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DOE OSTI2017

Fusion Materials Research at Oak Ridge National Laboratory in Fiscal Year 2017

Achieving fusion energy presents a formidable challenge to the field of materials science, to provide a suite of suitable materials in close integration to the plasma physics and other applied technologies. The technological challenge presented by the planned fusion power systems is the inability of current materials and components to withstand the harsh fusion nuclear environment. The overarching goal of the ORNL fusion materials program is to provide the applied materials science support and materials understanding to underpin the ongoing DOE Office of Science fusion energy program in parallel with developing materials for fusion power systems. In this effort the program continues to be integrated both with the larger U.S. and international fusion materials communities and with the international fusion design and technology communities. This long-running ORNL program continues to pursue development of low activation structural materials such as the Reduced Activation Ferritic/Martensitic Steels, higher strength/higher creep resistant/coolant compatible/radiation tolerant advanced steels, and Silicon Carbide Composites. Focus tasks within the steels portfolio are development of Castable Nanostructured Alloys, exploratory work on Bainitic steels and liquid metal compatibility evaluations. Parallel to this is the increased emphasis on radiation effects, high heat flux testing and the development of refractory metals. This includes the use of an ORNL Plasma Arc Lamp facility adapted for the thermal testing of irradiated materials, the development and evaluation of new tungsten materials, and the study and understanding of the irradiation performance of tungsten. In each case the materials are being developed in a design-informed fashion where properties improvements are led by fusion-relevant design studies and directed at advancing the Technology Readiness Level of the material systems.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Fusion Materials Research at Oak Ridge National Laboratory in Fiscal Year 2018

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

Achievement of ion temperatures in excess of 100 million degrees Kelvin in the compact high-field spherical tokamak ST40

Abstract Ion temperatures of over 100 million degrees Kelvin (8.6 keV) have been produced in the ST40 compact high-field spherical tokamak (ST). Ion temperatures in excess of 5 keV have not previously been reached in any ST and have only been obtained in much larger devices with substantially more plasma heating power. The corresponding fusion triple product is calculated to be n i 0 T i 0 τ E ≈ 6 ± 2 × 10 18 m − 3 keVs . These results demonstrate for the first time that ion temperatures relevant for commercial magnetic confinement fusion can be obtained in a compact high-field ST and bode well for fusion power plants based on the high-field ST.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

Ab initio study of tungsten-based alloys under fusion power-plant conditions

Tungsten (W) is considered a leading candidate for structural and functional materials in future fusion energy devices. The most attractive properties of tungsten for magnetic and inertial fusion energy reactors are its high melting point, high thermal conductivity, low sputtering yield, and low long-term disposal radioactive footprint. However, tungsten also presents a very low fracture toughness, primarily associated with intergranular failure and bulk plasticity, limiting its applications. In recent years, several families of tungsten-based alloys have been explored to overcome the aforementioned limitations of pure tungsten. These include tungsten-based high-entropy alloys (W-HEAs) and tungsten-based Self-passivating Metal Alloys with Reduced Thermo-oxidation or “SMART alloys” (W-SAs). Given their proximity to the plasma, it is crucial to understand how the exposure of these candidate plasma-facing materials (PFMs) to the neutron fluxes expected in fusion reactors impacts their material behavior over time. In this work, we present a computational approach that combines inventory codes and first-principles DFT electronic structure calculations to understand the behavior of transmuting tungsten-based PFMs. In particular, we calculate the changes in the chemical composition, production uncertainties, the elastic and ductility properties, and the density of states for five tungsten-based PFMs when exposed to EU-DEMO fusion first wall conditions for ten years.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2015

Fusion Materials Research at Oak Ridge National Laboratory in Fiscal Year 2015

The realization of fusion energy is a formidable challenge with significant achievements resulting from close integration of the plasma physics and applied technology disciplines. Presently, the most significant technological challenge for the near-term experiments such as ITER, and next generation fusion power systems, is the inability of current materials and components to withstand the harsh fusion nuclear environment. The overarching goal of the Oak Ridge National Laboratory (ORNL) fusion materials program is to provide the applied materials science support and understanding to underpin the ongoing Department of Energy (DOE) Office of Science fusion energy program while developing materials for fusion power systems. In doing so the program continues to be integrated both with the larger United States (US) and international fusion materials communities, and with the international fusion design and technology communities.This document provides a summary of Fiscal Year (FY) 2015 activities supporting the Office of Science, Office of Fusion Energy Sciences Materials Research for Magnetic Fusion Energy (AT-60-20-10-0) carried out by ORNL. The organization of this report is mainly by material type, with sections on specific technical activities. Four projects selected in the Funding Opportunity Announcement (FOA) solicitation of late 2011 and funded in FY2012-FY2014 are identified by “FOA” in the titles. This report includes the final funded work of these projects, although ORNL plans to continue some of this work within the base program.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2018

