Ablation of Fusion Materials Exposed to High Heat Flux in an Electrothermal Plasma Discharge as a Simulation for Hard Disruption

Electrothermal plasma sources operating in the confined controlled arc discharge regime produce heat fluxes in the range expected for hard disruptions in future large tokamaks. The radiative heat flux produced inside of the capillary discharge channel is from the formed high density (10²³–10²⁷/m³) plasma with heat fluxes of up to 125 GW/m² over a period of 100 μs, making such sources excellent simulators for ablation studies of plasma-facing materials in tokamaks during hard disruptions. Graphite, beryllium, lithium, stainless steel, tungsten, copper, and molybdenum are among the materials proposed for use in fusion reactors. Computational experiments with the ETFLOW code using heat fluxes between 10 and 125 GW/m² have shown low total erosion for the low-z materials Li, Be and C and higher erosion for high-z materials Fe, Cu, Mo and W. The time rate of material erosion for various ranges of heat fluxes shows increased erosion with time evolution over the 150 μs pulse length of the simulated disruption event. At the highest values of simulated heat flux, low-z materials were found to ablate almost identically. At all simulated values of heat flux, the ablation of high-z materials correlated positively with the z-number.     

Neutronics shielding analysis for the end plug of a tandem mirror fusion reactor

A neutronics analysis using the Monte Carlo method is carried out for the end-plug penetration and magnet system of a tandem mirror fusion reactor. Detailed penetration and the magnets' three-dimensional configurations are modeled. A method of position dependent angular source biasing is developed to adequately sample the DT fusion source in the central cell region and obtain flux contributions at the penetration components.To assure cryogenic stability, the barrier cylindrical solenoid is identified as needing substantial shielding of about 1 m of a steel-lead-boron-carbide-water mixture. Heating rates there would require a thermal-hydraulic design similar to that in the central cell blanket region. The transition coils, however, need a minimal 0.2 m thickness shield. The leakage neutron flux at the direct converters is estimated at 1.3×10¹⁵ n/(m²·s), two orders of magnitude lower than that reported at the neutral beam injectors for tokamaks around 10¹⁷ n/(m²·s) for a 1 MW/m² 14 MeV neutron wall loading. This result is obtained through a coupling between the nuclear and plasma physics designs in which hydrogen ions rather than deuterium atoms are used for energy injection at the end plug, to avoid creating a neutron source there. This lower and controllable radiation leakage problem is perceived as a potential major advantage of tandem mirrors compared to tokamaks and laser reactor systems.     

Computational Investigation of Nonideal Higher Ionization Carbon Plasma Under Simulated High Heat Fluxes in Fusion Tokamak Reactors

Plasma-facing materials in future large tokamaks will suffer from ablation due to expected hard disruptions, which affects the reactor interior lining tiles and the divertor modules. Ablation and surface evaporation due to the intense heat flux from disruption is associated with ionization of the evolved particulates. Generated ions at such plasma conditions may allow for higher ionization states such that the plasma at the boundary can be composed of electrons, ions (first, second and third ionization) and excited atoms. The boundary layer is dense and tends to be weakly nonideal. The NC State University electrothermal plasma code ETFLOW used to simulate the high heat flux conditions in which the carbon liner tested for simulated heat fluxes for transient discharge period of 100 μs, with FWHM of ~50 μs, to provide a wide range for obtaining reasonable good fits for the scaling laws. Transient events with ~10 MJ/m² energy deposition over short transient of 50–100 μs would produce heat fluxes of 100–200 GW/m². The heat flux range in this simulation is up to 288 GW/m² to explore the generation of carbon plasma up to the third ionization C⁺⁺⁺. The generation of such heat fluxes in the electrothermal plasma source requires discharge currents of up to 250 kA over a 100 μs pulse length with ~50 μs FWHM. The number density of the third ionization is six orders of magnitude less than the first ionization at the lowest heat flux and two orders of magnitude less at the highest heat flux. Plasma temperature varies from 31,600 K (2.722 eV) to 47,500 K (4.092 eV) at the lowest and highest heat fluxes, respectively. The plasma temperature and number density indicate typical high-density weakly nonideal plasma. The evolution of such high-density plasma particles into the reactor vacuum chamber will spread into the vessel and nucleate on the other interior components. The lifetime of the PFCs will shorten if the number of hard disruptions at such extreme heat fluxes would be increasing, resulting in major deterioration of the armor tiles.     

