Force-free Helical Reactor Research Articles

Operating scenario of 3GWth class FFHR power plant with bypass controlled supercritical CO2 gas turbine power generation system

Abstract In order to achieve high power generation efficiency, facility compactness, high safety, and high coexistence with fuel tritium in the 3 G Wth class FFHR(Force Free Helical Reactor) power plant, optimization, performance and operation scenario of the power generation system of the power plant model using the axial flow type uniaxial supercritical CO2 gas turbine power generation system were examined. As a result, the following conclusions were obtained. (1) The most efficient and compact axial flow uniaxial design supercritical CO2 gas turbine power generation system suitable for 3 G WthFFHR power plant has a maximum heat capacity of 1.5GWth×2, and the gas turbine conversion rate at that time was evaluated as 46.9 %(= 1,267MWe/2,700MWth). Therefore, split connection with two power generation systems is optimal for a 3 G Wth class FFHR power plant. Here, the operating speed of the turbo equipment was set to 3600 rpm in order to adapt this plant model to the power supply system in Japan. (2) Simultaneous supply of 818MWe/300 s electricity and 300MWth/30days heat is required at startup of the FFHR. (3) The in-house power during steady operation under the self-ignition condition was reduced to about 31.7MWe, and the amount of power generated at the transmission end was evaluated as 1,234MWe.

Visualization of the Plasma Shape in a Force Free Helical Reactor, FFHR

Visualization of the Plasma Shape in a Force Free Helical Reactor, FFHR

Mechanical Design Concept of Superconducting Magnet System for Helical Fusion Reactor

The conceptual design of a helical fusion reactor was studied at the National Institute for Fusion Science in collaboration with other universities. Two types of the force free helical reactor (FFHR) are FFHR-d1 and FFHR-c1. FFHR-d1 is a self-ignition demonstration reactor that operates with a major radius of 15.6 m at a magnetic field intensity of 4.7 T. FFHR-c1 is a compact subignition reactor that aims to realize steady electrical self-sufficiency. Compared to FFHR-d1, FFHR-c1 has a magnetic field intensity of 7.3 T and a geometrical scale of 0.7. The location of the superconducting coils in both types of FFHR is based on that of the Large Helical Device (LHD). LHD has a major radius of 3.9 m. According to the design of LHD, the deformation must be within the required value to compensate for the accuracy of the magnetic field. According to this concept, the magnet support structure of LHD was fabricated using thick Type 316 stainless steel to impart sufficient rigidity. Thus, the stress of the magnet system of LHD is sufficiently below the permissible stress. In the case of FFHR, from the viewpoint of the reactor, a large access port is required for the maintenance of the in-vessel components. The mechanical design of the support structure is conceptualized by considering the basic thickness of the material and residual aperture space by referencing the mechanical analysis results. Details of the design concepts of LHD and FFHR-d1/FFHR-c1 as well as the results of mechanical analyses are introduced in this paper.

Bi-level spatial subdivision based monte carlo ray tracing directly using CAD models

To achieve fast Monte Carlo ray tracing on CAD models, a Bi-level Spatial Subdivision based ray tracing method (BSS) is proposed in this paper. This method was implemented and integrated in Super Monte Carlo Program for Nuclear and Radiation Simulation (SuperMC) and tested on ITER Benchmark and Force-Free Helical Reactor (FFHR) model. The testing results on ITER Benchmark model demonstrate the correctness of this method. The testing results on FFHR demonstrate that this method can save most of the preprocessing work to greatly improve the Monte Carlo analysis efficiency. Both results reveal that this method achieves applicable calculation speed comparing to the traditional Constructive Solid Geometry (CSG) broadly used in Monte Carlo codes.

Conceptual Design of DC Power Supplies for FFHR Superconducting Magnet

The force-free helical reactor (FFHR) is a helical-type fusion reactor whose design is being studied at the National Institute for Fusion Science. The FFHR will use three sets of superconducting coils to confine the plasma. It is not a fusion plasma experimental device, and the magnetic field configuration will be optimized for burning plasma. This paper introduces a conceptual design for a dc power system to excite the superconducting coils of the FFHR. In this design, the poloidal coils are divided into a main part, which generates a magnetic field for steady-state burning, and a control part, which is used in the ignition process to control the magnetic axis. The feasibility of this configuration was studied using the Large Helical Device coil parameters, and the coil voltages required to sweep the magnetic axis were calculated. It was confirmed that the axis sweep could be performed without a high output voltage from the main power supply. Finally, the power supply ratings for the FFHR were estimated from the stored magnetic energy.

