Irradiation effects on CTR first wall materials

The selection of the controlled thermonuclear reactor (CTR) first wall material (refractory metallic alloy or austenitic stainless steel) will be a compromise based on a number of important nuclear, physical, thermal and mechanical properties, and safety, environmental, technological and economic factors. The correlations between helium production, irradiation, and helium embrittlement in the first wall depend mainly on the neutron spectrum, neutron fluence, irradiation temperature, wall material, reactor operating conditions, etc. The suppression of irradiation swelling by alloying elements (C, Si, Mo, and P) is effective and beneficial. The extent of change in mechanical properties of the first wall material due to neutron irradiation, thermal instabilities (thermal shock, thermal fatigue, crack initiation and propagation), corrosion effect, etc. remains undefined or not completely understood. A fabricated CTR first wall material must meet the requirements of design safety, weldment reliability and good operation performance.     

Radiation facilities for fusion reactor first wall and blanket structural materials development

Present and future irradiation facilities for the study of fusion reactor irradiation damage are reviewed. Present studies are centered on irradiation in accelerator-based neutron sources, fast- and mixed-spectrum fission reactors, and ion accelerators. The accelerator-based neutron sources are used to demonstrate damage equivalence between high-energy neutrons and fission reactor neutrons. Once equivalence is demonstrated, the large volume of test space available in fission reactors can be used to study displacement damage, and in some instances, the effects of high-helium concentrations and the interaction of displacement damage and helium on properties. Ion bombardment can be used to study the mechanisms of damage evolution and the interaction of displacement damage and helium. These techniques are reviewed, and typical results obtained from such studies are examined. Finally, future techniques and facilities for developing damage levels that more closely approach those expected in an operating fusion reactor are discussed.     

Depth profiling a near surface carbon contaminant in implanted first wall materials

Current attempts to generate stable plasmas in CTR devices are encountering severe material problems. Plasma species, interacting with the surface of the containment vessel, can desorb surface species which can contaminate the plasma or, interacting with surface impurities, can change the properties of the near surface region. Although methods to minimize these synergistic effects have been suggested, there exists only minimal information related to the fundamental surface and near surface (< 100 Å) processes involved. One process of interest, the interaction of energetic plasma species with adsorbed hydrocarbon contaminants has been examined using Imaging, Field-Desorption Mass Spectrometry. Angstrom resolved depth profiles of a carbon contaminant produced by ion-beam decomposition of an adsorbed surface hydrocarbon layer have been measured for tungsten, molybdenum and stainless steel specimens. The observed, 30 Angstrom penetration of carbon into the near surface region suggests that conventional first-wall cleaning techniques may be ineffective in completely removing this impurity.     

High temperature carbon implantation in SIMOX

The aim of this experiment was to explore the possibility to convert the Si-overlayer of a SIMOX wafer into 3C-SiC by carbon implantation. In a first attempt carbon was implanted at a temperature 1030°C and energy 100 keV to a dose of 2.5 × 10¹⁷ C⁺ cm⁻². The SIMOX was covered by a thick thermal oxide. Cross-section TEM observations on the implanted specimen reveal that carbon is concentrated mainly at the Si/SiO₂ interfaces at the front and back face of the Si-overlayer forming continuous but highly defected 3C-SiC layers which are in epitaxial relation with the Si matrix. The implanted carbon has the tendency to migrate from the SiO₂ and Si to the SiO₂/Si interfaces to form SiC there.     

