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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 541 records · Page 30
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Advanced Nodal P 3 /SP 3 Axial Transport Solvers for the MPACT 2D/1D Scheme

As part of its initiative to provide multiphysics simulations of nuclear reactor cores, the Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing the Virtual Environment for Reactor Applications Core Simulator (VERA-CS). The MPACT code, which is the primary neutron transport solver of VERA-CS, employs the two-dimensional/one-dimensional (2D/1D) method to solve 3-dimensional neutron transport problems and provide sub-pin-level resolution of the power distribution. While 2D method of characteristics is used to solve for the transport effects within each plane, 1D-nodal methods are used axially. There have been extensive studies of the 2D/1D method with a variety nodal methods, and the P 3 /SP 3 solver has proved to be an effective method of providing higher-fidelity solutions while maintaining a low computational burden.The current implementation in MPACT wraps a one-node nodal expansion method (NEM) kernel for each moment, iterating between them and performing multiple sweeps to resolve flux distributions. However, it has been observed that this approach is more sensitive to convergence problems. This paper documents the theory and application two new nodal P 3 /SP 3 approaches to be used within the 2D/1D method in MPACT. These two approaches aim to provide enhanced stability compared with the pre-existing one-node approach. Results from the HY-NEM-SP 3 solver show that the accuracy is consistent with the one-node formulations and provides improved convergence for some problems; but the solver has issues with cases in thin planes. Although the 2N-SENM-SP 3 solver is still under development, it is intended to resolve the issues with HY-NEM-SP 3 but it will incur some additional computational burden by necessitating an additional 1D-CMFD-P 3 solver to generate the second moment cell-averaged scalar flux.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Assessment of Pellet-Clad Interaction Indicators in Watts Bar Unit 1 using the VERA Framework

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity, multiphysics simulations of light water reactors (LWRs) by coupling a variety of codes within the Virtual Environment for Reactor Analysis (VERA). One of the primary goals of CASL is to predict local cladding failure through pellet-clad interaction (PCI). This capability is currently being pursued through several different approaches, such as with Tiamat, which is a simulation tool within VERA that more tightly couples the MPACT neutron transport solver, the CTF thermal hydraulics solver, and the MOOSE-based Bison-CASL fuel performance code. However, the process in this paper focuses on running fuel performance calculations with Bison-CASL to predict PCI using the multicycle output data from coupled neutron transport/thermal hydraulics simulations. In recent work within CASL, Watts Bar Unit 1 has been simulated over 12 cycles using the VERA core simulator capability based on MPACT and CTF. Using the output from these simulations, Bison-CASL results can be obtained without rerunning all 12 cycles, while providing some insight into PCI indicators. Multi-cycle Bison-CASL results are presented and compared against results from the FRAPCON fuel performance code. There are several quantities of interest in considering PCI and subsequent fuel rod failures, such as the clad hoop stress and maximum centerline fuel temperature, particularly as a function of time. Bison-CASL performs single-rod simulations using representative power and temperature distributions, providing high-resolution results for these and a number of other quantities. This will assist in identifying fuels rods as potential failure locations for use in further analyses.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

VERAIn

CASL's modeling and simulation technology, the Virtual Environment for Reactor Applications (VERA), incorporates coupled physics and science-based models, state-of-the-art numerical methods, modern computational science, integrated uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs), single-effect experiments, and integral tests. The computational simulation component of VERA is the VERA Core Simulator (VERA-CS). The core simulator is the specific collection of multi-physics computer codes used to model and deplete a LWR core over multiple cycles. The core simulator has a single common input file that drives all of the different physics codes. The parser code, VERAIn, converts VERA Input into an XML file that is used as input to different VERA codes.

Simunovic, Srdjan↗
DOE OSTI2016

VERA-CS Modeling and Simulation of PWR Main Steam Line Break Core Response to DNB

The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time step of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.

