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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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At least 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Short-time scale behavior modeling within long-time scale fuel cycle evaluations

Typically, short-time and long-time scales in nuclear energy system behavior are accounted for with entirely separate models. However, long-term changes in system characteristics do affect short-term transients through material variations. This paper presents an approach to consistently account for short-time scales within a nuclear system lifespan. The reported findings and developments are of significant importance for small modular reactors and other nuclear energy systems operating in autonomous modes. It is necessary to simulate the short time-scale kinetic behavior of the reactor as well as the long time-scale dynamics that occur with fuel burnup. The former is modeled using the point kinetics equations, while the latter is modeled by the Bateman equations. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

MARVEL Reactor Fuel Performance Report

The Microreactor Applications Research Validation and EvaLuation (MARVEL) project is producing a high temperature liquid metal-cooled nuclear test bed at Idaho National Laboratory (INL) to ultimately improve the integration of microreactors to end-user applications. This ambitious effort seeks to design, authorize, construct, test, and operate the reactor within five years. In order to construct and operate the MARVEL reactor in a timely manner, the system will utilize materials and component designs which have already been used, qualified, or licensed from previous reactors. The MARVEL reactor will be located at the INL Transient Reactor Test (TREAT) facility in the north high-bay equipment pit and will use the existing 304 stainless steel-clad U-ZrH1.6 pin-type fuel system developed by General Atomics and purchased from TRIGA International. This fuel has been previously qualified under the United States Department of Energy’s (US DOE) Reduced Enrichment for Research and Test Reactors (RERTR) Program. Even though the regulator of the MARVEL reactor is the US DOE, the standards and overall approach recommended by the Nuclear Regulatory Commission is well-defined and utilized here. Following NUREG-1537 regulatory guidance, this report documents the authorization case for the MARVEL fuel system’s application to MARVEL and establishes stable and predictable fuel performance during the most thermophysically unfavorable conditions achievable in the MARVEL reactor. To that end, this report provides a comprehensive survey of the known thermophysical properties, performance, and quantitative relationships associated with the MARVEL reactor fuel element and uses this information to determine its mechanical integrity and risk of reaching unacceptable conditions during the most extreme accident scenarios predicted for the reactor using the most conservative assumptions available. The information contained herein is compiled from a combination of historical reports and peer reviewed scientific publication manuscripts. Known mechanisms under which the fuel is susceptible to failure are highlighted and compared to conditions that could exist in the MARVEL reactor during an unanticipated transient or accident scenario. The two scenarios considered for analysis in this report are (1) an unprotected loss of flow accident and (2) a hypothetical unprotected loss of coolant accident during the loss of flow accident. Preliminary 2D steady-state analyses herein indicate that both fuel-cladding chemical interactions and fuel-cladding mechanical interactions are negligible throughout the fuel’s operational cycle under both normal and high temperature accident scenario conditions. Although higher fidelity 3D time-dependent modeling and simulations are planned, the following may be concluded presently. The MARVEL fuel element maintains its geometric stability and structural integrity during the most extreme accident scenarios predicted for the MARVEL reactor. The hoop stress during the unprotected loss of flow accident reaches about -1.27 MPa; this negative stress indicates that it is compressive rather than tensile. The compressive stress is a result of the NaK pressure on the outside of the fuel element, caused by the restricted thermal expansion of the NaK coolant, exceeding the internal pressures generated inside of the fuel element. The hoop stress generated in the cladding during the unprotected loss of coolant accident reaches a maximum of about approximately 10 MPa, which is nearly an order of magnitude less than the predicted yield strength of the cladding under high-temperature accident scenario conditions. Calculations were compared with results from high performance computational simulations using BISON and are in very close agreement. A conservative MARVEL fuel meat peak temperature limit of 900 °C is recommended presently, which is about 180 °C higher than the peak fuel temperature predicted to occur during the most extreme accident. Based on the known properties and behavior of the MARVEL fuel element, the fuel successfully meets its design and safety requirements under normal and most extreme accident conditions with a large safety margin.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Comparative urban drive cycle simulations of light-duty hybrid vehicles with gasoline or diesel engines and emissions controls

