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• The design basis accident (In-vessel LOCA) has been introduced. • According to the design of CFETR systems, the accident initial conditions, assumptions and analysis results are given. • The accident consequences are shown, and the advice of HCS header pipe design are given. For the helium-cooled ceramic breeder (HCCB) blanket concept improved in 2018 and the primary heat transfer system following China Fusion Engineering Test Reactor (CFETR) in 2019, an in-vessel loss of coolant accident (LOCA) has been investigated with the assumption of a break of the first wall (FW) coolant channels and a double-ended guillotine break of a helium cooling system (HCS) pipe in vacuum vessel (VV). The break leads to helium inject into the VV. After the accident occurs, the helium release leads a plasma disruption and pressure increase in the VV. This paper presents the analysis results of the in-vessel LOCA accidents in CFETR and gives the accident sequence compare with the break in the FW and header pipe. The accident analysis and influence compares can provide a reference for the iterative design. [ABSTRACT FROM AUTHOR]

• Loads that act on a breeding blanket during lifting have been calculated. • Multiple manipulator designs have been studied, and two parallel mechanisms and one linear mechanism have been developed and upgraded. • Hexapod design: Rigid, stable solution for handling blankets in tokamaks, eliminates the need for additional manipulator. • Tri-arm concept: fewer drives are needed, but the risk of collisions during seismic loads, requiring adjustment from neighbouring ports. • Linear LMS: easy to control but requires additional cables for guiding blankets in higher positions. Large port-based tokamaks for energy production are expected to be developed around breeding blanket technology, the purpose of which is to breed Tritium, provide shielding to the vessel walls, and carry heat transfer loops. These blankets are expected to be split into five segments in every torus section [1] , each weighing up to 60 tons, and will require periodic replacement through the Upper Port for maintenance. During replacement, the blankets need to be guided in the vacuum vessel through highly restricted kinematic paths to avoid collisions. A Two Port Remote Handling approach is therefore being investigated. In this approach, the Lower Manipulator System (LMS) is expected to operate through the Lower Port to guide the blankets and react to the severe loads arising from the offset between the blankets' Centre of Gravity (CoG) and the lifting point. Several mechanisms for the LMS have been developed to comprehensively explore blanket handling capabilities using this approach. This paper presents the different LMS concepts and discusses the respective salient features. [ABSTRACT FROM AUTHOR]

The Helium Cooled Pebble Bed (HCPB) breeding blanket is a candidate as driver blanket for the EU DEMO. The reference design of the HCPB is based on a cooling plate "sandwich" arrangement built in Multi-Module Segments. This architecture significantly improved the tritium breeding performance (TBR = 1.15) and the plant circulating power (≈130 MW) compared to the former ITER-like "beer-box"-like design (TBR<1.10, plant circulating power>200 MW). However, several issues remain with this design, in which (1) the still large power required per He circulator (beyond the state-of-the-art for these components) and (2) the large tritium inventory foreseen in Be have been identified as the most critical. After a basic research for a concept that mitigates these key issues, a fission-like configuration based on a hexagonal matrix of radial fuel-breeder pin assemblies has been found to solve many of the outstanding issues present in the reference design. This design has been developed for the latest EU-DEMO 2017 baseline. It features Be 12 Ti neutron multiplier, which largely reduces the tritium retention, swelling and reactivity issues compared to pure Be. The substitution of the relatively complex CPs by simpler fuel-breeder pins improves the heat transfer capability of the breeder zone (BZ), allowing 20°C higher blanket outlet temperature and drastically reducing an order of magnitude the pressure drop in the BZ. The paper reports the successful set of nuclear, thermo-hydraulic and thermo-mechanical performances, together with some notes about the manufacturing and assembly strategy, the improved RAMI features and emphasizes the reduction of the plant circulating power to a remarkably low figure (˜90 MW), enabling the use state-of-the-art He turbomachinery. Due to some inherent issues connected to the use of Be as neutron multiplier which cannot be simply solved by design, a parallel long research has identified molten lead as a cost-effective, low risk substitute for Be. Taking as basis the HCPB fuel-breeder pin concept, an alternative helium cooled Molten Lead Ceramic Breeder has been derived and some key performance figures are reported. The paper concludes with a discussion on the results and their implications on future development steps. [ABSTRACT FROM AUTHOR]

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]

• Based on the heat transfer theory and Thermal-Electric module, a simple method for quickly predicting the maximum temperature of water-cooled busbar with connector is proposed, which is convenient for engineering application. • Compared with V-T method and experimental results, the simple method in this paper has the advantages of global temperature distribution analysis and better temperature prediction accuracy, and can be used for the design of the same type of transmission lines. • The relationship among cooling water flow rate, contact resistance and maximum temperature of CRAFT busbar under rated working condition is determined, which can provide reliable guidance for actual operation of busbar. Due to the shortcomings of time-consuming and complicated operation of traditional fluid software simulation, a simple temperature prediction method that can quickly determine the reasonable working parameters of water-cooled busbar with connector is needed in the design of fusion power supplies. In this paper, considering the influence of contact resistance, the connector maximum temperature under same working conditions is studied by two kind of methods (combination method of heat transfer theory and Thermal-Electric module in ANSYS, Contact Voltage-Temperature (V-T) method), respectively. The comparison between simulation and experimental results shows that this method has the advantage of analyzing the global temperature distribution and predicting the maximum value more accurately than the latter. Based on this method, the relationship among contact resistance, water flow rate, and maximum temperature of the busbar connector used in Comprehensive Research Facility for Fusion Technology (CRAFT) is determined. This method can be applied to the rapid temperature analysis of fusion power busbar connector, which greatly reduces the design cycle. [ABSTRACT FROM AUTHOR]

