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1,079 records · Page 60
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Developing solid-surface plasma facing components for pilot plants and reactors with replenishable wall claddings and continuous surface conditioning. Part B: required research in present tokamaks

Abstract The companion part A paper (Stangeby et al 2022) reports a number of independent estimates indicating that high-duty-cycle DT tokamaks starting with pilot plants will likely experience rates of net erosion and deposition of solid PFC, plasma facing component, material in the range of 10 3 to 10 4 kg yr −1 , regardless of the material used. The subsequent redeposition of such large quantities of material has the potential for major interference with tokamak operation. Similar levels and issues will be involved if ∼continuous low-Z powder dropping is used for surface conditioning of DT tokamaks, independent of the material used for the PFC armor. In Stangeby et al (2022) (part A) it is proposed that for high-duty-cycle DT tokamaks, non-metallic low-Z refractory materials such as ceramics (graphite, SiC, etc) used as in situ replenishable, relatively thin—of order mm—claddings on a substrate which is resistant to neutron damage could provide a potential solution for protecting the main walls, while reducing the risk of degrading the confined plasma. Assessment of whether such an approach is viable will require information, much of which is not available today. Section 6 of part A identifies a partial list of major physics questions that will need to be answered in order to make an informed assessment. This part B report describes R&D needed to be done in present tokamaks in order to answer many of these questions. Most of the required R&D is to establish better understanding of low-Z slag generation and to identify means to safely manage it. Powder droppers provide a unique opportunity to carry out controlled studies on the management of low-Z slag in current tokamaks, independent of whether their protection tiles use low-Z or high-Z material.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Investigation of the neoclassical ambipolar electric field in ion-root plasmas on W7-X

The role of the radial electric field in high-performance ion-root plasmas on Wendelstein 7-X (W7-X) is examined and compared with neoclassical predictions herein. The W7-X stellarator is the world’s first large-scale optimized stellarator. One of the important targets chosen for optimization during the W7-X design process was the reduction of core neoclassical heat transport. This optimization was targeted for reactor-relevant high-density plasmas with $T_e ≈ T_i$ in which the neoclassical ambipolar radial electric field is expected to be negative throughout the plasmas core. Measurements of the core radial electric field $(E_r)$ have confirmed that ion-root conditions (negative $(E_r)$ in the plasma core) have been achieved in W7-X with high-density plasmas and central ECRH. These measured $(E_r)$ profiles agree well with the neoclassical ambipolar $(E_r)$ predicted by the code SFINCS . This good agreement provides confidence in the validity of neoclassical calculations in high-density ion-root conditions, and enables initial studies on the role of neoclassical transport in the optimized high-density regime of W7-X. Profile measurements of electron temperature $(T_e)$, ion temperature $(T_i)$ and electron density $(n_e)$ along with approximations for the average value of $Z$ eff have been used as inputs to the sfincs code to calculate the ambipolar $E_r$ profile along with neoclassical ion and electron energy flux profiles $(Q$ NCi , $Q$ NCe $)$. Finally the total experimental energy input to the plasma from ECRH heating has been compared to the neoclassical energy fluxes to provide a first estimate for the fraction of transport that can be attributed to neoclassical processes in reactorrelevant high-density ion-root conditions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Modeling of ExB effects on tungsten re-deposition and transport in the DIII-D divertor

Mixed-material DIVIMP-WallDYN modelling, now incorporating ExB drifts, is presented that simultaneously reproduces tungsten (W) erosion and deposition patterns observed during the DIII-D Metal Rings Campaign, in which a toroidally symmetric set of W-coated tiles were installed in the carbon (C) DIII-D divertor. Since most reactor plasma facing component (PFC) designs call for mixed-material environments, including ITER’s W/Be enviroment, the divertor targets will quickly evolve into reconstitued surfaces of multiple elements. This work identifies controlling physics that affects material migration patterns in the divertor, which impact PFC lifetimes and impurity leakage from the divertor to the core. These simulations indicate that radial and poloidal ExB transport dominates over parallel force balance for high-Z impurities such as W in the divertor region of DIII-D. It is demonstrated that ExB drifts are required to reproduce the experimental observation of non-local W and C co-accumulation in a band ~7-9 cm outboard of the outer-strike-point (OSP) W source, for attached Lmode conditions in the unfavorable ion grad-B drift direction. In addition, W gross erosion is localized to the region outboard of the OSP, as the formation of C co-deposits suppresses W erosion at the strike point. Time-dependent simulations with scaled ExB impurity drifts (60% of the OEDGE-calculated drift velocity) and W re-erosion quantitatively reproduce these features, including depth-resolved W/C ratios, within a factor of 2 over ~115 seconds of accumulated plasma exposure. The location of co-deposition regions is shown to be well represented by an analytic leakage model, driven largely by poloidal ExB drifts. Qualitative agreement is also found between campaign-integrated W deposition measurements and simulations for the favorable ion grad-B drift direction, the standard mode of operation for most tokamaks. Furthermore, these results imply that a longterm inward radial migration of material from the outer divertor through the private flux region may occur in future devices.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Understanding the limitations and potential of micro tensile testing of tungsten and needs for crystal plasticity modelling

