Search

The Idaho National Laboratory (INL) Fusion Safety Program (FSP) is the Office of Fusion Energy Sciences’ (FES) lead laboratory for Magnetic Fusion Energy (MFE) Safety. Our mission is to assist the US and international fusion communities in developing the inherent safety and environmental potential of fusion power by: 1) Developing fusion licensing data and analysis tools, 2) Participating in national and international collaborations and design studies, and 3) Assisting the US and international fusion community in licensing activities and guidance in operational safety. To achieve our mission, the FSP maintains core competencies in the several areas, two of which are: fusion safety code development and tritium retention and permeation in fusion plasma facing component (PFC) and blanket materials. This article details recent accomplishments of our program in these areas and future directions for fusion safety research and development at INL.

Detailed calculations using the multi-configuration Dirac–Fock (MCDF) method are carried out for the lowest 64 fine-structure levels of the 3s23p2, 3s23p3d, 3s3p3, 3s3p23d, 3s23d2, and 3p4 configurations in Si-like ions of La XLIII, Er LIV, Tm LV, and Yb LVI. Energies, oscillator strengths, wavelengths, line strengths, and radiative electric dipole transition rates are given for all ions. A parallel calculation using the many-body perturbation theory (MBPT) method is also carried out to assess the present energy levels accuracy. Comparisons are performed between these two sets of energy levels, as well as with other available results, showing that they are in good agreement with each other within 0.5%. These high accuracy results can be used to the modeling and the interpretation of astrophysical objects and fusion plasmas.

Baffle is a critical technique to suppress stray light. In order to reduce stray light of EAST Thomson scattering (TS) diagnostic system, a 60° cone angle double edged baffle array was installed for the system. Recently the 45° cone angle corrugation baffle was designed based on double edged baffle and simulation results indicate that the new design can reject 33.3% more stray light than the old one. Based on the mechanical structure and surface property of EAST TS diagnostic system, a stray light analysis model has been built, and using the model, stray light of the system was simulated with the two types of baffles added to the system respectively. Results show that the installation of 60° cone angle double edged baffle array in input tube can cut 92.1% of the stray light to the system, while using 45° cone angle corrugation baffle array in input tube and three 45° cone angle corrugation baffles in exit tube, additional 74.1% of the stray light will be removed.

Since impurities determine the plasma performance in magnetically-confined fusion plasmas, analyzing their behavior and controlling their confinement is significantly important for a stable and steady-state operation of fusion devices. A comparison of the impurity transport in the core region of large stellarators, such as Wendelstein 7-X and Large Helical Device (LHD) is quite useful to gain a better understanding of the impurity transport in reactor-relevant stellarator plasmas. Therefore, a Tracer-Encapsulated Solid Pellet (TESPEL) injection system has been newly installed for the operational phase OP1.2b of W7-X. The injector was designed and manufactured by NIFS to achieve a similar TESPEL deposition location as in LHD. Technical guidelines and functional specifications, addressing the specific conditions at W7-X had been implemented in close collaboration between NIFS and IPP. This resulted in a compact 3-stage differential pumping system considering spatial restrictions due to existing and future diagnostics, material requirements and limitations due to the magnetic stray field in the vicinity of the turbo pumps. This contribution reports on the design and the integration of the new TESPEL injection system to W7-X and first achievements during the commissioning of the system.

A modest approach to develop the ISPC enabling tool for fusion plasma chamber systems has been achieved. This high performance computing simulation addresses 3D physical phenomena in a complex and heterogeneous virtual fusion plasma chamber system and opens a new way for how one, such as DEMO/FNSF, ought to be designed and modeled. In the current approach, complex FNST scenarios were simulated and modeled through a community-built reflective middleware for simulation integrations involving multiple simulators. Example advancements are presented while issues and ideas are discussed to further expand the development of such a tool.

