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At least 181 records · Page 10

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DOE OSTI2025

Modeling resistive-inductive evolution of currents in Wendelstein 7-X

This research investigates the temporal evolution of the toroidal plasma current in the Wendelstein 7-X (W7-X) stellarator under different heating, fueling, and current drive scenarios. The THRIFT code has been modernized and its predictions of the evolution of the toroidal current have been compared against experimentally measured currents in W7-X. Good agreement is found with respect to the characteristic timescale between experimentally measured and simulated toroidal currents. The total bootstrap current is under-predicted owing to the applicability of the BOOTSJ model for the plasma collisionalities in question. Edge plasma resistivity is found to play an important role in the asymptotic behavior of the evolution of the current, indicating a possible limitation of the minimum plasma temperature when applying this model. Simulations of ECCD and heating power steps show THRIFT is capable of capturing the dynamical evolution of the current in response to changes in current sources. Future uses of THRIFT include validating and benchmarking other non-inductive current models.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2024

Achieving Higher Order Accuracy in Space in Hydrodynamic Simulations of Self-Gravitating Gas

Modern astrophysical simulation codes employ a variety of numerical algorithms capable of achieving higher-order accuracy in both space and time. Albeit they succeed in achieving an effective higher spatial resolution and in suppressing the numerical damping of waves, to our knowledge, all current astrophysical simulations invoking self-gravity are limited to second-order accuracy in space. If we can devise an algorithm to evaluate self-gravity with a higher-order spatial accuracy, we can better the evaluation of the gravitational acceleration and gravitational energy release which dictate the evolution of many astrophysical systems. Herein, we present a numerical algorithm for self-gravitating hydrodynamics capable of achieving fourth-order accuracy for a given density distribution on a Cartesian uniform grid. First, we derive the cell-averaged gravitational potential at fourth-order accuracy from the cell-averaged density by solving the Poisson equation. Next, we obtain the cell average of the product of the density and gravitational acceleration, which differs from the cell-averaged density multiplied by the cell-averaged gravitational acceleration. We then show the verification of the algorithm by applying it to critical test problems: (1) maintaining equilibria of self-gravitating slabs, even upon advection, (2) evolving a polytropic sphere with a massive power-law envelope, and (3) conservation of specific entropy during the propagation of a sound wave.

79 ASTRONOMY AND ASTROPHYSICS↗

DOE OSTI2023

Collective enhancement in the exciton model

The preequilibrium reaction mechanism is considered in the context of the exciton model. A modification to the one-particle–one-hole state density is studied which can be interpreted as a collective enhancement. The magnitude of the collective enhancement is estimated by simulating the Lawrence Livermore National Laboratory (LLNL) pulsed-spheres neutron-leakage spectra. Here the impact of the collective enhancement is explored in the context of the highly deformed actinide, 239 Pu. A consequence of this enhancement is the removal of fictitious levels in the distorted-wave Born approximation often used in modern nuclear reaction codes.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2019

QCOR

The QCOR compiler provides a mechanism for the expression and compilation of quantum code alongside standard, modern C++. It does so through integration with Clang and its robust abstract syntax tree analysis tools, as well as integration with the XACC infrastructure.

Maccaskey, AlexanderJ↗

DOE OSTI2018

Advanced Model Developments in SAM for Thermal Stratification Analysis during Reactor Transients

