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The management of digestate from agroforestry and food waste has become a challenge for the globally mitigation of economy and environmental impacts. This study is aimed to investigate the proximate and ultimate parameters of digestate sample obtained from zoo-technical digesters, fed with a mix of 70% bovine slurry, 30% corn and by-products. The analysis of characterization was carried out to determine moisture content, elemental composition (C, H, N, S content), metals content, anions content and thermogravimetric analysis. The research would provide a preliminary energy and environmental evaluation of digestate to choose an appropriate technology for its recovery. After characterization, the scientific innovation is represented, therefore, by using the digestate as an energy resource, for example to suggest the digestate conversion to biochar by means thermochemical process (pyrolysis, gasification) as a precursor for porous soil amendment or solid fuels. The main objective of this study was to experiment the potential reuse and sustainable disposal strategy of digestate by an eco-friendlier approach. The results obtained from the physicochemical characterization tests shown that the digestate have a high nutrient concentration as the macro elements (Na, K, Ca) and in average more concentration of microelements for liquid digestate., Proceedings of the 30th European Biomass Conference and Exhibition, 9-12 May 2022, Online, pp. 200-204

The detailed CFD simulation of the shell-side flow inside shell-and-tube heat exchangers results prohibitive for large-size components, due to the large number of tubes, together with the multi-scale nature of the flow features. To overcome this issue, the porous media approach can be adopted, substituting the tube bundle with a porous domain, where a suitable momentum sink term in the Navier-Stokes equation is provided. In this paper, a novel numerical approach to estimate the crossflow shell-side pressure drop is presented: a complete characterization of the tube bundle was performed by running 2D-CFD steady-state isothermal analyses, collecting pressure gradient magnitudes and directions for different mass flow rates and angles of attack of the incoming flow. The approach was validated against numerical results and compared to the methodologies currently available. Models, assumptions, and boundary conditions are herewith reported and thoroughly discussed, alongside the main results obtained.

An early attention to the layout of both plant site and its buildings is essential in a complex plant under preliminary design as DEMO in order to meet the assigned targets, namely i) the licensing requirements ii) a good availability in delivery electricity to the grid. The layout definition has to follow several criteria that become more complex and stringent for nuclear buildings, e.g. functional, maintenance, fire protection, safety, human factors, shielding, and remote handling. The criterion As Low As Reasonable Achievable with respect to the dose to the staff has to be applied in design, operation, maintenance and decommissioning phases. The tokamak building, where several complex systems have to converge to the torus to create and control the plasma, to take out its energy and to produce and extract tritium, provides the second and ultimate confinement barrier between the environment and the hazardous and radioactive materials present inside that might be mobilised, in case of accident, by the high energetic fluids stored in DEMO systems. The layout criteria are focused on avoiding any challenge to the safety functions: e.g. no common mode failures of the safety classified systems for all reference design basis events. Furthermore the safety classified equipment have to maintain their safety function all over the plant life in such challenging environmental conditions; an accurate layout might allow the qualification possible, making milder the environment, e.g. defining adequate shielding and areas where radiation dose, magnetic field and accidental environmental conditions are reasonable for sensible equipment. The experience of NPPs and ITER is also recognized as the basis of such design criteria for DEMO. The paper will outline the main design basis events and the layout criteria presenting some applications for the tokamak building that reflect the recent progress of the DEMO design.

Le forme della narrazione possono essere le più diverse: non solo parole, ma anche gesti e simboli, arte e poesia. I linguaggi dell’arte costituiscono da un punto di vista pedagogico un’area tematica importante che arricchisce ed amplia l’universo narrativo umano, in quanto permette di veicolare espressioni, sensazioni, significati e vissuti emotivi e, quindi, consente di imparare a “pensare coi propri sensi”. Nell’epoca della comunicazione digitale non si può prescindere dalla considerazione dei diversi medium attraverso cui i racconti sono prodotti, definendo nuove forme di narrazione, inedite costruzioni di significati, nuovi linguaggi che si contaminano a vicenda, ripercorrendo altresì le origini, attraverso l’utilizzo del segno. Il paper intende riportare un’esperienza laboratoriale il cui obiettivo principale è stato quello di fornire alle partecipanti un’esperienza di apprendimento e di espressione estetica, utilizzando tracce (da quella poetica a quella grafica a quella pittorica). The forms of narrative can be very different: not just words, but also gestures and symbols, art and poetry. From a pedagogical point of view, the languages of art represent an important issue, enriching and expanding human narrative universe, as they allow to convey expressions, feelings, emotional meanings and experiences and, consequently, lead to “think with our own senses”. In the age of digital communication, we cannot consider independently the different mediums through which tales are produced, defining new forms of storytelling, original constructions of meanings, new languages that mingle together, retracing the origins, through the use of the sign. This paper aims to bring a workshop experience, whose main objective was to provide the participants a learning experience and an aesthetic expression, using different traces (from poetry to graphics or painting).