Progress and Applications of MCAM: Monte Carlo Automatic Modeling Program for Particle Transport Simulation

MCAM (Monte Carlo Automatic Modeling program for particle transport simulation) was developed by FDS Team as a bi-directional interface program between general CAD systems and Monte Carlo particle transport simula- tion codes. The physics and material modeling and void space modeling functions were improved and the free-form surfaces processing function was developed recently. The applications to the ITER (International Thermonuclear Ex- perimental Reactor) building model and FFHR (Force Free Helical Reactor) model have demonstrated the feasibility, effectiveness and maturity of MCAM latest version for nuclear applications with complex geometry.

Hydrogen production via water splitting process in a molten-salt fusion breeder

This study presents the hydrogen production and fissile breeding potentials of Force-Free Helical Reactor (FFHR) fueled with the molten-salt mixtures. The sulfur–iodine (S–I) thermochemical water-splitting and high-temperature electrolysis cycles, which are the most promising water-splitting cycles, are selected to produce large-scale and pure hydrogen. The XSDRNPM/SCALE4.4a neutron transport code is used for the neutronic calculations. The analyses have been performed individually for four different molten-salt mixtures, (pure FLiBe, mixture of FLiBe and ThF 4, mixture of FLiBe and UF 4, and mixture of FLiBe, ThF 4 and 233UF 4). The numerical results bring out that the considered molten-salt fusion breeder reactor has a high neutronic performance and can produce a considerable amount of the hydrogen production (up to 40 kg/s), as well as the fissile fuel (up to 2.5 tons/yr).

Conceptual Design and Development of an Indirect-cooled Superconducting Helical Coil in the FFHR

The force-free helical reactor (FFHR) is a conceptual design of a heliotron fusion reactor being developed at the National Institute for Fusion Science. All the coils in the FFHR are made of superconductors. Several cooling schemes have been proposed for the helical coils; of these, indirect cooling is considered a good candidate. In this study, we investigated the possibility of using an indirect-cooled superconducting magnet for the FFHR. In parallel with this design study, we developed Nb3Sn superconductors, jacketed with an aluminum alloy, for use in an indirect-cooled magnet. The results of performance tests of a sub-scale superconductor showed good feasibility for application in the FFHR helical coil. Stress distribution in the helical coil was also analyzed, and the stress and strain were confirmed to be within the permissible range.

Stabilization of Unstable Low-Temperature and High-Density Operation in the Force-Free Helical Reactor

A new control method for the unstable operating point in the force-free helical reactor (FFHR) is proposed for low-temperature and high-density ignited operation. While in the stable ignition regime, the error of the fusion power of e’DT(Pf) = +(Pf0 - Pf) is used to obtain the desired fusion power with proportional-integral-derivative control of the fueling, we have discovered that in the unstable ignition regime, the error of the fusion power with an opposite sign of e’DT(Pf) = -(Pf0 - Pf) can stabilize the unstable operating point. Here, Pf0 is the fusion power set value, and Pf is the measured fusion power. Around the unstable operating point, excess fusion power (Pf0 < Pf) supplies fueling, increases the density, and then decreases the temperature. Less fusion power (Pf0 > Pf) in the subignited regime reduces the fueling, decreases the density, and then increases the temperature. While the operating point rotates to the clockwise direction in the stable ignition boundary, it rotates to the counterclockwise direction in the unstable ignition regime. Using this control algorithm, it is demonstrated that the operating point can reach the steady-state condition from an initial very low-temperature and low-density regime. The fusion power can also be shut down from the steady-state condition without any problems. Furthermore, characteristics of the stable and unstable ignition regimes are compared for the same fusion power, and control robustness to changes with various parameters has been studied.

Progress of design studies on an LHD-type steady state reactor

Progress of design studies on a heliotron reactor is reported. Recent experimental achievements in large helical device (LHD) are building up the physical basement. Integrated designs on LHD-type energy reactors have improved the base design of force free helical reactor and revealed economical feasibility. Large helical device is exploring a new horizon of net current free helical plasmas. Inherent characteristics demonstrate reactor relevant performances individually; steady state (1 h), high beta (5%), high density (1 × 10 21 m −3), etc. In particular, discovery of internal diffusion barrier provides a novel attractive scenario of super high density operation. It greatly facilitates physical requirements as well as mitigates engineering demands. New methods have been proposed to access and operate ignited plasmas; a long rise-up over 300 s to reduce the heating power to 30 MW and a new proportional-integration-derivative (PID) control of the fueling to handle the thermally unstable plasma at high-density operations.