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

MELCOR Accident Progression and Source Term Demonstration Calculations for a FHR

MELCOR is an integrated thermal hydraulics, accident progression, and source term code for reactor safety analysis that has been developed at Sandia National Laboratories for the United States Nuclear Regulatory Commission (NRC) since the early 1980s. Though MELCOR originated as a light water reactor (LWR) code, development and modernization efforts have expanded its application scope to include non-LWR reactor concepts. Current MELCOR development efforts include providing the NRC with the analytical capabilities to support regulatory readiness for licensing non-LWR technologies under Strategy 2 of the NRC's near- term Implementation Action Plans. Beginning with the Next Generation Nuclear Project (NGNP), MELCOR has undergone a range of enhancements to provide analytical capabilities for modeling the spectrum of advanced non-LWR concepts. This report describes the generic plant model developed to demonstrate MELCOR capabilities to perform fluoride-salt-cooled high-temperature reactor (FHR) safety evaluations. The generic plant model is based on publicly-available FHR design information. For plant aspects (e.g., reactor building leak rate and details of the cover-gas system) that are not described in the FHR references, the analysts made assumptions needed to construct a MELCOR full-plant model. The FHR model uses a TRi-structural ISOtropic (TRISO)-particle fuel pebble-bed reactor with a primary system rejecting heat to two coiled tube air heat ex changers. Three passive direct reactor auxiliary cooling systems provide heat removal to supplement or replace the emergency secondary system heat removal during accident conditions. Surrounding the reactor vessel is a low volume reactor cavity that insulates the reactor with fire bricks and thick concrete walls. A refractory reactor liner system provides water cooling to reduce the concrete wall temperatures. Example calculations are performed to show the plant response and MELCOR capabilities to characterize a range of accident conditions. The accidents selected for evaluation consider a range of degraded and failed modes of operation for key safety functions providing reactivity control, the primary system decay heat removal and also a piping leak of the line to the coolant drain tank.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

MELCOR Accident Progression and Source Term Demonstration Calculations for a Heat Pipe Reactor

MELCOR is an integrated thermal hydraulics, accident progression, and source term code for reactor safety analysis that has been developed at Sandia National Laboratories for the United States Nuclear Regulatory Commission (NRC) since the early 1980s. Though MELCOR originated as a light water reactor (LWR) code, development and modernization efforts have expanded its application scope to includ e non-LWR reactor concepts. Current MELCOR development efforts include providing the NRC with the analytical capabilities to support regulatory readiness for licensing non-LWR techno logies under Strategy 2 of the NRC?s near- term Implementation Action Plans. Beginning with the Next Generation Nuclear Project (NGNP), MELCOR has undergone a range of enha ncements to provide analytical capabilities for modeling the spectrum of advanced non-LWR concepts. This report describes the generic plant model developed to demonstrate MELCOR capabilities to perform heat pipe reactor (HPR) safety evaluations. The generic plant mode l is based on a publicly-available Los Alamos National Laboratory (LANL) Megapower design as modified in the Idaho National Laboratory (INL) Design A description. For plant aspects (e.g., reactor building size and leak rate) that are not described in the LANL and INL references , the analysts made assumptions needed to construct a MELCOR full-plant model. The HP R uses high assay, low-enrichment uranium (HALEU) fuel with steel cladding that uses heat pipes to transfer heat to a secondary Brayton air cycle. The core region is surrounded by a stainless-steel shroud, alumina reflector, core barrel and boron carbide neutron shield. The reactor is secured inside a below-grade cavity, with the operating floor located above the cavity. Example calculations are performed to show the plant response and MELCOR capabilities to characterize a range of accident conditions. The accidents selected for evaluation consider a range of degraded and failed modes of operation for key safety functions providing re activity control, the primary and secondary system heat removal, and the effectiveness of th e confinement natural circulation flow into the reactor cavity (i.e., a flow blockage).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

MELCOR Accident Progression and Source Term Demonstration Calculations for a HTGR

