Uncertainty Quantification for Full-core Steady-state PWR Core Simulation with SCALE and PARCS
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Zhou, Haining↗
SEARCH · Engineering Papers
Results for “core simulation”
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SUB-GRID MODEL IMPROVING SHRINKING CORE SIMULATIONS OF REDOX-SENSITIVE TECHNETIUM RELEASE FROM SLAG CEMENT WASTE FORMS - 17295
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Flach, G.↗
Uncertainty Quantification of Pressurized Water Reactor Coupled Core Simulation Using Stochastic Sampling Method
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Zeng, K.↗
Sub-Grid Model Improving Shrinking Core Simulations of Redox-Sensitive Technetium Release From Slag Cement Waste Forms - 17295
Numerical simulation of Tc transport in reducing cementitious materials undergoing oxidation is challenging because the effective solid-solute partition coefficient (K{sub d}) varies greatly between the reduced and oxidized states, assuming the Tc mass loading far exceeds that required to reach solubility in the aqueous phase, as is typical in waste disposal. In a conventional finite-volume formulation, the local aqueous concentration increases orders of magnitude when a computational cell switches from a reducing to oxidizing condition, which is an artifact of numerical discretization. In principal such artifacts can be reduced to an insignificant level by refining the computational grid. However, this approach is not feasible in typical facility scale applications due to computer runtime and storage limitations. In this paper an alternative sub-grid Tc release model is derived to approximately capture redox processes occurring at scales smaller than the computational mesh. This alternative approach introduces minimal additional computational expense, while significantly reducing numerical artifacts in the form of Tc concentration / flux spikes. The method is illustrated for two simple geometries and the Savannah River Site Saltstone waste Disposal Facility. The sub-grid model is shown to significantly reduce artificial spikes in Tc-99 release, resulting in more physically realistic numerical simulations. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Initial steady-state core simulation capability or thermal and pool-type molten salt reactors, coupling reactor physics, thermal-hydraulics, and evolving chemistry
This report presents the development and validation of an initial steady-state multiphysics capability for molten salt reactors (MSRs) under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in Fiscal Year 2025. The framework integrates neutronics, thermal-hydraulics, species transport, and thermochemistry to capture the coupled dynamics of liquid-fueled systems. Implementation and testing were performed on two representative designs: the Molten Salt Reactor Experiment (MSRE), a thermal-spectrum, channeled-flow reactor, and the Lotus Molten Salt Reactor (L-MSR), a fast-spectrum, pool-type reactor. The modeling suite employs Griffin for reactor physics and depletion, Pronghorn and SAM for thermal-hydraulics, Thermochimica for chemistry, and Saline for thermophysical properties, with benchmarking and validation carried out against historical MSRE data, experimental flow-loop measurements, and reference depletion calculations from Monte Carlo codes. The framework demonstrated the ability to reproduce key reactor behaviors including temperature feedback, reactivity losses, delayed neutron precursor transport, xenon poisoning, and redox potential evolution. The results confirm the feasibility and accuracy of the coupled models in predicting steady-state and selected transient MSR behaviors. This latter ones are used in this report as a proxy indicating that the steady-state models from which the transient starts are accurate. For MSRE, validation showed good agreement with pump start-up and natural circulation tests, while for the L-MSR, benchmarking confirmed hydraulic calibration and consistency of neutronics–thermal coupling. The tools also provided new insights into species transport, noble metal deposition, and salt solidification dynamics. On the Xenon transport front, the code is validated against the steady state Xenon poisoining measurement and showed good agreement with the experimental value. Identified areas for future work include advanced void transport modeling, three-dimensional simulations, improved alloy corrosion models, and tighter integration with high-fidelity Monte Carlo codes. These developments provide a foundation for high-fidelity MSR simulations that can support reactor design optimization, safety assessments, and long-term operational strategies.
