GW Of Fusion Power Research Articles

A DT fusion reactor design in field-reversed configuration using normal conductive coils

Assuming continued stability, favorable energy confinement time scaling, and an effective current drive and maintenance methods, a feasible DT fusion reactor design is proposed for field-reversed configuration (FRC) which uses normal conductive copper magnetic field coils at room temperature. The reactor has 3 GW fusion power, 1.5 MW m−2 neutron wall loading at the first wall, and thermal loading less than 1 MW m−2 at diverter plates. Plasma has an almost straight cylindrical shape of 40 m in length and 8.6 m in diameter. FRC can obtain very high beta (over 0.7 in average) and the magnetic field strength of the reactor will be 1.215 T, which can be produced by normal conductive coils having 70 m in length, 17.6 m in diameter, and 1.5 m in thickness with 0.6 effective conductive area ratio. Its Ohmic power loss is ~74 MW, which is less than 10% of the expected electric power output. A scenario to reach ignition from the initial formation is considered. At first, two FRCs are formed at the both ends of the reactor by fast theta-pinch with a negative bias magnetic field 6 m in length and 0.5 m in diameter. The FRCs are accelerated up to 250 km s−1 by the gradient of magnetic field strength towards the center of the burning region, collide with each other, and form a single large FRC. Their kinetic energy is converted to thermal energy, and the merged FRC is 10 m in length and 1.8 m in diameter. This FRC plasma is brought to ignition by intensive neutral beam injection (NBI) heating and particle supply. Given 200 s heating duration, the maximum NBI power is ~250 MW before alpha particle heating becomes significant. After ignition, NBI heating is not required, but there is a possibility that some part of equilibrium current must be supplied by the NBI in the MeV region.

New design of cable-in-conduit conductor for application in future fusion reactors

The China Fusion Engineering Test Reactor (CFETR) is a new tokamak device whose magnet system includes toroidal field, central solenoid (CS) and poloidal field coils. The main goal is to build a fusion engineering tokamak reactor with about 1 GW fusion power and self-sufficiency by blanket. In order to reach this high performance, the magnet field target is 15 T. However, the huge electromagnetic load caused by high field and current is a threat for conductor degradation under cycling. The conductor with a short-twist-pitch (STP) design has large stiffness, which enables a significant performance improvement in view of load and thermal cycling. But the conductor with STP design has a remarkable disadvantage: it can easily cause severe strand indentation during cabling. The indentation can reduce the strand performance, especially under high load cycling. In order to overcome this disadvantage, a new design is proposed. The main characteristic of this new design is an updated layout in the triplet. The triplet is made of two Nb3Sn strands and one soft copper strand. The twist pitch of the two Nb3Sn strands is large and cabled first. The copper strand is then wound around the two superconducting strands (CWS) with a shorter twist pitch. The following cable stages layout and twist pitches are similar to the ITER CS conductor with STP design. One short conductor sample with a similar scale to the ITER CS was manufactured and tested with the Twente Cable Press to investigate the mechanical properties, AC loss and internal inspection by destructive examination. The results are compared to the STP conductor (ITER CS and CFETR CSMC) tests. The results show that the new conductor design has similar stiffness, but much lower strand indentation than the STP design. The new design shows potential for application in future fusion reactors.

Liquid lithium loop system to solve challenging technology issues for fusion power plant

Steady-state fusion power plant designs present major divertor technology challenges, including high divertor heat flux both in steady-state and during transients. In addition to these concerns, there are the unresolved technology issues of long term dust accumulation and associated tritium inventory and safety issues. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving reactor plasma performance. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor concept and its variant, the active liquid lithium divertor concept, taking advantage of the enhanced or non-coronal Li radiation in relatively poorly confined divertor plasmas. To maintain the LL purity in a 1 GW-electric class fusion power plant, a closed LL loop system with a modest circulating capacity of ~1 l s−1 is envisioned. We examined two key technology issues: (1) dust or solid particle removal and (2) real time recovery of tritium from LL while keeping the tritium inventory level to an acceptable level. By running the LL-loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to the outside where the dust/impurities can be removed by relatively simple dust filter, cold trap and/or centrifugal separation systems. With ~1 l s−1 LL flow, even a small 0.1% dust content by weight (or 0.5 g s−1) suggests that the LL-loop could carry away nearly 16 tons of dust per year. In a 1 GW-electric (or ~3 GW fusion power) fusion power plant, about 0.5 g s−1 of tritium is needed to maintain the fusion fuel cycle assuming ~1% fusion burn efficiency. It appears feasible to recover tritium (T) in real time from LL while maintaining an acceptable T inventory level. Laboratory tests are being conducted to investigate T recovery feasibility with the surface cold trap concept.

