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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 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]

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]

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]

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]

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]

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]

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]

Two models are currently taken as reference for the breeding blanket (BB) of DEMO, the Helium Cooled Pebble Bed (HCPB) and the Water Cooled Lithium Lead (WCLL). For both the models two Balance of Plant (BOP) configurations are currently investigated. They are all based on four independent primary heat transfer systems (PHTSs). The largest PHTS is devoted to remove the BB thermal power, two PHTSs for the divertor (DIV) and one for the vacuum vessel (VV) heat removals. Both DIV loops and VV loop transfer their thermal power directly to the power conversion system (PCS). The two BOP configurations are investigated to cope with the pulsed operation of the reactor. The first configuration consists of an "indirect coupling" of the BB PHTS to the PCS by an Intermediate Heat Transfer System placed in between them. The second configuration consists in a "direct-coupling" of the BB PHTS to the PCS. Reliability, availability, maintainability and inspectability (RAMI) analyses for the four PHTS and PCS configurations are presented in this paper. The Failure Mode, Effect Analysis (FMEA) and the Reliability Block Diagram (RBD) methodologies have been used to perform the studies. As main result can be highlighted that even the low reliability of the systems to operate without failure and the large number of requests of corrective maintenance, all the systems are in compliance with an operational availability target of 30 %. [ABSTRACT FROM AUTHOR]

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]