Zobrazeno 1 - 10
of 14
pro vyhledávání: '"Fracture (geology)"'
Publikováno v:
Journal of Nuclear Materials. 498:117-128
The stress corrosion cracking (SCC) behavior of Alloy 690 with 0, 20% and 30% cold work (CW) was studied in supercritical water (SCW) environment with an emphasis on CW and creep on the CGRs (CGR). SCC and creep CGRs increased with %CW, which correla
Publikováno v:
Journal of Nuclear Materials. 496:313-324
The effect of prior single tensile overloading on the stress corrosion cracking behavior of cold rolled 316L in a simulated PWR water environment at 310 °C was investigated. SCC growth retardation by overloading was observed in cold rolled 316L spec
Autor:
M.J. Konstantinović
Publikováno v:
Journal of Nuclear Materials. 495:220-224
A probabilistic fracture model is applied to irradiation assisted stress corrosion cracking effect, assuming that the oxidized part of stainless steel sample plays an essential role in the crack initiation, propagation and the sample failure. The Wei
Publikováno v:
Journal of Nuclear Materials. 461:112-121
IASCC (Irradiation Assisted Stress Corrosion Cracking) crack initiation tests have been carried out on thimble tube material retrieved from a Belgian PWR. The crack initiation tests were carried out by constant load testing of thimble tube specimens
Publikováno v:
Journal of Nuclear Materials. 458:253-263
Stress corrosion cracking growth rates of Alloy 182 weld metals in T–S and T–L orientations in 288 °C pure water with various dissolved oxygen and hydrogen concentrations were measured. Extensive inter-dendritic stress corrosion cracking paths o
Publikováno v:
Journal of Nuclear Materials. 458:206-215
Stress corrosion cracking (SCC) susceptibility of 316Ti SS in supercritical water (SCW) was evaluated by Slow Strain Rate Tensile (SSRT) testing at four different temperatures from 380 °C up to 650 °C, at a pressure of 25 MPa and a strain rate of 9
Autor:
Hwanil Je, Akihiko Kimura
Publikováno v:
Journal of Nuclear Materials. 455:507-511
The stress corrosion cracking (SCC) susceptibility in supercritical pressurized water (SCPW) was investigated for the candidate structural steels of advanced fusion and fission nuclear system, which are SUS316L austenitic steel, F82H ferritic–marte
Publikováno v:
Journal of Nuclear Materials. 454:274-282
The effects of temperature on the tensile properties and stress corrosion cracking (SCC) susceptibility of austenitic stainless steels AL-6XN, HR3C, and 316Ti in deaerated supercritical water were studied by performing slow-strain-rate tensile (SSRT)
Autor:
Jianqiu Wang, Litao Zhang
Publikováno v:
Journal of Nuclear Materials. 446:15-26
Stress corrosion crack growth tests of a cold worked nuclear grade 316L stainless steel were conducted in simulated pressurized water reactor (PWR) primary water environment containing various dissolved oxygen (DO) contents but no dissolved hydrogen.
Publikováno v:
Journal of Nuclear Materials. 427:364-377
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 fai