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Tritium (T) retention characteristics in dust collected from the divertor in JET with ITER-like wall (JET-ILW) after the third campaign in 2015–2016 (ILW-3) have been examined in individual dust particles by combining radiography (tritium imaging plate technique) and electron probe micro-analysis. The results are summarized and compared with the data obtained after the first campaign in 2011–2012 (ILW-1). The dominant component in ILW-1 dust was carbon (C) originating from tungsten-coated carbon fibre composite (CFC) tiles in JET-ILW divertor and/or legacy of C dust after the JET operation with carbon wall. Around 85% of the total tritium retention in ILW-1 dust was attributed to the C dust. The retention in tungsten (W) and beryllium (Be) dominated particles was 100 times smaller than the highest T retention in carbon-based particles. After ILW-3 the main component contributing to the T retention was W. The number of small W particles with T increased, in comparison to ILW-1, most probably by the exfoliation and fragmentation of W coatings on CFC tiles though T retention in individual W particles was smaller than in C particles. The detection of only very few Be-dominated dust particles found after ILW-1 and ILW-3 could imply stable Be deposits on the divertor tiles.

The time dependence of remanence coercivity and thermal stability were investigated for hard/soft-stacked media consisting of a magnetically hard granular layer underneath a very thin soft layer with a large saturation magnetization, M s . The values of remanence coercivity at measurement times t ′=10 3 and 10 −5 s (pulse field) were measured, and defined as H r and H r P . The remanence coercivity on the recording time scale, H r (1 ns), and the energy barrier, Δ E / kT , were evaluated by fitting H r and H r P to Sharrock's equation taking into account the power law variation of the energy barrier, n . The value of H r (1 ns) for a (Co–Pt)–SiO 2 (9 nm)/Co–SiO 2 (2 nm) stacked medium with an interfacial coupling control layer was about 9 kOe, which was less than half of that of a (Co–Pt)–SiO 2 (9 nm) conventional medium (=21.3 kOe). The value of Δ E / kT for the stacked medium was about 111 ( n =0.7), and was not significantly different from the conventional medium. Moreover, no significant difference in the rate of decrease of H r with increasing temperature was observed between media with and without interlayers. These results indicate that the use of a thin soft layer with high M s was effective at significantly reducing H r with no notable change in thermal stability.

After installation of ferritic steel tiles, fast ion losses due to toroidal field ripple have been reduced by 1/2–1/3. The increase in absorbed power at same injection power can reduce the required number of neutral beam injector (NBI) units to sustain a given normalized beta, βN, resulting in a better flexibility of torque input by increasing the available combination of tangential NBI units. By making use of these advantages to sustain an internal transport barrier (ITB), the performance of long-pulse ELMy H-mode plasmas was improved in terms of sustained duration time for both high βN and high thermal confinement enhancement factor (HH98(y,2)). High βN > 2.3 together with HH98(y,2) ~ 1 was sustained for 23.1 s (~12τR, where τR is the current diffusion time) at q95 ~ 3.3, which also provide high βNHH98(y,2) ≥ 2.2 and a bootstrap current fraction of ≥40%. βNHH98(y,2) of 2.0 was sustained for 28.6 s, which is limited by the maximum injection period of 30s for NBI system. These long-pulse plasmas are possible candidates for ITER hybrid operation scenario. Improved confinement is characterized by the larger thermal components at a given density maintained by lower heating power than in previous experiments. The strength of the ITB depends on the pedestal temperature, which varies with edge density while keeping constant the edge pressure (limited by type I ELMs). The fact that co-toroidal rotation as a result of reduced fast ion losses provides better quality of Te-ITB also contributes the improvement of thermal plasma confinement. These long-pulse plasmas indicate that further investigation to establish high performance plasmas longer than the time scale of wall saturation (τW) with active particle control is essential to establish the operational scenarios for next step devices, where the wall pumping does not work.