Aneutronic/low radioactive fuel is the way to clean and cheap energy of the future. An alternative scheme using compact toroids – field-reversed configuration or spheromak – may be applied for the reactor based on any magnetic confinement system. Even more, any fusion concept, including hybrid magneto-inertial fusion might use advantages of D–3He fuel. Advanced fuel, including helium-3 – based fusion plasma and alternative systems are reviewed. Different schemes of reactors, near-term technology and non-electric applications are discussed.

In this work we present basic principle of treecode (TC) method, its advantages and drawbacks in its applications for laboratory and fusion plasmas. Particular task done in this work was to create a small but efficient PIC simulation program in order to obtain more data than currently possible with other such programs available today, and to compare results of our program with results obtained with other methods and packages. The program is intended to be used in plasma-wall investigations of laboratory, technology oriented and fusion plasmas.

The operation of ITER requires high-quality estimates of the plasma electron density over multiple regions in the plasma for plasma evaluation, plasma control and machine protection purposes. Although the density regimes of ITER are not very different from those of existing tokamaks (1018–1021 m−3), the severe conditions of the fusion plasma environment present particular challenges to implementing these density diagnostics. In this paper we present an overview of the array of ITER electron density diagnostics designed to measure over the entire ITER domain: plasma core, pedestal, edge, scrape-off layer and divertor. It will focus on the challenges faced in making these measurements, and the technical solutions of the current designs.

Large edge-localized mode (ELM) control techniques must be developed to help ensure the success of burning and ignited fusion plasma devices such as tokamaks and stellarators. In full performance ITER tokamak discharges, with QDT = 10, the energy released by a single ELM could reach ∼30 MJ which is expected to result in an energy density of 10–15 MJ/m2on the divertor targets. This will exceed the estimated divertor ablation limit by a factor of 20–30. A worldwide research program is underway to develop various types of ELM control techniques in preparation for ITER H-mode plasma operations. An overview of the ELM control techniques currently being developed is discussed along with the requirements for applying these techniques to plasmas in ITER. Particular emphasis is given to the primary approaches, pellet pacing and resonant magnetic perturbation fields, currently being considered for ITER.

The presence of dust particles in a fusion plasma is recognized as a serious issue for safe and efficient operation of the ITER tokamak. The paper presents an in situ laser assisted method for characterization of dust from thermal emission of the particles. The method was developed in the TEXTOR tokamak with the use of Thomson scattering (TS). The diagnostic is capable to detect single particles and measure the dust density profile along the laser probing axis, velocity distribution of dust particles along this axis as well as surface temperature and size of the detected particles.

Turbulence is one of the key problems of classical physics, and it has been the object of intense research in the last decades in a large spectrum of problems involving fluids, plasmas, and waves. In order to review some advances in theoretical and experimental investigations on turbulence a mini-symposium on this subject was organized in the Dynamics Days South America 2010 Conference. The main goal of this mini-symposium was to present recent developments in both fundamental aspects and dynamical analysis of turbulence in nonlinear waves and fusion plasmas. In this paper we present a summary of the works presented at this mini-symposium. Among the questions to be addressed were the onset and control of turbulence and spatio-temporal chaos.

The neutron diagnostics in tokamaks like Joint European Torus (JET) are essential in estimating fusion power. The neutron activation method, supported by neutron transport calculations, is particularly useful for the evaluation of the total neutron yield from a single plasma discharge. This paper presents the results of activation experiments and calculations carried out for JET plasmas, from the selection of the activation materials to their irradiations in the neutron field of JET discharges. Neutron transport calculations were performed, leading to activation coefficients for new materials. The results of the calculations were used to design new composite samples to obtain information on both the yield and the neutron spectrum. The neutron measurements using these new activation materials were performed during the last JET experimental campaigns. The results are compared with neutron transport calculations. Additionally, application of the cadmium difference method allows revelation of the part of thermal neutrons near the tokamak first wall. The advantages of new activation materials and benchmarking the activation method against neutron transport calculations are also discussed.