Tungsten is the leading plasma facing material candidate due to its exceptional properties. Understanding the response to neutron irradiation is crucial for the lifetime evaluation of tungsten. The limited space in nuclear reactors and the high levels of radioactivity of the specimens after irradiation are significant barriers to accurately measuring the mechanical properties after neutron irradiation. Testing of micro tensile specimens is one approach to reduce the total amount of irradiated material needed for a set of mechanical testing experiments. However, micro samples are not necessarily measuring the bulk material properties, as size effects produce higher measured mechanical properties than is observed in engineering size counterparts. Determining the minimal sample size for reliable bulk property measurement, combined with modeling efforts, is essential. We conducted room temperature tensile tests on tungsten micro tensile specimens fabricated with focused ion beam, plasma focused ion beam and femto-second laser ablation system to dimension of 2×2×7?µm³, 7×7×18?µm³, and 80×100×233?µm³ (width×thickness×gauge-length), respectively. Elevated tensile testing was performed up to 475°C on 5×5×18?µm³ specimens fabricated with focused ion beam. The smallest specimens exhibited a high degree of ductility and strength, whereas the largest specimens demonstrated behavior akin to bulk tungsten. To investigate the effects introduced by the micro specimen fabrication processes and to obtain the necessary bulk dislocation density for the crystal plasticity model, we employed X-ray diffraction depth profiling. This measurement was performed using different X-ray sources with varying penetration depths. Finally, the potential and limitations of micro mechanical tests for tungsten will be discussed.

36 - MATERIALS SCIENCE↗
DOE OSTI2022

Edge turbulence measurements in L-mode and I-mode at ASDEX Upgrade

The I-mode confinement regime is promising for future reactor operation due to high energy confinement without high particle confinement. However, the role of edge turbulence in creating I-mode's beneficial transport properties is still unknown. New measurements of edge turbulence (ρpol=0.9−1.0) in L-modes and I-modes at low and high densities at ASDEX Upgrade are presented in this paper. A high radial resolution correlation electron cyclotron emission radiometer measures the broadband turbulence throughout the L-mode and I-mode edge and pedestal. The weakly coherent mode (WCM) is measured in both L-mode and I-mode near the last closed flux surface with Te fluctuation levels of 2.3%–4.2%, with a frequency shift between the two phases related to a deeper Er well in I-mode. An neTe phase diagnostic captures a change of the WCM neTe phase between L-mode and I-mode from −171° to −143°. The thermal He beam diagnostic measures a WCM wavenumber range of −0.5 to −1.0 cm−1. A low-frequency edge oscillation (LFEO) appears in the I-mode phase of these discharges and displays coupling to the WCM, but the LFEO does not appear in the L-mode phase. Linear gyrokinetic simulations of the outer core and pedestal top turbulence indicate that while the dominant turbulent modes in the outer core are ion directed and electrostatic, the turbulence becomes increasingly electron directed and electromagnetic with increasing radius. Collisionality is not found to impact characteristics of the L-mode and I-mode edge turbulence with respect to the presence of the WCM; however, the quality of global confinement decreases with collisionality.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Thermo-Mechanical Distortion of Tungsten-Coated Steel During High Heat Flux Testing Using Plasma Arc Lamps