Nuclear Analysis of Blanket/Shield Design for D-D Tokamak Fusion Reactor

A study is made of the nuclear characteristics of blanket/shield design intended for the D-D tokamak reactor incorporating a cooperative effort by the University of Illinois, Brookhaven National Laboratory and Lawrence Livermore Laboratory. The reactor is characterized by high value of plasma beta (β=0.3) and low value of neutron wall loading (W _n =0.436 MW/m²). The 1 m-thick blanket is composed mainly of graphite, and the 1 m-thick shield is a combination of B₄C, Al and H₂O. Multi-dimensional calculations are carried out to confirm the results of one-dimensional calculation and to assess the problems inherent in the design.

Compared to blankets constituted of other materials, the graphite blanket possesses the characteristics of much smaller residual radioactivity, afterheat and biological hazard potential, which could possibly more than offset the lower nuclear heating that can be provided. The design requirements for the magnet shield are satisfied by the present combination and thicknesses of blanket/shield materials.     

Beam plasma neutron sources based on beam-driven mirror

The design and performance of a relatively low-cost, plasma-based, 14-MeV D-T neutron source for accelerated end-of-life testing of fusion reactor materials are described in this article. An intense flux (up to 5×10¹⁸ n/m²·s) of 14-MeV neutrons is produced in a fully-ionized high-density tritium target (n _e 3×10²¹ m^–3) by injecting a current of 150-keV deuterium atoms. The tritium plasma target and the energetic D⁺ density produced by D⁰ injection are confined in a column of diameter 0.16 m by a linear magnet set, which provides magnetic fields up to 12 T. Energy deposited by transverse injection of neutral beams at the midpoint of the column is conducted along the plasma column to the end regions. Longitudinal plasma pressure in the column is balanced by neutral gas pressure in the end tanks. The target plasma temperature is about 200 eV at the beam-injection position and falls to 5 eV or less in the end region. Ions reach the walls with energies below the sputtering threshold, and the wall temperature is maintained below 740 K by conventional cooling technology.     

Normal conducting steady-state toroidal magnet systems for ignited tokamaks

Normal conducting steady-state toroidal magnet systems are investigated, emphasis being placed on applications to large ignited next generation tokamaks. The study is based on water-cooled tape wound D coils. The data for the TF magnet systems are calculated in a consistent manner with a computer program including the plasma, shield, ohmic heating coil system, and geometric requirements for blanket modules, beam ducts, etc. An optimization procedure is used to find those TF coil systems which minimize cost-relevant quantities. The main results are that normal conducting TF coil systems of ignited next-generation tokamaks (following JET, TFTR, etc.) can be operated in a stationary mode and fed from the grid. Cost for electricity is a relatively small portion of the investment cost even in the case of long integral burn times (>10⁷ s).     

Impurity Transport in a Simulated Gas Target Divertor

Future generation fusion reactors and tokamaks will require dissipative divertors to handle the high particle and heat loads leaving the core plasma (100–400 MW/m² in ITER). A radiative divertor is proposed as a possible scenario, utilizing a hydrogen target gas to disperse the plasma momentum and trace impurity radiation to dissipate the plasma heat flux. Introducing an impurity into the target hydrogen gas enhances the radiative power loss but may lead to a significant impurity backflow to the main plasma. Thus, impurity flow control represents a crucial design concern. Such impurity flows are studied experimentally in this thesis. The PISCES-A linear plasma device (n 3 × 10¹⁹ m^–3, kT _e 20 eV) has been used to simulate a gas target divertor. To study the transport of impurities, a trace amount of impurity gas (i.e., neon and argon) is puffed near the target plate along with the hydrogen gas. Varying the hydrogen gas puffing rate permits us to study the effects of various background plasma conditions on the transport of impurities. A 1-1/2-D fluid code has been developed to solve the continuity and momentum equations for a neutral and singly ionized impurity in a hydrogen background plasma. The results indicate an axial reduction in the impurity concentration upstream from the impurity puffing source. Impurity entrainment is more effective for higher hydrogen target pressures (and for higher hydrogen plasma densities). However, if there is a reversal of the background plasma flow, impurity particles can propagate past the plasma flow reversal point and are then no longer entrained.     

