Publications

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Overview of Plasma Materials Interaction (PMI) activities are: (1) Hydrogen diffusion and trapping in metals - (a) Growth of hydrogen precipitates in tungsten PFCs, (b) Temperature dependence of deuterium retention at displacement damage, (c) D retention in W at elevated temperatures; (2) Permeation - (a) Gas driven permeation results for W/Mo/SiC, (b) Plasma-driven permeation test stand for TPE; and (3) Surface studies - (a) H-sensor development, (b) Adsorption of oxygen and hydrogen on beryllium surfaces.

Unbalanced double null ELMy H-mode configurations in DIII-D are used to simulate the situation in ITER high triangularity, burning plasma magnetic equilibria, where the second X-point lies close to the top of the vacuum vessel, creating a secondary divertor region at the upper blanket modules. The measured plasma conditions in the outer secondary divertor closely duplicated those projected for ITER. {sup 13}CH{sub 4} was injected into the secondary outer divertor to simulate sputtering there. The majority of the {sup 13}C found was in the secondary outer divertor. This material migration pattern is radically different than that observed for main wall {sup 13}CH{sub 4} injections into single null configurations where the deposition is primarily at the inner divertor. The implications for tritium codeposition resulting from sputtering at the secondary divertor in ITER are significant since release of tritium from Be co-deposits at the main wall bake temperature for ITER, 240 C, is incomplete. The principal features of the measured {sup 13}C deposition pattern have been replicated by the OEDGE interpretive code.

The mechanisms governing D loss in ErD2 films with and without a Mo diffusion barrier on kovar substrates were studied between 200 and 600 °C via in-situ Ion Beam Analysis (IBA). Significant intermixing between kovar and Er was observed above 450°C and between kovar and ErD2 above 500 °C. The D loss mechanism in ErD2 films was found to change from intermixing between kovar and ErD2 at low temperatures (< 500 °C) to thermal decomposition at higher temperatures (> 500 °C). Diffusion between kovar and ErD2 was measured isothermally at 500 and 550 °C. An activation energy of 2.1 eV and a pre-exponential factor of 0.071 cm2/s were determined. Diffusion between the kovar components and ErD2 film was inhibited by depositing a 200 nm Mo diffusion barrier between the kovar substrate and the ErD2 film. The processing of the Mo diffusion barrier was shown to impact its performance. Intermixing between the kovar / Mo / ErD2 stack becomes significant between 500 and 550 °C with a sputter deposited Mo diffusion barrier and between 550 and 600 °C for an electron-beam evaporated Mo diffusion barrier. Copyright © 2009 ASM International® All rights reserved.

The mechanisms governing D loss in ErD2 films with and without a Mo diffusion barrier on kovar substrates were studied between 200 and 600 °C via in-situ Ion Beam Analysis (IBA). Significant intermixing between kovar and Er was observed above 450°C and between kovar and ErD2 above 500 °C. The D loss mechanism in ErD2 films was found to change from intermixing between kovar and ErD2 at low temperatures (< 500 °C) to thermal decomposition at higher temperatures (> 500 °C). Diffusion between kovar and ErD2 was measured isothermally at 500 and 550 °C. An activation energy of 2.1 eV and a pre-exponential factor of 0.071 cm2/s were determined. Diffusion between the kovar components and ErD2 film was inhibited by depositing a 200 nm Mo diffusion barrier between the kovar substrate and the ErD2 film. The processing of the Mo diffusion barrier was shown to impact its performance. Intermixing between the kovar / Mo / ErD2 stack becomes significant between 500 and 550 °C with a sputter deposited Mo diffusion barrier and between 550 and 600 °C for an electron-beam evaporated Mo diffusion barrier. Copyright © 2009 ASM International® All rights reserved.

