Wall Material Properties
The main design concern related to the thermal power exhaust are the high power loads on divertor plates and the divertor material should withstand such loads preferably for the entire integrated operation time and be able to efficiently transmit the received power to the background structures. Therefore, these materials should have adequate thermo-mechanical properties (high thermal conductivity, low thermal expansion coefficient, good thermal shock resistance, high yield strength and toughness, low elastic moduli, long fatigue lifetime) and thermo-physical characteristics(high melting point, high evaporation heat, low erosion and low tritium inventory)
Material Properties of Beryllium
Based on an article by Dombrowski et al. 
Beryllium is foreseen to be used as a cladding for plasma facing components (PFC) and as a neutron multiplier in the tritium breeding blanket of a fusion reactor. The principal advantages which Be has in PFC applications is its ability to get oxygen and its low atomic number. These properties have been shown to eliminate density limit disruptions and to increase the time scale for magneto hydrodynamic instabilities. In the past, beryllium was used to absorb heat without active cooling. In future experiments actively cooled beryllium clad heat sinks or directly cooled beryllium components will be required to operate under steady-state conditions. Beryllium is the material of choice at the neutron multiplier through (n,2n) reactions in a solid breeder blanket design. The solubility and retention of hydrogen isotopes in beryllium both for PFC and blanket applications under normal and irradiated conditions should be understood.
|Melting Temperature||1289 °C|
|Boiling point at 1 atm||2470 °C|
|Heat of Fusion||1132 KJ/Kg|
|Heat of Vaporisation||24770 KJ/Kg|
|Thermal conductivity||204 W/m/K|
|Coefficient of thermal expansion||11.5<math>\times</math>10-6 /K|
|Specific heat||1.85 KW-sec/Kg/K|
|Viscosity at 1556 K||<math>\simeq</math>10-3 poises|
|Youngs modulus||<math>\simeq</math> 311 GPa|
|0.2% Yield Strength||<math>\simeq</math> 270 MPa|
Beryllium is an element with atomic number 4 and atomic weight of 9.012. It is a metal with hexagonal close packed structure. Beryllium in structural applications is notable for a combination of stiffness and low density. Excellent reflectivity for infrared radiation is observed a and beryllium is used for neutron reflector s and neutron multiplier due to its high neutron scattering cross-section, low thermal neutron capture cross-section and high (n,2n) neutron multiplication cross-section. It has a strong chemical affinity for oxygen and hence has a protective oxide layer. Its low specific modulus makes the distortion of components made out of beryllium minimized with respect to applied stress. The combination of high specific heat, high thermal conductivity, low coefficient of thermal expansion with the structural properties makes beryllium components dimensionally stable under transient thermal loads.
Beryllium, however, poses a health risk if mishandled. In its usual solid form as well as for finished parts and in most manufacturing operations, it is completely safe. However, breathing very fine particles may cause serious lung condition in a small percentage of individuals. Risk can be minimised with simple, proven and readily available engineering controls such as ventilation of operations producing fine dust. JET has operated safely with beryllium in large quantities and under conditions producing fine particles.
Types Beryllium Material
The mechanical properites of beryllium vary significantly from grade to grade and also strongly dependent upon specific fabrication technology.
- Power Metallurgy of Beryllium
- Ingot Metallurgy
- Refining Processes
- Commercial Grades
- Specific Heat
- Vapor Pressure
- Thermal Conductivity
- Coefficient of Thermal Expansion
- Electrical Resistivity
- Youngs Modulus
- Shear Modulus
- Poisson’s Ratio
- Ultimate Tensile Strength
- Yield Strength
- Uniform Elongation
- Total Elongation
- Reduction of Area
- Fracture Toughness
Properties after Neutron Irradiation
Many of the thermophysical properties of beryllium vary not only from grade to grade but also change substantially due to neutron irradiation. For exmaple, yield strength, ultimate tensile strength and ductility are expected to be influenced substantially by radiation damage. The effects of neutraon irradiation on the thermomechanical properties of beryllium are difficult to summarize since the testing has been done for widely differing combinations of irradiation conditions (flux, fluence, temperature, time), post irradiation testing conditions (temperature, pre-test annealing) and beryllium grades.
Beryllium has a strong affinity for oxygen which is usually present in all forms of beryllium exposed to air in the form of BeO. In a tokamak edge plasma beryllium has been shown to getter oxygen strongly. Hydrogen does not normally dissolve to any great extent in beryllium metal. However, the redeposition of beryllium atoms from the edge plasma onto PFC surfaces results in a co-deposition of hydrogen with beryllium. For DT pulsed and steady state devices the amount of tritium atoms trapped this way are crucial in respect of tritium inventory and accountability. The flammability of beryllium, in particular porous beryllium under accident conditions where upon hot beryllium is exposed to air or a steam air mixture must be investigated. Reactivity of beryllium with liquid metals such as sodium, potassium and lithium is low below 500 °C provided that the oxygen concentration of metals is low. Beryllium is attacked by hydrofluoric acid, hydrochloric acid, and alkali hydroxide solutions. It has acceptable corrosion properties in high purity cooling water at 100 °C in the pH range 5.5 to 6.5. In addition to high reactivity with oxygen, beryllium will react with carbon based materials to form beryllium carbide at temperatures as low as 750 °C. Beryllium reacts with most refractory metals such as tungsten to form intermetallic compounds.
