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At least 1,423 records · Page 79

Accurate Fe–He machine learning potential for studying He effects in BCC-Fe

Nanostructured Ferritic Alloys (NFAs) containing oxide nanoparticles are prospective advanced structural materials in future fusion reactors. Lacking fusion neutron sources with an adequate flux, modeling is critical to understand radiation damage in the NFAs, including helium effects. Machine learning interatomic potentials (MLPs) have been demonstrated to be capable of describing atomic interactions in chemically complex systems with comparable accuracy to density functional theory (DFT) but with much less computational costs. Hence, MLPs are chosen to study He bubble accumulation in chemically complex NFAs. As a preliminary step, we develop a Fe–He potential, based on ~10,000 atomic configurations. The developed MLP accurately predicts the bulk properties of BCC-Fe compared to DFT, including the lattice constant (2.832 Å), elastic constants (c 11 = 271 GPa, c 12 = 141, c 44 = 93), and phonon frequencies (maximum error < 3.5%). The MLP predicts He binding energies in He n V bubbles and He n clusters with a mean absolute error (MAE) of only ~70 meV, which is ~3 to ~5 times smaller than the MAE of binding energies estimated using empirical potentials. The MLP also accurately predicts the migration barrier of interstitial He (64 meV). Furthermore, the MLP correctly predicts the maximum number of He atoms (critical size) at which a He n V bubble (He n cluster) can kick-out a self-interstitial atom (SIA), forming a bigger He n V 2 bubble (He n V bubble). Subsequently, we use the MLP to perform molecular dynamics simulations to investigate the effect of temperature on the critical size of He n V bubbles and He n clusters. We find that the critical size decreases with increasing temperature — indicating that smaller bubbles and clusters can kick-out an SIA at higher temperatures, thereby escalating the He bubble growth. We report the current Fe–He MLP can be further developed to include all chemical interactions in NFAs.

36 MATERIALS SCIENCE↗

Laser-driven ionization mechanisms of aluminum for single particle aerosol mass spectrometry

Single particle aerosol mass spectrometry (SPAMS), an analytical technique for measuring the size and composition of individual micron-scale particles, is capable of analyzing atmospheric pollutants and bioaerosols much more efficiently and with more detail than conventional methods which require the collection of particles onto filters for analysis in the laboratory. Despite SPAMS’ demonstrated capabilities, the primary mechanisms of ionization are not fully understood, which creates challenges in optimizing and interpreting SPAMS signals. In this paper, we present a well-stirred reactor model for the reactions involved with the laser-induced vaporization and ionization of an individual particle. The SPAMS conditions modeled in this paper include a 248 nm laser which is pulsed for 8 ns to vaporize and ionize each particle in vacuum. The ionization of 1 μm, spherical Al particles was studied by approximating them with a 0-dimensional plasma chemistry model. The primary mechanism of absorption of the 248 nm photons was pressure-broadened direct photoexcitation to Al(y 2 D). Atoms in this highly excited state then undergo superelastic collisions with electrons, heating the electrons and populating the lower energy excited states. We found that the primary ionization mechanism is electron impact ionization of various excited state Al atoms, especially Al(y 2 D). Because the gas expands rapidly into vacuum, its temperature decreases rapidly. The rate of three-body recombination (e - + e - + Al + → Al + e - ) increases at low temperature, and most of the electrons and ions produced recombine within several μs of the laser pulse. The importance of the direct photoexcitation indicates that the relative peak heights of different elements in SPAMS mass spectra may be sensitive to the available photoexcitation transitions. We also discuss the effects of laser intensity, particle diameter, and expansion dynamics.

47 OTHER INSTRUMENTATION↗

GaAs growth rates of 528 μm/h using dynamic-hydride vapor phase epitaxy with a nitrogen carrier gas

Herein we demonstrate record GaAs growth rates approaching 530 μm/h using nitrogen carrier gas and 400 μm/h using hydrogen carrier gas in a dynamic-hydride vapor phase epitaxy reactor. We measured root mean square surface roughness below 1 nm using a 1 μm × 1 μm atomic force microscopy scan for GaAs growth rates up to 483 μm/h using a nitrogen carrier gas and 400 μm/h using a hydrogen carrier gas. We performed computational fluid dynamics modeling to study the effect of the carrier gas choice on the thermal profile within the reactor and how that influences the degree of AsH 3 decomposition. The modeling suggests that the lower thermal conductivity of the nitrogen carrier gas minimizes the amount of AsH 3 that thermally decomposes before reaching the wafer surface and the heavier atomic mass decreases the likelihood that AsH 3 will reach the heated reactor walls, leading to a growth rate enhancement relative to the hydrogen carrier case in a hydride-enhanced growth regime.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

