; Lefferts, Leonardus; van der Meer, Theodorus H.

    2012-01-01

    Carbon nanofibers (CNF), with fishbone and parallel wall structures, were grown by catalytic chemical vapor deposition on the surface of carbon foam and stainless steel foam, in order to improve their heat exchange performance. Enhancement in heat transfer efficiency between 30% and 75% was achieved

  7. Fatigue life response of ASME SA 106-B steel in pressurized water reactor environments

    Energy Technology Data Exchange (ETDEWEB)

    Terrell, J B [Materials Engineering Associates, Inc., Lanham, MD (USA)

    1989-01-01

    Fatigue strain-life tests were conducted on ASMESA 106-B piping steel base metal and weld metal specimens in 288{sup 0}C (550{sup 0}F) pressurized water reactor (PWR) environments as a function of strain amplitude, strain ratio, notch acuity, and cyclic frequency. Notched base metal specimens tested at 0.017 Hz in 0.001 part per million (ppm) dissolved oxygen environments nearly completely used up the margins of safety of 2 on stress and 20 on cycles incorporated into the ASMA Section III design curve for carbon steels. Tests conducted with smooth base metal and weld metal specimens at 1.0 Hz showed virtually no degradation in cycles to failure when compared to 288{sup 0}C air test results. In all cases, however, the effect of temperature alone reduced the margin of safety offered by the design curve in the low cycle regime for the test specimens. Comparison between the fatigue life results of smooth and notched specimens suggests that fatigue crack initiation is not significantly affected by 0.001 ppm dissolved oxygen, and that most of the observed degradation may be attributed to crack growth acceleration. These results suggest that the ASMA Section III methodology should be reviewed, taking into account the PWR environment variables which degrade the fatigue life of pressure-retaining components. (author).

  8. Fatigue life response of ASME SA 106-B steel in pressurized water reactor environments

    International Nuclear Information System (INIS)

    Terrell, J.B.

    1989-01-01

    Fatigue strain-life tests were conducted on ASMESA 106-B piping steel base metal and weld metal specimens in 288 0 C (550 0 F) pressurized water reactor (PWR) environments as a function of strain amplitude, strain ratio, notch acuity, and cyclic frequency. Notched base metal specimens tested at 0.017 Hz in 0.001 part per million (ppm) dissolved oxygen environments nearly completely used up the margins of safety of 2 on stress and 20 on cycles incorporated into the ASMA Section III design curve for carbon steels. Tests conducted with smooth base metal and weld metal specimens at 1.0 Hz showed virtually no degradation in cycles to failure when compared to 288 0 C air test results. In all cases, however, the effect of temperature alone reduced the margin of safety offered by the design curve in the low cycle regime for the test specimens. Comparison between the fatigue life results of smooth and notched specimens suggests that fatigue crack initiation is not significantly affected by 0.001 ppm dissolved oxygen, and that most of the observed degradation may be attributed to crack growth acceleration. These results suggest that the ASMA Section III methodology should be reviewed, taking into account the PWR environment variables which degrade the fatigue life of pressure-retaining components. (author)

  9. Effects of tempering on internal friction of carbon steels

    International Nuclear Information System (INIS)

    Hoyos, J.J.; Ghilarducci, A.A.; Salva, H.R.; Chaves, C.A.; Velez, J.M.

    2011-01-01

    Research highlights: → Time tempering dependent microstructure of two steels is studied by internal friction. → Internal friction indicates the interactions of dislocations with carbon and carbides. → Internal friction detects the first stage of tempering. → Precipitation hardening is detected by the decrease in the background. - Abstract: Two steels containing 0.626 and 0.71 wt.% carbon have been studied to determine the effects of tempering on the microstructure and the internal friction. The steels were annealed at 1093 K, quenched into water and tempered for 60 min at 423 K, 573 K and 723 K. The increase of the tempering time diminishes the martensite tetragonality due to the redistribution of carbon atoms from octahedrical interstitial sites to dislocations. Internal friction spectrum is decomposed into five peaks and an exponential background, which are attributed to the carbide precipitation and the dislocation relaxation process. Simultaneous presence of peaks P1 and P2 indicates the interaction of dislocations with the segregated carbon and carbide precipitate.

  10. Elevated-Temperature Ferritic and Martensitic Steels and Their Application to Future Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, RL

    2005-01-31

    In the 1970s, high-chromium (9-12% Cr) ferritic/martensitic steels became candidates for elevated-temperature applications in the core of fast reactors. Steels developed for conventional power plants, such as Sandvik HT9, a nominally Fe-12Cr-1Mo-0.5W-0.5Ni-0.25V-0.2C steel (composition in wt %), were considered in the United States, Europe, and Japan. Now, a new generation of fission reactors is in the planning stage, and ferritic, bainitic, and martensitic steels are again candidates for in-core and out-of-core applications. Since the 1970s, advances have been made in developing steels with 2-12% Cr for conventional power plants that are significant improvements over steels originally considered. This paper will review the development of the new steels to illustrate the advantages they offer for the new reactor concepts. Elevated-temperature mechanical properties will be emphasized. Effects of alloying additions on long-time thermal exposure with and without stress (creep) will be examined. Information on neutron radiation effects will be discussed as it applies to ferritic and martensitic steels.

  11. Swelling and swelling resistance possibilities of austenitic stainless steels in fusion reactors

    International Nuclear Information System (INIS)

    Maziasz, P.J.

    1983-01-01

    Fusion reactor helium generation rates in stainless steels are intermediate to those found in EBR-II and HFIR, and swelling in fusion reactors may differ from the fission swelling behavior. Advanced titanium-modified austenitic stainless steels exhibit much better void swelling resistance than AISI 316 under EBR-II (up to approx. 120 dpa) and HFIR (up to approx. 44 dpa) irradiations. The stability of fine titanium carbide (MC) precipitates plays an important role in void swelling resistance for the cold-worked titanium-modified steels irradiated in EBR-II. Futhermore, increased helium generation in these steels can (a) suppress void conversion, (b) suppress radiation-induced solute segregation (RIS), and (c) stabilize fine MC particles, if sufficient bubble nucleation occurs early in the irradation. The combined effects of helium-enhanced MC stability and helium-suppressed RIS suggest better void swelling resistance in these steels for fusion service than under EBR-II irradiation

  12. Microstructural investigations of fast reactor irradiated austenitic and ferritic-martensitic stainless steel fuel cladding

    International Nuclear Information System (INIS)

    Agueev, V.S.; Medvedeva, E.A.; Mitrofanova, N.M.; Romanueev, V.V.; Tselishev, A.V.

    1992-01-01

    Electron microscopy has been used to characterize the microstructural changes induced in advanced fast reactor fuel claddings fabricated from Cr16Ni15Mo3NbB and Cr16Ni15Mo2Mn2TiVB austenitic stainless steels in the cold worked condition and Cr13Mo2NbVB ferritic -martensitic steel following irradiation in the BOR-60, BN-350 and BN-600 fast reactors. The data are compared with the results obtained from a typical austenitic commercial cladding material, Cr16Ni15Mo3Nb, in the cold worked condition. The results reveal a beneficial effect of boron and other alloying elements in reducing void swelling in 16Cr-15Ni type austenitic steels. The high resistance of ferritic-martensitic steels to void swelling has been confirmed in the Cr13Mo2NbVB steel. (author)

  13. Evaluation of the Steel Creek ecosystem in relation to the proposed restart of L reactor

    International Nuclear Information System (INIS)

    Smith, M.H.; Sharitz, R.R.; Gladden, J.B.

    1981-10-01

    Information is presented on the following subjects: habitat and vegetation, the avifauna, semi-aquatic and terrestrial vertebrates, and aquatic communities of Steel Creek, species of special concern, and radiocesium in Steel Creek. Two main goals of the study were the compilation of a current inventory of the flora and fauna of the Steel Creek ecosystem and an assessment of the probable impacts of radionuclides, primarily 137 Cs, that were released into Steel Creek during earlier reactor operations. Although a thorough evaluation of the impacts of the L reactor restart is impossible at this time, it is concluded that the effects on the Steel Creek ecosystem will be substantial if no mitigative measures are taken

  14. In-reactor creep rupture of 20% cold-worked AISI 316 stainless steel

    International Nuclear Information System (INIS)

    Lovell, A.J.; Chin, B.A.; Gilbert, E.R.

    1981-01-01

    Results of an experiment designed to measure in-reactor stress-to-rupture properties of 20% cold-worked AISI 316 stainless steel are reported. The in-reactor rupture data are compared with postirradiation and unirradiated test results. In-reactor rupture lives were found to exceed rupture predictions of postirradiation tests. This longer in-reactor rupture life is attributed to dynamic point defect generation which is absent during postirradiation testing. The in-reactor stress-to-rupture properties are shown to be equal to or greater than the unirradiated material stress-to-rupture properties for times up to 7000 h. (author)

  15. Attenuation capability of low activation-modified high manganese austenitic stainless steel for fusion reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Eissa, M.M. [Steel Technology Department, Central Metallurgical Research and Development Institute (CMRDI), Helwan (Egypt); El-kameesy, S.U.; El-Fiki, S.A. [Physics Department, Faculty of Science, Ain Shams University, Cairo (Egypt); Ghali, S.N. [Steel Technology Department, Central Metallurgical Research and Development Institute (CMRDI), Helwan (Egypt); El Shazly, R.M. [Physics Department, Faculty of Science, Al-Azhar University, Cairo (Egypt); Saeed, Aly, E-mail: aly_8h@yahoo.com [Nuclear Power station Department, Faculty of Engineering, Egyptian-Russian University, Cairo (Egypt)

    2016-11-15

    Highlights: • Improvement stainless steel alloys to be used in fusion reactors. • Structural, mechanical, attenuation properties of investigated alloys were studied. • Good agreement between experimental and calculated results has been achieved. • The developed alloys could be considered as candidate materials for fusion reactors. - Abstract: Low nickel-high manganese austenitic stainless steel alloys, SSMn9Ni and SSMn10Ni, were developed to use as a shielding material in fusion reactor system. A standard austenitic stainless steel SS316L was prepared and studied as a reference sample. The microstructure properties of the present stainless steel alloys were investigated using Schaeffler diagram, optical microscopy, and X-ray diffraction pattern. Mainly, an austenite phase was observed for the prepared stainless steel alloys. Additionally, a small ferrite phase was observed in SS316L and SSMn10Ni samples. The mechanical properties of the prepared alloys were studied using Vickers hardness and tensile tests at room temperature. The studied manganese stainless steel alloys showed higher hardness, yield strength, and ultimate tensile strength than SS316L. On the other hand, the manganese stainless steel elongation had relatively lower values than the standard SS316L. The removal cross section for both slow and total slow (primary and those slowed down in sample) neutrons were carried out using {sup 241}Am-Be neutron source. Gamma ray attenuation parameters were carried out for different gamma ray energy lines which emitted from {sup 60}Co and {sup 232}Th radioactive sources. The developed manganese stainless steel alloys had a higher total slow removal cross section than SS316L. While the slow neutron and gamma rays were nearly the same for all studied stainless steel alloys. From the obtained results, the developed manganese stainless steel alloys could be considered as candidate materials for fusion reactor system with low activation based on the short life

  16. Reactor Materials Program electrochemical potential measurements by ORNL with unirradiated and irradiated stainless steel specimens

    International Nuclear Information System (INIS)

    Baumann, E.W.; Caskey, G.R. Jr.

    1993-07-01

    Effect of irradiation of stainless steel on electrochemical potential (ECP) was investigated by measurements in dilute HNO 3 and H 2 O 2 solutions, conditions simulating reactor moderator. The electrodes were made from unirradiated/irradiated, unsensitized/sensitized specimens from R-reactor piping. Results were inconclusive because of budgetary restrictions. The dose rate may have been too small to produce a significant radiolytic effect. Neither the earlier CERT corrosion susceptibility tests nor the present ECP measurements showed a pronounced effect of irradiation on susceptibility of the stainless steel to IGSCC; this is confirmed by the absence in the stainless steel of the SRS reactor tanks (except for the C Reactor tank knuckle area)

  17. Reactor Materials Program electrochemical potential measurements by ORNL with unirradiated and irradiated stainless steel specimens

    Energy Technology Data Exchange (ETDEWEB)

    Baumann, E.W.; Caskey, G.R. Jr.

    1993-07-01

    Effect of irradiation of stainless steel on electrochemical potential (ECP) was investigated by measurements in dilute HNO{sub 3} and H{sub 2}O{sub 2} solutions, conditions simulating reactor moderator. The electrodes were made from unirradiated/irradiated, unsensitized/sensitized specimens from R-reactor piping. Results were inconclusive because of budgetary restrictions. The dose rate may have been too small to produce a significant radiolytic effect. Neither the earlier CERT corrosion susceptibility tests nor the present ECP measurements showed a pronounced effect of irradiation on susceptibility of the stainless steel to IGSCC; this is confirmed by the absence in the stainless steel of the SRS reactor tanks (except for the C Reactor tank knuckle area).

  18. Electrochemical corrosion behavior of carbon steel with bulk coating holidays

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    With epoxy coal tar as the coating material, the electrochemical corrosion behavior of Q235 with different kinds of bulk coating holidays has been investigated with EIS (Electrochemical Impedance Spectroscopy) in a 3.5vol% NaCl aqueous solution.The area ratio of bulk coating holiday to total coating area of steel is 4.91%. The experimental results showed that at free corrosionpotential, the corrosion of carbon steel with disbonded coating holiday is heavier than that with broken holiday and disbonded & broken holiday with time; Moreover, the effectiveness of Cathodic Protection (CP) of carbon steel with broken holiday is better than that with disbonded holiday and disbonded & broken holiday on CP potential -850 mV (vs CSE). Further analysis indicated that the two main reasons for corrosion are electrolyte solution slowly penetrating the coating, and crevice corrosion at steel/coating interface near holidays. The ratio of impedance amplitude (Z) of different frequency to minimum frequency is defined as K value. The change rate of K with frequency is related to the type of coating holiday.

  19. Dynamic recrystallization behavior of a medium carbon vanadium microalloyed steel

    International Nuclear Information System (INIS)

    Wei, Hai-lian; Liu, Guo-quan; Xiao, Xiang; Zhang, Ming-he

    2013-01-01

    The dynamic recrystallization behavior of a medium carbon vanadium microalloyed steel was systematically investigated at the temperatures from 900 °C to 1100 °C and strain rates from 0.01 s −1 to 10 s −1 on a Gleeble-1500 thermo-simulation machine. The flow stress constitutive equation of hot deformation for this steel was developed with the activation energy Q being about 273 kJ/mol, which is in reasonable agreement with those reported before. Activation energy analysis showed that vanadium addition in microalloyed steels seemed not to affect the activation energy much. The effect of Zener–Hollomon parameter on the characteristic points of flow curves was studied using the power law relation, and the dependence of critical strain (stress) on peak strain (stress) obeyed a linear equation. Dynamic recrystallization is the most important softening mechanism for the experimental steel during hot compression. The dynamic recrystallization kinetics model of this steel was established based on flow stress and a frequently-used dynamic recrystallization kinetics equation. Dynamic recrystallization microstructure under different deformation conditions was also observed and the dependence of steady-state grain size on the Zener–Hollomon parameter was plotted

  20. Optimum alloy compositions in reduced-activation martensitic 9Cr steels for fusion reactor

    International Nuclear Information System (INIS)

    Abe, F.; Noda, T.; Okada, M.

    1992-01-01

    In order to obtain potential reduced-activation ferritic steels suitable for fusion reactor structures, the effect of alloying elements W and V on the microstructural evolution, toughness, high-temperature creep and irradiation hardening behavior was investigated for simple 9Cr-W and 9Cr-V steels. The creep strength of the 9Cr-W steels increased but their toughness decreased with increasing W concentration. The 9Cr-V steels exhibited poor creep rupture strength, far below that of a conventional 9Cr-1MoVNb steel and poor toughness after aging at 873 K. It was also found that the Δ-ferrite should be avoided, because it degraded both the roughness and high-temperature creep strength. Based on the results on the simple steels, optimized martensitic 9Cr steels were alloy-designed from a standpoint of enough thoughness and high-temperature creep strength. Two kinds of optimized 9Cr steels with low and high levels of W were obtained; 9Cr-1WVTa and 9Cr-3WVTa. These steels indeed exhibited excellent toughness and creep strength, respectively. The 9Cr-1WVTa steel exhibiting an excellent roughness was shown to be the most promising for relatively low-temperature application below 500deg C, where irradiation embrittlement is significant. The 9Cr-3WVTa steel was the most promising for high temperature application above 500deg C from the standpoint of enough high-temperature strength. (orig.)

  1. Ten years of Toarcian argillite - carbon steel in situ interaction

    Energy Technology Data Exchange (ETDEWEB)

    Dauzeres, Alexandre [IRSN, PRP-DGE/SRTG/LETIS, BP 17, 92262 Fontenay-aux-Roses cedex (France); Maillet, Anais [IRSN, PRP-DGE/SRTG/LETIS, BP 17, 92262 Fontenay-aux-Roses cedex (France); UMR CNRS 7285, IC2MP, Batiment B35 - 5, avenue Albert Turpain, 86022 Poitiers cedex (France); Gaudin, Anne [UMR CNRS 6112, LPGN, 2 rue de la Houssiniere, BP 92208, 44322 Nantes cedex 3 (France); El Albani, Abderrazak; Vieillard, Philippe [UMR CNRS 7285, IC2MP, Batiment B35 - 5, avenue Albert Turpain, 86022 Poitiers cedex (France)

    2013-07-01

    In situ interaction experiments over periods of 2, 6, and 10 years between Toarcian argillite and carbon steel discs were carried out in the Tournemire Underground Research Laboratory (URL), yielding a dataset of the materials' geochemical evolution under conditions representative of the future geological disposal of high-level long-lived radioactive wastes. The carbon steel discs were exposed to corrosion due to trapped oxygen. The corrosion rates indicate that the oxidizing transient lasted between 2 and 6 years. A systematic dissolution of calcium phases (Ca-smectite sheets in I/S and calcite) was observed in the iron diffusion halos. The iron release induced mineralogical dissolution and precipitation reactions, which partly clogged the argillite porosity. (authors)

  2. Hardness and adhesion performances of nanocoating on carbon steel

    Science.gov (United States)

    Hasnidawani, J. N.; Azlina, H. N.; Norita, H.; Bonnia, N. N.

    2018-01-01

    Nanocoatings industry has been aggressive in searching for cost-effective alternatives and environmental friendly approaches to manufacture products. Nanocoatings represent an engineering solution to prevent corrosion of the structural parts of ships, insulation and pipelines industries. The adhesion and hardness properties of coating affect material properties. This paper reviews ZnO-SiO2 as nanopowder in nano coating formulation as the agent for new and improved coating performances. Carbon steel on type S50C used as common substrate in nanocoating industry. 3wt% ZnO and 2wt% SiO2 addition of nanoparticles into nanocoating showed the best formulation since hardness and adhesion of nanocoating was good on carbon steel substrate. Incorporation of nanoparticles into coating increased the performances of coating.

  3. Irradiation induced tensile property change of SA 508 Cl.3 reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Chi, Se-Hwan; Hong, Jun-Hwa; Kuk, Il-Hiun

    1998-01-01

    Irradiation induced tensile property change of four kinds of reactor pressure vessel steels manufactured by different steel refining process was compared based on the differences in the unirradiated and irradiated microstructure. Microvickers hardness, indentation, and miniature tensile specimen tests were conducted for mechanical property measurement and optical microscope (OM) and transmission electron microscope (TEM) were used for microstructural characterization. Specimens were 2 irradiated to a neutron fluence of 2.7x10 19 n/cm 2 (E ≥ 1 MeV) at 288 deg. C. Investigation on the unirradiated microstructures showed largely a same microstructure in that tempered acicular bainite and ferrite with bainitic phase prevailing in the unirradiated condition. Band-shaped segregations were also clearly observed except a kind of materials. A large difference in the unirradiated microstructure appeared in the grain size and carbide microstructure. Of carbide microstructures, noticeable differences were observed in the size and distribution of cementite, and bainitic lath microstructures. No noticeable changes were observed in the optical and thin film TEM microstructures after irradiation. Complicated microstructural. state of heat treated bainitic low alloy microstructure prevents easy quantification of microstructural changes due to irradiation. Apparent differences, however, were observed in the results of mechanical testing. Results of tensile testing and hardness measurement show that a steel refined by vacuum carbon deoxidation(VCD) method exhibits the highest radiation hardening behavior. Some of mechanical testing results on irradiated materials were possible to understand based on the initial microstructure, but further investigations using a wide array of sophisticated tools (for example, SANS, APFIM) are required to understand and characterize irradiation induced defects that are responsible for irradiation hardening behavior but are not revealed by

  4. Irradiation induced tensile property change of SA 508 Cl. 3 reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Chi, Se Hwan; Hong, Jun Hwa; Kuk, Il Hiun

    1998-01-01

    Irradiation induced tensile property change of four kinds of reactor pressure vessel steels manufactured by different steel refining process was compared based on the differences in the miniature tensile specimen tests were conducted for mechanical property measurement and optical microscope (OM) and transmission electron microscope (TEM) were used for microstructural characterization. Specimens were irradiated to a neutron fluence of 2.7 x 10 19 n/cm 2 (E ≥ 1 MeV) at 288 deg C. Investigation on the unirradiated microstructures showed largely a same microstructure in that tempered acicular bainite and ferrite with bainitic phase prevailing in the unirradiated condition. Ban-shaped segregations were also clearly observed except a kind of materials. A large difference in the unirradiated microstructure appeared in the grain size and carbide microstructure. Of carbide microstructures, noticeable differences were observed in the size and distribution of cementite, and bainitic lath microstructures. No noticeable changes were observed in the optical and thin film TEM microstructures after irradiation. Complicated microstructural state of heat treated bainitic low alloy microstructure prevents easy quantification of microstructural changes due to irradiation. Apparent differences, however, were observed in the results of mechanical testing. Results of tensile testing and hardness measurement show that a steel refined by vacuum carbon deoxidation (VCD) method exhibits the highest radiation hardening behavior. Some of mechanical testing results on irradiated materials were possible to understand based on the initial microstructure, but further investigations using a wide array of sophisticated tools (for example, SANS, APFIM) are required to understand and characterize irradiation induced defects that are responsible for irradiation hardening behavior but are not revealed by conventional TEM. (author)

  5. Marine Atmospheric Corrosion of Carbon Steel: A Review

    OpenAIRE

    Alc?ntara, Jenifer; de la Fuente, Daniel; Chico, Bel?n; Simancas, Joaqu?n; D?az, Iv?n; Morcillo, Manuel

    2017-01-01

    The atmospheric corrosion of carbon steel is an extensive topic that has been studied over the years by many researchers. However, until relatively recently, surprisingly little attention has been paid to the action of marine chlorides. Corrosion in coastal regions is a particularly relevant issue due the latter’s great importance to human society. About half of the world’s population lives in coastal regions and the industrialisation of developing countries tends to concentrate production pl...

  6. Guide to the periodic inspection of nuclear reactor steel pressure vessels

    International Nuclear Information System (INIS)

    1969-01-01

    This Guide is intended to provide general information and guidance to reactor owners or operators, inspection authorities, certifying authorities or regulatory bodies who are responsible for establishing inspection procedures for specific reactors or reactor types, and for the preparation of national codes or standards. The recommendations of the Guide apply primarily to water-cooled steel reactor vessels which are at a sufficiently early stage of design so that recommendations to provide accessibility for inspection can be incorporated into the early stages of design and inspection planning. However, much of the contents of the Guide are also applicable in part to vessels for other reactor types, such as gas-cooled, pressure-tube, or liquid-metal-cooled reactors, and also to some existing water-cooled reactors and reactors which are in advanced stage of design or construction. 46 refs, figs, 1 tab

  7. Analysis of corrosion products of carbon steel in wet bentonite

    International Nuclear Information System (INIS)

    Osada, Kazuo; Nagano, Tetsushi; Nakayama, Shinichi; Muraoka, Susumu

    1992-02-01

    As a part of evaluation of the long-term durability for the overpack containers for high-level radioactive waste, we have conducted corrosion tests for carbon steel in wet bentonite, a candidate buffer material. The corrosion rates were evaluated by weight difference of carbon steel and corrosion products were analyzed by Fourier transform infrared spectroscopy (FT-IR) and colorimetry. At 40degC, the corrosion rate of carbon steel in wet bentonite was smaller than that in pure water. At 95degC, however, the corrosion rate in wet bentonite was much higher than that in pure water. This high corrosion rate in wet bentonite at 95degC was considered to result from evaporation of moisture in bentonite in contact with the metal. This evaporation led to dryness and then to shrinkage of the bentonite, which generated ununiform contact of the metal with bentonite. Probably, this ununiform contact promoted the local corrosion. The locally corroded parts of specimen in wet bentonite at 95degC were analyzed by Fourier transform infrared microspectroscopy (micro-FT-IR), and lepidocrocite γ-FeO(OH) was found as well as goethite α-FeO(OH). In wet bentonite at 95degC, hematite α-Fe 2 O 3 was identified by means of colorimetry. (author)

  8. Electrochemical and weight-loss study of carbon steel corrosion

    International Nuclear Information System (INIS)

    Thomas, V.J.; Olive, R.P.

    2007-01-01

    The Point Lepreau Generating Station (PLGS) will undergo an 18 month refurbishment project beginning in April, 2008. During this time, most of the carbon steel piping in the primary loop will be drained of water and dried. However, some water will remain during the shutdown due to the lack of drains in some lower points in the piping system. As a result, it is necessary to examine the effect of corrosion during the refurbishment. This study examined the effect of several variables on the corrosion rate of clean carbon steel. Specifically, the effect of oxygen in the system and the presence of chloride ions were evaluated. Corrosion rates were determined using both a weight-loss technique and electrochemical methods. The experiment was conducted at room temperature. The corrosion products from the experiment were analyzed using a Raman microscope. The results of the weight-loss measurements show that the corrosion rate of polished carbon steel is independent of both the presence of oxygen and chloride ions. The electrochemical method failed to yield meaningful results due to the lack of clearly interpretable data and the inherent subjectivity in the analysis. Lepidocricite was found to be the main corrosion product using the Raman microscope. (author)

  9. A study on the irradiation embrittlement and recovery characteristics of light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Chi, Se Hwan; Hong, Jun Hwa; Lee, Bong Sang; Oh, Jong Myung; Song, Sook Hyang; Milan, Brumovsky

    1999-03-01

    The neutron irradiation embrittlement phenomenon of light water RPV steels greatly affects the life span for safe operation of a reactor. Reliable evaluation and prediction of the embrittlement of RPV steels, especially of aged reactors, are of importance to the safe operation of a reactor. In addition, the thermal recovery of embrittled RPV has been recognized as an option for life extension. This study aimed to tracer/refine available technologies for embrittlement characterization and prediction, to prepare relevant materials for several domestic RPV steels of the embrittlement and recovery, and to find out possible remedy for steel property betterment. Small specimen test techniques, magnetic measurement techniques, and the Meechan and Brinkmann's recovery curve analysis method were examined/applied as the evaluation techniques. Results revealed a high irradiation sensitivity in YG 3 RPV steel. Further extended study may be urgently needed. Both the small specimen test technique for the direct determination of fracture toughness, and the magnetic measurement technique for embrittlement evaluation appeared to be continued for the technical improvement and data base preparation. Manufacturing process relevant to the heat treatment appeared to be improved in lowering the irradiation sensitivity of the steel. Further study is needed especially in applying the present techniques to the new structural materials under new irradiation environment of advanced reactors. (author)

  10. Mechanical Performance of Ferritic Martensitic Steels for High Dose Applications in Advanced Nuclear Reactors

    Science.gov (United States)

    Anderoglu, Osman; Byun, Thak Sang; Toloczko, Mychailo; Maloy, Stuart A.

    2013-01-01

    Ferritic/martensitic (F/M) steels are considered for core applications and pressure vessels in Generation IV reactors as well as first walls and blankets for fusion reactors. There are significant scientific data on testing and industrial experience in making this class of alloys worldwide. This experience makes F/M steels an attractive candidate. In this article, tensile behavior, fracture toughness and impact property, and creep behavior of the F/M steels under neutron irradiations to high doses with a focus on high Cr content (8 to 12) are reviewed. Tensile properties are very sensitive to irradiation temperature. Increase in yield and tensile strength (hardening) is accompanied with a loss of ductility and starts at very low doses under irradiation. The degradation of mechanical properties is most pronounced at martensitic steels exhibit a high fracture toughness after irradiation at all temperatures even below 673 K (400 °C), except when tested at room temperature after irradiations below 673 K (400 °C), which shows a significant reduction in fracture toughness. Creep studies showed that for the range of expected stresses in a reactor environment, the stress exponent is expected to be approximately one and the steady state creep rate in the absence of swelling is usually better than austenitic stainless steels both in terms of the creep rate and the temperature sensitivity of creep. In short, F/M steels show excellent promise for high dose applications in nuclear reactors.

  11. A study on the irradiation embrittlement and recovery characteristics of light water reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Chi, Se Hwan; Hong, Jun Hwa; Lee, Bong Sang; Oh, Jong Myung; Song, Sook Hyang; Milan, Brumovsky [NRI Czech (Czech Republic)

    1999-03-01

    The neutron irradiation embrittlement phenomenon of light water RPV steels greatly affects the life span for safe operation of a reactor. Reliable evaluation and prediction of the embrittlement of RPV steels, especially of aged reactors, are of importance to the safe operation of a reactor. In addition, the thermal recovery of embrittled RPV has been recognized as an option for life extension. This study aimed to tracer/refine available technologies for embrittlement characterization and prediction, to prepare relevant materials for several domestic RPV steels of the embrittlement and recovery, and to find out possible remedy for steel property betterment. Small specimen test techniques, magnetic measurement techniques, and the Meechan and Brinkmann's recovery curve analysis method were examined/applied as the evaluation techniques. Results revealed a high irradiation sensitivity in YG 3 RPV steel. Further extended study may be urgently needed. Both the small specimen test technique for the direct determination of fracture toughness, and the magnetic measurement technique for embrittlement evaluation appeared to be continued for the technical improvement and data base preparation. Manufacturing process relevant to the heat treatment appeared to be improved in lowering the irradiation sensitivity of the steel. Further study is needed especially in applying the present techniques to the new structural materials under new irradiation environment of advanced reactors. (author)

  12. Structure and creep of Russian reactor steels with a BCC structure

    Science.gov (United States)

    Sagaradze, V. V.; Kochetkova, T. N.; Kataeva, N. V.; Kozlov, K. A.; Zavalishin, V. A.; Vil'danova, N. F.; Ageev, V. S.; Leont'eva-Smirnova, M. V.; Nikitina, A. A.

    2017-05-01

    The structural phase transformations have been revealed and the characteristics of the creep and long-term strength at 650, 670, and 700°C and 60-140 MPa have been determined in six Russian reactor steels with a bcc structure after quenching and high-temperature tempering. Creep tests were carried out using specially designed longitudinal and transverse microsamples, which were fabricated from the shells of the fuel elements used in the BN-600 fast neutron reactor. It has been found that the creep rate of the reactor bcc steels is determined by the stability of the lath martensitic and ferritic structures in relation to the diffusion processes of recovery and recrystallization. The highest-temperature oxide-free steel contains the maximum amount of the refractory elements and carbides. The steel strengthened by the thermally stable Y-Ti nanooxides has a record high-temperature strength. The creep rate at 700°C and 100 MPa in the samples of this steel is lower by an order of magnitude and the time to fracture is 100 times greater than that in the oxide-free reactor steels.

  13. Reactor-vessel-sectioning demonstration

    International Nuclear Information System (INIS)

    Lundgren, R.A.

    1981-07-01

    A successful technical demonstration of simulated reactor vessel sectioning was completed using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel: an air arc gouger was chosen to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. After the simulated vessel was successfully cut from the SS side, another cut was made, starting from the carbon steel side. This cut was also successful. Cutting from the carbon steel side has the advantages of cost reduction since the air arc gouging step is eliminated and contamination controlled because the molten metal is blown inward

  14. Grain Refinement of Low Carbon Martensitic Steel by Heat Treatment

    Directory of Open Access Journals (Sweden)

    N. V. Kolebina

    2015-01-01

    Full Text Available The low-carbon steels have good corrosion and technological properties. Hot deformation is the main operation in manufacturing the parts from these steels. So one of the important properties of the material is a property of plasticity. The grain size significantly influences on the ductility properties of steel. The grain size of steel depends on the chemical composition of the crystallization process, heat treatment, and steel machining. There are plenty methods to have grain refinement. However, taking into account the large size of the blanks for the hydro turbine parts, the thermal cycling is an advanced method of the grain refinement adaptable to streamlined production. This work experimentally studies the heat treatment influence on the microstructure of the low-carbon 01X13N04 alloy steel and proposes the optimal regime of the heat treatment to provide a significantly reduced grain size. L.M. Kleiner, N.P. Melnikov and I.N. Bogachyova’s works focused both on the microstructure of these steels and on the influence of its parameters on the mechanical properties. The paper focuses mainly on defining an optimal regime of the heat treatment for grain refinement. The phase composition of steel and temperature of phase transformation were defined by the theoretical analysis. The dilatometric experiment was done to determine the precise temperature of the phase transformations. The analysis and comparison of the experimental data with theoretical data and earlier studies have shown that the initial sample has residual stress and chemical heterogeneity. The influence of the heat treatment on the grain size was studied in detail. It is found that at temperatures above 950 ° C there is a high grain growth. It is determined that the optimal number of cycles is two. The postincreasing number of cycles does not cause further reducing grain size because of the accumulative recrystallization process. Based on the results obtained, the thermal cycling

  15. Bainite formation kinetics in high carbon alloyed steel

    International Nuclear Information System (INIS)

    Luzginova, N.V.; Zhao, L.; Sietsma, J.

    2008-01-01

    In recent years, many investigations have been carried out on the modeling of the bainite formation. In the present work, a physical approach proposed in the literature is implemented to model the formation of lower bainite in high carbon steels (1 wt.% C). In this model, the carbon diffusion is assumed to control the kinetics of the bainite formation. Both the nucleation and the growth rates are considered in an Avrami type analysis. The effect of alloying elements is taken into account considering only the thermodynamics of the system. The results and the physical meaning of the model parameters are discussed. It is shown that the diffusional approach gives a reasonable description of bainite formation kinetics in high carbon steel. Only two fitting parameters are used: the first accounts for carbon grain-boundary diffusion and the second is the initial nucleation-site density. The model satisfactorily accounts for the effect of transformation temperature, but does not take into account the carbide precipitation during bainite formation and the effect of alloying elements on the diffusion coefficient of carbon

  16. European development of ferritic-martensitic steels for fast reactor wrapper applications

    International Nuclear Information System (INIS)

    Bagley, K.; Little, E.A.; Levy, V.; Alamo, A.

    1987-01-01

    9-12%Cr ferritic-martensitic stainless steels are under development in Europe for fast reactor sub-assembly wrapper applications. Within this class of alloys, attention is focussed on three key specifications, viz. FV448 and DIN 1.4914 (both 10-12%CrMoVNb steels) and EM10 (an 8-10%Cr-0.15%C steel), which can be optimized to give acceptably low ductile-brittle transition characteristics. The results of studies on these steels, and earlier choices, covering heat treatment and compositional optimization, evolution of wrapper fabrication routes, pre and post-irradiation mechanical property and fracture toughness behaviour, microstructural stability, void swelling and in-reactor creep characteristics are reviewed. The retention of high void swelling to displacement doses in excess of 100 dpa in reactor irradiations reaffirms the selection of 9-12%Cr steels for on-going wrapper development. Moreover, irradiation-induced changes in mechanical properties (e.g. in-reactor creep and impact behaviour), measured to intermediate doses, do not give cause for concern; however, additional data to higher doses and at the lower irradiation temperatures of 370 0 -400 0 C are needed in order to fully endorse these alloys for high burnup applications in advanced reactor systems

  17. Mineral CO2 sequestration by steel slag carbonation

    International Nuclear Information System (INIS)

    Huijgen, W.J.J.; Comans, R.N.J.; Witkamp, G.J.

    2005-12-01

    Mineral CO2 sequestration, i.e., carbonation of alkaline silicate Ca/Mg minerals, analogous to natural weathering processes, is a possible technology for the reduction of carbon dioxide emissions to the atmosphere. In this paper, alkaline Ca-rich industrial residues are presented as a possible feedstock for mineral CO2 sequestration. These materials are cheap, available near large point sources of CO2, and tend to react relatively rapidly with CO2 due to their chemical instability. Ground steel slag was carbonated in aqueous suspensions to study its reaction mechanisms. Process variables, such as particle size, temperature, carbon dioxide pressure, and reaction time, were systematically varied, and their influence on the carbonation rate was investigated. The maximum carbonation degree reached was 74% of the Ca content in 30 min at 19 bar pressure, 100C, and a particle size of <38 μm. The two must important factors determining the reaction rare are particle size (<2 mm to <38 μm) and reaction temperature (25-225C). The carbonation reaction was found to occur in two steps: (1) leaching of calcium from the steel slag particles into the solution; (2) precipitation of calcite on the surface of these particles. The first step and, more in particular, the diffusion of calcium through the solid matrix toward the surface appeared to be the rate-determining reaction step, The Ca diffusion was found to be hindered by the formation of a CaCO3-coating and a Ca-depleted silicate zona during the carbonation process. Research on further enhancement of the reaction rate, which would contribute to the development of a cost-effective CO2-sequestration process, should focus particularly on this mechanism

  18. Radiofrequency cold plasma nitrided carbon steel: Microstructural and micromechanical characterizations

    International Nuclear Information System (INIS)

    Bouanis, F.Z.; Bentiss, F.; Bellayer, S.; Vogt, J.B.; Jama, C.

    2011-01-01

    Highlights: → C38 carbon steel samples were plasma nitrided using a radiofrequency (rf) nitrogen plasma discharge. → RF plasma treatment enables nitriding for non-heated substrates. → The morphological and chemical analyses show the formation of a uniform thickness on the surface of the nitrided C38 steel. → Nitrogen plasma active species diffuse into the samples and lead to the formation of Fe x N. → The increase in microhardness values for nitrided samples with plasma processing time is interpreted by the formation of a thicker nitrided layer on the steel surface. - Abstract: In this work, C38 carbon steel was plasma nitrided using a radiofrequency (rf) nitrogen plasma discharge on non-heated substrates. General characterizations were performed to compare the chemical compositions, the microstructures and hardness of the untreated and plasma treated surfaces. The plasma nitriding was carried out on non-heated substrates at a pressure of 16.8 Pa, using N 2 gas. Surface characterizations before and after N 2 plasma treatment were performed by means of the electron probe microanalysis (EPMA), X-ray photoelectron spectroscopy (XPS) and Vickers microhardness measurements. The morphological and chemical analysis showed the formation of a uniform structure on the surface of the nitrided sample with enrichment in nitrogen when compared to untreated sample. The thickness of the nitride layer formed depends on the treatment time duration and is approximately 14 μm for 10 h of plasma treatment. XPS was employed to obtain chemical-state information of the plasma nitrided steel surfaces. The micromechanical results show that the surface microhardness increases as the plasma-processing time increases to reach, 1487 HV 0.005 at a plasma processing time of 8 h.

  19. Radiofrequency cold plasma nitrided carbon steel: Microstructural and micromechanical characterizations

    Energy Technology Data Exchange (ETDEWEB)

    Bouanis, F.Z. [Universite Lille Nord de France, F-59000 Lille (France); Unite Materiaux et Transformations (UMET), Ingenierie des Systemes Polymeres, CNRS UMR 8207, ENSCL, BP 90108, F-59652 Villeneuve d' Ascq Cedex (France); Bentiss, F. [Laboratoire de Chimie de Coordination et d' Analytique, Faculte des Sciences, Universite Chouaib Doukkali, B.P. 20, M-24000 El Jadida (Morocco); Bellayer, S.; Vogt, J.B. [Universite Lille Nord de France, F-59000 Lille (France); Unite Materiaux et Transformations (UMET), Ingenierie des Systemes Polymeres, CNRS UMR 8207, ENSCL, BP 90108, F-59652 Villeneuve d' Ascq Cedex (France); Jama, C., E-mail: charafeddine.jama@ensc-lille.fr [Universite Lille Nord de France, F-59000 Lille (France); Unite Materiaux et Transformations (UMET), Ingenierie des Systemes Polymeres, CNRS UMR 8207, ENSCL, BP 90108, F-59652 Villeneuve d' Ascq Cedex (France)

    2011-05-16

    Highlights: {yields} C38 carbon steel samples were plasma nitrided using a radiofrequency (rf) nitrogen plasma discharge. {yields} RF plasma treatment enables nitriding for non-heated substrates. {yields} The morphological and chemical analyses show the formation of a uniform thickness on the surface of the nitrided C38 steel. {yields} Nitrogen plasma active species diffuse into the samples and lead to the formation of Fe{sub x}N. {yields} The increase in microhardness values for nitrided samples with plasma processing time is interpreted by the formation of a thicker nitrided layer on the steel surface. - Abstract: In this work, C38 carbon steel was plasma nitrided using a radiofrequency (rf) nitrogen plasma discharge on non-heated substrates. General characterizations were performed to compare the chemical compositions, the microstructures and hardness of the untreated and plasma treated surfaces. The plasma nitriding was carried out on non-heated substrates at a pressure of 16.8 Pa, using N{sub 2} gas. Surface characterizations before and after N{sub 2} plasma treatment were performed by means of the electron probe microanalysis (EPMA), X-ray photoelectron spectroscopy (XPS) and Vickers microhardness measurements. The morphological and chemical analysis showed the formation of a uniform structure on the surface of the nitrided sample with enrichment in nitrogen when compared to untreated sample. The thickness of the nitride layer formed depends on the treatment time duration and is approximately 14 {mu}m for 10 h of plasma treatment. XPS was employed to obtain chemical-state information of the plasma nitrided steel surfaces. The micromechanical results show that the surface microhardness increases as the plasma-processing time increases to reach, 1487 HV{sub 0.005} at a plasma processing time of 8 h.

  20. Studies of fragileness in steels of vessels of BWR reactors; Estudios de fragilizacion en aceros de vasija de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA MARK lll reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A{sup 2}. (Author)

  1. Flaw evaluation of thermally aged cast stainless steel in light-water reactor applications

    International Nuclear Information System (INIS)

    Lee, S.; Kuo, P.T.; Wichman, K.; Chopra, O.

