The breeding blanket with integrated first wall (FW) is the key nuclear component for power extraction, tritium fuel sustainability, and radiation shielding in fusion reactors. The ITER device will ...address plasma burn physics and plasma support technology, but it does not have a breeding blanket. Current activities to develop “roadmaps” for realizing fusion power recognize the blanket/FW as one of the principal remaining challenges. Therefore, a central element of the current planning activities is focused on the question: what are the research and major facilities required to develop the blanket/FW to a level which enables the design, construction and successful operation of a fusion DEMO? The principal challenges in the development of the blanket/FW are: (1) the Fusion Nuclear Environment – a multiple-field environment (neutrons, heat/particle fluxes, magnetic field, etc.) with high magnitudes and steep gradients and transients; (2) Nuclear Heating in a large volume with sharp gradients – the nuclear heating drives most blanket phenomena, but accurate simulation of this nuclear heating can be done only in a DT-plasma based facility; and (3) Complex Configuration with blanket/first wall/divertor inside the vacuum vessel – the consequence is low fault tolerance and long repair/replacement time.
These blanket/FW development challenges result in critical consequences: (a) non-fusion facilities (laboratory experiments) need to be substantial to simulate multiple fields/multiple effects and must be accompanied by extensive modeling; (b) results from non-fusion facilities will be limited and will not fully resolve key technical issues. A DT-plasma based fusion nuclear science facility (FNSF) is required to perform “multiple effects” and “integrated” experiments in the fusion nuclear environment; and (c) the Reliability/Availability/Maintainability/Inspectability (RAMI) of fusion nuclear components is a major challenge and is one of the primary reasons why the blanket/FW will pace fusion development toward a DEMO.
This paper summarizes the top technical issues and elucidates the primary challenges in developing the blanket/first wall and identifies the key R&D needs in non-fusion and fusion facilities on the path to DEMO.
Due to the lack of external tritium sources, all fusion power plants must demonstrate a closed tritium fuel cycle. The tritium breeding ratio (TBR) must exceed unity by a certain margin. The key ...question is: how large is this margin and how high should the calculated TBR be? The TBR requirement is design and breeder-dependent and evolves with time. At present, the ARIES requirement is 1.1 for the calculated overall TBR of LiPb systems. The Net TBR during plant operation could be around 1.01. The difference accounts for deficiencies in the design elements (nuclear data evaluation, neutronics code validation, and 3D modeling tools). Such a low Net TBR of 1.01 is potentially achievable in advanced designs employing advanced physics and technology. A dedicated R&D effort will reduce the difference between the calculated TBR and Net TBR. A generic breeding issue encountered in all fusion designs is whether any fusion design will over-breed or under-breed during plant operation. To achieve the required Net TBR with sufficient precision, an online control of tritium breeding is highly recommended for all fusion designs. This can easily be achieved for liquid breeders through online adjustment of Li enrichment.
Developing a divertor concept for fusion power plants to be built after ITER is deemed to be an urgent task to meet the EU Fast Track scenario. This task is particularly challenging because of the ...wide range of requirements to be met, namely, the high incident peak heat flux, the blanket design with which the divertor has to be integrated, sputtering erosion of the plasma-facing material caused by the incident particles from the plasma, radiation effects on the properties of structural materials, and efficient recovery and conversion of the considerable fraction (∼15%) of the total fusion thermal power incident on the divertor.
This paper provides an overview of the development of different conceptual divertor designs (water-cooled, liquid metal-cooled, and helium-cooled types); their advantages and disadvantages and expected performance are outlined and discussed. Emphasis is placed on summarizing the status and progress of R&D associated with He-cooled divertor designs which have been proposed in most of conceptual plant models in Europe and USA.
The DCLL is an attractive breeding blanket concept that leads to a high-temperature (T∼700°C), high thermal efficiency (η>40%) blanket system. The key element of the concept is a flow channel insert ...(FCI) that serves as an electrical and thermal insulator to reduce the magnetohydrodynamic (MHD) pressure drop and to decouple the temperature-limited RAFM (reduced-activation ferritic/martensitic) steel wall from the flowing hot PbLi. The paper introduces the concept, reviews history of the development of the DCLL in the US and worldwide and then identifies critical R&D needs prior to fusion environment testing in four research areas important to the successful development of the DCLL concept: (1) PbLi MHD thermofluids, (2) fluid materials interaction, (3) tritium transport, and (4) FCI development and characterization. For these areas, the most important R&D results obtained in the US in the ITER DCLL TBM program (2005–2011) and more recently are reviewed, including experimental and computational studies of MHD PbLi flows, corrosion of RAFM, tritium permeation, and silicon carbide FCI fabrication and material qualification. We also discuss required features of non-fusion facilities for DCLL blanket testing, where current lab experiments and modeling could progress to multiple effects and partially-integrated studies that approach as nearly as possible prototypic, integrated blanket conditions prior to testing in a fusion environment.
