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A ternary beryllide was prepared by mixing Be 12 V with Be 12 Ti and using plasma sintering. After preparation at a pressure of 54 MPa for 20 min at 1200 °C, the changes in mechanical and thermal properties due to the addition of 1 wt.% Be 12 V and 10 wt.% Be 12 V were investigated. The Vickers hardness of both Be 12 Ti+1 wt.%Be 12 V and Be 12 Ti+10 wt.%Be 12 V were lower than Be 12 Ti. In addition, the fracture toughness was calculated using the IF method, and the calculation formula was calculated using various equations. According to the Anstis equation, it was 3.23 ± 0.13 MPa·m0.5 for Be 12 Ti, and it increased to 4.47 ± 0.13 MPa·m0.5 for Be 12 Ti+1 wt.%Be 12 V. As Be 12 V was added, it increased compared to Be 12 Ti. In the case of thermal properties, the thermal diffusivity of Be 12 Ti and Be 12 Ti+1 wt.%Be 12 V showed almost the same value, while that of Be 12 Ti+10 wt.%Be 12 V was lower than others. In the case of specific heat, moreover, it tended to decrease as the temperature increased in comparison to Be 12 Ti. This trend was particularly evident when 10 wt.% Be 12 V was added. Thus, the results of the thermal conductivity calculation indicated a decrease in tendency as the amount of addition increased in comparison to Be 12 Ti. Therefore, the addition of 1 wt.% Be 12 V seems to be more effective than the addition of 10 wt.% Be 12 V. [ABSTRACT FROM AUTHOR]

• Based on the SDC-IC standard, a integral analysis of thermal mechanical properties was conducted using a combination of elastic and plastic analysis methods • In ductile damage analysis, local plastic yielding occurs in the inner tube, but the overall load-bearing performance of the structure is good. • In non-ductile damage analysis, the T-tube unit have the immediate plastic flow localization, but there is no the immediate local fracture due to exhaustion. • From the perspective of transient cyclic thermal hydraulic analysis, the results show that there will be no ratcheting. To verify the mechanical properties of the helium-cooled T-tube divertor, based on the thermal hydraulic optimization of the T-tube divertor, a preliminary analysis was conducted through a combination of elastic and elasto-plastic analysis. The results show that for the M-type damage mode, the T-tube divertor exhibits a small-scale plastic yielding phenomenon, but this has no effect on the overall divertor. The T-tube divertor has local necking deformation but does not fail due to immediate local fracture, and the T-tube divertor has no creep rupture. For the C-type damage mode, the plastic strain of the T-tube divertor is not accumulated under cyclic loading, but in the fatigue analysis, the life of the divertor does not meet the design standards and further optimization is needed. The conclusion shows that the optimized helium-cooled T-tube divertor has good thermo-mechanical properties, and the safety of the helium-cooled T-tube divertor unit can be guaranteed. [ABSTRACT FROM AUTHOR]

• Re and Os production can be suppressed by depleting W of W-186. • The long term gamma activity can also be suppressed by depleting W of W-186. • Targeting depletion to specific regions of components can aid in the suppression of W-186, this could reduce manufacturing costs. • Discrete and functionally graded approaches are examples of methods for reducing the manufacturing cost of components with depleted regions. Tungsten (W) based materials are primary candidates for the plasma facing components (PFCs) in fusion power stations. A side effect of the use of W in neutron irradiation environments such as fusion is the transmutation to Rhenium (Re) and Osmium (Os), which are known to degrade the mechanical and thermal properties. In this study, neutron transport and nuclear inventory simulations were used to investigate a strategy of selective isotopic enrichment and/or depletion of W isotopes to supress the formation of Re and Os in a representative first-wall monoblock design. It was found that Re and Os production can be reduced by up to three orders of magnitude compared to natural W by depleting W-186 isotope to 100 %, as 186W(n,γ) 187W followed by beta decay is the main pathway for Re production. Two design methodologies were investigated that considered 1) an approach with two discrete, depleted in W-186 regions, with a range of depletion degree investigated and 2) a graded approach with layers of different levels of depletion to emulate a functionally graded material. [ABSTRACT FROM AUTHOR]

