Fusion Engineering and Design

Published by Elsevier
Print ISSN: 0920-3796
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For the future stellarator W7-X of IPP at Greifswald, a powerful electron cyclotron resonance heating facility is being built which will include 10 gyrotrons at 140 GHz with a CW power of 1 MW each plus 2 optional gyrotrons at 70 GHz. The millimetre wave power will be transmitted fully optically via two multiple-beam waveguides (MBWG) with a length of about 50 m. This paper presents calculations and low-power measurements of the transmission characteristics of a prototype MBWG, which promise a high transmission efficiency for the ECRH system under construction
 
The ARIES-II and ARIES-IV designs are the final two designs of the ARIES study. The ARIES-II design uses liquid lithium and vanadium for low activation, multiple barriers between the lithium and air and an inert cover gas to prevent lithium-air reactions. The ARIES-II reactor is passively safe (level of safety assurance=2) with a total l-km early dose of about 0.88 Sv. The ARIES-IV tokamak reactor has been designed to avoid the use of materials subject to neutron activation and materials that are energy sources for the release of those activation products which do occur. The coolant is helium, the breeder is lithium oxide, and the structure of the first wall, blanket and shield is silicon carbide. For the ARIES-IV design, beryllium metal is used for neutron multiplication. Since beryllium metal is combustible, releasing about 60 MJ/kg, the multiplier is the chief source of chemical energy for the release of activation products in the structure. We can argue that less than 1O% of the <sup>24</sup>Na inventory is likely to diffuse out of the SiC during a fire in which the Be neutron multiplier is consumed. Therefore, the offsite dose would the less that 2 Sv, and the reactor satisfies the condition for LSA=1
 
Transverse beam combining is a cost-saving option employed in many designs for induction linac heavy ion fusion drivers. But resultant transverse emittance increase, due predominantly to anharmonic space charge forces, must be kept minimal so as not to sacrifice focusability at the target. A prototype combining experiment has been built, using the MBE-4 experiment. Four sources produce four 4 mA Cs<sup>+</sup> beams at 200 keV. The ion sources are angled toward each other, so that the beams converge. Focusing upstream of the merge consists of 4 quadrupoles and a final combined-function element (quadrupole and dipole). All lattice elements are electrostatic. Due to the small distance between beams at the last element (~2 mm), the electrodes here are a cage of small wires, each at different voltage. The beams emerge into the 30 period transport lattice of MBE-4 where emittance growth due to merging, as well as the subsequent evolution of the distribution function, can be diagnosed. The combiner design, simulation predictions, and preliminary results from the experiment are presented
 
On the basis of a high bootstrap current fraction observation with JT-60, the concept of a Steady-State Tokamak Reactor, the SSTR, was conceived and was developed with the design activity of the SSTR at JAERI (Japan Atomic Energy Research Institue). Results of ITER/FER (International Thermonuclear Experimental Reactor/Fusion Experimental Reactor) design activities have enhanced the SSTR design. Moreover, the progress of R&D for fusion reactor engineering, especially in the development of superconducting coils and negative-ion-based NBI at JAERI, has promoted the SSTR conceptual design as a realistic power reactor. Although present fusion power reactor designs are currently considered to be too large and costly, results of the SSTR conceptual design suggest that an efficient and promising tokamak reactor will be feasible. The conceptual design of the SSTR provides a realistic reference for a demo tokamak reactor. Notably, a reduction of the circulating power of the power station facility, owing to spontaneous bootstrap current, may ensure the high Q around 50 necessary for a commercial power reactor
 
FIRE is a high-field, burning-plasma tokamak that is being studied as a possible option for future fusion research. Preliminary parameters for this machine are R<sub>0</sub>=2 m, a=0.5 m, B<sub>0</sub>=10 T, and I<sub>p</sub>=6 MA. Magnetic field coils are to be made of copper and pre-cooled with LN<sub>2</sub> before each shot. The flat-top pulse length desired is &ges;10 s. Ion cyclotron and lower hybrid rf systems will be used for heating and current drive. Present specifications call for 30 MW of ion cyclotron heating power, with 25 MW of lower hybrid power as an upgrade option
 
This paper identifies the requirements for a remotely operated precision laser ranging system for the International Thermonuclear Experimental Reactor. The inspection system is used for metrology and viewing, and must be capable of achieving submillimeter accuracy and operation in a reactor vessel that has high gamma radiation, high vacuum, elevated temperature, and magnetic field levels. A coherent, frequency modulated laser radar system is under development to meet these requirements. The metrology/viewing sensor consists of a compact laser-optic module linked through fiberoptics to the laser source and imaging units, located outside the harsh environment. The deployment mechanism is a remotely operated telescopic mast. Gamma irradiation up to 10<sup>7</sup> Gy was conducted on critical sensor components with no significant impact to data transmission, and analysis indicates that critical sensor components can operate in a magnetic field with certain design modifications. Plans for testing key components in a magnetic field are underway
 
This paper examines the effects of test specimen size on the cryogenic fracture toughness properties of a nitrogen-strengthened austenitic stainless steel for superconducting magnet structures in fusion energy systems. Single-specimen J-integral tests were performed on CT (compact tension) specimens with and without side-grooves in liquid helium at 4 K. The aspect ratio (specimen width to thickness) and thickness were varied. A three-dimensional finite element analysis was also conducted to investigate the effects of specimen thickness and side-groove on the through thickness distributions of the J-integral values. The results of the finite element analysis are used to supplement the experimental data.
 
This paper presents the Fusion Virtual Assembly System FVAS 1.0 which makes possible engineering application for assemblies of large-scale complex nuclear facilities. This system focuses on utilizing a desktop virtual environment tightly coupled with commercial Computer Aided Design (CAD) systems, and is designed to support assembly planning, evaluation, demonstration, and training, and to work on personal computer (PC) platform. Main features of FVAS include: (1) automatic assembly modeling, (2) separation of display scene and collision detection scene, (3) multi-way assembly planning, (4) records and replays of assembly processes. This paper describes the system architecture, features and capabilities of FVAS. And tests conducted using real-world engineering models are also described.
 