The main function of the ITER Erosion Deposition Monitor is to observe the erosion/deposition status of divertor vertical targets as well as the topology change and surface damage that may occur due to the plasma-wall interaction at these vertical targets. The main function of the optical box of this diagnostic system, which is mounted under the divertor dome of one of the divertor cassettes, is to provide a stiff base and multi-functional shielding for the optical components located inside of it. Since the optical box under the divertor dome is in a harsh environment, under high plasma, nuclear and EM loads, it becomes one of the critical parts of the optical system. The design challenges involve the tight space between the surrounding parts, the manufacturing of the complex cooling channels, the optical tolerances and adequate heat transfer between the different components. To produce a robust design and fulfill all the requirements, the optical box went through several design iterations and simulation stages. This paper outlines the challenges and the results of the development of the mechanical design of the components of the optical box of the Erosion Deposition Monitor, EDM , diagnostic system. The EDM is an optical diagnostic system whose accuracy depends on the integrity of the mechanical parts. This paper describes the challenges of the development of the optical box itself, the mirror supports and the shutters. We present design solutions for key components that decrease the temperature of the optical box and its internal parts. These components are the cooling channels, which have been integrated into the walls to increase the cooling of the box, and the redesigned mirror supports, which have better heat transfer between the water-cooled walls and the mirrors. [ABSTRACT FROM AUTHOR]

Experimental Advanced Superconducting Tokamak (EAST) is one of the important steady-state divertor Tokamak physical experiment platforms in the world. The main functions of the high-temperature and high-pressure Divertor Primary Heat Transfer System (DIV PHTS) are to provide high-pressure cooling water for the divertor to remove the heat load deposited on the divertor under the plasma operation mode and to provide high-temperature nitrogen gas for the divertor system to achieve vacuum condition during the baking mode. The stability and reliable operation of the DIV PHTS play an important role in the EAST tokamak plasma physics experiment. The high-temperature and high-voltage electrical insulation break are one of the important components of the system. The main function of the Electrical Insulation Break (EIB) is to break the closed circuit of the heat transfer system to avoid eddy current during plasma discharge affecting the magnetic field distribution characteristics of the tokamak. On the other way, it also needs to meet the requirement, there is no leak of the cooling water and nitrogen gas during cooling and baking operation mode. This paper proposes the design of an electrical insulating break (EIB) to meet the design requirements of EAST tokamak at first. Then the electric field analysis is carried out by using a finite element model to evaluate its electrical insulation characteristics. And the sealing structural characteristics are verified by applying the finite element method under two design modes. Finally, the sealing test result under the high-temperature baking and high-pressure cooling modes are presented to validate the design. [ABSTRACT FROM AUTHOR]

Indian Test Facility (INTF) is a test bed to characterize ITER Diagnostic Neutral Beam (DNB) system, is being constructed at ITER-India, IPR [1] . The operation is aimed to generate 18–20 A of 100 KeV, hydrogen neutral beam with a 3 s ON/20 s OFF duty cycle at 5 Hz modulation. The final focused beam will be characterized on a beam dump or a heat sink system named Second Calorimeter (SC), situated at a distance of 20.66 m from the Grounded Grid (GG) of the Beam Source (BS). It is designed to withstand a total beam power or heat load of 2.84MW, corresponding peak heat flux is ∼ 93.15MW/m 2 and the required number of thermal cycle of 6.13 × 10 4 . To handle such high heat flux, a V shaped design is adopted with a suitable apex angle. In INTF, First Calorimeter (FC) is placed just after Residual Ion Dump (RID) at a distance 5.95 m from the GG of BS [2] . In the present work a heat absorbing building block called hypervapotron was adopted from [3] and modified to meet the above mentioned heat load requirements. The paper presents the conceptualization of SC and a thermo-mechanical assessment of the heat transfer element by Finite Element method (FE), carried out using ANSYS. FE results are used to validate the HTE design as per Structural Design Criteria for In vessel Component (SDC-IC) for monotonic as well as cycling loading (Fatigue) damage criteria. This paper also reports the key challenges and usual solutions adopted in the design. [ABSTRACT FROM AUTHOR]

A set of novel design solutions for high performance cooling systems have been developed and tested by Consorzio RFX, achieving, with experimental tests, the challenging heat transfer conditions foreseen for Heating and Current Drive Systems of present and future nuclear fusion devices. The project, called Multi-design Innovative Cooling Research & Optimization (MICRO), has the triple objective of: • Verifying the full qualification of the manufacturing process and assess the cooling performance of the present solutions applied on the acceleration grids designed for ITER Heating Neutral Beam Injectors and MITICA experiment, which have similar heat loads as for DEMO1. • Identifying alternative designs realizable with presently qualified manufacturing processes for improving heat transfer mechanisms with limited pressure drops. • Developing further optimized solutions by means of the investigation of the interrelated effects of geometric parameters and thermofluid dynamic properties of the coolant fluid. The main aim of present work is to extend the fatigue life-cycle of high thermal stress components. In the first part of the paper the design description and constraints are presented together with the results obtained from numerical analyses carried out on the full-scaled model of the electrostatic grids; the outcomes of the experimental campaign, together with the characterization of the properties and performance of a set of alternative dielectric fluids, are shown in the second one. [ABSTRACT FROM AUTHOR]

• The preliminary Pb15.7Li Loop layout was shown in the paper. • The Magneto Hydro Dynamic effect induced in the flowing Pb-16Li inside BB and the pressure drops generated were analysed. • The developed anti-permeation and corrosion barriers inside the BB where shown. • The code developed to analyse the In-box Loca inside WCLL BB was introduced. • Preliminary integration of Pb-16Li loops inside the tokamak building was shown. Three of the four breeder blanket concepts currently under investigation for the European DEMO Reactor use the eutectic Pb-16Li as breeder material. Those are the Helium Cooled Lithium Lead (HCLL), Water Cooled Lithium Lead (WCLL) and Dual Coolant Lithium Lead (DCLL) blankets. Moreover, the WCLL is one of the blanket concepts that will be qualified in the ITER reactor, therefore the development and design of lead lithium loops and auxiliary systems is essential. The main functional requirements that Pb-16Li systems have to fulfill are: • to circulate the liquid Pb-16Li through the blanket and ancillaries; • to extract the tritium produced inside the breeder modules from Pb-16Li; • to control Pb-16Li chemistry and to remove accumulated impurities; The present work aims to describe the activities performed in order to achieve the following objectives: i) design and integration of the Pb-16Li loops inside the tokamak building, ii) development and characterization of antipermeation and anticorrosion coatings on structures in contact with Pb-16Li, iii) development and design of an activation products removal system, iv) design of a chemistry control system for Pb-16Li loops, v) performing magnetohydrodynamic analyses taking into account the impact on heat transfer and tritium transport in breeding blankets and performing safety analyses of water/Pb-16Li interaction due to LOCA inside the WCLL blanket. [ABSTRACT FROM AUTHOR]

