Steady-state Fusion Reactor Research Articles

3D toroidal physics: Testing the boundaries of symmetry breakinga)

Toroidal symmetry is an important concept for plasma confinement; it allows the existence of nested flux surface MHD equilibria and conserved invariants for particle motion. However, perfect symmetry is unachievable in realistic toroidal plasma devices. For example, tokamaks have toroidal ripple due to discrete field coils, optimized stellarators do not achieve exact quasi-symmetry, the plasma itself continually seeks lower energy states through helical 3D deformations, and reactors will likely have non-uniform distributions of ferritic steel near the plasma. Also, some level of designed-in 3D magnetic field structure is now anticipated for most concepts in order to provide the plasma control needed for a stable, steady-state fusion reactor. Such planned 3D field structures can take many forms, ranging from tokamaks with weak 3D edge localized mode suppression fields to stellarators with more dominant 3D field structures. This motivates the development of physics models that are applicable across the full range of 3D devices. Ultimately, the questions of how much symmetry breaking can be tolerated and how to optimize its design must be addressed for all fusion concepts. A closely coupled program of simulation, experimental validation, and design optimization is required to determine what forms and amplitudes of 3D shaping and symmetry breaking will be compatible with the requirements of future fusion reactors.

Visible wide angle view imaging system of KTM tokamak based on multielement image fiber bundle.

In the paper, new visible wide angle view imaging system of KTM tokamak is described. The system has been designed to observe processes inside of plasma and the processes occurring due to plasma-wall interactions through the long equatorial port. Imaging system is designed based on special image fiber bundle and entrance wide angle lens, which provide image of large section of the vacuum chamber, both poloidal half-section and divertor through the sufficiently long equatorial port. The system also consists of two video cameras: slow and fast with image intensifier. Commercial equipment had been used in design of the system that allowed reducing the cost and time for research and development. The paper also discusses advantages and disadvantages of the system in comparison with conventional endoscopes based on a lens system and considers its promising utilization in future tokamaks and future steady state fusion reactors.

Protection of tokamak plasma facing components by a capillary porous system with lithium

Development of plasma facing material (PFM) based on the Capillary-Porous System (CPS) with lithium and activity on realization of lithium application strategy are addressed to meet the challenges under the creation of steady-state tokamak fusion reactor and fusion neutron source. Presented overview of experimental study of lithium CPS in plasma devices demonstrates the progress in protection of tokamak plasma facing components (PFC) from damage, stabilization and self-renewal of liquid lithium surface, elimination of plasma pollution and lithium accumulation in tokamak chamber. The possibility of PFC protection from the high power load related to cooling of the tokamak boundary plasma by radiation of non-fully stripped lithium ions supported by experimental results. This approach demonstrated in scheme of closed loops of Li circulation in the tokamak vacuum chamber and realized in a series of design of tokamak in-vessel elements.

Critical Issues of Burning Plasma, Engineering, Economic and Environmental Assessments on Steady-State Fusion Reactors

For burning plasma simulation and reactor system analysis on steady-state high beta fusion reactors, TOTAL physics code and PEC engineering code have been developed. From TOTAL analysis, it is clarified that by choosing appropriate external current drive profile, high bootstrap-current fraction is achieved in steady-state. From PEC analysis, it is found that the current drive efficiency should be raised for cost of electricity (COE) and CO2 reductions in rather low-beta reactors. Newly derived scaling formulas on COE and life-cycle CO2 emission rate might contribute to the future reactor design projection.

Conceptual Design of Electron Density Measurement System for DEMO-Relevant Helical Plasmas

Electron density measurement remains indispensable to control fueling on a DEMO reactor. For steady-state operation of the DEMO reactor, density measurement should be highly reliable and accurate. A dispersion interferometer and a Faraday polarimeter are free from measurement errors caused by mechanical vibrations. Hence combination of the two diagnostics yields a suitable system for density measurement on future steady-state fusion reactors. A wavelength around 1 µm is one of the desirable candidates in terms of the fringe shift and the Faraday rotation angle, the variety of optical components, and the efficiency of frequency doubling for the dispersion interferometer. This paper presents a conceptual design for the dispersion interferometer and Faraday polarimeter with a 1 µm light source.

