Short wavelength interferometer for ITER

Numerical evaluation of experimental models to investigate the dynamic behavior of the ITER tokamak assembly

Over the past five years, four international parties, which include the European Communities, Japan, the Russian Federation and the United States, have been collaborating on the design and development of the International Thermonuclear Experimental Reactor (ITER), the next generation magnetic fusion energy device. During the ITER Engineering Design Activity (EDA), beryllium plasma spray technology was investigated by Los Alamos National Laboratory as a method for fabricating and repairing and the beryllium first wall surface of the ITER tokamak. Significant progress has been made in developing beryllium plasma spraying technology for this application. Information will be presented on the research performed to improve the thermal properties of plasma sprayed beryllium coatings and a method that was developed for cleaning and preparing the surface of beryllium prior to depositing plasma sprayed beryllium coatings. Results of high heat flux testing of the beryllium coatings using electron beam simulated ITER conditions will also be presented.

The International Thermonuclear Experimental Reactor (ITER) vacuum vessel (VV) is one of the critical components of the ITER tokamak fusion reactor. The first sector of the ITER VV was delivered to ITER Organization in 2020, and it is ready to assemble into the tokamak system. After manufacturing the ITER VV, an evaluation should ensure that the components are designed and manufactured to meet the functional requirements, such as vacuum leak tightness and structural integrity. The factory acceptance test (FAT) is essential for confirming acceptance in engineering and manufacturing. This paper introduces the engineering process and technical method of the FAT, which is applied explicitly to the first-of-a-kind ITER VV. We establish a visual inspection, pre-pumping assessment, pressure test, helium (vacuum) leak test, and final dimensional inspection for the FAT. The visual inspection revealed no blockages in the cooling channels of the double walls. The pre-pumping assessment conducted to check the vacuum level and residual gas condition, concluded that the inside of the VV was flawless and thus met the leak test requirements of 1 × 10−8 Pa m3 s−1. We confirmed no leakage or deformation through the pressure test under reduced pressure. The helium leak test demonstrated engineering soundness with leak tightness of 6.08 × 10−9 Pa m3 s−1, which is more stringent than the allowable limit. Furthermore, three-dimensional metrology was utilized to determine the as-built dimensions of the manufactured sector. Due to unavoidable weld deformation and tight tolerances, the as-built result does not perfectly meet the assigned tolerance level. Nevertheless, it can be considered as advanced information for assembly with in-vessel components and other sectors. Based on the conformance and suitability of the suggested FAT for the first ITER VV sector, we will determine the acceptability of the upcoming VV sectors, which will be manufactured and delivered by Korea shortly.

The DIII-D Advanced Tokamak (AT) program in the DIII-D tokamak [J. L. Luxon, Plasma Physics and Controlled Fusion Research, 1986, Vol. I (International Atomic Energy Agency, Vienna, 1987), p. 159] is aimed at developing a scientific basis for steady-state, high-performance operation in future devices. This requires simultaneously achieving 100% noninductive operation with high self-driven bootstrap current fraction and toroidal beta. Recent progress in this area includes demonstration of 100% noninductive conditions with toroidal beta, βT=3.6%, normalized beta, βN=3.5, and confinement factor, H89=2.4 with the plasma current driven completely by bootstrap, neutral beam current drive, and electron cyclotron current drive (ECCD). The equilibrium reconstructions indicate that the noninductive current profile is well aligned, with little inductively driven current remaining anywhere in the plasma. The current balance calculation improved with beam ion redistribution that was supported by recent fast ion diagnostic measurements. The duration of this state is limited by pressure profile evolution, leading to magnetohydrodynamic (MHD) instabilities after about 1s or half of a current relaxation time (τCR). Stationary conditions are maintained in similar discharges (∼90% noninductive), limited only by the 2s duration (1τCR) of the present ECCD systems. By discussing parametric scans in a global parameter and profile databases, the need for low density and high beta are identified to achieve full noninductive operation and good current drive alignment. These experiments achieve the necessary fusion performance and bootstrap fraction to extrapolate to the fusion gain, Q=5 steady-state scenario in the International Thermonuclear Experimental Reactor (ITER) [R. Aymar et al., Fusion Energy Conference on Controlled Fusion and Plasma Physics, Sorrento, Italy (International Atomic Energy Agency, Vienna, 1987), paper IAEA-CN-77/OV-1]. The modeling tools that have been successfully employed to both plan and interpret the experiment are used to plan future DIII-D experiments with higher power and longer pulse ECCD and fast wave and co- and counterneutral beam injection in a pumped double-null configuration. The models predict our ability to control the current and pressure profiles to reach full noninductivity with increased beta, bootstrap fraction, and duration. The same modeling tools are applied to ITER, predicting favorable prospects for the success of the ITER steady-state scenario.

