K.R. Umstadter

Codeposition of deuterium with ITER materials

The levels of retention in codeposited layers of each of the three ITER materials (C, Be and W) are compared. Scaling laws, based on the conditions during the codeposition process (surface temperature, incident particle energy and ratio of the depositing fluxes), are presented to allow prediction of expected retention under ITER conditions. Retention in carbon codeposits scales inversely with incident particle energy, whereas in the metallic codeposits the retention level scales proportionally to increasing particle energy. The differing scaling of retention with incident particle energy provides insights into which material may impact the global retention in ITER depending on where it may form codeposits. In addition to the amount of retention, the release behaviour of tritium from codeposits will influence the tritium accumulation rate within ITER. The thermal release behaviour of T (or D) from codeposits can be used to evaluate the effectiveness of baking at different temperatures as a means of tritium removal. Finally, the desorption kinetics from Be and W codeposits are contrasted. In the case of W codeposits, the duration of the baking cycle is important in determining the removal efficiency, whereas with Be codeposited layers, the maximum achievable bake temperature plays the leading role in determining removal efficiency.

Archimedes Plasma Mass Filter

Some 60,000 and 46,000 MT of sodium rich nuclear waste are now in storage in the US at Hanford a nd SR S facilities, respectively. We have de veloped a technology that uses the high sodium content to advantage: aqueous slurry wastes are first calcined into sodium hydroxide ( NaOH) melt slurries, then vaporized an d injected into a pl asma. The Archimedes Filter separates plasma ions into light and hea vy mass groups. For the first time, it is feasible to economically separate large amounts of material in a single-pass plasma device. Such a s eparation would s ubstantially decontaminate H igh Level Waste since most radionuclides partition to the h eavy fraction. The p lasma process is based o n setting up fast ExB rotation of a cylindrical plasma. At a certain critical rotational velocity E > B/2 ions are not confined by axial magnetic field and are lost radially. Because the critical rotational velocity depends on magnetic field the plasma and machine parameters can be set u p to separate heavy radionuclides from majority of the light elements in the plasma and, thus, accomplish waste cl ean up. The paper discusses the F ilter process, describes a de monstration device t hat has been constructed i n San Diego, USA, and presents the first experimental results.

Effects of transient heating events on tungsten plasma-facing materials in a steady-state divertor-plasma environment*

Nearly all heat-pulse tests of plasma-facing materials (PFMs) have been completed in vacuum environments without the presence of a background plasma. Thus, combined effects of thermal transients on materials undergoing plasma exposure need to be explored. Heat-pulse experiments have been conducted in the PISCES-A device using a pulsed laser in a divertor-like plasma background. The results indicate that the erosion of PFMs is enhanced as compared with transient-only or plasma-only experiments, and the threshold energy for material removal by a transient heat pulse in a steady-state plasma background is reduced. There appears to be a minimum ion fluence to the surface needed to cause these effects. Initial experiments with pre-loaded material samples exposed to plasmas and heat pulsed in the lower divertor of the DIII-D tokamak using the Divertor Material Evaluation System indicate that similar effects occur in confinement devices.

The effect of thermo-oxidation on plasma performance and in-vessel components in DIII-D

In April 2010, two thermo-oxidation experiments (?O-bakes?) were performed in the DIII-D tokamak. Internal surfaces of the tokamak, as well as a number of specimens inserted into the torus, were exposed to a mixture of 20% O2/80% He at a nominal pressure of 9.5?Torr (1.27?kPa) at a temperature of 350?360??C for a duration of 2?h. Three primary conclusions have been drawn from these experiments: (1) laboratory measurements on the release of deuterium from tokamak codeposits by oxidation have been duplicated in a tokamak environment, (2) no internal tokamak components or systems were adversely affected by the oxidation and (3) the recovery of plasma performance following oxidation was similar to that following regular torus openings.

Fuel retention in carbon materials under ITER-relevant mixed species plasma conditions

Samples of CFC NB41 and fine-grain graphite ATJ have been exposed to PISCES plasmas containing (i) pure deuterium, (ii) deuterium and beryllium, (iii) deuterium, beryllium and helium, and (iv) deut ...

Mixed Material Plasma‐Surface Interactions in ITER: Recent Results from the PISCES Group

This paper summarizes recent PISCES studies focused on the effects associated with mixed species plasmas that are similar in composition to what one might expect in ITER. Formation of nanometer scale whiskerlike features occurs in W surfaces exposed to pure He and mixed D/He plasmas and appears to be associated with the formation of He nanometer‐scaled bubbles in the W surface. Studies of Be‐W alloy formation in Be‐seeded D plasmas suggest that this process may be important in ITER all metal wall operational scenarios. Studies also suggest that BeD formation via chemical sputtering of Be walls may be an important first wall erosion mechanism. D retention in ITER mixed materials has also been studied. The D release behavior from beryllium co‐deposits does not appear to be a diffusion dominated process, but instead is consistent with thermal release from a number of variable trapping energy sites. As a result, the amount of tritium remaining in codeposits in ITER after baking will be determined by the maximum temperature achieved, rather than by the duration of the baking cycle.This paper summarizes recent PISCES studies focused on the effects associated with mixed species plasmas that are similar in composition to what one might expect in ITER. Formation of nanometer scale whiskerlike features occurs in W surfaces exposed to pure He and mixed D/He plasmas and appears to be associated with the formation of He nanometer‐scaled bubbles in the W surface. Studies of Be‐W alloy formation in Be‐seeded D plasmas suggest that this process may be important in ITER all metal wall operational scenarios. Studies also suggest that BeD formation via chemical sputtering of Be walls may be an important first wall erosion mechanism. D retention in ITER mixed materials has also been studied. The D release behavior from beryllium co‐deposits does not appear to be a diffusion dominated process, but instead is consistent with thermal release from a number of variable trapping energy sites. As a result, the amount of tritium remaining in codeposits in ITER after baking will be determined by the maxim...