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It has long been recognised that the shortage of external tritium sources for fusion reactors using D–T, the most promising fusion fuel, requires all such fusion power plants (FPP) to breed their own tritium. It is also recognised that the initial start-up of a fusion reactor will require several kilograms of tritium within a scenario in which radioactive decay, ITER and subsequent demonstrator reactors are expected to have consumed most of the known tritium stockpile. To circumvent this tritium fuel shortage and ultimately achieve steady-state operation for a FPP, it is essential to first accumulate sufficient tritium to compensate for loss due to decay and significant retention in the materials in order to start a new FPP. In this work, we propose to accumulate tritium starting from D–D fusion reactions, since D exists naturally in water, and to gradually build up the D–T plasma targeted in fusion reactor designs. There are two likely D–D fusion reaction channels, (1) D + D → T + p , and (2) D + D → He3 + n . The tritium can be generated via the reaction channel ‘(1)’ and the 2.45 MeV neutrons from ‘(2)’ react with lithium-6 in the breeding blanket to produce more tritium to be fed back into plasma fuel. Quantitative evaluations are conducted for two blanket concepts to assess the feasibility and suitability of this approach to FPP reactors. The preliminary results suggest that initial operation in D–D with continual feedback into the plasma of the tritium produced enables a fusion reactor designed solely for D–T operation to start-up in an acceptably short time-scale without the need for any external tritium source.

It has long been known that high temperature superconductors (HTS) could have an important role to play in the future of tokamak fusion research. Here we report on first results of the use of HTS in a tokamak magnet and on the progress in design and construction of the first fully-HTS tokamak.

The energy and power demands implied by start-up and attainment of operating conditions for a pulsed DEMO have been explored. The information gained has been used to establish the requirements both for transformer recharge and for an energy storage system designed to ensure the plant effectively maintains an uninterrupted 1 GW supply to the grid. It was determined that a minimum of approximately 150 MW of auxiliary heating would be required to access the desired operating conditions for the burn phase, and the current-drive potential of a heating system of this capacity was assessed to determine how the pulse length could be extended beyond purely inductive operation. The overall peak power and energy demand associated with the start-up phase of each pulse of output power was determined for a range of dwell times. These two quantities determine the costs of the subsystems required for the power provision during the transformer recharge phase and the energy storage needed for constant plant output. We are therefore able to produce a simple estimate of such costs as a function of dwell time. For falling dwell times, costs are found to rise sharply below around 500 s, but they remain relatively flat as dwell times increase beyond around 1000 s, over the range considered.

In current designs for the Conceptual Demonstration Fusion Power Plant (DEMO), a large proportion of the auxiliary power is consumed by the neutral beam systems used for heating and current drive. It is known that if the plasma is rotating in the same direction as the fast ions produced by the beams, then the efficiency of the current drive is reduced. However, if the plasma could be induced to rotate against the fast ions, the efficiency of the current drive would be increased, reducing the auxiliary power required by the neutral beam systems and/or the necessary beam acceleration energy. In this paper we estimate what power savings could be achieved, and examine possible methods for driving the counter-rotation. Keeping the plasma density (〈 n e 〉 = 0.96 × 10 20 ) m −3 , D – T ) and temperature ( T i0 = 40 keV) constant, if a counter-rotation of M = −0.2 could be achieved, it is estimated that this technique could lead to power savings of up to 15% of the wallplug power requirements of the deuterium neutral beams, reducing the injected power by 30 MW. Conversely, with the same counter-rotation and keeping the original injected power, beam energies could be reduced from 1.5 MeV to 1.0 MeV whilst driving the same current. Numerical simulations using the Monte Carlo TRANSP/NUBEAM code confirm these calculations. From the back electron current term, it is possible that a counter-injected high-Z beam using neon or argon would not only effectively induce rotation counter to the main beams, but also drive additional co-current. DEMO is expected to have a high radiative fraction to protect the divertor and consequently some method of seeding the core with radiating impurities is required in any case. Alternatively, rotation against the torque of the main beams could be induced by charging the plasma and generating a radial electric field. This could be achieved through the use of ICRH to encourage ion loss from the edge of the plasma, or by tangential beam injection at the edge of the plasma to suffer prompt ion orbit losses onto a recovery target. The power requirements of these methods and corresponding reductions in the main beam injector energy will be discussed.

