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This document is an editorial discussing advances in nuclear fusion energy and cross-cutting technologies. The current energy model, based on fossil fuels, is unsustainable and contributes to global warming. Nuclear fusion, specifically the use of breeding blankets, is a promising alternative energy source. The document highlights recent advancements in fusion technologies, such as the Helium-Cooled Pebble Bed (HCPB) and Water-Cooled Lithium Lead (WCLL) breeding blanket concepts. It also discusses safety systems, tritium permeation, and the development of fusion power plants. The editorial emphasizes the need for further research and development in these areas. [Extracted from the article]

The Stellarator Power Plant Studies Prospective R&D Work Package in the Eurofusion Programme was settled to bring the stellarator engineering to maturity, so that stellarators and particularly the HELIAS (HELical-axis Advanced Stellarator) configuration could be a possible alternative to tokamaks. However, its complex geometry makes designing a Breeding Blanket (BB) that fully satisfies the requirements for such a HELIAS configuration, which is a difficult task. Taking advantage of the acquired experience in BB design for DEMO tokamak, CIEMAT is leading the development of a Dual Coolant Lithium Lead (DCLL) BB for a HELIAS configuration. To answer the specific HELIAS challenges, new and advanced solutions have been proposed, such as the use of fully detached First Wall (FW) based on liquid metal Capillary Porous Systems (CPS). The proposed solutions have been studied in a simplified 1D model that can help to estimate the relative variations in Tritium Breeding Ratio (TBR) and displacement per atom (dpa) to verify their effectiveness in simplifying the BB integration and improving the machine availability while keeping the main BB nuclear functions (i.e., tritium breeding, heat extraction and shielding). This preliminary study demonstrates that the use of FW CPS would drastically reduce the radiation damage received by the blanket by 29% in some of the selected configurations along with a small decrease of 4.9% in TBR. This could even be improved to just a 3.8% TBR reduction by using a graphite reflector. Such an impact on the TBR is considered affordable, and the results presented, although preliminary in essence, have shown the existence of margins for further development of the FW CPS concept for HELIAS, as they have been not found, at least to date, to be significant showstoppers for the use of this technological solution. [ABSTRACT FROM AUTHOR]

The helium-cooled pebble bed (HCPB) breeding blanket (BB) is one of the two driver-blanket candidates for the European DEMO fusion reactor. Recent design activities were focused, among other objectives, on the achievement of an efficient shielding system to adequately protect the vacuum vessel (VV) and toroidal field coils (TFCs). Several shielding options have been studied in terms of architecture (e.g., in-BB shield and ex-BB shield) and materials (e.g., B4C, WC, WB, YHx, and ZrHx). In this study, the B4C material was selected as the most attractive option considering not only shielding performance but also availability, industrialization, experience, and cost factors. Subsequently, we performed a parametric study by implementing different thicknesses of a B4C external shield and reporting information of its effect on shielding performance, structural behavior, swelling and tritium breeding. Furthermore, a detailed structure for the VV was developed considering an internal layered configuration comprising steels/water with different boron contents. Corresponding shielding analyses were conducted regarding influence on neutron attenuation when implementing such a VV structure for both the baseline consolidated design of the HCPB and one of the previously developed and improved BSS configurations. The most critical responses (neutron flux and dpa) were fully established only using 10 cm B4C and an improved VV configuration. [ABSTRACT FROM AUTHOR]

As a possible long-term alternative to a tokamak fusion power plant, the stellarator concept offers salient physics features (no external current drive, no risk of plasma disruptions and low recirculating power, among others) that could be offset by the more complex configurations and challenging maintenance schemes. Very ffew conceptual design studies have been performed compared to those in tokamaks, and the enhancement of engineering aspects should follow. With the recent start of operation of Wendelstein 7-X, the helical-axis advanced stellarator (HELIAS) line has again raised interest in the scientific and technologic EUROfusion programme. The main aim at present is showing that stellarators (particularly helical-axis stellarators) are viable as potential fusion reactors. For the conceptualisation of a mature HELIAS power reactor, different engineering and technological aspects must be studied, improved and solved. To this end, starting from a very preliminary reactor design called HELIAS 5-B (5-field-period), with a fusion power of 3000 MW, a neutronic model has been developed and analysed introducing in the baseline the relevant components of breeding blankets (BB). The vast experience achieved at CIEMAT in BB designs for the DEMO tokamak has been exploited, adapting the dual-coolant lithium-lead BB design elaborated in the frame of the WPBB programme of EUROfusion/PPPT to the HELIAS configuration. Preliminary neutronic assessments have been performed focussing on tritium production, power density distributions and damage/shielding responses such as nuclear heating, neutron fluence, dpa and helium production. Particle transport calculations have been performed with MCNP5v1.6 Monte Carlo code. [ABSTRACT FROM AUTHOR]

