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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 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Acceleration of Thermochimica Calculations in Bison

The performance of the open-source thermochemical software library Thermochimica has been enhanced by developing a re-initialization algorithm to make use of data from previous calculations to reduce the number of convergence steps in the Gibbs energy minimization procedure. This algorithm has been tested in the context of stand-alone Thermochimica calculations, and speedups in the range of 2x-3x achieved for cases with chemistries resembling those of irradiated nuclear fuels. Routines to make use of this re-initialization procedure have been implemented in Bison, in which each node uses the results of the previous calculation at that node as the initial conditions for the following Thermochimica call. Examples based on calculating the diffusion of oxygen in UO$_2$ LWR fuel in 1D and 3D have been tested, with speedups up to 7.38x demonstrated.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Assessment of the BISON Metallic Fuel Performance Models

The US Department of Energy is leading a project to design and construct a fast spectrum test reactor called the Versatile Test Reactor (VTR). The BISON nuclear fuel performance code will be used to model VTR driver fuel, including looking at the effects of differences between the VTR driver fuel element design and the legacy fuel designs and experiments on which it is based. Simulations will be conducted to help determine whether the design’s behavior and performance are properly understood and to assess the margins to cladding failure and fuel melting relative to those predicted for past metallic fuel experiments. These predictions are expected to streamline VTR design and operation by helping inform the VTR driver fuel element design and by providing supplemental information for the fuel design safety basis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Development of bubble evolution model for new mechanistic transient fission gas release capability in BISON

This report summarizes efforts within NEAMS to investigate the mechanisms that govern fission gas behavior in UO 2 . In particular, the focus is on understanding how fission gas behavior causes transient fission gas release and fragmentation/pulverization of high burnup structure (HBS) in UO 2 . HBS forms in the periphery of the pellet where temperatures are relatively low. Previously, MD simulations were performed to determine the reaction energies for various Xe and U defects with bubbles, as a function of Xe to vacancy ratio or, equivalently, pressure. As had been shown in FY22, it was found that the unmodified version of the Simple Integrated Fission Gas Release and Swelling (SIFGRS) model within BISON greatly over-predicted the number gas atoms per vacancy in the bubbles in the outer rim of the pellet (a ratio of > 1 million). This was due to slow grain boundary vacancy diffusivity and not accounting for the pressure-dependent reaction energy for Xe interstitials with bubbles. The application of the pressure dependent reaction energies was able to restrict Xe to vacancy ratios to 2:1, which is far more realistic than those originally obtained from SIFGRS.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Development of Mechanistic Fission Gas Release and Swelling Models for UN Fuels in BISON

This report describes the work in NEAMS (Nuclear Energy Advanced Modeling and Simulation) to develop a mechanistic fission gas model for uranium nitride fuels in BISON. The existing Sifgrs (Simple integrated fission gas release and swelling) model tracks the average properties of two bubble populations in the bulk and at the grain boundaries. It was recognized that dislocations play a crucial role in the fission gas swelling of UN, and an irradiation-induced dislocation density model was needed, as well as a model describing how fission gas interacts with dislocations creating a third population of bubbles along dislocations. A mechanistic model that tracks dislocation bubbles was implemented in Sifgrs. This model was used to simulate fission gas swelling and release in UN. Lower-length-scale calculations and experimental observations from carbide fuel were leveraged to help populate the model with essential parameters. As a placeholder, an empirical function was formulated for the evolution of the dislocation network. Because of the difference in evolution of this network at different temperatures, the dislocation bubbles are able to capture behavior that the bulk intragranular bubbles cannot. It was found that the bulk bubbles dominate microscopic swelling at low temperatures, and the dislocation bubbles dominate at higher temperatures. Based on this model, the transition between the two bubble types is the main factor behind the breakaway swelling phenomenon in UN. In order to test the model, a couple of assessments were run. For the lower temperature JOYO pins, the model produces reasonable fission gas release and swelling values. For the higher temperature SP1 pin, the model dramatically underestimates the fission gas release. This issue is attributed to the gas being trapped inside dislocation bubbles, unable to escape to grain boundaries to cause fission gas release, and could be remedied by a mechanistic dislocation model that allows the dislocation density to decrease at very high temperatures. A preliminary mechanistic dislocation model was developed supported by first-principle calculations. These calculations provided valuable insight into how interstitial defects cluster in UN in the {110} orientation, and may eventually form dislocation loops, leading to the conclusion that dislocation loops may nucleate from these clusters. An estimate of the dislocation line energy in UN was obtained and will be improved in future work. The free-energy cluster dynamics code Centipede was used to track interstitial and vacancy absorption at dislocation loops and calculate their growth. In addition, improvements to Centipede were made including new convergence criteria for transient simulations and sink driving force updates.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Use and Assessment of BISON within VERA-CS in Support of PCI Challenge Problem

