Bulletin of the American Physical Society
64th Annual Meeting of the APS Division of Plasma Physics
Volume 67, Number 15
Monday–Friday, October 17–21, 2022; Spokane, Washington
Session PO03: Whole Device Modeling and ReactorsLive Streamed
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Chair: Christopher Holland, University of California, San Diego Room: Ballroom 100 C |
Wednesday, October 19, 2022 2:00PM - 2:12PM |
PO03.00001: Investigation of the sensitivity of engineering performance metrics to the plasma profiles for the ARC reactor. Katarzyna Borowiec, Jin Whan Bae, Vittorio Badalassi, Jin Myung Park The assessment of the fusion reactor design requires multi-physics high-fidelity modeling that characterizes the system's performance under design conditions. A complete system modeling environment is needed to streamline the design assessment and optimization. We coupled a core and edge plasma transport simulator (IPS-FASTRAN) and a general-purpose neutron and gamma transport code (MCNP) to explore the sensitivity of engineering quantities to various plasma profile characteristics. This is a part of a bigger effort for a full device modeling environment, FERMI. |
Wednesday, October 19, 2022 2:12PM - 2:24PM |
PO03.00002: Core-Edge Plasma Response to Variations in Advanced Tokamak Aspect Ratio using IPS-FASTRAN Plasma Simulator Ehab Hassan, Charles E Kessel, Philip B Snyder, Jin Myung Park, Rhea L Barnett, David L Green The core and edge plasma responses in advanced tokamaks for a range of aspect ratios (2 – 4) are studied using EPED and TGLF models in the IPS-FASTRAN plasma simulator with the implementation of different heating and current drive actuators (HCD). A systems code (SC) is used to search a manifold space of tokamak parameters, such as fusion gain = 4 – 12, toroidal magnetic field = 2 – 10.5 T, major radius = 1.5 – 5 m, plasma beta = 2 – 5.5, effective atomic charge number = 1.6 – 2.4, and Greenwald fraction = 0.8 – 1.1, for an optimum design point that provides a low to moderate electric power (PE ≥ 50 MW) and a large bootstrap fraction (fBS ≥ 0.7). The plasma parameters in a set of selected designs are used as inputs to the IPS-EPED framework to test the sensitivity of the pedestal height and width to the variations in the aspect ratio, major radius, plasma current, toroidal magnetic field, and Greenwald fraction. Then the sensitivity of the energy and particle fluxes in the core region to those parameters is tested in the IPS-FASTRAN framework while implementing different heating and current drive schemes, such as neutral-beam heating using NUBEAM, electron-cyclotron resonance heating using TORAY, helicon and lower-hybrid heating using GENRAY, and ion-cyclotron resonance heating using TORIC, is examined. Hence the fusion gain and power, heating efficiencies, heating power deposit locations, confinement time, bootstrap and Greenwald fractions, and plasma density and temperature profiles calculated in the IPS-FASTRAN framework can be used to update the inputs to the systems code to improve the results in the workflow. In the future, these manually operated workflows will be integrated into a single coupled workflow that can be operated iteratively until we achieve a steady-state scenario for the optimum design point. |
Wednesday, October 19, 2022 2:24PM - 2:36PM |
PO03.00003: Whole Device Modeling of ITER pedestal while including the physics of multiple gyrokinetic tungsten bundles Julien Dominski, Choongseok Chang, Robert Hager, Seung Hoe Ku, Martin O'Mullane, Vassili Parail, Pallavi Trivedi One of the main scientific objectives of the Exascale High-Fidelity Whole Device Modeling (WDM) project is the prediction of ITER pedestal height and shape. In view of the costly simulations, research has been conducted to model the whole device with core and edge regions coupled consistently [1]. On the other hand, ITER is a tungsten environment machine and tungsten ions are known to strongly affect the performance of the pedestal. An advanced stretch goal of the project