A high power operation of 1.2 MW for 4 s was obtained by a 110 GHz gyrotron with depressed collector on the JT-60 ECRF system, where the cathode voltage is −58 kV without voltage regulation and the acceleration voltage is +85 kV with regulation. The gyrotron is featured by a diamond output window and a RF absorber in its beam tunnel. The RF absorber suppresses the parasitic oscillation, which limited the performance of the gyrotron without the RF absorber less than 1 MW×2 s. The operation was done with the JT-60 tokamak plasma as a dummy load, where the EC wave was transmitted through a diamond antenna window on the corrugated waveguide. The edge temperature of the antenna window without cooling was in proportion to the pulse duration. The maximum efficiency of the gyrotron was about 42% in the depressed collector operation on the JT-60 ECRF system.
 
KSTAR (Korea Superconducting Tokamak Advanced Research) is a national tokamak aiming at the high beta operation based on AT (Advanced Tokamak) scenarios in Korea and ICRF (Ion Cyclotron Ranges of Frequency) is one of the essential heating and current drive tools to achieve this goal. The ICRF heating and current drive scenario requires 4 units of 2 MW transmitters with a frequency range from 25 to 60 MHz. The first KSTAR transmitter is a modified FMIT (Fusion Material Irradiation Test) transmitter consisting of four amplifier stages. An amplitude-modulated 1 mW frequency source drives a 500 W solid state wideband amplifier, which in turn drives three tuned triode/tetrode amplifier stages. The tube employed in the final power amplifier is a 4CM2500KG tetrode fabricated by CPI (Communications & Power Industries). After the fabrication of the cavity and power supply was completed in 2004, several failures of the tube during a factory and a site acceptance test occurred before eventually achieving 1.9 MW for 300 s at 33 MHz in 2007. The electrical efficiency of the FPA (Final Power Amplifier) is about 70%. Although this is a very encouraging result for the development of an ICRF transmitter for ITER (International Thermonuclear Experimental Reactor), continued efforts for a reliable operation are required to achieve the final goals of the KSTAR and ITER ICRF system.
 
The influence of blistering on deuterium retention in polycrystalline tungsten (W) has been studied at incident energies of tens of eV and a fixed incident flux of 1 × 1022 D/m2/s. The surface morphology was observed using scanning electron microscopy (SEM) and the deuterium retention in the irradiated samples measured using thermal desorption spectrometry (TDS). The results indicate that at substrate temperature around 400–500 K dense and high dome-shaped blisters appeared at the surface and the deuterium retention also reached its maxima. The blisters then became lower and sparse and finally disappeared at 900 K. Additionally, the D retention decreased to a quite low level above 700 K. Thermal desorption spectra of the deuterium in the irradiated samples showed four peak temperatures as a function of the substrate temperature during irradiation, indicating different types of trapping sites.
 
Power switching in RF heating systems is a delicate function as it is often linked to high power tube protection. In most RF systems, the end stage power tube is fed by a high voltage power supply (HVPS), which connection to the tube has to be interrupted in case of arc suspicion. The amount of energy that is allowable to be dissipated in the arc is in the range of 10–50 J, to limit the degradations observed on the tube structures. The protection function is usually performed by a crowbar. Furthermore, the HVPS is often shared by several power tubes, and the loss of all the power from the group of tubes is to be avoided to minimize the perturbation on the plasma experiment. A description of a 40 kV thyristor based crowbar and a 100 kV, 25 A MOSFET switch is given, as well as the contours of the existing components for high power switching applications.By combining small components, such as thyristors or MOSFET, in matrix, highly compact and reliable units have been built and implemented in Tore Supra RF systems.
 
Controllable power combination and distribution of multiple sources into multiple transmission lines may increase efficiency and flexibility of ECRH systems. In this work a narrow band diplexer–combiner is presented: it allows fast independent switching of power from two different gyrotrons in two transmission lines. It is based on a resonator made using suitable lengths of square corrugated waveguides (SCW) as beam splitters/combiners. A 105 GHz prototype was installed and tested in a low power configuration. Three SCW 2.5 m long with four mirrors are used to form two nested resonator loops. Two outputs are available and, depending on frequency, the desired amount of output power can be directed in either. Under ideal conditions, a worst-case 98%–2% power split would be achievable. Beam patterns were measured at the output of the first SCW in order to check similarity with the input beam and alignment. Frequency sweeps were made in order to guarantee that the resonant circuit was working properly. Finally, absolute power measurements were made with a quasi-optical bolometer to verify the simulations. Results of all tests are presented and discussed in this work, together with a feasibility study for insertion of the device in an existing ECRH transmission line.
 
Electron Cyclotron Resonance Heating (ECRH) is the main heating method for the Wendelstein 7-X stellarator (W7-X) which is presently under construction at IPP Greifswald. The mission of W7-X is to demonstrate the inherent steady state capability of stellarators at reactor relevant plasma parameters. A modular 10 MW ECRH-plant at 140 GHz with 1 MW CW-capability power for each module is also under construction to support the scientific objectives. The commissioning of the ECRH-plant is well under way; three gyrotrons are operational. The strict modular design allows to operate each gyrotron separately and independent from all others. The ECRH-plant consists of many devices such as gyrotrons and high voltage power supplies, superconductive magnets, collector sweep coils, gyrotron cooling systems with many water circuits and last but not least the quasi-optical transmission line for microwaves with remote controlled mirrors and further water cooled circuits. All these devices are essential for a CW operation. A steady state ECRH has specific requirements on the stellarator machine itself, on the microwave sources, transmission elements and in particular on the central control system. The quasi steady state operation (up to 30 min) asks for real time microwave power adjustment during the different segments of one stellarator discharge. Therefore, the ECRH-plant must operate with a maximum reliability and availability. A capable central control system is an important condition to achieve this goal. The central control system for the 10 MW ECRH-plant at W7-X comprises three main parts. In detail these are the voltage and current regulation of each gyrotron, the interlock system to prevent the gyrotrons from damages and the remote control system based on a hierarchy set of PLCs and computers. The architecture of this central control system is presented.
 