Electric hybridization is a very effective approach for reducing fuel consumption in light-duty vehicles. Lean combustion engines (including diesels) have also been shown to be significantly more fuel efficient than stoichiometric gasoline engines. Ideally, the combination of these two technologies would result in even more fuel efficient vehicles. However, one major barrier to achieving this goal is the implementation of lean-exhaust aftertreatment that can meet increasingly stringent emissions regulations without heavily penalizing fuel efficiency. We summarize results from comparative simulations of hybrid electric vehicles with either stoichiometric gasoline or diesel engines that include state-of-the-art aftertreatment emissions controls for both stoichiometric and lean exhaust. Fuel consumption and emissions for comparable gasoline and diesel light-duty hybrid electric vehicles were compared over a standard urban drive cycle and potential benefits for utilizing diesel hybrids were identified. Technical barriers and opportunities for improving the efficiency of diesel hybrids were identified.

Gao, Zhiming↗
DOE OSTI2011

Progress on Plant-Level Components for Nuclear Fuel Recycling: Commonality

Progress made in developing a common mathematical modeling framework for plant-level components of a simulation toolkit for nuclear fuel recycling is summarized. This ongoing work is performed under the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, which has an element focusing on safeguards and separations (SafeSeps). One goal of this element is to develop a modeling and simulation toolkit for used nuclear fuel recycling. The primary function of the SafeSeps simulation toolkit is to enable the time-dependent coupling of separation modules and safeguards tools (either native or third-party supplied) that simulate and/or monitor the individual separation processes in a separations plant. The toolkit integration environment will offer an interface for the modules to register in the toolkit domain based on the commonality of diverse unit operations. This report discusses the source of this commonality from a combined mathematical modeling and software design perspectives, and it defines the initial basic concepts needed for development of application modules and their integrated form, that is, an application software. A unifying mathematical theory of chemical thermomechanical network transport for physicochemical systems is proposed and outlined as the basis for developing advanced modules. A program for developing this theory from the underlying first-principles continuum thermomechanics will be needed in future developments; accomplishment of this task will enable the development of a modern modeling approach for plant-level models. Rigorous, advanced modeling approaches at the plant-level can only proceed from the development of reduced (or low-order) models based on a solid continuum field theory foundation. Such development will pave the way for future programmatic activities on software verification, simulation validation, and model uncertainty quantification on a scientific basis; currently, no satisfactory foundation exists for considering these activities. The exploitation of mathematical commonality for unit operations with the intent of developing generic and comprehensive software for nuclear reprocessing has not been considered to this author's knowledge. Past attention has been given to plant-level processes on an individual and isolated basis, which has led to various models and corresponding codes implementing non-systematic approaches based on elementary principles of chemical processing. This practice has built an initial knowledge base, but it was not fruitful in producing a lasting and extensible simulation capability. In contrast, a common, rigorous, mathematical modeling framework for all plant-level operations, as proposed here, has a tremendous practical and theoretical value because it will reduce the software implementation work, creates a well-defined modeling standard to compare past and future models, and, more importantly, opens the doors for scientific considerations of simulation fidelity; the latter has an obvious beneficial impact in supporting experimental validation programs. Therefore, the proposed framework is likely to generate a solid foundation for modeling plant-level processes for physicochemical nuclear (and non-nuclear) applications. Demonstration of concrete module implementation is the subject of future communications including prototypes for several modules, namely, voloxidation, dissolver, digester, accountability tank, and solvent extraction. Unit-operation commonality is a key aspect to be explored for a successful implementation of these modules aimed at realizing various flowsheets and plant configurations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Reactivity impact of accident tolerant claddings in an equilibrium PWR core

