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Klepper, C. Christopher

Publications and source records attributed to Klepper, C. Christopher.

At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Eliminating Signal Bias Caused by Vacuum System Backstreaming in the Diagnostic Residual Gas Analyzer of ITER

In fusion neutral gas analysis, such as with the Diagnostic Residual Gas Analyzer (DRGA) for ITER, the primary measurement range of interest comprises the low-amu species (1 to 6), especially deuterium and helium. The challenge in successfully obtaining accurate measurements is two-fold. First, the sensitivity of the method must be sufficient to resolve trace amounts accurately; typically, one percent or less. Second, the gas signal from the fusion processes must be free of bias caused by the latent presence (from system outgassing and/or vacuum backstreaming) of these gases to enable accurate interpretation of the measured signal. This latter criterion can be problematic for the lightest gases since there is a propensity for some fraction of the pumped gas load to undergo a phenomenon known as backstreaming. This behavior is manifested in pumping systems for gas properties related to relative atomic weight (lightest) and size (smallest). Backstreaming results in a significant amount of the pumped gas undertaking a reverse flow and re-entering the measurement region; thus, contaminating the forward, real-time measurement. To fully eliminate this adverse effect, a conductance-limiting device – or orifice – has been installed in the high-vacuum pumping system of the present ITER DRGA prototype. The system was already equipped with a secondary turbomolecular pump (TMP), but with limited effectiveness against backstreaming in the inter-pump volume (IPV). This orifice is placed within the suction inlet coupling of the secondary TMP, which is downstream of the IPV. Its objective is to eliminate the backstreaming phenomenon by increasing the back pressure in the IPV. However, the orifice sizing must take into consideration other factors, such as the diagnostic measurement objectives. For example, in the ITER DRGA, one of the measurement requirements is a dynamic response time of ~1s. Fortunately, an added benefit of the pumping restriction created by the orifice is that the upstream pressure increase is beneficial for the DRGA’s optical gas analysis (OGA) sensors. These sensors are attached to the IPV in the present design. The glow discharges, when used as an OGA light source, will typically have a brighter light emission with increasing plasma cell pressure. In addition to the fusion machine research sector, there are other potential applications of this pumping technique where the monitoring of lighter gas concentrations is essential, such as the photolithography process for the semiconductor fabrication of integrated circuits. This presentation will describe the vacuum system used to demonstrate a process to eliminate backstreaming as well as show test results to verify the accomplishment of this critical objective.

Marcus, Chris↗
DOE OSTI2024

Design Considerations to Ensure Robustness of the ITER Diagnostics Residual Gas Analyzer

Increasing robustness of the ITER diagnostic residual gas analyzer (DRGA) is critical for the potential control of plasma heating and fuel-cycle processing. Robustness is a requirement for a diagnostic to have a control function. The DRGA is a multisensor diagnostic system capable of resolving isotopic compositions of hydrogen and helium as well as other heavier elements and compounds. The divertor-specific DRGA system is intended to measure the composition of gases in the ITER subdivertor region and midplane. Its analysis station will be located in a port cell at the divertor level. From there, it will sample a slip stream of gas from the cryogenic pump duct. It will then exhaust into a shared roughing line where helium or other light gas impurities, some potentially from other diagnostic systems, are likely to be present. In order to provide reliable measurements, the DRGA must be robust in areas such as plasma optical emission source geometry for a compact design, mitigation of back-streaming from light gases and resilience to the ITER port cell environment, and radiation hardening of the DRGA electronics. By incorporating robustness into the design, areas of plasma heating and fuel-cycle control may be explored with the DRGA for ITER and next-generation fusion devices.