On the radiation damage characterization of candidate first wall materials in a fusion reactor using various molten salts

Evaluating radiation damage characteristics of structural materials considered to be used in fusion reactors is very crucial. In fusion reactors, the highest material damage occurs in the first wall because it will be exposed to the highest neutron, gamma ray and charged particle currents produced in the fusion chamber. This damage reduces the lifetime of the first wall material and leads to frequent replacement of this material during the reactor operation period. In order to decrease operational cost of a fusion reactor, lifetime of the first wall material should be extended to reactor’s lifetime. Using a protective flowing liquid wall between the plasma and first wall can decrease the radiation damage on first wall and extend its lifetime to the reactor’s lifetime. In this study, radiation damage characterization of various low activation materials used as first wall material in a magnetic fusion reactor blanket using a liquid wall was made. Various coolants (Flibe, Flibe + 4% mol ThF₄, Flibe + 8% mol ThF₄, Li₂₀Sn₈₀) were used to investigate their effect on the radiation damage of first wall materials. Calculations were carried out by using the code Scale4.3 to solve Boltzmann neutron transport equation. Numerical results brought out that the ferritic steel with Flibe based coolants showed the best performance with respect to radiation damage.     

聚变驱动次临界堆第一壁材料辐照损伤的初步研究

介绍了中子对材料的辐照损伤原理及化合物原子平均离位(DPA)截面计算方法;使用辐照损伤计算程序SPECTER计算了聚变驱动次临界堆(FDS-I)第一壁材料CLAM钢的辐照损伤参数,并将CLAM钢的辐照损伤计算结果与相同条件下316SS、SiC等聚变堆结构材料的计算结果进行了比较.     

High temperature silver ion implantation into glass

Soda-lime-silicate glass has been implanted with 55 keV Ag⁺ ions at five different temperatures: room temperature, 100, 225, 350 and 600°C. Rutherford backscattering spectroscopy was used to provide depth profiles for the implants. All samples show a low retention of silver and this varies with temperature. In the room temperature and 100°C implants the silver diffuses to, and is lost from the surface. This also occurs in the higher temperature implants, but in these cases there is a significant amount of inward diffusion by the silver. This diffusion extends to at least 500 nm: greatly in excess of the predicted range for a standard implant. Optical measurements on the samples show that those with the inward diffusion have formed an enhanced-index waveguide.     

Thermomechanical effects in a laser IFE first wall

Laser fusion chamber walls will experience large, pulsed heat loads at frequencies of several hertz. The heating, consisting of X-rays, neutrons, and ions, occurs over a few microseconds and is deposited volumetrically over the first few microns of the wall. For a reasonable chamber radius, the heating will be such that the surface temperature is a significant fraction of the melt temperature of the wall, and significant plasticity can be expected in ductile wall materials. This paper presents results for the transient temperatures and stresses in a tungsten-coated steel first wall for a laser fusion device. Failure analyses are carried out using both fatigue and fracture mechanics methodologies. The simulations predict that surface cracks are expected in the tungsten, but the cracks will arrest before reaching the substrate if the crack spacing is sufficiently small. In addition, the thermal and stress fields are compared for a laser fusion device with several simulation experiments. It is shown that the simulations can reproduce the peak surface temperatures, but the corresponding spatial distributions of the stress and temperature will be shallower than the reactor case.     

Safety study of primary coolants in fusion power reactor first wall design

This paper examines potential safety problems associated with the various primary coolant candidates currently considered for the EPR fusion blanket designs. The basic concern is the possibility of overheating and melting of the first wall and the blanket, induced by a malfunction in the primary coolant system. These accidents include the loss-of-coolant flow, the loss-of-heat removal, overpower transients, and the loss of coolant. Following a mechanistic safety for these four types of accident sequences and comparing helium and liquid metal cooling, it was found that helium has a more adverse effect on the first-wall heat up in the event of a loss-of-heat removal or a loss-of-coolant because its lack of thermal inertia.     

Electrical and optical surface degradation of silica due to superficial He implantation

Different oxides will be used in ITER and future fusion reactors for electrical insulation and optical components. The vacuum face of these materials will be subjected not only to neutron and gamma irradiation, but also to particle bombardment, due mainly to ionization of the residual gas and acceleration of the resulting ions by local electric fields. Previous work showed that silica suffers electrical and optical degradation when subjected to He bombardment with energies from 300 keV down to 27 keV. As the He ion energy may extend down to some few keV, or less, further work has been performed to study possible degradation for energies from 21 keV down to 5 keV. The results show that both surface and optical degradation occur at these low energies, more rapidly for the lowest energy (5 keV) ions. They also suggest that the superficial narrow implanted He profile plays an important role in the surface degradation.     

