Chang, J.Y.; Schrock, S.L.; Johnson, R.N.
The self-welding behavior of two similar materials, Stellite 6 and Stellite 156, in sodium are discussed. The materials were tested at temperatures from 850 to 1140 0 F for time periods up to six-months while immersed in flowing sodium. Contact stresses ranged from 6000 psi to 16,000 psi on contact areas from 0.35 to 0.47 square inches. All separation tests to determine the extent of self-welding were conducted in a tensile mode. The surface morphologies of the samples before and after each test were measured. At temperatures of 1115 0 F and above, almost all the Stellite 6 specimens indicated a significant tendency toward self-welding within a relatively short period of time (one week). Stellite 156 couples also developed a strong self-weld bond at 1060 0 F after six-month exposure
Huber, F.; Mattes, K.
To determine the parameters responsible for selfwelding, experimental investigations were carried out at the Karlsruhe Nuclear Research Center. These activities are related to the SNR 300 prototype sodium-cooled fast breeder reactor. The experimental equipment, test materials and conditions as well as the results obtained are described and an attempt is made to present a general applicable explanation of the self-welding phenomena
Kumar, Hemant, E-mail: firstname.lastname@example.org [Department of Metallurgical and Materials Engineering, Indian Institute of Technology Kharagpur, Kharagpur, 721 302 (India); Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Ramakrishnan, V.; Albert, S.K.; Bhaduri, A.K. [Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Ray, K.K. [Department of Metallurgical and Materials Engineering, Indian Institute of Technology Kharagpur, Kharagpur, 721 302 (India)
The self-welding susceptibility between NiCr-B coated 316LN stainless steel and the base metal, and that between NiCr-B hardfaced coatings has been evaluated in flowing sodium at 823 K for 90 and 135 days under contact stress of 8.0 and 11.0 MPa using a fabricated set-up. Neither any self-welding could be observed nor could any damage be detected on the specimen surfaces of the selected materials under the imposed experimental conditions, which indicate their satisfactory potential for applications in Fast Breeder Reactors.
Mizobuchi, Syotaro; Kano, Shigeki; Nakayama, Kohichi; Atsumo, Hideo
Self-Weldability of Various Materials in High Temperature Sodium. The self-welding behavior of various materials was evaluated by measuring the tensile breakaway force of the specimen which had been self-welded in high temperature sodium. Experiments were carried out to investigate the influence of the sodium temperature and the contact stress on the self-welding behavior. The results obtained are as follows: (1) The self-welding behavior in sodium was recognized to initiate by the diffusion of the principal element through the real contact area. (2) Remarkable self-welding behavior was observed for SUS 316 material at 650 0 C, and for 2 1/4Cr-1Mo steel at a sodium temperature of 600 0 C. The self-welding force acting on the real contact area corresponds to the tensile strength of each material. (3) Hard chrome plating or hardfacing material showed good self-weld resistance, but the different combinations of SUS 316 with either of these materials were observed to easily cause self-welding. (4) The self-weldability of Cr 3 C 2 /Ni-Cr material varied with the preparing methods, especially, with the distribution of the binder composition contained in this material. (5) A derived equation was proposed to evaluate the self-welding force. It was found that the measured breakaway force was relatively equal to the self-welding force derived from this equation. (author)
Degradation of materials exposed to Na in LMFBR service is reviewed. The degradation processes are discussed in sections on corrosion and mass transfer, erosion, wear and self welding, sodium--water reactions, and external corrosion. (JRD)
This invention concerns the coolant pumps of nuclear reactors. It resolves the problems of structures which have to withstand high temperatures, the difficulties in keeping the multiple bearings of the shaft aligned, the excessive fluid flows, the risks of scoring and seizing-up by self welding, the need for narrow machining tolerances and the difficulties of access for inspection and repairs [fr
Poorly controlled wear and friction affects energy conservation, material conservation, and the reliability and safety of mechanical systems, and is estimated to cost U.S. industries $16 billion/yr. ERDA's National Friction, Wear, and Self-Welding Program and its accomplishments are described. This program includes studies of wear and friction problems in high temperature and unusual environments, e.g., as experienced by LMFBR components, and common to much technology involving energy conversion using fossil-fuel, geothermal, nuclear, and solar resources. Program activities for tribology information handling and wear and friction testing are discussed
Johnson, R.N.; Farwick, D.G.
