Sample records for alloy-a-286

  1. Destructive examination of a cracked alloy A-286 Vent Valve Jackscrew from the Oconee unit 1 nuclear station

    International Nuclear Information System (INIS)

    Fyfitch, S.; Davidsaver, S.B.; Redmond, K.; Whitaker, D.E.; Doss, R.L.


    During the Fall 2012 refueling outage at Oconee Nuclear Station Unit 1, a reactor vessel internals video inspection revealed an abnormal condition on a reactor vessel vent valve. One of the Alloy A-286 jackscrews on this vent valve was visibly extended more than the other and bent, the lower section of the jackscrew threads were galled, and the lower barrel nuts were recessed. Furthermore, a circumferential crack-like indication was identified on the lower portion of the other jackscrew. The vent valve assembly was replaced during the refueling outage and the cracked portion of the jackscrew was submitted to Duke Energy's metallurgy lab for a failure mode determination. The failure investigation included metallography, scanning electronic microscopy (SEM), energy dispersive x-ray spectroscopy (EDS), and ASTM grain size determination. This paper determines that the failure most likely occurred from the resultant bending stress on the cracked jackscrew as the result of an impact load to the other jackscrew, and initiated and propagated by an intergranular stress corrosion cracking (IGSCC) mechanism. (authors)

  2. Embrittlement of nickel-, cobalt-, and iron-base superalloys by exposure to hydrogen (United States)

    Gray, H. R.


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

  3. Comprehensive Understanding of Ductility Loss Mechanisms in Various Steels with External and Internal Hydrogen (United States)

    Takakuwa, Osamu; Yamabe, Junichiro; Matsunaga, Hisao; Furuya, Yoshiyuki; Matsuoka, Saburo


    Hydrogen-induced ductility loss and related fracture morphologies are comprehensively discussed in consideration of the hydrogen distribution in a specimen with external and internal hydrogen by using 300-series austenitic stainless steels (Types 304, 316, 316L), high-strength austenitic stainless steels (HP160, XM-19), precipitation-hardened iron-based super alloy (A286), low-alloy Cr-Mo steel (JIS-SCM435), and low-carbon steel (JIS-SM490B). External hydrogen is realized by a non-charged specimen tested in high-pressure gaseous hydrogen, and internal hydrogen is realized by a hydrogen-charged specimen tested in air or inert gas. Fracture morphologies obtained by slow-strain-rate tensile tests (SSRT) of the materials with external or internal hydrogen could be comprehensively categorized into five types: hydrogen-induced successive crack growth, ordinary void formation, small-sized void formation related to the void sheet, large-sized void formation, and facet formation. The mechanisms of hydrogen embrittlement are broadly classified into hydrogen-enhanced decohesion (HEDE) and hydrogen-enhanced localized plasticity (HELP). In the HEDE model, hydrogen weakens interatomic bonds, whereas in the HELP model, hydrogen enhances localized slip deformations. Although various fracture morphologies are produced by external or internal hydrogen, these morphologies can be explained by the HELP model rather than by the HEDE model.

  4. Techniques developed to evaluate the fracture toughness offast breeder reactor duct. [Use of J-integral tests

    Energy Technology Data Exchange (ETDEWEB)

    Huang, F. H.; Wire, G. L.


    Large changes in strength and ductility of metals after irradiation are known to occur. The fracture toughness of irradiated metals, which is related to the combined strength and ductility of a material, may be significantly reduced and the potential for unstable crack extension increased. Therefore, the resistance of cladding and duct materials to fracture after exposure to fast neutron environments is of concern. Existing Type 316 stainless steel irradiated ducts are relatively thin and since this material retains substantial ductility, even after irradiation, the fracture behavior of the duct material cannot be analyzed by linear elastic fracture mechanics techniques. Instead, the multispecimen R-curve method and J-integral analysis were used to develop an experimental approach to evaluate the fracture toughness of thin breeder reactor duct materials irradiated at elevated temperatures. Alloy A-286 was chosen for these experiments because the alloy exhibits elastic/plastic behavior and the fracture toughness data of thicker (12 mm) specimens were available for comparison. Technical problems associated with specimen buckling and remote handling were treated in this work. The results are discussed in terms of thickness criterion for plane strain.

  5. Examination of a failed reactor coolant pump rotating assembly from Crystal River Unit 3

    International Nuclear Information System (INIS)

    Hayner, G.O.; Lubnow, T.; Clary, M.


    On January 18, 1989, the A reactor coolant pump rotating assembly at the Crystal River Unit 3 Nuclear Power Plant failed during operation. A rotating assembly from this pump had previously failed in 1986. The reactor coolant pump was fabricated by Byron Jackson Pump Division of Borg-Warner Ind. Products, Inc. from UNS S66286 superalloy (Alloy A286). A root cause failure analysis examination was performed on the pump shaft and other components. The failure analysis included shaft vibrational mode and stress analyses, pump clearance and alignment analyses, and detailed destructive examination of the shaft and hydrostatic bearing assemblies. Based on the detailed physical examination of the shaft it was concluded that cracks initiated in the pump shaft at two sites approximately 180 0 apart in a band of shallow, thermally induced fatigue cracks. The cracks initiated at the bottom edge of the motor end shrink fit pad under the shrink fit sleeve supporting the hydrostatic bearing journal. The band of thermally induced fatigue cracks was apparently caused by mixing of cold seal injection water and hot reactor coolant in gaps between the pump shaft and sleeve. The motor end shrink fit was apparently not effective in preventing introduction of the seal injection water to this area. Initial crack propagation occurred by fatigue due to lateral vibration; however, the majority of crack propagation occurred by abnormal torsional fatigue loading induced by contact and sticking between the rotating and stationary portions of the hydrostatic bearing. Final fracture of the shaft occurred by torsional overload. Metallurgical characteristics and mechanical properties of the shaft were within design specification and probably did not significantly influence the cracking process