    1997-01-01

    Cast stainless steel may be used in the fabrication of the primary loop piping, fittings, valve bodies, and pump casings in light-water reactors. However, this material is subject to embrittlement due to thermal aging at the reactor temperature, that is 290 o C (550 o F). The Argonne National Laboratory (ANL) recently completed a research program and the results indicate that the lower-bound fracture toughness of thermally aged cast stainless steel is similar to that of submerged arc welds (SAWs). Thus, the US Nuclear Regulatory Commission (NRC) staff has accepted the use of SAW flaw evaluation procedures in IWB-3640 of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code to evaluate flaws in thermally aged cast stainless steel for a license renewal evaluation. Alternatively, utilities may estimate component-specific fracture toughness of thermally aged cast stainless steel using procedures developed at ANL for a case-by-case flaw evaluation. (Author)

  2. Corrosion behavior of carbon steel in wet Na-bentonite medium

    International Nuclear Information System (INIS)

    Yeon, Jae-Won; Ha, Young-Kyoung; Choi, In-Kyu; Chun, Kwan-Sik

    1996-01-01

    Corrosion behaviors of carbon steel in wet Na-bentonite medium were studied. Corrosion rate of carbon steel in wet bentonite was measured to be 20 μm/yr at 25 deg C using the AC impedance technique. This value is agreed with that obtained by weight loss at 40 deg C for 1 year. The effect of bicarbonate ion on the corrosion of carbon steel in wet bentonite was also evaluated. The carbon steels in wet bentonite having 0.001, 0.01, and 0.1 M concentration of bicarbonate ion gave corrosion rates of 20, 8, and 0.2 μm/yr, respectively. Corrosion potentials of specimens were also measured and compared with the AC impedance results. Both results indicated that bicarbonate ion could effectively reduce the corrosion rate of carbon steels in bentonite due to the formation of protective layer on the carbon steel. (author)

  3. Electrochemical Reactor for Producing Oxygen From Carbon Dioxide, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — An electrochemical reactor is proposed by MicroCell Technologies, LLC to electrochemically reduce carbon dioxide to oxygen. In support of NASA's advanced life...

  4. Reticulated Vitreous Carbon Electrodes for Gas Phase Pulsed Corona Reactors

    National Research Council Canada - National Science Library

    Locke, B

    1998-01-01

    A new design for gas phase pulsed corona reactors incorporating reticulated vitreous carbon electrodes is demonstrated to be effective for the removal of nitrogen oxides from synthetic air mixtures...

  5. Reticulated Vitreous Carbon Electrodes for Gas Phase Pulsed Corona Reactors

    National Research Council Canada - National Science Library

    LOCKE, B

    1999-01-01

    A new design for gas phase pulsed corona reactors incorporating reticulated vitreous carbon electrodes is demonstrated to be effective for the removal of nitrogen oxides from synthetic air mixtures...

  6. Global impact of carbon-14 from nuclear power reactors

    International Nuclear Information System (INIS)

    Moghissi, A.A.; Carter, M.W.

    1977-01-01

    Carbon-14 is produced by nuclear power reactors, predominently as a result of the interaction of a neutron and nitrogen-14 both in the fuel and in the coolant. Several other reactions also contribute to the production of carbon-14. Present operational procedures, in general, for reactors and fuel reprocessing plants result in the release of carbon-14 into the environment. Combustion of fossil fuels and certain industrial operations contribute to the supply of CO 2 in the atmosphere and this contribution is essentially free of carbon-14. Future carbon-14 burdens by assuming a thorough mixing of all CO 2 in the atmosphere is predicted. Available data on electric power generation, fossil fuel combustion and certain other information are used to calculate the projected specific activity of carbon-14 by the year 2000 and the twenty-first century. According to these calculations, the global population dose from carbon-14 can be substantial. Also, carbon-14 in the vicinity of nuclear power reactors is considered. Because of the chemistry of carbon-14, it is shown that local problems may be more significant around BWR's as compared to PWR's. Based on environmental considerations of carbon-14, its increasing production and discharge into the atmosphere, and available control technology, it is recommended that nitrogen use and its presence be minimized in pertinent reactor components and operations

  7. High carbon microalloyed martensitic steel with ultrahigh strength-ductility

    Energy Technology Data Exchange (ETDEWEB)

    Qin, Shengwei; Liu, Yu; Hao, Qingguo [School of Materials Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240 (China); Wang, Ying [School of Mechanical Engineering, Shanghai Dianji University, Shanghai 200245 (China); Chen, Nailu, E-mail: nlchen@sjtu.edu.cn [School of Materials Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240 (China); Zuo, Xunwei; Rong, Yonghua [School of Materials Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240 (China)

    2016-04-29

    Based on the idea of rising the mechanical stability of retained austenite by the addition of Si in Fe-Mn based steels, an Fe-0.63C-1.52Mn-1.49Si-0.62Cr-0.036Nb was designed, then its hot rolled plate was successively tread by normalization process as pretreatment of novel quenching-partitioning-tempering (Q-P-T) process. Product of tensile and elongation (PSE) of 53.94 GPa% were obtained for this high carbon Q-P-T martensitic steel, and the PSE (40.18 GPa%) obtained by the conversion of tensile sample size using Oliver formula still is more excellent PSE than those of other microalloyed advanced high strength steels reported. The microstructural characterization reveals origin of ultrahigh PSE resulting from both the increase of considerable and dispersed carbon enriched retained austenite with relative high mechanical stability in volume fraction and the decrease of brittle twin-type martensite with the sensitivity of notch.

  8. Processing and refinement of steel microstructure images for assisting in computerized heat treatment of plain carbon steel

    Science.gov (United States)

    Gupta, Shubhank; Panda, Aditi; Naskar, Ruchira; Mishra, Dinesh Kumar; Pal, Snehanshu

    2017-11-01

    Steels are alloys of iron and carbon, widely used in construction and other applications. The evolution of steel microstructure through various heat treatment processes is an important factor in controlling properties and performance of steel. Extensive experimentations have been performed to enhance the properties of steel by customizing heat treatment processes. However, experimental analyses are always associated with high resource requirements in terms of cost and time. As an alternative solution, we propose an image processing-based technique for refinement of raw plain carbon steel microstructure images, into a digital form, usable in experiments related to heat treatment processes of steel in diverse applications. The proposed work follows the conventional steps practiced by materials engineers in manual refinement of steel images; and it appropriately utilizes basic image processing techniques (including filtering, segmentation, opening, and clustering) to automate the whole process. The proposed refinement of steel microstructure images is aimed to enable computer-aided simulations of heat treatment of plain carbon steel, in a timely and cost-efficient manner; hence it is beneficial for the materials and metallurgy industry. Our experimental results prove the efficiency and effectiveness of the proposed technique.

  9. Fluidized bed reactor for working up carbon coated particles

    International Nuclear Information System (INIS)

    Marschollek, M.; Simon, W.; Walter, C.

    1981-01-01

    A fluidized bed reactor is described for working up carbon coated particles, particularly nuclear fuel particles or fertile material particles consisting essentially of a cylindrical portion connected to a conical portion. Gas supply pipes, gas distribution space and gas distribution heads are provided within the conical reactor lower portion, the gas distribution members being arranged in at least two superimposed planes and distributed symmetrically over the cross-section of the reactor

  10. The effects of stainless steel radial reflector on core reactivity for small modular reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Jung Kil, E-mail: jkkang@email.kings.ac.kr; Hah, Chang Joo, E-mail: changhah@kings.ac.kr [KINGS, 658-91, Haemaji-ro, Seosaeng-myeon, Ulju-gun, Ulsan, 689-882 (Korea, Republic of); Cho, Sung Ju, E-mail: sungju@knfc.co.kr; Seong, Ki Bong, E-mail: kbseong@knfc.co.kr [KNFC, Daedeok-daero 989beon-gil, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

    2016-01-22

    Commercial PWR core is surrounded by a radial reflector, which consists of a baffle and water. Radial reflector is designed to reflect neutron back into the core region to improve the neutron efficiency of the reactor and to protect the reactor vessels from the embrittling effects caused by irradiation during power operation. Reflector also helps to flatten the neutron flux and power distributions in the reactor core. The conceptual nuclear design for boron-free small modular reactor (SMR) under development in Korea requires to have the cycle length of 4∼5 years, rated power of 180 MWth and enrichment less than 5 w/o. The aim of this paper is to analyze the effects of stainless steel radial reflector on the performance of the SMR using UO{sub 2} fuels. Three types of reflectors such as water, water/stainless steel 304 mixture and stainless steel 304 are selected to investigate the effect on core reactivity. Additionally, the thickness of stainless steel and double layer reflector type are also investigated. CASMO-4/SIMULATE-3 code system is used for this analysis. The results of analysis show that single layer stainless steel reflector is the most efficient reflector.

  11. Cast Steel Filtration Trials Using Ceramic-Carbon Filters

    Directory of Open Access Journals (Sweden)

    Lipowska B.

    2014-12-01

    Full Text Available Trials of cast steel filtration using two types of newly-developed foam filters in which carbon was the phase binding ceramic particles have been conducted. In one of the filters the source of carbon was flake graphite and coal-tar pitch, while in the other one graphite was replaced by a cheaper carbon precursor. The newly-developed filters are fired at 1000°C, i.e. at a much lower temperature than the currently applied ZrO2-based filters. During filtration trials the filters were subjected to the attack of a flowing metal stream having a temperature of 1650°C for 30 seconds.

  12. Corrosion fatigue initiation and short crack growth behaviour of austenitic stainless steels under light water reactor conditions

    International Nuclear Information System (INIS)

    Seifert, H.P.; Ritter, S.; Leber, H.J.

    2012-01-01

    Highlights: ► Corrosion fatigue in austenitic stainless steels under light water reactor conditions. ► Identification of major parameters of influence on initiation and short crack growth. ► Critical system conditions for environmental reduction of fatigue initiation life. ► Comparison with the environmental factor (F env ) approach. - Abstract: The corrosion fatigue initiation and short crack growth behaviour of different wrought low-carbon and stabilised austenitic stainless steels was characterised under simulated boiling water reactor and pressurised water reactor primary water conditions by cyclic fatigue tests with sharply notched fracture mechanics specimens. The special emphasis was placed to the behaviour at low corrosion potentials and, in particular, to hydrogen water chemistry conditions. The major parameter effects and critical conjoint threshold conditions, which result in relevant environmental reduction and acceleration of fatigue initiation life and subsequent short crack growth, respectively, are discussed and summarised. The observed corrosion fatigue behaviour is compared with the fatigue evaluation procedures in codes and regulatory guidelines.

  13. Requirements on cast steel for the primary coolant circuit of water cooled reactors

    International Nuclear Information System (INIS)

    The most important requirements placed on the structural components of water cooled nuclear reactors include corrosion resistance and mechanical materials properties. Intercrystalline corrosion resistance was tested using the Strauss Test in compliance with the DIN 50914 Standard. Following sensitization between 600 to 700 degC with a dwell time between 15 minutes and 100 hours, a specimen homogeneously annealed with the casting and rapidly water cooled showed no intercrystalline corrosion. Specimens cooled from 1050 degC at a rate of 100 degC per hour showed no unambiguous tendency for intercrystalline corrosion after sensitization; in some cases, however, an initial attack of intercrystalline corrosion was found. It was found that austenitic Cr-Ni cast steel containing 2.5% Mo and about 15% ferrite showed the sensitive intercrystalline corrosion range at higher temperatures and longer dwell times than rolled Cr-Ni steels. In plating the ferritic cast steel with a corrosion resistant plating material, annealing temperature after welding must not exceed 600 to 620 degC otherwise the resistance of the plated layer against intercrystalline corrosion would not be safeguarded, and following annealing for stress removal at a temperature of 600 to 620 degC all requirements must be satisfied by the weld metal and weld transition placed on the initial material. Martensite materials are used for the manufacture of components which are not used under pressure, such as alloys with 13% Cr and 1% to 6% Ni and alloys with 17% Cr and 4% Ni. Carbon content is maintained below 0.10% to guarantee good weldability and the highest corrosion resistance. Cast steels with 13% Cr and 4% Ni after a dwell of 2500 hours in fully desalinated water without oxygen and with 3600 ppm of boron at a test temperature of 95 to 300 degC showed a surface reduction of 0.005 mm annually. In identical conditions except for the water containing oxygen the reduction in surface was 0.05 mm per year. (J.B.)

  14. Survey of postirradiation heat treatment as a means to mitigate radiation embrittlement of reactor vessel steels

    International Nuclear Information System (INIS)

    Hawthorne, J.R.

    1979-01-01

    Nuclear-radiation service typically produces a progressive reduction in the notch ductility of low-alloy steels. The reduction is manifested by a decrease in Charpy-V (Csub(v)) upper-shelf energy level and by an elevation in temperature of the ductile-to-brittle transition. Post irradiation heat treatment (annealing) is being investigated as a method for the reversal of these detrimental radiation effects for reactor-vessel steels. This study was undertaken to analyze factors which could affect annealing response, report data available to qualify suspected influences on annealing, and summarize experimental results generated for many commercially produced reactor materials and companion materials produced in the laboratory

  15. Development of ODS (oxide dispersion strengthened) ferritic-martensitic steels for fast reactor fuel cladding

    International Nuclear Information System (INIS)

    Ukai, Shigeharu

    2000-01-01

    In order to attain higher burnup and higher coolant outlet temperature in fast reactor, oxide dispersion strengthened (ODS) ferritic-martensitic steels were developed as a long life fuel cladding. The improvement in formability and ductility, which are indispensable in the cold-rolling method for manufacturing cladding tube, were achieved by controlling the microstructure using techniques such as recrystallization heat-treatment and α to γ phase transformation. The ODS ferritic-martensitic cladding tubes manufactured using these techniques have the highest internal creep rupture strength in the world as ferritic stainless steels. Strength level approaches adequate value at 700degC, which meets the requirement for commercial fast reactors. (author)

  16. Comparison of material property specifications of ferritic steels in fast-breeder reactor technology

    International Nuclear Information System (INIS)

    Delporte, E.; Vanderborck, Y.

    1988-01-01

    The component fabrications for the fast breeder reactors request the use of ferritic steels specially appropriated for the construction of the equipments sustaining pressure and high temperature. The Activity Group nr 3 Materials of the FRCC has decided to make a study to compare the different norms related to the properties of somme ferritic steels used in the different European fast breeder projects. In particular, this study should allow in the different countries of the Community, to identify the designation of a specific steel and to compare its properties. Deviations between the different norms of a same material are mentioned to facilitate European standardization of this type of material

  17. Control of activation levels to simplify waste management of fusion reactor ferritic steel components

    International Nuclear Information System (INIS)

    Wiffen, F.W.; Santoro, R.T.

    1984-01-01

    The objective of this work is to examine the restrictions placed on the composition of steels to allow simplified waste management after service in a fusion reactor first wall. Decay of steel activity within tens of years could simplify waste disposal or even permit recycle. For material recycle, N, Al, Ni, Cu, Nb, and Mo must be excluded. For shallow land burial, initial concentration limits include (in at. ppM) Ni, <20,000; Mo, <3650; N, <3650; Cu, <2400; and Nb, <1.0. Other constituents of steels will not be limited

  18. Carbon and metal-carbon implantations into tool steels for improved tribological performance

    Science.gov (United States)

    Hirvonen, J.-P.; Harskamp, F.; Torri, P.; Willers, H.; Fusari, A.; Gibson, N.; Haupt, J.

    1997-05-01

    The high-fluence implantation of carbon and dual implantations of metal-metalloid pairs into steels with different microstructures are briefly reviewed. A previously unexamined system, the implantation of Si and C into two kinds of tool steels, M3 and D2, have been studied in terms of microstructure and tribological performance. In both cases ion implantation transfers a surface into an amorphous layer. However, the tribological behavior of these two materials differs remarkably: in the case of ion-implanted M3 a reduction of wear in a steel pin is observed even at high pin loads, whereas in the case of ion-implanted D2 the beneficial effects of ion implantation were limited to the lowest pin load. The importance of an initial phase at the onset of sliding is emphasized and a number of peculiarities observed in ion-implanted M3 steel are discussed.

  19. Corrosion of a carbon steel in simulated liquid nuclear wastes

    International Nuclear Information System (INIS)

    Saenz Gonzalez, Eduardo

    2005-01-01

    This work is part of a collaboration agreement between CNEA (National Atomic Energy Commission of Argentina) and USDOE (Department of Energy of the United States of America), entitled 'Tank Corrosion Chemistry Cooperation', to study the corrosion behavior of carbon steel A537 class 1 in different simulated non-radioactive wastes in order to establish the safety concentration limits of the tank waste chemistry at Hanford site (Richland-US). Liquid high level nuclear wastes are stored in tanks made of carbon steel A537 (ASTM nomenclature) that were designed for a service life of 20 to 50 years. A thickness reduction of some tank walls, due to corrosion processes, was detected at Hanford site, beyond the existing predicted values. Two year long-term immersion tests were started using non radioactive simulated liquid nuclear waste solutions at 40 C degrees. This work extends throughout the first year of immersion. The simulated solutions consist basically in combinations of the 10 most corrosion significant chemical components: 5 main components (NaNO 3 , NaCl, NaF, NaNO 2 and NaOH) at three concentration levels and 5 secondary components at two concentration levels. Measurements of the general corrosion rate with time were performed for carbon steel coupons, both immersed in the solutions and in the vapor phases, using weight loss and electrochemistry impedance spectroscopy techniques. Optic and scanning electron microscopy examination, analysis of U-bend samples and corrosion potential measurements, were also done. Localized corrosion susceptibility (pitting and crevice corrosion) was assessed in isolated short-term tests by means of cyclic potentiodynamic polarization curves. The effect of the simulated waste composition on the corrosion behavior of A537 steel was studied based on statistical analyses. The Surface Response Model could be successfully applied to the statistical analysis of the A537 steel corrosion in the studied solutions. General corrosion was not

  20. Reactor design considerations in mineral sequestration of carbon dioxide

    International Nuclear Information System (INIS)

    Ityokumbul, M.T.; Chander, S.; O'Connor, William K.; Dahlin, David C.; Gerdemann, Stephen J.

    2001-01-01

    One of the promising approaches to lowering the anthropogenic carbon dioxide levels in the atmosphere is mineral sequestration. In this approach, the carbon dioxide reacts with alkaline earth containing silicate minerals forming magnesium and/or calcium carbonates. Mineral carbonation is a multiphase reaction process involving gas, liquid and solid phases. The effective design and scale-up of the slurry reactor for mineral carbonation will require careful delineation of the rate determining step and how it changes with the scale of the reactor. The shrinking core model was used to describe the mineral carbonation reaction. Analysis of laboratory data indicates that the transformations of olivine and serpentine are controlled by chemical reaction and diffusion through an ash layer respectively. Rate parameters for olivine and serpentine carbonation are estimated from the laboratory data

  1. INFLUENCE OF ELECTRIC SPARK ON HARDNESS OF CARBON STEEL

    Directory of Open Access Journals (Sweden)

    I. O. Vakulenko

    2014-03-01

    Full Text Available Purpose. The purpose of work is an estimation of influence of an electric spark treatment on the state of mouldable superficial coverage of carbon steel. Methodology. The steel of fragment of railway wheel rim served as material for research with chemical composition 0.65% С, 0.67% Mn, 0.3% Si, 0.027% P, 0.028% S. Structural researches were conducted with the use of light microscopy and methods of quantitative metallography. The structural state of the probed steel corresponded to the state after hot plastic deformation. The analysis of hardness distribution in the micro volumes of cathode metal was carried out with the use of microhardness tester of type of PMT-3. An electric spark treatment of carbon steel surface was executed with the use of equipment type of EFI-25M. Findings. After electric spark treatment of specimen surface from carbon steel the forming of multi-layered coverage was observed. The analysis of microstructure found out the existence of high-quality distinctions in the internal structure of coverage metal, depending on the probed area. The results obtained in the process are confirmed by the well-known theses, that forming of superficial coverage according to technology of electric spark is determined by the terms of transfer and crystallization of metal. The gradient of structures on the coverage thickness largely depends on development of structural transformation processes similar to the thermal character influence. Originality. As a result of electric spark treatment on the condition of identical metal of anode and cathode, the first formed layer of coverage corresponds to the monophase state according to external signs. In the volume of coverage metal, the appearance of carbide phase particles is accompanied by the decrease of microhardness values. Practical value. Forming of multi-layered superficial coverage during electric spark treatment is accompanied by the origin of structure gradient on a thickness. The effect

  2. An assessment of carbon steel overpacks for radioactive waste disposal

    International Nuclear Information System (INIS)

    Marsh, G.P.; Bland, I.D.; Taylor, K.J.; Sharland, S.; Tasker, P.

    1986-01-01

    The report summarizes the results obtained at Harwell in the second phase of a project evaluating the corrosion behaviour of high-level waste overpacks in geological disposal. It has concentrated on the use of carbon steel in granitic and argillaceous environments, and has aimed at estimating the corrosion allowance required to achieve a 1000-year overpack life. Experimental and mathematical modelling studies have indicated that 200 mm of steel should be more than sufficient to prevent overpack penetration by general or localized corrosion. A theoretical assessment of the possible effects of micro-organisms on overpack corrosion has concluded that such species are likely to be found in repositories, but that only a fraction of their population should be corrosive towards carbon steel. Making the pessimistic assumption that all organic carbon in a 500 mm bentonite backfill is utilized by corrosive sulphate reducing bacteria, it has been estimated that this will result in an additional metal loss of 13 mm. One form of corrosion which cannot be dealt with by the corrosion allowance approach is stress corrosion cracking, since even at the lowest reported propagation rates, a metal thickness exceeding 3 m would be penetrated in 1000 years. It has been concluded that the possibility of stress corrosion cannot be dismissed, but, because the process requires a certain minimum stress level before it can occur, it should be possible to avoid the problem by giving the overpacks a stress relief heat treatment. Tests in a model environment have shown that a heat treatment designed to reduce fabrication stresses to 50% of the yield strengh should be sufficient to prevent cracking. It is recommended that this conclusion be substantiated by scaled-up experiments with model overpacks. The report draws further attention to degradation by hydrogen embrittlement

  3. The oxidation of titanium nitride- and silicon nitride-coated stainless steel in carbon dioxide environments

    International Nuclear Information System (INIS)

    Mitchell, D.R.G.; Stott, F.H.

    1992-01-01

    A study has been undertaken into the effects of thin titanium nitride and silicon nitride coatings, deposited by physical vapour deposition and chemical vapour deposition processes, on the oxidation resistance of 321 stainless steel in a simulated advanced gas-cooled reactor carbon dioxide environment for long periods at 550 o C and 700 o C under thermal-cycling conditions. The uncoated steel contains sufficient chromium to develop a slow-growing chromium-rich oxide layer at these temperatures, particularly if the surfaces have been machine-abraded. Failure of this layer in service allows formation of less protective iron oxide-rich scales. The presence of a thin (3-4 μm) titanium nitride coating is not very effective in increasing the oxidation resistance since the ensuing titanium oxide scale is not a good barrier to diffusion. Even at 550 o C, iron oxide-rich nodules are able to develop following relatively rapid oxidation and breakdown of the coating. At 700 o C, the coated specimens oxidize at relatively similar rates to the uncoated steel. A thin silicon nitride coating gives improved oxidation resistance, with both the coating and its slow-growing oxide being relatively electrically insulating. The particular silicon nitride coating studied here was susceptible to spallation on thermal cycling, due to an inherently weak coating/substrate interface. (Author)

  4. Friction stir processing on high carbon steel U12

    Energy Technology Data Exchange (ETDEWEB)

    Tarasov, S. Yu., E-mail: tsy@ispms.ru; Rubtsov, V. E., E-mail: rvy@ispms.ru [Institute of Strength Physics and Materials Science SB RAS, Tomsk, 634055 (Russian Federation); National Research Tomsk Polytechnic University, Tomsk, 634050 (Russian Federation); Melnikov, A. G., E-mail: melnikov-ag@tpu.ru [National Research Tomsk Polytechnic University, Tomsk, 634050 (Russian Federation)

    2015-10-27

    Friction stir processing (FSP) of high carbon steel (U12) samples has been carried out using a milling machine and tools made of cemented tungsten carbide. The FSP tool has been made in the shape of 5×5×1.5 mm. The microstructural characterization of obtained stir zone and heat affected zone has been carried out. Microhardness at the level of 700 MPa has been obtained in the stir zone with microstructure consisting of large grains and cementitte network. This high-level of microhardness is explained by bainitic reaction developing from decarburization of austenitic grains during cementite network formation.

  5. On the carbide formation in high-carbon stainless steel

    International Nuclear Information System (INIS)

    Mujahid, M.; Qureshi, M.I.

    1996-01-01

    Stainless steels containing high Cr as well as carbon contents in excess of 1.5 weight percent have been developed for applications which require high resistance erosion and environmental corrosion. Formation of carbides is one of important parameters for controlling properties of these materials especially erosion characteristics. Percent work includes the study of different type of carbides which from during the heat treatment of these materials. It has been found that precipitation of secondary carbides and the nature of matrix transformation plays an important role in determining the hardness characteristics of these materials. (author)

  6. Plasticity of low carbon steel in a hot state

    Energy Technology Data Exchange (ETDEWEB)

    Konovalov, V P; Rizol' , A I; Shram, N N [Ural' skij Nauchno-Issledovatel' skij Inst. Chernykh Metallov, Sverdlovsk (USSR)

    1977-07-01

    The hot ductility (in tapered-specimen piersing test and the in wedge-shaped specimen rolling test) is studied of the Armeo-type low carbon steel produced by vacuum induction and open hearth techniques. The variations of the chemical composition within specified ranges, particularly as regards sulphur, oxygen and the Mn/S ratio, have a marked effect on the processing ductility. The temperature range of brittle fracture and acceptable hot working reductions as functions of the chemical composition have been revealed.

  7. Plasticity of low carbon steel in a hot state

    International Nuclear Information System (INIS)

    Konovalov, V.P.; Rizol', A.I.; Shram, N.N.

    1977-01-01

    The hot ductility (in tapered-specimen piersing test and the in wedge-shaped specimen rolling test) is studied of the Armeo-type low carbon steel produced by vacuum induction and open hearth techniques. The variations of the chemical composition within specified ranges, particularly as regards sulphur, oxygen and the Mn/S ratio, have a marked effect on the processing ductility. The temperature range of brittle fracture and acceptable hot working reductions as functions of the chemical composition have been revealed

  8. Corrosion Inhibitor of Carbon Steel from Onion Peel Extract

    Directory of Open Access Journals (Sweden)

    Muhammad Samsudin Asep

    2018-01-01

    Full Text Available Carbon steels composed by two main elements, they are iron (Fe and carbon (C elements which widely used in industrial because of its resistance and more affordable than stainless steel, but their weakness is they have low corrosion resistance. One way to modify carbon steel is by coating them with antioxidant compounds that can delay, slow down, and prevent lipid oxidation process, which obtained from onion peel extract. Several studies on corrosion inhibitors have been performed. However, the efficiency was not reach the optimum. This study aims to examine the effect of onion peel extract concentration on the efficiency of corrosion inhibitor and characterization of the green corrosion inhibitor from onion peel extract. This research method begins by extracting onion peel to 200 ml solvent which we use aquadest and methanol and mixed with 5 grams of crushed onion peel, then let them be extracted for 60 minutes with room temperature. Once it was filtered and the solution obtained, followed by evaporating process with rotary evaporator to decrease the content of solvent. The product is ready to be used as a green corrosion inhibitor of carbon steel in 1 mol/L HCl. While the analysis used is HPLC qualitative analysis, and electroplatting process. The impedance is measured at a frequency of 100 kHz to 4 mHz with an AC current of 10mV. Inhibitor concentrations are vary between 2 ml and 4 ml of onion peel extract. Electroplatting is done within 30 minutes with 10 minutes each checking time. Furthermore, quantitative analysis was done for the analysis of corrosion rate and weight loss. Based on HPLC analysis, it is known that the extract of onion peel contains 1mg/L of quercetin, which is belong to flavonoid group as green inhibitor. While electroplatting process, aquadest solvent having average efficiency of 99,57% for 2 ml of extract, and 99,60% for 4 ml of extract. Methanol solvent having average efficiency of 99,52% for 2 ml of extract and 99

  9. Characterization of four prestressed concrete reactor vessel liner steels

    International Nuclear Information System (INIS)

    Nanstad, R.K.

    1980-12-01

    A program of fracture toughness testing and analysis is being performed with PCRV steels for HTGRs. This report focuses on background information for the base materials and results of characterization testing, such as tensile and impact properties, chemical composition, and microstructural examination. The steels tested were an SA-508 class 1 forging, two plates of SA-537 class 1, and one plate of SA-537 class 2. Tensile requirements in effect at the time of procurement are met by all four steels. However, the SA-537 class 2 plate would not meet the minimum requirement for yield strength. Drop-weight and Charpy impact tests verified that the RT/sub NDT/ is equal to the NDT for each steel. Charpy impact energies at the NDT range from 40 J (30 ft-lb) for one heat of SA-537 class 1 to 100 J (74 ft-lb) for the SA-537 class 2 plate; upper-shelf energies range from 170 to 310 J (125 to 228 ft-lb) for the same two steels, respectively. The onset of upper-shelf energy occurred at temperatures ranging from 0 to 50 0 C

  10. Study on Spheroidization and Related Heat Treatments of Medium Carbon Alloy Steels

    Directory of Open Access Journals (Sweden)

    Harisha S. R.

    2018-01-01

    Full Text Available The importance of medium carbon steels as engineering materials is reflected by the fact that out of the vast majority of engineering grade ferrous alloys available and used in the market today, a large proportion of them are from the family of medium carbon steels. Typically medium carbon steels have a carbon range of 0.25 to 0.65% by weight, and a manganese content ranging from 0.060 to 1.65% by weight. Medium carbon steels are more resistive to cutting, welding and forming as compared to low carbon steels. From the last two decades a number of research scholars reported the use of verity of heat treatments to tailor the properties of medium carbon steels. Spheroidizing is the novel industrial heat treatment employed to improve formability and machinability of medium/high carbon low alloy steels. This exclusive study covers procedure, the effects and possible outcomes of various heat treatments on medium carbon steels. In the present work, other related heat treatments like annealing and special treatments for property alterations which serve as pretreatments for spheroidizing are also reviewed. Medium carbon steels with property alterations by various heat treatment processes are finding increased responsiveness in transportation, aerospace, space, underwater along with other variegated fields. Improved tribological and mechanical properties consisting of impact resistance, stiffness, abrasion and strength are the main reasons for the increased attention of these steels in various industries. In the present scenario for the consolidation of important aspects of various heat treatments and effects on mechanical properties of medium carbons steel, a review of different research papers has been attempted. This review may be used as a guide to provide practical data for heat treatment industry, especially as a tool to enhance workability and tool life.

  11. Kinetics of electrochemical boriding of low carbon steel

    International Nuclear Information System (INIS)

    Kartal, G.; Eryilmaz, O.L.; Krumdick, G.; Erdemir, A.; Timur, S.

    2011-01-01

    In this study, the growth kinetics of the boride layers forming on low carbon steel substrates was investigated during electrochemical boriding which was performed at a constant current density of 200 mA/cm 2 in a borax based electrolyte at temperatures ranging from 1123 K to 1273 K for periods of 5-120 min. After boriding, the presence of both FeB and Fe 2 B phases were confirmed by the X-ray diffraction method. Cross-sectional microscopy revealed a very dense and thick morphology for both boride phases. Micro hardness testing of the borided steel samples showed a significant increase in the hardness of the borided surfaces (i.e., up to (1700 ± 200) HV), while the hardness of un-borided steel samples was approximately (200 ± 20) HV. Systematic studies over a wide range of boriding time and temperature confirmed that the rate of the boride layer formation is strongly dependent on boriding duration and has a parabolic character. The activation energy of boride layer growth for electrochemical boriding was determined as (172.75 ± 8.6) kJ/mol.

  12. Surface martensitization of Carbon steel using Arc Plasma Sintering

    Science.gov (United States)

    Wahyudi, Haris; Dimyati, Arbi; Sebayang, Darwin

    2018-03-01

    In this paper new technology of surface structure modification of steel by short plasma exposure in Arc Plasma Sintering (APS) device is presented. APS is an apparatus working based on plasma generated by DC pulsed current originally used for synthesizing materials via sintering and melting. Plasma exposure in APS was applied into the specimens for 1 and 3 seconds which generate temperature approximately about 1300-1500°C. The SUP9, pearlitic carbon steel samples were used. The hardness, hardening depth and microstructure of the specimens have been investigated by Vickers micro hardness test and Scanning Electron Microscopy (SEM) supported by Energy Dispersive X-Ray Spectroscopy (EDX). The results have showed that the mechanical property was significantly improved due to the formation of single martensitic structures as identified by SEM. The hardness of treated surface evaluated by Vickers hardness test showed significant improvement nearly three time from 190 VHN before to 524 VHN after treatment. Furthermore, EDX confirmed that the formation of martensite layer occurred without altering its composition. The APS also produced uniform hardened layer up to 250 μm. The experiment has demonstrated that arc plasma process was successfully improved the mechanical properties of steel in relatively very short time.

  13. Ductile austenitic steel for fuel cans and core components of sodium cooled reactors

    International Nuclear Information System (INIS)

    Schaefer, L.

    1995-01-01

    Two austenitic steel melts of a new composition have been studied after irradiation in the PFR fast neutron flux, in the BR2 reactor, and in the Harwell V.E. Cyclotron. The investigations were focussed on helium embrittlement and irradiation induced swelling. (orig.)

  14. Isothermal and thermal–mechanical fatigue of VVER-440 reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Fekete, Balazs, E-mail: fekete.mm.bme@gmail.com [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary); Department of Applied Mechanics, Budapest University of Technology and Economics, Muegyetem 5, Budapest H-1111 (Hungary); Trampus, Peter [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary)

    2015-09-15

    Highlights: • We aimed to determine the thermomechanical behaviour of VVER reactor steels. • Material tests were developed and performed on GLEEBLE 3800 physical simulator. • Coffin–Manson curves and parameters were derived. • High accuracy of the strain energy based evaluation was found. • The observed dislocation evolution correlates with the mechanical behaviour. - Abstract: The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of VVER-440 reactor pressure vessel under completely reserved total strain controlled low cycle fatigue tests were investigated. An advanced test facility was developed for GLEEBLE-3800 physical simulator which was able to perform thermomechanical fatigue experiments under in-service conditions of VVER nuclear reactors. The low cycle fatigue results were evaluated with the plastic strain based Coffin–Manson law, and plastic strain energy based model as well. It was shown that both methods are able to predict the fatigue life of reactor pressure vessel steels accurately. Interrupted fatigue tests were also carried out to investigate the kinetic of the fatigue evolution of the materials. On these samples microstructural evaluation by TEM was performed. The investigated low cycle fatigue behavior can provide reference for remaining life assessment and lifetime extension analysis.

  15. Positron annihilation and Moessbauer studies of neutron irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Brauer, G.; Matz, W.; Liszkay, L.; Molnar, B.

    1990-11-01

    Positron annihilation (lifetime, Doppler broadening) and Moessbauer studies on unirradiated, neutron irradiated and neutron irradiated plus annealed reactor pressure vessel steels (Soviet type 15Kh2NMFA) are presented. The role of microstructural properties and the formation of irradiation-induced precipitates is discussed. (orig.) [de

  16. Correlation between radiation damage and magnetic properties in reactor vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Kempf, R.A., E-mail: kempf@cnea.gov.ar [División Caracterización, GCCN, CAC-CNEA (Argentina); Sacanell, J. [Departamento Física de la Materia Condensada, GIyA, CAC-CNEA, CONICET (Argentina); Milano, J. [División Resonancias Magnéticas, CAB-CNEA, CONICET (Argentina); Guerra Méndez, N. [Departamento Física de la Materia Condensada, GIyA, CAC-CNEA, CONICET (Argentina); Winkler, E.; Butera, A. [División Resonancias Magnéticas, CAB-CNEA, CONICET (Argentina); Troiani, H. [División Física de Metales, CAB-CNEA and Instituto Balseiro (UNCU), CONICET (Argentina); Saleta, M.E. [División Resonancias Magnéticas, CAB-CNEA, CONICET (Argentina); Fortis, A.M. [Departamento Estructura y Comportamiento. Gerencia Materiales-GAEN, CAC-CNEA (Argentina)

    2014-02-01

    Since reactor pressure vessel steels are ferromagnetic, provide a convenient means to monitor changes in the mechanical properties of the material upon irradiation with high energy particles, by measuring their magnetic properties. Here, we discuss the correlation between mechanical and magnetic properties and microstructure, by studying the flux effect on the nuclear pressure vessel steel used in reactors currently under construction in Argentina. Charpy-V notched specimens of this steel were irradiated in the RA1 experimental reactor at 275 °C with two lead factors (LFs), 93 and 183. The magnetic properties were studied by means of DC magnetometry and ferromagnetic resonance. The results show that the coercive field and magnetic anisotropy spatial distribution are sensitive to the LF and can be explained by taking into account the evolution of the microstructure with this parameter. The saturation magnetization shows a dominant dependence on the accumulated damage. Consequently, the mentioned techniques are suitable to estimate the degradation of the reactor vessel steel.

  17. Study on microstructure and mechanical characteristics of low-carbon steel and ferritic stainless steel joints

    Energy Technology Data Exchange (ETDEWEB)

    Sarkari Khorrami, Mahmoud; Mostafaei, Mohammad Ali; Pouraliakbar, Hesam, E-mail: hpouraliakbar@alum.sharif.edu; Kokabi, Amir Hossein

    2014-07-01

    In this work, examinations on the microstructure and mechanical properties of plain carbon steel and AISI 430 ferritic stainless steel dissimilar welds are carried out. Welding is conducted in both autogenous and using ER309L austenitic filler rod conditions through gas tungsten arc welding process. The results indicate that fully-ferritic and duplex ferritic–martensitic microstructures are formed for autogenous and filler-added welds, respectively. Carbide precipitation and formation of martensite at ferrite grain boundaries (intergranular martensite) as well as grain growth occur in the heat affected zone (HAZ) of AISI 430 steel. It is found that weld heat input can strongly affect grain growth phenomenon along with the amount and the composition of carbides and intergranular martensite. Acquired mechanical characteristics of weld in the case of using filler metal are significantly higher than those of autogenous one. Accordingly, ultimate tensile strength (UTS), hardness, and absorbed energy during tensile test of weld metal are increased from 662 MPa to 910 MPa, 140 Hv to 385 Hv, and 53.6 J m{sup −3} to 79 J m{sup −3}, respectively by filler metal addition. From fracture surfaces, predominantly ductile fracture is observed in the specimen welded with filler metal while mainly cleavage fracture occurs in the autogenous weld metal.

  18. General corrosion of carbon steels in high temperature water

    International Nuclear Information System (INIS)

    Gras, J.M.

    1994-04-01

    This short paper seeks to provide a summary of the main knowledge about the general corrosion of carbon steels in high temperature water. In pure water or slightly alkaline deaerated water, steels develop a protective coating of magnetite in a double layer (Potter and Mann oxide) or a single layer (Bloom oxide). The morphology of the oxide layer and the kinetics of corrosion depend on the test parameters controlling the solubility of iron. The parameters exercising the greatest influence are partial hydrogen pressure and mass transfer: hydrogen favours the solubilization of the magnetite; the entrainment of the dissolved iron prevents a redeposition of magnetite on the surface of the steel. Cubic or parabolic in static conditions, the kinetics of corrosion tends to be linear in dynamic conditions. In dynamic operation, corrosion is at least one order of magnitude lower in water with a pH of 10 than in pure water with a pH of 7. The activation energy of corrosion is 130 kJ/mol (31 kcal/mol). This results in the doubling of corrosion at around 300 deg C for a temperature increase of 15 deg C. Present in small quantities (100-200 ppb), oxygen decreases general corrosion but increases the risk of pitting corrosion - even for a low chloride content - and stress corrosion cracking or corrosion-fatigue. The steel composition has probably an influence on the kinetics of corrosion in dynamic conditions; further work would be required to clarify the effect of some residual elements. (author). 31 refs., 9 figs., 2 tabs

  19. Distribution of radionuclides during melting of carbon steel

    Energy Technology Data Exchange (ETDEWEB)

    Thurber, W.C.; MacKinney, J.

    1997-02-01

    During the melting of steel with radioactive contamination, radionuclides may be distributed among the metal product, the home scrap, the slag, the furnace lining and the off-gas collection system. In addition, some radionuclides will pass through the furnace system and vent to the atmosphere. To estimate radiological impacts of recycling radioactive scrap steel, it is essential to understand how radionuclides are distributed within the furnace system. For example, an isotope of a gaseous element (e.g., radon) will exhaust directly from the furnace system into the atmosphere while a relatively non-volatile element (e.g., manganese) can be distributed among all the other possible media. This distribution of radioactive contaminants is a complex process that can be influenced by numerous chemical and physical factors, including composition of the steel bath, chemistry of the slag, vapor pressure of the particular element of interest, solubility of the element in molten iron, density of the oxide(s), steel melting temperature and melting practice (e.g., furnace type and size, melting time, method of carbon adjustment and method of alloy additions). This paper discusses the distribution of various elements with particular reference to electric arc furnace steelmaking. The first two sections consider the calculation of partition ratios for elements between metal and slag based on thermodynamic considerations. The third section presents laboratory and production measurements of the distribution of various elements among slag, metal, and the off-gas collection system; and the final section provides recommendations for the assumed distribution of each element of interest.

  20. Carbon steel protection in G.S. (Girlder sulfide) plants. CITROSOLV process influence. Pt. 6

    International Nuclear Information System (INIS)

    Lires, O.A.; Burkart, A.L.; Delfino, C.A.; Rojo, E.A.

    1988-01-01

    In order to protect carbon steel towers and piping of Girlder sulfide (G.S.) experimental heavy water plants against corrosion produced by the action of aqueous solutions of hydrogen sulfides, a method, previously published, was developed. Carbon steel, exposed to saturated aqueous solutions of hydrogen sulfide, forms iron sulfide scales. In oxygen free solutions evolution of corrosion follows the sequence: mackinawite → cubic ferrous sulfide → troilite → pyrrotite → pyrite. Scales formed by pyrrotite-pyrite or pyrite are the most protective layers (these are obtained at 130 deg C, 2 MPa, for periods of 14 days). CITROSOLV Process (Pfizer) is used to descaling and passivating stainless steel plant's components. This process must be used in mixed (carbon steel - stainless steel) circuits and may cause the formation of magnetite scales over the carbon steel. The influence of magnetite in the pyrrotite-pyrite scales formation is studied in this work. (Author) [es

  1. Effect of Cr and Mo on strain ageing behaviour of low carbon steel

    International Nuclear Information System (INIS)

    Pereloma, E.V.; Bata, V.; Scott, R.I.; Smith, R.M.

    2010-01-01

    This work explores the effects of Cr (0.26-0.74 wt%) and Mo (0.09-0.3 wt%) additions on the kinetics of strain ageing process in low carbon steel. The strain ageing behaviour of the steels was investigated by using tensile tests and transmission electron microscopy. The results have shown that Mo-alloyed steels undergo the same four stages of ageing as unalloyed low carbon steel, whereas Cr-alloyed steels exhibit only three stages of ageing. At the same time, the addition of Mo accelerates the ageing response, while alloying with Cr reduces the rate of strain ageing by ∼3 times in comparison with non-alloyed low carbon steel. It especially delays the offset of Stage III. This is explained by the reduction of carbon content in ferrite due to the enrichment of cementite with Cr leading to the reduction of its equilibrium solubility in ferrite.