Material Properties of Carbon and Carbon Composites
Based on an article by Burchell and Oku 
Carbon fiber composites (CFCs) or carbon-carbon (C/C) composites are a class of carbon materials widely used in tokamak devices. Carbon-carbon composites are comprised of two components, carbon fibers and a carbon matrix. Generally, C/C composites used in plasma facing components (PFCs) are heat treated during their production to a sufficient temperature ( > 2500 °C) to ensure that both the fiber and matrix are graphitized and the material is therefore actually a graphite/graphite composite. Carbon-carbon composites are stronger and tougher than graphites, yet retain the excellent machinability exhibited by graphites. Carbon-carbon composite materials are typically described as being uni-directional (1D), two-directional (2D) or three-directional (3D) indicating the number of fiber bundle directions that a composite may possess.
As a plasma facing component, the temperature dependence of the thermal conductivity is very important to prediction of the surface temperature of plasma-facing components, and the amount of erosion. It is desirable that the C/C composite has higher thermal conductivity at high temperatures, such as 300 W/m•K at 1000 °C. Though vapour pressure of the C/C composite depends upon the degree of graphitization and the degree of graphitization of the C/C composite materials is smaller than that of graphites, the difference of vapour pressure between graphites and C/C is expected to be small. Recently high thermal conductivity C/C composite development has been pursued worldwode and improvements in fiber and composite technology are driving the thermal conductivity of C/C composites toward that of pyrolytic graphites.
Mechanical properties of C/C composites are required for evaluating stresses, e.g., thermal stresses or electron-magnetic mechanical stress, induced in the PFC during disruption of the plasma.
- Young’s Modulus
- Bending strength
- Compression strength
- Shore hardness
- Electrical resistance
- Specific heat
- Thermal conductivity
- Thermal expansion coefficient
- Ash content
Neutron Irradiation Damage Mechanism
The carbon atoms in carbon fibers and matrix carbon are arranged at the atomic level in the graphite crystal structure, consisting of layers of carbon atoms; the atoms being (sp2) covalently bound in a hexagonal array within the graphite layer. The layers are stacked in an ABAB sequence with only weak secondary inter-layers bonding. The binding energy of a carbon atom in the graphite lattice is about 7 eV and the energy required to displace a carbon atom is ~ 25 to 60 eV.
The D-T reaction produces both 14.1 MeV neutrons and 3.5 MeV alpha particles. The majority of the alpha particles will deposit their energy in the plasma, but the neutrons will reach the first wall and displace carbon atoms. Moreover, the neutron energies are sufficient that transmutations such as 12C(n,<math>\alpha</math>)9Be and 12C(n,n’)3<math>\alpha</math> will occur and cause additional carbon atom displacements. Displaced carbon atoms will recoil through the graphite lattice, displacing other carbon atoms and leaving vacant lattice sites. Once moderated, the carbon atoms are able to diffuse interstitially between the graphite layer planes in two dimensions and a large fraction of them will recombine with lattice vacancies. Others will coalesce and eventually may form dislocation loops or new graphite planes. The interstitial clusters, on further irradiation, may themselves be destroyed by a fast neutron or displaced carbon atom – phenomenon known as radiation annealing.
Graphite crystal neutron damage induced dimensional changes will cause dimensional changes in C/C composite materials. The nature of the dimensional changes in the composites will depend upon both the irradiation conditions (temperature, fluence) and the microstructure of the carbon fiber and architecture of the composite. The latter (structural) parameters are important because they control the amount and location of porosity. The number of interstitial atoms and vacancies depends upon the neutron flux and energy, and the irradiation temperature controls the number of carbon atoms retained in the crystal lattice, and their distribution, because lattice vacancy mobility is temperature dependent.
The formation of neutron irradiation induced defect structures in the graphite lattice will adversely affect C/C composite materials physical properties. Phonons will be scattered by the defects, thus introducing additional thermal resistance and significantly reducing the thermal conductivity of an irradiated material. Similarly, electrons will be scattered by irradiation induced defects causing increased electrical resistivity. Pinning of dislocation lines by irradiation induced defects will increase the strength and elastic moduli of C/C composite materials.