The US DoE's Accident Tolerant Fuels (ATF) program aims to establish a process to evaluate the feasibility of several ATF concepts. The two main aspects of such a process are to evaluate the 3-D 'whole core' neutronics and fuel (thermal-mechanical) performance. Determining the best ATF candidate usually involves a compromise between these two performance aspects. The work presented in this paper focuses solely on the 3-D neutronics, looking specifically at the end of cycle (EOC) reactivity, and upon the establishment of a realistic and modern equilibrium PWR core to use as a test-bed. Both the lattice physics code CASMO-4 and the 3-D nodal simulator NESTLE were used in this study. Eventually, operational power histories are provided to our collaborating team to carry out fuel performance evaluations to be presented in a future article. Neutronically, a goal of a prospective ATF concept is to at least match the fuel cycle length of a core with standard Zircaloy/UO{sub 2} fuel, without having to significantly change the design of the fuel bundle; meaning, without changing pin pitch or outer clad diameter. Such a 'seamless' transition to another cladding type would greatly reduce the burden of all the nuclear energy stakeholders (regulators, vendors, utilities), from both an analysis and financial perspective. For cladding materials that are less transparent to neutrons, which for this study included all the ATF concepts, the two degrees of freedom with respect to fuel fabrication are the U-235 enrichment and cladding thickness. The latter involved changing only the fuel radius and inner clad radius while keeping the gap and outer clad radius constant. The thinnest cladding value was set to a generous estimate of the minimum iron-chromium-aluminum (FeCrAl) thickness that can be manufactured. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

FUEL CONSUMPTION AND COST SAVINGS OF CLASS 8 HEAVY-DUTY TRUCKS POWERED BY NATURAL GAS

We compare the fuel consumption and greenhouse gas emissions of natural gas and diesel heavy-duty (HD) class 8 trucks under consistent simulated drive cycle conditions. Our study included both conventional and hybrid HD trucks operating with either natural gas or diesel engines, and we compare the resulting simulated fuel efficiencies, fuel costs, and payback periods. While trucks powered by natural gas engines have lower fuel economy, their CO2 emissions and costs are lower than comparable diesel trucks. Both diesel and natural gas powered hybrid trucks have significantly improved fuel economy, reasonable cost savings and payback time, and lower CO2 emissions under city driving conditions. However, under freeway-dominant driving conditions, the overall benefits of hybridization are considerably less. Based on payback period alone, non-hybrid natural gas trucks appear to be the most economic option for both urban and freeway driving environments.

Gao, Zhiming↗
DOE OSTI2016

Contaminant Entrainment in a Liquid Fuel Fire with Multi-Component Evaporation Droplet Model

A number of fire scenarios can result in the release of hazardous chemical or radiological contaminants as sub-micron solid particulates, posing a significant danger of inhalation. These particulates may be entrained in the air around the fire through a variety of mechanisms including liquid droplet formation above the boiling surface of the contaminated fuel. The formation of both film drops from the rupture of a liquid film as a boiling bubble bursts and jet drops from the collapse of the cavity of a ruptured bubble may be involved. A 2015 study by Brown and Louie focused on such a fire scenario to investigate the physics of contaminant release. This paper extends that study by supplementing the existing physics models with a multicomponent evaporation mechanism, where liquid droplets formed near the boiling fuel surface experience evaporation of their various components at different rates based on distinct material properties (density, specific heat, heat of vaporization, etc.). As in the Brown and Louie study, the Sierra low-Mach module Fuego is used as the Lagrangian-Eulerian code for simulation of this scenario. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Organic Vacuum Pump Fluids for the Vacuum Pumping of Fusion Power Plants

Vacuum pumps are the heart of the fusion fuel cycle, but most currently proposed pumping technologies are not capable of handling the required flow rates and vacuum pressures. Oil-containing vacuum pumps can readily meet the flow requirements, but vacuum fluids will degrade in fusion-relevant environments due to contact with tritium and exposure to high-energy radiation. Here, in this study, we describe a methodology to screen vacuum fluid candidates and purify these candidate fluids post oxidation, post exposure to deuterium, and post exposure to gamma radiation (<7.5 MGy) to simulate a process in which vacuum fluids can be recovered and regenerated during the fusion fuel cycle. A series of oils, including a highly purified mineral oil, phenyl silicone oil, and a polyphenyl ether, are shown to be suitable candidates for vacuum pumping. Additionally, we describe a simple purification methodology to remove oxidized functionalities and the associated isotopologues induced by contact with deuterium from the candidate vacuum fluids This purification methodology can also be applied to radiological damage with moderate effect. Finally, we demonstrate that the sorbents can be regenerated through electromagnetic microwave digestion.

