Charles A. Gentile

Measurements of tritium retention and removal on the Tokamak Fusion Test Reactor

Recent experiments on the Tokamak Fusion Test Reactor have afforded an opportunity to measure the retention of tritium in a graphite limiter that is subject to erosion, codeposition, and high neutron flux. The tritium was injected by both gas puff and neutral beams. The isotopic mix of hydrogenic recycling was measured spectroscopically and the tritium fraction T/(H+D+T) transiently increased to as high as 75%. Some tritium was pumped out during the experimental run and some removed in a subsequent campaign using various clean‐up techniques. While the short term retention of tritium was high, various conditioning techniques were successful in removing ≊8000 Ci and restoring the tritium inventory to a level well below the administrative limit.

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.

Characterization of Carbon Tritide Particles in a Tokamak Fusion Reactor

ABSTRACT Amorphous tritiated carbon films are formed through co-deposition of the radioactive isotope tritium (3H or T) with carbon onto plasma facing surfaces in fusion plasmas. The Tokamak Fusion Test Reactor (TFTR), operated by the Princeton Plasma Physics Laboratory, was fueled by tritium and deuterium neutral beam injection and gas puffing. Tritium was co-deposited as amorphous hydrogenated carbon onto graphite tiles and stainless steel surfaces inside the reactor. Since termination of plasma operations, carbon tritide particles have remained in the air in the vessel. Dosimetric limits for occupational exposure to carbon tritide particles need to be established. The purpose of this study was to characterize carbon tritide particle samples inside the TFTR in terms of size, self-absorption of tritium beta, and dissolution rate in simulated lung fluid. Dose estimates of the inhaled carbon tritide particles can be calculated based on the dissolution rate, particle size, and self-absorption factor. The count median diameter and geometric standard deviation were 1.23 µm and 1.72, respectively, indicating that they are respirable particles and can stay suspended in the air for a longer time. The dissolution rate in the lung-simulated fluid was determined in a static system. The dissolution rate ranged from 10−1–10−3 per day in the first few hours, then it decreased to between 10−3 and 10−4. The retention curve of tritium in carbon indicated that >90% of the tritium remained in the particles after 110 d in the simulated lung fluid. This information is being used to support the establishment of respiratory protection requirements.

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.

Mathematical comparison of three tritium system effluent HTO cleanup systems

ABSTRACT It is important that air emissions from tritium systems be kept as low as reasonably achievable. Thus, over the years a number of gas detritiation systems have been developed. Recently there has been interest in lower-cost, simpler systems which do not convert HT to the much more hazardous HTO form. Examples of such systems are 1) a bubbler/dehumidifier, 2) a bubbler/collector, and 3) an adsorber/collector. A computer model of each configuration was written and run. Each system’s performance, including tritium buildup in liquid water, and tritium exhausted to the environment, are presented and compared.

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.

Oxidative decontamination of tritiated materials employing ozone gas

ABSTRACT The Princeton Plasma Physics Laboratory has developed a process by which to significantly reduce surface and near surface tritium contamination from various materials. The Oxidative Tritium Decontamination System (OTDS) reacts gaseous state ozone (accelerated by presence of catalyst), with tritium entrained/deposited on the surface of components (stainless steel, copper, plastics, ceramics, etc.) for the purpose of activity reduction by means of oxidation-reduction chemistry.1 In addition to removing surface and near surface tritium contamination from (high monetary value) components for re-use in non-tritium environments, the OTDS has the capability of removing tritium from the surfaces of expendable items, which can then be disposed of in a less expensive fashion. The OTDS can be operated in a batch mode by which up to approximately 20kg of tritium contaminated (expendable) items can be processed and decontaminated to levels permissible for free release (< 16.66Bq/100cm2). This paper will discuss the OTDS process, the level of tritium surface contamination removed from various materials, and a technique for “deep scrubbing” tritium from sub-surface layers.

