Helical reactor design FFHR-d1 and c1 for steady-state DEMO

NIFS launched the Fusion Engineering Research Project (FERP) in preparation for DEMO by starting the redesign of the LHD-type helical reactor FFHR-d1. In the first round, the main parameters were selected. The second round is preparing detailed three-dimensional (3D) design of the superconducting magnet support structures, and 3D neutronics analyses, where the diverter targets can be efficiently shielded from fast neutrons. A new Flinabe blanket mixed with metal powder was proposed. Fabrication of helical coils by connecting half-helical-pitch segments of 100 kA-class YBCO high-temperature superconductors is proposed as a promising method. Also in progress is improvement of the first round of the core plasma design, ignition start-up analyses, and fueling scenario. As a consequence, a multi-path strategy on FFHR-d1 has been introduced with versions of -d1A, -d1B, and -d1C, where design flexibility is expanded to include subignition with options FFHR-c1 for “before demo, compact, and component-test.”     

Design of structural components for the helical reactor FFHR-d1A

FFHR-d1 is a conceptual design of the helical reactor being developed at the National Institute for Fusion Science. The maintenance of in-vessel components is very important for the fusion demo reactor. In addition, sufficient pathways are needed for the divertor exhaust. To implement these, the vacuum vessel, coil support structure, and cryostat require large apertures. However, the coil support structure has to be sufficiently rigid to remain within soundness and deformation limits. A design combining the structural components in the FFHR-d1A was developed from mechanical and thermal viewpoints. Consequently, components having a sufficiently large port area were provided. An investigation of the maintenance and exhaust schemes has been planned on the basis of this fundamental design.     

Analysis of radiation environment at divertor in helical reactor FFHR-d1

Neutron transport calculations with a three-dimensional model of the helical reactor FFHR-d1 have been performed for the accurate analysis of neutronics environment in the divertor areas. Based on the obtained neutron spectra, magnitudes of irradiation damage, contact dose rates and decay heat have been evaluated mainly for Fe, W and Cu. Since divertors can be placed behind radiation shields in helical reactors, magnitudes of damage and radioactivation at the outboard divertors are almost two orders lower than those at blanket first walls. Cu materials could be used as a cooling channel material of the outboard divertors. In contrast, the magnitudes at the inboard divertors are only one order lower compared with those at the first walls due to the limited space at the inboard side. Damage on Cu is evaluated to be ∼10 dpa after 6 years operation. Further efforts in divertor development and reactor design could suppress the magnitude of damage to less than half for adoption of Cu materials for the inboard divertors.     

Contribution to DEMO reactor RH maintenance assessment

The scope of this paper is a preliminary assessment of the maintenance scheme in support of the European study for the next generation of fusion reactor: DEMO. Despite other fusion machine requiring remote handling maintenance operations, DEMO is supposed to work under steady state operational conditions. Therefore, requirement on the maintenance scheme is stronger. To target a good availability of the machine along machine operation plan, it is necessary to draw an adequate maintenance scheme. Indeed, due to the high fluxes generated by the plasma in the vacuum vessel, the first wall lifetime is limited, so the frequent replacement is necessary. On current fusion experimental machine, as first wall load conditions are less severe, DEMO condition implies high level of requirement on maintenance time. During DEMO lifetime, several full first wall replacements are expected. To provide access to the vacuum vessel machine for first wall removal, preparatory work is required to set the machine to adequate maintenance conditions and to open the machine properly, the same situation at the end of the maintenance period. Shutdown duration for first wall replacement should be as short as possible to reach the availability target of the machine. From this statement, the maintenance duration should not exceed 20% of the total lifetime of the reactor operation. First wall segmentation (i.e. total number of component to replace) has a high impact onto the replacement time. Considering the number of feasible designs for the first wall segmentation, we concentrate remote handling concept assessments one type of segmentation, the one minimizing the numbers of modules to replace [4], [5], [6]. Assumption on Divertor segmentation for these DEMO studies have similarities with Divertor ITER design; therefore ITER design output is relevant [1], [2]. We assume divertor removal performed in shadow time, while removing the other first wall modules.     

Effects of lithium burn-up on TBR in DEMO reactor SlimCS

Lithium in a breeding blanket is burned up through neutron nuclear reactions in fusion DEMO reactors. Effects of decrease of solid breeder materials due to lithium burn-up on tritium breeding ratio (TBR) are not systematically calculated in the past. For the SlimCS blanket design, TBR is calculated taking into account the lithium burn-ups by one dimensional Sn radiation transport calculation code ANISN in this study. The ⁶Li burn-ups are 8–79% after 10-year operation. TBR due to ⁶Li decreases to 40% of the initial one in some layer, while it increases in some layers. The TBR integrated over all the blanket decreases to around 96% of the initial one. The study makes it clear that the reduction of the TBR due to the lithium burn-up is not so large.     

