Products tagged with 'irradiation-assisted stress corrosion cracking'

View as

Sort by

Display per page

Available for download

Characterization Of Materials Showing Large Differences In IASCC Response

Product Number: ED22-18360-SG

Author: Anders Jenssen, Daniel Jädernäs, Peter Chou, Kristina Lindgren, Mattias Thuvander

Publication Date: 2022

$20.00

Irradiation-assisted stress corrosion cracking (IASCC) is a term used to describe cracking of austenitic materials in components subjected to a relatively high fast neutron flux. Like intergranular stress corrosion Paper No.18360 cracking (IGSCC), IASCC appears as intergranular cracks, but thermal sensitization of the grain boundaries is not required for the material to become susceptible to cracking. Service failures caused by IASCC have occurred in components such as core shrouds and top guides in boiling water reactors (BWRs) and baffle bolts in pressurized water reactors (PWRs).

Available for download

Crack Initiation Testing of Neutron-irradiated Type 347 Stainless Steel in Simulated PWR with LiOH or KOH

Product Number: ED22-17414-SG

Author: Michael Ickes, Lijin Dong, Abdullah Sinjlawi, Gary Was

Publication Date: 2022

$20.00

The immediate objective of this experiment is to investigate the IASCC initiation behavior of Type 347 stainless steel in lithium hydroxide and potassium hydroxide water chemistries across a range of irradiation damage and stress levels. A further objective is to provide data supporting improved predictive capabilities of IASCC failures by assessing the radiation dose dependence of IASCC initiation. In power plant components like the baffle-former bolts, the crack initiation step of IASCC is the rate limiting step, taking much longer than crack propagation as a fraction of time to failure. The results of this study will also be directly beneficial to the U.S. nuclear industry by providing an understanding of IASCC susceptibility in potassium hydroxide water chemistry, which may provide cost savings and more secure supply chains to nuclear power plants.

Available for download

The Effect Of Potassium Hydroxide Primary Water Chemistry On The IASCC Behavior Of 304 Stainless Steel

Product Number: ED22-17415-SG

Author: Michael Ickes, Peter Chou, Kai Chen, Gary Was

Publication Date: 2022

$20.00

The objective of this work was to compare irradiation-assisted stress corrosion cracking (IASCC) growth behavior in simulated pressurized water reactor (PWR) water with pH maintained with LiOH versus KOH. The U.S. nuclear industry is considering changing PWR primary water chemistries to use KOH in place of LiOH, as a means to ensure a stable supply chain and secure cost savings. This experiment will specifically investigate the impact of these alkali ions on the crack growth rate (CGR) and to examine the crack morphologies generated by the CGR experiment.

Products tagged with 'irradiation-assisted stress corrosion cracking'

Characterization Of Materials Showing Large Differences In IASCC Response

Crack Initiation Testing of Neutron-irradiated Type 347 Stainless Steel in Simulated PWR with LiOH or KOH

The Effect Of Potassium Hydroxide Primary Water Chemistry On The IASCC Behavior Of 304 Stainless Steel

Association for Materials Protection and Performance