oxidation↗
DOE OSTI2012

Addressing Facility Needs for Concrete Assessment Using Ultrasonic Testing: Mid-year Report

The UFD Gap Analysis to Support Extended Storage of Used Nuclear Fuel (June 30, 2011) emphasizes the need for the development of monitoring techniques and technologies for dry storage cask materials. A high priority is given to the development of 'systems for early detection of confinement boundary degradation.' This requires both new techniques for monitoring and inspection, as well as new measurable parameters to quantify mechanical degradation. The use of Nonlinear Elastic Wave Spectroscopy (NEWS) has been shown to provide sensitive parameters correlating to mechanical degradation in a wide variety of materials. Herein we report upon recent research performed to address the high priority of concrete degradation using a selection of these techniques and compare to a ASTM standard ultrasonic technique. Also reported are the near term plans to continue this research in the remaining FY and into the coming years. This research was conducted at Los Alamos National Laboratory (LANL) in the Acoustics Lab of the Geophysics group in the Earth and Environmental Sciences division, and in collaboration with the Laboratory for Nondestructive Evaluation at the University of the Mediterranean (Aix en Provence, France) and the Electrical Power Research Institute (EPRI). The objective of this research project was to determine the feasibility of using an NDE technique based on non-linear ultrasound for determining the depth and degree of microcracking in the near surface of concrete and to assess the degree of sensitivity of such technique. This objective is reached by the means of combining linear and nonlinear measurements, associated with numerical simulation. We first study the global effect of thermal damage on concrete's linear and nonlinear properties by resonance inspection techniques. We show that standard pulse wave speed techniques are not relevant to extract mechanical properties of concrete. The high sensitivity of measured nonlinearity is shown and serves as a validation tool for the rest of the study, i.e., probing the material nonlinearity at various depths through the use of Time Reversal Elastic Nonlinearity Diagnostic (TREND). The basic idea of probing the material nonlinearity at various depths by changing the frequency is validated by exhibiting a similar trend as nonlinear resonance measurements. We address at the end of this report, the potentialities of applying these procedures to real concrete structures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Mesoscale modeling of intergranular bubble percolation in nuclear fuels

Phase-field simulations are used to examine the variability of intergranular fission gas bubble growth and percolation on uranium dioxide grain boundaries on a mesoscopic length scale. Three key parameters are systematically varied in this study: the contact angle (or dihedral angle) defining the bubble shape, the initial bubble density on the grain boundary plane, and the ratio of the gas diffusivity on the grain boundary versus the grain interiors. The simulation results agree well with previous experimental data obtained for bubble densities and average bubble areas during coalescence events. Interestingly, the rate of percolation is found to be highly variable, with a large dependency on the contact angle and the initial bubble density and little-to-no dependency on the grain boundary gas diffusivity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Western Wind and Solar Integration Study Phase 2 (Presentation)

This presentation accompanies Phase 2 of the Western Wind and Solar Integration Study, a follow-on to Phase 1, which examined the operational impacts of high penetrations of variable renewable generation on the electric power system in the West and was one of the largest variable generation studies to date. High penetrations of variable generation can induce cycling of fossil-fueled generators. Cycling leads to wear-and-tear costs and changes in emissions. Phase 2 calculated these costs and emissions, and simulated grid operations for a year to investigate the detailed impact of variable generation on the fossil-fueled fleet. The presentation highlights the scope of the study and results.

14 SOLAR ENERGY↗
DOE OSTI2018

Large-eddy Simulations of Spray Variability Effects on Flow Variability in a Direct-injection Spark-ignition Engine Under Non-combusting Operating Conditions

