Menu
Search
Filters
Close
Menu

Products tagged with 'crack growth rate'

View as
Sort by
Display per page
Picture for 09296 Fracture Analysis of Slow Strain Rate Test for Stress Corrosion Cracking
Available for download

09296 Fracture Analysis of Slow Strain Rate Test for Stress Corrosion Cracking

Product Number: 51300-09296-SG
ISBN: 09296 2009 CP
Author: Russell D. Kane and Anand Venkatesh
Publication Date: 2009
$20.00

Recently Crack Growth Rate and Threshold Stress Intentisity Factor  analyses of data from Slow Strain Rate tests have been used for Stress COrrosion Cracking evaluation. This methodology has been discussed in detail based on different analytical techniques and the results from laboratory tests have also been presented.
Picture for Effect of Strain Rate on the Crack Growth Rate of 718 Under Cathodic Potentials
Available for download

51317--9669-Effect of Strain Rate on the Crack Growth Rate of 718 Under Cathodic Potentials

Product Number: 51317--9669-SG
ISBN: 9669 2017 CP
Author: Thodla Ramgopal
Publication Date: 2017
$20.00

Crack growth rate (CGR) behavior of UNS N07718 was investigated as a function of K-rate in two different environments under cathodic potentials, a mild environment containing 3.5wt% NaCl and a more aggressive environment containing 0.5M H2SO4.

 
Picture for Stress corrosion cracking behaviour of Ni-based welding Alloys 182 and 152 in PWR primary water environment
Available for download

51318-11444-Stress corrosion cracking behaviour of Ni-based welding Alloys 182 and 152 in PWR primary water environment

Product Number: 51318-11444-SG
Author: Yun Soo Lim / Seong Sik Hwang / Sung Woo Kim / Hong Pyo Kim
Publication Date: 2018
$20.00

Primary water stress corrosion cracking (PWSCC) testing of Alloy 182 and 152 welds was conducted using 1/2 compact tension specimens at 325 ℃ in simulated primary water environments of a pressurized water reactor.
Picture for Application of Hydrogen Service Fatigue Crack Growth Rate Models to a Simulated Remaining Life Assessment of Transmission Pipeline
Available for download

Application of Hydrogen Service Fatigue Crack Growth Rate Models to a Simulated Remaining Life Assessment of Transmission Pipeline

Product Number: 51323-19368-SG
Author: B.C. Rollins, B.N. Padgett, T.J. Prewitt, A. Chandra, S. Finneran, R. Thodla
Publication Date: 2023
$20.00

The hydrogen economy envisions the use of gaseous hydrogen (herein referred to as hydrogen) as an energy carrier for the reduction of carbon emissions. Transportation of hydrogen from the upstream source (generation location) to the end-user will be necessary to maximize the carbon reduction potential switching from natural gas to pure hydrogen or hydrogen blended natural gas products. A proposed, economically viable option is to utilize the existing and extensive natural gas pipeline infrastructure in the United States.
Picture for Crack Growth Behaviour Of Cobalt Based Alloy In High Temperature Water
Available for download

Crack Growth Behaviour Of Cobalt Based Alloy In High Temperature Water

Product Number: ED22-17294-SG
Publication Date: 2022
$20.00

Co-based alloy (30Cr-4W-Bal.Co), such as Stellite TM grade 6 (UNS No. ERCOCr-A) valve seat, is used extensively in applications where superior resistance to wear corrosion are required. But there are only rare data about crack growth behavior of the alloy in the high temperature water. It is difficult to extract specimen from as welded Stellite-6, because of the valve sheet did not have enough volume. So, to evaluate the crack growth behavior and its temperature dependence, crack growth rate measurements were performed using forged Co-based alloy bar (Stellite TM grade 6B) (UNS No. R30016) in a simulated PWR primary water at temperature ranging from 250 to 320°C using half-inch size compact tension specimens (1/2TCT). 
Picture for Cracking Behavior Of Decommissioned Baffle Plates In Light Water Reactor Environment
Available for download

Cracking Behavior Of Decommissioned Baffle Plates In Light Water Reactor Environment

Product Number: ED22-17301-SG
Author: Yiren Chen, Bogdan Alexandreanu, Appajosula S. Rao
Publication Date: 2022
$20.00

Most of the core internal components of light water reactors (LWRs) are made of austenitic stainless steels (SSs). Exposed to fast neutron irradiation, reactor core internal components are vulnerable to irradiation-induced or irradiation-enhanced degradations. At LWR-relevant temperatures and irradiation doses, significant microstructural and microchemical changes can take place in SSs, leading to deteriorated mechanical and corrosion properties.
Picture for Effect Of Water Chemistry On Crack Growth Rates In Neutron Irradiated X-750 And XM-19
Available for download

Effect Of Water Chemistry On Crack Growth Rates In Neutron Irradiated X-750 And XM-19