Bolometric imaging of detached plasmas in LHD

Detached plasmas can increase the radiated power fraction and thereby reduce the heat load on divertor tiles, which is important for steady-state fusion reactor operation. In LHD, divertor detachment is achieved through several means including gas puffing and the application of an m/n=1/1 magnetic island using external perturbation coils. A recently upgraded imaging bolometer with a tangential view provides unique images of the plasma radiation, which give essential information about the source, three-dimensional (3-D) structure and dynamics of the radiation belt in the edge region. When compared to the results of edge modelling (EMC3/EIRENE) they can verify the model assumptions and identify the radial location of the radiation region, which is important because it impacts the compatibility between heat load and core plasma confinement and is dependent on impurity transport and magnetic topological effects on energy transport which are included in the 3-D edge model.

Fusion materials irradiations at MaRIE'S fission fusion facility

Los Alamos National Laboratory's proposed signature facility, MaRIE, will provide scientists and engineers with new capabilities for modeling, synthesizing, examining, and testing materials of the future that will enhance the USA's energy security and national security. In the area of fusion power, the development of new structural alloys with better tolerance to the harsh radiation environments expected in fusion reactors will lead to improved safety and lower operating costs. The Fission and Fusion Materials Facility (F 3), one of three pillars of the proposed MaRIE facility, will offer researchers unprecedented access to a neutron radiation environment so that the effects of radiation damage on materials can be measured in situ, during irradiation. The calculated radiation damage conditions within the F 3 match, in many respects, that of a fusion reactor first wall, making it well suited for testing fusion materials. Here we report in particular on two important characteristics of the radiation environment with relevancy to radiation damage: the primary knock-on atom spectrum and the impact of the pulse structure of the proton beam on temporal characteristics of the atomic displacement rate. With respect to both of these, analyses show that F 3 has conditions that are consistent with those of a steady-state fusion reactor first wall.

Improved Resolution and Stability in a Dispersion Interferometer Using a Modulation Amplitude Ratio

On steady-state fusion reactors, reliable electron density measurement is required for fueling control. However, the signal-to-noise ratio (S/N) of conventional interferometers is affected by mechanical vibrations. In addition, fringe jump errors are possible at high densities. One possible approach to these problems is the use of a dispersion interferometer (DI) [1]. The DI is immune to mechanical vibrations, and hence does not need a vibration isolator. A two-color system is also not needed even if short-wavelength lasers (a CO2 laser and a YAG laser, for example) are used. The fringe jump errors are attributed to 2π ambiguity in the phase shift. If short-wavelength laser light with a phase shift smaller than 2π is selected, the interferometer becomes free from fringe jump errors in principle. However, the S/N is significantly degraded because the phase shift due to vibrations (∝ 1/λ, λ: wavelength) increases and that due to a plasma (∝ λ) decreases for conventional interferometers. In contrast, the DI has no issue even if the plasma term is smaller than 2π owing to its immunity to vibrations. Thus, the DI becomes free from fringe jump errors if an appropriate wavelength is selected. Techniques that improve the resolution of DIs are currently being introduced [2–5]. This paper reports improved resolution and long-time stability in a DI that adopts phase modulation [2] with a photoelastic modulator (PEM) and a new signal processing method that uses a modulation amplitude ratio [4, 5]. Figure 1 shows a schematic view of the present optical and electrical systems used for bench testing. The light source is a CO2 laser with an output power and wavelength of 8 W and 10.6 μm, respectively. The focused laser light is injected into a nonlinear crystal of AgGaSe2 to generate the frequency-doubled component (FD1). A mixed beam consisting of the fundamental and frequency-doubled components, whose polarizations are orthogonal, passes through a PEM with a drive frequency

Liquid lithium divertor system for fusion reactor

One of the most critical issues for the steady state fusion reactor is the heat flux in the divertor target. This paper proposes a liquid lithium divertor system to solve this problem. The proposed divertor system consists of a liquid lithium target, an evaporation chamber and a differential evacuation chamber. The heat coming from the fusion plasma along the divertor leg is removed by evaporation of lithium. The lithium vapor is condensed on the wall and is circulated with a pump. The coolant temperature for the wall is high enough to drive a power generator. Narrow slits along the divertor leg and the differential evacuation chamber reduce leakage of lithium vapor to the plasma chamber. A preliminary estimation predicts that the lithium ion density in the core plasma is lower than the plasma density.