Edge localized mode (ELM) measurements in many tokamaks, including ASDEX-Upgrade, DIII-D, JET, JT-60U and MAST, are reviewed, which includes progress in experimental observations at the plasma edge region by means of fast-time resolved diagnostics with high precision, such as scanning probe, radial interferometer chord, BES and tangentially viewing fast-gated camera at the midplane. ELM dynamics data show that the majority of the ELM particle and energy transport should be dominated by ion convection physics and associated timescales. Furthermore, recent diagnostic upgrades on many tokamaks reveal the ELM filament structure and their complex motion towards radial, poloidal and toroidal directions. Approaches to control the Type-I ELMs, in addition to the alternative scenarios to Type-I ELMy H-mode operation (so-called, small/no ELM regimes) are also a key area of research for current tokamaks, which demonstrated a high confinement (being comparable to that of Type-I ELMy H-mode plasmas at similar parameters) in the absence of large, ELM induced, transient heat/particle fluxes to the divertor targets. Although tolerable ELM regimes are obtained in existing devices, their application to ITER is uncertain. Issues of these regimes towards further experiments and power deposition on divertor targets and main chamber wall are discussed.

The 'Progress in the ITER Physics Basis' (PIPB) document is an update of the 'ITER Physics Basis' (IPB), which was published in 1999 [1]. The IPB provided methodologies for projecting the performance of burning plasmas, developed largely through coordinated experimental, modelling and theoretical activities carried out on today's large tokamaks (ITER Physics R&D). In the IPB, projections for ITER (1998 Design) were also presented. The IPB also pointed out some outstanding issues. These issues have been addressed by the Participant Teams of ITER (the European Union, Japan, Russia and the USA), for which International Tokamak Physics Activities (ITPA) provided a forum of scientists, focusing on open issues pointed out in the IPB. The new methodologies of projection and control are applied to ITER, which was redesigned under revised technical objectives. These analyses suggest that the achievement of Q > 10 in the inductive operation is feasible. Further, improved confinement and beta observed with low shear (= high βp = 'hybrid') operation scenarios, if achieved in ITER, could provide attractive scenarios with high Q (> 10), long pulse (>1000 s) operation with beta

A systematic study of the effect of the level of toroidal plasma rotation at the top of the ion temperature pedestal ( ) on the edge localised mode (ELM) characteristics in JT-60U has been performed. The level of toroidal plasma rotation was varied by using different combinations of tangential and perpendicular neutral beam injection (NBI). In the grassy ELM regime at high triangularity (δ) and high safety factor (q), the ELM frequency clearly increased up to ~1400 Hz, when counter (ctr) plasma rotation was increased. The response of the ELM frequency was independent of poloidal beta (βp) in the range 0.84 0.53. Even in non-rotating plasma with balanced-NBIs, a high ELM frequency of ~400 Hz was observed without a large energy loss. When the frequency of the plasma rotation in the co-direction of the plasma current became higher than ~1 kHz, type I ELMs with a frequency of ~20 Hz was observed. The achieved pedestal pressure and plasma confinement were similar both in plasmas with type I ELMs and in plasmas with grassy ELMs. The energy loss due to grassy ELMs was evaluated from the reduction in the electron temperature, and the ratio of the energy loss to the pedestal stored energy was less than 1%.

Modification for long pulse operation up to 65 s with a neutral beam heating power of 12 MW (30 s) has provided an opportunity for studies on plasma-wall interaction on a time scale of tens of seconds. During a long pulse, ELMy H-mode discharge, the net wall-pumping rate, evaluated by a particle balance analysis, gradually decreases, and subsequently, becomes zero in ~20 s after the neutral beam injection starts. This situation is interpreted as 'wall saturation'. Wall saturation is observed in repetitive long pulse discharges. Although the outgas due to the increase in the divertor plate temperature and increase in the static retention with pulses also contribute to the particle balance, the dynamic retention process dominates the particle balance and determines the wall saturation. Under the condition of wall saturation, the plasma density continuously increases due to insufficient divertor-pumping rate. This density rise results in detachment of the outer divertor plasma, followed by an X-point MARFE. Results from the calculation of a 2-dimensional plasma transport code indicate that higher divertor-pumping efficiency by a factor of 1.5 is required to avoid the detachment. Note that no negative effects on impurity behaviour such as carbon bloom, or significant dilution of the core plasma has been observed even in the case of the total input energy reaching 350 MJ.

