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Matthews, Christopher

Publications and source records attributed to Matthews, Christopher.

At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Atomistic modeling of out-of-pile xenon diffusion by vacancy clusters in UO 2

We report that diffusion of Xe atoms in UO 2 fuel is important for nuclear fuel performance, and is enabled by interaction with U and O vacancies. Previous work using atomistic calculations based on density functional theory (DFT) and empirical potentials (EP) focused on the role of small vacancy clusters (Xe U2Oy , y = 0, 1, 2) for Xe transport, but the model predictions failed to reproduce the experimental observations by consistently underestimating Xe diffusivity. In this work, where we focused on out-of-pile conditions (i.e. in the absence of irradiation), we have explored two of the uncertainties associated with the model: the DFT methodology and the types of clusters considered. Additionally, we found that using the energy barriers obtained with GGA + U DFT allows to reconcile simulation results and experimental observations in the intrinsic regime (i.e. out-of-pile conditions). The Xe UO cluster is the preferred configuration at high temperature, while the mobile Xe U2O cluster takes over below 1720 K. The latter cluster is also the main contributor to Xe diffusion and good agreement is obtained with current models empirically fitted to experiments for diffusion at high temperature. A simple expression that captures most of Xe diffusivity is proposed, relying on Xe UO and Xe U2O clusters only. We also determined the out-of-pile concentration and diffusivity of extended clusters Xe U z O y , 3 ≤ x ≤ 9, and determined that these do not contribute significantly to the overall diffusivity under intrinsic conditions. However, we speculate that the Xe U 4 O 3 and Xe U 8 O 9 clusters, which have relatively low migration barrier, may play a role under irradiation conditions, because the radiation-induced enhanced vacancy concentration would rapidly increase the fraction of large Xe-vacancy clusters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Task 1: Evaluation of M&S tools for micro-reactor concepts

The Westinghouse Electric Company (WEC) and Los Alamos National Laboratory (LANL) have been awarded a DOE ARPA-E MEITNER award for development of its eVinci microreactor design. Fundamentally, the attractiveness of heat pipe micro-reactors comes from the idea that the core is inherently self-regulating, simplifying the reactor design and requiring significantly fewer safety related components. While the WEC Design Team is focused on the experimental and fabrication tasks to ensuring self-regulation, the MEITNER Resource Team is focused on utilizing available codes to computationally show the core is indeed self-regulating. This requires the combination of fundamentally different codes to simulate the feedback mechanism between heat generation, temperature, and density.

42 ENGINEERING↗
DOE OSTI2019

A general constitutive framework for the combined creep, plasticity and swelling behavior of nuclear fuels in an implicit hypoelastic formulation

This work is motivated by the need to analyze the behavior of nuclear fuels which under normal operating conditions build up stresses due to non-homogeneous thermal expansion, fission gas and solid product swelling among other phenomena that are simultaneously relaxed by creep and plastic flow. This report details the stress and tangent update equations for combined J2 based rate independent plasticity, timehardening creep and fission gas induced swelling e ects in a fully implicit hypo-elastic formulation involving two cases: pure creep without plasticity where the yield criterion has not yet been met and the combined e ect of both creep and plasticity beyond yield. Closed form expressions for the consistent material tangent to be used in both cases are derived which can be used in implicit codes and is expected to help in obtaining optimal convergence rates.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Stress and tangent update equations for combined time-hardening creep and J 2 plasticity in an implicit hypo-elastic formulation

This work is motivated by the need to analyze the behavior of metallic nuclear fuels which under normal operating conditions build up stresses due to non-homogeneous thermal expansion, fission gas and solid product swelling among other phenomena that are simultaneously relaxed by creep and plastic flow. This report details the stress and tangent update equations for combined J 2 based rate independent plasticity and time-hardening creep effects in a fully implicit hypo-elastic formulation involving two cases: pure creep without plasticity where the yield criterion has not yet been met and the combined effect of both creep and plasticity beyond yield. Closed form expressions for the consistent material tangent to be used in both cases are derived which can be used in implicit codes and is expected to help in obtaining optimal convergence rates.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

