Deuterium Operation Research Articles

Operation of a lithium beam diagnostic in tritium and deuterium–tritium experiments at JET

Abstract A fast neutral lithium beam has been used at JET for more than 30 years to measure profiles of electron density in the pedestal and scrape off layer. The diagnostic was not designed for tritium operation. In this paper, we describe the modifications which were required to allow operation in tritium (T) and deuterium–tritium (D-T) fuelled pulses at JET in 2021 and 2023. Methods to allow remote startup and remote control of the beam were implemented. The vacuum pump and exhaust system had to be made tritium compatible. Radiation vulnerable electronics had to be removed from the torus hall. The observation system relied on regular in vessel mirror exchange. An in-vessel protection shield was installed close to the observation mirror to protect the mirror surface. Operation in T and D-T discharges, maintenance post tritium exposure, commissioning and the calibration of the diagnostic are discussed.

Performance of ITER pressure gauges during deuterium operation in the large helical device

Abstract During deuterium campaigns on the heliotron large helical device (LHD), ITER pressure gauges with different cathode materials were used to measure the neutral pressure in the sub-divertor region. Throughout these campaigns, it was observed that the performance of the LaB6 cathodes was unsatisfactory, credited to the generation of neutrons impacting the gauges. Conversely, measurements taken with pressure gauges with a ZrC cathode performed well throughout the deuterium pulses. The ITER pressure gauge with the ZrC cathode could be operated with a very high electron current of 800 µA, thus improving the lower detection limit of the neutral pressure in LHD. With this design it was also possible to avoid jumps in the ion current within strong magnetic fields, improving the accuracy of the measurement from 15% uncertainty to 5%. These features allowed very precise neutral pressure measurements to be made in a fusion device with magnetic confinement. The total running time in the magnetic field was around 60 hours, less than the cathode life time of 350 hours.

JET machine operations in T&D-T

JET, the world’s largest operating tokamak with unique Be/W wall and tritium handling capability, completed a Deuterium-Tritium (D-T) campaign in 2021 (Maggi et al 29th Fusion Energy Conf.) following a decade of preparatory experiments, dedicated enhancements, technical rehearsals and training (Horton et al 2016 Fusion Eng. Des. 109–111 925). Operation with tritium raises significant technical, safety and scientific challenges not encountered in standard protium or deuterium operation. This contribution describes the tritium operational requirements, pulses and technical preparations, new operating procedures, lessons learned and details on the achieved operational availability and performance. The preparation and execution of the recent JET tritium experiments benefitted from the previous experience in 1991 (Preliminary Tritium Experiment), 1997 (DTE1 campaign) and 2003 (Trace Tritium Campaigns) and consisted of the following five phases: technical rehearsals and scenario preparation, tritium commissioning, 100% tritium campaign, D-T campaign (DTE2), tritium clean-up. Following the clean-up JET resumed normal operation and is currently undertaking a further D-T campaign (DTE3).

Contributions of the extended ELISE and BATMAN Upgrade test facilities to the roadmap towards ITER NBI

ITER’s NBI systems are a first of its kind system with very challenging targets for the RF-driven ion source and the acceleration stage. In a step ladder approach, the ion source test facilities BATMAN Upgrade (BUG) and ELISE support the activities carried out at the Neutral Beam Test Facility, Padua, which is equipped with the ion source facility SPIDER and with MITICA being equivalent to the ITER Heating Neutral Beam injector (HNB), capable of operating at the full power and pulse length of the ITER HNBs. The contributions of the prototype ion source at BUG (1/8 scale) and the size scaling experiment ELISE (1/2 size ITER source) to the roadmap are manifold: for hydrogen operation the ion source performance is demonstrated in several sequential 1000 s pulses, whereas long pulse deuterium operation is limited by the heat load of the co-extracted electrons on the extraction grid. Measures like special magnetic filter field configurations or biasing of surfaces and improved Cs management are identified. Both facilities have recently been extended to full steady state compatibility and very first insights of the ion source performance with steady state extraction compared to the previously used beam blips (10 s extraction every 150 s) are already gained. A pulse length of 400 s, as required for the first deuterium campaigns at ITER, seems to be feasible soon, whereas the one hour pulse imposes the highest challenge to overcome. Investigations on beam divergence revealed a divergence at the upper limit of the acceptable value for the HNB. Measurements on the beam uniformity on the scale of beamlet groups and grid segments at ELISE demonstrated a uniformity of better than the required 90%. BUG and ELISE gave input to recent implementations at SPIDER; MITICA and ITER’s NBI. Still open points and challenges are addressed, for which a continuation of the step ladder approach is essential.