Large-eddy Simulations (LES) have been carried out to investigate spray variability and its effect on cycle-to-cycle flow variability in a direct-injection spark-ignition (DISI) engine under non-reacting conditions. Initial simulations were performed of an injector in a constant volume spray chamber to validate the simulation spray set-up. Comparisons showed good agreement in global spray measures such as the penetration. Local mixing data and shot-to-shot variability were also compared using Rayleigh-scattering images and probability contours. The simulations were found to reasonably match the local mixing data and shot-to-shot variability using a random-seed perturbation methodology. After validation, the same spray set-up with only minor changes was used to simulate the same injector in an optically accessible DISI engine. Particle Image Velocimetry (PIV) measurements were used to quantify the flow velocity in a horizontal plane intersecting the spark plug gap. The engine was operated in a skip-fired operating mode and comparisons focused on cycles that included fuel injection, but no spark event and therefore no combustion. 105 total LES engine cycles were simulated using a parallel cycle simulation approach and 3 different perturbation methods in an attempt to isolate the effects of shot-to-shot spray variability and the initial turbulent flow field as well as their interaction effects on overall engine CCVs. The experimental mean and standard deviations were reasonably well matched by the simulations, though quantitative comparisons near the injection event during the intake stroke were difficult due to the high uncertainty in the PIV measurements at these crank angles. The 3 simulation perturbation methods resulted in very similar results, though further analysis found the current parallel cycle approach may be limiting the ability of the simulations to isolate the spray and flow effects.

Cycle-to-cycle Variability↗
DOE OSTI2020

Use of the Manuela 3D/Radiation Scanning Tool at Nuclear Facilities - 20437

Licensees, facility operators and decommissioning contractors have a need for comprehensive, accurate and precise radiological and topographical data to plan and execute safe and efficient Decontamination and Decommissioning (D and D) projects. Traditional approaches rely greatly upon labor and administratively intensive data collection methods that require considerable time and effort to glean the relevant data to plan and perform work in a radiological work area. Orano Decommissioning Services has developed and utilized at French nuclear facilities a lightweight portable scanning instrument called MANUELA (Mobile Apparatus for Nuclear Expertise Localization Assistance). Recently Orano conducted demonstrations at two United States nuclear facilities using MANUELA that may prove useful in areas of the fuel cycle including: - ALARA (As Low As Reasonably Achievable) planning during facility operations; - Scoping and site characterization during the D and D planning phase; - Remedial action support during active decommissioning work; - Radiation and modeling planning for plant modifications; - Support Final Status Survey and demonstration of 'end state' compliance for license termination. MANUELA is currently being evaluated for use during decommissioning of nuclear facilities and could aide in demonstrating compliance with end state criteria after decommissioning. Of particular interest are decommissioning projects planned for various sites across the United States. Most of these sites present unique challenges and opportunities to safely and efficiently decommission in a regulatory compliant manner. The two demonstration site's current conditions of relatively low radiation levels inside containment proved to be ideal platforms to deploy the MANUELA instrument for evaluation of the complex geometries typically found in a nuclear facility. The MANUELA instrument was able to precisely identify local hot spots of the reactor system's primary equipment and was able to simultaneously overlay the dose information onto the scanned 3D images. Using data like this, a licensee would be able to make informed technical and personnel decisions on methods to remediate, shield or properly dispose of the radioactive source term. MANUELA can be used throughout the decommissioning process to validate progress and continue to refine the work activities to optimize the efforts of the crew which reduces downtime waiting for survey results, whereas traditional survey methodology relies on a physical survey to be performed and the data transcribed onto survey maps that are only two dimensional. The traditional format tends to be inaccurate and difficult to interpret the precise location(s) of hot spots or generalized dose rates. The resultant scan also allows the user to know the exact extent of the survey versus transposing information onto a map with only a portion of the map area scanned. MANUELA also compiles all of the relevant survey data into a downloadable table format which includes the measurement identification number, time stamp, point space location on the XYZ plane, and the dose rate value. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Separations and Waste Forms Research and Development: FY 2012 Accomplishments Report

The Separations and Waste Form Campaign (SWFC) under the U.S. Department of Energy (DOE) Fuel Cycle Research and Development Program (FCRD) is responsible for developing advanced separation and waste form technologies to support the various fuel cycle options defined in the DOE Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. The fiscal year (FY) 2012 accomplishments report highlights the results of the research and development (R&D) efforts performed within SWFC in FY 2012. Each section contains a high level overview of the activities, results, technical point of contact and applicable references. This report briefly outlines campaign management and integration activities, but the intent of the report is to highlight the many technical accomplishments made during FY 2012. The campaign continued to maintain a balanced effort between scientific investigations to underpin advanced separations and waste forms technologies and engineering advancement to maintain relevance and focus on implementable technologies. There was an increased focus on development of new tools and methods to develop a better fundamental understanding of separation processes and waste form behavior and this effort is closely tied to modeling and simulation activities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Mechanistic Multiscale Uncertainty Propagation in Support of Accelerated Fuel Qualification

