Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Bison”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 181 records · Page 10

Efficient Failure Probability Calculations and Modeling Interface Debonding in TRISO Particles with BISON

This document constitutes completion of the NEAMS milestone, which is titled Demonstrate efficiency improvements in TRISO failure probability calculations and the effect of debonding on fission product transport in TRISO fuel particles in Bison . In this report we present the development of the following: (1) the anisotropic elasticity model of pyrolytic carbon and extended material models using the local coordinate system for aspherical particle geometry; (2) a capability of modeling interface debonding in TRISO particles; and (3) an efficient high-fidelity approach to calculate the failure probability of TRISO particles. Benchmark problems and AGR-2 tests are used to verify the models and demonstrate these new capabilities.

36 MATERIALS SCIENCE↗

Implement and Test 3D Mortar Contact in BISON

We leverage the extension of the generation of mortar segment meshes to three dimensions in MOOSE’s framework to extend thermomechanical modeling capabilities to problems with three dimensions. A modular approach to gap heat transfer physics using the mortar finite element method was created and documented, mechanical contact was extended to three dimensions—including frictional behavior, performance and ease of use were improved, and steps towards scalability of solid mechanics problems involving contact were taken. Many of these new developments are demonstrated in the simulation of 3D light-water reactor (LWR) problems, where the thermomechanical interface problem is solved using the mortar finite element method. Usage of the mortar framework has improved convergence in 2D problems and has enabled employing friction in 3D problems, of which we show results of a short, local stack of 3D pellets. Consequently, the benefits of mortar in terms of solution convergence and quality are extended to three dimensions. Section 2 discusses fundamental developments that enabled the simulation of practical mortar problems in three dimensions and other general improvements, including the reduction of the derivative container size, the modification of dual basis computations when edge dropping (lack of secondary element projection) takes place, the improvement of conditioning when employing the VCP in-edge dropping conditions, and code usability and quality improvements. These latter code enhancements include the migration of tests using “old” mortar contact constraints to using dual mortar with a semi-smooth Newton solution strategy and the reuse of lower dimensional domains for straightforwardly setting up a mortar thermomechanical LWR problem, i.e. the MOOSE action is employed for mechanical contact and the thermal LWR action is employed to capture the gas conductance, contact, and radiation components of gap heat transfer physics. Independently of the mortar LWR thermal action, we developed a modular approach to gap heat transfer that resides in MOOSE and can be leveraged, e.g., in metallic fuel problems. This approach, whose code design based on MOOSE’s user objects to model specific physics was proposed by the maintenance activity, is detailed in Section 3. Based on the dual mortar finite element method, the frictional contact constraints were extended to three dimensions. A block sheared in two directions in and out of contact with a rigid plane is employed in Section 4 to show the way the approach handles changes in frictional states (e.g. stick to slip) within a competitive number of Newton iterations. Equations and numerical results on the use of the VCP with Cartesian Lagrange multipliers, whose combination enables their direct condensation, are described in Section 5.3. Two-dimensional and three-dimensional BISON LWR simulations are discussed in Section 6. Particularly, a stack of five eccentric pellets with a surface defect is simulated and the effect of pellet-cladding friction is assessed. Finally, conclusions are outlined in Section 7.

42 ENGINEERING↗

Axial Gas Communication Modeling with BISON [Slides]

This record provides an overview presentation of the methodology used for axial gas communication implementation in BISON. It incorporates work from a previous LDRD coupled with ongoing efforts in a technology commercialization fund (TCF) project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Coupling RELAP5-3D to BISON

This report details two approaches for coupling the BISON nuclear fuel performance code with RELAP5-3D. Both approaches are shown to work well. The first is a Python-based approach, and the second combines the two applications into one executable with data passed in memory from one library to the other. This second coupling approach forms the basis for analysis of complex loss of coolant scenarios and enables future calculations of modeling uncertainties.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

TRU BISON Simulation Overview [Slides]