Thermal effects and component cooling in Magnum-PSI

Magnum-PSI is a linear plasma generator, built at the FOM-Institute for Plasma Physics Rijnhuizen. Subject of study will be the interaction of plasma with a diversity of surface materials. The machine is designed to provide an environment with a steady state high-flux plasma (up to 10²⁴ H⁺ ions/m² s) in a 3 T magnetic field with an exposed surface of 80 cm² up to 10 MW/m². Magnum-PSI will provide new insights in the complex physics and chemistry that will occur in the divertor region of the future experimental fusion reactor ITER and reactors beyond ITER. The conditions at the surface of the sample can be varied over a wide range, such as plasma temperature, beam diameter, particle flux, inclination angle of the target, background pressure and magnetic field. An important subject of attention in the design of the machine was thermal effects originating in the excess heat and gas flow from the plasma source and radiation from the target.     

Experimental results of the erosion of molybdenum due to blistering by He+ bombardment

We report some preliminary measurement of the erosion rate of plasma dumps when bombarded with 100 keV He⁺ ions at high power density ( 1 MW/m²). The expected erosion rates, based on measurements of He blistering that were made at lower power density ( 0.3 MW/m²), indicate a potentially serious problem for fusion reactors. Our tests use a reactorlike power density and produce He blisters at a rate that is slower than predicted by 2 to 4 orders of magnitude, depending on the temperature of the molybdenum target.     

Characterization of Short Intense Pulsed Electrothermal Plasma Capillaries for Use as Fusion and Launchers Heat Flux Sources

Electrothermal plasma sources operating in the confined capillary arc regime are characterized by the magnitude and shape of the discharge current. The desired plasma parameters at the source exit, especially the pressure and heat flux, are highly dependent on the arc due to the effect of the arc radiant energy that ablates the inner wall of the source. These sources have applications in fusion as drivers for pellet injectors and as high heat flux sources for fusion materials studies. The high-pressure high heat flux flow is also of application in mass accelerators and launch technology systems. The 1-D, time-dependent ETFLOW capillary code models the plasma generation and flow inside the capillary discharges and determines the plasma parameters. The input file to the code is the discharge current density providing the Joule heating in the energy equation. A circuit module has been developed and incorporated in the code to generate desired current shapes and magnitudes. The current pulse length was varied between 5 and 100 μs at constant amplitude of 50 kA, and then the pulse amplitude was varied between 10 and 200 kA at a constant pulse length of 20 μs. Increasing the pulse length while maintaining its amplitude increases the plasma density and the total ablated mass, which have accumulation behavior by increasing the pulse length, and subsequently increases the exit pressure from 60 to 410 MPa in the cases studied herein. The pressure increase allows the thermalization of the plasma particles through more collisions, which reduces the plasma temperature by about 0.2 eV. The bulk velocity follows the trend of the plasma temperature, but at shorter pulse lengths the total ablated mass is lower and enables the plasma to carry the particles with increasing velocity. Increasing the pulse amplitude up to 200 kA increases the density to about 18 kg/m³ and the bulk velocity, which varies between 6.1 and 10.7 km/s. A sharp increase in most plasma parameters occurs as a result of the increase in the pulse amplitude.     