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{sm_bullet}Mixing from some thermal process steps thought to drive H,D,T loss - This does not appear to be a problem with the Mo/Er occluder stacks {sm_bullet}Diffusion barriers investigated to prevent mixing

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Deposition in next-step devices such as ITER will pose diagnostic challenges. Codeposition of hydrogen with carbon needs to be characterized and understood in the initial hydrogen phase in order to mitigate tritium retention and qualify carbon plasma facing components for DT operations. Plasma facing diagnostic mirrors will experience deposition that is expected to rapidly degrade their reflectivity, posing a challenge to diagnostic design. Some eroded particles will collect as dust on interior surfaces and the quantity of dust will be strictly regulated for safety reasons however, diagnostics of in-vessel dust are lacking. We report results from two diagnostics that relate to these issues. Measurements of deposition on NSTX with 4 Hz time resolution have been made using a quartz microbalance in a configuration that mimics that of a typical diagnostic mirror. Often deposition was observed immediately following the discharge suggesting that diagnostic shutters should be closed as soon as possible after the time period of interest. Material loss was observed following a few discharges. A novel diagnostic to detect dust particles on remote surfaces was commissioned on NSTX.

{sup 13}CH{sub 4} was injected with a toroidally-symmetric gas system into 22 identical lower-single-null L-mode discharges on DIII-D. The injection level was adjusted so that it did not significantly perturb the core or divertor plasmas, with a duration of {approx}3 s on each shot, for a total of {approx}300 T L of injected particles. The plasma shape remained very constant; the divertor strike points were controlled to {approx}1 cm at the divertor plate. At the beginning of the subsequent machine vent, 29 carbon tiles were removed for nuclear reaction analysis of {sup 13}C content to determine regions of carbon deposition. It was found that only the tiles inboard of the inner strike point had appreciable {sup 13}C above background. Visible spectroscopy measurements of the carbon injection and comparisons with modeling are consistent with carbon transport by means of scrape-off layer flow.

A major research goal of the national spherical torus experiment is establishing long-pulse, high beta, high confinement operation and its physics basis. This research has been enabled by facility capabilities developed during 2001 and 2002, including neutral beam (up to 7 MW) and high harmonic fast wave (HHFW) heating (up to 6 MW), toroidal fields up to 6 kG, plasma currents up to 1.5 MA, flexible shape control, and wall preparation techniques. These capabilities have enabled the generation of plasmas with {beta}{sub T} {triple_bond} <p>/(B{sub T0}{sup 2}/2{mu}{sub 0}) of up to 35%. Normalized beta values often exceed the no-wall limit, and studies suggest that passive wall mode stabilization enables this for H mode plasmas with broad pressure profiles. The viability of long, high bootstrap current fraction operations has been established for ELMing H mode plasmas with toroidal beta values in excess of 15% and sustained for several current relaxation times. Improvements in wall conditioning and fueling are likely contributing to a reduction in H mode power thresholds. Electron thermal conduction is the dominant thermal loss channel in auxiliary heated plasmas examined thus far. HHFW effectively heats electrons, and its acceleration of fast beam ions has been observed. Evidence for HHFW current drive is obtained by comparison of the loop voltage evolution in plasmas with matched density and temperature profiles but varying phases of launched HHFW waves. Studies of emissions from electron Bernstein waves indicate a density scale length dependence of their transmission across the upper hybrid resonance near the plasma edge that is consistent with theoretical predictions. A peak heat flux to the divertor targets of 10 MW m{sup -2} has been measured in the H mode, with large asymmetries being observed in the power deposition between the inner and outer strike points. Non-inductive plasma startup studies have focused on coaxial helicity injection. With this technique, toroidal currents up to 400 kA have been driven, and studies to assess flux closure and coupling to other current drive techniques have begun.