System analysis study for Korean fusion DEMO reactor

A conceptual design study for a steady-state Korean fusion DEMO reactor (K-DEMO) has been initiated. Two peculiar features need to be noted. First, the major radius is designed to be just below 6.5 m, considering practical engineering feasibilities. But still, high magnetic field at the plasma center around 8 T is expected to be achieved by using current state-of-the-art high performance Nb₃Sn strand technology. Second, a two-stage development plan is being considered. In the first stage, K-DEMO will demonstrate a net electricity generation but will also act as a component test facility. Then, after a major upgrade, K-DEMO is expected to show a net electric generation on the order of 300 MWe and the competitiveness in cost of electricity (COE). Feasibility of such a practical, near-future demonstration reactor is studied in this paper, based on a zero dimensional system analysis code study. It was shown that a net electric generation on the order of 300 MWe can be achieved below the optimistic β_N limit of 5. The elongation of K-DEMO is around 1.8 with single null configuration. Detailed optimization process and the resultant various plasma parameters are described.     

Research and development status on fusion DEMO reactor design under the Broader Approach

The main objective of DEMO design activity under the Broader Approach is to develop pre-conceptual design of DEMO options by addressing key design issues on physics, technology and system engineering. This paper describes the latest results of the design activity, including DEMO parameter study, divertor and remote maintenance. DEMO parameter study has recently started with “pulsed” DEMO having a major radius (R_p) of 9 m, and “steady state” DEMO of R_p = 8.2 m or more. Divertor design study has focused on a computer simulation of fully detached plasma under DEMO divertor conditions and the assessment of advanced divertor configuration such as super-X. Comparative study of various maintenance schemes for DEMO and narrowing down the schemes is in progress.     

A preliminary conceptual design study for Korean fusion DEMO reactor

As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy, so called fast-track approach. Korean strategy for fusion energy can be regarded as a fast-track approach and one special concept discussed in this paper is a two-stage development plan. At first, a steady-state Korean DEMO Reactor (K-DEMO) is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used as a component test facility. Then, at its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electric generation on the order of 300 MWe and the competitiveness in cost of electricity (COE). The major radius is designed to be just below 6.5 m, considering practical engineering feasibilities. By using high performance Nb₃Sn-based superconducting cable currently available, high magnetic field at the plasma center above 8 T can be achieved. A design concept for TF magnets and radial builds for the K-DEMO considering a vertical maintenance scheme, are presented together with preliminary design parameters.     

Design studies on three-dimensional issues for liquid blanket systems in helical reactor FFHR

A new design activity is under way for a helical type DEMO reactor FFHR-d1. The first stage of the activity involves the fundamental issues related to three-dimensional blanket design: (1) the minimum blanket space required for reactor parameter decisions, (2) the support method for the helical blanket system, and (3) the blanket module design. Investigations have been performed with neutronics and mechanical finite-element method calculations. Neutronics investigations indicate that a tungsten carbide radiation shield could reduce the minimum blanket space requirement by ～30 cm at the inboard region of FFHR-d1 compared with the blanket space of ～100 cm in the previous FFHR2 design. The investigations also showed that main shielding materials, ferritic steel and B₄C, could be used separately in a two-layered shielding configuration. The ferritic steel layer of the radiation shield is considered suitable to support the helical blanket system instead of relying on a thin vacuum vessel of the helical reactor. A size of a blanket module for a replacement process and the preferable cooling channel direction under a magnetic field are also discussed.     

Conceptual study of vertical sector transport maintenance for DEMO fusion reactor

The maintenance scheme is a critical issue for DEMO design, and requires high availability of the reactor. The SlimCS, designed in JAEA, adopts the horizontal sector transport hot cell maintenance scheme. In order to determine the most appropriate DEMO reactor maintenance scheme, it is important to assess the various maintenance schemes. In this paper the vertical sector transport maintenance concept is proposed for the first time. In the sector maintenance scheme, the amount of cutting/re-welding of the piping is minimized. The sector including blanket modules and a high-temperature shield was divided into 10° segments in a toroidal direction. The sectors are designed to be removed and re-inserted through upper alternate vertical maintenance ports. In the case of the vertical sector transport maintenance scheme, inter-coil structures could be adopted for use against turnover force in toroidal field (TF) coils. This shows an advantage in DEMO maintenance compared with horizontal sector transport.     

