Wataru Shu

Tritium Removal from Codeposits on Carbon Tiles by a Scanning Laser

Abstract A novel method for tritium release has been demonstrated on codeposited layers on graphite and carbon-fiber-composite tiles from the Tokamak Fusion Test Reactor. A scanning continuous wave Nd laser beam heated the codeposits to a temperature of 1200–2300 °C for 10–200 ms in an argon atmosphere. The temperature rise of the codeposit was significantly higher than that of the manufactured tile material (e.g. 1770 °C cf. 1080 °C). A major fraction of tritium was thermally desorbed with minimal change to the surface at a laser intensity of 80 W/mm 2 , peak temperatures above 1230 °C and heating duration 10–20 ms. In two experiments, 46% and 84% of the total tritium was released during the laser scan. The application of this method for tritium removal from a tokamak reactor appears promising and has significant advantages over oxidative techniques.

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.

Ablative removal of codeposits on JT-60 carbon tiles by an excimer laser

The codeposits on JT-60 tiles experienced hydrogen plasma burning were irradiated by focused beams of an excimer laser. The removal rate of the JT-60 codeposits was low when the laser energy density was smaller than the ablation threshold (1.0 J/cm 2 ), but reached to 1.1 μm/pulse at the laser energy density of 7.6 J/cm 2 . The effective absorption coefficient k in the JT-60 codeposits at ArF excimer laser wavelength was determined to be 1.9 μm -1 , which is almost one order smaller than the optical absorption coefficient at the same wavelength in graphite (16.4 μm -1 ). In the process of ablative removal of the codeposits, hydrogen was released predominantly in the form of hydrogen molecule and water formation could be ruled out. The temperature rise on the surface was measured on the basis of Plancks law of radiation, and the temperature during the irradiation at the laser energy density of 0.5 J/cm 2 decreased from 3570 K at the beginning of the irradiation to 2550 K at 1000th pulse of the irradiation.

Tritium permeation study through tungsten and nickel using pure tritium ion beam

Abstract Permeation behavior of tritium (T) implanted into tungsten and nickel has studied by using pure tritium ion beam. Tritium permeation characteristics of steady and transient states were obtained and were compared with the result of deuterium (D) under the same experimental conditions. It was concluded from the steady state results of T and D that the permeation was controlled by diffusion in implantation side – diffusion in permeation side process for tungsten and recombination at the implantation surface – diffusion in the permeation side process for nickel. Effective diffusivities were evaluated from the permeation behavior of pure T and D. Isotope effect on the diffusivity between D and T did not agree with the classical diffusion theory, but could be explained by a quantum statistical diffusion model for nickel. The results indicated that diffusivity of T could be larger than that of D in W and Ni, respectively.

OPERATIONAL RESULTS OF THE SAFETY SYSTEMS OF THE TRITIUM PROCESS LABORATORY OF THE JAPAN ATOMIC ENERGY AGENCY

Abstract The building and safety systems of the TPL (Tritium Process Laboratory) were constructed in 1984 and 1985. The safety systems in the TPL have operated with tritium since March 1988. The amount of tritium held in the TPL was 13 PBq in March 2007. The average tritium concentration in a stream from a stack of the TPL to the environment was 6.0 x 10-3 Bq/cm3 and is 1/100 smaller than that of the regulatory value for the concentration of HTO in the air in Japan. Safe operation with tritium has been demonstrated. A set of failure data for several main components of the TPL was also obtained as valuable data for a fusion tritium facility.

Tritium contamination and decontamination study on materials for ITER remote handling equipment

Abstract Several materials, lenses, dry bearings and cables were exposed to a tritiated moisture environment to study the behavior of tritium contamination on candidate materials for ITER remote handling equipment. To optimize the tritium removal procedure, decontamination experiments using a gas purge with three different moisture concentrations were also performed. The surface tritium concentrations of CeO 2 containing alkaline barium glass (NB), CeO 2 containing lead glass (LX) and synthetic quartz (Quartz) after the exposure experiments were 7.80, 10.94 and 0.67 Bq/cm 2 , respectively. It was found that the tritium concentration was influenced by the compositions of the materials. The concentrations of tritium on type 831 (solid lubrication material: graphite) and type 237 (solid lubrication material: tungsten disulfate) dry bearings after the exposure experiments were 89.80 and 31.78 Bq/cm 2 , respectively. The tritium concentration in an electric cable tested was 5014 Bq/g after HTO moisture exposure. The tritium concentrations of lenses, LX, as typical experimental results, decreased to 2.72, 4.42 and 3.89 Bq/cm 2 by purging with the moist air, dry air and dry N 2 , respectively. The tritium concentrations of dry bearing, type 831 dropped to 6.61, 9.42 and 10.16 Bq/cm 2 by the same three decontamination treatments, respectively. A large decontamination factor of 13.6 was achieved in the case of type 831 dry bearing with a moist air purge. The tritium concentration in the electric cable was 3236 Bq/g after a moist air purge, and the decontamination factor was as low as 1.6. Therefore, decontamination with a moist air purge is not so effective for the electric cable.