For contributing to the ITER project and promoting a possible early realization of DEMO, since 2007 the International Fusion Energy Research Centre (IFERC) shall perform the activities on (1) DEMO Design and R&D Coordination, (2) Computational Simulation Centre, and (3) ITER Remote Experimentation Centre in the framework under the agreement between the Government of Japan and the European Atomic Energy Community for the joint implementation of the Broader Approach Activities in the field of fusion energy. The DEMO Design activity aims at establishing a common basis for DEMO Design, including design features of DEMO, possible common concepts of DEMO, a roadmap for DEMO, and so on. Based on the common interest of EU and Japan toward DEMO, the DEMO R&D activities have been planned and carried out for the areas which are related to DEMO blanket development: SiC/SiC composites, tritium technology, materials engineering, advanced neutron multiplier, and advanced tritium breeders for DEMO blanket. In the activity of the Computational Simulation Centre, the objective is to provide and exploit a supercomputer for large scale simulation activities to analyse experimental data on fusion plasmas, prepare scenarios for ITER operation, predict the performance of the ITER facilities and contribute to the DEMO Design.

An optical flow method is applied to the study of several fusion plasma relevant issues, including plasma wall interactions. A multi-resolution coarse-to-fine procedure is used in order to cope with large displacements of objects between consecutive frames, characteristic of plasma images captured by JET fast visible camera. Occlusion modeling is also implemented. The method is able to provide good results for JET fast visible camera images which can be affected by saturation, discontinuous movement, reshaping of image objects, low gray-level in-depth resolution. Significant experimental cases concerning pellet injection, plasma filaments and MARFEs are analysed. The method is able to provide the real velocity for objects moving close to structures.

A time-of-flight (TOF) spectrometer for measurement of the 2.5-MeV neutron emission from fusion plasmas has been developed and put into use at the JET tokamak. It has been optimized for operation at high rates (TOFOR) for the purpose of performing advanced neutron emission spectroscopy (NES) diagnosis of deuterium plasmas with a focus on the fuel ion motional states for different auxiliary heating scenarios. This requires operation over a large dynamic range, including high rates of > 100 kHz with a maximum value of 0.5 MHz for the TOFOR design. This paper describes the design principles and their technical realization. The performance is illustrated with recent neutron TOF spectra recorded for plasmas subjected to different heating scenarios. A true event count rate of 39 kHz has been achieved at about a tenth of the expected neutron yield limit of JET, giving a projected maximum of 400 kHz at peak JET plasma yield. This means that the count rate capability for NES diagnosis of D plasmas has been improved more than an order of magnitude. Another important performance factor is the spectrometer bandwidth, where data have been acquired and analyzed successfully with a response function for neutrons over the energy range 1 to > 5 MeV. The implications of instrumental advancement represented by TOFOR are discussed. (C) 2008 Elsevier B.V. All rights reserved.

Elastic recoil detection analysis using heavy ions with a scanning nuclear microprobe was applied to determine the content of hydrogen isotopes in carbon material facing fusion plasma in the JET fu ...

We have investigated plasma turbulence at the edge of a tokamak plasma using data from electrostatic potential fluctuations measured in the Brazilian tokamak TCABR. Recurrence quantification analysis has been used to provide diagnostics of the deterministic content of the series. We have focused our analysis on the radial dependence of potential fluctuations and their characterization by recurrence-based diagnostics. Our main result is that the deterministic content of the experimental signals is most pronounced at the external part of the plasma column just before the plasma radius. Since the chaoticity of the signals follows the same trend, we have concluded that the electrostatic plasma turbulence at the tokamak plasma edge can be partially explained by means of a deterministic nonlinear system.