An experimental setup and a test section were designed and fabricated for high heat flux testing (HHFT) of neutron-irradiated specimens using water-wall plasma arc lamps. Because of the radiological considerations and limitations of reactor irradiation, the size of the test articles was limited to disks less than 10 mm in diameter. The specimen was clamped onto an actively cooled block, and clamping allowed the insertion of several thermocouples on the back surface of the specimen through a copper (Cu) block. Five vacuum plasma sprayed tungsten (W)–coated F82H steel specimens were subjected to HHFT. Surface profilometry measurements, which were conducted after HHFT, revealed central bowing of the top W surface. This type of residual distortion occurred for all of the specimens, and the larger the specimens were, the larger was the distortion.In an attempt to understand specimen distortion and address the science questions related to the testing of subsize specimens during HHFT, a simplified thermo-mechanical model was developed. By using a measured temperature in the Cu as an isothermal boundary condition, the model eliminated the need for coupling cooling fluid flow models with stress models, greatly simplifying the analysis. The main variable in the proposed model is hC, i.e., the thermal contact conductance between the F82H and the Cu washer. Inelastic properties, including hardening properties, were considered for F82H steel and Cu. Numerical simulation results demonstrated a buildup of residual deformation during HHFT and a very complex state of stress and deformation during typical heat flux (HF) cycling. Additionally, hoop stress evolution during a high heat flux cycle reveals that F82H at an interface with W would be mainly in compression during HF application and experienced a transition to a tension state during cooldown. Also, specimen distortion evolves during each HF cycle, as the specimen bows downward during HF application and upward during the cooldown period between HF cycles. The final specimen distortion, i.e., upward bowing of the specimen center, was qualitatively predicted for hC values of 4000 to 5000 W/(m 2 ·K). This hC range of values, for which bulging is obtained, is at the lower spectrum of the range of values for hC, consistent with the low thermal contact conductance expected from the unpolished F82H surface.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Multi-harmonic electron cyclotron heating and current drive scenarios for non-inductive start-up and ramp-up in high field ST-40 spherical tokamak

We report non-inductive start-up and ramp-up is an important topic for spherical tokamak reactor design as the central solenoid implementation is highly restrictive particularly for the low-aspect-ratio tokamak configuration. In the high field spherical tokamak (ST), ST-40 with B T0 ≤ 3 T, a preparation is underway for high power ECH and ECCD current start-up/ramp-up experiments utilizing two MW-class 140/105 GHz gyrotrons. Here, we explored various ECH/ECCD scenarios for a low-field-side (LFS) launch-angle steerable waveguide launcher placed near the mid-plane region. Due to the large toroidal field variation of ST configuration, multiple cyclotron harmonic resonance layers could exist within the plasma. In this start-up and ramp-up regime, both fundamental and second harmonic ECH resonances must be considered. We find that even with the presence of X-II resonance layer in the plasma, an efficient X-I ECH and ECCD regime can be accessed for the low electron temperature T e0 as low as 200 eV which is a typical starting temperature of ECH heated plasmas in an open-field-line configuration. The presence of X-II resonance could become significant at higher T e0 as X-II absorption increases with T e0 which could reduce the current ramp-up efficiency as the power reaching X-I is reduced. Finally for the pure X-I regime where the 2Ω e resonance is moved outside the plasma with B T0 ~ 3.4 T, we find that it is possible to reach the full current of I p ~ 1 MA fully non-inductively with the ECH power of ~1 MW at n e0 ~ 1.0 × 10 19 m -3 using 105 GHz frequency gyrotron. By reducing the outer limiter position R L ~ 78 cm to 70 cm, the pure X-I regime is recovered at the rated ST-40 magnetic field of B T0 ~ 3.0 T. This X-I regime is accessible with a relatively broad range of launched n II or the launching angles. A survey of X-mode X-II ECH and ECCD at higher density regimes is also shown for completeness.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Measurements and kinetic modeling of O 2 vibrational kinetics in O 2 –Ar mixtures partially dissociated by a Ns pulse discharge

Vibrational kinetics of O 2 is studied during the O atom recombination in an O 2 –Ar mixture, partially dissociated by a burst of ns discharge pulses in a heated plasma flow reactor. The time-resolved temperature in the discharge afterglow is determined by Rayleigh scattering. Time-resolved O atom number density is measured by ps Two-Photon absorption Laser Induced Fluorescence, calibrated in xenon. Time-resolved vibrational level populations of molecular oxygen, O 2 (v= 8–20), are measured by ps Laser Induced Fluorescence (LIF), with the absolute calibration by NO LIF. Time-resolved ozone number density is monitored by broadband UV absorption. The results are compared with the predictions of a state-specific kinetic model. The experimental data indicate a rapid initial decay of O 2 (v) populations generated by electron impact in the discharge, due to the vibration-translation (V–T) relaxation by O atoms. This is followed by a slower population reduction, on the time scale much longer compared to that for V–T relaxation or vibration-vibration (V–V) exchange. Both O atoms and the O 2 (v) populations decay on the same time scale, indicating that chemical reactions initiated by the O atom recombination result in the generation of vibrationally excited O 2 molecules. These trends are reproduced by the kinetic model, which shows that the reaction of O atoms with ozone is the dominant pathway of O 2 (v) generation at the present conditions. The predicted relative O 2 (v) populations are close to the experimental results, but absolute number densities differ from the experimental data. This is likely due to uncertainties in the absolute calibration of LIF measurements and in the spectroscopic model used in the data reduction. The present work demonstrates the capability for the absolute, time-resolved measurements of vibrationally excited O 2 in recombining gas flows, to quantify the energy partition in the recombination reactions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