Experimental data on permeability and diffusion of hydrogen through austenitic 18Cr–10Ni–0.65Ti steel with 0.08 and 0.12 wt.% of carbon have been obtained with special equipment designed and installed at the research reactor IVV-2M. It has been shown that parameters of hydrogen isotope transfer in the steel increased substantially during irradiation by fast neutrons of flux density ffast<1.8×1018 n m−2 per s, fluence F=9.5×1024 n m−2 (E>0.1 MeV) and absorbed gamma dose 3.6 W g−1 at temperatures of 573–1073 K. Under these irradiation conditions, even a small oxygen admixture content in hydrogen allowed creation of reliable surface barriers on the steel, which essentially decreased hydrogen permeability. At the same time, the physico-mechanical properties of the steel under study changed.
 
Efficient transmission and polarization in 110 GHz transmission line for JT-60U was studied. A 40-m run transmission test line (inner diameter of 31.75 mm) with eight miter bends, which included a pair of polarizers was designed and constructed. The transmission at 1 MW was performed and the transmission efficiency of 89% (without mirror optical unit (MOU)) was obtained. In basis of the good result, a 1 MW, 110 GHz electron cyclotron range of frequency (ECRF) system was constructed in JT-60U. Transmissions of 1 MW-2.0 s and 0.32 MW-5.0 s were performed in JT-60U ECRF system. The polarization of the transmitted wave in the transmission line was measured and the dependence of plasma electron heating on polarization was investigated. The optimum elliptical polarization of the incidence wave for the ECRF experiments in JT-60U was calculated using the cold dispersion relation with the measured polarization. The optimum polarization was applied to the electron heating experiment of JT-60U.
 
ECH power has proven capabilities for both heating and current drive in energetic plasmas. For the second phase of ECH power on DIII–D, there will be three 1 MW sources added to the existing three systems for a total power generating capacity of 5.1 MW. This upgrade is based on the use of the single disc chemical vapor deposition (CVD) diamond window, 1 MW diode gyrotron, recently developed by CPI. All gyrotrons are connected to the tokamak via a low-loss, windowless, evacuated transmission line system, using circular corrugated waveguide for propagation in the HE11 mode. Each waveguide system incorporates an in-vessel two-mirror launcher. The newest launcher can steer the rf beam poloidally from the center to the outer edge of the plasma and toroidally for either co- or counter-current drive. An overview of the total system, its integration with the DIII–D tokamak, and recent results will be discussed. The various new aspects of the upgrade ranging from building modifications to the use of the new steerable launcher will also be addressed.
 
This paper reports the test results of newly developed Varian tetrodes, X2242 and X2274, using a JT-60 ICRF amplifier under the US-Japan collaboration program. The JT-60 ICRF amplifier was designed to deliver 0.75 MW at 110 to 131 MHz with the Varian EIMAC 8973 tetrode. Although the new tetrodes are similar to the 8973 in all dimensions, they have pyrolitic graphite grids for higher screen and control grid dissipation capability. The new tetrodes require only the following amplifier system modifications: (a) new filament, bias, and screen grid power supplies, (b) a second output port for reduction of rf voltage in the output cavity. The objective for the tests are to confirm 1.5 MW output at 130 MHz for 5 seconds, and to check the reliability of both the tube and the amplifier with a mismatched load which simulates power transmission to an antenna coupled to the plasma. The first test with an X2242 demonstrates that excessive screen dissipation limits the output power. The second test with an X2274, whose improved screen grid reduces rf heating to 50% of that of the X2242, achieves 1.7 MW at 131 MHz for 5.4 seconds. This is not only a power higher than the objective but also the highest long pulse VHF power level for fusion research above 110 MHz. The modified amplifier with the X2274 also shows good, stable performance in the mismatched load tests. As the theory predicts, the maximum anode dissipation is 1.4 times higher with a VSWR = 1.5 than with the previous VSWR ≅ 1.0.
 
Successful gyrotron experiments at FZK employing a conventional hollow cylindrical waveguide cavity, a quasi-optical mode converter with dimple-type launcher, a broadband silicon nitride Brewster window and a single-stage depressed collector (SDC), gave up to 1.6 MW output power at efficiencies between 48 and 60% for all operating mode series in the frequency range from 114 to 166 GHz. Frequency tuning in 3.7 GHz steps has been achieved by slow variation (minutes) of the magnetic field strength in the cavity. A specific hybrid magnet system for fast frequency tuning (1 s) is being manufactured. These experiments confirm the preliminary results achieved with a prototype fused quartz glass Brewster window. Up to now, the pulse duration has been 1–5 ms. However a water-edge-cooled silicon nitride composite Brewster window with its thermal conductivity of k=60 W/mK, permittivity εr=7.85 (θBrewster=70.35°), tan δ=3.5×10−4, excellent mechanical properties and clear window aperture of 100 mm diameter, will allow long-pulse operation (1 s) in the 1 MW power range. By increasing the beam current to 70 A (70% of the limiting current!), a maximum power of 2.14 MW, the highest value ever generated by a weakly relativistic gyrotron, has been obtained at 140 GHz with an efficiency of 34% (53% with SDC). With a TE31,17 coaxial cavity gyrotron, also equipped with a quasi-optical mode converter and a SDC, a maximum RF-output power of 1.7 MW with an efficiency of 26.2% has been achieved at a beam current Ib=69.7 A and a cathode voltage Uc=93.2 kV. Around the nominal beam parameters (Ib=52.2 A, Uc=91.8 kV, Bcav=6.65 T) an RF-output power of 1.3 MW, with an efficiency of 27.2% (41% with SDC), has been measured. In experiments on frequency step tuning, also using the silicon nitride composite Brewster angle window, 19 different modes with ≈2.2 GHz frequency spacing have been excited with ≥1 MW power at frequencies in the range between 134 and 169.5 GHz.
 
