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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Reactor Pressure Vessel Integrity Assessments with the Grizzly Aging Simulation Code

Grizzly is a simulation tool being developed at Idaho National Laboratory (INL) as part of the US Department of Energy’s Light Water Reactor Sustainability program to provide improved safety assessments of systems, components, and structures in nuclear power plants subjected to age-related degradation. Its goal is to provide an improved scientific basis for decisions surrounding license renewal, which would permit operation of commercial nuclear power plants beyond 60 years. Grizzly is based on INL’s MOOSE framework, which enables multiphysics simulations in a parallel computing environment. It will address a wide variety of aging issues in nuclear power plant systems, components, and structures, modelling both the aging processes and the ability of age-degraded components to perform safely. The reactor pressure vessel (RPV) was chosen as the initial application for Grizzly. Grizzly solves tightly coupled equations of heat conduction and solid mechanics to simulate the global response of the RPV to accident conditions, and uses submodels to represent regions with pre-existing flaws. Domain integrals are used to calculate stress intensity factors on those flaws. A physically based empirical model is used to evaluate material embrittlement, and is used to evaluate whether crack growth would occur. Grizzly can represent the RPV in 2D or 3D, allowing it to evaluate effects that require higher dimensionality models to capture. Work is underway to use lower length scale models of material evolution to inform engineering models of embrittlement. This paper demonstrates an application of Grizzly to RPV failure assessment, and summarizes on-going work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Effect of Clad Fast Neutron Flux Distribution on Quarter-Core Fuel Performance Calculations with BISON

As part of the Consortium for Advanced Simulation of Light Water Reactors (CASL), the Virtual Environment for Reactor Applications (VERA) is being developed to provide high-fidelity multiphysics simulations of nuclear reactor cores. Over the past year, a one-way coupling capability has been developed, taking power and temperature data from MPACT/CTF calculations and running standalone BISON fuel rod cases. This has allowed CASL contributors to gain insight into the fuel performance characteristics while a more tightly coupled methodology between MPACT, CTF, and BISON has been under development. Many of the initial analyses have focused on the Watts Bar Unit 1 core using a fast flux factor—the default approach—which defines the clad fast flux as a linear function of the local linear heat rate. While this approach is more valid when considering single rod cases in isolation, in larger problems such as full-sized reactors, the local linear heat rate does not directly correlate to the clad fast neutron flux. For example, variations in assembly enrichment can lead to larger neutron fluxes in neighboring assemblies that may be at lower enrichment and power. Similarly, considering axial effects, spacer grids suppress power locally, but the fast neutron flux is significantly less affected, and the approach can lead to an underestimation of the fast neutron flux at that time. A poor estimation of the clad fast flux values can affect the clad creep rate, fuel-clad gap, and clad stresses. Additionally, the fast flux factor has been considered constant throughout the life of the rod. As the cycle progresses, the spectrum will harden as burnable absorbers and 235U deplete, so the fast flux would be expected to increase as the spectrum hardens. The effect on quantities of interest are highlighted for the first cycle of the Watts Bar Unit 1 core, comparing the results from an explicit clad fast flux representation from MPACT to the constant fast flux factor approach used in previous analyses. Future work will consider multicycle effects to assess the impact at higher burnups.

Stimpson, Shane G.↗
DOE OSTI2018

Pronghorn: A Porous Media Thermal-Hydraulics Core Simulator and its Validation with the SANA Experiments

Pebble bed High Temperature Reactors (HTRs) are characterized by many advantageous design features, such as excellent passive heat removal in accidents, large margins to fuel failure, and online refueling potential. However, a significant challenge in the core modeling of pebble bed reactors is the complex fuel-coolant structure. This paper presents a new porous media simulation code, Pronghorn, that aims to alleviate modeling challenges for pebble bed reactors by providing a fast-running, medium- fidelity core simulator. Pronghorn is intended to accelerate the design and analysis cycle for pebble bed and prismatic HTRs by permitting fast scoping studies and providing boundary conditions for systems-level analysis. Pronghorn is built on the Multiphysics Object-Oriented Simulation Environment (MOOSE) using modern software practices and a thorough testing framework. This paper describes the physical models used in Pronghorn and demonstrates Pronghorn’s capability for modeling gas-cooled pebble bed HTRs by presenting simulation results obtained for the German SANA pebble bed decay heat experiments. Within the limitations of the porous media approximation and existing available closure relationships, Pronghorn predicts the SANA experimental pebble temperatures well, expanding the code’s validation base. A brief code-to-code comparison shows a similar level of accuracy with other porous media simulation tools; this enhances Pronghorn’s advantages of an arbitrary equation of state, unstructured mesh capabilities, compressible flow models, high coupling potential to MOOSE fuels performance and systems-level thermal-hydraulics codes, and modern software design.