• The progresses of the Primary Heat Transfer System of the HCPB Breeding Blanket for the DEMO reactor are outlined. • The main components of Ex-Vessel PHTS has been sized to be compliant with the new In-Vessel requirements. • Results highlights that the 2017 HCPB BB PHTS has reached improved thermal-hydraulic and safety characteristics. • Further efforts are still needed to reduce PHTS pressure drops and volumes as well as to improve IHXs performance. In the frame of the activities promoted and encouraged by the EURO-fusion Power Plant Physics and Technology (PPPT) department aimed at developing the EU-DEMO fusion reactor, strong emphasis has been recently posed 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, a very important role is played by the Breeding Blanket (BB) Primary Heat Transfer System (PHTS) as it is responsible to extract more than 80% of the fusion plasma power. In this framework, University of Palermo, Ansaldo Nucleare and CREATE have focused their work to improve thermal-hydraulic, safety and integration features of the Ex-Vessel PHTS for the Helium-Cooled Pebble Bed (HCPB) BB concept of DEMO. Starting from the outcomes obtained in 2016 that have allowed to highlight some criticalities which needed design changes, the paper describes progress and developments of the 2017 HCPB PHTS as well as the goals which have been achieved. The results of the research activity carried out show, in fact, i) an increase of the thermal-hydraulic performances, since a reduction in both total coolant inventory and total pressure drop has been respectively reached, ii) a potential improvement of the overall safety characteristics of the system. Nevertheless a critical assessment of these key parameters reveals that some issues are still open in terms of design integration and feasibility of the whole DEMO BoP for the helium-cooled blanket option, indicating that additional efforts are required to make this technology more attractive. [ABSTRACT FROM AUTHOR]

• 3D CAD of the Primary Heat Transfer System of EU-DEMO fusion power plant is presented. • Helium-Cooled Pebble Bed (HCPB) Breeding Blanket (BB) option for 18-sectors tokamak is considered. • A first estimation of lengths and weights of the piping systems as well as their space occupancy is provided. • A rough estimation of the coolant inventory is also provided. This paper focuses on the 3D CAD implementation of the pipework and the main equipment of the Primary Heat Transfer System of EU-DEMO fusion power plant. In particular, the systems related to the Helium-Cooled Pebble Bed Breeding Blanket option are considered here. During the pulse operation, the breeding blanket modules will be the main thermal power source; Divertor and the Vacuum Vessel will contribute in the definition of the total reactor power. All the In-Vessel generated power is rejected to the Power Conversion System through a molten salt Intermediate Heat Transport System. The latter is equipped with an Energy Storage System to allow for continuous operation also during the dwell time. The layout of the pipework considered the major design constraints, such as the reinforced concrete structures and the dimensions of the main pieces of equipment. The size of the pipes was selected based on operating pressure, temperature and allowable velocity. Results provide a first estimation of length and weight of the piping systems as well as their space occupancy. These data are useful to evaluate the coolant inventory, as well as the expected pressure drops and plant cost. [ABSTRACT FROM AUTHOR]

The ITER Vacuum Vessel (VV) is supported by the nine VV gravity supports (VVGS) located on the cryostat toroidal pedestal. The VVGS is dual hinge type that fastened by dowel in the hinge-block hole. This paper presents the technical approach and result on the interference fitting process of the sleeve to the full-scaled VVGS mock-up. Since the sleeve and the hinge-block hole have under millimeter tolerance and around two-meter long length, the shrink fit method has been selected for the interference fitting of the sleeve into the hinge-block hole. The duration time for required shrinkage was expected by the simple analytical approach of natural convective heat transfer. The coolant charging method was suggested to secure a sufficient time for the process. The liquid nitrogen was charged to the handling fixture capped sleeve hole's cavity. As a result, the necessary contraction time was secured as thousands of seconds. Consequently, the interference fitting was successfully done for VVGS mock-up fabrication, and the technical solutions will be applied to the VVGS manufacturing. [ABSTRACT FROM AUTHOR]

• The paper concerns the development of a thermo-hydraulic model of the water-cooled EU DEMO BOP, using RELAP5/mod3.3 system code. • The thermo-hydraulic model is developed on the basis of WCLL BB 2016 design and related BOP, and DEMO 2015 baseline. • The research activity is carried out to validate components design of the PHTS. • Preliminary verifications on RELAP5 nodalization demonstrate that the code has the capability of simulating FW and BZ PHTSs and related main components. The conceptual design of the Primary Heat Transfer System (PHTS) of the water-cooled European (EU) DEMO foresees two independent cooling circuits, the breeding zone PHTS and the first wall PHTS. During the pulse time (120 min) the first delivers thermal power to the turbine, the latter delivers thermal power to the Intermediate Heat Transfer System (IHTS) equipped by an Energy Storage System (ESS). The IHTS delivers partially thermal power to the turbine in pulse so that to accumulate a suitable amount of energy in ESS to operate the turbine during the dwell time (10 min) at almost constant load, despite the EU DEMO pulsation of the generated thermal power. A dynamic model of the primary systems of water-cooled EU DEMO is developed using RELAP5/Mod3.3 system code to verify components sizing and to investigate code predictive capabilities. The model includes the primary and secondary side of the breeding zone and first wall, and in particular: in-vessel (i.e. breeding blanket) and ex-vessel components (i.e. main collectors, hot and cold legs, heat exchangers, steam generators, pumps). Preliminary assessments of the nodalization have been carried out, in particular checking pressure drops along the systems and heat exchanger performances. [ABSTRACT FROM AUTHOR]