The company JEMA has designed and manufactured two High Voltage Switching Mode Power Supplies (HVSMPS), rated at 130 kV dc and 130 A, each of which will feed the accelerator grids of two Positive Ion Neutral Injector (PINI) loads, to be installed at the Joint European Torus (EFDA-JET facility located at Culham, UK). The solution designed by JEMA includes two matching transformers which adapt the 36 kV of the JET AC power distribution network to the required 670 V at the secondary side. Additionally, such transformers provide a 30° phase shift which is required by a 30 000 A 12 pulse thyristor rectifier. The obtained and stabilised 650 V feed 120 IGBT invertors, which operate at 2778 Hz with modulated square waveform. Each invertor feeds a High Insulation High Frequency Transformer. The 120 transformers corresponding to one power supply are arranged in three oil filled tanks and provide the main insulation from the low voltage to the high voltage side. The square waveform obtained at the secondary of each transformer is rectified by means of a diode bridge. The connection in series of the 120 diode bridges provides the required 130 kV d.c. at the output. In order to protect the load, a redundant solid state crowbar has been designed. Such short circuiting device is composed of 26 Light Triggered Thyristors (LTTs), connected in series. Electrical simulations have been carried out in order to ensure that the system complies with the requirements of high accuracy and adequate protection of the load. The critical design of the High Voltage-High Frequency Transformers has also required electrostatic simulations of the electric field distribution
 
The design features, on-site testing, commissioning and operation are described of two new 130 kV/130 A HV power supply units serving four upgraded 130 kV/60 A positive ion neutral injectors (PINIs) on JET. Both units were factory tested at full power and pulse length into dummy resistive load. Following on-site installation, the factory tests were repeated. The transition from dummy-load testing to PINI operation required full integration of the HVPS within the overall JET control system, and rigorous testing of the co-ordinated actions and protections of all PINI power supplies (filament and arc for plasma source and negative suppression grid). The implementation of these functions is described. Extensive use was made of parasitic integrated test pulses, where the other PINIs could be operated normally, with the HVPS energised under full remote control together with the corresponding PINI plasma sources, but with the HVPS connected to dummy load. The amount of NB operation time dedicated to commissioning was thereby minimised, yet gave a high degree of confidence of readiness for HV energisation of the PINI, and first beam operation followed less than 24 h from HV connection to the PINI. The routine operating experience and performance, including load protection characteristics of the new HVPS units are also described.
 
Neutron spectrum measurements in a cylindrical assembly of 316L stainless steel were carried out for DT neutrons to assess the multigroup neutron cross section libraries and the calculational method for fusion neutronics. Energy spectra of scalar flux in the assembly above 0.8 MeV were measured at the positions of 8 to 88 cm in depth using a small spherical NE213 liquid scintillator with an unfolding code FORIST. The measured spectra were analyzed with the two-dimensional transport code DOT3.5 by using three multigroup cross section libraries originated from the ENDF/B-III, -IV, and JENDL-2 and -3PR1 nuclear data files. The calculated spectra with the 42-group library from ENDF/B-III are not consistent with the measured ones in shape. Calculation with the 125-group library from ENDF/B-IV overestimates the measured spectra in the whole energy region at every position, because of the uncertainty in the non-elastic cross section for Cr and the elastic and non-elastic cross sections for Fe. The best agreement is obtained between the measurement and the calculation using the 125-group library from JENDL-2 and -3PR1.
 
Substantial degradation of the electrical resistivity of ceramic insulators may occur during the operation of fusion reactors. This is caused by radiation-induced conductivity (RIC) resulting from high energy neutrons with energies up to 14 MeV. Although data on the RIC resulting from neutrons of energy 14 MeV are required, no measurements have been performed so far. The first in situ measurements of the electrical resistivity during irradiation with neutrons of energy 14 MeV was carried out for Al2O3 in the present study using the Fusion Neutronics Source (FNS) at the Japan Atomic Energy Research Institute. An irradiation chamber was developed which could realize appropriate measurements of RIC for the specimens in the neutron field of the FNS. Using the irradiation chamber, the RIC of Al2O3 resulting from irradiation by neutrons of energy 14 MeV was measured at room temperature in the neutron flux range 1012−1015 n m −2 s−1. If the obtained relationship between the RIC and the neutron dose rate can be extrapolated to fusion-reactor-relevant neutron dose rates, then the electrical conductivity of Al2O3 increases to a level of 10−7−10−6 Ω−1 m−1. Such a level of degradation of the electrical resistivity may be accommodated with appropriate fusion reactor design.
 
It is now widely recognized that further progress toward a commercial fusion reactor critically depends on the availability of low activated materials to be operated for many years in the fusion neutron environment without degradation of their properties. Therefore, high power neutron source (NS) for extensive material tests is in great demand. The realistic NS can be built on the basis of the gas dynamic trap (GDT) which is one of the simplest system for magnetic plasma confinement. GDT is an axisymmetric mirror machine of the Budker–Post type, but with a high mirror ratio (R>10), and operated with warm and relatively dense plasma. Thus, due to frequent collisions, the plasma confined in a trap is very close to isotropic Maxwellian, and hence many instabilities are not excited and plasma behavior is similar to a classical one.
 
Silicon carbide (SiC) in the form of ceramic matrix is a low activation structural material proposed for fusion reactors. Its development is pursued in the European Fusion Technology Program. A SiC block (457×457×711 mm3), borrowed from JAERI, was irradiated with 14 MeV neutrons at the FNG facility of ENEA Frascati. Activation reaction rates, neutron fluxes and spectra, as well as nuclear heating were measured in four selected experimental positions inside the block. The experimental analysis was performed using the Monte Carlo transport code MCNP-4C and point-wise cross sections derived from FENDL-2.0, EFF-2.4 and EFF-3.0 evaluated nuclear data files. Deterministic transport calculations were also performed using the discrete ordinates code DORT. The sensitivity and uncertainty analysis were performed as well using the SUSD3D code. Results indicate that calculation based on EFF-3.0 nuclear data file estimates the neutron flux and spectra with a reasonable uncertainty which is still lower than ±30% for all measured quantities.
 
