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All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

ICME and In-Situ Process Monitoring for Rapid Qualification of Components Made by Laser-based Powder Bed Additive Manufacturing Processes for Nuclear Structural Applications

Additive manufacturing (AM) through selective powder bed melting of successive layers is being considered as disruptive technology for rapid production of low-cost nuclear reactor internal components with complex geometries. However, there is a potential for non-uniform distribution of physical features, such as porosity or microstructural differences, due to variations of temperature across a component build. These heterogenities make the deployment of traditional material qualification and non-destructive evaluation of AM components difficult. This research project explored the feasibility of using in-situ process monitoring methods and integrated computational materials engineering (ICME) principles as an alternate qualification methodology and approach. This project included six individual tasks: (i) design of artifacts relevant to the nuclear power industry, (ii) evaluation of laser processing and in-situ measurements, (iii) computational modeling, (iv) ex-situ microstructural characterization, (v) evaluation of scaling the methodology for large-scale structures, and (vi) development of a data package to codes and standards organizations. As a part of this research, laser powder bed fusion of metals (316L, Alloy 718, and Ti6Al4V) was explored. This document provides an overview of the research performed over three years and serves as the final U.S. Department of Energy report for this project. In the third year, the project focused on the following activities: (a) in-situ infrared and optical imaging of each layer during processing of representative stainless steel geometries; (b) development of methodologies to analyze the data; and (c) application of ICME methodologies and other experiments to estimate the relevance of defects and microstructure to tensile properties.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Modeling turbulent impurity transport in the SOL of DIII-D with a reduced model

A novel impurity transport model that approximates SOL turbulence as a fluctuating poloidal electric field is shown to be an acceptable replacement for the traditional approach of assigning an arbitrary radial diffusion coefficient to the impurity ions. The model is implemented in the DIVIMP impurity transport code and applied to an L-Mode tungsten divertor experiment on DIII-D. The poloidal electric field is represented as fluctuating between ±1000 V m –1 based on previous measurements. The resulting intermittent v r = E θ × B T transport causes ions to transport both into the core as well as into the far-SOL. Simultaneous agreement with estimates of the W density just inside the separatrix as well as in the far-SOL is obtained (n W ~ 10 14 m –3 and n W ~ 10 12 m –3 , respectively). Prompt re-deposition of the W ions was necessary to obtain agreement (f redep ~ 99%). Here, we conclude that simulating impurity transport using a physics-based approximation for turbulence in the SOL, versus arbitrarily assigning diffusion coefficients, may enable better reactor scale predictions of core impurity contamination.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Additive Manufacturing of Fe-based Alloys for Nuclear Reactor Environments

This report is a Milestone 2 deliverable in FY2025, under work package MT-25AN090211 to support research and qualification activities supported by the Advanced Materials and Manufacturing Technologies (AMMT) program. It provides an update on the work done on the laser based additive manufacturing of steels for nuclear applications, conducted collaboratively by Argonne National Laboratory and Pacific Northwestern National Laboratory. Building upon FY23 and FY24 efforts, the focus of FY25 includes continued optimization of process parameters for the alloys: A709 & G92. The work package also includes fabricating test samples to conduct a thorough microstructural analysis and perform preliminary mechanical testing. The major accomplishments of this work package are summarized below, applicable to both Laser Powder Bed Fusion (LPBF) processes conducted at Argonne National Laboratory and Laser Powder Directed Energy Deposition (LP-DED) processes conducted at Pacific Northwest National Laboratory.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Development of Monitoring Techniques for Laser Powder Bed Additive Manufacturing of Metal Structures (Progress Report)

The Transformational Challenge Reactor (TCR) program is leveraging additive manufacturing (AM) technologies to fabricate nuclear grade components to be assembled into a fully functional microreactor core. Compared to traditional manufacturing technologies, additive manufacturing technologies allow (1) observation of the manufacturing process at a much higher resolution in real time using in situ monitoring technologies to capture the sensor signature that scientifically describes each event occurring over time and space, and (2) validation of the manufacturing process quality using domain-informed data analytics techniques as a potential qualification and certification methodology for the final component. This report provides an update on the program work on laser powder bed fusion in-situ process monitoring and associated data analytics results. Examples are provided to illustrate the progress. Elements of the Digital Thread and data management are discussed in the main document, and an extensive supplemental material section is provided detailing the Digital Platform, as well as its implementation and components. In conclusion the path forward for the next fiscal year is discussed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Investigation of radiated-power for low aspect ratio fusion plasmas