Tritium decontamination of TFTR carbon tiles employing ultra violet light

Tritium decontamination on the surface of Tokamak Fusion Test Reactor (TFTR) bumper limiter tiles used during the Deuterium-Deuterium (D-D) phase of TFTR operations was investigated employing an ultra violet light source with a mean wavelength of 172 nm and a maximum radiant intensity of 50 mW/cm 2 . The partial pressures of H 2 , HD, C and CO 2 during the UV exposure were enhanced more than twice, compared to the partial pressures before UV exposure. In comparison, the amount of O 2 decreased during the UV exposure and the production of a small amount of O 3 was observed when the UV light was turned on. Unlike the decontamination method of baking in air or oxygen, the UV exposure removed hydrogen isotopes from the tile to vacuum predominantly in forms of gases of hydrogen isotopes. The tritium surface contamination on the tile in the area exposed to the UV light was reduced after the UV exposure. The results show that the UV light with a wavelength of 172 nm can remove hydrogen isotopes from carbon-based tiles at the very surface.

Tritium Decontamination of TFTR D-T Plasma Facing Components Using an Ultra Violet Laser

ABSTRACT Tritium decontamination of the surface of plasma facing components used during the deuterium-tritium (D-T) phase of the Tokamak Fusion Test Reactor (TFTR) was investigated using an ultra violet (UV) laser with a wavelength of 193 nm, a pulse energy of 200 mJ, a pulse duration of 25 ns and a beam size of 2.3 cm by 0.7 cm. Tritium was released immediately after the samples were irradiated by the UV laser. An initial spike of tritium release was observed within 40 seconds for each of three types of TFTR D-T plasma facing components. Most of the decrease in surface tritium concentration occurred in the first minute of UV laser irradiation. In a second experiment, the UV laser was focused to irradiate the deposited layers on JT-60 graphite tile that had experienced hydrogen plasma operation. The effective absorption coefficient and the ablation threshold for the JT-60 codeposits irradiated by the UV laser were determined to be 1.9 µm−1 and 1.0 J/cm2, respectively. An erosion rate of 1.1 µm/pulse was reached at a laser energy density of 7.6 J/cm2.

Isotope Effect of Hydrogen Rapidly Supplied from the Metal Storage Bed

Abstract In order to establish a proper control method of the DT fuel isotope balance in ITER, isotopic composition of hydrogen, which was released rapidly from the metal hydride bed by vacuum pump, was investigated using a ZrCo bed (1/10 scale of ITER fuel storage & delivery system) as functions of initially stored H/D ratio and temperature. The equilibrium pressure (P) of hydrogen – metal system has large isotope effect such as PH2 < PD2 < PT2 for ZrCo, however, the difference of H,D isotope fractions was within about 5%, during rapid delivery of about 90% hydrogen gases at 623 K and initial H:D of 1:1. In cases of initial H:D of 9:1 or 1:9, the differences of H,D isotope fractions were rather small of a few %. Even if the fluctuation of the isotope ratio is less than 5%, depending on the requirements from plasma physics experiments and fuel accountancy of tritium plant, batch fuel delivery from adequate gas tanks after isotope composition adjustment will be preferable to direct rapid delivery from storage bed.

Tritium Decontamination from Co-deposited Layer on Tungsten Substrate by Ultra Violet Lamp and Laser

Tritium decontamination using ultra violet (UV) lamp and laser was performed. Simulated co-deposited layer on tungsten substrate was deposited by C2H2 or C2D2 glow discharge. The co-deposited layer was irradiated to UV lights from a xenon excimer lamp (172 nm) or ArF excimer laser (193 nm) and the in-situ decontamination behavior was evaluated by a mass spectrometer. After the UV irradiation, the hydrogen concentration in the co-deposited layer was evaluated by elastic recoil detection analysis (ERDA) and the depth profile was analyzed by secondary ion mass spectrometry (SIMS). For the co-deposited layer formed by C2D2 glow discharge, it was found that M/e 3 (HD) gas was released mainly during the UV lamp irradiation while both M/e 3 (HD) and M/e 4 (D2) gases were detected during the UV laser irradiation. Though the co-deposited layer was not removed by UV lamp irradiation, almost all the co-deposited layer was removed by UV laser irradiation within 1 min. The ratio of hydrogen against carbon in the co-deposited layer was estimated to be 0.53 by ERDA and the number of photon needed for removing 1 fim thick co-deposited layer was calculated to be 3.7×1018 cm-2 for the UV laser by SIMS measurement. It is concluded that C-H (C-D) bond on the co-deposited layer were dissociated by irradiation of UV lamp while the co-deposited layer itself was removed by the UV laser irradiation.