MELCOR is an integrated thermal hydraulics, accident progression, and source term code for reactor safety analysis that has been developed at Sandia National Laboratories for the United States Nuclear Regulatory Commission (NRC) since the early 1980s. Though MELCOR originated as a light water reactor (LWR) code, development and modernization efforts over the past decades have expanded its application scope to include non-LWR reactor concepts. Current MELCOR development efforts include providing the NRC with the analytical capabilities to support regulatory readiness for licensing non-LWR technologies under Strategy 2 of the NRC's near-term Implementation Action Plans. Beginning with the Next Generation Nuclear Project (NGNP), MELCOR ha s undergone a range of enhancements to provide analytical capabilities for modeling the spectrum of advanced non-LWR concepts. This report describes the generic plant model developed to demonstrate MELCOR capabilities to perform high-temperature gas reactor (HTGR) safety evaluations. The generic plant model is based on publicly available PMBR-400 design information. For plant aspects (e.g., reactor building size and leak rate) that are not described in the PBMR-400 references, the analysts made assumptions needed to construct a MELCOR full-plant model. The HTGR model uses a TRi-structural ISOtropic (TRISO)-particle fuel pebble-bed reactor with a primary system rejecting heat to a recuperative heat exchange r. Surrounding the reactor vessel is a reactor cavity contained within a confinement room cooled by the Reactor Cavity Cooling System (RCCS). Example calculations are performed to show the plant response and MELCOR capabilities to characterize a range of accident conditions. The accidents selected for evaluation consider a range of degraded and failed modes of operation for key safety functions providing reactivity control, primary system heat removal and reactor vessel decay heat removal, and confinement cooling.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Achieving Higher Order Accuracy in Space in Hydrodynamic Simulations of Self-Gravitating Gas

Modern astrophysical simulation codes employ a variety of numerical algorithms capable of achieving higher-order accuracy in both space and time. Albeit they succeed in achieving an effective higher spatial resolution and in suppressing the numerical damping of waves, to our knowledge, all current astrophysical simulations invoking self-gravity are limited to second-order accuracy in space. If we can devise an algorithm to evaluate self-gravity with a higher-order spatial accuracy, we can better the evaluation of the gravitational acceleration and gravitational energy release which dictate the evolution of many astrophysical systems. Herein, we present a numerical algorithm for self-gravitating hydrodynamics capable of achieving fourth order accuracy for a given density distribution on a Cartesian uniform grid. First, we derive the cell-averaged gravitational potential at fourth-order accuracy from the cell-averaged density by solving the Poisson equation. Next, we obtain the cell average of the product of the density and gravitational acceleration, which differs from the cell-averaged density multiplied by the cell-averaged gravitational acceleration. We then show the verification of the algorithm by applying it to critical test problems: (1) maintaining equilibria of self-gravitating slabs, even upon advection, (2) evolving a polytropic sphere with a massive power-law envelope, and (3) conservation of specific entropy during the propagation of a sound wave.

79 ASTRONOMY AND ASTROPHYSICS↗

DOE OSTI2026

SAM Finite Volume Method Development Status Update: GCR Application, Restart, and MultiApp

The System Analysis Module (SAM) is being developed as a modern system analysis code for advanced non-light-water-reactor safety analysis under the U.S. DOE NEAMS program. Previous feasibility studies have demonstrated that a staggered-grid finite volume method (SG-FVM), implemented under the MOOSE framework, can deliver more than an order of magnitude speedup over the existing continuous Galerkin finite element method (CG-FEM) solver for liquid-cooled, incompressible but thermally expandable flow systems. This work extends the previous effort to compressible, gas-cooled reactor applications, where pressure couples directly into the mass equation adding additional nonlinearity into the equation system. New code capabilities are implemented for pebble bed high-temperature gas-cooled reactor (PB-HTGR) analysis, including a pebble bed CoreChannel component, built-in pebble bed effective thermal conductivity model and channel-to-channel crossflow model. The capabilities are tested, benchmarked, and demonstrated for problems with increased level of model and physical complexities, including the HTTU effective thermal conductivity test, the SANA passive cooling test, and a demonstration case using the GPBR200 reactor design covering steady-state operation, DLOFC and PLOFC transients. Across all cases, the SG-FVM solver demonstrated strong robustness and efficiency, and the solutions agree well with reference results and data. The finding of this work proves that SG-FVM is a viable and efficient solver pathway for compressible, gas-cooled reactor system analysis in SAM. In addition, work has been done to successfully support SAM-FVM recover/restart code feature that is essential to reactor safety analysis applications, and MultiApp code feature that is essential to multi-scale and multi-physics simulations. In summary, this work continued from previous feasibility studies, and further demonstrated that the SG-FVM will serve as a strong foundation for SAM’s advanced solver algorithm for future deployment.