The decreasing availability of energy and the concern about climate change necessitate the development of novel sustainable energy sources. Fusion energy is such a source. The ultimate potential of fusion energy is very high and badly needed. A major step forward in the development of fusion energy is the decision to construct the experimental test reactor ITER. ITER will stimulate research in many areas of science. This article discusses research opportunities in the context of plasma–surface interaction. The fusion plasma with a typical temperature of 10 keV has to be brought into contact with a physical wall in order to remove the helium produced and drain the excess energy in the fusion plasma. The fusion plasma is far too hot to be brought into direct contact with a physical wall. It would degrade the wall and the debris from the wall would extinguish the plasma. Therefore, schemes are developed to cool down the plasma. The resulting plasma–surface interaction concerned in ITER is facing several challenges including surface erosion, material redeposition and tritium retention. In this article we introduce how plasma–surface interaction relevant for ITER can be studied in a smaller scale experiment: Magnum-PSI.

To tackle successfully the difficult problem of turbulence and transport in fusion plasmas one requires a multidisciplinary approach which uses advances made in related fields, and to foster support for fusion because of the contributions it has made and will make to the understanding of turbulence. Progress in the theory of fusion plasmas turbulence is also a matter of extending modeling power by new analytical techniques instead of confining the methods to some few classical paradigms. In this article, we will present and discuss some approaches that have been developed in astrophysics, condensed matter physics and field theory and which could be applied to solve some complex issues of plasma turbulence. To cite this article: S. Benkadda, C. R. Physique 7 (2006).

The present level of understanding of the physics of the formation and sustainment of edge transport barriers in magnetically confined fusion plasmas is presented. The formation of edge transport barriers is studied through evolution of mechanisms which can suppress plasma turbulence and so reduce turbulent driven transport, such as E × B flow shear stabilization of turbulence. Comparisons of theoretical studies with experimental results are described including investigations of zonal flows, which are considered important for saturation and self-regulation of turbulence and turbulence-driven transport. Processes that affect the dynamics and spatial structure of the edge barrier are described with emphasis on the width of the transport barrier. To cite this article: P. Gohil, C. R. Physique 7 (2006).

One powerful way to investigate transport is to study the dynamic plasma response to externally applied small perturbations. This article describes the experimental techniques and basic theory underlying the perturbative approach. It then reviews the most recent results on electron heat transport and discusses their interpretation based on key concepts from electrostatic turbulence theory. Finally it presents some hints for future extensions of the work. To cite this article: P. Mantica, F. Ryter, C. R. Physique 7 (2006).

This introduction presents the main instabilities responsible for turbulence in tokamak plasmas, and the prominent features of the resulting transport. The usual techniques to construct reduced transport models are described. These models can be tested by analysing steady state and transient regimes. Another way to test the theory is to use a similarity principle, similar to the one used in fluid mechanics. Finally, the physics involved in the formation and sustainment of transport barriers is presented. To cite this article: X. Garbet, C. R. Physique 7 (2006).

A view of the mechanisms underlying particle and energy transport in fusion plasmas devices is presented. Heat and energy transport in fusion plasmas is partly due to turbulent processes having an intermittent character and presenting some universal features. The improved understanding of intermittent turbulent transport is changing the standard picture of transport, based on local diffusive mechanisms. The development of new diagnostic methods and the improvement of modeling tools are spawning a new era in fusion plasma research, in which the non-equilibrium aspects of the systems under study is carefully taken into account in order to unravel the global picture connecting transport, gradients and flows. To cite this article: C. Hidalgo et al., C. R. Physique 7 (2006).

A reduction of turbulent transport is an important task on the way of an economic fusion power production. The knowledge gathered on the driving forces and the dynamics of plasma turbulence helps to identify different techniques, which have been successfully applied in fusion experiments to improve confinement. The techniques can be divided into two classes, those related to sheared plasma flows and the others related to an optimisation of the magnetic configuration. The paper gives an overview of the different approaches and aims at an intuitive understanding of the mechanisms leading to reduced transport. To cite this article: U. Stroth, M. Ramisch, C. R. Physique 7 (2006).

