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113 records · Page 7

Active Measurement of Plasma Parameters in Over-Dense MFE Reactor Plasma Using High Frequency Beat-Wave Resonantly Generated Thermal Bernstein Waves

UC Davis was funded to explore the combination of beat-wave excitation and laser scattering to make completely non-invasive determination of local plasma conditions. This involves the use of two high frequency mm-wave sources of differing frequencies (the pump waves) to excite plasma waves at the difference of the frequencies, which is known as the beat frequency. The difference frequency or the frequency of the beat wave is chosen so that we produce plasma waves whose fluctuating electron density is a sensitive function of plasma parameters at the location where the source waves overlap. This fluctuation can be detected from outside the plasma, by measuring the deflection of a probe beam directed at the overlap region between the two pump waves (i.e., collective scattering). This work is an extension of mm-wave frequencies of a 10 GHz beat-wave experiment conducted on the DDT toroidal device (see Fig. 1) in a magnetized plasma where the excited Langmuir wave is replaced by the magnetized upper hybrid wave. Here, two high power (> 1 kW) ~94 GHz pump sources will be employed to match the upper hybrid resonance. The primary envisioned experiment is to drive the electron Bernstein wave (EBW) in the interior of a magnetized plasma. The EBW is a hot-plasma electrostatic wave, which cannot propagate in vacuum. Roots of the EBW dispersion relation occur at frequencies slightly higher than harmonics of the electron cyclotron frequency. Use of the EBW has particular advantages for measuring plasma magnetic field in plasmas with relatively high density and low magnetic field, as is the case in spherical tokamaks. This is the “over dense” condition of future efficient high beta magnetic confinement fusion reactors such as the spherical tokamak (STs). The overarching goal of the proposed project is to demonstrate beat wave EBW excitation in low density and temperature parameters that are directly scalable to over-dense spherical tokamaks such as NSTX or MAST, where measurement of magnetic field by electron cyclotron emission (ECE) is experimentally complex, and requires models of plasma profiles in order to interpret results. In such experiments, availability of point measurements of magnetic field strength provided by a beat-wave diagnostic would serve as a powerful independent check on the ECE results, serving to validate the assumed profiles and wave-propagation models used. The first step was to assemble/modify the target chamber in which the beat-wave plasma experiments will take place. Here, the target chamber of the CTIX compact-toroid (spheromak) injector was chosen. At the time that the proposal was submitted, the aim was to conduct the experiments in the CTIX experimental facilities located at the Livermore campus of UC Davis using microwave/mm-wave equipment brought from the main campus., However, by the time DoE funds for this project were awarded, the CTIX experimental facilities located at the Livermore campus of UC Davis were in the process of being shut down due to the closure of the Livermore campus by Lawrence Livermore National Laboratory (LLNL). Consequently, the equipment needed for this project was crated up and shipped from Livermore to Davis as part of the laboratory shutdown after space on campus was secured and College of Engineering approval secured. Once it arrived in Davis on December 20, 2018, work on the project began in earnest with the initial efforts directed toward developing a new plasma target source. Because of the shutdown of the CTIX facility, the scope of the project thus had to be reduced due to the abbreviated amount of time allocated to the project once the shutdown/move had been completed. Chamber modifications have been completed, including the fabrication and installation of a 38 GHz interferometer, as described in Sec. 3. The thermionic plasma source has been tested and characterized. The magnetic field required for the EBW experiments will be generated by permanent magnets mounted on plates positioned above and below the chamber. Magnetic simulations have been performed, and show that a 3-ring magnet arrangement on each plate, with a plate separation of 50 cm, is required to achieve the ~100 G uniform magnetic field in the interaction region that we feel is necessary to ensure good EBW coupling. We proved unable to successfully revive either of Prof. Luhmann’s two high power (≥20 kW) 94 GHz millimeter-wave sources after their extended time in storage and decided to postpone further efforts and concentrate on slightly lower power sources. Consequently, we were successful in reviving a ~2 kW extended klystron amplifier (EIK), which together with a ~2 kW extended interaction oscillator (EIO) will provide the two pump sources required for the project. Safe operation of the EIO requires extensive modifications to one of Prof. Luhmann’s pulse modulators - these modulator modifications are ongoing, which has unfortunately delayed the planned beat-wave experiments. Details of this work are provided in Sec. 3. Even though project funding has now ended, we still plan to proceed to complete the EIO pulse modulator modifications, install the two ~94 GHz pump wave sources on the target chamber, and conduct unmagnetized tests of the beat-wave concept which involve coupling to unmagnetized electron plasma waves rather than magnetized EBWs. Should these experiments prove successful, we plan to submit a follow-up proposal for additional funding which would allow us to undertake the magnetized EBW experiments.

