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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. 
Picture for K / Size Effects on the Crack Growth Rate of Irradiated Stainless Steels of Boiling Water Reactors
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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 Nanoscale Characterization of the Oxide Formed on a Neutron-Irradiated Zr-2.5Nb Alloy in High Temperature Water
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Nanoscale Characterization of the Oxide Formed on a Neutron-Irradiated Zr-2.5Nb Alloy in High Temperature Water

Product Number: ED22-18364-SG
Author: S.Adil, M. Topping, K. Daub, S.Y. Persaud, M.R. Daymond
Publication Date: 2022
$20.00

The interaction of metals and alloys with aqueous environments is ubiquitous, leading to oxide formation (passivity) or corrosion in many cases. Although these phenomena have significant importance across various industries and domains of materials science, the fundamental atomic-scale mechanisms by which corrosion and oxide formation operate are still unclear. Oxide films can have complex chemistry and texture, especially at the metal-oxide interface which acts as the primary barrier from solution interaction. The Zr-H2O system has industrial and academic interest due to its use in nuclear reactors.
Picture for Mechanistic Understanding Of Irradiation Assisted Stress Corrosion Cracking
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Role of Grain Boundary Oxidation in Crack Initiation of Neutron-irradiated Stainless Steel

Product Number: ED22-17333-SG
Author: Srinivasan Swaminathan, Donghai Du, Gary S. Was
Publication Date: 2022
$20.00

Irradiation assisted stress corrosion cracking (IASCC) continues to be a major concern for the
structural integrity of core internals in both pressurized water reactors (PWRs) and boiling water
reactor (BWRs). While factors such as stress, an irradiated microstructure and a high temperature
water environment are required for IASCC, a better understanding of the underlying mechanism
has become a subject of intense long-term research. In the last two decades, much progress has
been made in understanding IASCC susceptibility, though a clear cause-and-effect has yet to be
established on the mechanism of intergranular cracking in highly neutron irradiated stainless steels
in the PWR environment.
Picture for The Complexity And Challenges Of Irradiation-Induced Phase Instability Phenomenon In Nuclear Power Plant Components
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The Complexity And Challenges Of Irradiation-Induced Phase Instability Phenomenon In Nuclear Power Plant Components

Product Number: ED22-17322-SG
Author: D.A. Merezhko, M.N. Gussev, M.S. Merezhko, T.M. Rosseel, F. A. Garner
Publication Date: 2022
$20.00

Austenitic stainless steels (AuSS) are widely used as structural materials for nuclear reactor vessel internals (RVIs), as well as for fuel cladding and pressurizer components. Some of these components cannot be removed and replaced, and therefore the irradiation performance of the steel determines the lifetime of each reactor component. Typical irradiation-induced detrimental effects in light water-cooled power reactors include embrittlement, accelerated creep, and radiation-altered corrosion. Some second-order effects such as void swelling, hydrogen accumulation, and radiation-induced phase instability might be slowly evolving to first-order importance, especially as Western nuclear power plants are being considered for lifetime extensions to 60 and possibly 80 total years.
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