We present impact-parameter classical trajectory Monte-Carlo and molecular close-coupling calculations for total and partial cross sections for Ne10+, Ar18 + H(1s) collisions, which have recently became of interest in fusion plasma research.

In Langmuir-probe diagnostics of fusion plasmas, collisions between charged and neutral particles are often neglected. We present a simple analytical theory of the plane probe including these collisions. Our analytical results are confirmed by high-resolution particle-in-cell (PIC) simulations covering the whole range of the Volt–Ampere characteristic.

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.

Dust is produced in fusion devices by energetic plasma–surface interactions. As the amount of dust increases, potential safety and operational concerns arise. The dust may contain tritium, may be radioactive from activation products, and may be chemically reactive and/or toxic. Possible accidents in large fusion reactors could mobilize the dust and threaten public safety. Dust also poses potential problems to device operation. For example, plasma startup could be impeded, particulate injected from flaking deposits may disrupt the fusion plasma, and tritium retention in dust will affect fuel recovery systems. The current understanding of dust's role in fusion devices is reviewed in this paper by discussing mechanisms of dust production, considering ways dust impacts device safety and operation, and comparing characteristics of dust collected from existing fusion plasma research devices.

The helium cooled pebble bed (HCPB) blanket is one of the two European DEMO blanket concepts proposed for testing in international thermonuclear experimental reactor (ITER). The purpose of the tests is to validate the design principles and the operational feasibility for the demonstration blanket system. This includes the basic support functions like tritium extraction, helium cooling and heat transport, and helium purification. In addition, the basic properties and operating characteristics of the system's materials will be validated. Safety, reliability, maintenance and dismantling will be equally addressed. To assess these qualities, the ITER horizontal ports will be used to provide a relevant fusion plasma and the appropriate nuclear environment. At conclusion of the ITER feat EDA phase (July 2001), a revised design of the HCPB test blanket system have been completed to adapt the previous design (for ITER FDR) to the new operational conditions of ITER and to a new strategy for the blanket testing in this reactor. Design description, performance and safety analyses are presented in this paper.

New injectors for steady state fusion plasma refuelling by solid hydrogen isotope pellets and an injector of impurity pellets encapsulated into hydrogen shells for particle transport diagnostics have been designed at the PELIN Laboratory, a branch of the PELIN, Inc. (Canada), to meet requirements of LHD, TORE SUPRA, and ITER. A review of the injector designs and results obtained to date is presented.

A pulse compression method is proposed to simulate the propagation of a pulse in an inhomogeneous plasma. It allows very fast computations compared to the usual time-dependent code. The characteristics as well as the limitations of this method are discussed. In particular the phase shift of the pulse frequency components is analyzed for different kinds of density profiles. The validity of using this method under typical conditions in fusion plasmas is then discussed.

The paper presents data on radiotechnical characteristics of antennae and reflectors for mm and sub-mm wavelength ranges, made of carbon-graphite materials (CGM). It is shown that the reflectivity value of plane reflectors made of different kinds of CGM depends mostly on the surface roughness being close to that of metallic reflectors, and does not change significantly under neutron irradiation. Characteristics of CGM-made antennae are not much different from those for antennae made of metal. Pyramidal antennae made of CGM maintains its properties under imitated plasma disruption much better than a similar metallic one would.

Oscillator strengths for 2s2p3P-2s3d3D and 2s2p3P-2s3s3S fine structure transitions in beryllium-like ions (Z = 4–36) have been calculated using both the Relativistic Quantum Defect Orbital (RQDO) method, with and without explicit inclusion of core-valence correlation, and the multiconfiguration Dirac-Fock (MCDF) approach. Our f-values, which are of interest in astrophysics and fusion plasma research, are compared with other theoretical and experimental results for the cases for which these are available.