Investigating radiated power in fusion plasmas is of utmost importance to understand the effect of undesired impurities, such as metals, in present devices, or also desired impurities, such as noble gasses, to purposefully radiate a large fraction of power in future devices. These studies are especially important for high Z impurities which will play a crucial role in future generation fusion pilot plants. In this work, we have developed a power radiation analysis module, which is used to investigate 2D distributions of impurity densities and radiated power asymmetries caused due to both plasma rotation and the cooling rate dependence on temperature profile for the cases of experimental NSTX plasmas and designed scenarios for the spherical tokamak advanced reactor (STAR), both being low aspect ratio tokamaks. Two different atomic databases have been tested during this work to study their impact on the radiated power distribution, especially for high Z impurities. Also in PRAM, self-consistent calculations of two-dimensional electron, main ion, and impurity ion densities are carried out using one-dimensional input density, temperature, and rotation profiles. In the case of NSTX, discharges with high rotation of ∼ 170 km s, measured with charge exchange recombination spectroscopy, have been investigated. Rotation-induced charge separation, leading to an electrostatic potential, is calculated iteratively to a self-consistent solution while testing high Z impurities to observe any 2D asymmetry in the core radiated power due to centrifugal forces. The STAR design, being much larger (R = 4 m), is projected to have a much lower rotation, and is shown to have low rotation-induced asymmetries, on the order of ten percent or less, between the low field and high field sides. However, another effect not due to rotation but to the dependence of impurity cooling rates on temperature can lead to radiation peaking off-axis, near the plasma edge. This effect is noticeable for argon in NSTX, for example, but can also be enhanced for certain impurities at much higher temperatures projected for STAR (T e0 ~ 32 keV), for example for undesired tungsten or possibly desired xenon.

NSTX↗
DOE OSTI2020

Improvement on the plasma performances via application of flowing lithium limiters in EAST tokamak

Three generation of flowing liquid lithium (FLiLi) limiters have been developed and successfully tested in EAST since 2014. It was confirmed that liquid Li could be driven by an innovative in-vessel DC electromagnetic J × B pump to form a recirculating loop. The FLiLi design was continuously upgraded, resulting in improvement of the liquid Li surface coverage (>80%), elimination of surface erosion, and the increase of heat exhaust capability. Promising results are obtained during FLiLi operations, including a reduction of impurities and recycling to improve plasma confinement, and a mitigation of ELM activity in H-mode plasmas. The FLiLi limiter program has shown basic compatibility with H-mode discharges with auxiliary heating power up to 8.3 MW. These efforts will expand to flowing liquid Li PFC in high heat flux zones, e.g. near the divertor strike point in reactors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Experimental, numerical and analytical evaluation of j × B -thrust for fast-liquid-metal-flow divertor systems of nuclear fusion devices

Abstract Divertor systems of fusion devices are exposed to intense heat loads from plasmas, which degrade solid plasma-facing components. Fast liquid metal (LM) flow divertors may be more advantageous for this purpose but have risk of piling due to intense magnetohydrodynamic (MHD) drag. However, severe deceleration of the flow could be countered with the injection of currents that are transverse to external magnetic fields, allowing to thrust the flow with j × B (Lorentz) forces. Given that the injection of currents as an approach to propel LM-divertor flows has remained experimentally understudied, this article focuses on the evaluation of j × B -thrust and finding its drawbacks. j × B -thrust was experimentally tested with free-surface-LM flows, a vertical magnetic field and an externally applied current. Experiments were reviewed with a theoretical model, showing agreement in the trends of theory and experiments. Full 3D-MHD-free-surface-flow simulations were also performed with FreeMHD and confirmed the sensitivity to unstable flow behavior in LM systems when applying external currents. Furthermore, excessive power requirements are expected for the implementation of j × B -thrust at the reactor scale, making these systems inefficient for commercial devices. This paper evidences that the simple operation of a LM-flow divertor with j × B -thrust, without any of the instabilities caused from reactor plasmas or parasitic currents, already presents intrinsic challenges.