Optimization activities on design studies of LHD-type reactor FFHR

Recent activities on optimizing the base design of the large helical device (LHD)-type helical reactor FFHR (force free helical reactor) are presented. Three candidates to secure the blanket space are proposed with the aim of reactor size optimization without deteriorating α-heating efficiency and by taking cost analyses into account. In this way the key engineering aspects are investigated; from 3D blanket designs, it is demonstrated that the peaking factor of the neutron wall loading is 1.2–1.3 and a blanket covering ratio of over 90% is possible by proposing discrete pumping with a semi-closed shield (DPSS) concept. Helical blanket shaping along the divertor field lines is the next big issue. For large superconducting magnet systems under the maximum nuclear heating of 200 W/m 3, cable-in-conduit conductor (CICC) and alternative conductor designs are proposed with a robust design of cryogenic support posts. For access to ignited plasmas, new methods are proposed, in which a long rise-up time over 300 s reduces the heating power to 30 MW and a new proportional-integration-derivative (PID) control of the fueling can handle the thermally unstable plasma at high-density operation. This paper focuses on FFHR2m1, which is a modified version of FFHR.

Mechanical behaviour analysis of superconducting magnet in LHD-type reactor FFHR

The force-free helical reactor (FFHR) is a conceptual design of a steady state fusion reactor that has been studied to demonstrate a LHD-type fusion power plant. The helical coil of the FFHR has a major radius of 14 m, a magnetic energy of 120 GJ and a maximum field of 13 T. An aluminium-alloy jacketed Nb3Sn superconductor and indirect cooling using cooling panels within the coil was proposed as a candidate magnet system for the helical coil. Due to the complicated three-dimensional structure of the helical coil winding, it is very important to clarify the mechanical behaviour of the magnet, by considering not only the overall force and deformation but also the detailed stress and strain behaviour in the cross section of the coil. In this study, we evaluated the mechanical behaviour of the helical coil using a 3D axisymmetric finite element method model by considering the non-axisymmetric electromagnetic force.

Neutronics investigation of advanced self-cooled liquid blanket systems in the helical reactor

Neutronics investigations have been conducted in the design activity of the helical-type reactor Force Free Helical Reactor (FFHR2) adopting Flibe-cooled and Li-cooled advanced liquid blanket systems. In this study, comprehensive investigations and geometry modifications related to the tritium breeding ratios (TBRs), neutron shielding performance and neutron wall loading on the first walls in FFHR2 have been performed by improving the three-dimensional (3D) neutronics calculation system developed for non-axisymmetric helical designs. The total TBRs obtained after modifying the blanket dimensions indicated that all the advanced blanket systems proposed for FFHR2 would achieve adequate tritium self-sufficiency by dimension adjustment and optimization of structures in the breeder layers. However, it appeared that the most important neutronics issue in the present helical blanket configuration was suppression of neutron streaming through the divertor pumping areas and reflection from support structures for protection of poloidal and helical coils. Evaluation of neutron wall loading on the first walls indicated that the peaking factor would be moderated as low as 1.2 by the toroidal and helical effect of the helical-shaped plasma distribution in the helical reactor.

Tritium Control for Flibe/V-Alloy Blanket System

One of the critical issues of Flibe/V-alloy blanket with REDOX control by Be is a large tritium inventory in V-alloy structures. Among the possible solutions to this issue would be to control REDOX not by Be but by addition of MoF6 or WF6 enhancing the reaction from T2 to TF. The present study investigated feasibility of this procedure by thermodynamic and neutronics calculations. Using the blanket dimensions of Force Free Helical Reactor (FFHR), tritium inventory in V-alloy structure and Flibe were estimated based on the calculated equilibrium partial pressures of T2 and TF in various cases of REDOX control by MoF6 or WF6. Also carried out were neutronics examinations for the impact of Mo or W doping in the blanket. The results showed that the tritium inventory in the blanket area would be less than 100g at the TF level of 0.1 and 1 ppm in Flibe with addition of WF6 and MoF6, respectively. WF6 doping is far more advantageous than MoF6 doping for low activation purposes.

New Control Method of the Unstable Operating Point in the FFHR Helical Reactor

A new and simple control method of the unstable operating point in the Force Free Helical Reactor (FFHR) is proposed for the ignited operation with high-density plasma. Proportional-integration-derivative (PID) control of the fueling has been used to obtain the desired fusion power with the fusion power error of e(Pf) = (Pfo − Pf) in the stable operating point. We have discovered that in the unstable regime the error of the fusion power with an opposite sign of e(Pf) = −(Pfo − Pf) can stabilize the unstable operating point. Around the unstable operating point, excess fusion power supplies fueling and then increases the density and decreases the temperature. Less fusion power in the sub-ignited regime reduces the fueling, decreases the density, and increases the temperature. The operating point approaches the final unstable operating point as oscillation is damped away.