This report summarizes Type I edge localized mode (ELM) dynamics measurements from a number of tokamaks, including ASDEX-Upgrade, DIII-D, JET, JT-60U and MAST, with the goal of providing guidance and insight for the development of ELM simulation and modelling. Several transport mechanisms are conjectured to be responsible for ELM transport, including convective transport due to filamentary structures ejected from the pedestal, parallel transport due to edge ergodization or magnetic reconnection and turbulent transport driven by the high edge gradients when the radial electric field shear is suppressed. The experimental observations are assessed for their validation, or conflict, with these ELM transport conjectures.

Compatibility of advanced tokamak plasmas with high density and high radiation loss has been investigated in both reversed shear (RS) plasmas and high βp H-mode plasmas with a weak positive shear on JT-60U. In the RS plasmas, the operating regime is extended to high density above the Greenwald density (nGW) with high confinement (HHy2 > 1) and high radiation loss fraction (frad > 0.9) by tailoring the internal transport barriers (ITBs). With a small plasma-wall gap, the radiation loss in the main plasma (inside the magnetic separatrix) reaches 80% of the heating power due to metal impurity accumulation. However, high confinement of HHy2 = 1.2 is sustained even with such a large radiation loss in the main plasma. By neon seeding, the divertor radiation loss is enhanced from 20% to 40% of the total radiation loss. In the high βp H-mode plasmas, high confinement (HHy2 = 0.96) is maintained at high density ( ) with high radiation loss fraction (frad ~ 1) by utilizing high-field-side pellets and argon (Ar) injection. The high is attributed to the formation of strong density ITB. Strong core-edge parameter linkage for confinement improvement is observed, where the pedestal pressure and the core plasma confinement increase together. The measured radiation profile including contributions from all impurities in the main plasma is peaked, and the central radiation is ascribed to the contribution from Ar accumulated inside the ITB. Impurity transport analyses indicate that the Ar density profile, twice as peaked as the electron density profile, which is the same level as that observed in the high βp H-mode plasma, can yield an acceptable radiation profile even with a peaked density profile in a fusion reactor.

Results of coherent edge fluctuation measurements using three diagnostics (a reciprocating Langmuir probe, a two channel O-mode reflectometer, and fast magnetic probes) in H-mode discharges on JFT-2M are presented. In discharges in which a high recycling steady (HRS) H-mode phase is obtained through a transient phase with slightly enhanced Dα intensity, two types of coherent fluctuations are observed. The higher frequency mode (around 300 kHz) is the high frequency mode (HFM) observed in the HRS H-mode (Kamiya K et al 2003 9th IAEA Tech. Meeting H-mode Workshop Topic B-14). The lower frequency mode has a frequency of around 80 kHz. The HFM is detected by a Langmuir probe over a wide region in the SOL, as well as by the reflectometer and magnetic probes. However, the HFM is not detected by the higher frequency (38 GHz) channel of the reflectometer after the HRS transition, suggesting that the HFM is not located deeply inside the plasma. The 80 kHz mode is detected by both channels of the reflectometer and by a Langmuir probe, but not by magnetic probes, suggesting that it is an electrostatic mode. In contrast to the HFM, the 80 kHz mode is detected by the Langmuir probe only near the separatrix during the transient phase, which leads to either the HRS phase or the ELMy phase, and is similar to the fluctuations reported in Shinohara K et al (1998 J. Plasma Fusion Res. 74 607).