Magnetohydrodynamics↗
DOE OSTI2023

Laser Ablation Plasmas and Spectroscopy for Nuclear Applications

The development of measurement methodologies to detect and monitor nuclear-relevant materials remains a consistent and significant interest across the nuclear energy, nonproliferation, safeguards, and forensics communities. Optical spectroscopy of laser-produced plasmas is becoming an increasingly popular diagnostic technique to measure radiological and nuclear materials in the field without sample preparation, where current capabilities encompass the standoff, isotopically resolved and phase-identifiable (e.g., UO and UO[Formula: see text]) detection of elements across the periodic table. These methods rely on the process of laser ablation (LA), where a high-powered pulsed laser is used to excite a sample (solid, liquid, or gas) into a luminous microplasma that rapidly undergoes de-excitation through the emission of electromagnetic radiation, which serves as a spectroscopic fingerprint for that sample. This review focuses on LA plasmas and spectroscopy for nuclear applications, covering topics from the wide-area environmental sampling and atmospheric sensing of radionuclides to recent implementations of multivariate machine learning methods that work to enable the real-time analysis of spectrochemical measurements with an emphasis on fundamental research and development activities over the past two decades. Background on the physical breakdown mechanisms and interactions of matter with nanosecond and ultrafast laser pulses that lead to the generation of laser-produced microplasmas is provided, followed by a description of the transient spatiotemporal plasma conditions that control the behavior of spectroscopic signatures recorded by analytical methods in atomic and molecular spectroscopy. High-temperature chemical and thermodynamic processes governing reactive LA plasmas are also examined alongside investigations into the condensation pathways of the plasma, which are believed to serve as chemical surrogates for fallout particles formed in nuclear fireballs. Laser-supported absorption waves and laser-induced shockwaves that accompany LA plasmas are also discussed, which could provide insights into atmospheric ionization phenomena from strong shocks following nuclear detonations. Furthermore, the standoff detection of trace radioactive aerosols and fission gases is reviewed in the context of monitoring atmospheric radiation plumes and off-gas streams of molten salt reactors. Finally, concluding remarks will present future outlooks on the role of LA plasma spectroscopy in the nuclear community.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Demonstration of RELAP5-3D for transient analysis of a dual coolant lead lithium fusion blanket concept

Our research aims to build a multiphysics framework for transient analysis of the Dual- Coolant Lead-Lithium (DCLL) blanket design of the proposed Fusion Nuclear Science Facility (FNSF) for all Onticipated Operational Occurrences (AOOs) using Reduced Order Models (ROMs). This paper presents an assessment of using RELAP5-3D for transient thermal-hydraulic blanket analysis of the DCLL design. Preliminary validation studies for the RELAP5-3D properties were conducted using a representative vertical flow loop model and a comparison study using experimental data from the thermal convection corrosion loops at ORNL. These studies inspected the thermal hydraulic response of Lead-Lithium Eutectic (PbLi) systems within the RELAP5-3D code; mainly pressure drop and heat transfer, as compared to analytical calculations and experimental data. The Magnetohydrodynamic (MHD) pressure drop effect for liquid metal under the effects of a uniform magnetic field was implemented into our models using an equivalent forms loss coefficient based on the phasic momentum equations. We verified our implementation of MHD pressure drop within RELAP5-3D using a systematic approach consisting of two studies. Models were developed to perform comparison studies based on experimental data from the Argonne Liquid metal EXperiment (ALEX) facility and the Magnetohydrodynamic PbLi Experiment (MaPLE) facility. Using the developed validation basis, we built a simplified DCLL channel model within RELAP5-3D, inspected MHD pressure drop and performed transient temperature analysis. The DCLL channel model used radial heating data from the International Thermonuclear Experimental Reactor (ITER) which was normalized to the FNSF design. Here, a representative startup transient was performed using a relative power curve from the initial ramp up of a long-term power pulse of the Tore Supra reactor. This study demonstrated that RELAP5-3D ROMs are capable of accurate transient analysis of the DCLL blanket. We have since utilized MCNP to develop heating profiles based on the FNSF design and are working toward implementing them into our DCLL model.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Progress Towards a Titanium-Zirconium-Molybdenum Alloy Coreblock Prototype Using Powder Bed Fusion

Microreactor designs make use of a coreblock for structural stability allowing for these devices to provide powder generation while being portable and inherently safe. In order to maximize energy output of the microreactors, materials must be able to withstand high temperature without melting and without substantial decreases in mechanical properties. As simultaneous projects looking at steels, graphite, and other materials are on-going, this work package will investigate titanium – zirconium – molybdenum alloy (TZM) as it allows for potentially higher microreactor operating temperatures than traditional reactor materials such as stainless steel. Notably, this alloy was developed in the 1960s for use in rockets and has a very high melting point (2623°C) but is difficult to form, machine and join. With limited needs in aerospace, literature is appropriately sparce in looking at the AM of TZM (examples include [1,2]). However, the properties appear to be an excellent match for the needs of the microreactor coreblock leading to the current project. In addition, small washers form the bulk of TZM parts needed in aerospace while the microreactor community was looking at larger, more complex parts. Therefore, AM of TZM is being explored to enable near – net shape of components with the needed complexity to impact the nuclear energy field.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Radiation effects on stress evolution and dimensional stability of large fusion energy structures