Control of tritium in FFHR-2 self-cooled Flibe blanket

The application of Flibe (2LiF + BeF 2 mixed molten salt) to a blanket material of force-free helical reactor (FFHR-2) was discussed based on our recent experiments on redox control of Flibe by Be and on hydrogen permeation through Flinak (0.465LiF + 0.115NaF + 0.42 KF mixed molten salt) as a fluid simulant of Flibe. The two experimental works were carried out under the collaboration work of JUPITER-II. To maintain Flibe in the blanket loop under reduction condition is a key issue to transform TF to T 2 with a fast reaction rate. In order to achieve that, proper rates of Be dissolution and reduction reaction of TF to T 2 in Flibe have to be established in the self-cooled blanket. The overall material balance of tritium in the FFHR-2 system was discussed in terms of the tritium permeation through Flibe and the tritium recovery rate.

Blanket neutronics of Li/vanadium-alloy and Flibe/vanadium-alloy systems for FFHR

Neutronics calculations were carried out to characterize self-cooled Li and Flibe blankets with vanadium alloy (V–4Cr–4Ti) structure, as advanced options for the Force Free Helical Reactor (FFHR-2). In the present paper, the parameter dependence of Tritium Breeding Ratio (TBR) and neutron shield performance was examined for a Li blanket (FFHR-LV) and a Flibe blanket (FFHR-FV). The TBR was independent from the Flibe composition between 30 and 45 mol% of BeF 2 in the case of the FFHR-FV. Due to the increase in the temperature range of the coolant Flibe as a result of the change of the structural materials from JLF-1 (low activation ferritic steel) to V–4Cr–4Ti, the fraction of BeF 2 can be changed from 43 to 31 mol% without changing the margin to coagulation of Flibe. The compositional change of Flibe resulted in a significant decrease in its viscosity. The decrease of TBR by the effects of MHD insulator coating with Er was negligibly small with the expected doping level of Er in the FFHR-LV. The increase in the B 4C in the shielding material resulted in the decrease in the fast neutron flux at the superconducting magnet by 60–70% in both FFHR-LV and FFHR-FV. Optimum neutronics performance for the two blanket systems was derived by the calculation.

05/02873 Rejuvenation of spent fuel in the force-free helical reactor (FFHR)

05/02873 Rejuvenation of spent fuel in the force-free helical reactor (FFHR)

05/01520 Implications of convective scrape-off layer transport for fusion reactors with solid and liquid walls

The blanket design for a force-free helical reactor (FFHR) is presented, which is a demo-relevant heliotron-type D-T fusion reactor based on the first all-superconducting-coils device, LHD (large helical device) under construction in NIFS at present. For the goal of a self-ignited reactor of 3 GW thermal output, the design parameters at the first stage for concept definition of FFHR have been investigated. The main feature of FFHR is a force-free-like configuration of helical coils, which makes it possible to simplify the coil supporting structure and to use a high magnetic field instead of high plasma beta. The other feature is the selection of molten-salt FLiBe as a self-cooling tritium breeder for mainly safety reasons owing to the low tritium inventory, low reactivity with air and water, low pressure operation, and low MHD resistance compatible with a high magnetic field. In particular, as common issues in fusion reactors, the FLiBe blanket system in FFHR is expressed in detail by showing engineering possibilities to overcome key issues on tritium permeation, material corrosion, heat transfer, operation pressure, etc. The basic design for maintenance and repair of the blanket is also discussed.

Rejuvenation of spent fuel in the force-free helical reactor (FFHR)

This study considers spent fuel rejuvenation potential of the force-free helical reactor (FFHR), which is a relevant heliotron-type D-T fusion reactor. For this purpose, three different spent fuels were selected, i.e.: (1) CANDU, (2) PWR-UO 2, and (3) PWR-MOX spent fuels. The spent fuel (volume fraction of 60%), spherically prepared and cladded with SiC (volume fraction of 10%), was located in the fuel zone (FZ) in the blanket of the FFHR. The FZ was cooled with high-pressured helium gas (volume fraction of 30%) for the nuclear heat transfer. The neutronic calculations were performed by solving the Boltzmann transport equation with the help of the neutron transport code XSDRNPM-S/SCALE4.3. The calculations of the time dependent atomic densities of the isotopes were performed for an operation period (OP) of up to 4 years by 75% plant factor ( η) under a first-wall neutron load ( P) of 1.5 MW/m 2. The temporal variations of the atomic densities of the isotopes in the spent fuel composition and other physical parameters were calculated for a discrete time interval (Δ t) of 1/12 year (one month), by using the interface program (code). In all investigated spent fuel cases, the tritium self-sufficiency is maintained for DT fusion driver along the OP. The CANDU spent fuel becomes usable in a conventional CANDU reactor after a regeneration time of ∼6 months. The CFFE value approaches 3.5% in the blanket fuelled with the PWR-UO 2 and -MOX spent fuels after 40 and 34 months, respectively. The plutonium component can never reach a nuclear weapon grade quality during spent fuel rejuvenation. Consequently, the blanket has high neutronic performance for the rejuvenation of the spent fuels.