CTF↗
DOE OSTI2017

Improvement of the 2D/1D Method in MPACT Using the Sub-Plane Scheme

Oak Ridge National Laboratory and the University of Michigan are jointly developing the MPACTcode to be the primary neutron transport code for the Virtual Environment for Reactor Applications (VERA). To solve the transport equation, MPACT uses the 2D/1D method, which decomposes the problem into a stack of 2D planes that are then coupled with a 1D axial calculation. MPACT uses the Method of Characteristics for the 2D transport calculations and P3 for the 1D axial calculations, then accelerates the solution using the 3D Coarse mesh Finite Dierence (CMFD) method. Increasing the number of 2D MOC planes will increase the accuracy of the alculation, but will increase the computational burden of the calculations and can cause slow convergence or instability. To prevent these problems while maintaining accuracy, the sub-plane scheme has been implemented in MPACT. This method sub-divides the MOC planes into sub-planes, refining the 1D P3 and 3D CMFD calculations without increasing the number of 2D MOC planes. To test the sub-plane scheme, three of the VERA Progression Problems were selected: Problem 3, a single assembly problem; Problem 4, a 3x3 assembly problem with control rods and pyrex burnable poisons; and Problem 5, a quarter core problem. These three problems demonstrated that the sub-plane scheme can accurately produce intra-plane axial flux profiles that preserve the accuracy of the fine mesh solution. The eigenvalue dierences are negligibly small, and dierences in 3D power distributions are less than 0.1% for realistic axial meshes. Furthermore, the convergence behavior with the sub-plane scheme compares favorably with the conventional 2D/1D method, and the computational expense is decreased for all calculations due to the reduction in expensive MOC calculations.

Graham, Aaron M↗
DOE OSTI2017

Subplane-based Control Rod Decusping Techniques for the 2D/1D Method in MPACT

The MPACT transport code is being jointly developed by Oak Ridge National Laboratory and the University of Michigan to serve as the primary neutron transport code for the Virtual Environment for Reactor Applications Core Simulator. MPACT uses the 2D/1D method to solve the transport equation by decomposing the reactor model into a stack of 2D planes. A fine mesh flux distribution is calculated in each 2D plane using the Method of Characteristics (MOC), then the planes are coupled axially through a 1D NEM-P$_3$ calculation. This iterative calculation is then accelerated using the Coarse Mesh Finite Difference method. One problem that arises frequently when using the 2D/1D method is that of control rod cusping. This occurs when the tip of a control rod falls between the boundaries of an MOC plane, requiring that the rodded and unrodded regions be axially homogenized for the 2D MOC calculations. Performing a volume homogenization does not properly preserve the reaction rates, causing an error known as cusping. The most straightforward way of resolving this problem is by refining the axial mesh, but this can significantly increase the computational expense of the calculation. The other way of resolving the partially inserted rod is through the use of a decusping method. This paper presents new decusping methods implemented in MPACT that can dynamically correct the rod cusping behavior for a variety of problems.

Graham, Aaron M↗
DOE OSTI2017

Transactive Energy Simulation Platform

TESP combines existing domain simulators in the electric power grid, with new transactive agents, growth models and evaluation scripts. The existing domain simulators include GridLAB-D for the distribution grid and single-family residential buildings, MATPOWER for transmission and bulk generation, and EnergyPlus for large buildings. More are planned for subsequent versions of TESP. The new elements are: TEAgents - simulate market participants and transactive systems for market clearing. Some of this functionality was extracted from GridLAB-D and implemented in Python for customization by PNNL and others; Growth Model - a means for simulating system changes over a multiyear period, including both normal load growth and specific investment decisions. Customizable in Python code; and Evaluation Script - a means of evaluating different transactive systems through customizable post-processing in Python code. TESP provides a method for other researchers and vendors to design transactive systems, and test them in a virtual environment. It allows customization of the key components by modifying Python code.

Source record↗
DOE OSTI2016

VERAView

In order to support engineering analysis of Virtual Environment for Reactor Analysis (VERA) model results, the Consortium for Advanced Simulation of Light Water Reactors (CASL) needs a tool that provides visualizations of HDF5 files that adhere to the VERAOUT specification. VERAView provides an interactive graphical interface for the visualization and engineering analyses of output data from VERA. The Python-based software provides instantaneous 2D and 3D images, 1D plots, and alphanumeric data from VERA multi-physics simulations.

