AMPP Store - Products tagged with 'environmental degradation 2022'

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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.

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SCC Behavior Of Alloy 52M Weld Butter On SA-508 LAS From The Double-Repair Dissimilar Metal Weld In A PWR Environment

Product Number: ED22-17283-SG

Author: Bogdan Alexandreanu, Yiren Chen

Publication Date: 2022

$20.00

This paper presents the results of a confirmatory research program conducted with the purpose of evaluating the susceptibility of Nickel-base Alloy 690 and 52/152 and variant welds to stress corrosion cracking (SCC) with a focus on the SCC response of Alloy 52/152 weldments with repairs.Nickel-base 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, repairs of existing components as well as for designs of advanced nuclear reactors because of their increased resistance to stress corrosion cracking (SCC) relative to Alloys 600, 82, and 182.

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SCC Growth In BWR-NWC Of A Low-Alloy Steel With Intermediate S-Content Tested At K Between 50 And 70 Mpa√M

Product Number: ED22-17305-SG

Author: Martin Bjurman, Elena Calota, Björn Forssgren, Johan Stjärnsäter

Publication Date: 2022

$20.00

Stress corrosion cracking (SCC) of RPV steels has shown fairly quick initiation and high crack growth rates (CGRs) in simulated normal water chemistry (NWC) autoclave tests. Still the operating experience shows no known cases that reflect this high sensitivity. The bulk of these tests have been conducted on either high sulfur material, with significant dynamic loading and/or in high sulphate or chloride environments. Recent studies at PSI and GE have shown increased CGRs at 3-5 ppb chloride. This led to the limit for normal operating conditions in the EPRI BWR water Chemistry Guidelines [3,4] to be reduced from 5 to 3 ppb of chloride during the course of this project. The effects on the in-crack chemistry of test specimens vs. those of real cracks, and the effect of cladding on cracking in LAS have been debated.

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SCC Initiation and Growth Rate Response of A-286 in B/Li and B/K Solutions 340°C

Product Number: ED22-17262-SG

Author: P.L. Andresen, P. Chou

Publication Date: 2022

$20.00

Stress corrosion cracking (SCC) initiation and growth rate testing was performed using Alloy A-286, a high-strength iron-base alloy, to evaluate for possible differences in response in boric acid solutions containing KOH vs. LiOH. PWRs are considering a switch from LiOH to KOH because of the uncertain future availability and high cost of Li. To achieve the same pH at temperature (pHT) in typical PWR primary water, the same molar concentration of Li and K is needed. The atomic weight of K is 39.1, which is 5.63 higher than the atomic weight of Li at 6.94, so 1 ppm Li yields the same pHT as 5.63 ppm K. The conductivity differs somewhat because of difference in the mobility of Li+ and K+.

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SCC of Alloy 152/52 Welds Defects, Repairs, Dilution Zones and HAZs in PWR Water

Product Number: ED22-17243-SG

Author: Peter L. Andresen, Chris Wax

Publication Date: 2022

$20.00

Stress corrosion cracking (SCC) has been observed for over six decades in light water reactors structural components, with wide variations in the rate of SCC initiation and crack growth. Newer materials have been adopted in the last three decades, primarily the ~30% Cr Alloy 690 (UNS N06690) and its weld metals, Alloy 52 (UNS W86052) and Alloy 152 (UNS W86152). These materials were initially viewed as immune to SCC, but are now recognized to be susceptibility to SCC, and can exhibit high growth rates in some conditions.

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SCC Of Alloy 82 Weld Metal In BWR Water

Product Number: ED22-17242-SG

Author: Peter L. Andresen, Katsuhiko Kumagai

Publication Date: 2022

$20.00

There is extensive evidence from laboratory data and plant experience of the SCC susceptibility of Alloy 82 weld metal in both BWR and PWR environments. Two international expert panels evaluated laboratory data under PWR conditions and created disposition curves to address the effects of stress intensity factor (K), temperature, and other factors. Another expert panel is creating a disposition
curve under BWR conditions for K, temperature, corrosion potential, impurities, and other factors. Nickel alloy weld metals at lower Cr levels (~15% Cr for Alloy 182 and ~20% Cr for Alloy 82) are more susceptible to SCC than weld metals of higher Cr content (~30% Cr for Alloy 52/152). This paper focuses on on-thefly effects on SCC growth rate of Alloy 82 weld metal in BWR environments of corrosion potential, water
purity and temperature.