On hybrid scenarios in KSTAR

We report the status of hybrid scenario experiments in Korea Superconducting Tokamak Advanced Research (KSTAR). The hybrid scenario is defined as stationary discharges with ${\beta _{\rm{N}}}$ $ \geqslant $ 2.4 and ${H_{89}}$ $ \geqslant $ 2.0 at ${q_{95}}$ < 6.5 without or with very mild sawtooth activities in KSTAR. It is being developed towards reactor-relevant conditions. High performance of ${\beta _{\rm{N}}} \lesssim 3.0$, ${H_{89}} \lesssim 2.4$ and G-factor ($ \equiv {\beta _{\rm{N}}}{H_{89}}/q_{95}^2$) $ \lesssim 0.4$6 has been achieved and sustained for $ \gtrsim 40{{\tau }_{\text{E}}}$ at ne/nGW ~ 0.7 with heating power of $ \lesssim $5 MW. Some KSTAR hybrid discharges exhibit a unique feature of a slow transition from conventional H-mode to hybrid mode after the third neutral beam injection. The reason for the confinement enhancement is extensively studied in this transition period of a representative discharge exhibiting a common feature of KSTAR hybrid scenarios. 0D performance analysis with magnetohydrodynamic activities, 1D kinetic profile dynamics, power balance analysis, linear gyro-kinetic analysis and edge pedestal stability analysis were conducted. The enhancement is thought to be from both the core and the pedestal. The improvement in the core region of the ion energy channel is observed from the linear gyro-kinetic analysis considering the electromagnetic, the fast ion, the Shafranov shift, ${\it{\omega }_{E \times B}}$, and the magnetic shear effect. The electromagnetic finite $\beta $ stabilisation plays a role in the inner core region at ${\rho _{\rm{tor}}}\sim0.35$ together with the fast ion effect. The alpha stabilisation effect is also found at ${\rho _{\rm{tor}}}\sim0.5$. ${\omega _{E \times B}}$, which could reduce the linear growth of the ion temperature gradient mode in the outer core region at ${\rho _{\rm{tor}}}\sim0.5 - 0.7$ with the highest contribution from the toroidal rotation. Furthermore, regarding the improvement in the pedestal, Shafranov shift broadens the stability boundary of the pedestal in support of the diamagnetic effect. The pedestal height and width could be reproduced by the EPED model, while a realistic current profile is used to calculate the internal inductance for Shafranov shift. Based on these findings, a comprehensive confinement enhancement mechanism has been proposed by considering the core-edge interplay.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Overview of the TCV tokamak program: Scientific progress and facility upgrades

The TCV tokamak is augmenting its unique historical capabilities (strong shaping, strong electron heating) with ion heating, additional electron heating compatible with high densities, and variable divertor geometry, in a multifaceted upgrade program designed to broaden its operational range without sacrificing its fundamental flexibility. The TCV program is rooted in a three-pronged approach aimed at ITER support, explorations towards DEMO, and fundamental research. A 1 MW, tangential neutral beam injector (NBI) was recently installed and promptly extended the TCV parameter range, with record ion temperatures and toroidal rotation velocities and measurable neutral-beam current drive. ITER-relevant scenario development has received particular attention, with strategies aimed at maximizing performance through optimized discharge trajectories to avoid MHD instabilities, such as peeling-ballooning and neoclassical tearing modes. Experiments on exhaust physics have focused particularly on detachment, a necessary step to a DEMO reactor, in a comprehensive set of conventional and advanced divertor concepts. The specific theoretical prediction of an enhanced radiation region between the two X-points in the low-field-side snowflake-minus configuration was experimentally confirmed. Fundamental investigations of the power decay length in the scrape-off layer (SOL) are progressing rapidly, again in widely varying configurations and in both D and He plasmas; in particular, the double decay length in L-mode limited plasmas was found to be replaced by a single length at high SOL resistivity. Experiments on disruption mitigation by massive gas injection and electron-cyclotron resonance heating (ECRH) have begun in earnest, in parallel with studies of runaway electron generation and control, in both stable and disruptive conditions; a quiescent runaway beam carrying the entire electrical current appears to develop in some cases. Developments in plasma control have benefited from progress in individual controller design and have evolved steadily towards controller integration, mostly within an environment supervised by a tokamak profile control simulator. TCV has demonstrated effective wall conditioning with ECRH in He in support of the preparations for JT-60SA operation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Experimental demonstration of transient CHI start-up using a floating biased electrode configuration