42 ENGINEERING↗

Design, construction, and testing of no-insulation small subscale solenoids for compact tokamaks

Fusion energy systems studies (FESS) for next-step devices based on the most promising magnetic configurations indicate that high magnetic fields and high current density for magnet coil systems may reduce device size and lower the cost. High current density and radiation resistant fusion magnets are particularly beneficial for low cost, low aspect ratio compact reactor designs such as Fusion Nuclear Science Facility (FNSF), Fusion Pilot Plant (FPP) of low-aspect ratio spherical tokamak (ST) or compact stellarators. Unlike typical high field magnets for nuclear magnetic resonance (NMR), magnetic resonance imaging (MRI), or high energy physics research applications, neutron irradiation damage to organic insulations in the coil winding pack is a critical issue for next generation fusion reactors where orders of magnitude higher neutron fluence than those in present experimental reactors such as ITER are expected. Moreover, a high coil winding pack current density is needed for high field magnets in low cost, compact radial build next step fusion reactors. The slow current charging time, however, is an issue in fully non-insulated coils. In this work, we present the design, construction and testing of subscale Nb3Sn solenoids, up to half the diameter of the central solenoid (CS) for National Spherical Torus Experiment (NSTX), with and without inter-layer insulation for enhancing radiation resistance while improving the winding pack current density and coil performance. The major radius for the NSTX/NSTX-Upgrade (reference for compact ST) is 0.854/0.934 meter, and 4.8 meters for FNSF. The magnetic field for NSTX/NSTX-U at the plasma center is 0.6/1 Tesla and 9 Tesla for FNSF. The inner diameter of the central solenoid coil is 0.2/0.4 meter for NSTX/NSTX-U and 1.2 meters for FNSF. The current charging and discharging behavior of the prototype solenoids was investigated to quantify performance advantages of a simplified coil fabrication without error-prone vacuum pressure impregnation (VPI) for the removal of epoxy organic insulation. Scalability of the no insulation concept for fusion can be addressed via engineered coating for novel insulation on contact resistance or intra-layer no insulations. Radiation resistance and coil winding efficiency were significantly improved in the no insulation coils for next step compact fusion reactors.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

Disruption avoidance via island suppression: the crucial roles of DIII-D and foundational research

The FESAC long range plan calls out disruption avoidance and mitigation as key remaining technical gaps. In discussing the roles of DIII-D and NSTX-U, the FESAC long range plan says “Additional research on these facilities, in combination with private and international collaborations, continuing support of existing university tokamak programs, and utilization of US expertise in theory and simulation, is needed to find solutions to remaining technical gaps. These gaps include disruption prediction, avoidance, and mitigation …”. Disruptions pose an existential threat to ITER and to FPPs. For a fusion reactor, unplanned shutdowns caused by disruptions will be a significant barrier to connecting such a reactor to the electric grid, even if disruption mitigation is successful. Disruption studies for ITER in recent years have largely focused on disruption mitigation (e.g., pellet injection), motivated by near-term deadlines for finalizing the design of the mitigation hardware. It is recognized, however, that mitigation alone will not suffice. The 2022 U.S. ITER Research Needs Workshop Report states that ”[d]isruptions are considered the largest threat to the ITER Research Program”, and that “[m]itigation should be a last resort”. As we discuss below, there are unresolved foundational issues that play a critical role in avoidance, and DIII-D is an ideal device for generating the data needed to address these issues.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Active Lithium Injection for a Real Time Control of the Divertor Heat Flux for Fusion Devices