Tritium decontamination of TFTR D-T graphite tiles employing ultra violet light and a Nd:YAG laser

The use of an ultra violet light source (wavelength=172 nm) and a Nd:YAG laser for the decontamination of TFTR D-T tiles will be investigated at PPPL. The development of this form of tritium decontamination may be useful for future D-T burning fusion devices which employ carbon plasma facing components on the first wall. Carbon tiles retain hydrogen isotopes, and the in-situ tritium decontamination of carbon can be extremely important in maintaining resident in-vessel tritium inventory to a minimum. A test chamber has been designed and fabricated at PPPL. The chamber has the ability to be maintained under vacuum, be baked to 200/spl deg/C. and provides sample ports for gas analyses. Tiles from TFTR that have been exposed to D-T plasmas will be placed within the chamber and exposed to either UV light source or the ND:YAG Laser. The experiment will determine the effectiveness of these two techniques for the removal of tritium. In addition, exposure rates and scan times for the UV light source and/or Nd:YAG laser will be determined for tritium removal optimization from D-T tiles.

High heat flux interactions and tritium removal from plasma facing components by a scanning laser

A new technique for studying high heat flux interactions with plasma facing components is presented. The beam from a continuous wave 300 W neodymium laser was focussed to 80 W/mm/sup 2/ and scanned at high speed over the surface of carbon tiles. These tiles were previously used in the TFTR inner limiter and have a surface layer of amorphous hydrogenated carbon that was codeposited during plasma operations. Laser scanning released up to 84% of the codeposited tritium. The temperature rise of the codeposit on the tiles was significantly higher than that of the manufactured material. In one experiment, the codeposit surface temperature rose to 1,770/spl deg/C while for the same conditions, the manufactured surface increased to only 1,080/spl deg/C. The peak temperature did not follow the usual square-root dependence on heat pulse duration.

SURFACE CHARACTERIZATION OF TFTR FIRST WALL GRAPHITE TILES USED DURING DT OPERATIONS

ABSTRACT Surface characterization studies were performed on graphite tiles used as first wall materials during DT operation of the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. These ex situ analysis studies revealed a number of interesting and unexpected features. In this work we examined the spatial and (where possible) the depth distribution of impurity species deposited onto the plasma facing surfaces using X-ray Photo-electron Spectroscopy (XPS) and Secondary Ion Mass Spectrometry (SIMS). This work determined that beyond the predominant species of carbon and oxygen, common impurities included silicon, boron, lithium and sulfur. Oxygen content in the plasma facing tile surfaces ranged from 20 to 50 atomic percent [excluding H-isotopes], clearly indicating an extensive zone of oxidized carbon. By contrast, carbon tile surfaces not exposed to the plasma have surface oxygen contents ranging from 2 to 6 atomic percent. Analytical measurements of the secondary impurities (B, Li, Si, S) levels were on the order of 1–4 atomic percent, (boron and lithium were injected for wall conditioning in TFTR.) The core level binding energies of these impurity species were consistent with the presence of common oxides or hydroxides (e.g., BxOy, Li2O, LiOH, Silicates). XPS measurements performed in concert with depth profiling indicated that the tile oxidized zone was significantly deeper than 1 micrometer into the (averaged) surface. Surface analytical results clearly indicate that plasma operations clearly redeposit injected impurities (Li, B) and the depth profiles and distributions of the hydrogen isotopes may be impactedand/or influenced by this deposition process. An attempt at determining hydrogen isotope concentration distributions was made using positive ion SIMS. Specific regions of some surfaces clearly indicated the presence of m/z=3 (HD, T) and m/z=15 (CH3, CHD, CT). Preliminary data examination using positive ion SIMS examination indicates that these mass markers are substantially higher in the near surface region when compared with spectra recorded deeper in the surface region. The deuterium and tritium concentrations were; however, sufficiently low or compromised bycommon isobaric interferencesthat accurate isotopic distributions using SIMS were not possible. These findings are in agreement with results reported by others. [Morimoto et al, Sun et al, reference 3 Haasz et al]