42 ENGINEERING↗
DOE OSTI2016

Solution of the Hot Zero Power BEAVRS Using a Pin-Cell Homogenized Model

The validation of a reactor physics code system is an essential part of the code development process. It allows for the evaluation of the code's ability to predict and model the complex physical processes that take place in a nuclear reactor under various operational conditions. Idaho National Laboratory (INL) has been developing a set of reactor physics analysis tools within the MOOSE framework. MAMMOTH is the main reactor physics application, which integrates other MOOSE applications to solve complex multiphysics problems. Results from this study demonstrate that the MOOSE based reactor physics tools, currently under development at INL, have the flexibility to model with high accuracy the HZP condition for a standard PWR. The DRAGON5-Rattlesnake full core pin-cell homogenized simulation produces eigenvalues and detector signals that are comparable to those generated with OpenMC and nTRACER. However, many improvements are necessary to achieve higher fidelity results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Elk: A New MOOSE Framework Application for Radio-frequency Electromagnetics

A new tool in-development for general electromagnetic (EM) simulation for modeling radio-frequency (rf) systems is presented. The primary motivation for development of this tool is to leverage the flexibility and high-performance-computing capability of the Multiphysics Object-Oriented Simulation Environment (MOOSE) Framework developed by Idaho National Laboratory (INL) for complex EM problems. A MOOSE application (or MOOSE App) enables the end user to create a fully-functional simulation scenario without needing explicit knowledge of the underlying finite-element weak form implementation. MOOSE combines the libMesh library and the PetSc suite as the finite element and solver libraries, giving a robust platform for solving highly non-linear, coupled sets of PDEs such as those found in EM and plasma physics problems. Initial validation efforts have included standard waveguide and antenna benchmarks with the current focus being field validation against the well-known capacitively-coupled GEC Reference Cell (GEC-CCP). The results of this effort will be outlined and discussed. Future work plans including cold-plasma rf wave propagation studies and coupling to the plasma fluid MOOSE App, Zapdos [Lindsay et al 2016 J. Phys. D: Appl. Phys. 49 235204], will also be discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

FAST.Farm Response to Varying Wind Inflow Techniques: Preprint

FAST.Farm is a newly developed multiphysics, midfidelity engineering tool that can be used to predict turbine power and structural loads of wind turbines in a wind farm. Previous studies have shown the similarities and differences between FAST.Farm and large-eddy simulations (LES) using the same LES-precursor-generated ambient wind inflow. The ability to generate ambient wind inflow using a synthetic turbulence engineering model (e.g., TurbSim or the Mann model) has recently been integrated into FAST.Farm to potentially enable more computationally efficient usage of the tool. This work aims to: 1) compare FAST.Farm simulations using LES-generated inflow to those using synthetically generated inflow from TurbSim, and 2) establish guidelines by which TurbSim-generated inflow should be generated for wind farm analysis. It was found that properly setting spatial coherence parameters for the transverse wind velocity components is necessary to accurately predict wake meandering. It is shown that TurbSim-generated inflow can be used in FAST.Farm to accurately predict thrust, power, speed, and torque for waked and unwaked turbines; wake meandering behavior across different atmospheric conditions; and averaged wake-deficit advection and evolution effects.