Simultaneous fast ELM measurements using a reflectometer, interferometers, Dα intensity and a magnetic probe reveal the detailed characteristics of type I ELMs. From the phase signal of the reflectometer indicating the radial movement of the cut-off layer, four different phases in the ELM event, i.e. a precursor phase, collapse phase, recovery phase and a relaxation phase, were observed. In the collapse phase, the radial extent of the collapse of the density pedestal reached up to twice the pedestal width. A fast drop of the horizontal interferometer signal was observed, while the vertical interferometer on the high-field side edge plasma only exhibited a small and delayed reduction. These features of the plasma response due to ELMs might indicate a non-uniform collapse of the density pedestal in the poloidal direction, localized near the low-field side (LFS) midplane. Expelled particles from the LFS midplane were measured by using scrape-off layer Mach probes, and the heat flux to the divertor target was measured using a fast infrared TV camera. The time constant of the peak heat load was comparable to that of the enhancement of the plasma flow towards the divertor target. The normalized ELM energy loss by pedestal stored energy in JT-60U was found to be less than 10% of the pedestal stored energy. It shows weak dependence on the pedestal parameters such as electron collisionality, parallel ion loss time and the ratio of the pedestal density to the Greenwald density limit.

Recent progress in experimental and theoretical studies of edge localized mode (ELM) physics is reviewed for the reactor relevant plasma regimes, namely the high confinement regimes, that is, H-modes and advanced scenarios.Theoretical approaches to ELM physics, from a linear ideal magnetohydrodynamic (MHD) stability analysis to non-linear transport models with ELMs are discussed with respect to experimental observations, in particular the fast collapse of pedestal pressure profiles, magnetic measurements and scrape-off layer transport during ELMs.High confinement regimes with different types of ELMs are addressed in this paper in the context of development of operational scenarios for ITER. The key parameters that have been identified at present to reduce the energy losses in Type I ELMs are operation at high density, high edge magnetic shear and high triangularity. However, according to the present experimental scaling for the energy losses in Type I ELMs, the extrapolation of such regimes for ITER leads to unacceptably large heat loads on the divertor target plates exceeding the material limits. High confinement H-mode scenarios at high triangularity and high density with small ELMs (Type II), mixed regimes (Type II and Type I) and combined advanced regimes at high βp are discussed for present-day tokamaks. The optimum combination of high confinement and small MHD activity at the edge in Type II ELM scenarios is of interest to ITER. However, to date, these regimes have been achieved in a rather narrow operational window and far from ITER parameters in terms of collisionality, edge safety factor and βp.The compatibility of the alternative internal transport barrier (ITB) scenario with edge pedestal formation and ELMs is also addressed. Edge physics issues related to the possible combination of small benign ELMs (Type III, Type II ELMs, quiescent double barrier) and high performance ITBs are discussed for present-day experiments (JET, JT-60U, DIII-D) in terms of their relevance for ITER. Successful plasma edge control, at high triangularity (~0.5) and high density (~0.7nGR), in ITB scenarios in JET is reported.Active control of ELMs by edge current, pellet injection, impurities and external magnetic perturbations creating an ergodic zone localized at the separatrix are discussed for present-day experiments and from the perspective of future reactors.

A substantial reduction of the L-H transition threshold power was observed under the W-shaped divertor in JT-60U, in comparison with the open divertor. Radiation power in the main region was also decreased in this case. These differences of radiation power from the main plasma between open and W-shaped divertor were not large enough to account for the apparent reduction of threshold power. In this paper, the emphasis is on the edge plasma parameters which relate to the L-H transition. It was found that edge plasma pressure just before the L-H transition in the W-shaped divertor case became smaller than that in the open divertor case. After the modification of divertor geometry, edge ion collisionality just before the L-H transition also became lower and it was established with lower auxiliary power input. It was suggested that reduction of edge ion collisionality arose from neutral particles near the X-point by the analysis of poloidal profile of neutral density. Therefore, neutral particles near the X-point prevented the threshold power for the L-H transition from more reduction at low plasma density, and caused the shift of plasma density which gave minimum threshold power.

The B2-EIRENE code has been compared with plasma edge measurements from Alcator C-MOD, DIII-D, JET, JT-60U in L-mode and H-mode regimes. Similarities in L-mode discharges between JET and JT-60U are well reproduced by the code as well as the sudden approach to detachment seen in JT-60U when a 1% chemical sputtering yield for carbon is assumed. Comparisons between similar L-mode regimes in JET, JT-60U and Alcator C-MOD reveal the larger importance of anomalous energy convection in Alcator C-MOD with regard to larger tokamaks. H-mode discharges of DIII-D and JET can be satisfactorily modelled with similar values of the transport coefficients (approximately an order magnitude smaller than in L-mode) when the power flow into the scrape-off layer (SOL) is dominated by the ion channel, as previous EDGE2D simulations for JET suggested.