The high-density Z-pinch as a pulsed fusion neutron source for fusion nuclear technology and materials testing

The dense Z-pinch (DZP) is one of the earliest and simplest plasma heating and confinement schemes. Recent experimental advances based on plasma initiation from hair-like (10s m in radius) solid hydrogen filaments have so far not encountered the usually devastating MHD instabilities that plagued early DZP experimenters. These encouraging results along with the debut of a number of proof-of principle, high-current (1–2 MA in 10–100 ns) experiments have prompted consideration of the DZP as a pulsed source of DT fusion neutrons of sufficient strength (S _N 10¹⁹ n/s) to provide uncollided neutron fluxes in excess ofI _w = 5–10 MW/m² over test volumes of 10–30 liters or greater. While this neutron source would be pulsed (100s ns pulse widths, 10–100 Hz pulse rate), giving flux time compressions in the range 10⁵–10⁶, its simplicity, near-term feasibility, low cost, high-Q operation, and relevance to fusion systems thatmay provide a pulsed commercial end-product, e.g., inertial confinement or the DZP itself, together create the impetus for preliminary consideration as a neutron source for fusion nuclear technology and materials testings. The results of a preliminary parametric systems study (focusing primarily on physics issues), conceptual design, and cost vs. performance analyses are presented. The DZP promises an inexpensive and efficient means to provide pulsed DT neutrons at an average rate in excess of 10¹⁹ n/s, with neutron currents I_w10 MW/m² over volumes V_exp 30 liter using single-pulse technologies that differ little from those being used in present-day experiments.Work supported by U.S. DOE.     

Initial Exploration of High-Field Pulsed Stellarator Approach to Ignition Experiments

In the framework of fusion energy research based on magnetic confinement, pulsed high-field tokamaks such as Alcator and FTU have made significant scientific contributions, while several others have been designed to reach ignition, but not built yet (IGNITOR, FIRE). Equivalent stellarator concepts, however, have barely been explored. The present study aims at filling this gap by: (1) performing an initial exploration of parameters relevant to ignition and of the difficulties for a high-field stellarator approach, and, (2) proposing a preliminary high-field stellarator concept for physics studies of burning plasmas and, possibly, ignition. To minimize costs, the device is pulsed, adopts resistive coils and has no blankets. Scaling laws are used to estimate the minimum field needed for ignition, fusion power and other plasma parameters. Analytical expressions and finite-element calculations are used to estimate approximate heat loads on the divertors, coil power consumption, and mechanical stresses as functions of the plasma volume, under wide-ranging parameters. Based on these studies, and on assumptions on the enhancement-factor of the energy confinement time and the achievable plasma beta, it is estimated that a stellarator of magnetic field B?~?10 T and 30 m³ plasma volume could approach or reach ignition, without encountering unsurmountable thermal or mechanical difficulties. The preliminary conceptual device is characterised by massive copper coils of variable cross-section, detachable periods, and a lithium wall and divertor.     

Preparation and properties of CVD-W coated W/Cu FGM mock-ups

Tungsten was coated on a W/Cu functionally graded material (FGM) by chemical vapor deposition technique (CVD), and then the tungsten coated tile was brazed on the CuCrZr heat sink with a cooling channel. The thickness of CVD-W was 2 mm deposited by a fast rate of about 0.7 mm/h. The features of the CVD-W coating including morphology, element composition and thermal properties were characterized. A tungsten coating with high density, purity and thermal conductivity is achieved. The bonding strength between the CVD-W layer and FGM was measured using bonding tensile tests. Thermal screening and fatigue tests were performed on the CVD-W mock-ups under fusion relevant conditions using an electron beam device. Experimental results showed that the CVD-W mock-up failed by melting of Cu beneath the tungsten layer under a high heat load of 14.5 MW/m² and 30 s pulse duration. Thermal fatigue tests showed that the CVD-W mock-up could sustain 1000 cycles at a heat load of 11.7 MW/m² absorbed power density and 15 s pulse duration without visible failure.     

Heat flux capabilities of first-wall tube arrays for an experimental fusion reactor

A preliminary design study has been made of some of the thermomechanical problems of water and helium cooling for the first wall of a near-term experimental fusion reactor. The first wall is envisioned as an array of 316 stainless steel tubes between the plasma and the blanket modules to intercept a heat flux from the plasma estimated to be between 0.25 and 1.0 MW/m². Evaluations have been made of the maximum allowable heat fluxes for constraints imposed on the tube wall temperature, the cyclic stresses, the quasi-steady stresses and energy recovery from the coolant. For tubes with 2 meter long heated sections, 10 mm inside diameter and 1 mm wall thickness, water cooling was found to be more than adequate for plasma heat fluxes over 1 MW/m² with a fatigue life of 10⁶ cycles; for a 2 mm wall thickness, at least 0.7 MW/m² can be handled for the same fatigue life. Helium-cooled tubes can also handle heat fluxes up to about 1 MW/m² with a 1 mm tube wall thickness and over 0.5 MW/m² with a 2 mm tube wall thickness, but the required pumping powers tend to be high. The problems of plasma disruptions and erosion by energetic plasma ions are also discussed briefly.     