Quinlan, Brendan↗
DOE OSTI2024

Complex multicomponent spectrum analysis with Deep Neural Network

Here, in this paper, we present the use of deep neural networks to estimate physical parameters from complex optical emission spectra of the D β /H β transition. Specifically, we focus on estimating the radio frequency electric field vector of the lower hybrid wave and isotope ratio within the scrape-off-layer plasma of the WEST tokamak. Fitting the spectral data using a traditional non-linear least squares analysis requires many free parameters and is computationally expensive, rendering the data unusable for real-time control. By implementing relatively small neural networks, the physical parameters can be directly extracted from the spectral data with reasonable accuracy in a few milliseconds. The deep neural network prediction can serve as input for a reduced model using least-squares fitting or for real-time control. We show that deep neural networks can be an effective tool for analyzing complex multicomponent spectra, providing a speedup of more than 10 5 times compared to least residual analysis, with an accuracy of 0.5% for the isotope ratio, and 0.09 kV/cm and 0.38 kV/cm for the RF radial and poloidal electric field respectively.

47 OTHER INSTRUMENTATION↗
DOE OSTI2023

Progress with applications of three-ion ICRF scenarios for fusion research: A review

The viability of magnetic confinement fusion as an energy source depends on achieving the high ion temperatures required for D-T fusion. Among the available techniques, plasma heating with waves in the ion cyclotron range of frequencies (ICRF) is a prominent method for bulk ion heating in fusion plasmas. Furthermore, a detailed understanding of the non-linear physics of alpha heating and the complex impact of MeV-range fast ions on plasma dynamics becomes progressively more important. This paper provides a comprehensive overview of recent developments with the three-ion ICRF scenarios on Alcator C-Mod, ASDEX Upgrade and JET tokamaks. The results demonstrate the flexibility of these novel scenarios for heating bulk ions in D-T ≈ 50%-50% plasmas and efficient generation of MeV-range fast ions in multi-ion species plasmas. In conclusion, several key results relevant for ITER and future fusion reactors are highlighted.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

ICRH operations and experiments during the JET-ILW tritium and DTE2 campaigns

2021 has culminated with the completion of the JET-ILW DTE2 experimental campaign. This contribution summarizes Ion Cyclotron Resonance Heating (ICRH) operations from system and physics point of view. Improvements to the (ICRH) system, to operation procedures and to real time RF power control were implemented to address specific constraints from tritium and deuterium-tritium operations and increase the system reliability and power availability during D-T pulses. ICRH was operated without the ITER-Like Antenna (ILA) because water leaked from an in-vessel capacitor into the vessel on day-2 of the D-T campaign. Three weeks were required to identify and isolate the leak and resume plasma operations. Dedicated RF-Plasma Wall Interaction (PWI) experiments were conducted; tritium plasmas exhibit a higher level of Be sputtering on the outer wall and impurity content when compared to deuterium or hydrogen plasmas. The JET-DTE2 campaigns provided the opportunity to characterize ICRH schemes foreseen for the ITER operation, in the ITER like wall environment in ELMy H-mode scenarios aiming at maximizing fusion performance. The second harmonic tritium resonance heating and to a lesser extent minority 3He heating (ITER D-T ICRH reference schemes) lead to improved ion temperature and fusion performance when compared to hydrogen minority ICRH. However, these discharges suffered from a lack of stationarity and gradual impurity accumulation potentially because of a deficit of ICRH power when using JET antennas at lower frequencies. Fundamental deuterium ICRH was used in tritium-rich plasmas and with deuterium Neutral Beam Heating; this ICRH scheme proved to be very efficient boosting ion temperature and fusion performance in these plasmas.