Taking a nuclear fuel concept through the research, development, and qualification stages has historically taken on the order of 20 to 25 years because of extensive irradiation tests required for a variety of conditions. The concept of accelerated fuel qualification (AFQ) has been proposed to increase the innovation pace for nuclear fuels. The goal of AFQ is not to replace the traditional qualification approach but rather to reduce the total number of experiments required to ensure approval from the regulatory authority. Of the many AFQ approaches being explored, advanced modeling—and, in particular, mechanistic modeling—is in a uniquely cross-cutting position to reduce the number of required integral tests through the inclusion of separate-effects testing, while helping to extrapolate reactor performance during rare events. We make the case that propagation of uncertainty through various computational length scales helps contextualize mechanistic modeling. We will utilize UO 2 fission gas diffusion predictions from the atomistically informed cluster dynamics code Centipede to inform fuel performance rodlet simulations using the BISON finite element code as the metric for showing how multiscale mechanistic uncertainty quantification can help reduce uncertainty in fuel performance. In conclusion, by quantifying uncertainty and its reduction through multiscale modeling, the qualification process may be accelerated through the reduction of costly irradiation experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Determination of the spent fuel decay heat with the VERA core simulator

VERA core simulator capabilities to predict spent fuel decay heat have been explored. The comparison with the Serpent2 and NEWT/TRITON sequence from the SCALE package, performed on the NPP Krško fuel test cases, has enabled independent verification of the successful implementation of the VERA depletion/decay calculation sequence applied on the extended nuclide set. Since the VERA is a 3-D core simulator, the focus in the rest of the paper was on the investigation of the 3-D effects and non-linearity of the soluble boron concentration averaging process, which are difficult to assess with most of available computational tools. A difference in the decay heat prediction between the 3-D and 2-D model approaches was analyzed. In addition, the impact of the Inconel grids on the decay heat was determined. Averaging the concentration of soluble boron over the fuel irradiation period has been found to be a viable approach, since it results in slight decay heat overprediction, which is considered conservative for most applications. Finally, demonstration on the Watts Bar Unit 1 cycle 1 depletion has shown formidable VERA capabilities to accurately predict spent fuel decay heat.

Kromar, Marjan↗
DOE OSTI2022

Uncertainty Quantification of Calculated Temperatures for the AGR 5/6/7 Experiment

This report documents the quantification of uncertainty of the calculated temperature data for the Advanced Gas Reactor (AGR) 5/6/7 fuel irradiation experiment conducted in the Advanced Test Reactor at Idaho National Laboratory in support of the Advanced Reactor Technologies? research and development program. Recognizing uncertainties inherent in physics and thermal simulations of the AGR 5/6/7 capsules, the results of the numerical simulations are used in combination with statistical analysis methods to improve qualification of measured data. The calculated fuel temperatures for AGR tests are also used for validation of the fission product transport and fuel performance simulation models. These crucial roles of the calculated fuel temperatures in ensuring achievement of the AGR experimental program objectives require accurate determination of the model temperature uncertainties. This report covers temperature uncertainty results for each of the five AGR 5/6/7 capsules. To quantify the uncertainty of calculated temperatures determined using the ABAQUS finite element heat transfer code, this study identifies and analyzes model parameters of potential importance to the calculated temperatures of fuel compacts and thermocouples. The selection of input parameters for uncertainty quantification is based on the ranking of their influences upon temperature predictions. Thus, selected input parameters include those with high sensitivity and those with the largest uncertainty. Propagation of model parameter uncertainty and sensitivity is then used to quantify the overall uncertainty of calculated temperatures. Measurement uncertainty, analysis of modeling assumptions, and expert judgment are used as the basis to quantify the uncertainty range for selected input parameters. The input uncertainties are dynamic, accounting for the effect of unplanned events and changes in thermal properties of capsule components over extended exposure to high temperatures and fast neutron irradiation. The sensitivity analysis performed in this work went beyond the traditional local sensitivity. Using experimental design, analysis of pairwise interactions of model parameters was performed to establish sufficiency of the time dependent first order (linear) expansion terms in constructing the temperature response surface. To achieve completeness, uncertainty propagation made use of pairwise noise correlations of model parameters. Furthermore, using an interpolation scheme over the input parameter domain, the analysis obtains time dependent sensitivity over the test campaign duration. This allows computation of uncertainty for the calculated fuel temperatures and the calculated graphite temperatures at thermocouple locations during the entire irradiation period.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Predictive Maturity of Multi-Scale Simulation Models for Fuel Performance

