Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “core simulation”

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
All sourcesNASA NTRSDOE OSTI
NASA NTRS2013

Reactor Simulator Testing Overview

OBJECTIVE: Integrated testing of the TDU components TESTING SUMMARY: a) Verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. b) Thermal test heat regeneration design aspect of a cold trap purification filter. c) Pump performance test at pump voltages up to 150 V (targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V). TESTING HIGHLIGHTS: a) Gas and vacuum ground support test equipment performed effectively for NaK fill, loop pressurization, and NaK drain operations. b) Instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. c) Cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. d) ALIP produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

Schoenfeld, Michael P.↗
DOE OSTI2020

Confining Liquids inside Carbon Nanotubes: Accelerated Molecular Dynamics with Spliced, Soft-Core Potentials and Simulated Annealing

Understanding emergent phenomena of fluids under physical confinement requires the development of advanced tools for rapid and accurate simulation of their physiochemical properties. Simulating liquid molecules commensurate in size with the nanoscale enclosures that confine them is a key challenge. In this work, we demonstrate an accelerated molecular dynamics simulation technique that combines soft-core potentials (SCP) and simulated annealing (SA) to analyze confined liquids. This integrated SCP/SA method relies on a new spliced soft-core potential (SSCP), which enables tunable accuracy with respect to the target hard-core potential (HCP). SCP/SA enables the packing of enclosures with bulk material in a controlled, thermodynamically consistent manner. The enhanced SSCP accuracy is a critical feature of SCP/SA, enabling a smooth transition between the SCP and the HCP at a desired SCP hardness. We applied SCP/SA to the problem of filling a carbon nanotube (CNT) in periodic boundary conditions with a popular ionic liquid (IL), 1- butyl-3-methylimidazolium hexafluorophosphate $[\text{BMIM}^+][\text{PF}^–_6]$. We performed a series of triplicate simulations on systems with varying CNT diameter and charge to demonstrate SCP/SA’s versatility. Beyond this IL/CNT system, the SCP/SA simulation framework has a broad range of potential applications, not limited to nanoscale enclosures and interfaces, including both solid-state and biological systems.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Transition Core Modeling for Extended Enrichment & Accident-Tolerant Fuels Using Polaris/PARCS

Commercial light water reactor (LWR) operators and fuel vendors are currently interested in increasing the low-enriched uranium (LEU) fuel enrichments from the current limit of 5.0 $^w/_o$ $^{235}U$ up to 10 $^ w/_o$ $^{235}U$ (referred to as "LEU+") in their current fleets; they are also interested in using accident-tolerant fuel (ATF) with both LEU and LEU+ fuel. This report aims to identify modeling challenges and accuracy concerns in transition core analysis using the SCALE Polaris lattice physics code and U.S. Nuclear Regulatory Commission core simulator PARCS. At the time this study was started, no publicly available LEU+ core designs existed for boiling water reactor (BWR) or pressurized water reactor (PWR) systems. Therefore, fuel lattices were shuffled within a multi-assembly model to mimic neutronically challenging lattice combinations seen in transition cores, such as a fresh LEU+ lattice next to depleted LEU lattices. In addition to multi-assembly models, whole-core BWR transition core calculations were performed for ATF and LEU+ fuel using an existing Hatch-1 Cycle 3 core model. A whole-core BWR model was chosen due to the more heterogeneous core designs compared to those for a PWR core. Since the original core is an old checkerboard core design and no core or fuel design optimization was performed for the modeled fuel types, these core calculations were intended only to provide: (1) Comparisons of core characteristics of interest, such as the pin power distributions and peaking factors, Doppler temperature coefficients (DTCs), and control blade worths (CBWs) under challenging core designs, (2) Identification of reactor physics challenges in modeling LEU+ and ATF cores, and (3) A stress test for the Polaris/PARCS two-step modeling approach, including characterization of the relative accuracy for predicting characteristics of interest such as pin power distributions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2016