• Material property variations in the nominally pure industrial tungsten materials. • Observe thermal conductivity decrease with impurity content while hardness correlates with the grain size. • Material properties of W, W-TiC, and W-Y 2 O 3 fabricated by spark plasma sintering method. • Improvement in stability against heat flux through the stabilized fine-grained microstructure of W-TiC and W-Y 2 O 3. In this study, we investigated thermal and microstructural properties of pure and dispersion strengthened tungsten fabricated by Spark Plasma Sintering (SPS) method in comparison to the tungsten materials produced by different commercial suppliers. We measured the thermal conductivity of the examined materials and compared them with chemical composition, density and microstructure of the samples. Thermal conductivity up to 900 ℃ was derived from thermal diffusivity and specific heat measurement and chemical composition was analyzed through glow discharge mass spectroscopy. Thermal conductivity values show a close correlation with content of compositional element. On the other hand, Vickers hardness shows correlation with the average grain size of the samples. The sample produced by different fabrication process shows different density and thermal conductivity. While TiC and Y 2 O 3 added tungsten fabricated by SPS method shows improved stability in microstructure against recrystallization after exposure to high heat flux, the reduction of thermal conductivity is analyzed in comparison to the amount of contained additive elements. [ABSTRACT FROM AUTHOR]

The results of a study of the chemical composition, vacuum and thermal properties of boron carbide ceramics, which is planned to be used in ITER diagnostic ports, are presented. The measured boron carbide content in sintered and hot pressed ceramics is 99,0÷99,7%, they do not contain unauthorized impurities, which in accordance to ITER requirements. The outgassing rate of sintered and hot pressed boron carbide produced by Virial Ltd meets the requirements of the ITER Vacuum Handbook. Sintered ceramic is cheaper than hot pressed, and it is produced in large volumes, which it would be preferable for use in ITER ports, but with its massive use may create a significant gas load on the vacuum system that requires accurate analysis and outgassing tests on all candidated ceramics. [ABSTRACT FROM AUTHOR]

• The thermal conductivity of Li 2 TiO 3 pebble bed has been successfully measured by the laser flash method. • The thermal conductivity of Li 2 TiO 3 pebble bed until 500 °C increased with increasing measuring temperature. • The thermal conductivity of Li 2 TiO 3 pebble bed at more than 500 °C was saturated to 1.3 W/m/K. • The specific heat and thermal diffusivity of Li 2 TiO 3 pebble bed were adopted up to 700 °C due to chemical reaction and pebble connection. In order to verify the performance and safety of the solid-type breeding blanket concept, the thermal conductivity of the pebble bed is required as an essential parameter for design and analysis of the blanket. The hot wire, hot disk, guarded hot plate, and laser flash method are considered as measurement technique for the thermal conductivity of pebble bed. This study aims at the provision of the thermal conductivity of Li 2 TiO 3 pebble bed, which is one of the candidates for tritium breeder material in the solid-type breeding blanket. The laser flash method was adopted due to its advantages compared with other methods, such as high accuracy and repeatability, easy sample preparation, absolute measurement technique, and so on. In this method, the thermal diffusivity is directly measured by the laser flash system, and then the thermal conductivity is obtained by multiplying density and specific heat to the measured thermal diffusivity. The thermal conductivity of the Li 2 TiO 3 pebble bed with packing factor of 57% increased with increasing temperature up to 500 °C and was saturated to 1.3 W/m/K at more than 500 °C. [ABSTRACT FROM AUTHOR]