Plasma window performance and scaling for an accelerator-based neutron source.

A "plasma window" was developed for use with a SHINE accelerator-based neutron source. In this work, the design of the plasma window is presented along with results demonstrating successful operation in deuterium over a range of aperture diameters (5-12mm), gas flow rates (3.5 to 12.5 slm), and input powers (10-55 kW). An empirically determined relationship between the aperture diameter, deuterium gas flow rate, input power, and plasma window pressure differential is also presented.

Tritium removal from JET-ILW after T and D–T experimental campaigns

After the second Deuterium–Tritium Campaign (DTE2) in the JET tokamak with the ITER-Like Wall (ILW) and full tritium campaigns that preceded and followed after the DTE2, a sequence of fuel recovery methods was applied to promote tritium removal from wall components. The sequence started with several days of baking of the main chamber walls at 240 °C and at 320 °C. Subsequently, baking was superimposed with Ion-Cyclotron Wall Conditioning (ICWC) and Glow Discharge Conditioning (GDC) cleaning cycles in deuterium. Diverted plasma operation in deuterium with different strike point configurations, including a Raised Inner Strike Point (RISP) configuration, and with different plasma heating—Ion Cyclotron Resonance Frequency (ICRF) and Neutral Beam Injection (NBI)—concluded the cleaning sequence. Tritium content in plasma and in the pumped gas was monitored throughout the experiment. The applied fuel recovery methods allowed reducing the residual tritium content in deuterium NBI-heated plasmas to about 0.1% as deduced from neutron rate measurements. This value is well below the requirement of 1% set by the maximum 14 MeV fusion neutron budget allocated in the ensuing deuterium plasma campaign. The quantified tritium removal over the course of the experiment was 13.4±0.7×1022 atoms or 0.67±0.03 g with ∼58% attributed to baking, ∼12.5% to ICWC, ∼26% to GDC, and ∼3.5% to first low power RISP plasmas. The experimentally estimated amount of removed tritium is in good agreement with long-term tritium accounting by the JET tritium reprocessing plant, in which the unaccounted amount was reduced by 0.71 g after the cleaning experiment.

Tritium Operation of the JET Neutral Beam Systems and Tritium NBI Power Calculations

Neutral beam injection (NBI) is a very flexible auxiliary heating method for tokamak plasmas, capable of being efficiently coupled to various plasma configurations required in the tritium and deuterium–tritium experimental campaign (DTE2) to be undertaken in the Joint European Torus (JET) device. In particular, experiments for high fusion yield and alpha particle studies require high-power NBI heating, and for maximum performance and optimum fuel mixture control in deuterium–tritium (D–T) plasmas, it is necessary to operate the JET NBI systems in both deuterium and tritium. The technical aspects of the JET NBI systems for compatibility with T operation are discussed, and the associated commissioning is described. The characterization of the JET NBI system in the tritium gas mode will be presented, with particular focus on the power and species mix measurements; this will be the first time that such data have been collected and analyzed for tritium neutral beams. Deuterium operation in the tritium gas mode was successfully carried out in 2019 with no loss in reliability. In this period of operation, the NBI power has been measured using beamline diagnostics and corroborated with plasma measurements. The species mix of the beam has been measured on the neutral beam test bed (NBTB) and also corroborated by plasma diagnostics on JET. These results will be presented alongside tritium NBI results allowing comparison of possible JET NBI performance between deuterium and tritium. Measurements of the NBI power in tritium show that there will be a higher neutralized fraction than in deuterium and a higher full energy fraction. When the effect of particle mass is also accounted for, this will lead overall to reduced beam penetration and lower particle flux per MW.