  2. Effect of Cr and Mo on strain ageing behaviour of low carbon steel

    Energy Technology Data Exchange (ETDEWEB)

    Pereloma, E.V., E-mail: elenap@uow.edu.au [School of Mechanical, Materials and Mechatronic Engineering, University of Wollongong, Northfields Avenue, Wollongong, NSW 2522 (Australia); Bata, V. [Department of Materials Engineering, Monash University (Australia); Scott, R.I.; Smith, R.M. [BlueScope Steel Limited, Port Kembla (Australia)

    2010-04-25

    This work explores the effects of Cr (0.26-0.74 wt%) and Mo (0.09-0.3 wt%) additions on the kinetics of strain ageing process in low carbon steel. The strain ageing behaviour of the steels was investigated by using tensile tests and transmission electron microscopy. The results have shown that Mo-alloyed steels undergo the same four stages of ageing as unalloyed low carbon steel, whereas Cr-alloyed steels exhibit only three stages of ageing. At the same time, the addition of Mo accelerates the ageing response, while alloying with Cr reduces the rate of strain ageing by {approx}3 times in comparison with non-alloyed low carbon steel. It especially delays the offset of Stage III. This is explained by the reduction of carbon content in ferrite due to the enrichment of cementite with Cr leading to the reduction of its equilibrium solubility in ferrite.

  3. Efficiency of inhibitor for biocorrosion influenced by consortium sulfate reducing bacteria on carbon steel

    Energy Technology Data Exchange (ETDEWEB)

    Mahat, Nur Akma; Othman, Norinsan Kamil [School of Applied Physics, Faculty of Science and Technology, Universiti Kebangsaan Malaysia, 43600 Bangi, Selangor (Malaysia); Sahrani, Fathul Karim [School of Environment and Natural Resources Science, Faculty of Science and Technology, Universiti Kebangsaan Malaysia, 43600 Bangi, Selangor (Malaysia)

    2015-09-25

    The inhibition efficiency of benzalkonium chloride (BKC) in controlling biocorrosion on the carbon steel surfaces has been investigated. The carbon steel coupons were incubated in the presence of consortium SRB (C-SRB) with and without BKC for the difference medium concentration. The corrosion rate and inhibition efficiency have been evaluated by a weight loss method. The morphology of biofilm C-SRB on the steel surfaces were characterized with variable pressure scanning electron microscopy (VPSEM). The results revealed that BKC exhibits a low corrosion rate, minimizing the cell growth and biofilm development on the carbon steel surfaces.

  4. Efficiency of inhibitor for biocorrosion influenced by consortium sulfate reducing bacteria on carbon steel

    International Nuclear Information System (INIS)

    Mahat, Nur Akma; Othman, Norinsan Kamil; Sahrani, Fathul Karim

    2015-01-01

    The inhibition efficiency of benzalkonium chloride (BKC) in controlling biocorrosion on the carbon steel surfaces has been investigated. The carbon steel coupons were incubated in the presence of consortium SRB (C-SRB) with and without BKC for the difference medium concentration. The corrosion rate and inhibition efficiency have been evaluated by a weight loss method. The morphology of biofilm C-SRB on the steel surfaces were characterized with variable pressure scanning electron microscopy (VPSEM). The results revealed that BKC exhibits a low corrosion rate, minimizing the cell growth and biofilm development on the carbon steel surfaces

  5. Efficiency of inhibitor for biocorrosion influenced by consortium sulfate reducing bacteria on carbon steel

    Science.gov (United States)

    Mahat, Nur Akma; Othman, Norinsan Kamil; Sahrani, Fathul Karim

    2015-09-01

    The inhibition efficiency of benzalkonium chloride (BKC) in controlling biocorrosion on the carbon steel surfaces has been investigated. The carbon steel coupons were incubated in the presence of consortium SRB (C-SRB) with and without BKC for the difference medium concentration. The corrosion rate and inhibition efficiency have been evaluated by a weight loss method. The morphology of biofilm C-SRB on the steel surfaces were characterized with variable pressure scanning electron microscopy (VPSEM). The results revealed that BKC exhibits a low corrosion rate, minimizing the cell growth and biofilm development on the carbon steel surfaces.

  6. Analysis of corrosion products of carbon steel in wet bentonite

    International Nuclear Information System (INIS)

    Osada, K.; Nagano, T.; Kozai, N.; Nakashima, S.; Nakayama, S.; Muraoka, S.

    1991-01-01

    The following conclusions were obtained; (1) At 40degC, the average corrosion rate of SS41 carbon steel in wet bentonite was 0.025 mm/y. This is smaller than the value of 0.042 mm/y obtained in pure water at 40degC. However, at 95degC, the corrosion rate of SS41 carbon steel in wet bentonite was 0.27 mm/y, which is much larger than that in pure water at 95degC. (2) At 95degC, γ-FeO(OH) (lepidocrocite) was formed only in wet bentonite, and it was absent in pure water. Evaporation of moisture resulted in the formation of partial covering of bentonite, which promoted local corrosion. Consequently, γ-FeO(OH) was considered to be formed. (3) In wet bentonite at 95degC, α-Fe 2 O 3 (hematite) can be identified by means of colorimetry. The color of corrosion products is orangish, indicating the contribution of α-Fe 2 O 3 in iron hydroxides. (author)

  7. Microstructures and mechanical properties of duplex low carbon steel

    Science.gov (United States)

    Alfirano; Eben, U. S.; Hidayat, M.

    2018-04-01

    The microstructures behavior of duplex cold-rolled low carbon steel for automotive applications has been investigated. Intercritical annealing treatment is commonly used to develop a duplex low carbon steel containing ferrite and martensite. To get a duplex phase ferrite and martensite, the specimens were heated at inter-critical annealing temperature of 775°C - 825°C, for heating time up to 20 minutes, followed by water-quenched. The hardness of specimens was studied. The optical microscopy was used to analyze the microstructures. The optimal annealing conditions (martensite volume fraction approaching 20%) at 775°C with a heating time of 10 minutes was achieved. The highest hardness value was obtained in cold-rolled specimens of 41% in size reduction for intercritical annealing temperature of 825°C. In this condition, the hardness value was 373 HVN. The correlation between intercritical annealing temperature and time can be expressed in the transformation kinetics as fγ/fe = 1-exp(-Ktn) wherein K and n are grain growth rate constant and Avrami’s exponent, respectively. From experiment, the value of K = 0.15 and n = 0.461. Using the relationship between temperatures and heating time, activation energy (Q) can be calculated that is 267 kJ/mol.

  8. Marine Atmospheric Corrosion of Carbon Steel: A Review.

    Science.gov (United States)

    Alcántara, Jenifer; Fuente, Daniel de la; Chico, Belén; Simancas, Joaquín; Díaz, Iván; Morcillo, Manuel

    2017-04-13

    The atmospheric corrosion of carbon steel is an extensive topic that has been studied over the years by many researchers. However, until relatively recently, surprisingly little attention has been paid to the action of marine chlorides. Corrosion in coastal regions is a particularly relevant issue due the latter's great importance to human society. About half of the world's population lives in coastal regions and the industrialisation of developing countries tends to concentrate production plants close to the sea. Until the start of the 21st century, research on the basic mechanisms of rust formation in Cl - -rich atmospheres was limited to just a small number of studies. However, in recent years, scientific understanding of marine atmospheric corrosion has advanced greatly, and in the authors' opinion a sufficient body of knowledge has been built up in published scientific papers to warrant an up-to-date review of the current state-of-the-art and to assess what issues still need to be addressed. That is the purpose of the present review. After a preliminary section devoted to basic concepts on atmospheric corrosion, the marine atmosphere, and experimentation on marine atmospheric corrosion, the paper addresses key aspects such as the most significant corrosion products, the characteristics of the rust layers formed, and the mechanisms of steel corrosion in marine atmospheres. Special attention is then paid to important matters such as coastal-industrial atmospheres and long-term behaviour of carbon steel exposed to marine atmospheres. The work ends with a section dedicated to issues pending, noting a series of questions in relation with which greater research efforts would seem to be necessary.

  9. Marine Atmospheric Corrosion of Carbon Steel: A Review

    Science.gov (United States)

    Alcántara, Jenifer; de la Fuente, Daniel; Chico, Belén; Simancas, Joaquín; Díaz, Iván; Morcillo, Manuel

    2017-01-01

    The atmospheric corrosion of carbon steel is an extensive topic that has been studied over the years by many researchers. However, until relatively recently, surprisingly little attention has been paid to the action of marine chlorides. Corrosion in coastal regions is a particularly relevant issue due the latter’s great importance to human society. About half of the world’s population lives in coastal regions and the industrialisation of developing countries tends to concentrate production plants close to the sea. Until the start of the 21st century, research on the basic mechanisms of rust formation in Cl−-rich atmospheres was limited to just a small number of studies. However, in recent years, scientific understanding of marine atmospheric corrosion has advanced greatly, and in the authors’ opinion a sufficient body of knowledge has been built up in published scientific papers to warrant an up-to-date review of the current state-of-the-art and to assess what issues still need to be addressed. That is the purpose of the present review. After a preliminary section devoted to basic concepts on atmospheric corrosion, the marine atmosphere, and experimentation on marine atmospheric corrosion, the paper addresses key aspects such as the most significant corrosion products, the characteristics of the rust layers formed, and the mechanisms of steel corrosion in marine atmospheres. Special attention is then paid to important matters such as coastal-industrial atmospheres and long-term behaviour of carbon steel exposed to marine atmospheres. The work ends with a section dedicated to issues pending, noting a series of questions in relation with which greater research efforts would seem to be necessary. PMID:28772766

  10. Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL

    2010-08-01

    Materials behaviors caused by neutron irradiation under fission and/or fusion environments can be little understood without practical examination. Easily accessible material information system with large material database using effective computers is necessary for design of nuclear materials and analyses or simulations of the phenomena. The developed Embrittlement Data Base (EDB) at ORNL is this comprehensive collection of data. EDB database contains power reactor pressure vessel surveillance data, the material test reactor data, foreign reactor data (through bilateral agreements authorized by NRC), and the fracture toughness data. The lessons learned from building EDB program and the associated database management activity regarding Material Database Design Methodology, Architecture and the Embedded QA Protocol are described in this report. The development of IAEA International Database on Reactor Pressure Vessel Materials (IDRPVM) and the comparison of EDB database and IAEA IDRPVM database are provided in the report. The recommended database QA protocol and database infrastructure are also stated in the report.

  11. Evolution of carbon steel corrosion in feedwater conditions reproduce by the Fortrand loop

    International Nuclear Information System (INIS)

    Delaunay, Sophie; Bescond, Aurelien; Mansour, Carine; Bretelle, Jean-Luc

    2012-09-01

    Fouling and tubes support plate blockage of steam generators (SG) are major problems in the secondary circuit of pressurized water reactor (PWR) plants. Corrosion products (CP) responsible of these phenomena are mainly constituted of magnetite. Limit the amount of these CP, generated in the feedwater system and transported to SG, constitutes one way to limit fouling and blockage of SGs. This work requires the understanding of CP behaviour in the feedwater system conditions. A specific experimental circulating water loop, FORTRAND, was built at EDF to follow the formation, the transport and the deposition of iron oxides in representative conditions of the secondary circuit feedwater system. The test section operating at high temperature (up to 250 deg. C) is made in carbon steel and includes three removable segments while all the other parts of the loop are made in stainless steel. First results confirm the formation of iron oxides on carbon steel and stainless steel surface in the conditions of PWR secondary circuits. The surface characterizations show that magnetite is the corrosion product formed on carbon steel and stainless steel at 220 deg. C and goethite is formed at room temperature on stainless steel. The aim of the most recent tests performed in FORTRAND loop was to follow the evolution of corrosion in the feedwater conditions. Tests were performed in one-phase flow conditions at 150 L.h -1 with a linear velocity of 0.82 m/s at 220 deg. C in morpholine/ammonia/hydrazine medium, at pH 25C equal to 9.2. To conduct this study, a removable segment constituted by ten tubes was added to the loop. Several tests were performed to follow the deposit thickness, the iron lost in solution and the oxide morphology with time from two to nine hundred sixty hours. Chemical conditions were controlled and the reproducibility of the results was confirmed by the observation of three tubes at each test. SEM pictures present kinetics with three steps: after the first hours the

  12. 75 FR 69125 - Certain Seamless Carbon and Alloy Steel Standard, Line, and Pressure Pipe From China

    Science.gov (United States)

    2010-11-10

    ... with material injury by reason of imports from China of certain seamless carbon and alloy steel standard, line, and pressure pipe (``seamless SLP pipe''), provided for in subheadings 7304.19.10, 7304.19... Seamless Carbon and Alloy Steel Standard, Line, and Pressure Pipe From China Determination On the basis of...

  13. RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, Stephanie; Van-Duysen, Jean Claude

    2005-01-01

    Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called 'Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, ...) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program

  14. Mitigating the Risk of Stress Corrosion of Austenitic Stainless Steels in Advanced Gas Cooled Reactor Boilers

    International Nuclear Information System (INIS)

    Bull, A.; Owen, J.; Quirk, G.; G, Lewis; Rudge, A.; Woolsey, I.S.

    2012-09-01

    Advanced Gas-Cooled Reactors (AGRs) operated in the UK by EDF Energy have once-through boilers, which deliver superheated steam at high temperature (∼500 deg. C) and pressure (∼150 bar) to the HP turbine. The boilers have either a serpentine or helical geometry for the tubing of the main heat transfer sections of the boiler and each individual tube is fabricated from mild steel, 9%Cr1%Mo and Type 316 austenitic stainless steel tubing. Type 316 austenitic stainless steel is used for the secondary (final) superheater and steam tailpipe sections of the boiler, which, during normal operation, should operate under dry, superheated steam conditions. This is achieved by maintaining a specified margin of superheat at the upper transition joint (UTJ) between the 9%Cr1%Mo primary superheater and the Type 316 secondary superheater sections of the boiler. Operating in this mode should eliminate the possibility of stress corrosion cracking of the Type 316 tube material on-load. In recent years, however, AGRs have suffered a variety of operational problems with their boilers that have made it difficult to maintain the specified superheat margin at the UTJ. In the case of helical boilers, the combined effects of carbon deposition on the gas side and oxide deposition on the waterside of the tubing have resulted in an increasing number of austenitic tubes operating with less than the specified superheat margin at the UTJ and hence the possibility of wetting the austenitic section of the boiler. Some units with serpentine boilers have suffered creep-fatigue damage of the high temperature sections of the boiler, which currently necessitates capping the steam outlet temperature to prevent further damage. The reduction in steam outlet temperature has meant that there is an increased risk of operation with less than the specified superheat margin at the UTJ and hence stress corrosion cracking of the austenitic sections of the boiler. In order to establish the risk of stress

  15. A Study of the Effect of Interrupted Quenches on a Thermomechanically Processed High Carbon Steel.

    Science.gov (United States)

    1982-10-01

    steel . Successful martempering requires a cooling rate sufficient to avoid the nose of the C- curve and thus prevent significant bainite formation. When...STUDY OF THE EFFECT OF INTERRUPTED QUENCHES ON A THERMONECHANICALLY PROCESSED HIGH CARBON STEEL by Steven A. Barton October 1982 Thesis Advisor: T.R...unlimited. A Study of the Effect of Interrupted Quenches on a Thermomechanically Processed High Carbon Steel by Steven A. Barton Lieutenant, United

  16. Compatibility of steels for fast breeder reactor in high temperature sodium

    International Nuclear Information System (INIS)

    Yuhara, Shunichi

    1981-01-01

    In recent years, considerable progress has been made and experience has been obtained for material applicability in sodium-cooled fast breeder reactors. In this report, materials, principal dimensions and sodium conditions for the reactor system components, which include fuel pin cladding, intermediate heat exchangers, steam generators and pipings, are reviewed with emphasis on the thin-walled, high temperature and high strength components. The corrosion, mechanical and tribological behavior in sodium of important materials used for the reactor components, such as Types 304 and 316 stainless steel and 2 1/4Cr-1Mo steel, are discussed on the basis of characteristic testing results. Furthermore, material requirements concerned with compatibility in sodium are summarized from this review and discussion. (author)

  17. Investigation of neutron irradiated reactor vessel steels using post-irradiation annealing techniques

    Energy Technology Data Exchange (ETDEWEB)

    Nakata, Hayato; Fukuya, Koji [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2001-09-01

    The matrix damage is known to be a major factor that contributes to embrittlement and hardening of irradiated reactor vessel steels, and is assumed to be composed of the point defect clusters. However field emission gun scanning transmission electron microscopy (FEGSTEM) and atom probe (AP) could not detect any evidence of the matrix damage. In this study, post irradiation annealing experiments combining positron annihilation lineshape analysis (PALA) and hardness experiments were applied to an actual surveillance test specimen and a sample of reactor vessel steel irradiated in a material test reactor (MTR), in order to investigate the matrix damage recovery behavior and its contribution to hardening. It was confirmed that higher fluence increased the hardness and the volume fraction of open volume defects and that higher flux decreased the thermal stability of matrix damage and the effect on hardening. The contribution of matrix damage to hardening could be estimated to be below 30%. (author)

  18. Apparent embrittlement saturation and radiation mechanisms of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Pachur, D.

    1981-01-01

    The irradiation and annealing results of three different reactor pressure vessel steels are reported. Steel A, a basic material according to ASTM A-533 B having 0.15 percent vanadium; and Steel C contained 3.2 percent nickel. The steels were irradiated at 150, 300, and 400 degree C with neutron fluxes of 6 multiplied by 10 11 and 3 multiplied by 10 13 neutrons (n)/cm 2 /s. An apparent saturation-in-irradiation effect was found within certain neutron fluence ranges. During the annealing, various recovery processes occur in different temperature ranges. These are characterized by various activation energies. The individual processes were determined by the different time dependencies at various temperatures. Two causes for the apparent saturation were discovered from the behavior of the annealing curves

  19. Study of reactions between nuclear fuel and cladding (316 stainless steel) in reactors. Influence of oxygen

    International Nuclear Information System (INIS)

    Otter, Monique.

    1980-12-01

    We have studied oxidation of 316 steel in close contact with oxides (Usub(0,74)Pusub(0,26)O 2 or UO 2 ), the stoichiometry of oxygen ranging from 2.00 to 2.5. Experiments are carried out either in a closed isothermal system or in an opened isothermal system with a fixed oxygen potential of uranium oxide. We have realized a potentiostatic device using a solid state electrotyte galvanic cell. In a closed system, the sensitized austenitic steel shows intergranular and volume oxidation probably enhanced by migration of steel components towards the fuel. Evidence of the usefulness of passivation have been obtained. We conclude that in a fast reactor sensitized cladding steel is oxydized by the constant potential of oxygen of UPuO 2 . Deposits observed in fuel can be explain by evaporation and cyclic transport phenomena that can be differents from VAN-ARKEL mechanism taking place through fission products [fr

  20. Behavior of Type 316 stainless steel under simulated fusion reactor irradiation

    International Nuclear Information System (INIS)

    Wiffen, F.W.; Maziasz, P.J.; Bloom, E.E.; Stiegler, J.O.; Grossbeck, M.L.

    1978-05-01

    Fusion reactor irradiation response in alloys containing nickel can be simulated in thermal-spectrum fission reactors, where displacement damage is produced by the high-energy neutrons and helium is produced by the capture of two thermal neutrons in the reactions: 58 Ni + n → 59 Ni + γ; 59 Ni + n → 56 Fe + α. Examination of type 316 stainless steel specimens irradiated in HFIR has shown that swelling due to cavity formation and degradation of mechanical properties are more severe than can be predicted from fast reactor irradiations, where the helium contents produced are far too low to simulate fusion reactor service. Swelling values are greater and the temperature dependence of swelling is different than in the fast reactor case

  1. Raising the Corrosion Resistance of Low-Carbon Steels by Electrolytic-Plasma Saturation with Nitrogen and Carbon

    Science.gov (United States)

    Kusmanov, S. A.; Grishina, E. P.; Belkin, P. N.; Kusmanova, Yu. V.; Kudryakova, N. O.

    2017-05-01

    Structural features of the external oxide layer and internal nitrided, carbonitrided and carburized layers in steels 10, 20 and St3 produced by the method of electrolytic plasma treatment are studied. Specimens of the steels are tested for corrosion in a naturally aerated 1-N solution of sodium chloride. The condition of the metal/sodium chloride solution interface is studied by the method of electrochemical impedance spectroscopy. It is shown that the corrosion resistance of low-carbon steels can be raised by anode electrolytic-plasma saturation with nitrogen and carbon. Recommendations are given on the choice of carbonitriding modes for structural steels.

  2. A study of the condition for the passivation of carbon steel in bentonite

    International Nuclear Information System (INIS)

    Taniguchi, Naoki; Morimoto, Masataka; Honda, Akira

    1999-01-01

    It is important to study the corrosion behavior of materials to be used for overpack for high-level radioactive waste disposal. Carbon steel is one of the candidate materials. The type of corrosion on carbon steel depends on whether the carbon steel is passivated or not. In this study, the condition for the passivation of carbon steel was studied using bentonite as the buffer material. Anodic polarization in bentonite and the measurements of pH of porewater in bentonite was measured. The results of these experiments showed that the possibility of passivation is small in highly compacted bentonite in groundwater in Japan. Therefore, localized corrosion on carbon steel due to the breakdown of passive film is unlikely in bentonite. In other words, general corrosion seems to be the most probable type of corrosion under repository condition in Japan. (author)

  3. Inelastic Cyclic Deformation Behaviors of Type 316H Stainless Steel for Reactor Pressure Vessel of Sodium-Cooled Fast Reactor at Elevated Temperatures

    International Nuclear Information System (INIS)

    Yoon, Ji-Hyun; Hong, Seokmin; Koo, Gyeong-Hoi; Lee, Bong-Sang; Kim, Young-Chun

    2015-01-01

    Type 316H stainless steel is a primary candidate material for a reactor pressure vessel of a sodium-cooled fast (SFR) reactor which is under development in Korea. The reactor pressure vessel for a SFR is subjected to inelastic deformation induced by cyclic thermal stress. Fully reversed cyclic testing and ratcheting testing at elevated temperatures were performed to characterize the inelastic cyclic deformation behaviors of Type 316H stainless steel at the SFR operating temperature. It was found that cyclic hardening of Type 316H stainless steel was enhanced, and the accumulation of ratcheting deformation of Type 316H stainless steel was retarded at around the SFR operating temperature. The results of the tensile testing and the microstructural investigation for dislocated structures after the inelastic deformation testing showed that dynamic strain aging affected the inelastic cyclic deformation behavior of Type 316 stainless steel at around the SFR operating temperature.

  4. Reactor scale modeling of multi-walled carbon nanotube growth

    International Nuclear Information System (INIS)

    Lombardo, Jeffrey J.; Chiu, Wilson K.S.

    2011-01-01

    As the mechanisms of carbon nanotube (CNT) growth becomes known, it becomes important to understand how to implement this knowledge into reactor scale models to optimize CNT growth. In past work, we have reported fundamental mechanisms and competing deposition regimes that dictate single wall carbon nanotube growth. In this study, we will further explore the growth of carbon nanotubes with multiple walls. A tube flow chemical vapor deposition reactor is simulated using the commercial software package COMSOL, and considered the growth of single- and multi-walled carbon nanotubes. It was found that the limiting reaction processes for multi-walled carbon nanotubes change at different temperatures than the single walled carbon nanotubes and it was shown that the reactions directly governing CNT growth are a limiting process over certain parameters. This work shows that the optimum conditions for CNT growth are dependent on temperature, chemical concentration, and the number of nanotube walls. Optimal reactor conditions have been identified as defined by (1) a critical inlet methane concentration that results in hydrogen abstraction limited versus hydrocarbon adsorption limited reaction kinetic regime, and (2) activation energy of reaction for a given reactor temperature and inlet methane concentration. Successful optimization of a CNT growth processes requires taking all of those variables into account.

  5. Development of austenitic stainless steel plate (316MN) for fast breeder reactors

    International Nuclear Information System (INIS)

    Nakazawa, Takanori; Abo, Hideo; Tanino, Mitsuru; Komatsu, Hazime.

    1989-01-01

    High creep-fatigue resistance is required for the structural materials for fast breeder reactors. As creep-fatigue life is closely related to creep-rupture ductility, the effects of C, N and Mo on creep-rupture properties were investigated with a view to improving the creep-fatigue resistance of stainless steel. Strengthening by the addition of C has a great adverse effect on rupture ductility, but N can strengthen the steel without decreasing rupture ductility. Strengthening by Mo decreases rupture ductility but this effect is small. The low-C-medium-N (0.01%C - 0.07%N) stainless steel 316 MN developed based on the findings described above exhibits only a small decrease in creep-rupture strength in long-time periods compared with the conventional 316 steel. This steel offers excellent rupture ductility and the 10,000-hour rupture strength which is about 1.2 times that of conventional steel. Moreover, this steel exhibits excellent properties in creep fatigue test. (author)

  6. Special Advanced Studies for Pollution Prevention. Delivery Order 0017: Sol-Gel Surface Preparation for Carbon Steel and Stainless Steel Bonding

    National Research Council Canada - National Science Library

    Zheng, Haixing

    1997-01-01

    The objective of this program is to study the feasibility of using sol-gel active alumina coatings for the surface preparation of carbon steel and stainless steel for adhesive bonding, and to optimize...

  7. Relationships between Charpy impact shelf energies and upper shelf Ksub(IC) values for reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Witt, F.J.

    1983-01-01

    Charpy shelf data and lower bound estimates of Ksub(IC) shelf data for the same steels and test temperatures are given. Included are some typical reactor pressure vessel steels as well as some less tough or degraded steels. The data were evaluated with shelf estimates of Ksub(IC) up to and exceeding 550 MPa√m. It is shown that the high shelf fracture toughness representative of tough reactor pressure vessel steels may be obtained from a knowledge of the Charpy shelf energies. The toughness transition may be obtained either by testing small fracture toughness specimens or by Charpy energy indexing. (U.K.)

  8. Antimony (Sb) sorption studies on zircaloy, carbon steel (CS) and magnetite coated CS (MCS) surfaces in aqueous medium at pH 10.2 and 280℃

    International Nuclear Information System (INIS)

    Keny, S.J.; Kumbhar, A.G.; Achary, S.N.; Basu, Saibal

    2014-01-01

    Antimony sorption studies on zircaloy, CS and magnetite coated carbon steel (MCS) at primary heat transport temperature (290℃) of pressurised heavy water reactor (PHWR) are of direct relevance in investigating Sb activity problem faced in Indian PHWRs. Sb impregnated PHT pump carbon bearing releases Sb to reactor core. This Sb activates, and redeposit on out-of-core surfaces and results in exposure and apparent high decontamination factors. This Sb is not amenable to normal decantation. The form and state of deposited Sb is not yet fully known. This works attempts for this

  9. 76 FR 66893 - Certain Circular Welded Carbon Steel Pipes and Tubes From India, Thailand, and Turkey; Final...

    Science.gov (United States)

    2011-10-28

    ...] Certain Circular Welded Carbon Steel Pipes and Tubes From India, Thailand, and Turkey; Final Results of... circular welded carbon steel pipes and tubes from India, Thailand, and Turkey, pursuant to section 751(c..., Thailand, and Turkey. See Antidumping Duty Order; Certain Welded Carbon Steel Standard Pipes and Tubes from...

  10. 78 FR 16832 - Corrosion-Resistant Carbon Steel Flat Products From Germany and the Republic of Korea: Revocation...

    Science.gov (United States)

    2013-03-19

    ...] Corrosion-Resistant Carbon Steel Flat Products From Germany and the Republic of Korea: Revocation of... ``ITC'') that revocation of the antidumping duty (``AD'') orders on corrosion-resistant carbon steel... (``Sunset'') Review, 77 FR 85 (January 3, 2012). \\2\\ See Corrosion-Resistant Carbon Steel Flat Products From...

  11. Irradiation effects on material properties of steels used in nuclear reactors: a literature review

    International Nuclear Information System (INIS)

    Gerceker, N.; Dara, I. H.

    2001-01-01

    The structural materials of a nuclear power plant are of vital importance as they provide mechanical strength, structural support and physical containment for the primary reactor components as well as the nuclear power plant itself. These structural materials comprise mainly of metals and their alloys, ceramics and cermets. However, metals and their alloys are the most widely used materials and the irradiation effects are more pronounced on metallic materials as of their high temperature properties are more sensitive (with respect to ceramics and cermets) to any kind of external effects. The wholesale creation of effects on material properties has been studied for over four decades and it is not realistic to attempt to represent even a small part of the field in single poster paper. In the present contribution, a literature review of the irradiation effects on the material properties of different types of steel alloys will be given because steels are widely used as structural materials in reactors and therefore the irradiation effects on steels may be of paramount importance for reactor design, operation and safety concepts which will be discussed about radiation effects on material properties of steels will provide highlights to better understanding of the origins and development of radiation effects in materials

  12. Behaviour and microstructure of stainless steels irradiated in the french fast breeder reactors

    International Nuclear Information System (INIS)

    Dubuisson, P.; Gilbon, D.

    1991-01-01

    The burn-up of Fast Breeder Reactors is limited by the irradiation induced dimensional changes and mechanical properties of structural materials used for replaceable in-core components. This paper describes the behaviour improvements and also the radiation-induced microstructures of the different steels used for fuel pin cladding and wrapper tubes in French reactors. Materials of fuel pin cladding are austenitic steels whose main problem is swelling. Improvements in swelling resistance by cold-working, titanium additions and modifications of matrix (Fe-Cr-Ni) from SA 316 to CW 15-15 Ti are shown. These improvements are correlated with a best stability of microstructure under irradiation. Beneficial effects of phosphorus addition and multistabilisation (NbVTi) on radiation induced microstructure and swelling resistance are also shown. Austenitic steels used for wrapper tubes are limited both by swelling and by void embrittlement. The ferritic F17 (17Cr), ferritic-martensitic EM12 (9Cr-2MoNbV) and martensitic EM10 (9Cr-1Mo) steels present high swelling resistance. Nevertheless radiation-induced embrittlement is observed in EM12 and especially in F17. This embrittlement results from a fine and uniform radiation enhanced precipitation in ferrite grains. By contrast, the microstructure of fully martensitic EM10 steel is mush more stable and its ductile-brizzle transition temperature stays below 0 0 C. 12 figs

  13. Recent Advances on Carbon Molecular Sieve Membranes (CMSMs and Reactors

    Directory of Open Access Journals (Sweden)

    Margot A. Llosa Tanco

    2016-08-01

    Full Text Available Carbon molecular sieve membranes (CMSMs are an important alternative for gas separation because of their ease of manufacture, high selectivity due to molecular sieve separation, and high permeance. The integration of separation by membranes and reaction in only one unit lead to a high degree of process integration/intensification, with associated benefits of increased energy, production efficiencies and reduced reactor or catalyst volume. This review focuses on recent advances in carbon molecular sieve membranes and their applications in membrane reactors.

  14. Improved corrosion resistance of cast carbon steel in sulphur oxides by Alonizing

    International Nuclear Information System (INIS)

    Holtzer, M.; Dzioba, Z.

    1992-01-01

    The results of studies on the Alonizing of cast steel and of testing the corrosion resistance of this cast steel in an atmosphere containing 5 to 6% SO 2 + 50% SO 3 at 853 K are described and compared with the results obtained with unalonized cast carbon steel and high-alloy 23Cr-8Ni-2Mo cast steel. The duration of the corrosion tests was 336 hours. The aluminium diffusion layer on cast carbon steel was obtained by holding the specimens in a mixture containing 99% of powered Fe-Al and 1% of NH 4 Cl at 1323 ± 20 K. The holding time was 10 and 20 hours, respectively. The aluminium layer formed on the cast carbon steel was examined by optical microscopy and an X-ray microanalysis. After Alonizing for 10 h the layer had reached a thickness of 950 μm, and contained up to 35% Al. In a mixture of sulphur oxides corrosion rate of the alonized cast carbon steel was by about 600 times lower than of the unalonized cast carbon steel, and by about 50 times lower than that of the 23Cr-8Ni-2Mo cast steel. (orig.) [de

  15. Control of activation levels to simplify waste management of fusion reactor ferritic steel components

    International Nuclear Information System (INIS)

    Wiffen, F.W.; Santoro, R.T.

    1983-01-01

    Activation characteristics of a material for service in the neutron flux of a fusion reactor first wall fall into three areas: waste management, reactor maintenance and repair, and safety. Of these, the waste management area is the most likely to impact the public acceptance of fusion reactors for power generation. The decay of the activity in steels within tens of years could lead to simplified waste disposal or possibly even to materials recycle. Whether or not these can be achieved will be controlled by (1) selection of alloying elements, (2) control of critical impurity elements, and (3) control of cross contamination from other reactor components. Several criteria can be used to judge the acceptability of potential alloying elements in iron, and to define the limits on content of critical impurity elements. One approach is to select and limit alloying additions on the basis of the activity. If material recycle is a goal, N, Al, Ni, Cu, Nb, and Mo must be excluded. If simplified waste storage by shallow land burial is the goal, regulations limit the concentration of only a few isotopes. For first-wall material that will be exposed to 9 MW-y/m 2 service, allowable initial concentration limits include (in at. ppM) Ni < 20,000; Mo < 3650; N < 3650, Cu < 2400; and Nb < 1.0. The other constituent elements of ferritic steels will not be limited. Possible substitutes for the molybdenum normally used to strengthen the steels include W, Ta, Ti, and V

  16. Electrochemical Corrosion Behavior of Carbon Steel and Hot Dip Galvanized Steel in Simulated Concrete Solution with Different pH Values

    Directory of Open Access Journals (Sweden)

    Wanchen XIE

    2017-08-01

    Full Text Available Hot dip galvanizing technology is now widely used as a method of protection for steel rebars. The corrosion behaviors of Q235 carbon steel and hot galvanized steel in a Ca(OH2 solution with a pH from 10 to 13 was investigated by electrode potential and polarization curves testing. The results indicated that carbon steel and hot galvanized steel were all passivated in a strong alkaline solution. The electrode potential of hot dip galvanized steel was lower than that of carbon steel; thus, hot dip galvanized steel can provide very good anodic protection for carbon steel. However, when the pH value reached 12.5, a polarity reversal occurred under the condition of a certain potential. Hot dip galvanized coating became a cathode, and the corrosion of carbon steel accelerated. The electrochemical behaviors and passivation abilities of hot dip galvanized steel and carbon steel were affected by pH. The higher the pH value was, the more easily they were passivated.DOI: http://dx.doi.org/10.5755/j01.ms.23.3.16675

  17. Hot ductility of medium carbon steel with vanadium

    International Nuclear Information System (INIS)

    Lee, Chang-Hoon; Park, Jun-Young; Chung, JunHo; Park, Dae-Bum; Jang, Jin-Young; Huh, Sungyul; Ju Kim, Sung; Kang, Jun-Yun; Moon, Joonoh; Lee, Tae-Ho

    2016-01-01

    Hot ductility of medium carbon steel containing 0.52 wt% of carbon and 0.11 wt% of vanadium was investigated using a hot tensile test performed up to fracture. The hot ductility was evaluated by measuring the reduction of area of the fractured specimens, which were strained at a variety of test temperatures in a range of 600–1100 °C at a strain rate of 2×10"−"3/s. The hot ductility was excellent in a temperature range of 950–1100 °C, followed by a decrease of the hot ductility below 950 °C. The hot ductility continued to drop as the temperature was lowered to 600 °C. The loss of hot ductility in a temperature range of 800–950 °C, which is above the Ae_3 temperature, was due to V(C,N) precipitation at austenite grain boundaries. The further decline of hot ductility between 700 °C and 750 °C resulted from the transformation of ferrite films decorating austenite grain boundaries. The hot ductility continued to decrease at 650 °C or less, owing to ferrite films and the pearlite matrix, which is harder than ferrite. The pearlite was transformed from austenite due to relatively high carbon content.

  18. Two new techniques for the remote evaluation of reactor steel condition - microscopic removal and surface examination

    International Nuclear Information System (INIS)

    Clayton, R.

    Much reactor inspection work involves an assessment of the condition of structural steel. This paper reviews two different techniques which provide information for such an assessment. The first - micro-sample removal (for the measurement of surface oxide thickness and chemical composition) - requires contact with the steel surface, whereas the second - a 'teach and learn' photographic technique (in which a special photogrammatic camera is used to obtain high-quality close-up photographs, to assess surface condition and corrosion growth) can obtain surface information on inaccessible components. (author)

  19. Damage mechanism of piping welded joints made from austenitic Steel for the type RBMK reactor

    International Nuclear Information System (INIS)

    Karzov, G.; Timofeev, B.; Gorbakony, A.; Petrov, V.; Chernaenko, T.

    1999-01-01

    In the process of operation of RBMK reactors the damages were taking place on welded piping, produced from austenitic stainless steel of the type 08X18H10T. The inspection of damaged sections in piping has shown that in most cases crack-like defects are of corrosion and mechanical character. The paper considers in details the reasons of damages appearance and their development for this type of welded joints of downcomers 325xl6 mm, which were fabricated from austenitic stainless steel using TlG and MAW welding methods. (author)

  20. Effects of nickel on irradiation embrittlement of light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    2005-06-01

    This TECDOC was developed under the IAEA Coordinated Research Project (CRP) entitled Effects of Nickel on Irradiation Embrittlement of Light Water Reactor Pressure Vessel (RPV) Steels. This CRP is the sixth in a series of CRPs to determine the influence of the mechanism and quantify the influence of nickel content on the deterioration of irradiation embrittlement of reactor pressure vessel steels of the Ni-Cr-Mo-V or Mn-Ni-Cr-Mo types. The scientific scope of the programme includes procurement of materials, determination of mechanical properties, irradiation and testing of specimens in power and/or test reactors, and microstructural characterization. Eleven institutes from eight different countries and the European Union participated in this CRP and six institutes conducted the irradiation experiments of the CRP materials. In addition to the irradiation and testing of those materials, irradiation experiments of various national steels were also conducted. Moreover, some institutes performed microstructural investigations of both the CRP materials and national steels. This TECDOC presents and discusses all the results obtained and the analyses performed under the CRP. The results analysed are clear in showing the significantly higher radiation sensitivity of high nickel weld metal (1.7 wt%) compared with the lower nickel base metal (1.2 wt%). These results are supported by other similar results in the literature for both WWER-1000 RPV materials, pressurized water reactor (PWR) type materials, and model alloys. Regardless of the increased sensitivity of WWER-1000 high nickel weld metal (1.7 wt%), the transition temperature shift for the WWER-1000 RPV design fluence is still below the curve predicted by the Russian code (standard for strength calculations of components and piping in NPPs - PNAE G 7-002-86). For higher fluence, no data were available and the results should not be extrapolated. Although manganese content was not incorporated directly in this CRP

  1. Corrosion behaviour of dissimilar welds between martensitic stainless steel and carbon steel from secondary circuit of candu npp

    International Nuclear Information System (INIS)

    Popa, L.; Fulger, M.; Tunaru, M.; Velciu, L.; Lazar, M.

    2015-01-01

    Corrosion damages of welds occur in spite of the fact that the proper base metal and filler metal have been correctly selected, industry codes and standards have been followed and welds have been realized with full weld penetration and have proper shape and contour. It is not unusual to find that, although the base metal or alloy is resistant to corrosion in a particular environment, the welded counterpart is not resistant. In secondary circuit of a Nuclear Power Station there are some components which have dissimilar welds. Our experiments were performed in chloride environmental on two types of samples: non-welded (420 martensitic steel and 52.2k carbon steel) and dissimilar welds (dissimilar metal welds: joints beetween 420 martensitic steel and 52.2k carbon steel). To evaluate corrosion susceptibility of dissimilar welds was used electrochemical method (potentiodynamic method) and metallography microscopy (microstructural analysis). The present paper follows the localized corrosion behaviour of dissimilar welds between austenitic stainless steel and carbon steel in solutions containing chloride ions. We have been evaluated the corrosion rates of samples (welded and non-welded) by electrochemically. (authors)

  2. High-carbon chromium steel resistance to small plastic deformation

    International Nuclear Information System (INIS)

    Gajduchenya, V.F.; Madyanov, S.A.; Apaev, B.A.; Kirillov, Yu.V.; Sokolov, L.D.

    1978-01-01

    The phase composition of a steel with 1.08% C and 2.1% Cr, and the variation in the level of microstresses in the matrix as related to the annealing temperature in the range of 400-600 deg C and in the applied compression stress were investigated. To study the phase composition, and chromium content in the α-solution and the carbide phases, magnetic, chemical, and X-ray spectrum analyses were carried out. The change in the level of microstresses was determined roentgenographically. During the stress relaxation test at temperatures of 20-180 deg C, the mechanism of plastic deformation near the yield point was investigated. It is shown that three dislocation mechanisms operate in high-carbon chromium steel under the conditions at hand: overcoming the Pierls-Nabarro barriers by the dislocations, overcoming the stress fields of coherent carbide particles by dislocations, and circumvention of second-phase particles by dislocations. The dependence of the realization of the different plastic deformation mechanisms on the number of carbide particles and the chromium concentration in the matrix was established. The thermally activated nature of the motion of the dislocations under conditions of stress relaxation at an elevated temperature is noted

  3. Crystallographic features of lath martensite in low-carbon steel

    International Nuclear Information System (INIS)

    Kitahara, Hiromoto; Ueji, Rintaro; Tsuji, Nobuhiro; Minamino, Yoritoshi

    2006-01-01

    Electron backscattering diffraction with field-emission scanning electron microscopy was used to analyze crystallographically the lath martensite structure in a 0.20% carbon steel. The crystallographic features of the lath martensite structure, of the order of the prior austenite grain size or larger, were clarified. Although the orientations of the martensite crystals were scattered around the ideal variant orientations, the martensite in this steel maintained the Kurdjumov-Sachs (K-S) orientation relationship. The procedures of the crystallographic analysis of the martensite (ferrite) phase with the K-S orientation relationship were explained in detail. Variant analysis showed that all 24 possible variants did not necessarily appear within a single prior austenite grain and that all six variants did not necessarily appear within each packet. Specific combinations of two variants appeared within local regions (sub-blocks), indicating a strict rule for variant selection. Prior austenite grain boundaries and most of the packet boundaries were clearly recognized. However, it was difficult to determine the block boundaries within the sub-blocks

  4. The kinetics of pitting corrosion of carbon steel

    International Nuclear Information System (INIS)

    Marsh, G.P.; Taylor, K.J.; Sooi, Z.

    1988-02-01

    The development of an improved statistical method for analysing pit growth data to take account of the difference in area of laboratory specimens and full sized high level nuclear waste containers is described. Statistical analysis of data from pit growth experiments with large area (460 cm 2 ) plates of BS 4360 steel have indicated that the depth distributions correlate most closely with a limited distribution function. This correlation implies that previous statistical analyses to estimate the maximum pit depths in full size containers, which were made using unlimited distribution functions, will be pessimistic. An evaluation of the maximum feasible pitting period based on estimating the period during which the oxygen diffusion flux is sufficient to stabilise a passive film on carbon steel containers has indicated that this is of the order of 125 years rather than the full 1000 year container life. The estimate is sensitive to the value of the leakage current assumed to flow through the passive film, and therefore work is planned to measure this accurately in relevant granitic environments. (author)

  5. The corrosion behaviour of carbon steel in Portland cement

    International Nuclear Information System (INIS)

    Grauer, R.