Tungsten is a promising candidate for plasma-facing materials in a fusion reactor, and several research studies have been conducted to improve the mechanical properties of pure W. However, the thermal properties are also important characteristics of plasma-facing materials. In this study, the thermal properties of pure W and its alloys were measured, and the effect of alloying on the thermal properties of pure W was investigated. Potassium-bubble (K-bubble) dispersion, which is one of the major methods utilized to improve the mechanical properties of W, did not affect the thermal diffusivity and conductivity of pure W. On the other hand, the presence of rhenium, which is major alloying element of W, affected the absolute values and the temperature dependence of thermal diffusivity and conductivity. The effect of alloying on specific heat and the anisotropy in thermal diffusivity of pure W and its alloys were also investigated. Measurements of the specific heat showed that K-bubble dispersion and Re addition had insignificant effects on the values obtained for pure W. Anisotropy in thermal diffusivity was not observed, and the effect of anisotropic grain structure and alloying was insignificant. [ABSTRACT FROM AUTHOR]

For the testing of helium cooled plasma facing components a homogeneous surface heat flux distribution has to be generated with up to 500 kW/m 2 in the case of the ITER Test Blanket Module. One option is to use infrared radiation heaters which consist of several quartz glass (fused silica) tubes with tungsten filaments inside to generate the required heat flux. This paper introduces the numerical model of the heater which has been investigated in the SIRHEX (“Surface Infrared Radiation Heating Experiment”) facility at KIT. The model uses a transient simulation to assess the heat flux distribution on the surface of the test mock-up. To validate the model, the numerical results are compared with the latest results from SIRHEX. Using the now qualified model, the heater set-up is optimized to generate a homogeneous distribution for a future test of a blanket first wall. [ABSTRACT FROM AUTHOR]

Helium is considered as coolant in the plasma facing first wall of several blanket concepts for DEMO fusion reactors, due to the favorable properties of flexible temperature range, chemical inertness, no activation, comparatively low effort to remove tritium from the gas and no chemical corrosion. Existing blanket designs have shown the ability to use helium cooled first walls with heat flux densities of 0.5 MW/m 2 . Average steady state heat loads coming from the plasma for current EU DEMO concepts are expected in the range of 0.3 MW/m 2 . The definition of peak values is still ongoing and depends on the chosen first wall shape, magnetic configuration and assumptions on the fraction of radiated power and power fall off lengths in the scrape off layer of the plasma. Peak steady state values could reach and excess 1 MW/m 2 . Higher short-term transient loads are expected. Design optimization approaches including heat transfer enhancement, local heat transfer tuning and shape optimization of the channel cross section are discussed. Design points to enable a helium cooled first wall capable to sustain heat flux densities of 1 MW/m 2 at an average shell temperature lower than 500 °C are developed based on experimentally validated heat transfer coefficients of structured channel surfaces. The required pumping power is in the range of 3–5% of the collected thermal power. The FEM stress analyses show code-acceptable stress intensities. Several manufacturing methods enabling the application of the suggested heat transfer enhanced first wall channels are explored. An alternative cooling technology is suggested, extending the tolerable heat loads up to 3 MW/m 2 . [ABSTRACT FROM AUTHOR]

The quality of CuCrZr/316L joint is crucial for the safety of ITER hypervapotron cooling structure and hot isostatic pressing (HIP) is an important bonding technique for this structure. In this paper, the authors present a finding that the CuCrZr/316L HIP-joint can be enhanced by nickel electroplating on 316L and CuCrZr. A Charpy Impact Value as high as 111.5 ± 3.3 J/cm 2 , which is more than two times the value in a published article, is obtained. The influence of nickel electroplating is twofold: (1) it can prevent the occurrence of nickel-poor region in 316L and the formation of ferrite; (2) it can protect CuCrZr from oxidation during the heating stage of HIP. However, tensile test is not as effective as Charpy Impact Test in characterizing the bonding quality of the CuCrZr/316L HIP-joint. The surface treatment employed in this study is amenable to batch-scale industrial manufacturing at low cost. [ABSTRACT FROM AUTHOR]

