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It is necessary to take effective cooling methods from the divertor which can sustain one-sided high heat fluxes. In order to ensure the reliability and security of the divertor in fusion reactor, the heat transfer and thermo-mechanical assessment of W/CuCrZr monoblock divertor were conducted. In this paper, the heat transfer enhancements of the subcooled flow boiling in the vertically upward screw tubes, the tubes with twisted tape inserts and the plain tubes were examined by the Fluent software. Then, the thermo-mechanical analyses of the three cooling tubes were performed in this paper. The 3S m rule, Bree-diagram rule, fatigue, pipe material temperatures and the assessment of critical heat flux (CHF) were compared and analyzed for the three kinds of cooling tubes, which can guide the optimum for the heat transfer and structural analyses of the monoblock divertor. As a result, compared with the screw tube and plain tube, the tube with twisted tape inserts can enhance the heat transfer. Among the three kinds of cooling tubes, secondary stress of tungsten and CuCrZr in the tube with twisted tape inserts were the smallest. Thermal fatigue analyses of the tube with twisted tape inserts were calculated and the tube owned a fatigue life of approximately 0.427 years under the current working conditions. [ABSTRACT FROM AUTHOR]

Highlights • A CO 2 -cooled FW has been proposed in this paper. • A high heat flux with radiation heat flux and particle heat flux has been applied on FW. • Different wall roughness position of channel has been shown. • Different wall roughness value has been given. • Sensitivity particle heat flux has been studied in this paper. Abstract The Demonstration Fusion Reactor in Europe (EU-DEMO) is considered to be the last step before building a commercial fusion power plant. Up to now, He has been chosen as gas coolant for three of the four candidate breeding blankets proposed for the EU-DEMO, being the Helium Cooled Pebble Bed (HCPB) among them. The choice of He coolant is supported, among other features, by its superior heat transfer capabilities, its transparency to neutrons and chemical inertness. However, the use of this coolant in DEMO poses some feasibility questions, like the large coolant inventory, resulting in possible safety issues, as well as the large circulating power. On the other side, CO 2 has been the fluid coolant choice in the nuclear fission industry for gas-cooled reactors since the 1950s. The key advantage of this gas is its larger density, which results in more compact components, lower coolant inventories, lower circulating powers and thus better plant thermal efficiency. This, coupled also with a relatively high molecular stability at moderate temperatures (up to 600 °C), its larger availability and lower price make CO 2 an attractive coolant for a pebble bed breeding blanket, as an alternative to He. As a Plasma Facing Component, the First wall (FW) is one of the components under the most challenging environment. Currently, the FW region at the upper port inboard location is expected to suffer the most critical heat flux loads. Indeed, a peak charged particle heat flux 0.99 MW/m2 together with a radiation heat flux of 0.18 MW/m2 is foreseen at this location. This paper reports the thermal-hydraulic analyses carried out to find the practical heat flux limits by using CO 2 as coolant. The analyses have been performed by using the commercial computational fluid dynamics (CFD) code ANSYS CFX. Different augmented technical wall roughness areas have been considered, for a best heat transfer-to-pressure drop ratio. Sensitivity analyses have been performed as well for higher peak particle heat fluxes, evidencing that a CO 2 -cooled FW can dissipate peak heat fluxes up to 1.8 MW/m2 keeping the circulator power in a reasonable range. [ABSTRACT FROM AUTHOR]

The present paper deals with the surface heat flux estimation with thermocouples (TC) and fiber Bragg grating (FBG) embedded in the plasma facing components (PFC) of the WEST tokamak. A 2D heat transfer model combined with the conjugate gradient method and the adjoint state is used to estimate the plasma heat flux deposited on the PFC. The plasma heat flux is characterized by the time evolution of its amplitude and spatial shape on the target (heat flux decay length λ q t , power spreading in the private flux region S t and the strike point location x 0). As a first step, five ohmic pulses have been investigated with different magnetic configuration and divertor X-point height varying from 44 to 68 mm from the surface. Despite an outboard shift, the relative displacements of the outer strike point as well as the heat flux decay length derived from the TC/FBG systems are consistent with the magnetic equilibrium reconstruction. [ABSTRACT FROM AUTHOR]