DEMO is planned to be a prototype fusion power plant capable of demonstrating production of electricity at the level of a few hundred MW. DEMO is considered to be an intermediate step between the ITER experimental reactor and a commercial power plant. Design and assessment studies on the European (EU) DEMO are carried out by the EUROfusion consortium. The Primary Heat Transfer System (PHTS) transfers heat from the breeding blanket (BB) and other reactor heat sources (divertor and vacuum vessel) to the secondary circuit. Two main BB concepts, and the respective PHTSs, for EU-DEMO are considered: the Helium Cooled Pebble Bed (HCPB) BB and the Water Cooled Lithium Lead (WCLL) BB. Two variants for each concept are possible: with or without the Intermediate Heat Transfer System (IHTS), including the Energy Storage System (ESS), between the PHTS and the Power Conversion System (PCS), which generates electricity. The role of IHTS and ESS is to prevent energy pulses obtained by BB PHTS during the plasma burns to be directly transferred to PCS. In 2017 we proposed the first concept of the PCS configuration for the option WCLL BB with the ESS, which enabled almost constant generation of electricity during both plasma pulse and dwell phases. However, we observed some disadvantages of this concept, which had to be cured, i.e., excessive temperature differences between the pulse and dwell phases in several heat exchangers. In the present paper, we describe the GateCycle model of the improved PCS configuration for the WCLL BB with the ESS option, which ensures not only almost constant electricity production during both plasma pulse and dwell phases, but also acceptably small temperature fluctuations | T pulse - T dwell | in all the circuit components, which should prevent their failure due to excessive thermal stress. [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 thermal-hydraulic analyses on the prototype and test model are performed. • The PPIRT and scaling criteria for LOFA are identified. • The basic design scheme of test model is proposed. The water Cooled Ceramic Breeder (WCCB) blanket, as a candidate for Chinese Fusion Engineering Test Reactor (CFETR), must demonstrate its safety before engineering application. Enough data are necessary and the test model should ensure that the data are convincing within reasonable expenses. In this paper, scaling criteria are identified for loss of flow accident (LOFA) in WCCB blanket coolant system, which guides the design scheme of test model that the length scaling factor of 1 and volume scaling factor of 1/120. Scaling analyses are performed to develop similarity criteria for pump coastdown and loop natural circulation in three temporal phases identified by LOFA simulation. The test model only preserves a branch at the sector level and sub-module level, which results in different volume scaling factors at levels. Typical heat-transferring cells, constituted by blanket components, are preserved for heat transfer similarity. By computational simulation of an ideal model, it is testified that the test model can reproduce thermal-hydraulic responses within acceptable range of the distortion. [ABSTRACT FROM AUTHOR]

The identification of structures, systems and components (SSCs) performing safety functions is of paramount importance for an EU DEMO development in all phases of the project: design, fabrication, commissioning, operation, maintenance, inspections and tests. Early identification of safety relevant SSCs contributes in the correct definition of systems requirements, preliminary layout and design integration. It may also be exploited as a supporting criterion for selection among the plant variants, e.g. alternative blanket concepts. Those SSCs assigned to a safety importance class (SIC) will receive adequate attention during the different developing phases. A graduation of safety important SSCs is adopted for DEMO on the basis of the IAEA Guide No. SSG-30, on SSCs classification. Three safety importance classes (SIC-1, SIC-2 and SIC-3) are assigned on the basis of: a) the consequences the loss of the safety function could lead to, b) the involvement of the SSCs in preventing, detecting or mitigating incident or accident and, c) the involvement of the SSCs in bringing and maintaining the plant in a safe state. In this paper, the provisional safety classification of the Primary Heat Transfer System (PHTS) related to the Helium Cooled Pebble Bed (HCPB) and Water Cooled Lithium Lead (WCLL) blankets of DEMO is presented. [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]

The Experimental Advanced Superconducting Tokamak (EAST) had been under major upgrading from 2020 to 2021 in China. In this campaign of major upgrading, the new lower tungsten divertor is successfully installed to achieve the targets of long pulse and high β H-mode plasma operation. High temperature and high pressure Divertor primary heat transfer system (DIV PHTS) is also developed to remove the plasma deposition heat load on the Divertor. This paper presents a description of the design and construction of DIV PHTS for EAST Tokamak. Starting from the requirements specified in the Technical Specifications of DIV PHTS, the process to meet the Divertor cooling and baking requirement is designed and validated by the supported thermal-hydraulic analysis. Then, the piping stress analysis under different types of load combinations is performed to verify the structural design requirements. General layout design is also performed based on the component requirement and the allocated space in the EAST Tokamak hall. Finally, the procurement and test of the system main equipment is also given. [ABSTRACT FROM AUTHOR]

• Dynamic response mechanism of the EAST baking system in the baking process is developed. • The theoretical calculation model for transient heat transfer in the baking system is established. • The model can be applied for prediction of the baking system operation behavior. Before long-pulse plasma operation, wall conditioning is addressed as a key element to achieve steady state long-pulse operation with high performance, which is one of the major challenges for EAST (Experimental Advanced Superconducting Tokamak). The vacuum vessel and in-vessel components are continuously heated with hot nitrogen for a few weeks, namely the baking process, and the beginning of wall conditioning. It is difficult to accurately predict the specific temperature distribution of the baking system in the baking process under the current analysis and experimental conditions. Therefore, a simplified calculation model has been established in this paper, which is helpful to reveal the dynamic response mechanism and the heat loss to guide the input power of the baking system. The heat transfer relationship has been given by the theoretical method, and the calculation results of two cases are obtained by using Matlab/Simulink. The results gained by the heat transfer model were compared with the actual experimental data, which verified the correctness of the model. This heat transfer calculation model can be used in the future EAST experiment to predict the baking system operation behavior. [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 Plasma must be heated by external heating for ignition in the future fusion reactor, ICRF heating is a favourable high-density plasma heating method since the fast wave launched from ICRF antenna can be transmitted to plasma core even in high-density plasma. While heating power increasing, the exposed antenna surface endures higher heat load. Faraday Shield (FS), as one of the key components, faces plasma and undertakes high heat load during plasma discharge. High stress and deformation on the FS is produced without removing the heating, which drastically reduces the antenna performance, even damages FS structure to threat RF antenna safety. To explore high-efficiency heat transfer enhancement technology to remove heat on the FS, one finite element analysis was used to analyse heat and temperature distribution on FS by different heat transfer ways, optimizing the FS structure. Based on the analysis results, several novel heat transfer enhancement way are proposed. Some workpieces with proposed heat transfer enhancement structure were designed and manufactured for test. Afterwards, two mock-ups with one quarter dimension of EAST ICRF antenna were manufactured without and with optimum heat transfer enhancement technology based on former test results, they were tested under same high heat load, significant results are obtained and presented in the paper, results are shown that heat transfer enhancement technology are drastically improved heat transfer performance of FS, which lay the foundation to high-power, steady-state operation ICRF antenna design in the future. [ABSTRACT FROM AUTHOR]