Master Curve Fracture Toughness Characterization of Eurofer97 Using Miniature Multi-Notch Bend Bar Specimens for Fusion Applications

Eurofer97 is one of leading candidates of reduced activation ferritic martensitic (RAFM) steels for first wall structural materials of early demonstration fusion power plants. During fusion plant operation, high neutron irradiation damage on first wall materials can cause irradiation embrittlement and reduce the fracture toughness of RAFM steels. Therefore, it is critical to select proper testing techniques to characterize the fracture toughness of RAFM steels with high fidelity. In this manuscript, we present the feasibility study of using pre-cracked miniature multi-notch bend bar specimens (M4CVN) with a dimension of 45mm (length) × 3.3mm (width) × 1.65mm (thickness) to characterize the transition fracture toughness of Eurofer97 steel based on the ASTM E1921 Master Curve method. The testing yielded a provisional Master Curve reference temperature ToQ of −89°C of unirradiated Eurofer97 steel heat J362A in the normalized and tempered condition. The results are within the normal scatter range of Master Curve reference temperature T0 for Eurofer97 steel, indicating suitability of applying M4CVN specimens for characterizing the transition fracture toughness of Eurofer97 steel.

Chen, Xiang↗

DOE OSTI2023

PFPO plasma scenarios for exploration of long pulse operation in ITER

Long Pulse Scenarios (LPS) in ITER foreseen during the Pre-Fusion Power Operation (PFPO) phase of the ITER Research Plan (IRP) are assessed using 1.5D transport simulations within the ASTRA framework. Such assessment is required to predict the operational space for LPS operation in PFPO, as well as to evaluate which physics processes for LPS operation during Fusion Power Operation (FPO) could be studied during PFPO. An important aspect in the development of LPSs in PFPO is to minimize lifetime consumption of the Central Solenoid (CS) for these scenarios. The maximum pulse length achievable for LPSs in PFPO with no consumption of CS lifetime (currents in CS coils $&LessSlantEqual;$30 kA per turn) has been assessed for a range of heating schemes and heating mixes, confinement regimes (L-mode and H-mode) and for helium and hydrogen plasmas. The operational space of LPS and pulse length has been explored through density scans with the Heating and Current Drive mix required for the FPO Q $&GreaterSlantEqual;$ 5 steady-state plasma scenario (namely Neutral Beam Injection and Electron Cyclotron Heating) including acceptable shine through losses on the first wall for both helium and hydrogen plasmas. Fast particle physics aspects that are common between FPO plasmas and LPS PFPO H-mode plasmas at low densities are studied including MHD stability analysis with the KINX code and non-perturbative critical gradient model based on high-n Toroidal Alfven Eigenmodes (TAE) stability kinetic ballooning code HINST calculations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2026

Stellarator Design Exploration Using Symbolic-Regression Neutronics Surrogates

Systems codes require fast, simplified models to rapidly evaluate fusion power plant concepts, but neutronics analyses are often a computational bottleneck. Here, to address this, surrogate models for key neutronics responses have been developed using 3-D neutronics-ready models built with the open-source code ParaStell from a database of stellarator equilibria. Neutronics responses such as tritium breeding ratio (TBR), nuclear heating, and neutron-induced radiation damage displacements per atom (dpa) were simulated using OpenMC. Through sensitivity analysis and symbolic regression (SR), simple power-law formulas were derived connecting these neutronics responses to global stellarator parameters, including fusion power, plasma surface area, and plasma elongation. Validation shows these formulas can predict the simulation results with low error, enabling quick and accurate assessment of neutronics requirements in stellarator design exploration activities with systems codes.