Global shutdown dose rate maps for a DEMO conceptual design

For the calculations of highly reliable shutdown dose rate (SDR) maps in fusion devices like a DEMO plant, the Rigorous-2-step (R2S) method is nowadays routinely applied using high-resolution decay gamma sources from initial high-resolution neutron flux meshes activating all materials in the system. This approach has been utilized in the present paper with the objective to provide SDR results relevant for RH systems of a conceptual DEMO design developed in the EU.The primary objective was to assess specific locations of interest for RH equipment inside the vessel and along the extension of maintenance ports. To this end, a provisional DEMO MCNP model has been used, featuring HCLL-type blankets, tungsten/copper divertor, manifolds, vacuum vessel with ports and toroidal field coils. The operational scenario assumed 2.1GW fusion power and a life-time of 20 years with plant availability of 30%, where removable parts will be extracted after 5.2 years. Results of absorbed dose rate distributions for several relevant materials are presented and discussed in terms of the different contributions from the various activated components.

Depth Profile of Tritium Production in Concrete Around a Fusion Device

The depth profile of tritium in concrete near a fusion device has been calculated from an energy spectrum data of neutron which can be obtained by means of a radiation transport calculation code. The production of tritium increases gradually up to 8 cm depth in concrete, and the concentration at this depth is 1.5 times of that near the surface. The concentration will be 12 kBq/cm3 at the maximum point in concrete after 30 years operation at 3 GW fusion power. These profile data of tritium in concrete are useful for decommissioning of a facility.

Integrated plasma physics modelling for the Culham steady state spherical tokamak fusion power plant

Integrated modelling of important plasma physics issues related to the design of a steady-state spherical tokamak (ST) fusion power plant is described. The key is a steady-state current drive, and 92% of this is provided by a combination of bootstrap and diamagnetic currents, both of which have a substantial toroidal component in a ST. The remaining current is to be provided by either neutral beam injection or radio-frequency waves, and various schemes for providing this are discussed and quantified. The desire to achieve a high bootstrap current drives the design to high plasma pressure, β (normalized to the magnetic field pressure), and high elongation. Both these requirements have implications for ideal magneto-hydrodynamic instability which are discussed. Confinement is addressed both through comparison with the recent scaling laws developed from the conventional tokamak database and self-consistent one-dimensional modelling of the transport processes. This modelling shows that the power required for the current drive (∼50 MW) is sufficient to heat the plasma to a regime where more than 3 GW of fusion power is produced, taking into account the dilution due to He ash and prompt α-particle losses, which are small. A preliminary study of the micro-instabilities, which may be responsible for the turbulent transport is provided. Given assumptions about the particle confinement, we make estimates of the fuelling requirements to maintain the steady state. Finally, the power loading due to the exhaust is derived using theory-based scalings for the scrape-off layer width.

Concept of a Helias ignition experiment

The Helias ignition experiment is an upgraded version of the Wendelstein 7-X experiment. The magnetic configuration is a four-period Helias configuration (major radius 18 m, plasma radius 2.0 m, B = 4.5 T), which presents a more compact option than the five-period configuration. Much effort has been focused on two versions of the four-period configuration. One option is the power reactor HSR4/18 providing at least 3 GW of fusion power and the second is the ignition experiment HSR 4/18i aiming at a minimum of fusion power and the demonstration of self-sustaining burn. The design criteria of the ignition experiment HSR 4/18i are the following: The experiment should demonstrate a safe and reliable route to ignition; self-sustained burn without external heating; steady-state operation during several hundred seconds; reliability of the technical components and tritium breeding in a test blanket. The paper discusses the technical issues of the coil system and describes the vacuum vessel and the shielding blanket. The power balance will be modelled with a transport code and the ignition conditions will be investigated using current scaling laws of energy confinement in stellarators. The plasma parameters of the ignition experiment are: peak density 2–3×1020 m−3, peak temperature 11–15 keV, average beta 3.6% and fusion power 1500–1700 MW.