42 - ENGINEERING↗
Taming Wild Horses: The Need for Virtual Time-based Scheduling of VMs in Network Simulations
The next generation of scalable network simulators employ virtual machines (VMs) to act as high-fidelity models of traffic producer/consumer nodes in simulated networks. However, network simulations could be inaccurate if VMs are not scheduled according to virtual time, especially when many VMs are hosted per simulator core in a multi-core simulator environment. Since VMs are by default free-running, on the outset, it is not clear if, and to what extent, their untamed execution affects the results in simulated scenarios. Here, we provide the first quantitative basis for establishing the need for generalized virtual time scheduling of VMs in network simulators, based on an actual prototyped implementations. To exercise breadth, our system is tested with multiple disparate applications: (a) a set of message passing parallel programs, (b) a computer worm propagation phenomenon, and (c) a mobile ad-hoc wireless network simulation. We define and use error metrics and benchmarks in scaled tests to empirically report the poor match of traditional, fairness-based VM scheduling to VM-based network simulation, and also clearly show the better performance of our simulation-specific scheduler, with up to 64 VMs hosted on a 12-core simulator node.
99 GENERAL AND MISCELLANEOUS↗
An assessment of coupling algorithms for nuclear reactor core physics simulations
This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss–Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton–Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Numerical simulations demonstrating the efficiency of JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.
71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
Accuracy Enhancement of Nuclear Power Plant Simulators Utilizing High Accuracy Simulation Predictions
More recently, reactor core simulators for core designs associated with commercial nuclear power plants that utilize what is believed to be higher fidelity models have been developed. Features such as neutronics models that utilize transport equation solvers with fine spatial meshes and many energy-groups, thermal-hydraulic models that utilize sub-channel solvers with fine spatial mesh and capable of treating a wide range of fluid conditions, and fuel-coolant chemistry interaction models capable of treating CRUD deposition are to be found in these higher fidelity core simulators. These reactor core simulators require access to higher performance computers, characterized by many processors, cores and large memory. So associated with utilization of these simulators is access to high performance computers and ability to accommodate in one’s workflow longer execution times. By contrast, currently used core simulators by the nuclear industry can execute on engineering workstations and have execution times of seconds to minutes. The desirability for having short execution times is not only desired for support of time critical tasks but supports the mental process of decision making by engineers. The goal of the work reported upon here has the objective of retaining the fidelity of higher fidelity models while retaining the ability to utilize engineering workstations. Beyond the core simulator goal, additional goals of this work include incorporating the just described core simulator capability into a Nuclear Steam Supply System (NSSS) simulator, and to incorporate the resulting capability into an environment supportive of design and operational decision making associated with nuclear power stations. The model selected for the core neutronics model is the NESTLE code, for the core thermal-hydraulic model is the CTF code utilizing coarse mesh, and for the NSSS model is the RELAP5-3D code. WSC’s proprietary 3KEYMASTERTM platform is being used to provide software coupling, user interface, visualization, and reporting. The NESTLE core neutronics simulator was first integrated with the CTF core thermal-hydraulic simulator using CTF developed communication commands which are also used for CTF to communicate with RELAP5-3D under WSC’s proprietary 3KEYMASTERTM platform. To assure NESTLE prediction consistency with higher fidelity core neutronic simulators, buffer codes have been created to automatically generate from output files written by the VERA core simulator the NESTLE nodal neutronic parameter’ library, geometry, and pin-power reconstruction input files, thereby avoiding a number of challenges associated with utilizing lattice physics codes and providing consistency with VERA predictions. To treat absorber rod effects a multi-set library is utilized, where a set refers to a specific absorber rod fully inserted pattern. A coarse spatial mesh CTF model was developed with features added that support using CTF as envisioned in the engineering quality simulator. A hybrid meshing approach was implemented to allow for automated construction of models with mixed levels of refinement. Specifically, a core model could resolve some assemblies at a nodal level (4 subchannels per assembly) and others at a pin-resolution (one subchannel per coolant subchannel in the assembly). The intention is that this will allow for better resolution of limiting conditions such as DNBR and PCT, which are based on local rod and subchannel conditions. Further development was done of features that enhance the capabilities for the envisioned engineering quality simulator that has been developed, but now for RELAP-3D. The RELAP5-3D code development includes ability to model more than 999 components and the addition of the cross-channels turbulence mixing model and the void drift model that are implemented in CTF, aiming to achieve closer prediction agreement of the two codes for transient simulations, specifically, more accurate matches of the overall mass, momentum, and energy exchanges of both the liquid and gas phases between the neighboring core assemblies. Graphics were also developed for the Instructor Station for this project under WSC’s proprietary 3KEYMASTERTM platform to facilitate design and operational decision making.