Conceptual design for the superconducting magnet system of a pulsed DEMO reactor

A methodology has been developed to consistently investigate, taking into account main reactor components, possible magnet solutions for a pulsed fusion reactor aiming at a large solenoid flux swing duration within the 2–3 h range. In a conceptual approach, investigations are carried out in the equatorial plane, taking into account the radial extension of the blanket-shielding zone, of the toroidal field magnet system inner leg and of the central solenoid for estimation of the pulsed swing.Design criteria are presented for the radial extension of the superconducting magnets, which is mostly driven by the structures (casings and conductor jacket). Typical available cable current densities are presented as a function of the magnetic field and of the temperature margin.The magnet design criteria have been integrated into SYCOMORE, a code for reactor modeling presently in development at CEA/IRFM in Cadarache, using the tools of the EFDA Integrated Tokamak Modeling task force.Possible solutions are investigated for a 2 GW fusion power reactor with different aspect ratios.The final adjustment of the DEMO pulsed reactor parameters will have to be consistently done, considering all reactor components, when the final goals of the machine will be completely clarified.     

Overview of EU DEMO design and R&D activities

One important objective of the EU fusion roadmap Horizon 2020 is to lay the foundation of a Demonstration Fusion Power Reactor (DEMO) to follow ITER, with the capability of generating several 100 MW of net electricity to the grid and operating with a closed fuel-cycle by 2050. This is currently viewed by many of the nations engaged in the construction of ITER as the remaining crucial step towards the exploitation of fusion power. This paper outlines the DEMO design and R&D approach that is being adopted in Europe and presents some of the preliminary design options that are under evaluation as well as the most urgent R&D work that is expected to be launched in the near-future. The R&D on materials for a near-term DEMO is discussed in detail elsewhere.     

Study of safety features and accident scenarios in a fusion DEMO reactor

After the Fukushima Dai-ichi nuclear accident, a need for assuring safety of fusion energy has grown in the Japanese (JA) fusion research community. DEMO safety research has been launched as a part of Broader Approach DEMO Design Activities (BA-DDA). This paper reports progress in the fusion DEMO safety research conducted under BA-DDA. Safety requirements and evaluation guidelines have been, first of all, established based on those established in the Japanese ITER site invitation activities. The radioactive source terms and energies that can mobilize such source terms have been assessed for a reference DEMO concept. This concept employs in-vessel components that are cooled by pressurized water and built of a low activation ferritic steel (F82H), contains solid pebble beds made of lithium-titanate (Li₂TiO₃) and beryllium–titanium (Be₁₂Ti) for tritium breeding and neutron multiplication, respectively. It is shown that unlike the energies expected in ITER, the enthalpy in the first wall/blanket cooling loops is large compared to the other energies expected in the reference DEMO concept. Reference accident event sequences in the reference DEMO in this study have been analyzed based on the Master Logic Diagram and Functional Failure Mode and Effect Analysis techniques. Accident events of particular concern in the DEMO have been selected based on the event sequence analysis and the hazard assessment.     

Neutronic analyses of the HCPB DEMO reactor using a consistent integral approach

This work is devoted to nuclear design analyses of the new HCPB-type DEMO reactor developed in the frame of the EFDA PPPT program. The neutronic simulations were carried out with the MCNP5 code using a full scale 3D torus sector model of the DEMO reactor. The model was generated with the McCad conversion tool from available CAD models using a consistent integral approach. The neutronic analyses addressed the tritium breeding performance, the nuclear power generation and the shielding capabilities of the reactor. Although tritium self-sufficiency was shown, the tritium breeding performance of the current design calls for further design improvements to arrive at a higher uncertainty margin. The shielding performance of the reactor is close to the limit. Sufficient shielding can be easily provided by a slight increase of the inboard shield thickness.     

Evaluation of remote maintenance schemes by plasma equilibrium analysis in Tokamak DEMO reactor

The remote maintenance schemes in a DEMO reactor are categorized by insertion direction, blanket segmentation, and divertor maintenance scheme, and are quantitatively evaluated by analysing the plasma equilibrium. The positions of the poloidal field (PF) coil are limited by the size of the toroidal field (TF) coil and the maintenance port layout of each remote maintenance scheme. Because the PF coils are located near the larger TF coil and far from the plasma surface, the horizontal sector transport maintenance scheme requires the largest part of total PF coil current, 25% larger than that required for separated sector transport using vertical maintenance ports with segmented divertor maintenance (SDM). In the unsegmented divertor maintenance (UDM) scheme, the total magnetic stored energy in the PF coils at plasma equilibrium is about 30% larger than that stored in the SDM scheme, but the time required for removal and installation of all the divertor cassettes in the UDM scheme is roughly a third of that required in the SDM scheme because the number of divertor cassettes in the UDM scheme is a third of that in the SDM scheme. From the viewpoint of simple maintenance operations, the merit of the UDM scheme has more merit than the SDM scheme.     

