Advancements in the behavioral modeling of fuel elements and related structures

Billone, M.C.; Montgomery, Robert; Rashid, Y.R.; Head, J.L.

doi:10.1016/0029-5493(92)90005-g

Cited by 8 publications

(5 citation statements)

References 8 publications

(8 reference statements)

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“…Common biaxial flexural strength tests include ring-on-ring, piston-on-3-balls, or the ball-on-ring, all of which work to reduce friction during testing [11]. Several studies [11][12][13] have used Equation ( 1) to calculate the transverse stress (σ) and obtain the TRS of similar ceramics:…”

Section: Biaxial Flexure Testmentioning

confidence: 99%

Fracture Strength Determination Methods for Ceramic Materials Applied to Uranium Dioxide

Nelson¹,

Silva

Lupercio

et al. 2020

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Section: Biaxial Flexure Testmentioning

confidence: 99%

Fracture Strength Determination Methods for Ceramic Materials Applied to Uranium Dioxide

Nelson¹,

Silva

Lupercio

et al. 2020

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“…Because of the central role of the fuel in the nuclear steam supply system (NSSS), fuel elements of various types and levels of complexity are incorporated in the NSSS codes. [67] Those codes, arranged in ascending order of complexity of the fuel elements models contained in them, include system thermal hydraulics codes, core simulation codes, loss of coolant accident codes, and fuel performance codes. This shows the important role of a fuel behavior code within an advanced simulation initiative as the one considered here, which aims at simulating all aspects of this NSSS.…”

Section: Via Backgroundmentioning

confidence: 99%

“…Due to the extreme complexity of solving that system of equations in a complete form, different levels of approximations have to be made to enable the modeling effort. Those approximations range from simple 1-D thermal analysis models to a full thermo-mechanical 2-D finite elements representations [67].…”

Section: -mentioning

confidence: 99%

“…Large numbers of computer codes that simulate fuel behavior are available. Generally, those codes can be divided into three major categories which include simple correlation fuel performance codes, fuel model codes, and mechanistic fuel performance codes [67,72]. The latest category is of interest here, since the former categories provide simplistic representation of the fuel behavior that are limited to certain applications and cannot be used to extrapolate the fuel behavior.…”

Section: Vic Current Tools and Approachmentioning

confidence: 99%

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Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.

Palmiotti¹,

Cahalan²,

Pfeiffer³

et al. 2006

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This report presents requirements for advanced simulation of nuclear reactor and chemical processing plants that are of interest to the Global Nuclear Energy Partnership (GNEP) initiative. Justification for advanced simulation and some examples of grand challenges that will benefit from it are provided. An integrated software tool that has its main components, whenever possible based on first principles, is proposed as possible future approach for dealing with the complex problems linked to the simulation of nuclear reactor and chemical processing plants. The main benefits that are associated with a better integrated simulation have been identified as: a reduction of design margins, a decrease of the number of experiments in support of the design process, a shortening of the developmental design cycle, and a better understanding of the physical phenomena and the related underlying fundamental processes. For each component of the proposed integrated software tool, background information, functional requirements, current tools and approach, and proposed future approaches have been provided. Whenever possible, current uncertainties have been quoted and existing limitations have been presented. Desired target accuracies with associated benefits to the different aspects of the nuclear reactor and chemical processing plants were also given. In many cases the possible gains associated with a better simulation have been identified, quantified, and translated into economical benefits. Results reported in the AFCI series of technical memoranda frequently are preliminary in nature and subject to revision. Consequently, they should not be quoted or referenced without the author's permission

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“…The processes that control the behavior of fuel and cladding materials during irradiation consist of a complex interaction between pre-existing material microstructural features and the changes caused by the imposed thermal, mechanical, chemical, and irradiation conditions. Traditional fuel behavior modeling capabilities are built on a suite of empirical or semi-empirical material property or behavior models [3]. In most cases, the empirical models consider that the mechanisms of irradiation creep, irradiation growth, and thermal creep are uncoupled and contribute independently to the overall response of the cladding.…”

Section: Introductionmentioning

confidence: 99%

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

Montgomery

Tomé

Liu³

et al. 2017

Journal of Computational Physics

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Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical constitutive models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. To improve upon this approach, a microstructurally-based zirconium alloy mechanical deformation analysis capability is being developed within the United States Department of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL). Specifically, the viscoplastic self-consistent (VPSC) polycrystal plasticity modeling approach, developed by Lebensohn and Tomé [1], has been coupled with the BISON engineering scale fuel performance code to represent the mechanistic material processes controlling the deformation behavior of light water reactor (LWR) cladding. A critical component of VPSC is the representation of the crystallographic nature (defect and dislocation movement) and orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. A future goal is for VPSC to obtain information on reaction rate kinetics from atomistic calculations to inform the defect and dislocation behavior models described in VPSC. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON and provides initial results utilizing the coupled functionality.

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Advancements in the behavioral modeling of fuel elements and related structures

Cited by 8 publications

References 8 publications

Fracture Strength Determination Methods for Ceramic Materials Applied to Uranium Dioxide

Fracture Strength Determination Methods for Ceramic Materials Applied to Uranium Dioxide

Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

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