Research on the spherical torus (or spherical tokamak) (ST) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect ratio devices, such as the conventional tokamak. The ST experiments are being conducted in various US research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium sized ST research facilities: PEGASUS at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta ({beta}), non-inductive sustainment, Ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values {beta}{sub T} of up to 35% with a near unity central {beta}{sub T} have been obtained. NSTX will be exploring advanced regimes where {beta}{sub T} up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for non-inductive sustainment in NSTX is the high beta poloidal regime, where discharges with a high non-inductive fraction ({approx}60% bootstrap current+NBI current drive) were sustained over the resistive skin time. Research on radio-frequency (RF) based heating and current drive utilizing high harmonic fast wave and electron Bernstein wave is also pursued on NSTX, PEGASUS, and CDX-U. For non-inductive start-up, the coaxial helicity injection, developed in HIT/HIT-II, has been adopted on NSTX to test the method up to I{sub p} {approx} 500 kA. In parallel, start-up using a RF current drive and only external poloidal field coils are being developed on NSTX. The area of power and particle handling is expected to be challenging because of the higher power density expected in the ST relative to that in conventional aspect-ratio tokamaks. Due to its promise for power and particle handling, liquid lithium is being studied in CDX-U as a potential plasma-facing surface for a fusion reactor.

As part of a study of carbon-tritium co-deposition, we carried out an experiment on DIII-D involving a toroidally symmetric injection of {sup 13}CH{sub 4} at the top of a LSN discharge. A Monte Carlo code, DIVIMP-HC, which includes molecular breakup of hydrocarbons, was used to model the region near the puff. The interpretive analysis indicates a parallel flow in the SOL of M {parallel} {approx} 0.4 directed toward the inner divertor. The CH{sub 4} is ionized in the periphery of the SOL and so the particle confinement time, T{sub C}, is not high, only {approx} 5 ms, and about 4X lower than if the CH{sub 4} were ionized at the separatrix. For such a wall injection location, however, approximately 60-75% of the CH{sub 4} gets ionized to C{sup +}, C{sup 2+}, etc., and is efficiently transported along the SOL to the inner divertor, trapping hydrogen by co-deposition there.

The DIII-D research program is developing the scientific basis for advanced tokamak (AT) modes of operation in order to enhance the attractiveness of the tokamak as an energy producing system. Since the last international atomic energy agency (IAEA) meeting, we have made significant progress in developing the building blocks needed for AT operation: (1) we have doubled the magnetohydrodynamic (MHD) stable tokamak operating space through rotational stabilization of the resistive wall mode; (2) using this rotational stabilization, we have achieved {beta}{sub N}H{sub 89} {ge} 10 for 4{tau}{sub E} limited by the neoclassical tearing mode (NTM); (3) using real-time feedback of the electron cyclotron current drive (ECCD) location, we have stabilized the (m, n) = (3, 2) NTM and then increased {beta}{sub T} by 60%; (4) we have produced ECCD stabilization of the (2, 1) NTM in initial experiments; (5) we have made the first integrated AT demonstration discharges with current profile control using ECCD; (6) ECCD and electron cyclotron heating (ECH) have been used to control the pressure profile in high performance plasmas; and (7) we have demonstrated stationary tokamak operation for 6.5 s (36{tau}{sub E}) at the same fusion gain parameter of {beta}{sub N}H{sub 89}/q{sub 95}{sup 2} {approx_equal} as ITER but at much higher q{sub 95} = 4.2. We have developed general improvements applicable to conventional and AT operating modes: (1) we have an existence proof of a mode of tokamak operation, quiescent H-mode, which has no pulsed, edge localized modes (ELM) heat load to the divertor and which can run for long periods of time (3.8 s or 25{tau}{sub E}) with constant density and constant radiated power; (2) we have demonstrated real-time disruption detection and mitigation for vertical disruption events using high pressure gas jet injection of noble gases; (3) we have found that the heat and particle fluxes to the inner strike points of balanced, double-null divertors are much smaller than to the outer strike points. We have made detailed investigations of the edge pedestal and scrape-off layer (SOL): (1) atomic physics and plasma physics both play significant roles in setting the width of the edge density barrier in H-mode; (2) ELM heat flux conducted to the divertor decreases as density increases; (3) intermittent, bursty transport contributes to cross field particle transport in the SOL of H-mode and, especially, L-mode plasmas.