Highlights • The Primary Heat Transfer System of the HCPB Breeding Blanket for the DEMO reactor has been preliminary designed. • A thermo-hydraulic assessment of Ex-Vessel PHTS has been carried out in order to size its main components. • Results show that HCPB BB PHTS is characterized by large IHXs heat transfer area, high coolant inventory, huge pumping power. • Further efforts are needed to reduce PHTS pressure drops and volumes as well as to improve IHXs performance. Abstract The European DEMO conceptual design foresees four Breeding Blanket (BB) concepts that rely on different cooling and breeding technologies. As DEMO has been conceived to deliver net electricity to the grid, the choice of the blanket coolant plays a pivotal role in the reactor design having a strong influence on plant operation, safety and maintenance. Moreover, the machine pulsed operation makes the BB Primary Heat Transfer System (PHTS) the main hub of the DEMO Balance of Plant (BoP). Within this framework, a study has been carried out to design the Ex-Vessel PHTS of the Helium-Cooled Pebble Bed (HCPB) BB concept. The paper describes criteria and rationale followed with the aim to achieve a simple PHTS design based on the adoption of easy-to-manufacture main components. Results of preliminary thermal-hydraulic calculations carried out to size heat exchangers, piping and compressors are presented and critically discussed. Finally, the evaluation of PHTS key parameters as total pressure drops and total coolant inventory is reported to provide input relevant to the assessment of the impact of helium technology on the design integration and feasibility of DEMO BoP. [ABSTRACT FROM AUTHOR]

Highlights • EU- DEMO shall be strongly Balance of Plant oriented. • Two BOP options are considered for both WCLL BB and HCPB BB BoP with the aim to minimize technological risks. • Preliminary design of HCPB/WCLL BB PHTS, IHTS, ESS and conceptual design of PCS have been perfomed. • BB PHTS, IHTS, ESS and PCS are considered to be feasible, but further investigations need before drawing any firm conclusions. • The direct coupling of the BB PHTS to PCS is under analysis with the help of the Industry. Abstract DEMO in Europe is considered to be the nearest-term reactor design to follow ITER and capable of demonstrating production of electricity, operating with a closed fuel-cycle and to be a facilitating machine between ITER and a commercial reactor. The aim of this paper is to show the design progress of the complex system's "chain" devoted to the extraction of the plasma generated pulsed thermal power and its conversion into electricity delivered to the grid, including the Primary Heat Transport System (PHTS), the Power Conversion System (PCS) and the Intermediate Heat Transport System (IHTS) – provided by an Energy Storage System (ESS) – in between PHTS and PCS, which is introduced for the scope of smoothing the generated pulsed plasma power removed by PHTS transmitted to PCS for a more continuous conversion/production of electricity and to guaranty plant lifetime. This will be done with reference to two breeding blanket concepts: the Helium Cooled Pebble Bed (HCPB) Breeding Blanket (BB) and the Water Cooled Lithium Lead (WCLL) BB. The PHTS design criteria and the system design finalization through a descriptive 3D CAD model which shows the system inside the tokamak building are presented. The design and operational challenges as well as the on-going development plans to investigate possible design simplifications (e.g. the direct coupling of Breading Blanket PHTS to PCS) and improved technology readiness actions potentially leading to better plant reliability and cost minimisation are briefly described. [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]

Highlights • Analysis of thermo-hydraulic behavior in helium-cooled divertor module with nozzle-to-target distance. • Revealing the detrimental effect with the change of cartridge diameter due to fluidic behavior. • Suggestion of thermal design guideline for nozzle-to-target distance in helium-cooled divertor. Abstract The power exhaust from high temperature plasma at the level of divertor is one of the biggest technological challenges in the path to establish fusion power reactors. At the Karlsruhe Institute of Technology (KIT), a novel helium-cooled divertor using jet impingement cooling array has been investigated as one of the possible solutions for a DEMO tokamak. The concept, consisting of tungsten armour slabs fixed on a tungsten-laminate tube inside which a jet distribution cartridge is placed, has undergone an extensive optimization process. Among other things, the heat transfer coefficient (HTC) and pressure drop over a divertor unit have been thoroughly looked upon in order to improve the divertor thermohydraulic performance. The present paper is presenting such an optimization study in which the effect on the divertor performances of the distance between the cartridge and the heat loaded surface is investigated. Given the geometrical constraints, the nozzle-to-target surface distance is changed either by increasing the diameter of the inlet manifold, or, by varying the cartridge position, starting from a coaxial configuration, to various positions closer to the impingement surface. [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]