Gynogenetic plants of pot gerbera (Gerbera jamesonii) were successfully produced from cultures of unpollinated ovulesin vitro. Genotypic variations in the number of ovules that formed callus were found among the lines tested. One particularly responsive genotype was found among 17 genotypes tested where the frequency of callus-forming ovules was 17.5%. Four genotypes formed no callus at all. The frequency of shoot formation from the callus varied from 0-19.6% in nine genotypes. Ploidy was determined by flow cytometry, and 37 (80.4%) regenerants were haploid, seven (15.2%) were diploid, and two (4.3%) were mixoploid with both haploid and diploid cells. The doubling of chromosomes was achieved by treatment with 0.05% colchicine for 2-6 din vitro, and 24.2-34.1% of treated haploid plants were found to have been diploidized.

Determination of the profiles of the ion temperature and the plasma toroidal rotation has been accomplished by charge exchange recombination spectroscopy in PBX-M. The angular momentum and the thermal ion energy transport have been studied mainly during the H mode phase of a high βp discharge (Ip ≈ 330 kA, 3.5 × 1019 ≤ ne ≤ 6.5 × 1019 m-3) having different heating beam configurations (combination of two perpendicular and two tangential neutral beam injections, abbreviated as 2 perp. NBI and 2 parall. NBI). The toroidal rotation velocity V rises substantially in the region of r/a ≥ 0.5 after the L-H transition, and the V profile (peakedness) is more highly dependent on the beam configuration than the Ti profile. The angular momentum confinement time varies from 147 ms (rigid rotation for 2 perp. NBI) to 39 ms (viscous rotation for 2 parall. NBI). In contrast, the thermal energy confinement time is 44-48 ms and is almost independent of the configuration. The transport analysis shows that the radial angular momentum diffusion is caused mainly by the viscous losses and that the angular momentum diffusivity χ is reduced substantially in the outer minor radius region during the 2 perp. NBI H mode. The neoclassical friction effect between the bulk ions and the impurities may influence the χ profiles locally, where the ion temperature gradient is steep

Relative sensitivities of surface barrier diode detectors used for soft x‐ray imaging of plasma emission were calibrated using synchrotron radiation as a calibration source. The variation of responses among 33 detectors against monochromatized light in the soft x‐ray region were measured at the calibration beam line BL5B of UVSOR. The response of detectors against pulsed incident radiation (20 μs) was stored and averaged by a boxcar integrator. The calibration factors obtained at this experiment were applied to correct the soft x‐ray emission profiles measured from a high‐temperature plasma.

In high density and low temperature peripheral plasmas of JT‐60 U, i.e. detached divertor plasmas, C III and C IV lines were observed by a visible and VUV spectrometers in order to investigate dominant radiators, radiation power and particle balance between the radiators. An emission peak was found between the inner strike and the X‐point. With increasing electron density, the emission peak moved to the X‐point with a constant electron temperature of ∼7 eV. In the case the emission peak was located on the X‐point, the dominant radiators in the emission peak were C2+ and C3+, which contributed 30% and 60% to the total radiative power. It was found that C3+ was produced by the ionization of C2+ and the volume recombination of C4+ at a similar rates. However, the loss flux of C3+ was lower by two orders of magnitude than the C3+ production flux, indicating that another loss mechanism such as transport loss around the X‐point was significant.

H-mode plasmas with high poloidal beta values in the PBX-M tokamak often exhibit periods of edge localized mode (ELM) activity, with each ELM preceded by a short duration (≤ 350 μs) burst of high frequency (200-250 kHz) magnetic activity. The burst grows on a time-scale of 10 μs and disappears just before the rapid increase in the Dα emission that is characteristic of the ELM. The burst of activity is observed at all poloidal locations, with the largest amplitudes at coils on the inner major radius side, indicating that the mode is not outward ballooning in character. Stability calculations indicate that this high frequency ELM precursor is probably the pressure driven ideal kink mode.