    1988-01-01

    The production of hydrogen can cause problems in a repository for low- and intermediate-level waste. Since gas production is mainly due to the corrosion of carbon steel, it is important to have as reliable data as possible on the corrosion rate of steel in anaerobic cement. A review of the literature shows that the corrosion current densities lie in the range 0.01 to 0.1 μA/cm 2 (corresponding to corrosion rates between 0.1 and 1.2 μm/a). This implies hydrogen production rates between 0.022 and 0.22 mol/(m 2 .a). Corrosion rates of this order of magnitude are technically irrelevant, with the result that there is very little interest in determining them accurately. Furthermore, their determination entails problems of measurement technique. Given the current situation, it would appear somewhat risky to accept the lower value for hydrogen production as proven. Proposals are made for experiments which would reduce this element of uncertainty. (author) 10 figs., 35 refs

  6. Environmental review of options for managing radioactively contaminated carbon steel

    International Nuclear Information System (INIS)

    1996-10-01

    The U.S. Department of Energy (DOE) is proposing to develop a strategy for the management of radioactively contaminated carbon steel (RCCS). Currently, most of this material either is placed in special containers and disposed of by shallow land burial in facilities designed for low-level radioactive waste (LLW) or is stored indefinitely pending sufficient funding to support alternative disposition. The growing amount of RCCS with which DOE will have to deal in the foreseeable future, coupled with the continued need to protect the human and natural environment, has led the Department to evaluate other approaches for managing this material. This environmental review (ER) describes the options that could be used for RCCS management and examines the potential environmental consequences of implementing each. Because much of the analysis underlying this document is available from previous studies, wherever possible the ER relies on incorporating the conclusions of those studies as summaries or by reference

  7. Hot corrosion of pack cementation aluminized carbon steel

    International Nuclear Information System (INIS)

    Waheed, A.F.; Mohamed, K.E.; Abd El-Azim, M.E.; Soliman, H.M.

    1998-01-01

    Low carbon steel was aluminized by the pack cementation technique at various aluminizing temperatures and times in or der to have different aluminide coatings. The aluminized specimens were sprayed at the beginning of the hot corrosion experiments with Na C 1+Na 2 SO 4 solution. The hot corrosion tests were carried out by thermal cycling at 850 degree C in air. The results were evaluated by, corrosion kinetics based on weight change measurements, scanning electron microscopy and energy dispersive X-ray analysis. It was found that the maximum corrosion resistance to this corrosive environment is achieved by aluminizing at 900 degree C for 19 h or 950 degree C for >4 h. These aliminizing conditions lead to formation of thick aluminide coatings with sufficient aluminium concentration (>15 wt%) at their outer surface necessary for continuous formation of protective Al 2 O 3 scale. The tested materials are used in protection of some components used in electric power stations (conventional or nuclear)

  8. Effect of Nanoparticles on Wettability of Nanocoating on Carbon Steel

    Directory of Open Access Journals (Sweden)

    Norhasnidawani Johari

    2016-12-01

    Full Text Available Nanocoatings plays an important role in coating industry. The solution was being prepared through copolymerization of epoxy resin hardener and with the incorporation of metal oxide nanoparticles, Zinc Oxide (ZnO and Silica (SiO2. ZnO and SiO2 were synthesized using sol-gel. Epoxy hardener acted as host while the metal oxide nanoparticles as guest components. The formulation of nanocoatings with excellent adhesion strength and corrosion protection of carbon steel was studied. The performance of wetting ability with different medium was analysed using contact angle. Water medium showed the addition of 3wt% of hybrid between ZnO and SiO2 was the best nanocoating to form hydrophobic surface and was also the best nanocoating surface to form hydrophilic surface with vacuum oil dropping. In oil dropping, the contact angle was smaller than 90° and the water drop tends to spreads on surface.

  9. Microbiologically induced corrosion of carbon steel under continuous flow conditions

    International Nuclear Information System (INIS)

    Tunaru, Mariana; Dragomir, Maria; Voicu, Anca

    2008-01-01

    Microbiologically induced corrosion is the label generally applied to corrosion involving the action of bacteria on metal surfaces. While different combinations of bacterial species, materials and chemical constituents are interrelated factors, stagnant water is the factor most often mentioned in reported cases. This paper presents the results obtained regarding the testing of microbiologically induced corrosion of carbon steel under continuous flow conditions in the presence of iron-oxidizing bacteria. The tests were performed on coupons of SA106gr.B exposed both in stagnant conditions and in flow conditions. The surfaces of these coupons were studied by metallographic technique, while the developed biofilms were analysed using microbiological technique. The correlation of all the results which were obtained emphasized that the minimizing the occurrence of stagnant or low-flow conditions can prove effective in reducing the risk of microbiologically induced corrosion in plant cooling-water systems. (authors)

  10. Stress corrosion cracking of A515 grade 60 carbon steel

    International Nuclear Information System (INIS)

    Moore, E.L.

    1971-01-01

    An investigation was conducted to evaluate the effect of welding method plate thickness, and subsequent stress relief treatment on the stress corrosion cracking propensity of ASTM A515 Grade 60 carbon steel plate exposed to a 5 M NaNO 3 solution at 190 0 F for eight weeks. It was found that all weld coupons receiving no thermal stress relief treatment cracked within eight weeks; all weld coupons given a vibratory stress relief cracked within eight weeks; two of the eight weld coupons stress relieved at 600 0 F for one hour cracked within eight weeks; none of the weld coupons stress relieved at 1100 0 F for one hour cracked within eight weeks; and that cracking was generally more severe in coupons fabricated from 7/8 inch plate by shielded metal arc welding than it was in coupons fabricated by other welding methods. (U.S.)

  11. Polyaspartic acid as a green corrosion inhibitor for carbon steel

    Energy Technology Data Exchange (ETDEWEB)

    Cui, R. [Department of Chemistry, Hebei Normal University, Shijiazhuang 050016 (China); Department of Chemistry and Materials Engineering, Changshu Institute of Technology, Changshu 215500 (China); Gu, N.; Li, C. [Department of Chemistry, Hebei Normal University, Shijiazhuang 050016 (China)

    2011-04-15

    The inhibitor effect of the environmentally friendly corrosion inhibitor polyaspartic acid (PASP) on the corrosion of carbon steel in 0.5 M H{sub 2}SO{sub 4} was investigated by weight loss, potentiodynamic polarization, electrochemical impedance spectroscopy (EIS), and scanning electron microscopy (SEM). Polarization curve results clearly reveal the fact that PASP is a good anode-type inhibitor. EIS results confirm its corrosion inhibition ability. The inhibition efficiency increases with increasing PASP concentration, and the maximum inhibition efficiency was 80.33% at 10 C. SEM reveals that a protective film forms on the surface of the inhibited sample. The adsorption of this inhibitor is found to follow the Freundlich adsorption isotherm. A mechanism is proposed to explain the inhibitory action of the corrosion inhibitor. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  12. Investigation of the benzotriazole as addictive for carbon steel phosphating

    International Nuclear Information System (INIS)

    Annies, V.; Cunha, M.T.; Rodrigues, P.R.P.; Banczek, E.P.; Terada, M.

    2010-01-01

    This work studied the viability of substitution of sodium nitrite (NaNO 2 ) for benzotriazole (BTAH) in the zinc phosphate bath (PZn+NaNO 2 ) for phosphating of carbon steel (SAE 1010). The characterization of the samples was carried out by Scanning Electron Microscopy, Optical Microscopy and X-ray diffraction. The chemical composition was evaluated by Energy Dispersive Spectroscopy. The corrosion behavior of the samples was investigated by Open Circuit Potential, Electrochemical Impedance Spectroscopy and Anodic Potentiodynamic Polarization Curves in a 0.5 mol L -1 NaCl electrolyte. The experimental results showed that the phosphate layer obtained in the solution with benzotriazole (PZn+BTAH) presented better corrosion resistance properties than that obtained in sodium nitrite. The results demonstrated that the sodium nitrite NaNO 2 can be replaced by benzotriazole (BTAH) in zinc phosphate baths. (author)

  13. Stability of ferritic steel to higher doses: Survey of reactor pressure vessel steel data and comparison with candidate materials for future nuclear systems

    International Nuclear Information System (INIS)

    Blagoeva, D.T.; Debarberis, L.; Jong, M.; Pierick, P. ten

    2014-01-01

    This paper is illustrating the potential of the well-known low alloyed clean steels, extensively used for the current light water Reactor Pressure Vessels (RPV) steels, for a likely use as a structural material also for the new generation nuclear systems. This option would provide, especially for large components, affordable, easily accessible and a technically more convenient solution in terms of manufacturing and joining techniques. A comprehensive comparison between several sets of surveillance and research data available for a number of RPV clean steels for doses up to 1.5 dpa, and up to 12 dpa for 9%Cr steels, is carried out in order to evaluate radiation stability of the currently used RPV clean steels even at higher doses. Based on the numerous data available, positive preliminary conclusions are drawn regarding the eventual use of clean RPV steels for the massive structural components of the new reactor systems. - Highlights: • Common embrittlement trend between RPV and advanced steels till intermediate doses. • For doses >1.5 dpa, damage rate saturation tendency is observed for RPV steels. • RPV steels might be conveniently utilised also outside their foreseen dose range

  14. The role of carbon in the breakaway oxidation of mild steel in high pressure carbon dioxide

    International Nuclear Information System (INIS)

    Surman, P.L.; Brown, A.M.

    1974-01-01

    The rate controlling step in the oxidation of iron and mild steel in CO 2 is the diffusion of iron across the inner of two layers of magnetite scale. Cation diffusion is directed towards available oxidant and hence tends to produce fresh oxide in freely available space. The initial oxidation process is thus protective and stress-free. As oxidation proceeds the gaseous reaction product, carbon monoxide, tends to accumulate at the oxide/metal interface. Eventually this leads to simultaneous carbon deposition and oxide formation. This carbon contamination allows oxidant access to oxide crystallite 'jacking points', and hence volume expansion and stressed breakaway corrosion can occur. Experiments designed to simulate the promotion, propagation and healing of breakaway oxidation and to define the conditions for carbon deposition are reported. (author)

  15. Modelling hydrogen permeation in a hydrogen effusion probe for monitoring corrosion of carbon steels

    International Nuclear Information System (INIS)

    Santiwiparat, P.; Rirksomboon, T.; Steward, F.R.; Lister, D.H.; Cook, W.G.

    2015-01-01

    Hydrogen accumulation inside carbon steel and stainless steel devices shaped like cylindrical cups attached to a pipe containing hydrogen gas was modelled with MATLAB software. Hydrogen transfer around the bottom of the cups (edge effect) and diffusion through the cup walls (material effect) were accounted for. The variation of hydrogen pressure with time was similar for both materials, but the hydrogen plateau pressures in stainless steel cups were significantly higher than those in carbon steel cups. The geometry of the cup also affected the plateau pressure inside the cup. (author)

  16. Investigating pitting in X65 carbon steel using potentiostatic polarisation

    Science.gov (United States)

    Mohammed, Sikiru; Hua, Yong; Barker, R.; Neville, A.

    2017-11-01

    Although pitting corrosion in passive materials is generally well understood, the growth of surface pits in actively-corroding materials has received much less attention to date and remains poorly understood. One of the key challenges which exists is repeatedly and reliably generating surface pits in a practical time-frame in the absence of deformation and/or residual stress so that studies on pit propagation and healing can be performed. Another pertinent issue is how to evaluate pitting while addressing general corrosion in low carbon steel. In this work, potentiostatic polarisation was employed to induce corrosion pits (free from deformation or residual stress) on actively corroding X65 carbon steel. The influence of applied potential (50 mV, 100 mV and 150 mV vs open circuit potential) was investigated over 24 h in a CO2-saturated, 3.5 wt.% NaCl solution at 30 °C and pH 3.8. Scanning electron microscopy (SEM) was utilised to examine pits, while surface profilometry was conducted to measure pit depth as a function of applied potential over the range considered. Analyses of light pitting (up to 120 μm) revealed that pit depth increased linearly with increase in applied potential. This paper relates total pit volume (measured using white light interferometry) to dissipated charge or total mass loss (using the current response for potentiostatic polarisation in conjunction with Faraday's law). By controlling the potential of the surface (anodic) the extent of pitting and general corrosion could be controlled. This allowed pits to be evaluated for their ability to continue to propagate after the potentiostatic technique was employed. Linear growth from a depth of 70 μm at pH 3.8, 80 °C was demonstrated. The technique offers promise for the study of inhibition of pitting.

  17. Carbon nanotubes: from nano test tube to nano-reactor.

    Science.gov (United States)

    Khlobystov, Andrei N

    2011-12-27

    Confinement of molecules and atoms inside carbon nanotubes provides a powerful strategy for studying structures and chemical properties of individual molecules at the nanoscale. In this issue of ACS Nano, Allen et al. explore the nanotube as a template leading to the formation of unusual supramolecular and covalent structures. The potential of carbon nanotubes as reactors for synthesis on the nano- and macroscales is discussed in light of recent studies.

  18. Environmentally-Assisted Cracking of Low-Alloy Reactor Pressure Vessel Steels under Boiling Water Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P.; Ritter, S

    2002-02-01

    The present report summarizes the experimental work performed by PSI on the environmentally-assisted cracking (EAC) of low-alloy steels (LAS) in the frame of the RIKORR-project during the period from January 2000 to August 2001. Within this project, the EAC crack growth behaviour of different low-alloy reactor pressure vessel (RPV) steels, weld filler and weld heat-affected zone materials is investigated under simulated transient and steady-state BWR/NWC power operation conditions. The EAC crack growth behaviour of different low-alloy RPV steels was characterized by slow rising load (SRL) / low-frequency corrosion fatigue (LFCF) and constant load tests with pre-cracked fracture mechanics specimens in oxygenated high-temperature water at temperatures of either 288, 250, 200 or 150 C. These tests revealed the following important interim results: Under low-flow and highly oxidizing (ECP >= 100 mV SHE) conditions, the ASME XI 'wet' reference fatigue crack growth curve could be significantly exceeded by cyclic fatigue loading at low frequencies (<0.001 Hz), at high and low load-ratios R, and by ripple loading near to DKth fatigue thresholds. The BWR VIP 60 SCC disposition lines may be significantly or slightly exceeded (even in steels with a low sulphur content) in the case of small load fluctuations at high load ratios (ripple loading) or at intermediate temperatures (200 -250 C) in RPV materials, which show a distinct susceptibility to dynamic strain ageing (DSA). (author)

  19. Decontamination and decarburization of stainless and carbon steel by melt refining

    International Nuclear Information System (INIS)

    Mizia, R.E.; Worcester, S.A.; Twidwell, L.G.; Webber, D.; Paolini, D.J.; Weldon, T.A.

    1996-01-01

    With many nuclear reactors and facilities being decommissioned in the next ten to twenty years the concern for handling and storing Radioactive Scrap Metal (RSM) is growing. Upon direction of the DOE Office of Environmental Restoration and Waste Management, Lockheed Idaho Technology Company (LITCO) is developing technologies for the conditioning of spent fuels and high-level wastes for interim storage and repository acceptance, including the recycling of Radioactive Scrap Metals (RSM) for beneficial reuse with the DOE complex. In February 1993, Montana Tech of the University of Montana was contracted to develop and demonstrate technologies for the decontamination of stainless steel RSM. The general objectives of the Montana Tech research program included conducting a literature survey, performing laboratory scale melt refining experiments to optimize decontaminating slag compositions, performing an analysis of preferred melting techniques, coordinating pilot scale and commercial scale demonstrations, and producing sufficient quantities of surrogate-containing material for all of the laboratory, pilot and commercial scale test programs. Later on, the program was expanded to include decontamination of carbon steel RSM. Each research program has been completed, and results are presented in this report

  20. Evaluation on Safety of Stainless Steels in Chemical Decontamination Process with Immersion Type of Reactor Coolant Pump for Nuclear Reactor

    International Nuclear Information System (INIS)

    Kim, Seong Jong; Han, Min Su; Jang, Seok Ki; Kim, Ki Joon

    2011-01-01

    Due to commercialization of nuclear power, most countries have taken interest in decontamination process of nuclear power plant and tried to develop a optimum process. Because open literature of the decontamination process are rare, it is hard to obtain skills on decontamination of foreign country and it is necessarily to develop proper chemical decontamination process system in Korea. In this study, applicable possibility in chemical decontamination for reactor coolant pump (RCP) was investigated for the various stainless steels. The stainless steel (STS) 304 showed the best electrochemical properties for corrosion resistance and the lowest weight loss ratio in chemical decontamination process with immersion type than other materials. However, the pitting corrosion was generated in both STS 415 and STS 431 with the increasing numbers of cycle. The intergranular corrosion in STS 431 was sporadically observed. The sizes of their pitting corrosion also increased with increasing cycle numbers

  1. Underwater cutting of stainless steel plate and pipe for dismantling reactor pressure vessels

    International Nuclear Information System (INIS)

    Hamasaki, M.; Tateiwa, F.; Kanatani, F.; Yamashita, S.

    1982-01-01

    A consumable electrode water jet cutting technique is described. Satisfactory underwater cutting of 80mm stainless steel plate using a current of 2000A and at a water depth of 200mm has been demonstrated. The electrical requirements for this arc welding method applied to cutting were found to be approximately one third those required for conventional plasma arc cutting for the same thickness plate. An application of this technique might be found in the dismantling of atomic reactor pressure vessels, and parts of commercial atomic reactors. (author)

  2. Progress in Investigation of WWER-440 Reactor Pressure Vessel Steel by Gamma and Moessbauer Spectroscopy

    International Nuclear Information System (INIS)

    Hascik, J.; Slugen, V.; Lipka, J.; Hinca, R.; Toth, I.; Groene, R.; Uvacik, P.; Kupca, L.

    1998-01-01

    Gamma spectroscopic analyse and first experimental results of original irradiated reactor pressure vessel surveillance specimens are discussed in. In 1994, the new ''Extended Surveillance Specimen Program for nuclear Reactor Material Study'' was started in collaboration with the nuclear power plants (NPP) V-2 Bohunice (Slovakia). The first batch of MS samples (after 1 year, which is equivalent to 5 years of loading RPV-steel) was measured and interpreted using the new four components approach with the aim to observe microstructural changes due to thermal and neutron treatment resulting from operating conditions in NPP. The systematic changes in the relative areas of Moessbauer spectra components were observed. (author)

  3. Energy use and carbon dioxide emissions in the steel sector in key developing countries

    Energy Technology Data Exchange (ETDEWEB)

    Price, L.K.; Phylipsen, G.J.M.; Worrell, E.

    2001-04-01

    Iron and steel production consumes enormous quantities of energy, especially in developing countries where outdated, inefficient technologies are still used to produce iron and steel. Carbon dioxide emissions from steel production, which range between 5 and 15% of total country emissions in key developing countries (Brazil, China, India, Mexico, and South Africa), will continue to grow as these countries develop and as demand for steel products such as materials, automobiles, and appliances increases. In this report, we describe the key steel processes, discuss typical energy-intensity values for these processes, review historical trends in iron and steel production by process in five key developing countries, describe the steel industry in each of the five key developing countries, present international comparisons of energy use and carbon dioxide emissions among these countries, and provide our assessment of the technical potential to reduce these emissions based on best-practice benchmarking. Using a best practice benchmark, we find that significant savings, in the range of 33% to 49% of total primary energy used to produce steel, are technically possible in these countries. Similarly, we find that the technical potential for reducing intensities of carbon dioxide emissions ranges between 26% and 49% of total carbon dioxide emissions from steel production in these countries.

  4. Influence of Heat Treatments on the Corrosion Resistance of Medium -Carbon Steel using Sulfuric Spring Water

    Directory of Open Access Journals (Sweden)

    Ikhlas Basheer

    2015-02-01

    Full Text Available The corrosion is one of the important problems that may be occur to the parts of machinery and equipment after manufactured and when used as a result of exposure to corrosive media. Plain-carbon steel is considered as one of the most common minerals used in industrial applications. Some of heat treatments can have direct effect on the corrosion rate of steel by building up galvanic corrosion cells between its microscopic phases. Therefore, to adopt one of kinds of the plain-carbon steel and the most commonly used in industry to be study subject, that is medium carbon steel and took samples of this steel has been treated thermally in three methods which the normalising, annealing, and hardening .The corrosive media used in the research is Sulfuric Spring, it contains many chemical compounds to show its influence on the corrosion of steel. The weight loss method is used to determine corrosion rate and to compare between the results obtained, show that the greatest corrosion resistance of the annealed steel and the corrosion resistance of the hardened steel is the lowest while the corrosion  resistance of the normalised steel is in-between them.         Calcium carbonate was formed on the metal surface which acts as an isolating layer which decrease corrosion rate with time

  5. Analysis of stirred-tank carbonation reactors

    International Nuclear Information System (INIS)

    Sheppard, N.F.; Rizo-Patron, R.C.; Sun, W.H.

    1978-01-01

    The removal of CO 2 from air in a calcium hydroxide slurry-agitated reactor was investigated to aid the design of such vessels. Gas-liquid interfacial areas were calculated using theoretical rate expression and experimental data at specific operating conditions. A correlation for interfacial areas was then determined as a function of impeller speed, impeller diameter, gas flow rate, and concentration of the slurry. Decontamination factors were also determined

  6. Mechanism of fatigue crack initiation in austenitic stainless steels in light water reactor environments

    International Nuclear Information System (INIS)

    Chopra, O.K.; Shack, W.J.; Muscara, J.

    2003-01-01

    This paper examines the mechanism of fatigue crack initiation in austenitic stainless steels (SSs) in light water reactor (LWR) coolant environments. The effects of key material and loading variables on the fatigue lives of wrought and cast austenitic SSs in air and LWR environments have been evaluated. The influence of reactor coolant environments on the formation and growth of fatigue cracks in polished smooth SS specimens is discussed. The results indicate that the fatigue lives of these steels are decreased primarily by the effects of the environment on the growth of cracks <200 μm and, to a lesser extent, on enhanced growth rates of longer cracks. The fracture morphology in the specimens has been characterized. Exploratory fatigue tests were conducted to study the effects of surface micropits or minor differences in the surface oxide on fatigue crack initiation. (author)

  7. Research and development of austenitic stainless steels for fusion reactors, (1)

    International Nuclear Information System (INIS)

    1984-11-01

    In the alloy development for the first wall of blanket structure of the fusion experimental reactor and a subsequent reactor of Tokamak type, the prime candidate alloy (PCA) and reference steels were melted and examined on fundamental materials properties under a contract between JAERI and iron and steel companies, and under NRIM-JAERI collaborative work during the fiscal years of 1981 and 1982. All the alloys showed reasonable performance on mechanical properties, phase stability at elevated temperatures and weldability. The PCA has been proved to be used in controlled water-coolant environment. As to the welding of the PCA, welding rods suitable for TIG and covered arc welding have been selected from several candidate rods. (author)

  8. Analysis of lime-slurry stirred tank carbonation reactor

    International Nuclear Information System (INIS)

    McAleese, J.P.; Belt, B.A.; Datesh, J.R.; Shaeffer, M.C.

    1977-01-01

    Gas residence time distributions were determined for a stirred tank carbonation reactor. Empirical correlations for the first and second moments of the residence time distribution (RTD) curves as functions of flow rates and impeller speeds were obtained. Decontamination factors for 85 Kr were measured

  9. Carbon dioxide hydrate formation in a fixed-bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fan, S.; Lang, X. [South China Univ. of Technology, Guangzhou (China). Key Laboratory of Enhanced Heat Transfer and Energy Conservation; Wang, Y.; Liang, D. [Chinese Academy of Sciences, Guangzhou (China). Guangzhou Inst. of Energy Conversion and Guangzhou Center of Natural Gas Hydrate; Sun, X.; Jurcik, B. [Air Liquide Laboratories, Tsukuba (Japan)

    2008-07-01

    Gas hydrates are thermodynamically stable at high pressures and near the freezing temperature of pure water. Methane hydrates occur naturally in sediments in the deep oceans and permafrost regions and constitute an extensive hydrocarbon reservoir. Carbon dioxide (CO{sub 2}) hydrates are of interest as a medium for marine sequestration of anthropogenic carbon dioxide. Sequestering CO{sub 2} as hydrate has potential advantages over most methods proposed for marine CO{sub 2} sequestration. Because this technique requires a shallower depth of injection when compared with other ocean sequestration methods, the costs of CO{sub 2} hydrate sequestration may be lower. Many studies have successfully used different continuous reactor designs to produce CO{sub 2} hydrates in both laboratory and field settings. This paper discussed a study that involved the design and construction of a fixed-bed reactor for simulation of hydrate formation system. Water, river sands and carbon dioxide were used to simulate the seep kind of hydrate formation. Carbon dioxide gas was distributed as small bubbles to enter from the bottom of the fixed-bed reactor. The paper discussed the experimental data and presented a diagram of the gas hydrate reactor system. The morphology as well as the reaction characters of CO{sub 2} hydrate was presented in detail. The results were discussed in terms of experimental phenomena and hydrate formation rate. A mathematical model was proposed for describing the process. 17 refs., 7 figs.

  10. Analysis of barium hydroxide and calcium hydroxide slurry carbonation reactors

    International Nuclear Information System (INIS)

    Patch, K.D.; Hart, R.P.; Schumacher, W.A.

    1980-05-01

    The removal of CO 2 from air was investigated by using a continuous-agitated-slurry carbonation reactor containing either barium hydroxide [Ba(OH) 2 ] or calcium hydroxide [Ca(OH) 2 ]. Such a process would be applied to scrub 14 CO 2 from stack gases at nuclear-fuel reprocessing plants. Decontamination factors were characterized for reactor conditions which could alter hydrodynamic behavior. An attempt was made to characterize reactor performance with models assuming both plug flow and various degrees of backmixing in the gas phase. The Ba(OH) 2 slurry enabled increased conversion, but apparently the process was controlled under some conditions by phenomena differing from those observed for carbonation by Ca(OH) 2 . Overall reaction mechanisms are postulated

  11. Fatigue Life Estimation of Medium-Carbon Steel with Different Surface Roughness

    Directory of Open Access Journals (Sweden)

    Changyou Li

    2017-03-01

    Full Text Available Medium-carbon steel is commonly used for the rail, wire ropes, tire cord, cold heading, forging steels, cold finished steel bars, machinable steel and so on. Its fatigue behavior analysis and fatigue life estimation play an important role in the machinery industry. In this paper, the estimation of fatigue life of medium-carbon steel with different surface roughness using established S-N and P-S-N curves is presented. To estimate the fatigue life, the effect of the average surface roughness on the fatigue life of medium-carbon steel has been investigated using 75 fatigue tests in three groups with average surface roughness (Ra: 0.4 μm, 0.8 μm, and 1.6 μm, respectively. S-N curves and P-S-N curves have been established based on the fatigue tests. The fatigue life of medium-carbon steel is then estimated based on Tanaka-Mura crack initiation life model, the crack propagation life model using Paris law, and material constants of the S-N curves. Six more fatigue tests have been conducted to validate the presented fatigue life estimation formulation. The experimental results have shown that the presented model could estimate well the mean fatigue life of medium-carbon steel with different surface roughness.

  12. Development of ferritic steels for steam generators of fast breeder reactors

    International Nuclear Information System (INIS)

    Nguyen-Thanh; Vigneron, G.; Vanderschaeghe, A.

    1988-01-01

    STEIN INDUSTRIE, a manufacturer of equipment for the conventional and nuclear power industry, has built up expertise in the use of Cr-Mo steels used at high temperatures. The main ferritic steels developed were 10 CD 9-10 (AFNOR), Z10 CDNb V 9-2 (AFNOR), X 20 Cr Mo V 12-1 (DIN) and ASTM Grade 9.1. For the fast breeder reactor system, STEIN INDUSTRIE proposes the use of these steels in the construction of steam generators. The wide programme of development undertaken by STEIN INDUSTRIE is aimed at the following main subjects: - characterization of materials - welding and bending tests - studies of special junctions. This article reports the results obtained

  13. Some aspects of the utilization of zicaloy and austenitic steel as cladding material for PWR reactor fuel rods

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.; Perrotta, J.A.

    1985-01-01

    The behaviour under irradiation of fuel rods for light water reactors was simulated by using fuel performance codes. Two types of cladding were analyzed: zircaloy and austenitic stainless steel. The fuel performance codes, originally made for zircaloy cladding, were adapted for austenitic stainless steel. The simulation results for the two types of cladding are presented, compared and discussed. (F.E.) [pt

  14. SURFACE ROUGHNESS AND CUTTING FORCES IN CRYOGENIC TURNING OF CARBON STEEL

    Directory of Open Access Journals (Sweden)

    T. C. YAP

    2015-07-01

    Full Text Available The effect of cryogenic liquid nitrogen on surface roughness, cutting forces, and friction coefficient of the machined surface when machining of carbon steel S45C in wet, dry and cryogenic condition was studied through experiments. The experimental results show that machining with liquid nitrogen increases the cutting forces, reduces the friction coefficient, and improves the chips produced. Beside this, conventional machining with cutting fluid is still the most suitable method to produce good surface in high speed machining of carbon steel S45C whereas dry machining produced best surface roughness in low speed machining. Cryogenic machining is not able to replace conventional cutting fluid in turning carbon steel.

  15. Experimental study of neutron streaming through steel-walled annular ducts in reactor shields

    International Nuclear Information System (INIS)

    Toshimas, M.; Nobuo, S.

    1983-01-01

    For the purpose of providing experimental data to assess neutron streaming calculations, neutron flux measurements were performed along the axes of the steel-walled annular ducts set up in a water shield of the pool-type reactor JRR-4. An annular duct simulated the air gap around the main coolant pipe. Another duct simulated the streaming path around the primary circulating pump of the integrated-type marine reactor. A 90-deg bend annular duct was also studied. In a set of measurements, the distance Z between the core center and the duct axis and the annular gap width delta were taken as parameters, that is, Z = 0, 80, and 160 cm and delta = 2.2, 4.7, and 10.1 cm. The reaction rates and the fluxes measured by the activation method are given in terms of absolute magnitude within an accuracy of + or - 30%. An empirical formula is derived based on those measured data, which describes the axial distribution of the neutron flux in the steel-walled annular duct in reactor shields. It is expressed by a simple function of the axial distance in units of the square root of the line-of-sight area, S /SUB l/ . The accuracy of the formula is examined by taking into account the duct location with respect to the reactor core, the neutron energy, the steel wall thickness, and the media outside of the steel wall. The accuracy of the formula is, in general, <30% in the axial distance between 3√S /SUB l/ and 30√S /SUB l/

  16. Effect of decontamination on oxidation of austenitic stainless steel in reactor conditions

    International Nuclear Information System (INIS)

    Starkman, T.

    1984-07-01

    Austenitic stainless steels were oxidized in static autoclaves in light water reactor conditions. After the autoclave treatments the specimens were decontaminated with the aid of alkaline potassium permanganate (AP) and oxalic and citric acid (CITROX) as well as electrochemically in H 3 PO 4 . Alternating oxidation and decontamination tests were performed. An elemental analysis of the surfaces of the specimens was carried out by electron spectroscopy. Changes in structures and thicknesses of the oxide layers were observed. (author)

  17. Reactor pressure vessel steels[1997 Scientific Report of the Belgian Nuclear Research Centre

    Energy Technology Data Exchange (ETDEWEB)

    Van De Velde, J.; Fabry, A.; Van Walle, E.; Chaouuadi, R.

    1998-07-01

    Research and development activities related to reactor pressure vessel steels during 1997 are reported. The objectives of activities of the Belgian Nuclear Research Centre SCK/CEN in this domain are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate a methodology on a broad database; (3) to achieve regulatory acceptance and industrial use.

  18. Study on prediction model of irradiation embrittlement for reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Wang Rongshan; Xu Chaoliang; Huang Ping; Liu Xiangbing; Ren Ai; Chen Jun; Li Chengliang

    2014-01-01

    The study on prediction model of irradiation embrittlement for reactor pres- sure vessel (RPV) steel is an important method for long term operation. According to the deep analysis of the previous prediction models developed worldwide, the drawbacks of these models were given and a new irradiation embrittlement prediction model PMIE-2012 was developed. A corresponding reliability assessment was carried out by irradiation surveillance data. The assessment results show that the PMIE-2012 have a high reliability and accuracy on irradiation embrittlement prediction. (authors)

  19. Compatibility of 316L stainless steel with tritium breeders for fusion reactors

    International Nuclear Information System (INIS)

    Broc, M.; Fauvet, P.; Flament, T.; Sannier, J.

    1986-06-01

    Compatibility problems with structural materials are a concern for the choice of the tritium breeder for fusion reactors. In the frame of the European Programme on Fusion Technology, two types of blankets are considered: liquid (eutectic lithium-lead alloy at 0.68 wt % Li: 17Li83Pb) and solid (lithium aluminate or silicate) breeders. This paper is devoted to compatibility studies of 316L stainless steel with 17Li83Pb alloy and γ-LiA10 2 ceramic

  20. Comparative Analysis of Carbon Plasma in Arc and RF Reactors

    International Nuclear Information System (INIS)

    Todorovic-Markovic, B.; Markovic, Z.; Mohai, I.; Szepvolgyi, J.

    2004-01-01

    Results on studies of molecular spectra emitted in the initial stages of fullerene formation during the processing of graphite powder in induction RF reactor and evaporation of graphite electrodes in arc reactor are presented in this paper. It was found that C2 radicals were dominant molecular species in both plasmas. C2 radicals have an important role in the process of fullerene synthesis. The rotational-vibrational temperatures of C2 and CN species were calculated by fitting the experimental spectra to the simulated ones. The results of optical emission study of C2 radicals generated in carbon arc plasma have shown that rotational temperature of C2 species depends on carbon concentration and current intensity significantly. The optical emission study of induction RF plasma and SEM analysis of graphite powder before and after plasma treatment have shown that evaporation of the processed graphite powder depends on feed rate and composition of gas phase significantly. Based on the obtained results, it was concluded that in the plasma region CN radicals could be formed by the reaction of C2 species with atomic nitrogen at smaller loads. At larger feed rate of graphite powder, CN species were produced by surface reaction of the hot carbon particles with nitrogen atoms. The presence of nitrogen in induction RF plasma reduces the fullerene yield significantly. The fullerene yield obtained in two different reactors was: 13% in arc reactor and 4.1% in induction RF reactor. However, the fullerene production rate was higher in induction RF reactor-6.4 g/h versus 1.7 g/h in arc reactor

  1. Further evaluation of creep-fatigue life prediction methods for low-carbon nitrogen-added 316 stainless steel

    International Nuclear Information System (INIS)

    Takahashi, Y.

    1999-01-01

    Low-carbon, medium-nitrogen 316 stainless steel is a principal candidate for a main structural material of a demonstration fast breeder reactor plant in Japan. A number of long-term creep tests and creep-fatigue tests have been conducted for four products of this steel. Two representative creep-fatigue life prediction methods, i.e., time fraction rule and ductility exhaustion method were applied. Total stress relaxation behavior was simulated well by an addition of a viscous strain term to the conventional (primary plus secondary) creep strain, but only the letter was assumed to contribute to creep damage in the ductility exhaustion method. The present ductility exhaustion approach was found to have very good accuracy in creep-fatigue life prediction for all materials tested, while the time fraction rule tended to overpredict failure life as large as a factor of 30. Discussion was made on the reason for this notable difference

  2. Fluidized bed reactor for processing particles coated with carbon

    International Nuclear Information System (INIS)

    Marschollek, M.; Simon, W.; Walter, C.

    1978-01-01

    The carbon coating of production returns of these particles first has to be removed before the heavy metal core released can be reprocessed. For reasons of criticality, removal of burnt-up particles downwards must be possible in the fluidized bed reactor even if the reactor diameter is greater than 800 mm, and the material temperatures must not exceed 650 0 C. It consists of an upper cylindrical and a lower conical part, where, according to the invention, the gas distributor heads in the conical part are situated in several planes above one another for the fluidisation and combustion gas and where they are evently distributed over the reactor crossection, so that an even flow profile is achieved over the reactor cross section. (HP) [de

  3. Neutron irradiation embrittlement of reactor pressure vessel steel 20 MnMoNi55 weld

    International Nuclear Information System (INIS)

    Ghoneim, M.M.

    1987-05-01

    The effect of neutron irradiation on the mechanical and fracture properties of an 'improved' 20 MnMoNi 55 Pressure Vessel Steel (PVS) weld was investigated. In addition to very low residual element content, especially Cu (0.035 wt.%), and relatively higher Ni content (0.9 wt.%), this steel has higher strength (30% more) than the steels used currently in nuclear reactor pressure vessels. The material was irradiated to 3.5x10 19 and 7x10 19 n/cm 2 (E > 1 Mev) at 290 0 C and 2.5x10 19 n/cm 2 (E > 1 MeV) at 160 0 C in FRJ-1 and FRJ-2 research reactors at KFA, Juelich, F.R.G. Test methods used in the evaluation included instrumented impact testing of standard and precracked Charpy specimens, tensile, and fracture toughness testing. Instrumented impact testing provided load and energy vs. time (deflection) data in addition to energy absorption data. The results indicated that the investigated high strength improved steel is more resistant to irradiation induced embrittlement than conventional PVSs. (orig./IHOE)

  4. Comparison of material property specifications of austenitic steels in fast breeder reactor technology

    International Nuclear Information System (INIS)

    Vanderborck, Y.; Van Mulders, E.

    1985-01-01

    Austenitic stainless steels are very widely used in components for European Fast Breeder Reactors. The Activity Group Nr.3 ''Materials'', within Working Group ''Codes and Standards'' of the Fast Reactor Co-Ordination Committee of the European Communities, has decided to initiate a study to compare the material property specifications of the austenitic stainless steel used in the European Fast Breeder Technology. Hence, this study would allow one to view rapidly the designation of a particular steel grade in different European countries and to compare given property values for a same grade. There were dissimilarities, differences or voids appear, it could lead to an attempt to complete and/or to uniformize the nationally given values, so that on a practical level interchangeability, availability and use ease design and construction work. A selection of the materials and of their properties has been made by the Working Group. Materials examined are Stainless Steel AISI 304, 304 L, 304 LN, 316, 316 L, 316 LN, 316''Ti stab.'', 316''Nb stab''., 321, 347

  5. Acoustic emission during the elastic-plastic deformation of low alloy reactor pressure vessel steels. I

    International Nuclear Information System (INIS)

    Holt, J.; Goddard, D.J.

    1980-01-01

    Measurements of the acoustic emission behaviour of A533B and C-Mn low alloy reactor pressure vessel steels subjected to uniaxial tensile deformation are described. The effects on the emission activity of the rolling plane orientation and the carbide morphology were examined. Detailed discussions are given of the stress dependence of the emission activity below yield and of its recovery by annealing at the stress relief temperature. It is shown that the dominant emission source is the same in both steels and is associated with inclusions, such as MnS, elongated by the rolling process, the carbide morphology being relatively unimportant. A criterion for the occurrence of an emission is obtained which is directly analogous to the general criterion for yielding. It is also shown that a large fraction, at least, of the emission activity arises from a recoverable process such as localized yielding around inclusions or limited inclusion decohesion and not from inclusion fracture. Low activity in C-Mn steel taken from reactor pressure vessels, previously attributed to spheroidization of carbides, is shown to be due to the limited acoustic recovery of these relatively high sulphur content steels when annealed at the stress relief temperature. It is concluded that the limited amplitudes of these emissions during deformation severely restrict their potential application in practice. (Auth.)

  6. Isothermal and thermal-mechanical fatigue of VVER-440 reactor pressure vessel steels

    Science.gov (United States)

    Fekete, Balazs; Trampus, Peter

    2015-09-01

    The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of VVER-440 reactor pressure vessel under completely reserved total strain controlled low cycle fatigue tests were investigated. An advanced test facility was developed for GLEEBLE-3800 physical simulator which was able to perform thermomechanical fatigue experiments under in-service conditions of VVER nuclear reactors. The low cycle fatigue results were evaluated with the plastic strain based Coffin-Manson law, and plastic strain energy based model as well. It was shown that both methods are able to predict the fatigue life of reactor pressure vessel steels accurately. Interrupted fatigue tests were also carried out to investigate the kinetic of the fatigue evolution of the materials. On these samples microstructural evaluation by TEM was performed. The investigated low cycle fatigue behavior can provide reference for remaining life assessment and lifetime extension analysis.

  7. Ultra-large size austenitic stainless steel forgings for fast breeder reactor 'Monju'

    International Nuclear Information System (INIS)

    Tsukada, Hisashi; Suzuki, Komei; Sato, Ikuo; Miura, Ritsu.

    1988-01-01

    The large SUS 304 austenitic stainless steel forgings for the reactor vessel of the prototype FBR 'Monju' of 280 MWe output were successfully manufactured. The reactor vessel contains the heart of the reactor and sodium coolant at 530 deg C, and its inside diameter is about 7 m, and height is about 18 m. It is composed of 12 large forgings, that is, very thick flanges and shalls made by ring forging and an end plate made by disk forging and hot forming, using a special press machine. The manufacture of these large forgings utilized the results of the basic test on the material properties in high temperature environment and the effect that the manufacturing factors exert on the material properties and the results of the development of manufacturing techniques for superlarge forgings. The problems were the manufacturing techniques for the large ingots of 250 t class of high purity, the hot working techniques for stainless steel of fine grain size, the forging techniques for superlarge rings and disks, and the machining techniques of high precision for particularly large diameter, thin wall rings. The manufacture of these large stainless steel forgings is reported. (Kako, I.)

  8. Sustainable Steel Carburization by Using Snack Packaging Plastic Waste as Carbon Resources

    Directory of Open Access Journals (Sweden)

    Songyan Yin

    2018-01-01

    Full Text Available In recent years, the research regarding waste conversion to resources technology has attracted growing attention with the continued increase of waste accumulation issues and rapid depletion of natural resources. However, the study, with respect to utilizing plastics waste as carbon resources in the metals industry, is still limited. In this work, an environmentally friendly approach to utilize snack packaging plastic waste as a valuable carbon resources for steel carburization is investigated. At high temperature, plastic waste could be subject to pyrolytic gasification and decompose into small molecular hydrocarbon gaseous products which have the potential to be used as carburization agents for steel. When heating some snack packaging plastic waste and a steel sample together at the carburization temperature, a considerable amount of carbon-rich reducing gases, like methane, could be liberated from the plastic waste and absorbed by the steel sample as a carbon precursor for carburization. The resulting carburization effect on steel was investigated by optical microscopy, scanning electron microscopy, electron probe microanalyzer, and X-ray photoelectron spectrometer techniques. These investigation results all showed that snack packaging plastic waste could work effectively as a valuable carbon resource for steel carburization leading to a significant increase of surface carbon content and the corresponding microstructure evolution in steel.

  9. Galvanic Interaction between Chalcopyrite and Pyrite with Low Alloy and High Carbon Chromium Steel Ball

    Directory of Open Access Journals (Sweden)

    Asghar Azizi

    2013-01-01

    Full Text Available This study was aimed to investigate the galvanic interaction between pyrite and chalcopyrite with two types of grinding media (low alloy and high carbon chromium steel ball in grinding of a porphyry copper sulphide ore. Results indicated that injection of different gases into mill altered the oxidation-reduction environment during grinding. High carbon chromium steel ball under nitrogen gas has the lowest galvanic current, and low alloy steel ball under oxygen gas had the highest galvanic current. Also, results showed that the media is anodic relative to pyrite and chalcopyrite, and therefore pyrite or chalcopyrite with a higher rest potential acted as the cathode, whilst the grinding media with a lower rest potential acted as the anode, when they are electrochemically contacted. It was also found that low alloy steel under oxygen produced the highest amount of EDTA extractable iron in the slurry, whilst high carbon chromium steel under nitrogen atmosphere led to the lowest amount.

  10. Hydrogen degradation of 21-6-9 and medium carbon steel by disc pressure test

    International Nuclear Information System (INIS)

    Zhou, D.H.; Zhou, W.X.; Xu, Z.L.