Product Number: ED22-17287-SG
Author: John Jackson, Drew Johnson, Robert G. Carter, Michael Heighes, Peter L. Andresen
Publication Date: 2022
$20.00

As the current reactor fleet continues to age, with many reactors wanting to extend operational licenses beyond their initial 40 lifetime, it is becoming increasingly important to understand how structural materials in these reactor environments will degrade over time. A critical material degradation mode which can limit the lifetime of reactor components is irradiation assisted stress corrosion cracking (IASCC). As the name implies, a material must be subjected to both a corrosive environment as well as mechanical stresses while the original material microstructure has been affected by irradiation.
Picture for K / Size Effects on the Crack Growth Rate of Irradiated Stainless Steels of Boiling Water Reactors
Available for download

K / Size Effects on the Crack Growth Rate of Irradiated Stainless Steels of Boiling Water Reactors

Product Number: ED22-17304-SG
Author: Masato Koshiishi, Ryoji Obata, Shigeaki Tanaka
Publication Date: 2022
$20.00

Irradiation assisted stress corrosion cracking (IASCC) is a phenomenon caused by neutron irradiation of austenitic stainless steels (SSs). The crack growth rates (CGRs) of IASCC for boiling water reactor (BWR) components are needed for assessments to ensure component integrity. The CGR formula has been proposed as a function of the stress intensity factor (K).
Picture for Proposed Crack Growth Rate Disposition Curves For Stress Corrosion Cracking Of Alloy 82 In BWR Environments
Available for download

Proposed Crack Growth Rate Disposition Curves For Stress Corrosion Cracking Of Alloy 82 In BWR Environments

Product Number: ED22-18350-SG
Author: P.L. Andresen, D.J. Shim, T. Terachi, T. Yamada, Y. Nomura
Publication Date: 2022
$20.00

A database of SCC growth rates in commercial austenitic stainless steels exposed to pressurized water reactor (PWR) primary water environments was developed and analyzed from international data in high temperature water, with an emphasis on deaerated or hydrogenated water while also including water containing oxygen. Crack growth rate (CGR) disposition equations were derived to reflect the effects of stress intensity factor (K), temperature, Vickers hardness (HV, to represent retained deformation), with enhancement factors for oxygen-containing, high corrosion potential conditions. The tolerance to chloride and sulfate impurities in PWR primary water was also evaluated.
Review of Techniques to Evaluate SCC Susceptibility of Carbon Steels Used in Construction of Hanford Nuclear Reservation Waste Tanks
Available for download

Review Of Techniques To Evaluate SCC Susceptibility Of Carbon Steels Used In Construction Of Nuclear Waste Tanks

Product Number: 51322-18062-SG
Author: B.C. Rollins, S. Chawla, J. Gerst, J.A. Beavers, J. Page, C. Girardot, S. Campbell, N. Sridhar
Publication Date: 2022
$20.00

The Hanford Nuclear Reservation contains radioactive and chemically hazardous wastes arising mostly from weapons production, beginning with World War II and continuing through the Cold War. The wastes are stored in 177 carbon steel underground storage tanks, of which 149 are single-shell tanks (SSTs) and the remaining are double-shell tanks (DSTs). The U.S. Department of Energy, Office of River Protection is responsible for retrieving the tank wastes, treating them in order to encapsulate them in glass logs, and then permanently closing the tanks and associated facilities.
Picture for SCC Growth Behavior Of Stainless Steels And Nickel-Base Alloys During Chemical Transients In BWR NWC Environments
Available for download

SCC Growth Behavior Of Stainless Steels And Nickel-Base Alloys During Chemical Transients In BWR NWC Environments

Product Number: 51321-17008-SG
Author: Xiaoyuan Lou; Robert Carter; Raj Pathania
Publication Date: 2021
$20.00
Picture for SCC Initiation And Growth In PWR Primary Water Containing KOH Vs. Lioh
Available for download

SCC Initiation and Growth Rate Response of A-286 in B/Li and B/K Solutions 340°C

Product Number: ED22-17262-SG
Author: P.L. Andresen, P. Chou
Publication Date: 2022
$20.00

Stress corrosion cracking (SCC) initiation and growth rate testing was performed using Alloy A-286, a high-strength iron-base alloy, to evaluate for possible differences in response in boric acid solutions containing KOH vs. LiOH. PWRs are considering a switch from LiOH to KOH because of the uncertain future availability and high cost of Li. To achieve the same pH at temperature (pHT) in typical PWR primary water, the same molar concentration of Li and K is needed. The atomic weight of K is 39.1, which is 5.63 higher than the atomic weight of Li at 6.94, so 1 ppm Li yields the same pHT as 5.63 ppm K. The conductivity differs somewhat because of difference in the mobility of Li+ and K+.
Categories
Industry
Popular tags
View all

Association for Materials Protection and Performance

© AMPP All Rights Reserved.
Membership Terms of Content Use
Contact Customer Support