In the 14 MeV neutron streaming experiment performed at the Frascati Neutron Generator (FNG) a mock-up of a module of a fusion reactor shielding system was irradiated. A void channel with a high aspect ratio (l/a>10) was penetrating through the shield and a cavity was placed at the end. Several nuclear quantities (activation reaction rates) were measured at different positions. In this work the experiment has been analysed using discrete ordinates (SN) code DORT and the sensitivity/uncertainty code package SUSD3D. The uncertainty of the neutron flux at two measurement positions (A — inside the cavity and B — 87.6 cm in the shielding block) due to the iron cross-sections was found to be very low (e.g. the for the flux above 10 MeV between ∼0.2% and 3%). The uncertainty analysis is normally useful to explain the discrepancies between the calculated (C) and the measured (E) nuclear quantities, or at least to eliminate some possible sources for the discrepancy and to quantify uncertainty in design calculation. In this case the uncertainty due to the iron cross-sections seems to be too low to explain the discrepancy between C and E.
 
The quasi-optical beam launching antenna for the Electron Cyclotron Resonance Heating (ECRH) experiment at 140 GHz, 1.6 MW to the plasma, on the Frascati Tokamak Upgrade (FTU) has been successfully exploited in experimental operations. It provides four beams 400 kW each, with independent poloidal and toroidal steering capability and a maximum power density of 60 kW/cm2 at the plasma edge. The beam radius in the plasma is ≈20 mm, allowing a very high localisation of the absorbed power. The main characteristics of the antenna are:•Four movable launching mirrors under vacuum, far from the plasma edge (no movable parts near the plasma).•Oblique toroidal injection capability at fixed angles obtained with reflections on two fixed stainless steel plates, gold plated, inserted at the sides of the port.•Vacuum gate separating the main vacuum from the appendix containing the two matching mirrors and the barrier window (for safety and maintenance).•All movements are transferred outside the vacuum chamber through bellows and linear displacement. Actuators and encoders are in air.•Full capability to adjust small movements of the machine during cool-down and warm-up, preserving the alignment. Pattern measurements at low power before installation were performed to characterise the effects of the real system on the ideal shape and the polarisation of the launched beams, with particular attention to diffraction effects. Final control of the shape of the beams and the alignment of the mirrors in the system is performed after installation, under vacuum and with the tokamak at liquid nitrogen temperature, with the aid of a retractable probe and using a thermopile as the RF detector.
 
For the development of a 1 MW, 140 GHz gyrotron for CW operation which will be installed at the stellarator facility Wendelstein 7-X at IPP Greifswald, a collaboration between different European research institutes and an industrial company has been established. In order to prove the proper functioning of the millimeter wave components installed in the gyrotron — such as the cavity, the waveguide taper and the quasioptical mode converter — these components should be cold tested, preferably before installation. However, due to lack of time as well as long delivery times, this was not possible. Therefore, two units of the quasioptical mode converter and the cavity were fabricated with identical geometry, one of those being used for measurements on the low power test device. To perform these cold tests for tapers and mode converters, the gyrotron cavity output mode has to be simulated. This means that a high order rotating mode (TE28,8 mode) must be generated at low power. This can be achieved by means of a mode generator consisting of two mirrors and a coaxial cavity with a perforated outer wall. Before applying the mode generator to the components, its proper behavior and the accurate alignment of the system must be verified either by radiation pattern measurements or k-spectrometer measurements. As the coupling through the holes of the k-spectrometer is extremely low, a special vector network analyzer with a dynamic range of at least 100 dB had to be developed. This has been achieved by integration of a phase locked backward-wave oscillator with a line width of 100 Hz and an output power of 10 mW. A non-destructive measurement of the resonance frequency and the quality factor of the cavity does not seem possible. The second cavity will be prepared for the cold measurement by drilling a small radial hole into its wall in the plane of the field maximum. This hole is then used for the input coupling. The accuracy required for this hole is rather critical. The coupling coefficient must be high for sufficient excitation of the rf field, but on the other hand it must neither change the frequency nor the quality factor strongly. The transmission is measured by a probe at the output of the uptaper.
 
In current tokamaks and, in particular, in future larger devices such as ITER, the control of neo-classical tearing modes (NTM) is essential for achieving high performance in terms of the β limit. A commonly used scheme for NTM stabilization consists in driving a helical current at the resonance surface of interest with electron-cyclotron-current-drive. Depending on the ratio between the magnetic island size and the RF beam width, complete stabilization of the NTM will only be achieved with deep RF power modulation in phase with the mode. In the frame of the European development program of high power sources for ECRH applications between Forschungszentrum Karlsruhe, IPP Garching/Greifswald, EPFL Lausanne, IPF Stuttgart, CEA Cadarache and Thales Electron Devices, the modulation capabilities of the 140 GHz/1 MW gyrotron have been experimentally investigated. RF-power modulation depths higher than 80% at a frequency of 50 kHz with cathode modulation and 1.5 kHz with depression voltage modulation have been obtained. The limitations in frequency were given by the corresponding power supplies and not by the gyrotron itself. Detailed analysis of the collector loading with respect to the modulation scheme will be presented and the intrinsic gyrotron limitations for long-pulse operation with deep modulation will be discussed.
 
For the future stellarator W7-X of IPP at Greifswald, a powerful electron cyclotron resonance heating (ECRH) facility is under design, which will include ten gyrotrons at 140 GHz with a CW power of 1 MW each plus two optional gyrotrons at 70 GHz. The millimetre wave power will be transmitted via two multi-beam waveguides (MBWG) with a length of about 50 m. This paper shows results for the optimum shape of the mirrors and the mode conversion of the MBWG transmission system. The design which will be used for water-cooled mirrors and cooling structures is discussed. Concepts for the back reflectors, which will be mounted at the inner side of the vacuum vessel of W7-X to redirect the non-absorbed part of the beam back into the plasma are presented. A test facility to test the transmission properties of quasi-optical lines, including a measurement system for ohmic loss of mirrors, a reflectometer to detect possible surface deformations and an alignment control system is described.
 