High temperature high dose C ion implantation in epitaxial SiGe

Si_{1 − x}Ge_x epitaxial layers fully strained (x = 0.27) and relaxed (x = 0.55) have been implanted with C ions at 500°C. Implantation energy and doses were selected to obtain the C peak in the central region of the SiGe layer, with a concentration similar to the Ge content. The implanted layers have been analyzed by Raman scattering, X-ray diffraction, transmission electron microscopy and secondary ion mass spectroscopy. The data obtained show the direct synthesis of β-SiC precipitates aligned in relation to the SiGe lattice after the implantation, as well as a Ge enrichment and stress relaxation of the SiGe lattice. For the relaxed layer a significant Ge redistribution from the implanted region is observed.     

Thermohydraulic analysis of high-Prandtl-number fluid in complex duct simulating first wall in fusion reactor

For fusion reactors, molten salt is one of the candidates for coolant materials. Molten salt is a high-Prandtl-number fluid; thus, it is necessary to enhance the heat transfer coefficient. It is proposed that rods are inserted into a duct to enhance the heat transfer coefficient. The flow field behind the rod in the duct is visualized to compare experimental data with simulation results. The trends and distributions in the numerical simulation are the same as those in the experiment, and furthermore, the magnitudes of the time and space scales in the numerical simulation are of the same order as those in the experiment. Thermohydraulic numerical analysis confirmed that the heat transfer coefficient is improved by inserting the rod when the fluid is a high-Prandtl-number fluid and the flow field is in the turbulent region. However, it is necessary for the rods to be arranged in the streamwise direction.     

High temperature physical properties of fast reactor materials

Physical properties of materials of interest in fast reactor safety are presented. These include enthalpy and heat capacity, vapor pressure, density, surface tension, speed of sound, viscosity, and thermal conductivity. The emphasis is on saturated fuel materials and coolant. Much of the data is the result of direct experimental measurement while the remainder was obtained by extrapolative techniques from data at lower temperatures.     

Fatigue cracking of a bare steel first wall in an inertial confinement fusion chamber

Inertial confinement fusion power plants will deposit high energy X-rays onto the outer surfaces of the first wall many times a second for the lifetime of the plant. These X-rays create brief temperature spikes in the first few microns of the wall, which cause an associated highly compressive stress response on the surface of the material. The periodicity of this stress pulse is a concern due to the possibility of fatigue cracking of the wall. We have used finite element analyses to simulate the conditions present on the first wall in order to evaluate the driving force of crack propagation on fusion-facing surface cracks.Analysis results indicate that the X-ray induced plastic compressive stress creates a region of residual tension on the surface between pulses. This tension film will likely result in surface cracking upon repeated cycling. Additionally, the compressive pulse may induce plasticity ahead of the crack tip, leaving residual tension in its wake. However, the stress amplitude decreases dramatically for depths greater than 80–100 μm into the fusion-facing surface. Crack propagation models as well as stress-life estimates agree that even though small cracks may form on the surface of the wall, they are unlikely to propagate further than 100 μm without assistance from creep or grain erosion phenomena.     

First wall fusion blanket temperature variation — slab geometry

The first wall of a fusion blanket is approximated by a slab, with the surface facing the plasma subjected to an applied heat flux, while the rear surface is convectively cooled. The relevant parameters affecting the heat transfer during the early phases of heating as well as for large times are established. Analytical solutions for the temperature variation with time and space are derived. Numerical calculations for an aluminum and stainless steel slab are performed for a wall loading of 1 MW(th)/m². Both helium and water cooling are considered.     

