S. Langish

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.

DETRITIATION OF THE JET CARBON TILES BY FLAME HEATING

Operation with tritium plasma led to contamination of the JET in-vessel components with tritium at a level exceeding 12kBq/g, which is the Low Level Waste (LLW) threshold in the UK. Carbon tiles used at JET for protecting the pumped divertor and inner wall against heat and neutron flux create one of the Intermediate Level Waste (ILW) streams to deal with during the JET decommissioning in the future. To reduce quantity and cost of ILW disposal from JET, the study has been initiated for development of detritiation techniques. This paper presents a brief description of the study of the JET carbon tiles detritiation using flame heating.

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.

Tritiated Dust Levitation by Beta Induced Static Charge

Abstract Tritiated particles have been observed to spontaneously levitate under the influence of a static electric field. Tritium-containing codeposits were mechanically scraped from tiles that had been used in the Tokamak Fusion Test Reactor (TFTR) inner limiter during the deuterium-tritium campaign and were placed in a glass vial. On rubbing the plastic cap of the vial, a remarkable “fountain” of particles was seen inside the vial. Particles from an unused tile or from a TFTR codeposit that formed during deuterium discharges did not exhibit this phenomenon. It appears that tritiated particles are more mobile than other particles, and this should be considered in assessing tokamak accident scenarios and in occupational safety.

Tritium Removal by Laser Heating and Its Application to Tokamaks

ABSTRACT A novel laser heating technique has recently been applied to removing tritium from carbon tiles that had been exposed to deuterium-tritium plasmas in the Tokamak Fusion Test Reactor (TFTR). A continuous wave neodymium laser, of power up to 300 watts, was used to heat the surface of the tiles. The beam was focussed to an intensity, typically 8 kW/cm2, and rapidly scanned over the tile surface by galvanometer driven scanning mirrors. Under the laser irradiation, the surface temperature increased dramatically, and temperatures up to 2,300 °C were recorded by an optical pyrometer. Tritium was released and circulated in a closed loop system to an ionization chamber that measured the tritium concentration. Most of the tritium (up to 84%) could be released by the laser scan. This technique appears promising for tritium removal in a next step DT device as it avoids oxidation, the associated deconditioning of the plasma facing surfaces, and the expense of processing large quantities of tritium oxide. Some engineering aspects of the implementation of this method in a next step fusion device will be discussed.

Comparison and evaluation of various tritium decontamination techniques and processes

In support of fusion energy development, various techniques and processes have been developed over the past two decades for the removal and decontamination of tritium from a variety of items, surfaces, and components. The motivational force for tritium decontamination by chemical, physical, mechanical, or a combination of these methods, is driven by two underlying forces. The first of these motivational forces is safety. Safety is paramount to the established culture associated with fusion energy. The second of these motivational forces is cost. In all aspects, less tritium contamination equals lower operational and disposal costs. This paper will discuss and evaluate the various processes employed for tritium removal and decontamination.

Studies of tritiated co-deposited layers in TFTR

Plasma facing components in TFTR contain an important record of plasma wall interactions in reactor grade DT plasmas. Tiles, flakes, wall coupons, a stainless steel shutter and dust samples have been retrieved from the TFTR vessel for analysis. Selected samples have been baked to release tritium and assay the tritium content. The in-vessel tritium inventory is estimated to be 0.56 g and is consistent with the in-vessel tritium inventory derived from the difference between tritium fueling and tritium exhaust. The distribution of tritium on the limiter and vessel wall showed complex patterns of co-deposition. Relatively high concentrations of tritium were found at the top and bottom of the bumper limiter, as predicted by earlier BBQ modeling.