Under the ambitious EUROfusion mission of "bringing the stellarator line to technological maturity", the development of a Dual Coolant Lithium–Lead (DCLL) Breeding Blanket (BB) for a Helical-Axis Advanced Stellarator (HELIAS) started. The BB is the crucial systems to achieve tritium self-sufficiency in fusion power plants. The DCLL has potentialities to answer the challenges posed by complex HELIAS configuration, having liquid breeder and decoupled first wall (FW) and breeding zone circuits. Both characteristics are being empowered to simplify remote maintenance and integration of the BB segments. Hence, a "quasi-toroidal" BB segmentation has been proposed to minimize the magnetohydrodynamic pressure drop eschewing complex electrical insulating components while hindering maintenance. Such controversy motivated the use of a detached FW small panels manageable through ports. On this basis, a decoupled Capillary Porous System FW has been assessed under thermal-hydraulics and neutronics point of view. Moreover, ad-hoc tools have been developed for the parametrization of models and convergence toward a viable DCLL HELIAS design, speeding-up the coupling of computer-aided design modelling and analyses. Such novel strategies are described in view of a potential acceleration of the stellarator reactor engineering and by providing an overview of the stellarator challenges inside a global context. • A Dual Coolant Lithium–Lead Breeding Blanket is explored for an HELIAS. • A "quasi-toroidal" BB segmentation has been proposed to minimize MHD pressure drop. • A more complex maintenance motivated the use of a detached FW based on CPS. • Thermal-hydraulics and neutronics assessments performed. • Ad-hoc tools developed for parametrized neutronic models. [ABSTRACT FROM AUTHOR]

Highlights • TBR and nuclear heating have been studied for the last version of DCLL BB for DEMO. • Upgrades of this baseline are analyzed under neutronic point of view. • The impact on nuclear performances of different FW designs is addressed. • Two FCI designs with different strategies for their neutronic simulation also affect the TBR result. • The divertor compositions is also demonstrated to have impact on TBR. • NBI integration consequences on TBR and NH are evaluated. Abstract On the way towards a comprehensive design of DEMO, step by step all the systems and components must be introduced as their definition or refinement progresses, in order to demonstrate the viability of a design on larger scale, i.e. leaving fewer margins to undetermined questions. Among the EUROfusion Programme, new aspects have been recently fixed or further developed as the Divertor, the First Wall (FW) and the Flow Channel Inserts (FCI) designs. Furthermore, the integration of Heating and Current Drive (H&CD) systems, as the Neutral Beam Injector (NBI), has started. The introduction or modification of these systems and components could seriously jeopardize the nuclear behaviour of an initially validated Breeding Blanket (BB) DEMO concept, since many neutronics criteria - among others - could be no more fulfilled. Since the design of DEMO is a continuous upgrade under iterative process, as the advances push on, most of the studies have to be repeated to demonstrate that criteria are still respected in a fully integrated design. The consequences of these upgrades over the neutronic responses are addressed in this paper. Among others, the influence on Tritium Breeding Ratio (TBR) of a new design of detached FW protecting BB from high heat fluxes is investigated. The impact of different typologies of FCIs is assessed also according to the degree of detail in the neutronic description. The divertor composition also reveals to have strong impact on responses apparently not related with its design, as the tritium production in the BB. Besides, the integration of NBI minimizing its invasiveness in the BB is verified by neutronic analyses concerning the main BB functions: fuel breeding and heat generation. Accordingly, TBR and Nuclear Heating (NH) are assessed. The study is performed for a Dual Coolant Lead Lithium (DCLL) BB DEMO although can be extrapolated to other BB concepts. [ABSTRACT FROM AUTHOR]