We give the results of a study using Monte Carlo ion interaction codes to simulate and optimize elastic recoil detection analysis for {sup 3}He buildup in tritide films. Two different codes were used. The primary tool was MCERD, written especially for simulating ion beam analysis using optimizations and enhancements for greatly increasing the probabilities for the creation and the detection of recoil atoms. MPTRIM, an implementation of the TRIMRC code for a massively parallel computer, was also used for comparison and for determination of absolute yield. This study was undertaken because of a need for high-resolution depth profiling of 3He and near-surface light impurities (e.g. oxygen) in metal hydride films containing tritium.

Cross-sections for the elastic recoil of hydrogen isotopes, including tritium, have been measured for {sup 4}He{sup 2+} ions in the energy range of 9.0-11.6 MeV. These cross-sections have been measured at a scattering angle of 30{sup o} in the laboratory frame. Cross-sections were measured by allowing a {sup 4}He{sup 2+} beam to fall incident on solid targets of ErH{sub 2}, ErD{sub 2} and ErT{sub 2}, each of 500 nm nominal thickness and known areal densities of H, D, T and Er. The uncertainty in each cross-section is estimated to be {+-}3.2%.

Hydrogen isotope thin film standards have been manufactured at Sandia National Laboratories for use by the materials characterization community. Several considerations were taken into account during the manufacture of the ErHD standards, with accuracy and stability being the most important. The standards were fabricated by e-beam deposition of Er onto a Mo substrate and the film stoichiometrically loaded with hydrogen and deuterium. To determine the loading accuracy of the standards two random samples were measured by thermal desorption mass spectrometry and atomic absorption spectrometry techniques with a stated combined accuracy of {approx}1.6% (1{sigma}). All the standards were then measured by high energy RBS/ERD and RBS/NRA with the accuracy of the techniques {approx}5% (1{sigma}). The standards were then distributed to the IBA materials characterization community for analysis. This paper will discuss the suitability of the standards for use by the IBA community and compare measurement results to highlight the accuracy of the techniques used.

Mechanisms of H release from Mg-doped, p-type GaN were investigated in vacuum, in N{sub 2} and O{sub 2} gases, and in electron-cyclotron-resonance N{sub 2} plasmas. Replacing grown-in protium with deuterium (D) and employing sensitive nuclear-reaction analysis allowed the retained concentration to be followed quantitatively over two decades during isothermal heating, illuminating the kinetics of controlling processes. Oxidation attending the O{sub 2} exposures was monitored through nuclear-reaction analysis of {sup 18}O. N{sub 2} gas at atmospheric pressure increases the rate of D release appreciably relative to vacuum. The acceleration produced by O{sub 2} gas is much greater, but is diminished in later stages of the release by oxidation. The N{sub 2} plasma employed in these studies had no resolvable effect. We argue that surface desorption is rate controlling in the D release, and that it occurs by D-D recombination and the formation of N-D and O-D species. Our results are quantitatively consistent with a theoretical model wherein the bulk solution is in equilibrium with surface states from which desorption occurs by processes that are both first and second order in surface coverage.

Proposed next-step devices for development of fusion energy present a major increase in the energy content and duration of plasmas far beyond those encountered in existing machines. This increases the importance of controlling interactions between the fusion plasma and first-wall materials. These interactions change the wall materials and strongly affect the core plasma conditions. Two critical processes are the erosion of materials by the plasma, and the redeposition of eroded material along with hydrogen isotopes from the plasma. These impact reactor design through the lifetime of plasma-facing components and the inventory of tritium retained inside the vessel. Ion beam analysis has been widely used to investigate these complex plasma-material interactions in most of the large fusion plasma experiments. The design and choice of plasma-facing materials for next-step machines rely on knowledge obtained from these studies. This paper reviews the use of ion beam analysis for fusion energy research, and shows how these studies have helped to guide the design and selection of materials for a next-step machine.

