Sample records for alloy-a-286

  1. Fracture mechanics behaviour of neutron irradiated Alloy A-286

    International Nuclear Information System (INIS)

    The effect of fast-neutron irradiation on the fatigue-crack propagation and fracture toughness behaviour of Alloy A-286 was characterized using fracture mechanics techniques. The fracture toughness was found to decrease continuously with increasing irradiation damage at both 24 deg. C and 427 deg. C. In the unirradiated and low fluence conditions, specimens displayed appreciable plasticity prior to fracture, and equivalent Ksub(Ic) values were determined from Jsub(Ic) fracture toughness results. At high irradiation exposure levels, specimens exhibited a brittle Ksub(Ic) fracture mode. The 427 deg. C fracture toughness fell from 129 MPa√m in the unirradiated condition to 35 MPa√m at an exposure of 16.2 dpa (total fluence of 5.2x1022n/cm2). Room temperature fracture toughness values were consistently 40 to 60 percent higher than the 427 deg. C values. Electron fractography revealed that the reduction in fracture resistance was attributed to a fracture mechanism transition from ductile microvoid coalescence to channel fracture. Fatigue-crack propagation tests were conducted at 427 deg. C on specimens irradiated at 2.4 dpa and 16.2 dpa. Crack growth rates at the lower exposure level were comparable to those in unirradiated material, while those at the higher exposure were slightly higher than in unirradiated material. (author)

  2. Babcock ampersand Wilcox experience with alloy A-286 reactor vessel internal bolting

    International Nuclear Information System (INIS)

    Multiple reactor vessel internal bolt failures were discovered during the 1981 and 1982 in service inspections performed at three PWR nuclear power plants. All the failures were limited to bolts that fastened the lower portion of the reactor vessel internal thermal shield to the lower grid assembly. Subsequent examinations during 1982, 1983 and 1984 revealed bolt failures at four additional plants. These failures included bolts that fastened the core barrel to the core support shield and lower grid assembly. Additional failures were also discovered in the bolts used to join the surveillance specimen holder tube to the thermal shield. All the affected fasteners were fabricated from Alloy A-286 (ASTM A453 Grade 660) material. Alloy A-286 is a high strength precipitation hardened austenitic stainless steel containing a nominal Cr and Ni content of 15% and 25%, respectively. As a result of these bolt failures, the Babcock ampersand Wilcox Co., under the direction of the B ampersand W Owners Group, performed extensive evaluations of Alloy A-286 reactor vessel internal fasteners. The principal conclusions obtained from this investigation are given below. 1. Internals bolting failures have been observed at nominal peak calculated stress levels of greater than or equal to 690 MPa (100 ksi). The number of failures generally increases with increasing stress. Variations in this correlation are postulated to be the result of scatter in the calculated peak stress data. 2. A variety of material conditions including the use of highly cold worked barstock in the fabrication of some of the bolts, degree of annealing and hot forging may have contributed to the bolt failures. 3. No specific upset environmental conditions were found that could be judged to be a leading cause of the bolt failures. 4 refs., 2 tabs

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

    International Nuclear Information System (INIS)

    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)

  4. Eddy-Current Detection of Weak Bolt Heads (United States)

    Messina, C. P.


    Electronic test identifies flawed units passing hardness tests. Eddy-current test detects weakness in head-to-shank junctions of 1/4-28 cup-washer lock bolts. Developed for alloy A286 steel bolts in Space Shuttle main engine fuel turbo-pump. Test examines full volume of head, including head-to-shank transition and nondestructively screens out potentially defective units. Test adapts to any other alloys.

  5. 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.

  6. Techniques developed to evaluate the fracture toughness offast breeder reactor duct

    International Nuclear Information System (INIS)

    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