Il presente documento è stato preparato nel corso della terza annualità dell'AdP ENEA-MiSE nell'ambito dell'obiettivo C (SPERIMENTAZIONE E CALCOLO IN APPOGGIO AGLI STUDI SULLA SICUREZZA) della Linea Progettuale 1 (Sviluppo competenze scientifiche nel campo della sicurezza nucleare). Esso riporta i risultati dell’attività di validazione e verifica dei codici di sistema TRACE e CATHARE2 sulla base della campagna sperimentale condotta sull’impianto SPES 2., peer-reviewed

Il presente documento è stato preparato nel corso della seconda annualità dell'AdP ENEA-MSE nell'ambito dell'obiettivo C (SPERIMENTAZIONE E CALCOLO IN APPOGGIO AGLI STUDI SULLA SICUREZZA) della Linea Progettuale 1 (Sviluppo competenze scientifiche nel campo della sicurezza nucleare). Esso riporta i principali risultati dell’attività di validazione e verifica dei codici di sistema TRACE e CATHARE sulla base della campagna sperimentale condotta sull’impianto SPES 2., peer-reviewed

Distribuzione libera - Collocazione contrattuale: Accordo di programma ENEA-MSE. Piano annuale di realizzazione 2011. Linea progettuale 2: Studi di sicurezza sugli impianti nucleari - Note: Il presente documento è stato preparato con il contributo del personale ENEA e CIRTEN: F. Donato, C. Lombardo (ENEA); P. Chiovaro, S. D'Amico, P.A. Di Maio, F. Mascari, I. Moscato, G. Velia (Università di Palermo); Sigla documento rif. :CIRTEN-UNIPA RL- PAR 2011 1215/1250 - Convalida a cura di: F. De Rosa - Approvazione a cura di: P. Meloni, Il presente documento riferisce sulle attività svolte nel corso della prima annualità dell'AdP ENEA-MSE nell'ambito della linea progettuale 2 (LP2), obiettivo D4 (Adeguamento dello SPES2 per prove di sicurezza). In particolare, il documento riporta i principali risultati del lavoro preliminare necessario per la simulazione di un incidente tipo Fukushima sulla "facility" integrale SPES-2 di SIET. Il lavoro sviluppato ha riguardato la messa a punto del modello TRACE della "facility' SPES-2, utilizzato per la simulazione del transitorio di perdita totale di alimentazione elettrica (Station Black Out SBO). I risultati ottenuti, descritti nel presente rapporto, essendo in accordo con la fisica del fenomeno, dimostrano l'idoneità globale dell'impianto alla realizzazione dei test di interesse. L'approfondimento di alcuni fenomeni locali, che determinano asimmetrie in alcune componenti della "facility", mette in luce la necessità di studi approfonditi da condurre attraverso successive analisi. Inoltre, il modello CATHARE sviluppato nell'ambito della stessa attività potrebbe essere impiegato, in futuro, per la simulazione dello stesso transitorio e per l'approfondimento a cui sopra brevemente accennato., peer-reviewed

Currently, for the EU DEMO, two Breeding Blankets (BBs) have been selected as potential candidates for the integration in the reactor. They are the Water Cooled Lithium Lead and the Helium Cooled Pebble Bed BB concepts. The two BB variants together with the associated ancillary systems drive the design of the overall plant. Therefore, a holistic investigation of integration issues derived by the BB and the installation of its ancillary systems has been performed. The issues related to the water activation due to the 16N and 17N isotopes and the impact on the primary heat transfer systems have been investigated providing guidelines and dedicated solution for the integration of safety devices as isolation valves. The tritium retention and the permeation rates through the blanket and its ancillary systems have been also assessed taking into account different operating points both for the BB and ancillaries and comparing, when possible, the releases with the operating and safety limits. Moreover, the issues related to the tritium start-up inventory as well as the uncertainties on the Tritium Breeding Ratio (TBR) due to the integration of the auxiliary systems within the Vacuum Vessel have been also studied. Finally, the impact of the BB concepts on the safety systems like the Vacuum Vessel Pressure Suppression System is described with a particular focus on the different measures that should be implemented according to the considered concept. All these aspects are then taken into account to drive future developments during the Concept Design Phase., peer-reviewed