concerns the inclusion of tungsten physics in the WDM simulation. This research has been initiated with the inclusion of impurities in XGC [2] and the implementation of tungsten bundles [3] to model the 74 ionization levels of tungsten ions with a few gyrokinetic bundles (~4 to 7), as it decreases the computational cost of such simulations. It was found that the particle fluxes depend strongly on the ionization state of tungsten and associated poloidal asymmetries [3]. The low-Z tungsten ions move inwardly from the scrape-off-layer to the core while high-Z tungsten ions move outwardly from the core to the pedestal. The overall dynamics leading to an accumulation of tungsten ion in the pedestal. We started to implement atomic interactions between tungsten bundles [4]. Preliminary results including atomic interactions between these tungsten bundles in a pedestal will be presented. |
Wednesday, October 19, 2022 2:36PM - 2:48PM |
PO03.00004: Feasibility of a Full Cycle D-D Fusion Reactor Shawn Simko, Benedikt Geiger, Thomas Pütterich Full cycle deuterium-deuterium (D-D) fusion reactors benefit from the burning of D-D reaction products providing additional power via the deuterium-tritium (D-T) and deuterium-helium-3 (D-He3) fusion. However, the high temperatures needed to attain significant D-D fusion reactivity require consideration of relativistic effects and additional radiative loss mechanisms, which are insignificant in D-T reactors. Here, we present a 0-D model for the Lawson criterion of a generalized D-D reactor accounting for high energy radiative loss mechanisms. The confinement of impurity and fusion ash is considered using a set of fixed ratios of particle to energy confinement times. The supra-thermal He3 and tritium fusion products, produced at energies of a few MeV, have enhanced reactivity. This effect is included by adding on slowing down distributions for fast ions, enhancing ion burn-up by about 5 percent, which is beneficial for overall plasma performance. Finally, the 0-D model is extended to a simple 1-D model, including temperature and density profiles calculated using empirical scaling laws. This is iteratively solved to obtain steady-state conditions, allowing for the exploration of required reactor parameters. |
Wednesday, October 19, 2022 2:48PM - 3:00PM |
PO03.00005: The Scattering Interactions of Neutrons with 6Li and 7Li at Incident Neutron Energies of 14 MeV Using an Inertial Confinement Fusion Platform Chad J Forrest, Udo Schroeder, James P Knauer, Radha Bahukutumbi, Vladimir Y Glebov, S. P Regan, Arnold K Schwemmlein, Christian Stoeckl Lithium isotopes are primary components of the blankets proposed for continuous, on-site fuel regeneration in DT fusion reactor facilities. The 6Li(n, n’)dα and 7Li(n, n’)tα processes are presumed the main breeding channels for the deuterium-tritium fuel. As generator of the radioactive tritium, the latter process is of crucial importance. Knowledge of the particle production cross sections and the spectra of scattered and secondary neutrons from interactions of 14-MeV neutrons (> 100 keV) in thick targets is required for simulations of neutronics and, specifically, of the tritium breeding rates. In this work, cross sections were measured of neutron-induced processes in targets of isotopically enriched lithium isotopes. The experiments used an intense source generating 1011 14-MeV neutrons in pulses of 100-ps duration. Such bright neutron flashes are generated in inertial-confinement (ICF) DT implosions1 driven by the University of Rochester pulsed Laser System OMEGA. Secondary neutrons were measured in forward-angle geometry (θ = 0 to 7.4°) with a sensitive, neutron time-of‑flight spectrometer of high dynamic range. Total and double-differential cross sections of inelastic neutron scattering are inferred from the following reactions 6Li(n,n')dα and 7Li(n,n')tα and will be presented. This material is based upon work supported by the Department of Energy National Nuclear Security Administration under Award Number DE-NA0003856.