17 WIND ENERGY↗
DOE OSTI2018

Development of Preliminary VERA-CS Crud Induced Localized Corrosion Modeling Capability

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing multiphysics core-simulator software for light water reactors (LWRs) known as VERA-CS in order to improve the state of the art in modeling and simulation of challenge problems that are limiting to the nuclear industry. One of these challenge problems includes fuel rod crud deposition, which can lead to crud-induced power shift (CIPS) and crud-induced localized corrosion (CILC). This paper documents work that was performed to develop a preliminary CILC-modeling capability in VERA-CS in support of the crud challenge problem. The CILC capabilities were developed by coupling VERA-CS to the CASL-developed Cicada code, which provides 1D and 3D clad conduction and oxide growth modeling tools, as well as coupling to the CASL-developed MAMBA code, which is used for modeling clad crud deposition. An approach called rod thermal-hydraulic reconstruction (ROTHCON) was developed and integrated into VERA-CS. This allows the modeler to capture spacer-grid turbulence and heat transfer effects in the CTF subchannel code so that the spatial resolution of crud and oxide rod surface growth could be better resolved. After implementing these capabilities, several assessments were performed to ensure that the capabilities function as expected, and a pin-resolved quarter-core simulation was run as a demonstration.

Salko Jr, Robert↗
DOE OSTI2019

Chapter 4 - Research Tools: Microstructure, Mechanical Properties, and Computational Thermodynamics

The purpose of this Chapter is to provide an overview of the research tools that are available to evaluate the response of materials that are subject to atomic displacements caused by fast particle irradiation. The major focus is on metals employed in structural components in operating or planned nuclear fission and fusion reactors. Understanding, predicting and characterising the changes in material properties under irradiation requires the ability to characterise the irradiation-induced microstructure at and above the atomic scale, and the ability to relate such changes to the observed degradation in performance sustaining properties. Thus, any discussion of research tools necessarily involves microstructural characterization techniques and mechanical property characterization techniques. In addition, computational thermodynamics (and related multiscale, multiphysics modeling) can also be considered characterization tools since they enable the study of the thermodynamically and kinetically mediated mechanisms (influenced by the irradiation environment) that affect microstructural evolution. Hence, this chapter also contains a brief review of computational thermodynamics.

English, Colin A.↗
DOE OSTI2019

Validation Studies using ELK and the Open-Source MOOSE Framework Application Zapdos for Electromagnetic Coupled Plasma Simulation

An open source simulation platform that captures electromagnetics, plasma chemistry, and fluid treatment of the plasma domain would be a powerful tool for studying low temperature plasma (LTP) phenomena. Within the Multiphysics Object-Oriented Simulation Environment (MOOSE) open source framework, ELK, CRANE, and Zapdos are combined to provide open source simulation capability for the LTP community. This talk presents continued validation of the general electromagnetic (EM) simulation tool ELK (Electromagnetic Library for Kinetics & fluids). Previous validation efforts focused on simulating standard waveguide and antenna benchmarks in vacuum and linear dielectrics, but ELK has since been coupled with the MOOSE-based plasma fluid application Zapdos [Lindsay et al 2016 J. Phys. D: Appl. Phys. 49 235204] to perform radio-frequency (rf) capacitively coupled plasma (CCP) system simulation. Plasma validation efforts initially focused on electrostatic benchmarks to demonstrate ELK-Zapdos code coupling and then moved to electromagnetic benchmarks based on interesting geometries and conditions (VHF effects in a CCP reactor, microwave heating, etc.). Results and progress in these efforts will be summarized and discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Thermally Anisotropic Composites for Heat Redirection and Thermal Management in Building Envelopes

In building envelope systems, thermal management is important from energy conservation and thermal comfort perspectives, and is typically done via insulation, thermal mass, solar control and shading, etc. Thermal anisotropy is commonly utilized in electronics for heat dissipation. This article investigates the feasibility of utilizing thermally anisotropic composites (TACs) to reduce unwanted heat gains and losses through the building envelope by redirecting the heat transfer to or from appropriate heat sinks and sources.Numerical simulations of building envelope components using COMSOL Multiphysics and whole building simulations using EnergyPlus (E+) were performed to estimate the energy savings potential of TACs used in external walls of residential buildings. Application of TACs was compared with baseline walls with cavity insulation as well as with and without exterior continuous insulation (CI). The COMSOL simulations indicated 75% or higher reductions in heat gains when using a TAC coupled with a heat sink compared to an exterior CI case. In terms of energy savings in a hot climate, E+ models estimated cooling energy savings of 19% with a "TAC + heat sink" compared to a baseline wall with only cavity insulation. Further, a small-scale test sample with a TAC-heat sink system was built and tested in a multi-transducer heat flow meter (HFM). The heat sink was a copper tube circulating cold water. The experiments verified the re-direction of heat to a sink and significant reduction in heat transfer through the test specimen. The experiments showed that the heat redirection only occurred in the presence of both thermal anisotropy and water circulation; if either element was missing, the system behaved like a traditional envelope component.