This report documents the strategies for verification and validation of the codes LSP and ICARUS used for simulating the operation of the neutron tubes used in all modern nuclear weapons. The codes will be used to assist in the design of next generation neutron generators and help resolve manufacturing issues for current and future production of neutron devices. Customers for the software are identified, tube phenomena are identified and ranked, software quality strategies are given, and the validation plan is set forth.

This LDRD is aimed to place Sandia at the forefront of GaN-based technologies. Two important themes of this LDRD are: (1) The demonstration of novel GaN-based devices which have not yet been much explored and yet are coherent with Sandia's and DOE's mission objectives. UV optoelectronic and piezoelectric devices are just two examples. (2) To demonstrate front-end monolithic integration of GaN with Si-based microelectronics. Key issues pertinent to the successful completion of this LDRD have been identified to be (1) The growth and defect control of AlGaN and GaN, and (2) strain relief during/after the heteroepitaxy of GaN on Si and the separation/transfer of GaN layers to different wafer templates.

We have used ion channeling to examine the lattice configuration of deuterium in Mg doped GaN grown by MOCVD. The deuterium is introduced by exposure to gas phase or ECR plasmas. A density functional approach including lattice relaxation, was used to calculate total energies for various locations and charge states of hydrogen in the wurtzite Mg doped GaN lattice. Results of channeling measurements are compared with channeling simulations for hydrogen at lattice locations predicted by density functional theory. © 2000 Materials Research Society.

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The diffusion, uptake, and release of H in p-type GaN are modeled employing state energies from density-function theory and compared with measurements of deuterium uptake and release using nuclear-reaction analysis. Good semiquantitative agreement is found when account is taken of a surface permeation barrier.

Plasma facing components in TFTR contain an important record of plasma wall interactions in reactor grade DT plasmas. Tiles, flakes, wall coupons 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 0.5 g and is consistent with the in-vessel tritium inventory derived from the difference between tritium fueling and tritium exhaust. Relatively high concentrations of tritium were found at the top and bottom of the bumper limiter, as predicted by earlier BBQ modeling. The distribution of tritium on the limiter and vessel wall showed complex patterns of co-deposition.

The ability to withstand disruptions makes carbon-based materials attractive for use as plasma-facing components in divertors. However, such materials suffer high erosion rates during attached plasma operation which, in high power long pulse machines, would give short component lifetimes and high tritium inventories. The authors present results from recent experiments in DIII-D, in which the Divertor Materials Evaluation System (DiMES) was used to examine erosion and deposition during short exposures to well defined plasma conditions. These studies show that during operation with detached plasmas, produced by gas injection, net erosion is suppressed everywhere in the divertor. Net deposition of carbon with deuterium was observed at the inner and outer strikepoints and in the private-flux region between strikepoints. For these low temperature plasmas (T{sub e} < 2eV), physical sputtering is eliminated. These results show that with detached plasmas, the location of carbon net erosion and the carbon impurity source, probably lies outside the divertor. Physical or chemical sputtering by charge-exchange neutrals or ions in the main plasma chamber is a probable source of carbon under these plasma conditions.

The National Spherical Torus Experiment (NSTX) started plasma operations in February 1999, and promptly achieved high current, inner wall limited, double null, and single null plasma discharges, initial Coaxial Helicity Injection, and High Harmonic Fast Wave results. NSTX is designed to study the physics of Spherical Tori (ST) in a device that can produce non-inductively sustained high-{beta} discharges in the 1 MA regime and to explore approaches toward a small, economical high power density ST reactor core. As expected, discharge reproducibility and performance were strongly affected by wall conditions. In this paper, the authors describe the internal geometry, and initial plasma discharge, impurity control, wall conditioning, erosion, and deposition results.

Formation energies and vibrational frequencies for H in wurtzite GaN were calculated from density functional theory and used to predict equilibrium state occupancies and solid solubilities for p-type, intrinsic, and n-type material. The solubility of deuterium (D) was measured at 600--800 C as a function of D{sub 2} pressure and doping and compared with theory. Agreement was obtained by reducing the H formation energies 0.2 eV from ab-initio theoretical values. The predicted stretch-mode frequency for H bound to the Mg acceptor lies 5% above an observed infrared absorption attributed to this complex. It is concluded that currently recognized H states and physical processes account for the equilibrium behavior of H examined in this work.