1 T. R. Boehly et al., Opt. Commun. 133, 495 (1997).
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Wednesday, October 19, 2022 3:00PM - 3:12PM |
PO03.00006: Novel Hybrid Reactor Concepts Based on Ignitor Technology and Physics Gilberto Faelli, Bruno Coppi, Matteo Salvetti, Renato Spigler, Ignitor Program Members A development of the Ignitor program, aimed at making fusion energy of near term relevance, is that of starting from technology and physics advances on which the Ignitor effort is based to conceive novel hybrid reactors. High field compact machines have produced record high density plasmas with excellent con?nement properties that can be utilized as neutron sources for power producing reactors with Thorium as its ?ssile component (E. P. Velikhov 2019). The Columbus concept [1], that had been studied as a follow-up to Ignitor in order to investigate the burn conditions of Tritium deprived plasmas, has been reconsidered as a neutron source to start by taking into account that approaching ignition is not the objective of it but other requirements have to be complied with. In this context, C. Bolton (2020) has suggested adopting pure-D plasmas for which the high field approach is appropriate and this attractive option is being analyzed including relevant advances in material science and ?ssion reactor engineering. |
Wednesday, October 19, 2022 3:12PM - 3:24PM |
PO03.00007: Non-thermal (``Cool'') Fusion Burning Plasma Regimes Alessandro Cardinali, Bruno Coppi, Bamandas Basu, Valeria Ricci Radially localized (``captive'') ballooning modes [1], capable of sustaining the transfer of energy from fusion reaction products to the reacting nuclei, have been identified. These modes, which involve high power transfers with acceptable particle density fluctuation levels, can lead to so-called ``cool fusion'' scenarios with considerably lower temperatures of the fusing nuclei than those associated with simple Maxwellian distributions. For DT plasmas the appropriate frequency of these modes is close to the deuterium cyclotron frequency for $k_{\perp}d_{i}{\sim}1$ and $d_{i}=c/{\omega_{pi}}$. Thus, the needed electron temperatures have to be adequate (e.g., around the ideal ignition temperature or higher) in order to avoid significant electron damping by the relevant ballooning mode-particle resonances [1]. These findings, which are consistent with recent experimental observations [2], suggest that evidence for non-thermal fusion should continue to be looked for and a serious effort should be devoted to exploit the advantages of this kind of fusion burning processes. |
Wednesday, October 19, 2022 3:24PM - 3:36PM |
PO03.00008: The importance of core fueling for the Q ≥ 5 inductive mission for the Burning Experimental Superconducting Tokamak (BEST) Tim Slendebroek, Joseph T McClenaghan, Andrea M. Garofalo, Lang L Lao, Siye Ding Self-consistent integrated modeling using STEP in OMFIT has shown that for the inductive scenario in BEST the fusion gain Q ≥ 5 mission can be achieved using moderately sized D-T fueling pellets. BEST, a new superconducting tokamak, is currently under design by ASIPP in China with 0D parameters R0=3.6 [m], a=1.1 [m], BT0=6.15 [T], Paux=41 [MW], including 10 [MW] NBI, 15 [MW] ICRF, 10 [MW] LHW and 6 [MW] ECH. The baseline standard H-mode operation scenario will be operating at Ip of 7.3 [MA], Zeffective of 2 and fgw~ 0.9 at the pedestal. For this scenario, the integrated modeling predicts a fusion performance of Q = 1.5, with a promising path to better performance by using core fueling pellets. To predict the density source of the pellets, the flexible Pellet Ablation Module (PAM) has been used. PAM has been previously validated on DIII-D experiments1. The injection angle, position, size, and velocity are constrained by engineering limitations. Therefore, pellets of 2mm - 4.5mm in radius, injected from the low field side at velocities of 1.5 km/s – 3.0 km/s were evaluated in the STEP workflow. The resulting fusion gain 1.5 < Q < 5.9 increases with pellet size and velocity as the deposition depth 0.3 < ρdep < 0.7 and density peaking 1.6 < ne0/neped < 2.8 increases at the same neped. |
Wednesday, October 19, 2022 3:36PM - 3:48PM |
PO03.00009: Machine learning combining simulation and experimental data for high-qmin scenario control development at DIII-D Joseph A Abbate, Rory Conlin, Egemen Kolemen, Emiliano Fable, Giovanni Tardini, Mark D Boyer We attempt to build on decades of experiments, theory, and simulations to more accurately predict and control systems. As a concrete use-case for such a model, we consider development of a model-predictive controller for achieving higher betan in DIII-D's high-qmin scenario. We use experimental data combined with simulations from the MHD stability code DCON (for tearing mode betan limit calculations), the transport codes ASTRA and TRANSP (for flux-driven profile evolution), and the fast-ion distribution code RABBIT (helpful for Alfven eigenmode analysis and predictions). We compare the accuracy and robustness of predictors and controllers trained on just simulation data, just on experimental data, and with both pieces of information. |