The Engineering Design Activities (EDA) for an International Thermonuclear Experimental Reactor (Iter) were established on 21 July 1992 by an agreement between the European Atomic Energy Community, the Government of Japan, the Government of the Russian Federation and the Government of the USA. The overall programmatic objective is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. The engineering objectives are to establish the design of Iter in such a way that controlled ignition and extended burn of a fusion plasma can be demonstrated, essential reactor technologies can be demonstrated in an integrated system, and integrated tests of high heat flux and nuclear components can be performed. This article outlines the fusion process and the tokamak concept used for Iter and summarises the potential advantages of fusion energy. The objectives of the EDA are presented together with the organisational structure to translate them into a design. Finally the Iter parameters and design status are reported; and a brief overview of time planning and cost is given.

Although a considerable volume of valuable data on the behaviour of tritium in beryllium and carbon-based armours exposed to hydrogenic fusion plasmas has been compiled over the past years both from operation of present-day tokamaks and from laboratory simulations, knowledge is far from complete and tritium inventory predictions for these materials remain highly uncertain. In this paper we elucidate the main mechanisms responsible for tritium trapping and release in next-step D-T tokamaks, as well as the applicability of some of the presently known data bases for design purposes. Owing to their strong anticipated implications on tritium uptake and release, attention is focused mainly on the interaction of tritium with neutron damage induced defects, on tritium codeposition with eroded carbon and on the effects of oxide and surface contaminants. Some preliminary quantitative estimates are presented based on most recent experimental findings and latest modelling developments as well. The influence of important working conditions such as target temperature, loading particle fluxes, erosion and redeposition rates, as well as material characteristics such as the type of morphology of the protection material (i.e. amorphous plasma-sprayed beryllium vs. solid forms), and design dependent parameters are discussed in this paper. Remaining issues which require additional effort are identified.

The concept of the reduction of the helium ash concentration in a fusion plasma, based on selective pumping of helium ions, is discussed. Where this procedure is employed together with the pumping of a divertor, the helium ash concentration may be considerably reduced. Thus, this method might be helpful to sustain the burning plasma state or the ignition condition. The pumping condition required for the selective pumping metal is also discussed.

Electron excitation in H + -He(2 1 S) collisions in discussed. Excitation cross sections for the set of He 3 1 L states are determined in a many-state one-electron close-coupling model in the energy range 6–150 keV. These turn out to deviate by an order of magnitude from the only other existing theory prediction. Excitation to He 2 1 P states is determined for the first time, in a one-center two-electron model. At present there is much interest in a reliable assessment of these processes for fusion plasma heating and diagnostic purposes.

The current limit and the power consumption of a Multiple Electrostatic Quadrupole Array Linear Accelerator (MEQALAC) and a Radio Frequency Quadrupole (RFQ) accelerator are compared theoretically. Design studies are presented for 1 MeV, 24 mA N + accelerators for ion-implantation purposes and for 1.5 MeV, 100 mA D − accelerators. The latter are specified to be used for fusion plasma diagnostics.

Recent measurements of radiative capture reactions of protons and deuterons on light nuclei at low energies are summarized. Applications of these measurements to various problems in astrophysical nucleosynthesis, fusion plasma diagnostics and near-surface quantitative analysis of low mass elements are described. The results of our investigation of the possible enhancement of deuteron-induced neutron-transfer reactions on light nuclei are presented.