A high-resolution neutron spectroscopic camera for the SPARC tokamak based on the Jet European Torus deuterium-tritium experience.

Dedicated nuclear diagnostics have been designed, developed, and built within EUROFUSION enhancement programs in the last ten years for installation at the Joint European Torus and capable of operation in high power Deuterium-Tritium (DT) plasmas. The recent DT Experiment campaign, called DTE2, has been successfully carried out in the second half of 2021 and provides a unique opportunity to evaluate the performance of the new nuclear diagnostics and for an understanding of their behavior in the record high 14MeV neutron yields (up to 4.7 × 1018 n/s) and total number of neutrons (up to 2 × 1019n) achieved on a tokamak. In this work, we will focus on the 14MeV high resolution neutron spectrometers based on artificial diamonds which, for the first time, have extensively been used to measure 14MeV DT neutron spectra with unprecedented energy resolution (Full Width at Half Maximum of ≈1% at 14 MeV). The work will describe their long-term stability and operation over the DTE2 campaign as well as their performance as neutron spectrometers in terms of achieved energy resolution and high rate capability. This important experience will be used to outline the concept of a spectroscopic neutron camera for the SPARC tokamak. The proposed neutron camera will be the first one to feature the dual capability to measure (i) the 2.5 and 14MeV neutron emissivity profile via the conventional neutron detectors based on liquid or plastics scintillators and (ii) the 14MeV neutron spectral emission via the use of high-resolution diamond-based spectrometers. The new opportunities opened by the spectroscopic neutron camera to measure plasma parameters will be discussed.

Pushing the limits of existing plasma focus towards 1016 fusion neutrons with Q = 0.01

Existing conventional megajoule plasma focus machines with 2–3 MA are producing fusion neutron yields of several times 1011 in deuterium operation, the fusion yields predominantly being the beam-gas target. Increasing the current to 10 MA and using 50%–50% D-T mixture will scale the neutron yield towards 1016 D-T fusion neutrons. In this work, we derive the Lawson criterion for plasma focus devices with a beam-target fusion neutron mechanism, so that we may glimpse what future technological advancements are needed for a break-even Q = 1 plasma focus. We perform numerical experiments with a present-day feasible 0.9 MV, 8.1 MJ, 11 MA machine operating in 100 Torr in 50%–50% D-T mixture. The Lee Code simulation gives a detailed description of the plasma focus dynamics through each phase, and provides plasma and yield parameters which show that out of 1.1 × 1019 fast beam ions produced in the plasma focus pinch, only 1.24 × 1014 ions take part in beam-target fusion reactions within the pinch, producing the same number of D-T neutrons. The remnant beam ions, numbering at least 1019, exit the focus pinch at 1.9 MeV, which is far above the 115 keV ion energy necessary for an optimum beam-target cross-section. We propose to regain the lost fusion rates by using a high-pressure D-T-filled drift-tube to attenuate the energy of the remnant beam ions until they reach the energy for the optimum fusion cross-section. Such a fusion enhancement tube would further harvest beam-target fusion reactions by increasing the interaction path length (1 m) at increased interaction density (6 atm). A gain factor of 300 is conservatively estimated, with a final yield of 3.7 × 1016 D-T neutrons carrying kinetic energy of 83.6 kJ, demonstrating Q = 0.01.

On the effect of biased surfaces in the vicinity of the large extraction area of the ELISE test facility

The ion source at the ELISE test facility is an intermediate step towards the operation of the ITER NBI source and it demonstrated fulfillment of the ITER NBI requirements for accelerated negative current density in hydrogen. However, in deuterium operation the co-extracted electron current density (je) is higher and much more unstable and limits the source performance. In the standard setup of ELISE, je is reduced by a positive potential applied to the plasma grid (PG) with respect to the source body and the bias plate (BP). To further reduce and stabilize je in deuterium, an alternative scheme by biasing the BP is investigated. Measurements of the BP currents and of the extracted currents, combined with probe diagnostics in the vicinity of the PG for short-pulse in deuterium operation are presented. Biasing the BP, leads to a change in the distribution of the plasma potential in front of the extraction area, affecting the charged particle fluxes towards the BP and the PG and a strong reduction of je.

