AMPP Store - Products tagged with 'environmental degradation 2022'

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The Effect Of Specimen Size On The Stress Corrosion Crack Growth Rate Of 10% Cold Worked Type 316L In BWR Environment

Product Number: ED22-17319-SG

Author: Johan Stjärnsäter, Anders Jenssen, Elena Calota, Hannah Johansson

Publication Date: 2022

$20.00

Fracture mechanical specimens of the compact tension (CT) type are normally used in tests that study stress corrosion crack (SCC) growth rates (CGR). Normally, the width, W, is twice the thickness, B (W=2B), and B for common specimen sizes is 12.5 or 25 mm. The specimen size can be changed by scaling its dimensions.

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The Effect Of Strain On Surface Residual Stress In Alloy 600 Tensile Specimens

Product Number: ED22-17258-SG

Author: Karyn Cooper, William Brayshaw, Mark Chatterton, Sonya Pemberton, Stuart Medway, Jennifer Borg

Publication Date: 2022

$20.00

Nickel based Alloy 600 is used within the nuclear industry in structural components due to its good mechanical properties and general corrosion resistance, however upon exposure to primary water environments at elevated temperatures it can be affected by Primary Water Stress Corrosion Cracking (PWSCC). Nickel Based Alloy (NBA) susceptibility to PWSCC is dependent on a number of factors that include material type, condition and microstructure, as well as fabrication method, and can be investigated by uniaxial initiation testing in a primary water environment, where specimens are held at constant load under an elevated temperature.

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The SCC Initiation Behavior Of Alloy 600 And SS 316L In Simulated PWR Primary Water Environments With A Low Dissolved Hydrogen

Product Number: ED22-18330-SG

Author: Hsiang-Ling Shih, Mei-Ya Wang, Tsung-Kuang Yeh, Liang-Hsuan Chen

Publication Date: 2022

$20.00

Alloy 600 and SS 316L are common materials used for structural components of pressurized water reactors (PWRs). However, as PWRs age, incidents of general corrosion and stress corrosion cracking (SCC) are more likely to be found in the structural components. One of the major material degradation problems is primary water stress corrosion cracking (PWSCC).

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.

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Thermo-Hygro-Mechanical Modeling Of ASR-Induced Degradation In A Reinforced Concrete Nuclear Containment Vessel

Product Number: ED22-18345-SG

Author: Amit Jain, Benjamin W. Spencer, Sudipta Biswas

Publication Date: 2022

$20.00

Alkali-silica reaction (ASR) induced damage is one of the main causes of degradation in reinforced concrete (RC) structures, especially in the high relative humidity environmental conditions. ASR involves complex dissolution-precipitation reactions in concrete that take place in the presence of alkali ions, silica, and moisture. Alkali ions diffuse into the porous aggregate through the concrete pore solution and start
the dissolution of silica by breaking silanol and siloxane bonds in the reactive aggregates.

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Transition to Online Test / Verification of Cable Insulation Condition

Product Number: ED22-17269-SG

Author: S.W. Glass, L. S. Fifield, M. Spencer, A. Sriraman

Publication Date: 2022

$20.00

Under current energy-market conditions, the nuclear industry must innovate and move towards a more economically viable approach for many operations and maintenance (O&M) activities. One O&M change can include minimizing labor involved and frequency of O&M activities, including shifting some manual tests to automated online testing. Changing test practices must not only consider costs but also risks and efficacy of new test approaches compared to current practice.

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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.

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ulti-Axial Fatigue Initiation Testing and Specimen Characterization of 304 Stainless Steel

Product Number: ED22-18372-SG

Author: Eleanor M Grieveson, Peter Gill, Paul Onwuarolu, Ben Coult, Alistair Garner

Publication Date: 2022

$20.00

Current fatigue assessments for the fatigue life of a plant component are usually based on methodologies that use uniaxial fatigue test data (i.e. ASME Section III, and are intended to be conservative for design and fitness-for-purpose assessments when applied to plant components and loading. This data is generated through cyclic loading of specimens at a constant amplitude, and failure is usually defined as when there is a load drop of 25% from steady state stress under strain-controlled conditions (or specimen separation for stress control). The corresponding number of cycles is then used as the definition of fatigue life for a particular strain amplitude. It is known that there are differences between fatigue behaviour in an idealised laboratory setting and in-service components which can contribute to excessive conservatism in plant assessments.

Understanding the Effect of Water Chemistry and Surface Grinding on the Stress Corrosion Cracking Initiation of Filler Metal 82

Product Number: ED22-17143-SG

Author: N.SUPORNPAIBUL, M.G. BURKE, J. DUFF, F. SCENINI, Y.WANG

Publication Date: 2022

$20.00

SCC of Ni-base filler metal (FM) 82 has been reported in the nozzles and other components in Light Water Reactors (LWRs). The typical characteristics of stress corrosion cracking (SCC) of Ni-base alloys are a long incubation time followed by slow propagation, which can suddenly transition to fast propagation. Whilst there has been considerable effort expended to develop an SCC mechanism that can explain and predict SCC in Alloy 600, fewer studies have investigated SCC of FM 82. The Preferential Intergranular Oxidation (PIO) SCC mechanism of Alloy 600 proposed by Bertali et al. which is an evolution of the Selective Internal Oxidation SCC mechanism proposed by Scott and Le Calvar is considered one of the most representative primary water SCC mechanisms for Alloy 600.

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Understanding The Effects Of Zn Injection And OLNC-Treatment On 316 Stainless Steel Oxide Under Simulated BWR Conditions

Product Number: ED22-17227-SG

Author: Dora Capone, Liberato Volpe, Ben F. Spencer, Tsuyoshi Ito, Jonathan Duff, Tsuyoshi Ito, M. Grace Burke, Yoichi Wada, Fabio Scenini, Makoto Nagase

Publication Date: 2022

$20.00

Austenitic stainless-steel alloys are widely used as structural components in light water reactors (LWR) coolant systems, due to their passivity in high temperature water solutions. After initial passivation, subsequent development and dissolution rates of the protective film are very low. Nevertheless, metal cations and colloidal particles that are generated by superficial corrosion of structural materials, can be activated and generate radioactive isotopes that are responsible for radiation source as they circulate through the reactor core. Specifically, the radioactive 60Co, generated by neutron activation of the inactive 59Co (constituent of the naturally occurring Co), releases high-energy γ rays with a half-life of 5.3 years and is the main radiation source in boiling water reactor plants. Mitigating the incorporation of 60Co into stainless-steel oxide depends on understanding the phenomenon of oxide growth and development as a
function of the water chemistry employed, which involves thermodynamic and kinetic considerations.

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Weldability And Corrosion Properties Of Austenitic Stainless Steel And Low Alloy Steel Manufactured Using PM-HIP Method

Product Number: ED22-18347-SG

Author: Gidong Kim, Seunghyun Kim, Kang-myung Seo, Yong-joon Kang, Sang Woo Song

Publication Date: 2022

$20.00

Powder metallurgy with hot isostatic pressing (PM-HIP), which enables to densify powders in a furnace at high pressure and temperature is a newly developed manufacturing process for pressure retaining part in nuclear industry. It has advantages such as weld free near-net shape products, improved material properties for castings, short process time, saving in raw material costs, etc. EPRI have conducted several tests such as tensile test, impact test, chemical composition analysis on the manufactured samples for pressure retaining applications.

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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.

Products tagged with 'environmental degradation 2022'

Association for Materials Protection and Performance