Deuterium Plasma Experiments Research Articles

Evaluation of Gamma-Ray Dose Distribution in the Toroidal Direction of LHD Vacuum Vessel from Radionuclides Generated in Deuterium Plasma Experiments

The evaluation of activities from radionuclides inside a vacuum vessel is required for the safe maintenance and estimation of the activation level of fusion devices. The Large Helical Device (LHD), Toki, Japan, has been performing deuterium plasma experiments since 2017, and radionuclides have accumulated in components such as vacuum vessels. In this work, the gamma-ray spectrum and activity of each detected radionuclide in an LHD vacuum vessel were evaluated using portable high-purity germanium (P-HPGe) and numerical calculations of photon transport for comparison with the removed armor tiles to evaluate the occupational exposure in the maintenance of LHD experiments. Measurements using the P-HPGe detector were performed at 20 positions along the toroidal direction of the LHD, at each O port, and between each pair of adjacent O ports. Using the estimated efficiency of the photoelectronic effects inside the P-HPGe detector calculated by Monte Carlo N-particle code version 6.1, the activities of the radionuclides in the armor tiles attached to the vacuum vessel were evaluated. The estimated activity concentrations were compared with the measured activities of the armor tiles removed from the vacuum vessel. Gamma rays from 58Co, 54Mn, and 60Co were detected, and the measured activities of 58Co and 54Mn generated by fast neutrons in the armor tile were almost consistent with those obtained from the removed armor tiles. The radioactivity of 60Co in the armor tile was underestimated compared with the activity obtained from the removed armor tiles. The measurement of activation by fast neutrons is feasible, whereas activation by thermal neutrons requires more accurate calculations and surveys to be measured using this method.

Divertor power load investigations with deuterium and tritium in type-I ELMy H-mode plasmas in JET with the ITER-like wall

Divertor power load is a major challenge towards a burning plasma in a next-step tokamak. Here, the first results of a divertor power load characterisation in tritium plasmas in type-I ELMy H-mode, obtained in the JET deuterium-tritium campaign (DTE2) performed in 2021, are presented. It is demonstrated that both, transient loads due to type-I ELMs as well as the power fall-off length, do not exhibit an explicit ion mass dependence, with remarkably similar values in the tritium plasmas and in the deuterium references. This gives an improved credence to published scaling law predictions, solely based on deuterium plasma experiments. Moreover, the type-I ELM impact on the inner divertor target is studied in deuterium discharges. A slightly increased parallel energy fluence on the inner target with a factor of 1.08 compared to the outer target is observed. This is explained by the smaller major radius of the inner target.

Tritium decontamination in vacuum vessel after deuterium plasma experiment due to air exposure

The deuterium plasma experiment using Large Helical Device (LHD) has been conducted since 2017. A small amount of tritium was produced in the vacuum vessel and a part of tritium remained in the plasma-facing materials. After the deuterium plasma experimental campaign, the vacuum vessel was opened for the LHD maintenance activity. Then, the plasma-facing material was exposed to room air and remained tritium was released at ambient temperature. The tritium release rate increased rapidly immediately after the vacuum vessel was opened to the atmosphere, but it turned to decrease soon after. After several weeks of opening the vacuum vessel, the tritium release rate became low and almost constant. The total tritium decontamination rate, defined as the ratio of tritium released by air exposure to tritium produced during the plasma experimental campaign, was achieved at several percent. The air exposure operation is expected to have a decontamination effect due to isotope exchange reactions on the surface of the plasma-facing material.

ECRH experiments on Tokamaks SST-1 & Aditya-U and ECRH upgradation plan for SST-1