    1986-01-01

    This paper reports the method of disc pressure test and the results for 21-6-9 stainless steel and medium carbon steel in hydrogen gas with different pressures and time of storage. The results show the hydrogen induced degradation of these two kinds of steel. An attempt was made to establish an index which uses variation of area of deformed disc to determine the degradation of ductility in a hydrogen environment. (orig.)

  11. Stress state evaluation in low carbon and TRIP steels by magnetic permeability

    International Nuclear Information System (INIS)

    Kouli, M.-E.; Giannakis, M

    2016-01-01

    Magnetic permeability is an indicative factor for the steel health monitoring. The measurements of magnetic permeability lead to the evaluation of the stress state of any ferromagnetic steel. The magnetic permeability measurements were conducted on low carbon and TRIP steel samples, which were subjected to both tensile and compressive stresses. The results indicated a direct correlation of the magnetic permeability with the mechanical properties, the stress state and the microstructural features of the examined samples. (paper)

  12. Carbonation of steel slag for CO2 sequestration: Leaching of products and reaction mechanisms

    NARCIS (Netherlands)

    Huijgen, W.J.J.; Comans, R.N.J.

    2006-01-01

    Carbonation of industrial alkaline residues can be used as a CO2 sequestration technology to reduce carbon dioxide emissions. In this study, steel slag samples were carbonated to a varying extent. Leaching experiments and geochemical modeling were used to identify solubility-controlling processes of

  13. Reinforcement steel corrosion in passive state and by carbonation: Consideration of galvanic currents and interface steel - concrete defaults

    International Nuclear Information System (INIS)

    Nasser, A.

    2010-01-01

    This thesis aims to study the durability of nuclear waste deep storage structures. The work carried out is essentially an experimental study, and focuses on the corrosion of steel in the passive state with aerated or non-aerated conditions on the one hand, and the corrosion of steel in carbonated concrete during the propagation phase on the other hand. Indeed, the pore solution of concrete in contact with the metal is alkaline (pH between 12 and 13). Under these conditions, steel reinforced concrete remains passive by forming a stable and protective oxide layer (corrosion of steel in the passive state). This passive layer limits the steel corrosion rate at very low values (negligible on a short life time) but not null. For the nuclear waste storage structures due to a very long life time (up to several hundred years), this low corrosion rate can become a risk. Therefore, it is necessary to study the evolution of the oxide layer growth over time. The objectives of the thesis are to study the influence of the steel-concrete interface quality on reinforcement corrosion in passive and active state, and the possible occurrence of galvanic corrosion currents between different reinforcement steel areas. (author)

  14. Role of cavity formation in SCC of cold worked carbon steel in high-temperature water. Part 2. Study of crack initiation behavior

    International Nuclear Information System (INIS)

    Yamada, Takuyo; Aoki, Masanori; Miyamoto, Tomoki; Arioka, Koji

    2013-01-01

    To consider the role of cavity formation in stress corrosion cracking (SCC) of cold worked (CW) carbon steel in high-temperature water, SCC and creep growth (part 1) and initiation (part 2) tests were performed. The part 2 crack initiation tests used blunt notched compact tension (CT) type specimens of CW carbon steel exposed under the static load condition in hydrogenated pure water and in air in the range of temperatures between 360 and 450°C. Inter-granular (IG) crack initiation was observed both in water and in air even in static load condition when steel specimens had been cold worked. 1/T type temperature dependencies of initiation times were observed for CW carbon steel, and the crack initiation times in an operating pressurized heavy water reactor, PHWR (Pt Lepreau) seemed to lie on the extrapolated line of the experimental results. Cavities were identified at the grain boundaries near the bottom of a notch (highly stressed location) before cracks initiated both in water and air. The cavities were probably formed by the condensation of vacancies and they affected the bond strength of the grain boundaries. To assess the mechanism of IGSCC initiation in high temperature water, the diffusion of vacancies driven by stress gradients was studied using a specially designed CT specimen. As a model for IGSCC in CW carbon steel in high temperature water, it was concluded that the formation of cavities from the collapse of vacancies offers the best interpretation of the present data. (author)

  15. Measurement of carbon activity in sodium by Fe-Mn 20% alloy, and by strainless austenitic steel 304L and 316L

    International Nuclear Information System (INIS)

    Oberlin, C.; Saint Paul, P.; Baque, P.; Champeix, L.

    1980-01-01

    Precise knowledge of carbon activity in sodium used as coolant in fast breeder reactors, is essential for continuous survey of carburization-decarburization processes. Carbon activity can be periodically surveyed by measuring the carbon concentration or by hot trap like metal alloy strip placed in sodium loop. In fact, in equilibrium, activity of carbon in sodium is equal to the activity in metal alloy. Thus if the relation between concentration of carbon and it activity in the alloy is known, it is possible to estimate the activity of carbon in sodium. Materials to be used should have high solubility in carbon at the needed temperature. They should quickly attain equilibrium with sodium and they should not contain impurities that can affect the results. Materials chosen according to these criteria were Fe-Mn 20%, stainless austenitic steel AISI 304L and 316L

  16. Electrochemical formation of carbonated corrosion products on carbon steel in deaerated solutions

    International Nuclear Information System (INIS)

    Refait, Ph.; Bourdoiseau, J.A.; Jeannin, M.; Nguyen, D.D.

    2012-01-01

    Highlights: ► Green rust is electro-generated at low NaHCO 3 concentration (0.003 mol dm −3 ). ► Chukanovite and carbonated green rust are obtained in NaHCO 3 + Na 2 SO 4 deaerated electrolytes. ► The mechanisms of formation of carbonated corrosion products of carbon steel are specified. - Abstract: To investigate the nature and properties of carbonated rust layers, carbon steel electrodes were polarised anodically at a potential ∼100–200 mV higher than the open circuit potential in NaHCO 3 solutions (0.003, 0.1 and 1 mol dm −3 ) continuously deaerated by an argon flow. X-ray diffraction and μ-Raman spectroscopy were used to identify the electro-generated compounds. GR(CO 3 2− ) (=Fe II 4 Fe III 2 (OH) 12 CO 3 ·4H 2 O) is observed at 0.003 and 0.1 mol dm −3 NaHCO 3 whereas FeCO 3 is obtained at the largest concentration (1 mol dm −3 ). GR(CO 3 2− ) is accompanied by magnetite Fe 3 O 4 at the lowest NaHCO 3 concentration. The current density decreases to negligible values in each case, indicating that a passive film also forms independently of the nature of the carbonated compound. Experiments were performed similarly in solutions of NaHCO 3 and Na 2 SO 4 . Chukanovite Fe 2 (OH) 2 CO 3 could be obtained in solutions containing 0.03 mol dm −3 of each salt. In contrast with the results obtained in the solutions free of sulphate, the current density remains important during the formation of the rust layer

  17. 78 FR 16252 - Certain Hot-Rolled Carbon Steel Flat Products From India, Indonesia, and Thailand: Final Results...

    Science.gov (United States)

    2013-03-14

    ... Indonesia P.T. Krakatau Steel 10.21 All Others 10.21 Thailand Sahaviriya Steel Industries Public Company...] Certain Hot-Rolled Carbon Steel Flat Products From India, Indonesia, and Thailand: Final Results of... products (``HR steel'') from India, Indonesia, and Thailand pursuant to section 751(c) of the Tariff Act of...

  18. Carbon-14 in reactor plant water

    International Nuclear Information System (INIS)

    Knowles, G.K.

    1979-01-01

    The method for the analysis of 14 C in reactor plant water and various waste streams previously used at the Idaho National Engineering Laboratory has been shown to be ineffective for samples which contain organic compounds. The previous method consisted of acidification and refluxing of the sample, precipitation of the liberated CO 2 , and subsequent analysis by the liquid scintillation method. The method was simple but it did not convert all compounds containing 14 C in the sample to CO 2 . The new method, while it is based on the previous method, has been improved by employing a strong oxidant, potassium persulfate and silver nitrate, for more complete oxidation of the organics to CO 2 . The new method yields 14 C values that have typically been one to two orders of magnitude higher than the values obtained using the former method. This indicates that most of the 14 C present in the current reactor water samples being analyzed is associated with trace amounts of organics

  19. Fatigue of carbon and low-alloy steels in LWR environments

    International Nuclear Information System (INIS)

    Chopra, O.K.; Michaud, W.F.; Shack, W.J.

    1994-01-01

    Fatigue tests have been conducted on A106-Gr B carbon steel and A533-Gr B low-alloy steel to evaluate the effects of an oxygenated-water environment on the fatigue life of these steels. For both steels, environmental effects are modest in PWR water at all strain rates. Fatigue data in oxygenated water confirm the strong dependence of fatigue life on dissolved oxygen (DO) and strain rate. The effect of strain rate on fatigue life saturates at some low value, e.g., between 0.0004 and 0.001%/s in oxygenated water with ∼0.8 ppm DO. The data suggest that the saturation value of strain rate may vary with DO and sulfur content of the steel. Although the cyclic stress-strain and cyclic-hardening behavior of carbon and low-alloy steels is distinctly different, the degradation of fatigue life of these two steels with comparable sulfur levels is similar. The carbon steel exhibits pronounced dynamic strain aging, whereas strain-aging effects are modest in the low-alloy steel. Environmental effects on nucleation of fatigue crack have also been investigated. The results suggest that the high-temperature oxygenated water has little or not effect on crack nucleation

  20. Nondestructive characterization of embrittlement in reactor pressure vessel steels -- A feasibility study

    International Nuclear Information System (INIS)

    McHenry, H.I.; Alers, G.A.

    1998-01-01

    The Nuclear Regulatory Commission recently initiated a study by NIST to assess the feasibility of using physical-property measurements for evaluating radiation embrittlement in reactor pressure vessel (RPV) steels. Ultrasonic and magnetic measurements provide the most promising approaches for nondestructive characterization of RPV steels because elastic waves and magnetic fields can sense the microstructural changes that embrittle materials. The microstructural changes of particular interest are copper precipitation hardening, which is the likely cause of radiation embrittlement in RPV steels, and the loss of dislocation mobility that is an attribute of the ductile-to-brittle transition. Measurements were made on a 1% copper steel, ASTM grade A710, in the annealed, peak-aged and overaged conditions, and on an RPV steel, ASTM grade A533B. Nonlinear ultrasonic and micromagnetic techniques were the most promising measures of precipitation hardening. Ultrasonic velocity measurements and the magnetic properties associated with hysteresis-loop measurements were not particularly sensitive to either precipitation hardening or the ductile-to-brittle transition. Measurements of internal friction using trapped ultrasonic resonance modes detected energy losses due to the motion of pinned dislocations; however, the ultrasonic attenuation associated with these measurements was small compared to the attenuation caused by beam spreading that would occur in conventional ultrasonic testing of RPVs

  1. Design and analysis of reactor containment of steel-concrete composite laminated shell

    International Nuclear Information System (INIS)

    Ichikawa, K.

    1977-01-01

    Reinforced and prestressed concrete containments for reactors have been developed in order to avoid the difficulties of welding of steel containments encountered as their capacities have become large: growing thickness of steel shells gave rise to the requirement of stress relief at the construction sites. However, these concrete vessels also seem to face another difficulty: the lack of shearing resistance capacity. In order to improve the shearing resistance capacity of the containment vessel, while avoiding the difficulty of welding, a new scheme of containment consisting of steel-concrete laminated shell is being developed. In the main part of a cylindrical vessel, the shell consists of two layers of thin steel plates located at the inner and outer surfaces, and a layer of concrete core into which both the steel plates are anchored. In order to validate the feasibility and safety of this new design, the results of analysis on the basis of up-to-date design loads are presented. The results of model tests in 1:30 scale are also reported. (Auth.)

  2. Embrittlement recovery due to annealing of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Eason, E.D.; Wright, J.E.; Nelson, E.E.; Odette, G.R.; Mader, E.V.

    1996-01-01

    Embrittlement of reactor pressure vessels (RPVs) can be reduced by thermal annealing at temperatures higher than the normal operating conditions. Although such an annealing process has not been applied to any commercial plants in the United States, one US Army reactor, the BR3 plant in Belgium, and several plants in eastern Europe have been successfully annealed. All available Charpy annealing data were collected and analyzed in this project to develop quantitative models for estimating the recovery in 30 ft-lb (41 J) Charpy transition temperature and Charpy upper shelf energy over a range of potential annealing conditions. Pattern recognition, transformation analysis, residual studies, and the current understanding of the mechanisms involved in the annealing process were used to guide the selection of the most sensitive variables and correlating parameters and to determine the optimal functional forms for fitting the data. The resulting models were fitted by nonlinear least squares. The use of advanced tools, the larger data base now available, and insight from surrogate hardness data produced improved models for quantitative evaluation of the effects of annealing. The quality of models fitted in this project was evaluated by considering both the Charpy annealing data used for fitting and the surrogate hardness data base. The standard errors of the resulting recovery models relative to calibration data are comparable to the uncertainty in unirradiated Charpy data. This work also demonstrates that microhardness recovery is a good surrogate for transition temperature shift recovery and that there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes

  3. Corrosion Performance of Carbon Steel in Simulated Pore Solution in the Presence of Micelles

    NARCIS (Netherlands)

    Hu, J.; Koleva, D.A.; De Wit, J.H.W.; Kolev, H.; Van Breugel, K.

    2011-01-01

    This study presents the results on the investigation of the corrosion behavior of carbon steel in model alkaline medium in the presence of very low concentration of polymeric nanoaggregates [0.0024 wt % polyethylene oxide (PEO)113-b-PS70 micelles]. The steel electrodes were investigated in chloride

  4. 76 FR 64312 - Light-Walled Welded Rectangular Carbon Steel Tubing From Taiwan: Final Results of the Expedited...

    Science.gov (United States)

    2011-10-18

    ... Rectangular Carbon Steel Tubing From Taiwan: Final Results of the Expedited Sunset Review of the Antidumping... the antidumping duty order on light-walled welded rectangular carbon steel tubing from Taiwan pursuant... steel tubing from Taiwan pursuant to section 751(c) of the Act. See Initiation of Five-Year (``Sunset...

  5. 75 FR 29976 - Certain Cut-to-Length Carbon-Quality Steel Plate Products From Italy: Extension of the Final...

    Science.gov (United States)

    2010-05-28

    ... DEPARTMENT OF COMMERCE International Trade Administration [A-475-826] Certain Cut-to-Length Carbon-Quality Steel Plate Products From Italy: Extension of the Final Results of Antidumping Duty Administrative...-quality steel plate products from Italy. See Certain Cut-to-Length Carbon-Quality Steel Plate Products...

  6. 78 FR 29113 - Certain Cut-to-Length Carbon-Quality Steel Plate Products From the Republic of Korea: Final...

    Science.gov (United States)

    2013-05-17

    ...-Quality Steel Plate Products From the Republic of Korea: Final Results of Antidumping Duty Administrative... administrative review of the antidumping duty order on certain cut-to-length carbon-quality steel plate products... duty order on certain cut-to-length carbon-quality steel plate products from the Republic of Korea...

  7. 78 FR 4385 - Certain Cut-to-Length Carbon-Quality Steel Plate Products From the Republic of Korea: Preliminary...

    Science.gov (United States)

    2013-01-22

    ...-Quality Steel Plate Products From the Republic of Korea: Preliminary Results of Antidumping Duty... the antidumping duty order on certain cut-to- length carbon-quality steel plate products (CTL plate... Carbon-Quality Steel Plate Products from the Republic of Korea'' dated concurrently with this notice...

  8. 75 FR 33578 - Certain Welded Carbon Steel Standard Pipes and Tubes from India: Preliminary Results of...

    Science.gov (United States)

    2010-06-14

    ... Shrimp From Brazil, 69 FR 76910 (December 23, 2004), and accompanying Issues and Decision Memorandum at... Eleventh Administrative Review of the Antidumping Duty Order on Certain Corrosion-Resistant Carbon Steel...

  9. RPV-1: a first virtual reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, St.

    2005-01-01

    The presented work was aimed at building a first VTR (virtual test reactor) to simulate irradiation effects in pressure vessel steels of nuclear reactor. It mainly consisted in: - modeling the formation of the irradiation induced damage in such steels, as well as their plasticity behavior - selecting codes and models to carry out the simulations of the involved mechanisms. Since the main focus was to build a first tool (rather than a perfect tool), it was decided to use, as much as possible, existing codes and models in spite of their imperfections. - developing and parameterizing two missing codes: INCAS and DUPAIR. - proposing an architecture to link the selected codes and models. - constructing and validating the tool. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or transmit data. A user friendly Python interface facilitates the running of the simulations and the visualization of the results. RPV-1 relies on many simplifications and approximations and has to be considered as a prototype aimed at clearing the way. According to the functionalities targeted for RPV-1, the main weakness is a bad Ni and Mn sensitivity. However, the tool can already be used for many applications (understanding of experimental results, assessment of effects of material and irradiation conditions,....). (O.M.)

  10. Neutron irradiation effects in reactor pressure vessel steels and weldments. Working document

    International Nuclear Information System (INIS)

    1998-10-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. A separate abstract was prepared for the introduction and for each of the eleven chapters, which are: 1. Reactor Pressure Vessel Design, 2. Reactor Pressure Materials, 3. WWER Pressure Vessels, 4. Determination of Mechanical Properties, 5. Neutron Exposure, 6. Methodology of Irradiation Experiments, 7. Effect of Irradiation on Mechanical Properties, 8. Mechanisms of Irradiation Embrittlement, 9. Modelling of Irradiation Damage, 10. Annealing of Irradiation Damage, 11. Safety Assessment using Surveillance Programmes and Data Bases

  11. The metrological problems of irradiation embrittlement of reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Vodenicharov, S.; Kamenova, Ts.

    1993-01-01

    Neutron irradiation of reactor pressure vessel steels increases the T k -values of transition temperature from ductile to brittle fracture. This effect is very important in emergency situations, when the water cooling injection in the reactor results in high thermal gradients. In such cases there is a risk from the appearance of a brittle fracture with catastrophic crack propagation speed at relatively low stresses. That is why the T k -value determination is very important for the safe operation of the reactor systems. Some advanced experimental methods for T k -testing and control have been discussed in the present article and the standards of different countries have been compared. The methods applying subsize specimens and welding-restored specimens have been reviewed. (author)

  12. Emerging surface characterization techniques for carbon steel corrosion: a critical brief review

    OpenAIRE

    Dwivedi, D.; Lepkova, K.; Becker, T.

    2017-01-01

    Carbon steel is a preferred construction material in many industrial and domestic applications, including oil and gas pipelines, where corrosion mitigation using film-forming corrosion inhibitor formulations is a widely accepted method. This review identifies surface analytical techniques that are considered suitable for analysis of thin films at metallic substrates, but are yet to be applied to analysis of carbon steel surfaces in corrosive media or treated with corrosion inhibitors. The rev...

  13. Effect of small addition of Cr on stability of retained austenite in high carbon steel

    Energy Technology Data Exchange (ETDEWEB)

    Hossain, Rumana; Pahlevani, Farshid, E-mail: f.pahlevani@unsw.edu.au; Sahajwalla, Veena

    2017-03-15

    High carbon steels with dual phase structures of martensite and austenite have considerable potential for industrial application in high abrasion environments due to their hardness, strength and relatively low cost. To design cost effective high carbon steels with superior properties, it is crucial to identify the effect of Chromium (Cr) on the stability of retained austenite (RA) and to fully understand its effect on solid-state phase transition. This study addresses this important knowledge gap. Using standard compression tests on bulk material, quantitative X-ray diffraction analysis, nano-indentation on individual austenitic grains, transmission electron microscopy and electron backscatter diffraction–based orientation microscopy techniques, the authors investigated the effect of Cr on the microstructure, transformation behaviour and mechanical stability of retained austenite in high carbon steel, with varying Cr contents. The results revealed that increasing the Cr %, altered the morphology of the RA and increased its stability, consequently, increasing the critical pressure for martensitic transformation. This study has critically addressed the elastoplastic behaviour of retained austenite – and provides a deep understanding of the effect of small additions of Cr on the metastable austenite of high carbon steel from the macro- to nano-level. Consequently, it paves the way for new applications for high carbon low alloy steels. - Highlights: • Effect of small addition of Cr on metastable austenite of high carbon steel from the macro- to nano-level • A multi-scale study of elastoplastic behaviour of retained austenite in high carbon steel • The mechanical stability of retained austenite during plastic deformation increased with increasing Cr content • Effect of grain boundary misorientation angle on hardness of individual retained austenite grains in high carbon steel.

  14. CO2 laser cladding of VERSAlloyTM on carbon steel with powder feeding

    International Nuclear Information System (INIS)

    Kim, Jae-Do; Kweon, Jin-Wook

    2007-01-01

    Laser cladding processing with metal powder feeding has been experimented on carbon steel with VERSAlloy TM . A special device for the metal powder feeding was designed and manufactured. By adopting proper cladding parameters, good clad layers and sound metallurgical bonding with the base metal were obtained. Analysis indicates that the micro hardness of clad layer and the heat-affected zone increased with increasing of cladding speed. The experimental results showed that VERSAlloy TM cladded well with carbon steel

  15. Zn-10.2% Fe coating over carbon steel atmospheric corrosion resistance. Comparison with zinc coating

    International Nuclear Information System (INIS)

    Arnau, G.; Gimenez, E.; Rubio, M.V.; Saura, J.J.; Suay, J.J.

    1998-01-01

    Zn-10.2% Fe galvanized coating versus hot galvanized coating over carbon steel corrosion performance has been studied. Different periods of atmospheric exposures in various Valencia Community sites, and salt spray accelerated test have been done. Carbon steel test samples have been used simultaneously in order to classify exposure atmosphere corrosivity, and environmental exposure atmosphere characteristics have been analyzed. Corrosion Velocity versus environmental parameters has been obtained. (Author) 17 refs

  16. Electroslag welding of rotor steels produced with vacuum-carbon reduction

    International Nuclear Information System (INIS)

    Roshchin, M.B.; Modzhuk, M.D.; Izvekov, B.V.

    1985-01-01

    Metallurgical processes of electroslag welding of rotor steels, melted with vacuum-carbon deoxidation, have been considered. It is established, that during electroslag welding of steels with carbon content 0.20...0.30%, suppression of welding bath boiling and production of dense weld metal with a high impact strength can be ensured at oxygen concentration in soldered on metal not exceeding 0.01% and silicon content 0.06...0.10%

  17. Feasibility analysis of modified AL-6XN steel for structure component application in supercritical water-cooled reactor

    Institute of Scientific and Technical Information of China (English)

    Xinggang LI; Qingzhi YAN; Rong MA; Haoqiang WANG; Changchun GE

    2009-01-01

    Modified AL-6XN austenite steel was patterned after AL-6XN superaustenitic stainless steel by introducing microalloy elements such as zirconium and titanium in order to adapt to recrystallizing thermo-mechanical treatment and further improve crevice corrosion resistance. Modified AL-6XN exhibited comparable tensile strength, and superior plasticity and impact toughness to commercial AL-6XN steel. The effects of aging behavior on corrosion resistance and impact toughness were measured to evaluate the qualification of modified AL-6XN steel as an in-core component and cladding material in a supercritical water-cooled reactor. Attention should be paid to degradation in corrosion resistance and impact toughness after aging for 50 hours when modified AL-6XN steel is considered as one of the candidate materials for in-core components and cladding tubes in supercritical water-cooled reactors.

  18. Evaluation of the Steel Creek ecosystem in relation to the proposed restart of the L-reactor

    International Nuclear Information System (INIS)

    Smith, M.H.; Sharitz, R.R.; Gladden, J.B.

    1982-10-01

    This report summarizes the findings of slightly more than one year's study of the Steel Creek ecosystem. Generally, the findings have allowed us to refine our understanding of the structural and functional organization of the Steel Creek ecosystem which is an essential prerequisite for predicting the impacts associated with L-reactor restart. Reanalysis of the Steel Creek plant community relationships using 1981 aerial photography revealed that this component of the delta ecosystem continues to change as a result of natural successional processes. The major detectable changes have occurred on the more elevated portions of Steel Creek delta where coverage by woody species (especially willow) is continuing to increase. This successional woody community is invading areas previously dominated by persistent herbaceous species such as cut grass. Eleven vegetation associations were identified in the Steel Creek delta area, including two associations that were not apparently affected by the earlier reactor operations

  19. Effects of irradiation at lower temperature on the microstructure of Cr-Mo-V-alloyed reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Grosse, M; Boehmert, J; Gilles, R [Hahn-Meitner-Institut Berlin GmbH (Germany)

    1998-10-01

    The microstructural damage process due to neutron irradiation [1] proceeds in two stages: - formation of displacement cascades - evolution of the microstructure by defect reactions. Continuing our systematic investigation about the microstructural changes of Russian reactor pressure vessel steel due to neutron irradiation the microstructure of two laboratory heats of the VVER 440-type reactor pressure vessel steel after irradiation at 60 C was studied by small angle neutron scattering (SANS). 60 C-irradiation differently changes the irradiation-induced microstructure in comparison with irradiation at reactor operation temperature and can, thus, provide new insights into the mechanisms of the irradiation damage. (orig.)

  20. Updated embrittlement trend curve for reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Kirk, M.; Santos, C.; Eason, E.; Wright, J.; Odette, G.R.

    2003-01-01

    The reactor pressure vessels of commercial nuclear power plants are subject to embrittlement due to exposure to high energy neutrons from the core. Irradiation embrittlement of RPV belt-line materials is currently evaluated using US Regulatory Guide 1.99 Revision 2 (RG 1.99 Rev 2), which presents methods for estimating the Charpy transition temperature shift (ΔT30) at 30 ft-lb (41 J) and the drop in Charpy upper shelf energy (ΔUSE). A more recent embrittlement model, based on a broader database and more recent research results, is presented in NUREG/CR-6551. The objective of this paper is to describe the most recent update to the embrittlement model in NUREG/CR-6551, based upon additional data and increased understanding of embrittlement mechanisms. The updated ΔT30 and USE models include fluence, copper, nickel, phosphorous content, and product form; the ΔT30 model also includes coolant temperature, irradiation time (or flux), and a long-time term. The models were developed using multi-variable surface fitting techniques, understanding of the ΔT30 mechanisms, and engineering judgment. The updated ΔT30 model reduces scatter significantly relative to RG 1.99 Rev 2 on the currently available database for plates, forgings, and welds. This updated embrittlement trend curve will form the basis of revision 3 to Regulatory Guide 1.99. (author)

  1. Embrittlement recovery due to annealing of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Eason, E.D.; Wright, J.E.; Nelson, E.E.; Odette, G.R.; Mader, E.V.

    1998-01-01

    The irradiation embrittlement of nuclear reactor pressure vessels (RPV) can be reduced by thermal annealing at temperatures higher than the normal operating conditions. The objective of this work was to analyze the pertinent data and develop quantitative models for estimating the recovery in 41 J (30 ft-lb) Charpy transition temperature (TT) and Charpy upper shelf energy (USE) due to annealing. An analysis data base was developed, reviewed for completeness and accuracy, and documented as part of this work. Models were developed based on a combination of statistical techniques, including pattern recognition and transformation analysis, and the current understanding of the mechanisms governing embrittlement and recovery. The quality of models fitted in this project was evaluated by considering both the Charpy annealing data used for fitting and a surrogate hardness data base. This work demonstrates that microhardness recovery is a good surrogate for shift recovery and that there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes. (orig.)

  2. Regulatory aspects of radiation embrittlement of reactor vessel steels

    International Nuclear Information System (INIS)

    Randall, P.N.

    1979-01-01

    One purpose of this conference, is to re-examine the conventional wisdom about neutron radiation embrittlement and the methods used to counteract embrittlement in reactor vessels. Perhaps, there have been sufficient advances in fracture mechanics, core physics, dosimetry, and physical metallurgy to permit a forward step in the quantitative treatment of the subject. Certainly this would be consistent with the position of the U.S. Nuclear Regulatory Commission (the NRC) in general. ''There has been a continued evolution toward increased specificity.'' This statement appeared in the response prepared by the staff to a request from the Commission to explain how the staff decides to apply a new requirement and to whom, i.e., to back-fit or forward-fit-only or whatever. Pressure for increased specificity, i.e., for fleshing out general design criteria, comes from ''technical surprises'' in the form of operating experiences or from research information, and from attempts to improve our confidence in the safety of plants, especially new plants. Our goal is to have anticipated and evaluated all possible modes of failure with sufficient quantitativeness that the probability of failure can be estimated with some accuracy. Failing this, regulators demand large margins of safety to cover our ignorance

  3. Corrosion by concentrated sulfuric acid in carbon steel pipes and tanks: state of the art

    Energy Technology Data Exchange (ETDEWEB)

    Panossian, Zehbour; Almeida, Neusvaldo Lira de; Sousa, Raquel Maria Ferreira de [Instituto de Pesquisas Tecnologicas (IPT), Sao Paulo, SP (Brazil); Pimenta, Gutemberg de Souza [PETROBRAS S.A., Rio de Janeiro, RJ (Brazil). Centro de Pesquisas e Desenvolvimento (CENPES); Marques, Leandro Bordalo Schmidt [PETROBRAS Engenharia, Rio de Janeiro, RJ (Brazil)

    2009-07-01

    PETROBRAS, allied to the policy of reduction of emission of pollutants, has been adjusting the processes of the new refineries to obtain products with lower sulfur content. Thus, the sulfur dioxide, extracted from the process gases of a new refinery to be built in the Northeast, will be used to produce sulfuric acid with concentration between (94-96) %. This acid will be stored in carbon steel tanks and transported through a buried 8-km carbon steel pipe from the refinery to a pier, where it will be loaded onto ships and sent to the consumer markets. Therefore, the corrosion resistance of carbon steel by concentrated acid will become a great concern for the mentioned storage and transportation. When the carbon steel comes into contact with concentrated sulfuric acid, there is an immediate acid attack with the formation of hydrogen gas and ferrous ions which, in turn, forms a protective layer of FeSO{sub 4} on the metallic surface. The durability of the tanks and pipes made of carbon steel will depend on the preservation of this protective layer. This work presents a review of the carbon steel corrosion in concentrated sulfuric acid and discusses the preventive methods against this corrosion, including anodic protection. (author)

  4. Design of aging-resitant martensitic stainless steels for pressurized water reactors

    International Nuclear Information System (INIS)

    Cozar, R.; Meyzaud, Y.

    1983-06-01

    With the exception of AISI 403 or 410 grades, the use of high strength martensitic stainless steels in PWR is poorly developped because these materials, like ferritic stainless steels, become embrittled by the precitation of a b.c.c. chromium-rich phase during aging at the operating temperature (290 to 350 0 C). The influence of alloying elements and microstructure on the aging behavior of forged low-carbon martensitic stainless steels containing 12 to 16% Cr, 0 to 2% Mo and 0 to 8% Ni was determined during accelerated aging at 450 0 C. Quantitative relationships were derived between the maximum increase in hardness, the maximum shift in CVN transition temperature and the chemical composition (Cr, Mo, C) and microstructure

  5. Influence of the impurities on the depth of penetration with carbon steel weldings

    Directory of Open Access Journals (Sweden)

    O. Savytsky

    2014-04-01

    Full Text Available In this paper the results of the research about the influence of the impurities on the depth of penetration with carbon steels weldings of different chemical composition are presented. These data suggest that presence of those impurities, such as sulphure and oxygen, in the steel, increases the depth of penetration to 1,3 - 1,5 times compared to welding refined steels. Applying activating fluxes for welding high tensile steels, provides an increase in the depth of penetration of 2 - 3 times.

  6. Intragranular ferrite morphologies in medium carbon vanadium-microalloyed steel

    Directory of Open Access Journals (Sweden)

    Fadel A.

    2013-01-01

    Full Text Available The aim of this work was to determine TTT diagram of medium carbon V-N micro-alloyed steel with emphasis on the development of intragranular ferrite morphologies. The isothermal treatment was carried out at 350, 400, 450, 500, 550 and 600°C. These treatments were interrupted at different times in order to analyze the evolution of the microstructure. Metallographic evaluation was done using optical and scanning electron microscopy (SEM. The results show that at high temperatures (≥ 500°C polygonal intragranulary nucleated ferrite idiomorphs, combined with grain boundary ferrite and pearlite were produced and followed by an incomplete transformation phenomenon. At intermediate temperatures (450, 500°C an interloced acicular ferrite (AF microstructure is produced, and at low temperatures (400, 350°C the sheave of parallel acicular ferrite plates, similar to bainitic sheaves but intragranularly nucleated were observed. In addition to sheaf type acicular ferrite, the grain boundary nucleated bainitic sheaves are observed. [Projekat Ministartsva nauke Republike Srbije, br. OI174004

  7. The anaerobic corrosion of carbon steel in concrete

    International Nuclear Information System (INIS)

    Naish, C.C.; Balkwill, P.H.; O'Brien, T.M.; Taylor, K.J.; Marsh, G.P.

    1990-11-01

    The report describes the work of a two year programme investigating the anaerobic corrosion of carbon steel embedded in a range of candidate repository cements and concretes at laboratory ambient temperatures. The factors investigated in the study were the rate of the anaerobic corrosion reaction, the effect of hydrogen overpressure on the reaction rate and the form of the corrosion product. Both electrochemical and sample weight loss corrosion rate measurements were used. The cements and concretes used were prepared both with and without small additions of chloride (2% by weight of mix water). The results indicate that the corrosion rate is low, <1 μm/year, the effect of hydrogen overpressure is not significant over the range of pressures investigated, 1-100 atmospheres, and that the corrosion product is dependent on the cement used to cast the samples. Magnetite was identified in the case of blast furnace slag replacement cements but for pulverised fuel ash and ordinary Portland cements no corrosion product was evident either from X-ray diffraction or laser Raman measurements. (Author)

  8. The anaerobic corrosion of carbon steel in concrete

    International Nuclear Information System (INIS)

    Naish, C.C.; Balkwill, P.H.; O'Brien, T.M.; Taylor, K.J.; Marsh, G.P.

    1990-11-01

    The report describes the work of a two year programme investigating the anaerobic corrosion of carbon steel embedded in a range of candidate repository cements and concretes at laboratory temperatures. The factors investigated in the study were the rate of the anaerobic corrosion reaction, the effect of hydrogen overpressure on the reaction rate and the form of the corrosion product. Both electrochemical and sample weight loss corrosion rate measurements were used. The cements and concretes used were prepared both with and without small additions of chloride (2% by weight of mix water). The results indicate that the corrosion rate is low, < 1 μm/year, the effect of hydrogen overpressure is not significant over the range of pressures investigated, 1-100 atmospheres, and that the corrosion product is dependent on the cement used to cast the samples. Magnetite was identified in the case of blast furnace slag replacement cements but for pulverised fuel ash and ordinary Portland cements no corrosion product was evident either from X-ray diffraction or laser Raman measurements. Further work is presently underway to investigate the effects of elevated temperatures and chloride levels on the anaerobic corrosion reaction and the rate of hydrogen gas production. (author)

  9. Investigation of the pitting corrosion of low carbon steel containers

    International Nuclear Information System (INIS)

    Mughabghab, S.F.; Sullivan, T.M.

    1988-01-01

    The present study was undertaken because the prediction of the degradation rate of low carbon steel contains over long time frames is one of the crucial elements in the development of a source term model for low-level shallow land burial. The principal data base considered is that of the NBS corrosion measurements of ferrous materials buried in the ground for periods of up to 18 years. In this investigation, the maximum penetration in mils, hm, due to pitting corrosion was found to conform closely to the relation h m = kt n where it is the exposure time of the sample in years, κ is the pitting parameter in mil/(years) n , and n > O is a parameter related to the aeration property of the soil. The central objective of the present investigation is the determination of the dependence of the pitting parameters κ and n on the soil properties. The result of a detailed linear correlation analysis of κ on one hand, the pH value and the resistivity of the soil on the other hand revealed that κ is principally influenced by the pH value of the soil. The resistivity of the soil is found to play a minor role

  10. Decrudding and chemical cleaning of carbon steel components - an evaluation

    International Nuclear Information System (INIS)

    Gaonkar, K.B.; Elayathu, N.S.D.; Shibad, P.R.; Gadiyar, H.S.

    1982-01-01

    Corrosion and accumulation of corrosion products on the surfaces of structural components and plant equipments can cause se vereoperational problems during service. An illustration is the heat exchanger systems in nuclear power stations. Development and standardisation of appropriate chemical cleaning and decontamination procedures and their evaluation hence merit serious consideration. A number of chemical cleaning procedures using formulations based on hydrochloric and citric acid solutions have been examined to study their crud dissolving and derusting ability in addition to the attack on base material. The compositions were chosen: (1) along with complexing agents EDTA and ammonium citrate, (2) with pH control, and (3) with the use of inhibitors acridine, rhodine, hexamine and phenyl-thiourea. The evaluations have been made at 28 and 60 deg C. Rusted carbon steel coupons having a rust of 10-12 mg/cm 2 on the surface have been used for the purpose of the above evaluations. Data on corrosion rates of monel and cupronickel (70:30) in the descaling solutions have also been presented. Results on the above evaluation studies have been discussed. (author)

  11. Effects of hydrogen on carbon steels at the Multi-Function Waste Tank Facility

    International Nuclear Information System (INIS)

    Carlos, W.C.

    1995-01-01

    Concern has been expressed that hydrogen produced by corrosion, radiolysis, and decomposition of the waste could cause embrittlement of the carbon steel waste tanks at Hanford. The concern centers on the supposition that the hydrogen evolved in many of the existing tanks might penetrate the steel wall of the tank and cause embrittlement that might lead to catastrophic failure. This document reviews literature on the effects of hydrogen on the carbon steel proposed for use in the Multi-Function Waste Tank Facility for the time periods before and during construction as well as for the operational life of the tanks. The document draws several conclusions about these effects. Molecular hydrogen is not a concern because it is not capable of entering the steel tank wall. Nascent hydrogen produced by corrosion reactions will not embrittle the steel because the mild steel used in tank construction is not hard enough to be susceptible to hydrogen stress cracking and the corrosion product hydrogen is not produced at a rate sufficient to cause either loss in tensile ductility or blistering. If the steel intended for use in the tanks is produced to current technology, fabricated in accordance with good construction practice, postweld heat treated, and operated within the operating limits defined, hydrogen will not adversely affect the carbon steel tanks during their 50-year design life. 26 refs

  12. Microstructural stability of fast reactor irradiated 10 to 12% Cr ferritic-martensitic stainless steels

    International Nuclear Information System (INIS)

    Little, E.A.; Stoter, L.P.

    1982-01-01

    The strength and microstructural stability of three 10 to 12% Cr ferritic-martensitic stainless steels have been characterized following fast reactor irradiation to damage levels of 30 displacements per atom (dpa) at temperatures in the range 380 to 615 0 C. Irradiation results in either increases or decreases in room temperature hardness depending on the irradiation temperature. These strength changes can be qualitatively rationalized in terms of the combined effects of irradiation-induced interstitial dislocation loop formation and recovery of the dislocation networks comprising the initial tempered martensite structures. Precipitate evolution in the irradiated steels is associated with the nonequilibrium segregation of the elements nickel, silicon, molybdenum, chromium and phosphorus, brought about by solute-point defect interactions. The principal irradiation-induced precipitates identified are M 6 X, intermetallic chi and sigma phases and also α' (Cr-rich ferrite). The implications of the observed microstructural changes on the selection of martensitic stainless steels for fast reactor wrapper applications are briefly considered

  13. Pre-Combustion Carbon Dioxide Capture by a New Dual Phase Ceramic-Carbonate Membrane Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Jerry Y. S. [Arizona State Univ., Tempe, AZ (United States)

    2015-01-31

    This report documents synthesis, characterization and carbon dioxide permeation and separation properties of a new group of ceramic-carbonate dual-phase membranes and results of a laboratory study on their application for water gas shift reaction with carbon dioxide separation. A series of ceramic-carbonate dual phase membranes with various oxygen ionic or mixed ionic and electronic conducting metal oxide materials in disk, tube, symmetric, and asymmetric geometric configurations was developed. These membranes, with the thickness of 10 μm to 1.5 mm, show CO2 permeance in the range of 0.5-5×10-7 mol·m-2·s-1·Pa-1 in 500-900°C and measured CO2/N2 selectivity of up to 3000. CO2 permeation mechanism and factors that affect CO2 permeation through the dual-phase membranes have been identified. A reliable CO2 permeation model was developed. A robust method was established for the optimization of the microstructures of ceramic-carbonate membranes. The ceramic-carbonate membranes exhibit high stability for high temperature CO2 separations and water gas shift reaction. Water gas shift reaction in the dual-phase membrane reactors was studied by both modeling and experiments. It is found that high temperature syngas water gas shift reaction in tubular ceramic-carbonate dual phase membrane reactor is feasible even without catalyst. The membrane reactor exhibits good CO2 permeation flux, high thermal and chemical stability and high thermal shock resistance. Reaction and separation conditions in the membrane reactor to produce hydrogen of 93% purity and CO2 stream of >95% purity, with 90% CO2 capture have been identified. Integration of the ceramic-carbonate dual-phase membrane reactor with IGCC process for carbon dioxide capture was analyzed. A methodology was developed to identify optimum operation conditions for a

  14. The Effects of Cr and Al Addition on Transformation and Properties in Low‐Carbon Bainitic Steels

    OpenAIRE

    Junyu Tian; Guang Xu; Mingxing Zhou; Haijiang Hu; Xiangliang Wan

    2017-01-01

    Three low‐carbon bainitic steels were designed to investigate the effects of Cr and Al addition on bainitic transformation, microstructures, and properties by metallographic method and dilatometry. The results show that compared with the base steel without Cr and Al addition, only Cr addition is effective for improving the strength of low‐carbon bainitic steel by increasing the amount of bainite. However, compared with the base steel, combined addition of Cr and Al has no significant effect o...

  15. Estimation of fast neutron fluence in steel specimens type Laguna Verde in TRIGA Mark III reactor; Estimacion de la fluencia de neutrones rapidos en probetas de acero tipo Laguna Verde en el reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Aguilar H, F., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    The main purpose of this work is to obtain the fluence of fast neutrons recorded within four specimens of carbon steel, similar to the material having the vessels of the BWR reactors of the nuclear power plant of Laguna Verde when subjected to neutron flux in a experimental facility of the TRIGA Mark III reactor, calculating an irradiation time to age the material so accelerated. For the calculation of the neutron flux in the specimens was used the Monte Carlo code MCNP5. In an initial stage, three sheets of natural molybdenum and molybdenum trioxide (MoO{sub 3}) were incorporated into a model developed of the TRIGA reactor operating at 1 M Wth, to calculate the resulting activity by setting a certain time of irradiation. The results obtained were compared with experimentally measured activities in these same materials to validate the calculated neutron flux in the model used. Subsequently, the fast neutron flux received by the steel specimens to incorporate them in the experimental facility E-16 of the reactor core model operating at nominal maximum power in steady-state was calculated, already from these calculations the irradiation time required was obtained for values of the neutron flux in the range of 10{sup 18} n/cm{sup 2}, which is estimated for the case of Laguna Verde after 32 years of effective operation at maximum power. (Author)

  16. The CCT diagrams of ultra low carbon bainitic steels and their impact toughness properties

    International Nuclear Information System (INIS)

    Lis, A.K.; Lis, J.; Jeziorski, L.

