AMPP Store - Products tagged with 'environmental degradation 2022'

Available for download

Effect Of Water Chemistry On Crack Growth Rates In Neutron Irradiated X-750 And XM-19

Product Number: ED22-17287-SG

Author: John Jackson, Drew Johnson, Robert G. Carter, Michael Heighes, Peter L. Andresen

Publication Date: 2022

$20.00

As the current reactor fleet continues to age, with many reactors wanting to extend operational licenses beyond their initial 40 lifetime, it is becoming increasingly important to understand how structural materials in these reactor environments will degrade over time. A critical material degradation mode which can limit the lifetime of reactor components is irradiation assisted stress corrosion cracking (IASCC). As the name implies, a material must be subjected to both a corrosive environment as well as mechanical stresses while the original material microstructure has been affected by irradiation.

Available for download

Elucidating the Effect of Build Location and Powder Morphology on the Corrosion of Selective Laser Sintered Additively Manufactured 316L Stainless Steel

Product Number: ED22-17456-SG

Author: E.Getto, E.Farnan, M.G. Koul, R.J. Santucci, T.Gray, J. S. Gibbs, J.McLaughlin

Publication Date: 2022

$20.00

Additive manufacturing (AM) has allowed for the rapid prototyping of parts and the ability to control the structure of a material and, to a lesser extent, the microstructure. AM as applied to metals, as stainless steel is especially promising as it provides the ability to produce complex shaped components rapidly. Laser-powder bed fusion (L-PBF) is one such technique in which the part is formed from the base up by fusing successive layers of powder. As each layer is fused, the print plate moves down and a new layer of metallic powder is swept over the top. A laser then welds this layer to the top of the existing piece and the process repeats.1 However, AM can result in non-equilibrium microstructures, porosity, and residual stresses that could affect the longevity of the material and performance in corrosive conditions.

Available for download

Environmental Degradation Of Materials Considered For Accident Tolerant Fuels In Light Water Reactors.

Product Number: ED22-17252-SG

Author: Michael D. Worku, Russ M. Fawcett, Rajnikant V. Umretiya, Dan R. Lutz, AndrewK. Hoffman, Evan J. Dolley, Liang Yin, Liang Yin, Raul B. Rebak

Publication Date: 2022

$20.00

In recent years and months some countries started to label nuclear energy as clean energy because it does not increase the carbon footprint in the planet. Figure 1 shows that nuclear energy is the largest contribution of clean (or green) energy in the USA. This clean energy comes from the commercial operation of boiling water reactors (BWR) and pressurized water reactors (PWR). The total number of nuclear power reactors in the USA is slowly decreasing in time because they became non-economical to operate compared to the burning of natural gas. The International Atomic Energy Agency (IAEA) reported that in 2013 there were 102 light water reactors (LWR) producing electricity in the USA, but in 2020 the total LWR number decreased to 94 due to the decommissioning of eight reactors.

Available for download

EPRI’S Approach To Longer Term Operations

Product Number: ED22-17191-SG

Author: Emma L. Wong, Andrew Mantey, Samuel Johnson, Jean Smith

Publication Date: 2022

$20.00

EPRI has been supporting the nuclear industry over the last several decades to provide the technical bases and research to support the operation of the current fleet of nuclear power plants beyond their initial licensing period (typically 30-40 years of operation). Hundreds of technical reports and guidance documents have been issued on topics ranging from developing and implementing aging management programs, identification, and evaluation of degradation mechanisms, and remaining useful life of key passive components (e.g., reactor vessel internals, cables, and concrete). A previous 2019 ANS Environmental Degradation Conference paper discussed the research goals and results of EPRI up to 2019 for concrete and cables. These research results provide a living technical basis as these results are supplemented regularly with industry operating experience, inspection results, and condition monitoring or non-destructive evaluations.