Fission Surface Power Technology Demonstration Unit Test Results

The Fission Surface Power (FSP) Technology Demonstration Unit (TDU) is a system-level demonstration of fission power technology intended for use on manned missions to Mars. The Baseline FSP systems consists of a 190 kWt UO2 fast-spectrum reactor cooled by a primary pumped liquid metal loop. This liquid metal loop transfers heat to two intermediate liquid metal loops designed to isolate fission products in the primary loop from the balance of plant. The intermediate liquid metal loops transfer heat to four Stirling Power Conversion Units (PCU), each of which produce 12 kWe (48 kW total) and reject waste heat to two pumped water loops, which transfer the waste heat to titanium-water heat pipe radiators. The FSP TDU simulates a single leg of the baseline FSP system using an electrically heater core simulator, a single liquid metal loop, a single PCU, and a pumped water loop which rejects the waste heat to a Facility Cooling System (FCS). When operated at the nominal operating conditions (modified for low liquid metal flow) during TDU testing the PCU produced 8.9 kW of power at an efficiency of 21.7 percent resulting in a net system power of 8.1 kW and a system level efficiency of 17.2 percent. The reduction in PCU power from levels seen during electrically heated testing is the result of insufficient heat transfer from the NaK heater head to the Stirling acceptor, which could not be tested at Sunpower prior to delivery to the NASA Glenn Research Center (GRC). The maximum PCU power of 10.4 kW was achieved at the maximum liquid metal temperature of 875 K, minimum water temperature of 350 K, 1.1 kg/s liquid metal flow, 0.39 kg/s water flow, and 15.0 mm amplitude at an efficiency of 23.3 percent. This resulted in a system net power of 9.7 kW and a system efficiency of 18.7 percent.

Facility Cooling System↗
NASA NTRS2016

Fission Surface Power Technology Demonstration Unit Test Results

The Fission Surface Power (FSP) Technology Demonstration Unit (TDU) is a system-level demonstration of fission power technology intended for use on manned missions to Mars. The Baseline FSP systems consists of a 190 kWt UO2 fast-spectrum reactor cooled by a primary pumped liquid metal loop. This liquid metal loop transfers heat to two intermediate liquid metal loops designed to isolate fission products in the primary loop from the balance of plant. The intermediate liquid metal loops transfer heat to four Stirling Power Conversion Units (PCU), each of which produce 12 kWe (48 kW total) and reject waste heat to two pumped water loops, which transfer the waste heat to titanium-water heat pipe radiators. The FSP TDU simulates a single leg of the baseline FSP system using an electrically heater core simulator, a single liquid metal loop, a single PCU, and a pumped water loop which rejects the waste heat to a Facility Cooling System (FCS). When operated at the nominal operating conditions (modified for low liquid metal flow) during TDU testing the PCU produced 8.9 kW of power at an efficiency of 21.7% resulting in a net system power of 8.1 kW and a system level efficiency of 17.2%. The reduction in PCU power from levels seen during electrically heated testing is the result of insufficient heat transfer from the NaK heater head to the Stirling acceptor, which could not be tested at Sunpower prior to delivery to GRC. The maximum PCU power of 10.4 kW was achieved at the maximum liquid metal temperature of 875 K, minimum water temperature of 350 K, 1.1 kg/s liquid metal flow, 0.39 kg/s water flow, and 15.0 mm amplitude at an efficiency of 23.3%. This resulted in a system net power of 9.7 kW and a system efficiency of 18.7 %.

fission power technology↗
DOE OSTI2022

CTF Improved Drag Model and Flow Regime Transition Criteria

The demand for accurate prediction of two-phase flow behavior in a boiling water reactor (BWR) requires a comprehensive understanding of flow regime, void fraction, heat transfer, and pressure drop. The CTF subchannel code, which is used for the Thermal/Hydraulic (T/H) solution in the Consortium for Advanced Simulation of Light Water Reactors (CASL)-developed Virtual Environment for Reactor Application (VERA) core simulator, is being further developed for BWR applications. In support of this goal, the present work highlights some of the two-phase closure model developments towards improving the CTF void fraction prediction, especially for subcooled boiling. The drift-flux approach has been well-developed for upward dispersed two-phase flows and proven to be accurate in predicting void fraction in bubbly and slug flow regimes. In this work, these kinematic constitutive relations for the drift-flux velocity have been implemented into CTF to describe the interfacial drag of bubbly flow as an alternative to the existing model for better void fraction prediction. The success of these constitutive relations also relies on a good flow regime map that accounts for flow conditions and channel geometry. A more reliable flow regime transition criteria that account for the flow condition has also been implemented in this study for modeling the flow regime transition criteria. The newly implemented models are shown to give improved void fraction predictions in comparison to experimental data.

