AMPP Store - Products tagged with 'environmental degradation 2022'

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Long-Term Corrosion Behavior Of Borated Stainless Steel In Simulated Spent Fuel Pool Environment

Product Number: ED22-17318-SG

Author: Daehyeon Park, Yunju Lee, Junhyuk Ham, Seung Chang Yoo, Kiyoung Kim, Donghee Lee, Yongdeog Kim, Ji Hyun Kim

Publication Date: 2022

$20.00

Most of countries store spent nuclear fuels in pools (SFP) which are built in nuclear power plants. As number of nuclear power plants and corresponding number of spent fuels increased, density in SFP storage rack also increased. In this regard, maintain subcriticality of spent nuclear fuels was raised as an issue and BSS was selected as structural material and neutron absorber for high density storage rack. BSS has better mechanical properties than other neutron absorbers which were fabricated based on aluminum alloys.

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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.

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Long-Term Performance Of High-Cr, Nickel-Based Weld Alloys In LWR Environments

Product Number: ED22-17285-SG

Author: Bogdan Alexandreanu, Yiren Chen, Xuan Zhang

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.

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Low Voltage Cable Condition Monitoring To Detect Global Degradation And Localization Techniques For Aging Management At Nuclear Power Plants

Product Number: ED22-17298-SG

Author: Andrew Mantey, Sarajit Banerjee, Pierre Simo, David Rouison, Jackson Bornath, Lindsay Vasilak

Publication Date: 2022

$20.00

A cornerstone of aging management programs for commercial nuclear reactors is the condition monitoring techniques used to determine insulation degradation of cables. Improved condition monitoring methods has been the focus of research especially for low voltage cables. There are many effective methods available such as elongation at break, indenter modulus, oxidation induction, etc.

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LWR Aging Management For Life Beyond 80

Product Number: ED22-17299-SG

Author: Thomas M. Rosseel, Xiang Chen, Elena Tajuelo Rodriguez, Mikhail Sokolov, Maxim Gussev, Yann Le Pape, Frank Garner, Leo S. Fifield, S. W. Glass

Publication Date: 2022

$20.00

Based on US Energy Information Administration (EIA) Annual Energy Outlook (AOE) predictions, by 2050, the US nuclear capacity for electricity generation may decrease to ~ 80% of 2021 levels, Moreover, by 2050, nearly 50% of the existing US LWR fleet will be within 10 years of 80 years of operation suggesting that without a Life Beyond 80 (LBE) plan and limited new builds and advanced reactors, the US could lose up to 50% of its nuclear capacity resulting in a ~30-gigawatt (GW) capacity shortage by 2060. These numbers could change dramatically depending upon oil and gas supplies, and the growth of renewables.

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Material and Manufacturing Study to Enable Coreblocks for Microractor Technology

Product Number: ED22-17136-SG

Author: Lindsay B. O’Brien, Holly R. Trellue, Amber N. Black, Ryan M. Mier, Andrew N. Duffield, Colt J. Montgomery, John. S. Carpenter

Publication Date: 2022

$20.00

Microreactor technology has the potential to provide efficient, modular, and inherently safe baseload power that can be used in regions that are too remote to support the larger, light water reactor (LWR) technology that dominates today’s nuclear energy landscape. To generate enough power and thermal efficiency to be attractive, the microreactors must be operated at higher temperatures (approximately 1112-1652°F or 600-900°C) than traditional LWR’s, and therefore are cooled using technologies such as heat pipes with gas, sodium, or molten salt coolants.

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Microstructural Changes in He Irradiated Zircaloy-4 and Influence On Corrosion Kinetics

Product Number: ED22-18343-SG

Author: Alistair Garner, Alexander Carruthers, Philipp Frankel, Sam Armson, Conor Gillen, Sarah Sherry, Mark Fenwick, Aidan Cole-Baker

Publication Date: 2022

$20.00

The corrosion of Zircaloy-4 under autoclave conditions without the presence of radiation is relatively well understood, with the development of cyclic corrosion kinetics that are well simulated by correlative predictive models (1) (2). Under irradiation in a PWR environment, however, the corrosion kinetics of Sn-containing Zr alloys are severely accelerated and although early corrosion behaviour is unchanged, after an oxide thickness of ~5 μm, accelerations of up to 40 x out-of-pile behaviour are observed (3) (4). Among the likely contributors to this accelerated corrosion are neutron irradiation damage to both the substrate and oxide, gamma irradiation, radiolysis, and hydrogen effects.