The friction, wear, and corrosion performance of several metallurgical coatings in 200 to 650 0 C sodium are reviewed. Emphasis is placed on those coatings which have successfully passed the qualification tests necessary for acceptance in breeder reactor environments. Tests include friction, wear, corrosion, thermal cycling, self-welding, and irradiation exposure under as-prototypic-as-possible service conditions. Materials tested were coatings of various refractory metal carbides in metallic binders, nickel-base and cobalt-base alloys and intermetallic compounds such as the aluminides and borides. Coating processes evaluated included plasma spray, detonation gun, sputtering, spark-deposition, and solid-state diffusion
Borgstedt, H.U.; Mattes, K.; Wild, E.
Control rod guides and fuel element duct load pads have to be fabricated from materials exhibiting optimum slide behaviour. Galling or self-welding under static conditions should not be tolerated. Given bearing clearances have to be maintained constant and loop contamination, caused by wear particles, have to be prevented. Since high friction between contacting pads may impose severe limitations on core compaction, for the duct load pads a maximum friction coefficient of 0.5 is acceptable. The effect of sodium corrosion should not impair the friction and wear behaviour of the materials applied. This report covers the work performed to optain appropriate mechanical design data. (orig.) [de
Mizobuchi, Syotaro; Kano, Shigeki; Nakayama, Kohichi; Atsumo, Hideo
Friction and self-welding test were conducted on several materials used for the contacting and sliding components of a sodium cooled fast breeder reactor. In the present study, the friction and self-welding characteristics of each material were evaluated through measuring the kinetic and breakaway friction coefficients. The influence of oscillating rotation and vertical reciprocating motion on the friction mode was also investigated. The results obtained are as follows: (1) Colmonoy No.6, the nickel base hardfacing alloy, indicated the lowest kinetic friction coefficient of all the materials in the present study. Also, Cr 3 C 2 /Ni-Cr material prepared by a detonation gun showed the most stable friction behavior. (2) The breakaway friction coefficient of each material was dependent upon dwelling time in a sodium environment. (3) The friction behavior of Cr 3 C 2 /Ni-Cr material was obviously related with the finishing roughness of the friction surface. It was anticipated that nichrome material as the binder of the chrome carbide diffused and exuded to the friction surface by sliding in sodium. (4) The friction coefficient in sliding mode of vertical reciprocating was lower than that of oscillating rotation. (author)
Under normal high-temperature gas-cooled reactor (HTGR) operating conditions, faying surfaces of metallic components under high contact pressure are prone to friction, wear, and self-welding damage. Component design calls for coatings for the protection of the mating surfaces. Anticipated operating temperatures up to 850 to 950 0 C (1562 to 1742 0 F) and a 40-y design life require coatings with excellent thermal stability and adequate wear and spallation resistance, and they must be compatible with the HTGR coolant helium environment. Plasma and detonation-gun (D-gun) deposited chromium carbide-base and stabilized zirconia coatings are under consideration for wear protection of reactor components such as the thermal barrier, heat exchangers, control rods, and turbomachinery. Programs are under way to address the structural integrity, helium compatibility, and tribological behavior of relevant sprayed coatings. In this paper, the need for protection of critical metallic components and the criteria for selection of coatings are discussed. The technical background to coating development and the experience with the steam cycle HTGR (HTGR-SC) are commented upon. Coating characterization techniques employed at General Atomic Company (GA) are presented, and the progress of the experimental programs is briefly reviewed. In characterizing the coatings for HTGR applications, it is concluded that a systems approach to establish correlation between coating process parameters and coating microstructural and tribological properties for design consideration is required
Naghdi, R.; Sheibani, Sh.; Tamizifar, M.
The capsules containing radioactive materials as brachytherapy sources are used for implanting into some target organs for malignant disorders treatments, such as prostate, eyes, and brain cancers. The conventional method for sealing the tubes is to weld them using a laser beam which is now a part of tube melting methods (self welding). The purpose of this study was to seal miniature titanium tubes containing radioactive materials in the form of capsules. This study introduced a new method based on melting process. A piece of commercially pure titanium grade 2 in the form of disk was used for the experiment. The sample was melted at the top of the tube by a Tungsten Inert Gas welding device for a short time duration. After completion of the melting, the disk in the form of a drop was mixed with a small part of it and both were solidified and hence closed the tube. We evaluated the tubes for the metallurgical properties and seal process which took place by Tungsten Inert Gas in different zones, including the heat affected zone, fusion zone, and interface of the joint of the drop to the tube. Finally, the produced samples were tested according to the ISO2919 and ISO9978 and the results confirmed the Disk and Tungsten Inert Gas procedure.