The present paper addresses the impact of the divertor option on the nuclear performances of the Demonstration fusion reactor (DEMO). As the effect of the number and size of the divertor has been already evaluated, in this work the focus has been posed on the composition in terms of amount of cooling inside the divertor cassette. Transport responses, as the Tritium breeding ratio (TBR), neutron and gamma fluxes and spectra inside the plasma chamber, as well as activation responses such as shutdown dose rate, decay gamma fluxes and heating have been evaluated for two different blanket concepts of the future European DEMO reactor: DCLL and WCLL. Three different divertor compositions have been tested demonstrating the importance of this component not only locally but in the global radiation field. The transport analysis has been performed with the Monte Carlo code MCNP5 and the JEFF3.1.1 and JEFF3.2 data libraries. The activation responses calculated using Advanced D1S method have been recently assessed and summarized in the present paper. [ABSTRACT FROM AUTHOR]

In large fusion machines, as the foreseen DEMO, the high energy neutrons produced will cause the transmutation of the interacting materials which will become a source of radioactive waste. Besides the main constituents of a material that could transmute but which presence is essential, the impurities often give rise to significant additional activation compared to the base material. Thus, once identified the nuclides generating the dominant activation products for the DCLL (Dual Coolant Lithium-Lead) concept of Breeding Blanket (BB) for the European DEMO, the objective of the study has been to determine the impurity limits for such materials in order to minimize the radiological impact of the waste produced. This will enable to consider it as Low Level Waste and dispose it in the Spanish near-surface repository of El Cabril. For both the original compositions as well as the revised one with reduced amount of impurities, suggested to mitigate the waste impact, activation calculations have been performed. Hence, total beta-gamma activity, alfa activity, specific activity for different nuclides, decay heat and surface gamma dose rate have been analyzed with reference to the IAEA and SEAFP-2 standards for waste classification and to the specific regulations of El Cabril. [ABSTRACT FROM AUTHOR]

The present paper is devoted to the shutdown dose rate assessment, one of the most important safety issues in fusion power plant design and operation. The characterization of the radiation environment after the shutdown is fundamental to plan safe operation and maintenance in a fusion machine in order to guarantee the dose limits are not exceeded. In this paper the shutdown dose due to the radionuclides generated by neutron activation of reactor components have been assessed for the design of the European DCLL v3.0 developed during 2016 integrated into the generic DEMO baseline 2015 fusion reactor. The shutdown dose rate calculations from 1 day to 1 year after shutdown have been performed through the Advanced D1S method coupling the particle transport MCNP5 and ACAB inventory codes. 3D maps of shutdown dose rates, decay fluxes for the whole reactor and decay gamma heating for Eurofer structures have been provided as well as specific values at relevant positions. Results are presented and discussed also in terms of the different nuclides contributions from the various activated components. Two different divertor compositions have been also used demonstrating the importance of this component not only locally but in the global radiation field inside the plasma chamber. [ABSTRACT FROM AUTHOR]

In the frame of the newly established EUROfusion WPBB Project for the period 2014–2018, four breeding blanket options are being investigated to be used in the fusion power demonstration plant DEMO. CIEMAT is leading the development of the conceptual design of the Dual Coolant Lithium Lead, DCLL, breeding blanket. The primary role of the blanket is of energy extraction, tritium production, and radiation shielding. With this aim the DCLL uses LiPb as primary coolant, tritium breeder and neutron multiplier and Eurofer as structural material. Focusing on the achievement of the fundamental neutronic responses a preliminary blanket model has been designed. Thus detailed 3D neutronic models of the whole blanket modules have been generated, arranged in a specific DCLL segmentation and integrated in the generic DEMO model. The initial design has been studied to demonstrate its viability. Thus, the neutronic behaviour of the blanket and of the shield systems in terms of tritium breeding capabilities, power generation and shielding efficiency has been assessed in this paper. The results demonstrate that the primary nuclear performances are already satisfactory at this preliminary stage of the design, having obtained the tritium self-sufficiency and an adequate shielding. [ABSTRACT FROM AUTHOR]