Removing of Mixed Coatings by Plasma Discharges

Next generation tokamaks offer the possibility of highly efficient energy generation from the fusion reaction of hydrogen isotopes. In tokamak operation, the core plasma interaction with the wall materials could produce tiles erosion. Redeposition of the eroded materials (C–W–Be) leads to an increase in the allowable tritium load if the coatings are not periodically removed. Amongst removal methods, plasma based techniques employing Ar, H₂ gas have been investigated. Plasma cleaning has been carried out on hydrogenated carbon and carbon–tungsten coatings. It has been shown that at a RF power density of 1.3 W/cm² (pressure of 1 Pa), the plasma cleaning was effective in removing the coatings. Details of further work in this research activity will be presented.     

Deuterium retention in plasma spray tungsten coatings exposed to low-energy, high flux D plasma

Two types of porous plasma spray tungsten coatings deposited onto stainless steel and graphite substrates were exposed to low-energy (76 eV ), high-flux (10²² D/m² s) D plasma to ion fluences of (3-4) × 10²⁶ D/m² at various temperatures. Deuterium retention in the W coatings was examined by thermal desorption spectroscopy and the D(³He,p)⁴He nuclear reaction, allowing determination of the D concentration at depths up to 7 μm. The relatively high D concentration (above 0.1 at.%) at depths of several micrometers observed after D plasma exposure at 340-560 K can be related to accumulation of D₂ molecules in pores, while at temperatures above 600 K deuterium is accumulated mainly in the form of D atoms chemisorbed on the inner pore surfaces. At exposure temperatures above 500 K, the D retention in the plasma spray W coating on graphite substrate increases significantly due to trapping of diffusing D atoms at carbon dangling bonds located at the edge of a graphite crystallite.     

The impact of thermal fatigue and other design limits on pulsed commercial tokamak reactor design

Methods of analysis for fusion first-wall design are developed. Several design limits have been evaluated and combined to present tradeoffs in the form of design windows. These considerations include limits related to thermal fatigue, primary membrane strength, displacement under loading, ratcheting, radiation damage, and plasma-wall interactions. Special emphasis is placed on the investigation of thermal fatigue using a two-dimensional treatment of a tubular first-wall configuration. The work is motivated by the proposal of the Ultra Long Pulse Commercial Reactor (ULTR), a machine capable of delivering plasma burn pulses of up to 24 hr in length. The present work looks in detail at the impact of pertinent characteristics of the first-wall design, such as pulse length, coolant pressure, first-wall thickness, and first-wall lifetime on the structural effects considered. Computer programs are developed and are used to consider several major structural effects on a cylindrical first-wall element for both 316 stainless steel and vanadium alloy. Results indicate that short pulse lengths (greater than a few minutes) can be tolerated in tokamak operation. For stainless steel this is true for heat depositions up to 1 MW/m², while vanadium can tolerate heat depositions as high as 2 MW/m². Long pulse operation can be used to increase modestly the allowable heat deposition or to increase useful wall thickness by 1–2 mm. It appears that irradiation swelling and embrittlement, not fatigue, ultimately limits the first-wall design.     

