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3445 total products found.
Picture for Long-Term Performance Of Alloy 690 In LWR Environment
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Long-Term Performance Of Alloy 690 In LWR Environment

Product Number: ED22-17284-SG
Author: Bogdan Alexandreanu, Yiren Chen, Xuan Zhang, Wei-Ying Chen
Publication Date: 2022
$20.00

Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their increased resistance to stress corrosion cracking (SCC) relative to Alloys 600, 82, and 182. Many of these reactors are expected to operate for 40-80 years. Likewise, advanced water-cooled small modular reactors (SMRs) will use Ni-Cr alloys in their primary systems and are expected to receive initial operating licenses for 60 years.
Picture for Evaluating Irradiation Damage In CANDU Components By Characterizing Surrogate Material Harvested From The Decommissioned NRU Reactor At Chalk River Laboratories- Application To The CANDU Calandria Tubesheet Degradation
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Evaluating Irradiation Damage In CANDU Components By Characterizing Surrogate Material Harvested From The Decommissioned NRU Reactor At Chalk River Laboratories- Application To The CANDU Calandria Tubesheet Degradation

Product Number: ED22-17290-SG
Author: Mitchell Mattuchi, Qiang Wang, Brian Langelier, Nicolas Huin, Brian Connolly, Grace Burke
Publication Date: 2022
$20.00

The CANDU® reactor is a pressurized heavy water moderated (D2O) reactor, fueled with natural uranium. Nuclear fission of natural uranium occurs in the fuel channels and the heat generated from the fission is transported from the fuel channel to the steam generator, via the primary heat transport system. The nuclear reaction is moderated by heavy water that surrounds the fuel channels contained within the Calandria Vessel. 
Picture for Influence Of A Strain Gradient On The Stress Corrosion Cracking Behavior Of 304 Stainless Steel In High Temperature Water
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Effect of Complex Deformation on Corrosion Fatigue and Stress Corrosion Cracking of 304/304L Stainless Steel Elbows in Aerated and Deaerated Water

Product Number: ED22-17115-SG
Author: William C. Moshier, Kevin B. Fisher, Justin R. Hesterberg, and Denise J. Par
Publication Date: 2022
$20.00

Stress corrosion cracking (SCC) growth in 300-series stainless steels (SS) exposed to high temperature water is known to generally increase with increasing levels of cold work. The influence of cold work on SCC has been reported for both oxygenated boiling water reactor (BWR) normal water chemistry as well as for hydrogenated pressurized water reactor (PWR) water chemistry. 
Picture for Assessing Environmentally Assisted Cracking Of Structural Alloys In Molten Salts

Assessing Environmentally Assisted Cracking Of Structural Alloys In Molten Salts

Product Number: ED22-17120-SG
Author: George Young, Jake Quincey, Andrew Brittan, Pit Shulze, Stella Tsotsos, Lars Parrington, Josef Parrington, Anne Demma, Micah Hackett, Samuel Briggs and Julie D. Tucker
Publication Date: 2022
$20.00

The potential for structural alloys to undergo environmentally assisted cracking in molten salts is relatively unexplored due to their limited industrial application. However, fluoride salts are of prime interest to many advanced reactors including the Kairos Power FHR reactors. Table I summarizes literature studies of EAC in molten fluoride salts. For the ten studies shown, seven are for Ni-Mo-Cr family of alloys (INOR-8 / Hastelloy N or variants) that were used in the Molten Salt Reactor Experiment (MSRE), two studies investigate austenitic stainless steels, and there is one report of EAC in oxygen free high conductivity (OFHC) copper.
Picture for Crack Growth Behaviour Of Cobalt Based Alloy In High Temperature Water
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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 Elucidating The Effect Of Build And Build Direction On The Corrosion Of Selective Laser Sintered Additively Manufactured 316L Stainless Steel
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Elucidating the Effect of Build Location and Powder Morphology on the Corrosion of Selective Laser Sintered Additively Manufactured 316L Stainless Steel

Product Number: ED22-17456-SG
Author: E.Getto, E.Farnan, M.G. Koul, R.J. Santucci, T.Gray, J. S. Gibbs, J.McLaughlin
Publication Date: 2022
$20.00

Additive manufacturing (AM) has allowed for the rapid prototyping of parts and the ability to control the structure of a material and, to a lesser extent, the microstructure. AM as applied to metals, as stainless steel is especially promising as it provides the ability to produce complex shaped components rapidly. Laser-powder bed fusion (L-PBF) is one such technique in which the part is formed from the base up by fusing successive layers of powder. As each layer is fused, the print plate moves down and a new layer of metallic powder is swept over the top. A laser then welds this layer to the top of the existing piece and the process repeats.1 However, AM can result in non-equilibrium microstructures, porosity, and residual stresses that could affect the longevity of the material and performance in corrosive conditions.
Picture for The Temperature Functionality Of Sensitized 304 Stainless Steel SCC Growth In Deaerated Water

The Temperature Functionality of Sensitized Stainless Steel SCC Growth in Deaerated Water