Biswas, Kaushik↗
DOE OSTI2017

Surrogate Models for TREAT Transient Calculations

The Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is currently preparing for resumption of operation after more than 20 years in a long-term standby state. In parallel with preparation for operations and a new transient testing program, the DOE-NE Advanced Modeling and Simulation program (NEAMS) has been supporting the development of three-dimensional transient multi-physics modeling capabilities within the reactor physics application MAMMOTH [1]. Recently, the MAMMOTH reactor physics team has begun development of uncertainty and sensitivity analysis for the TREAT core. This research will be used to support the analysis of reactor operation and experiment planning. Traditional approaches, i.e., Monte Carlo simulations, for such tasks require a large number of model evaluations under both steady state and transient conditions. However, exhaustive sets of sampling runs are simply not practical here since evaluations of each model can take a substantial amount of time. A solution to this problem is to build a surrogate model employing a smaller number of model evaluations. The surrogate model is designed to provide an accurate and fast approximation to the detailed TREAT model. In this manner, the surrogate model can be used in place of the TREAT transient model for uncertainty and sensitivity analysis at a fraction of the computational cost. The high dimensional model representation (HDMR) [2] method is employed here for approximating the responses of interest in terms of functions of lower dimensions. This method has been implemented in RAVEN [3] and all TREAT transient calculations are performed using the multiphysics code MAMMOTH. In addition, a code interface has also been developed in order to couple RAVEN with MAMMOTH. The surrogate model is then tested and compared for the MAMMOTH model of the TREAT startup testing program Transient Test #15 core [4], and its accuracy is measured by its ability to estimate the mean, coefficient of variance, skewness, excess kurtosis and probability density function of the outputs of Transient Test #15 model, as well as root mean square error metric.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Comparison between PARFUME and Bison Using the AGR-2 Irradiation Experiment

This report documents comparisons between Fuel Model (PARFUME) model predictions versus Bison for selected compacts from the second irradiation test of the Advanced Gas Reactor (AGR) program that occurred from June 2010 to October 2013 in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). PARFUME is a fuel performance analysis and modeling code, used for evaluating gas-reactor tristructural isotropic (TRISO) coated particle fuel for prismatic, pebble bed, plate, and cylindrical type fuel geometries. PARFUME is an integrated mechanistic computer code that evaluates the thermal, mechanical, and physico-chemical behavior of TRISO coated-fuel particles and the probability for fuel failure given the particle-to-particle statistical variations in physical dimensions and material properties that arise during the fuel fabrication process. Bison is a nuclear fuel performance application built using the Multiphysics Object-Oriented Simulation Environment (MOOSE) finite element library (Permann 2020). Bison is capable of modeling multiple fuel forms in a wide variety of dimensions and geometries. It solves coupled nonlinear partial differential equations, including heat conduction, mechanics, fission product species transport etc., in a fully implicit manner. Comparisons between PARFUME and Bison were performed using four compacts from the AGR-2 experiment. Selected outputs were chosen based on their impact on the probability of the SiC layer failing. In general there was good agreement between PARFUME and Bison with the exception of predicting the gap formed between the buffer and IPyC. Further comparisons between PARFUME and Bison are planned to further develop Bison’s capabilities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Tomographic Reconstruction of the Neutron Time-Energy Spectrum from a Dense Plasma Focus

A dense plasma focus (DPF) nuclear fusion generator is an attractive pulsed neutron source in many applications, due to its relatively large neutron yield, produced in a short time duration. In order to design a DPF that generates neutrons within a specified time profile, or generates a neutron energy spectrum with specific properties, it is necessary to understand and be able to fully characterize the results of the fusion process releasing the neutrons. The time-energy spectrum of fusion neutrons is an ideal quantity to use to validate multiphysics codes that simulate the pinch and fusion processes, because it is a quantity that requires the physics of each of the stages leading up to and ending in the fusion reaction to be simulated correctly. In particular, since DPF fusion neutrons are not monoenergetic – and there can often be several fusion pinches creating neutrons – a computer simulation matching high quality neutron spectrum measurements provides great confidence in the fidelity of the simulation. In order to make such a comparison, it is first necessary to have quality measurements from which to infer the spectrum. In this work we pose neutron spectroscopy as the classical tomographic inverse problem from neutron time of flight data at multiple distances, but enhanced by using detector pairs designed to reduce the scatter background and with an additional measurement of the time profile of the fusion pinch near its source. The detector pairs enhance the quality of the time of flight measurements and the additional constraint posed by the measured time profile allows for temporal resolution in the spectrum reconstruction that is equal to the measurement resolution, which can be subnanosecond, and energy resolution in the 10’s of keV, without the problem being underconstrained. We present results from a Deuterium-fueled DPF at the U.S. Department of Energy’s Nevada National Security Site and show that we can infer the time-energy spectrum from our measurements for both single and multi-pinch fusion reactions with equal fidelity.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