Abstract The state of EU DEMO Balance of Plant as for the end of 2016 is described in this paper. Starting with 6 outboard and 3 inboard cooling loops, the thermal energy of the helium is transferred by an intermediate heat exchanger to the solar salt, which transports it out of the tokamak building to the energy storage system, which is actually an industrial 2-tanks design. Heat is transferred from solar salt to the feedwater in the steam generator where it boils and becomes superheated before entering the steam turbine. The steam turbine design follows state-of-the-art concept, but it is not finally optimized, since further DEMO BoP design changes still cannot be excluded. The other 3 usable heat sources (two from divertor and one from vacuum vessel) are used to heat-up the feedwater on the return line to the steam generator, thus contributing to the efficient utilization of heat produced during DEMO operation. Taking into account pulse/dwell times of 2.0/0.5 h, DEMO can operate permanently under ∼80% power thus allowing flexible and continuous plant operation. EBSILON® results are presented for steady-state conditions including both pulsed and dwell time periods. [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]

This paper deals with cooling of the first wall of Helium-Cooled-Pebble-Bed Test Blanket Module (HCPB TBM) for ITER. The first wall cooling is an important investigation issue due to an extreme asymmetry of heat loads: heat flux on the plasma facing side is several times stronger than the one on the side which faces breeding units. Our preliminary 3D CFD analysis revealed that under such conditions the heat transfer coefficient is significantly lower than predicted by common heat transfer correlations (see Ilić et al., 2006). For an experimental validation of these results HETRA (HEat TRAnsfer) test section has been designed and built at the Institute for Neutron Physics and Reactor Technology in Karlsruhe Institute of Technology. The HETRA test section involves in full scale one U-pass of the cooling channel in the first wall of HCPB TBM Version 1.1 (see Meyder et al., 2005). The HCPB TBM relevant experimental conditions have been provided: test channel made of Eurofer steel, helium coolant at pressure of 8 MPa and inlet temperature of 300 °C and heat flux of 270 kW/m 2 at the channel surface representing plasma facing side of the first wall. Test channels with hydraulically smooth and with hydraulically rough walls have been built. At each test channel the temperature of Eurofer walls has been measured at ∼60 positions. For numerical investigations the 3D CFD modelling with the code STAR CD has been applied. This paper is the first report on this study and presents the development of the test section and of the 3D CFD model. The analyses of the obtained experimental and computational results are presented in the second report (see Ilić et al., 2014). [ABSTRACT FROM AUTHOR]

• Baseline design for DONES SHRS has been selected to be similar to the IFMIF-EVEDA design. • From the key parameters (mass flow, temperature, basic properties of the heat transfer fluids) only the mass flow has been changed, according to the heat load difference between DONES and IFMIF. • The system (component selection, layout) has been designed to minimize the re-building effort during DONES – IFMIF change. The present paper summarizes the current status of the Secondary Heat Removal System (SHRS) of the IFMIF-DONES (International Fusion Material Irradiation Facility - Demo Oriented Neutron Source). As part of the Lithium systems (LS) in IFMIF-DONES, SHRS is the responsible sub-system for the removal of the heat which develops in the LS-Test Assembly (TA) during the Li-DT reaction. In this way, it has an important role in maintaining the stable lithium flow by providing controlled temperature at the inlet of TA. In SHRS, a somewhat conventional hot oil system serves as an intermediate heat removal sub-system between the main liquid lithium loop and the water cooling system. The latter serves as the final heat sink. During the preliminary design phase, the concept developed during the IFMIF-EVEDA (Engineering Validation and Engineering Design Activities) phase served as a basis, but its applicability was studied with special regard to the effect of the difference in the transferred heat of IFMIF and IFMIF-DONES on the selection of the thermal layout, from point of view of cost, complexity and safety. The selected thermal layout consists of two intermediate oil loops with three heat exchangers and two different organic oils as heat transfer mediums. Following the definition of the thermal layout, the main units were selected, a conceptual pipeline layout was designed which corresponds to the current building configuration and a preliminary thermal–mechanical assessment was performed in order to assist the design of the supporting structures. [ABSTRACT FROM AUTHOR]

An analytical estimate of the effective thermal conductivity of a pebble bed with mono-sized pebbles in the presence of a stagnant gas is presented. The Discrete Element Method (DEM) is used to simulate granular assemblies under uniaxial compression and Resistor Network (R-N) model is applied for estimating the effective thermal conductivity numerically. In this paper, two heat transfer paths, i.e., conduction through pebble–pebble contact and pebble–gas–pebble interface are considered. The effect of the cutoff range for gap and the strain on the effective conductivity is studied. The validity of the analytical model is verified with the results from the R-N Model for different solid-to-gas thermal conductivity ratios and for different packing fractions. The analytical and numerical models show a good agreement with the experimental results of the effective thermal conductivity for lithium orthosilicate pebble beds in the presence of helium and air at different temperatures. [ABSTRACT FROM AUTHOR]

In this paper, a Discrete Element Method (DEM) for the evaluation of the effective thermal conductivity of pebble beds in fusion blankets is presented. Pebble beds are multiphase materials in which both the solid and the gas phase filling the voids between particles coexist. The effective thermal conductivity of a pebble bed depends on the thermal properties of the two phases as well as on the system properties (e.g. gas pressure, temperature etc.). In particular, the pressure of the system is a key parameter for the heat transfer in a packed granular assembly since the thermal conductivity of a confined gas decreases with decreasing pressure (known as Smoluchowski effect). In this work, the influence of the gas pressure on the effective thermal conductivity in the Knudsen domain was implemented, to our knowledge, for the first time in a DEM code. The heat transfer mechanisms implemented are: when two particles touch each other the conduction through the contact area between them and, in any case, the conduction through the gas phase in the gap between neighbouring solid particles, may they be touching or not. These mechanisms are expected to dominate the heat transfer in a fusion breeder packed bed. Parametric studies were carried out to investigate the influence of the solid and gas materials, temperature, pressure and compression state. Numerical results were compared with existing experimental literature data and recent experiments carried out at Karlsruhe Institute of Technology (KIT). [ABSTRACT FROM AUTHOR]