Wednesday, October 19, 2022 3:48PM - 4:00PM |
PO03.00010: Preparing for ITER plasma operation Michael L Walker, Jayson L Barr, Ivo Carvalho, David Eldon, Andrey Lvovskiy, George Sips, Zichuan A Xing, Brett E Chapman, Vincent R Graber, Eugenio Schuster, James J Yang, Joyeeta Sinha, Peter C de Vries, Hyun-Tae Kim We report on progress made in a sequence of experiments at DIII-D addressing key questions and testing operational scenarios in preparation for early ITER operation. These experiments build on one another so their execution is divided roughly equally across the 2022-23 campaigns. The work informs how to best achieve and maintain the desired plasma conditions needed for ITER first plasma and PFPO-1 objectives. One pair of experiments is designed to improve the physics understanding of the evolution of plasma parameters in EC-assisted plasma initiation and the potential for generation of runaway electrons during Ohmic initiation. Another set of experiments is designed to first add capability to the DIII-D PCS to enable emulation of ITER plant control systems and then use this capability to accurately emulate the execution of planned ITER scenarios. The latter work has a particular focus on development and validation of algorithms for handling exceptional events, i.e., those events that require either an asynchronous change of control algorithm/objective or a switch to an alternate scenario and its associated control algorithms designed to mitigate the consequences of the event. |
Wednesday, October 19, 2022 4:00PM - 4:12PM |
PO03.00011: Kinetic and magnetic control of fusion power fluctuations in EU-DEMO Mattia Siccinio, Pierre David, Luigi E Di Grazia, Emiliano Fable, Massimiliano Mattei, Francesco Maviglia, Christos Tsironis, Dirk Van Eester, Pietro Vincenzi, Wu Chuanren, Hartmut Zohm Plasma heating during the burn phase of the European demonstration reactor EU-DEMO will be determined by fusion-generated alpha particles. This occurrence leads to control issues which are absent in present tokamaks, where plasma heating is driven without any feedback from the plasma on the heating itself. Currently, a so-called “tokamak flight simulator” is being developed in EUROfusion, with the purpose of simulating all dynamic phases of the EU-DEMO operation. In particular, the 1.5D transport code ASTRA, which models the plasma and its magnetic equilibrium, has been coupled to the commercial control environment Simulink. This model has then been equipped with magnetic control blocks developed with the code CREATE-NL. The resulting tool models the action of kinetic and magnetic actuators, as well as the corresponding plasma dynamic response, in a self-consistent way. In this work, the code is employed to investigate the control strategy to counteract fusion power fluctuations during burn phase in EU-DEMO, which could lead to major operational issues if uncontrolled. The considered actuators are H&CD auxiliaries, mass injection systems and plasma shape and position control coils. Their interplay is investigated, towards the definition of a control strategy for the burn phase. |
Wednesday, October 19, 2022 4:12PM - 4:24PM |
PO03.00012: Interpretive modelling of fusion performance in JET DTE2 discharges with TRANSP Žiga Štancar, Krassimir Kirov, Fulvio Auriemma, Hyun-Tae Kim, Ridhima Sharma, Rita Lorenzini, Michal Poraziński, Paula Sirén, Ernesto Lerche, Mikhail Maslov, Dirk Van Eester, Yevgen Kazakov, Paola Mantica, Michael Fitzgerald, James Oliver, Zamir Ghani, Marina Gorelenkova, Francesca M Poli, Edward Litherland-Smith, Sheena Menmuir, Ephrem Delabie, Francis J Casson, Jeronimo Garcia Olaya Fusion power measurements are an essential tokamak operational parameter, therefore it is important to develop and validate integrated modelling tools capable of interpreting the fusion performance of current experiments, and extrapolate to ITER-like conditions. A range of plasma scenarios with varying fusion performance were tested in JET's recent deuterium-tritium campaign (DTE2), providing an opportunity to benchmark integrated modelling codes. We present an overview of interpretive modelling of over 80 JET DTE2 discharges using the TRANSP code. Our main aim is to assess the capability of reproducing the fusion performance of various plasma scenarios using different external heating and DT mixtures. We compare neutron rates measured by fission chambers and calculated ones, finding a strong dependency of the match between the two and absolute neutron rate. The calculations are found to agree with measurements for higher performing discharges with larger external heating power, while low-neutron shots have an average discrepancy of around + 40 %. A similar trend is found for the ratio between thermal and beam-target fusion, where larger discrepancies are seen in shots with beam-driven performance. We compare the observations to studies of JET's D and DTE1 campaigns, and assess uncertainties stemming from input diagnostics data. |
