Construction Of DEMO Research Articles

Online capability-based resource allocation for on-site construction operations utilizing digital twin models

Planning and controlling on-site construction operations are complex and dynamic procedures, mainly manually executed without algorithmic decision support. An initial challenge is to allocate available resources to construction processes based on required and available capabilities. Due to the dynamic nature of construction projects (e.g., redesigns, resource failure, unpredictable restrictions), there is a demand for frequent reallocation of resources. In recent years, researchers studied capability-based resource allocation approaches by defining ontologies to describe the capabilities of resources. However, since most of the existing approaches focus on ontologies for resources in production environments (e.g., industrial robots), the modeling and application of the models for online allocation in dynamic construction environments remain unsolved. In this study, an ontology-based Digital Twin model, adopted from a production engineering background, is used to enable online capability-based resource allocations for construction-specific approaches. The Digital Twin model can be updated by a lightweight, publish-subscribe network, triggering an update of capability-based feasibility checks for resource allocations. The resulting framework is tested on a demo construction project from the research project “Internet of Construction (IoC)”. The results contribute to the automation of planning and controlling resource allocations for dynamic on-site construction operations. Using machine-readable ontologies, the transition from manually performed activities to robotically supported tasks is enabled.

Preliminary design and structural assessment of DTT in-vessel coil supports

The present work regards the design methodologies for a crucial component of the new thermonuclear fusion energy machine named “Divertor Tokamak Test facility” (DTT). The purpose of this experimental facility is the investigation of the thermal exhaust issue, a critical aspect in the view of DEMO construction, and future nuclear fusion reactors for commercial use. The construction of DTT machine is starting in the ENEA Site, Frascati, Italy. The poloidal magnet system in the proposed DTT includes a central solenoid divided into 6 independent modules, plus 6 external coils. The poloidal field system also includes six copper in-vessel coils, namely two in-vessel coils for radial and vertical plasma stabilization and control, and four out of six in-vessel divertor coils for magnetic control of scrape-off layer and strike point sweeping. This work aims to provide a preliminary design of the supports for the 6 in-vessel coils considering the Electromagnetic (EM) loads due to a plasma Downward (DW) Vertical Displacement Event (VDE) at the flat-top of plasma reference single null scenario. During a DW VDE a huge amount of current flows in the in-vessel coils (circumferential direction), whose interaction with the poloidal EM field generates strong EM forces. The electromagnetic loads have been calculated with MAXFEA, a two-dimensional finite elements code for the simulation of plasma equilibrium, and then applied to homogenized in-vessel coils in the structural analysis.

Neutronics analyses for the bio-shield and liners of the IFMIF-DONES test cell

IFMIF-DONES (International Fusion Material Irradiation Facility- DEMO Oriented NEutron Source) is an accelerator based irradiation facility which is aimed to provide the irradiation data needed for the construction of DEMO. The Test Cell (TC) is the central room of IFMIF-DONES enclosing the target and the test module. This work presents the neutronics analyses conducted to provide the relevant nuclear responses for the bio-shield of the TC, including heating, radiation damage and helium production. The results for the TC steel liners covering the inner TC surface have been obtained with an unstructured mesh based interpolation approach. The results show that the softening of the neutron spectrum at the thin layer of the inner wall increases the nuclear heating as well as the helium production in SS316 L. The helium production might be of major concern if re-welding of the liner is required during the lifetime. The impact of a reduced beam footprint size on the TC nuclear responses has been analyzed and found to be insignificant comparing with the normal footprint.

Japanese Fusion Materials Development Path to DEMO

This paper reports Japanese strategy for developing blanket structural materials for DEMO. In the Japanese program, the candidate materials are categorized into Primary Option (RAFM) and Advanced Options (V-alloy, SiC/SiC, ODS-RAFM etc.). A staged development is planned corresponding to the three decision-making points (DPs), DP1: Intermediate check and review (C&R), DP2: Decision of transition of research and development (R&D) focus to DEMO, and DP3: Decision of DEMO construction. The near-term D-Li neutron source (A-FNS) and IFMIF are regarded as key facilities for the development. The strategy emphasizes “standardization” as an important step toward DEMO design qualification and licensing. The procedure to standard materials specifications by way of establishing structural design criteria and materials property requirements, and the procedure’s interaction with the schedule of irradiation data acquisition are discussed.

Developing structural, high-heat flux and plasma facing materials for a near-term DEMO fusion power plant: The EU assessment