Hizoum, Belgacem↗
DOE OSTI2020

Coupled fuel performance calculations in VERA and demonstration on Watts Bar unit 1, cycle 1

As the core simulator capabilities in the Virtual Environment for Reactor Applications (VERA) have become more mature and stable, increased attention has been focused on coupling Bison to provide fuel performance simulations. This technique has been a very important driver for the pellet-clad interaction challenge problem being addressed by the Consortium for Advanced Simulation of Light Water Reactors (CASL). Here, two coupling approaches are demonstrated on quarter core problems based on Watts Bar Nuclear Plant Unit 1, Cycle 1: (1) an inline approach in which a one-way coupling is used between neutronics/thermal hydraulics (through MPACT/CTF) and fuel performance (through Bison) but no fuel temperature information is passed back to MPACT/CTF, and (2) a two-way approach (coupled) in which the fuel temperature is passed from Bison to MPACT/CTF. In both approaches, power and temperature distributions from MPACT/CTF are used to inform the Bison simulations for each rod in the core.The demonstrations presented here are the first integrated fuel performance simulations in VERA, which opens many possibilities for future work, including applications to accident-tolerant fuel efforts and transient simulations, which are of critical importance to CASL. These demonstrations also highlight the potential to move away from the current Bison-informed fuel temperature lookup table approach, which is the default in MPACT/CTF simulations, if performance improvements are made in the near future.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

MPACT 4.4 Theory Manual

MPACT is a three-dimensional (3D) full-core neutron transport code capable of calculating subpin power distributions. Calculations are based on the Boltzmann transport equation for neutron fluxes for problems in which the detailed geometrical configuration of fuel components such as the pellet and cladding are explicitly retained. The cross-section data needed for the neutron transport calculation are obtained directly from a multigroup cross section library, which has traditionally been used by lattice physics codes to generate few-group homogenized cross sections for nodal core simulators. Hence, MPACT assumes neither a priori homogenization nor group condensation for the full core spatial solution. The 3D MPACT transport solution can be obtained using the method of characteristics (MOC), which employs discrete ray tracing within each fuel pin. However, for practical reactor applications, the direct application of MOC to 3D core configurations requires an excessive amount of memory and computing time due to the very large number of rays. For practical 3D full-core calculations, MPACT commonly uses an approximate “2D/1D” method that treats the radial (x and y) variables differently from the axial (z) variable. In particular, the radial dependence of the solution is calculated using transport theory, and the axial dependence is calculated using diffusion or P 3 theory. The 2D/1D method requires the core to be divided into a vertical stack of axial slices with a thickness of Δ z ≈ 5–10 cm. Each axial slice is divided radially into coarse spatial cells with boundaries that usually constitute the pin cell boundaries, for which Δ x = Δ y ≈ 1.5 cm. Then, each coarse radial cell (pin cell) is divided into 50–100 fine radial cells, which resolve the angular flux in the fuel, cladding, and moderator regions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2000

Numerical Propulsion System Simulation: An Overview

The cost of implementing new technology in aerospace propulsion systems is becoming prohibitively expensive and time consuming. One of the main contributors to the high cost and lengthy time is the need to perform many large-scale hardware tests and the inability to integrate all appropriate subsystems early in the design process. The NASA Glenn Research Center is developing the technologies required to enable simulations of full aerospace propulsion systems in sufficient detail to resolve critical design issues early in the design process before hardware is built. This concept, called the Numerical Propulsion System Simulation (NPSS), is focused on the integration of multiple disciplines such as aerodynamics, structures and heat transfer with computing and communication technologies to capture complex physical processes in a timely and cost-effective manner. The vision for NPSS, as illustrated, is to be a "numerical test cell" that enables full engine simulation overnight on cost-effective computing platforms. There are several key elements within NPSS that are required to achieve this capability: 1) clear data interfaces through the development and/or use of data exchange standards, 2) modular and flexible program construction through the use of object-oriented programming, 3) integrated multiple fidelity analysis (zooming) techniques that capture the appropriate physics at the appropriate fidelity for the engine systems, 4) multidisciplinary coupling techniques and finally 5) high performance parallel and distributed computing. The current state of development in these five area focuses on air breathing gas turbine engines and is reported in this paper. However, many of the technologies are generic and can be readily applied to rocket based systems and combined cycles currently being considered for low-cost access-to-space applications. Recent accomplishments include: (1) the development of an industry-standard engine cycle analysis program and plug 'n play architecture, called NPSS Version 1, (2) A full engine simulation that combines a 3D low-pressure subsystem with a 0D high pressure core simulation. This demonstrates the ability to integrate analyses at different levels of detail and to aerodynamically couple components, the fan/booster and low-pressure turbine, through a 3D computational fluid dynamics simulation. (3) Simulation of all of the turbomachinery in a modern turbofan engine on parallel computing platform for rapid and cost-effective execution. This capability can also be used to generate full compressor map, requiring both design and off-design simulation. (4) Three levels of coupling characterize the multidisciplinary analysis under NPSS: loosely coupled, process coupled and tightly coupled. The loosely coupled and process coupled approaches require a common geometry definition to link CAD to analysis tools. The tightly coupled approach is currently validating the use of arbitrary Lagrangian/Eulerian formulation for rotating turbomachinery. The validation includes both centrifugal and axial compression systems. The results of the validation will be reported in the paper. (5) The demonstration of significant computing cost/performance reduction for turbine engine applications using PC clusters. The NPSS Project is supported under the NASA High Performance Computing and Communications Program.