Micro-engineered first wall tungsten armor for high average power laser fusion energy systems

The high average power laser program is developing an inertial fusion energy demonstration power reactor with a solid first wall chamber. The first wall (FW) will be subject to high energy density radiation and high doses of high energy helium implantation. Tungsten has been identified as the candidate material for a FW armor. The fundamental concern is long term thermo-mechanical survivability of the armor against the effects of high temperature pulsed operation and exfoliation due to the retention of implanted helium. Even if a solid tungsten armor coating would survive the high temperature cyclic operation with minimal failure, the high helium implantation and retention would result in unacceptable material loss rates. Micro-engineered materials, such as castellated structures, plasma sprayed nano-porous coatings and refractory foams are suggested as a first wall armor material to address these fundamental concerns. A micro-engineered FW armor would have to be designed with specific geometric features that tolerate high cyclic heating loads and recycle most of the implanted helium without any significant failure. Micro-engineered materials are briefly reviewed. In particular, plasma-sprayed nano-porous tungsten and tungsten foams are assessed for their potential to accommodate inertial fusion specific loads. Tests show that nano-porous plasma spray coatings can be manufactured with high permeability to helium gas, while retaining relatively high thermal conductivities. Tungsten foams where shown to be able to overcome thermo-mechanical loads by cell rotation and deformation. Helium implantation tests have shown, that pulsed implantation and heating releases significant levels of implanted helium. Helium implantation and release from tungsten was modeled using an expanded kinetic rate theory, to include the effects of pulsed implantations and thermal cycles. Although, significant challenges remain micro-engineered materials are shown to constitute potential candidate FW armor materials.     

Investigation of some properties of technical materials envisaged for the first wall of textor

First results of a programme started in the KFA Jülich to investigate through simulation experiments possible mechanisms responsible for the early appearance of impurities in Tokamak discharges are described. They stress the dominating influence of manufacturing and prehandling procedures on the surface properties and vacuum behaviour of technical materials frequently used and often considered for the first wall of large experimental devices. In addition, measurements of the tritium retention in carbon indicate that particular attention should be paid to the possible role of such reactions on the tritium inventory when low-Z materials are envisaged for the first wall.     

Growth of carbon impurity density and adequate divertor wall temperature for fusion experimental reactors

Sudden increase of carbon impurity called carbon bloom has terminated the energy breakeven condition in the present large tokamak. In order to lengthen the burning plasma state in the next device, carbon bloom has to be well suppressed. The temporal evolution of carbon impurity density is analytically examined by using a simple one-point kinetic or zero-dimensional model including the effects of graphite erosions due to oxygen and ion, and gettering for oxygen due to boron or beryllium. The growth of carbon bloom due to radiation-enhanced sublimation is discussed based on the effective self-sputtering of carbon. Even when the self-sputtering yield is less than unity, carbon density is observed to continuously increase with the discharge time if the oxygen gettering action is not perfectly conducted. From the present analysis and data on the erosion of carbon materials, and the evaporation of gettering materials, it is suggested that the divertor wall temperature has to be kept less than approximately 900–1000°C to avoid the continuous growth of the carbon density.     

Vacancy-type defects near Al surface studied by slow positron annihilation spectroscopy before and after He implantation

The vacancy-type defects He_nV_m near Al surface before and after He⁺ implantation and their evolutions with annealing temperatures and aging time have been investigated by mono-energy slow positron annihilation spectroscopy (SPAS) with S parameters. The results show that many vacancies are produced during the sample preparation process, which can be re-occupied by Al atoms during annealing, Al⁺ and MeV He⁺ implantation. S parameters denote the concentration and size of He_nV_m clusters induced by He⁺ implantation in Al. The higher fluence of He implanted, the larger S parameters will be, indicating more He_nV_m clusters produced. S parameters decrease with the increase of annealing temperatures until the fastest change temperature, and then an opposite or minor change occurs depending on the fluence of He implanted in Al, showing that the concentration and size of He_nV_m clusters will vary with the annealing temperatures. Aged at RT for some time, the concentration and mean size of He_nV_m clusters in Al will get smaller and larger, respectively, resulting in the decrease of S parameters with the aging time. In conclusions, the evolution of vacancy-type defects He_nV_m near Al surface after He⁺ implantation depends on the annealing temperatures, He concentration and aging time.