The behavior of H in p-GaN(Mg) at temperatures >400 C is modeled by using energies and vibrational frequencies from density-functional theory to parameterize transport and reaction equations. Predictions agree semiquantitatively with experiment for the solubility, uptake, and release of the H when account is taken of a surface barrier. Hydrogen is introduced into GaN during growth by metal-organic chemical vapor deposition (MOCVD) and subsequent device processing. This impurity affects electrical properties substantially, notably in p-type GaN doped with Mg where it reduces the effective acceptor concentration. Application of density-functional theory to the zincblende and wurtzite forms of GaN has indicated that dissociated H in interstitial solution assumes positive, neutral, and negative charge states. The neutral species is found to be less stable than one or the other of the charged states for all Fermi energies. Hydrogen is predicted to form a bound neutral complex with Mg, and a local vibrational mode ascribed to this complex has been observed. The authors are developing a unified mathematical description of the diffusion, reactions, uptake, and release of H in GaN at the elevated temperatures of growth and processing. Their treatment is based on zero-temperature energies from density functional theory. One objective is to assess the consistency of theory with experiment at a more quantitative level than previously. A further goal is prediction of H behavior pertinent to device processing. Herein is discussed aspects relating to p-type GaN(Mg).

A technique has been developed for producing calibrated metal hydride films for use in the measurement of high-energy (5--15 MeV) particle reaction cross sections for hydrogen and helium isotopes on hydrogen isotopes. Absolute concentrations of various hydrogen isotopes in the film is expected to be determined to better than {+-}2% leading to the capacity of accurately measuring various reaction cross sections. Hydrogen isotope concentrations from near 100% to 5% can be made accurately and reproducibly. This is accomplished with the use of high accuracy pressure measurements coupled with high accuracy mass spectrometric measurements of each constituent partial pressure of the gas mixture during loading of the metal occluder films. Various techniques are used to verify the amount of metal present as well as the amount of hydrogen isotopes; high energy ion scattering analysis, PV measurements before, during and after loading, and thermal desorption/mass spectrometry measurements. The most appropriate metal to use for the occluder film appears to be titanium but other occluder metals are also being considered. Calibrated gas ratio samples, previously prepared, are used for the loading gas. Deviations from this calibrated gas ratio are measured using mass spectrometry during and after the loading process thereby determining the loading of the various hydrogen isotopes. These techniques are discussed and pertinent issues presented.

We estimate the total in-vessel deuterium retention in Alcator C-Mod from a run campaign of about 1090 plasmas. The estimate is based on measurements of deuterium retained on 22 molybdenum tiles from the inner wall and divertor. The areal density of deuterium on the tiles was measured by nuclear reaction analysis. From these data, the in-vessel deuterium inventory is estimated to be about 0.1 gram, assuming the deuterium coverage is toroidally symmetric. Most of the retained deuterium is on the walls of the main plasma chamber, only about 2.5% of the deuterium is in the divertor. The D coverage is consistent with a layer saturated by implantation with ions and charge-exchange neutrals from the plasma. This contrasts with tokamaks with carbon plasma-facing components (PFC's) where long-term retention of tritium and deuterium is large and mainly in the divertor due to codeposition with carbon eroded by the plasma. The low deuterium retention in the C-Mod divertor is mainly due to the absence of carbon PFC's in C-Mod and the low erosion rate of Mo.

This section reviews physical processes involved in the implantation of energetic hydrogen into plasma facing materials and its subsequent diffusion, release, or immobilization by trapping or precipitation within the material. These topics have also been discussed in previous reviews. The term hydrogen or H is used here generically to refer to protium, deuterium or tritium.