A 42GHz-500kW ECRH system [1-6] is used to carry out various experiments related to plasma breakdown and ECR heating on tokamaks SST-1 and Aditya-U. The system has been upgraded with new anode modulator power supply to launch two ECRH pulses to carry out breakdown and heating simultaneously. In SST-1, ECRH system is used routinely for plasma breakdown at fundamental harmonic, approximately 150kW power is launched for 70ms to 150ms duration and consistent plasma start-up is achieved in SST-1. In the recent experiments, second EC pulse is also launched at the plasma flat-top to heat the plasma, some heating signatures are seen but more experiments will be carried out to confirm the plasma heating with ECRH. In Aditya-U tokamak, simultaneous plasma breakdown and heating experiments are carried out successfully [2]. In the first pulse around 100kW power in fundamental O-mode is launched for 70ms duration for the breakdown at low-loop voltage and around 150kW ECRH power for 50ms duration is launched in second EC pulse to heat the plasma. In case of Aditya-U, plasma heating is observed clearly as soft X-ray signal increases sharply with ECRH. In AdityaU tokamak, deuterium plasma experiments have been carried out and ECRH launched at the flat-top of deuterium plasma current. In deuterium plasma also ECR heating is observed as soft X-ray signal increases with ECH power. For SST-1, ECRH system is being upgraded with another 82.6GHz system, this system would be used to carry out plasma heating and start-up at second harmonic. The 82.6GHz system is already connected with the SST-1 tokamak, the old 82.6GHz-200kW Gyrotron will be upgraded to 400kW system to carry out effective heating experiments on SST-1 at higher ECRH power. The paper discusses the recent results of ECRH experiments carried out on tokamaks SST-1 & Aditya-U and presents the upgradation plan of EC system for SST-1.

TRITIUM CONCENTRATION IN MONTHLY PRECIPITATION NEAR THE FUSION TEST FACILITY IN JAPAN BEFORE AND AFTER THE DEUTERIUM PLASMA EXPERIMENT.

In Japan, the deuterium plasma experiment using the Large Helical Device was started at the National Institute for Fusion Science (NIFS) in March 2017 to investigate high-temperature plasma physics and hydrogen isotope effects in research leading towards the realisation of fusion energy. The deuterium plasma experiment produces small amount of tritium by fusion reactions. To understand any impacts by the experiment to the surrounding environment, monthly precipitation samples have been collected at the NIFS site since November 2013 to assess the relationship between isotope composition and chemical species in precipitation including tritium. By comparing data before and after the deuterium plasma experiment start, it was found that tritium released from the main stack of the fusion test facility had no impact on the environment surrounding NIFS.

LEVELS OF ATMOSPHERIC TRITIUM IN THE SITE OF FUSION TEST FACILITY.

In the deuterium plasma experiment using Large Helical Device at the National Institute for Fusion Science (NIFS), a small amount of tritium is produced by the D-D fusion reaction. Then, a part of produced tritium is discharged into the environment via a stack. Thus, the atmospheric tritium in the site of NIFS has been monitored before starting the deuterium plasma experiment. The atmospheric tritium concentrations at NIFS were indicated to be background levels in Japan. To investigate the impact of tritium discharged from the stack, the correlation between the atmospheric tritium concentration and the tritium concentration observed in the stack was evaluated, and no significant correlation was found. In addition, the atmospheric tritium concentration at NIFS ranged within the background levels in Japan. Therefore, the impact of discharged tritium from the stack would be negligible in the environment at NIFS.

Evaluation of Induced Radioactivity Generated during LHD Deuterium Plasma Experiments

Evaluation of Induced Radioactivity Generated during LHD Deuterium Plasma Experiments

Diagnostic of Fusion Neutrons on EAST Tokamak Using 4H-SiC Detector

Due to significant radiation hardness, thermal properties, low sensitivity to gamma rays, and fast response, the 4H-SiC detector exhibits promising properties in measuring the neutron production in the high-power fusion devices. To demonstrate the potential value of the 4H-SiC detectors as a neutron flux monitor and a neutron spectrometer on experimentally advanced superconducting tokamak (EAST) during the 2021 EAST deuterium–deuterium plasma experimental campaigns, the 4H-SiC detector covered with a 25- <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\mu \text{m}$ </tex-math></inline-formula> -thick <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">6</sup> LiF film 95% enriched in <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">6</sup> Li and the 4H-SiC detector covered with a 1.5-mm-thick polyethylene layer were installed on a viewport located at the mid-plane of port F outside the EAST vacuum vessel and operated continuously during the whole experimental campaign with the goal to measure both the total and the time-dependent neutron emission from the plasma. The results are compared with those obtained with detectors routinely working at EAST ( <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">235</sup> U Fission Chambers). Despite their small active volumes, the 4H-SiC detectors were able to measure the total and the time-dependent neutron emission with good reliability and stability. The results of this work not only demonstrated the satisfactory performance of the 4H-SiC detector in the harsh environment of a mixed neutron–gamma high-radiation field and strong electromagnetic field, but also its applicability as a neutron spectrometer and a neutron flux monitor for fusion deuterium–plasma experiments.