When the local heat flux exceeds specified flux limit, tungsten PFC surfaces can be damaged, which is not acceptable for a reliable reactor operations. The divertor PFCs are typically designed for a specific heat flux limit usually assuming an average steady-state heat flux which is typically 5–10 MW/m 2 . However, in addition to steady-state heat flux, fusion reactor divertor PFCs could experience transient heat fluxes such as ELMs and/or other magnetic reconnection events which can deposit large transient heat fluxes onto the divertor PFCs. The transient divertor heat flux could be significantly larger than the steady-state heat flux which could damage the solid PFC surfaces. The divertor heat flux can be subjected to additional complications such as the uncertainties in the the divertor strike point heat flux projection. Moreover, there are additional experimental observations of non-axisymmetric power flux which can occur under non-axisymmetric magnetic perturbations. The liquid lithium (LL) PFCs is more resilient against such transient heat fluxes as they could evaporate LL as needed and the lost LL can be then replenished afterward. In this paper, we analyze a case for a transient divertor heat pulse of 1 MJ in 10 ms for a ITER-size reactor. This is a small perturbation (~ 0.1%) to the expected plasma stored energy compared to the previously analyzed case of 20 MJ heat pulse. Even with this relatively modest heat pulse, the LL surface undergoes ~ 100 °C temperature rise. However, the resulting LL surface heating without rapid cooldown mechanism could lead to excessive LL evaporation continuing well after the transient heat flux resulting in a significant Li injection of ~ 0.6 mol in about a 200 ms period. This amount of Li injection could cause plasma dilution and performance degradation. On the other hand, an active Li injection capability if optimized could prevent the LL surface temperature rise and thus reducing subsequent Li evaporation into the plasma by a factor of 7 compared to the passive LL PFC case. A crucial tool of active Li injection is a rapid response pellet injector which could inject light impurity pellets before the excessive heat flux could reach the divertor plate causing serious damage. A simple pellet ablation model suggests a favorable pellet deposition profile for smaller ~ 0.1 mm radius pellet with ~ 10–20 m/s velocity. Moreover, if it is possible to inject from the private flux region, the pellet injection efficiency into the high heat flux strike point region can be as high as 80% compared to ~ 50% for the injection from outer radius region. The pellet deposition efficiency can be further improved by designing a shell-pellet which can burst when a certain ablation fraction is reached. Finally, a possible implementation technique using an inductive pellet injector with a rapid time response of a few msec is proposed here which can be tested in NSTX-U.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Dynamics of filaments during the edge-localized mode crash on NSTX

Edge localized modes (ELMs) are routinely observed in H-mode plasma regimes of the National Spherical Torus Experiment (NSTX). Due to the explosive nature of the instability, only diagnostics with high temporal and spatial resolution could provide a detailed insight into the dynamics associated with the ELMs. Gas-puff imaging at NSTX provides 2D measurements of the magnetic field aligned fluctuations (e.g., ELM filaments) in the scrape-off layer and at the plasma edge with 2.5 μs temporal and 10 mm optical resolution. A novel analysis technique was developed to estimate the frame-by-frame velocities and the spatial parameters of the dominant structures associated with the ELMs. The analysis was applied to single ELM events to characterize the ELM crash dynamics and then extended to a database of 159 ELM events. Statistical analysis was performed in order to find the characterizing dynamics of the ELM crash. The results show that on average, an ELM crash consists of a filament with a circular cross section, which is propelled outward with a characterizing peak radial velocity of ~3.3 km/s. The radial velocity was found to be linearly dependent on the distance of the filament from the separatrix, which has never been seen before. The ELM filament is characterized by propagation in the ion-diamagnetic direction poloidally with a peak velocity of 11.4 km/s. The ELM crash lasts for approximately 100 μs until the radial propulsion settles back to the pre-ELM level. Furthermore, the experimental findings were compared with analytical theory. Two possible mechanisms were identified for explaining the observations: the curvature interchange model and the current–filament interaction model.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