Tungsten is a candidate material for the International Thermonuclear Experimental Reactor (ITER) as well as other future magnetic fusion energy devices. Tungsten is well suited for certain fusion applications in that it has a high threshold for sputtering as well as a very high melting point. As with all materials to be used on the inside of a tokamak or similar device, there is a need to know the behavior of hydrogen isotopes embedded in the material. With this need in mind, the Tritium Plasma Experiment (TPE) has been used to examine the retention of tritium in tungsten exposed to high fluxes of 100 eV tritons. Both tungsten and tungsten containing 1% lanthanum oxide were used in these experiments. Measurements were performed over the temperature range of 423-973 K. After exposure to the tritium the samples were transferred to an outgassing system containing an ionization chamber for detection of the tritium. The samples were outgassed using linear ramps from room temperature up to 1473 K. Unlike most other materials exposed to energetic tritium, the tritium retention in tungsten reaches a maximum at intermediate with low retention at both high and low temperatures. For the very high triton fluences used (>1025 T/m2), the fractional retention of the tritium was below 0.02% of the incident particles. This report presents not only the results of the tritium retention, but also includes the modeling of the results and the implication for ITER and other future fusion devices where tungsten is used.

Experiments were performed on Alcator C-Mod with electron cyclotron resonance (ECR) plasmas to help determine their applicability to a fusion reactor. Strong radial inhomogeneity of the plasma density was measured, decreasing by a factor of ten a few centimeters inside the resonance location, but remaining approximately constant (ne≈1016 m-3) outside the resonance location. Electron temperature remained mostly constant outside the resonance location, Te≈10 eV; ion temperature increased outside the resonance location from Ti≈2 eV to 10 eV. Toroidal asymmetries in ion saturation current density were observed, indicating local toroidal plasma flow. The ECR plasma was used to remove a diamond-like carbon coating from a stainless-steel sample. Removal rates peaked at 4.2±0.4 nm/h with the sample a few centimeters outside the resonance location. Removal rates decreased inside and further outside the resonance location. The plasma did not remove the carbon from the sample uniformly, possibly due to plasma flow. Yields were calculated (Y≈10-3) to be lower than other published results for chemical sputtering of deuterium ions on carbon, possibly due to toroidally asymmetric plasma conditions.

The International Thermonuclear Experimental Reactor (ITER) is envisioned to be the next major step in the world`s fusion program from the present generation of tokamaks and is designed to study fusion plasmas with a reactor relevant range of plasma parameters. During normal operation, it is expected that a fraction of the unburned tritium, that is used to routinely fuel the discharge, will be retained together with deuterium on the surfaces and in the bulk of the plasma facing materials (PFMs) surrounding the core and divertor plasma. The understanding of he basic retention mechanisms (physical and chemical) involved and their dependence upon plasma parameters and other relevant operation conditions is necessary for the accurate prediction of the amount of tritium retained at any given time in the ITER torus. Accurate estimates are essential to assess the radiological hazards associated with routine operation and with potential accident scenarios which may lead to mobilization of tritium that is not tenaciously held. Estimates are needed to establish the detritiation requirements for coolant water, to determine the plasma fueling and tritium supply requirements, and to establish the needed frequency and the procedures for tritium recovery and clean-up. The organization of this paper is as follows. Section 2 provides an overview of the design and operating conditions of the main components which define the plasma boundary of ITER. Section 3 reviews the erosion database and the results of recent relevant experiments conducted both in laboratory facilities and in tokamaks. These data provide the experimental basis and serve as an important benchmark for both model development (discussed in Section 4) and calculations (discussed in Section 5) that are required to predict tritium inventory build-up in ITER. Section 6 emphasizes the need to develop and test methods to remove the tritium from the codeposited C-based films and reviews the status and the prospects of the most attractive techniques. Section 7 identifies the unresolved issues and provides some recommendations on potential R and D avenues for their resolution. Finally, a summary is provided in Section 8.