Plasma-wall interactions and plasma behaviour in fusion devices with liquid lithium plasma facing components

The application of lithium as a self-recovery and renewable material of plasma facing components (PFC) can be used to solve steady state fusion reactor PFC problems. This paper is a survey of liquid Li use in current tokamaks. Liquid Li as tokamak limiter material has been tested in T-11 M tokamak (TRINITI, RF), in FTU (Italy) and in CDX-U (USA). The idea of T-11 M and FTU liquid Li limiters is based on the Capillary–Pore System (CPS) concept. The main feature of CDX-U toroidal limiter was free liquid Li surface. The crucial issue of tokamak is impurity contamination. Lithium experiments in tokamaks discovered that poor lithium penetration into hot plasma core from its periphery (lithium screening) and the development close to plasma boundary lithium non-coronal irradiative blanket. Lithium screening can be physical ground of lithium ‘emitter–collector’ limiter concept with irradiated blanket and PFC prevention from a high local power load up to level of ITER parameters.

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.

Current Drive

Certain devices aimed at magnetic confinement of thermonuclear plasma rely on the steady flow of an electric current in the plasma. In view of the dominant place it occupies in both the world magnetic-confinement fusion effort and the author’s own activity, the tokamak toroidal configuration is selected as prototype for discussing the question of how such a current can be maintained. Tokamaks require a stationary toroidal plasma current, this being traditionally provided by a pulsed magnetic induction which drives the plasma ring as the secondary of a transformer. Since this mechanism is essentially transient, and steadystate fusion reactor operation has manifold advantages, significant effort is now devoted to developing alternate steady-state means of generating toroidal current. These methods are classed under the global heading of “noninductive current drive” or simply “current drive”, generally, though not exclusively, employing the injection of waves and/or toroidally directed particle beams. In what follows we highlight the physical mechanisms underlying surprisingly various approaches to driving current in a tokamak, downplaying a number of practical and technical issues. When a significant data base exists for a given method, its experimental current drive efficiency and future prospects are detailed.

Stability and Confinement Studies of High-Performance NBI Plasmas in the Large Helical Device Toward a Steady-State Helical Fusion Reactor

Recent progress in plasma performance and the understanding of the related physics in the Large Helical Device is overviewed. The volume-averaged beta value is increased with an increase in the neutral beam injection (NBI) heating power, and it reached 5.0% of the reactor-relevant value. In high-β plasmas, the plasma aspect ratio should be controlled so that the Shafranov shift would be reduced, mainly to suppress transport degradation and the deterioration of the NBI heating efficiency. The operational regime of a high-density plasma with an internal diffusion barrier (IDB) has been extended, and the IDB, which was originally found using the local island divertor, has been realized in the helical divertor configuration. The central density was recorded as high as 1 × 1021 m-3, and the central pressure reached 130 kPa. Based on these high-density plasmas with the IDB, a new ignition scenario has been proposed. This should be a scenario specific to the helical fusion reactor, in which the helical ripple transport would be mitigated. A low-energy positive-NBI system was newly installed for an increase in the direct ion heating power. As a result, the ion temperature (Ti) exceeded 5.2 keV at a density of 1.2 × 1019 m-3 in a hydrogen plasma. Transport analysis shows improvement of ion transport, and the Ti-increase tends to be accompanied by a large toroidal rotation velocity of the order of 50 km/s in the core region. The plasma properties in the extended operational regime are discussed from the perspective of a steady-state helical fusion reactor.