• The weight of OVT target is about 1.85 t, which is suitable for the development of RH manipulator, reduces the weight and improves the reliability of RH maintenance. • RH channels, main bolt and main water pipe RH compatible structures are designed on the front face of the OVT target, realizing RH process planning. The OVT target can be easily removed or positioned with only slight rotation, realizing the reliable RH maintenance of front-face target and effectively improving RH maintenance efficiency. • Advanced materials are selected. KW has high recrystallization temperature and strong neutron irradiation resistance. ODS-Cu has good high-temperature performance and is an ideal high-temperature heat transfer material. The structural material is RAFM with neutron activation resistance. • The thermal-hydraulic analysis was carried out, which verified that the OVT target has heat removal ability to withstand heat flux of 20 MW/m2 and neutron radiation heat under 1.5 GW fusion power. • The work completed will promote the structural analysis and optimization of OVT. The divertor is a key in-vessel component of the China Fusion Engineering Test Reactor (CFETR), which is mainly used to remove high heat load and exclude particles. In addition, the CFETR divertor also needs to meet requirements of Remote handling (RH) maintenance due to activation of materials by neutrons. CFETR proposed the concept of hybrid divertor-blanket to meet the requirements mentioned above and improve tritium breeding ratio (TBR). The new concept cancels the cassette and adopts a separate structure. Space below the divertor is occupied by the breeding blanket to improve TBR, and targets are directly fixed on the breeding blanket to improve RH maintenance efficiency. The outer vertical target (OVT) is a high heat load component of the divertor, which needs to remove the heat flux of 20 MW/m2 and neutron radiation heat at 1.5 GW fusion power and meet the requirements of RH maintenance. The structure design, RH compatible structure design, RH process and thermal-hydraulic analysis of OVT are completed in this paper. The relevant design effectively reduces the weight and complexity of RH maintenance and improves RH efficiency of OVT. RH process for OVT proves to have great engineering feasibility. Advanced low activation materials are adopted for OVT. Thermal-hydraulic analysis verifies that OVT can remove heat flux of 20 MW/m2 and neutron radiation heat at 1.5 GW fusion power, which confirms the rationality of the cooling system and effectively prompts the structural analysis of OVT. [ABSTRACT FROM AUTHOR]

Highlights • Novel swirl tube design for enhancing the heat transfer performance of pole shields of EAST-NBI. • Optimizing for the cooling scheme and the key structural parameters. • The design scheme effectively reduces the surface temperature of pole shields of EAST-NBI. Abstract The Neutral Beam Injection (NBI) auxiliary heating system for Experimental Advanced Superconducting Tokamak (EAST) is designed with 180 ° magnetic field deflection to deflect the particles that are not neutralized during the beam transmission. For protecting the deflection magnet from the influence of the divergent beam, the pole shields are designed to be placed on both sides of its optical path, and the highest power density of the above deposit will be as high as several MW/m2. In order to ensure reliable operation of 100s and even 1000s long pulse EAST-NBI system, it is very important to enhance the heat transfer performance of these shields. A novel heat transfer enhancement design of pole shield with swirl tube for EAST-NBI system is presented in this paper. By using a verified subcooled boiling model, the performance optimization of the key structural parameters is completed according to the operation limit of the magnetic pole shield and the cooling water system, and also the final structural optimization scheme is checked. The results show that heat transfer performance of the proposed scheme can well satisfy the requirements of the long pulse design. This study provides a potential method for the heat transfer enhancement design of other high heat flux components in EAST-NBI system, which is of great theoretical and practical significance for the future operation with long pulse (over 1000 s) and high power 2˜4 MW. [ABSTRACT FROM AUTHOR]