PUMA: Plutonium Modeling and Analysis Theory Manual Rev 1.0

The high rate of alpha decay of the plutonium-238 isotope lends the material to be used for many unique space applications where the reach of solar radiation is either too weak or unavailable. With a power density of 0.54 watts per gram, plutonium-238 has been used to power space exploration satellites from mars rovers to deep space probes. Los Alamos National Laboratory (LANL) has remained the primary location for fabrication of the plutonium oxide pellets. In an effort to apply modern techniques and tools towards supporting LANL’s fabrication responsibilities, PUMA was created under an LDRD funded program to gain understanding of plutonium oxide heat source pellets during fabrication.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

Summary and Assessment of Metal Fuel Capabilities in Bison

The work described in this report was performed under funding from the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. This report is issued in satisfaction of the level 2 milestone M2MS-18IN0201012 on Release Bison with Initial Metallic Fuel Performance Capability. Metallic fuel development began in Bison a few years after light water reactor development began (~ 2009). The initial development was funded by the Advanced Fuels Campaign. With recent interest from industry, the Nuclear Regulatory Commission, and the Versatile Test Reactor program, the Nuclear Energy Modeling Simulation Program has increased funding for metallic fuel development and validation in FY18. The material models with the most mature development are binary uranium-zirconium and ternary uranium-plutonium-zirconium alloys for fuel and HT9 for cladding. The first material models were incorporated from open literature sources, which include mechanical and thermal material properties that are functions of temperature, porosity, and zirconium concentration, swelling, fission gas release, zirconium diffusion, creep, and sodium coolant channel boundary conditions. These material models have been utilized to simulate ERB-II fuel pins and the results compared to measurements. The EBR-II experiment measurements mostly consist of cladding strain and zirconium redistribution. As such, comparisons to EBR-II measurements are currently limited to these two figures of merit. Simulations and comparisons to experiment measurements have also been done for transient testing of EBR- II fuel pins in TREAT, which included temperature measurements. Preliminary comparisons between Bison calculations and experiment measurements are favorable. These comparisons have also highlighted the importance of the fuel swelling and fission gas release models on cladding strain. Also, Bison metallic fuel simulations are currently being run in support of ATR experiments and VTR exploratory design calculations. Further development and evaluation of individual material models and fuel system models are planned for FY19.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Bison Improvements for Robustness and Speed

Early in its development, the primary emphasis of efforts to improve the Bison fuel performance code were focused on initial development of new capabilities. As Bison has matured and become adopted by a wider set of users, it has become clear that a number of robustness issues needed to be addressed to ensure that the code can reliably produce a converged solution. This need became particularly evident when a recent effort by Westinghouse Electric Company to simulate all of the fuel rods in a reactor using Bison resulted in approximately 30% of the models failing to run to completion because of lack of convergence or other errors that were encountered. It is important to emphasize that there was no evidence that Bison generated results that were incorrect for the set of models exercised. In many cases, issues such as those encountered could be addressed by adjusting solver-related parameters. However, there is no question that for Bison to become widely adopted for production work, it needs to be inherently far more robust than it was. Because of this, a concerted effort was undertaken to improve Bison's robustness, or its ability to reliably obtain converged solutions under a wider variety of conditions. As a result of this work, a number of high-priority areas for improvement for modeling light water reactor (LWR) fuel rods have been identified and addressed. As a result of this work, the number of failing Bison analyses in the aforementioned set of simulations for a full core has been reduced to a fraction of a percent. In addition, the robustness of Bison models using the VERA core simulator has been significantly improved. This is important because all of the Bison models representing individual fuel rods must run to completion successfully for the full core model to run to completion. This report provides summaries of the identified sources of robustness issues in Bison and the development that was done to resolve these issues. It also summarizes the addition of new capabilities to address deficiencies in Bison's ability to include needed aspects of LWR fuel behavior, and improvements that were made to improve Bison's solution efficiency.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

FY17 Progress in Modeling of Lanthanide Transport in Metallic Fuels

A critical review of fuel-clad-chemical interactions along with modelling requirements is published. The mechanism of lanthanide transport is studied experimentally (NEUP collaboration) and using simulations and initial results are published in Refs.

