Magnetic Field and Force of Helical Coils for Force Free Helical Reactor (FFHR)

The electromagnetic force on a helical coil becomes smaller by decreasing the coil pitch parameter which is the angle of the coil to the toroidal direction. This makes it possible to enlarge the central toroidal field or to simplify the supporting structures of the coil. The plasma minor radius, however, becomes smaller with the pitch parameter, and a higher field is necessary to attain the same plasma performance. Another important item in a helical reactor is the distance between the helical coil and the plasma to gain enough space for blankets. In order to reduce the mass of the coil supports, a lower aspect ratio is advantageous, and an optimum value of the pitch parameter will exist around 1.2 and 1.0 for the helical systems of the pole numbers of 2 and 3, respectively

Recent Fusion Research in the National Institute for Fusion Science

The National Institute for Fusion Science (NIFS), which was established in 1989, promotes academic approaches toward the exploration of fusion science for steady-state helical reactor and realizes the establishment of a comprehensive understanding of toroidal plasmas as an inter-university research organization and a key center of worldwide fusion research. The Large Helical Device (LHD) Project, the Numerical Simulation Science Project, and the Fusion Engineering Project are organized for early realization of net current free fusion reactor, and their recent activities are described in this paper. The LHD has been producing high-performance plasmas comparable to those of large tokamaks, and several new findings with regard to plasma physics have been obtained. The numerical simulation science project contributes understanding and systemization of the physical mechanisms of plasma confinement in fusion plasmas and explores complexity science of a plasma for realization of the numerical test reactor. In the fusion engineering project, the design of the helical fusion reactor has progressed based on the development of superconducting coils, the blanket, fusion materials and tritium handling

Design Concept of Supercritical CO2 Gas Cooled Divertors in FFHR Series Fusion Reactors

In the FFHR power reactor equipped with a supercritical CO2 gas turbine power generation system, an divertor cooling system is connected to this power generation system [S. Ishiyama et al., Prog. Nucl. Energy 50, No.12-6, 325 (2008) [1]]. In this paper, for the purpose of developing a diverter by supercritical CO2 gas cooling that can cope with a neutron heavy irradiation environment with a heat load of 15 MW/m2 or more, CFD heat transfer flow analysis was carried out for performance evaluation and its design optimization by a structural analysis models of a supercritical CO2 gas cooled divertors. As a result, in the supercritical CO2 gas cooled tungsten mono-block divertors (50 × 50 mm × 5 channel × 5,000 mL) with a flow path length of 5 m or less, the engineering designable range of these advanced diverters having the same cooling performance as the water cooling divertor was clarified, and its practicality is extremely high from the feature that the structural model has an extremely low risk during operation as compared with the water cooled divertor

Design of a Closed Helical Divertor in LHD and the Prospect for Helical Fusion Reactors

A new closed helical divertor configuration for efficient particle control and reduction of the heat load on the divertor plates is proposed. The closed divertor configuration practically utilizes an ergodic layer and magnetic field line configuration on divertor legs in helical systems. For optimization of the design of the closed divertor, the distribution of the strike points is calculated in various magnetic configurations in the Large Helical Device (LHD). It suggests that the installation of the closed divertor components in the inboard side of the torus under an inward shift configuration (Rax=3.60m) is the best choice for achieving the above two purposes. This divertor configuration does not interfere with plasma heating and diagnostic systems installed in outer ports. The prospect of the closed divertor configuration to a helical fusion reactor is investigated using a three-dimensional neutral particle transport simulation code with a one-dimensional plasma fluid calculation on the divertor legs. The investigation shows efficient particle pumping from the in board side and reduction of the heat load due to the combined effect of the optimized closed divertor geometry, ergodized divertor legs, and low electron temperature in the ergodic layer. It indicates a promising closed divertor configuration for helical fusion reactors

Microstructure of Oxide Insulator Coating before and after Thermal Cycling Test

Erbium oxide (Er2O3) was shown to be a high potential candidate for tritium permeation barrier and electrical insulator coating for advanced breeding blanket systems such as liquid Li, Li-Pb or molten-salt blankets. Recently, we succeeded to form Er2O3 coating layer on large interior surface area of metal pipe using Metal Organic Chemical Vapor Deposition (MOCVD) process. In this paper, we investigated the microstructure of Er2O3 coating layer on stainless steel 316 (SUS 316) plate before and after heat treatments with hydrogen or argon gases. From the results of TEM observations, we confirmed that Er2O3 coating layer with 700 nm thickness was formed on the SUS 316 plate and this layer was identified to poly-crystal phase because the diffraction fleck which was arranged like a ring was observed in the selected electron diffraction pattern. No macroscopic defects such as crack and peeling in Er2O3 coating layer were observed before and after thermal cycling test. The change of microstructure of the Er2O3 coating layer on before and after heat cycling test was reported

NITA Coil—Innovation for Enlarging the Blanket Space in the Helical Fusion Reactor

An innovative idea is proposed for enlarging the blanket space on the inboard side of the torus for the helical fusion reactor FFHR-d1. A set of sub-helical coils, named NITA coils, with opposite-directed current outside the main helical coils, effectively reduces the helical pitch parameter and enlarges the blanket space. Dependence of the blanket space and plasma volume on the effective helical pitch parameter is examined. The obtained magnetic surfaces and their properties are compared with that of the original configuration

Design Window Analysis for the Helical DEMO Reactor FFHR-d1

Conceptual design activity for the LHD-type helical DEMO reactor FFHR-d1 has been conducted at the National Institute for Fusion Science under the Fusion Engineering Research Project since FY2010. In the first step of the conceptual design process, design window analysis was conducted using the system design code HELIOSCOPE by the “Design Integration Task Group”. On the basis of a parametric scan with the core plasma design based on the DPE (Direct Profile Extrapolation) method, a design point having a major radius of 15.6 m and averaged magnetic field strength at the helical coil winding center of 4.7 T was selected as a candidate. The validity of the design was confirmed through the analysis by the related task groups (in-vessel component, blanket, and superconducting magnet)

Multiscale Stress Analysis and 3D Fitting Structure of Superconducting Coils for the Helical Fusion Reactor

Conceptual design studies for the Large-Helical-Device-type helical reactor, i.e., FFHR-d1, are being conducted in the National Institute for Fusion Science. Three different cooling schemes and conductor types have been proposed for the superconducting magnet system. A multiscale structural analysis is used to assess the mechanical characteristics of the magnet structure, taking into account the types of cooling schemes and superconductors. Multiscale analysis evaluates both the stress distribution in the coil support structure and local stress in the constituents of the superconductors without rebuilding a finite-element model of the support structure. Concerning a segmented fabrication of the helical coils using a high-temperature superconductor, the feasibility of segment installation is confirmed using a three-dimensional printing model, which identifies the maximum segment length and the necessary gap in the coil casing to install a segment