42 ENGINEERING↗
Development of a New 47-Group Library for the CASL Neutronics Simulators
The CASL core simulator MPACT is under development for the neutronics and thermal-hydraulics coupled simulation for the pressurized light water reactors. The key characteristics of the MPACT code include a subgroup method for resonance self-shielding, and a whole core solver with a 1D/2D synthesis method. The ORNL AMPX/SCALE code packages have been significantly improved to support various intermediate resonance self-shielding approximations such as the subgroup and embedded self-shielding methods. New 47-group AMPX and MPACT libraries based on ENDF/B-VII.0 have been generated for the CASL core simulator MPACT of which group structure comes from the HELIOS library. The new 47-group MPACT library includes all nuclear data required for static and transient core simulations. This study discusses a detailed procedure to generate the 47-group AMPX and MPACT libraries and benchmark results for the VERA progression problems.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Initial full core SMR simulations with NekRS
This document describes the completion of a recent milestone of the ExaSMR program concerning "Full core simulations". The ExaSMR project is developing tools for coupling Monte Carlo (MC) radiation transport solvers to a computational fluid dynamics (CFD) solver. Work is based on the Shift MC, OpenMC MC, and Nek5000/NekRS CFD codes. As part of this milestone, a novel set of pin-resolved CFD full-core simulations have been performed for the first time. These simulations represent a significant increase in capability in what is now possible with CFD on pre-Exascale systems. The simulations have been performed with the spectral element solvers NekRS on the supercomputer Summit. Multiple simulations campaigns have been conducted: (1) LES simulations in bare bundle, (2) RANS simulations in bare bundles, (3) RANS simulations with momentum sources and (4) Conjugate heat transfer calculations.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Relation between gamma-ray family and EAS core: Monte-Carlo simulation of EAS core
Preliminary results of Monte-Carlo simulation on Extensive Air Showers (EAS) (Ne=100,000) core is reported. For the first collision at the top of the atmosphere, high multiplicity (high rapidity, density) and a large Pt (1.5GeV average) model is assumed. Most of the simulated cores show a complicated structure.
Yanagita, T.↗
Multidimensional Simulations of Core Convection
The cores of main sequence intermediate- and high-mass stars are convective. Mixing at the radiative–convective boundary, waves excited by the convection, and magnetic fields generated by convective dynamos all influence the main sequence and post-main sequence evolution of these stars. These effects must be understood to accurately model the structure and evolution of intermediate- and high-mass stars. Unfortunately, there are many challenges in simulating core convection due to the wide range of temporal and spatial scales, as well as many important physics effects. In this review, we describe the latest numerical strategies to address these challenges. We then describe the latest state-of-the-art simulations of core convection, summarizing their main findings. These simulations have led to important insights into many of the processes associated with core convection. Two outstanding problems with multidimensional simulations are, 1. it is not always straightforward to extrapolate from simulation parameters to the parameters of real stars; and 2. simulations using different methods sometimes appear to arrive at contradictory results. To address these issues, next generation simulations of core convection must address how their results depend on stellar luminosity, dimensionality, and turbulence intensity. Furthermore, code comparison projects will be essential to establish robust parameterizations that will become the new standard in stellar modeling.
79 ASTRONOMY AND ASTROPHYSICS↗
Comparison of simulation results with sea-level experimental data on 10(14) - 10(16) air shower cores
Simulation predictions for the Leeds 35 sq m horizontal discharge chamber array for proton primaries with a approx. E sup 2.7 spectrum extrapolated from balloon data to 10 to the 16th power eV give power law rho (r)-spectra with constant slope approx. -2 consistent with the experimental data up to the point at which they steepen but overshooting them at higher densities, and at high shower sizes predicted cores which are significantly steeper than those observed. Further comparisons with results for heavy nuclei primaries (up to A = 56) point to the inadequacy of changes in primary composition to account for the observed density spectra and core flattening, and the shower size spectrum together, and point, therefore, to the failure of the scaling interaction model at approx. 10 to the 15th power eV primary energy.