The project proposed to provide a Predictive Maturity Framework with its companion metrics that (1) introduce a formalized, quantitative means to communicate information between interested parties, (2) provide scientifically dependable means to claim completion of Validation and Uncertainty Quantification (VU) activities, and (3) guide the decision makers in the allocation of Nuclear Energy’s resources for code development and physical experiments. The project team proposed to develop this framework based on two complimentary criteria: (1) the extent of experimental evidence available for the calibration of simulation models and (2) the sophistication of the physics incorporated in simulation models. The proposed framework is capable of quantifying the interaction between the required number of physical experiments and degree of physics sophistication. The project team has developed this framework and implemented it with a multi-scale model for simulating creep of a core reactor cladding. The multi-scale model is composed of the viscoplastic self-consistent (VPSC) code at the meso-scale, which represents the visco-plastic behavior and changing properties of a highly anisotropic material and a Finite Element (FE) code at the macro-scale to represent the elastic behavior and apply the loading. The framework developed takes advantage of the transparency provided by partitioned analysis, where independent constituent codes are coupled in an iterative manner. This transparency allows model developers to better understand and remedy the source of biases and uncertainties, whether they stem from the constituents or the coupling interface by exploiting separate-effect experiments conducted within the constituent domain and integral-effect experiments conducted within the full-system domain. The project team has implemented this procedure with the multi- scale VPSC-FE model and demonstrated its ability to improve the predictive capability of the model. Within this framework, the project team has focused on optimizing resource allocation for improving numerical models through further code development and experimentation. Related to further code development, we have developed a code prioritization index (CPI) for coupled numerical models. CPI is implemented to effectively improve the predictive capability of the coupled model by increasing the sophistication of constituent codes. In relation to designing new experiments, we investigated the information gained by the addition of each new experiment used for calibration and bias correction of a simulation model. Additionally, the variability of ‘information gain’ through the design domain has been investigated in order to identify the experiment settings where maximum information gain occurs and thus guide the experimenters in the selection of the experiment settings. This idea was extended to evaluate the information gain from each experiment can be improved by intelligently selecting the experiments, leading to the development of the Batch Sequential Design (BSD) technique. Additionally, we evaluated the importance of sufficiently exploring the domain of applicability in experiment-based validation of high-consequence modeling and simulation by developing a new metric to quantify coverage. This metric has also been incorporated into the design of new experiments. Finally, we have proposed a data-aware calibration approach for the calibration of numerical models. This new method considers the complexity of a numerical model (the number of parameters to be calibrated, parameter uncertainty, and form of the model) and seeks to identify the number of experiments necessary to calibrate the model based on the level of sophistication of the physics. The final component in the project team’s work to improve model calibration and validation methods is the incorporation of robustness to non-probabilistic uncertainty in the input parameters. This is an improvement to model validation and uncertainty quantification stemming beyond the originally proposed scope of the project. We have introduced a new metric for incorporating the concept of robustness into experiment-based validation of numerical models. This project has accounted for the graduation of two Ph.D. students (Kendra Van Buren and Josh Hegenderfer) and two M.S. students (Matthew Egeberg and Parker Shields). One of the doctoral students is now working in the nuclear engineering field and the other one is a post-doctoral fellow at the Los Alamos National Laboratory. Additionally, two more Ph.D. students (Garrison Stevens and Tunc Kulaksiz) who are working towards graduation have been supported by this project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