Successfully addressing the PCI Challenge Problem in CASL could provide significant economic benefit to industry by allowing them to gain better insight into their systems and avoiding economic penalties incurred by possible fuel failures or risk mitigation strategies such as slow power ramp rates during transients. This effort requires integrating BISON into VERA-CS effectively and efficiently, demonstrating its use and continuing to further develop the capabilities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

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

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.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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 OSTI2019

Modeling fuel assembly distortion in light water reactors using BISON

Individual fuel rods in a light water reactor fuel assembly (FA) often have different power profiles that result in a spatial temperature gradient in the FA. This temperature gradient causes some rods to elongate more than others which could in turn distort the entire FA. Such FA distortion has been known to cause operational difficulties with refueling and with the insertion of control rods [1]. Numerical modeling of FA distortion due to differential elongation of fuel rods requires: (i) modeling the thermal elongation as well as the creep and irradiation induced elongation of the fuel rods, and (ii) modeling the bending response of the FA due to the differential elongation of the fuel rods. Past studies on this topic have been focused on the mechanical deformation of the FA with simplistic modeling of fuel rod elongation. In this paper, a simulation approach is presented that directly couples 2D axisymmetric fuel performance models of the individual fuel rods in a FA with structural beam models of the fuel rods in a FA using the Bison code [2]. This allows all of the mechanisms internal to the fuel rod leading to cladding elongation to be accounted for in the FA distortion model. To demonstrate this capability, the bowing of a simplified and idealized 6x6 FA (Fig. 1a), modeled using beam elements, is analyzed under different spatially varying power profiles. For this study, the guide tubes are rigidly connected to the spacer grid. Also, the fuel rods are rigidly connected to the spacer grid in all directions except the vertical direction (Y) along which slipping can occur between the fuel rods and spacer grids, depending on the friction coefficient between them. Coupling the thermo-mechanical response of individual fuel rods to the global mechanical behavior of the FA can shed light on the mechanisms that cause distortion of the FA and potential solutions to address this issue.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Pressurized Water Reactor Control Rod Ejection Analysis Using PARCS, RELAP5-3D, and BISON for High Burnup Fuel

This paper describes safety assessment of high burnup fuel (HBF) during loss of coolant accidents (LOCA), reactivity-initiated accidents (RIA), and anticipated operational occurrences (AOOs) for Southern Nuclear’s Vogtle Electric Generating plant (VEGP). A goal of this project is to create a multiphysics modeling approach in aspects of thermal hydraulics, neutronics, and fuel performance, as well as to use it in combination with experimental studies (Figure 1). In support of the project, this paper seeks to address fuel performance of a HBF pin during steady state conditions and during a control rod ejection (CRE) scenario using a loose coupling of PARCS, RELAP5-3D, and BISON.. The primary goal of this work was to analyze high burnup fuel behavior during a CRE scenario. The results indicate fuel failure is not expected to occur under the conditions presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

BISON modeling of reactivity-initiated accident experiments in a static environment