Development of a 56 GHz ECH system for deuterium plasma experiments of a low magnetic field in LHD

We have completed establishing a 56GHz electron cyclotron heating (ECH) system with a new gyrotron to realize high β plasma experiments with pure deuterium gas in the Large Helical Device (LHD). This new ECH system made it possible to conduct the relatively low magnetic field experiment around 1T with pure deuterium gas because tangential neutral beam injection (NBI) systems of deuterium beams did not have enough power to initiate plasma in the LHD. We succeeded in initiating and sustaining pure deuterium plasma in the magnetic field of 1T by using the 56GHz ECH system and the deuterium NBI (D-NBI). This new ECH system contributed to expanding the new experimental regime in the LHD.

Time-resolved secondary triton burnup 14 MeV neutron measurement by a new scintillating fiber detector in middle total neutron emission ranges in deuterium large helical device plasma experiments

A middle-sensitivity scintillating fiber detector (hereafter middle Sci-Fi detector) that works at a deuterium-tritium neutron flux of ~105-107 cm−2s−1 was utilized to measure secondary deuterium-tritium neutron emission rates with high temporal resolution at a total neutron emission rate of 1013 to 1015 n/s, where strong magnetohydrodynamic (MHD) instabilities were observed in the large helical device deuterium plasma experiments. The gain and angular characteristics of the middle Sci-Fi detector were evaluated in an accelerator-based deuterium-tritium neutron source in the intense 14 MeV neutron source facility at Osaka University. Observation of 1 MeV triton transport due to MHD instability was performed by a middle Sci-Fi detector whose deuterium-tritium neutron counting rate was approximately 20 times higher than that of the conventional Sci-Fi detector. Fusion-born triton transport due to energetic-particle-driven MHD instability was observed using the middle Sci-Fi detector due to its high detection efficiency and high discrimination ability of deuterium-tritium neutrons from the sea of deuterium-deuterium neutrons.

Removal of tritium from vacuum vessel by RF heated plasmas in LHD

The tritium removal from the vacuum vessel is one of the key issues for the realization of a fusion reactor. As in situ tritium removal techniques from the plasma facing materials, helium glow discharge wall conditioning (GDWC) and RF heated plasma due to electron cyclotron wall conditioning (ECWC) and ion cyclotron wall conditioning (ICWC) are applied with Large Helical Device (LHD) which remains tritium produced by deuterium plasma experiment. As the result, it suggests that ICWC could remove the tritium more efficiently than the ECWC and the GDWC. In the ICWC operation, the high energy particles induced by the minority ion heating mode as the standard ICRF heating operation may play an important role in tritium removal from the plasma-facing materials due to the particle’s bombardment.

Estimation of tritium inventory in exhaust detritiation system for fusion test device in the initial tritium recovery operation

As one of the radiation safety issues at nuclear fusion facilities, the development of the tritium handling technique is indispensable. A small amount of tritium is produced by the deuterium plasma experiment in a large fusion test device and is exhausted from the vacuum vessel. From the viewpoint of radiation safety and public acceptance, tritium in the exhaust gas could be recovered by the exhaust detritiation system (EDS). Most of the tritium recovered by EDS is stored in the wastewater tanks as tritiated water. Then they are fractionated into the wastewater containers for delivery to the Japan Radioisotope Association once a year. However, some of the tritiated water would have remained in EDS. The residual tritium would be the issue of the dismantlement of EDS in the future. Thus, the amount of residual tritium in EDS was estimated in the initial tritium recovery operation. Since most of the wastewater containing tritium remains in the storage tanks, the residual wastewater was the main tritium inventory. Also, the tritium in the molecular sieve used in EDS was estimated to be less than 0.1 GBq. Thus, the decontamination of molecular sieves would be preferred when dismantling the EDS.