• For the first time a preliminary design of the DEMO CMS is presented. • A multiphysical (i.e., thermal-structural) finite-element simulation was employed. • Final design was achieved by means of shape parameterization with RBF mesh morphing. • Used as a starting configuration a simple baseline geometry from ITER. • Heat load reduced from 10 W to 5 W and stress intensity from 765 MPa to 698 MPa. The Cold Mass Support (CMS) is a basic structural component of the feeder system of the EU DEMO tokamak. The function of this structural element is not only to provide support to the containment duct – wiring the cryogenic lines, electrical bus bars and instrumentation to the magnets – in terms of both gravity loads and Lorentz forces, but also to isolate the low-temperature elements from the ambient, minimising the (conductive and radiative) heat loads. In this work, we describe a simulation and optimisation procedure adopted for the preliminary design of the EU DEMO CMS, which makes use of finite-element analyses and a shape parameterisation based on Radial Basis Functions (RBF). Structural and thermal numerical grids are morphed from a baseline configuration, and complex shape variations are achieved by means of an advanced meshless procedure based on CAD parametric surfaces, respecting the given geometrical constraints. A response surface-based optimisation is finally performed, in order to obtain a preliminary CMS candidate able to fulfil both structural and thermal targets. [ABSTRACT FROM AUTHOR]

Korea has developed an advanced reduced-activation alloy (ARAA) as a structural material for helium-cooled ceramic reflector test blanket module (HCCR-TBM) applications. The present paper describes the history of alloy development and the properties of ARAA, which has been produced at a 5 t scale using vacuum induction melting and electro-slag re-melting methods. ARAA is a 9Cr–1.2W based ferritic–martensitic steel with 0.01 wt.% Zr. The mechanical properties, thermal properties and physical and magnetic properties of ARAA show similar temperature dependencies to those observed for Eurofer 97. However, ARAA exhibits a much longer creep–rupture time than conventional RAFM steel, which suggests a positive effect on Zr addition. The enhanced creep strength of ARAA by the addition of Zr is attributed to the reduced temperature-dependence of the yield strength. [ABSTRACT FROM AUTHOR]

This paper introduces tape casting as a new route for the production of isotropic and fine-grained tungsten components. Microstructural and thermal properties of tape cast tungsten samples are determined. Thermal shock behavior according to the thermo-cyclic loading of standing X-ray anodes is investigated and compared to the behavior of a rolled tungsten grade. The development of surface roughness during the thermal shock loading is discussed in relation to the development of the grain structure and crack pattern. The fine-grained and stable microstructure of the tape cast material exhibits less roughening under such test conditions. [ABSTRACT FROM AUTHOR]

The thermal performance of a carbon fibre composite-copper monoblock, a sub-component of a fusion reactor divertor, was investigated by finite element analysis. High-accuracy simulations were created using an emerging technique, image-based finite element modelling, which converts X-ray tomography data into micro-structurally faithful models, capturing details such as manufacturing defects. For validation, a case study was performed where the thermal analysis by laser flash of a carbon fibre composite-copper disc was simulated such that computational and experimental results could be compared directly. Results showed that a high resolution image-based simulation (102 million elements of 32 μm width) provided increased accuracy over a low resolution image-based simulation (0.6 million elements of 194 μm width) and idealised computer aided design simulations. Using this technique to analyse a monoblock mock-up, it was possible to detect and quantify the effects of debonding regions at the carbon fibre composite-copper interface likely to impact both component performance and expected lifetime. These features would not have been accounted for in idealised computer aided design simulations. [ABSTRACT FROM AUTHOR]