The highly indented plasmas of the PBX‐M tokamak experiment [Plasma Physics and Controlled Nuclear Fusion Research (IAEA, Vienna, 1989), Vol. 1, p. 97] have reached plasma regimes of both high volume‐averaged beta (βt), and high‐beta poloidal (βp), and show evidence of the suppression of external surface modes by the passive stabilizing system. Values of βt up to 4.0 I/aB (% MA/m T) with Ti(0)≊4 keV have been obtained. A magnetohydrodynamic analysis of plasmas with βp=2.0 indicates that these plasmas are near the threshold of the second stability regime. A value of βt of 6.8% has been reached with Ti(0)>5 keV and an indentation of 28%. Control of plasma shape is accomplished with a feedback system that uses a moment expansion about a single equilibrium and is augmented by time‐dependent waveforms to redefine plasma shape. Diagnostics to measure the safety factor q have been developed and used to make accurate measurements of q(r) and to verify changes made in q(0).

Progress, since the ITER Physics Basis publication (ITER Physics Basis Editors et al 1999 Nucl. Fusion 39 2137-2664), in understanding the processes that will determine the properties of the plasma edge and its interaction with material elements in ITER is described. Experimental areas where significant progress has taken place are energy transport in the scrape-off layer (SOL) in particular of the anomalous transport scaling, particle transport in the SOL that plays a major role in the interaction of diverted plasmas with the main-chamber material elements, edge localized mode (ELM) energy deposition on material elements and the transport mechanism for the ELM energy from the main plasma to the plasma facing components, the physics of plasma detachment and neutral dynamics including the edge density profile structure and the control of plasma particle content and He removal, the erosion of low- and high-Z materials in fusion devices, their transport to the core plasma and their migration at the plasma edge including the formation of mixed materials, the processes determining the size and location of the retention of tritium in fusion devices and methods to remove it and the processes determining the efficiency of the various fuelling methods as well as their development towards the ITER requirements. This experimental progress has been accompanied by the development of modelling tools for the physical processes at the edge plasma and plasma-materials interaction and the further validation of these models by comparing their predictions with the new experimental results. Progress in the modelling development and validation has been mostly concentrated in the following areas: refinement in the predictions for ITER with plasma edge modelling codes by inclusion of detailed geometrical features of the divertor and the introduction of physical effects, which can play a major role in determining the divertor parameters at the divertor for ITER conditions such as hydrogen radiation transport and neutral-neutral collisions, modelling of the ion orbits at the plasma edge, which can play a role in determining power deposition at the divertor target, models for plasma-materials and plasma dynamics interaction during ELMs and disruptions, models for the transport of impurities at the plasma edge to describe the core contamination by impurities and the migration of eroded materials at the edge plasma and its associated tritium retention and models for the turbulent processes that determine the anomalous transport of energy and particles across the SOL. The implications for the expected performance of the reference regimes in ITER, the operation of the ITER device and the lifetime of the plasma facing materials are discussed.

The introduction of Ion Bernstein Wave Heating (IBWH) experiments in PBX‐M have demonstrated significant bulk ion heating and ion temperature (Ti) profile modification. Ti profiles were obtained using a multi‐channel, charge‐exchange recombination spectroscopy system (the CHERS diagnostic). With IBWH, there are density increases; so to produce comparative discharges when no IBWH is present, additional gas must be injected. For ne, of ∼3×1013 cm−3, the central ion temperature, Ti0, was observed to increase by 350 eV when 0.6 MW of IBWH was applied for ∼300 msec. The product ne Ti0, a measure of the ion stored energy, increases linearly with IBWH power. The Ti profile broadened with IBWH and the ion temperature gradient steepened near the predicted power deposition layer, providing evidence of localized heating. Examination of ΔTi profile evolution indicated that the IBWH power was, in fact, deposited in a localized, off‐axis region.