Modeling↗

DOE OSTI2014

Neutronics for diagnostics in ITER - Paper 121

ITER, currently under construction in the south of France, will be the first tokamak producing large amounts of neutrons and having a fusion energy gain factor, which is the ratio between the fusion power produced and the input power necessary to maintain the plasma, greater than 10. ITER will be able to generate 500 MW of fusion power, from which 400 MW are 14.1 MeV neutrons from the Deuterium-Tritium fusion reactions in the thermonuclear plasma. For such nuclear machine neutronics are of paramount importance, being a design driver for many of its components, in particular its Ports and its numerous Diagnostics, for which the influence of neutronics is especially important in three main areas: - Shielding: there is a strong need to provide nuclear shielding to areas where human access is foreseen for maintenance. Also, to minimize any impact in the primary shielding components (blankets) in order to avoid significant increases in the heating of the Vacuum Vessel (VV) and Superconductive Coils. - Sensors and diagnostic components: most of the diagnostics have sensitive components that need to be shielded against radiation to survive. Efficient shielding is not always possible, and the radiation levels are usually the decisive factor for selecting the most appropriate component. - Neutron measurements: a large set of neutron diagnostics will be implemented to measure the neutron production and its emissivity profile. Although the scale and performance of ITER will not be very far from a commercial reactor, it is ultimately an experiment to improve the understanding of reactor-grade tokamak plasmas and their effect on plasma facing components. To accomplish this task, large efforts are being dedicated to design over 45 different diagnostics systems. This paper gives an overview of the main activities related to neutronics within the ITER Diagnostics Division. They include the neutron streaming problems in the Ports and the common neutronics issues of the diagnostics housed inside them, as well as the challenges of the in vessel diagnostics. Also the special calculations needed for the neutron diagnostics. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2025

Neutronics analysis of spin-polarized fuel in spherical tokamaks

The effect of spin-polarized deuterium–tritium fuel on the blanket, magnets, and surrounding structures is analyzed for a spherical tokamak fusion power plant. The focus of this paper is the neutron flux distribution in the fusion device and the corresponding engineering considerations, such as tritium breeding ratio (TBR) and magnet fluence. Significant benefits were found for anti-aligned polarization: a 2.7% increase in the total TBR (which could increase significantly with further optimization) and an approximately 68% increase in the magnet lifetime. If the inboard breeding blanket can be replaced with extra shielding and magnets, then it could significantly increase fusion power with no loss of magnet lifetime. This would be accomplished by increasing the toroidal field strength.

fusion↗

DOE OSTI2022

The LIBRA Experiment: Investigating Robust Tritium Accountancy in Molten FLiBe Exposed to a D-T Fusion Neutron Spectrum

The Liquid Immersion Blanket: Robust Accountancy (LIBRA) experiment will be a first-of-a-kind experiment to explore and develop the liquid immersion blanket (LIB) concept. The LIB is a radically simple molten–LiF-BeF2 (FLiBe)–salt tritium breeding blanket for deuterium-tritium (D-T)–fueled fusion power plants (FPPs) achieving a high tritium breeding ratio (TBR) in neutronics models. However, tritium breeding in FLiBe is inherently difficult to study experimentally. As a result, the coupled issues of FLiBe radiochemistry and tritium (T) transport are poorly understood. LIBRA approaches this challenge by simulating an FPP blanket environment using a D-T neutron generator and ≈ 1000 kg of FLiBe. LIBRA will investigate T breeding, containment, and extraction, coupled with FLiBe redox control and radiochemistry. The primary goal of LIBRA is to demonstrate robust T accountancy in blanket prototypical conditions. Here, T accountancy encompasses accurate predictions of T breeding in the FLiBe; detection and measurement of all T bred in LIBRA; and speciation of the T extracted from the FLiBe. Initial neutronics simulations of LIBRA indicate that a global TBR of ~ 1 is possible, where the TBR is defined as the number of tritons bred and extracted from FLiBe relative to the number of neutrons produced by D-T fusion reactions in the neutron generator. In this paper, we present the LIBRA concept and its scientific goals in the context of T breeding experiments. We also consider the potential impact of the LIB on the future fusion power industry, motivating further development of FLiBe-based T breeding research activities such as LIBRA.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2014

