M. Nishi

Ramp-Rate limitation due to current imbalance in a large cable-in-conduit conductor consisting of chrome-plated strands

The current distribution in the conductor, consisting of chrome-plated strands, was analysed assuming asymmetric strand transposition. The results show the circulation current is induced through the electrical joints at both ends of the conductor and electrical contact among the strands in the conductor. The current imbalance is produced as a result of the superimposition of the circulation and transport currents and becomes larger as the ramping rate increases. It was also found that the large current induced in the strands during a pulse charge cannot sufficiently be reduced at normal generation because of the induced voltage on these strands. The current flowing in the normal-state strands becomes larger for faster ramping. In addition, the effect of the non-uniform current distribution on the stability was experimentally investigated. The stability margin deteriorated when the current distribution in the conductor was not uniform. Moreover, the quench process in the ramp-rate limitation was considered. Since the coolant temperature is locally raised around the normal-state strands in the laminar-state coolant flow, the generation of the laminar flow region affects the ramp-rate limitation as a result of the current imbalance. From these results, it can be concluded that the current imbalance in the conductor has a very strong influence on the ramp-rate limitation.

The water detritiation system of the ITER tritium plant

Abstract The water detritiation system (WDS) of tritium plant for the International Thermonuclear Experimental Reactor (ITER) was designed. The concept of the Combined Electrolysis Catalytic Exchange (CECE) process was selected for the WDS. The design conditions are (a) tritium concentration of waste water: 3.7 × 1010∼3.7 × 1011 Bq/kg, (b) waste water flow rate: 20 kg/h (1100 mol/h), a net working rate: 300 days, annual capacity: 150 tons/year (c) tritium concentration in the H2 discharged to environment: 6.5 x 101 Bq/m3, (d) tritium concentration in the H2O vapor discharged to environment: 3.7 x 103 Bq/m3, (e) tritium concentration in the electrolyzer: ∼ 1.85 × 1013 Bq/kg. Tritium concentration in the electrolyzer is determined considering the lifetime of the electrolyzer which depends on tritium concentration. Design value of height of a unit (30cm) of water-hydrogen isotopic exchange column and the correlation between the column internal flow rates and the column diameter were determined based on similar system for Japanese advanced thermal reactor (Fugen) moderated with heavy water.

The tritium fuel cycle of ITER-FEAT

The Tritium Plant of ITER-FEAT is essential for the operation of the machine after the initial hydrogen phase, as tritium will be produced from DD fusion reactions. Within the fuel cycle of the Tokamak deuterium and later also tritium will be provided to the Fuelling Systems, and the unburned DT fraction recovered from the exhaust gases. The design of the tritium fuel cycle has to be based upon well proven technology to assure the safe handling of tritium along with credible accountancy, low tritium inventory, low generation of wastes and a high reliability of all components throughout the lifetime of ITER-FEAT.

Experimental results on instability caused by non-uniform current distribution in the 30 kA NbTi demo poloidal coil (DPC-U) conductor

Abstract Two 30 kA, NbTi Demo Poloidal Coils, DPC-U1 and DPC-U2, were fabricated and tested in the Demo Poloidal Coil project at the Japan Atomic Energy Research Institute. DPC-U1 and -U2 have a large current, forced flow cooling, cable-in-conduit conductor, which is composed of 486 strands. The strand surfaces are insulated by formvar to reduce coupling losses between the strands. DPC-U1 and -U2 reached their design current, but exhibited instability during charge, in many cases resulting in a coil quench. Such a quench occurred even at a current one-tenth of the conductor critical current. To clarify the cause of the instability, a detailed investigation on the quench current and normal voltage behaviour was carried out by charging the coil in several ways to the coil quench, and by measuring the stability of the coil at a current of 16–21.5 kA. These experimental results revealed the existence of non-uniformity of current distribution among the strands in the conductor, even under slow charging. This non-uniformity of current distribution caused the instability of the coil. The time constant of current redistribution is very large due to the insulation between the strands. However, if part of the conductor can be forced to go normal without coil quench occurring, a redistribution of current takes place and the current distribution becomes more uniform. It was then demonstrated that the current distribution could become uniform by applying heat to the conductor to generate intentional normalcy. Consequently, the possibility of stable operation of the DPC-U was suggested.

Ion species control in high flux deuterium plasma beams produced by a linear plasma generator

The ion species ratios in low energy high flux deuterium plasma beams formed in a linear plasma generator were measured by a quadrupole mass spectrometer. And the species control in the plasma generator was evaluated by changing the operational parameters like neutral pressure, arc current, and axial magnetic confinement to the plasma column. The measurements reveal that the lower pressures prefer to form more D+ ions, and the medium magnetic confinement at the higher pressures results in production of more D2+, while the stronger confinement and/or larger arc current are helpful to D2+ conversion into D3+. Therefore, the ion species can be controlled by adjusting the operational parameters of the plasma generator. With suitable adjustment, we can achieve plasma beams highly enriched with a single species of D+, D2+, or D3+, to a ratio over 80%. It has been found that the axial magnetic configuration played a significant role in the formation of D3+ within the experimental pressure range.