Abstract The vertical targets in fusion divertors are the main component which are exposed to very high heat flux generally exceeding 10 MW/m2. In addition, since the heat is non-uniformly imposed to the cooling channels by unique one-side heating, the local heat flux can easily reach up to 20 MW/m2. It is a very challenging engineering target even when considering the latest cooling technology. In this study, CFD-based shape optimization is applied in order to find the optimal channel cross-sectional shape for minimizing local heat flux on the cooling surface. For simplicity, 2-D thermal analysis is adopted as the first step and it estimates preliminary optimized shapes. In this 2-D analysis, heat transfer model is simply reflected on the boundary conditions by analytic method using heat transfer correlations. Then, the optimized shapes are validated using detailed 3-D CFD simulations based on Eulerian two-phase flow models with/without boiling heat transfer. Maximum heat flux on the cooling channel surface is defined as the objective function in the optimization process. Finally, it is found that the maximum heat flux and temperature in the cooling channel can be significantly reduced compared to the original circular channel shape by simple modification to the ellipse-like cross-section. Effects of boiling heat transfer and manufacturability are also discussed in this paper. [ABSTRACT FROM AUTHOR]

Abstract The geometrical freedom that additive manufacturing (AM) provides enables realisation of formerly impossible design concepts for high heat flux components aimed at use in a fusion reactor. This paper demonstrates a number of these advantages using a tile-based divertor target with two examples of novel cooling geometry. To showcase the potential benefits of high temperature operation using materials newly available for AM processing, tantalum was used as the structural material. Rather than present optimised designs, these two concepts highlight features enabled by this new technology. Simple quantitative analysis shows heat transfer improvements of 25% for a multiple small pipe concept compared to single pipe designs. Finite element analysis for both designs, including a tungsten armour, also demonstrates a more uniform temperature distribution, reducing thermal stresses below elastic limits for 5 MWm−2 heat flux cases, even with 600 °C coolant. At 10 MW m−2 yield is exceeded but the expectation is further design optimisation should enhance structural integrity. Overall, there are modest gains in heat flux handling, lifetime, and thermal efficiency compared to existing concepts, but there are significant improvements in waste reduction and remote handling mass, using typically 80% less material. Moreover, integrated manifolding and careful location of coolant connection is suggested to facilitate repair and replaceability, which should also reduce risk associated with qualifying joints. [ABSTRACT FROM AUTHOR]

Several novel design solutions for high performance cooling systems have been developed by Consorzio RFX, permitting to simulate the challenging heat transfer conditions foreseen in the future fusion devices. The project, called Multi-design Innovative Cooling Research & Optimization (MICRO), aims, on one hand, to verify the present solution applied inside the NBI accelerator grids (i.e. MITICA experiment) and, on the other, to perform further improvements in the heat transfer process with an acceptable pressure drop and reliable manufacturing process. The main advantages rely on the possibility to extend the fatigue life-cycle of different high thermal stress components and to investigate the possibility to employ alternative dielectric fluids instead of water. Such design solutions would in fact allow the exploitation of less performing fluids in terms of cooling capability. This is particularly relevant in view of DEMO and future power plants characterized by higher efficiency and reliability. The paper gives a description of the Computation Fluid Dynamics (CFD) analysis and the samples manufacturing process. [ABSTRACT FROM AUTHOR]

Within the EU-DEMO first wall protection framework, work on limiter's plasma-facing component design has started under plasma disruptive events (Richiusa et al., 2022). Starting from the rationale behind the TARTIFL&TTE software (Richiusa et al., 2022), this companion paper describes the progress on the engineering modelling of the plasma-facing material phase change under high heat flux, with the aid of COMSOL Multiphysics® software. The aim is to develop a reliable technique which can be used by designers for predicting how much of the solid armour undergoes phase change. This helps satisfy requirements for actively cooled components, such as at which armour depth safely locating the cooling system, and if its design can safely handle the heat transfer in the resulting component configuration after the disruption is extinguished. A few changes to the driving idea in Richiusa et al. (2022) will be also highlighted. The multiphysics software allows us to implement a 1D model which can be extended to 2D/3D geometries subjected to both uniform and non-uniform heat flux. It also offers the capability of conjugated heat transfer in solids and liquids by coupling the two different domains. Although this multiphysics approach is also investigated, no effort in melting layer motion modelling is done. Therefore, the equivalent way of validating this approach while reducing its computational time is working with one single solid domain, within which any liquid phase changes are tracked by an apparent heat capacity formulation. The vapour domain is not modelled. The material removal due to the evaporative mass flux is modelled by means of moving mesh frames which push the recessing liquid interface backwards according to gas kinetics-driven boundary conditions. The melt pool is not removed during the transient. Mass balance considerations drive the liquid-to-vapour interface velocity. The 3D Multiphisics implementation (3D-TARTIFL&TTE), is here supported by a benchmark activity and an application to the limiter's plasma-facing armour, whose preferred chosen thickness is 20 mm. • Limiters are proposed for protecting the EU-DEMO first wall against plasma-wall contacts. • Sacrificial limiter's plasma-facing components can be damaged under disruptions, whose severity and probability should be minimized by the plasma scenario design. • Limiters would ideally face multiple events, as long as their cooling system integrity is safely preserved. • The limiter design should consider phase changes to estimate the minimum safest armour thickness undergoing plasma impacts. • 3D-TARTIFL&TTE is a multiphysics approach for modelling heat transfer in solid components under phase change. [ABSTRACT FROM AUTHOR]