Klepper, C. Christopher↗
DOE OSTI2022

Developments and Challenges in the Design of the ITER DRGA

We report the ITER Diagnostic Residual Gas Analyzer (DRGA) will measure the distribution of gas species, i.e., deuterium (D), tritium (T), and impurities, in the divertor exhaust stream and in the plasma periphery, with time resolution relevant to fusion plasma–wall particle dynamics. The uniqueness of the DRGA, over previous implementations of plasma dynamics residual gas analysis, is an integrated approach, combining mass and low-temperature plasma-activated optical spectroscopy, in a differentially pumped analysis station. A further unique feature of the ITER divertor-specific DRGA is an ~8-m separation of the analysis station from the sampled pumping duct, while still maintaining a ~1-s response time for hydrogen isotopic concentrations. ITER DRGA final design activities are strongly benefiting from testing of prototypical DRGA components and methods on present fusion devices, most currently on JET and W7-X. DRGA systems are implemented on both these devices and include sensors (and pumping methods) that are directly relevant to the ITER DRGA design. The recent JET-DTE2 campaign has provided the first experience on operating the combined ITER DRGA sensors with D-T plasmas. While enhancing system design for ITER, this experience on operating devices has also revealed additional engineering challenges, which further guide the continuing final design project. Meanwhile, the recent determination that the ITER DRGA, with slight optimization, will resolve the helium isotopes well enough to support an ITER pre-DT, He-3-based heating scheme, has greatly increased ITER Research Program interest in the DRGA and its implementation well ahead of the DT phase.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Developing high performance RF heating scenarios on the WEST tokamak

High power experiments, up to 9.2 MW with LHCD and ICRH, have been carried out in the full tungsten tokamak WEST. Here, quasi non inductive discharges have been achieved allowing to extend the plasma duration to 53 s with stationary conditions in particular with respect to tungsten contamination. Transitions to H mode are observed, and H-modes lasting up to 4 s have been obtained. The increase in stored energy is weak since the power crossing the separatrix is close to the L–H threshold. Hot L mode plasmas (central temperature exceeding 3 keV) with a confinement time following the ITER L96 scaling law are routinely obtained. The weak aspect ratio dependence of this scaling law is confirmed. Tungsten accumulation is generally not an operational issue on WEST. Difficulty of burning through tungsten can prevent the discharge from accessing to a hot core plasma in the ramp-up phase, or can lead to rapid collapse of the central temperature when radiation is enhanced by a slight decrease of the temperature. Except a few pulses post-boronization, the plasma radiation is rather high (P rad /P tot ~ 50%) and is dominated by tungsten. This fraction does not vary as the RF power is ramped up and is quite similar in ICRH and/or LHCD heated plasmas. An estimate of the contribution of the RF antennas to the plasma contamination in tungsten is given.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Characterizing W sources in the all-W wall, all-RF WEST tokamak environment

In this work, experimental data, together with interpretive modeling tools, are examined to study trends in the tungsten (W) source in the all-W environment of the WEST tokamak, both from the divertor and from the main chamber. In particular, a poloidal limiter protecting an ion cyclotron resonance heating (ICRH) antenna is used as proxy for main chamber sourcing. The key study is carried out by stepping up lower hybrid current drive (LHCD) power, as the only auxiliary power source. Limiter and divertor W sources exhibit a qualitatively similar proportionality to the total power crossing the separatrix, P SEP , although the main chamber source remains substantially lower than the divertor source, for the range of P SEP accessible in the experiments. Intepretive modeling of the limiter source is carried out with a particle-in-cell (PIC) sheath model coupled to a surface sputtering model. Oxygen is used as a proxy for all light impurity species allowing for characterization of the critical W erosion regions. To get a good quantitative match to the data, it is necessary to assume that the oxygen arrives at the surface mostly at high ionization stages (4+ and above). A separate simulation with SOLEDGE-EIRENE, constrained to measured upstream scrape-off-layer plasma profiles, gives oxygen fractional abundances that are compatible with the PIC simulation result. This is understood to arise from transport processes that dominate over recombination. Substituting the LHCD by ICRH, in an equivalent experiment, the local W source exhibits a 3× enhancement. This can be matched by the simulation, by assuming local RF electric field rectification, based on ~100 eV peak-to-peak, near-antennna electric field. This work has highlighted the particular importance of understanding the ion charge state balance of light impurities as these are most likely the dominant sputtering species in fusion devices with high-Z walls.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Initial results from boron powder injection experiments in WEST lower single null L-mode plasmas