36 MATERIALS SCIENCE↗
DOE OSTI2017

BISON

The Idaho National Laboratory (INL) fuel performance code BISON is a finite element code used to simulate nuclear fuel during irradiation. It contains models for different types of fuel and cladding such as fundamental thermo-physical properties, creep, swelling, porosity, and mechanical interaction. The authors contribute to the code through the development of some of these models, such as newly formulated thermal conductivity, phase redistribution, and thermal expansion models. In addition, solver methods are contributed where appropriate to facilitate calculations. The code is controlled by INL through an export-controlled license similar to other DOE nuclear related codes.

Matthews, Christopher↗
DOE OSTI2017

Radiation Re-solution Calculation in Uranium-Silicide Fuels

The release of fission gas from nuclear fuels is of primary concern for safe operation of nuclear power plants. Although the production of fission gas atoms can be easily calculated from the fission rate in the fuel and the average yield of fission gas, the actual diffusion, behavior, and ultimate escape of fission gas from nuclear fuel depends on many other variables. As fission gas diffuses through the fuel grain, it tends to collect into intra-granular bubbles, as portrayed in Figure 1.1. These bubbles continue to grow due to absorption of single gas atoms. Simultaneously, passing fission fragments can cause collisions in the bubble that result in gas atoms being knocked back into the grain. This so called “re-solution” event results in a transient equilibrium of single gas atoms within the grain. As single gas atoms progress through the grain, they will eventually collect along grain boundaries, creating inter-granular bubbles. As the inter-granular bubbles grow over time, they will interconnect with other grain-face bubbles until a pathway is created to the outside of the fuel surface, at which point the highly pressurized inter-granular bubbles will expel their contents into the fuel plenum. This last process is the primary cause of fission gas release. From the simple description above, it is clear there are several parameters that ultimately affect fission gas release, including the diffusivity of single gas atoms, the absorption and knockout rate of single gas atoms in intra-granular bubbles, and the growth and interlinkage of intergranular bubbles. Of these, the knockout, or re-solution rate has an particularly important role in determining the transient concentration of single gas atoms in the grain. The re-solution rate will be explored in the following sections with regards to uranium-silicide fuels in order to support future models of fission gas bubble behavior.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Enhancements to BISON U-Zr Metallic Fuel X447 Example Problem

As development of a metallic fuel modeling capability in BISON has progressed, the need for an example problem used as a comparison basis was observed. Collaborative work between researchers at Los Alamos National Laboratory (LANL) and Idaho National Laboratory (INL) then proceeded to determine a viable rod to use as the basis and create a BISON input deck utilizing as many metallic fuel models as feasible. The basis chosen was what would be considered a generic rod from subassembly X447, an assembly irradiated in EBR-II towards the end of its operating life, heavily based on reported data for fuel pin DP11. Thus, the approach was adopted to use flow characteristics from subassembly X447 as a basis for the convective heat transfer solution, power history and axial power profiles that are representative of rod DP11 from subassembly X447. The rod simulated is a U-10Zr wt% (U-22.5Zr at%) composition. A 2D-RZ mesh would be used to capture axial thermal hydraulic effects, axial swelling and stress-strain calculations over the full length of the rod. After initial work was invested, a refinement of the various models and input parameters was conducted to ensure consistency between operator-declared conditions, model input requirements and those represented in the example problem. This report serves as a synopsis of the enhancements and refinements to the example problem conducted throughout the 2016 fiscal year.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