The strongest magnetic barrier in the DIII-D tokamak and comparison with the ASDEX UG

The maximum operational high beta in single-null divertor (SND) long pulse tokamak discharges in the DIII-D tokamak with a cross-sectional shape similar to the proposed International Thermonuclear Experimental Reactor (ITER) device is found to be limited by the onset of resistive instabilities that have the characteristics of neoclassically destabilized tearing modes. There is a soft limit due to the onset of an m/n=3/2 rotating tearing mode that saturates at low amplitude and a hard limit at slightly higher beta due to the onset of an m/n=2/1 rotating tearing mode that grows, slows down and locks. By operating at higher density and thus collisionality, the practical beta limit due to resistive tearing modes approaches the ideal magnetohydrodynamic (MHD) limit

Kinetic particle-in-cell (PIC) method is a reliable technique for the laser–plasma interactions simulation based on particles’ statistical behavior. This method can be applied in simulating physical phenomena involved in optical diagnostics tools, such as the interferometry-polarimetry (IP) system. IP is one of the significant laser diagnostic methods suggested in the large tokamaks, such as international thermonuclear experimental reactor (ITER), as to improve the accuracy of density and magnetic field measurements. In this article, two separate simulations of the poloidal IP system are run by applying a 2-D/3-V XOOPIC code to observe the phase-shift and Faraday effect in a magnetized plasma with input density and poloidal magnetic field of ${n}_{e} =3.00\times 10^{19}\,\,\mathrm {m}^{-3}$ and ${B}_{p}=1.00$ T, respectively. For this purpose, a gaseous $\mathrm {(CO}_{2})$ linear-polarized far-infrared laser beam of $\boldsymbol {\lambda }_{i}=118\,\,\mu \text{m}$ wavelength, and about ${30}~{\text {W}}/ {\text {m}}^{2}$ intensity is passed through the plasma, parallel to the magnetic field. The density and poloidal magnetic field are computed from laser phase-shift and Faraday rotation theories, at ${n}_{e,\text {com}}=2.99\times {10}^{19}\,\,\text {m}^{-3}$ and ${B}_{p,\text {com}}=0.99$ T, which correspond to the input values of these simulations. The obtained results indicate the competence and potency of the PIC method in simulating role-playing phenomena in ITER diagnostic devices like the IP system.

Heat removal from the structural elements of a compact tokamak with warm coils intended for the transmutation of minor actinides is an especially critical problem. This paper is concerned with the thermal processes on the first-wall elements and the toroidal field windings, ignoring the most stressed component of JUST-T, i.e. the blanket. The analysis has shown that the peak load on the first-wall elements in JUST-T does not exceed a similar load in the International Thermonuclear Experimental Reactor tokamak. The two-dimensional numerical code for calculation of the current distribution density, Ohmic heating and heat transfer due to heat conductivity and extraction of heat by the water flux has been developed.

In magnetically confined fusion experiments interferometry is commonly used to measure plasma electron density. This because interferometry is a very reliable technique, not affected by calibration problem. The main drawbacks of interferometers are the integral characteristic of the measurement and the fringe loosing problem. Scanning beam interferometry can always overcome the former problem and in some cases the second as well. The advantages of the scanning beam technique can better exploited in two colours vibration compensated medium infra-red (MIR) interferometers. Indeed the short wavelength of the probing beams provide small beam sizes allowing plasma density measurement along many non-overlapping paths, furthermore the vibration compensation system relaxes the requirement of massive interferometer structures embracing the fusion device, actually allowing to install back-reflecting mirror directly attached to the experimental machine.As an example of application of this technique the two-color medium-infra-red-compensated scanning beam interferometer installed on the Frascati Tokamak Upgrade (FTU) experiment is presented. We present also a preliminary design of a scanning interferometer for the new proposed Fusion Advanced Studies Torus (FAST) experiment. The interferometer uses retroreflectors installed inside the vessel to back reflect the scanning beams; this will allow realizing scanning beams measuring from the plasma edge. This feature in principle can solve the fundamental problem of interferometers: the fringe loosing problem.