In conjunction with the restart of the TREAT reactor and the design of test vehicles, modeling and simulation efforts are being used to model the response of Accident Tolerant Fuel (ATF) concepts under reactivity-initiated accident (RIA) conditions. The purpose of this work is to model a baseline case of a 10 cm long UO{sub 2}- Zircaloy fuel rodlet in the Multi-SERTTA capsule in PWR conditions, using BISON and RELAP5 over a range of energy depositions and with varying reactor power pulse widths. The results provide an initial best estimate of fuel performance modeling during a RIA event in TREAT with a focus on the effect of varying the pulse width and energy deposition on both thermal and mechanical parameters that are important for predicting failure of the fuel rodlet. The combined BISON/RELAP5 model captures coupled thermal and mechanical effects on the fuel-to-cladding gap conductance, cladding-to-coolant heat transfer coefficient and water temperature and pressure that would not be capable in each code individually. These combined effects allow for a more accurate modeling of the thermal and mechanical response in the fuel rodlet and thermal-hydraulics of the test vehicle. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Time-Dependent Sensitivity Analysis of OECD Benchmark using BISON and RAVEN

Developments in uncertainty quantification for nuclear simulations have decreased the computational cost required to perform accurate sensitivity analysis. This work demonstrates implementation of these methods in the RAVEN framework, allowing for time-dependent sensitivity analysis. By demonstration we consider an OECD benchmark case. We propagate uncertainties using RAVEN operating on the BISON fuels performance code. We then consider the time-evolution of output response sensitivity to uncertain inputs. We perform sensitivity analysis using time-based stochastic collocation for generalized polynomial chaos (SCgPC) and high-dimension model reduction (HDMR). While uncertainty quantification applied to end-of-cycle statistics is useful, we demonstrate that dominant sources of uncertainty for figures of merit may fluctuate throughout the simulation. Time-dependent sensitivity analysis allows more insight on key design parameters for fuel. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Improvements to Modeling Capabilities of ATF Concepts in the BISON Fuel Performance Code

Research into accident tolerant materials for the fuel and cladding in Light Water Reactors has been at the forefront of the nuclear fuel research community since the events that occurred at the Fukushima Daiichi nuclear power plant in March 2011. Accident tolerant materials are defined as those that provide significantly increased response time in the event of an accident while providing similar or improved performance as the conventional UO{sub 2}/zirconium-based cladding fuel rods during normal operation. In particular, qualitative attributes for materials with enhanced accident tolerance include improved reaction (e.g., oxidation) kinetics with steam, resulting in lower hydrogen (or other combustible gases) generation rate, while maintaining acceptable thermo-mechanical properties, fuel-clad interactions, and fission-product behavior. Through it's Office of Nuclear Energy, the United States Department of Energy (U.S. DOE) has accelerated research in this area through the Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC). The goal of the AFC Accident Tolerant Fuel (ATF) program is to guide selection of promising concepts for insertion into a commercial reactor as part of a lead test rod or assembly in 2022. Given the aggressive development schedule, it is impossible to perform a comprehensive set of experiments to provide material characterization data. Therefore, the AFC is utilizing computational analysis tools in an effort to understand the proposed accident tolerant materials. The two materials of interest in this work are iron-chromium-aluminum (FeCrAl) cladding and U{sub 3}Si{sub 2}. From a fuel performance standpoint, the correlations available to describe the behavior of these two materials is limited. Therefore, as additional experimental data becomes available and correlations are updated, the fuel performance models must be updated as well to assure that the material models reflect the latest experimental findings. In this paper we demonstrate the importance of utilizing the latest state-of-the-art models for fuel performance simulations by comparing the change in the updated thermal conductivity model for U{sub 3}Si{sub 2} fuel and oxidation model for FeCrAl cladding alloys using the BISON fuel performance code under normal operating conditions. The results indicate that the latest thermal conductivity model for U{sub 3}Si{sub 2} results in fuel centerline temperatures that are on average approximately 3% higher than the previous model. For the oxidation behavior of FeCrAl alloys the new model predicts observable but small oxide thickness (6 μm) compared to the previous model which predicted essentially zero oxide scale formation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Development of mechanistic, microstructure-informed BISON models for fission product-induced failure mechanisms in advanced nuclear fuels