• One optimized hypervapotron module with longitudinal and horizontal internal ribs was designed, and the rationality of the structure has been confirmed by some design criterions. • CuCrZr alloy channel was confirmed to experience plastic cycle and fatigue life prediction was calculated to be 1.249 year combining with CFETR operating conditions. • This paper provides a systematic method for fatigue life prediction and reliability analysis for hypervapotron of divertor. The engineering design of China Fusion Engineering Testing Reactor (CFETR) will be finished in the end of this year. Compared with Experimental Advanced Superconducting Tokamak (EAST), this is a large development which means closer to commercial fusion reactors. The design objective of CFETR divertor requires a steady-state thermal load of 10–20MW/m2 exhaust. There are several design options for the CFETR divertor. Hypervapotron with its superior performance has been researched for several decades as one important candidate for future fusion device divertor. In this paper, a new design of hypervapotron plasma facing unit (PFU) was developed. Horizontal and longitudinal ribs were adopted to generate more vaporized cores to enhance heat transfer. Some optimizations were performed based on the allowable max heat flux (MAHF) to get the final design. Rensselaer Polytechnic Institute (RPI) boiling model of Eulerian Multiphase fluid and k-epsilon model of Viscous were chosen to deal with sub-cooled boiling. 3Sm rule and Bree-diagram was used to confirm the rationality of the structure after thermal-mechanical analysis. Numerical simulation results suggested that CuCrZr alloy will experience cyclic plastic strain. Based on this, fatigue life prediction of the hypervapotron PFU at a steady-state heat flux of 10MW/m2 were carried out through local stress-strain method, neuber curve and design fatigue life curve. Eventually, a fatigue life of approximately 1.2489 year was verified for CFETR divertor. This paper provides a systematic method for fatigue life prediction and reliability analysis of hypervapotron. [ABSTRACT FROM AUTHOR]

• The divertor adopts a separate design, which means the IVT, the Dome, and the OVT are respectively connected with the corresponding matching interface structure through the support components and the RH compatible structure. This concept realizes RH maintenance of the IVT, Dome and OVT independently and aims at realizing rapid and reliable RH maintenance on the front of the target plate, reducing the weight level of RH, improving the reliability of RH maintenance and improving the interchangeability of the Dome. • Special PFUs and cooling system are designed to adapt to the RH, to shadow them from the direct bombardment of plasma and provide enough heat removal capability. • Thermal-hydraulic analysis was performed to verify the reasonableness of the cooling system design and heat transfer capacity of Dome. The maximum temperature of the materials are all within the allowable temperature range under the heat flux and nuclear heat. • The results will provide supports to the electromagnetic and structure analysis of Dome and promote the final design of CFETR water-cooled divertor. The main function of China Fusion Engineering Test Reactor (CFETR) Divertor Dome is to isolate neutral particles and impurity particles as well as improve the ability of particle exclusion. For the CFETR Divertor Dome, a new integral hybrid divertor-blanket structure was proposed in previous study. The new Dome should also be compatible with the Remote handling (RH) requirements due to neutron activation of in-vessel components, thus special RH channels are designed on the front of the plasma-facing units (PFUs). Due to direct bombardment or sputtering of plasma, the Dome will suffer heat flux up to 10 MW/m2 and nuclear heat under 1.5 GW fusion power during steady-state operation, which puts forward critical and tricky requests for the design of cooling system. While for the RH channels, the situation is even direr due to its location and poor cooling conditions. In this paper, cooling system of the Dome is proposed, and cooling channels with special thermal transfer structure are designed in the PFUs to guarantee them with heat removal ability up to 10 MW/m2. Special PFUs and cooling system are designed to adapt to the RH, to shadow them from the direct bombardment of plasma and provide enough heat removal capability. Thermal-hydraulic analysis is carried out to verify the heat transfer ability of the Dome cooling system. The results will provide supports to the electromagnetic and structure analysis of Dome and promote the final design of CFETR water-cooled divertor. [ABSTRACT FROM AUTHOR]