Product Number: ED22-17110-SG
Author: David Morton, Elaine West, Tyler Moss and George Newsome
Publication Date: 2022
$20.00

The SCC of stainless steels has been an issue facing light water reactors (LWRs) since 1965 when sensitized components failed in the Dresden boiling water reactor (BWR). Numerous experimental efforts have been performed to characterize the SCC of stainless steel in LWRs in the last several decades and many of these efforts have been reported at each of the prior Environmental Degradation of Materials Conferences. Recent research has focused on characterizing SCCGR dependencies in hydrogen deaerated water. Testing of cold worked (CW) stainless steel has shown that heavily CW stainless steel has Arrhenius temperature functionality with a thermal activation energy of roughly 75 kJ/mol . In moderately to low CW stainless steel, a departure from Arrhenius temperature functionality is observed due to high temperature SCCGR retardation (HTR). This paper further extends this research and describes tests which were conducted to characterize the SCCGR temperature dependency of sensitized and CW 304 SS in hydrogenated water.
Picture for Simulation Of Neutron Irradiation Effects On Zr(Fe,Cr)2 Precipitates In Zircaloy-4 Using A Two-Step Proton Irradiation
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Zr(Fe,Cr)2 Precipitate Amorphization and Iron Redistribution in Zircaloy-4 Using a Two-Step Proton Irradiation

Product Number: ED22-17211-SG
Author: Peng Wang, Gary S. Was
Publication Date: 2022
$20.00

The corrosion of zirconium-based alloys is a service life-limiting factor in fuel rod performance. Mechanistic understanding of the corrosion process under reactor irradiation conditions still alludes to the nuclear industry. Pre-transition corrosion behavior of Zircaloy-4 has been reported to show a minimal effect from the irradiation environment, and the in-reactor corrosion kinetics is athermal and similar to the ex-situ autoclave corrosion exposure. However, the post-transition in-reactor corrosion kinetics depends on temperature and neutron flux. As discussed by Kammenzind et al. in Ref., the long-term post-transition corrosion rates of Zircaloy-4 are significantly accelerated in a PWR radiation environment over that observed with non-irradiated specimens in an autoclave environment.
Picture for U.S. Perspective On Harvesting And Archiving Of Aged Components To Improve Materials Degradation Knowledge And Inform Aging Management
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U.S. Perspective On Harvesting And Archiving Of Aged Components To Improve Materials Degradation Knowledge And Inform Aging Management

Product Number: ED22-17113-SG
Author: M.Hiser, P. Xu, R.Tregoning, J. Smith, C. Judge
Publication Date: 2022
$20.00


Historically, regulators, industry and other research organizations have performed research on materials harvested from a broad range of components, including the reactor pressure vessel (RPV), internals, and piping. Harvesting has included both service-aged materials as well as components from unfinished reactors. This harvesting and associated research has provided valuable insights into materials performance, such as flaw populations, materials properties, aging effects and non-destructive evaluation effectiveness. 
Picture for Use Of Halophytes To Manage Soil Characteristics In Joint Use Environments
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Use Of Halophytes To Manage Soil Characteristics In Joint Use Environments

Product Number: 51321-16550-SG
Author: Jon H. Brasher/ Evelyn Boone/Dr. Dennis Delaney/ Dr. Audrey Gamble/ Dr. Dale Monks/ Selina Bruckner
Publication Date: 2021
$20.00
Laying the Foundation for an Engineered and Integrated Approach to Pipeline External Corrosion Protection
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Laying the Foundation for an Engineered and Integrated Approach to Pipeline External Corrosion Protection

Product Number: 51322-17962-SG
Author: Keith Parker, Trey Johnston, Dale Lindemuth, Alfonso Garcia, Stephen B. Gibson
Publication Date: 2022
$20.00

Enbridge is proposing to develop a program that utilizes state-of-the-art technologies and proven inspection methods to prescribe interventions related to external corrosion mitigation using a predictive, integrated approach. This new program embraces complex problems by collecting, analyzing, and integrating environmental, pipeline integrity, and corrosion control data to predict external corrosion risk with sound engineering models (mechanistic, reliability and risk) to anticipate, prevent, and contain unexpected events.
Picture for Assessment Of Neutron Irradiation Damage In Aggregates
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Assessment Of Neutron Irradiation Damage In Aggregates

Product Number: ED22-17265-SG
Author: Elena Tajuelo Rodriguez, Lawrence M. Anovitz, Michael Cheshire, Yann Le Pape, Jan Ilavsky, Thomas M. Rosseel
Publication Date: 2022
$20.00

The required electrical power in the United States has led the utilities and the US Nuclear Regulatory Commission to evaluate second license renewals for operating light-water reactors, and some extensions have already been reviewed for extended operation to 80 years. As these plants were licensed to operate for 40 years with options for additional 20 year extensions, the extended operation raised questions in terms of materials performance under extreme conditions and extended time. The effects of prolonged irradiation must be understood and evaluated to predict and ensure the reliability of plant components. 

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