Particle trapping, size-filtering, and focusing in the nonthermal plasma synthesis of sub-10 nanometer particles

Low-pressure nonthermal flowing plasmas are widely used for the gas-phase synthesis of nanoparticles and quantum dots of materials that are difficult or impractical to synthesize using other techniques. To date, the impact of temporary electrostatic particle trapping in these plasmas has not been recognized, a process that may be leveraged to control particle properties. Furthermore, we present experimental and computational evidence that, during their growth in the plasma, sub-10 nm silicon particles become temporarily confined in an electrostatic trap in radio-frequency excited plasmas until they grow to a size at which the increasing drag force imparted by the flowing gas entrains the particles, carrying them out of the trap. We demonstrate that this trapping enables the size filtering of the synthesized particles, leading to highly monodisperse particle sizes, as well as the electrostatic focusing of the particles onto the reactor centerline. Understanding of the mechanisms and utilization of such particle trapping will enable the design of plasma processes with improved size control and the ability to grow heterostructured nanoparticles.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Survey Corrosion Testing Methodologies for Additively Manufactured Materials – PNNL

The Advanced Materials and Manufacturing Technology (AMMT) Program intends to develop cross-cutting technologies in support of a broad range of nuclear reactor technologies, and to maintain U.S. leadership in materials and manufacturing technologies for nuclear energy applications. The overarching vision of the AMMT Program is to accelerate the development, qualification, demonstration, and deployment of advanced materials and manufacturing technologies to enable reliable and economical nuclear energy.

36 MATERIALS SCIENCE↗

Development of processes for the production of solar grade silicon from halides and alkali metals, phase 1 and phase 2

High temperature reactions of silicon halides with alkali metals for the production of solar grade silicon are described. Product separation and collection processes were evaluated, measure heat release parameters for scaling purposes and effects of reactants and/or products on materials of reactor construction were determined, and preliminary engineering and economic analysis of a scaled up process were made. The feasibility of the basic process to make and collect silicon was demonstrated. The jet impaction/separation process was demonstrated to be a purification process. The rate at which gas phase species from silicon particle precursors, the time required for silane decomposition to produce particles, and the competing rate of growth of silicon seed particles injected into a decomposing silane environment were determined. The extent of silane decomposition as a function of residence time, temperature, and pressure was measured by infrared absorption spectroscopy. A simplistic model is presented to explain the growth of silicon in a decomposing silane enviroment.

Dickson, C. R.↗

Study of Electro-Cyclonic Filtration and Pneumatic Transfer of Lunar Regolith Simulants under 1/6-g and 1-g Gravity Conditions

NASA has built a prototype oxygen production plant to process the lunar regolith using the hydrogen reduction chemical process. This plant is known as "ROxygen - making oxygen from moon rocks". The ROxygen regolith transfer team has identified the flow and transfer characteristics of lunar regolith simulant to be a concern for lunar oxygen production efforts. It is important to ISRU lunar exploration efforts to develop hardware designs that can demonstrate the ability to flow and transfer a given mass of regolith simulant to a desired vertical height under lunar gravity conditions in order to introduce it into a reactor. We will present results obtained under both 1/6-g and 1-g gravity conditions for a system that can pneumatically convey 16.5 kg of lunar regolith simulant (NU-LHT-2M, Mauna Kea Tephra, and JSC-1A) from a flat-bottom supply hopper to a simulated ISRU reactor (dual-chambered receiving hopper) where the granular material is separated from the convey gas (air) using a series of cyclone separators, one of which is an electrically enhanced cyclone separator (electrocyclone). The results of our study include (1) the mass flow rate as a function of input air pressure for lunar regolith simulants that are conveyed pneumatically as a dusty gas in a vertical direction against gravity under lunar gravity conditions (for NU-LHT-2M and Mauna Kea Tephra), and under earth gravity conditions (for NU-LHT-2M, Mauna Kea Tephra and JSC-1A), and (2) the efficiency of the cyclone/electrocyclone filtration system in separating the convey gas (air) from the granular particulates as a function of particle size.