The paper focuses on the nuclear analyses of the ITER Equatorial Port Visible/Infrared Wide Angle Viewing System (VIS/IR WAVS). This instrument comprises of viewing systems in the 4 Equatorial Ports (EP) 3, 9, 12 and 17. The main mission of this diagnostic is to support the operation of the tokamak by providing visible and infrared viewing and temperature data of the first wall to protect it from damage. Its design is driven by both the tokamak severe environment and the high performances required for machine protection. New nuclear studies have been carried out for the development of the diagnostic and for test purposes under ITER-like irradiation conditions in order to choose the most appropriate materials for the optical components. Thus, three neutronic analyses have been carried out: the first in order to verify if silver could be used as a covering material for the optical components in different location of the Interspace area; the second in order to establish the irradiation time required in a Co-60 gamma facility (at CIEMAT) for testing purposes of the sapphire vacuum window; and the third to give more detailed specifications for the irradiation campaigns under gamma (in the Co-60 facility) and neutrons (at SCK·CEN BR2 reactor), about the time required to achieve the same dose than the one accumulated in ITER at the end-of-life (EOL) in the different components of the Port Area for the materials proposed as substrate and coatings. The neutronic and activation calculations have been performed using the Monte Carlo code MCNP5, the activation code ACAB and the cross section libraries FENDL 2.1 and EAF-2007. The description of the simulations and their results are presented in this paper. [ABSTRACT FROM AUTHOR]

Abstract: A conceptual design for a DEMO fusion reactor based on a dual coolant He/Pb15.7Li breeding blanket (DCLL) is being developed within the Spanish Breeding Blanket Technology Programme: TECNO_FUS. The production of tritium and activation products of LiPb might be a concern from the radiological safety point of view. Thus, in this contribution, an accidental release in atmosphere of radionuclides from LiPb breeder has been studied. Activation calculations have been performed with ACAB code assuming an irradiation scenario of 5 FPY for the maximum neutron fluence rate in the equatorial breeding zone. The results in terms of specific activity, surface gamma dose rate and committed effective dose (CED) due to inhalation at different times have been used to chose the potentially more hazardous radionuclides. Dispersion of the selected radionuclides has been modeled with HOTSPOT code using the Gaussian plume model and two different atmospheric conditions. Offsite dose (for external irradiation and inhalation) due to an accidental release of 1kg of activated LiPb has been calculated after 5 FPY of irradiation (shutdown) using HOTSPOT atmospheric dispersion in class D weather conditions. According to the results, fulfilling the dose requirement for no evacuation would permit to release up to 40kg of activated LiPb, without taking into account the possible isotopic purification and detritiation systems. This value can be compared with the actually released amount in a given accidental scenario for designing the reactor confinement barriers and safety systems. [Copyright &y& Elsevier]

• Different alternatives to upgrade the EU DCLL MMS design to a SMS architecture have been evaluated. • An electrically resistive PbLi-compatible ceramic is proposed for the BZ, enclosed by a steel case including the FW. • CO 2 is a fairly good alternative to He, although possible radiological issues must be clarified. • The preliminary analyses of two proposed supporting systems point out towards the mechanical feasibility of the design. • The double wall compromises the tritium self-sufficiency. Some variants of the baseline are promising to enhance the TBR. In the last years, CIEMAT worked in the development of a low temperature DCLL breeding blanket design aimed at using conventional materials and technologies. It followed the multi-module segment (MMS) approach, which allows operating within the thermal range tolerated by EUROFER while keeping a low bulk velocity of the self-cooled liquid breeder (PbLi) in the breeding zone. Thinking in the commercial exploitation of fusion, it must be taken into account that there is risk that the foreseen availability goal in fusion reactors is not achieved. In that case, the limited net efficiency of low temperature blankets could not be sufficient to obtain a competitive cost of the electricity. For that reason, more advanced blanket solutions are being explored with the objective of developing a simpler, more reliable and more efficient design based on a single-module segment (SMS) architecture and the use of self-cooled incompressible liquid breeder. In this work, different strategies to solve interrelated aspects like the topology and electrical insulation of the breeder circuits, the first wall integration and the main neutronic responses are analysed and discussed. [ABSTRACT FROM AUTHOR]

