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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.

TSP and atmospheric 210Pb concentrations were observed at Tsukuba, Japan during July 2001 to March 2005. TSP concentrations ranged from 14.5 to 152.4 µg m−3. 210Pb concentrations ranged from 0.04 to 0.86 mBq m−3 which was low in summer and high in winter to spring. Although TSP concentrations changed annually, 210Pb concentrations had almost the same concentration level and seasonal variation. The correlation coefficient between 210Pb concentration and NO3− concentration was slightly higher than for other ion species. This result indicated that both had been attached to the same particle size and transported to the sampling site.

In a large fusion test facility, when a deuterium-plasma experiment is conducted a small amount of tritium is produced by the d(d, p)t reaction. From the viewpoints of radiation management and public acceptance, the tritium monitoring and recovery systems were developed and installed for the fusion test device. As for the tritium monitoring equipment, an expiratory test system of tritium was utilized for the internal dose assessment of workers. Active tritium samplers were operated continuously to monitor the amount of tritium released from the stack. As for the tritium recovery equipment, an exhaust detritiation system (EDS) for the plasma experiment has been developed and installed at the downstream of the vacuum pumping system in the fusion test device. All of the exhausted tritium from the vacuum vessel was treated by the EDS during the deuterium-plasma experimental campaign. Then, the tritium recovery rate achieved was more than 95%.

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.

Deuterium plasma operations using a large fusion test device have been carried out since 2017 at the National Institute for Fusion Science. A small amount of tritium was produced by the fusion reaction, d(d, p)t. Then, a part of the tritium was released into the environment. Thus, monitoring the level of tritium in the environment around the fusion test facility is important. This is done before starting the deuterium plasma experiment. The environmental tritium concentrations indicated that they are at background levels in Japan. After starting the deuterium plasma experiment, the environmental tritium around the fusion test facility was within the range of environmental variation. This suggests that there was no impact of tritium on the environment during the first deuterium plasma experimental campaign.

A water bubbler system that can distinguish chemical forms of tritium was proposed for long-term tritium monitoring of the exhaust gas of a large fusion test device. The characteristics and performance of the water bubbler system were evaluated under operational conditions and confirmed to be suitable for tritium monitoring. For the tritium measurements, the water bubbler system determined the tritium activity and distinguished the chemical forms of tritium. The tritium activity and chemical forms in the exhaust gas provided helpful information to understand the tritium behavior in the large fusion test device.

The deuterium plasma experiment was started using the Large Helical Device (LHD) at the National Institute for Fusion Science (NIFS) in March 2017 to investigate high-temperature plasma physics and the hydrogen isotope effects towards the realization of fusion energy. In order to clarify any experimental impacts on precipitation, precipitation has been collected at the NIFS site since November 2013 as a means to assess the relationship between isotope composition and chemical species in precipitation containing tritium. The tritium concentration ranged from 0.10 to 0.61 Bq L&minus, 1 and was high in spring and low in summer. The stable isotope composition and the chemical species were unchanged before and after the deuterium plasma experiment. Additionally, the tritium concentration after starting the deuterium plasma experiment was within three sigma of the average tritium concentration before the deuterium plasma experiment. These results suggested that there was no impact by tritium on the environment surrounding the fusion test facility.

The Large Helical Device (LHD) started the deuterium plasma experiment on March 7, 2017. Approximately 6.4 GBq of tritium was produced in the first deuterium plasma experiment campaign and were utilized as tracers for the investigation of release behavior and the balance of tritium in the LHD vacuum vessel. To determine the tritium balance in LHD, the tritium release from the vacuum vessel was continually observed during the plasma experiment period and the vacuum vessel maintenance activities. The tritium exhaust rate was approximately 32.8% at the end of the plasma experiment. After the plasma experiment, the vacuum vessel was ventilated by room air for the maintenance activity and the tritium release from the in-vessel components was observed. The tritium release rate gradually decreased and became constant after four-month in spite of water vapor concentration. It is suggested that the tritium release mechanism from the vacuum vessel is a diffusion-limited process from the bulk. The tritium release amount during the maintenance activity for one year was approximately 5.0%. Considering the decrease of tritium decay for 1.5 years, tritium inventory in LHD was estimated to be approximately 3.66 GBq (57.2%) at the end of maintenance activity.

