AMPP Store - Products tagged with 'environmental degradation 2022'

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Multi-Modal Characterization of Aggregates in Concrete via EDS and Raman Imaging

Product Number: ED22-17288-SG

Author: Krishna C. Polavaram, Elena Tajuelo Rodriguez, Amani Cheniour, Yann Le Pape, Nishant Garg

Publication Date: 2022

$20.00

The potential extension of the lifetime of nuclear power plants has cultivated an interest in the long-term aging behaviour of materials such as concrete. Since concrete is a complex material and its properties evolve with time, the effect of prolonged radiation exposure is of high interest and needs to be understood. Cracking and radiation-induced volumetric expansion (RIVE)(Le Pape et al., 2020) of the mineral components in aggregates occur as a result of neutron radiation and depends on several factors including the chemical nature and mineralogical characteristics of the aggregates such as composition, crystallinity, grain size, and phase distribution.

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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.

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Non-Arrhenius Stress Corrosion Crack Growth Behavior Of 304 Stainless Steel In Deaerated Water

Product Number: ED22-17245-SG

Author: Tyler Moss, Justin Hesterberg, Dave Morton, Denise Paraventi

Publication Date: 2022

$20.00

The Naval Nuclear Laboratory (NNL) has performed evaluations of SCC in 304/304L stainless steel since 2005 with the goal of developing an empirical equation. Testing has focused on the effects of temperature, stress intensity factor, material cold work, orientation, and sulfur content on SCC in hydrogenated water. Non-Arrhenius growth, termed herein as high temperature retardation (HTR), was observed in several studies where the SCC growth rate was found to slow at elevated temperature at low cold work levels in 316 and 304/304L stainless steel.

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Oxygen Consumption As Sensitive Measure Of Polymer Electrical Cable Insulation Aging

Product Number: ED22-17274-SG

Author: Yelin Ni, Katarzyna Grubel, Jonathan D. Egbert, Leonard S. Fifield

Publication Date: 2022

$20.00

It is of sustained interest to estimate the remaining useful lifetime of polymers used as cable insulation in nuclear power plants for cable aging management and license extension purposes. Studies have been focused on a range of topics from mechanism of degradation process, kinetic modeling, effects on chemical signatures and mechanical properties, and accelerated aging techniques for lifetime prediction.

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Performance Of Laser Additively Manufactured Alloys In (Light Water) Reactor Environments

Product Number: ED22-17282-SG

Author: Bogdan Alexandreanu, Xuan Zhang, Yiren Chen

Publication Date: 2022

$20.00

Nuclear power has been the largest source of carbon-free power in the U.S. (and much of the developed world) for almost a half century. As such, in the U.S. today, nuclear power plants of the Light Water Reactor (LWR) design generate 20% of all electricity, comprising over half of carbon-free electricity generation. In order to meet the short-term 2030 greenhouse gas emission reduction target, the existing nuclear fleet will play an important role, while the development and deployment of advanced reactors such as the small modular reactors (SMR) of the LWR design can be accelerated.

Phase Field Modeling Of Oxide Growth In An Aged Fe-Cr

Product Number: ED22-17141-SG

Author: Yongfeng Zhang, Líney Árnadóttir, Julie Tucker, Burkan Isgor, Kofi Oware Sarfo, Larry Aagesen

Publication Date: 2022

$20.00

Duplex stainless steels (DSS) are widely used as structural alloys in marine and energy industries because of their excellent combination of mechanical properties and corrosion resistance. In light water reactor (LWR) power plants, these alloys find their applications in piping and internal structural components. With a currently designed lifetime of 40 years, these DSS components show little degradation in their mechanical properties. However, most current and future nuclear power plants are expected to operate beyond 60 years. This prolonged service period challenges the integrity of materials and components in the reactor. DSS component lifetime in the reactor is subjected to elevated temperatures, internal pressures, and corrosive environments.

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Potassium Hydroxide For PWR Primary Coolant pH Control: EPRI Program

Product Number: ED22-17394-SG

Author: Keith Fruzzetti, Dennis Hussey, Peter Chou, David Perkins

Publication Date: 2022

$20.00

This paper outlines and summarizes the robust testing and assessment program developed and implemented by the Electric Power Research Institute (EPRI), following upon an initial feasibility evaluation completed in 2015. A multi-year, multi-discipline program has been developed, incorporating significant industry input, to address the identified technical gaps in materials, fuels, chemistry, and radiation safety that need evaluation to support a plant demonstration in a Western-design PWR.

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Preliminary Results From Assessment Of Advanced Manufacturing Technologies (AMT) For Reactor Materials And Components

Product Number: ED22-17118-SG

Author: Mark Yoo, Matthew Hiser, Amy Hull

Publication Date: 2022

$20.00

The Nuclear Regulatory Commission’s (NRC’s) approach to preparing to regulate and review industry proposals for using advanced manufacturing technologies (AMTs) in commercial nuclear applications focuses on identifying differences with AMT relative to conventional manufacturing. Initial AMTs based on industry interest include laser powder bed fusion (LPBF) and laser-directed energy deposition (L-DED) additive manufacturing (AM) methods, powder metallurgy-hot isostatic pressing (PM-HIP), electron beam welding (EBW), and cold spray (CS).

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

Product Number: ED22-18349-SG

Author: Peter L. Andresen, Robert G. Carter, Katsuhiko Kumagai

Publication Date: 2022

$20.00

SCC in Fe- and Ni-base alloys has been observed in high temperature water, both in the laboratory tests and in BWRs. SCC results from complex interactions of ~10 primary variables and hundreds of secondary variables, broadly categorized in terms of stress, environment and microstructure.

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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.

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PWSCC Crack Growth Rate Testing Of Wrought And HIP Alloy 625 In A PWR Environment.

Product Number: ED22-18522-SG

Author: J. Borg, S. Medway, F. Pickering, C. Smith

Publication Date: 2022

$20.00

This paper will discuss the crack growth rates measured for four different heats of HIP material and discuss possible relationships with hardness and stress intensity factor, along with considerations of grain size and features observed on the fracture surface.

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Regulatory Research On The Aging Management Of Structures, Systems, And Components In Nuclear Power Plants Supporting Long Term Operation

Product Number: ED22-18639-SG

Author: Mark Yoo, Matthew Hiser, Eric Focht, Carol Moyer, Amy Hull, Robert Tregoning

Publication Date: 2022

$20.00

This paper will focus on ongoing regulatory research related to aging management of reactor vessel internals, including measuring stress corrosion cracking initiation and growth rates, and developing a mechanistic understanding of other potential degradation modes, with a particular focus on issues that may be more important for operation beyond 80 years of life.

Products tagged with 'environmental degradation 2022'

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