AMPP Store - Products tagged with 'conference papers 2022'

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A Novel Approach To Simulate Realistic Concrete Microstructures Under Irradiation

Product Number: ED22-17280-SG

Author: Amani Cheniour, Yujie Li, Elena Tajuelo Rodriguez, Shane C. Henderson, Yann Le Pape, José David Arregui Mena, Thomas M. Rosseel

Publication Date: 2022

$20.00

The concrete biological shields (CBSs) of light water reactors are affected by neutron and gamma irradiation at high radiation doses, resulting in the degradation of the concrete’s material properties. Several studies in the literature focused on evaluating both the expansion of aggregate-forming minerals and the resulting loss of mechanical properties. Modeling efforts have been carried out to predict the
radiation-induced volumetric expansion (RIVE) and damage using different numerical methods such as the finite element method or fast-Fourier transform (FFT).

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Acceleration Of Long-Range Ordering in NiCrFe Alloys Using Proton Irradiation

Product Number: ED22-18352-SG

Author: Cole Evered, Nicholas Aerne, David J. Sprouster, Julie D. Tucker

Publication Date: 2022

$20.00

The extension of nuclear reactor lifetimes beyond 40 years requires the qualification of plant components to ensure performance past their initial design requirements. Nickel-based alloys containing chromium (NiCr) are of concern at these extended lifetimes, as these types of alloys form an embrittling precipitate phase. Below a critical temperature—which is above the normal 300-400⁰C reactor operating temperatures—NiCr alloys form a stable, fully coherent MoPt2-typelong-range ordered (LRO) phase with stoichiometry Ni2Cr.

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Alloy 182 Models for the Prediction of the Susceptibility to SCC of BMI Nozzle J-welds

Product Number: ED22-17286-SG

Author: Thierry Couvant

Publication Date: 2022

$20.00

Stress Corrosion Cracking (SCC) models are important for engineering and regulatory assessments. The SCC time to the growth of a crack of engineering scale is the main fraction of component life prior to failure and is therefore of significant interest for modeling. However, the stochastic characteristics of early crack development is challenging for model development and validation.

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Application Of Digital Image Correlation (DIC) And Infrared Thermography Techniques To Evaluate Fuel Cladding Behavior During Accident Conditions

Product Number: ED22-18327-SG

Author: Samuel Bell, Ben Garrison, Nathan Capps, Ryan Sweet, Kenneth Kane

Publication Date: 2022

$20.00

Accident scenarios, such as a loss-of-coolant accident (LOCA), subject claddings to rapid thermal transients, internal loading, and a high temperature steam environment. Understanding cladding behavior in this dynamic setting allows for better assessment of safety concerns such as coolant flow blockage and fuel relocation and dispersal. Improvement in model predictability and multi-physics fuel performance codes such as BISON are at the forefront of cladding related research. Particularly, efforts aim at addressing model accuracy to support burnup extension and increases in fuel cycle lengths.

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Assessing Stainless Steel Compatibility in Flowing Fluoride Salts

Product Number: ED22-18337-SG

Author: B. A. Pint, C. G. Parker, Y.-F. Su, D. Sulejmanovic, M. J. Lance, R. Pillai

Publication Date: 2022

$20.00

There is considerable interest in molten halide salts for several applications including thermal storage and next generation nuclear reactors. While molten salt as a working fluid and/or fuel media offers advantages, salt compatibility with structural and functional materials is a concern. Various reports in the literature suggest that chloride and fluoride salts can be highly corrosive to structural alloys but do not always clearly describe how the salt was handled and dried/purified prior to and during the corrosion experiment.

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Assessment Of Neutron Irradiation Damage In Aggregates

Product Number: ED22-17265-SG

Author: Elena Tajuelo Rodriguez, Lawrence M. Anovitz, Michael Cheshire, Yann Le Pape, Jan Ilavsky, Thomas M. Rosseel

Publication Date: 2022

$20.00

The required electrical power in the United States has led the utilities and the US Nuclear Regulatory Commission to evaluate second license renewals for operating light-water reactors, and some extensions have already been reviewed for extended operation to 80 years. As these plants were licensed to operate for 40 years with options for additional 20 year extensions, the extended operation raised questions in terms of materials performance under extreme conditions and extended time. The effects of prolonged irradiation must be understood and evaluated to predict and ensure the reliability of plant components.