xSDK: Building an ecosystem of highly efficient math libraries for exascale

Current efforts to build increasingly powerful computer architectures are opening up new avenues for more complex and higher fidelity simulations coupled with data analytics and learning, leading to new scientific insights and deeper understanding. At one extreme, exascale computers will be much faster than previous computer generations (performing 10 18 operations per second—that is, 1,000 times faster than petascale). To achieve these performance improvements, computer architectures are becoming increasingly complex, with deep memory hierarchies, very high node and core counts, and heterogeneous features such as graphics processing units (GPUs). Such architectural changes impact the full breadth of computing scales, as heterogeneity pervades even current-generation laptops, workstations, and moderate-sized clusters. While emerging advanced architectures provide unprecedented opportunities, they also present significant challenges for developers of scientific applications, such as multiphysics and multiscale codes, who must adapt their software to handle disruptive changes in architectures and new programming models that have not yet stabilized. Developers must consider increasing concurrency while reducing communication and synchronization, and other complexities such as the potential for using mixed precision to leverage the compute power available in low-precision tensor cores. On one hand, developers must implement new scientific capabilities, which in turn increase code complexity. On the other hand, the codes must be ported to new architectures, requiring the inclusion of new programming models and the restructuring of code to achieve good performance. Addressing these issues is beyond the capability of any single person or team—leading to the need for collaboration among many teams, who encapsulate their expertise in reusable software and work together to create sustainable software ecosystems.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

Challenges in simulating ground interacting nuclear explosions

This paper summarizes recent above-ground nuclear explosion simulations as part of a broader effort to better characterize conditions within a fireball that may influence the chemical evolution of bomb materials and other materials entrained from the local explosion environment. A critical component of this work is validation against historic footage of atmospheric testing, requiring that we understand how the frequency-dependent sensitivity of the utilized film footage influences data captured in such images. We focus first on the early physics of a nuclear explosion in the atmosphere before discussing some of the technical challenges we seek to capture in late-time models that include more complex emplacement conditions and subsurface features. We discuss required physics packages (compressible hydrodynamics, radiation transport, as well as necessary ancillary tables such as equations of state (EOS) and opacities). Additionally, we note reasonable “shortcuts” one may make and their limitations, e.g., using ideal gas EOS, replacing spectrally resolved radiation with spectrally averaged radiation, and exchanging deterministic transport with diffusion. We then discuss an approach to achieving an equilibrated initial stress state for problems where buoyancy and subsurface lithostatic stress are important. Our methodology is presented in the context of LLNL’s ALE3D multiphysics code but may readily be implemented in other codes. In this paper, we start with a description of the challenges of NUDET simulations, followed by a presentation of the simulated intensity (flux) as it would appear on an analysis of the Dixie test. We then progressively introduce additional complexity in subsequent sections (near-surface burst and gravity initialization) before concluding.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2021

An HPC-Based Hydrothermal Finite Element Simulator for Modeling Underground Response to Community-Scale Geothermal Energy Production