Using a recently installed impurity powder dropper (IPD), boron powder (<150 μm) was injected into lower single null (LSN) L-mode discharges in WEST. IPDs possibly enable real-time wall conditioning of the plasma-facing components and may help to facilitate H-mode access in the full-tungsten environment of WEST. The discharges in this experiment featured I p = 0.5 MA, B T = 3.7 T, q 95 = 4.3, t pulse = 12–30 s, n e,0 ~ 4 × 10 19 m –2 , and P LHCD ~ 4.5 MW. Estimates of the deuterium and impurity particle fluxes, derived from a combination of visible spectroscopy measurements and their corresponding S/XB coefficients, showed decreases of ~50% in O + , N + , and C + populations during powder injection and a moderate reduction of these low-Z impurities (~50%) and W (~10%) in the discharges that followed powder injection. Along with the improved wall conditions, WEST discharges with B powder injection observed improved confinement, as the stored energy W MHD , neutron rate, and electron temperature T e increased significantly (10%–25% for W MHD and 60%–200% for the neutron rate) at constant input power. Notably, these increases in confinement scale up with the powder drop rate and are likely due to the suppression of ion temperature gradient (ITG) turbulence from changes in Z eff and/or modifications to the electron density profile.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Improvements on the Diagnostic Residual Gas Analyzer at Wendelstein 7-X

Exhaust gas analysis provides key information on fusion processes, divertor operation, and wall state in fusion experiments and future reactors. The diagnostic residual gas analyzer (DRGA) concept has been developed for ITER with a focus on fast helium and hydrogen isotope detection. The first operation of the prototype DRGA (P-DRGA) at the stellarator Wendelstein 7-X showed potential for improvement in terms of magnetic sensor shielding, data acquisition automation, and potential new additions to the cluster of sensors on the P-DRGA. More recently, a Monte Carlo simulation of the flow of the mixed gas species effluent from the pressure-reducing orifice, down to about 8-m sampling tube and into the analysis region of the sensors, has been found to generally agree with previous calculations and measurements but revealed potential back-streaming effects for light gases, with impact on detection limits both for the prototype and for the ITER DRGA currently in design. For the upcoming campaign of the prototype, an enhanced soft iron shield will safeguard the gauges against magnetic stray field influence. The newly introduced shielding has been tested for its effect on magnetic stray fields and found to reduce the inside residual field by about two orders of magnitude.

Schlisio, G↗
DOE OSTI2021

Simultaneous H/D/T and 3He/4He absolute concentration measurements with an optical Penning gauge on JET

The Optical Penning (or Species-Selective Penning) Gauge technique is used for the measurement of the H2/D2/T2 fuel isotopic composition and He/D2 concentration in the neutralized particle exhaust of fusion devices. An Optical Penning detection system has already been used in JET during its first deuterium-tritium campaigns (DTE1) and has been chosen by ITER. Recent improvements to the sub-divertor diagnostic suite as well as the optical instrumentation, in preparation for the next DT campaign, have improved the optical throughput of the measurement and enabled 3He/D2 detection down to 0.1 %. For the first time, the 3He/4He isotopic ratio can be measured simultaneously with the H/D/T ratio. This technique is of particular interest for ion cyclotron resonance frequency (ICRF) heating scenarios.

Vartanian, S.↗
DOE OSTI2021

WEST actively cooled load resilient ion cyclotron resonance heating system results

Three identical new WEST ion cyclotron resonance heating (ICRH) antennas have been designed, assembled then commissioned on plasma from 2013 to 2019. The WEST ICRH system is both load-resilient and compatible with long-pulse operations. The three antennas have been successfully operated together on plasma in 2019 and 2020, with up to 5.8 MW of coupled power. In this work, the load resilience capability has been demonstrated and the antenna feedback controls for phase and matching have been developed. The breakdown detection systems have been validated and successfully protected the antennas. The use of ICRH in combination with lower hybrid has triggered the first high confinement mode transitions identified on WEST.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