For plasma heating by ECR in the Stellarator W7-X under construction, 140 GHz gyrotrons with 1 MW cw output power are under development. These tubes have a voltage depressed collector for electron energy recovery. Each gyrotron is fed by two high-voltage sources: a high-power supply for driving the electron beam and a precision low-power supply for beam acceleration. In addition, a protection system with a thyratron crowbar for fast power removal in case of gyrotron arcing is installed. The low-power high-voltage source for beam acceleration is realized by a high-voltage servo-amplifier driving the depression voltage such that the influence of the voltage noise of the main high-power supply on the acceleration voltage is suppressed by feed-back control of the amplifier. Design and simulation of the servo-amplifier by PSpice is presented.
 
Electron cyclotron resonance heating (ECRH) has proven to be one of the most attractive heating schemes for stellarators. Therefore, ECRH was chosen to be the main heating method for the Wendelstein 7-X stellarator (W7-X) now under construction at IPP Greifswald, Germany. A 10 MW ECRH system with continuous wave (CW) possibilities, operating at 140 GHz will be built up to meet the scientific goals of the stellarator. Two prototype gyrotrons with an output power of 1 MW were developed in collaboration between European research laboratories and European industry (Thales Electron Devices, France). The gyrotrons are equipped with a single-stage depressed collector, an optimised quasi-optical mode converter and a CVD-diamond window. The prototypes have been successfully tested at FZK. With the second one, an output power of 0.89 MW at a pulse duration of 3 min and an output power of 0.54 MW for about 15 min have been obtained.
 
The power supply of the ASDEX Upgrade (AUG) tokamak consists of 11 thyristor converter modules which feed the poloidal field coils. All converters installed are powered by 10.5 kV flywheel generators starting at a frequency of 110 Hz, running down to 85 Hz during a pulse. For quasi-stationary advanced tokamak experiments with enlarged flat-top phase, the power supply system must be extended. The variable frequency of the pulsed network, fast load changes, together with the different parameters of the load coils require a very sophisticated converter design with reduced reactive power consumption. The paper describes the design and testing of the modular Thyristor Converter Group 6 with neutral control and four quadrant possibilities. It presents the various configurations available for operation on AUG magnetic coils, analyses the results of measurements obtained during commissioning, compares them to the calculated (design) values and reports on the performance achieved in fast four quadrant operation, improving the possibilities of the AUG feedback control of plasma shape and position.
 
A sensitivity analysis has been performed for a 14 MeV neutron benchmark on an iron assembly, typical for a fusion neutronic integral experiment. Probabilistic and deterministic computational methods have been used in the sensitivity calculations with the main objective to check and validate the novel Monte Carlo technique for calculating point detector sensitivities. Good agreement has been achieved between the Monte Carlo and the deterministic approaches for the individual calculated sensitivity profiles, the uncertainties and the neutron flux spectra. It is thus concluded that the Monte Carlo technique for calculating point detector sensitivities and related uncertainties as being implemented in MCSEN, a local version of the MCNP4A code with the capability to calculate point detector sensitivities, is well qualified for sensitivity and uncertainty analyses of integral experiments.
 
Under ITER/EDA R&D Task T-218, an experiment on nuclear heating was conducted at the Fusion Neutronics Source Facility (FNS) in JAERI in order to provide experimental data of nuclear heating for structural materials and to validate the data and methods relevant to ITER design calculations. Probe materials, SS-316, Cu, W, Zr, Be, Cr and graphite were tested in a copper centered experimental assembly bombarded with 14 MeV neutrons. A calorimetric method and TLDs were employed for total heating and γ-heating measurements, respectively. The measurements were carried out along a central axis of a copper centered experimental assembly. The total heating rate in the probe material was derived from a temperature rise measured by a thermistor attached to the probe. MCNP4A with JENDL-3.2, JENDL-Fusion File and FENDL-1 libraries were used for experimental analyses. Nuclear heating was derived with the use of KERMA factors based on JENDL-3.2 and FENDL-1. The adequacy of the KERMA data is discussed based on nuclear heating ratios of calculation to experiment (C/E).
 
Lithium zirconate, Li2ZrO3, is known as a candidate blanket material in a fusion reactor. Various neutronics benchmark experiments for zirconium have thus been carried out so far. According to the independent benchmark studies by two parties, the neutron spectrum calculations show fairly large overestimation for most evaluated nuclear data libraries. However, the reason has not yet been made clear up to now. The author's group expects it would be due to a problem of evaluation for the natZr(n,2n) reaction cross-section, because the cross-section measurement is basically not possible with the foil activation method for zirconium isotopes except for 90Zr.In the present study, two neutrons emitted from natZr(n,2n) reaction have been measured directly to investigate the reason for the above overestimation. The measurement was done with our own special technique of detecting angle-correlated neutrons by the coincidence detection technique and the pencil-beam DT neutron source of FNS, JAEA. Angle-correlated energy differential cross-sections for natZr(n,2n) reaction were successfully measured. The obtained total cross-section above the emitted neutron energy of 800 keV was fairly larger than the one evaluated in JENDL-3.3. The total cross-section of natZr(n,2n) reaction was estimated by extrapolating the spectrum down to zero energy taking into account the nuclear temperature. The estimated cross-section value with the nuclear temperature of 1 MeV, which is larger than the one adopted in JENDL-3.3, was in acceptable agreement with JENDL-3.3. It is suggested from the result that the disagreement pointed out in the previous benchmark studies may be due to inappropriate nuclear temperature used in the evaluation. Further investigation of the nuclear temperature employed in the nuclear data evaluation should thus be carried out once again.
 
A large scale superconducting magnet system is indispensable to realize a fusion reactor and researches on fusion neutron irradiation effects on the magnet materials must be performed continuously and systematically to promote an establishment of reasonable design and fabrication of a fusion reactor. Since pure 14 MeV neutrons are expected to give clear results easily understood in comparison with energy-distributed neutrons, a new cryogenic target system has been installed in Fusion Neutronics Source at Japan Atomic Energy Research Institute under collaboration among Universities, National Institutes and JAERI. The new system has the potential to perform the electrical measurement at cryogenic temperature without warm-up of the samples.This report describes the design and installation of the cryogenic target system for 14 MeV neutron irradiation and some performance test results. The cold target is able to be cooled down within 2 h and there is no temperature rise during a 3 h neutron irradiation.
 