DEMO design activities in the broader approach under Japan/EU collaboration

The DEMO design studies in the BA (broader approach in the field of fusion energy) are being conducted by the DEMO Design Activity unit of International Fusion Energy Research Centre for the broader approach (BA) and the Home Teams in EU and Japan since 2011. The activity covers most of the critical issues on the DEMO design. Emphasis during the last two years was on studies to develop the best embodiment of a tokamak as a power reactor consistent with credible operating scenarios and feasible engineering solutions to critical design issues. The technical activities have focused on, for example, plasma physics for DEMO plants, divertor physics and technology, in-vessel components, maintenance schemes and safety research.     

Design of a brazing connector for DEMO in-vessel components

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.     

Neutronic design analyses for a dual-coolant blanket concept: Optimization for a fusion reactor DEMO

The generation of design specifications for a DEMO reactor, including breeding blanket (BB), vacuum vessel (VV) and magnetic field coils (MFC), requires a consistent neutronic optimization of structures between plasma and MFC. This work targets iteratively to generate these neutronic specifications for a Dual-Coolant He/Pb15.7Li breeding blanket design. The iteration process focuses on the optimization of allowable space between plasma scrapped-off-layer and VV in order to generate a MFC/VV/BB/plasma sustainable configuration with minimum global system volumes. Two VV designs have been considered: (1) a double-walled option with light-weight stiffeners and (2) a thick massive one. The optimization process also involves VV materials, looking to warrant radiation impact operational limits on the MFC. The resulting nuclear responses: peak nuclear heating in toroidal field (TF) coil, tritium breeding ratio (TBR), power amplification factor and helium production in the structural material are provided.     

Parametric neutronic analysis of HCLL blanket for DEMO fusion reactor utilizing vacuum vessel ITER FDR design

Three-dimensional parametric neutronics calculations using the Monte Carlo code MCNP-4C have been performed for a DEMO-type reactor based on the Helium-Cooled Lithium-Lead (HCLL) blanket. The aim of the analysis was to minimize the radial blanket thickness, while ensuring tritium self-sufficiency and to assess the shielding performance of the reactor in terms of the radiation loads to the super-conducting toroidal field (TF) coils. It was found that tritium self-sufficiency can be achieved with a breeder zone thickness reduced to no more than 55 cm at a ⁶Li enrichment of 90%. Assuming a ⁶Li enrichment of 60%, a breeder zone thickness of 60 cm is required to achieve the target TBR of 1.10 which is assumed to be sufficient to cover potential tritium losses and uncertainties. With regard to the shielding performance it was found that the design limits for the radiation loads to the TF-coil can be met with radial blanket thicknesses of 75 cm, 60 cm and 55 cm utilizing a two-component shield of Eurofer steel and tungsten carbide between the breeder zone and the vacuum vessel. The blanket variants with larger radial breeder zone show better shielding performances due to the reduced Eurofer shielding material acting as gamma radiation emitter in between the breeder zone and the vacuum vessel. In particular the radiation dose absorbed in the Epoxy insulator was shown to be the most critical quantity in this regard.     

Potential and limits of water cooled divertor concepts based on monoblock design as possible candidates for a DEMO reactor

In this paper water-cooled divertor concepts based on tungsten monoblock design identified in previous studies as candidate for fusion power plant have been reviewed to assess their potential and limits as possible candidates for a DEMO concept deliverable in a short to medium term (“conservative baseline design”). The rationale and technology development assumptions that have led to their selection are revisited taking into account present factual information on reactor parameters, materials properties and manufacturing technologies.For that purpose, main parameters impacting the divertor design are identified and their relevance discussed. The state of the art knowledge on materials and relevant manufacturing techniques is reviewed. Particular attention is paid to material properties change after irradiation; phenomenon thresholds (if any) and possible operating ranges are identified (in terms of temperature and damage dose). The suitability of various proposed heat sink/structural and sacrificial layer materials, as proposed in the past, are re-assessed (e.g. with regard to the possibility of reducing peak heat flux and/or neutron radiation damages). As a result, potential and limits of various proposed concepts are highlighted, ranges in which they could operate (if any) defined and possible improvements are proposed.Identified missing point in materials database and/or manufacturing techniques knowledge that should be uppermost investigated in future R&D activities are reported.This work has been carried out in the frame of EFDA PPPT Work Programme activities.