Difference of co-extracted electron current and beam acceleration in a negative ion source with hydrogen-isotope ions

Improvement of the performance on a hydrogen/deuterium negative ion source for a nuclear fusion device is reported. In particular, the suppression of the co-extracted electron current, Ie, is an important issue to ensure the stable beam acceleration. Improvement of the Ie has been confirmed by optimizing the magnetic field of the electron deflection magnet in the extraction grid. Two other new methods for reduction of the Ie were validated. The first was an electron fence whose rods were set between the rows of apertures on a plasma grid. The electron and negative ion current ratio, approximately Ie/I acc, was greatly improved from 0.7 to 0.25 in deuterium. The second was an outer iron yoke which enhanced the magnetic flux density 19% inside the arc discharge chamber. The Ie/Iacc using the outer yoke decreased by 0.1 compared with using a normal magnetic filter in a deuterium operation. These attempts have improved the total deuterium injection beam power of 8.4 MW by three negative ion based NBIs.

On filamentation in the dense plasma focus

Striking pictures showing filamentary structures in the plasma focus have intrigued researchers from the early days of plasma focus research. A definitive understanding of their occurrence, origin, structure, and role in plasma focus physics is still not in sight as summarized in a recent comprehensive review. This is because they are often not observed in a “standard mode” of plasma focus operation with pure deuterium, particularly in large installations, but are found in smaller experiments or those with gaseous admixtures. This has led to the suspicion that filaments are not a native feature of the plasma focus phenomenon. Recent success in observation of filaments in PF-1000 in pure deuterium operation by novel modification of the interferometer system that allows simultaneous interferometry and schlieren photography changes this situation. This Letter looks at the implications of this development in the larger context of plasma focus physics. Conceptualization of filamentation as a native feature of the traveling current distribution behind an ionizing strong shock wave is shown to be a feasible paradigm that can be formulated as a computable model for filamentation in the plasma focus.

Recent development of neutron and energetic-particle diagnostics for LHD deuterium discharges

An integrated set of neutron diagnostics developed for the deuterium operation of the Large Helical Device (LHD) has been revealing behavior of energetic ions in three-dimensional plasmas, together with energetic-particle diagnostics. In order to obtain deeper understanding of physics related to energetic ions in the LHD, development of plasma diagnostics that can provide energy distribution of energetic ions is now being accelerated. Recent advances in development of the D-D neutron energy spectrometer, a neutral particle analyzer based on a single-crystal chemical vapor deposition diamond, and a tangential fast-ion Dα diagnostic for deuterium discharges of the LHD are described.

Employing of ZrCo as a fuel source in a discharge-type fusion neutron source operated in self-sufficient mode

A discharge-type fusion neutron source generates neutrons by fusion reactions of hydrogen isotope atoms. In order to operate the fusion device based on the deuterium–tritium (D–T) fusion reaction, the tritium inventory is required to be decreased. In the present work, a self-sufficient system was installed into the fusion device for reducing the amount of hydrogen isotope fuel gas. The fuel gas was supplied and recovered with an intermetallic compound ZrCo in a sealed chamber in this system. A deuterium–deuterium operation was maintained for more than 60 min with stable discharge voltages, and the temperature of the ZrCo bed was changed to improve the neutron production rate. Factors that influenced the pressure inside the chamber were determined and optimized. Gas analysis using a quadrupole mass spectrometer indicates a dilution of the deuterium fuel was caused by hydrogen isotope exchange between the supplied deuterium gas and the absorbed protium on the chamber and electrodes surface. This system's hydrogen isotope control method and the amount of fuel were proposed for a D–T operation.