Lee, Ronald W.↗
DOE OSTI2017

Development and Assessment of CTF for Pin-resolved BWR Modeling

CTF is the modernized and improved version of the subchannel code, COBRA-TF. It has been adopted by the Consortium for Advanced Simulation for Light Water Reactors (CASL) for subchannel analysis applications and thermal hydraulic feedback calculations in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS). CTF is now jointly developed by Oak Ridge National Laboratory and North Carolina State University. Until now, CTF has been used for pressurized water reactor modeling and simulation in CASL, but in the future it will be extended to boiling water reactor designs. This required development activities to integrate the code into the VERA-CS workflow and to make it more ecient for full-core, pin resolved simulations. Additionally, there is a significant emphasis on producing high quality tools that follow a regimented software quality assurance plan in CASL. Part of this plan involves performing validation and verification assessments on the code that are easily repeatable and tied to specific code versions. This work has resulted in the CTF validation and verification matrix being expanded to include several two-phase flow experiments, including the General Electric 3 3 facility and the BWR Full-Size Fine Mesh Bundle Tests (BFBT). Comparisons with both experimental databases is reasonable, but the BFBT analysis reveals a tendency of CTF to overpredict void, especially in the slug flow regime. The execution of these tests is fully automated, analysis is documented in the CTF Validation and Verification manual, and the tests have become part of CASL continuous regression testing system. This paper will summarize these recent developments and some of the two-phase assessments that have been performed on CTF.

Salko, Robert K↗
DOE OSTI2017

SFU Hacking for Non-Hackers v. 1.005

The program provides a limited virtual environment for exploring some concepts of computer hacking. It simulates a simple computer system with intentional vulnerabilities, allowing the user to issue commands and observe their results. It does not affect the computer on which it runs.

Carter, David James↗
DOE OSTI2017

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

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

Stimpson, Shane G.↗
DOE OSTI2018

Watts Bar I Ex-Core Analyses using VERA

The integration and direct coupling of the Oak Ridge National Laboratory (ORNL) Shift Monte Carlo code within the Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) offers unique capabilities that combine high-fidelity in-core and ex-core radiation transport. In this recent development activity, the deterministic neutronics code MPACT performs the incore radiation transport with temperature feedback using COBRA-TF (CTF), and provides the fission source to Shift for a follow-on fixed source radiation transport calculation to determine ex-core quantities of interest. The coupling of MPACT to Shift allows the neutron source for the ex-core transport to be calculated from the time-dependent, fully coupled MPACT solution. This paper presents the first use of VERA to perform coupon irradiation for validation of its ex-core capabilities.

Davidson, Eva↗
DOE OSTI2018

VERA Infrastructure

The Virtual Environment for Reactor Applications (VERA) Infrastructure provides the infrastructure to build, configure, test, install, and package for release the VERAS suite of codes that is being developed by the Consortium for the Advanced Simulation of Light-water reactors (CASL). It includes the list of codes that can be built within VERA, the location on the configuration-controlled CASL development computer, the version number used in any particular software installation, the location of the third-party libraries used for testing VERA on a suite of the CASL testing and development computers, and the scripts to configure, build, test, install, and package for release a particular version of VERA. The VERA Infrastructure contains a list of scripts, configuration options, and utilities that are specific to CASL, but contain no physics or numeric, nuclear technology, or encryption technology.

Baird, Mark↗
DOE OSTI2019

Development of Molten Salt Reactor Modeling and Simulation Capabilities in VERA

In recent years, the nuclear community has seen a resurgence of interest in molten salt reactors (MSRs). Because of their molten, flowing fuel, MSRs present several new challenges not present in many other reactors, necessitating tightly coupled multiphysics modeling and simulation for accurate analysis. The Virtual Environment for Reactor Applications (VERA) has been developed to provide a high-fidelity multiphysics framework for reactor analysis. Though it was originally developed for light water reactor applications, VERA’s multiphysics capabilities make for a natural extension to MSR analysis. Recent efforts have focused on capturing the effects of moving fuel. A lumped depletion capability has been added that accounts for the moving and mixing of the fuel, as well as time spent outside the active core. A general multi-phase species transport module has also been added to track the movement and evolution of salt constituents as the salt flows through the core and around the primary loop. This paper presents a discussion of the new MSR capabilities in VERA, along with results for the lumped depletion and species transport capabilities for the Molten Salt Reactor Experiment, demonstrating VERA’s effectiveness in multiphysics simulations of MSRs.

Graham, Aaron↗
DOE OSTI2019

Inference of Crud Model Parameters from Plant Data

Over the past several years, the Virtual Environment forReactor Analysis, VERA, has provided a novel capability forthe Consortium for Advanced Simulation of Light Water Re-actors (CASL) to modelcrud-induced power shift (CIPS) inreactors operated by the Tennessee Valley Authority (TVA).While this developed capability in VERA is unique, multiplekey parameters in the crud physics have remained unknown.The methodology employed here provides the capability toinfer these key model parameters from measured plant data.Once the model parameters have been determined, a forwardVERA computation is performed and results are comparedto measured flux map data to determine the ecacy of thecalibrated crud model.