The European Research Roadmap to the Realisation of Fusion Energy foresees that the DEMO reactor is going to succeed ITER in the pathway towards the exploitation of nuclear fusion, achieving long plasma operation time, demonstrating tritium self-sufficiency and producing net electric output on an industrial scale. Therefore, its design must be more oriented towards the Balance of Plant (BoP) than it is in ITER. Since the early pre-conceptual phase of the DEMO project, emphasis has been laid on identifying the main requirements affecting the overall architecture of the BoP. For instance, specific efforts and proper solutions have been envisaged to cope with the pulsed nature of the heat source. Furthermore, the current development of two blanket concepts calls for two separate BoP options to be conceived. This paper summarizes the main alternatives outlined at the end of DEMO pre-conceptual design phase for the BoP concepts based on both the Helium-Cooled Pebble Bed (HCPB) and Water-Cooled Lithium Lead (WCLL) Breeding Blanket (BB) technologies. Then, the assumed reference configurations of both the BoP concepts are described in detail, highlighting the main features and the most relevant engineering aspects. Attention will be focussed on technological challenges, integration constraints and other open issues, highlighting pros and cons of the chosen BoP options to be further investigated in the next design phase.

Within the framework of the activities foreseen by the EUROfusion action on the cooling water activation assessment for a DEMO reactor equipped with a Water Cooled Lithium Lead Breeding Blanket (WCLL BB), the University of Palermo is involved in the assessment of dose rates induced by the decay of nitrogen radioisotopes produced by water activation, nearby the main components (e.g. isolation valves) of both First Wall (FW) and Breeder Zone (BZ) cooling circuits. In particular, the aim of this work is to evaluate the spatial distribution of nitrogen isotopes (16N and 17N) in the WCLL BB cooling circuits. To this purpose, a coupled neutronic/fluid-dynamic problem is solved following a theoretical-numerical approach and adopting an integrated computational tool mainly relying on the use of MCNP6 and ANSYS CFX codes. The operative procedure adopted foresees the assessment of the production rate distributions of nitrogen isotopes within FW and BZ cooling channels and tubes by means of totally heterogeneous neutronic analyses. A fully 3-D fluid-dynamic approach is, then, used to compute the nitrogen isotope concentrations within the In-Vessel complex flow domain, while a 1-D lumped parameters approach is adopted to calculate their distribution along the Ex-Vessel BB Primary Heat Transfer System. The results obtained, herewith presented and critically discussed, provided the necessary data to perform dedicated neutronic and photonic transport analyses and, hence, to assess the dose rates in the aforementioned target locations.

In the frame of the EUROfusion consortium activities the Helium Cooled Pebble Bed (HCPB) and the Water Cooled Lithium Lead (WCLL) concepts are being developed as possible candidates to become driver Breeding Blanket (BB) for the EU DEMO, which aims at the tritium self-sufficiency and net electricity production. The two BB design options encompass water or helium as coolants and solid ceramic with beryllium/beryllides or PbLi as tritium breeder and neutron multipliers. The BB segments have evolved towards a more stable conceptual design taking into account multiple feasibility aspects and requirements imposed by interfacing systems. Possible solutions to improve shielding capabilities of Helium cooled BB are investigated and the impact of water coolant activation is assessed by studying the spatial distribution of 16N and 17N isotopes dose rates, in particular in proximity of isolation valves. The reference and back-up technologies for the Tritium Extraction and Removal (TER) from the helium purge gas and the PbLi are developed addressing key feasibility aspects and implications on the tokamak layout and with considerable R&D efforts. As the BB internals offer an ideal environment (high temperatures, thin structural material) to promote the tritium permeation, studies are devoted to establish a tritium balance in the different systems during operation, with special care to the permeation rate and inventory in the coolant. Those are the key parameters for the feasibility assessment and technology selection for the Coolant Purification Systems (CPS).

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.