Construction of ITER common test facility for CS model coil

Japan Atomic Energy Research Institute is constructing the International Thermonuclear Experimental Reactor common test facility for the Central Solenoid Model Coil which is around 180 tons, a forced-flow cooled magnet with the maximum pulsed operation of 2 T/s and generates the rated magnetic field of 13 T at 48 kA with stored energy of 668 MJ. The test facility consists of a coil vacuum chamber, a cryogenic system with the 5-kW refrigerator and 500-g/s cryogenic pump, two pairs of 50-kA current leads, two DC power supplies (50 kA and 60 kA) and two JT-60 pulsed power supplies (50 kA, /spl plusmn/4.5 kV and /spl plusmn/40 kA, /spl plusmn/1.5 kV). The facility will be demonstrating the refrigeration and operation of a fusion pulsed magnet and the design and construction will accumulate experience towards the construction of ITER.

Design of the ITER tritium plant, confinement and detritiation facilities

Abstract This paper describes the design of the ITER tritium plant subsystems, layout in the tritium building and the construction plan. The tritium plant comprises tokamak fuel cycle processing systems, as well as tritium confinement and detritation systems. The plant processes tritiated gases received from the tokamak and other sources to produce the D, T gas streams for fuelling, and detritiates various waste streams including tritiated water before discharge to the environment. The plant has been designed to meet not only all anticipated plasma operation scenarios in the DD and DT phases with a wide range of burn pulse durations from short pulse (450 s) and long pulse (3000 s), but also safety requirements (minimization of equipment tritium inventory and environmental tritium release from different accidental events in tokamak and tritium processing subsystems, and reduction of workers’ tritium exposure, etc).

Extraction of Hydrogen from Water Vapor by Hydrogen Pump Using Ceramic Protonic Conductor

Abstract A blanket tritium recovery system that uses an electrochemical hydrogen pump with a protonic conductor membrane is proposed. One of the advantages of this system is the potential for processing the blanket sweep gas without fractionation of hydrogen isotopes and water vapor. In this work, hydrogen in a water molecule is extracted by a hydrogen pump using a Perovskite-type ceramic such as SrCe0.95Yb0.05O3-α. The threshold, which corresponds to the energy of H2O decomposition, for hydrogen extraction from the water molecule is 500 to 600 mV at 873 K. The threshold becomes smaller with increases of the partial pressure of the water vapor. In the case of pumping of the H2-H2O mixture gas, transportation of H2 precedes H2O decomposition below the threshold (H2O decomposition voltage), and the threshold becomes larger. In order to process the blanket sweep gas without fractionation of hydrogen isotope and water vapor, comparatively high applied voltage is required.

Analysis of hydrogen isotopes with a micro gas chromatograph

Abstract In the fuel cycle system of fusion reactors, analysis of hydrogen isotopes is very important from the view point of system control. The gas chromatograph (GC) with cryogenic separation column (cryogenic GC) is one of the most extensively used methods for the analysis of hydrogen isotopes. The micro GC with cryogenic column is expected to improve analysis time, that is a major disadvantage of conventional GC. The present authors have modified the micro GC to use its separation column at cryogenic temperature for H2, HD and D2 mixture analysis. Obtained retention time of H2, HD and D2 was about 85, 100 and 130 s, respectively. Peak resolution between H2 and HD, these are nearest each other, was about 1.0. These result suggests that the column developed in this work attained the practical level for the separation of hydrogen isotopes without tritium. Present detection limit of hydrogen isotopes was about 100–200 p.p.m., and it can be improved further by adjustment of separation column.

Strand production and benchmark testing for the ITER model coils

As part of the technology demonstration for the main features of the ITER Tokamak superconducting coils, two model coils, characteristic bore 2/spl divide/3 m, will be manufactured jointly by the four ITER partners. The coils will require a total of 26 tonnes of Nb/sub 3/Sn strand, supplied equally by each of the partners. The procurement of the strand is proceeding in stages, with performance and continuous quality demonstrated first on about 1t from each party underway since Sept. 93 and due for completion by Oct. 94. The strand uses both the bronze and internal tin routes, achieving jc(noncopper) in the range 550-700 A/mm/sup 2/ at 12 T and 4.2 K, with hysteresis losses from 200 to 600 mJ/cc(nonCu) for +/- 3 T cycle. Unit lengths >1.5 km are required with diameters about 0.8 mm. The status and parameters achieved in the production is reported. One of the first steps in confirming the strand quality has been to establish consistent testing procedures through a benchmark activity using strand exchange between all parties. The first block of testing was completed in May 94 and a second round is now underway. The results of the two rounds and the steps taken to standardise the testing are described.<<ETX>>