A study has confirmed the feasibility of designing, fabricating and installing resonant magnetic field perturbation (RMP) coils in JET 1 with the objective of controlling edge localized modes (ELM). A system of two rows of in-vessel coils, above the machine midplane, has been chosen as it not only can investigate the physics of and achieve the empirical criteria for ELM suppression, but also permits variation of the spectra allowing for comparison with other experiments. These coils present several engineering challenges. Conditions in JET necessitate the installation of these coils via remote handling, which will impose weight, dimensional and logistical limitations. And while the encased coils are designed to be conventionally wound and bonded, they will not have the usual benefit of active cooling. Accordingly, coil temperatures are expected to reach 350 °C during bakeout as well as during plasma operations. These elevated temperatures are beyond the safe operating limits of conventional OFHC copper and the epoxies that bond and insulate the turns of typical coils. This has necessitated the use of an alternative copper alloy conductor C18150 (CuCrZr). More importantly, an alternative to epoxy had to be found. An R&D program was initiated to find the best available insulating and bonding material. The search included polyimides and ceramic polymers. The scope and status of this R&D program, as well as the critical engineering issues encountered to date are reviewed and discussed.

Cryogenic pellet injection has been used in START and it is under commissioning for MAST for fuelling and diagnostics. In START NBI, heated plasmas fuelled with deuterium pellets, high normalized density (Greenwald number NG≤1.1) with good energy confinement time (τE/τEelmfITER97≤1.3) both at high beta (βT≤25%, βN≤4.9) have been obtained for time scales close to the energy confinement time (∼τE). In addition, high fuelling efficiency (∼100%) have also been simultaneously observed. A complete pellet injection programme is currently under development in MAST, including the commissioning of a eight pellet injector, technical aspects of launching pellets through guide tubes, dedicated diagnostics, and modelling.

In this paper we describe three novel high resolution diagnostic systems on the Culham tokamaks. The first is a pair of high dispersion, short sampling time visible light spectrometers, employing a fast charge-coupled device (CCD) and a multianode photomultiplier as detectors, allowing unusually high bandwidth ion drift velocity measurements. Secondly, the application of an X-ray CCD to a Johann spectrometer for ion and electron core parameter characterisation is outlined. Lastly, a very high throughput, 30-channel Thomson scattering spectrometer system is described, which exploits the low f number lens design techniques, coated fibres and third-generation image intensifiers now available.

Edge localised modes (ELMs), are presently common in high performance tokamak plasmas, but must be controlled or avoided in larger future tokamaks. They are quasiperiodic instabilities with a frequency that correlates with the plasma's energy confinement and the heat fluxes from ELMs to material surfaces, and have at least two distinct types that can be distinguished by the response of their frequency to heating. An improved characterisation of these instabilities can place constraints on theoretical models, and has the potential to reduce the experimental time presently required for the classification of ELMs and the development of scenarios. It can also provide new insights into the processes responsible for them. In a probability density function (pdf) for the "waiting" time intervals between ELMs was rigorously derived, with simple experimentally motivated assumptions leading to a pdf with the specific form of a Weibull distribution. The Weibull distribution arose from a simple model with a power law form, a more detailed model when evaluated would lead to a different pdf. To test the Weibull model a tool was developed that can detect ELMs in real-time and for non-steadystate data using the light radiation associated with ELMs. The method uses a single dimensionless threshold to determine whether an ELM has occured, which is set independently of the data and in advance of the analysis. Consequently large data sets can be studied in a quick and easy but rigorous manner. Because the derivation of the Weibull pdf assumed steady-state data, a database of JET plasmas with between 3 and 6 seconds of steady H-mode was formed.

