Rod internal pressure quantification and distribution analysis using Frapcon

Bratton, Ryan N.; Jessee, Matthew Anderson; Wieselquist, William

doi:10.2172/1235837

Cited by 10 publications

(9 citation statements)

References 6 publications

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“…The calculations documented in Ref. 21 were consistent with the database documented in Ref. 20 for standard PWR fuel rods.…”

Section: Relevant Ranges Of Internal Pressures and Hoop Stressessupporting

confidence: 78%

“…Characterization of the sister rods will result in up to 25 valuable data points for EOL RIP for standard PWR fuel rods. In addition, Oak Ridge National Laboratory used FRAPCON to predict gas pressures for over 68,000 fuel rods irradiated during Cycles 1−12 in the Watts Bar Nuclear Unit 1 reactor [21]. The calculations documented in Ref.…”

Section: Relevant Ranges Of Internal Pressures and Hoop Stressesmentioning

confidence: 99%

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Effects of Radial Hydrides on PWR Cladding Ductility following Drying and Storage

Billone

Burtseva

García-Infanta

2017

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The purpose of this research effort is to determine the effects of canister and/or cask drying, transfer and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperature (PCT) values and hoop stresses. Extensive radial-hydride precipitation could lower failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods, at temperatures below the ductility transition temperature (DTT), which has been referred to as the ductile-to-brittle transition temperature (DBTT) in previous reports. During post-storage transport, hoop stresses will be induced by "pinch-type" loading at grid spacers. Such loading is considered to be mild for normal conditions of transport and more severe for hypothetical accident conditions.

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“…The calculations documented in Ref. 21 were consistent with the database documented in Ref. 20 for standard PWR fuel rods.…”

Section: Relevant Ranges Of Internal Pressures and Hoop Stressessupporting

confidence: 78%

Section: Relevant Ranges Of Internal Pressures and Hoop Stressesmentioning

confidence: 99%

Effects of Radial Hydrides on PWR Cladding Ductility following Drying and Storage

Billone

Burtseva

García-Infanta

2017

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“…Pin internal pressurisation was neglected as this was considered a secondary effect. This was because fission gas release was assumed to be low during early fuel cycles not exposed to accident conditions; Bratton et al have shown that the rod internal pressure reaches no more than 6-7 MPa by end of life [49]. Given that the mechanical response is of concern here, a simple linear temperature profile was applied as a boundary condition to the nodes.…”

Section: Test Geometry Loads and Boundary Conditionsmentioning

confidence: 99%

Preliminary modelling of crack nucleation and propagation in SiC/SiC accident-tolerant fuel during routine operational transients using peridynamics

Haynes

Shepherd

Wenman

2020

Journal of Nuclear Materials

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Silicon carbide fibre in silicon carbide matrix composites (SiC/SiC) are a promising cladding for use in accident tolerant fuels (ATF) in current light water reactor (LWR) designs. However, as they are a radically different material from current metal clads, current thermomechanical simulation methods struggle to accurately predict their behaviour, especially regarding the potential development of cracks. Thus, a new peridynamic model for SiC/SiC cladding has been developed in the Abaqus finite element code. The material model was isotropic and considers matrix cracking and fibre pull-out. The thermal expansion, swelling and the degradation of the thermal conductivity are modelled under typical LWR irradiation conditions. The swelling on the outer surface is predicted to be greater than the inner surface due to the lower irradiation temperature, causing a tensile stress on the inside of the cladding; tension being more challenging for a ceramic than a metal. This stress increases during the decrease in power at the start of a typical pressurised water reactor refuelling outage and causes microcracking of the matrix on the cladding inner surface. In models without fibres, cracks would propagate through the cladding. If fibres are modelled, matrix cracking will extend to a depth of around 20% through the cladding from the inner surface, which is unlikely to be an acceptable design. If an inner monolith of SiC is additionally modelled, cracking propagates through the monolith and acts as a stress raiser for matrix cracking in the composite, and therefore does not constitute a design improvement. If an outer SiC monolith is modelled, fibre pull-out strain on the inner surface of the cladding was increased by just under 70%. No cracks are predicted in an outer monolith which may therefore remain gas-tight and thus a more suitable design. These predictions are consistent with experimental findings. Highlights• Irradiation of accident tolerant SiC/SiC cladding is modelled using peridynamics.• Due to swelling, a tensile stress is predicted on the inside of the cladding. • In the absence of a monolith, micro-cracking extends to 20% from the inner surface.• Cracks in an inner monolith are a stress raiser in the composite.• No cracks are predicted in an outer monolith, hence it's the most promising design.