Saturation in deuterium retention of CFC graphite targets exposed to beryllium-seeded plasmas on PISCES-B

ITER strike-plates are foreseen to be of carbon-fiber-composite (CFC). In this study the CFC bulk deuterium retention in ITER-relevant conditions is investigated. DMS 701 (Dunlop) CFC targets were exposed to plasma in PISCES-B divertor plasma simulator. Samples were exposed to both pure deuterium plasma and beryllium-seeded plasma at high fluences (up to ) and high surface temperature (1070 K). The deuterium contents of the exposed samples have been measured using both thermal-desorption-spectrometry (TDS) during baking at 1400 K and ion beam nuclear reaction analysis (NRA). The total deuterium inventory has been obtained from TDS while NRA measured the deuterium depth distribution. In the analysed fluence range at target temperature of 1070 K, no fluence dependence was observed. The measured released deuterium is . In the case of target exposure with beryllium-seeded plasma no change in the released amount of deuterium was found. The deuterium concentration inside the samples is almost constant until the probed depth of ?m, except in the first 1 μm surface layer, where it is 5 times higher than in the bulk. No C erosion/redeposition was observed in the Be-seeded plasma cases. The measured retention, applied to 50 m² of ITER CFC surface, would imply a tritium saturated value of 0.3 gT, much lower than the ITER safety limit of 350 g.     

D-T experiments on TFTR

Temperatures, densities and confinement of deuterium plasmas confined in tokamaks have been achieved within the last decade that are approaching those required for a D-T reactor. As a result, the unique phenomena present in a D-T reactor plasma (D-T plasma confinement, alpha confinement, alpha heating and possible alpha driven instabilities) can now be studied in the laboratory. Recent experiments on the Tokamak Fusion Test Reactor (TFTR) have been the first magnetic fusion experiments to study plasmas with reactor fuel concentrations of tritium. The injection of 20 MW of tritium and 14 MW of deuterium neutral beams into the TFTR produced a plasma with a T/D density ratio of 1 and yielded a maximum fusion power of 9.2 MW. The fusion power density in the core of the plasma was 1.8 MW m^–3 approximating that expected in a D-T fusion reactor. In other experiments TFTR has produced 6.4 MJ of fusion energy in one pulse satisfying the original 1976 goal of producing 1 to 10 MJ of fusion energy per pulse. A TFTR plasma with T/D density ratio of 1 was found to have 20% higher energy confinement time than a comparable D plasma, indicating a confinement scaling with average ion mass, A, of _E. The core ion temperature increased from 30 keV to 37 keV due to a 35% improvement of ion thermal conductivity. Using the electron thermal conductivity from a comparable deuterium plasma, about 50% of the electron temperature increase from 9 keV to 10.6 keV can be attributed to electron heating by the alpha particles. At fusion power levels of 7.5 MW, fluctuations at the Toroidal Alfvén Eigenmode frequency were observed by the fluctuation diagnostics. However, no additional alpha loss due to the fluctuations was observed. These D-T experiments will continue over a broader range of parameters and higher power levels.Work supported by U.S. Department of Energy Contract No. DE-AC02-76-CHO-3073.     

Plasma MHD generator for modelling energy conversion in pulsed thermonuclear reactors

Conclsuions The construction of an experimental model for studying MHD energy conversion from a pulsed thermonuclear reactor is a realistic technical task at the present time. Doing this would permit development of a large scale MHD generator module for the typical parameters of the heated working medium in a pulsed thermonuclear reactor.In principle it is possible to obtain an efficiency of at least about 40% with a linear plasma MHD generator. The efficiency of the whole plant might be increased further by utilization of the thermal energy at the outlet of the MHD channel in traditional methods.When such an MHD generator is built difficulties with the behavior of supersonic plasma streams undergoing strong velocity reduction in a channel and the associated gasdynamic problems can clearly be solved successfully by active modification of the boundary layer and appropriate profiling of the MHD channel. Some complications may arise if a regime with time varying magnetic braking is used. Also important is the problem of the behavior of the plasma stream at large magnetic Reynolds numbers (Re_m1).The basic technological problems are these: materials for the MHD channel, cooling arrangements for the channel (especially the critical cross section of the flow path), and pumping off the boundary layer at the electrodes and preventing lithium condensation on the channel walls. Because of the small magnetic field required, construction of the magnet system will clearly not present substantial technical difficulties associated with its size.The most important physical questions as well as a number of technological questions characteristic of this problem may be investigated on a fairly simple model MHD generator with an output power level of 300–500 MW, a pulse duration of 10–20 msec, and a lithium plasma source.Translated from Atomnaya Énergiya, Vol. 39, No. 6, pp. 387–391, December, 1975.