Transient coaxial helicity injection (CHI) current start-up in a spherical tokamak (ST) has been demonstrated for the first time using a reactor-relevant floating single biased (FSB) electrode configuration on QUEST. Unlike previous applications of CHI on HIT-II and on NSTX in which the entire ST vessel was split in two using two toroidal ceramic rings as vacuum breaks, in the FSB electrode configuration on QUEST, only the lower outer divertor plate is electrically insulated from the rest of the vessel. This is the CHI injector region. A CHI discharge is generated by driving current, using an external power supply, along magnetic field lines that connect the inner and outer divertor plates in the CHI injector region. The CHI plasma evolution and closed flux formation on QUEST are very similar to those on NSTX and have generated toroidal currents of 150 kA with 100 kA of current flowing on closed flux surfaces. An alternate, force-free model approach has been used to derive the CHI scaling relations. The resulting CHI scaling relations from the model are consistent with the conventional CHI scaling relations, which state that the CHI generated toroidal current is proportional to the CHI injector flux and the CHI injector current is inversely proportional to the toroidal field. These results from QUEST bode well for the application of CHI in a new generation of compact high field STs and low aspect ratio tokamaks in which the space for the central solenoid is very restricted.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Expedited Walkdowns and Preparation of Cost Estimates for Decontamination and Demolition of Excess Facilities - 20335

Lawrence Livermore National Security, LLC (LLNS) is the prime contractor responsible for managing and operating Lawrence Livermore National Laboratory (LLNL). The mission of the LLNL is to strengthen security of the United States through development and application of world-class science and technology to enhance the nation's defense, reduce the global threat from terrorism and weapons of mass destruction, and respond with vision, quality, integrity, and technical excellence to scientific issues of national importance. To accomplish its mission, LLNL must plan and manage its campus space and facilities to optimize use of its relatively small one-square-mile footprint. LLNL currently has numerous excess facilities that require decontamination and demolition (D and D). The LLNL Legacy Facility Program is responsible for stewardship and risk reduction programs to effectively manage these excess assets at LLNL. The Legacy Facility Program identified 10 excess facilities at LLNL that were candidates for demolition based on risks from further degradation of the facilities, potential exposure to hazardous materials, or need for removal to make way for new program facilities. The buildings range from a structure built in 1943 to serve as a US Navy drill hall during World War II, to a facility that provided neutrons to study material properties for the fusion energy program. The National Nuclear Security Administration (NNSA) appropriated FY19 funding for the development of Class 3 estimates for these 10 excess facilities that would be used to develop a Program Management Plan for funding consideration. Class 3 estimates are not conceptual, nor overly detailed, but adequate to be used for budget and appropriation purposes. Using existing strategic sourcing agreements, LLNL was able to organize a team of senior subject matter experts (SMEs) with diverse technical backgrounds across the Department of Energy (DOE) Complex to complete extensive process knowledge (PK) reviews and walk-downs of each excess facility and to deliver accurate and complete estimates under an expedited schedule (i.e., within eight weeks). To facilitate the expedited schedule, LLNL was fully prepared for the arrival of the JGMS team, which consisted of the following companies: J.G. Management Systems, Inc. (JGMS); Strata-G, LLC; and Michael Baker International. LLNL provided background and PK information on the excess facilities, pictures, supporting historical documentation, and radiological survey history as well as escorted access to the facilities. The JGMS team has previous experience supporting Program Management Plans for excess facilities at the Y-12 National Security Complex. The team mobilized to LLNL and performed the facility walk-downs from January 22-25, 2019. Using a methodical evaluation process, expertise, and experience, the team was able to deliver walk-down draft reports and estimates to LLNL before their February 8, 2019 deadline. LLNL then in turn reviewed, completed the estimates by adding waste management costs and other LLNL costs, and delivered the estimates to NNSA by February 15, 2019. The completed final reports, including draft D and D estimates, were delivered to LLNL before their March 6, 2019 deadline. The work was successfully completed ahead of the expedited schedule required by LLNL, resulting in a commendation from the client. This paper describes how the partnership and collaboration as a fully integrated team comprised of the prime contractor, LLNS, and the JGMS team successfully met the needs of the US government, resulting in the performance of expedited site walk-downs, the development of summary evaluation reports, and the preparation of draft Class 3 estimates and preliminary schedules for the D and D of 10 excess facilities at LLNL. The efforts of the team will enable the D and D of the first excess facility in the 2020-2021 time frame. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Stable heat and particle flux detachment with efficient particle exhaust in the island divertor of Wendelstein 7-X