ITER materials properties documentation is extended to weld metals used for welding Type 316L(N) steel, i.e. the structural material retained for manufacturing ITER major components, such as the vacuum vessel. The data presented here are mainly for the Type 16-8-2 and complete those already reported for the low temperature (Type 316L) and the high temperature (Type 19-12-2) filler metals.The weld metal properties data for Type 16-8-2 filler metal and its joints are collected, sorted and analysed according to the French design and construction rules for nuclear components (RCC-MR). Particular attention is paid to the type of weld metal (e.g. wire for TIG, covered electrode for manual arc, flux wire for automatic welding), as well as, to the weld geometry and welding position. Design allowables are derived from validated data for each category of weld and compared with those of the base metal. In most cases, the analyses performed are extended beyond the conventional analyses required for codes to cover specific needs of ITER. These include effects of exposures to high temperature cycles during component fabrication, e.g. HIPing and low dose neutron irradiation at low and medium temperatures.The ITER Materials Properties Handbook (MPH) is, here, enriched with files for physical and mechanical properties of Type 16-8-2 weld metal. These files, combined with the codification and inspection files, are part of the documentation required for ITER licensing needs. They show that all three weld-metals satisfy the code requirements, provided compositions and types of welds used correspond to those specified in RCC-MR.
 
A practical multi-gyrotron oscillation system, using collector-potential depression, composed of six gyrotron tubes and 3 U of power supplies, was designed, fabricated and tested. This system was designed to generate power levels of 3 MW for pulse duration of 1-s at ≈168 GHz for electron cyclotron heating of LHD at the National Institute for Fusion Science. The all-solid-state power supply unit can drive a maximum of three gyrotrons by equipping the collector power supply with three pairs of the anode and body power supplies. The gyrotrons used a TE31,8,1-mode interaction cavity. A single-stage depressed collector with sweeping coils was employed to increase system efficiency and reduce the heat flux to the collector surface. An internal converter produced a flattened Gaussian profile at a single-disk silicon-nitride window. The output mode was reconverted into the HE11 mode by an MOU. We reconstructed a main circuit of the power supply unit because of stray capacitors in the actual circuit. There were some differences between the designed and measured output wave profiles. The tubes were tested for 1-s pulse with power levels of 500 kW; system efficiencies were 30% at the peak and 28% at the average and temperatures of the windows were ≈200°C.
 
This paper describes a part of the work performed on the determination of the thermophysical properties of alloys in the binary system LiPb. The preparation of the eutectic alloy from the pure elements and its characterization using chemical analysis, metallography, thermal and thermal differential analysis is described. Results of the measurements of the following properties are presented: latent heat of fusion, specific heat, density, thermal expansion, thermal and electric conductivity and viscosity. The wetting behaviour of Li(17)Pb(83) against SS 316 is discussed in terms showing the influence of especially of oxygen on the wetting angle in this system.
 
The report presents the results of development of the 170 GHz/1 MW gyrotrons in Russia. The main problems of powerful gyrotrons at present (formation of a high quality helical electron beam, selection of a high efficiency cavity mode, decreasing of inner converter losses, reliable collector and output window) are discussed. The design versions of the main parts of gyrotrons including the short-pulse prototype, first and second versions of industrial models and test set-up are presented. Advanced 170 GHz ITER gyrotrons were also investigated including a single-stage depressed collector versions (CPD) as well as an improved window (CVD diamond window from FZK). The possibility to make high efficient 170 GHz/1 MW gyrotron for ITER is proved by calculations and experiments.
 
High current and high voltage DC circuit breakers have already been used in the field of fusion technology, in the Joint European Torus (JET) and JT60 Tokamaks, with pulsed currents up to 90 kA or more, but were not able to simultaneously carry steady state currents of 60 kA and to interrupt current under high voltage. Therefore, R&D on high current (60 kA) and high voltage (20 kV) switches, especially on DC circuit breakers, has been undertaken in order to ensure the protection of the superconducting coils of the ITER Tokamak in case of a quench. For this purpose, the coil magnetic energy (in the range of several GJ) is discharged in a large resistor connected in parallel with the DC circuit breaker. Most of the coil circuits operate at 60 kA maximum DC current. Nevertheless, the Central Solenoid (CS) coil of the ITER Tokamak was designed, till the middle of 1998, to be operated at 168 kA. The assessment of such a circuit breaker concept, using parallel operation of 60 kA rated mechanical switches, will be developed in this paper. This theoretical analysis assumes the use of pure mechanical switches connected in parallel without any use of semi-conductor devices.
 
An electrical resistivity monitor for the detection of composition changes in the lithium-lead eutectic alloy, Pb-17Li, has been developed. A miniature electromagnetic pump is used to sample alloy continuously from a pool or loop system and force it through a capillary section, within which the necessary resistance measurement are made, prior to its return to the bulk source.To calibrate the monitor, detailed resistivity-temperature and resistivity-composition data have been determined for Pb-Li alloys at temperatures from 600 to 800K and composition from 0 to 20.5 at% Li. The resistivity increases with both temperature and composition; for Pb-17Li at 723 K, dϱ/dT = 0.054×10−8ωK−1 and dϱ/d[K] = 1.27×10−8ωm(at% Li)−1.The sensitivity of the monitor is such that changes in composition of as little as ±0.05 at% Li can be detected and its response time is limited solely by the rate of sampling.
 
The compatibility between the lithium-lead eutectic and different CrMn steels has been studied in the temperature range 723 K in capsules in a rotating furnace for times up to 9500 h and in a thermal convention loop for 4200 h with a ΔT of 50 K.The corroded specimens have been examined by metallographic and SEM analysis.It is shown that the corrosion mechanism consists essentially of the dissolution of Mn in Pb-17Li with the formation of a porous ferrite layer in which penetration of Pb and Li has been observed.No grain boundary attack has been observed. The behaviour of the different steels is reported together with semiquantitative analysis of corrosion layers.
 