Within the framework of the activities foreseen by the EUROfusion action on the cooling water activation assessment for a DEMO reactor equipped with a Water Cooled Lithium Lead Breeding Blanket (WCLL BB), the University of Palermo is involved in the investigation of the absorbed dose induced by the decay of nitrogen radioisotopes produced by water activation, in the main components (e.g. isolation valves) of both First Wall (FW) and Breeder Zone (BZ) cooling circuits. The aim of this work is to assess the spatial distribution of the absorbed dose in the DEMO Upper Pipe Chase (UPC), focusing the attention on the space neighbouring a typical isolation valve of the Primary Heat Transfer System (PHTS), taking into account a period of 7 full power years. To this end, a computational approach has been followed adopting MCNP5 Monte Carlo code. In particular, a totally heterogeneous neutronic model of a portion of the UPC has been set up, including the valve and the main FW and BZ PHTS piping, and the spatial distribution of nitrogen isotopes concentrations, previously assessed, have been used to model the photonic and neutronic sources. The results obtained, herewith presented and critically discussed, provided some information on the nuclear issues of the WCLL BB PHTS, to be considered as hints for the blanket design optimization.

The Pre-Concept Design (PCD) of the Balance of Plant (BoP) systems of the EU-DEMO power plant is described in this paper for both breeding blanket (BB) concepts under assessment, namely the Water Cooled Lithium Lead (WCLL) BB and the Helium Cooled Pebble Bed (HCPB) BB. Moreover, the results of a preliminary evaluation of a number of BoP variants are discussed. This paper outlines the steps of the BoP design development, highlighting the project objectives and the strategy for their achievement under the very challenging requirements which include, among others, the intermittent nature of the DEMO plasma heat source. The main achievements during the PCD Phase will be reported together with the development plan for the Concept Design (CD) Phase to reach a mature (feasible) BoP concept for DEMO., peer-reviewed

DEMO will be a fusion power plant demonstrating the integration into the grid architecture of an electric utility grid. The design of the power conversion chain is of particular importance, as it must adequately account for the specifics of nuclear fusion on the generation side and ensure compatibility with the electric utility grid at all times. One of the special challenges the foreseen pulsed operation, which affects the operation of the entire heat transport chain. This requires a time-dependant analysis of different concept design approaches to ensure proof of reliable operation and efficiency to obtain nuclear licensing. Several architectures of Balance of Plant were conceived and developed during the DEMO Pre-Concept Design Phase in order to suit needs and constraints of the in-vessel systems, with particular regard to the different blanket concepts. At this early design stage, emphasis was given to the achievement of robust solutions for all essential Balance of Plant systems, which have chiefly to ensure feasible and flexible operation modes during the main DEMO operating phases – Pulse, Dwell and ramp-up/down – and to adsorb and compensate for potential fusion power fluctuations during plasma flat-top. Although some criticalities, requiring further design improvements were identified, these preliminary assessments showed that the investigated cooling system architectures have the capability to restore nominal conditions after any of the abovementioned cases and that the overall availability could meet the DEMO top-level requirements. This paper describes the results of the studies on the tokamak coolant and Power Conversion System (PCS) options and critically highlights the aspects that require further work.

In the frame of the activities promoted and encouraged by the EUROfusion Consortium aimed at developing the EU-DEMO fusion reactor, great emphasis has been placed at a very early stage of the design to incorporate the provisions needed to improve the overall plant safety and reliability performances as well as to analyse possible mitigation actions. In this framework, the research activity has been focused on the representative and safety relevant cooling loop of the Helium Cooled Pebble Bed (HCPB) Breeding Blanket (BB) Primary Heat Transfer System (PHTS), purposely selected by the safety team, in order to assess its thermal-hydraulic behaviour during normal operational conditions (ramp up/down and steady state) and to preliminarily investigate the consequences of an ex-vessel LOCA accidental scenario ensuing a Double-Ended Guillotine (DEG) break in the hot leg. The research activity has been carried out following a theoretical-computational approach based on the finite volume method adopting the RELAP5−3D system code along with the ANSYS CFX computational fluid dynamic code, which were implicitly integrated to achieve a more detailed and realistic simulation of the EU-DEMO reactor thermal-hydraulics. Models, assumptions and outcomes of this preliminary study are herein presented and critically discussed.

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.

The European DEMOnstration power plant (DEMO) is considered to be the nearest-term fusion reactor capable of producing several hundred MWs of net electricity, operating with a closed tritium fuel-cycle (achieving the tritium self-sufficiency), and qualifing technological solutions for a fusion power plant with different breeding blanket (BB) concepts that are under investigation. The BB is a key component for the development of the DEMO plant design and, in particular, of those systems having the responsibility to remove the plasma-generated thermal power and its conversion in the electrical energy. This study deals with the preliminary thermo-mechanical design of the heat exchangers and steam generator components. The design criteria and the main design analyses were discussed to achieve a simple but robust design. The proposed preliminary thermo-mechanical designs, both for heat exchangers and for steam generators, are considered feasible.