The island divertor concept is an innovative and promising idea to handle heat and particle exhaust in stellarators. At the Wendelstein 7-X (W7-X) stellarator, this divertor concept plays a central role in the device mission to demonstrate reactor relevant plasma confinement for steady-state time scales of up to 30 minutes in the high-performance campaign (OP2) starting in 2022. During the recently concluded first campaign with the inertially cooled island divertor, a large step in the experimental qualification of this divertor concept has been made. In discharges heated with electron cylotron resonance heating of 5-6 MW, central densities in the range of 0.7-1.2 × 10 20 m -3 have been reached in combination with full divertor heat flux detachment. Also, significant neutral gas pressures and neutral compression ratios were shown for the first time in combination with reduced divertor particle flux. The divertor heat loads drop by an order of magnitude from >5 MW m -2 to below 0.5 MW m-2 with increasing density, and substantial compression of neutrals reaching neutral pressure in the sub-divertor volume of >6.0 × 10 -4 mbar was seen. These elevated neutral pressure levels can be obtained and maintained with an up to 80% reduction of the particle fluxes onto the divertor target tiles. This discharge scenario was held stably detached for up to 28 seconds, which is equivalent to several hundred energy confinement times τ E and longer than the time scales for current relaxation. No impurity accumulation was seen at constant Z eff ≈ 1.5 and the stored energy stayed constant at levels of W dia >600 kJ. The level of neutral pressure and compression reached in this scenario extrapolates well to the steady-state particle exhaust requirements for high-performance steady-state operation in OP2, in which the fully actively cooled high-heat-flux divertor will be available. Finally, an overview of this recently discovered divertor regime is given and the status of the physics understanding based on modeling of these regimes with the EMC3-EIRENE code is presented.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Laser Powder Bed Fusion Manufactured Alloys 625 and 282 and Properties Relevant to Advanced Nuclear Components

Under the Advanced Materials and Manufacturing Technologies program, two Ni-based alloys fabricated by laser powder bed fusion (LPBF) have been evaluated: γ′-strengthened Haynes 282 and solution-strengthened Inconel 625. Large printing defects were observed in the LPBF 282 alloy fabricated using a Renishaw AM250 machine, likely due to particle spattering during printing. Annealing at 1,180°C for 1 h followed by 4 h at 800°C in a high density of 50 nm γ′-strengthening precipitates as well as partial recrystallization resulted in a bimodal grain distribution. Creep testing performed at 750°C revealed lower creep life and ductility for the LPBF 282 compared with wrought 282. X-ray computed tomography combined with optical and scanning electron microscopy microstructural characterization revealed crack formation during creep testing, initiated either from printing defects or from creep cavitation at grain boundaries. Printing defects were the likely reason for the lower creep performance of LPBF 282 and could be suppressed by optimizing the build configuration. Printing 625 on an EOS M290 machine using the recommended EOS parameters resulted in a very low defect density. Superior creep strength at 725°C was observed for the as-printed LPBF 625 along the build direction when compared with wrought 625. No defect larger than ~50 µm could be detected by X-ray computed tomography in the as-printed conditions and after creep testing for 1,000 h at 725°C and 150 MPa. Small, needle-like δ-phase precipitates were observed after creep testing for 100 h at 725°C and 200 MPa. The rapid formation of the γ phase is directly related to the initial Nb and Mo segregation in the LPBF 625 cell walls, and its effect on the alloy’s high-temperature performance will be evaluated. Solution annealing was carried out at 1,150°C for 1 h and led to full recrystallization of the alloy. A decrease in strength and increase in ductility were observed after solution annealing, and the alloy recrystallization resulted in isotropic tensile properties at room temperature, contrary to what was observed for the as-printed LPBF 625.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Molten Salt Corrosion Tests of Additively Manufactured Stainless Steel 316H