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“…With two additional stretch goals: 1) compare results to FRAPCON results obtained by Ryan Bratton [16], a student at Pennsylvania State University working remotely from ORNL, who obtained these results as part of his thesis work; and 2) perform a mesh sensitivity study to assess appropriate refinement level.…”

Section: Milestone Objectivesmentioning

confidence: 99%

Standalone BISON Fuel Performance Results for Watts Bar Unit 1, Cycles 1-3

Clarno

Pawlowski

Stimpson

et al. 2016

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The Consortium for Advanced Simulation of Light Water Reactors (CASL) is moving forward with more complex multiphysics simulations and increased focus on incorporating fuel performance analysis methods. The coupled neutronics/thermal-hydraulics capabilities within the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) have become relatively stable, and major advances have been made in analysis efforts, including the simulation of twelve cycles of Watts Bar Nuclear Unit 1 (WBN1) operation. While this is a major achievement, the VERA-CS approaches for treating fuel pin heat transfer have well-known limitations that could be eliminated through better integration with the BISON fuel performance code. Several approaches are being implemented to consider fuel performance, including a more direct multiway coupling with Tiamat, as well as a more loosely coupled one-way approach with standalone BISON cases. Fuel performance typically undergoes an independent analysis using a standalone fuel performance code with manually specified input defined from an independent core simulator solution or set of assumptions. This milestone covers efforts to facilitate the use of VERA for core simulation and fuel performance to operate in this standalone mode of execution by using power distribution and moderator temperature data from VERA-CS. This approach can be used for several CASL fuel performance applications: 1) as a screening tool for the pellet-clad interaction (PCI) challenge problem to identify rods that require further analysis, 2) to analyze reactivity insertion accidents (RIAs) and departures from nucleate boiling (DNBs) which require time-dependent full core results to establish the boundary conditions for the high-fidelity BISON simulations, and 3) to improve fuel temperature models in VERA-CS by providing higher fidelity comparisons of core-wise temperature distributions. 4) as a high fidelity fuel performance methodology (using explicit rather than bounding core operating conditions) to evaluate and screen fuel and core design concepts for a comprehensive set of design criteria including pin powers, PCI, pin burnups, corrosion, oxidation, margin to fuel melt, and rod internal pressure. This report summarizes the improvements made since the initial milestone to execute BISON from VERA-CS output. Many of these improvements were prompted through tighter collaboration with the BISON development team at Idaho National Laboratory (INL). A brief description of WBN1 and some of the VERA-CS data used to simulate it are presented. Data from a small mesh sensitivity study are shown, which helps justify the mesh parameters used in this work. The multi-cycle results are presented, followed by the results for the first three cycles of WBN1 operation, particularly the parameters of interest to PCI screening (fuel-clad gap closure, maximum centerline fuel temperature, maximum/minimum clad hoop stress, and cumulative damage index). Once the mechanics of this capability are functioning, future work will target cycles w...

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Rod internal pressure quantification and distribution analysis using Frapcon

Cited by 10 publications

References 6 publications

Effects of Radial Hydrides on PWR Cladding Ductility following Drying and Storage

Effects of Radial Hydrides on PWR Cladding Ductility following Drying and Storage

Preliminary modelling of crack nucleation and propagation in SiC/SiC accident-tolerant fuel during routine operational transients using peridynamics

Standalone BISON Fuel Performance Results for Watts Bar Unit 1, Cycles 1-3

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