Here we assess the effects of neutron irradiation on the deformation and stress evolution of large-scale fusion energy structures. This is accomplished through non-linear finite element structural analysis of the coupled thermal and mechanical fields at the Beginning-Of-Life (BOL), at 45 dpa, and at 90 dpa. Radiation effects include volumetric swelling and the influence of radiation on the mechanical properties. The system studied here is a large section of a full inboard module of an integrated structure comprising the First Wall and Blanket (FW/B) of a Dual Cooled Lithium-Lead (DCLL) energy conversion unit in the Fusion Nuclear Science Facility (FNSF). The structural material is the ferritic/martensitic steel F82H. We analyze several radiation effects phenomena that can lead to significant impact on the mechanical design and lifetime of the structure. These include volumetric swelling effects on the assembly and disassembly of modules and the impact of spatial gradients in neutron damage and helium generation on deformation, stress evolution, and plastic strain accumulation. We show that the stress state is strongly influenced by the accumulation of swelling strains, radiation hardening and softening, and by spatial gradients in displacement damage and helium gas generation. Several key regions in the FW/B structure are identified where severe plastic strains accumulate and may be potential sites for failure, especially during cyclic reactor operations. We finally show that minimum gaps between Inboard (IB) blanket sectors must be present during assembly so as to avoid excessive stress and strain in FW/B modules critical regions and raising chances of failure. A range of structural displacements are shown, from as low as 4 mm at BOL for free side walls to as high as 46 mm at 90 dpa for fully constrained side walls.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Passive deconfinement of runaway electrons using an in-vessel helical coil

A helical coil designed to passively generate non-axisymmetric fields during a plasma disruption is shown (via electromagnetic analysis, linear MHD modeling, and relativistic drift orbit tracing) to be effective at deconfining runaway electrons (REs) on a time scale significantly faster than the plasma current quench. Magnetic equilibria from DIII-D RE-producing scenarios are used to calculate the toroidal electric field generated during the current quench phase of a disruption, which in turn drives current in the proposed n = 1 in-vessel helical coil, without the need for any external power supplies or disruption detection or prediction techniques. Simulations of the plasma evolution using the TokSys GS Evolve code predict the inductive coupling of coil currents up to 12% of the pre-disruption plasma current into the helical coil. The coil geometry is parametrically varied to maximize both the non-resonant and resonant components of the 3D magnetic perturbation, resulting in δB/B ≈ 10 –2 and a vacuumisland overlapwidth of up to 0.7ψ N . The REORBIT module of the MARS-F code is used to model the full non-axisymmetric magnetic field and trace RE drift orbits to determine the effect on RE deconfinement, with up to 70% of the RE orbits lost after 0.2 ms. A two-stage evolution of the RE orbit loss fraction is observed to be caused by resonant trapping between multiple magnetic island chains. Finally, electromagnetic and thermal stresses on the coil are calculated to be within operational limits for installation in DIII-D, and scale favorably to a reactor-size device. Furthermore, these findings motivate future experimental study of the helical coil concept in DIII-D or other tokamaks.

3D coil↗
DOE OSTI2023

Robust Avoidance of Edge-Localized Modes alongside Gradient Formation in the Negative Triangularity Tokamak Edge

In a series of high performance diverted discharges on DIII-D, we demonstrate that strong negative triangularity (NT) shaping robustly suppresses all edge-localized mode (ELM) activity over a wide range of plasma conditions: ⟨n⟩ = 0.1 − 1.5 × 1020 m−3, Paux = 0 − 15 MW and |Bt| = 1 − 2.2 T, corresponding to Ploss/PLH08 ∼ 8. The full dataset is consistent with the theoretical prediction that magnetic shear in the NT edge inhibits access to ELMing H-mode regimes; all experimental pressure profiles are found to be at or below the infinite-n ballooning stability limit. Our present dataset also features edge pressure gradients in strong NT that are closer to an H–mode than a typical L-mode plasma, supporting the consideration of NT for reactor design.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS2003

Pulsed Magnetic Field Driven Gas Core Reactors for Space Power & Propulsion Applications

The present results indicated that: 1. A pulsed magnetic driven fission power concept, PMD-GCR is developed for closed (NER) and semi-open (NTR) operations. 2. In power mode, power is generated at alpha less than 1 for power levels of hundreds of KW or higher 3. IN semi open NTR mode, PMD-GCR generates thrust at I(sub sp) approx. 5,000 s and jet power approx. 5KW/Kg. 4. PMD-GCR is highly subcritical and is actively driven to critically. 5. Parallel path with fusion R&D needs in many areas including magnet and plasma.