    1998-01-01

    The CCT diagrams of ULCB N i steels, HN3MV, HN3MVCu having 5.1% Ni and 3.5% Ni and Cu bearing steels; HN3M1.5Cu, HSLA 100 have been determined. The reduced carbon concentration in steel, in order to prevent the formation of cementite, allowed for using nickel, manganese, chromium and molybdenum to enhance hardenability and refinement of the bainitic microstructures by lowering B S temperature. Copper and microadditions of vanadium and niobium are successfully used for precipitation strengthening of steel both in thermomechanically or heat treated conditions. Very good fracture toughness at low temperatures and high yield strength properties of HN3MVCu and HN3MV steels allowed for fulfillment of the requirements for steel plates for pressure vessels and cryogenic applications. (author)

  17. A quality approach to maintain the properties of S235 JR structural carbon steel in Lebanon

    International Nuclear Information System (INIS)

    Sidawi, J.A.; Al Khatib, H.

    2004-01-01

    Full text.S235JR carbon steel is one of the most popular steels used in Lebanon. It is imported by steel dealers and is widely used by all fabricators and manufacturers of steels for many structural purposes and applications. This kind of steel has good ductile properties as well as excellent weldability. It is still known by its previous designation St 37-2 or E 24-2. S235JR is produced in many shapes and thicknesses such as steel plates, sheets, angles and different other geometric shapes. Standard chemical and mechanical tests were conducted and reported on S235JR hot-rolled structural low-carbon mild steel specimens collected from Lebanese steel market. The main objective of this work is to assure the compliance of these properties with those set by the steel manufacturer. The above mentioned tests were performed at the laboratories of the Industrial Research Institute (IR) in Lebanon to assure the quality and credibility of the results. related European and American standards were presented as references and compared with the achieved results. Discussion was presented to show the similarities and differences between S235JR steel samples and standard requirements. Some of the reasons for such differences were discussed. Sufficient data was furnished through this work for the public and mainly for the Lebanese Standard Organization LIBNOR to easily adopt and implement the EN 10025:1993 European standard that can be applied in Lebanon concerning the most commonly used hot rolled low carbon structural steel. A follow up concerning adopting and implementing EN 10025:1993 will be briefed

  18. Ultra-Low Carbon Bainitic Steels for Heavy Plate Applications

    Science.gov (United States)

    1990-12-01

    these steels. The CCT diagrams 7 of steels typical of the HY grades indicate that the nose of the proeutectoid ferrite/pearlite reactions is located...austenite, carbides, and martensite. An example of the type of CCT diagram for one of the steels used in this investigation is presented in Figure 12...introduce a "bay" of unstable austenite which acts to separate the ferrite "nose" from the bainite/martensite regions on TTT or CCT diagrams , see Figure

  19. Flux effect on neutron irradiation embrittlement of reactor pressure vessel steels irradiated to high fluences

    International Nuclear Information System (INIS)

    Soneda, N.; Dohi, K.; Nishida, K.; Nomoto, A.; Iwasaki, M.; Tsuno, S.; Akiyama, T.; Watanabe, S.; Ohta, T.

    2011-01-01

    Neutron irradiation embrittlement of reactor pressure vessel (RPV) steels is of great concern for the long term operation of light water reactors. In particular, the embrittlement of the RPV steels of pressurized water reactors (PWRs) at very high fluences beyond 6*10 19 n/cm 2 , E > 1 MeV, needs to be understood in more depth because materials irradiated in material test reactors (MTRs) to such high fluences show larger shifts than predicted by current embrittlement correlation equations available worldwide. The primary difference between the irradiation conditions of MTRs and surveillance capsules is the neutron flux. The neutron flux of MTR is typically more than one order of magnitude higher than that of surveillance capsule, but it is not necessarily clear if this difference in neutron flux causes difference in mechanical properties of RPV. In this paper, we perform direct comparison, in terms of mechanical property and microstructure, between the materials irradiated in surveillance capsules and MTRs to clarify the effect of flux at very high fluences and fluxes. We irradiate the archive materials of some of the commercial reactors in Japan in the MTR, LVR-15, of NRI Rez, Czech Republic. Charpy impact test results of the MTR-irradiated materials are compared with the data from surveillance tests. The comparison of the results of microstructural analyses by means of atom probe tomography is also described to demonstrate the similarity / differences in surveillance and MTR-irradiated materials in terms of solute atom behavior. It appears that high Cu material irradiated in a MTR presents larger shifts than those of surveillance data, while low Cu materials present similar embrittlement. The microstructural changes caused by MTR irradiation and surveillance irradiation are clearly different

  20. Borated stainless steel storage project to the spent fuel of the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Rodrigues, Antonio Carlos Iglesias; Madi Filho, Tufic; Ricci Filho, Walter

    2013-01-01

    The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks to the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating circumstances, the storage will have capacity for approximately six years. Whereas the estimated useful life of the IEA-R1 is around twenty years, it will be necessary to increase the storage capacity for the spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. This work objective is the project of high capacity storage for spent fuel elements, using borated stainless steel, to answer the Reactor IEA-R1 demand and the security requirements of the International Atomic Energy Agency. (author)

  1. Borated stainless steel storage project to the spent fuel of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, Antonio Carlos Iglesias; Madi Filho, Tufic; Ricci Filho, Walter, E-mail: acirodri@ipen.br, E-mail: tmfilho@ipen.br, E-mail: wricci@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks to the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating circumstances, the storage will have capacity for approximately six years. Whereas the estimated useful life of the IEA-R1 is around twenty years, it will be necessary to increase the storage capacity for the spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. This work objective is the project of high capacity storage for spent fuel elements, using borated stainless steel, to answer the Reactor IEA-R1 demand and the security requirements of the International Atomic Energy Agency. (author)

  2. Nanostructure evolution of neutron-irradiated reactor pressure vessel steels: Revised Object kinetic Monte Carlo model

    Energy Technology Data Exchange (ETDEWEB)

    Chiapetto, M., E-mail: mchiapet@sckcen.be [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium); Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Messina, L. [DEN-Service de Recherches de Métallurgie Physique, CEA, Université Paris-Saclay, F-91191 Gif-sur-Yvette (France); KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-114 21 Stockholm (Sweden); Becquart, C.S. [Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Olsson, P. [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-114 21 Stockholm (Sweden); Malerba, L. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium)

    2017-02-15

    This work presents a revised set of parameters to be used in an Object kinetic Monte Carlo model to simulate the microstructure evolution under neutron irradiation of reactor pressure vessel steels at the operational temperature of light water reactors (∼300 °C). Within a “grey-alloy” approach, a more physical description than in a previous work is used to translate the effect of Mn and Ni solute atoms on the defect cluster diffusivity reduction. The slowing down of self-interstitial clusters, due to the interaction between solutes and crowdions in Fe is now parameterized using binding energies from the latest DFT calculations and the solute concentration in the matrix from atom-probe experiments. The mobility of vacancy clusters in the presence of Mn and Ni solute atoms was also modified on the basis of recent DFT results, thereby removing some previous approximations. The same set of parameters was seen to predict the correct microstructure evolution for two different types of alloys, under very different irradiation conditions: an Fe-C-MnNi model alloy, neutron irradiated at a relatively high flux, and a high-Mn, high-Ni RPV steel from the Swedish Ringhals reactor surveillance program. In both cases, the predicted self-interstitial loop density matches the experimental solute cluster density, further corroborating the surmise that the MnNi-rich nanofeatures form by solute enrichment of immobilized small interstitial loops, which are invisible to the electron microscope.

  3. Evolution of manganese–nickel–silicon-dominated phases in highly irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Wells, Peter B.; Yamamoto, Takuya; Miller, Brandon; Milot, Tim; Cole, James; Wu, Yuan; Odette, G. Robert

    2014-01-01

    Formation of a high density of Mn–Ni–Si nanoscale precipitates in irradiated Cu-free and Cu-bearing reactor pressure vessel steels could lead to severe unexpected embrittlement. Models long ago predicted that these precipitates, which are not treated in current embrittlement prediction models, would emerge only at high fluence. However, the mechanisms and variables that control Mn–Ni–Si precipitate formation, and their detailed characteristics, have not been well understood. High flux irradiations of six steels with systematic variations in Cu and Ni contents were carried out at ∼295 °C to high and very high neutron fluences of ∼1.3 × 10 20 and ∼1.1 × 10 21 n cm −2 . Atom probe tomography shows that significant mole fractions of Mn–Ni–Si-dominated precipitates form in the Cu-bearing steels at ∼1.3 × 10 20 n cm −2 , while they are only beginning to develop in Cu-free steels. However, large mole fractions of these precipitates, far in excess of those found in previous studies, are observed at 1.1 × 10 21 n cm −2 at all Cu contents. At the highest fluence, the precipitate mole fractions primarily depend on the alloy Ni, rather than Cu, content. The Mn–Ni–Si precipitates lead to very large increases in measured hardness, corresponding to yield strength elevations of up to almost 700 MPa

  4. Properties of welded joints of 2,25Cr-1Mo steel with various carbon content

    International Nuclear Information System (INIS)

    Vornovitskij, I.N.; Brodetskaya, E.Z.; Pozdnyakova, A.S.

    1980-01-01

    Properties of welded joints of 2,25 Cr - 1 Mo steel pipelines with different carbon content are considered. It is shown that application of electrodes developed in some countries for welding permits in many cases to exclude heat treatment of welded joints owing to high ductility of weld deposited metal. To improve the ductility, it is necessary to limit both carbon content down to 0,03-0,06% and detrimental elements (sulfur, phosphorus). Heat affected zone hardness may be increased at the expense of carbon. Weld deposited metal possesses the highest long-term strength at the given test temperature; in this case long-term strength of welded joints and base metal is practically the same. The long-term strength of high-carbon steel is higher at the test temperature of 565 deg C as compared to mean-carbon and low-carbon steels, whose long-term strength is practically equal at this temperature. The long-term strength of high-carbon and mean-carbon steels is practically the same and higher as compared with low-carbon one at the test temperature of 510 deg C

  5. Spectroscopic investigations of plasma nitriding processes: A comparative study using steel and carbon as active screen materials

    Science.gov (United States)

    Hamann, S.; Burlacov, I.; Spies, H.-J.; Biermann, H.; Röpcke, J.

    2017-04-01

    Low-pressure pulsed DC H2-N2 plasmas were investigated in the laboratory active screen plasma nitriding monitoring reactor, PLANIMOR, to compare the usage of two different active screen electrodes: (i) a steel screen with the additional usage of CH4 as carbon containing precursor in the feeding gas and (ii) a carbon screen without the usage of any additional gaseous carbon precursor. Applying the quantum cascade laser absorption spectroscopy, the evolution of the concentration of four stable molecular species, NH3, HCN, CH4, and C2H2, has been monitored. The concentrations were found to be in a range of 1012-1016 molecules cm-3. By analyzing the development of the molecular concentrations at variations of the screen plasma power, a similar behavior of the monitored reaction products has been found for both screen materials, with NH3 and HCN as the main reaction products. When using the carbon screen, the concentration of HCN and C2H2 was 30 and 70 times higher, respectively, compared to the usage of the steel screen with an admixture of 1% CH4. Considering the concentration of the three detected hydrocarbon reaction products, a combustion rate of the carbon screen of up to 69 mg h-1 has been found. The applied optical emission spectroscopy enabled the determination of the rotational temperature of the N2+ ion which has been in a range of 650-900 K increasing with the power in a similar way in the plasma of both screens. Also with power the ionic component of nitrogen molecules, represented by the N2+ (0-0) band of the first negative system, as well as the CN (0-0) band of the violet system increase strongly in relation to the intensity of the neutral nitrogen component, i.e., the N2 (0-0) band of the second positive system. In addition, steel samples have been treated with both the steel and the carbon screen resulting in a formation of a compound layer of up to 10 wt. % nitrogen and 10 wt. % carbon, respectively, depending on the screen material.

  6. Corrosion of Carbon Steel and Corrosion-Resistant Rebars in Concrete Structures Under Chloride Ion Attack

    Science.gov (United States)

    Mohamed, Nedal; Boulfiza, Mohamed; Evitts, Richard

    2013-03-01

    Corrosion of reinforced concrete is the most challenging durability problem that threatens reinforced concrete structures, especially structures that are subject to severe environmental conditions (i.e., highway bridges, marine structures, etc.). Corrosion of reinforcing steel leads to cracking and spalling of the concrete cover and billions of dollars are spent every year on repairing such damaged structures. New types of reinforcements have been developed to avoid these high-cost repairs. Thus, it is important to study the corrosion behavior of these new types of reinforcements and compare them to the traditional carbon steel reinforcements. This study aimed at characterizing the corrosion behavior of three competing reinforcing steels; conventional carbon steel, micro-composite steel (MMFX-2) and 316LN stainless steel, through experiments in carbonated and non-carbonated concrete exposed to chloride-laden environments. Synthetic pore water solutions have been used to simulate both cases of sound and carbonated concrete under chloride ions attack. A three-electrode corrosion cell is used for determining the corrosion characteristics and rates. Multiple electrochemical techniques were applied using a Gamry PC4™ potentiostat manufactured by Gamry Instruments (Warminster, PA). DC corrosion measurements were applied on samples subjected to fixed chloride concentration in the solution.

  7. Accelerated Carbonation of Steel Slag Compacts: Development of High-Strength Construction Materials

    Energy Technology Data Exchange (ETDEWEB)

    Quaghebeur, Mieke; Nielsen, Peter, E-mail: peter.nielsen@vito.be; Horckmans, Liesbeth [Sustainable Materials Management, VITO, Mol (Belgium); Van Mechelen, Dirk [RECMIX bvba, Genk (Belgium)

    2015-12-17

    Mineral carbonation involves the capture and storage of carbon dioxide in carbonate minerals. Mineral carbonation presents opportunities for the recycling of steel slags and other alkaline residues that are currently landfilled. The Carbstone process was initially developed to transform non-hydraulic steel slags [stainless steel (SS) slag and basic oxygen furnace (BOF) slags] in high-quality construction materials. The process makes use of accelerated mineral carbonation by treating different types of steel slags with CO{sub 2} at elevated pressure (up to 2 MPa) and temperatures (20–140°C). For SS slags, raising the temperature from 20 to 140°C had a positive effect on the CO{sub 2} uptake, strength development, and the environmental properties (i.e., leaching of Cr and Mo) of the carbonated slag compacts. For BOF slags, raising the temperature was not beneficial for the carbonation process. Elevated CO{sub 2} pressure and CO{sub 2} concentration of the feed gas had a positive effect on the CO{sub 2} uptake and strength development for both types of steel slags. In addition, the compaction force had a positive effect on the strength development. The carbonates that are produced in situ during the carbonation reaction act as a binder, cementing the slag particles together. The carbonated compacts (Carbstones) have technical properties that are equivalent to conventional concrete products. An additional advantage is that the carbonated materials sequester 100–150 g CO{sub 2}/kg slag. The technology was developed on lab scale by the optimization of process parameters with regard to compressive strength development, CO{sub 2} uptake, and environmental properties of the carbonated construction materials. The Carbstone technology was validated using (semi-)industrial equipment and process conditions.

  8. Overview of 9Cr steels properties for structural application in sodium fast reactors

    International Nuclear Information System (INIS)

    Cabet, Celine; Courouau, Jean-Louis; Dalle, France; Desgranges, Clara; Forest, Laurent; Martinelli, Laure; Sauzay, Maxime

    2015-01-01

    A research and development programme has been launched by CEA, EDF and AREVA for the choice and qualification of material for sodium fast reactor (SFR) structural components. The requirements on steam generator (SG) are demanding, with operating temperatures ranging from 240 deg. C to 530 deg. C in water/steam and in sodium for an extended design life of several decades. The selection of the SG materials is based on many characteristics: fabrication, welding, thermal properties, mechanical strength at low and high temperature, environmental resistance. 9%Cr steels which are relevant candidate alloys for different designs of SGs have been extensively studied in the past decade. The objective of this paper is to review some advances made at CEA on determining properties of the X10CrMoVNb9-1 steel (hereafter named 'grade 91'): welding, modelling of cyclic softening, modelling of long-term creep, compatibility with liquid sodium, corrosion in steam. (authors)

  9. Stress corrosion cracking of L-grade stainless steels in boiling water reactor (BWR) plants

    International Nuclear Information System (INIS)

    Suzuki, Shunichi; Fukuda, Toshihiko; Yamashita, Hironobu

    2004-01-01

    L-grade stainless steels as 316NG, SUS316L and SUS304L have been used for the BWR reactor internals and re-circulation pipes as SCC resistant materials. However, SCC of the L-grade material components were reported recently in many Japanese BWR plants. The detail investigation of the components showed the fabrication process such as welding, machining and surface finishing strongly affected SCC occurrence. In this paper, research results of SCC of L-grade stainless steels, metallurgical investigation of core shrouds and re-circulation pipings, and features of SCC morphology were introduced. Besides, the structural integrity of components with SCC, countermeasures for SCC and future R and D planning were introduced. (author)

  10. Laser cutting of thick steel plates and simulated steel components using a 30 kW fiber laser

    International Nuclear Information System (INIS)

    Tamura, Koji; Ishigami, Ryoya; Yamagishi, Ryuichiro

    2016-01-01

    Laser cutting of thick steel plates and simulated steel components using a 30 kW fiber laser was studied for application to nuclear decommissioning. Successful cutting of carbon steel and stainless steel plates up to 300 mm in thickness was demonstrated, as was that of thick steel components such as simulated reactor vessel walls, a large pipe, and a gate valve. The results indicate that laser cutting applied to nuclear decommissioning is a promising technology. (author)

  11. Applications of controlled thermonuclear reactor (CTR) fusion power in the steel industry

    International Nuclear Information System (INIS)

    Jordan, R.K.; Steinberg, M.

    1975-03-01

    A review of the process and economics of basic steel production is presented for the purpose of indicating where CTR fusion energy may be applicable. The present conventional air blown blast furnace produces a relatively low Btu value top gas with limited usefulness. The industry consumes relatively large amounts of natural gas for reheating ingots, plates, etc. A concept is presented wherein oxygen is used in the blast furnace which would double the capacity of the furnace and produce a rich carbon monoxide gas stream useful as synthesis gas for methanol and ammonia production. A CTR supplying high energy radiation in a blanket would disproportionate carbon dioxide to carbon monoxide and oxygen which could be used at high temperatures in the blast furnace in place of an oxygen supply stream. Coke would be used in this scheme. In a second scheme the oxygen is separated from the disproportioned CO 2 stream and CO is used in a direct reduction furnace which is followed by an electric furnace to refine the reduced product. Other schemes include iron ore reduction with electrolytic hydrogen and the use of thermal energy for reforming coal with steam or CO 2 for production of reducing gas. The electrosmelting of scrap metal using CTR power could become an important operation in the future. A complex of steel, fertilizer, fuel and chemical production is presented. Steel capacity and power requirement data are presented and projected to the year 2020. (U.S.)

  12. Chemistry conditions in crevices of carbon steel and stainless steel: a comparative study

    International Nuclear Information System (INIS)

    Pushpalata, R.; Veena, S.; Chandran, Sinu; Mohan, T.V.K.; Rangarajan, S.; Narasimhan, S.V.

    2008-01-01

    Occurrence of crevice corrosion in the steam generator tubes of nuclear power plants may lead to transport of radioactivity to the secondary side. It is expected that effect of crevice corrosion will be more pronounced in a passive material like stainless steel (SS) as compared to carbon steel (CS). Theoretical modeling of the dynamics of crevice chemistry calls for experimental data with respect to various water chemistry parameters like pH, conductivity and concentrations of the ionic species in typical crevices of different geometry (aspect ratio of length and width). This paper presents the experimental results obtained with crevices in CS -106 B, SS-304 (nano grain) and SS 316 blocks (varying dimensions) exposed to a medium containing 1 ppm of lithium and chloride ion each for 10 days in static autoclave at 245 deg C. The bulk solution pH showed a reduction in alkalinity and slight increase in conductivity. In case of CS about 58 times increase in Cl - was observed in the smaller crevice of dimension 1 mm (width) x 25 mm (depth) whereas it was only ∼ 12 times in the bigger crevice (2 mm x 39 mm). Other anionic impurities like SO 4 2- and Br - present as impurities in NaCI were also found to be concentrated in the crevices whereas not much increase in cationic impurities was observed. In a similar experiment with SS blocks with crevice dimension comparable to diffusion layer thickness, appreciable increase in chloride concentration was observed. Electrochemical experiments were also carried out in deaerated NaCI (3.5%) solution at 25 deg C with CS, SS-304 (nano grain) and SS-316 (normal-grain) coupons. The OCP was -297 mV for SS-316 whereas for SS-304 coupon the OCP was -339 mV. Potentiodynamic anodic polarization curve showed a passive behavior up to 0.0V and then a sudden increase in anodic current. On nano-grained SS, a yellowish film on the surface was observed with a large number of pits whereas severe general corrosion was observed in the normal

  13. 77 FR 37711 - Circular Welded Carbon-Quality Steel Pipe From India, Oman, the United Arab Emirates, and Vietnam...

    Science.gov (United States)

    2012-06-22

    ...)] Circular Welded Carbon-Quality Steel Pipe From India, Oman, the United Arab Emirates, and Vietnam...-fair-value imports from India, Oman, the United Arab Emirates, and Vietnam of circular welded carbon... respect to circular welded carbon-quality steel pipe from Oman and the United Arab Emirates being sold in...

  14. 75 FR 55745 - Corrosion-Resistant Carbon Steel Flat Products From the Republic of Korea: Preliminary Results...

    Science.gov (United States)

    2010-09-14

    ... Products covered by this order are certain corrosion-resistant carbon steel flat products from Korea. These... DEPARTMENT OF COMMERCE International Trade Administration [C-580-818] Corrosion-Resistant Carbon... review of the countervailing duty (CVD) order on corrosion-resistant carbon steel flat products (CORE...

  15. 78 FR 19210 - Corrosion-Resistant Carbon Steel Flat Products From the Republic of Korea: Final Results of...

    Science.gov (United States)

    2013-03-29

    .... Scope of the Order Products covered by this order are certain corrosion-resistant carbon steel flat... DEPARTMENT OF COMMERCE International Trade Administration [C-580-818] Corrosion-Resistant Carbon... countervailing duty (CVD) order on corrosion-resistant carbon steel flat products from the Republic of Korea for...

  16. 78 FR 55241 - Corrosion-Resistant Carbon Steel Flat Products From the Republic of Korea: Preliminary Results of...

    Science.gov (United States)

    2013-09-10

    ... merchandise covered by this Order \\2\\ is certain corrosion- resistant carbon steel flat products from Korea... DEPARTMENT OF COMMERCE International Trade Administration [C-580-818] Corrosion-Resistant Carbon... the countervailing duty (CVD) order on corrosion-resistant carbon steel flat products (CORE) from the...

  17. Calculation of gas-flow in plasma reactor for carbon partial oxidation

    Science.gov (United States)

    Bespala, Evgeny; Myshkin, Vyacheslav; Novoselov, Ivan; Pavliuk, Alexander; Makarevich, Semen; Bespala, Yuliya

    2018-03-01

    The paper discusses isotopic effects at carbon oxidation in low temperature non-equilibrium plasma at constant magnetic field. There is described routine of experiment and defined optimal parameters ensuring maximum enrichment factor at given electrophysical, gas-dynamic, and thermodymanical parameters. It has been demonstrated that at high-frequency generator capacity of 4 kW, supply frequency of 27 MHz and field density of 44 mT the concentration of paramagnetic heavy nuclei 13C in gaseous phase increases up to 1.78 % compared to 1.11 % for natural concentration. Authors explain isotopic effect decrease during plasmachemical separation induced by mixing gas flows enriched in different isotopes at the lack of product quench. With the help of modeling the motion of gas flows inside the plasma-chemical reactor based on numerical calculation of Navier-Stokes equation authors determine zones of gas mixing and cooling speed. To increase isotopic effects and proportion of 13C in gaseous phase it has been proposed to use quench in the form of Laval nozzle of refractory steel. The article represents results on calculation of optimal Laval Nozzle parameters for plasma-chemical reactor of chosen geometry of. There are also given dependences of quench time of products on pressure at the diffuser output and on critical section diameter. Authors determine the location of quench inside the plasma-chemical reactor in the paper.

  18. The possibility of tribopair lifetime extending by welding of quenched and tempered stainless steel with quenched and tempered carbon steel

    Directory of Open Access Journals (Sweden)

    V. Marušić

    2015-04-01

    Full Text Available In the conditions of tribocorrosion wear, extending of parts lifetime could be achieved by using stainless steel,which is hardened to sufficiently high hardness. In the tribosystem bolt/ bushing shell/link plate of the bucket elevator transporter conveyor machine, the previously quenched and tempered martensitic stainless steel for bolts is hardened at ≈47 HRC and welded with the quenched and tempered high yield carbon steel for bolts. Additional material, based on Cr-Ni-Mo (18/8/6 is used. The microstructure and hardness of welded samples are tested. On the tensile tester, resistance of the welded joint is tested with a simulated experiment. Dimensional control of worn tribosystem elements was performed after six months of service.

  19. Mass attenuation coefficients, effective atomic and electron numbers of stainless steel and carbon steels with different energies

    International Nuclear Information System (INIS)

    Mohd Fakarudin Abdul Rahman; Mohd Iqbal Saripan; Nor Paiza Mohamad Hasan; Ismail Mustapha

    2011-01-01

    The total mass attenuation coefficients (μ/ ρ) of stainless steel (SS316L) and carbon steel (A516) that are widely used as petrochemical plant components, such as distillation column, heat exchanger, boiler and storage tank were measured at 662, 1073 and 1332 keV of photon energies. Measurements of radiation intensity for various thicknesses of steel were made by using transmission method. The γ-ray intensity were counted by using a Gamma spectrometer that contains a Hyper-pure Germanium (HPGe) detector connected with Multi Channel Analyzer (MCA). The effective numbers of atomic (Z eff ) and electron (N eff ) obtained experimentally were compared by those obtained through theoretical calculation. Both experimental and calculated values of Z eff and N eff were in good agreement. (author)

  20. Interface Analyses Between a Case-Hardened Ingot Casting Steel and Carbon-Containing and Carbon-Free Refractories

    Science.gov (United States)

    Fruhstorfer, Jens; Dudczig, Steffen; Rudolph, Martin; Schmidt, Gert; Brachhold, Nora; Schöttler, Leandro; Rafaja, David; Aneziris, Christos G.

    2018-06-01

    Corrosion tests of carbon-free and carbon-containing refractories were performed. The carbon-free crucibles corroded, whereas the carbon-containing crucibles were negligibly attacked. On them, inclusions were attached. This study investigates melt oxygen contents, interface properties, and steel compositions with their non-metallic inclusions in order to explore the inclusion formation and deposition mechanisms. The carbon-free crucibles were based on alumina, mullite, and zirconia- and titania-doped alumina (AZT). The carbon-containing (-C) ones were alumina-C and AZT-C. Furthermore, nanoscaled carbon and alumina additives (-n) were applied in an AZT-C-n material. In the crucibles, the case-hardened steel 17CrNiMo7-6 was remelted at 1580 °C. It was observed that the melt and steel oxygen contents were higher for the tests in the carbon-free crucibles. Into these crucibles, the deoxidizing alloying elements Mn and Si diffused. Reducing contents of deoxidizing elements resulted in higher steel oxygen levels and less inclusions, mainly of the inclusion group SiO2-core-MnS-shell (2.5 to 8 μ m). These developed from smaller SiO2 nuclei. The inclusion amount in the steel was highest after remelting in AZT-C-n for 30 minutes but decreased strongly with increasing remelting time (60 minutes) due to inclusions' deposition on the refractory surface. The Ti from the AZT and the nanoadditives supported inclusion growth and deposition. Other inclusion groups were alumina and calcium aluminate inclusions. Their contents were high after remelting in carbon- or AZT-containing crucibles but generally decreased during remelting. On the AZT-C-n crucible, a dense layer formed from vitreous compositions including Al, Ca, Mg, Si, and Ti. To summarize, for reducing forming inclusion amounts, mullite is recommended as refractory material. For capturing formed inclusions, AZT-C-n showed a high potential.

  1. Irradiation, Annealing, and Reirradiation Effects on American and Russian Reactor Pressure Vessel Steels

    International Nuclear Information System (INIS)

    Chernobaeva, A.A.; Korolev, Y.N.; Nanstad, R.K.; Nikolaev, Y.A.; Sokolov, M.A.

    1998-01-01

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. Even though a postirradiation anneal may be deemed successful, a critical aspect of continued RPV operation is the rate of embrittlement upon reirradiation. There are insufficient data available to allow for verification of available models of reirradiation embrittlement or for the development of a reliable predictive methodology. This is especially true in the case of fracture toughness data. Under the U.S.-Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS), Working Group 3 on Radiation Embrittlement, Structural Integrity, and Life Extension of Reactor Vessels and Supports agreed to conduct a comparative study of annealing and reirradiation effects on RPV steels. The Working Group agreed that each side would irradiate, anneal, reirradiate (if feasible ), and test two materials of the other. Charpy V-notch (CVN) and tensile specimens were included. Oak Ridge National Laboratory (ORNL) conducted such a program (irradiation and annealing, including static fracture toughness) with two weld metals representative of VVER-440 and VVER-1000 RPVs, while the Russian Research Center-Kurchatov Institute (RRC-KI) conducted a program (irradiation, annealing, reirradiation, and reannealing) with Heavy-Section Steel Technology (HSST) Program Plate 02 and Heavy-Section Steel Irradiation (HSSI) Program Weld 73W. The results for each material from each laboratory are compared with those from the other laboratory. The ORNL experiments with the VVER welds included irradiation to about 1 x 10 19 n/cm 2 (>1 MeV), while the RRC-KI experiments with the U.S. materials included irradiations from about 2 to 18 x 10 19 n/cm 2 (>l MeV). In both cases, irradiations were conducted at ∼290 C and annealing treatments were conducted at ∼454 C. The ORNL and RRC

  2. Paying the full price of steel – Perspectives on the cost of reducing carbon dioxide emissions from the steel industry

    International Nuclear Information System (INIS)

    Rootzén, Johan; Johnsson, Filip

    2016-01-01

    This study examines the impacts felt downstream of carbon pricing and investments made in CO_2 abatement within the steel industry. Using the supply of steel to a passenger car as a case study, the effects of a steel price increase on cost structures and price at each step of the supply chain were assessed. Since the prices of emission allowances under the European Union Emissions Trading System fall well below those required to unlock investments in low-CO_2 production processes in the integrated steelmaking industry this paper seeks to pave the way for a discussion on complementary policy options. The results of the analysis suggest that passing on the compliance costs of the steel industry would have only marginal impacts on costs and prices for the end-use sectors (e.g., on the production cost or selling price of the passenger car). Under the assumptions made herein, at a carbon price of 100 €/tCO_2, the retail price of a mid-sized European passenger car would have to be increased by approximately 100–125 €/car (<0.5%) to cover the projected increases in steel production costs. - Highlights: • Examines impacts downstream of investments in CO_2 abatement in the steel industry. • Show how investing in low-CO_2 processes have marginal impacts in end-user stage. • Increase in the retail price of a mid-sized passenger car would be well below 1%. • Open up for complementary policies, financing mechanisms or new business models.

  3. Transmutation and activation of stainless steel 316 SS in a thermal fusion reactor blanket

    International Nuclear Information System (INIS)

    Gruber, J.; Schneider, J.

    1977-10-01

    Using the program MATEXP (matrix exponential method) the influence of neutron flux is calculated for stainless steel 3s16 SS which is used as a structural material in a fusion reactor blanket (CTRD-I). The transmutations, activations and γ-dose rates are determined for an operation time of 20 years. Investigating the decay behaviour after operation time, we found that the long term activity and dose rate was mainly influenced by five nuclides: Fe55, Ni63, Ni59, Co60 and Nb94. (orig.) [de

  4. Pressure vessels for reactors made from structural steel with limited tensile strength

    International Nuclear Information System (INIS)

    Machatti, H.

    1973-01-01

    The reactor pressure vessel is prestressed in several directions with prestressing elements fabricated of steel with a high yielding point. This design allows a substantial reduction of wall thickness or an increase of the inner diameter at equal wall thickness. The prestress of the prestressing elements is designed to achieve a maximum stress release of the vessel walls at normal operating conditions and to fully utilize the maximum load of the vessel walls. For safety reasons the cross section of the prestressing elements is constructed in a way that strain is always 20 % lower the yield point. (P.K.)

  5. IRRADIATION CREEP AND MECHANICAL PROPERTIES OF TWO FERRITIC-MARTENSITIC STEELS IRRADIATED IN THE BN-350 FAST REACTOR

    International Nuclear Information System (INIS)

    Porollo, S. I.; Konobeev, Yu V.; Dvoriashin, A. M.; Budylkin, N. I.; Mironova, E. G.; Leontyeva-Smirnova, M. V.; Loltukhovsky, A. G.; Bochvar, A. A.; Garner, Francis A.

    2002-01-01

    Russian ferritic/martensitic steels EP-450 and EP-823 were irradiated to 20-60 dpa in the BN-350 fast reactor in the form of pressurized creep tubes and small rings used for mechanical property tests. Data derived from these steels serves to enhance our understanding of the general behavior of this class of steels. It appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures less then 420 degrees C, but may be camouflaged somewhat by precipitation-related densification. The irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels, and that the loss of strength at test temperatures above 500 degrees C is a problem generic to all F/M steels. This conclusion is supported by post-irradiation measurement of short-term mechanical properties. At temperatures below 500 degrees C both steels retain their high strength (yield stress 0.2=550-600 MPa), but at higher test temperatures a sharp decrease of strength properties occurs. However, the irradiated steels still retain high post-irradiation ductility at test temperatures in the range of 20-700 degrees C.

  6. Preparation of diamond like carbon thin film on stainless steel and ...

    Indian Academy of Sciences (India)

    Diamond-like carbon; buffer layer; plasma CVD; surface characterization; biomedical applications. Abstract. We report the formation of a very smooth, continuous and homogeneous diamond-like carbon DLC thin coating over a bare stainless steel surface without the need for a thin Si/Cr/Ni/Mo/W/TiN/TiC interfacial layer.

  7. Corrosion of carbon steel under waste disposal conditions

    International Nuclear Information System (INIS)

    Marsh, G.

    1990-01-01

    The corrosion of carbon steel has been studied in the United Kingdom under granitic groundwater conditions, with pH between 5 and 10 and possibly substantial amounts of Cl - , SO 4 2- and HCO 3 - /CO 3 2- . Corrosion modes considered include uniform corrosion under both aerobic and anaerobic conditions; passive corrosion; localized attack in the form of pitting or crevice corrosion; and environmentally assisted cracking - hydrogen embrittlement or stress corrosion cracking. Studies of these processes are being carried out in order to predict the metal thicknesses required to give container lifetimes of 500 to 1000 years. A simple uniform corrosion model predicts a corrosion rate of around 13.4 μm/a at 20C, rising to 69 μm/a at 50C and 208 μm/a at 90C. A radiation dose of 10 5 rad/h and a G-value of 2.8 for the production of oxidizing species would account for an increase in corrosion rate of 7 μm/a. This model overestimates slightly the results actually achieved for experimental samples exposed for two years, the difference being due to a protective film formed on the samples. These corrosion rates predict that the container must be 227 mm thick to withstand uniform corrosion; however, they predict very high levels of hydrogen production. Conditions will be favourable for localized or pitting corrosion for about 125 years, leading to a maximum penetration of 160 mm. Since the exposure environment cannot be predicted precisely, one cannot state that stress corrosion cracking is impossible. Thus the container must be stress relieved. Other corrosion mechanisms such as microbial corrosion and hydrogen embrittlement are not considered significant

  8. Effect of Mo Content on Microstructure and Property of Low-Carbon Bainitic Steels

    Directory of Open Access Journals (Sweden)

    Haijiang Hu

    2016-07-01

    Full Text Available In this work, three low-carbon bainitic steels, with different Mo contents, were designed to investigate the effects of Mo addition on microstructure and mechanical properties. Two-step cooling, i.e., initial accelerated cooling and subsequent slow cooling, was used to obtain the desired bainite microstructure. The results show that the product of strength and elongation first increases and then shows no significant change with increasing Mo. Compared with Mo-free steel, bainite in the Mo-containing steel tends to have a lath-like morphology due to a decrease in the bainitic transformation temperature. More martensite transformation occurs with the increasing Mo, resulting in greater hardness of the steel. Both the strength and elongation of the steel can be enhanced by Mo addition; however, the elongation may decrease with a further increase in Mo. From a practical viewpoint, the content of Mo could be ~0.14 wt. % for the composition design of low-carbon bainitic steels in the present work. To be noted, an optimal scheme may need to consider other situations such as the role of sheet thickness, toughness behavior and so on, which could require changes in the chemistry. Nevertheless, these results provide a reference for the composition design and processing method of low-carbon bainitic steels.

  9. A computational model for the carbon transfer in stainless steel sodium systems

    International Nuclear Information System (INIS)

    Casadio, S.; Scibona, G.

    1980-01-01

    A method is proposed of computing the carbon transfer in the type 316, 304 and 321 stainless steels in sodium environment as a function of temperature, exposure time and carbon concentration in the sodium. The method is based on the criteria developed at ANL by introducing some simplifications and takes also into account the correlations obtained at WARD. Calculated carbon profiles are compared both with experimental data and with the results available by the other computer methods. The limits for quantitative predictions of the stainless steel carburization or decarburization exposed in a specific environment are discussed. (author)

  10. Can Thermally Sprayed Aluminum (TSA) Mitigate Corrosion of Carbon Steel in Carbon Capture and Storage (CCS) Environments?

    Science.gov (United States)

    Paul, S.; Syrek-Gerstenkorn, B.

    2017-01-01

    Transport of CO2 for carbon capture and storage (CCS) uses low-cost carbon steel pipelines owing to their negligible corrosion rates in dry CO2. However, in the presence of liquid water, CO2 forms corrosive carbonic acid. In order to mitigate wet CO2 corrosion, use of expensive corrosion-resistant alloys is recommended; however, the increased cost makes such selection economically unfeasible; hence, new corrosion mitigation methods are sought. One such method is the use of thermally sprayed aluminum (TSA), which has been used to mitigate corrosion of carbon steel in seawater, but there are concerns regarding its suitability in CO2-containing solutions. A 30-day test was carried out during which carbon steel specimens arc-sprayed with aluminum were immersed in deionized water at ambient temperature bubbled with 0.1 MPa CO2. The acidity (pH) and potential were continuously monitored, and the amount of dissolved Al3+ ions was measured after completion of the test. Some dissolution of TSA occurred in the test solution leading to nominal loss in coating thickness. Potential measurements revealed that polarity reversal occurs during the initial stages of exposure which could lead to preferential dissolution of carbon steel in the case of coating damage. Thus, one needs to be careful while using TSA in CCS environments.

  11. The Influence of Calcium Carbonate Composition and Activated Carbon in Pack Carburizing Low Carbon Steel Process in The Review of Hardness and Micro Structure

    Science.gov (United States)

    Hafni; Hadi, Syafrul; Edison

    2017-12-01

    Carburizing is a way of hardening the surface by heating the metal (steel) above the critical temperature in an environment containing carbon. Steel at a temperature of the critical temperature of affinity to carbon. Carbon is absorbed into the metal form a solid solution of carbon-iron and the outer layer has high carbon content. When the composition of the activator and the activated charcoal is right, it will perfect the carbon atoms to diffuse into the test material to low carbon steels. Thick layer of carbon Depending on the time and temperature are used. Pack carburizing process in this study, using 1 kg of solid carbon derived from coconut shell charcoal with a variation of 20%, 10% and 5% calcium carbonate activator, burner temperature of 950 0C, holding time 4 hours. The test material is low carbon steel has 9 pieces. Each composition has three specimens. Furnace used in this study is a pack carburizing furnace which has a designed burner box with a volume of 1000 x 600 x 400 (mm3) of coal-fired. Equipped with a circulation of oxygen from the blower 2 inches and has a wall of refractory bricks. From the variation of composition CaCO3, microstructure formed on the specimen with 20% CaCO3, better diffusion of carbon into the carbon steel, it is seen by the form marten site structure after quenching, and this indicates that there has been an increase of or adding carbon to in the specimen. This led to the formation of marten site specimen into hard surfaces, where the average value of hardness at one point side (side edge) 31.7 HRC

  12. Lower Length Scale Model Development for Embrittlement of Reactor Presure Vessel Steel

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yongfeng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schwen, Daniel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Chakraborty, Pritam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, Xianming [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    This report summarizes the lower-length-scale effort during FY 2016 in developing mesoscale capabilities for microstructure evolution, plasticity and fracture in reactor pressure vessel steels. During operation, reactor pressure vessels are subject to hardening and embrittlement caused by irradiation induced defect accumulation and irradiation enhanced solute precipitation. Both defect production and solute precipitation start from the atomic scale, and manifest their eventual effects as degradation in engineering scale properties. To predict the property degradation, multiscale modeling and simulation are needed to deal with the microstructure evolution, and to link the microstructure feature to material properties. In this report, the development of mesoscale capabilities for defect accumulation and solute precipitation are summarized. A crystal plasticity model to capture defect-dislocation interaction and a damage model for cleavage micro-crack propagation is also provided.

  13. Fast reactor shield sensitivity studies for steel--sodium--iron systems

    International Nuclear Information System (INIS)

    Oblow, E.M.; Weisbin, C.R.

    1977-01-01

    A study was made of the adequacy of the current ENDF/B-IV sodium and iron neutron cross section data files for fast reactor shield design work. Experimental data from 21 fast reactor shield configurations containing large thicknesses of steel, sodium, and iron were analyzed with discrete ordinates calculations and sensitivity methods to assess the data files. This study represents the largest full-scale sensitivity analysis of benchmark quality experimental data to date. Included in the sensitivity studies were the results of the new cross section adjustment algorithms added to the FORSS code system. Conclusions were drawn about the need for more accurate data for sodium and iron elastic and discrete inelastic cross sections above 1 MeV and the values of the total cross section in the vicinity of important minima

  14. Creep crack growth in a reactor pressure vessel steel at 360 deg C

    Energy Technology Data Exchange (ETDEWEB)

    Rui Wu; Seitisleam, F.; Sandstroem, R. [Swedish Institute for Metals Research, Stockholm (Sweden)

    1998-12-31

    Plain creep (PC) and creep crack growth (CCG) tests at 360 deg C and post metallography were carried out on a low alloy reactor pressure vessel steel (ASTM A508 class 2) with different microstructures. Lives for the CCG tests were shorter than those for the PC tests and this is more pronounced for simulated heat affected zone microstructure than for the parent metal at longer lives. For the CCG tests, after initiation, the cracks grew constantly and intergranularly before they accelerated to approach rupture. The creep crack growth rate is well described by C*. The relations between reference stress, failure time and steady crack growth rate are presented for the CCG tests. It is demonstrated that the failure stress due to CCG is considerably lower than the yield stress at 360 deg C. Consequently, the CCG will control the static strength of a reactor vessel. (orig.) 17 refs.