Lytle, John K.↗
DOE OSTI2018

Establishing a neutronics design and equilibrium cycle analysis for the I 2 S-LWR reactor with UO 2 and U 3 Si 2 Fuel

The Integral Inherently Safe (I 2 S-LWR) reactor was developed not only to enhance passive reactor safety performance, but also to improve reactor economics. This paper describes the steady-state and depletion analysis of the I 2 S-LWR core using both UO 2 and U 3 Si 2 fuel designs. The development and validation of the core models are described, to include the generation of HELIOS cross sections and comparison with the WEC lattice code PARAGON and the Monte Carlo code SERPENT. The few group homogenized cross sections were then prepared for the U.S. NRC PARCS core simulator using the GenPMAXS code. The multicycle capability in PARCS was used to determine the equilibrium core condition for three batch cores loaded with the UO 2 and then with the U 3 Si 2 fuel designs. Comparing the results from the WEC code ANC at beginning, middle, and end of cycle showed good agreement. Core conditions at beginning and end of cycle were then used to determine the converged steady-state conditions of a coupled RELAP5/PARCS model, which provided the initial condition for various transient accident simulations which are reported in a companion paper.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2012

MCNP LWR Core Generator

The reactor core input generator allows for MCNP input files to be tailored to design specifications and generated in seconds. Full reactor models can now easily be created by specifying a small set of parameters and generating an MCNP input for a full reactor core. Axial zoning of the core will allow for density variation in the fuel and moderator, with pin-by-pin fidelity, so that BWR cores can more accurately be modeled. LWR core work in progress: (1) Reflectivity option for specifying 1/4, 1/2, or full core simulation; (2) Axial zoning for moderator densities that vary with height; (3) Generating multiple types of assemblies for different fuel enrichments; and (4) Parameters for specifying BWR box walls. Fuel pin work in progress: (1) Radial and azimuthal zoning for generating further unique materials in fuel rods; (2) Options for specifying different types of fuel for MOX or multiple burn assemblies; (3) Additional options for replacing fuel rods with burnable poison rods; and (4) Control rod/blade modeling.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Development of CTF Capability for Modeling Reactor Operating Cycles with Crud Growth

CTF is a thermal-hydraulics simulation tool designed for modeling both fluid and solid regions inside light water reactor vessels. It forms the thermal-hydraulic package of the Virtual Environment for Reactor Applications Core Simulator being developed by the Consortium for Advanced Simulation of Light water reactors. The project documented by this report has been undertaken to add a new crud-modeling capability to CTF. Achieving this goal required the completion of three specific tasks: (1) developing a method for running CTF simulations of reactor operating cycles , (2) developing and implementing a basic crud modeling “surrogate”, and (3) developing and implementing a method for inputting detailed core power distributions. These three tasks have been completed and have resulted in a new feature that allows users to run long-time-scale simulations with CTF as a collection of individual, steady-state points that comprise a larger transient. The user is able to drive inlet mass flow rate, inlet temperature, outlet pressure, and core power distributions to new values for each state in the simulation. The crud modeling tool is used at each of these state points to grow crud as a function of current thermal hydraulic conditions of the previously solved state. The developed crud modeling tool makes many simplifications that allow it to make a crude approximation of general crud behavior. It is envisioned that the infrastructure that has resulted from this project will be utilized in the future to facilitate coupling to more advanced crud chemistry codes. This document details work completed during this project and demonstrates the verification of this new capability.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Development of the MPACT 69-group Library for Magnox Reactor Analysis using CASL VERA