The BISON fuel performance code has been used to model minor actinide (MA) bearing metallic ternary fuel. This presentation gives an overview on the progress made so far on this effort, focusing on the modeling of EBR-II experiment X501-G591. Comparisons with PIE have also been made.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Model of U3Si2 Fuel System using BISON Fuel Code

This research considers the proposed advanced fuel system: U3Si2 combined with an advanced cladding. U3Si2 has a number of advantageous thermophysical properties, which motivate its use as an accident tolerant fuel. This preliminary model evaluates the behavior of U3Si2 using available thermophysical data to predict the cladding-fuel pellet temperature and stress using the fuel performance code: BISON. The preliminary results obtained from the U3Si2 fuel model describe the mechanism of Pellet-Clad Mechanical Interaction for this system while more extensive testing including creep testing of U3Si2 is planned for improved understanding of thermophysical properties for predicting fuel performance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

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 insertion accident (RIA) conditions. The purpose of this work is to model a baseline case of a 10 cm long UO2-Zircaloy fuel rodlet using BISON and RELAP5 over a range of energy depositions and with varying reactor power pulse widths. The results show 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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

SCANAIR-BISON BENCHMARK ON CIP0-1 RIA TEST

In the frame of their research programs on fuel safety, the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN) and the Idaho National Laboratory (INL) have developed respectively SCANAIR and BISON codes to describe the thermo-mechanical behavior of irradiated fuel rods during Reactivity Initiated Accidents (RIA) in LWR. A RIA, characterized by a very rapid increase of reactivity and power in some rods of the reactor, can be schematically represented in two main phases. First, the energy deposition leads to a rapid rise of the fuel temperature which induces thermal swelling of the fuel pellets. During this phase, the Pellet Clad Mechanical Interaction (PCMI) leads to clad deformation and potentially to failure depending on fuel enthalpy increase and on the level of clad embrittlement. After the PCMI phase, the increase of clad temperature can lead to the boiling crisis of the water surrounding the rod. During the film boiling phase, the clad to coolant heat transfer becomes very low and the clad can reach high temperature (>700°C). Depending on the internal gas pressure, the ductile clad can undergo large deformation and possible failure. The RIA codes have to predict properly these two phases of the transient in order to be used, in particular, to assess current fuel safety criteria but also to be able to predict the behavior of upcoming fuel (such as accident tolerant fuel). In this paper, codes computation comparison against experimental data obtained on CIP0-1 test is done for the first phase of the transient, the PCMI phase, in order to point out the strengths and weaknesses of the codes. The CIP0-1 experiment was performed with UO2 high burnup fuel in the CABRI sodium loop facility, at 280°C and low pressure (~3 bars) without boiling crisis and no failure of the fuel rod. The initial state of the rod calculated by irradiation codes are compared with measurements before the transient. Then global rod thermal behavior during the CIP0-1 test is assessed with the sodium coolant temperature measurements. The rod mechanical behavior is analyzed with the fuel and clad elongation and the clad residual hoop strain. Finally comparison of the fission gas behavior between codes is performed and compared to the measurements of the fission gas released in the experiment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

MULTIPHYSICS STEADY-STATE SIMULATION OF THE HIGH TEMPERATURE TEST REACTOR WITH MAMMOTH, BISON AND RELAP-7

In this work we present a new approach to solving a multiphysics steady-state simulation for the High Temperature Test Reactor (HTTR) using one Super Homogenization (SPH) corrected homogenized 3-D neutronics model in MAMMOTH, multiple heterogeneous 2-D axisymmetric fuel pin models in BISON and several 1-D annular flow thermohydraulics models in RELAP-7. The coupling of the different physics is complicated by the fundamentally disparate models, potentially affecting the stability of the global system. A pseudo-transient to converge the temperature fields is used to circumvent this difficulty. The benefit of doing so as opposed to having fully 3-D heterogenous models for all physics is that the computational burden for reactors as complex as HTTR may quickly become unbearable while coarse models may not capture all the essential phenomena of the problem. The driving purpose of this work is to establish a self-sustaining steady-state solution that can be used to start safety transients.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