Nuclear Analysis of ITER - Paper 129

The world record fusion power production is 16.1 MW (5.7x10{sup 18} n/s). This was achieved at JET in 1995. The neutron budget during that experimental campaign was 2.5x10{sup 20} neutrons. In the first few seconds of an ITER full power pulse, the neutron production will exceed that the total from the entire history of previous controlled fusion experiments. It is envisaged that ITER should produce as much as 700 MW of fusion power. (This is the maximum envisaged power production from the plasma and does not include power multiplication in the blankets or shielding). This equates to the production of 2.48x10{sup 20} 14-MeV neutrons/s which will give an uncollided flux at the first wall of approximately 4x10{sup 13} n/cm{sup 2}/s and a total with the addition of the collided flux of some 10{sup 14} n/cm{sup 2}/s. For safety and licensing purposes it is the responsibility of ITER to satisfy the host state regulations and must limit doses to the public and staff. Furthermore, sufficient shielding must be provided to ensure that all components function adequately; and for decommissioning, ITER has a responsibility to reduce the volume and radio-toxicity of waste. The construction, operational and decommissioning phases of ITER are expected to last in excess of 30 years and the nuclear issues will be of critical importance throughout this time. The huge leap in neutron production in fusion experiments means that extensive nuclear analysis is required for ITER and it is the thesis of this paper that it is required at an unprecedented level. This is because of the combination of the large physical size of the installation, the variety of radiation sources, the attenuation required, the complexity of the geometry and the complexity of the sources and level of detail required. As a result of this unprecedented challenge the latest techniques for analysis, variance reduction and data display are being employed and in some cases being specifically developed. In this paper a few of the many representative examples of nuclear analysis will illustrate the thesis. The paper includes an examination of areas where further development could be made in the areas of radiation shield and nuclear analysis. Extensive nuclear analysis is required at ITER and the combination of the large physical size of the installation, the variety of radiation sources, the attenuation required, the complexity of the geometry, and the complexity of the sources mean that analysis is required at an unprecedented level of detail and quality. The latest techniques for analysis, variation reduction and data display are being employed and in some cases being specifically developed. Developments are required in areas of analysis, data visualization and experimental verification. (author)

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2019

Overview of the JET preparation for deuterium–tritium operation with the ITER like-wall

For the past several years, the JET scientific programme (Pamela et al 2007 Fusion Eng. Des. 82 590) has been engaged in a multi-campaign effort, including experiments in D, H and T, leading up to 2020 and the first experiments with 50%/50% D–T mixtures since 1997 and the first ever D–T plasmas with the ITER mix of plasma-facing component materials. For this purpose, a concerted physics and technology programme was launched with a view to prepare the D–T campaign (DTE2). This paper addresses the key elements developed by the JET programme directly contributing to the D–T preparation. This intense preparation includes the review of the physics basis for the D–T operational scenarios, including the fusion power predictions through first principle and integrated modelling, and the impact of isotopes in the operation and physics of D–T plasmas (thermal and particle transport, high confinement mode (H-mode) access, Be and W erosion, fuel recovery, etc). This effort also requires improving several aspects of plasma operation for DTE2, such as real time control schemes, heat load control, disruption avoidance and a mitigation system (including the installation of a new shattered pellet injector), novel ion cyclotron resonance heating schemes (such as the three-ions scheme), new diagnostics (neutron camera and spectrometer, active Alfvèn eigenmode antennas, neutral gauges, radiation hard imaging systems...) and the calibration of the JET neutron diagnostics at 14 MeV for accurate fusion power measurement. The active preparation of JET for the 2020 D–T campaign provides an incomparable source of information and a basis for the future D–T operation of ITER, and it is also foreseen that a large number of key physics issues will be addressed in support of burning plasmas.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2012

The neutronics studies of fusion fission hybrid power reactor

In this paper, a series of neutronics analysis of hybrid power reactor is proposed. The ideas of loading different fuels in a modular-type fission blanket is analyzed, fitting different level of fusion developments, i.e., the current experimental power output, the level can be obtained in the coming future and the high-power fusion reactor like ITER. The energy multiplication of fission blankets and tritium breeding ratio are evaluated as the criterion of design. The analysis is implemented based on the D-type simplified model, aiming to find a feasible 1000MWe hybrid power reactor for 5 years' lifetime. Three patterns are analyzed: 1) for the low fusion power, the reprocessed fuel is chosen. The fuel with high plutonium content is loaded to achieve large energy multiplication. 2) For the middle fusion power, the spent fuel from PWRs can be used to realize about 30 times energy multiplication. 3) For the high fusion power, the natural uranium can be directly used and about 10 times energy multiplication can be achieved.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Radiation Damage Analysis of FNSF Components Using McCad and MCNP