Erratum Erratum to ‘‘Analytic cross sections for electron impact collisions with nitrogen molecules’’ [At. Data Nucl. Data Tables 92 (2006) 375–406] Tatsuo Tabata a,b, Toshizo Shirai c,1, Masao Sataka d, Hirotaka Kubo e,∗ a Osaka Prefecture University, 1-1 Gakuen-cho, Naka-ku, Sakai, Osaka 599-8531, Japan b Institute for Data Evaluation and Analysis, 198-51 Kami, Nishi-ku, Sakai, Osaka 593-8311, Japan c Department of Fusion Plasma Research, Japan Atomic Energy Research Institute, Naka-shi, Ibaraki 319-0193, Japan d Tokai Research and Development Center, Nuclear Science Institute, Japan Atomic Energy Agency, 2-4 Shirakata-shirane, Tokai-mura, Ibaraki 319-1195, Japan e Fusion Research and Development Directorate, Japan Atomic Energy Agency, 801 -1 Mukoyama, Naka-shi, Ibaraki 319-0193, Japan

The status of atomic physics research in the United States relevant to magnetically confined fusion plasmas is briefly reviewed. Program support is concentrated on performing benchmark experiments to confirm theory which will be used for a large fraction of the required data base. Recent advances and present needs in electron-ion collisions, low energy collisions in ion sources, and atomic structure are emphasized.

A two photon excitation scheme for detection of neutral hydrogen atoms in a fusion plasma is described. The approach has the advantages of both the Lyman α and Balmer α scattering methods presently being used. Further, it is shown that existing laser technology produces the required laser power.

Atoms constituting the first wall will be eroded off and then re-deposited on other portions of the wall through the plasma-wall interaction. The re-deposited film is considered to incorporate with many gaseous atoms and will evaporate and exfoliate easily during the operation thereafter, resulting in a deleterious contamination of fusion plasma. Here, we studied preliminarily the behavior of sputter-deposited TiC films which incorporated with inert Ar gas, at elevated temperatures. The film deposited with lower bias voltage contained fewer Ar atoms but was easy to evaporate and had weak adhesive strength to its substrate. The film deposited with higher bias voltage resisted to the thermal evaporation and had good adhesive strength but contained more Ar atoms. The incorporated Ar atoms formed blisters and caused the film exfoliation.

A neutron spectrometer based on a double interaction and time-of-flight technique is under development. Preliminary experimental results are presented and a special application of the techniques as a diagnostic instrument for extended samples such as

The internal surfaces of a toroidal magnetically confined fusion reactor control the impurity injection and gas recycling into the fusion plasma. Coating of internal surfaces may provide a desirable and possibly necessary design flexibility for achieving the temperatures, ion densities and containment times necessary for net energy production from fusion reactions. In this paper the reactor environments experienced by various components are reviewed along with possible materials responses. Characteristics of coating-substrate systems that are important to fusion applications are delineated and the present status of coating development for fusion applications is reviewed. Coating development for fusion applications is just beginning and poses a unique and important challenge for materials development.

Trapping and release of keV deuterium in pyrolytic graphite has been the subject of several investigations because of its relevance to the problem of interaction of a hot fusion plasma with the solid wall. In the present work the influence of He-induced damage on implantation profiles for deuterium in pyrolytic graphite is investigated for either a smaller (9 keV D + , 3 keV He + ) or larger He range (1.5 keV D + , 13 keV He + ). During annealing, the implanted D is found to be retained more strongly at depths where He-damage has been introduced. The results show that the trapping sites of deuterium in pyrolytic graphite are at least partly connected with radiation damage.

Much progress has been made in the last two years in the U.S. magnetic fusion technology program. Notable advancements have been made in superconducting magnets, gyrotrons, reduced activation materials, pellet injectors, and safe handling of tritium to name a few. Advancements in fusion plasma conditions and economic, environmentally attractive reactor designs will continue to require significant improvements in the technology of magnets, heating, fueling, materials, and safety. The ITER process will provide a forum and an opportunity to accelerate the development of fusion technology through new international collaborations, and we should take advantage of that opportunity.

The almost simultaneous appearance of calculated and measured dielectronic recombination cross sections for singly and few-times charged target ions has produced an interesting situation. The large disparity between theory and experiment has provoked speculation about a possible “electric field effect”. The implications of this idea are examined, in the context of distorted-wave theory, and using the isolated-resonance-approximation.