The Consortium for Advanced Simulation of Light Water Reactors (CASL) has developed the CASL toolset, Virtual Environment for Reactor Analysis (VERA), for pressurized water reactor (PWR) analysis. Recently the CASL VERA was improved for Magnox reactor analysis, which required the development of a new cross section library and new geometrical and thermal feedback capabilities for graphite-moderated Magnoxreactors. The MPACT neutronics module of the CASL core simulator is a 3D whole core transport code, which requires a new cross section library with a different energy group structure due to the different neutronic characteristics of Magnox compared with PWR. A new 69-group structure was developed based on the MPACT 51-group structure to have more thermal energy groups and to be a subset of the SCALE 252-group structure. The ENDF/B-VII.1 MPACT 69-group library was developed for Magnox reactor analysis using the SCALE/AMPX and VERA-XSTools for which a super-homogenization method was applied, and transport cross sections were generated for graphite using a neutron leakage conservation method. Benchmark results show that new MPACT 69-group library works reasonably well for Magnox reactor analysis.

Kim, Kang Seog↗
DOE OSTI2014

High-resolution three-dimensional simulations of core-collapse supernovae in multiple progenitors

Three-dimensional (3D) simulations of core-collapse supernovae (CCSNe) are granting new insight into the as-yet-uncertain mechanism that drives successful explosions. While there is still debate about whether explosions are obtained more easily in 3D than in 2D, it is undeniable that there exist qualitative and quantitative differences between the results of 3D and 2D simulations. We present an extensive set of high-resolution 1D, 2D, and 3D CCSN simulations with multispecies neutrino leakage carried out in two different progenitors. Our simulations confirm the results of Couch indicating that 2D explodes more readily than 3D. We argue that this is due to the inadequacies of 2D to accurately capture important aspects of the 3D dynamics. We find that without artificially enhancing the neutrino heating rate, we do not obtain explosions in 3D. We examine the development of neutrino-driven convection and the standing accretion shock instability (SASI) and find that, in separate regimes, either instability can dominate. We find evidence for growth of the SASI for both 15 M {sub ☉} and 27 M {sub ☉} progenitors; however, it is weaker in 3D exploding models. The growth rate of both instabilities is artificially enhanced along the symmetry axis in 2D as compared with our axis-free 3D Cartesian simulations. Our work highlights the growing consensus that CCSNe must be studied in 3D if we hope to solve the mystery of how the explosions are powered.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2013

Novel fine flux integration methods for the DIABOLO simplified transport solver in COCAGNE

As part of its activity, EDF R and D is developing a new nuclear core simulation code named COCAGNE. This code relies on DIABOLO, a Simplified P{sub N} (SP{sub N}) method to compute the neutron flux inside the core for k{sub eff} eigenvalue problems. In order to complete complex simulations involving a large number of successive eigenvalue calculations within acceptable CPU times, the DIABOLO solver uses computational meshes that contain a restricted number of cells (e.g. 4x4 cells per assembly). Solutions obtained on these relatively coarse computational meshes are then spatially interpolated to obtain physical quantities such as neutron production rates on finer meshes (e.g. pin-by-pin power production). This paper describes two novel methods, named Straw-hat and Poisson, that allow the neutron flux to be integrated over fine mesh cells. Unlike the original method used in COCAGNE, these new methods compute accurate fine flux integrals trough the interpolation of the current DOFs of DIABOLO'S mixed dual RTk finite elements. Assessed on a set of 3-D PWR realistic core configurations, these methods are shown to bring about significant improvements over the original integration scheme: for smooth enough flux distribution, the Poisson method improves the pin-by-pin production rates accuracy by as much as one order of magnitude. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2016

Interactive Visualization of Multi-Group Cross Sections on High-Fidelity Spatial Meshes