First lower hybrid current drive experiments on the WEST tokamak

The first lower hybrid current drive experiments in the full tungsten WEST tokamak are reported. Good wave coupling is found at rather low plasma current (q 95 ~4.3) and medium density (n e ~3×10 19 m -3 ). Reflection coefficient is in agreement with the expectation from the linear theory of coupling. With low reflection coefficients, 5MW was coupled for 2 seconds. High central electron temperature, up to 5keV, is achieved at ne = 3-4×10 19 m -3 . Flat and even hollow profiles of tungsten density are derived from the bolometry diagnostic. The stored thermal energy follows the H96-P scaling law with very slight degradation with density. The current drive efficiency has been assessed in low loop voltage (V L ~0.15V) discharges. Low plasma current operation and rather high effective charge (Z eff ~3) lead to modest current drive efficiency (η = 0.5-0.65 × 10 19 A.W -1 m -2 ). Long pulse operation (~30s) with high LHCD power (P LH =2.7MW) is achieved with stationary parameters, in particular electron and impurity densities, with the upper water-cooled tungsten divertor.

Goniche, M.↗
DOE OSTI2020

Fast response, optical radiance measurements of low intensity impurity emission in WEST plasmas with staggered wavelength filters

Fast response acquisition of atomic line emission through the use of a filterscope has been implemented on the WEST tokamak at multiple poloidal locations. Filterscopes consist of a fiber-optic-based transmission from the tokamak to specifically engineered optical bandpasses and photomultiplier tubes (PMT) that collect the emission intensity to measure a radiance, L [W/cm2/str], over the specified bandpass. This diagnostic records plasma-wall interaction (PWI) properties (e.g. impurity emission and recycling of main fuel ions) up to a max acquisition rate of 100 kHz. These PMT radiometric measurements are calibrated into a line-normalized radiance, LN [photons/sec/cm2/str], similar to traditional spectrometers, which can later be converted to a particle flux. Low intensity emission peaks are difficult to quantify due to often comparable continuum levels, thus a secondary filter shifted to a judiciously-selected, line-free region (~1 nm away) allows for background subtraction. The system currently installed on WEST targets tungsten (W) gross sputtering specifically by monitoring, with a filter pair, the spectral region near the neutral W line emission (W I 400.9 nm). The line and background filters of the pair are centered at 400.6 nm and 403.1 nm, respectively. Through a set beam splitters on each sight line, the filter pair are measuring from the same location. The two L signals near W I are then subtracted from each other yielding only the W I LN. W I LN data from a recent WEST experimental campaign is presented and compared with plasma parameters and traditional spectrometer measurements of W I line emission to demonstrate the capabilities of this staggered-filter filterscope method. Specifically, the choice of WEST-specific bandpass curves for the W I LN will be illustrated based on this latter comparison.

Neff, Anton↗
DOE OSTI2020

Neutral gas analysis for JET DT operation

Neutral gas analysis, the measurement and evaluation of total and partial pressures, is a key technique to study the impact of neutral gas dynamics on retention, recycling and release processes of fuel or impurity species in fusion devices. At JET, the experiment closest to ITER in terms of operating parameters and size, various detectors and techniques for partial pressure and total pressure measurements are deployed together to characterise neutral gas dynamics during and after plasma operation on various toroidal and poloidal locations. An extensive modification of JET's sub-divertor neutral gas diagnostic system aims at retaining and extending established measurement capabilities in the forthcoming Deuterium-Tritium (DT) experiments (DTE2). To achieve DT compatibility, a separation of radiation-sensitive electronics from the sensor and adequate radiation shielding is required, as well as utilisation of a DT compatible differential pumping system with adjustable throughput to account for the strong pressure variation in the sub-divertor region. Finally, the sub-divertor neutral gas diagnostic will be equipped with multiple Residual Gas Analysers (RGAs), utilising quadrupole mass spectrometry and electrostatic ion-trap-principles, all operating with remote electronics located behind the biological radiation shield. These RGAs will record data in a fast selected discrete mass mode during plasma pulses (cycle ~2 s) and will automatically switch back to continuous data recording (cycle ~100 s) afterwards. They will be complemented by a newly improved Penning gauge spectroscopy configuration in particular supporting the He and D 2 separation relevant for DT operation. The distance between these devices and their associated control unit is typically 15m . A newly developed RGA with a cable length of 80 m, compatible with the ITER environment, will also be employed for the first time. Finally, this set-up and its operation in DTE2 will provide vital input to the development of the ITER divertor RGA in the most relevant environment currently available.