One of the generic designs of the nuclear fusion DEMO reactor proposed by the EUROfusion consortium foresees the development of a tritium breeding blanket (BB) relying on the use of the liquid-metal PbLi eutectic alloy as both neutron multiplier and tritium breeder, namely the water-cooled lithium lead (WCLL) BB, whose strengths and weaknesses are well known. This paper focuses the attention on one of the possible disadvantages of such a technology: the production of the highly radiotoxic radionuclide 210Po, which could become a safety issue to be accounted for. The 210Po concentration within the PbLi circuit has been assessed by solving a modified version of Bateman’s equations to consider the alloy circulation, so a one-dimensional convective fluid-dynamic model has been set up. Nuclear quantities have been evaluated by Monte Carlo neutron transport analyses using MCNP code and adopting a fully heterogeneous model of DEMO equipped with the WCLL BB. Moreover, rough sensitivity analyses have been performed to assess the influence on the results of the uncertainties related to the 209Bi radiative-capture cross section and the initial concentration of this nuclide which is present in the PbLi as an impurity. Results obtained have been critically discussed and some safety issues have been addressed to evaluate the possible hazard in case of a leak of PbLi accident.

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.

The helium-cooled pebble bed (HCPB) blanket is one of the two concepts proposed as a driver blanket for the European Union Demonstration Fusion Power Reactor (EU DEMO). In contrast to past conceptual design studies, in the frame of the current Power Plant Physics and Technology of the EUROfusion Consortium, the ongoing EU DEMO preconceptual design activities have adopted a holistic and integrated (i.e., systems engineering) design approach. As a consequence of this new approach, many interfaces and requirements have been identified, some of them driving the design of the blankets. This paper shows the advancements in the HCPB breeding blanket and describes the lessons learned after implementing the new approach. This new set of requirements has led us to reconsider fundamental aspects of the HCPB blanket design, especially in the way of how the heat is extracted from the blanket. Among others, the requirement to achieve a mature balance of plant (BOP) system plays a central role as a key design driver and has forced us to reduce pressure drops in the breeding blanket. In this regard, the blanket has been redesigned, leading to an enhanced concept based on single-module segments with a hexagonal matrix of fuel-breeder pins. Both the fuel-breeder pins and the first wall (FW) are equipped with turbulence promoters (augmented wall roughness in the fuel-breeder pins and V-ribs in the FW), following a similar idea as in the past MAGNOX, Advanced Gas Reactor (AGR), and Gas Cooled Reactor (GCR) programs in fission. This has led to minimizing the pressure drops while maximizing the heat transfer. Also, the blanket outlet temperature has been extended to 520°C, following the same principle as in Generation IV’s GCRs of maximizing the temperature difference across the core to minimize the reactor mass flow rate and thus the circulating power. All these features have led to a remarkably low plant circulating power (80 to 90MW) and the required power per helium blower (5 to 6MW), which potentially solves the key long-standing problem of the BOP technology readiness level for an ≈2.4-GW(thermal) helium-cooled DEMO reactor.

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 (TBR200 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.

In Europe (EU), in the frame of the EUROfusion consortium activities, four Breeding Blanket (BB) concepts are being developed with the aim of fulfilling the performances required by a near-term fusion power demonstration plant (DEMO) in terms of tritium self-sufficiency and electricity production. The four blanket options cover a wide range of technological possibilities, as water and helium are considered as possible coolants and solid ceramic breeder in combination with beryllium and PbLi as tritium breeder and neutron multipliers. The strategy for the BB selection and operation has to account for the challenging schedule of the EU DEMO, the ambitious operational requirements of the BBs and the still large development needed to have a BB qualified and licensed for operating in DEMO. In parallel to the continuous design efforts on the four blanket concepts, their integration in-vessel and ex-vessel has started. On the one hand it has become clear that despite the numerous systems to be integrated in-vessel the protection of the blanket first wall has to be addressed with highest priority. On the other hand the ex-vessel interfaces and the requirements imposed by the blanket to the primary heat transfer system and to the PbLi loop components have a considerable impact on the whole DEMO Plant layout. The aim of this paper is: to present the strategy for the DEMO BB down selection and BB operation in DEMO; to describe the status of the design evolution of the four EU BB concepts; to provide an overview of the challenges of the in-vessel and ex-vessel integration of the main systems interfacing the BBs and describe their design status.