Geothermal heat, as renewable energy, shows great advantage with respect to its environmental impact due to its significantly lower CO2 emissions than conventional fossil fuel. Open and closed-loop geothermal heat pumps, which utilize shallow geothermal systems, are an efficient technology for cooling and heating buildings, especially in urban areas. Integrated use of geothermal energy technologies for district heating, cooling, and thermal energy storage can be applied to optimize the subsurface for communities to provide them with multiple sustainable energy and community resilience benefits. The utilization of the subsurface resources may lead to a variation in the underground environment, which might further impact local environmental conditions. However, very few simulators can handle such a highly complex set of coupled computations on a regional or city scale. We have developed high-performance computing (HPC) based hydrothermal finite element (FE) simulator that can simulate the subsurface and its hydrothermal conditions at a scale of tens of km. The HPC simulator enables us to investigate the subsurface thermal and hydrologic response to the built underground environment (such as basements and subways) at the community scale. In this study, a coupled hydrothermal simulator is developed based on the open-source finite element library deal.II. The HPC simulator was validated by comparing the results of a benchmark case study against COMSOL Multiphysics, in which Aquifer Thermal Energy Storage (ATES) is modeled and a process of heat injection into ATES is simulated. The use of an energy pile system at the Treasure Island redevelopment site (San Francisco, CA, USA) was selected as a case study to demonstrate the HPC capability of the developed simulator. The simulator is capable of modeling multiple city-scale geothermal scenarios in a reasonable amount of time.

Sun, Xiang↗
DOE OSTI2022

Initial study on cross section generation requirements for a PBR equilibrium core

A Serpent model of the HTR-PM equilibrium core was developed for use in cross section prepa-ration studies in order to guide methods development for the Griffin reactor multiphysics applica-tion. The model includes detailed isotopics for 10 distinct pebble burnup groups in 126 core zoneswith unique fuel and moderator temperatures obtained from a coupled neutronics-thermal-fluidsequilibrium core calculation using Griffin-Pronghorn. A sensitivity study of the fuel and mod-erator temperatures for various core regions was performed with the MOOSE stochastic tools.The results show that the uncertainties are, not unexpectedly, dominated by the value of the fluidtemperature and that the power level, heat transfer coefficient and effective conduction to neigh-boring pebbles and fluid constitute, at best, second order effects. The temperature uncertaintyrange varies from 28 K to 57 K at the core entry and exit planes, respectively, but these val-ues are probably higher. We still have to quantify the significance of these uncertainties in thepreparation of cross sections, which will be postponed for future work. In addition, we verifythat the single effective pebble approach works well for the preparation of region averaged crosssections in the infinite domain approximation. Nevertheless, there are significant discrepanciesin the cross sections when compared to the multi-pebble model. This could affect the predictionof peak values and in the depletion calculation. We conclude that is highly desirable for futurestudies with Griffin to be able to handle both the ?effective? pebble approximation and the multi-pebble approach for the various pebble burnup groups. This enables Griffin users the flexibilityto perform higher-fidelity studies. Finally, we initiate the preparation of cross sections for variouscore regions from the full core Serpent reference model. We quantify the differences in 26 groupcross sections from infinite domain models. These reference cross sections will serve to validatethe double heterogeneity, self-shielding, and spectrum-correction methods in Griffin.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Initial Developments in Modeling Graphite Behavior

Graphite is used in High Temperature Reactor designs as a moderator, reflector and core material responsible for protecting the fuel and maintaining structural stability. From a safety perspective, it is imperative to be able to predict how graphite will behave under reactor operating conditions which may compromise the core’s structural integrity or fuel’s safety performance. The following report summarizes recent modeling work undertaken at Idaho National Laboratory which focuses on graphite degradation behavior. One phenomenon which is explored in this report and can cause significant degradation to graphite properties is oxidation. Oxidation behavior in graphite is highly temperature dependent. At low temperatures, oxygen can fully penetrate a component and cause a homogeneous damage profile. At high temperatures, only the graphite near the surface of a component is affected. Understanding and modeling this temperature dependence is essential to predicting graphite behavior. In this report, three graphite models are discussed. The first model is used to investigate strength loss after low temperature oxidation. It does this by generating an approximate graphite microstructure then determining the required load to cause crack propagation. It has been shown that the model reproduces the quasi brittle stress versus strain relationship observed in graphite. Strength loss results from the model are shown to match experimental values. The second model investigates strength loss under high temperature oxidation conditions. This model is more applicable to full scale graphite components where the oxidation damage is often inhomogeneous throughout a component. The third model computes stresses in graphite under reactor conditions. Specifically, the model incorporates the effect of creep, irradiation dose, thermal strains, and oxidation in order to predict internal stresses in a component. All of these models are implemented in the Multiphysics Object-Oriented Simulation Environment (MOOSE), an open source, parallel finite element framework.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