The first wall (FW) of DEMO is a component with high thermal loads. The cooling of the FW has to comply with the material's upper and lower temperature limits and requirements from stress assessment, like low temperature gradients. Also, the cooling has to be integrated into the balance-of-plant, in a sense to deliver exergy to the power cycle and require a limited pumping power for coolant circulation. This paper deals with the basics of FW cooling and proposes optimization approaches. The effectiveness of several heat transfer enhancement techniques is investigated for the use in helium cooled FW designs for DEMO. Among these are wall-mounted ribs, large scale mixing devices and modified hydraulic diameter. Their performance is assessed by computational fluid dynamics (CFD), and heat transfer coefficients and pressure drop are compared. Based on the results, an extrapolation to high heat fluxes is tried to estimate the higher limits of cooling capabilities. [ABSTRACT FROM AUTHOR]

The Water Cooled Ceramic Breeder (WCCB) blanket based on Pressurized Water Reactor (PWR) condition is one of the blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR). The first wall (FW) which plays an important part in the blanket design must remove the high heat flux radiated from plasma and nuclear heat deposition on the structure in any operating conditions. In this paper, the characteristics of enhanced heat transfer and friction for the FW with the inserted screw blade are studied by the numerical method. After the comparison between the numerical and experimental results, the standard k – ε turbulent model is selected to do the numerical calculation. The numerical results show that the peak temperature of RAFM steel could be reduced by decreasing screw pitch or increasing screw diameter, while accompanied with ascending flow resistance. Besides, among all of the chosen calculation cases compared with the smooth channel, the maximum value of temperature reduction is 10 °C under the conditions of heat flux of 0.5 MW/m 2 as well as screw pitch of 18 mm and screw diameter of 6 mm. The maximum increment ratio of the friction factor is 257% under the conditions of screw pitch of 10 mm and screw diameter of 4 mm. Furthermore, screw blade of 74 mm pitch and 4 mm diameter presents the highest overall performance evaluation criterion (PEC) value of 0.93 under Reynolds number of 270 000 conditions, and shows the best overall heat transfer enhancement performance. [ABSTRACT FROM AUTHOR]

• This paper use CFD method to simulate the cooling of the high-temperature divertor components. • This paper gives a clear overview of heat transfer and impingement jet profile in HEMJ divertor. • Our work shows the optimizing results of H/D ratio of divertor design. As part of the EU-DEMO prospective design studies, a He-cooled divertor concept has been developed by Karlsruhe Institute of Technologies (KIT). Using multi-jet cooling, the aim is to provide a robust solution that is capable of coping with heat fluxes as high as 10 MW/m2 under stationary loading conditions. Using the CFD simulations the cooling performance was evaluated and a configuration providing the lowest pipe temperature was identified. [ABSTRACT FROM AUTHOR]