First wall issues related with energetic particle deposition in a tokamak fusion power reactor

Energetic particle deposition to the wall due to toroidal magnetic field (TF) ripple was assessed for a 2 GW fusion power reactor. When the present allowance for the loss is applied, the alpha particle flux to the wall can be as high as 2×10 18 m −2 s −1 in the reactor, eroding tungsten by ∼20 μm per year. The peak particle fluence over a 2-year operation cycle can reach 10 26 m −2, probably being larger than a critical fluence for blister formation. The result suggests that, for the steady-state tokamak fusion reactor, we should introduce a new design methodology of determining an acceptable level of TF ripple on the basis of particle fluence to the wall, instead of the present one based on a tolerable heat flux.

DEMO plant design beyond ITER

A conceptual design of a DEMO reactor for the first integrated demonstration of generating fusion plant is made under the assumption that its design and construction would be started in 2020s and its operation in 2030s. A steady-state tokamak is minimized to have 5.8 m of major radius with 2.3 GW of fusion power with energy amplification Q exceeding 30. Modest extrapolation of improved plasma physics as well as technology development such as superconducting magnet, blanket and reduced activation material are assumed. It is suggested that demonstrating the fusion as a viable source of energy in plant scale will be possible based on the successful operation of ITER, while a number of technical issues must be solved in coherent programs paralleling ITER.

Blanket and first wall conceptual design issues of the spherical tokamak power plant

This paper presents aspects of the conceptual design of the vessel and in-vessel components of the spherical tokamak power plant. Some of the critical parameters for power plant operation and design of the first wall, blanket and shield, are the sufficient production of tritium needed for sustaining the fusion reaction, the neutron and photon heating rates of the various components and the displacement damage of the structural material. The concept under consideration is of the double null configuration, operating at 4.0 GW of fusion power, with an aspect ratio of 1.4, and a wall loading at the midplane of 5 MW m −2. The structural material is assumed to be 9% Cr–martensitic stainless steel, and the coolant helium gas, with the exception of the copper alloy central column which is pressurised water cooled and uses vanadium alloy as the structural material. The neutron multiplier and tritium generating materials are varied to optimise tritium production, the preferred option being a combination of beryllium and lithium oxide. Taking into consideration the relatively severe thermal and neutron induced loading in the spherical tokamak, the main features of this design are the first wall structure which is independent of the blanket, and the radial ribs that occupy all the space under the toroidal field coils. Detailed neutronics calculations have been carried out, in both one (radial, infinite cylinders) and two (radial-poloidal, toroidally axisymmetric) dimensions, to obtain the neutron spectrum, heating rates, tritium generation and displacement damage at various locations around the machine. In one-dimensional calculations the tritium generation ratio values are in the range of 1–1.5 depending on several factors, such as the first wall thickness and multiplier to tritium generating materials ratio in the blanket. The two-dimensional calculations, with a mixture ratio of beryllium to lithium oxide 4 : 1, result in a tritium generation ratio of 1.34, which is within the design target. Of importance are the materials used as a shield for the central column. Various shields have been analysed and one concept that gave good results was the combination of tungsten–carbide, and vanadium alloy with helium cooling. The results of neutronic calculations for the displacement damage give a maximum dose at the first wall of 55 dpa year −1 for the martensitic steel and 2.1 dpa year −1 for the outer layer of the copper alloy central column at the midplane. These dose values give in-service lifetimes within the minimum design target of 2 years for the structural material and 6 years for the copper alloy in the central column.

Power plant design study of a high aspect ratio Tokamak using a SiC composite structure

The DREAM (DRastically EAsy Maintenance) Tokamak is a fusion power plant which is designed from the viewpoint of maintenance feasibility. For this purpose, the DREAM reactor uses a plasma with a very high aspect ratio ( A) and adopts SiC as a structural material. The choice of SiC affects the design of the core plasma, i.e. large inboard shield thickness, low synchrotron radiation reflectivity, and small plasma elongation for positional stability. The objectives of this study are to explore the feasibility of a high- A device, such as a power plant, and to clarify the technological impact of SiC material on the plasma design. Plasma size is optimized by the physics guidelines similar to ITER. The plasma major and minor radii of DREAM are 16 m and 2 m, respectively, and the average neutron wall load is 2.5 MW m −2, the maximum toroidal field is 20 T, and the fusion power is 5.5 GW. Steady-state operation is obtained with 50 MW of external current-drive power and 90% bootstrap current. The divertor heat load is estimated to be about 10 MW m −2. A radiative divertor concept is adopted to achieve a low divertor plasma temperature. The DREAM Tokamak concept is found to be a possible candidate for a future power plant with more than 5 GW of fusion power and an acceptable divertor condition.