Ash, A. G.↗
A study of the core module simulator floor capability
The floor of the Core Module Simulator (CMS) is required to support various combinations of dead load and live load during the testing process. Even though there is published data on the structural capability of the grating, it is not always evident if the combined loadings with joint loads will cause structural failure.
Foreman, James W.↗
Data accompanying the publication "Sensitivity of grass fires burning in marginal conditions to atmospheric turbulence" in the Journal of Geophysical Research-Atmospheres
Data accompanying Jonko, Yedinak, Conley, and Linn (2021): "Sensitivity of grass fires burning in marginal conditions to atmospheric turbulence", JGR-Atmospheres. LA-UR-21-24832. The following files were generated from simulations performed with Los Alamos National Laboratory coupled atmosphere-fire behavior model FIRETEC v1.3: -area_burned.csv contains the area burned per second in m^2 for 45 CORE simulations, 20 INIT simulations and 20 BOUND simulations (85 columns), and 600 seconds of simulation (600 rows). -fire_perimeter.csv containing the fire perimeters in m for 45 CORE simulations, 20 INIT simulations and 20 BOUND simulations (85 columns), and 600 seconds of simulation (600 rows). -consumption_per_second.csv containing consumption in kg/s for 45 CORE simulations, 20 INIT simulations and 20 BOUND simulations (85 columns), and 600 seconds of simulation (600 rows). -rate_of_spread.csv containing the distance of the fire front from the ignition line in m for 45 CORE simulations, 20 INIT simulations and 20 BOUND simulations (85 columns), and 600 seconds of simulation (600 rows). -CORE1_u_10m_300_sec.hdf, INIT3_u_10m_300_sec.hdf, and BOUND3_u_10m_300_sec.hdf contain u wind speeds at 10 m height for 300 seconds (3D: time, x, y) of simulation for simulations CORE1, INIT3, and BOUND3, respectively. -CORE.final.fuel.density.hdf contains final fuel densities in the surface layer of the model (3D: ensemble member, x, y) for all 45 CORE simulations. -CORE{1-45}_3D_u_120_sec.hdf: 45 files containing vertically and horizontally resolved u winds for 120 seconds of simulation (4D: time, x, y, z) for the 45 CORE ensemble members. For additional information about this data, please contact the corresponding author at ajonko@lanl.gov.
54 ENVIRONMENTAL SCIENCES↗
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↗
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↗
Heat‐Generating Nuclear Waste in Salt: Field Testing and Simulation
Core Ideas A field‐scale experiment and numerical simulations confirm salt backfill behavior. Simulations closely match temperature around and under the piled salt backfill. Results indicate limited dissolution–precipitation reactions around the heat source. Alteration of backfill is unlikely if the drift is allowed to dry before emplacement. Investigations relating to in‐drift disposal of heat‐generating nuclear waste in salt have raised questions about heat–brine interactions in the unsaturated run‐of‐mine (RoM) salt pile used as backfill. These interactions have the potential to change the structure of the RoM salt surrounding the canister, possibly altering long‐term containment of the source. An experiment is in progress at the Waste Isolation Pilot Plant (WIPP), New Mexico, in which a heated canister was placed on the floor of an open drift, covered in a pile of RoM salt, and energized with 1000 W. Temperature in the RoM salt pile had stabilized after about 15 d, allowing evaluation of the heat‐up period of the ongoing experiment. Using a multiphase porous flow simulator that has been modified to handle salt‐specific coupled processes, we examined coupled thermal–hydrological–chemical behavior in the RoM salt pile. Our simulations suggest that for the relatively dry cases examined, porosity changes within RoM salt in a generic salt repository are likely to be minor in the period between waste emplacement and plastic closure of the drift. The primary sensitivity for porosity change is to the early moisture content of the RoM salt used to cover the canister. Secondary influences include moisture availability from the disturbed rock zone (DRZ) surrounding the drift and the capillary pressure ratio between the DRZ and the RoM salt. Early changes in porosity and permeability may be affected by moisture content, but this was not observed in the test. Such changes would be most likely to occur when using damp RoM salt or if waste is emplaced in a drift immediately following opening of the drift before evaporative dewatering of the drift walls occurs.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