Irradiation Induced Dimensional Changes
The property degradation due to oxidation of nuclear graphites and carbons is of importance. The oxidation produces changes in the pore and micro structures of the materials: generally many new pores are generated in the structure due to oxidation and the bulk density decreases, that is the porosity increases. The oxidation rates of C/C composites were found to follow the Arrhenius type equation: k = A exp (-Ea/RT)
Graphite and C/C composites have very high specific surface areas, typically in the range 0.25-1 m2/g. Consequently, carbon materials can adsorb large quantities of gases, resulting in the possible retention of substantial amounts of hydrogen isotopes. This plays an important role in determining the recycling of fuel in the plasma-wall interface. Moreover, the adsorption of tritium may cause an unacceptably high tritium inventory for a fusion device during tritium operation. There are four processes by which tritium can be retained in or on graphites and C/C composites: saturation of the implanted area; codeposition with carbon on surfaces; adsorption on internal porosity and transgranular diffusion with trapping. Neutron irradiation has been shown to increase the amount of deuterium and tritium retained in graphite and C/C composites. Crystal lattice damage, in the form of interstitial loops, vacancies and vacancy loops provide tritium trapping sites.
Material Properties of High-Z Elements
Based on an article by T. Tanabe 
High Z materials such as Mo and W were once used as plasma facing materials (PFM) in early tokamaks because of their refractory properties for future use in DT burning machines as well as enough experiences as vacuum high temperature materials. Owing to their high radiation loss, however, the plasma at that time was difficult to be heated and often collapsed by the accumulation of high Z impurities in the plasma center. Instead of the high Z, introduction of carbon based low Z materials have advanced drastically in recent large tokamaks. Ironically, the advances of the plasma parameters result in extraordinary high heat load to the plasma facing materials. Hence, evaporation, erosion, and thermal stress (shock) of the PFM make the future usage of the low Z materials doubtful. In addition neutron irradiation is found to greatly deteriorate thermo-mechanical properties of the carbon based materials.
Thus, high Z materials come back in to the consideration. For high temperature use of high Z materials as PFM, there are two important issues; (1) thermal response and thermal shock resistance to extraordinary high heat load in a very short period, and (2) recrystallization or thermal annealing for long term. Most of the high Z materials are well known to be brittle at room temperature while they may tolerate the thermal stress if they are used above ductile-brittle transition temperature. Therefore the response to the thermal shock or high heat load test should be examined in terms of operating temperature and temperature gradient in the material.
Apart from the PFM use, high Z materials are also the candidate for heat sink or structure materials of plasma facing component (PFC), owing to their excellent thermal properties and capability of cooling. Especially their higher melting temperature and higher thermal conductivity compared with stainless steels, the standard structure material, make a higher heat load as well as higher temperature operation possible. For this application of the high Z, additional issues such as thermal stress due to large temperature gradient between front surface to cooling channel and corrosion by coolant should be taken into account.
Characteristics of high Z refractory metals
Because of the highest melting point, the lowest vapour pressure and evaporation rate, and the largest thermal conductivity, W is expected to be the best candidate as a PFM. However, the poorest ductility of W makes the utilization of W bulk very difficult. Thermal radiation and emissivity are also important factors for the use of high Z materials at elevated temperatures.
The high threshold energy of hydrogen sputtering owing to their high mass is one of the advantages of high Z materials, whereas the low threshold energy by self-sputtering could cause impurity accumulation in a plasma center by sputter run-away. Control of plasma edge temperature below the self-sputtering threshold is a critical issue for the application of the high Z materials as PFM.
Different from the physical sputtering, chemical sputtering or erosion has basically no energy threshold but depends on the temperature of a target material. Since Mo and W are easily oxidized to form volatile oxides, enhanced erosion by the chemical sputtering of oxygen (the main impurity in plasma) is one of the concerns. It is seen that the erosion of Mo and W by oxygen is enhanced at intermediated temperatures and high incident energies but does not exceed self-sputtering.
Interaction with hydrogen
Hydrogen adsorption on clean W surface is the one of the most extensively studied system in surface physics and the adsorption sites and states of hydrogen on a clean surface of W single crystals are rather well understood.
Diffusivity, permeability and solubility
Corrosion by coolant
Behavior of high Z materials in plasma and high heat load tests
Results of high heat load tests
Behavior of high Z materials in a plasma
- ↑ D. E. Dombrowski, E. B. Deksnis and M. A. Pick, Thermomechanical Properties of Beryllium , Atomic and Plasma-Material Interaction Data for Fusion, Vol. 5, P. 19 (1994)
- ↑ T. D. Burchell and T. Oku, Material properties data for fusion reactor plasma-facing carbon-carbon composites , Atomic and Plasma-Material Interaction Data for Fusion, Vol. 5, P. 77 (1994)
- ↑ T. Tanabe High-Z candidate plasma facing materials , Atomic and Plasma-Material Interaction Data for Fusion, Vol. 5, P. 129 (1994)