ITER reference accidents

Abstract Reference accidents are selected to envelope possible accident sequences to confirm the adequacy of ITER's safety design. Analyses of ITER reference accidents were carried out starting from postulated initiating events (PIE) to all consequential failures including environmental releases, if any. In case of plasma control malfunctioning, up to 3.3 GW of fusion power can be reached transiently, however, no radiological consequence is expected from those transients. Assuming multiple first wall pipe failures, the pressure suppression system limits the maximum pressure within the vacuum vessel to below 0.5 MPa. Long-term decay heat removal is assured by the vacuum vessel cooling system operating in natural circulation. In case of ex-vessel coolant loss with continued plasma burn, the plasma facing components can potentially reach large temperatures. This is avoided by the fusion power shutdown system that terminates plasma burn. A small in-vessel coolant leak with failure of one vacuum vessel penetration line and several confinement options are investigated to show compliance with release limits. Failures in the magnet system are analyzed and their impact on other systems such as damage to neighboring confinement barriers are investigated by postulating air, water and helium ingress into the cryostat. Failures in the tritium plant are analyzed to demonstrate the adequacy of confinement. Radioactive releases for all reference accidents are shown to be well below conservatively defined design guidelines.

Plasma Control Requirements and Concepts for ITER

Plasma control requirements for the International Thermonuclear Experimental Reactor (ITER) are identified, and an overview of proposed ITER plasma control concepts is presented, ITER will operate with a burning deuterium-tritium plasma to produce 1.5 GW of fusion power for durations of 1000 s or more. Key plasma control requirements to achieve these objectives encompass (a) plasma scenario and sequencing: plasma initiation, current rampup, divertor formation, auxiliary heating, ignition and burn, deignition (fusion power shutdown), and current rampdown and termination; (b) plasma magnetics control: plasma current and shape (R0, a, κ, δ) versus time, plus control of critical plasma-to-first-wall clearance gaps, including ion-cyclotron coupling gap and divertor magnetic configuration, during the diverted heating/ignition/burn/deignition phase of the plasma scenario; (c) plasma kinetics and divertor control: core plasma density and/or fusion power, core impurity content and/or radiated power fraction; core profile control (auxiliary heating and/or current drive), and divertor control (pumping, in-divertor gas and/or impurity injection and magnetic configuration optimization for divertor performance); and (d) fast plasma shutdown: fusion power and current shutdown by means of impurity injection. Physics and hardware concepts are presented as to how these plasma control functions will be implemented. Diagnostic measurements needed for plasma control are summarized. The relationship of plasma control to machine protection and public safety is also addressed.

ITER-EDA Physics Design Requirements and Plasma Performance Assessments

Physics design guidelines, plasma performance estimates, and sensitivity of performance to changes in physics assumptions are presented for the ITER-EDA Interim Design. The overall ITER device parameters have been derived from the performance goals using physics guidelines based on the physics R&D results. The ITER-EDA design has a single-null divertor configuration (divertor at the bottom) with a nominal plasma current of 21 MA, magnetic field of 5.68 T, major and minor radius of 8.14 m and 2.8 m, and a plasma elongation (at the 95% flux surface) of {approximately}1.6 that produces a nominal fusion power of {approximately}1.5 GW for an ignited burn pulse length of {ge}1000 s. The assessments have shown that ignition at 1.5 GW of fusion power can be sustained in ITER for 1000 s given present extrapolations of H-mode confinement ({tau}{sub E} = 0.85 {times} {tau}{sub ITER93H}), helium exhaust ({tau}*{sub He}/{tau}{sub E} = 10), representative plasma impurities (n{sub Be}/n{sub e} = 2%), and beta limit [{beta}{sub N} = {beta}(%)/(I/aB) {le} 2.5]. The provision of 100 MW of auxiliary power, necessary to access to H-mode during the approach to ignition, provides for the possibility of driven burn operations at Q = 15. This enables ITER to fulfill its mission of fusion power ({approximately} 1--1.5 GW) and fluence ({approximately}1 MWa/m{sup 2}) goals if confinement, impurity levels, or operational (density, beta) limits prove to be less favorable than present projections. The power threshold for H-L transition, confinement uncertainties, and operational limits (Greenwald density limit and beta limit) are potential performance limiting issues. Improvement of the helium exhaust ({tau}*{sub He}/{tau}{sub E} {le} 5) and potential operation in reverse-shear mode significantly improve ITER performance.