Progress on Integrated Neutron Diagnostics for Deuterium Plasma Experiments and Energetic Particle Confinement Studies in the Large Helical Device During the Campaigns from FY2017 to FY2019

To achieve a steady-state fusion burning plasma, energetic particle confinement studies have been performed in fusion devices. Deuterium plasma experiments in the Large Helical Device starting from March 2017 have expanded the energetic ion confinement study toward a helical-type fusion reactor. To conduct this study, integrated neutron diagnostics, such as a wide dynamic range neutron flux monitor composed of three sets of detectors, a neutron activation system with two irradiation ends, three vertical neutron cameras, four types of scintillating fiber detectors, and a fast time response neutron fluctuation detector, were installed based on the plan and were working stably, as designed. Moreover, an energetic particle confinement study was advanced by utilizing integrated neutron diagnostics. Furthermore, the results of the energetic particle confinement study using neutron diagnostics obtained from FY2017 to FY2019 are reviewed in this paper.

Extending the total neutron emission rate of steady-state deuterium large helical plasma guided by a data-driven approach

A trial for extending the total neutron emission rate (Sn) was performed in the Large Helical Device (LHD) deuterium plasma experiment using a gas puff in order to show the way to realize a steady-state fusion reactor. In the 22nd experimental campaign of the LHD, a high Sn discharge was performed by the experimental scenario guided by a data-driven approach. A regression analysis of Sn was conducted beforehand using the externally controllable parameters based on the database of the previous three campaigns. Then, the high Sn record was successfully updated to be 3.7 × 1015 n/s in a steady-state discharge during the 22nd campaign in line with the regression expression. This study demonstrates that such a data-driven approach is effective for the gradual extension of Sn beyond the employed database.

Investigation on tritium retention and surface properties on the first wall in the large helical Device

• Examined tritium (T) retention in material probes on the first wall after the LHD 2017 campaign. • Studied the relationships between surface microstructures and T retention in material probes. • Much larger amounts of T remained in probes located at deposition-dominant area. • Among probes located at erosion-dominant area, remaining T in the probe on which energetic tritons impinged was much larger than others. In the Large Helical Device (LHD), the first deuterium plasma experiment was conducted in 2017. To investigate tritium migration in the LHD vacuum vessel, long-term material probes were installed on the first wall before the deuterium plasma experiment. After the experiment, the microstructure and amount of tritium remaining in each probe were analyzed. The results showed that a relatively large amount of tritium remained in the probes on the first wall, forming a thick deposition layer, rather than in the probes located in the erosion-dominant area. In the deposition layers on the probes, the dominant element is carbon, which can be generated on the divertor tiles made of graphite. The result of orbit calculation of the energetic tritons in the case of the standard magnetic configuration in the LHD showed that approximately 40% of the tritons generated by deuterium–deuterium fusion reactions were promptly lost mainly to the divertor. Thermalized tritons also flew to the divertor along with the background plasma. The divertor tiles, on which the tritons impinged, were eroded by the divertor plasma, and carbon atoms and tritiated hydrocarbon molecules were generated and deposited on the first wall. This can be the dominant mechanism of tritium retention in the first wall. Among the material probes located in the erosion-dominant area, the amount of tritium remaining in the probe on which the energetic tritons impinged was relatively large. The results of the tritium balance analysis show that the first wall is not the dominant reservoir of tritium in the LHD.

Performance of exhaust detritiation system for a fusion test device in the initial phase of the operation

Abstract The exhaust detritiation system (EDS) has been operating for the deuterium plasma experiment in large fusion test device since 2016. The EDS consists of two systems: the molecular sieves (MS) type for plasma exhaust gas, the polymer membrane (PM) type for vacuum vessel purge gas during the maintenance activity. The tritium removal performance of the EDS was evaluated for four years. As the operation results, the maximum detritiation factor (DF) of the MS system was achieved to be more than 103 even though the average tritium concentration in the process gas was less than 104 Bq/m3. On the other hand, the DF of the PM system was less than 103, because the average tritium concentration during the vacuum vessel maintenance activity was at the order of less than 103 Bq/m3. Throughout the initial phase of the deuterium plasma experiment, the tritium concentration at the outlet of EDS was maintained to be less than the order of 102 Bq/m3. Also, a failure analysis was performed to evaluate the reliability of the EDS in the initial phase of the operation. Over the four years of operation, the overall failure rate of the MS system was on the order of 10−5 events/hour.