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Microstructure And Mechanical Properties Of Additively Manufactured 316L Stainless Steel.

Product Number: ED22-18491-SG

Author: Karyn Cooper, Stuart Medway, Jennifer Borg

Publication Date: 2022

$20.00

Additive manufacturing is a term that encompasses a number of technologies that manufacture structures by building material up, layer by layer, and which are attractive due to a number of factors, such as the ability to rapidly produce complex components with controlled microstructures in a single step with reduced post processing requirements. Laser-powder bed fusion (L-PBF) is an additive manufacturing technique where a laser continuously melts successive layers of powder material, building up from a horizontal build plate.

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Microstructure Characterization Of Alloy 600TT Steam Generator Tubing

Product Number: ED22-17239-SG

Author: Amanda Jenks, Chuck Marks, Brent Capell

Publication Date: 2022

$20.00

This paper describes the evolution of production standards for Alloy 600 tubing, the historical performance of steam generator tubing, and the results of microstructural analyses of archive and pulled tubing samples from commercial PWRs to address these issues. Alloy 600 is a corrosion-resistant nickel-base alloy that is used in a variety of applications that require good resistance to general corrosion, high strength, and good formability. It has been used extensively for steam generator tubing in commercial nuclear power plants, and this experience led to the use of several different types of Alloy 600 material.

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Micro-Tensile Testing of Chromium Coated Zirconium Cladding

Product Number: ED22-17270-SG

Author: David Frazer, Cameron Howard, Yachun Wang, Daniel Murray, Fei Teng, Colin Judge, Jorie Walters, Benjamin Maier, Allan Jaworski, Jonathan Wright, Luke Olson

Publication Date: 2022

$20.00

After the Fukushima accident there has been a large push globally for accident tolerant fuels (ATF) to increase the grace period during an accident, that is, the time during which operators may be able to avoid major consequences by undertaking mitigating actions. At Fukushima, the oxidation of the Zircaloy cladding produced hydrogen gas, that contributed to the failure of the primary containment. A concept for ATF is to coat zirconium-based cladding with chromium to inhibit the oxidation of the cladding and reduce hydrogen production.

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Modeling Stress Corrosion Cracking With Explicit 3D Grain Microstructure

Product Number: ED22-17160-SG

Author: Benjamin S Anglin, Thomas W Web

Publication Date: 2022

$20.00

Stress corrosion cracking (SCC) of Type 304 stainless steel (304 SS) in elevated temperature (288 °C) high purity water is typically an intergranular (IG) process with cracks propagating along grain boundaries, which are mesoscopic entities relevant on the grain scale. It follows then that the nature of the grain boundaries plays a significant role in SCC. In fact, for IG SCC to occur three things must be present: 1) stress; 2) a corrosive environment; and 3) susceptible grain boundaries. SCC growth rate (SCCGR) equations for 304SS in high temperature, high purity water, test orientation, temperature, material composition, and sensitization.

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Modelling Of IGSCC With Adaptive Framework

Product Number: ED22-17225-SG

Author: Michal Sedlak Mosesson, Pål Efsing

Publication Date: 2022

$20.00

Intergranular Stress Corrosion Cracking (IG-SCC) plays an important role as one of the most recognized degradation phenomena in Nuclear Power Plants (NPP). SCC is both multi-disciplinary with many parameters that are dependent on each other. This study was based on developing a multi-physics finite element model for IG-SCC prediction in unirradiated structural materials for non-pressure vessel components in NPPs. The environment considered was boiling water reactor (BWR) with normal water chemistry (NWC), containing approx. 200ppb oxidant (O2 + H2O2) and varying aggressive ions Cl-. The model was focused on the slip-oxidation model, where a crack is advancing by anodic dissolution, passivation, and oxide rupture at the crack tip. The rupture of the oxide film is due to the constant stresses applied creating slips in the bulk material which fractures the oxide.

Products tagged with 'environmental degradation 2022'

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