Net erosion rates of carbon target plates have been measured in situ for the DIII-D lower divertor. The principal method of obtaining this data is the DiMES sample probe. Recent experiments have focused on erosion at the outer strike-point of two divertor plasma conditions: (1) attached (Te > 40 eV) ELMing plasmas and (2) detached (Te < 2 eV) ELMing plasmas. The erosion rates for the attached cases are > 10 cm/year, even with incident heat flux < 1 MW/m{sup 2}. In this case, measurements and modeling agree for both gross and net carbon erosion, showing the near-surface transport and redeposition of the carbon is well understood and that effective sputtering yields are > 10%. In ELM-free discharges, this erosion rate can account for the rate of carbon accumulation in the core plasma. Divertor plasma detachment eliminates physical sputtering, while spectroscopically measured chemical erosion yields are also found to be low (Y(C/D{sup +}) {le} 2.0 {times} 10{sup {minus}3}). This leads to suppression of net erosion at the outer strike-point, which becomes a region of net redeposition ({approximately} 4 cm/year). The private flux wall is measured to be a region of net redeposition with dense, high neutral pressure, attached divertor plasmas. Leading edges intercepting parallel heat flux ({approximately} 50 MW/m{sup 2}) have very high net erosion rates ({approximately} 10 {micro}m/s) at the OSP of an attached plasma. Leading edge erosion, and subsequent carbon redeposition, caused by tile gaps can account for half of the deuterium codeposition in the DIII-D divertor.

The exposure of ATG graphite sample to DIII-D divertor plasma was provided by the DiMES (Divertor Material Evaluation System) mechanism. The graphite sample arranged to receive the parallel heat flux on a small region of the surface was exposed to 600ms of outer strike point plasma. The sample was constructed to collect the eroded material directed downward into a trapping zone onto s Si disk collector. The average heat flux onto the graphite sample during the exposure was about 200W/cm{sup 2}, and the parallel heat flux was about 10 KW/cm{sup 2}. After the exposure the graphite sample and Si collector disk were analyzed using SEM, NRA, RBS, Auger spectroscopy. IR and Raman spectroscopy. The thermal desorption was studied also. The deposited coating on graphite sample is amorphous carbon layer. Just upstream of the high heat flux zone the redeposition layer has a globular structure. The deposition layer on Si disk is composed also from carbon but has a diamond-like structure. The areal density of C and D in the deposited layer on Si disk varied in poloidal and toroidal directions. The maximum D/C areal density ratio is about 0.23, maximum carbon density is about 3.8 {times} 10{sup 18}cm{sup {minus}2}, maximum D area density is about 3 {times} 10{sup 17}cm{sup 2}. The thermal desorption spectrum had a peak at 1,250K.

The mission of the Divertor Materials Evaluation System (DiMES) in DIII-D is to establish an integrated data base from measurements in the divertor of a tokamak in order to address some of the ITER and fusion power reactor plasma material interaction issues. Carbon and metal coatings of Be, W, V, and Mo were exposed to the steady-state outer strike point on DIII-D for 4--18 s. These short exposure times ensure controlled exposure conditions, and the extensive arrays of DIII-D divertor diagnostics provide a well-characterized plasma for modeling efforts. Post-exposure analysis provides a direct measure of surface material erosion rates and the amount of retained deuterium. For carbon, these results match closely with the results of accumulated carbon deposition and erosion, and the corresponding deuterium retention of long term exposure tiles in DIII-D. Under the carbon-contaminated background plasma of DIII-D, metal coatings of Be, V, Mo, and W were exposed to the steady-state outer strike point under ELMing and ELM-free H-mode discharges. The rate of material erosion and deuterium retention were measured. As expected, W shows the lowest erosion rate at 0.1 mm/s and the lowest deuterium uptake of 2 {times} 10{sup 20}/m{sup 2}.

The results of the investigation of retention and thermal desorption of hydrogen isotopes of B{sub 4}C coated RGT (a recrystallized graphite with high thermal conductivity, 600 W/mK) after the exposure to high heat flux in the divertor strike point region of DIII-D using the DiMES sample exchange system are reported. It is shown that the material is very promising for plasma facing elements of tokamaks.