The Fusion Energy System Studies Fusion Nuclear Science Facility (FESS-FNSF) concept represents a transitional step between ITER and a commercial fusion power plant. The FNSF is a conceptualized D-T fueled tokamak with 518 MW of fusion power that has been extensively used to explore and optimize design features. The energetic 14.1-MeV neutrons can produce significant localized heating and activations, and can cause damage to plasma-facing components, which can determine maintenance/outage scheduling needs and also impact the lifetime of the device as a whole. This study illustrates a neutronics analysis that was conducted on a 22.5-degree symmetric sector of the FNSF with the goal of understanding the neutron heating and radiation damage that can be characterized by quantifying the displacements per atom (dpa). Concurrently, this study also focused on the development of analysis capabilities by converting a three-dimensional computer-aided design model of the FNSF into MCNP6.2 input using the McCad code. Accordingly, some confirmatory results on tritium production and the tritium breeding ratio (TBR) are provided to support model validation. The results produced by MCNP6.2 simulations showed that the highest heating and damage occurred in the outboard region, which concentrated approximately 290 MW of the total nuclear heating, in contrast to 97 MW within the inboard region. These results are consistent with previous studies that employed earlier versions of the FNSF concept and different modeling approaches. Furthermore, this study also provides additional details on neutron wall loading, as well as total heating from neutrons and gammas, results which show the total heating of the device (16 sectors) is approximately 477.83 ± 0.80% MW, indicating a neutron energy multiplication factor of 1.15. Additionally, the capability to calculate hydrogen and helium production, as well as dpa, is illustrated. Finally, the neutronics effects of using alternative materials to tungsten carbide were evaluated for the vacuum vessel, low-temperature shield, and structural ring components, which showed that compounds like YH 2 , Mg(BH 4 ) 2 , and ZrH 2 could reduce the total heating on the magnet and also reduce the TBR.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2019

Chapter 3 - Overview of Reactor Systems and Operational Environments for Structural Materials in Fusion Reactors

This chapter presents an overview of the conditions that structural materials being developed for fusion power reactor applications would encounter. We begin by briefly describing the relevant plasma physics conditions that dictate the fusion environment, which determines the loads driving microstructure evolution of fusion structural alloys. The fundamentals of radiation damage in the fusion environment are discussed, but the reader is referred to a much more detailed treatment of this topic given in Chapter 5. Our intent is to highlight the most important degradation mechanisms operating in the fusion environment. We compare and contrast the fusion environment with that found in existing and proposed advanced fission energy systems, highlighting in particular the role that gaseous and solid transmutation products have on property evolution. The majority of the chapter is devoted to a presentation of the principal structural material choices for major subsystems of a fusion power reactor such as the first-wall/blanket, divertor, and vacuum vessel. Our aim is to highlight the essential features of the material choices for each subsystem, and not to delve into the details, since later chapters address each class of structural material in greater depth. The emphasis is on the implication of the environment to the basic physical and mechanical properties requirements of these materials and challenges they face specific to the fusion environment.

Kurtz, Richard J.↗

DOE OSTI2019

Retrospective of the ARPA-E ALPHA Fusion Program

This paper provides a retrospective of the ALPHA (Accelerating Low-cost Plasma Heating and Assembly) fusion program of the Advanced Research Projects Agency–Energy (ARPA-E) of the U.S. Department of Energy. ALPHA’s objective was to catalyze research and development efforts to enable substantially lower-cost pathways to economical fusion power. To do this in a targeted, focused program, ALPHA focused on advancing the science and technology of pulsed, intermediate-density fusion approaches, including magneto-inertial fusion and Z-pinch variants, that have the potential to scale to commercially viable fusion power plants. Here, the paper includes a discussion of the origins and framing of the ALPHA program, a summary of project status and outcomes, a description of associated technology-transition activities, and thoughts on a potential follow-on ARPA-E fusion program.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2024

Simultaneous access to high normalized density, current, pressure, and confinement in strongly-shaped diverted negative triangularity plasmas

Strongly-shaped diverted negative triangularity (NT) plasmas in the DIII-D tokamak demonstrate simultaneous access to high normalized density, current, pressure, and confinement. NT plasmas are shown to exist across an expansive parameter space compatible with high fusion power production, revealing surprisingly good core stability properties that compare favorably to conventional positive triangularity plasmas in DIII-D. Non-dimensionalizing the key parameters, expanded operating spaces featuring edge safety factors below 3, normalized betas above 3, Greenwald density fractions above 1, and high-confinement mode (H-mode) confinement qualities above 1 are observed, even simultaneously, and all with a robustly stable edge free from deleterious edge-localized mode instabilities. Scaling of the confinement time with engineering parameters reveals at least a linear dependence on plasma current although with significant power degradation, both in excess of expected H-mode scalings. These results increase confidence that NT plasmas are a viable approach to realize fusion power and open directions for future detailed study.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