Available for download

Evaluating Irradiation Damage In CANDU Components By Characterizing Surrogate Material Harvested From The Decommissioned NRU Reactor At Chalk River Laboratories- Application To The CANDU Calandria Tubesheet Degradation

Product Number: ED22-17290-SG

Author: Mitchell Mattuchi, Qiang Wang, Brian Langelier, Nicolas Huin, Brian Connolly, Grace Burke

Publication Date: 2022

$20.00

The CANDU® reactor is a pressurized heavy water moderated (D2O) reactor, fueled with natural uranium. Nuclear fission of natural uranium occurs in the fuel channels and the heat generated from the fission is transported from the fuel channel to the steam generator, via the primary heat transport system. The nuclear reaction is moderated by heavy water that surrounds the fuel channels contained within the Calandria Vessel.

Available for download

Evaluating the Use of Phosphate to Mitigate Caustic-PbSCC of Alloy 690TT

Product Number: ED22-17293-SG

Author: Giulia B. Mazzei, Jonathan Duff,D.A. Horner, M.Grace Burke, Fabio Scenini

Publication Date: 2022

$20.00

PbSCC of Ni-base alloys is active over a wide range of environmental conditions but for the higher Cr content Alloys 800 and 690 only under abnormal crevice environment of high or low pH that can occur in the secondary side of Pressurized Water Nuclear Reactors (PWRs). Several experimental campaigns have aimed at understanding this phenomenon, concluding that PbSCC can develop in both acidic and caustic solutions, for low and high concentration of Pb, across a wide electrochemical potential range and in presence of chlorides contaminants.

Available for download

Evaluation Of Copper Catalytic Effects In Cable Insulation Polymers

Product Number: ED22-17260-SG

Author: Trevor Toll, Casey Sexton, Codi Ferree, Patrick Ward

Publication Date: 2022

$20.00

This paper summarizes work performed to evaluate a phenomenon that can occur in electrical cable insulation polymers during the aging process. This phenomenon, the copper catalytic effect, occurs because of diffusion of copper ions from the conductor into the insulation polymers during the aging process. In this research, the copper catalytic effects observed in cross-linked polyethylene, cross-linked polyolefin, and ethylene propylene rubber insulation subjected to thermal accelerated aging at both 120˚C and 130 ˚C were evaluated. In addition, the insulation polymers from cables removed from service in operating nuclear power plants were also evaluated to determine if this effect is prevalent for naturally aged materials. The results acquired from this work were used to characterize the copper catalytic effects observed in these polymers, analyze how this phenomenon affects the degradation process of the materials, and determine the impact that the copper catalytic effect has on condition monitoring data acquired during the aging process.

Available for download

Evaluation Of Specific Energy Absorption Rate In Aged EPR-Insulated Cables Using Finite Element Analysis

Product Number: ED22-17312-SG

Author: Sean O’Brien, Brian Hinderliter, Margaret Elmer-Dixon, Melissa Maurer-Jones, Robert Duckworth

Publication Date: 2022

$20.00

Medium voltage (MV) cables, which typically operate in the range of 2 kV to 35 kV, are commonly used in nuclear power plants (NPPs) throughout the world. These cables support the safety and wellbeing of NPPs by providing supplementary power for safety systems to continue operating during emergency events such as natural disasters or human-induced outages. This allows for uninterrupted reactor operations for a short period of time until the primary safety systems can be brought back online. Given their critical importance to the operation of NPPs, MV cables are often installed in locations such as underground concrete ducts or electrical conduits that limit cable exposure to environmental stressors such as moisture and temperature. Despite the fact that these cables are not operating continuously given the overall rarity of NPP emergency events, they must still satisfy reliability and lifetime performance requirements of cables used in primary NPP operations.