ITER cooling water system is composed by several cooling loops, the primary heat transfer loops that form the Tokamak Cooling Water System (TCWS), the secondary heat transfer loops that form the Component Cooling Water System (CCWS) and the Chilled Water System (CHWS) and a tertiary heat transfer loop which is the Heat Rejection System (HRS). The CCWS is further divided into CCWS-1, CCWS-2A, CCWS-2B, CCWS-2C, CCWS-2D depending on the water chemistry needs of clients and wetted area material. The component cooling water system loop 2B (CCWS-2B) has the function to remove heat load from coil power supply component, Neutral Beam Injectors (NBIs) system component and diagnostic system which are located in different buildings. As the total number of the client connections for the loop is a few hundreds, simplified thermal hydraulic analysis model has been developed to perform thermal analysis on the component cooling water system loop 2B. The curve of the cooling water temperature at client inlet and outlet during one cycle of the most demanding plasma operation scenario was obtained and the cooling water flow rate can meet the thermal removal requirement of client was also confirmed from this analysis. In addition, operation behaviour of the main heat exchanger for CCWS-2B in this thermal analysis was depicted for main heat exchanger selection purposes. This study has been carried out with the AFT Fathom code. [ABSTRACT FROM AUTHOR]

This paper presents the simulation results of 3D finite element analysis (FEA) of laser welding processes for the ITER correction coil case welding; predicts the temperature distribution and compares it with the experimental result to evaluate the impact to the properties of winding pack during the welding process. A specimen of coil case was modeled and simulated by using specialized welding simulation software SYSWELD, Modeling used austenitic stainless steel 316LN as the specimen material and a 3D Conical Gaussian was used as a heat source model. A plate sample was welded before the FE modeling in order to obtain the laser welding parameters and the Gaussian parameters of molten pool. To verify the simulation results, a coil case sample was welded using laser welding with welding parameters that matched the model, and the corresponding temperature values were measured using thermocouples. Compared with the FEA results, it was found that the FEA temperature distribution shows good agreement with the experimental measurements and the laser welding process will not harm the winding pack. [ABSTRACT FROM AUTHOR]

Paper describes the newly established high heat flux test facility (HHFTF) at IPR in India. The facility is being used to simulate typical steady state and transient thermal load conditions for testing Divertor plasma facing components using a 200 kW Electron Beam as heat source. After successful installation and commissioning of the HHFTF, two straight tungsten mono-block type test mock-ups are tested for several hundred thermal cycles with absorbed heat flux up to 11 MW/m 2 (corresponding incident heat flux 20 MW/m 2 ). Maximum temperature on tungsten surface observed using a two-color pyrometer is found to be 1150 °C for absorbed heat flux up to 11 MW/m 2 (corresponding incident heat flux 20 MW/m 2 ). Present paper gives a brief introduction to the HHFTF followed by experimental details and some results of high heat flux tests were performed on two straight tungsten mono-blocks. [ABSTRACT FROM AUTHOR]

Jet impinging technique is an advance divertor concept for the design of future fusion power plants. This technique is extensively used due to its high heat removal capability with reasonable pumping power and for safe operation. In this design, plasma-facing components are fabricated with numerous fingers cooled by helium jets to reduce the thermal stresses. The present study is focused towards finding an optimum performance of one such finger mock-up through systematic computational fluid dynamics (CFD) studies. Heat transfer characteristics of jet impingement have been numerically investigated with sectorial extended surfaces (SES). The result shows that addition of SES enhances heat removal potential with minimum pumping power. Detailed parametric studies on critical parameters that influence thermal performance of the finger mock-up have been analyzed. Thermo-mechanical analysis has been carried out through finite element based approach to know the state of stress in the assembly as a result of large temperature gradients. It is seen that the stresses are within the permissible limits for the present design. The whole numerical simulation has been carried out using general-purpose CFD software (ANSYS FLUENT, Release 14.0, User Guide, Ansys, Inc., 2011). Benchmark validation studies have been performed against high-heat flux experiments (B. Končar, P. Norajitra, K. Oblak, Appl. Therm. Eng., 30, 697–705, 2010) and a good agreement is noticed between the present simulation and the reported results. [ABSTRACT FROM AUTHOR]