The US continues to prioritize commercial deployment of advanced reactors—particularly those utilizing U-Zr metallic and TRISO particle fuels. Fission product-induced failure mechanisms for these systems include fuel–cladding chemical interaction in metallic fuel rods, which threatens cladding integrity, and Pd penetration in TRISO particles, which degrades SiC layer properties. Empirical models can be applied within the bounds of existing irradiation databases, but their utility is limited when considering new designs or investigating fundamental material behaviors. The Nuclear Energy Advanced Modeling and Simulation Program is therefore developing mechanistic models for these behaviors, which are expected to aid in fuel design, assist in development of failure mitigation strategies, and provide support for qualification and licensing. These efforts leverage multiscale capabilities to capture the microstructural features and processes that govern these behaviors. This talk summarizes past and ongoing efforts to develop and validate these models for the BISON fuel performance code.

FCCI↗
DOE OSTI2025

Serpent - Bison - THM Preliminary Multiphysics Modeling of a Nuclear Thermal Propulsion System Fuel Assembly

This work demonstrates the Monte Carlo neutronic and Thermo-Hydraulic coupling scheme using the Serpent code and MOOSE application Bison and Thermo Hydraulic Module. The coupling scheme is then applied to the reference BWX Technologies Nuclear Ther- mal Propulsion system at he fuel assembly level where it’s used to perform an analysis of the isothermal material coefficients and potential material reactivity worth. A method is developed to isolate which feedback effects should be considered for proceeding with reduced order deterministic neutronic modeling where branch off analysis must be con- ducted. The convergence behavior of the coupling scheme is demonstrated where it fol- lows the standard Picard iteration approach. Verification studies for the method of deduc- ing relevant feedback effects is also demonstrated.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Impact of temperature variations on BISON predictions of Ag release in AGR-1 and AGR-2 experiments

Understanding and quantifying the release of fission products like silver (Ag) from TRistructural ISOtropic (TRISO) fuel particles is important to assess the safe operation of advanced high temperature reactors. Although the silicon carbide (SiC) layer of TRISO particles is effective as the main fission product barrier, Ag can be released from intact TRISO particles. A mechanistic model for the effective Ag diffusivity, D eff , was previously developed as a function of temperature and microstructure variables informed by atomistic modeling of Ag diffusivity on the mesoscale. Here in this study, we use this model to explore how experimental temperature uncertainties impact the overall predicted Ag release. This analysis shows that temperature uncertainties have a significant impact on the overall Ag release predictions. Furthermore, we show that the time average volume average temperature (TAVA) temperature is not an appropriate proxy for temperature histories to predict fission product release. We attribute this to the Arrhenius dependence of Ag diffusivity with respect to temperature. The detailed temperature histories, therefore, provide the most accurate results are are of most importance for modeling efforts. Overall, this work shows the importance of considering the experimental uncertainty of the temperature on computational predictions of fission product transport and release and the need for more accurate temperature histories from future experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Early Implementation of SiC Cladding Fuel Performance Models in BISON

SiC-based ceramic matrix composites (CMCs) [5–8] are being developed and evaluated internationally as potential LWR cladding options. These development activities include interests within both the DOE-NE LWR Sustainability (LWRS) Program and the DOE-NE Advanced Fuels Campaign. The LWRS Program considers SiC ceramic matrix composites (CMCs) as offering potentially revolutionary gains as a cladding material, with possible benefits including more efficient normal operating conditions and higher safety margins under accident conditions [9]. Within the Advanced Fuels Campaign, SiC-based composites are a candidate ATF cladding material that could achieve several goals, such as reducing the rates of heat and hydrogen generation due to lower cladding oxidation rates in HT steam [10]. This work focuses on the application of SiC cladding as an ATF cladding material in PWRs, but these work efforts also support the general development and assessment of SiC as an LWR cladding material in a much broader sense.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Transient Multiphysics with MAMMOTH and Bison

This is a poster for the Young Research Association poster presentation happening July 19 in EROB. The poster is about my PhD research and my internship project.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