Investigation of the distribution of remaining tritium in divertor in LHD

Abstract To reveal the triton transport and the tritium migration in a deuterium plasma experiment in the Large Helical Device (LHD), the distribution of the remaining tritium in divertor tiles made of graphite after the first deuterium plasma experimental campaign in 2017 was investigated. In this study, tritium contents in divertor tiles have been measured by using a full-combustion method. The asymmetric tritium retention in divertor tiles located at symmetric positions, which was found in the previous study by the surface tritium measurement using an imaging plate technique, has been confirmed by the results of the full-combustion method. The asymmetry is considered to be attributed to the asymmetric distribution of lost-points of energetic tritons in divertor. A depth profile of remaining tritium in a divertor tile estimated by using a combination of the imaging plate technique and a sputtering treatment shows that the peak of the profile locates at several micro-meters from the surface. This result suggests that the majority of the remaining tritium impinged upon the divertor tile as energetic tritons. In this study, a distribution of lost-points of energetic tritons has been calculated by using a Lorentz orbit following code (LORBIT) with taking into account divertor components. The obtained distribution has been compared with measured tritium distributions on divertor tiles. The measured and calculated distributions are similar to each other, but they are not the same. The difference between them can be attributed to plasma exposures during and after the deuterium plasma campaign.

Observations with fast visible cameras in high power Deuterium plasma experiments in the JET ITER-like wall tokamak

Abstract High speed visible imaging allows to visualize the 2-dimensional dynamics of fast phenomena in the boundary layer of fusion plasmas. Here we describe the two high speed visible cameras currently operating in the JET tokamak, enabling the simultaneous observation of a large fraction of the JET tokamak vacuum vessel with the appropriate configuration for many different plasma phenomena. As an example we discuss recent observations on the spatial and temporal dynamics of visible emission at the outer wall during Edge Localized Modes (ELMs) and the penetration of Shattered Injected Pellet (SPI) into the plasma.

Thermal neutron measurement by single crystal CVD diamond detector applied with the pulse shape discrimination during deuterium plasma experiment in LHD

Abstract The Pulse shape discrimination (PSD) technique for single crystal CVD diamond detector (SDD) was applied to the real-time thermal neutron measurement during the 3rd campaign of Deuterium-Deuterium (D-D) plasma experiment in the Large Helical Device (LHD). The PSD method is based upon the different shape of electrical pulses produced in diamond by gamma-ray (triangular-shaped pulse) and energetic ions (rectangular-shaped pulse), respectively. An improved PSD was developed which makes use of the full width at half maximum (FWHM) and the full width at 2/3 of the peak height (FW2/3 PH). These two parameters reflect the difference in the pulse shape produced by gamma-rays and energetic ions. The performance of this improved PSD technique was first tested irradiating the SDD by gamma-rays from 60Co and alphas from 241Am radiation sources, respectively. Then, using a 6LiF thermal neutron converter, a successful discrimination between gamma-rays and alpha particles and tritons generated by the 6Li(n,α)3H reaction induced by neutrons from 252Cf spontaneous fission neutron source was achieved. Thanks to this PSD processing technique, the thermal neutron detection efficiency increased about 1.7 times compared to previous measurements performed without the PSD technique. The real-time thermal neutron measurement by SDD with this enhanced PSD processing was also used during D-D plasma discharges in the Large Helical Device (LHD). It was found that both the time integrated counts and the time evolution count rate measured by the SDD reproduced well that of the fission chamber used as official LHD neutron monitor. The capability of the SDD detector using the PSD technique of precise real-time thermal neutron measurement in LHD was demonstrated.

Tritium Balance in Large Helical Device during and after the First Deuterium Plasma Experiment Campaign

Tritium Balance in Large Helical Device during and after the First Deuterium Plasma Experiment Campaign