The behavior of deuterium in thermal convection loops with molten Pb17Li was investigated in the temperature range from 300 to 610°C, and in the range of deuterium partial pressures from 0.05 to 1000 mbar. Dissolution and desorption are controlled by diffusion through a 0.02 mm thick LM boundary layer at the interface, no chemical reactions are involved in the rate determining step. This boundary layer is also effective in case of permeation through membranes, if one side is covered by the LM. The permeation through 0.6 mm iron was reduced by a factor of 100. However in case of a fusion reactor blanket this boundary layer will not be important, because the wall thickness of the components is much larger. For the 2 mm stainless steel of the thermal convection loops with a downstream oxide layer, no effect of the boundary layer could be seen. The amount of oxides in the loop had no influence on the results. Furthermore an excess of H2 at low partial pressures did not change transfer rates of deuterium.The solubility of deuterium in the LM was determined from the kinetics of loading and degassing. The found values are one order of magnitude smaller than the lowest values so far published.The transport behavior of the rare gases He, Ne, Ar, Kr and Xe was investigated. The solubility of helium was found five orders of magnitude lower than that of deuterium, those for Ne, Ar, Kr and Xe even lower than that for helium. Helium-bubble formation has to be considered if the flow rate of the LM in a blanket is small, or in case of static irradiation experiments. On the other hand argon can be used as covergas for a fusion reactor blanket. Because of the low solubility in the LM, the Ar-41 activity will be much smaller than in sodium cooled reactors.
 
Within the framework of the studies carried out for the development of a gas–liquid alloy contactor for the extraction of hydrogen from Pb–17Li, the behaviour of a 800 mm high packed column has been investigated on the Melodie loop. The previous contactor technology, a structured packing supplied by the Sulzer Company, has been retained since it had shown satisfying efficiency, likely due to the beneficial effect, on the mass transfer, of the liquid flow division that it involves. The best results of the present study have been achieved via a reduction of the liquid load on the packing: an efficiency of up to 30% was reached at 673 K for an inlet hydrogen pressure in Pb–17Li of 1000 Pa. The impact of the hydrogen pressure in the inlet Pb–17Li flow and on the extraction efficiency has been experimentally assessed: this study allowed us to evaluate the potential of the process in terms of packing height. Finally, a future experimental facility, which should allow us to observe the hydraulic behaviour of liquid mercury (simulating Pb–17Li) on the packing is presented.
 
For self-cooled liquid metal blankets, the magnetohydrodynamic (MHD) pressure drop and velocity distributions are considered as critical issues. This paper summarizes MHD work performed for a DEMO-related Pb-17Li blanket, where the coolant flows downwards in rear poloidal ducts; turns around by 180° at the blanket bottom; is diverted from poloidal ducts into short radial channels which feed to toroidal First wall coolant ducts; flows through the subsequent radial channels; is collected again in poloidal channels and leaves the blanket segment at the blanket top. To reduce the pressure drop and to decouple electrically parallel channels, flow channel inserts are used for all the ducts except the first wall ducts. A previous pressure drop assessment resulted in significant values for duct geometries with flow distribution or collection, and multichannel effects for the system of U-bends. As a result of the uncertainty of these assessments, corresponding investigations were carried out recently. Characteristic results are presented in this paper. It is shown that, for both geometries, the pressure drops are considerably lower than those previously assessed. First results from experiments on the velocity distribution in a radial-toroidal-radial U-bend are also presented. Here, it is shown that, with an increasing interaction parameter, the liquid preferentially flows close to the First wall. Additionally, a pair of strong vortices was observed in a toroidal duct. Both effects are supposed very favourable for heat transfer.
 
The high gas leak rate through SiCf/SiC makes such ceramic composites candidate material for tritium extraction. The main issues for the viability of a SiCf–SiC/Pb–17Li extractor are identified. The potential advantages of the proposed extractor concept are demonstrated.
 
In the water-cooled blanket concept designed for fusion reactors, the liquid Pb-17Li alloy surrounding the water pipes is used to produce the tritium required for a self-sustaining fusion reaction. During operation, the steel box which acts as the liquid metal container can be corroded by the lithium–lead alloy flowing at low velocity. The corrosion products dissolved in the alloy are then transported with the flow. In some regions, characterised for example by low temperatures, these corrosion products can crystallise and form aggregates which can be deposited on the wall and can contribute to the plugging of the ducts. The mechanisms involved in corrosion, formation of aggregates and deposition depend on several factors such as hydrodynamics, solubility, kinetics of exchange at the solid/liquid interface, roughness of the wall, … In addition, they can also be affected by the high external magnetic field used to confine the plasma. The present paper gives an overview of the different types of possible magnetic field effects on the corrosion and deposition processes in the flowing liquid alloy.
 
A blanket with Pb-17Li molten alloy as breeder material, low activation (LA) martensitic steel as structural material and water as coolant has been considered for the SEAFP reference plant model (SEAFP Safety and Environmental Assessment of Fusion Power). In this report the compatibility aspects of this system are examined. The corrosion mechanism of martensitic steels is based on dissolution of metallic components of the material in the liquid alloy and on their precipitation in areas of reduced temperature, where the solubility limits of the metallic elements are exceeded. The corrosion rate of martensitic steel is therefore controlled by a number of physico-chemical parameters, such as the temperature at the metal-Pb-17Li alloy interface, the temperature of the cold zone, the temperature difference ΔT, the velocity of molten Pb-17Li alloy, the diffusion rates of dissolved metallic elements, etc. The existing data demonstrate that the corrosion is homogeneous and the weight loss is linear with time. The corrosion rate can be considered acceptable from the engineering point of view. Moreover, the use of aluminide coatings to decrease tritium permeability seems to increase appreciably the corrosion resistance.
 
Top-cited authors
L.V. Boccaccini
  • Karlsruhe Institute of Technology
Neil B. Morley
  • University of California, Los Angeles
Gianfranco Federici
  • Fusion for Energy
M. A. Abdou
  • Alexandria University
Sergey Smolentsev
  • University of California, Los Angeles