• Activation of water coming from PHTS of WCLL/HCPB concepts to out-cryostat regions of the DEMO building assessed. • Neutron flux and accumulated dose in silicon calculated for electronics. • Biological dose rate due to ACPs incorporated into the water flow which emit gammas after the shutdown is calculated. • Different configurations of the building shielding materials and components studied. • Benchmark with a similar case on Upper pipe chase. The radiation protection issues linked to the activated cooling water coming from the in-vessel components of a fusion reactor to the ex-vessel regions are important design and safety constraints in the building layout definition, as demonstrated at ITER. To tackle these issues more rigorously already in the pre-conceptual design of DEMO some relevant activation and neutronic studies have been performed for the re-definition of a preliminary architecture of the tokamak complex. The design conditions of the Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) concept have been considered as well as those of the Helium Cooled Pebble Bed (HCPB). The radiation environment in those areas of the building hosting the pipes carrying activated water devoted to the cooling of the First Wall/limiter and divertor has been characterized in terms of decay photons and secondary neutron emissions, which are emitted by N16 and N17 radionuclides. In particular, the feasibility of placing radiation sensitive equipment near flowing activated water is evaluated by calculating accumulated dose in silicon and neutron flux. Viability for manual maintenance activities is assessed by calculating biological dose rate for workers due to Activated Corrosion Products (ACP). The values achieved are then compared to applicable limits. Identification of problematic areas stimulated the examination of some design proposals to improve shielding materials and components. [ABSTRACT FROM AUTHOR]

• The most relevant aspects of the design of the European DCLL MMS design adapted to DEMO 2017 are described. • Several channels configurations have been assessed in order to select the most adequate for the whole FW. • Specific models have been created to study the impact of the change to DEMO 2017 on thermal-hydraulics at module level. • The presence of important heat flux peaks on the FW is translated into a poorer performance of the whole blanket. • The overall TBR has been kept in a significant value of 1.173. The main shielding responses are fulfilled. The capability of the breeding blanket to efficiently fulfill its main requirements: tritium breeding, power extraction and neutron shielding is strongly sensitive on the variation of the geometrical constraints imposed by plasma physics and the integration with the rest of the plant systems. Within the Power Plant Physics and Technology Programme, EUROfusion is developing a Dual-Coolant Lithium-Lead (DCLL) breeding blanket based on the multi-module segment approach. In the current DEMO reactor specification released in 2017, the outboard blanket thickness has been decreased with the aim of providing better stability to the plasma and increasing its elongation. Besides, the first wall has been shaped in order to optimize the heat flux loads from the plasma. This work describes the main aspects of the adaptation of the EU DCLL breeding blanket to the characteristics of DEMO 2017 and assesses the influence of the changes on its thermal-hydraulic and neutronic performances. [ABSTRACT FROM AUTHOR]

• Radiation field due to activated water flowing in pipes is assessed for a generic building outside DEMO reactor. • The focus is on the feasibility for electronic equipment location and hands-on maintenance in a representative room containing those pipes, as well as in a generic adjacent room. • A compendium of different configurations varying a wide range of parameters is studied. • Additional configurations to improve the shielding of electronic equipment and workers are analyzed. This paper is devoted to provide generic guidelines to facilitate the characterization of an important contribution to the radiation environment during operation and after the shutdown. This is fundamental to plan safe operation and maintenance of the European DEMO Fusion Power Plant. The study is focused on the estimation of activation responses to electronics and workers, in representative rooms of the DEMO building housing those pipes carrying activated water coming from the water cooled breeding blankets (WCLL BB). The prediction of some radiation-related quantities is needed in order to evaluate feasibility for functioning of rad-sensible components (e.g. electronic and electrical equipment) as well as for manual maintenance planning. The feasibility for placing rad-sensible equipment near the activated water pipes is determined by calculating the accumulated dose in silicon (due to both neutrons and photons), and neutron fluence rate. Besides, the viability for manual maintenance activities is evaluated by calculating residual dose rate for workers. These quantities are assessed in a representative room containing the pipes as well as in an adjacent room. In view of future developments on DEMO designs a parametric study has been performed and a compendium of different configurations is considered: several numbers of pipes, pipes dimensions and thickness of the wall separating both rooms are examined, as well as different delay times. Additionally, a study dedicated to the shielding improvement of both rooms is presented considering such parametric variations. [ABSTRACT FROM AUTHOR]