Neutron/Gamma separation in [formula omitted] thick single crystal diamonds

Single Crystal Diamond (sCD) are candidate neutron detectors for the ITER Radial Neutron Camera (RNC) due to their capability of performing both neutron counting and spectroscopy while withstanding high levels of radiation. This paper presents a study on methods of separation between neutron and gamma events, through the analysis of the pulse shapes of a 500μm thick sCD. Methods for gamma rejection may be beneficial especially during ITER deuterium–deuterium operation to extract the 2.5 MeV neutron signal. Four neutron/gamma separation algorithms for sCD neutron detectors are presented. Three algorithms rely on pulses shape information and one on the frequency analysis. The final experimental implementation based on multi-step separation approach achieved above 90% of correctly discriminated pulses.

Extension of high power deuterium operation of negative ion based neutral beam injector in the large helical device.

Second deuterium operation of the negative ion based neutral beam injector was performed in 2018 in the large helical device. The electron and ion current ratio improves to Ie/Iacc(D) = 0.31 using the short extraction gap distance of 7 mm between the plasma grid (PG) and the extraction grid (EG). The strength of the magnetic field by the electron deflection magnet installed in the EG increases by 17% at the PG ingress surface, which effectively reduces the electron component in the negative ion rich plasma in the vicinity of PG apertures. The reduction of the electron current made it possible to operate at a high power arc discharge and beam extraction. Then, the deuterium negative ion current increases to 55.4 A with the averaged current density of 233 A/m2. The thermal load on the EG using 7 mm gap distance is 0.6 times smaller than the thermal load using a 8 mm gap caused by the reduction of coextracted electron current. The injection beam power increases to 2.9 MW in the beam line BL3, and the total beam injection power increases to 7 MW by three beam lines in the second deuterium campaign.

Energetic ion confinement studies using comprehensive neutron diagnostics in the Large Helical Device

Understanding energetic particle (EP) confinement is one of the critical issues in realizing fusion reactors. In stellarator/helical devices, the research on EP confinement is one of the key topics to obtain better confinement by utilizing the flexibility of a 3D magnetic field. A study of EP transport in the Large Helical Device (LHD) has been performed by means of escaping EP diagnostics in hydrogen plasma operation. By starting deuterium operation of the LHD, the confinement study of EPs has progressed remarkably using newly developed comprehensive neutron diagnostics providing information for EPs confined in the core region. The total neutron emission rate (Sn) increases due to the relatively low deviation of the beam ion orbit from the flux surface with the inward shift of the magnetic axis. The Sn has a peak around the electron density of 2 × 1019 m−3 to 3 × 1019 m−3, as predicted. It is found that the fraction of beam–beam components in Sn is evaluated to be approximately 20% by the Fokker–Planck models TASK/FP in the plasma with both co- and counter-neutral beam injections. The equivalent fusion gain in DT plasma achieved 0.11 in a negative-ion-based neutral beam heated plasma. Time evolution of Sn following the short pulse neutral beam injection into the electron–cyclotron-heated low-beta plasma is reproduced by drift kinetic simulation, indicating that transport of a beam ion injected by a short pulse neutral beam can be described with neoclassical models in magnetohydrodynamic quiescent low-beta plasmas. The vertical neutron camera works successfully, demonstrating that in the co-neutral beam-injected plasma, the neutron emission profile shifts according to the magnetic axis position. The shift of the neutron emission profile is reproduced by orbit-following models. The triton burnup study is performed for the first time in a stellarator/heliotron to understand the alpha particle confinement. It is found that the triton burnup ratio, which largely increases at inward-shifted configurations due to the better triton orbit and better plasma performance in the inward-shifted configuration, is similar to that measured in a tokamak having a similar minor radius to the LHD. We study the confinement capability of EPs toward a helical reactor in the magnetohydrodynamic quiescent region and expansion of the energetic ion physics study in toroidal fusion plasmas.