Mantovani, James G.↗

Irradiation Testing of Nuclear Flux Sensors - Presentation for 2021 ASI Annual Webinar

This power point presentation is a summary of the work produced in FY2021 under the Irradiation Testing of Neutron Flux Sensors project funded by DOE's Advanced Sensor Initiative, and the outlook for FY22. The objective of this portion of the ASI program is to "Test and demonstrate in-pile instrumentation in conditions similar to those expected to be seen in service, i.e., the conditions they would see in either in irradiation experiments supporting advanced reactors, or ultimately, in advanced reactors themselves". The presentation describes how neutron flux sensors were tested in four INL reactors plus Idaho State University's research reactor during FY2021, and the results obtained from these tests.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Assessment of UTP A 2133 Mn as a Matching Filler Metal for Alloy 800H in Section III, Division 5 Applications

Six alloys are qualified in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) for elevated-temperature nuclear components. Due to reactor design alloy restrictions and motivation to improve operational efficiency and plant lifetimes, it is necessary to maximize the usefulness of qualified alloys. One of these alloys is Alloy 800H, which is qualified for a maximum temperature of 760°C and a maximum 300,000-hour service life. However, the welding filler metals qualified to join Alloy 800H have stress rupture factors that reduce the allowable weldment strength by factors as low as 0.59 times the Alloy 800H base metal. In an effort to improve the weldment creep-rupture performance, or to increase the stress reduction factor, non-code qualified filler metals are under investigation by the Department of Energy?s Advanced Reactor Technologies program. For this investigation, UTP A 2133 Mn filler metal (Fe-Cr-Ni-Mn-Nb) was used to join 0.5 in. thick Alloy 800H plate to demonstrate passing an ASME BPVC Section IX weld qualification. The multi-pass, pulsed GTAW process showed successful weld qualification results, and additional property measurements were conducted to compare previous analyses of Alloy 82 and Alloy 617 filler metals. Future work will evaluate the cross-weld creep-rupture performance of the UTP A 2133 Mn in comparison to Alloy 82 and Alloy 617 filler metals.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Verification of Bison fission product species conservation under TRISO reactor conditions

When assessing the reliability and predictive capabilities of a simulation tool, code verification is used to ensure that the implemented numerical algorithm is a faithful representation of its underlying mathematical model, including partial differential or integral equations, initial and boundary conditions, and auxiliary relationships. During this process, numerical results in a discrete solution are compared to the analytical solution of the mathematical model. Here, in this paper, the code verification process is applied to one-dimensional spatiotemporal problems that exercise partial differential equation governing the conservation of fission product species (or mass diffusion). Numerical experiments were performed in the Bison fuel performance code to evaluate its predictive capability under various TRISO reactor conditions such as base irradiation and safety heating test conditions for either short- or long-lived fission product species, as well as a case concerning evaporation from the outer surface of a particle. The code predictions were compared with the expected exact results obtained from the analytical expressions, and the fact that they demonstrate the correct analytical behavior provides strong evidence of proper numerical algorithm implementation.

07 ISOTOPE AND RADIATION SOURCES↗

Demonstration and Verification of Thermo-Mechanical Bowing in a Limited Free-Bow SFR Concept Using MOOSE

Core bowing due to thermal gradients and irradiation induced swelling and creep introduces significant reactivity feedback effects in liquid metal-cooled fast spectrum reactors. During startup, normal operations, and transient events, differential thermal and flux gradients cause expansion, creep, and swelling, which results in bowing in preferential directions depending on local material properties and load-pad and restraint-ring design. The bowing phenomenon produces negative reactivity during accident conditions provided the restraint system has been properly designed to optimally guide the deformation outwards in the active core region. Under the Department of Energy Nuclear Energy Advanced Modeling and Simulation program, a Multiphysics Object Oriented Simulation Environment (MOOSE)-based multiphysics approach to model core bowing is being developed. The present work expands on previous modeling of simpler International Atomic Energy Agency (IAEA) verification problems with these tools by modeling IAEA Verification Problem 4, which involves a symmetric sector of a reactor core with ducted assemblies undergoing differential thermal expansion due to thermal gradients bowing outward with duct-to-duct contact. This model is available on the National Reactor Innovation Center Virtual Test Bed repository. This example was verified against benchmark participant results, which includes bowing deformation evaluation and duct-to-duct interactions at load pads with mechanical contact.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

A process to verify numerical models for seismic fluid-structure interaction in advanced reactor vessels