Design and Optimization of Support Posts for Cryogenic Components in FFHR

FFHR is a concept design of a steady-state fusion reactor that has been studied to demonstrate a large helical device (LHD)-type fusion power plant. The weight and thermal contraction of cryogenic components are sustained by support posts. A folded multi-plate-type post adopted in the LHD could be feasible for the FFHR. The dimensions of the post were determined according to the buckling load estimation against a gravitational load. Using this fundamental design, the flexibility and stress distribution of the post were calculated by a finite element method. The flexibility against the radial displacement was 24 kN/mm, and the maximum stress for a carbon-fiber-reinforced plastic and stainless steel plate was 155 and 544 MPa, respectively, which were below their allowable levels. The heat loads were calculated as 10.5 kW at 80 K and 0.34 kW at 4 K; the results revealed that the heat load at 4 K was almost 1/20 compared with that on a post made entirely of stainless steel. Natural frequencies were analyzed to assess safety against external loads such as an earthquake. The results showed that the LHD-type support post was suitable for the FFHR from the mechanical and thermal viewpoints.

Steady-state operation and high energy particle production of MeV energy in the Large Helical Device

Achieving steady-state plasma operation at high plasma temperatures is one of the important goals of worldwide magnetic fusion research. High temperatures of approximately 1–2 keV, and steady-state plasma sustainment operations have been reported. Recently the steady-state operation regime was greatly extended in the Large Helical Device (LHD). A high-temperature plasma was created and maintained for 54 min with 1.6 GJ in the 2005FY experimental programme. The three-dimensional heat-deposition profile of the LHD helical divertor was modified, and during long-pulse discharges it effectively dispersed the heat load using a magnetic axis swing technique developed at the LHD. A sweep of only 3 cm in the major radius of the magnetic axis position (less than 1% of the major radius of the LHD) was enough to disperse the divertor heat load. The steady-state plasma was heated and sustained mainly by hydrogen minority ion heating using ion cyclotron range of frequencies and partially by electron cyclotron of fundamental resonance frequency. By accumulating the small flux of charge-exchanged neutral particles during the long-pulse operation, a high energy ion tail which extended up to 1.6 MeV was observed. This is the first experimental evidence of high energetic ion confinement of MeV range in helical devices. The long-pulse operations lasted until a sudden increase in radiation loss occurred, presumably because of metal wall flakes dropping into the plasma. The sustained line-averaged electron density and temperature were approximately 0.8 × 1019 m−3 and 2 keV, respectively, at a 1.3 GJ discharge (#53776) and 0.4 × 1019 m−3 and 1 keV at a 1.6 GJ discharge (#66053). The average input power was 680 kW and 490 kW, and the plasma duration was 32 min and 54 min, respectively. These successful long operations show that the heliotron configuration has a high potential as a steady-state fusion reactor.

Application of visible bremsstrahlung to a density monitor in steady state fusion reactor

The measurement of electron density is still important in a steady state fusion reactor in the future in addition to neutron and magnetic measurements in order to control the fusion output and to sustain stable plasma discharges. In the present toroidal fusion devices, FIR interferometer and Thomson scattering diagnostics have been used for the electron density measurement as an active diagnostic method. However, it is very difficult to operate such instruments continuously over a year. The use of visible bremsstrahlung as the passive diagnostics is, then, proposed as the electron density monitor instead of such active diagnostics. Since the effective charge, Z eff, is considered to be much less than 2 in the high-density fusion reactor, the bremsstrahlung intensity, which is proportional to the square of electron density, becomes a typical indicator of the electron density. In order to check the feasibility, the line-averaged density and the square root of visible bremsstrahlung signals were compared using two kinds of high-density LHD discharges in a range of (5–10) × 10 13 cm −3, such as H 2 gas puff and repetitively injected H 2 ice pellets discharges. A ratio of the square root of the visible bremsstrahlung signal to the density was a constant in both discharges. The visible bremsstrahlung profile was also compared with the density profile. The results obtained in the present study strongly suggested the use of the visible bremsstrahlung to the density monitor in the future reactor diagnostics.