The software and hardware architecture of the real-time protection of in-vessel components in JET-ILW

For the first time, the JET operation in deuterium–tritium (D–T) plasma, which is scheduled to take place on JET in 2020, will be performed in the ITER mix of plasma-facing component materials. In view of the preparation of the DT campaign (DTE2), several aspects of the plasma operation require significant improvements, such as a real-time protection of the first wall. The risk of damaging the metallic PFCs caused by beryllium melting or cracking of tungsten owing to thermal fatigue required a new reliable D–T compatible active protection system. Therefore, the future development of the JET real time first wall protection is focused on the D–T campaign and the ITER relevant conditions which may cause failure of camera electronics within the Torus hall. In addition to the technological aspect, the intensive preparation of the diverse software tools and real time algorithms for hot spot detection as well as alarm handling strategy required for the wall protection is in progress.This contribution describes the improved design, implementation, and operation of the near infrared (NIR) imaging diagnostic system of the JET-ILW plasma experiment and its integration into the existing JET protection architecture. To provide the reliable wall protection during the DTE2, two more sensitive logarithmic NIR camera systems equipped with new optical relays to take images and cameras outside of the biological shield have been installed on JET-ILW and calibrated with an in-vessel calibration light source (ICLS). Additionally, post-pulse data visualization and advanced analysis of all types of imaging data is provided by the new software framework JUVIL (JET users video imaging library). The formation of hot spots is recognized as a significant threat due to rapid surface temperature rise. Because it could trigger the protection system to stop a pulse, it is important to identify the mechanisms and conditions responsible for the formation of such hot spots. To address this issue the new software tool ‘Hotspot Editor’ has been developed.

Exploring deuterium beam operation and the behavior of the co-extracted electron current in a negative-ion-based neutral beam injector

The achievements of the deuterium beam operation of a negative-ion-based neutral beam injector (N-NBI) in the large helical device (LHD) are reported. In beam operation in LHD-NBIs, both hydrogen (H) and deuterium (D) neutral beams were generated by changing the operation gas using the same accelerator. The maximum accelerated deuterium negative-ion current () reaches 46.2 A from two beam sources with the averaged current density being 190 A m−2 for 2 s, and the extracted electron to accelerated ion current ratio () increases to 0.39 using 5.6 V high bias voltage in the first deuterium operation in 2017. An increase of electron density in the vicinity of the plasma grid (PG) surface, which is considered the main reason for the increase of co-extracted electrons in a beam, is confirmed by the half-size research negative-ion source in the neutral beam test stand at the National Institute for Fusion Science (NIFS). The deuterium negative-ion density is also larger than the hydrogen negative-ion density in the vicinity of the PG surface using the same discharge conditions. In the latest experimental campaign in 2018, increases to 55.4 A with the averaged current density being 233 A m−2 for 1.5 s using the shot extraction gap length. The low of 0.31 can be maintained by using high discharge power. The various parameters mentioned above are defined in detail below.

Achievements of the ELISE test facility in view of the ITER NBI

Neutral Beam Injection (NBI) for ITER shall deliver in total 33 MW heating power to the plasma with two injectors at a beam energy of 1 MV. Taking neutralization efficiency and all losses along the beam path into account a negative ion current density of 329 A/m2 (H− for 1000s) and 286 A/m2 (D− for 3600 s) has to be extracted from each ion source (size 1 × 2 m2) with a beam uniformity >90%. For the optimization of the radio frequency driven (RF) negative ion source and its extraction system the ELISE test facility has been set up at IPP as a size scaling experiment with a source area of 0.9 × 1 m2 and an extraction area of 0.1 m2.Since its experimental start in 2013 continuous progress was achieved improving the system technically for long pulse operation at high RF power (total 300 kW for 4 drivers) as well as approaching ITER´s physics requirements for the ion beam. For hydrogen ˜90% of the required ion current density can be extracted for long pulses (1000s); the accelerated current density even exceeded the ITER target. For deuterium operation the main challenge is the amount, temporal in-stability and spatial asymmetry of co-extracted electron current which limit the applicable RF power and the source performance.Shaping the magnetic field topology and controlling the electric fields inside the source together with an improved caesium conditioning procedure have been identified as possible tools to improve the source performance.