A new type of active tritium sampler that can discriminate between chemical forms in a fusion test facility without the use of combustion gases was developed. The proposed tritium sampler was operated using water vapour instead of combustion gases. To test the operation and performance of the device when water vapour is used, we evaluated the catalytic oxidation properties, and the evaporation and collection of water vapour under actual sampling conditions. The properties of the added water mass and the operation temperature of catalysts in the proposed sampling system were then determined. Thereafter, we carried out air sampling for tritium monitoring. The levels of tritium concentration measured by the proposed tritium sampling system were similar to the values measured by the conventional sampling system. Our findings show that the proposed tritium sampling system without combustion gases is a good replacement for the conventional tritium sampling system in a fusion test facility.

This study evaluates expiratory testing as a simple bioassay method for estimating the effective dose equivalent for individuals working inside radiation control areas in a large helical device (LHD) building at the National Institute for Fusion Science (NIFS) in Toki City, Japan. The expiratory testing apparatus is configured according to similar devices reported previously. The core components of the apparatus include the sample collection device, a water bubbler column, exhalation airflow controls, and a solution injection component. Exhaled water vapor from individual test subjects is collected and tritium levels are measured using a liquid scintillation counter. Tritium radioactivity measurements are then compared to the effective dose according to evaluation criteria estimated using the effective dose equation for tritium.

From the viewpoints of radiation management and public acceptance, the monitoring of tritium concentration in the stack is one of the key issues for the deuterium plasma experiment in the Large Helical Device (LHD). Since 2012, an active tritium sampler was employed in the stack to monitor the background levels of tritium and discriminate between its chemical forms. However, the operation procedure became complicated, so we developed a simplified active tritium sampler to reduce the work after sampling. In this sampler system, there was no discrimination between tritium chemical forms, instead, all tritium chemical forms were collected in one process. To verify the performance of the simplified active tritium sampler, tritium concentration data was compared with the existing tritium active sampler and environmental tritium data at the NIFS site. The results revealed a correlation between these tritium data. Therefore, the simplified active tritium sampler was applied in the stack with a primary monitoring system for the deuterium plasma experiment in the LHD.

To deal with endless data streams acquired in LHD steady-state experiments, the LHD data acquisition system was designed with a simple concept that divides a long pulse into a consecutive series of 10-s “subshots”. Latest digitizers applying high-speed PCI-Express technology, however, output nonstop gigabyte per second data streams whose subshot intervals would be extremely long if 10-s rule was applied. These digitizers need shorter subshot intervals, less than 10-s long. In contrast, steady-state fusion plants need uninterrupted monitoring of the environment and device soundness. They adopt longer subshot lengths of either 10 min or 1 day. To cope with both uninterrupted monitoring and ultra-fast diagnostics, the ability to vary the subshot length according to the type of operation is required. In this study, a design modification that enables variable subshot lengths was implemented and its practical effectiveness in LHD was verified., 2012 The Japan Society of Plasma Science and Nuclear Fusion Research

Because the LHD project has been operating with the cooperation of many institutes in Japan, the remote participation facilities play an important role. Therefore, NIFS has been introducing these facilities to its remote participants. Because the authors regard Web services as essential tools for the current Internet communication, Web services for remote participation have been developed. However, because these services are dispersed among several servers in NIFS, users cannot find the required services easily. Therefore, the authors developed a portal Web server to list the existing and new Web services for the LHD experiment. The server provides services such as summary graph, plasma movie of the last plasma discharge, daily experiment logs, and daily experimental schedules. One of the most important information from these services is the summary graph. Usually, the plasma discharges of the LHD experiment are executed every three minutes. Between the discharges, the summary graph of the last plasma discharge is displayed on the front screen in the control room soon after the discharge is complete. The graph is useful in evaluating the last discharge, which is important information for determining the subsequent experiment schedule. Therefore, it is required to display the summary graph, which plots more than 10 data diagnostics, as soon as possible. On the other hand, the data-appearance time varies from one diagnostic to another. To display the graph faster, the new system retrieves the data asynchronously; several data retrieval processes work simultaneously, and the system plots the data all at once.