47 OTHER INSTRUMENTATION↗
DOE OSTI2019

RF sheath modeling of experimentally observed plasma surface interactions with the JET ITER-Like Antenna

Waves in the Ion Cyclotron Range of Frequencies (ICRF) enhance local Plasma-Surface Interactions (PSI) near the wave launchers and magnetically-connected objects via Radio-Frequency (RF) sheath rectification. ITER will use 20MW of ICRF power over long pulses, questioning the long-term impact of RF-enhanced localized erosion on the lifetime of its Beryllium (Be) wall. Recent dedicated ICRF-heated L-mode discharges documented this process on JET for different types of ICRF antennas. Using visible spectroscopy in JET ICRF-heated L-mode discharges, poloidally-localized regions of enhanced (by ~2–4x) Be I and Be II light emission were observed on two outboard limiters magnetically connected to the bottom of the active ITER-Like Antenna (ILA). The observed RF-PSI induced by the ILA was qualitatively comparable to that induced by the JET standard, type-A2 antennas, for similar strap toroidal phasing and connection geometries. The Be II line emission was found more intense when powering the bottom half of the ILA rather than its top half. Conversely, more pronounced SOL density modifications were observed with only top array operation, on field lines connected to the top half of the ILA. So far the near-field modeling of the ILA with antenna code TOPICA (Torino Polytechnic Ion Cyclotron Antenna), using curved antenna model, was partially able to reproduce qualitatively the observed phenomena. A quantitative discrepancy persisted between the observed Be source amplification and the calculated, corresponding increases in E // field at the magnetically connected locations to the ILA when changing from only top to only bottom half antenna operation. This paper revisits these current drive phased and half-ILA powered cases using for the new simulations flat model of the ILA and more realistic antenna feeding to calculate the E// field maps with TOPICA code. Further, the Self-consistent Sheaths and Waves for Ion Cyclotron Heating Slow Wave (SSWICH-SW) code, which couples slow wave evanescence with DC Scrape-Off Layer (SOL) biasing, is used to estimate the poloidal distribution of rectified RF-sheath Direct Current (DC) potential V DC in the private SOL between the ILA poloidal limiters. The approach so far was limited to correlating the observed, enhanced emission regions at the remote limiters to the antenna near-electric fields, as calculated by TOPICA. The present approach includes also a model for the rectification of these near-fields in the private SOL of the ILA. With the improved approach, when comparing only top and only bottom half antenna feeding, we obtained good qualitative correlation between all experimental measurements and the calculated local variations in the E // field and V DC potential.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Laboratory-based validation of the baseline sensors of the ITER diagnostic residual gas analyzer

The divertor-specific ITER Diagnostic Residual Gas Analyzer (DRGA) will provide essential information relating to DT fusion plasma performance. This includes pulse-resolving measurements of the fuel isotopic mix reaching the pumping ducts, as well as the concentration of the helium generated as the ash of the fusion reaction. In the present baseline design, the cluster of sensors attached to this diagnostic's differentially pumped analysis chamber assembly includes a radiation compatible version of a commercial quadrupole mass spectrometer, as well as an optical gas analyzer using a plasma-based light excitation source. This paper reports on a laboratory study intended to validate the performance of this sensor cluster, with emphasis on the detection limit of the isotopic measurement. This validation study was carried out in a laboratory set-up that closely prototyped the analysis chamber assembly configuration of the baseline design. This includes an ITER-specific placement of the optical gas measurement downstream from the first turbine of the chamber's turbo-molecular pump to provide sufficient light emission while preserving the gas dynamics conditions that allow for \textasciitilde 1 s response time from the sensor cluster [1].

Biewer, Theodore↗