Molten salt reactors (MSRs) have drawn considerable interest due to their favorable safety features, high thermal efficiency, and compatibility with different fuel cycles. Yet, the success of MSRs hinges critically on the performance of structural materials to be used in these aggressive molten salt environments, where corrosion and material compatibility remain primary challenges to long-term reliability. Additively manufactured (AM) nuclear structural materials prompt the use of novel geometries and compositions to enhance material performance and reduce costs of constructing MSRs. The rapid solidification conditions inherent to AM processing impart distinctive microstructural features, including cellular sub-structures, dislocation densities, residual stress, and oxide inclusions, which can influence material performance in MSR components. While the mechanical properties of AM stainless steels have been widely studied, their corrosion behavior, particularly in molten salt environments, has received far less attention. Addressing these needs, the Advanced Materials and Manufacturing Technologies (AMMT) program provides a framework for systematically evaluating how unique microstructures produced by AM processes influence the performance of these materials in these demanding environments and for developing reproducible testing workflows that can support future code qualification efforts and standards development. Bridging this knowledge gap is essential for assessing the viability of AM alloys in MSRs and informing qualification strategies. A further challenge is the absence of standardized protocols for molten salt corrosion testing. Accordingly, this report provides an account of the corrosion evaluation of AM 316H stainless steel in NaCl 2 -MgCl 2 molten salt at 550 °C, with exposure times of 100 and 500 hours. It documents the experimental procedures implemented under the AMMT program, including salt preparation, exposure protocols, and post-test characterization methods, to establish reproducibility and transparency. Importantly, the study examines AM 316H samples in the as-fabricated condition, directly reflecting the surface state most relevant to engineering applications, and compares their behavior to machine-cut surfaces. Overall, preliminary evaluations have noted that surface conditions (e.g. morphology, contamination, etc.) have a noticeable impact on the corrosion resiliency. The impact of the corrosion is difficult to detect at 100h, unless, in the case of AM 316H, the specimen surface is decontaminated. After 500 h, as-fabricated surfaces of AM and wrought 316H display evidence of general versus preferential corrosion attack, respectively. Both AM and wrought 316H machine-cut surfaces exhibit a continuous Cr depletion zone, evident of general corrosion. While the estimated extent of corrosion appears within the same order of magnitude regardless of the surface condition, it is apparent that more predictable behavior is observed on machine-cut surfaces. Nonetheless, further investigation is necessary to fully elucidate the corrosion mechanism under these conditions.

36 - MATERIALS SCIENCE↗
DOE OSTI2023

Development of Analysis Methods that Integrate Numeric and Textual Equipment Reliability Data

Within the Light Water Reactor Sustainability (LWRS) program, the Risk-Informed Systems Analysis (RISA) Pathway is performing collaborative research on the development and deployment of technologies designed to assist operating nuclear power plants (NPPs) to reduce operating costs improve plant reliability and availability. One of the RISA research areas is focusing on the development of methods and tools designed to optimize plant operations (e.g., maintenance/replacement schedules, optimal maintenance postures for plant structures, systems, and components [SSCs]) in a manner that is more cost effective than current approaches and makes better use of available SSC health data. The Risk-Informed Asset Management (RIAM) project targets this research area by creating a direct bridge between component equipment reliability (ER) data and system engineer decision making regarding maintenance activity scheduling and component aging management. In this respect, one challenge that NPP system engineers are facing is that the amount of ER data being continuously generated is not only extremely large in size, but it comes in different forms: textual (e.g., condition or maintenance reports) and numeric (e.g., generated by monitoring systems). All these data elements provide them with valuable insights and information regarding: 1) the discovery of anomalous behaviors or degradation trends, 2) the identification of the possible causes behind such behaviors/trends, and 3) the prediction of their direct consequences. However, several challenges have proved to be roadblocks to this process. While some of these challenges are technical in nature (i.e., data are often distributed over several physical servers/databases), others are conceptual in nature: data elements come in different formats (e.g., numeric or textual), and measured values have different scales (e.g., vibration spectra and oil temperature). The activities performed by the RIAM project during FY23 directly tackles the need to simultaneously integrate the analysis of ER data in all its forms, numeric and textual. Note that such task has never been performed before due to the complexity of the systems under consideration but, most importantly, because of the technical challenges behind the harmonization of ER data formats and the lack of adequate computational methods to analyze them. Our approach borrows ideas and concepts from the medical field where integration of several data sources is vital to assist medical practitioners to perform correct diagnosis and indicate optimal treatments. In our view a NPP asset is equivalent to a patient in a medical context. The main difference is the complexity of a human body is a magnitude more complex when compared to typical assets commonly present in NPPs (e.g., centrifugal pumps, or motor operated valves). This simplifies our first requirement when analyzing heterogenous ER data formats: to put data into “context”. Context is here intended as the additional piece of information that is needed by ER data analysis tools to understand what these data elements are referring to, i.e., which king of knowledge they are generating. In our context, this knowledge can be translated into models that capture the form and functional architecture of assets/systems, their dependencies, and how they interact. These models actually emulate the knowledge that that NPP system engineers possess about assets and systems; this is their key of success when analyzing ER data, their challenge is ability to handle large amount of data. Here, we employ model-based system engineering (MBSE) models of systems and assets to represent and capture their architecture and functional, i.e. cause-effect, relations. Then, ER data elements are processed by identifying first of all which elements of the developed MBSE elements they are referring to. For numeric ER data this task is fairly easy since it is possible to precisely pinpoint what MBSE elements the corresponding sensor are observing (e.g., bearing temperature of a centrifugal pump). Task is much harder for textual data since the information contained in issue or maintenance reports needs to “be understood” by a computational tool. Here we called this process as “knowledge extraction”. Once again, we borrow the experience in the medical field where methods to extract knowledge from textual data have been developed in the past decade. The missing element for us is the availability of a complete dictionary of NPP related concepts (in addition to the MBSE models presented earlier) that can put “text into context”. In FY23, such dictionary has been developed along with all the computational elements required for knowledge extraction. Lastly, once numeric and textual ER data elements have been processed and “understood”, then the last step is the discovery of possible cause-effect relations among them. This is performed by observing if a logical connection through the MBSE models exists, and if the