Wednesday, October 19, 2022 4:24PM - 4:36PM |
PO03.00013: Modeling of global impurity transport in the Proto-Material Plasma Exposure eXperiment (Proto-MPEX) during electron and ion cyclotron heating scenarios Atul Kumar, Cornwall H Lau, Timothy R Younkin, Juergen Rapp Radiofrequency based Ion cyclotron heating (ICH) and electron cyclotron heating (ECH) in the Material Plasma Exposure eXperiment (MPEX) are used to heat the ions and electrons independently and provide fusion divertor conditions ranging from sheath-limited to fully detached divertor regimes. In this report, predictive modeling using the Global Impurity TRansport (GITR) code [1-3] has been employed to study the erosion, and redeposition of Al impurities from helicon window in the presence of RF rectified sheath and their global transport up to the target in Proto-MPEX. To understand the role of electron heating on impurity transport, a sensitivity scan of electron temperature, density, and flow has been performed. During these scans, it has been found that electron heating shifts the peak of the impurity density profiles at the target, thereby reducing the impurity transport to the target. Radial impurity transport has also been consistent with the classical collisional impurity transport. However, modeling suggests that ionization also plays a crucial role in radial impurity transport. To understand the role of ICH, a similar sensitivity scan has been performed at the Alumina ICH antenna. Results of this scan will also be presented in this work. |
Wednesday, October 19, 2022 4:36PM - 4:48PM |
PO03.00014: Uncertainty Quantification for Model Predictive Control of Tokamak Plasma Boundary Simulations with SOLPS-ITER Sebastian De Pascuale, Vitaly Zankin, Jeremy D Lore, Ben Russo, Paul Laiu, Birdy Phathanapirom, Steven L Brunton, J. Nathan Kutz We deploy the SOLPS-ITER multi-fluid plasma and kinetic neutrals coupled transport solver to simulate dynamics of the tokamak plasma boundary in response to strategies for heat-flux mitigation on the divertor target. Several system identification approaches are implemented in the reduced modeling required for real-time predictive control. We merge Bayesian inference with the sparse identification of governing equations to develop a framework for model evaluation and selection. Analytic expressions for the steady-state correlations between upstream and downstream conditions are used as a reference for coupling scalar ODEs of the SOLPS-ITER computational domain. Training and testing over a series of gas puff actuation sequences for detachment control on the tokamak divertor are tuned to parameters for realistic integration with experimental devices. We demonstrate the successful application of adaptable and interpretable models in the feedback control of SOLPS-ITER simulations within the IPS integrated modeling environment. |
Wednesday, October 19, 2022 4:48PM - 5:00PM |
PO03.00015: OEDGE modeling of far-SOL tungsten impurity sources and screening in WEST Jake B Maeker, Jacob H Nichols, David C Donovan, Alex GROSJEAN, Jamie P Gunn, Nicolas Fedorczak, Christophe Guillemaut, Chris Klepper, Ezekial A Unterberg The WEST tokamak contains entirely W PFCs, and so represents an ideal environment to study W impurity transport. This work demonstrates that the divertor target experiences the highest gross W erosion in the device but is well screened for impurities penetrating into the main Scrape-Off Layer (SOL) plasma, while other plasma-facing component (PFC) regions like the pumping baffle experience lower erosion but can have higher proportional contamination of the near-SOL. The background plasma and impurity transport are modeled using the OEDGE code suite (DIVIMP, EIRENE, and OSM). Target data from divertor Langmuir probes and upstream data from interferometry/reflectometry and reciprocating Langmuir probes are used to create a background plasma solution with OSM and EIRENE. DIVIMP uses a Monte Carlo method to track individual impurities and simulate transport. W-I spectroscopic data from WEST at the lower divertor and baffle is used to estimate the sputtered W flux. Several shots over a power scan with between 1.5 – 2.35 MW reaching the SOL are modelled. Impurity densities near the separatrix are found for each modeled shot. Transport mechanisms for the migration of far-SOL impurities into the near-SOL are analyzed in terms of the force balances in the plasma models. Results indicate that increasing power does not necessarily increase SOL contamination from the baffle source but instead hints at a complex relationship that requires the assessment of how the power scaling affects the balance of friction forces and ion temperature gradient forces on impurities in the SOL. [Maeker, JNM, 2022]. |
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