Core simulation tools utilizing Monte Carlo transport methods represent the 'Holy Grail' for nuclear reactor analysis. However, deterministic transport methods pose a number of advantages to Monte Carlo from a computational perspective, which seems likely to make these methods more viable in the near to intermediate term. Deterministic methods almost universally discretize the energy domain in a few to hundreds of energy groups. One of the key challenges for deterministic transport-based methods to be viable is accurate, reactor agnostic multi-group cross section generation. This paper describes tools for data processing and visualization of multigroup cross sections generated on high-fidelity spatial meshes from the OpenMC Monte Carlo code. Next generation deterministic transport reactor core modeling tools will require multi-group cross section (MGXS) generation on high-fidelity spatial meshes. High-fidelity MGXS are necessary for the accurate simulation of neutron physics in nuclear reactors with a high degree of spectral heterogeneity - e.g., large localized neutron flux gradients throughout a reactor core. Although many local (e.g., pin cell) approximations have been developed to model the scalar flux to compute spectral-averaged MGXS, these techniques cannot account for complex intra-pin and intra-assembly spatial self-shielding effects. Monte Carlo presents the most accurate method for reactor agnostic multi-group cross section generation since it does not require the use of local approximations to the flux. Monte Carlo has been proposed as a highly desirable framework for reactor agnostic cross section generation for many years. Monte Carlo methods can effectively embed all of the relevant spectral effects into the MGXS through the random particle tracking and interaction sampling processes. Previous work has applied MCNP and Serpent to compute few-group constants for diffusion calculations, and OpenMC to compute multi-group cross sections for transport-based methods. The primary difficulty with using Monte Carlo for MGXS generation is the tremendous computational resources required to adequately converge multi-group cross sections in each spatial region and energy group. Recent development of the OpenMC Monte Carlo code for high-performance computing platforms has made multigroup cross section generation from Monte Carlo more tractable in the foreseeable future. The recent release of OpenMC v0.7.1 includes a new Python-based module for MGXS generation. The MGXS generation module leverages many of the algorithms and features for 'big data' analysis of large tally datasets on complex spatial meshes. This paper describes the data management workflow and visualization tools which have been developed to explore the plethora of MGXS data generated with OpenMC. It is the authors' hope that the development of the visualization tool described in this paper will ultimately enable the design of machine learning algorithms to accelerate the convergence rate of high-fidelity multi-group cross section generated with Monte Carlo. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

ROMUSE 2.0 User Manual

ROMUSE (Reduced Order Modeling Based Uncertainty/Sensitivity Estimator) is an effort within the Consortium for Advanced Simulation of Light water reactors (CASL) to provide an analysis tool to be used in conjunction with reactor core simulators, especially the Virtual Environment for Reactor Applications (VERA). ROMUSE is written in C++ and is currently capable of performing various types of parameters perturbations, uncertainty quantification, surrogate models construction and subspace analysis. Version 2.0 has the capability to interface with DAKOTA which gives ROMUSE access to the various algorithms implemented within DAKOTA. ROMUSE is mainly designed to interface with VERA and the Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design (SCALE) [1,2,3], however, ROMUSE can interface with any general model (e.g. python and matlab) with Input/Output (I/O) format that follows the Hierarchical Data Format 5 (HDF5). In this brief user manual, the use of ROMUSE will be overviewed and example problems will be presented and briefly discussed. The algorithms provided here range from algorithms inspired by those discussed in Ref.[4] to nuclear-specific algorithms discussed in Ref. [3].

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2023

Improved Fast Reactor Capability of Griffin in FY23

Griffin is a MOOSE based reactor multiphysics analysis application jointly developed by Idaho National Laboratory and Argonne National Laboratory under the DOE-NE NEAMS program. In FY23, we enhanced capabilities required for fast reactor analysis. This effort included primarily updating the cross-section generation workflow using MC2-3 for various reactor configurations, such as homogeneous, duct-heterogeneous, ring-heterogeneous, and fully-heterogeneous geometries. In addition, we initiated the implementation of a multi-cycle depletion and shuffling capability. To support fast reactor simulation capabilities, we significantly improved the performance of the DFEM-SN-based R-Z transport solver to efficiently solve ultrafine group (over 1000 groups) transport problems. Additionally, the performance of HFEM-PN was improved by introducing red-black iteration, the cmfd acceleration technique, and various optimizations. We also completed the pin power reconstruction capability to support multiphysics simulations while identifying and addressing issues associated with SPH equivalence parameter approach. These enhanced capabilities for fast reactor core simulations, specially HFEM-PN and pin power reconstruction features, were applied to benchmark problems involving ABTR and ABR-1000. These applications showcased excellent agreement with Monte Carlo and other code solutions in terms of eigenvalue, control rod worth, and assembly and pin powers.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