  15. Creep crack growth in a reactor pressure vessel steel at 360 deg C

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Rui; Seitisleam, F; Sandstroem, R [Swedish Institute for Metals Research, Stockholm (Sweden)

    1999-12-31

    Plain creep (PC) and creep crack growth (CCG) tests at 360 deg C and post metallography were carried out on a low alloy reactor pressure vessel steel (ASTM A508 class 2) with different microstructures. Lives for the CCG tests were shorter than those for the PC tests and this is more pronounced for simulated heat affected zone microstructure than for the parent metal at longer lives. For the CCG tests, after initiation, the cracks grew constantly and intergranularly before they accelerated to approach rupture. The creep crack growth rate is well described by C*. The relations between reference stress, failure time and steady crack growth rate are presented for the CCG tests. It is demonstrated that the failure stress due to CCG is considerably lower than the yield stress at 360 deg C. Consequently, the CCG will control the static strength of a reactor vessel. (orig.) 17 refs.

  16. A STUDY OF CORROSION AND STRESS CORROSION CRACKING OF CARBON STEEL NUCLEAR WASTE STORAGE TANKS

    International Nuclear Information System (INIS)

    BOOMER, K.D.

    2007-01-01

    The Hanford reservation Tank Farms in Washington State has 177 underground storage tanks that contain approximately 50 million gallons of liquid legacy radioactive waste from cold war plutonium production. These tanks will continue to store waste until it is treated and disposed. These nuclear wastes were converted to highly alkaline pH wastes to protect the carbon steel storage tanks from corrosion. However, the carbon steel is still susceptible to localized corrosion and stress corrosion cracking. The waste chemistry varies from tank to tank, and contains various combinations of hydroxide, nitrate, nitrite, chloride, carbonate, aluminate and other species. The effect of each of these species and any synergistic effects on localized corrosion and stress corrosion cracking of carbon steel have been investigated with electrochemical polarization, slow strain rate, and crack growth rate testing. The effect of solution chemistry, pH, temperature and applied potential are all considered and their role in the corrosion behavior will be discussed

  17. Low cycle thermomechanical fatigue of reactor steels: Microstructural and fractographic investigations

    Energy Technology Data Exchange (ETDEWEB)

    Fekete, Balazs, E-mail: fekete.mm.bme@gmail.com [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary); Department of Applied Mechanics, Budapest University of Technology and Economics, Muegyetem 5, Budapest H-1111 (Hungary); Kasl, Josef; Jandova, Dagmar [Výzkumný a zkušební ústav Plzeň s.r.o., Tylova 1581/46, 316 00 Plzen (Czech Republic); Jóni, Bertalan [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary); Eötvös Loránd University, Egyetem tér 1-3, Budapest H-1053 (Hungary); Misják, Fanni [Centre for Energy Research, Institute of Technical Physics and Materials Science, Konkoly-Thege M. 29-33, Budapest H-1121 (Hungary); Trampus, Peter [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary)

    2015-07-29

    The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of a VVER-440 reactor pressure vessel were investigated under fully reversed total strain controlled low cycle fatigue tests. The measurements were carried out in isothermal conditions at 260 °C and with thermal-mechanical conditions in the range 150–270 °C using a GLEEBLE-3800 servo-hydraulic thermal-mechanical simulator. The low cycle fatigue results were evaluated with the Coffin–Manson law, and the parameters of the Ramberg–Osgood stress–strain relation were investigated. Fracture mechanics behavior was observed using scanning electron microscopic analysis of the crack shapes and fracture surfaces. Crack propagation was assessed in relation to the actual crack size and the loading level. Interrupted fatigue tests were also carried out to investigate the kinetics of the fatigue evolution of the materials. Microstructural evaluation of the samples was performed using light, scanning and transmission electron microscopy as well as X-ray diffraction, and measurement of dislocations was completed using TEM and XRD. The course of dislocation density in relation to cumulative usage factor was similar for both steels. However, the nature and distribution of dislocations were different in the individual steels and this resulted in different mechanical behaviors. The nature of the fracture surfaces of both steels appeared similar despite differences in dislocation arrangement. The distances between striation lines initially increased with increasing crack length and then became saturated. The low cycle fatigue behavior investigated can provide a reference for the remaining life assessment and lifetime extension analysis of nuclear power plant components.

  18. 76 FR 31938 - Certain Hot-Rolled Carbon Steel Flat Products From India: Notice of Preliminary Results of 2009...

    Science.gov (United States)

    2011-06-02

    ... the File from Christopher Hargett, International Trade Compliance Analyst, through Melissa Skinner... Skinner, Office Director, concerning ``Certain Hot-Rolled Carbon Steel Flat Products from India: Customs...

  19. Nano-scaled iron-carbon precipitates in HSLC and HSLA steels

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    This paper studies the composition, quantity and particle size distribution of nano-scaled precipitates with size less than 20 nm in high strength low carbon (HSLC) steel and their effects on mechanical properties of HSLC steel by means of mass balance calculation of nano-scaled precipitates measured by chemical phase analysis plus SAXS method, high-resolution TEM analysis and thermodynamics calculation, as well as temper rapid cooling treatment of ZJ330. It is found that there existed a large quantity of nano-scaled iron-carbon precipitates with size less than 18 nm in low carbon steel produced by CSP and they are mainly Fe-O-C and Fe-Ti-O-C precipitates formed below temperature A1. These precipitates have ob- vious precipitation strengthening effect on HSLC steel and this may be regarded as one of the main reasons why HSLC steel has higher strength. There also existed a lot of iron-carbon precipitates with size less than 36 nm in HSLA steels.

  20. Nano-scaled iron-carbon precipitates in HSLC and HSLA steels

    Institute of Scientific and Technical Information of China (English)

    FU Jie; WU HuaJie; LIU YangChun; KANG YongLin

    2007-01-01

    This paper studies the composition, quantity and particle size distribution of nano-scaled precipitates with size less than 20 nm in high strength Iow carbon (HSLC) steel and their effects on mechanical properties of HSLC steel by means of mass balance calculation of nano-scaled precipitates measured by chemical phase analysis plus SAXS method, high-resolution TEM analysis and thermodynamics calculation, as well as temper rapid cooling treatment of ZJ330. It is found that there existed a large quantity of nano-scaled iron-carbon precipitates with size less than 18 nm in Iow carbon steel produced by CSP and they are mainly Fe-O-C and Fe-Ti-O-C precipitates formed below temperature A1. These precipitates have obvious precipitation strengthening effect on HSLC steel and this may be regarded as one of the main reasons why HSLC steel has higher strength. There also existed a lot of iron-carbon precipitates with size less than 36 nm in HSLA steels.

  1. Ferrite morphology and residual phases in continuously cooled low carbon steels

    International Nuclear Information System (INIS)

    Dunne, D.P.

    1999-01-01

    Although much research has been conducted on the isothermal transformation products of medium to high carbon hardenable steels, relatively little has been reported for transformation of low carbon structural steels under continuous cooling conditions. The trend towards reduced carbon levels (less than about 0.1 wt% C) has been driven by demands for formability and weldability, challenging steel designers to maintain strength by microalloying and/or thermomechanical controlled processing. Although control of the ferritic products formed in low carbon steels after hot rolling, normalising and welding is essential in order to ensure adequate strength and toughness, understanding of the microstructures formed on continuous cooling is still limited. In addition, transformation mechanisms remain controversial because of polarisation of researchers into groups championing diffusional and displacive theories for the transformation of austenite over a wide range of cooling rates. The present review compares and draws together the main ferrite classification schemes, and discusses some critical issues on kinetics and mechanisms, in an attempt to rationalise the effects of cooling rate, prior austenite structure and composition on the resulting ferrite structure and its mechanical properties. It is concluded that with increasing cooling rate the ferritic product becomes finer, more plate-like, more dislocated, more carbon supersaturated, more likely to be formed by a displacive mechanism, harder and stronger. Other conclusions are that: (i) 'bainitic ferrite', which is a pervasive form of ferrite in continuously cooled low carbon steels, is different from the conventional upper and lower bainites observed in higher carbon steels, insofar as the co-product 'phase' is typically martensite-austenite islands rather than cementite; and (ii) low carbon bainite rather than martensite is the dominant product at typical fast cooling rates (<500K/s) associated with commercial

  2. Controlling corrosion of carbon steel in cooling water applications -- A novel environmentally acceptable approach

    International Nuclear Information System (INIS)

    Banerjee, G.; Miller, A.E.

    1998-01-01

    Cr(VI) containing salts have been in use for a long time as one of the best inhibitors for minimizing corrosion of carbon steel in cooling water applications. Irrespective of the type of system, i.e., once through, open recirculating, pressurized water reactor power plants, etc. and irrespective of the conductivity of water, i.e., low or high, Cr(VI) salts always have proven to be very effective inhibitors. However, the toxicity of chromate compounds and the consequential disposal difficulties have made it essential to look for an alternate treatment. It is however, imperative that the alternate system must provide the matching efficiency as that provided by Cr(VI) salts and that it should also be easy to maintain and be economical. While many researchers have been trying to find a suitable chromate free inhibitor system, the present authors have explored the possibility of formulating an inhibitor system containing Cr(VI) at a concentration below the safety limit for drinking water as suggested by EPA/OSHA. This is based on the assumption that EPA (Environmental Protection Agency) and OSHA (Occupational Safety and Health Administration) only regulate the discharge and exposure limits of chromium above which it is found harmful. Therefore, any new formulation containing Cr(VI) well below these safety limits should be acceptable environmentally. If such a formulation can perform similar to ones with high concentration of Cr(VI), it will also be commercially acceptable. The authors will discuss the preliminary results of such a strategy

  3. Evaluation of elastic-plastic fracture of toughness and fracture resistance of carbon steel STS42

    International Nuclear Information System (INIS)

    Kobayashi, Hideo; Nakamura, Haruo; Kashiwagi, Kohmei

    1987-01-01

    The elastic-plastic fracture toughness (J Ic ) and fracture resistance (J-R curve) of a carbon steel, STS42, used for piping in a nuclear reactor were evaluated according to the several evaluating methods recommended or proposed so far, to discuss their applicability and utility. The results obtained are as follows: (1) In evaluating J Ic , the multiple specimen method recommended by the Japan Society for Mechanical Engineers (JSME standard S001) gives the most reliable results by using smaller sized specimens. (2) The single-specimen methods by using the compliance technique, adopted in the ASTM standards (E813, E813 modified, Tentative test procedure for determining the plain strain J-R curve), do not give an accurate J-R curve or J Ic , due to an error in the calculated crack length. (3) In evaluating the J-R curve, it is necessary to account for crack extension in calculating the J-integral. (4) According to the above results, a new standard method for determining the J-R curve including the J Ic test method should be poprosed. (author)

  4. Chemistry of the aqueous medium - Determining factor of corrosion in carbon steel components of secondary circuit

    International Nuclear Information System (INIS)

    Radulescu, M.; Pirvan, I.; Dinu, A.; Velciu, L.

    2003-01-01

    The interplay of chemistry of aqueous medium and corrosion processes followed by deposition and/or release of corrosion products determines both formation and growth of superficial films as well as the kinetics of ion release from materials into the aqueous medium. Material corrosion in the secondary circuit of a NPP can be minimized by choosing the materials of the components and by a rigorous inspection of the chemistry of aqueous agent. The chemical inspection helps in minimizing: - the corrosion of the components immersed in feedwater and vapor and of Steam Generator components; - 'dirtying' of the systems particularly of the surfaces implied in heat transfer; - the amount of insoluble chemical species resulting in corrosion process and carried along the circuit; - the corrosion of secondary circuit components during revisions or outages. An important role among the chemical parameters of the fluids circulated in NPP tubing appears to be the pH. In CANDU reactors it must be kept within the range of 8.7 to 9.4 by treating the medium with volatile amines (morpholine and cyclohexylamine). A plot is presented giving the corrosion rate of carbon steels as a function of the pH of the medium. Besides, the oxygen concentration dissolved in the aqueous medium must be maintained under 5 μg per water kg. Other factors determining the corrosion rates are also discussed. The paper gives the results of the experiments done with various materials, solutions and analysis methods

  5. A new method for surface modifications of carbon steels and alloys

    Directory of Open Access Journals (Sweden)

    Valeriy Dondokovich Dugar-Zhabon

    2012-12-01

    Full Text Available A three-dimensional treatment method involving implantation of ions into solids immersed in a high voltage pulse discharge ignited on the left-hand-branch of the Paschen curve was elaborated about fifteen years ago. This method, named 3DII for short, has been used in the equipment JUPITER (Joint Universal Plasma and Ion Technologies Experimental Reactor for practical purposes. Hereafter, the need for better means to improve the metal surface protection against aggressive media prompted an elaboration of the MOSMET concept which is based on a hybrid treatment involving the processes of implantation and deposition. It is significant that the processes can be set into action simultaneously or separately. In this article, the conditions of hybrid treatment of AISI SAE 1010, 1020 y 1045 carbon steels, their subsequent electrochemical diagnostics and corrosion test results are described. The corrosion rate of the samples treated by titanium hybrid discharge is found approximately an order of magnitude smaller as compared to the non-treated samples.

  6. Formation of fouling deposits on a carbon steel surface from Colombian heavy crude oil under preheating conditions

    Science.gov (United States)

    Muñoz Pinto, D. A.; Cuervo Camargo, S. M.; Orozco Parra, M.; Laverde, D.; García Vergara, S.; Blanco Pinzon, C.

    2016-02-01

    Fouling in heat exchangers is produced by the deposition of undesired materials on metal surfaces. As fouling progresses, pressure drop and heat transfer resistance is observed and therefore the overall thermal efficiency of the equipment diminishes. Fouling is mainly caused by the deposition of suspended particles, such as those from chemical reactions, crystallization of certain salts, and some corrosion processes. In order to understand the formation of fouling deposits from Colombian heavy oil (API≈12.3) on carbon steel SA 516 Gr 70, a batch stirred tank reactor was used. The reactor was operated at a constant pressure of 340psi while varying the temperature and reaction times. To evaluate the formation of deposits on the metal surfaces, the steel samples were characterized by gravimetric analysis and Scanning Electron Microscopy (SEM). On the exposed surfaces, the results revealed an increase in the total mass derived from the deposition of salt compounds, iron oxides and alkaline metals. In general, fouling was modulated by both the temperature and the reaction time, but under the experimental conditions, the temperature seems to be the predominant variable that controls and accelerates fouling.

  7. Variation of transition temperatures from upper to lower bainites in plain carbon steels

    International Nuclear Information System (INIS)

    Oka, M.; Okamoto, H.

    1995-01-01

    Experimental results and explanations for the transition temperature from upper to lower bainites in carbon steels containing from 0.20 to 1.80 wt%C were presented metallographically and kinematically. The experimental results are summarized as follows: (1) Lower bainite is not formed in steels with less than 0.35 wt%C and no transition from upper to lower bainite occurs. (2) The transition temperature of steels containing from 0.54 to 1.10 %C indicates a constant temperature of 350 C and does not depend on the carbon content. It is important to note that a transition temperature of 350 C corresponds to the Ms temperature of a 0.55%C steel being the boundary of the martensite morphology between a lath and a plate. (3) Transition temperatures of steels with more than 1.10%C decrease along the a line below about 65 C from T 0 -composition line. The bainitic transformation is essentially a kind of the martensitic one and its nucleation site is considered to be a carbon depleted zone in austenite by the thermal fluctuation of carbon atom at an isothermal holding temperature. The supercooling of about 65 C below the T 0 -composition line at the carbon range more than 1.10 wt%C is attributed to the non-chemical free energy for the displacive growth of lower bainite. (orig.)

  8. The effects of bacteria on the corrosion behavior of carbon steel in compacted bentonite

    International Nuclear Information System (INIS)

    Nishimura, T.; Wada, R.; Nishimoto, H.; Fujiwara, K.; Taniguchi, N.; Honda, A.

    1999-10-01

    As a part of evaluation of corrosion life of carbon steel overpack, the experimental studies have been performed on the effects of bacteria on the corrosion behavior of carbon steel in compacted bentonite using iron bacteria (IB) as a representative oxidizing bacteria and sulphur reducing bacteria (SRB) as a representative reducing bacteria. The results of the experimental studies showed that; The activity of SRB was low in compacted bentonite in spite of applying suitable condition for the action of bacteria such as temperature and nutritious solution. Although the corrosion behavior of carbon steel was affected by the existence of bacteria in simple solution, the corrosion rates of carbon steel in compacted bentonite were several μ m/year -10 μ m/year irrespective of coexistence of bacteria and that the corrosion behavior was not affected by the existence of bacteria. According to these results, it was concluded that the bacteria would not affect the corrosion behavior of carbon steel overpack under repository condition. (author)

  9. Evaluation of carbon diffusion in heat treatment of H13 tool steel under different atmospheric conditions

    Directory of Open Access Journals (Sweden)

    Maziar Ramezani

    2015-04-01

    Full Text Available Although the cost of the heat treatment process is only a minor portion of the total production cost, it is arguably the most important and crucial stage on the determination of material quality. In the study of the carbon diffusion in H13 steel during austenitization, a series of heat treatment experiments had been conducted under different atmospheric conditions and length of treatment. Four austenitization atmospheric conditions were studied, i.e., heat treatment without atmospheric control, heat treatment with stainless steel foil wrapping, pack carburization heat treatment and vacuum heat treatment. The results showed that stainless steel foil wrapping could restrict decarburization process, resulting in a constant hardness profile as vacuum heat treatment does. However, the tempering characteristic between these two heat treatment methods is different. Results from the gas nitrided samples showed that the thickness and the hardness of the nitrided layer is independent of the carbon content in H13 steel.

  10. Corrosion of stainless and carbon steels in molten mixtures of industrial nitrates

    Energy Technology Data Exchange (ETDEWEB)

    Goods, S.H.; Bradshaw, R.W. [Sandia National Labs., Livermore, CA (United States); Prairie, M.R.; Chavez, J.M. [Sandia National Labs., Albuquerque, NM (United States)

    1994-03-01

    Corrosion behavior of two stainless steels and carbon steel in mixtures of NaNO{sub 3} and KNO{sub 3} was evaluated to determine if impurities found in commodity grades of alkali nitrates aggravate corrosivity as applicable to an advanced solar thermal energy system. Corrosion tests were conducted for 7000 hours with Types 304 and 316 stainless steels at 570C and A36 carbon steel at 316C in seven mixtures of NaNO{sub 3} and KNO{sub 3} containing variations in impurity concentrations. Corrosion tests were also conducted in a ternary mixture of NaNO{sub 3}, KNO{sub 3}, and Ca(NO{sub 3}){sub 2}. Corrosion rates were determined by descaled weight losses while oxidation products were examined by scanning electron microscopy, electron microprobe analysis, and X-ray diffraction. The nitrate mixtures were periodically analyzed for changes in impurity concentrations and for soluble corrosion products.

  11. A fracture mechanics approach for estimating fatigue crack initiation in carbon and low-alloy steels in LWR coolant environments

    International Nuclear Information System (INIS)

    Park, H. B.; Chopra, O. K.

    2000-01-01

    A fracture mechanics approach for elastic-plastic materials has been used to evaluate the effects of light water reactor (LWR) coolant environments on the fatigue lives of carbon and low-alloy steels. The fatigue life of such steel, defined as the number of cycles required to form an engineering-size crack, i.e., 3-mm deep, is considered to be composed of the growth of (a) microstructurally small cracks and (b) mechanically small cracks. The growth of the latter was characterized in terms of ΔJ and crack growth rate (da/dN) data in air and LWR environments; in water, the growth rates from long crack tests had to be decreased to match the rates from fatigue S-N data. The growth of microstructurally small cracks was expressed by a modified Hobson relationship in air and by a slip dissolution/oxidation model in water. The crack length for transition from a microstructurally small crack to a mechanically small crack was based on studies on small crack growth. The estimated fatigue S-N curves show good agreement with the experimental data for these steels in air and water environments. At low strain amplitudes, the predicted lives in water can be significantly lower than the experimental values

  12. Influence of crack depth on the fracture toughness of reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Theiss, T.J.; Bryson, J.W.

    1991-01-01

    The Heavy Section Steel Technology Program (HSST) at Oak Ridge National Laboratory (ORNL) is investigating the influence of flaw depth on the fracture toughness of reactor pressure vessel (RPV) steel. Recently, it has been shown that, in notched beam testing, shallow cracks tend to exhibit an elevated toughness as a result of a loss of constraint at the crack tip. The loss of constraint takes place when interaction occurs between the elastic-plastic crack-tip stress field and the specimen surface nearest the crack tip. An increased shallow-crack fracture toughness is of interest to the nuclear industry because probabilistic fracture-mechanics evaluations show that shallow flaws play a dominant role in the probability of vessel failure during postulated pressurized-thermal-shock (PTS) events. Tests have been performed on beam specimens loaded in 3-point bending using unirradiated reactor pressure vessel material (A533 B). Testing has been conducted using specimens with a constant beam depth (W = 94 mm) and within the lower transition region of the toughness curve for A533 B. Test results indicate a significantly higher fracture toughness associated with the shallow flaw specimens compared to the fracture toughness determined using deep-crack (a/W = 0.5) specimens. Test data also show little influence of thickness on the fracture toughness for the current test temperature (-60 degree C). 21 refs., 5 figs., 3 tabs

  13. The tensile and fatigue properties of type 1.4914 ferritic steel for fusion reactor applications

    International Nuclear Information System (INIS)

    Marmy, P.; Victoria, M.; Ruan, Y.

    1989-08-01

    Martensitic steels have received considerable attention as structural materials in fusion reactor applications. In present designs, fusion reactors are expected to operate in a cyclic mode, thus producing cyclic thermal stresses in the first wall. Due to its thermal expansion coefficient and very low swelling rate, 1.4914 martensitic steel is a suitable candidate for the first wall with high neutron loadings. This paper presents the preirradiation results obtained with subsize-specimens designed to be irradiated with a proton beam in the PIREX facility at the Paul Scherrer Institute (PSI) of Wuerenlingen. Both tensile and low cycle fatigue tests were performed in vacuum in the region from 300 K to 870 K (720 K in the case of fatigue tests). Tensile tests on the subsize specimens (0.33 mm thick) compared well to those on bulk specimens, showing a minimum in ductility at around 620 K. The fatigue tests, performed on tubular specimens (3.4 mm external diameter, 0.35 mm wall thickness) showed substantial softening setting in at a low number of cycles. The initial microstructure observed in transmission microscopy consists of fine martensite laths. As cyclic deformation proceeds, dislocation cells form, that gradually replace the martensitic laths. (author) 19 figs., 5 tabs., 16 refs

  14. On flux effects in a low alloy steel from a Swedish reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Boåsen, Magnus, E-mail: boasen@kth.se [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Efsing, Pål [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Ehrnstén, Ulla [VTT Technical Research Centre of Finland Ltd, PO Box 1000, FI-02044 VTT (Finland)

    2017-02-15

    This study aims to investigate the presence of Unstable Matrix Defects in irradiated pressure vessel steel from weldments of the Swedish PWR Ringhals 4 (R4). Hardness tests have been performed on low flux (surveillance material) and high flux (Halden reactor) irradiated material samples in combination with heat treatments at temperatures of 330, 360 and 390 °C in order to reveal eventual recovery of any hardening features induced by irradiation. The experiments carried out in this study could not reveal any hardness recovery related to Unstable Matrix Defects at relevant temperatures. However, a difference in hardness recovery was found between the low and the high flux samples at heat treatments at higher temperatures than expected for the annihilation of Unstable Matrix Defects–the observed recovery is here attributed to differences of the solute clusters formed by the high and low flux irradiations. - Highlights: • Hardness testing is combined with post irradiation annealing at 330, 360 and 390 °C. • Unstable matrix defects is studied in a reactor pressure vessel steel. • Comparison between surveillance material and accelerated irradiation. • No evidence of unstable matrix defects, i.e. not present in studied material. • Difference in hardness recovery between irradiation conditions found at 390 °C.

  15. Effects of thermal annealing and reirradiation on toughness of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Nanstad, R.K.; Iskander, S.K.; Sokolov, M.A.

    1996-01-01

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes recent experimental results from work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response, or open-quotes recovery,close quotes of several irradiated RPV steels; it also includes recent results from both ORNL and the Russian Research Center-Kurchatov Institute (RRC-KI) on a cooperative program of irradiation, annealing and reirradiation of both U.S. and Russian RPV steels. The cooperative program was conducted under the auspices of Working Group 3, U.S./Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS). The materials investigated are an RPV plate and various submerged-arc welds, with tensile, Charpy impact toughness, and fracture toughness results variously determined. Experimental results are compared with applicable prediction guidelines, while observed differences in annealing responses and reirradiation rates are discussed

  16. Effects of thermal annealing and reirradiation on toughness of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Nanstad, R.K.; Iskander, S.K.; Sokolov, M.A.

    1997-01-01

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes recent experimental results from work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response, or open-quotes recovery,close quotes of several irradiated RPV steels; it also includes recent results from both ORNL and the Russian Research Center-Kurchatov Institute (RRC-KI) on a cooperative program of irradiation, annealing and reirradiation of both U.S. and Russian RPV steels. The cooperative program was conducted under the auspices of Working Group 3, U.S./Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS). The materials investigated are an RPV plate and various submerged-arc welds, with tensile, Charpy impact toughness, and fracture toughness results variously determined. Experimental results are compared with applicable prediction guidelines, while observed differences in annealing responses and reirradiation rates are discussed

  17. Statistical properties of material strength for reliability evaluation of components of fast reactors. Austenitic stainless steels

    International Nuclear Information System (INIS)

    Takaya, Shigeru; Sasaki, Naoto; Tomobe, Masato

    2015-03-01

    Many efforts have been made to implement the System Based Code concept of which objective is to optimize margins dispersed in several codes and standards. Failure probability is expected to be a promising quantitative index for optimization of margins, and statistical information for random variables is needed to evaluate failure probability. Material strength like tensile strength is an important random variable, but the statistical information has not been provided enough yet. In this report, statistical properties of material strength such as creep rupture time, steady creep strain rate, yield stress, tensile stress, flow stress, fatigue life and cyclic stress-strain curve, were estimated for SUS304 and 316FR steel, which are typical structural materials for fast reactors. Other austenitic stainless steels like SUS316 were also used for statistical estimation of some material properties such as fatigue life. These materials are registered in the JSME code of design and construction of fast reactors, so test data used for developing the code were used as much as possible in this report. (author)

  18. Mechanical Properties of Advanced Gas-Cooled Reactor Stainless Steel Cladding After Irradiation

    Science.gov (United States)

    Degueldre, Claude; Fahy, James; Kolosov, Oleg; Wilbraham, Richard J.; Döbeli, Max; Renevier, Nathalie; Ball, Jonathan; Ritter, Stefan

    2018-05-01

    The production of helium bubbles in advanced gas-cooled reactor (AGR) cladding could represent a significant hazard for both the mechanical stability and long-term storage of such materials. However, the high radioactivity of AGR cladding after operation presents a significant barrier to the scientific study of the mechanical properties of helium incorporation, said cladding typically being analyzed in industrial hot cells. An alternative non-active approach is to implant He2+ into unused AGR cladding material via an accelerator. Here, a feasibility study of such a process, using sequential implantations of helium in AGR cladding steel with decreasing energy is carried out to mimic the buildup of He (e.g., 50 appm) that would occur for in-reactor AGR clad in layers of the order of 10 µm in depth, is described. The implanted sample is subsequently analyzed by scanning electron microscopy, nanoindentation, atomic force and ultrasonic force microscopies. As expected, the irradiated zones were affected by implantation damage (steel cladding is retained despite He2+ implantation.

  19. Effect of Ethanol Chemistry on SCC of Carbon Steel

    Science.gov (United States)

    2011-02-22

    Pipeline companies have a keen interest in assessing the feasibility of transporting fuel grade ethanol (FGE) and ethanol blends in existing pipelines. Previous field experience and laboratory research, funded by PRCI and API, has shown that steel ca...

  20. Modeling flow stress constitutive behavior of SA508-3 steel for nuclear reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Sun Mingyue, E-mail: mysun@imr.ac.cn [Shenyang National Laboratory for Materials Science, Institute of Metal Research, Chinese Academy of Sciences, 72 Wenhua Road, Shenyang 110016 (China); Luhan, Hao; Shijian, Li; Dianzhong, Li; Yiyi, Li [Shenyang National Laboratory for Materials Science, Institute of Metal Research, Chinese Academy of Sciences, 72 Wenhua Road, Shenyang 110016 (China)

    2011-11-15

    Highlights: > A series of flow stress constitutive equations for SA508-3 steel were successfully established. > The experimental results under different conditions have validated the constitutive equations. > An industrial application of the model was present to simulate a large conical shell forging process. - Abstract: Based on the measured stress-strain curves under different temperatures and strain rates, a series of flow stress constitutive equations for SA508-3 steel were firstly established through the classical theories on work hardening and softening. The comparison between the experimental and modeling results has confirmed that the established constitutive equations can correctly describe the mechanical responses and microstructural evolutions of the steel under various hot deformation conditions. We further represented a successful industrial application of this model to simulate a forging process for a large conical shell used in a nuclear steam generator, which evidences its practical and promising perspective of our model with an aim of widely promoting the hot plasticity processing for heavy nuclear components of fission reactors.

  1. Modeling flow stress constitutive behavior of SA508-3 steel for nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Sun Mingyue; Hao Luhan; Li Shijian; Li Dianzhong; Li Yiyi

    2011-01-01

    Highlights: → A series of flow stress constitutive equations for SA508-3 steel were successfully established. → The experimental results under different conditions have validated the constitutive equations. → An industrial application of the model was present to simulate a large conical shell forging process. - Abstract: Based on the measured stress-strain curves under different temperatures and strain rates, a series of flow stress constitutive equations for SA508-3 steel were firstly established through the classical theories on work hardening and softening. The comparison between the experimental and modeling results has confirmed that the established constitutive equations can correctly describe the mechanical responses and microstructural evolutions of the steel under various hot deformation conditions. We further represented a successful industrial application of this model to simulate a forging process for a large conical shell used in a nuclear steam generator, which evidences its practical and promising perspective of our model with an aim of widely promoting the hot plasticity processing for heavy nuclear components of fission reactors.

  2. Characterization of matrix damage in ion-irradiated reactor vessel steel

    International Nuclear Information System (INIS)

    Fujii, Katsuhiko; Fukuya, Koji

    2004-01-01

    Exact nature of the matrix damage, that is one of radiation-induced nano-scale microstructural features causing radiation embrittlement of reactor vessel, in irradiated commercial steels has not been clarified yet by direct characterization using transmission electron microscopy (TEM). We designed a new preparation method of TEM observation samples and applied it to the direct TEM observation of the matrix damage in the commercial steel samples irradiated by ions. The simulation irradiation was carried out by 3 MeV Ni 2+ ion to a dose of 1 dpa at 290degC. Thin foil specimens for TEM observation were prepared using the modified focused ion beam method. A weak-beam TEM study was carried out for the observation of matrix damage in the samples. Results of this first detailed observation of the matrix damage in the irradiated commercial steel show that it is consisted of small dislocation loops. The observed and analyzed dislocation loops have Burgers vectors b = a , and a mean image size and the number density are 2.5 nm and about 1 x 10 22 m -3 , respectively. In this experiment, all of the observed dislocation loops were too small to determine the vacancy or interstitial nature of the dislocation loops directly. Although it is an indirect method, post-irradiation annealing was used to infer the loop nature. Most of dislocation loops were stable after the annealing at 400degC for 30 min. This result suggests that their nature is interstitial. (author)

  3. X-ray photoelectron spectroscopy characterization of high dose carbon-implanted steel and titanium alloys

    Science.gov (United States)

    Viviente, J. L.; García, A.; Alonso, F.; Braceras, I.; Oñate, J. I.

    1999-04-01

    A study has been made of the depth dependence of the atomic fraction and chemical bonding states of AISI 440C martensitic stainless steel and Ti-6Al-4V alloy implanted with 75 keV C + at very high doses (above 10 18 ions cm -2), by means of X-ray photoelectron spectroscopy combined with an Ar + sputtering. A Gaussian-like carbon distribution was observed on both materials at the lowest implanted dose. More trapezoidal carbon depth-profiles were found with increasing implanted doses, and a pure carbon layer was observed only on the titanium alloy implanted at the highest dose. The implanted carbon was combined with both base metal and carbon itself to form metallic carbides and graphitic carbon. Furthermore, carbon-enriched carbides were also found by curve fitting the C 1s spectra. The titanium alloy showed a higher carbidic contribution than the steel implanted at the same C + doses. A critical carbon concentrations of about 33 at.% and 23 at.% were measured for the formation of C-C bonds in Ti-6Al-4V and steel samples, respectively. The carbon atoms were bound with metal to form carbidic compounds until these critical concentrations were reached; when this C concentration was exceeded the proportion of C-C bonds increased and resulted in the growth of carbonaceous layers.

  4. Yttrium implantation effects on extra low carbon steel and pure iron

    Energy Technology Data Exchange (ETDEWEB)

    Caudron, E.; Buscail, H. [Clermont-Ferrand-2 Univ., Le Puy en Velay (France). Lab. Vellave d`Elaboration; Jacob, Y.P.; Stroosnijder, M.F. [Institute for Advanced Materials, Joint Research Center, The European Commission, 21020, Ispra (Vatican City State, Holy See) (Italy); Josse-Courty, C. [Laboratoire de Recherche sur la Reactivite des Solides, UMR 56-13 CNRS, UFR Sciences et Techniques, 9 Avenue A. Savary, B.P. 400, 21011, Dijon Cedex (France)

    1999-05-25

    Extra low carbon steel and pure electrolytic iron samples were yttrium implanted using ion implantation technique. Compositions and structures of pure iron and steel samples were investigated before and after yttrium implantation by several analytical and structural techniques (RBS, SIMS, RHEED and XRD) to observe the yttrium implantation depth profiles in the samples. This paper shows the different effects of yttrium implantations (compositions and structures) according to the implanted sample nature. (orig.) 23 refs.

  5. 75 FR 42782 - Hot-Rolled Flat-Rolled Carbon-Quality Steel Products From Brazil, Japan, and Russia

    Science.gov (United States)

    2010-07-22

    ...)] Hot-Rolled Flat-Rolled Carbon-Quality Steel Products From Brazil, Japan, and Russia AGENCY: United... Brazil and Japan, and the suspended investigation on hot-rolled steel from Russia. SUMMARY: The... Japan, and the suspended investigation on hot-rolled steel from Russia would be likely to lead to...

  6. 75 FR 62566 - Hot-Rolled Flat-Rolled Carbon-Quality Steel Products From Brazil, Japan, and Russia

    Science.gov (United States)

    2010-10-12

    ...)] Hot-Rolled Flat-Rolled Carbon-Quality Steel Products From Brazil, Japan, and Russia AGENCY: United... antidumping duty investigation on hot-rolled steel from Russia. SUMMARY: The Commission hereby gives notice of... suspended investigation on hot-rolled steel from Russia would be likely to lead to continuation or...

  7. Evolution of precipitation in reactor pressure vessel steel welds under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Kristina, E-mail: kristina.lindgren@chalmers.se [Department of Physics, Chalmers University of Technology, SE-412 96 Göteborg (Sweden); Boåsen, Magnus [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Stiller, Krystyna [Department of Physics, Chalmers University of Technology, SE-412 96 Göteborg (Sweden); Efsing, Pål [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Vattenfall Ringhals AB, SE-430 22 Väröbacka (Sweden); Thuvander, Mattias [Department of Physics, Chalmers University of Technology, SE-412 96 Göteborg (Sweden)

    2017-05-15

    Reactor pressure vessel steel welds are affected by irradiation during operation. The irradiation results in nanometre cluster formation, which in turn affects the mechanical properties of the material, e.g. the ductile-to-brittle transition temperature is shifted to higher levels. In this study, cluster formation is characterised in high Ni (1.58%) low Cu (0.04%) steel welds identical to Ringhals R4 welds, using atom probe tomography in both surveillance material and in material irradiated at accelerated dose rates. Clusters containing mainly Ni and Mn, but also some Si and Cu were observed in all of the irradiated materials. Their evolution did not change drastically during irradiation; the clusters grew and new clusters were nucleated. Hence, both the cluster number density and the average size increased with irradiation time. Some flux effects were observed when comparing the high flux material and the surveillance material. The surveillance material has a lower cluster number density, but larger clusters. The resulting impact on the mechanical properties of these two effects cancel out, resulting in a measured hardness that seems to be on the same trend as the high flux material. The dispersed barrier hardening model with an obstacle strength factor of 0.15 was found to reproduce the increase in hardness. In the investigated high flux materials, the clusters' Cu content was higher. - Highlights: •Clustering in a low Cu, high Ni reactor pressure vessel steel weld is studied. •The clusters nucleate and grow during irradiation, and consist of Ni, Mn, Si, and Cu. •High flux neutron irradiated material is compared to surveillance material. •High flux was found to result in smaller clusters with a larger number density. •Hardness follows the same dependence on fluence, independent of flux.

  8. Effect of High-Temperature Thermomechanical Treatment on the Brittle Fracture of Low-Carbon Steel

    Science.gov (United States)

    Smirnov, M. A.; Pyshmintsev, I. Yu.; Varnak, O. V.; Mal'tseva, A. N.

    2018-02-01

    The effect of high-temperature thermomechanical treatment (HTMT) on the brittleness connected with deformation-induced aging and on the reversible temper brittleness of a low-carbon tube steel with a ferrite-bainite structure has been studied. When conducting an HTMT of a low-alloy steel, changes should be taken into account in the amount of ferrite in its structure and relationships between the volume fractions of the lath and the acicular bainite. It has been established that steel subjected to HTMT undergoes transcrystalline embrittlement upon deformation aging. At the same time, HTMT, which suppresses intercrystalline fracture, leads to a weakening of the development of reversible temper brittleness.

  9. Effects of LWR environments on fatigue life of carbon and low-alloy steels

    International Nuclear Information System (INIS)

    Chopra, O.K.; Shack, W.J.

    1995-03-01

    SME Boiler and Pressure Vessel Code provides construction of nuclear power plant components. Figure I-90 Appendix I to Section III of the Code specifies fatigue design curves for structural materials. While effects of environments are not explicitly addressed by the design curves, test data suggest that the Code fatigue curves may not always be adequate in coolant environments. This paper reports the results of recent fatigue tests that examine the effects of steel type, strain rate, dissolved oxygen level, strain range, loading waveform, and surface morphology on the fatigue life of A 106-Gr B carbon steel and A533-Gr B low-alloy steel in water

  10. Interim fatigue design curves for carbon, low-alloy, and austenitic stainless steels in LWR environments

    International Nuclear Information System (INIS)

    Majumdar, S.; Chopra, O.K.; Shack, W.J.

    1993-01-01

    Both temperature and oxygen affect fatigue life; at the very low dissolved-oxygen levels in PWRs and BWRs with hydrogen water chemistry, environmental effects on fatigue life are modest at all temperatures (T) and strain rates. Between 0.1 and 0.2 ppM, the effect of dissolved-oxygen increases rapidly. In oxygenated environments, fatigue life depends strongly on strain rate and T. A fracture mechanics model is developed for predicting fatigue lives, and interim environmentally assisted cracking (EAC)-adjusted fatigue curves are proposed for carbon steels, low-alloy steels, and austenitic stainless steels

  11. The hot working characteristics of a boron bearing and a conventional low carbon steel

    International Nuclear Information System (INIS)

    Stumpf, Waldo; Banks, Kevin

    2006-01-01

    Constitutive hot working constants were determined for an 11 ppm boron low carbon strip steel and compared from 875 to 1140 deg. C and strain rates of 0.001-2.5 s -1 to a high nitrogen low carbon strip steel. The boron steel showed a different hot working behaviour than the conventional steel with the steady state flow stress about 50-60% higher, the peak strain more than 50% higher and the eventual ferrite grain size about 40% smaller, if compared at the same temperature compensated strain rates or Z values. This difference persisted where the soaking temperature before compression was varied between 1140 and 1250 deg. C, proving that undissolved AlN in the boron-bearing steel was not responsible. With systematically varied linear cooling rates after hot working, the final ferrite grain size in the boron steel is finer and is independent of the two Z values applied during hot working. Retarded softening by dynamic recrystallisation during hot working in the boron containing steel is probably caused by boron solute drag of moving grain boundaries

  12. Behaviour of carbon steel and chromium steels in CO2 environments

    International Nuclear Information System (INIS)

    Lefebvre, B.; Bounie, P.; Guntz, G.; Prouheze, J.C.; Renault, J.J.

    1984-01-01

    The behavior in aqueous CO 2 environments of steel with chromium content between 0 and 22% has been studied by autoclave tests. The influence of chromium and molybdenum contents has been investigated particularly on 13 Cr steel. Conventional electrochemical test results are related to the CO 2 autoclave test results. The influence of the environment: temperature, chloride concentration, partial pressure of CO 2 and some amount of H 2 S on the corrosion resistance are discussed

  13. Influence of boron on strain hardening behaviour and ductility of low carbon hot rolled steel

    International Nuclear Information System (INIS)

    Deva, Anjana; Jha, B.K.; Mishra, N.S.

    2011-01-01

    Highlights: → Unique feature of low strain hardening exponent (n) with high total elongation has been discussed in industrially produced low carbon boron containing steel. → n has been correlated with the micro structural changes occurring during deformation of steel. → This feature of low n and high % elongation has potential for higher cold reducibility. → The work is being reported for the first time on industrially produced low carbon boron containing steel. - Abstract: The beneficial effect of boron on mechanical properties of low carbon Al-killed steel has been reported in recent past. However, the effect of boron on strain hardening exponent (n) and ductility has not been fully understood. This aspect has been discussed in present work. The results of mill trials with reference to n and ductility with boron added steel are compared to those for commercial grade. The lowering of 'n' with increased total elongation in boron bearing steel has been related to the microstructural evolution as a result of boron addition.

  14. The Mechanism of High Ductility for Novel High-Carbon Quenching-Partitioning-Tempering Martensitic Steel

    Science.gov (United States)

    Qin, Shengwei; Liu, Yu; Hao, Qingguo; Wang, Ying; Chen, Nailu; Zuo, Xunwei; Rong, Yonghua

    2015-09-01

    In this article, a novel quenching-partitioning-tempering (Q-P-T) process was applied to treat Fe-0.6C-1.5Mn-1.5Si-0.6Cr-0.05Nb hot-rolled high-carbon steel and the microstructures including retained austenite fraction and the average dislocation densities in both martensite and retained austenite were characterized by X-ray diffraction, scanning electron microscopy, and transmission electron microscopy, respectively. The Q-P-T steel exhibits high strength (1950 MPa) and elongation (12.4 pct). Comparing with the steel treated by traditional quenching and tempering (Q&T) process, the mechanism of high ductility for high-carbon Q-P-T steel is revealed as follows. Much more retained austenite existing in Q-P-T steel than in Q&T one remarkably enhances the ductility by the following two effects: the dislocation absorption by retained austenite effect and the transformation-induced plasticity effect. Besides, lower dislocation density in martensite matrix produced by Q-P-T process plays an important role in the improvement of ductility. However, some thin plates of twin-type martensite embedded in dislocation-type martensite matrix in high-carbon Q-P-T steel affect the further improvement of ductility.