BISON Methods and Contact Algorithms

This paper reviews five critical algorithms for BISON’s simulation of nuclear fuel: the burnup model, the coolant channel model, fuel cracking models, mechanical contact, and thermal contact.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

MULTIPHYSICS STEADY-STATE SIMULATION OF THE HIGH TEMPERATURE TEST REACTOR WITH MAMMOTH, BISON AND RELAP-7

In this work we present a new approach to solving a multiphysics steady-state simulation for the High Temperature Test Reactor (HTTR) using one Super Homogenization (SPH) corrected homogenized 3-D neutronics model in MAMMOTH, multiple heterogeneous 2-D axisymmetric fuel pin models in BISON and several 1-D annular flow thermohydraulics models in RELAP-7. The coupling of the different physics is complicated by the fundamentally disparate models, potentially affecting the stability of the global system. A pseudo-transient to converge the temperature fields is used to circumvent this difficulty. The benefit of doing so as opposed to having fully 3-D heterogenous models for all physics is that the computational burden for reactors as complex as HTTR may quickly become unbearable while coarse models may not capture all the essential phenomena of the problem. The driving purpose of this work is to establish a self-sustaining steady-state solution that can be used to start safety transients.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

MULTIPHYSICS STEADY-STATE SIMULATION OF THE HIGH TEMPERATURE TEST REACTOR WITH MAMMOTH, BISON AND RELAP-7

In this work we present a new approach to solving a multiphysics steady-state simulation for the High Temperature Test Reactor (HTTR) using one Super Homogenization (SPH) corrected homogenized 3-D neutronics model in MAMMOTH, multiple heterogeneous 2-D axisymmetric fuel pin models in BISON and several 1-D annular flow thermohydraulics models in RELAP-7. The coupling of the different physics is complicated by the fundamentally disparate models, potentially affecting the stability of the global system. A pseudo-transient to converge the temperature fields is used to circumvent this difficulty. The benefit of doing so as opposed to having fully 3-D heterogenous models for all physics is that the computational burden for reactors as complex as HTTR may quickly become unbearable while coarse models may not capture all the essential phenomena of the problem. The driving purpose of this work is to establish a self-sustaining steady-state solution that can be used to start safety transients.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

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 oc- curred 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 UO2/zirconium-based cladding fuel rods dur- ing normal operation [1]. In particular, qualitative attributes for materials with enhanced accident tolerance include im- proved reaction (e.g., oxidation) kinetics with steam, resulting in slower hydrogen (or other combustible gases) generation rate, while maintaining acceptable thermo-mechanical prop- erties, fuel-clad interactions, and fission-product behavior [1]. Through its Office of Nuclear Energy the United States De- partment 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 se- lection 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 impossi- ble 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 U3Si2. 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 to represent the current state of the art understanding of these materials. In this paper we demonstrate the importance of utilizing the latest state-of-the-art models for fuel performance simula- tions by comparing old and new thermal conductivity models for U3Si2 fuel and old and new oxidation models for FeCrAl cladding alloys using the BISON [2, 3] fuel performance code under normal operating conditions. The results indicate that the latest thermal conductivity model for U3Si2 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 ox- ide thickness (~6 µm) compared to the previous model which predicted zero oxide scale formation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Development of Accelerated Steady-state Test Capsule Experiments to Replicate EBR-II Fuel Behavior Using BISON Fuel Performance Analysis

Here in this work, BISON fuel performance calculations were performed to predict the fuel behavior of accelerated burnup U-Pu-Zr fuel, with temperature operation conditions of the fuel and the cladding mirroring conditions within EBR-II fuel pins. The temperature operating conditions within the FAST accelerated burnup rods were aimed at replicating EBR-II X447/X447A fuel surface and inner cladding surface temperatures. Due to the FAST capsule design, these temperatures can be replicated with fission rate densities being significantly increased. fuel performance modeling has not been assessed for novel experiments such as accelerated burnup utilizing the FAST capsule within ATR. This is an important step in understanding accelerated irradiation methods as many performance models are empirical models conforming to the results of PIE but do not always include physical models that would represent the changes in irradiation tests.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