ASDEX Upgrade will be evolved by installing a new upper divertor with two in-vessel coils allowing the investigation of alternative magnetic configurations. The in-vessel coils will be bended inside the torus. The co-axial conductor consists of a Tefzel insulated water cooled copper core embedded into a stainless steel protection shield. To test the conductor concept and to qualify the electrical and thermal properties of the Tefzel insulated conductor (TIC), prototype conductors of 3 and 8 m length were used. These conductors undergo electrical and thermal tests accompanied by FEM calculations. The electrical tests include high voltage and partial discharge tests. For thermo-mechanical tests the prototype conductors were fed with up to 13 kA for 5 s. During the test the copper conductor was heated up by 60 K. The Tefzel cladding and the protection tube are warmed up with a delay given by the thermal heat conduction and capacity. The thermo-mechanical behaviour of the TIC during the heating and cooling phase could be confirmed by the FEM-models. As part of the qualification, the conductor was thermally overloaded by a factor of 3, without degradation. The paper presents the electrical and thermal tests as well as the results of the related FEM calculations. [ABSTRACT FROM AUTHOR]

• The eigen equation of the 155/157Gd areal density and thermal neutron shielding rate was established by MCNP. • The novel Gd/316L thermal neutron shielding composite was fabricated by SPS at lower temperature. • The microstructure and composition of transition layer were studied. • The transition layer (FeNi 3 , Gd 3 Ni, and GdNi) dissolves firstly during the corrosion process. The constitutive equation of the 155/157Gd areal density and thermal neutron shielding rate was established by Monte Carlo N Particle Transport Code (MCNP) and used to design the novel Gd/316L stainless steel thermal neutron shielding materials. When the 155/157Gd areal density is greater than 0.01 g/cm2, the thermal neutron absorption rate of the Gd/316L composite reaches approximately 99%. The novel thermal neutron shielding materials with different Gd (1 wt.%, - 5 wt.%,) contents were prepared at 950 °C by spark plasma sintering. The transition layer composed of FeNi 3 , GdNi, and Gd 3 Ni can be formed during the sintering process. Meanwhile, the mechanical properties of Gd/316L composites decrease with the increase of the Gd content. In H 3 BO 3 solution, pitting corrosion and galvanic corrosion occur at the interface between Gd and 316L matrix. [ABSTRACT FROM AUTHOR]

• Long-term reliable vacuum tests of boron carbide ceramics have been performed. • Experiments were carried out on the activation of boron carbide ceramics by neutrons. • Boron carbide ceramics meet all ITER requirements in terms of chemical composition, vacuum and thermal properties. • The use of large quantities of boron carbide ceramics in vacuum to protect ITER ports is acceptable. The results of long-term outgassing test of large quantities of ceramics, which is planned to be used in ITER diagnostic ports, are presented. The outgassing rate of boron carbide ceramics produced by Virial Ltd meets the requirements of the ITER Vacuum Handbook. Massive use of ceramics in ITER may create a significant gas load on the vacuum system that requires accurate analysis. Calculations of gas emission of Equatorial Port #11 were performed and it was demonstrated that a significant amount of ceramics can be used for neutron protection. Experiments on activation of ceramics and stainless steels by fast neutrons were carried out, which showed a rapid reduction of residual radioactivity and absence of hazardous impurities. [ABSTRACT FROM AUTHOR]

• A DEM-CFD model considering contact conduction was used to study the effective thermal conductivity (ETC) of pebble beds. • The ETC predicted by the current model is consistent with experimental data in a wide range of k s / k f from 1 to 10000. • If k s / k f is larger than 20, the difference of the ETC for the contact model and the non-contact model is greater than 10%. Pebble beds are main schemes for tritium breeding zones in the solid blanket for fusion reactors. The effective thermal conductivity of pebble beds (k e f f) is an important thermal property used in the thermal hydraulic design of solid blankets. In this study, a 3D DEM-CFD one-way coupling method considering contact conduction is applied to analyze the influence of solid contact conduction and solid-fluid-solid conduction on the effective thermal conductivity of pebble beds. The current numerical model gives a k e f f consistent with experimental data in a wide range of solid-to-fluid conductivity ratio k s / k f from 1 to 10,000. It can be concluded that when the k s / k f is more than 20, contact conduction should be taken into account to make a good prediction for k e f f. [ABSTRACT FROM AUTHOR]