For a two week period during the Joint European Torus (JET) 2012 experimental campaign, the same high confinement plasma was repeated 151 times. The dataset was analysed to produce a probability density function (pdf) for the waiting times between edge-localised plasma instabilities ("ELMs"). The result was entirely unexpected. Instead of a smooth single peaked pdf, a succession of 4­5 sharp maxima and minima uniformly separated by 7­8 millisecond intervals was found. Here we explore the causes of this newly observed phenomenon, and conclude that it is either due to a self-organised plasma phenomenon or an interaction between the plasma and a real-time control system. If the maxima are a result of "resonant" frequencies at which ELMs can be triggered more easily, then future ELM control techniques can, and probably will, use them. Either way, these results demand a deeper understanding of the ELMing process.

Single null X-point discharges on COMPASS-D show a gradual L to H transition. Frequent boronization gives long ELM-free H-mode periods, duration > 100 ms, with continuously rising density to > 1.5 × 1010 m−3. The LH transition threshold behaviour is more complex than that predicted by present simple scaling laws, with a well-defined low-density limit being observed and sensitivity to small perturbations at the plasma boundary. Density and recycling behaviour is analysed. In discharges refuelled by recycling alone, the coefficient of recycling as well as the particle confinement time is higher in H-mode periods.

This paper presents a conceptual Mars Exploration Vehicle (MEV) architecture, which includes two unmanned Mars Lander Transfer Vehicles (MLTVs) and a Mars Crew Transfer Vehicle (MCTV) with a crew of four. The MLTVs and MCTV are assembled in low Earth orbit (LEO) from modules launched by four Space Launch System (SLS) and five Delta IV Heavy rockets. The MLTVs and MCTV individually escape from LEO, transit to Mars, brake into Mars orbit using propulsion and aerobraking, and rendezvous and dock in low Mars orbit (LMO). Each MLTV includes an Earth Departure stage (EDS), Mars Transfer Stage (MTS), Lander Service Module (LSM), and two landers: A Mars Personnel Lander (MPL) provides two-way crew transportation between LMO and the surface. Three unmanned Mars Cargo Landers (MCLs) provide one-way cargo transportation and the functionalities of habitats (MCL-H) (2) and rover (MCL-R). The landers rendezvous and assemble on the surface to form a base. The MCTV includes two EDS, two MTS, and the following: (1) The Multi-Purpose Crew Vehicle (MPCV) transports the crew from Earth to LEO, provides propulsion, and returns the crew to Earth after nominal mission completion or in aborts. (2) Three Deep Space Vehicles (DSVs) provide life support consumables, passive biological radiation shielding, crew habitation space, and propulsion. The DSV design was derived from the MCL-H. (3) An Artificial Gravity Module (AGM) allows the MCTV to rotate and generate artificial gravity for the crew and provides photo-voltaic power generation and deep space communications. A miniature magnetosphere (Mini-Mag), a potential key enabler for human interplanetary exploration, is electromagnetically generated on the AGM and provides active crew biological radiation shielding. The MEV architecture is based on many existing or near-term technologies. It incorporates significant modularity and could provide an economical approach to achieve progressively more ambitious stepping stone missions along a flexible path for human solar system exploration: starting with test flights in Earth and lunar orbit and progressing through missions to nearEarth asteroids and the moons of Mars, and culminating in the Mars landing mission.

The COMPASS-D tokamak has recently obtained H-mode plasmas in a single-null X-point configuration with no additional heating. Clear H-mode signatures have been observed including reduction in Dalpha light and the unmistakable appearance of ELMs and ELM-free periods. Recycling and gas puff can play crucial roles in achieving and terminating the H-modes. The emergence of distinct ELMs, from the apparently-random fluctuations, of the background Dalpha light, is usually one of progression, taking some 10s of ms rather than a sudden discontinuity. The relatively slow evolution allows for detail examination of important phenomena, such as improvements to confinement and fluid rotation. The H-mode quality improves with density, unless impurity accumulation becomes important. Improvements to the energy and particle confinement are observed and are accompanied by positive increases in the radial electric field at approximately=4 cm inside the last closed flux surface (LCFS). The ohmic powers in discharges with clearly defined H-modes have not been less, so far, than the power obtained from power threshold scalings from larger machines.