The thermal-hydraulic design of the breeding blanket, as the main thermal source for power conversion, reveals itself as a key issue to counteract the influence of the foreseen low overall plant availability on the cost of electricity. In the case of the Dual Coolant Lithium-Lead (DCLL), the decreased contribution of helium (non-breeding coolant) as thermal source in comparison with lithium-lead (breeding coolant) presents clear advantages, like less dependence on the long-term availability of He and lower recirculating power (recompression). The short operational range of temperature (300–550 °C) imposed by the use of RAFM steel is handled by adopting the Multi-Module Segment concept. This allows lower PbLi velocities by arranging in parallel the circuits of different modules. In consequence, the magnetohydrodynamic pressure drop and corrosion rates may be diminished. On the other hand, the high Péclet numbers validate the use of simpler computational codes to couple thermally both coolants, taking advantage of assuming that the heat transfer between the structure and the fluids is one-dimensional. A thermal-hydraulic one-dimensional code (PLATOON) has been developed for sensitivity analyses. The paper also addresses an assessment of the cooling of the radial stiffening plates and a preliminary study of the effects of the high heat generation gradient in the front poloidal PbLi channels. [ABSTRACT FROM AUTHOR]

The cost of electricity generated by fusion power will be strongly conditioned by the availability of future reactors. One key issue is the developing of feasible quick pipe connectors for the connection/disconnection of critical in-vessel components during maintenance operations. Brazing is a widely used joining technique which produces leak-proof high strength joints, with excellent stress distribution, little distortion and minimum oxidation. This work presents a design of a self-brazing/debrazing connector to be used with helium, lead–lithium and water pipes in DEMO. The remote handling compatible design includes an induction heating system, a brazing atmosphere supply, an inspection system (leak testing), a bolted/clamped union to provide stiffness against disruptions and thermal loads, and a positioning and alignment system. [ABSTRACT FROM AUTHOR]

Tritium recovery is one of the major issues of a future DEMO reactor, in order to comply with the requirements of tritium self-sufficiency. Within the EUROfusion Programme the permeation against vacuum (PAV) technique has been considered as baseline for those blankets which use PbLi as breeder. A conceptual design of a rectangular multi-channel PAV for its implementation in an experimental PbLi loop, under construction at CIEMAT, has been produced. A comparison between vanadium/niobium/tantalum and α-Fe membranes has been performed in terms of costs, machining, mechanical resistance and efficiency resulting in a design based on vanadium membranes and a stainless steel structure. Structural calculations are also presented, paying special attention to the interface between the membranes and the main structure in order to avoid leakages. Other important aspects such as keeping an adequate vacuum level have also been considered. [ABSTRACT FROM AUTHOR]

One of the most important issues in future fusion power plants is the extraction of tritium generated in the breeders in order to achieve self-sufficiency. When the breeder is a liquid metal one of the most promising techniques is the Permeation Against Vacuum, whose principle is based on tritium diffusion through a permeable membrane in contact with the liquid metal carrier and its further extraction by a vacuum pump. A conceptual design of permeator has been developed, taking into account the features of a DEMO reactor with a Dual Coolant Lithium Lead (DCLL) breeder blanket. The study is based on the analysis of different membranes and geometries aiming at the overall efficiency (extraction capability) of the device, as well as its compatibility with the breeder material. The permeator is based on a rectangular section multi-channel distribution where the liquid metal channels and vacuum channels are alternated in order to maximize the contact area and therefore to promote tritium transport from the bulk to the walls. The resulting permeator design has an excellent estimated extraction efficiency, of 80%, in a relatively compact device. [ABSTRACT FROM AUTHOR]