Seismic design and qualification of a liquid-filled advanced nuclear reactor will have to account for fluid-structure interaction (FSI). Interaction between the tank, internal components, and contained liquid will rely on analysis of numerical models that must be verified and validated. Here this study demonstrates a verification process for models of a base-supported cylindrical tank by comparing numerical predictions and analytical solutions. The numerical models are consistent with the assumptions made to derive analytical solutions, namely, either a rigid or a linear elastic tank, ideal fluid, and small-amplitude, unidirectional, horizontal inputs. One software platform is used to illustrate the process. Seismic FSI analysis is performed using the Arbitrary Lagrangian-Eulerian (ALE) and Incompressible Computational Fluid Dynamics (ICFD) solvers in LS-DYNA. Reported responses are those used for design, including hydrodynamic pressures on the tank wall, shear forces and moments at the tank base, and wave heights of the contained liquid. The accuracy of the numerical results is discussed. The numerical models are verified for calculating the pressures on the tank wall and reactions at its base. Accurate simulation of wave action is challenging for both solvers. Recommendations for modeling, code development, and steps for verification are provided. Although focused on reactor vessels and one software platform, the verification process described herein is broadly applicable to liquid-filled vessels and other finite element codes.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Multiphysics Running-In Simulations for Pebble-Bed Reactors with Griffin

Griffin, a Multiphysics Object-Oriented Simulation Environment (MOOSE)–based application targeting transient modeling of advanced reactors, has been used recently to model pebble-bed reactors (PBRs). The modeling effort has focused thus far on equilibrium core calculations. A new capability to simulate the running-in phase of PBR operation has been added to Griffin. This work demonstrates the new capability with a coupled multiphysics running-in simulation. Griffin computes power densities in the core at each time step of the running-in simulation and passes these to Pronghorn, which models fluid flow and heat transfer to calculate pebble surface temperatures. These surface temperatures are used along with the power densities in a heat conduction model to compute average fuel and moderator temperatures, which are passed back to Griffin and accounted for with temperature-dependent cross sections. This work also describes a novel methodology for determining appropriate pebble feed rates and control rod positioning during the running-in simulation. Furthermore, the RZ-geometry model used in this work requires minimal computational resources and can be used for optimization and uncertainty studies in future works.

Griffin↗

Development of the Underwater Disc Saw Used to Segment the Internals of Taiwan Research Reactor - 20038

Taiwan Research Reactor will be dismantled in the period of 2022 to 2027 according to the approved dismantling plan. Based on the sampling and analyzing results, some internals of Taiwan Research Reactor must be segmented under the water because of their higher radiation, and the calandria is one of them. To dismantle this calandria, Institute of Nuclear Energy Research develops an underwater disc saw that is driven by a hydraulic pump and will be used to segment the shell of the calandria. In this article, Taiwan Research Reactor and the structure of the calandria are briefly depicted, and the segmentation processes of this calandria and the design of this disc saw are also included. Besides, the trial of cutting the shell of the simplified full size calandria mockup is briefly demonstrated in this article. Further tests will be performed in the near future to figure out which parts or functions of this disc saw need to be modified to get better cutting performance. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Domain decomposition in the GPU-accelerated Shift Monte Carlo code

The GPU solver within the Shift continuous-energy Monte Carlo neutron transport code has been extended to provide domain decomposition in addition to domain replication to enable the solution of problems with memory requirements exceeding the capacity of a single GPU. The strategy follows the Multiple Set, Overlapping Domain (MSOD) approach that is used in Shift’s CPU solver and integrates into the event-based algorithm used for Shift’s GPU solver. Furthermore, the ability to assign processors to spatial domains non-uniformly has been maintained. In this work, two different approaches for communicating particle data between domains are considered, and multiple criteria for load balancing problems have been investigated. Numerical results are presented for both fresh and depleted small modular nuclear reactor (SMR) cores. A parallel efficiency of approximately 80% was achieved with up to 16 spatial domains measured relative to full domain replication. A scaling study on the Summit supercomputer demonstrates a weak scaling parallel efficiency of over 90% on over 24000 GPUs.