Low‐aspect‐ratio tokamaks offer both the economic advantage of smaller size and a number of physics advantages which are not available at conventional aspect ratio. The Small Tight Aspect Ratio Tokamak (START) [Fusion Technology 1990, edited by B. E. Keen, M. Huguet, and R. Hemsworth (North‐Holland, Amsterdam, 1991), Vol. 1, p. 353] was conceived as a first substantial test of tokamak plasma behavior at low aspect ratio. It has achieved plasma currents up to 200 kA, peak densities of ∼2×1020 m−3 and central electron temperatures of ∼500 eV at an aspect ratio of 1.3–1.5. Central beta values of ∼13% have been measured and the volume‐averaged beta 〈β〉 can approach the Troyon limit. Plasmas are naturally elongated (κ≲2.0) and are vertically stable without feedback control. Major disruptions have not been observed at low aspect ratios (A≤2.0).

Experimental results from the COMPASS-C tokamak reveal a sharp threshold in amplitude above which externally applied static resonant magnetic perturbations (RMPs) induce stationary magnetic islands. Such islands (in particular, m=2, n=1 islands) give rise to a significant degradation in energy and particle confinement, suppression of the sawtooth oscillation and a large change in the impurity ion toroidal velocity. The observed threshold for inducing stationary (2,1) islands is consistent with a phenomenological resistive MHD model which takes into account plasma rotation (including poloidal flow damping) and externally applied resonant fields. Broadly similar results are found for applied fields other than m=2, n=1. Other results from RMP experiments are also discussed, such as the stabilization of rotating MHD activity, stimulated disruptions and extensions to the disruptive density limit. Finally, the likely effect of field errors on large tokamaks is briefly examined in the light of the COMPASS-C results

Low aspect ratio tokamaks may provide a low cost route to developing magnetic fusion energy and may offer simpler and more economic fusion reactors. The first test of this concept with hot plasmas has been conducted with the Small Tight Aspect Ratio Tokamak (START) experiment now in operation at Culham Laboratory. The novel method of plasma formation is described and observations of plasmas at aspect ratios as low as 1.3 are presented, indicating high temperature discharges with good global stability properties and typical tokamak characteristics. This represents a new and significant extension of the parameter range of the conventional tokamak

Otherwise stable discharges can become nonlinearly unstable to disruptive locked modes when subjected to a resonant m=2, n=1 error field from irregular poloidal field coils, as in DIII‐D [Nucl. Fusion 31, 875 (1991)], or from resonant magnetic perturbation coils as in COMPASS‐C [Proceedings of the 18th European Conference on Controlled Fusion and Plasma Physics, Berlin (EPS, Petit‐Lancy, Switzerland, 1991), Vol. 15C, Part II, p. 61]. Experiments in Ohmically heated deuterium discharges with q≊3.5, n ≊ 2 × 1019 m−3 and BT ≊ 1.2 T show that a much larger relative error field (Br21/BT ≊ 1 × 10−3) is required to produce a locked mode in the small, rapidly rotating plasma of COMPASS‐C (R0 = 0.56 m, f≊13 kHz) than in the medium‐sized plasmas of DIII‐D (R0 = 1.67 m, f≊1.6 kHz), where the critical relative error field is Br21/BT ≊ 2 × 10−4. This dependence of the threshold for instability is explained by a nonlinear tearing theory of the interaction of resonant magnetic perturbations with rotating plasmas that p...

The question of the influence of nonaxisymmetric field perturbations on tokamaks is investigated. Recent experiments in the COMPASS‐C tokamak [in Proceedings of the 15th Symposium on Fusion Technology, Utrecht (North‐Holland, Amsterdam, 1989), Vol. 1, p. 361] with externally applied helical fields reveal that magnetic islands do not appear until the applied field exceeds a certain value, when plasma rotation and confinement are affected. A new resistive magnetohydrodynamic model including plasma rotation now provides an explanation of this threshold, and is quantitatively consistent with experimental results in Ohmic plasmas. The results indicate the tolerable error fields in future tokamaks. The effects of perturbations with various poloidal and toroidal mode numbers have been studied.