The problem of source term qualification is one of the most important topics in order to predict possible releases of the Activation Products (APs) and Tritium from the DEMO Fusion reactor. The prevention of any possible consequence, which can affect the environment and the population, is the mission of Fusion technology. In the frame of the EUROfusion Work Package of Safety and Environment (WPSAE) a methodology to scale and evaluate the source terms assessed for ITER and the Fusion Power Plant Conceptual Study (PPCS) has been studied. This paper refers to the activity currently conducted for the DEMO source terms assessment and the preliminary results obtained. During activities in the task, the methodology was developed for the evaluation of Tritium and APs concentration inventory. The methodology is explained in detail for the prediction of the Tritium and APs concentration in Vacuum Vessel (VV) and in the Breeding Blanket (BB) starting from the DEMO current design data and the inventories assumed in ITER, PPCS and SEAFP programs. These results refer to the Helium Cooled Lead Liquid (HCLL) and the Helium Cooled Pebble Bed (HCPB) concepts. The approach is based on the foundations, set in the fission technology safety analysis of the Design Basis Accidents (DBA), Design Extension Conditions (DEC) and Beyond Design Basis Accident (BDBA). [ABSTRACT FROM AUTHOR]

Safety analysis for the design basis accident (DBA) is essential to support DEMO blanket concept design. It is necessary to study the pressure behaviour in the blanket and the connected systems during the loss of coolant accident (LOCA) in a blanket module, as well as the temperature evolution in the coolant flow and the associated structures. For the Helium Cooled Pebble Bed (HCPB) blanket concept (version 2014) three representative accidental sequences of LOCA have been simulated using system code MELCOR 1.8.6 for fusion. The LOCA is identified to be the failure of cooling channels in the stiffening grid, in the FW or in the breeder unit. Simulation results are discussed in this paper. [ABSTRACT FROM AUTHOR]

Since the stress due to the thermal load on breeding blanket structure is one of the main design driver, the thermal conductivity is necessary input data for thermal-structural and thermo-hydraulic analyses performed in order to understand the heat transfer phenomena and estimate the thermal stress. Since for the functional materials of solid type breeding blanket a pebble-bed form is mainly adopted instead of a bulk form such as a block or a disk, it should be needed to measure the thermal conductivity of pebble-bed. In this study, the effective thermal conductivity of pebble-bed is measured by laser flash method, which is one of the various thermal conductivity measurement methods, because this method has several advantages such as a wide thermal conductivity range of the measurement and a small amount of pebbles. A sample holder considering the heat transfer mechanism from the laser source to pebble-bed has been specially designed in order to apply the laser flash technique to the pebble-bed sample and it has been validated by the experiments. This paper introduces preliminary results of the effective thermal conductivity on the pebble-bed using this sample holder. [ABSTRACT FROM AUTHOR]

The blanket concepts investigated under the EUROfusion program rely on water or helium as primary or secondary coolant medium; main duty of the coolant is to recover the thermal power from the first wall and the blanket units and drive it into the Primary Heat Transfer System (PHTS). The coolant path goes through three different systems: the breeder, the tritium plant and the PHTS. In the breeder region, part of the produced tritium can permeate into the coolant loop, from the coolant into the steam generator and thus reach the environment. The Coolant Purification System (CPS), located inside the tritium plant, is responsible to recover tritium from the coolant and to control the coolant chemistry. The paper provides the status of the CPS design: the technology review was completed last year, while now the CPS interfaces and requirements are defined. Main issues arise from the definition of the amount of coolant to be routed inside the CPS due to the large uncertainties related to the amount of tritium permeation rate from blanket into coolant and to the presence and effectiveness of anti-permeation barriers. [ABSTRACT FROM AUTHOR]

Increased cooling performance is eagerly required for many cutting edge engineering and industrial technologies. Nanofluids have attracted considerable interest due to their potential to enhance the thermal performance of conventional heat transfer fluids. However, heat transfer in nanofluids is a controversial research theme, since there is yet no conclusive answer to explain the underlying heat transfer mechanisms. This study investigates the physics behind the heat transfer behavior of Al 2 O 3 –H 2 O nanofluids under natural convection. A high spatial resolution flow velocimetry method – Particle Image Velocimetry – is employed in dilute nanofluids inside a Rayleigh-Benard configuration with appropriate optical access. The resulting mean velocity and flow structures of pure water and nanofluids are reported and their overall heat transfer performances are compared for Rayleigh numbers, Ra , of the order of 10 9 . This paper aims to identify the contribution of the suspended nanoparticles on the heat and mass transfer mechanisms in low flow velocity applications, as those occurring during natural convection. The outcome of this work is a first step towards the evaluation of the applicability of nanofluids in applications where more complex heat transfer modes, namely boiling and Critical Heat Flux, are involved that are of great importance for the cooling of Fusion reactors. [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]

The subcooled flow boiling heat transfer of water has many applications, such as cooling devices of nuclear power systems and electronics, and the applicability of correlations of subcooled flow boiling heat transfer to water needs to be assessed. A number of correlations for subcooled flow boiling heat transfer were proposed. However, their prediction accuracy for water was not systematically investigated. This paper presents a comprehensive review of correlations of subcooled flow boiling heat transfer. A large database of subcooled flow boiling heat transfer of water containing 1184 data points is compiled from 14 published articles, based on which 21 correlations of subcooled flow boiling heat transfer are assessed, and their prediction accuracy is analyzed for different channel sizes, channel shapes, and flow directions. The results are helpful for understanding the prediction methods of subcooled flow boiling heat transfer and choosing a proper correlation for a given application. [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]

• Radiation analysis of SMARTEX-C to estimate the maximum temperature on different parts. • Detailed thermal hydraulic analysis of cooling and baking scheme attached externally around the vessel and over the flanges of SMARTEX-C. • A possible cooling scheme has been designed to resolve the high temperature of different parts of vessel with limited external surface area. • Pressure drop, heat transfer coefficient and temperature have been evaluated for the designed cooling and baking scheme. In this paper, a unified cooling and baking scheme has been designed and analyzed, which aids in maintaining the vacuum levels of an experimental system SMARTEX-C. It is a SM all A spect R atio T oroidal Ex periment in a 'C' (SMARTEX-C) shaped geometry. It is a trap meant for confining electron plasma i.e. non-neutral plasma. A vacuum ∼ (1.5 ± 0.5) x 10−10 mbar is maintained inside the vessel. Electrons are emitted from a heated tungsten filament, which eventually deteriorates the vacuum of the vessel with time. To perform reproducible electron plasma experiments it is essential to maintain vacuum levels constant. The designed cooling scheme helps in maintaining it. Additionally, flow of hot Nitrogen has been incorporated in this scheme which serves baking of the vacuum vessel making the dedicated baking system redundant and makes the operation of the system hassle-free. The scheme is modeled in CATIA© and analyzed in ANSYS CFX©. The thermal-hydraulic performance has been analyzed and parametric results in terms of temperature, heat transfer coefficient and pressure drop have been provided for the cooling and baking of the vessel. [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]