Collins, Benjamin↗
DOE OSTI2019

Economic Assessment of Nuclear Hybrid Energy Systems: Nuclear-Renewable-Water Integration in Arizona

One of the goals of the Nuclear-Renewable Hybrid Energy Systems (N-R HES) modeling and simulation (M&S) project is to assess the economic viability of N-R HES in a market that contains Variable Renewable Energy (VRE) sources like solar. The N-R HES M&S project is a multi-laboratory effort, led by Idaho National Laboratory (INL) and supported by Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL). Development of the N-R HES software framework was initiated at INL in 2016 [1] and has reached some level of maturity such that it can be applied to more than simple demonstration cases, e.g. real industry problems. Nevertheless, more capabilities are constantly added to accommodate the special needs of challenging, real-life problems. The N-R HES framework is built on the Risk Analysis Virtual ENvironment (RAVEN) code [2], which it uses as a driver and workflow manager for all calculations. There are four main cornerstones of the N-R HES simulation framework: 1) The generation of stochastic synthetic time series. 2) A set of probabilistic analysis and optimization algorithms available in RAVEN. 3) A set of models for representing the physical behavior of N-R HES. These can be reduced-order models or high-fidelity Modelica [3] models. 4) The RAVEN plug-in called CashFlow [4] that maps physical performance into economic performance. Within this framework, a broad spectrum of questions related to N-R HES can be addressed. One of the challenges currently of high interest is the increasing penetration of VRE altering the profile of the net demand (demand after removing all non-curtailable renewable energy sources, e.g. rooftop solar). The N-R HES software framework is capable of analyzing the potential feasibility of mitigating the resultant volatility in the net electricity demand. One possible solution to manage net demand volatility is adding stabilizing loads to the grid. These loads can be external industrial processes that are able to ramp production up and down quickly, or for regulated markets, adding variable internal loads that will effectively increase the baseload of the power plant and therefore reduce the risk of internal demand that is lower than the capacity of the power plant. The latter is a possible solution currently being considered by Arizona Public Service (APS), the operating owner of the Palo Verde (PV) Nuclear Power Plant (NPP).

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2018

Automatic Selection of High-Fidelity Models and Surrogates for Uncertainty Analysis

Surrogate models (SMs) with distinct features developed in the last decade provide an efficient substitute to the high-fidelity physical simulations. However, it is generally challenging to quantify the accuracy of surrogate models in a region of interest and examine reliability in the overall domain without additional system evaluations. This paper introduces a new scheme, automated adaptive model selection, to adaptively construct and quantify a set of surrogate models and automatically switch between the surrogates and a high-fidelity code. Cross-validation approaches, such as KFold, Leave One Out (LOO) and Leave P Out (LPO), are employed to assess the global accuracy of the surrogates. The switching strategy is based on local validation metrics, crowding distance, that are able to provide an assessment on the goodness of the prediction made by the surrogates. This decision metric decides whether the surrogates are good enough for a certain realization or whether the high-fidelity code needs to be executed. This new scheme has been implemented in Risk Analysis in Virtual ENvironment (RAVEN) software, which can be employed to accelerate the Probabilistic Risk Assessment (PRA) and Uncertainty Quantification (UQ) processes.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2022

Uncertainty Quantification in High-Low Dynamic System Coupling using RAVEN and TRANSFORM

This work demonstrates new functionality and applications stemming from the development of high-fidelity to low-fidelity (high-low) coupling for system simulations and to further explore the capabilities of the Risk Analysis Virtual Environment (RAVEN) in the performance of uncertainty quantification in this kind of high-low coupled system models. The work builds from previous work on high-low coupling that utilized COBRA-TF (CTF), the high-fidelity subchannel analysis code, with a low fidelity model built in ORNL’s TRANSFORM, the system analysis code, utilizing the Functional Mock-Up Interface (FMI). Steady-state and transient analysis examples using the high/low coupled models generated from CTF and TRANSFORM/FMI are investigated. The workflows for both steady-state and transient coupled simulations are described. A steady-state parameter sweep and uncertainty analysis of the primary flow rates and reactor power are demonstrated. Likewise, a transient pump trip and power ramp sensitivity studies are also demonstrated. This work elucidates some of the potential benefits and future needs of using RAVEN for high/low system coupling analysis of energy systems. It also shows some of the difficulties that can be encountered in coupling system simulations.

Williams, Wesley↗