In the week before the 13th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion Research in Washington, DC, 45 participants from 16 nations gathered in Arlington, VA, for a three day meeting on Research Using Small Tokamaks under the chairmanship of D. Banner (IAEA). One of the aims of this series of meetings is to provide a forum for countries with small or developing tokamak programmes to present their recent results and, perhaps more importantly, to exchange ideas. There are usually some representatives from the more advanced fusion programmes, for example from some of the several small tokamaks just starting operation in the USA. The large number of small tokamaks represented is listed in Table I. There are, of course, many other devices that were not represented at this meeting. Most of the papers were on experimental and theoretical programmes developed within a single laboratory, but there were some co-operative experiments between small and large laboratories. Each talk was accompanied by a written paper which will appear, as usual, in an IAEA Technical Document.

JET supports ITER both in development of technology and in torus operations. The latter include the study of ITER-like scenarios, operation in tritium (a trace tritium campaign has recently been completed), mitigation of ELMs and disruptions, investigation of tritium retention, real time control of plasma parameters, and control of extreme plasma shapes. Operation of highly shaped ITER scenarios has been demonstrated. Heating systems are being upgraded, with the enhanced NB system in full operation, an ITER-like RF antenna under manufacture, coupling of LHCD over large distances achieved and external conjugate-T matching of an RF antenna demonstrated. New diagnostics including halo current probes and neutron and alpha particle detector are being installed.. In-vessel installation is being carried out remotely using up-graded remote handling tools. Analysis of tritiated samples from the 1997 tritium campaign, and the development detritiation methods for hard and soft waste continues. Substantial enhancement of JET to better exploit its capabilities in the preparations for ITER are presently being implemented.

The START experiment (Small Tight Aspect Ratio Tokamak) is a low-budget device under construction that is specifically intended to investigate MHD behavior at extremely tight aspect ratios (as low as R/a-1.2) as well as the effectiveness of a major radius compression technique to produce high toroidal current in such plasmas. The main components of the START assembly are described along with the mode of operation. >

START continues to provide key results on the 'spherical tokamak' concept. This paper outlines the engineering solutions adopted to overcome problems of space restrictions in a very low aspect ratio tokamak, whilst providing high performance at low cost.

A Mega Amp Spherical Tokamak (known as MAST) is proposed to be constructed at Culham Laboratory. The aim is to continue the highly successful tight aspect ratio work pioneered on the START machine at Culham.

The highly successful tight aspect ratio tokamak research pioneered on the START machine at Culham, together with the attractive possibilities of the concept, suggest a larger device should be considered. The design of a Mega Amp Spherical Tokamak is described, operating at much higher currents and over longer pulses than START and compatible with strong additional heating.

Two coils are to be installed inside the COMPASS ‘D’ shaped vacuum vessel for use with experiments on the control of plasma instabilities. The coils will carry a maximum current of 10kA for short periods and are to be energised, using a power amplifier with a closed loop control system, to attempt to maintain correct plasma control and position, particularly during a disruption. A crucial factor in deciding the location of the coils is the need to give near optimum control of the plasma without obscuring diagnostic lines of sight. The coils are made from 310 stainless steel as a compromise between mechanical strength and rigidity and electrical resistivity. A combination of magnetic field design codes and finite element analysis was used to evaluate the electromagnetic forces on the coils and the resulting stresses and reactions in the vacuum vessel and supporting arrangement. Each coil is made in 4 rolled sections and is supported by 8 braces with conducting joints between the sections and insulated connections to the braces. The current is supplied to the coils by means of specially designed coaxial vacuum feed-throughs which are connected by very low inductance feeders to the power supply. The coils can be connected independently or linked in series or parallel to provide a combination of vertical and radial position control.