Available for download

Evaluation On Applicability Of Nickel-Base Alloy Weld Materials With High Chromium Content In BWR Environment

Product Number: ED22-17130-SG

Author: Takahiro Hayashi, Yoshinori Katayama, Daiki Tanaka, Yumiko Tsuchiya, and Hidenori Takahashi

Publication Date: 2022

$20.00

Ni-base alloy weld material has been widely used for primary reactor components of BWR. Stress corrosion cracking (SCC) in Ni-base alloy welds is of an increased importance and an ongoing subject in the industry to secure material reliability of the components especially for long-term operation of light water reactors. Although alloy 82 has shown excellent service performance in BWR applications, it is known that alloy 82 exhibit SCC susceptibility in laboratory tests under simulated BWR environment with a combination of particular, severe test conditions such as high level of material cold work and highly accelerated environment. In addition, few experiences with SCCs in the welds associated with alloy 82 have been recently reported in the operating BWR plants.

Available for download

Examination Of IASCC In 304 Stainless Steel Core Shroud From A Commercial Boiling Water Reactor (BWR)

Product Number: ED22-18366-SG

Author: Matthew Olszta, Karen Kruska, Danny Edwards, Daniel Schreiber, Peter Chou

Publication Date: 2022

$20.00

Nuclear reactors inherently operate under extreme environments, and hence the materials and alloys utilized for their design are required to withstand unique conditions. Not only do these materials need to stand up to corrosion at high temperatures, but also in certain components must they resist microstructural and physical property changes due to radiation. Two of the major effects of radiation on reactor alloys and radiation-induced segregation (RIS) and precipitation, which have been observed in reactor pressure vessel (RPV) steels, ferritic-martensitic steels, nickel base alloys, and austenitic stainless steels.

Available for download

Experimental Characterization of Grain Boundary Fracture Properties in a FeCr12Ni26Si3 Austenitic Stainless Steel Oxidized in PWR Environment

Product Number: ED22-17126-SG

Author: Rachma Azihari, Jérémy Hure, Marc Legros, Benoît Tanguy

Publication Date: 2022

$20.00

Austenitic stainless steels are used for the core internal structures (bolts, baffles, formers) in Pressurized Water Reactors (PWR). During operational service, baffle to former bolts have been observed to undergo Irradiation-Assisted Stress Corrosion Cracking (IASCC), which is characterized by intergranular cracking. IASCC results from the material corrosion susceptibility, the microstructural changes induced by irradiation, the corrosive media and the mechanical loading. Numerous studies have been conducted to evaluate the complex interplay between the different factors, mostly focusing on InterGranular Stress Corrosion Cracking (IGSCC) of pre-irradiated samples in PWR environment. In particular, the oxidation behavior of grain boundaries and the mechanical loading of grain boundaries have been assessed in details. Depending on the oxidation time and the GB nature, oxide penetration along GB has been observed. The intergranular oxide is composed of (Nix,Fe1-x)Cr2O4 spinels. However, all grain boundaries (GBs) do not have the same oxidation behavior, and it has been reported that high angle grain boundaries show higher oxidation susceptibility than special grain boundaries. Radiation induced segregation at grain boundaries might also lead to higher susceptibility to intergranular oxidation. Irradiation also modifies the deformation mechanisms in austenitic steels resulting in strain localization which is believed to be an important factor in IASCC initiation as it can lead to local increase of the stress due to dislocation pile-ups at GB.

Available for download

Fatigue Crack Initiation In Zr-2.5Nb Tested In Heavy-Water At 300°C

Product Number: ED22-17233-SG

Author: H.M. Nordin, A.J. Phillion, Q. Wang, M. Lenzi, H.Yu, T.M. Karlsen

Publication Date: 2022

$20.00

Pressure tubes, manufactured from Zr-2.5Nb, are used within a CANDU power reactor to contain the fuel bundles and coolant forming the primary pressure boundary in the core with an expected operating lifetime of 25 years. As part of the Canadian Standards Association (CSA) fitness-for-service requirements for pressure tubes, flaws or stress risers that potentially lead to cracks cannot exist in pressure tubes in operating reactors. Flaws may include: fuel bundle scratches, crevice corrosion marks, fuel bundle bearing pad fretting flaws and debris fretting flaws.

Products tagged with 'environmental degradation 2022'

Association for Materials Protection and Performance