Zou, Ling↗

NASA NTRS1983

Airloads on bluff bodies, with application to the rotor-induced downloads on tilt-rotor aircraft

The aerodynamic characteristics of airfoils with several flap configurations were studied theoretically and experimentally in environments that simulate a wing immersed in the downwash of a hovering rotor. Special techniques were developed for correcting and validating the wind tunnel data for large blockage effects, and the test results were used to evaluate two modern blockage effects, and the test results were used to evaluate two modern computational aerodynamics codes. The combined computed and measured results show that improved flap and leading-edge configurations can be designed which will achieve large reductions in the downloads of tilt-rotor aircraft, and thereby improve their hover efficiency.

Mccroskey, W. J.↗

NASA NTRS1984

Computerized aerodynamic design of a transonically 'quiet' blade

The high noise levels produced by helicopters are major sources of concern. There are many sources of the noise, but during high-speed forward flight, impulsive noise dominates the noise spectrum. The cause of the high-speed impulsive noise is the propagation into the far field of shock waves that form on the advancing blade. This mechanism has been labeled 'delocalization'. It has been shown, however, that by judicious design of the blade-tip planform, delocalization can be prevented. The objective of the present study is to illustrate how blade-tip configurations (both planform and airfoil shape) can be systematically varied to identify shapes that avoid delocalization and simultaneously improve aerodynamic performance. This has been done using the latest version of the ROT22 transonic, full-potential, quasi-steady, rotor flow-field code. A hypothetical modern rotor blade was postulated, and tip modifications consisting of taper, sweep, and airfoil section alterations were investigated. Planform modifications were found to be most effective in eliminating delocalization.

Tauber, M. E.↗

NASA NTRS1985

Airloads on bluff bodies, with application to the rotor-induced downloads on tilt-rotor aircraft

Mccroskey, W. J.↗

NASA NTRS1990

Experimental studies of 90 deg corner cascades in the National Full-Scale Aerodynamic Complex

Two 90 deg, fixed geometry, turning vane cascades were replaced in the 40- by 80-foot wind tunnel circuit of the National Full-Scale Aerodynamic Complex as a part of a major facility upgrade. Prior to the replacement of these vanes, small-scale, two-dimensional tests were conducted to assess the aerodynamic performance of the original, 1930's technology vane geometry and a new vane geometry which was developed using a modern cascade design code. The small scale tests were conducted in the 1/10th-Scale Cascade Test Facility and included measurements of total pressure, static pressure, and flow angularity distributions in the wake of the cascade and static pressure distributions along the vane profiles. Test results and a detailed comparison of the two vanes designs are presented. Based on these results, the new design was chosen and subsequently installed in the wind tunnel. During the integrated systems test of the upgraded facility, two of the 52 vanes in the first cascade were instrumented near midspan. The full-scale test results are presented, these show the cascade pressure loss coefficient to be slightly larger than obtained in the small-scale tests. Elements of the three-dimensional flow field upstream of the cascade, which effect can impact cascade performance, are discussed and recommendations for future research are given.