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Stress Corrosion Cracking Behaviour Of Thermally Aged Alloy 182

Product Number: ED22-17291-SG

Author: Siddharth Suman, Hans-Peter Seifert, Philippe Spätig, Stefan Ritter

Publication Date: 2022

$20.00

Alloy 182 is an austenitic (FCC) nickel base Ni-Cr-Fe-Mn weld metal that is used as a weld filler or weld pad metal to join stainless steel reactor internals, reactor instrumentation penetrations and main coolant piping to the low-alloy steel reactor pressure vessel. Stress corrosion cracking (SCC) in alloy 182 dissimilar welds is one of the most important material degradation problems and an ongoing issue in boiling (BWR) and pressurized water reactors (PWR) world-wide with potential safety concerns.
Thermally activated preferential local ordering (unlike atom pairs is greater than that in a random solution) of elements within a lattice, over spatial dimensions that are typically on the order of a few nearest neighbor spacing 20 to 50 Å (2 to 5 nm) is referred as short range ordering (SRO). SRO causes lattice contraction and induces additional stress which is claimed as the driving force for SCC in the alloy 600, alloy 690, and alloy 182.

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Stress Corrosion Cracking Testing Of Laboratory Processed 304/304L Stainless Steel Elbows And Pipe

Product Number: ED22-18591-SG

Author: Denise J Paraventi, Justin R Hesterberg, Tyler E Moss, and Stephen C Healey

Publication Date: 2022

$20.00

Stress corrosion cracking (SCC) of austenitic stainless steels, while not as prevalent as that in nickelbased alloys such as Alloy 600 and Alloys 82/182, has been observed in the primary system of commercial pressurized water reactors. These instances of SCC have been associated with water chemistry issues and/or occluded regions; however, in many cases high levels of cold work were also present in the material as well.

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Study On Hold Time Effects On Standard Fatigue Specimens And Component Relevant Loading Conditions At Welds Using Bending Specimens From Welded Pipes

Product Number: ED22-18531-SG

Author: Michael Grimm, Matthias Herbst, Armin Roth, Jürgen Rudolph

Publication Date: 2022

$20.00

This paper summarizes the results of a research project on environmental effects on the environmentally assisted fatigue lifetime of laboratory specimens made of austenitic stainless steel type 316L and corresponding welds. In particular, investigations on the effect of hold-times, applied during testing in high-temperature water and tests on the fatigue welds were performed.

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Tests On Chloride-Induced Stress Corrosion Cracking In Austenitic Stainless Steel Welds For Dry Storage Canister Application

Product Number: ED22-17131-SG

Author: Seunghyun Kim, Dong-Jin Oh, Chang-Young Oh, Gidong Kim, and Sang Woo Song

Publication Date: 2022

$20.00

Chloride-induced stress corrosion cracking (CISCC) is a degradation phenomenon hindering structural integrity of a dry storage canister for interim storage of spent nuclear fuel. Owing to materials susceptibility, residual stress and corrosive environments, pitting corrosion and evolution of CISCC occur. Previous workers on CISCC have figured out that austenitic stainless steels is susceptible to CISCC due to its microstructural characteristics. In chloride-containing media, pits are formed at the surface of austenitic stainless steels and theses pits play a role as CISCC initiation sites. However, due to its complexity, fundamental mechanism of CISCC at various temperatures and relative humidity (RH) values is still in-debate.

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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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The Effect Of Potassium Hydroxide Primary Water Chemistry On The IASCC Behavior Of 304 Stainless Steel

Product Number: ED22-17415-SG

Author: Michael Ickes, Peter Chou, Kai Chen, Gary Was

Publication Date: 2022

$20.00

The objective of this work was to compare irradiation-assisted stress corrosion cracking (IASCC) growth behavior in simulated pressurized water reactor (PWR) water with pH maintained with LiOH versus KOH. The U.S. nuclear industry is considering changing PWR primary water chemistries to use KOH in place of LiOH, as a means to ensure a stable supply chain and secure cost savings. This experiment will specifically investigate the impact of these alkali ions on the crack growth rate (CGR) and to examine the crack morphologies generated by the CGR experiment.

Products tagged with 'environmental degradation 2022'

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