  15. Corrosion Behaviour of Nickel Plated Low Carbon Steel in Tomato Fluid

    Directory of Open Access Journals (Sweden)

    Oluleke OLUWOLE

    2010-12-01

    Full Text Available This research work investigated the corrosion resistance of nickel plated low carbon steel in tomato fluid. It simulated the effect of continuous use of the material in a tomato environment where corrosion products are left in place. Low carbon steel samples were nickel electroplated at 4V for 20, 25, 30 and 35 mins using Watts solution.The plated samples were then subjected to tomato fluid environment for for 30 days. The electrode potentials mV (SCE were measured every day. Weight loss was determined at intervals of 5 days for the duration of the exposure period. The result showed corrosion attack on the nickel- plated steel, the severity decreasing with the increasing weight of nickel coating on substrate. The result showed that thinly plated low carbon steel generally did not have any advantage over unplated steel. The pH of the tomato solution which initially was acidic was observed to progress to neutrality after 4 days and then became alkaline at the end of the thirty days test (because of corrosion product contamination of the tomatocontributing to the reduced corrosion rates in the plated samples after 10 days. Un-plated steel was found to be unsuitable for the fabrication of tomato processing machinery without some form of surface treatment - thick nickel plating is suitable as a protective coating in this environment.

  16. Corrosion-resistant Foamed Cements for Carbon Steels

    Energy Technology Data Exchange (ETDEWEB)

    Sugama T.; Gill, S.; Pyatina, T., Muraca, A.; Keese, R.; Khan, A.; Bour, D.

    2012-12-01

    The cementitious material consisting of Secar #80, Class F fly ash, and sodium silicate designed as an alternative thermal-shock resistant cement for the Enhanced Geothermal System (EGS) wells was treated with cocamidopropyl dimethylamine oxide-based compound as foaming agent (FA) to prepare numerous air bubble-dispersed low density cement slurries of and #61603;1.3 g/cm3. Then, the foamed slurry was modified with acrylic emulsion (AE) as corrosion inhibitor. We detailed the positive effects of the acrylic polymer (AP) in this emulsion on the five different properties of the foamed cement: 1) The hydrothermal stability of the AP in 200 and #61616;C-autoclaved cements; 2) the hydrolysis-hydration reactions of the slurry at 85 and #61616;C; 3) the composition of crystalline phases assembled and the microstructure developed in autoclaved cements; 4) the mechanical behaviors of the autoclaved cements; and, 5) the corrosion mitigation of carbon steel (CS) by the polymer. For the first property, the hydrothermal-catalyzed acid-base interactions between the AP and cement resulted in Ca-or Na-complexed carboxylate derivatives, which led to the improvement of thermal stability of the AP. This interaction also stimulated the cement hydration reactions, enhancing the total heat evolved during cement’s curing. Addition of AP did not alter any of the crystalline phase compositions responsible for the strength of the cement. Furthermore, the AP-modified cement developed the porous microstructure with numerous defect-free cavities of disconnected voids. These effects together contributed to the improvement of compressive-strength and –toughness of the cured cement. AP modification of the cement also offered an improved protection of CS against brine-caused corrosion. There were three major factors governing the corrosion protection: 1) Reducing the extents of infiltration and transportation of corrosive electrolytes through the cement layer deposited on the underlying CS

  17. A study on the impediment of thickness diminution of Carbon steel tube by using a applied magnetic field

    International Nuclear Information System (INIS)

    Kim, Jong Oh; Kim, Jong Hui; Cho, Wan Sik; Hong, Sung Min; Park, Yun Won

    2001-03-01

    Magnetic properties of the carbon steel tube which is used as the pipe laying of cooling water in nuclear power plant were measured to research the impediment of thickness diminution of carbon steel tube. Magnetic field distribution of carbon steel tube in the applied magnetic field was simulated by computer program. On the basis of the simulation results, Alnico 5DG and Alnico 5 were selected as the permanent magnets applicable to the carbon steel tube. Sm2Co17 magnet was used to compare the performance of permanent magnets. The experimental apparatus similar to the draining environment of cooling water in nuclear power plant was also manufactured in order to research the impediment of thickness diminution of carbon steel carbon tube

  18. Tribological performance of hard carbon coatings on 440C bearing steel

    Energy Technology Data Exchange (ETDEWEB)

    Kustas, F M; Misra, M S; Shepard, D F; Froechtenigt, J F [Martin Marietta Astronautics Group, Denver, CO (United States)

    1991-11-01

    Hard carbon coatings such as amorphous carbon, diamond and diamond-like carbon have received considerable attention for tribological applications owing to their high hardness, high modulus and desirable surface properties. Unfortunately, most of the deposition techniques induce high substrate temperatures that would temper traditional bearing steels and reduce the substrate load-carrying capability. Therefore, to effectively use these desirable coatings, a lower temperature deposition technique is required. Ion beam deposition can provide essentially ambient temperature conditions, accurate control of process parameters and good coating-substrate adhesion. To use these attributes, a test program was initiated to deposit mass-analyzed, high purity C{sup +} and CH{sub 4}{sup +} ions on molybdenum and 440C bearing steel for subsequent characterization by Raman spectroscopy and friction-wear tests. Results for a coating deposited from a carbon monoxide source showed an amorphous carbon-microcrystalline graphtie structure which exhibited very high microhardness and a three fold reduction in coefficient of friction for unlubricated tests compared to untreated 440C steel. In addition, incrementally increasing the applied load (by up to a factor of 5) resulted in progressively lower coefficients of friction, which conforms to solid lubrication theory. End-of-travel wear debris and some limited coating delamination were observed within thinner areas of the coating. Therefore an amorphous carbon-graphite coating applied to 440C steel at ambient temperature exhibits solid lubricating film characteristics with high load-carrying capability. (orig.).

  19. Anodic Oxidation of Carbon Steel at High Current Densities and Investigation of Its Corrosion Behavior

    Science.gov (United States)

    Fattah-Alhosseini, Arash; Khan, Hamid Yazdani

    2017-06-01

    This work aims at studying the influence of high current densities on the anodization of carbon steel. Anodic protective coatings were prepared on carbon steel at current densities of 100, 125, and 150 A/dm2 followed by a final heat treatment. Coatings microstructures and morphologies were analyzed using X-ray diffraction (XRD) and scanning electron microscope (SEM). The corrosion resistance of the uncoated carbon steel substrate and the anodic coatings were evaluated in 3.5 wt pct NaCl solution through electrochemical impedance spectroscopy (EIS) and potentiodynamic polarization measurements. The results showed that the anodic oxide coatings which were prepared at higher current densities had thicker coatings as a result of a higher anodic forming voltage. Therefore, the anodized coatings showed better anti-corrosion properties compared to those obtained at lower current densities and the base metal.

  20. Effect of microstructural variation on the Cu/CK45 carbon steel friction weld joint

    Energy Technology Data Exchange (ETDEWEB)

    Lee, W.B.; Jung, S.B. [Advanced Materials and Process Research Center for IT, Sungkyunkwan Univ., Gyounggi-do (Korea)

    2003-12-01

    The mechanical properties of friction-welded pure Cu/CK45 carbon steel joints have been studied. The joint strength increased with increasing upset pressure till it reached a critical value. However, the joint strength was fixed at a low strength with increasing friction time, compared to that of the Cu base metal. The hardness near the interface at the Cu side was softer than that of the base metal due to the dynamically recrystallized and annealed grain. The width of the softened region became wider with increasing friction time and decreasing upset pressure. But the hardness of the CK45 carbon steel side showed a slightly higher value than that of the base metal. This result was explained by the formation of martensite structure at the CK45 carbon steel side during the welding process. (orig.)

  1. Microbiologically influenced corrosion of carbon steel in the presence of sulphate reducing bacteria

    International Nuclear Information System (INIS)

    Tunaru, M.; Velciu, L.; Mihalache, M.; Laurentiu, P.

    2016-01-01

    Sulphate-reducing bacteria (SRB) are the most important organisms in microbiologically induced corrosion. In this context, the paper presents an assessment (by experimental tests) of the behaviour of carbon steel samples (SA106gr.B) in SRB media. Some of samples were immersed in microbial environment in order microbiological analysis of their surface and another part was used to perform accelerated electrochemical tests to determine electrochemical parameters for the system carbon steel / microbial medium (corrosion rate, the polarization resistance of the surface, susceptibility to pitting corrosion). The surfaces of the tested samples were analyzed using the optical and electronic microscope, and emphasized the role of bacteria in the development of biofilms under which appeared characteristics of corrosion attack. The correlation of all results confirmed that SRB accelerated the localized corrosion of the surfaces of SA 106gr.B carbon steel. (authors)

  2. Carbon steel protection in G.S. [Girldler sulphide] plants: Pt. 7

    International Nuclear Information System (INIS)

    Lires, Osvaldo; Delfino, Cristina; Rojo, Enrique.

    1989-01-01

    In order to protect carbon steel towers and piping of a GS experimental heavy water plant against corrosion produced by the action of aqueous solutions of hydrogen sulphide, a method, elsewhere published, was developed. Carbon steel exposed to saturated aqueous solutions of hydrogen sulphide forms iron sulphide scales. In oxygen free solutions, evolution of corrosion follows the sequence mackinawite → cubic ferrous sulphide → troilite → pyrrotite → pyrite. Scales formed by pyrrotite and pyrite are the most protective layers (these are obtained at 130 deg C, 2 MPa for a period of 14 days). During a plant shutdown procedures, the carbon steel protected with those scales is exposed to water and highly humid air; under such conditions oxidation is unavoidable. Later, treatment in plant conditions does not regenerate scales because the composition of regenerated scales involves more soluble iron sulphides such as mackinawite and troilite. Therefore, it is not recommendable to expose the protective scales to atmospherical conditions. (Author)

  3. Influence of ultraviolet light irradiation on the corrosion behavior of carbon steel AISI 1015

    Science.gov (United States)

    Riazi, H. R.; Danaee, I.; Peykari, M.

    2013-03-01

    Corrosion of carbon steel in sodium chloride solution was studied under ultraviolet illumination using weight loss, polarization, electrochemical impedance spectroscopy and current transient tests. The polarization test revealed an increase in the corrosion current density observed under UV illumination. The impedance spectroscopy indicated that the charge transfer resistance of the system was decreased by irradiation of UV light on a carbon steel electrode. The weight loss of carbon steel in solution increased under UV light, which confirms the results obtained from electrochemical measurements. We propose that the main effect of UV irradiation is on the oxide film, which forms on the surface. Thus, in presence of UV, the conductivity of oxide film might increase and lead to higher metal dissolution and corrosion rate.

  4. Effect of Biodiesel Concentration on Corrosion of Carbon Steel by Serratia marcescens

    Directory of Open Access Journals (Sweden)

    Pusparizkita Yustina M

    2018-01-01

    Full Text Available Biodiesel come into being used as an alternative source of energy as the diminishing of petroleum reserves. This fuel is typically stored in tanks that are commonly made from carbon steel, which is easily corroded by microorganisms. Recent studies have shown that bacteria aside from SRB may also be involved in corrosion. Therefore, this research was aimed to evaluate the effect of biodiesel concentration (15%, 20% and 30% v/v mixed in diesel oil on the corrosion of carbon steel by S. marcescens that dominate biocorrosion on hydrocarbon products. In this study, the corrosion process was investigated by evaluation of biofilm morphology and composition, the rate of corrosion and the corrosion product of carbon steel which was exposed in the mixture of hydrocarbons and the presence of S. marcescens. It can be concluded that higher concentration of biodiesel in diesel oil leads to higher growth of bacteria in the biofilm and higher corrosion rate.

  5. A Study on Effect of Local Wall Thinning in Carbon Steel Elbow Pipe on Elastic Stress Concentration

    International Nuclear Information System (INIS)

    Kim, Jong Sung; Seo, Jae Seok

    2009-01-01

    Feeder pipes that connect the inlet and outlet headers to the reactor core in CANDU nuclear power plants are considered as safety Class 1 piping items. Therefore, fatigue of feeder pipes should be assessed at design stage in order to verify structural integrity during design lifetime. In accordance with the fatigue assessment result, cumulative usage factors of some feeder pipes have significant values. The feeder pipes made of SA-106 Grade B or C carbon steel have some elbows and bends. An active degradation mechanism for the carbon steel outlet feeder piping is local wall thinning due to flow-accelerated corrosion. Inspection results from plants and metallurgical examinations of removed feeders indicated the presence of localized thinning in the vicinity of the welds in the lower portion of outlet feeders, such as Grayloc hub-to-bend weld, Grayloc hub-to-elbow weld, elbow-to-elbow, and elbow-to-pipe weld. This local wall thinning can cause increase of peak stress due to stress concentration by notch effect. The increase of peak stress results in increase of cumulative usage factor. However, present fatigue assessment doesn't consider the stress concentration due to local wall-thinning. Therefore, it is necessary to assess the effect of local wall thinning on stress concentration. This study investigates the effect of local wall thinning geometry on stress concentration by performing finite element elastic stress analysis

  6. Quenching and partitioning treatment of a low-carbon martensitic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Tsuchiyama, Toshihiro, E-mail: toshi@zaiko.kyushu-u.ac.jp [Department of Materials Science and Engineering, Graduate School of Engineering, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395 (Japan); Tobata, Junya; Tao, Teruyuki [Graduate School of Engineering, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395 (Japan); Nakada, Nobuo; Takaki, Setsuo [Department of Materials Science and Engineering, Graduate School of Engineering, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395 (Japan)

    2012-01-15

    Highlights: Black-Right-Pointing-Pointer The amount of retained austenite was increased by Q and P treatment in 12Cr-0.1C steel. Black-Right-Pointing-Pointer Ideal carbon concentrations in austenite and ferrite were calculated assuming CCE condition. Black-Right-Pointing-Pointer The optimum partitioning treatment condition for 12Cr-0.1C steel was found. Black-Right-Pointing-Pointer The strength-ductility balance of 12Cr-0.1C steel was improved by TRIP effect. - Abstract: Quenching and partitioning (Q and P) treatment was applied to a commercial low-carbon martensitic stainless steel, AISI Type 410 (Fe-12Cr-0.1C). The quench interruption temperature was optimized with consideration of the ideal carbon concentration in untransformed austenite after partitioning to lower the Ms temperature to room temperature. After partitioning at an appropriate temperature, a significant fraction of austenite was retained through the enrichment of carbon into the untransformed austenite. It was also suggested that the addition of silicon is not necessarily required for the Q and P treatment of 12Cr steel because of the retardation of carbide precipitation at the partitioning temperature owing to the large amount of chromium. Tensile testing revealed that the Q and P-treated material exhibited a significantly improved strength-ductility balance compared with conventional quench-and-tempered materials due to the transformation-induced plasticity (TRIP) effect by the retained austenite.

  7. Quenching and partitioning treatment of a low-carbon martensitic stainless steel

    International Nuclear Information System (INIS)

    Tsuchiyama, Toshihiro; Tobata, Junya; Tao, Teruyuki; Nakada, Nobuo; Takaki, Setsuo

    2012-01-01

    Highlights: ► The amount of retained austenite was increased by Q and P treatment in 12Cr–0.1C steel. ► Ideal carbon concentrations in austenite and ferrite were calculated assuming CCE condition. ► The optimum partitioning treatment condition for 12Cr–0.1C steel was found. ► The strength–ductility balance of 12Cr–0.1C steel was improved by TRIP effect. - Abstract: Quenching and partitioning (Q and P) treatment was applied to a commercial low-carbon martensitic stainless steel, AISI Type 410 (Fe–12Cr–0.1C). The quench interruption temperature was optimized with consideration of the ideal carbon concentration in untransformed austenite after partitioning to lower the Ms temperature to room temperature. After partitioning at an appropriate temperature, a significant fraction of austenite was retained through the enrichment of carbon into the untransformed austenite. It was also suggested that the addition of silicon is not necessarily required for the Q and P treatment of 12Cr steel because of the retardation of carbide precipitation at the partitioning temperature owing to the large amount of chromium. Tensile testing revealed that the Q and P-treated material exhibited a significantly improved strength–ductility balance compared with conventional quench-and-tempered materials due to the transformation-induced plasticity (TRIP) effect by the retained austenite.

  8. Development of carbon steel with superior resistance to wall thinning and fracture for nuclear piping system

    International Nuclear Information System (INIS)

    Rhee, Chang Kyu; Lee, Min Ku; Park, Jin Ju

    2010-07-01

    Carbon steel is usually used for piping for secondary coolant system in nuclear power plant because of low cost and good machinability. However, it is generally reported that carbon steel was failed catastrophically because of its low resistance to wall thinning and fracture toughness. Especially, flow accelerated corrosion (FAC) is one of main problems of the wall thinning of piping in the nuclear power plant. Therefore, in this project, fabrication technology of new advanced carbon steel materials modified by dispersion of nano-carbide ceramics into the matrix is developed first in order to improve the resistance to wall thinning and fracture toughness drastically compared to the conventional one. In order to get highly wettable fine TiC ceramic particles into molten metal, the micro-sized TiC particles were first mechanically milled by Fe (MMed TiC/Fe) in a high energy ball mill machine in Ar gas atmosphere, and then mixed with surfactant metal elements (Sn, Cr, Ni) to obtain better wettability, as this lowered surface tension of the carbon steel melt. According to microscopic images revealed that an addition of MMed TiC/Fe-surfactant mixed powders favorably disperses the fine TiC particles in the carbon steel matrix. It was also found that the grain size refinement of the cast matrix is achieved remarkably when fine TiC particles were added due to the fact that they act as nucleation sites during the solidification process. As a results, a cast carbon steel dispersed with fine TiC particles shows improved mechanical properties such as hardness, tensile strength and cavitation resistance compared to that of without particles. However, the slight decrease of toughness was found

  9. Characterization of carbon ion implantation induced graded microstructure and phase transformation in stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Feng, Kai; Wang, Yibo [Shanghai Key laboratory of Materials Laser Processing and Modification, School of Materials Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240 (China); Li, Zhuguo, E-mail: lizg@sjtu.edu.cn [Shanghai Key laboratory of Materials Laser Processing and Modification, School of Materials Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240 (China); Chu, Paul K. [Department of Physics and Materials Science, City University of Hong Kong, Tat Chee Avenue, Kowloon, Hong Kong (China)

    2015-08-15

    Austenitic stainless steel 316L is ion implanted by carbon with implantation fluences of 1.2 × 10{sup 17} ions-cm{sup −} {sup 2}, 2.4 × 10{sup 17} ions-cm{sup −} {sup 2}, and 4.8 × 10{sup 17} ions-cm{sup −} {sup 2}. The ion implantation induced graded microstructure and phase transformation in stainless steel is investigated by X-ray diffraction, X-ray photoelectron spectroscopy and high resolution transmission electron microscopy. The corrosion resistance is evaluated by potentiodynamic test. It is found that the initial phase is austenite with a small amount of ferrite. After low fluence carbon ion implantation, an amorphous layer and ferrite phase enriched region underneath are formed. Nanophase particles precipitate from the amorphous layer due to energy minimization and irradiation at larger ion implantation fluence. The morphology of the precipitated nanophase particles changes from circular to dumbbell-like with increasing implantation fluence. The corrosion resistance of stainless steel is enhanced by the formation of amorphous layer and graphitic solid state carbon after carbon ion implantation. - Highlights: • Carbon implantation leads to phase transformation from austenite to ferrite. • The passive film on SS316L becomes thinner after carbon ion implantation. • An amorphous layer is formed by carbon ion implantation. • Nanophase precipitate from amorphous layer at higher ion implantation fluence. • Corrosion resistance of SS316L is improved by carbon implantation.

  10. Evaluation of carbon-14 life cycle in reactors VVER-1000

    International Nuclear Information System (INIS)

    Lysakova, Katerina; Neumann, Jan; Vonkova, Katerina

    2012-09-01

    This work is aimed at the evaluation of carbon-14 life cycle in light water reactors VVER-1000. Carbon-14 is generated as a side product in different systems of nuclear reactors and has been an issue not only in radioactive waste management but mainly in release into the environment in the form of gaseous effluents. The principal sources of this radionuclide are in primary cooling water and fuel. Considerable amount of C-14 is generated by neutron reactions with oxygen 17 O and nitrogen 14 N present in water coolant and fuel. The reaction likelihood and consequently volume of generated radioisotope depends on several factors, especially on the effective cross-section, concentrations of parent elements and conditions of power plant operating strategies. Due to its long half-life and high capability of integration into the environment and thus into the living species, it is very important to monitor the movement of carbon-14 in all systems of nuclear power plant and to manage its release out of NPP. The dominant forms of radioactive carbon-14 are the hydrocarbons owing to the combinations with hydrogen used for absorption of radiolytic oxygen. These organic compounds, such as formaldehyde, methyl alcohol, ethyl alcohol and formic acid can be mostly retained on ion exchange resins used in the system for purifying primary cooling water. The gaseous carbon compounds (CH 4 and CO 2 ) are released into the atmosphere via the ventilation systems of NPP. Based on the information and data obtained from different sources, it has been designed a balance model of possible carbon-14 pathways throughout the whole NPP. This model includes also mass balance model equations for each important node in system and available sampling points which will be the background for further calculations. This document is specifically not to intended to describe the best monitoring program attributes or technologies but rather to provide evaluation of obtained data and find the optimal way to

  11. Improvement of deposition efficiency and control of hardness for cold-sprayed coatings using high carbon steel/mild steel mixture powder

    International Nuclear Information System (INIS)

    Ogawa, Kazuhiro; Amao, Satoshi; Yokoyama, Nobuyuki; Ootaki, Kousuke

    2011-01-01

    In this study, in order to make high carbon steel coating by cold spray technique, spray conditions such as carrier gas temperature and pressure etc. were investigated. And also, in order to improve deposition efficiency and control coating hardness of cold-sprayed high carbon steel, high carbon and mild steel mixed powder and its mechanical milled powder were developed and were optimized. By using the cold-spray technique, particle deposition of a high carbon steel was successful. Moreover, by applying mixed and mechanical milled powders, the porosity ratio was decreased and deposition efficiency was improved. Furthermore, using these powders, it is possible to control the hardness value. Especially, when using mechanical milled powder, it is very difficult to identify the interface between the coating and the substrate. The bonding between the coating and the substrate is thus considered to be excellent. (author)

  12. Study of corrosion behavior of carbon steel under seawater film using the wire beam electrode method

    International Nuclear Information System (INIS)

    Liu, Zaijian; Wang, Wei; Wang, Jia; Peng, Xin; Wang, Yanhua; Zhang, Penghui; Wang, Haijie; Gao, Congjie

    2014-01-01

    Corrosion behavior of carbon steel under seawater film with various thickness was investigated by the wire beam electrode (WBE) method. It was found that the corrosion rate of carbon steel increased significantly under thin seawater film than it was immersed in seawater. The current variation under seawater film indicated that the thickness of diffusion layer of oxygen was about 500 μm, and the maximal current appeared around 40 μm, at which corrosion rate transited from cathodic control to anodic control. The results suggest that WBE method is helpful to study the corrosion process under thin electrolyte film

  13. Surface modifications induced by yttrium implantation on low manganese-carbon steel

    Energy Technology Data Exchange (ETDEWEB)

    Caudron, E.; Buscail, H. [Univ. Blaise Pascal Clermont-Fd II, Le Puy en Velay (France). Lab. Vellave d' Elaboration et d' Etude des Materiaux; Haanapel, V.A.C.; Jacob, Y.P.; Stroosnijder, M.F. [Institute for Health and Consumer Protection, Joint Research Center, The European Commission, 21020, Ispra (Italy)

    1999-12-15

    Low manganese-carbon steel samples were ion implanted with yttrium. Sample compositions and structures were investigated before and after yttrium implantations to determine the yttrium distribution in the sample. Yttrium implantation effects were characterized using several analytical and structural techniques such as X-ray photoelectron spectroscopy, reflection high energy electron diffraction, X-ray diffraction, glancing angle X-ray diffraction and Rutherford backscattering spectrometry. In this paper it is shown that correlation between composition and structural analyses provides an understanding of the main compounds induced by yttrium implantation in low manganese-carbon steel. (orig.)

  14. A discussion for stabilization time of carbon steel in atmospheric corrosion

    Science.gov (United States)

    Zhang, Zong-kai; Ma, Xiao-bing; Cai, Yi-kun

    2017-09-01

    Stabilization time is an important parameter in long-term prediction of carbon steel corrosion in atmosphere. The range of the stabilization time of carbon steel in atmospheric corrosion has been published in many scientific literatures. However, the results may not precise because engineering experiences is dominant. This paper deals with the recalculation of stabilization time based on ISO CORRAG program, and analyzes the results and makes a comparison to the data mentioned above. In addition, a new thinking to obtain stabilization time will be proposed.

  15. The development of RFT technique for carbon steel tubes in balance-of-plant heat exchangers

    International Nuclear Information System (INIS)

    Kim, Chang Soo; Kim, Han Jong; Moon, Yong Sick; Kim, Jae Dong; Kim, Wang Bae; Nam, Min Woo

    2005-01-01

    The NDT method of carbon steel tubes is applied RFT technique. As other NDT methods, It is surprising that RFT has been rapidly developed over the past decade. These improvements have resulted in multi-frequency system, dual driver probes and development of analysis technique. Also these improvements give some profit to power plants as well as general industry. Therefore, the purpose of this study is to improve the reliability of RFT technique for carbon steel tubes. To uplift RFT technique, probes, calibration standards and specimen was developed.

  16. Hot ductility behavior of a low carbon advanced high strength steel (AHSS) microalloyed with boron

    OpenAIRE

    Mejía, Ignacio; Bedolla Jacuinde, Arnoldo; Maldonado, Cuauhtémoc; Cabrera Marrero, José M.

    2011-01-01

    The current study analyses the influence of boron addition on the hot ductility of a low carbon advanced high strength NiCrVCu steel. For this purpose hot tensile tests were carried out at different temperatures (650, 750, 800, 900 and 1000 ◦C) at a constant true strain rate of 0.001 s−1. Experimental results showed a substantial improvement in hot ductility for the low carbon advanced high strength steel when microalloyed with boron compared with that without boron addition. Nevertheless,...

  17. Meso-scale magnetic signatures for nuclear reactor steel irradiation embrittlement monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Suter, J. D., E-mail: pradeep.ramuhalli@pnnl.gov; Ramuhalli, P., E-mail: pradeep.ramuhalli@pnnl.gov; Hu, S.; Li, Y.; Jiang, W.; Edwards, D. J.; Schemer-Kohrn, A. L.; Johnson, B. R. [Pacific Northwest National Laboratory, 902 Battelle Blvd, Richland, WA 99352 (United States); McCloy, J. S., E-mail: john.mccloy@wsu.edu; Xu, K., E-mail: john.mccloy@wsu.edu [Washington State University, PO Box 642920, Pullman, WA 99164 (United States)

    2015-03-31

    Verifying the structural integrity of passive components in light water and advanced reactors will be necessary to ensure safe, long-term operations of the existing U.S. nuclear fleet. This objective can be achieved through nondestructive condition monitoring techniques, which can be integrated with plant operations to quantify the “state of health” of structural materials in real-time. While nondestructive methods for monitoring many classes of degradation (such as fatigue or stress corrosion cracking) are relatively advanced, this is not the case for degradation caused by irradiation. The development of nondestructive evaluation technologies for these types of degradation will require advanced materials characterization techniques and tools that enable comprehensive understanding of nuclear reactor material microstructural and behavioral changes under extreme operating environments. Irradiation-induced degradation of reactor steels causes changes in their microstructure that impacts their micro-magnetic properties. In this paper, we describe preliminary results of integrating advanced material characterization techniques with meso-scale computational models. In the future, this will help to provide an interpretive understanding of the state of degradation in structural materials. Microstructural data are presented from monocrystalline Fe and are correlated with variable-field magnetic force microscopy and micro-magnetic measurements. Ongoing research is focused on extending the measurements and models on thin films to gain insights into the structural state of irradiated materials and the resulting impact on magnetic properties. Preliminary conclusions from these correlations are presented, and next steps described.

  18. Stress corrosion cracking tests on electron beam welded carbon steel specimens in carbonate-bicarbonate solution

    International Nuclear Information System (INIS)

    Parkins, R.N.

    1985-04-01

    Stress corrosion cracking tests have been performed on tapered carbon steel test pieces containing electron beam welds with a view to defining susceptibility to such cracking in a carbonate-bicarbonate solution at 90 C and an appropriate electrode potential. The tests involved applying cyclic loads to the specimens and it is shown that the threshold stress for cracking reduces linearly with increase in the magnitude of the cyclic load component. Extrapolation of these trends to zero fluctuating stress indicates static load threshold stresses in the vicinity of the yield stress (i.e. about 300 N/mm 2 for parent plate without a weld, 400 N/mm 2 for specimens with welds on one side only and 600 N/mm 2 for specimens having welds penetrating through the thickness of the specimen). The averages of the maximum crack velocities observed were least for parent plate material and greatest for weld metal, the former being essentially intergranular in morphology and the latter mostly transgranular, with heat affected zone material being intermediate between these extremes. (author)

  19. 77 FR 31877 - Corrosion-Resistant Carbon Steel Flat Products From Germany and Korea; Scheduling of Full Five...

    Science.gov (United States)

    2012-05-30

    ... INTERNATIONAL TRADE COMMISSION [Investigation Nos. 701-TA-350 and 731-TA-616 and 618 (Third Review)] Corrosion-Resistant Carbon Steel Flat Products From Germany and Korea; Scheduling of Full Five-Year Reviews... corrosion-resistant carbon steel flat products from Korea and the antidumping duty orders on corrosion...

  20. 75 FR 18153 - Corrosion-Resistant Carbon Steel Flat Products from the Republic of Korea: Extension of Time...

    Science.gov (United States)

    2010-04-09

    ... DEPARTMENT OF COMMERCE International Trade Administration [C-580-818] Corrosion-Resistant Carbon Steel Flat Products from the Republic of Korea: Extension of Time Limit for Preliminary Results of... countervailing duty order on corrosion-resistant carbon steel flat products (CORE) from Korea. See Countervailing...

  1. 77 FR 16810 - Corrosion-Resistant Carbon Steel Flat Products From the Republic of Korea: Extension of Time...

    Science.gov (United States)

    2012-03-22

    ... DEPARTMENT OF COMMERCE International Trade Administration [C-580-818] Corrosion-Resistant Carbon Steel Flat Products From the Republic of Korea: Extension of Time Limit for Preliminary Results of... Register the countervailing duty order on corrosion-resistant carbon steel flat products (CORE) from Korea...

  2. 76 FR 20954 - Corrosion-Resistant Carbon Steel Flat Products From the Republic of Korea: Extension of Time...

    Science.gov (United States)

    2011-04-14

    ... DEPARTMENT OF COMMERCE International Trade Administration [C-580-818] Corrosion-Resistant Carbon Steel Flat Products From the Republic of Korea: Extension of Time Limit for Preliminary Results of... Register the countervailing duty order on corrosion-resistant carbon steel flat products (CORE) from Korea...

  3. Pathways to a low-carbon iron and steel industry in the medium-term : the case of Germany

    NARCIS (Netherlands)

    Arens, Marlene; Worrell, Ernst; Eichhammer, Wolfgang; Hasanbeigi, Ali; Zhang, Qi

    2017-01-01

    The iron and steel industry is a major industrial emitter of carbon dioxide globally and in Germany. If European and German climate targets were set as equal proportional reduction targets (referred to here as “flat” targets) among sectors, the German steel industry would have to reduce its carbon

  4. 77 FR 73674 - Circular Welded Carbon-Quality Steel Pipe From India, Oman, The United Arab Emirates, and Vietnam

    Science.gov (United States)

    2012-12-11

    ...)] Circular Welded Carbon-Quality Steel Pipe From India, Oman, The United Arab Emirates, and Vietnam... welded carbon-quality steel pipe from India, Oman, the United Arab Emirates, and Vietnam, provided for in... from India, Oman, the United Arab Emirates, and Vietnam were subsidized and/or dumped within the...

  5. 76 FR 68208 - Circular Welded Carbon-Quality Steel Pipe From India, Oman, United Arab Emirates, and Vietnam...

    Science.gov (United States)

    2011-11-03

    ... (Preliminary)] Circular Welded Carbon-Quality Steel Pipe From India, Oman, United Arab Emirates, and Vietnam... carbon-quality steel pipe from India, Oman, United Arab Emirates, and Vietnam, provided for in... Governments of India, Oman, United Arab Emirates, and Vietnam. Unless the Department of Commerce extends the...

  6. 76 FR 78313 - Circular Welded Carbon-Quality Steel Pipe From India, Oman, the United Arab Emirates, and Vietnam

    Science.gov (United States)

    2011-12-16

    ... (Preliminary)] Circular Welded Carbon-Quality Steel Pipe From India, Oman, the United Arab Emirates, and... India, Oman, the United Arab Emirates, and Vietnam of circular welded carbon- quality steel pipe... the Governments of India, Oman, the United Arab Emirates, and Vietnam.\\2\\ \\1\\ The record is defined in...

  7. 77 FR 25405 - Corrosion-Resistant Carbon Steel Flat Products From the Republic of Korea: Extension of Time...

    Science.gov (United States)

    2012-04-30

    ... antidumping duty order on corrosion-resistant carbon steel flat products from the Republic of Korea, covering... DEPARTMENT OF COMMERCE International Trade Administration [A-580-816] Corrosion-Resistant Carbon Steel Flat Products From the Republic of Korea: Extension of Time Limit for the Preliminary Results of...

  8. 76 FR 21332 - Corrosion-Resistant Carbon Steel Flat Products From the Republic of Korea: Extension of Time...

    Science.gov (United States)

    2011-04-15

    ... antidumping duty order on corrosion-resistant carbon steel flat products from the Republic of Korea, covering... DEPARTMENT OF COMMERCE International Trade Administration [A-580-816] Corrosion-Resistant Carbon Steel Flat Products From the Republic of Korea: Extension of Time Limits for the Preliminary Results of...

  9. 75 FR 25841 - Corrosion-Resistant Carbon Steel Flat Products From the Republic of Korea: Extension of Time...

    Science.gov (United States)

    2010-05-10

    ... antidumping duty order on corrosion-resistant carbon steel flat products from the Republic of Korea, covering... DEPARTMENT OF COMMERCE International Trade Administration [A-580-816] Corrosion-Resistant Carbon Steel Flat Products From the Republic of Korea: Extension of Time Limits for the Preliminary Results of...

  10. 77 FR 27438 - Certain Corrosion-Resistant Carbon Steel Flat Products From Korea: Final Results of Expedited...

    Science.gov (United States)

    2012-05-10

    ... DEPARTMENT OF COMMERCE International Trade Administration [C-580-818] Certain Corrosion-Resistant... order on certain corrosion-resistant carbon steel flat products (``CORE'') from the Republic of Korea.... Scope of the Order The merchandise covered by the order includes flat-rolled carbon steel products, of...

  11. 77 FR 25141 - Corrosion-Resistant Carbon Steel Flat Products From Germany and South Korea: Extension of Time...

    Science.gov (United States)

    2012-04-27

    ...-Resistant Carbon Steel Flat Products From Germany and South Korea: Extension of Time Limits for Preliminary...) orders on corrosion-resistant carbon steel flat products (CORE) from Germany and South Korea (Korea... from Germany and South Korea: Adequacy Redetermination Memorandum,'' (April 20, 2012). The preliminary...

  12. 77 FR 15718 - Circular Welded Carbon-Quality Steel Pipe From India, the Sultanate of Oman, the United Arab...

    Science.gov (United States)

    2012-03-16

    ...-811] Circular Welded Carbon-Quality Steel Pipe From India, the Sultanate of Oman, the United Arab... Oman (Oman), the United Arab Emirates (UAE), and the Socialist Republic of Vietnam (Vietnam). See Circular Welded Carbon-Quality Steel Pipe From India, the Sultanate of Oman, the United Arab Emirates, and...

  13. 77 FR 5240 - Light-Walled Welded Rectangular Carbon Steel Tubing From Taiwan: Continuation of Antidumping Duty...

    Science.gov (United States)

    2012-02-02

    ... Rectangular Carbon Steel Tubing From Taiwan: Continuation of Antidumping Duty Order AGENCY: Import... revocation of the antidumping duty order on light-walled welded rectangular carbon steel tubing from Taiwan would likely lead to a continuation or recurrence of dumping and material injury to an industry in the...

  14. 75 FR 1335 - Circular Welded Carbon Steel Pipes and Tubes from Taiwan; Extension of Time Limit for Preliminary...

    Science.gov (United States)

    2010-01-11

    ... DEPARTMENT OF COMMERCE International Trade Administration [A-583-008] Circular Welded Carbon Steel... review of the antidumping duty order on circular welded carbon steel pipes and tubes from Taiwan.\\1\\ On... review within the original time frame because we require additional time to obtain information from the...

  15. Direct gas-solid carbonation kinetics of steel slag and the contribution to in situ sequestration of flue gas CO(2) in steel-making plants.

    Science.gov (United States)

    Tian, Sicong; Jiang, Jianguo; Chen, Xuejing; Yan, Feng; Li, Kaimin

    2013-12-01

    Direct gas-solid carbonation of steel slag under various operational conditions was investigated to determine the sequestration of the flue gas CO2 . X-ray diffraction analysis of steel slag revealed the existence of portlandite, which provided a maximum theoretical CO2 sequestration potential of 159.4 kg CO 2 tslag (-1) as calculated by the reference intensity ratio method. The carbonation reaction occurred through a fast kinetically controlled stage with an activation energy of 21.29 kJ mol(-1) , followed by 10(3) orders of magnitude slower diffusion-controlled stage with an activation energy of 49.54 kJ mol(-1) , which could be represented by a first-order reaction kinetic equation and the Ginstling equation, respectively. Temperature, CO2 concentration, and the presence of SO2 impacted on the carbonation conversion of steel slag through their direct and definite influence on the rate constants. Temperature was the most important factor influencing the direct gas-solid carbonation of steel slag in terms of both the carbonation conversion and reaction rate. CO2 concentration had a definite influence on the carbonation rate during the kinetically controlled stage, and the presence of SO2 at typical flue gas concentrations enhanced the direct gas-solid carbonation of steel slag. Carbonation conversions between 49.5 % and 55.5 % were achieved in a typical flue gas at 600 °C, with the maximum CO2 sequestration amount generating 88.5 kg CO 2 tslag (-1) . Direct gas-solid carbonation of steel slag showed a rapid CO2 sequestration rate, high CO2 sequestration amounts, low raw-material costs, and a large potential for waste heat utilization, which is promising for in situ carbon capture and sequestration in the steel industry. Copyright © 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  16. Image analysis of corrosion pit initiation on ASTM type A240 stainless steel and ASTM type A 1008 carbon steel

    Science.gov (United States)

    Nine, H. M. Zulker

    The adversity of metallic corrosion is of growing concern to industrial engineers and scientists. Corrosion attacks metal surface and causes structural as well as direct and indirect economic losses. Multiple corrosion monitoring tools are available although those are time-consuming and costly. Due to the availability of image capturing devices in today's world, image based corrosion control technique is a unique innovation. By setting up stainless steel SS 304 and low carbon steel QD 1008 panels in distilled water, half-saturated sodium chloride and saturated sodium chloride solutions and subsequent RGB image analysis in Matlab, in this research, a simple and cost-effective corrosion measurement tool has identified and investigated. Additionally, the open circuit potential and electrochemical impedance spectroscopy results have been compared with RGB analysis to gratify the corrosion. Additionally, to understand the importance of ambiguity in crisis communication, the communication process between Union Carbide and Indian Government regarding the Bhopal incident in 1984 was analyzed.

  17. Practical domain for ultrasonic testing of stainless steel over plain carbon steel layered components using M21 waves

    International Nuclear Information System (INIS)

    Grewal, D.S.; Bray, D.E.

    1995-01-01

    The first higher order mode of the Rayleigh wave was discussed by Sezawa in the early part of this century in context of seismological wave studies. These Sezawa, or M 21 , or first higher order mode Rayleigh waves, have subsequently been used in the field of nondestructive testing of layered materials based on the development of the seismological model of the Sezawa waves by others. In this paper the study of the Tiersten formulation in context with slow speed over high speed materials, e.g. stainless steel overlay on plain carbon steel, the limitations and applicability of that formulation is reported. This study illustrates the practical bounds for testing such layered media, using numerical analysis of this formulation for the first higher-order mode to establish theoretical limits, and corroboration of these bounds by experimental results

  18. Does carbonation of steel slag particles reduce their toxicity? An in vitro approach.

    Science.gov (United States)

    Ibouraadaten, Saloua; van den Brule, Sybille; Lison, Dominique

    2015-06-01

    Mineral carbonation can stabilize industrial residues and, in the steel industry, may contribute to simultaneously valorize CO2 emissions and slag. We hypothesized that, by restricting the leaching of metals of toxicological concern such as Cr and V, carbonation can suppress the toxicity of these materials. The cytotoxic activity (WST1 assay) of slag dusts collected from a stainless and a Linz-Donawitz (LD) steel plant, before and after carbonation, was examined in J774 macrophages. The release of Cr, V, Fe, Mn and Ni was measured after incubation in artificial lung fluids mimicking the extracellular and phagolysosomal milieu to which particles are confronted after inhalation. LD slag had the higher Fe, Mn and V content, and was more cytotoxic than stainless steel slag. The cytotoxic activity of LD but not of stainless dusts was reduced after carbonation. The cytotoxic activity of the dusts toward J774 macrophages necessitated a direct contact with the cells and was reduced in the presence of inhibitors of phagocytosis (cytochalasin D) or phagolysosome acidification (bafilomycin), pointing to a key role of metallic constituents released in phagolysosomes. This in vitro study supports a limited reduction of the cytotoxic activity of LD, but not of stainless, steel dusts upon carbonation. Copyright © 2015 Elsevier Ltd. All rights reserved.

  19. Factors affecting stress assisted corrosion cracking of carbon steel under industrial boiler conditions

    Science.gov (United States)

    Yang, Dong

    Failure of carbon steel boiler tubes from waterside has been reported in the utility boilers and industrial boilers for a long time. In industrial boilers, most waterside tube cracks are found near heavy attachment welds on the outer surface and are typically blunt, with multiple bulbous features indicating a discontinuous growth. These types of tube failures are typically referred to as stress assisted corrosion (SAC). For recovery boilers in the pulp and paper industry, these failures are particularly important as any water leak inside the furnace can potentially lead to smelt-water explosion. Metal properties, environmental variables, and stress conditions are the major factors influencing SAC crack initation and propagation in carbon steel boiler tubes. Slow strain rate tests (SSRT) were conducted under boiler water conditions to study the effect of temperature, oxygen level, and stress conditions on crack initation and propagation on SA-210 carbon steel samples machined out of boiler tubes. Heat treatments were also performed to develop various grain size and carbon content on carbon steel samples, and SSRTs were conducted on these samples to examine the effect of microstructure features on SAC cracking. Mechanisms of SAC crack initation and propagation were proposed and validated based on interrupted slow strain tests (ISSRT). Water chemistry guidelines are provided to prevent SAC and fracture mechanics model is developed to predict SAC failure on industrial boiler tubes.

  20. Comparison of four NDT methods for indication of reactor steel degradation by high fluences of neutron irradiation

    Czech Academy of Sciences Publication Activity Database

    Tomáš, Ivan; Vértesy, G.; Pirfo Barroso, S.; Kobayashi, S.

    2013-01-01

    Roč. 265, DEC (2013), s. 201-209 ISSN 0029-5493 Institutional support: RVO:68378271 Keywords : neutron irradiation * steel degradation * nuclear reactor pressure vessel * magnetic NDT * magnetic minor hysteresis loops * Magnetic Barkhausen Emission Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 0.972, year: 2013 http://www.sciencedirect.com/science/article/pii/S0029549313004664