Abstract: Efficient heat transfer technologies are essential for magnetically confined fusion reactors; this applies to both the current generation of experimental reactors as well as future power plants. A number of High Heat Flux devices have therefore been developed specifically for this application. One of the most promising candidates is the HyperVapotron, a water cooled device which relies on internal fins and boiling heat transfer to maximise the heat transfer capability. Over the past 30years, numerous variations of the HyperVapotron have been built and tested at fusion research centres around the globe resulting in devices that can now sustain heat fluxes in the region of 20–30MW/m2 in steady state. Until recently, there had been few attempts to model or understand the internal heat transfer mechanisms responsible for this exceptional performance with the result that design improvements have been traditionally sought experimentally which is both inefficient and costly. This paper presents the successful attempt to develop an engineering model of the HyperVapotron device using customisation of commercial Computational Fluid Dynamics software. To establish the most appropriate modelling choices, in-depth studies were performed examining the turbulence models (within the Reynolds Averaged Navier Stokes framework), near wall methods, grid resolution and boiling submodels. Comparing the CFD solutions with HyperVapotron experimental data suggests that a RANS-based, multiphase model is indeed capable of predicting performance over a wide range of geometries and boundary conditions if suitable sub-models are developed. Whilst a definitive set of design improvements is not defined here, it is expected that the methodologies and tools developed will enable designers of future High Heat Flux devices to perform significant virtual prototyping before embarking on the more costly build and test programmes. [Copyright &y& Elsevier]

Abstract: The divertor concept for DEMO fusion reactor is based on modular design cooled by multiple impinging jets. Such divertor should be able to withstand a surface heat flux of at least 10MW/m2 at an acceptable pumping power. To reduce the thermal loads the plasma-facing side of the divertor is build up of numerous small cooling fingers. Each cooling finger is cooled by an array of jets blowing through the holes on the steel cartridge. The size, number and arrangement of jets on the cartridge influences the heat transfer and pressure drop characteristics of the divertor. Five different cartridge designs are analyzed in the paper. The most critical parameters, such as structure temperature, heat removal ability, pressure drop, cooling efficiency and thermal stress loadings in the cooling finger are predicted for each cartridge design. A combined computational fluid dynamics and structural model was used to perform the necessary numerical analyses. The results have shown that the cartridge design with the best heat transfer and pressure drop characteristics is not also the most favorable choice from the point of view of minimum stress peaks. [Copyright &y& Elsevier]

Abstract: Several advanced He-cooled W-alloy divertor concepts have been considered recently for power plant applications. They range in scale from a plate configuration with characteristic dimension of the order of 1m, to the ARIES-CS T-tube configuration with characteristic dimension of the order of 10cm, to the EU FZK finger concept with characteristic dimension of the order of 1.5cm. The trend in moving to smaller-scale units is aimed at minimizing the thermal stress under a given heat load; however, this is done at the expense of increasing the number of units, with a corresponding impact on the reliability of the system. The possibility of optimizing the design by combining different configurations in an integrated design, based on the anticipated divertor heat flux profile, also has been proposed. Several heat transfer enhancement schemes have been considered in these designs, including slot jet, multi-hole jet, porous media and pin arrays. This paper summarizes recent US efforts in this area, including optimization and assessment of the different concepts under power plant conditions. Analytical and experimental studies of the concepts and cooling schemes are presented. Key issues are identified and discussed to help guide future R&D, including fabrication, joining, material behavior under the fusion environment and impact of design choice on reliability. [Copyright &y& Elsevier]

In the frame of the activities promoted and encouraged by the EUROfusion Power Plant Physics and Technology (PPPT) department aimed at developing the EU-DEMO fusion reactor, strong emphasis has been recently addressed to the whole Balance of Plant (BoP) which represents the set of systems devoted to convert the plasma generated thermal power into electricity and to deliver it to the grid. Among these systems, the Divertor Primary Heat Transfer Systems (PHTSs) are intended to feed coolant to the two main components of the Divertor assembly, namely the Plasma Facing Components (PFCs) and the Cassette Body (CB). Since the DEMO Divertor must withstand high heat flux loads together with a considerable neutron deposited power, very tight tolerances may be allowed to the coolant inlet conditions. Therefore, the design of reliable PHTSs is of the utmost importance towards the development of an EU-DEMO fusion reactor. Within this framework, a study has been jointly carried out by University of Palermo, Ansaldo Nucleare and CREATE to design the Ex-Vessel PHTSs of both the PFCs and the CB for a DEMO reactor equipped with a Helium Cooled Pebble Bed Breeding Blanket. The paper describes criteria and rationale followed with the aim to achieve simple PHTS designs based on the adoption of easy-to-manufacture main components avoiding too much extrapolation from the state-of-art technology. Results of preliminary thermal-hydraulic calculations carried out to size heat exchangers, pressurizers, piping and pumps are presented and critically discussed, with particular attention to those integration, safety and feasibility constraints that may deeply affect the design of such components. Finally, the evaluation of PHTS key parameters as total pumping power and coolant inventory is reported. [ABSTRACT FROM AUTHOR]