The findings of the EU ‘Materials Assessment Group’ (MAG), within the 2012 EU Fusion Roadmap exercise, are discussed. MAG analysed the technological readiness of structural, plasma facing and high heat flux materials for a DEMO concept to be constructed in the early 2030s, proposing a coherent strategy for R&D up to a DEMO construction decision.A DEMO phase I with a ‘Starter Blanket’ and ‘Starter Divertor’ is foreseen: the blanket being capable of withstanding ⩾2MWyrm−2 fusion neutron fluence (∼20dpa in the front-wall steel). A second phase ensues for DEMO with ⩾5MWyrm−2 first wall neutron fluence. Technical consequences for the materials required and the development, testing and modelling programmes, are analysed using: a systems engineering approach, considering reactor operational cycles, efficient maintenance and inspection requirements, and interaction with functional materials/coolants; and a project-based risk analysis, with R&D to mitigate risks from material shortcomings including development of specific risk mitigation materials. The DEMO balance of plant constrains the blanket and divertor coolants to remain unchanged between the two phases. The blanket coolant choices (He gas or pressurised water) put technical constraints on the blanket steels, either to have high strength at higher temperatures than current baseline variants (above 650°C for high thermodynamic efficiency from He-gas coolant), or superior radiation-embrittlement properties at lower temperatures (∼290–320°C), for construction of water-cooled blankets. Risk mitigation proposed would develop these options in parallel, and computational and modelling techniques to shorten the cycle-time of new steel development will be important to achieve tight R&D timescales. The superior power handling of a water-cooled divertor target suggests a substructure temperature operating window (∼200–350°C) that could be realised, as a baseline-concept, using tungsten on a copper-alloy substructure. The difficulty of establishing design codes for brittle tungsten puts great urgency on the development of a range of advanced ductile or strengthened tungsten and copper compounds.Lessons learned from Fission reactor material development have been included, especially in safety and licensing, fabrication/joining techniques and designing for in-vessel inspection. The technical basis of using the ITER licensing experience to refine the issues in nuclear testing of materials is discussed.Testing with 14MeV neutrons is essential to Fusion Materials development, and the Roadmap requires acquisition of ⩾30dpa (steels) 14MeV test data by 2026. The value and limits of pre-screening testing with fission neutrons on isotopically- or chemically-doped steels and with ion-beams are evaluated to help determine the minimum14MeV testing programme requirements.

Progress in the physics basis of a Fusion Nuclear Science Facility based on the Advanced Tokamak concept

Physics based integrated modelling of the baseline scenario for a Fusion Nuclear Science Facility based on the Advanced Tokamak concept (FNSF-AT) (Chan et al 2010 Fusion Sci. Technol. 57 66) has found steady-state equilibria with good stability and controllability properties at the fusion performance required to accomplish FNSF's nuclear science mission with margin. 2D divertor analysis for this baseline scenario predicts that peak heat flux <10 MW m−2 can be obtained even with scrape-off layer power width ∼1 mm. Using this baseline fusion performance, high fidelity and high-resolution 3D neutronics calculations show acceptable cumulative end-of-life organic insulator dose levels in all the device coils, and TBR >1. Two current drive scenarios, two divertor configurations, and two blanket concepts have been analysed. FNSF-AT would complement ITER in addressing science and technology gaps to a commercially attractive DEMO, and could enable a DEMO construction decision triggered by the achievement of Q = 10 in ITER.

A Fusion Nuclear Science Facility for a fast-track path to DEMO

An accelerated fusion energy development program, a “fast-track” approach, requires proceeding with a nuclear and materials testing program in parallel with research on burning plasmas, ITER. A Fusion Nuclear Science Facility (FNSF) would address many of the key issues that need to be addressed prior to DEMO, including breeding tritium and completing the fuel cycle, qualifying nuclear materials for high fluence, developing suitable materials for the plasma-boundary interface, and demonstrating power extraction. The Advanced Tokamak (AT) is a strong candidate for an FNSF as a consequence of its mature physics base, capability to address the key issues, and the direct relevance to an attractive target power plant. The standard aspect ratio provides space for a solenoid, assuring robust plasma current initiation, and for an inboard blanket, assuring robust tritium breeding ratio (TBR) >1 for FNSF tritium self-sufficiency and building of inventory needed to start up DEMO. An example design point gives a moderate sized Cu-coil device with R/a=2.7m/0.77m, κ=2.3, BT=5.4T, IP=6.6 MA, βN=2.75, Pfus=127MW. The modest bootstrap fraction of ƒBS=0.55 provides an opportunity to develop steady state with sufficient current drive for adequate control. Proceeding with a FNSF in parallel with ITER provides a strong basis to begin construction of DEMO upon the achievement of Q∼10 in ITER.

Neutronics analysis of the conceptual design of a component test facility based on the spherical tokamak

One of the crucial aspects of fusion research is the optimisation and qualification of suitable materials and components. To enable the design and construction of DEMO in the future, ITER is taken to demonstrate the scientific and technological feasibility and IFMIF will provide rigorous testing of small material samples. Meanwhile, a dedicated, small-scale components testing facility (CTF) is proposed to complement and extend the functions of ITER and IFMIF and operate in association with DEMO so as to reduce the risk of delays during this phase of fusion power development. The design of a spherical tokamak (ST)-based CTF is being developed which offers many advantages over conventional machines, including lower tritium consumption, easier maintenance, and a compact assembly. The neutronics analysis of this system is presented here. Based on a three-dimensional neutronics model generated by the interface programme MCAM from CAD models, a series of nuclear and radiation protection analyses were carried out using the MCNP code and FENDL2.1 nuclear data library to assess the current design and guide its development if needed. The nuclear analyses addresses key neutronics issues such as the neutron wall loading (NWL) profile, nuclear heat loads, and radiation damage to the coil insulation and to structural components, particularly the stainless steel vessel wall close to the NBI ports where shielding is limited. The shielding of the divertor coil and the internal Poloidal Field (PF) coil, which is introduced in the expanded divertor design, are optimised to reduce their radiation damage. The preliminary results show that the peak radiation damage to the structure of martensitic/ferritic steel is about 29 dpa at the mid-plane assuming a life of 12 years at a duty factor 33%, which is much lower than its ∼150 dpa limit. In addition, TBMs installed in 8 mid-plane ports and 6 lower ports, and 60% 6Li enrichment in the Li 4SiO 4 breeder, the total tritium generation is ∼3.7 kg in the whole lifetime of CTF.