The DIII-D tokamak research program is carried out by General Atomics (GA) for the U.S. Department of Energy (DOE). The DIII-D is the most flexible tokamak in the world. The primary goal of the DIII-D tokamak research program is to provide data to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. In doing so, the DIII-D program provides physics and technology R&D outputs to aid the Tokamak Physics Experiment (TPX) and the International Thermonuclear Experimental Reactor (ITER). Specific DIII-D objectives include the steady-state sustainment of plasma current as well as demonstrating techniques for microwave heating, divertor heat removal, fuel exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion with high beta and with good confinement. The long-range plan is organized into two major thrusts; the development of an advanced divertor and the development of advanced tokamak concepts. These two thrusts have a common goal: an improved DEMO reactor with lower cost and smaller size than the present DEMO which can be extrapolated from the conventional ITER operational scenario. In order to prepare for the long-range program, in FY93 the DIII-D research program concentrated on three major areas: Divertor and Boundary Physics, Advanced Tokamak Studies, and Tokamak Physics. The major goals of the Divertor and Boundary Physics studies are the control of impurities, efficient heat removal and understanding the strong role that the edge plasma plays in the global energy confinement of the plasma. The advanced tokamak studies initiated the investigation into new techniques for improving energy confinement, controlling particle fueling and increasing plasma beta. The major goal of the Tokamak Physics Studies is the understanding of energy and particle transport in a reactor relevant plasma.

The DIII-D tokamak research program is managed by General Atomics (GA) for the US Department of Energy (DOE). Major program participants include GA, Lawrence Livermore National Laboratory (LLNL), Oak Ridge National Laboratory (ORNL), and the University of California together with several other national laboratories and universities. The DIII-D is a moderate sized tokamak with great flexibility and extremely capable subsystems. The primary goal of the DIII-D tokamak research program is to provide data for development of a conceptual physics blueprint for a commercially attractive fusion power plant. In so doing, the DIII-D program provides physics and technology R&D output to aid the International Thermonuclear Experimental Reactor (ITER) and the Princeton Tokamak Physics Experiment (TPX) projects. Specific DIII-D objectives include the achievement of steady-state plasma current as well as the demonstration of techniques for radio frequency heating, divertor heat removal, particle exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion in plasmas with high beta and with high confinement. The long-range plan is organized with two principal elements, the development of an advanced divertor and the development of advanced tokamak concepts. These two elements have a common goal: an improved demonstration reactor (DEMO) with lower cost and smaller size than present DEMO concepts. In order to prepare for this long-range development, in FY94 the DIII-D research program concentrated on three major areas: Divertor and Boundary Physics, Advanced Tokamak studies, and Tokamak Physics.

DIII-D edge physics data base

A tangential viewing, 10.59 microm CO(2) laser polarimeter for electron density measurements based on plasma induced Faraday rotation has been installed on DIII-D. The system uses colinear right- and left-hand circularly polarized beams with a difference frequency of 40 MHz to generate the necessary signal for heterodyne phase detection. The high-resolution phase information required to adequately resolve degree level polarization rotation is obtained using an all-digital "real-time" phase demodulation scheme based on modern digital signal processing techniques. Initial application of the system to DIII-D disruption mitigation experiments utilizing "massive gas jet" injection exhibits reliable operation and excellent agreement with CO(2) interferometer measurements; interestingly, the obtained Faraday rotation angles are in the range of those expected in ITER plasmas.

The density in a tokamak can be determined by measuring the Faraday rotation of a laser directed tangent to the toroidal field. If there is a horizontal array of such beams, then ne(R) can be readily obtained with a simple Abel inversion about the center line of the tokamak. For a large machine, such as ITER, TPX, or JT-60, a 10.6 μm laser would be appropriate. If the machine operated at a full field of 10–50 T m and a peak density of 2.5×1020/m3, the rotation angle would be quite large—about 15°–75° per pass. An elegant measurement system can be made up from a single laser beam diffracted off a moving grating to form a fan of ∼10 probe beams. With the addition of a few optical components to the system, the return beams can be recombined and sent to a single detector. In the detector there is a separate frequency component for both the right and left hand component of each ray. These can be separated electronically to provide a reference and probe signal for each ray; the difference in phase between the two signals is twice the Faraday rotation angle.