There is increasing interest in the behaviour of tokamaks with tight aspect ratios, A (A=Ro/a), where toroidal effects become very important [ 1 ], World experience is of tokamaks with A>2.5, but recent theoretical studies predict that devices such as the ‘spherical torus’ reactor may have significant advantages in the search for economic fusion power. As yet however there are no experimental results with high temperature plasmas, in this area. START is specifically intended to study the physics behaviour at extremely tight aspect ratios, as low as 1.2.

Here we describe a new experiment to test the shielding concept of a dipole-like magnetic field and plasma, surrounding a spacecraft forming a 'mini magnetosphere'. Initial laboratory experiments have been conducted to determine the effectiveness of a magnetized plasma barrier to be able to expel an impacting, low beta, supersonic flowing energetic plasma representing the solar wind. Optical and Langmuir probe data of the plasma density, the plasma flow velocity and the intensity of the dipole field clearly show the creation of a narrow transport barrier region and diamagnetic cavity virtually devoid of energetic plasma particles. This demonstrates the potential viability of being able to create a small 'hole' in a solar wind plasma, of the order of the ion Larmor orbit width, in which an inhabited spacecraft could reside in relative safety. The experimental results have been quantitatively compared with a 3D particle-in-cell 'hybrid' code simulation that uses kinetic ions and fluid electrons, showing good qualitative agreement and excellent quantitative agreement. Together the results demonstrate the pivotal role of particle kinetics in determining generic plasma transport barriers.

Experiments are described which investigate the parametric variation of particle confinement in heated and unheated decaying plasmas in the Culham Superconducting Levitron. The parameter range investigated, expressed in terms of a collisionality parameter, is similar to that found in the current generation of reverse-field pinch experiments and in the edge regime of present-day tokamak experiments. The variation of the particle confinement time with electron temperature, electron-density and magnetic field strength was found to be approximately of the form τp ∝ T½ B2/n which is that expected for classical diffusion . The measured confinement times were, however, shorter than those predicted by classical diffusion by a factor of 3–10, depending on the magnetic shear. The diffusion rate was found to be approximately proportional to where s is the mean shear length. Low-frequency spectra of floating-potential fluctuations were measured, and it was found that the amplitude of these fluctuations was reduced by increasing magnetic shear. The resistive-g mode has been suggested as a possible explanation for the low-frequency fluctuations, and the diffusion coefficient predicted for this mode (D = γ/k2) has a scaling similar to that observed.

The observation of a plasma current generated by the absorption of rf waves at th electron cyclotron resonance is reported. The current flows in opposite directions on opposite sides of the electron cyclotron resonance position confirming that the current is driven by asymmetric heating of the electron distribution function. The current (approx.30 mA/W) varies linearly with microwave power and electron temperature and inversely with electron density.

Using a neutral-beam injection power of 3.4 M W, volume-averaged toroidal betas of up to βT = 4.5% have been obtained in low-toroidal-field, low-qψ, vertically elongated discharges in the Doublet III tokamak. This level of βT is above the minimum level required for a tokamak reactor, thus demonstrating that reactor level values of βT are possible in a tokamak device. The observed enhancement of βT with vertical elongation lends confidence in the design of future devices which rely on vertical elongation.

Power from a 28 GHz gyrotron has been applied to the Culham Levitron. Up to 130 kW has been used at both second and fundamental resonance field conditions. At the second harmonic, absorption is weak due to low electron temperature and can be explained by non-resonant collisional damping. At the fundamental the average electron density reached is such that ωp2/ω2 = 0.9 but the electron temperature remains low (~ 10 eV). During the discharges N2 neutrals are evolved from cryogenic surfaces and as a result power lost by radiation increases. It is expected that linear conversion at the upper hybrid resonance will be the main absorption process at the fundamental but under these conditions Bernstein waves generated at the hybrid resonance will be collisionally damped in a distance short compared to the plasma dimensions. The measured plasma profiles are consistent with the power being deposited in that manner.