Over the last decades a lot of experimental researches have been done to increase the reliability of passive decay heat removal systems implementing in-pool immersed heat exchanger. In this framework, a domestic research program on innovative safety systems was carried out leading the design and the development of the PERSEO facility at the SIET laboratories. The configuration of the system consists of an heat exchanger contained in a small pool which is connected both at the bottom and at the top to a large water reservoir pool. Within the frame of a national research program funded by the Italian minister of economic development, the DEIM department of the University of Palermo in cooperation with ENEA has developed a computational model of the PERSEO facility in order to simulate its behaviour during an integrated test. The analysis here presented has been performed by using the best-estimate TRACE code and - in order to highlight the capabilities and limits of the TRACE model in reproducing qualitatively and quantitatively the experimental trends - the main results have been compared with the experimental data. The comparison shows that the model is able to predict the overall behaviour of the plant during the meaningful phases of the transient analysed. Nevertheless, some improvements in the modelling of certain components in which take place complex three-dimensional phenomena are suggested in order to reduce some discrepancies observed between code results and test measurements., peer-reviewed

The EU-DEMO conceptual design is being conducted among research institutions and universities from 26 countries of European Union, Switzerland and Ukraine. Its mission is to realise electricity from nuclear fusion reaction by 2050. As DEMO has been conceived to deliver net electricity to the grid, the choice of the Breeding Blanket (BB) coolant plays a pivotal role in the reactor design having a strong influence on plant operation, safety and maintenance. In particular, due to the pulsed nature of the heat source, the Primary Heat Transfer System (PHTS) becomes a very important actor of the Balance of Plant (BoP) together with the Power Conversion System (PCS). Moreover, aiming to mitigate the potential negative impact of plasma pulsing on BoP equipment, for the DEMO plant is also being investigated a “heat transfer chain” option which envisages an Intermediate Heat Transfer System (IHTS) equipped with an Energy Storage System (ESS) between PHTS and PCS. Within this framework, a preliminary study has been carried out to analyse the thermal dynamic behaviour of the IHTS system for the Helium-Cooled Pebble Bed (HCPB) BB concept during pulse/dwell transition which should be still considered as the normal operating mode of a fusion power plant. Starting from preliminary thermal-hydraulic calculations made in order to size the main BoP components, the global performances of DEMO BoP have been quantitatively assessed focusing the attention on the attitude of the whole IHTS to smooth the sudden power variations which come from the plasma. The paper describes criteria and rationale followed to develop a numerical model which manages to simulate simple transient scenarios of DEMO BoP. Results of numerical simulations are presented and critically discussed in order to point out the main issues that DEMO BoP has to overcome to achieve a viable electricity power output.

The present paper deals with the investigation of the evolution and consequences of a Station Black-Out (SBO) initiating event transient in the SPES3 facility [1]. This facility is an integral simulator of a small modular reactor being built at the SIET laboratories, in the framework of the R&D program on nuclear fission funded by the Italian Ministry of Economic Development and led by ENEA. The SBO transient will be simulated by using the RELAP5 and TRACE nodalizations of the SPES3 facility. Moreover, the analysis will contribute to study the differences on the code predictions considering the different modelling approach with one and/or three-dimensional components and to compare the capability of these codes to describe the SPES3 facility behaviour., peer-reviewed

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.

This paper describes the status of the pre-conceptual design activities in Europe to advance the technical basis of the design of a DEMOnstration Fusion Power Plant (DEMO) to come in operation around the middle of this century with the main aims of demonstrating the production of few hundred MWs of net electricity, the feasibility of operation with a closed-tritium fuel cycle, and maintenance systems capable of achieving adequate plant availability. This is expected to benefit as much as possible from the ITER experience, in terms of design, licensing, and construction. Emphasis is on an integrated design approach, based on system engineering, which provides a clear path for urgent R&D and addresses the main design integration issues by taking account critical systems interdependencies and inherent uncertainties of important design assumptions (physics and technology). A design readiness evaluation, together with a technology maturation and down selection strategy are planned through structured and transparent Gate Reviews. By embedding industry experience in the design from the beginning it will ensure that early attention is given to technology readiness and industrial feasibility, costs, maintenance, power conversion, nuclear safety and licensing aspects.