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2024

Overview of ASDEX upgrade results in view of ITER and DEMO

Experiments on ASDEX Upgrade (AUG) in 2021 and 2022 have addressed a number of critical issues for ITER and EU DEMO. A major objective of the AUG programme is to shed light on the underlying physics of confinement, stability, and plasma exhaust in order to allow reliable extrapolation of results obtained on present day machines to these reactor-grade devices. Concerning pedestal physics, the mitigation of edge localised modes (ELMs) using resonant magnetic perturbations (RMPs) was found to be consistent with a reduction of the linear peeling-ballooning stability threshold due to the helical deformation of the plasma. Conversely, ELM suppression by RMPs is ascribed to an increased pedestal transport that keeps the plasma away from this boundary. Candidates for this increased transport are locally enhanced turbulence and a locked magnetic island in the pedestal. The enhanced D-alpha (EDA) and quasi-continuous exhaust (QCE) regimes have been established as promising ELM-free scenarios. Here, the pressure gradient at the foot of the H-mode pedestal is reduced by a quasi-coherent mode, consistent with violation of the high-n ballooning mode stability limit there. This is suggestive that the EDA and QCE regimes have a common underlying physics origin. In the area of transport physics, full radius models for both L- and H-modes have been developed. These models predict energy confinement in AUG better than the commonly used global scaling laws, representing a large step towards the goal of predictive capability. A new momentum transport analysis framework has been developed that provides access to the intrinsic torque in the plasma core. In the field of exhaust, the X-Point Radiator (XPR), a cold and dense plasma region on closed flux surfaces close to the X-point, was described by an analytical model that provides an understanding of its formation as well as its stability, i.e., the conditions under which it transitions into a deleterious MARFE with the potential to result in a disruptive termination. With the XPR close to the divertor target, a new detached divertor concept, the compact radiative divertor, was developed. Here, the exhaust power is radiated before reaching the target, allowing close proximity of the X-point to the target. No limitations by the shallow field line angle due to the large flux expansion were observed, and sufficient compression of neutral density was demonstrated. With respect to the pumping of non-recycling impurities, the divertor enrichment was found to mainly depend on the ionisation energy of the impurity under consideration. In the area of MHD physics, analysis of the hot plasma core motion in sawtooth crashes showed good agreement with nonlinear 2-fluid simulations. This indicates that the fast reconnection observed in these events is adequately described including the pressure gradient and the electron inertia in the parallel Ohm’s law. Concerning disruption physics, a shattered pellet injection system was installed in collaboration with the ITER International Organisation. Thanks to the ability to vary the shard size distribution independently of the injection velocity, as well as its impurity admixture, it was possible to tailor the current quench rate, which is an important requirement for future large devices such as ITER. Progress was also made modelling the force reduction of VDEs induced by massive gas injection on AUG. The H-mode density limit was characterised in terms of safe operational space with a newly developed active feedback control method that allowed the stability boundary to be probed several times within a single discharge without inducing a disruptive termination. Regarding integrated operation scenarios, the role of density peaking in the confinement of the ITER baseline scenario (high plasma current) was clarified. The usual energy confinement scaling ITER98(p,y) does not capture this effect, but the more recent H20 scaling does, highlighting again the importance of developing adequate physics based models. Advanced tokamak scenarios, aiming at large non-inductive current fraction due to non-standard profiles of the safety factor in combination with high normalised plasma pressure were studied with a focus on their access conditions. A method to guide the approach of the targeted safety factor profiles was developed, and the conditions for achieving good confinement were clarified. Based on this, two types of advanced scenarios (‘hybrid’ and ‘elevated’

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