For the tokamak magnetic confinement concept, theory has predicted a distinct relationship between the plasma shape, the internal magnetic structure, and the presence or absence of fast ion losses in the presence of plasma instabilities. We have, for the first time, carried out measurements of the magnetic safety factor profile, [ital q]([ital r]), in plasmas unstable to a specific instability, the so-called fishbone'' mode. The experimental equilibria reconstructed from these data have been used to demonstrate that when the plasma is unstable to fishbones, the fast ion confinement properties depend strongly on the radius of the magnetic surface where [ital q]([ital r])=1.

From the intensity ratios of the three He I lines measured at 20 kHz, the temporal evolutions of the electron temperature and density during and after the power and the particle flow into the divertor plasma caused by edge localized modes are determined. The electron temperature increases from 70 eV to 80 eV with increasing Dα intensity. Then, at the peak of Dα intensity, the electron temperature starts decreasing down to 60 eV. The electron density increases from 0.1 × 1019 m−3 to 0.3 × 1019 m−3 with increasing Dα intensity, and then starts to decrease more gradually compared with the electron temperature after the peak of Dα intensity. It is interpreted that the increase of the electron temperature is ascribed to the power and the particle flow into the divertor plasma, and that the decrease of the electron temperature and the increase of the electron density are ascribed to the ionization of the recycled neutrals, which consumes the electron energy and produces electrons.

Plasmas exhibiting the ``fishbone`` instability studied on the PBX-M tokamak show a distinct relationship between the plasma shape, the internal magnetic structure, and the presence or absence of fast ion losses associated with the fishbone mode. We have, for the first time, carried out measurements of the magnetic safety factor profile in fishbone-unstable plasmas, and used the knowledge of the associated experimental equilibria to compare the stability and fast ion loss properties of these plasmas with experimental observations.

The design progress in a compact low aspect ratio (low A) DEMO reactor, ‘SlimCS’, and its design issues are reported. The design study focused mainly on the torus configuration including the blanket, divertor, materials and maintenance scheme. For continuity with the Japanese ITER-TBM, the blanket is based on a water-cooled solid breeder blanket. For vertical stability of the elongated plasma and high beta access, the blanket is segmented into replaceable and permanent blankets and a sector-wide conducting shell is arranged inbetween these blankets. A numerical calculation indicates that fuel self-sufficiency can be satisfied when the blanket interior is ideally fabricated. An allowable heat load to the divertor plate should be 8 MW m−2 or lower, which can be a critical constraint for determining a handling power of DEMO.

The authors studied 34 patients undergoing abdominal total hysterectomy in order to evaluate whether epidural clonidine added to lidocaine solution could alter the requirements of sedatives during epidural anesthesia. Patients were randomly assigned to one of four groups; 18 ml of 2% lidocaine with 1:200,000 clonidine (n = 6), 1:100,000 clonidine (n = 7), 1:200,000 epinephrine (n = 13), or neither (plain, n = 8). The requirements of sedatives (diazepam, thiamylal) and analgesic (butorphanol) prior to the second epidural injection were compared among the four groups. The dose of intravenous diazepam or thiamylal required for sedation in the patients receiving lidocaine with 1:100,000 clonidine had a tendency to be smaller as compared with those in other three groups. There was a significant difference (P less than 0.05) in the requirement of diazepam between the patients given lidocaine with 1:100,000 clonidine and those given plain lidocaine. The present results suggest that the addition of clonidine to lidocaine solution could reduce the requirements of sedatives in epidural anesthesia.

A case of an inadvertent massive infusion of water via the humidifier module into the trachea during the use of cascade type heated humidifier is described. A Fisher and Paykel anesthetic humidifier was employed in the exhalation side of Jackson Rees type breathing circuit between the anesthesia machine and patient's endotracheal tube. During anesthesia for inguinal hernia, fresh gas flow was 6 liters per minute in the breathing circuit. Humidifier module was inadvertently upside down, and the water in the module was abruptly injected into the patient's trachea and we could not ventilate for a short period of time. This water infusion into the trachea was caused by a large supply of fresh gas flow into the breathing circuit and by narrowing of tube diameter. It is essential, when using heated humidifier, to examine it carefully before anesthesia.