• A model of the ITER chiller water system (CHWS-1B) has developed using RELAP5 mod3.3. • Several scenarios of decay heat exchanger leak/break in the drain tank room have investigated. • Event evolution has analysed, required measures to isolate the affected zone have defined. • Accident management strategy is included in the analysis effort as well as a definition of the instrumentation and control fulfilling a relevant role in the event detection and mitigation. Chilled Water System (CHWS-1) is an ITER safety cooling system comprised of two independent and segregated trains (CHWS-1A and CHWS-1B), which provide cooling water to components located in Tokamak and Hot Cell Complexes. VV-PHTS is the only component that, during water baking mode, operates at high pressure than CHWS-1. In case of leaks, it represents a risk in terms of over pressurization and contamination of the CHWS-1 (due to tritiated water and activated corrosion products (ACPs)). Therefore, this paper postulates CHWS-1 leak events at one interface: Vacuum Vessel Primary Heat Transfer System. (VV-PHTS). Three scenarios have been analysed: an incident event (heat exchanger pipe leakage) and two accident events (heat exchanger tube break and multiple tubes break). A simplified RELAP5–3D model of CHWS-1B was developed to perform thermal hydraulic transient analysis as well as to define and size system operation and control. A postulated accident management strategy is included while also defining CHWS-1 instrumentation and control (I&C), which play a relevant role in event detection and mitigation procedures. The simulation results are studied to analyze and verify the system behavior as well as the required actions to reach and maintain a safe state after the given events. With respect to the events studied, this paper reports and confirms the system design to properly and safely bring CHWS-1B to a shutdown, as per corresponding ITER project requirements. This operating scenario has been studied for academic purposes; it is not related to the actual CHWS-1B design. During VV-PHTS baking mode, decay HX is not foreseen to be operative. In order to avoid affecting the redundancy of the system, this study supports this decision. [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]

Abstract: This paper presents an experimental test section which has been developed at the Institute for Neutron Physics and Reactor Technique in Karlsruhe Institute of Technology for investigations of heat removal from the first wall (FW) of the Helium-Cooled-Pebble-Bed Test Blanket Module (HCPB TBM). The test section named HETRA is attached to HEBLO facility which can provide Helium parameters relevant for HCPB TBM conditions–pressure of 8MPa and inlet temperature of 300°C. The purpose of planned investigations is to verify numerically found decrease of heat transfer coefficient due to a pronounced asymmetry of heat loads at the first wall of HCPB TBM. This paper describes how HETRA test section has been built: from its design over fabrication to the assembling. The planned measurements are also explained by presentation of a detailed measuring scheme. Finally, the first results concerning experimental and computational determination of pressure losses in HETRA cooling channel are presented. [Copyright &y& Elsevier]

Abstract: The total power loads applied to inertially cooled components are commonly calculated starting from calorimetric measurements as cooling fluid temperature increase and flow rate during and after the power load application. In this paper, some generalizations to the standard models of measure interpretation are introduced. A method is presented to take into account the effect of thermal conduction and radiation between the component and the surrounding material. It consists in evaluating the conduction and convection thermal resistances and the radiation heat flux by Computational Fluid Dynamics (CFD) numerical models. An analytical model permits to use these evaluations to obtain, from the measured convection flux, the corresponding fluxes by conduction and radiation. Sometimes the durations of pulses and duty cycles are such that the calorimetric measurements after one pulse are actually influenced by the previously applied pulse. A correction factor is introduced to cancel this systematic error. The cooling down phase can be long compared to the duty cycles between the pulses, and consequently the signals cannot be completely recorded till the thermal equilibrium. In order to evaluate the total heat flux on a component also in this case, a model is presented which considers a proper extrapolation curve to estimate the integral of the energy exhausted by the cooling systems. As an example, the models introduced in the paper have been used to interpret experimental measurements carried out on the RADI experiment at IPP Garching and some results are presented in the paper. [Copyright &y& Elsevier]

Abstract: Within the magnetic fusion energy program in the US, a program called APEX is investigating the use of free flowing liquid surfaces to form the inner surface of the chamber around the plasma. As part of this work, the APEX Team has investigated several possible design implementations and developed a specific engineering concept for a fusion reactor with liquid walls. Our approach has been to utilize an already established design for a future fusion reactor, the ARIES-RS, for the basic chamber geometry and magnetic configuration, and to replace the chamber technology in this design with liquid wall technology for a first wall and divertor and a blanket with adequate tritium breeding. This paper gives an overview of one design with a molten salt (a mixture of lithium, beryllium and sodium fluorides) forming the liquid surfaces and a ferritic steel for the structural material of the blanket. The design point is a reactor with 3840MW of fusion power of which 767MW is in the form of energetic particles (alpha power) and 3073MW is in the form of neutrons. The alpha plus auxiliary power total 909MW of which 430MW is radiated from the core mostly onto the first wall and the balance flows into the edge plasma and is distributed between the first wall and the divertor. In pursuing the application of liquid surfaces in APEX, the team has developed analytical tools that are significant achievements themselves and also pursued experiments on flowing liquids. This work is covered elsewhere, but the paper will also note several such areas to indicate the supporting science behind the design presented. Significant new work in modeling the plasma edge to understand the interaction of the plasma with the liquid walls is one example. Another is the incorporation of magneto-hydrodynamic (MHD) effects in fluid modeling and heat transfer. [Copyright &y& Elsevier]