Summary The first Neutral Injection System Beamline became operational on the Doublet III tokamak in September 1981. Power delivery tests and system debugging proceeded through December and initial experiments on plasma heating were started in January 1982. This first beamline has been running well and as of February 1982 is routinely delivering about 2.0 MW of neutral hydrogen power to various circular and elongated dee-shaped deuterium plasmas. The second beamline is now being brought on-line and we expect to have 5–6 MW of heating power available for definitive β studies in mid 1982. We find that the change in the total plasma stored energy caused by the beams is essentially independent of toroidal field and plasma current, but increases weakly with density. The total stored energy increases with current and density, but is independent of toroidal field. A volume-average βT= 2.2% is obtained in low toroidal field discharges.

Non-inductive current drive using second harmonic ECRH at both 28 GHz and 60 GHz has been studied in the CLEO tokamak. At 60 GHz, RF driven currents of up to 5 kA have been observed at e = 4 × 1018m−3 for 185 kW of injected power, indicating an efficiency of η ≡ eIRFR0/PRF = 0.001 (1020 m−3, A, m, W−1). The RF driven current scaled linearly with total plasma current in the range of 5-15 kA and was maximized when the cyclotron resonance was located near to the centre of the plasma. Sawtooth activity was normally strongly affected and transient sawtooth stabilization was often observed. With detailed theoretical studies it is possible to reproduce both the high absorption efficiencies and the scaling of RF driven current with resonance position seen in the 60 GHz experiments. However, the magnitude of the observed current is a factor of about three below that theoretically predicted. At 28 GHz, no evidence of RF driven current could be detected. Possible reasons for this are discussed.

A human exploration mission to a near-Earth object (NEO) asteroid is currently being considered by NASA as a stepping stone on a "flexible path" for human exploration of the solar system. This mission may occur in the mid 2020’s and could be humanity’s first longduration mission into deep space. It would subject the crew to the harsh interplanetary environment for six to seven months. The NEO exploration mission could provide humankind with valuable experience operating in deep space, as an intermediate step between short duration Apollo lunar missions and missions to Mars which could last nearly three years. The purpose of this paper is to present a candidate vehicle concept to accomplish the NEO exploration mission. This design will highlight key features that will be needed to keep the crew healthy and safe during a sixto seven-month duration mission into interplanetary space, such as sufficient volume for human habitation, artificial gravity, and protection from solar and cosmic radiation. The NEO Exploration Vehicle (NEV) is modular and assembled in earth orbit. It is composed of three major subassemblies: (1) The MultiPurpose Crew Vehicle (MPCV) transports the crew from earth to low earth orbit (LEO). It is launched separately and docks with the other two subassemblies already in LEO. The MPCV also provides for return of the crew to earth after successful completion of the mission or aborts. (2) A Deep Space Vehicle (DSV) provides main propulsion, habitation volume, life support consumables, and passive biological radiation shielding for extended missions into deep space. It is based on a design for a Mars habitat/cargo lander. (3) An artificial gravity module (AGM) links the MPCV and DSV together and allows the NEV to rotate to generate artificial gravity. The AGM also provides crew access between the MPCV and DSV, photo-voltaic power generation, and deep space communications. A miniature magnetosphere (Mini-Mag) is generated on the AGM and provides active crew biological radiation shielding. The NEV would demonstrate key technologies for future missions to Mars as well as the Mini-Mag radiation shield, which could be a key enabler for human interplanetary exploration. The NEV design is applicable to other deep space missions.

The helical distortion of a compact tokamak plasma with currents in modular Furth-Hartman type coils is investigated with respect to the impact on the magnetic field structure and to the ideal magnetohydrodynamic stability. A sequence of configurations, in which the current in the helical coils is increased up to the level of that of the main toroidal coils, is generated. The magnetic field structure acquires a ripple along the field lines when the currents in both sets of coils becomes comparable. The magnitude of this ripple depends on the modulation of the Furth-Hartman coils on their winding surface. The local ideal ballooning stability properties deteriorate with the increasing helical deformation of the plasma shape because the normal magnetic field line curvature becomes larger in regions where the helical contribution enhances the toroidicity component while the local magnetic shear vanishes. The global external kink stability properties, in contrast, improve because the normal curvature on average does not change while the local magnetic shear acquires a noticeable helical modulation.