Thirty-minute plasma sustainment by real-time magnetic-axis swing for effective divertor-load-dispersion in the Large Helical Device

Achieving steady-state plasma operation at high plasma temperatures is one of the important goals of worldwide magnetic fusion research. A high temperature of approximately 2keV, and steady-state plasma-sustainment operation of the Large Helical Device (LHD) [O. Motojima, K. Akaishi, H. Chikaraishi et al., Nucl. Fusion 40, 599 (2000)] is reported. High-temperature plasmas were created and maintained for more than 30min with a world record injected heating power of 1.3GJ. The three-dimensional heat-deposition profile of the LHD helical divertor was modified and during long-pulse discharges it effectively dispersed the heat load using a magnetic-axis swing technique developed at the LHD. A sweep of only 3cm of the major radius of the magnetic axis position (less than 1% of the major radius of the LHD) was enough to disperse the divertor heat load. The modification of the heat-load profile was explained well by field-line tracing. The steady-state plasma was heated and sustained mainly by hydrogen minority ion heating using ion cyclotron range of frequencies. The operation lasted until a sudden increase of radiation loss occurred, presumably because of wall metal flakes dropping into the plasma. The sustained line-averaged electron density was approximately 0.7–0.8×1019m−3. The average input power was 680kW, and the plasma duration was 31min 45s. This successful long operation shows that the heliotron configuration has a high potential as a steady-state fusion reactor.

Estimation of the recycled power associated with the cryogenic refrigeration power of a fusion reactor based on TORE SUPRA experiment and ITER design

The refrigeration power associated with the superconducting magnets and cryopumps of a steady-state fusion reactor is not negligible. The power has to be minimized because it plays a role in the power station global efficiency and in the required amplification factor Q. On the one hand, the long plasma discharges obtained in December 2003 on TORE SUPRA give an insight of the cryogenic losses that might be expected for a steady-state fusion reactor equipped with superconducting magnets. The superfluid bath of the windings in TORE SUPRA allows a simple calorimetric estimation of the cryogenic losses through the temperature evolution of the bath during the long discharge. The different kinds of losses in TORE SUPRA are estimated, discussed and explained. Not all of them will be present in a real reactor. On the other hand, in the framework of ITER preparation, the magnet system and the associated refrigerator have been dimensioned taking into account again all kinds of cold losses. This exercise is important because ITER, by its size, could be relevant to the steady-state reactor situation regarding refrigeration. Based on TORE SUPRA experiment and ITER design it is, therefore, possible to propose for the first time a preliminary figure for the cryoplant power of a steady-state reactor. The order of magnitude of the cryoplant power is ten times lower than that of the fusion reactor recycled power which can be considered acceptable.

Current Drive

Certain devices aimed at magnetic confinement of thermonuclear plasma rely on the steady flow of an electric current in the plasma. In view of the dominant place it occupies in both the world magnetic-confinement fusion effort and the author’s own activity, the tokamak toroidal configuration is selected as prototype for discussing the question of how such a current can be maintained. Tokamaks require a stationary toroidal plasma current, this being traditionally provided by a pulsed magnetic induction which drives the plasma ring as the secondary of a transformer. Since this mechanism is essentially transient, and steady-state fusion reactor operation has manifold advantages, significant effort is now devoted to developing alternate steady-state means of generating toroidal current. These methods are classed under the global heading of “noninductive current drive” or simply “current drive”, generally, though not exclusively, employing the injection of waves and/or toroidally directed particle beams. In what follows we highlight the physical mechanisms underlying surprisingly various approaches to driving current in a tokamak, downplaying a number of practical and technical issues. When a significant data base exists for a given method, its experimental current drive efficiency and future prospects are detailed.

Possible use of the DEMO-S ceramic blanket for neptunium transmutation

The DEMO-S steady-state demonstration fusion reactor uses a ceramic blanket for energy generation and tritium breeding. The possible transmutation of minor actinides of spent nuclear fuel in the helium-cooled ceramic blanket is examined. The reactor tritium breeding ratio, reaction rates and heat deposition in the module materials were analysed. These values were estimated as a function of the fuel element diameter with neptunium, the number of fuel elements in the cassette, and the presence or absence of a moderator around the cassettes. A three-dimensional Monte Carlo calculation using the MCNP code was performed for the full heterogeneous model of the blanket and fuel assemblies. After six years of reactor operation, for the cassette containing 19 fuel elements with NpN, 82% of the neptunium burnt out in the case of the water-cooled cassette and 72% in the case of the lead-cooled cassette. This corresponds to a transmutation of ∼100 kg neptunium in one module.