Meyn, Larry A.↗

DOE OSTI2017

Development of a Reduced-Order Three-Dimensional Flow Model for Thermal Mixing and Stratification Simulation during Reactor Transients

Mixing, thermal-stratification, and mass transport phenomena in large pools or enclosures play major roles for the safety of reactor systems. Depending on the fidelity requirement and computational resources, various modeling methods, from the 0-D perfect mixing model to 3-D Computational Fluid Dynamics (CFD) models, are available. Each is associated with its own advantages and shortcomings. It is very desirable to develop an advanced and efficient thermal mixing and stratification modeling capability embedded in a modern system analysis code to improve the accuracy of reactor safety analyses and to reduce modeling uncertainties. An advanced system analysis tool, SAM, is being developed at Argonne National Laboratory for advanced non-LWR reactor safety analysis. While SAM is being developed as a system-level modeling and simulation tool, a reduced-order three-dimensional module is under development to model the multi-dimensional flow and thermal mixing and stratification in large enclosures of reactor systems. This paper provides an overview of the three-dimensional finite element flow model in SAM, including the governing equations, stabilization scheme, and solution methods. Additionally, several verification and validation tests are presented, including lid-driven cavity flow, natural convection inside a cavity, laminar flow in a channel of parallel plates. Based on the comparisons with the analytical solutions and experimental results, it is demonstrated that the developed 3-D fluid model can perform very well for a wide range of flow problems.

Hu, Rui↗

DOE OSTI2020

Simulation of Particle-Material Interactions

This paper gives an overview of the particle transport theory essentials, the basics of particle-material interaction simulation, physical quantities needed to simulate particle transport and interactions in materials, Monte Carlo simulation flow, response of additive detectors, statistical weights and other techniques to minimize statistical errors. Effects in materials under irradiation, materials response related to component lifetime and performance are considered with a focus on high-energy and high-power accelerator applications. Implementation of simulation of particle-material interactions in the modern Monte Carlo codes along with the code s main features and results of recent benchmarking are described.

43 PARTICLE ACCELERATORS↗

DOE OSTI2016

Ab Initio Molecular Dynamics Analysis of Xenon Incorporation in Uranium Dioxide

The fission process in nuclear fuel produces volatile fission product gases, such as xenon, that migrate out of the fuel and into surrounding structures. Consequently, these fission product gases may affect thermal performance and integrity of the fuel. Fission gas release models, used in modern fuel performance codes such as BISON, are currently under investigation to improve their accuracy and ability to simulate the fission gas release phenomenon and its consequences. For uranium dioxide (UO{sub 2}), the most commonly used fuel in commercial power reactors, fission gas release models are known to have significant uncertainties. In the interest of improving such models, several complex diffusional mechanisms have been examined using a variety of techniques such as Nudge Elastic Band (NEB), utilizing both Density Functional Theory (DFT) and semi-empirical inter-atomic potential functions. One proposed diffusion mechanism in UO{sub 2} is through tetra-vacancy (i.e., 2U{sub V} + 2O{sub V}) structures, which is expected to provide a suitable means for the incorporation and diffusion of xenon in the fuel crystal. In this work, a model was developed to study the incorporation of xenon in UO{sub 2} via a tetra-vacancy using ab initio molecular dynamics (AIMD). This is a first principles technique that abandons the use empirical force fields as usually is the case of classical MD analysis. Furthermore, in contrast with static transition analysis tools such as NEB, AIMD is able to capture the dynamics of a given process. The translation of xenon within a tetra-vacancy has been investigated using ab-initio molecular dynamics. The transition frequency, barrier energy, and incorporation energy were extracted from the model, demonstrating its ability to capture a variety of parameters and its potential for use in studying fission gas diffusive behavior. Future refinements to the model will seek to develop additional insight into Xe and Kr mobility and will likely involve executing more elaborate AIMD simulations to investigate this phenomenon. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Status and Ongoing Developments for the GeN-Foam Multi-Physics Solver

An open-source multi-physics solver for nuclear reactor analysis, named GeN-Foam (Generalized Nuclear Foam), is being developed by the Laboratory for Reactor Physics and System Behavior at the PSI and at the EPFL. The objective is to supplement available legacy codes with a modern, flexible tool making use of the most recent numerical methodologies. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