97 MATHEMATICS AND COMPUTING↗

Microstructural Engineering of Cu-Rich Nanoprecipitate formation in NiCoFeCrCu0.12 High-Entropy Alloy via Severe Plastic Deformation for Enhanced Irradiation Tolerance

This study demonstrates a defect-engineering approach for controlling Cu-rich precipitates in FeNiCrCoCu0.2 high-entropy alloys (Cu-HEAs), delivering a novel pathway for next-generation nuclear reactor materials with superior irradiation resistance. This work establishes that severe plastic deformation (SPD) processing via Shear Assisted Processing and Extrusion (ShAPE) and Friction Stir Layer Deposition (FSLD) creates dense dislocation networks and subgrain boundaries that fundamentally alter precipitation behavior under identical thermal treatments. Atom probe tomography (APT) indicates that SPD produces a metastable, atomically homogeneous solid solution that, upon moderate heat treatment (500°C/10 hour), develops remarkedly stronger Cu clustering than the as-cast counterpart. High-temperature exposure (800°C/100 h) produces near-pure Cu precipitates (~90 at% Cu) with significantly enhanced defect-sink efficacy in SPD-processed alloys: precipitate sizes of 50-60 nm and number densities of 2.7-3.8 × 10¹7 m?³, compared to 89 nm and 0.44 × 10¹7 m?³ in as-cast materials. Collectively, the findings establish defect-mediated precipitation control as a scalable, high-impact route to tailor sink density and distribution in HEAs, enabling microstructures optimized for irradiation tolerance and mechanical robustness in nuclear reactor environments.

Meher, Subhashish↗

Characterization of radiation damage in 3D printed SiC

The SiC fuel matrix for advanced gas-cooled high temperature reactors as part of the Transformational Challenge Reactor program serves as the fuel particle container structure, a barrier to fission gas release, and a heat transfer medium. Its performance is particularly important because the fuel matrix must demonstrate good structural stability and thermal behaviors. An additive manufacturing methodology combining a binder jet 3D printing process with chemical vapor infiltration (CVI) for the production of SiC was recently developed. In this study, post irradiation examination by transmission electron microscopy shows that defect accumulation within the printed particles is very similar to other forms of high-purity SiC. However, damage accumulation was not directly observed in the CVI matrix because black spot damage and dislocation loops are difficult to image within the nanoscale highly faulted CVI matrix and because interstitial defects may rapidly annihilate at the stacking faults. Therefore, electron energy loss spectroscopy (EELS) analysis was used to analyze defect swelling in both the printed particles and the CVI matrix. The EELS analysis helped reveal that the radiation-induced swelling in the CVI matrix is similar to that of the printed SiC particles. This work shows that 3D printed SiC has behavior that is comparable to SiC processed by other means and that 3D printing could serve as a suitable processing technique for high-purity SiC for nuclear applications.

36 MATERIALS SCIENCE↗

Alchemy: A Model-Based Approach for 2D to 3D Autonomous Nuclear System Design

Engineering design of nuclear power plant (NPP) piping and equipment systems frequently bypasses crucial 2D system planning, instead moving straight to 3D modeling. This often leads to designs that exceed building envelope constraints, forcing expensive and time-consuming redesigns. When 2D modeling is employed, it typically involves labor-intensive manual workflows that convert 2D drawings into 3D models, resulting in inefficiencies and errors across design iterations. These workflows further suffer from poor software interoperability and dependence on proprietary software ecosystems, thereby contributing to schedule delays and cost overruns. This paper presents Alchemy, an autonomous framework that transforms 2D system definitions into Industry Foundation Classes (IFC)-compliant 3D building information models (BIMs) for expediting nuclear facility design at the conceptual preliminary phase. Using a model-based approach, the framework treats the 2D system diagram as the central reference model employed to automatically generate all subsequent outputs, ensuring consistency between the system definition and the resulting physical design. A web-based interface enables engineers to define hierarchical system topologies including associated equipment, geometric properties, and connectivity requirements. A two-phase equipment layout optimization algorithm automatically computes collision-free spatial configurations within predefined building envelopes. An artificial intelligence (AI)-assisted pipe routing module then generates orthogonal, collision-free routing paths, allowing the user to select either an A* search-based method or an Ant Colony Optimization (ACO)-based method. All outputs are authored natively in IFC format, relying on open-source technologies and standardized formats in order to ensure extensibility and eliminate proprietary software dependencies. The proposed framework is validated on two representative pressurized-water reactor (PWR)-based case studies, for which it autonomously generates IFC-compliant 3D models in minutes, drastically reducing workflows that typically require hours of manual effort. The generated model demonstrates topologically correct equipment placement, physically plausible spatial relationships, and collision-free pipe routing consistent with known PWR loop configurations. This work represents a foundational step toward digital engineering for nuclear facility preliminary design, with future ongoing development targeting design code compliance and expanded system complexity.

97 - MATHEMATICS AND COMPUTING↗