Impact of H-mode power threshold on ITER performance and operation

H-mode operation is projected to be needed in ITER in order to achieve its objectives of sustained burn for 1000 s with 1.5 GW of fusion power. However, experiments have shown that a minimum level of input power is required to access and sustain the H-mode regime. This paper studies the impact on ITER operation of the presently available scalings of this minimum input or threshold power for both the direct L- to H- and reverse H- to L-mode transitions. We present scenarios for access into the H-mode, for maintaining the plasma in H-mode and for the fusion power shutdown with the reverse transition into L-mode. The commonly used empirical scaling for the L- to H-mode transition with a reduction of a factor of two for the reverse transition has severe consequences on ITER operation and performance and various methods to quantify this impact are presented. We conclude, however, that the ITER mission and physics objectives remain compatible with such a scaling.

Wall conditioning in ITER

This article presents plans for wall conditioning for ITER. Typical machine parameters are 1.5 GW of fusion power and 1000 s of DT burn duration. To achieve these plasma, the wall conditioning is required in ITER. For baseline methods, glow discharge cleaning for zero-magnetic field, and electron cyclotron resonance discharge cleaning and divertor active pumping for high-magnetic fields are considered. Alternative methods are also considered.

ITER blanket designs

The ITER first wall, blanket, and shield system is being designed to handle 1.5 ± 0.3 GW of fusion power and 3 MWa m −2 average neutron fluence. In the basic performance phase of ITER operation, the shielding blanket uses austenitic steel structural material and water coolant. The first wall is made of bimetallic structure, austenitic steel and copper alloy, coated with beryllium and it is protected by beryllium bumper limiters. The choice of copper first wall is dictated by the surface heat flux values anticipated during ITER operation. The water coolant is used at low pressure and low temperature. A breeding blanket has been designed to satisfy the technical objectives of the Enhanced Performance Phase of ITER operation for the Test Program. The breeding blanket design is geometrically similar to the shielding blanket design except it is a self-cooled liquid lithium system with vanadium structural material. Self-healing electrical insulator (aluminum nitride) is used to reduce the MHD pressure drop in the system. Reactor relevancy, low tritium inventory, low activation material, low decay heat, and a tritium self-sufficiency goal are the main features of the breeding blanket design.

Estimation of the tritium production and inventory in beryllium

Abstract Beryllium has been proposed as a candidate material for the neutron multiplier in fusion blanket designs. Tritium will be produced and will accumulate in beryllium under neutron irradiation. The tritium production and inventories under 1.5 and 3.0 GW fusion power operation were calculated for a layered pebble bed blanket with lithium oxide (Li 2 O) breeder and beryllium (Be) multiplier. Neutronics calculations were carried out using the one-dimensional transport code ANISN, and the tritium production due to direct reaction of 9 Be ( n , T ) 7 Li and the two-step reactions 9 Be ( n , α ) 6 Li ( n , α ) T was taken into account. The tritium production due to the two-step reaction was calculated to be 50% of the total tritium production after 1 year full power operation (FPY). The tritium inventory was estimated by considering three kinetic parameters, the permeability from the breeder region, diffusivity in a beryllium matrix, and solubility. Tritium permeation from the breeder region to the beryllium region through a 316SS wall was as much as 3 g h −1 , which is 30% of the tritium production (9.6 g h −1 ) in the breeder region. Using the diffusion coefficient of beryllium with no oxide layer on its surface, the total tritium inventory was calculated to be 7 g FPY −1 , mainly owing to solubility. The content of beryllium oxide significantly affects the effective diffusion coefficient. Using a diffusion coefficient for beryllium with beryllium oxide layer on its surface, the tritium inventory was found to be equal to the amount produced.