Abstract: The ARIES-AT power core was evolved with the overall objective of achieving high performance while maintaining attractive safety features, credible maintenance and fabrication processes, and reasonable design margins as a rough indication of reliability. The blanket and divertor designs are based on Pb-17Li as coolant and breeder, and low-activation SiCf/SiC as structural material. Flowing Pb-17Li in series through the divertor and blanket is appealing since it simplifies the coolant routing and minimize the number of cooling systems. However, Pb-17Li provides marginal heat transfer performance in particular in the presence of MHD effects and the divertor design had to be adapted to accommodate the peak design heat flux of 5MW/m2. The blanket flow scheme enables operating Pb-17Li at a high outlet temperature (about 1100°C) for high power cycle efficiency while maintaining SiCf/SiC at a substantially lower temperature consistent with allowable limits. Waste minimization and additional cost savings are achieved by radially subdividing the blanket into two zones: a replaceable first zone and a life of plant second zone. Maintenance methods have been investigated which allow for end-of-life replacement of individual components. This paper summarizes the results of the design study of the ARIES-AT power core focusing on the blanket and divertor and including a discussion of the key parameters influencing the design, such as the SiCf/SiC properties and the MHD effects, and a description of the design configuration, analysis results and reference operating parameters. [Copyright &y& Elsevier]

• Assessment of Hypervapotron elements for high heat flux application. • Application FEA, CFD tools and experiments for evaluation of heat transfer phenomena. • Understanding on applicability of empirical correlation used for subcooled boiling regimes for Hypervapotron geometry. Indian Test Facility (INTF) is a testbed to carry out neutral beam experiments where beam with peak power densities reach about 66 MW/m2 and incident on calorimeters. To survive such thermal fluxes, Hypervapotron (HV) elements are devised in designing calorimeters. The HV element has a finned surface inside and working in a subcooled nucleate boiling regime with water as a coolant. HV elements are efficient heat transfer technologies made from high thermal conductive materials (CuCrZr – Copper alloy) and capable of removing heat flux densities about 10−30 MW/m2 w.r.t various coolant flow conditions. In this paper, efforts are undergone to evaluate heat transfer performance of these elements in three approaches – a) 3D FEA using Empirical correlations available in literature, b) 2D CFD in ANSYS CFX, c) Conducting high heat flux experiments on the HV element and compared the methodologies and estimate the safe operating boundaries for INTF operation. [ABSTRACT FROM AUTHOR]

• A 3D FEM code for the heat flux analysis of the tungsten monoblock divertor in the EAST has been successfully developed. • The 3D FEM code can calculate the heat flux on the divertor with the surface temperature obtained by the IR camera. • The heat flux evolution and its distribution on the divertor due to ELM losses in the EAST is discussed. The calculation of heat flux on the divertor is critical to study the characteristics of energy transport in controlled fusion plasma. Since 2014, the upper divertor of Experimental Advanced Superconducting Tokamak (EAST) has been upgraded into water-cooled W/Cu monoblock, and therefore the original two-dimensional(2D) heat calculation code does not apply to the heat flux calculation on the upper divertor due to the asymmetry toroidal geometry. Thus, it is urgent to develop a Three-dimension(3D) heat flux calculation code based on Finite Element Method(FEM) to determine the 2D heat flux distribution at the divertor surface in EAST. This paper presents a detailed design and development of this 3D FEM heat flux calculation method with the full consideration the monoblock geometry of the upper divertor. Based on the surface temperature captured from the IR camera, and the effect of a thin surface layer on the divertor, the heat transmission coefficient was determined in the solution of heat conduction equation. To benchmark this code, the calculation results are compared with the results using ANSYS software with the same conditions(constant temperature boundary and constant heat flux boundary). The comparison results show good agreements that the maximum fractional error between the two methods is less than 4 %. Finally, the characteristics of heat flux distribution on the divertor due to edge localized modes(ELMs) losses on EAST are also demonstrated. [ABSTRACT FROM AUTHOR]