A Fusion Nuclear Science Facility (FNSF) has been recognized in the fusion community as a necessary facility to resolve the critical technology issues of in-vessel components prior to the ...construction of a DEMO reactor (Abdou et al., 1996) 1. Among these components, development of a reliable, low-cost and safe blanket system that provides self-sufficient tritium breeding and efficient conversion of the extracted fusion energy to electricity, while meeting all material, design and configuration limitations is among the most important but still challenging goals. In the recent FNSF study in the US (Kesel et al., 2015) 2, a Dual-Coolant Lead-Lithium (DCLL) blanket has been selected as the main breeding blanket concept. This paper summarizes the most important details of the proposed DCLL blanket design, presents the MHD thermohydraulic analysis for the PbLi flows in the blanket conduits and introduces supporting R&D studies, which are presently ongoing at UCLA. We also discuss the required pre-FNSF R&D in the area of MHD Thermofluids to support the further work on the DCLL blanket design & analysis and its integration into the fusion facility.
The dual-coolant (DC) blanket—characterised by its simple construction, simple function, and high thermal efficiency—is one of the EU advanced blanket concepts to be investigated in the frame of the ...long-term power plant conceptual study (PPCS). Its basic concept is based on the use of helium-cooled ferritic steel structure, the self-cooled Pb–17Li breeding zone, and SiC/SiC flow channel inserts, serving as electrical and thermal insulators. The present work on PPCS is drawn extensively on the preparatory study on plant availability carried out in 1999 with an objective to perform the conceptual design of the DC blanket concept where some details are to be selected in accordance with the overall strategy, which allows an extrapolation of the present knowledge between the near-term solutions (helium-cooled pebble bed (HCPB), water-cooled lead–lithium (WCLL) blanket concepts), and the very advanced self-cooled Pb–17Li SiC/SiC (SCLL) blanket concept. In the PPCS the reactor power is adapted to a typical size of commercial reactors of 1500 MWe which requires iterative calculations between the blanket layout and the system code analysis. The results of the first iteration are reported. This work is under the coordination of FZK in co-operation with CEA, EFET, IBERTEF, UKAEA, VTT Processes and VR.
A computational suite called TRANSMAG has been developed to address corrosion of ferritic/martensitic steels and associated transport of corrosion products in the eutectic alloy PbLi as applied to ...blankets of a fusion power reactor. The computational approach is based on simultaneous solution of flow, energy and mass transfer equations with or without a magnetic field, assuming mass transfer controlled corrosion and uniform dissolution of iron in the flowing PbLi. First, the new tool is applied to solve an inverse mass transfer problem, where the saturation concentration of iron in PbLi at temperatures up to 550°C is reconstructed from the experimental data on corrosion in turbulent flows without a magnetic field. As a result, a new correlation for the saturation concentration CS has been obtained in the form CS=e13.604–12975/T, where T is the temperature of PbLi in K and CS is in wppm. Second, the new correlation is used in the computations of corrosion in laminar flows in a rectangular duct in the presence of a strong transverse magnetic field. As shown, the mass loss increases with the magnetic field such that the corrosion rate in the presence of a magnetic field can be a few times higher compared to purely hydrodynamic flows. In addition, the corrosion behavior was found to be different between the side wall of the duct (parallel to the magnetic field) and the Hartmann wall (perpendicular to the magnetic field) due to formation of high-velocity jets at the side walls. The side walls experience a stronger corrosion attack demonstrating a mass loss up to 2–3times higher compared to the Hartmann walls. Also, computations of the mass loss are performed to characterize the effect of the temperature (400–550°C) and the flow velocity (0.1–1m/s) on corrosion in the presence of a strong 5T magnetic field prototypic to the outboard blanket conditions.
A few of the key highlights covered in this paper include:
•The FNSF maintenance approach establishes guidance for future fusion plants.•The horizontal sector maintenance approach is preferred to the ...vertical approach.•Autonomous robotic maintenance will be a mainstream technology when FNSF is built.•Power Core maintenance times are estimated to yield an acceptable plant availability.•Hot Cell requirements and functionality are conceptualized for a fusion power plant.
This paper addresses an approach to maintain and sustain the high-intensity Fusion Nuclear Science Facility (FNSF) experimental fusion facility that would enable it to have extended operations exceeding several exchanges of the primary power core components. The maintenance approach must be safe, quick, reliable, repeatable and precise. Moreover, this approach and equipment should be able to be extrapolated to the future fusion demonstration power plant (DEMO). A preliminary evaluation is presented for the underlying maintenance and safety requirements and the key design approaches that will define the FNSF and shape future high-power fusion facilities.
The proposed Fusion Nuclear Science Facility (FNSF) represents the first facility to enter the complex fusion nuclear regime, and its technical mission and attributes are being developed. The FNSF ...represents one part of the fusion energy development pathway to the first commercial power plant with other major components being the pre-FNSF research and development, research in parallel with the FNSF, pre-DEMO research and development, and the demonstration power plant (DEMO). The Fusion Energy Systems Studies group is developing the technical basis for the FNSF in order to provide a better understanding of the demands on the fusion plasma and fusion nuclear science programs.
The pre-conceptual design for the US DEMO inboard dual-coolant lead–lithium breeding blanket is introduced for the first time followed by the assessment for the most important magnetohydrodynamic ...issues for the blanket module itself and access ducts. The considered issues include: (i) the magnetohydrodynamic pressure drop, (ii) electric insulation in poloidal flows using the silicon carbide flow channel insert, and (iii) countercurrent flows in the access ducts.
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