Anghaie, Samim↗
DOE OSTI2023

Observations of xenon spectra on the EAST x-ray crystal spectrometer for high-temperature plasma diagnostics

The Xe 44+ 2.7203 Å line, which has been proposed as one of the diagnostic lines for the x-ray imaging crystal spectrometer on ITER, is observed on the EAST tokamak together with its several satellite lines. The observations are made under high electron temperature (T e ) conditions (core T e > 5 keV). Most of the observed xenon lines are identified by comparing the experiment results with the atomic simulation results. The first ion temperature measurements made by the xenon spectra on EAST are also reported in this article. These xenon spectra observations contribute to the justification for using xenon as the diagnostic impurity in x-ray crystal spectrometers in future reactor-scale high-temperature plasmas.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Direct optimization of neoclassical ion transport in stellarator reactors

Abstract We directly optimize stellarator neoclassical ion transport while holding neoclassical electron transport at a moderate level, creating a scenario favorable for impurity expulsion and retaining good ion confinement. Traditional neoclassical stellarator optimization has focused on minimizing ϵ eff , the geometric factor that characterizes the amount of radial transport due to particles in the 1 / ν regime. Under expected reactor-relevant conditions, core electrons will be in the 1 / ν regime and core fuel ions will be in the ν regime. Traditional optimizations thus minimize electron transport and rely on the radial electric field ( E r ) that develops to confine the ions. This often results in an inward-pointing E r that drives high- Z impurities into the core, which may be troublesome in future reactors. In this work, we increase the ratio of the thermal transport coefficients L 11 e / L 11 i , which previous research has shown can create an outward-pointing E r . This effect is very beneficial for impurity expulsion. We obtain self-consistent density, temperature, and E r profiles at reactor-relevant conditions for an optimized equilibrium. This equilibrium is expected to enjoy significantly improved impurity transport properties.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Synergistic effects of W–Ta alloys exposed to various ELM-like fusion events

Abstract Recently tungsten alloys have shown capabilities in suppressing morphological changes when exposed to harsh fusion conditions. Morphological changes due to He + and D + ions at high temperatures (1000–2000 K) can lead to easier erosion and sputtering of tungten (W) particles towards quenching the fusion plasma. This study investigated W–3Ta alloy as potential plasma facing materials (PFMs) exposed to different edge-localized mode (ELM)-like energies and ion loadings. Performance of W–3Ta has been tested as a function of ELM-like energies (0.6–1.5 GW m −2 ) and varying He + and D + ion concentrations (100% He + , 50% He + :50% D + ,10% He + :90% D + (similar ELM-like reactor condition), and 100% D + ions). Early-stage fuzz formation and surface pores were observed during 100% He + ion irradiation and mixed ion loadings but not observed during 100% D + ion loadings. Pores decreased in quantity and increased in size with increasing transient heat loading. Grain boundaries acted as trapping sites of helium and deuterium atoms during dual beam ion irradiations, as observed through increased pore size along the grain boundary. Qualitative erosion estimates were measured using witness plates. Increased erosion was observed with increasing D + ion concentration, which is attributed to D supersaturation in the near surface region affecting thermal and mechanical properties. More erosion was observed for W–3Ta samples compared to pure W, evident through large (>10 µ m) cracks on the surface, and an increase in W metal x-ray photoelectron spectroscopy peaks under similar loading conditions. The slight enhancement in erosion is attributed to the slightly decreased thermal properties, ∼10% thermal diffusivity drop, of W–3Ta at high temperatures. The damage threshold was impacted by ion loading, where distinct cracking was observed for 100% He + at a lower heat loading. This matrix study using reactor-like loading supports the need for extensive studies on future W alloy PFMs before selecting one as an alternative.

Physics↗
NASA NTRS1972

Post impact behavior of mobile reactor core containment systems.

In the future, nuclear assemblies containing fission products will be transported at high speeds. An example is a reactor supplying power to a large subsonic airplane. In this case an accident can occur resulting in a ground impact at speeds up to 1000 ft/sec. This paper analyzes the containment vessel temperatures after impact and attempts to understand the design variables that affect the post impact survival of the system. The heat transfer analysis includes conduction, radiation, and convection in addition to the core material heats of fusion and vaporization under partial-burial conditions. Also, included is the fact that fission products vaporize and transport radially outward and condense on cooler surfaces, resulting in a moving heat source.

Puthoff, R. L.↗