This report describes progress in three HEDL programs supported by the U.S. Department of Energy's Division of Magnetic Fusion Energy. They are (A) Irradiation Effects Analysis, (B) Mechanical Performance of MFE Materials, and (C) Preparation and Presentation of Design Data. (A) Interatomic potentials are being developed for use in simulating displacement damage in binary alloys. A computer code is being written that derives A-A, A-B, and B-B potentials from macroscopic data on A 3 B alloys of L1 2 symmetry. The potentials are the Moliere type at small-to-intermediate separations and fitted cubics at large separations. Helium production cross sections for isotopes of Fe, Ni, and Cr, calculated with the HAUSER*4 code, are tabulated at 15 MeV. Agreement with measurements on Al and Cu was demonstrated previously. The energy dependence of the (n,α), (n,αn), and (n,nα) cross sections in the 13 to 20 MeV range are plotted for 56 Fe. (B) A computer code has been developed for calculating the energy deposition by 0.5 to 25 MeV protons incidents on a cylindrical metal specimen. (C) New additions to the Nuclear Systems Materials Handbook include low cycle fatigue of Type 304 stainless steel; swelling correlation for 30% cold-worked Type 316 stainless steel; friction, wear, and self-welding of Tribaloy 700; process guidelines on cleaning and cleanliness; physical properties of stainless steels; and fatigue-crack growth behavior of Inconel 600 and 718
Kishore, G. V. K.; Kumar, Anish; Chakraborty, Gopa; Albert, S. K.; Rao, B. Purna Chandra; Bhaduri, A. K.; Jayakumar, T.
Nickel base Ni-Cr alloy variants are extensively used for hardfacing of austenitic stainless steel components in sodium cooled fast reactors (SFRs) to avoid self-welding and galling. Considerable difference in the compositions and melting points of the substrate and the Ni-Cr alloy results in significant dilution of the hardface deposit from the substrate. Even though, both the deposit and the substrate are non-magnetic, the diluted region exhibits ferromagnetic behavior. The present paper reports a systematic study carried out on the variations in microstructures and magnetic behavior of American Welding Society (AWS) Ni Cr-C deposited layers on 316 LN austenitic stainless steels, using atomic force microscopy (AFM) and magnetic force microscopy (MFM). The phase variations of the oscillations of a Co-Cr alloy coated magnetic field sensitive cantilever is used to quantitatively study the magnetic strength of the evolved microstructure in the diluted region as a function of the distance from the deposit/substrate interface, with the spatial resolution of about 100 nm. The acquired AFM/MFM images and the magnetic property profiles have been correlated with the variations in the chemical compositions in the diluted layers obtained by the energy dispersive spectroscopy (EDS). The study indicates that both the volume fraction of the ferromagnetic phase and its ferromagnetic strength decrease with increasing distance from the deposit/substrate interface. A distinct difference is observed in the ferromagnetic strength in the first few layers and the ferromagnetism is observed only near to the precipitates in the fifth layer. The study provides a better insight of the evolution of ferromagnetism in the diluted layers of Ni-Cr alloy deposits on stainless steel.
Kishore, G.V.K.; Kumar, Anish, E-mail: email@example.com; Chakraborty, Gopa; Albert, S.K; Rao, B. Purna Chandra; Bhaduri, A.K.; Jayakumar, T.
Nickel base Ni–Cr alloy variants are extensively used for hardfacing of austenitic stainless steel components in sodium cooled fast reactors (SFRs) to avoid self-welding and galling. Considerable difference in the compositions and melting points of the substrate and the Ni–Cr alloy results in significant dilution of the hardface deposit from the substrate. Even though, both the deposit and the substrate are non-magnetic, the diluted region exhibits ferromagnetic behavior. The present paper reports a systematic study carried out on the variations in microstructures and magnetic behavior of American Welding Society (AWS) Ni Cr–C deposited layers on 316 LN austenitic stainless steels, using atomic force microscopy (AFM) and magnetic force microscopy (MFM). The phase variations of the oscillations of a Co–Cr alloy coated magnetic field sensitive cantilever is used to quantitatively study the magnetic strength of the evolved microstructure in the diluted region as a function of the distance from the deposit/substrate interface, with the spatial resolution of about 100 nm. The acquired AFM/MFM images and the magnetic property profiles have been correlated with the variations in the chemical compositions in the diluted layers obtained by the energy dispersive spectroscopy (EDS). The study indicates that both the volume fraction of the ferromagnetic phase and its ferromagnetic strength decrease with increasing distance from the deposit/substrate interface. A distinct difference is observed in the ferromagnetic strength in the first few layers and the ferromagnetism is observed only near to the precipitates in the fifth layer. The study provides a better insight of the evolution of ferromagnetism in the diluted layers of Ni–Cr alloy deposits on stainless steel. - Highlights: • Study of evolution of ferromagnetism in Comonoy-6 deposit on austenitic steel. • Magnetic force microscopy (MFM) exhibited ferromagnetic matrix in first two layers. • The maximum MFM