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Characterization of 304 Stainless Steel Crack Tips Exhibiting High Temperature Retardation in Deaerated Water

Product Number: ED22-18392-SG

Author: Robert Morris, Tyler Moss, David Morton, Kevin Fisher

Publication Date: 2022

$20.00

Understanding and mitigating stress corrosion cracking (SCC) in stainless steels used in light water reactors is important, and experimental efforts to characterize this behavior have been performed over the last several decades. While SCC growth has been shown to follow an Arrhenius temperature functionality, a departure from this functionality has been observed due to high temperature SCC growth rate retardation (HTR). This paper characterizes observed trends between different cold work levels and temperature effects on cracking behavior and crack tip morphologies in 304 stainless steel.

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Characterization Of Grain Boundary Cavity Evolution During Long-Term Crack Initiation Testing Of Cold Worked Alloy 690 In Simulated PWR Primary Water

Product Number: ED22-17324-SG

Author: Ziqing Zhai, Mychailo Toloczko, Stephen Bruemmer

Publication Date: 2022

$20.00

Determining the resistance of high-Cr Ni-base Alloy 690 to environmental degradation during long-term pressurized water reactor (PWR) exposure is needed to confirm its viability as the replacement material for Alloy 600 and help establish a quantitative factor of improvement for stress corrosion crack (SCC) initiation. SCC initiation testing on cold-worked (CW) Alloy 600 materials in PWR primary water has demonstrated that intergranular (IG) attack is the precursor to SCC initiation in this material. In comparison, an equivalent degradation and cracking process does not exist in CW Alloy 690.

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Characterization Of Materials Showing Large Differences In IASCC Response

Product Number: ED22-18360-SG

Author: Anders Jenssen, Daniel Jädernäs, Peter Chou, Kristina Lindgren, Mattias Thuvander

Publication Date: 2022

$20.00

Irradiation-assisted stress corrosion cracking (IASCC) is a term used to describe cracking of austenitic materials in components subjected to a relatively high fast neutron flux. Like intergranular stress corrosion Paper No.18360 cracking (IGSCC), IASCC appears as intergranular cracks, but thermal sensitization of the grain boundaries is not required for the material to become susceptible to cracking. Service failures caused by IASCC have occurred in components such as core shrouds and top guides in boiling water reactors (BWRs) and baffle bolts in pressurized water reactors (PWRs).

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Chemistry Change Of CRUD With Various Metal Ion Concentration Conditions In PWR Fuel Cladding

Product Number: ED22-17313-SG

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

Publication Date: 2022

$20.00

Since 1960s, CRUD induced problems become an issue in PWR power plant. To enhance economic efficiency of PWR, the burn-ups of nuclear fuel have been increased. In the reactor core, some unexpected power shifts, which shows negative axial offset from prediction, have been observed. This phenomenon is called by AOA (Axial offset Anomaly) and it seems to be caused by deposits on the upper side of fuel cladding surface which is called CRUD.

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Cold Spray for Repair of Nuclear Reactor System Components

Product Number: ED22-17267-SG

Author: Catherine Cmar, Michael Ickes, Arash Parsi

Publication Date: 2022

$20.00

Supersonic particle deposition also known as cold dynamic gas spray, or “Cold Spray” is a materials deposition process. During the cold spray process a gas stream, typically helium or nitrogen, is split into two streams where 90% of the gas is sent to an electric heater and 10% is sent to a powder feeder. The powder feeder contains powder composed of small metallic particles, or blends of metallic and non-metallic particles, ranging from 5 to 100 μm in size.

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Comparing Stress Corrosion Cracking Initiation Times of Sensitized 304 Stainless Steel in KOH And LiOH Primary Water Chemistries with High Levels Of Dissolved Oxygen

Product Number: ED22-18341-SG

Author: Stuart Medway, Peter Chou, David Tice

Publication Date: 2022

$20.00

The testing described in this paper is part of a wider initiative by the Electric Power Research Institute (EPRI) to perform a due-diligence assessment to support possible application to plants and demonstrate the use of potassium hydroxide (KOH) in western Pressurized Water Reactors (PWRs). Lithium hydroxide (LiOH) is used in the primary coolant loop of PWRs to modify the pH of the coolant water. LiOH is most commonly used as the alkalizing substance as 7Li is already present in the water
chemistry as a by-product of the neutron reaction with boron (10B). To reduce the risk of accelerated corrosion of the Zircaloy fuel cladding material, there is an upper limit of 3.5 ppm for 7Li, although standard starting chemistry is typically 2 ppm. The amount of lithium is subsequently reduced during operation as the required level of boron is reduced due to fuel burnup through the fuel lifecycle.

Products tagged with 'conference papers 2022'

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